Application of pulsed multi-ion irradiations in radiation damage research: A stochastic cluster dynamics simulation study

NASA Astrophysics Data System (ADS)

Hoang, Tuan L.; Nazarov, Roman; Kang, Changwoo; Fan, Jiangyuan

2018-07-01

Under the multi-ion irradiation conditions present in accelerated material-testing facilities or fission/fusion nuclear reactors, the combined effects of atomic displacements with radiation products may induce complex synergies in the structural materials. However, limited access to multi-ion irradiation facilities and the lack of computational models capable of simulating the evolution of complex defects and their synergies make it difficult to understand the actual physical processes taking place in the materials under these extreme conditions. In this paper, we propose the application of pulsed single/dual-beam irradiation as replacements for the expensive steady triple-beam irradiation to study radiation damages in materials under multi-ion irradiation.

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

NASA Astrophysics Data System (ADS)

Daum, Eric

2000-12-01

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

Project of electro-cyclotron resonance ion source test-bench for material investigation.

PubMed

Kulevoy, T V; Chalykh, B B; Kuibeda, R P; Kropachev, G N; Ziiatdinova, A V

2014-02-01

Development of new materials for future energy facilities with higher operating efficiency is a challenging and crucial task. However, full-scale testing of radiation hardness for reactor materials is quite sophisticated and difficult as it requires long session of reactor irradiation; moreover, induced radioactivity considerably complicates further investigation. Ion beam irradiation does not have such a drawback; on the contrary, it has certain advantages. One of them is high speed of defect formation. Therefore, it provides a useful tool for modeling of different radiation damages. Improved understanding of material behavior under high dose irradiation will probably allow to simulate reactor irradiation close to real conditions and to make an adequate estimation of material radiation hardness. Since 2008 in Institute for Theoretical and Experimental Physics, the ion beam irradiation experiments are under development at the heavy ion radio frequency quadrupole linac and very important results are obtained already [T. V. Kulevoy et al., in Proceedings of the International Topical Meeting on Nuclear Research Applications and Utilization of Accelerators, IAEA Vienna, Austria, 2009, http://www.pub.iaea.org/MTCD/publications/PDF/P1433_CD/darasets/papers/ap_p5_07.pdf]. Nevertheless, the new test bench based on electro-cyclotron resonance ion source and high voltage platform is developed. The project of the test bench is presented and discussed.

Project of electro-cyclotron resonance ion source test-bench for material investigation

NASA Astrophysics Data System (ADS)

Kulevoy, T. V.; Chalykh, B. B.; Kuibeda, R. P.; Kropachev, G. N.; Ziiatdinova, A. V.

2014-02-01

Development of new materials for future energy facilities with higher operating efficiency is a challenging and crucial task. However, full-scale testing of radiation hardness for reactor materials is quite sophisticated and difficult as it requires long session of reactor irradiation; moreover, induced radioactivity considerably complicates further investigation. Ion beam irradiation does not have such a drawback; on the contrary, it has certain advantages. One of them is high speed of defect formation. Therefore, it provides a useful tool for modeling of different radiation damages. Improved understanding of material behavior under high dose irradiation will probably allow to simulate reactor irradiation close to real conditions and to make an adequate estimation of material radiation hardness. Since 2008 in Institute for Theoretical and Experimental Physics, the ion beam irradiation experiments are under development at the heavy ion radio frequency quadrupole linac and very important results are obtained already [T. V. Kulevoy et al., in Proceedings of the International Topical Meeting on Nuclear Research Applications and Utilization of Accelerators, IAEA Vienna, Austria, 2009, http://www.pub.iaea.org/MTCD/publications/PDF/P1433_CD/darasets/papers/ap_p5_07.pdf]. Nevertheless, the new test bench based on electro-cyclotron resonance ion source and high voltage platform is developed. The project of the test bench is presented and discussed.

Dibasic calcium phosphate dihydrate, USP material compatibility with gamma radiation

NASA Astrophysics Data System (ADS)

Betancourt Quiles, Maritza

Gamma radiation is a commonly used method to reduce the microbial bioburden in compatible materials when it is applied at appropriate dose levels. Gamma irradiation kills bacteria and mold by breaking down the organism’s DNA and inhibiting cell division. The purpose of this study is to determine the radiation dosage to be used to treat Dibasic Calcium Phosphate Dihydrate, USP (DCPD) and to evaluate its physicochemical effects if any, on this material. This material will be submitted to various doses of gamma radiation that were selected based on literature review and existing regulations that demonstrate that this method is effective to reduce or eliminate microbial bioburden in natural source and synthetic materials. Analytical testing was conducted to the DCPD exposed material in order to demonstrate that gamma radiation does not alter the physicochemical properties and material still acceptable for use in the manufacture of pharmaceutical products. The results obtained through this study were satisfactory and demonstrated that the gamma irradiation dosages from 5 to 30 kGy can be applied to DCPD without altering its physicochemical properties. These are supported by the Assay test data evaluation of lots tested before and after gamma irradiation implementation that show no significant statistical difference between irradiated and non irradiated assay results. The results of this study represent an achievement for the industry since they provide as an alternative the use of Gamma irradiation technology to control the microbial growth in DCPD.

Mechanical properties and microstructural change of W–Y2O3 alloy under helium irradiation

PubMed Central

Tan, Xiaoyue; Luo, Laima; Chen, Hongyu; Zhu, Xiaoyong; Zan, Xiang; Luo, Guangnan; Chen, Junling; Li, Ping; Cheng, Jigui; Liu, Dongping; Wu, Yucheng

2015-01-01

A wet-chemical method combined with spark plasma sintering was used to prepare a W–Y2O3 alloy. High-temperature tensile tests and nano-indentation microhardness tests were used to characterize the mechanical properties of the alloy. After He-ion irradiation, fuzz and He bubbles were observed on the irradiated surface. The irradiation embrittlement was reflected by the crack indentations formed during the microhardness tests. A phase transformation from α-W to γ-W was investigated by X-ray diffraction (XRD) and transmission electron microscopy (TEM). Polycrystallization and amorphization were also observed in the irradiation damage layer. The W materials tended to exhibit lattice distortion, amorphization, polycrystallization and phase transformation under He-ion irradiation. The transformation mechanism predicted by the atomic lattice model was consistent with the available experimental observations. These findings clarify the mechanism of the structural transition of W under ion irradiation and provide a clue for identifying materials with greater irradiation resistance. PMID:26227480

Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrani, Kurt A; Karlsen, T. M.; Yamamoto, Yukinori

2015-09-01

A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, exceptmore » for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.« less

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

1976-04-01

Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teysseyre, Sebastien Paul

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.

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

NASA Astrophysics Data System (ADS)

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

2018-03-01

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

Radiation testing of composite materials, in situ versus ex situ effects

NASA Technical Reports Server (NTRS)

Kurland, R. M.; Thomasson, J. F.; Beggs, W. C.

1981-01-01

The effect of post irradiation test environments on tensile properties of representative advanced composite materials (T300/5208, T300/934, C6000/P1700) was investigated. Four ply (+ or - 45 deg/+ or - 45 deg) laminate tensile specimens were exposed in vacuum up to a bulk dose of 1 x 10 to the 10th power rads using a mono-energetic fluence of 700 keV electrons from a Van de Graaff accelerator. Post irradiation testing was performed while specimens were being irradiated (in situ data), in vacuum after cessation of irradiation (in vacuo data), and after exposure to air (ex situ data). Room temperature and elevated temperature effects were evaluated. The radiation induced changes to the tensile properties were small. Since the absolute changes in tensile properties were small, the existance of a post irradiation test environment effect was indeterminate.

Status and Planned Experiments of the Hiradmat Pulsed Beam Material Test Facility at CERN SPS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Charitonidis, Nikolaos; Efthymiopoulos, Ilias; Fabich, Adrian

2015-06-01

HiRadMat (High Irradiation to Materials) is a facility at CERN designed to provide high-intensity pulsed beams to an irradiation area where material samples as well as accelerator component assemblies (e.g. vacuum windows, shock tests on high power targets, collimators) can be tested. The beam parameters (SPS 440 GeV protons with a pulse energy of up to 3.4 MJ, or alternatively lead/argon ions at the proton equivalent energy) can be tuned to match the needs of each experiment. It is a test area designed to perform single pulse experiments to evaluate the effect of high-intensity pulsed beams on materials in amore » dedicated environment, excluding long-time irradiation studies. The facility is designed for a maximum number of 1016 protons per year, in order to limit the activation of the irradiated samples to acceptable levels for human intervention. This paper will demonstrate the possibilities for research using this facility and go through examples of upcoming experiments scheduled in the beam period 2015/2016.« less

Postirradiation thermocyclic loading of ferritic-martensitic structural materials

NASA Astrophysics Data System (ADS)

Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

Flexural strength of proof-tested and neutron-irradiated silicon carbide

NASA Astrophysics Data System (ADS)

Price, R. J.; Hopkins, G. R.

1982-08-01

Proof testing before service is a valuable method for ensuring the reliability of ceramic structures. Silicon carbide has been proposed as a very low activation first-wall and blanket structural material for fusion devices, where it would experience a high flux of fast neutrons. Strips of three types of silicon carbide were loaded in four-point bending to a stress sufficient to break about a third of the specimens. Groups of 16 survivors were irradiated to 2 × 10 26n/ m2 ( E>0.05 MeV) at 740°C and bend tested to failure. The strength distribution of chemically vapor-deposited silicon carbide (Texas Instruments) was virtually unchanged by irradiation. The mean strength of sintered silicon carbide (Carborundum Alpha) was reduced 34% by irradiation, while the Weibull modulus and the truncated strength distribution characteristic of proof-tested material were retained. Irradiation reduced the mean strength of reaction-bonded silicon carbide (Norton NC-430) by 58%, and the spread in strength values was increased. We conclude that for the chemically vapor-deposited and the sintered silicon carbide the benefits of proof testing to eliminate low strength material are retained after high neutron exposures.

The effects of space radiation on a chemically modified graphite-epoxy composite material

NASA Technical Reports Server (NTRS)

Reed, S. M.; Herakovich, C. T.; Sykes, G. F.

1986-01-01

The effects of the space environment on the engineering properties and chemistry of a chemically modified T300/934 graphite-epoxy composite system are characterized. The material was subjected to 1.0 x 10 to the 10th power rads of 1.0 MeV electron irradiation under vacuum to simulate 30 years in geosynchronous earth orbit. Monotonic tension tests were performed at room temperature (75 F/24 C) and elevated temperature (250 F/121 C) on 4-ply unidirectional laminates. From these tests, inplane engineering and strength properties (E sub 1, E sub 2, Nu sub 12, G sub 12, X sub T, Y sub T) were determined. Cyclic tests were also performed to characterize energy dissipation changes due to irradiation and elevated temperature. Large diameter graphite fibers were tested to determine the effects of radiation on their stiffness and strength. No significant changes were observed. Dynamic-mechanical analysis demonstrated that the glass transition temperature was reduced by 50 F(28 C) after irradiation. Thermomechanical analysis showed the occurrence of volatile products generated upon heating of the irradiated material. The chemical modification of the epoxy did not aid in producing a material which was more radiation resistant than the standard T300/934 graphite-epoxy system. Irradiation was found to cause crosslinking and chain scission in the polymer. The latter produced low molecular weight products which plasticize the material at elevated temperatures and cause apparent material stiffening at low stresses at room temperature.

Feasibility Study on S-Band Microwave Radiation and 3D-Thermal Infrared Imaging Sensor-Aided Recognition of Polymer Materials from End-of-Life Vehicles

PubMed Central

Huang, Jiu; Zhu, Zhuangzhuang; Tian, Chuyuan; Bian, Zhengfu

2018-01-01

With the increase the worldwide consumption of vehicles, end-of-life vehicles (ELVs) have kept rapidly increasing in the last two decades. Metallic parts and materials of ELVs can be easily reused and recycled, but the automobile shredder residues (ASRs), of which elastomer and plastic materials make up the vast majority, are difficult to recycle. ASRs are classified as hazardous materials in the main industrial countries, and are required to be materially recycled up to 85–95% by mass until 2020. However, there is neither sufficient theoretical nor practical experience for sorting ASR polymers. In this research, we provide a novel method by using S-Band microwave irradiation together with 3D scanning as well as infrared thermal imaging sensors for the recognition and sorting of typical plastics and elastomers from the ASR mixture. In this study, an industrial magnetron array with 2.45 GHz irradiation was utilized as the microwave source. Seven kinds of ELV polymer (PVC, ABS, PP, EPDM, NBR, CR, and SBR) crushed scrap residues were tested. After specific power microwave irradiation for a certain time, the tested polymer materials were heated up to different extents corresponding to their respective sensitivities to microwave irradiation. Due to the variations in polymer chemical structure and additive agents, polymers have different sensitivities to microwave radiation, which leads to differences in temperature rises. The differences of temperature increase were obtained by a thermal infrared sensor, and the position and geometrical features of the tested scraps were acquired by a 3D imaging sensor. With this information, the scrap material could be recognized and then sorted. The results showed that this method was effective when the tested polymer materials were heated up to more than 30 °C. For full recognition of the tested polymer scraps, the minimum temperature variations of 5 °C and 10.5 °C for plastics and elastomers were needed, respectively. The sorting efficiency was independent of particle sizes but depended on the power and time of the microwave irradiation. Generally, more than 75% (mass) of the tested polymer materials could be successfully recognized and sorted under an irradiation power of 3 kW. Plastics were much more insensitive to microwave irradiation than elastomers. With this method, the tested mixture of the plastic group (PVC, ABS, PP) and the mixture of elastomer group (EPDM, NBR, CR, and SBR) could be fully separated with an efficiency of 100%. PMID:29702564

Feasibility Study on S-Band Microwave Radiation and 3D-Thermal Infrared Imaging Sensor-Aided Recognition of Polymer Materials from End-of-Life Vehicles.

PubMed

Huang, Jiu; Zhu, Zhuangzhuang; Tian, Chuyuan; Bian, Zhengfu

2018-04-27

With the increase the worldwide consumption of vehicles, end-of-life vehicles (ELVs) have kept rapidly increasing in the last two decades. Metallic parts and materials of ELVs can be easily reused and recycled, but the automobile shredder residues (ASRs), of which elastomer and plastic materials make up the vast majority, are difficult to recycle. ASRs are classified as hazardous materials in the main industrial countries, and are required to be materially recycled up to 85⁻95% by mass until 2020. However, there is neither sufficient theoretical nor practical experience for sorting ASR polymers. In this research, we provide a novel method by using S-Band microwave irradiation together with 3D scanning as well as infrared thermal imaging sensors for the recognition and sorting of typical plastics and elastomers from the ASR mixture. In this study, an industrial magnetron array with 2.45 GHz irradiation was utilized as the microwave source. Seven kinds of ELV polymer (PVC, ABS, PP, EPDM, NBR, CR, and SBR) crushed scrap residues were tested. After specific power microwave irradiation for a certain time, the tested polymer materials were heated up to different extents corresponding to their respective sensitivities to microwave irradiation. Due to the variations in polymer chemical structure and additive agents, polymers have different sensitivities to microwave radiation, which leads to differences in temperature rises. The differences of temperature increase were obtained by a thermal infrared sensor, and the position and geometrical features of the tested scraps were acquired by a 3D imaging sensor. With this information, the scrap material could be recognized and then sorted. The results showed that this method was effective when the tested polymer materials were heated up to more than 30 °C. For full recognition of the tested polymer scraps, the minimum temperature variations of 5 °C and 10.5 °C for plastics and elastomers were needed, respectively. The sorting efficiency was independent of particle sizes but depended on the power and time of the microwave irradiation. Generally, more than 75% (mass) of the tested polymer materials could be successfully recognized and sorted under an irradiation power of 3 kW. Plastics were much more insensitive to microwave irradiation than elastomers. With this method, the tested mixture of the plastic group (PVC, ABS, PP) and the mixture of elastomer group (EPDM, NBR, CR, and SBR) could be fully separated with an efficiency of 100%.

In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

NASA Astrophysics Data System (ADS)

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

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

DOE PAGES

Garrison, L. M.; Zenobia, Samuel J.; Egle, Brian J.; ...

2016-08-01

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

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

PubMed

Garrison, L M; Zenobia, S J; Egle, B J; Kulcinski, G L; Santarius, J F

2016-08-01

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

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

NASA Astrophysics Data System (ADS)

Garrison, L. M.; Zenobia, S. J.; Egle, B. J.; Kulcinski, G. L.; Santarius, J. F.

2016-08-01

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA-500 μA; the typical current used is 72 μA, which is an average flux of 9 × 1014 ions/(cm2 s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

The accomplishments of lithium target and test facility validation activities in the IFMIF/EVEDA phase

NASA Astrophysics Data System (ADS)

Arbeiter, Frederik; Baluc, Nadine; Favuzza, Paolo; Gröschel, Friedrich; Heidinger, Roland; Ibarra, Angel; Knaster, Juan; Kanemura, Takuji; Kondo, Hiroo; Massaut, Vincent; Saverio Nitti, Francesco; Miccichè, Gioacchino; O'hira, Shigeru; Rapisarda, David; Sugimoto, Masayoshi; Wakai, Eiichi; Yokomine, Takehiko

2018-01-01

As part of the engineering validation and engineering design activities (EVEDA) phase for the international fusion materials irradiation facility IFMIF, major elements of a lithium target facility and the test facility were designed, prototyped and validated. For the lithium target facility, the EVEDA lithium test loop was built at JAEA and used to test the stability (waves and long term) of the lithium flow in the target, work out the startup procedures, and test lithium purification and analysis. It was confirmed by experiments in the Lifus 6 plant at ENEA that lithium corrosion on ferritic martensitic steels is acceptably low. Furthermore, complex remote handling procedures for the remote maintenance of the target in the test cell environment were successfully practiced. For the test facility, two variants of a high flux test module were prototyped and tested in helium loops, demonstrating their good capabilities of maintaining the material specimens at the desired temperature with a low temperature spread. Irradiation tests were performed for heated specimen capsules and irradiation instrumentation in the BR2 reactor at SCK-CEN. The small specimen test technique, essential for obtaining material test results with limited irradiation volume, was advanced by evaluating specimen shape and test technique influences.

Temperature Effects of Ultraviolet Irradiation on Material Degradation

NASA Astrophysics Data System (ADS)

Mori, Kazuyuki; Ishizawa, Junichiro

Ultraviolet rays (UV) cause organic materials to deteriorate. UV irradiation ground testing is therefore important to understand the “adequate lifetime assessment” and the “end-of-life (EOL) characteristic” of materials used in space. In previous experiments, high temperatures were found to accelerate the UV degradation of cross-linked ethylene tetrafluoroethylene (X-ETFE). This causes concern of potentially similar effects in other materials. In this study, we evaluated UV degradation at high temperatures and subsequently determined materials usable in space that had shown accelerated degradation due to UV irradiation at high temperatures.

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

DOE PAGES

Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; ...

2017-06-27

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

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley

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

Water corrosion of F82H-modified in simulated irradiation conditions by heat treatment

NASA Astrophysics Data System (ADS)

Lapeña, J.; Blázquez, F.

2000-12-01

This paper presents results of testing carried out on F82H in water at 260°C with 2 ppm H 2 and the addition of 0.27 ppm Li in the form of LiOH. Uniform corrosion tests have been carried out on as-received material and on specimens from welded material [TIG and electron beam (EB)]. Stress corrosion cracking (SCC) tests have been carried out in as-received material and in material heat treated to simulate neutron irradiation hardening (1075°C/30' a.c. and 1040°C/30' + 625°C/1 h a.c.) with hardness values of 405 and 270 HV30, respectively. Results for uniform corrosion after 2573 h of testing have shown weight losses of about 60 mg/dm 2. Compact tension (CT) specimens from the as-received material tested under constant load have not experienced crack growth. However, in the simulated irradiation conditions for a stress intensity factor between 40 and 80 MPa√m, crack growth rates of about 7×10 -8 m/s have been measured.

Influence of electron radiation and temperature on the cyclic, matrix dominated response of graphite-epoxy

NASA Technical Reports Server (NTRS)

Reed, Susan M.; Herakovich, Carl T.; Sykes, George F., Jr.

1987-01-01

The effects of electron radiation and elevated temperature on the matrix-dominated cyclic response of standard T300/934 and a chemically modified T300/934 graphite-epoxy are characterized. Both materials were subjected to 1.0 x 10 to the 10th rads of 1.0 MeV electron irradiation, under vacuum, to simulate 30 years in geosynchronous orbit. Cyclic tests were performed at room temperature and elevated temperature (121 C) on 4-ply unidirectional laminates to characterize the effects associated with irradiation and elevated temperature. Both materials exhibited energy dissipation in their response at elevated temperature. The irradiated modified material also exhibited energy dissipation at room temperature. The combination of elevated temperature and irradiation resulted in the most severe effects in the form of lower proportional limits, and greater energy dissipation. Dynamic-mechanical analysis demonstrated that the glass transition temperature, T(g), of the standard material was lowered 39 C by irradiation, wereas the T(g) of the modified material was lowered 28 C by irradiation. Thermomechanical analysis showed the occurrence of volatile products generated upon heating of the irradiated materials.

10 CFR 36.1 - Purpose and scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... area subject to irradiation are contained within a device and are not accessible by personnel), medical radiology or teletherapy, radiography (the irradiation of materials for nondestructive testing purposes), gauging, or open-field (agricultural) irradiations. ...

10 CFR 36.1 - Purpose and scope.

Code of Federal Regulations, 2012 CFR

2012-01-01

... area subject to irradiation are contained within a device and are not accessible by personnel), medical radiology or teletherapy, radiography (the irradiation of materials for nondestructive testing purposes), gauging, or open-field (agricultural) irradiations. ...

10 CFR 36.1 - Purpose and scope.

Code of Federal Regulations, 2013 CFR

2013-01-01

... area subject to irradiation are contained within a device and are not accessible by personnel), medical radiology or teletherapy, radiography (the irradiation of materials for nondestructive testing purposes), gauging, or open-field (agricultural) irradiations. ...

10 CFR 36.1 - Purpose and scope.

Code of Federal Regulations, 2011 CFR

2011-01-01

... area subject to irradiation are contained within a device and are not accessible by personnel), medical radiology or teletherapy, radiography (the irradiation of materials for nondestructive testing purposes), gauging, or open-field (agricultural) irradiations. ...

Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. L. Davis; D. L. Knudson; J. L. Rempe

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status ofmore » INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garrison, L. M., E-mail: garrisonlm@ornl.gov; Egle, B. J.; Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706

2016-08-15

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1997-06-01

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{sup 7} Gy in form of fast neutrons and {gamma}-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination ofmore » 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.« less

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

The effect of low dose rate irradiation on the tensile properties and microstructure of austenitic stainless steel.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allen, T. R.; Tsai, H.; Cole, J. I.

2002-09-17

To assess the effects of long-term, low-dose-rate neutron exposure on mechanical strength and ductility, tensile properties were measured on 12% and 20% cold-worked Type 316 stainless steel. Samples were prepared from reactor core components retrieved from the EBR-II reactor following final shutdown. Sample locations were chosen to cover a dose range of 1-56 dpa at temperatures from 371-440 C and dose rates from 0.5-5.8 x10{sup -7} dpa/s. These dose rates are approximately an order of magnitude lower than those of typical EBR-II test sample locations. The tensile tests for the 12% CW material were performed at 380 C and 430more » C while those for the 20% CW samples were performed at 370 C. In each case, the tensile test temperature approximately matched the irradiation temperature. To help understand the tensile properties, microstructural samples with similar irradiation history were also examined. The strength and loss of work hardening increase the fastest as a function of irradiation dose for the 12% CW material irradiated at lower temperature. The decrease in ductility with increasing dose occurs more rapidly for the 12% CW material irradiated at lower temperature and the 20% cold-worked material. Post-tensile test fractography indicates that at higher dose, the 20% CW samples begin a shift in fracture mode from purely ductile to mainly small facets and slip bands, suggesting a transition toward channel fracture. The fracture for all of the 12% cold-worked samples was ductile. For both the 12% and 20% CW materials, the yield strength increases correlate with changes in void and loop density and size.« less

14 MeV Neutron Irradiation Effect on Superconducting Magnet Materials for Fusion Device

NASA Astrophysics Data System (ADS)

Nishimura, A.; Hishinuma, Y.; Seo, K.; Tanaka, T.; Muroga, T.; Nishijima, S.; Katagiri, K.; Takeuchi, T.; Shindo, Y.; Ochiai, K.; Nishitani, T.; Okuno, K.

2006-03-01

As a large-scale plasma experimental device is planned and designed, the importance of investigations on irradiation effect of 14 MeV neutron increases and an experimental database is desired to be piled up. Recently, intense streaming of fast neutron from ports are reported and degradation of superconducting magnet performance is anticipated. To investigate the pure neutron effect on superconducting magnet materials, a cryogenic target system was newly developed and installed at Fusion Neutronics Source in Japan Atomic Energy Research Institute. Although production rate of 14 MeV neutron is not large, only 14 MeV neutron can be supplied to irradiation test without gamma ray. Copper wires, superconducting wires, glass fiber reinforced composites are irradiated and the irradiation effects are characterized. At the same time, sensors for measuring temperature and magnetic field are irradiated and their performance was investigated after irradiation. This paper presents outline of the cryogenic target system and some irradiation test results.

Towards a programme of testing and qualification for structural and plasma-facing materials in ‘fusion neutron’ environments

NASA Astrophysics Data System (ADS)

Stork, D.; Heidinger, R.; Muroga, T.; Zinkle, S. J.; Moeslang, A.; Porton, M.; Boutard, J.-L.; Gonzalez, S.; Ibarra, A.

2017-09-01

Materials damage by 14.1MeV neutrons from deuterium-tritium (D-T) fusion reactions can only be characterised definitively by subjecting a relevant configuration of test materials to high-intensity ‘fusion-neutron spectrum sources’, i.e. those simulating closely D-T fusion-neutron spectra. This provides major challenges to programmes to design and construct a demonstration fusion reactor prior to having a large-scale, high-intensity source of such neutrons. In this paper, we discuss the different aspects related to these ‘relevant configuration’ tests, including: • generic issues in materials qualification/validation, comparing safety requirements against those of investment protection; • lessons learned from the fission programme, enabling a reduced fusion materials testing programme; • the use and limitations of presently available possible irradiation sources to optimise a fusion neutron testing program including fission-neutron irradiation of isotopically and chemically tailored steels, ion damage by high-energy helium ions and self-ion beams, or irradiation studies with neutron sources of non-fusion spectra; and • the different potential sources of simulated fusion neutron spectra and the choice using stripping reactions from deuterium-beam ions incident on light-element targets.

Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux

NASA Astrophysics Data System (ADS)

Petrie, Christian M.; Koyanagi, Takaaki; McDuffee, Joel L.; Deck, Christian P.; Katoh, Yutai; Terrani, Kurt A.

2017-08-01

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300-350 °C under a representative high heat flux (∼0.66 MW/m2) during one cycle of irradiation in an un-instrumented ;rabbit; capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb the expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. The success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.

Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petrie, Christian M.; Koyanagi, Takaaki; McDuffee, Joel L.

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300–350 °C under a representative high heat flux (~0.66 MW/m 2) during one cycle of irradiation in an un-instrumented “rabbit” capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb themore » expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. Furthermore, the success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.« less

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

An exploratory study to determine applicability of nano-hardness and micro-compression measurements for yield stress estimation

NASA Astrophysics Data System (ADS)

Hosemann, P.; Swadener, J. G.; Kiener, D.; Was, G. S.; Maloy, S. A.; Li, N.

2008-03-01

The superior properties of ferritic/martensitic steels in a radiation environment (low swelling, low activation under irradiation and good corrosion resistance) make them good candidates for structural parts in future reactors and spallation sources. While it cannot substitute for true reactor experiments, irradiation by charged particles from accelerators can reduce the number of reactor experiments and support fundamental research for a better understanding of radiation effects in materials. Based on the nature of low energy accelerator experiments, only a small volume of material can be uniformly irradiated. Micro and nanoscale post irradiation tests thus have to be performed. We show here that nanoindentation and micro-compression testing on T91 and HT-9 stainless steel before and after ion irradiation are useful methods to evaluate the radiation induced hardening.

Space radiation resistant transparent polymeric materials

NASA Technical Reports Server (NTRS)

Giori, C.; Yamauchi, T.

1977-01-01

A literature search in the field of ultraviolet and charged particle irradiation of polymers was utilized in an experimental program aimed at the development of radiation stable materials for space applications. The rationale utilized for material selection and the synthesis, characterization and testing performed on several selected materials is described. Among the materials tested for ultraviolet stability in vacuum were: polyethyleneoxide, polyvinylnaphthalene, and the amino resin synthesized by the condensation of o-hydroxybenzoguanamine with formaldehyde. Particularly interesting was the radiation behavior of poly(ethyleneoxide), irradiation did not cause degradation of optical properties but rather an improvement in transparency as indicated by a decrease in solar absorptance with increasing exposure time.

Design and Testing for a New Thermosyphon Irradiation Vehicle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Felde, David K.; Carbajo, Juan J.; McDuffee, Joel Lee

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heatmore » loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.« less

Tensile and fatigue data for irradiated and unirradiated AISI 310 stainless steel and titanium - 5 percent aluminum - 2.5 percent tin: Application of the method of universal slopes

NASA Technical Reports Server (NTRS)

Debogdan, C. E.

1973-01-01

Irradiated and unirradiated tensile and fatigue specimens of AISI 310 stainless steel and Ti-5Al-2.5Sn were tested in the range of 100 to 10,000 cycles to failure to determine the applicability of the method of universal slopes to irradiated materials. Tensile data for both materials showed a decrease in ductility and increase in ultimate tensile strength due to irradiation. Irradiation caused a maximum change in fatigue life of only 15 to 20 percent for both materials. The method of universal slopes predicted all the fatigue data for the 310 SS (irradiated as well as unirradiated) within a life factor of 2. For the titanium alloy, 95 percent of the data was predicted within a life factor of 3.

Status of Wrought FeCrAl-UO 2 Capsules Irradiated in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Harp, J.; Core, G.

2017-07-01

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction,more » and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.« less

Review on the EFDA programme on tungsten materials technology and science

NASA Astrophysics Data System (ADS)

Rieth, M.; Boutard, J. L.; Dudarev, S. L.; Ahlgren, T.; Antusch, S.; Baluc, N.; Barthe, M.-F.; Becquart, C. S.; Ciupinski, L.; Correia, J. B.; Domain, C.; Fikar, J.; Fortuna, E.; Fu, C.-C.; Gaganidze, E.; Galán, T. L.; García-Rosales, C.; Gludovatz, B.; Greuner, H.; Heinola, K.; Holstein, N.; Juslin, N.; Koch, F.; Krauss, W.; Kurzydlowski, K. J.; Linke, J.; Linsmeier, Ch.; Luzginova, N.; Maier, H.; Martínez, M. S.; Missiaen, J. M.; Muhammed, M.; Muñoz, A.; Muzyk, M.; Nordlund, K.; Nguyen-Manh, D.; Norajitra, P.; Opschoor, J.; Pintsuk, G.; Pippan, R.; Ritz, G.; Romaner, L.; Rupp, D.; Schäublin, R.; Schlosser, J.; Uytdenhouwen, I.; van der Laan, J. G.; Veleva, L.; Ventelon, L.; Wahlberg, S.; Willaime, F.; Wurster, S.; Yar, M. A.

2011-10-01

All the recent DEMO design studies for helium cooled divertors utilize tungsten materials and alloys, mainly due to their high temperature strength, good thermal conductivity, low erosion, and comparably low activation under neutron irradiation. The long-term objective of the EFDA fusion materials programme is to develop structural as well as armor materials in combination with the necessary production and fabrication technologies for future divertor concepts. The programmatic roadmap is structured into four engineering research lines which comprise fabrication process development, structural material development, armor material optimization, and irradiation performance testing, which are complemented by a fundamental research programme on "Materials Science and Modeling". This paper presents the current research status of the EFDA experimental and testing investigations, and gives a detailed overview of the latest results on fabrication, joining, high heat flux testing, plasticity, modeling, and validation experiments.

Effects of ionizing radiation on properties of monolayer and multilayer flexible food packaging materials

NASA Astrophysics Data System (ADS)

Riganakos, K. A.; Koller, W. D.; Ehlermann, D. A. E.; Bauer, B.; Kontominas, M. G.

1999-05-01

Volatile compounds produced in flexible food packaging materials (LDPE, EVAc, PET/PE/EVOH/PE) during electron beam irradiation were isolated by purge and trap technique and identified by combined gas chromatography-mass spectrometry (GC/MS), after thermal desorption and concentration. For comparison purposes non-irradiated films were also studied. Film samples were irradiated at low (5 kGy, corresponding to cold pasteurization), intermediate (20 kGy, corresponding to cold sterilization) and high (100 kGy) doses. It was observed that a number of volatile compounds are produced after irradiation in all cases. Furthermore the amounts of all volatile compounds increase with increasing irradiation dose. Both primary (methyl-derivatives etc.) as well as secondary i.e. oxidation products (ketones, aldehydes, alcohols, carboxylic acids etc.) are produced upon irradiation. These products may affect organoleptic properties and thus shelf-life of prepackaged irradiated foods. No significant changes were observed in the structure of polymer matrices as exhibited by IR spectra after irradiation of the materials at doses tested. Likewise, no significant changes were observed in O 2, H 2O and CO 2 permeability values of plastic packaging materials after irradiation.

HiRadMat at CERN SPS - A test facility with high intensity beam pulses to material samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

Charitonidis, N.; Fabich, A.; Efthymiopoulos, I.

2015-07-01

HiRadMat (High Irradiation to Materials) is a facility at CERN designed to provide high-intensity pulsed beams to an irradiation area where material samples as well as accelerator component assemblies (e.g. vacuum windows, shock tests on high power targets, collimators) can be tested. The beam parameters (SPS 440 GeV protons with a pulse energy of up to 3.4 MJ, or alternatively lead/argon ions at the proton equivalent energy) can be tuned to match the needs of each experiment. It is a test area designed to perform single pulse experiments to evaluate the effect of high-intensity pulsed beams on materials in amore » dedicated environment, excluding long-time irradiation studies. The facility is designed for a 10{sup 16} maximum number of protons per year, in order to limit the activation to acceptable levels for human intervention. This paper will demonstrate the possibilities for research using this facility and showing examples of upcoming experiments scheduled in the beam period 2014/2015. (authors)« less

Testing of Candidate Rigid Heatshield Materials at LHMEL for the Entry, Descent, and Landing Technology Development Project

NASA Technical Reports Server (NTRS)

Sepka, Steven; Gasch, Matthew; Beck, Robin A.; White, Susan

2012-01-01

The material testing results described in this paper were part of a material development program of vendor-supplied, proposed heat shield materials. The goal of this program was to develop low density, rigid material systems with an appreciable weight savings over phenolic-impregnated carbon ablator (PICA) while improving material response performance. New technologies, such as PICA-like materials in honeycomb or materials with variable density through-the-thickness were tested. The material testing took place at the Wright-Patterson Air Force Base Laser Hardened Materials Laboratory (LHMEL) using a 10.6 micron CO2 laser operating with the test articles immersed in a nitrogen-gas environment at 1 atmosphere pressure. Test measurements included thermocouple readings of in-depth temperatures, pyrometer readings of surface temperatures, weight scale readings of mass loss, and sectioned-sample readings of char depth. Two laser exposures were applied. The first exposure was at an irradiance of 450 W/cm2 for 50 or 60 seconds to simulate an aerocapture maneuver. The second laser exposure was at an irradiance of 115 W/cm2 for 100 seconds to simulate a planetary entry. Results from Rounds 1 and 2 of these screening tests are summarized.

The effect of cyclic loading on the irradiation hardening of type 316L stainless steel

NASA Astrophysics Data System (ADS)

Scholz, R.

1997-01-01

Strain controlled fatigue tests have been performed in torsion on annealed type 316L stainless steel irradiated with 19 MeV deuterons at 400°C for shear strain ranges between 0.95% and 1.4%. The irradiation hardening of the material was suppressed to a great extent for continuous cycling conditions in comparison to hold time tests.

Impact Properties of Irradiated HT9 from the Fuel Duct of FFTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byun, Thak Sang; Maloy, S; Toloczko, M

2012-01-01

This paper reports Charpy impact test data for the ACO-3 duct material (HT9) from the Fast Flux Test Facility (FFTF) and its archive material. Irradiation doses for the specimens were in the range of 3 148 dpa and irradiation temperatures in the range of 378 504 oC. The impact tests were performed for the small V-notched Charpy specimens with dimensions of 3 4 27 mm at an impact speed of 3.2 m/s in a 25J capacity machine. Irradiation lowered the upper-shelf energy (USE) and increased the transition temperatures significantly. The shift of transition temperatures was greater after relatively low temperaturemore » irradiation. The USE values were in the range of 5.5 6.7 J before irradiation and decreased to the range of 2 5 J after irradiation. Lower USEs were measured for lower irradiation temperatures and specimens with T-L orientation. For the irradiated specimens, the dose dependences of transition temperature and USE were not significant because of the radiation effect on impact behavior nearly saturated at the lowest dose of about 3 dpa. A comparison showed that the lateral expansion of specimens showed a linear correlation with absorbed impact energy, but with large scatter in the results. The size effect was also discussed to clarify the differences in the impact data of subsize and standard specimens.« less

10 CFR 36.1 - Purpose and scope.

Code of Federal Regulations, 2014 CFR

2014-01-01

... area subject to irradiation are contained within a device and are not accessible by personnel), medical radiology or teletherapy, radiography (the irradiation of materials for nondestructive testing purposes), gauging, or open-field (agricultural) irradiations. [58 FR 7728, Feb. 9, 1993, as amended at 78 FR 17007...

Experimental demonstration of radiation effects on the performance of a stirling-alternator convertor and candidate materials evaluation

NASA Astrophysics Data System (ADS)

Mireles, Omar R.

Free-piston Stirling power convertors are under consideration by NASA for service in the Advanced Stirling Radioisotope Generator (ASRG) and Fission Surface Power (FSP) systems to enable aggressive exploration missions by providing a reliable and constant power supply. The ASRG must withstand environmental radiation conditions, while the FSP system must tolerate a mixed neutron and gamma-ray environment resulting from self-irradiation. Stirling-alternators utilize rare earth magnets and a variety of organic materials whose radiation limits dominate service life estimates and shielding requirements. The project objective was to demonstrate the performance of the alternator, identify materials that exhibit excessive radiation sensitivity, identify radiation tolerant substitutes, establish empirical dose limits, and demonstrate the feasibility of cost effective nuclear and radiation tests by selection of the appropriate personnel and test facilities as a function of hardware maturity. The Stirling Alternator Radiation Test Article (SARTA) was constructed from linear alternator components of a Stirling convertor and underwent significant pre-exposure characterization. The SARTA was operated at the Sandia National Laboratories Gamma Irradiation Facility to a dose of over 40 Mrad. Operating performance was within nominal variation, although modestly decreasing trends occurred in later runs as well as the detection of an electrical fault after the final exposure. Post-irradiation disassembly and internal inspection revealed minimal degradation of the majority of the organic components. Radiation testing of organic material coupons was conducted since the majority of the literature was inconsistent. These inconsistencies can be attributed to testing at environmental conditions vastly different than those Stirling-alternator organics will experience during operation. Samples were irradiated at the Texas A&M TRIGA reactor to above expected FSP neutron fluence. A thorough materials evaluation followed and results indicate that the majority of material properties experienced minimal statistically significant change.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley

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

Effects of anisotropy and irradiation on the deformation behavior of Zircaloy 2. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pelloux, R.M.; Ballinger, R.; Lucas, G.

1979-01-01

An experimental program investigated the effects of texture anisotropy and irradiation on the mechanical behavior of Zircaloy-2. Short time and time dependent mechanical behavior were considered. Irradiation effects were simulated through the use of 4.75 MeV protons. The temperature ranges investigated were 298/sup 0/K and 573 to 673/sup 0/K. Both cold worked-stress relieved and annealed material were used in this experimental program. Short time yield behavior of different crystallographic textures was determined by uniaxial and plane strain tests in the temperature range 298/sup 0/K and 573 to 673/sup 0/K. Monotonic flow loci were constructed for each texture. Yield behavior ismore » a strong function of the crystallographic texture number f at all temperatures investigated. The rotation of texture with increasing plastic strain was investigated as a function of initial texture at 298/sup 0/K and 623/sup 0/K. The rate of texture rotation df/epsilon/sub p/ was found to be a unique function of the initial texture for plastic strains less than 0.08. Time dependent mechanical behavior was investigated in the range 573 to 673/sup 0/K using constant load creep and stress relaxation tests. The tensile creep strength is proportional to the resolved fraction of basal poles in the test direction. In variable stress and temperature tests, the time-hardening rule was found to be inapplicable. The strain-hardening rule was applied with success to data obtained at temperatures less than or equal to 648/sup 0/K. Irradiation creep tests were conducted in vacuum at 598/sup 0/K and 102 to 241 MPa on 80..mu..m thick Zircaloy-2 foil specimens in both the recrystallized and cold worked-stress relieved condition. In the irradiation creep tests irradiation hardening and enhanced irradiation creep were observed. Radiation hardening effects were significant in annealed material but were attenuated in cold worked-stress relieved material.« less

Effect of different surface treatments on tensile bond strength of silicone-based soft denture liner.

PubMed

Akin, Hakan; Tugut, Faik; Mutaf, Burcu; Akin, Gulsah; Ozdemir, A Kemal

2011-11-01

Failure of the bond between the acrylic resin and resilient liner material is commonly encountered in clinical practice. The purpose of this study was to investigate the effect of different surface treatments (sandblasting, Er:YAG, Nd:YAG, and KTP lasers) on tensile bond strength of silicone-based soft denture liner. Polymethyl methacrylate test specimens were fabricated and each received one of eight surface treatments: untreated (control), sandblasted, Er:YAG laser irradiated, sandblasted + Er:YAG laser irradiated, Nd:YAG laser irradiated, sandblasted + Nd:YAG laser irradiated, KTP laser irradiated, and sandblasted + KTP laser irradiated. The resilient liner specimens (n = 15) were processed between two polymethyl methacrylate (PMMA) blocks. Bonding strength of the liners to PMMA were compared by tensile test with the use of a universal testing machine at a crosshead speed of 5 mm/min. Kruskal-Wallis and Wilcoxon tests were used to analyze the data (α = 0.05). Altering the polymethyl methacrylate surface by Er:YAG laser significantly increased the bond strengths in polymethyl methacrylate/silicone specimens, however, sandblasting before applying a lining material had a weakening effect on the bond. In addition, Nd:YAG and KTP lasers were found to be ineffective for increasing the strength of the bond.

High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guillen, Donna; Greenwood, Lawrence R.; Parry, James

2014-06-22

A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated formore » up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.« less

Modeling and testing miniature torsion specimens for SiC joining development studies for fusion

DOE PAGES

Henager, Jr., C. H.; Nguyen, Ba N.; Kurtz, Richard J.; ...

2015-08-05

The international fusion community has designed a miniature torsion specimen for neutron irradiation studies of joined SiC and SiC/SiC composite materials. For this research, miniature torsion joints based on this specimen design were fabricated using displacement reactions between Si and TiC to produce Ti 3SiC 2 + SiC joints with SiC and tested in torsion-shear prior to and after neutron irradiation. However, many miniature torsion specimens fail out-of-plane within the SiC specimen body, which makes it problematic to assign a shear strength value to the joints and makes it difficult to compare unirradiated and irradiated strengths to determine irradiation effects.more » Finite element elastic damage and elastic–plastic damage models of miniature torsion joints are developed that indicate shear fracture is more likely to occur within the body of the joined sample and cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. The model results are compared and discussed with regard to unirradiated and irradiated test data for a variety of joint materials. The unirradiated data includes Ti 3SiC 2 + SiC/CVD-SiC joints with tailored joint moduli, and includes steel/epoxy and CVD-SiC/epoxy joints. Finally, the implications for joint data based on this sample design are discussed.« less

Inverse Analysis of Irradiated NuclearMaterial Gamma Spectra via Nonlinear Optimization

NASA Astrophysics Data System (ADS)

Dean, Garrett James

Nuclear forensics is the collection of technical methods used to identify the provenance of nuclear material interdicted outside of regulatory control. Techniques employed in nuclear forensics include optical microscopy, gas chromatography, mass spectrometry, and alpha, beta, and gamma spectrometry. This dissertation focuses on the application of inverse analysis to gamma spectroscopy to estimate the history of pulse irradiated nuclear material. Previous work in this area has (1) utilized destructive analysis techniques to supplement the nondestructive gamma measurements, and (2) been applied to samples composed of spent nuclear fuel with long irradiation and cooling times. Previous analyses have employed local nonlinear solvers, simple empirical models of gamma spectral features, and simple detector models of gamma spectral features. The algorithm described in this dissertation uses a forward model of the irradiation and measurement process within a global nonlinear optimizer to estimate the unknown irradiation history of pulse irradiated nuclear material. The forward model includes a detector response function for photopeaks only. The algorithm uses a novel hybrid global and local search algorithm to quickly estimate the irradiation parameters, including neutron fluence, cooling time and original composition. Sequential, time correlated series of measurements are used to reduce the uncertainty in the estimated irradiation parameters. This algorithm allows for in situ measurements of interdicted irradiated material. The increase in analysis speed comes with a decrease in information that can be determined, but the sample fluence, cooling time, and composition can be determined within minutes of a measurement. Furthermore, pulse irradiated nuclear material has a characteristic feature that irradiation time and flux cannot be independently estimated. The algorithm has been tested against pulse irradiated samples of pure special nuclear material with cooling times of four minutes to seven hours. The algorithm described is capable of determining the cooling time and fluence the sample was exposed to within 10% as well as roughly estimating the relative concentrations of nuclides present in the original composition.

Neutron-irradiation creep of silicon carbide materials beyond the initial transient

DOE PAGES

Katoh, Yutai; Ozawa, Kazumi; Shimoda, Kazuya; ...

2016-06-04

Irradiation creep beyond the transient regime was investigated for various silicon carbide (SiC) materials. Here, the materials examined included polycrystalline or monocrystalline high-purity SiC, nanopowder sintered SiC, highly crystalline and near-stoichiometric SiC fibers (including Hi-Nicalon Type S, Tyranno SA3, isotopically-controlled Sylramic and Sylramic-iBN fibers), and a Tyranno SA3 fiber–reinforced SiC matrix composite fabricated through a nano-infiltration transient eutectic phase process. Neutron irradiation experiments for bend stress relaxation tests were conducted at irradiation temperatures ranging from 430 to 1180 °C up to 30 dpa with initial bend stresses of up to ~1 GPa for the fibers and ~300 MPa for themore » other materials. Initial bend stress in the specimens continued to decrease from 1 to 30 dpa. Analysis revealed that (1) the stress exponent of irradiation creep above 1 dpa is approximately unity, (2) the stress normalized creep rate is ~1 × 10 –7 [dpa –1 MPa –1] at 430–750 °C for the range of 1–30 dpa for most polycrystalline SiC materials, and (3) the effects on irradiation creep of initial microstructures—such as grain boundary, crystal orientation, and secondary phases—increase with increasing irradiation temperature.« less

TUNABLE IRRADIATION TESTBED

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Asner, David M.

PNNL has developed and continues to develop innovative methods for characterizing irradiated materials from nuclear reactors and particle accelerators for various clients and collaborators around the world. The continued development of these methods, in addition to the ability to perform unique scientific investigations of the effects of radiation on materials could be greatly enhanced with easy access to irradiation facilities. A Tunable Irradiation Testbed with customized targets (a 30 MeV, 1mA cyclotron or similar coupled to a unique target system) is shown to provide a much more flexible and cost-effective source of irradiating particles than a test reactor or isotopicmore » source. The configuration investigated was a single shielded building with multiple beam lines from a small, flexible, high flux irradiation source. Potential applications investigated were the characterization of radiation damage to materials applicable to advanced reactors, fusion reactor, legacy waste, (via neutron spectra tailored to HTGR, molten salt, LWR, LMR, fusion environments); 252Cf replacement; characterization of radiation damage to materials of interest to High Energy Physics to enable the neutrino program; and research into production of short lived isotopes for potential medical and other applications.« less

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

NASA Astrophysics Data System (ADS)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-06-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between -170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.

Development of a remote-controlled fatigue test machine using a laser extensometer for investigation of irradiation effect on fatigue properties

NASA Astrophysics Data System (ADS)

Yonekawa, M.; Ishii, T.; Ohmi, M.; Takada, F.; Hoshiya, T.; Niimi, M.; Ioka, I.; Miwa, Y.; Tsuji, H.

2002-12-01

In order to investigate effects of neutron irradiation on fatigue properties of nuclear materials, a remote-controlled high temperature fatigue test machine was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR) in the Japan Atomic Energy Research Institute (JAERI). A small-sized fatigue specimen having double blades to measure strain with a laser extensometer was designed for this machine. A strain amplitude in fatigue tests of a completely reversed push-pull type using a triangular wave was controlled with an accuracy of ±3% of the total strain range during test. Low cycle fatigue tests of type 304 stainless steel irradiated in JMTR at 823 K up to a fast neutron fluence of 1×10 25 n/m 2 ( E>1 MeV) were performed in total strain ranges of 0.7-1.4% at 823 K using the designed small-sized specimens.

University of Wisconsin Ion Beam Laboratory: A facility for irradiated materials and ion beam analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, K. G.; Wetteland, C. J.; Cao, G.

2013-04-19

The University of Wisconsin Ion Beam Laboratory (UW-IBL) has recently undergone significant infrastructure upgrades to facilitate graduate level research in irradiated materials phenomena and ion beam analysis. A National Electrostatics Corp. (NEC) Torodial Volume Ion Source (TORVIS), the keystone upgrade for the facility, can produce currents of hydrogen ions and helium ions up to {approx}200 {mu}A and {approx}5 {mu}A, respectively. Recent upgrades also include RBS analysis packages, end station developments for irradiation of relevant material systems, and the development of an in-house touch screen based graphical user interface for ion beam monitoring. Key research facilitated by these upgrades includes irradiationmore » of nuclear fuels, studies of interfacial phenomena under irradiation, and clustering dynamics of irradiated oxide dispersion strengthened steels. The UW-IBL has also partnered with the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) to provide access to the irradiation facilities housed at the UW-IBL as well as access to post irradiation facilities housed at the UW Characterization Laboratory for Irradiated Materials (CLIM) and other ATR-NSUF partner facilities. Partnering allows for rapid turnaround from proposed research to finalized results through the ATR-NSUF rapid turnaround proposal system. An overview of the UW-IBL including CLIM and relevant research is summarized.« less

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

Anisotropic swelling and microcracking of neutron irradiated Ti 3AlC 2-Ti 5Al 2C 3 materials

DOE PAGES

Ang, Caen K.; Silva, Chinthaka M.; Shih, Chunghao Phillip; ...

2015-12-17

M n + 1AX n (MAX) phase materials based on Ti–Al–C have been irradiated at 400 °C (673 K) with fission neutrons to a fluence of 2 × 10 25 n/m 2 (E > 0.1 MeV), corresponding to ~ 2 displacements per atom (dpa). We report preliminary results of microcracking in the Al-containing MAX phase, which contained the phases Ti 3AlC 2 and Ti 5Al 2C 3. Equibiaxial ring-on-ring tests of irradiated coupons showed that samples retained 10% of pre-irradiated strength. Volumetric swelling of up to 4% was observed. Phase analysis and microscopy suggest that anisotropic lattice parameter swelling causedmore » microcracking. Lastly, variants of titanium aluminum carbide may be unsuitable materials for irradiation at light water reactor-relevant temperatures.« less

Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures

NASA Astrophysics Data System (ADS)

Bailey, Nathan A.; Stergar, Erich; Toloczko, Mychailo; Hosemann, Peter

2015-04-01

Oxide dispersion strengthened (ODS) alloys are meritable structural materials for nuclear reactor systems due to the exemplary resistance to radiation damage and high temperature creep. Summarized in this work are atom probe tomography (APT) investigations on a heat of MA957 that underwent irradiation in the form of in-reactor creep specimens in the Fast Flux Test Facility-Materials Open Test Assembly (FFTF-MOTA) for the Liquid Metal Fast Breeder Reactor (LMFBR) program. The oxide precipitates appear stable under irradiation at elevated temperature over extended periods of time. Nominally, the precipitate chemistry is unchanged by the accumulated dose; although, evidence suggests that ballistic dissolution and reformation processes are occurring at all irradiation temperatures. At 412 °C-109 dpa, chromium enrichments - consistent with the α‧ phase - appear between the oxide precipitates, indicating radiation induced segregation. Grain boundaries, enriched with several elements including nickel and titanium, are observed at all irradiation conditions. At 412 °C-109 dpa, the grain boundaries are also enriched in molecular titanium oxide (TiO).

Influence of thermal and radiation effects on microstructural and mechanical properties of Nb-1Zr

NASA Astrophysics Data System (ADS)

Leonard, Keith J.; Busby, Jeremy T.; Zinkle, Steven J.

2011-07-01

The microstructural changes and corresponding effects on mechanical properties, electrical resistivity and density of Nb-1Zr were examined following neutron irradiation up to 1.8 dpa at temperatures of 1073, 1223 and 1373 K and compared with material thermally aged for similar exposure times of ˜1100 h. Thermally driven changes in the development of intragranular and grain boundary precipitate phases showed a greater influence on mechanical and physical properties compared to irradiation-induced defects for the examined conditions. Initial formation of the zirconium oxide precipitates was identified as cubic structured plates following a Baker-Nutting orientation relationship to the β-Nb matrix, with particles developing a monoclinic structure on further growth. Tensile properties of the Nb-1Zr samples showed increased strength and reduced elongation following aging and irradiation below 1373 K, with the largest tensile and hardness increases following aging at 1098 K. Tensile properties at 1373 K for the aged and irradiated samples were similar to that of the as-annealed material. Total elongation was lower in the aged material due to a strain hardening response, rather than a weak strain softening observed in the irradiated materials due in part to an irregular distribution of the precipitates in the irradiated materials. Though intergranular fracture surfaces were observed on the 1248 K aged tensile specimens, the aged and irradiated material showed uniform elongations >3% and total elongation >12% for all conditions tested. Cavity formation was observed in material irradiated to 0.9 dpa at 1073 and 1223 K. However, since void densities were estimated to be below 3 × 10 17 m -3 these voids contributed little to either mechanical strengthening of the material or measured density changes.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dr. Mohit Jain; Dr. Ganesh Skandan; Dr. Gordon E. Khose

Generation IV Very High Temperature power generating nuclear reactors will operate at temperatures greater than 900 oC. At these temperatures, the components operating in these reactors need to be fabricated from materials with excellent thermo-mechanical properties. Conventional pure or composite materials have fallen short in delivering the desired performance. New materials, or conventional materials with new microstructures, and associated processing technologies are needed to meet these materials challenges. Using the concept of functionally graded materials, we have fabricated a composite material which has taken advantages of the mechanical and thermal properties of ceramic and metals. Functionally-graded composite samples with variousmore » microstructures were fabricated. It was demonstrated that the composition and spatial variation in the composition of the composite can be controlled. Some of the samples were tested for irradiation resistance to neutrons. The samples did not degrade during initial neutron irradiation testing.« less

Development of indigenous insulation material for superconducting magnets and study of its characteristics under influence of intense neutron irradiation

NASA Astrophysics Data System (ADS)

Sharma, Rajiv; Tanna, V. L.; Rao, C. V. S.; Abhangi, Mitul; Vala, Sudhirsinh; Sundaravel; Varatharajan, S.; Sivakumar, S.; Sasi, K.; Pradhan, S.

2017-02-01

Epoxy based glass fiber reinforced composites are the main insulation system for the superconducting magnets of fusion machines. 14MeV neutrons are generated during the DT fusion process, however the energy spectra and flux gets modified to a great extent when they reach the superconducting magnets. Mechanical properties of the GFRP insulation material is reported to degrade up to 30%. As a part of R & D activity, a joint collaboration with IGCAR, Kalpakkam has been established. The indigenous insulation material is subjected to fast neutron fluence of 1014 - 1019 n/m2 (E>0.1 MeV) in FBTR and KAMINI Reactor, India. TRIM software has been used to simulate similar kind of damage produced by neutrons by ion irradiation with 5 MeV Al ions and 3 MeV protons. Fluence of the ions was adjusted to get the same dpa. We present the test experiment of neutron irradiation of the composite material (E-glass, S-glass fiber boron free and DGEBA epoxy). The test results of tensile, inter laminar shear and electrical breakdown strength as per ASTM standards, assessment of micro-structure surface degradation before and after irradiation will be presented. MCNP simulations are carried out for neutron flux, dose and damages produced in the insulation material.

Cryogenic electrical properties of irradiated cyanate ester/epoxy insulation for fusion magnets

NASA Astrophysics Data System (ADS)

Li, X.; Wu, Z. X.; Li, J.; Xu, D.; Liu, H. M.; Huang, R. J.; Li, L. F.

2017-12-01

The insulation materials used in high field fusion magnets require excellent mechanical properties, high electrical breakdown strength, good thermal conductivity and high radiation tolerance. Previous investigations showed that cyanate ester/epoxy (CE/EP) insulation material, a candidate insulation for fusion magnets, can maintain good mechanical performance at cryogenic temperature after 10 MGy irradiation and has a much longer pot life than traditional epoxy insulation material. In order to quantify the electrical properties of the CE/EP insulation material at low temperature, a cryogenic electrical property testing system cooled by a G-M cryocooler was developed for this study. An insulation material with 40% cyanate ester and 60% epoxy was subjected to 60Co γ-ray irradiation in air at ambient temperature with a dose rate of 300 Gy/min, and total doses of 1 MGy, 5 MGy and 10 MGy. The electrical breakdown strength of this CE/EP insulation material was measured before and after irradiation. The results show that cryogenic temperature has a positive effect on the electrical breakdown strength of this composite, while the influence of 60Co γ-ray irradiation is not obvious at 6.1 K.

Preparation and characterization of electron-beam treated HDPE composites reinforced with rice husk ash and Brazilian clay

NASA Astrophysics Data System (ADS)

Ortiz, A. V.; Teixeira, J. G.; Gomes, M. G.; Oliveira, R. R.; Díaz, F. R. V.; Moura, E. A. B.

2014-08-01

This work evaluates the morphology, mechanical and thermo-mechanical properties of high density polyethylene (HDPE) composites. HDPE reinforced with rice husk ashes (80:20 wt%), HDPE reinforced with clay (97:3 wt%) and HDPE reinforced with both rice husk ashes and clay(77:20:3 wt%) were obtained. The Brazilian bentonite chocolate clay was used in this study. This Brazilian smectitic clay is commonly used to produce nanocomposites. The composites were produced by melting extrusion process and then irradiation was carried out in a 1.5 MeV electron-beam accelerator (room temperature, presence of air). Comparisons using the irradiated and non-irradiated neat polymer, and the irradiated and non-irradiated composites were made. The materials obtained were submitted to tensile, flexural and impact tests. Additionally HDT, SEM and XRD analyses were carried out along with the sol-gel analysis which aimed to assess the cross-linking degree of the irradiated materials. Results showed great improvement in most HDPE properties and a high cross-linking degree of 85% as a result of electron-beam irradiation of the material.

The tensile and fatigue properties of DIN 1.4914 martensitic stainless steel after 590 MeV proton irradiation

NASA Astrophysics Data System (ADS)

Marmy, P.; Victoria, M.

1992-09-01

Tensile and low cycle fatigue subsize specimens of DIN 1.4914 martensitic steel (MANET) have been irradiated with 590 MeV protons to doses up to 1 dpa and at temperatures between 363 and 703 K. The helium produced by spallation reactions was measured as 130 appm/dpa. A strong radiation hardening is found, which decreases as the irradiation temperature increases. The tensile elongation is reduced after irradiation, but the fracture mode is always ductile and transgranular. The radition hardening produced at low irradiation temperatures is recovered after annealing at higher temperatures. Continous softening is observed during low cycle fatigue testing. The rate of softening of the irradiated material is stonger than that of the unirradiated material and tends to reach the saturation level of the latter. The irradiation badly affects the fatigue life, particularly in the temperature domain of dynamic strain ageing between 553 and 653 K.

Radiation treatment for sterilization of packaging materials

NASA Astrophysics Data System (ADS)

Haji-Saeid, Mohammad; Sampa, Maria Helena O.; Chmielewski, Andrzej G.

2007-08-01

Treatment with gamma and electron radiation is becoming a common process for the sterilization of packages, mostly made of natural or synthetic plastics, used in the aseptic processing of foods and pharmaceuticals. The effect of irradiation on these materials is crucial for packaging engineering to understand the effects of these new treatments. Packaging material may be irradiated either prior to or after filling. The irradiation prior to filling is usually chosen for dairy products, processed food, beverages, pharmaceutical, and medical device industries in the United States, Europe, and Canada. Radiation effects on packaging material properties still need further investigation. This paper summarizes the work done by different groups and discusses recent developments in regulations and testing procedures in the field of packaging technology.

Spallation modeling in the Charring Material Thermal Response and Ablation (CMA) computer program

NASA Astrophysics Data System (ADS)

Sullivan, J. M.; Kobayashi, W. S.

1987-06-01

It has been observed during tests of certain laminated composite materials exposed to relatively high continuous wave laser irradiation, that the heated surface will spall. To model this phenomenon, the Charring Material Thermal Response and Ablation code has been updated. In addition to temperature response, in-depth decomposition, and surface recession, thermal and mechanical stresses are calculated. Spall is modeled as a discrete mass removal event occurring when the stresses exceed the ultimate strength of the char through a critical depth. Comparisons are made with test data for a carbon phenolic cylinder exposed to a shock tube environment and for a flat plate Kevlar epoxy test specimen exposed to high intensity laser irradiation. Good agreement is shown; however, the results indicate a requirement for more comprehensive elevated-temperature material properties for further validation.

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henager, Jr., C. H.; Nguyen, Ba N.; Kurtz, Richard J.

The international fusion community has designed a miniature torsion specimen for neutron irradiation studies of joined SiC and SiC/SiC composite materials. For this research, miniature torsion joints based on this specimen design were fabricated using displacement reactions between Si and TiC to produce Ti 3SiC 2 + SiC joints with SiC and tested in torsion-shear prior to and after neutron irradiation. However, many miniature torsion specimens fail out-of-plane within the SiC specimen body, which makes it problematic to assign a shear strength value to the joints and makes it difficult to compare unirradiated and irradiated strengths to determine irradiation effects.more » Finite element elastic damage and elastic–plastic damage models of miniature torsion joints are developed that indicate shear fracture is more likely to occur within the body of the joined sample and cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. The model results are compared and discussed with regard to unirradiated and irradiated test data for a variety of joint materials. The unirradiated data includes Ti 3SiC 2 + SiC/CVD-SiC joints with tailored joint moduli, and includes steel/epoxy and CVD-SiC/epoxy joints. Finally, the implications for joint data based on this sample design are discussed.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.

The international fusion community has designed a miniature torsion specimen for neutron irradiation studies of joined SiC and SiC/SiC composite materials. Miniature torsion joints based on this specimen design were fabricated using displacement reactions between Si and TiC to produce Ti3SiC2 + SiC joints with CVD-SiC and tested in torsion-shear prior to and after neutron irradiation. However, many of these miniature torsion specimens fail out-of-plane within the CVD-SiC specimen body, which makes it problematic to assign a shear strength value to the joints and makes it difficult to compare unirradiated and irradiated joint strengths to determine the effects of themore » irradiation. Finite element elastic damage and elastic-plastic damage models of miniature torsion joints are developed that indicate shear fracture is likely to occur within the body of the joined sample and cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. The model results are compared and discussed with regard to unirradiated and irradiated joint test data for a variety of joint materials. The unirradiated data includes Ti3SiC2 + SiC/CVD-SiC joints with tailored joint moduli, and includes steel/epoxy and CVD-SiC/epoxy joints. The implications for joint data based on this sample design are discussed.« less

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

T. R. Allen; J. B. Benson; J. A. Foster

2009-05-01

To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities ismore » granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team projects and faculty/staff exchanges. In June of 2008, the first week-long ATR NSUF Summer Session was attended by 68 students, university faculty and industry representatives. The Summer Session featured presentations by 19 technical experts from across the country and covered topics including irradiation damage mechanisms, degradation of reactor materials, LWR and gas reactor fuels, and non-destructive evaluation. High impact research results from leveraging the entire research infrastructure, including universities, industry, small business, and the national laboratories. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. Current partner facilities include the MIT Reactor, the University of Michigan Irradiated Materials Testing Laboratory, the University of Wisconsin Characterization Laboratory, and the University of Nevada, Las Vegas transmission Electron Microscope User Facility. Needs for irradiation of material specimens at tightly controlled temperatures are being met by dedication of a large in-pile pressurized water loop facility for use by ATR NSUF users. Several environmental mechanical testing systems are under construction to determine crack growth rates and fracture toughness on irradiated test systems.« less

Pre-irradiation testing of actively cooled Be-Cu divertor modules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linke, J.; Duwe, R.; Kuehnlein, W.

1995-09-01

A set of neutron irradiation tests is prepared on different plasma facing materials (PFM) candidates and miniaturized components for ITER. Beside beryllium the irradiation program which will be performed in the High Flux Reactor (HFR) in Petten, includes different carbon fiber composites (CFQ) and tungsten alloys. The target values for the neutron irradiation will be 0.5 dpa at temperatures of 350{degrees}C and 700{degrees}C, resp.. The post irradiation examination (PIE) will cover a wide range of mechanical tests; in addition the degradation of thermal conductivity will be investigated. To determine the high heat flux (HHF) performance of actively cooled divertor modules,more » electron beam tests which simulate the expected heat loads during the operation of ITER, are scheduled in the hot cell electron beam facility JUDITH. These tests on a selection of different actively cooled beryllium-copper and CFC-copper divertor modules are performed before and after neutron irradiation; the pre-irradiation testing is an essential part of the program to quantify the zero-fluence high heat flux performance and to detect defects in the modules, in particular in the brazed joints.« less

Microwave-immobilized polybutadiene stationary phase for reversed-phase high-performance liquid chromatography.

PubMed

Lopes, Nilva P; Collins, Kenneth E; Jardim, Isabel C S F

2004-03-19

Polybutadiene (PBD) has been immobilized on high-performance liquid chromatography (HPLC) silica by microwave radiation at various power levels (52-663 W) and actuation times (3-60 min). Columns prepared from these reversed-phase HPLC materials, as well as from similar non-irradiated materials, were tested with standard sample mixtures and characterized by elemental analysis (%C) and infrared spectroscopy. A microwave irradiation of 20 min at 663 W gives a layer of immobilized PBD that presented good performance. Longer irradiation times give thicker immobilized layers having less favorable chromatographic properties.

Space Reflector Materials for Prometheus Application

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Nash; V. Munne; LL Stimely

2006-01-31

The two materials studied in depth which appear to have the most promise in a Prometheus reflector application are beryllium (Be) and beryllium oxide (BeO). Three additional materials, magnesium oxide (MgO), alumina (Al{sub 2}O{sub 3}), and magnesium aluminate spinel (MgAl{sub 2}O{sub 4}) were also recently identified to be of potential interest, and may have promise in a Prometheus application as well, but are expected to be somewhat higher mass than either a Be or BeO based reflector. Literature review and analysis indicates that material properties for Be are largely known, but there are gaps in the properties of Be0 relativemore » to the operating conditions for a Prometheus application. A detailed preconceptual design information document was issued providing material properties for both materials (Reference (a)). Beryllium oxide specimens were planned to be irradiated in the JOY0 Japanese test reactor to partially fill the material property gaps, but more testing in the High Flux Isotope Reactor (HFIR) test reactor at Oak Ridge National Laboratory (ORNL) was expected to be needed. A key issue identified for BeO was obtaining material for irradiation testing with an average grain size of {approx}5 micrometers, reminiscent of material for which prior irradiation test results were promising. Current commercially available material has an average grain size of {approx}10 micrometers. The literature indicated that improved irradiation performance could be expected (e.g., reduced irradiation-induced swelling) with the finer grain size material. Confirmation of these results would allow the use of historic irradiated materials test results from the literature, reducing the extent of required testing and therefore the cost of using this material. Environmental, safety and health (ES&H) concerns associated with manufacturing are significant but manageable for Be and BeO. Although particulate-generating operations (e.g., machining, grinding, etc.) involving Be-bearing materials require significant controls, handling of clean, finished products requires only modest controls. Neither material was initially considered to be viable as a structural material, however, based on improved understanding of its unirradiated properties, Be should be evaluated due to having potentially acceptable structural properties in the unirradiated condition, i. e., during launch, when loads might be most limiting. All three of the alternative materials are non-hazardous, and thus do not engender the ES&H concerns associated with use of Be or BeO. Aluminum oxide is a widely available ceramic material with well characterized physical properties and well developed processing practices. Although the densest (3.97 g/cm{sup 3} versus Be: 1.85, BeO: 3.01, MgO: 3.58, and MgAl{sub 2}O{sub 4}: 3.60, all theoretical density), and therefore the heaviest, of all the materials considered for this application, its ease of fabrication, mechanical properties, availability and neutronic characteristics warrant its evaluation. Similarly, MgO is widely used in the refractory materials industry and has a large established manufacturing base while being lighter than Al{sub 2}O{sub 3}. Most of the commercially available MgO products incorporate additives or a second phase to avoid the formation of Mg(OH){sub 2} due to spontaneous reaction with ambient humidity. The hygroscopicity of MgO makes it a more difficult material to work with than Al{sub 2}O{sub 3} or MgAl{sub 2}O{sub 4}. Magnesium aluminate spinel, although not as widely available as either Al{sub 2}O{sub 3} or MgO, has the advantage of a density almost as low as MgO without being hygroscopic, and shares comparable neutronic performance characteristics in the reflector application.« less

BWRVIP-140NP: BWR Vessel and Internals Project Fracture Toughness and Crack Growth Program on Irradiated Austenitic Stainless Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilman, J

2005-03-15

To prepare for this project, EPRI and BWRVIP conducted a workshop at Ponte Vedra Beach, Florida during February 19-21, 2003 (EPRI report 1007822). Attendees were invited to exchange relevant information on the effects of irradiation on austenitic materials in light water reactors and to produce recommendations for further work. EPRI reviewed the data, recommendations, and conclusions derived from the workshop and developed prioritized test matrices defining new data needs. Proposals were solicited, and selected proposals are the basis for the program described in this report. Results The planned test matrix for fracture toughness testing includes 21 tests on 5 materials.

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

NASA Astrophysics Data System (ADS)

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.

Testing of solar cell covers and encapsulants conducted in a simulated space environment

NASA Technical Reports Server (NTRS)

Russell, D. A.

1981-01-01

The materials included in the evaluation were 0211 micro-sheet, FEP-A used as a cover and as an adhesive, DC 93-500 adhesive, PFA "hard coat" used as a cover, GE 615/UV-24 used as a cover, GR 650 used as a cover, and electrostatically bonded 7070 glass. The test environments were 1 MeV electron irradiation interspersed with thermal cycling, 0.5 MeV proton irradiation interspersed with thermal cycling and UV exposure interspersed with thermal cycling. Summary data is given describing the response of the test materials both visually and electrically to the three different environments.

Fabrication of recyclable and durable superhydrophobic materials with wear/corrosion-resistance properties from kaolin and polyvinylchloride

NASA Astrophysics Data System (ADS)

Qu, Mengnan; Liu, Shanshan; He, Jinmei; Feng, Juan; Yao, Yali; Ma, Xuerui; Hou, Lingang; Liu, Xiangrong

2017-07-01

In this study, mechanically stable and recyclable superhydrophobic materials were prepared from polyvinylchloride (PVC) and kaolin nanoparticles modified by stearic acid using a simple and low-cost drop-coating. The obtained materials displayed liquid-repellent toward water and several other liquids of daily life (such as orange juice, coffee, milk, coca cola and ink). These superhydrophobic materials showed remarkable robustness against sandpaper abrasion, UV-irradiation and ultrasonication test, while retaining its superhydrophobicity even after 60 abrasion cycles loaded of 500 g with sandpaper, 7 days UV-irradiation or 120 min ultrasonication test. The excellent durability against complex conditions was attributed to the hierarchical structure and strong interfacial adhesion of the materials. More significantly, the materials used in the coating could be recycled and reconstructed without losing its superhydrophobicity. The current superhydrophobic materials tolerate rigorous environment, opening a new avenue to a variety of practical applications.

Determination of contamination character of materials in space technology testing

NASA Technical Reports Server (NTRS)

Haynes, D. L.; Coulson, D. M.

1972-01-01

The contamination character of selected materials used in space technology testing is presented. Many of these materials contain components that become volatile in a space environment. Most previous data were limited to weight loss or vapor pressure. However, these parameters are not necessarily a direct measure of the contamination character of these materials. Selected materials were exposed to a thermal-vacuum environment, and the degree of contamination was measured by collecting the outgases from these materials on a cold test mirror surface. The degradation of reflectivity of the mirror was measured over a spectral range from 1100 A to 2.5 microns. Half the mirror's surface was also exposed to UV irradiation to determine its effects on the contaminative character of the depositing outgases. The amount of deposit per unit area was measured by microbalances mounted near the mirror; the sensor of one microbalance was UV irradiated. A quadrupole mass spectrometer was used to determine the composition of the outgases.

Experimental investigations on thermo mechanical behaviour of aluminium alloys subjected to tensile loading and laser irradiation

NASA Astrophysics Data System (ADS)

Jelani, Mohsan; Li, Zewen; Shen, Zhonghua; Sardar, Maryam; Tabassum, Aasma

2017-05-01

The present work reports the investigation of the thermal and mechanical behaviour of aluminium alloys under the combined action of tensile loading and laser irradiations. The two types of aluminium alloys (Al-1060 and Al-6061) are used for the experiments. The continuous wave Ytterbium fibre laser (wavelength 1080 nm) was employed as irradiation source, while tensile loading was provided by tensile testing machine. The effects of various pre-loading and laser power densities on the failure time, temperature distribution and on deformation behaviour of aluminium alloys are analysed. The experimental results represents the significant reduction in failure time and temperature for higher laser powers and for high load values, which implies that preloading may contribute a significant role in the failure of the material at elevated temperature. The reason and characterization of material failure by tensile and laser loading are explored in detail. A comparative behaviour of under tested materials is also investigated. This work suggests that, studies considering only combined loading are not enough to fully understand the mechanical behaviour of under tested materials. For complete characterization, one must consider the effect of heating as well as loading rate.

Discuss the testing problems of ultraviolet irradiance meters

NASA Astrophysics Data System (ADS)

Ye, Jun'an; Lin, Fangsheng

2014-09-01

Ultraviolet irradiance meters are widely used in many areas such as medical treatment, epidemic prevention, energy conservation and environment protection, computers, manufacture, electronics, ageing of material and photo-electric effect, for testing ultraviolet irradiance intensity. So the accuracy of value directly affects the sterile control in hospital, treatment, the prevention level of CDC and the control accuracy of curing and aging in manufacturing industry etc. Because the display of ultraviolet irradiance meters is easy to change, in order to ensure the accuracy, it needs to be recalibrated after being used period of time. By the comparison with the standard ultraviolet irradiance meters, which are traceable to national benchmarks, we can acquire the correction factor to ensure that the instruments working under accurate status and giving the accurate measured data. This leads to an important question: what kind of testing device is more accurate and reliable? This article introduces the testing method and problems of the current testing device for ultraviolet irradiance meters. In order to solve these problems, we have developed a new three-dimensional automatic testing device. We introduce structure and working principle of this system and compare the advantages and disadvantages of two devices. In addition, we analyses the errors in the testing of ultraviolet irradiance meters.

Spallation Neutron Source Materials Studies

NASA Astrophysics Data System (ADS)

Sommer, W. F.

1998-04-01

Operation of accelerator facilities such as Los Alamos Neutron Science Center (LANSCE), ISIS at Rutherford Appleton Laboratory, the Swiss Institute Neutron Source (SINQ) at Paul Scherrer Institute, and others has provided valuable information on materials performance in high energy particle beams and high energy neutron environments. The Accelerator Production of Tritium (APT) project is sponsoring an extensive series of tests on the effect of spallation neutron source environments to physical and mechanical properties of candidate materials such as nickel-based alloys, stainless steel alloys, aluminum alloys and solid target materials such as tungsten. Measurements of corrosion rates of these candidate materials during irradiation and while in contact with flowing coolant water are being made. The APT tests use the irradiation facility in the beam stop area of the LANSCE accelerator using 800 MeV protons as well as the neutron flux-spectrum generated as these protons interact with targets. The initial irradiations were completed in summer 1997, exposing materials to a fluence approaching 4-6 x 10^21 protons/cm^2. Sample retrieval is now underway. Mechanical properties measurements are being conducted at several laboratories. Studies on components used in service have also been initiated.

Severe Accident Test Station Design Document

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snead, Mary A.; Yan, Yong; Howell, Michael

The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure tomore » provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.« less

NSUF Irradiated Materials Library

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cole, James Irvin

The Nuclear Science User Facilities has been in the process of establishing an innovative Irradiated Materials Library concept for maximizing the value of previous and on-going materials and nuclear fuels irradiation test campaigns, including utilization of real-world components retrieved from current and decommissioned reactors. When the ATR national scientific user facility was established in 2007 one of the goals of the program was to establish a library of irradiated samples for users to access and conduct research through competitively reviewed proposal process. As part of the initial effort, staff at the user facility identified legacy materials from previous programs thatmore » are still being stored in laboratories and hot-cell facilities at the INL. In addition other materials of interest were identified that are being stored outside the INL that the current owners have volunteered to enter into the library. Finally, over the course of the last several years, the ATR NSUF has irradiated more than 3500 specimens as part of NSUF competitively awarded research projects. The Logistics of managing this large inventory of highly radioactive poses unique challenges. This document will describe materials in the library, outline the policy for accessing these materials and put forth a strategy for making new additions to the library as well as establishing guidelines for minimum pedigree needed to be included in the library to limit the amount of material stored indefinitely without identified value.« less

Ion irradiation testing and characterization of FeCrAl candidate alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew

2014-10-29

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commerciallymore » available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.« less

Low Activation Joining of SiC/SiC Composites for Fusion Applications: Modeling Miniature Torsion Tests with Elastic and Elastic-Plastic Models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.

2015-03-01

The use of SiC and SiC-composites in fission or fusion environments requires joining methods for assembling systems. The international fusion community designed miniature torsion specimens for joint testing and irradiation in test reactors with limited irradiation volumes. These torsion specimens fail out-of-plane when joints are strong and when elastic moduli are within a certain range compared to SiC, which causes difficulties in determining shear strengths for joints or for comparing unirradiated and irradiated joints. A finite element damage model was developed that indicates fracture is likely to occur within the joined pieces to cause out-of-plane failures for miniature torsion specimensmore » when a certain modulus and strength ratio between the joint material and the joined material exists. The model was extended to treat elastic-plastic joints such as SiC/epoxy and steel/epoxy joints tested as validation of the specimen design.« less

Laser-accelerated particle beams for stress testing of materials.

PubMed

Barberio, M; Scisciò, M; Vallières, S; Cardelli, F; Chen, S N; Famulari, G; Gangolf, T; Revet, G; Schiavi, A; Senzacqua, M; Antici, P

2018-01-25

Laser-driven particle acceleration, obtained by irradiation of a solid target using an ultra-intense (I > 10 18  W/cm 2 ) short-pulse (duration <1 ps) laser, is a growing field of interest, in particular for its manifold potential applications in different domains. Here, we provide experimental evidence that laser-generated particles, in particular protons, can be used for stress testing materials and are particularly suited for identifying materials to be used in harsh conditions. We show that these laser-generated protons can produce, in a very short time scale, a strong mechanical and thermal damage, that, given the short irradiation time, does not allow for recovery of the material. We confirm this by analyzing changes in the mechanical, optical, electrical, and morphological properties of five materials of interest to be used in harsh conditions.

Ultrasonic Transducer Irradiation Test Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daw, Joshua; Palmer, Joe; Ramuhalli, Pradeep

2015-02-01

Ultrasonic technologies offer the potential for high-accuracy and -resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other ongoing efforts include an ultrasonic technique to detect morphology changesmore » (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. For this reason, the Pennsylvania State University (PSU) was awarded an ATR NSUF project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 10 21 n/cm 2. The goal of this research is to characterize and demonstrate magnetostrictive and piezoelectric transducer operation during irradiation, enabling the development of novel radiation-tolerant ultrasonic sensors for use in Material Testing Reactors (MTRs). As such, this test is an instrumented lead test and real-time transducer performance data is collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers. To date, one piezoelectric transducer and two magnetostrictive transducers have demonstrated reliable operation under irradiation. The irradiation is ongoing.« less

Microhardness evaluations of CAD/CAM ceramics irradiated with CO2 or Nd:YAP laser

PubMed Central

Rocca, Jean Paul; Fornaini, Carlo; Medioni, Etienne; Brulat-Bouchard, Nathalie

2017-01-01

Background and aims The aim of this study was to measure the microhardness values of irradiated computer-aided design/computer-aided manufacturing (CAD/CAM) ceramics surfaces before and after thermal treatment. Materials and Methods Sixty CAD/CAM ceramic discs were prepared and grouped by material, i.e. lithium disilicate ceramic (Emax CAD) and zirconia ceramic (Emax ZirCAD). Laser irradiation at the material surface was performed with a carbon dioxide laser at 5 Watt (W) or 10 W power in continuous mode (CW mode), or with a neodymium:yttrium aluminum perovskite (Nd:YAP) laser at 10 W on graphite and non-graphite surfaces. Vickers hardness was tested at 0.3 kgf for lithium disilicate and 1 kgf for zirconia. Results Emax CAD irradiated with CO2 at 5 W increased microhardness by 6.32 GPa whereas Emax ZirCAD irradiated with Nd:YAP decreased microhardness by 17.46 GPa. Conclusion CO2 laser effectively increases the microhardness of lithium disilicate ceramics (Emax CAD). PMID:28740324

Impact properties of irradiated HT9 from the fuel duct of FFTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byun, Thak Sang; Lewis, W. Daniel; Toloczko, Mychailo B.

2012-02-01

This paper reports Charpy impact test data for the ACO-3 duct material (HT9) from the Fast Flux Test Facility (FFTF) and its archive material. Irradiation doses for the specimens were in the range of 3– 148 dpa and irradiation temperatures in the range of 378–504 *C. The impact tests were performed for the small V-notched Charpy specimens with dimensions of 3 * 4 * 27 mm at an impact speed of 3.2 m/s in a 25 J capacity machine. Irradiation lowered the upper-shelf energy (USE) and increased the transition temperatures significantly. The shift of ductile–brittle transition temperatures (DDBTT) was greatermore » after relatively low temperature irradiation. The USE values were in the range of 5.5–6.7 J before irradiation and decreased to the range of 2–5 J after irradiation. Lower USEs were measured for lower irradiation temperatures and specimens with T-L orientation. The dose dependences of transition temperature and USE were not significant because of the radiation effect on impact behavior nearly saturated at the lowest dose of about 3 dpa. A comparison showed that the lateral expansion of specimens showed a linear correlation with absorbed impact energy, but with large scatter in the results. Size effect was also discussed to clarify the differences in the impact property data from subsize and standard specimens as well as to provide a basis for comparison of data from different specimens. The USE and DDBTT data from different studies were compared.« less

Impact properties of irradiated HT9 from the fuel duct of FFTF

NASA Astrophysics Data System (ADS)

Byun, Thak Sang; Daniel Lewis, W.; Toloczko, Mychailo B.; Maloy, Stuart A.

2012-02-01

This paper reports Charpy impact test data for the ACO-3 duct material (HT9) from the Fast Flux Test Facility (FFTF) and its archive material. Irradiation doses for the specimens were in the range of 3-148 dpa and irradiation temperatures in the range of 378-504 °C. The impact tests were performed for the small V-notched Charpy specimens with dimensions of 3 × 4 × 27 mm at an impact speed of 3.2 m/s in a 25 J capacity machine. Irradiation lowered the upper-shelf energy (USE) and increased the transition temperatures significantly. The shift of ductile-brittle transition temperatures (ΔDBTT) was greater after relatively low temperature irradiation. The USE values were in the range of 5.5-6.7 J before irradiation and decreased to the range of 2-5 J after irradiation. Lower USEs were measured for lower irradiation temperatures and specimens with T-L orientation. The dose dependences of transition temperature and USE were not significant because of the radiation effect on impact behavior nearly saturated at the lowest dose of about 3 dpa. A comparison showed that the lateral expansion of specimens showed a linear correlation with absorbed impact energy, but with large scatter in the results. Size effect was also discussed to clarify the differences in the impact property data from subsize and standard specimens as well as to provide a basis for comparison of data from different specimens. The USE and ΔDBTT data from different studies were compared.

Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. H. Jackson; S. P. Teysseyre

2012-10-01

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials ofmore » interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.« less

Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. H. Jackson; S. P. Teysseyre

2012-02-01

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials ofmore » interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.« less

Optical and color stabilities of paint-on resins for shade modification of restorative resins.

PubMed

Arikawa, Hiroyuki; Kanie, Takahito; Fujii, Koichi; Ban, Seiji; Homma, Tetsuya; Takahashi, Hideo

2004-06-01

The purpose of this study was to examine the optical and color stabilities of the paint-on resin used for shade modification of restorative resins. Three shades of paint-on resin and two crown and bridge resins were used. The light transmittance characteristics of the materials during accelerated aging tests such as water immersion, toothbrush abrasion, ultraviolet (UV) light irradiation, and staining tests were measured. Discolorations of materials resulting from tests were also determined. There were no significant effects of water immersion, toothbrush abrasion and UV light irradiation on the light transmittance and visible color change of paint-on resins, whereas the staining tests significantly decreased the light transmittance and increased color change of the translucent shades of materials. Our results indicate that the paint-on resins exhibit stable optical properties and color appearance, which are at least as good as the crown and bridge resins.

Critical processes and parameters in the development of accident tolerant fuels drop-in capsule irradiation tests

DOE PAGES

Barrett, K. E.; Ellis, K. D.; Glass, C. R.; ...

2015-12-01

The goal of the Accident Tolerant Fuel (ATF) program is to develop the next generation of Light Water Reactor (LWR) fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. An irradiation test series has been defined to assess the performance of proposed ATF concepts under normal LWR operating conditions. The Phase I ATF irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL). Design, analysis, and fabrication processes formore » ATR drop-in capsule experiment preparation are presented in this paper to demonstrate the importance of special design considerations, parameter sensitivity analysis, and precise fabrication and inspection techniques for figure innovative materials used in ATF experiment assemblies. A Taylor Series Method sensitivity analysis approach was used to identify the most critical variables in cladding and rodlet stress, temperature, and pressure calculations for design analyses. The results showed that internal rodlet pressure calculations are most sensitive to the fission gas release rate uncertainty while temperature calculations are most sensitive to cladding I.D. and O.D. dimensional uncertainty. The analysis showed that stress calculations are most sensitive to rodlet internal pressure uncertainties, however the results also indicated that the inside radius, outside radius, and internal pressure were all magnified as they propagate through the stress equation. This study demonstrates the importance for ATF concept development teams to provide the fabricators as much information as possible about the material properties and behavior observed in prototype testing, mock-up fabrication and assembly, and chemical and mechanical testing of the materials that may have been performed in the concept development phase. Special handling, machining, welding, and inspection of materials, if known, should also be communicated to the experiment fabrication and inspection team.« less

Heavy-section steel technology program. Semiannual progress report for period ending February 28, 1973

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1974-02-01

The materials investigations under the HSST program are divided into studies of unirradiated materials and studies of irradiation effects. The studies of unirradiated materials, which include inspection, characterization, metallurgy, variability determinations, transition temperature investigations, fracture mechanics studies, and fatigue-crack propagation tests, are discussed. The investigations of irradiated materials include studies of radiation effects on A-533-B steel. Results of studies on thick pressure vessels and pipes of ASTM A508 steel are also reported along with results of studies on Mode III crack extension in reactor piping. (JRD)

Nuclear Radiation Tolerance of Single Crystal Aluminum Nitride Ultrasonic Transducer

NASA Astrophysics Data System (ADS)

Reinhard, Brian; Tittmann, Bernhard R.; Suprock, Andrew

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models, (Rempe et al., 2011; Kazys et al., 2005). These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2. The irradiation is also supported by a multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET ASI) program. The results from this irradiation, which started in February 2014, offer the potential to enable the development of novel radiation tolerant ultrasonic sensors for use in Material Testing Reactors (MTRs). As such, this test is an instrumented lead test and real-time transducer performance data is collected along with temperature and neutron and gamma flux data. Hence, results from this irradiation offer the potential to bridge the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers. To date, very encouraging results have been attained as several transducers have continued to operate under irradiation. The irradiation is ongoing and will continue to approximately mid-2015.

Development of COPVS for High pressure, In-Space, Cryogenic Fuel Storage

NASA Technical Reports Server (NTRS)

DeLay, Tom; Schneider, Judy; Dyess, Mark; Hastings, Chad; Noorda, Ryan; Noorda, Jared; Patterson, James

2008-01-01

Polymeric composite overwrapped pressure vessels (COPVs) provide an attractive material system to support developing commercial launch business and alternate fuel ventures. However to be able to design with these materials, the mechanical behavior of the materials must be understood with regards to processing, performance, damage tolerance, and environment. For the storage of cryogenic propellants, it is important to evaluate the materials performance and impact damage resistance at cryogenic temperatures in order to minimize weight and to ensure safety and reliability. To evaluate the ultimate performance, various polymeric COPV's have been statically burst tested at cryogenic conditions before and after exposure to irradiation. Materials selected for these COPVs were based on the measured mechanical properties of candidate resin systems and fibers that were also tested at cryogenic conditions before and after exposure to irradiation. The correlation of COPV burst pressures with the constituent material properties has proven to be a valuable screening method for selection of suitable candidate materials with resistance to material degradation due to exposure to temperature and radiation.

NEET In-Pile Ultrasonic Sensor Enablement-FY 2012 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

JE Daw; JL Rempe; BR Tittmann

2012-09-01

Several Department Of Energy-Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development, Advanced Reactor Concepts, Light Water Reactor Sustainability, and Next Generation Nuclear Plant programs, are investigating new fuels and materials for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials when irradiated. The Nuclear Energy Enabling Technology (NEET) Advanced Sensors and Instrumentation (ASI) in-pile instrumentation development activities are focused upon addressing cross-cutting needs for DOE-NE irradiation testing by providing higher fidelity, real-time data, with increased accuracy and resolution from smaller, compact sensors that are lessmore » intrusive. Ultrasonic technologies offer the potential to measure a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes, under harsh irradiation test conditions. There are two primary issues associated with in-pile deployment of ultrasonic sensors. The first is transducer survivability. The ability of ultrasonic transducer materials to maintain their useful properties during an irradiation must be demonstrated. The second issue is signal processing. Ultrasonic testing is typically performed in a lab or field environment, where the sensor and sample are accessible. Due to the harsh nature of in-pile testing, and the range of measurements that are desired, an enhanced signal processing capability is needed to make in-pile ultrasonic sensors viable. This project addresses these technology deployment issues.« less

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.« less

Comparison of irradiation behaviour of HTR graphite grades

NASA Astrophysics Data System (ADS)

Heijna, M. C. R.; de Groot, S.; Vreeling, J. A.

2017-08-01

The INNOGRAPH irradiations were executed in the High Flux Reactor (HFR) in Petten by NRG supported by the European Framework programs HTR-M, RAPHAEL, and ARCHER to generate data on the irradiation behaviour of graphite grades for High Temperature Reactor (HTR) application available at that time. Samples of the graphite grades NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, and IG-110 have been irradiated at 750 °C and 950 °C. The inherent scatter induced by the probabilistic material behaviour of graphite requires uncertainty and scatter induced by test conditions and post-irradiation examination to be minimized. The INNOGRAPH irradiations supplied an adequate number of irradiated samples to enable accurate determination of material properties and their evolution under irradiation. This allows comparison of different graphite grades and a qualitative assessment of their appropriateness for HTR applications, as a basis of selection, design and core component lifetime. The results indicate that coarse grained graphite grades exhibit more favourable behaviour for application in HTRs due to their low dimensional anisotropy and fracture propagation resilience.

A polycrystal plasticity model of strain localization in irradiated iron

NASA Astrophysics Data System (ADS)

Barton, Nathan R.; Arsenlis, Athanasios; Marian, Jaime

2013-02-01

At low to intermediate homologous temperatures, the degradation of structural materials performance in nuclear environments is associated with high number densities of nanometric defects produced in irradiation cascades. In polycrystalline ferritic materials, self-interstitial dislocations loops are a principal signature of irradiation damage, leading to a mechanical response characterized by increased yield strengths, decreased total strain to failure, and decreased work hardening as compared to the unirradiated behavior. Above a critical defect concentration, the material deforms by plastic flow localization, giving rise to strain softening in terms of the engineering stress-strain response. Flow localization manifests itself in the form of defect-depleted crystallographic channels, through which all dislocation activity is concentrated. In this paper, we describe the formulation of a crystal plasticity model for pure Fe embedded in a finite element polycrystal simulator and present results of uniaxial tensile deformation tests up to 10% strain. We use a tensorial damage descriptor variable to capture the evolution of the irradiation damage loop subpopulation during deformation. The model is parameterized with detailed dislocation dynamics simulations of tensile tests up to 1.5% deformation of systems containing various initial densities of irradiation defects. The coarse-grained simulations are shown to capture the essential details of the experimental stress response observed in ferritic alloys and steels. Our methodology provides an effective linkage between the defect scale, of the order of one nanometer, and the continuum scale involving multiple grain orientations.

Effects of gamma ray and electron beam irradiation on the mechanical, thermal, structural and physicochemical properties of poly (ether-block-amide) thermoplastic elastomers.

PubMed

Murray, Kieran A; Kennedy, James E; McEvoy, Brian; Vrain, Olivier; Ryan, Damien; Cowman, Richard; Higginbotham, Clement L

2013-01-01

Both gamma ray and electron beam irradiation are widely used as a means of medical device sterilisation. However, it is known that the radiation produced by both processes can lead to undesirable changes within biomedical polymers. The main objective of this research was to conduct a comparative study on the two key radiosterilisation methods (gamma ray and electron beam) in order to identify the more detrimental process in terms of the mechanical, structural, chemical and thermal properties of a common biomedical grade polymer. Poly (ether-block-amide) (PEBA) was prepared by injection moulding ASTM testing specimens and these were exposed to an extensive range of irradiation doses (5-200 kGy) in an air atmosphere. The effect of varying the irradiation dose concentration on the resultant PEBA properties was apparent. For instance, the tensile strength, percentage elongation at break and shore D hardness can be increased/decreased by controlling the aforementioned criteria. In addition, it was observed that the stiffness of the material increased with incremental irradiation doses as anticipated. Melt flow index demonstrated a dramatic increase in the melting strength of the material indicating a sharp increase in molecular weight. Conversely, modulated differential scanning calorimetry established that there were no significant alterations to the thermal transitions. Noteworthy trends were observed for the dynamic frequency sweeps of the material, where the crosslink density increased according to an increase in electron beam irradiation dose. Trans-vinylene unsaturations and the carbonyl group concentration increased with an increment in irradiation dose for both processes when observed by FTIR. The relationship between the irradiation dose rate, mechanical properties and the subsequent surface properties of PEBA material is further elucidated throughout this paper. This study revealed that the gamma irradiation process produced more adverse effects in the PEBA material in contrast to the electron beam irradiation process. Copyright © 2012 Elsevier Ltd. All rights reserved.

NEET In-Pile Ultrasonic Sensor Enablement-Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Daw; J. Rempe; J. Palmer

2014-09-01

Ultrasonic technologies offer the potential to measure a range of parameters during irradiation of fuels and materials, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes under harsh irradiation test conditions. There are two primary issues that currently limit in-pile deployment of ultrasonic sensors. The first is transducer survivability. The ability of ultrasonic transducer materials to maintain their useful properties during an irradiation must be demonstrated. The second issue is signal processing. Ultrasonic testing is typically performed in a lab or field environment, where the sensor and sample are accessible. The harsh nature ofmore » in-pile testing and the variety of desired measurements demand that an enhanced signal processing capability be developed to make in-pile ultrasonic sensors viable. To address these issues, the NEET ASI program funded a three year Ultrasonic Transducer Irradiation and Signal Processing Enhancements project, which is a collaborative effort between the Idaho National Laboratory, the Pacific Northwest National Laboratory, the Argonne National Laboratory, and the Pennsylvania State University. The objective of this report is to document the objectives and accomplishments from this three year project. As summarized within this document, significant work has been accomplished during this three year project.« less

In situ measurements of a homogeneous to heterogeneous transition in the plastic response of ion-irradiated <111> Ni microspecimens

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Xinyu; Strickland, Daniel J.; Derlet, Peter M.

We report on the use of quantitative in situ microcompression experiments in a scanning electron microscope to systematically investigate the effect of self-ion irradiation damage on the full plastic response of <111> Ni. In addition to the well-known irradiationinduced increases in the yield and flow strengths with increasing dose, we measure substantial changes in plastic flow intermittency behavior, manifested as stress drops accompanying energy releases as the driven material transits critical states. At low irradiation doses, the magnitude of stress drops reduces relative to the unirradiated material and plastic slip proceeds on multiple slip systems, leading to quasi-homogeneous plastic flow.more » In contrast, highly irradiated specimens exhibit pronounced shear localization on parallel slip planes, which we ascribe to the onset of defect free channels normally seen in bulk irradiated materials. Our in situ testing system and approach allows for a quantitative study of the energy release and dynamics associated with defect free channel formation and subsequent localization. As a result, this study provides fundamental insight to the nature of interactions between mobile dislocations and irradiation-mediated and damage-dependent defect structures.« less

In situ measurements of a homogeneous to heterogeneous transition in the plastic response of ion-irradiated <111> Ni microspecimens

DOE PAGES

Zhao, Xinyu; Strickland, Daniel J.; Derlet, Peter M.; ...

2015-02-11

We report on the use of quantitative in situ microcompression experiments in a scanning electron microscope to systematically investigate the effect of self-ion irradiation damage on the full plastic response of <111> Ni. In addition to the well-known irradiationinduced increases in the yield and flow strengths with increasing dose, we measure substantial changes in plastic flow intermittency behavior, manifested as stress drops accompanying energy releases as the driven material transits critical states. At low irradiation doses, the magnitude of stress drops reduces relative to the unirradiated material and plastic slip proceeds on multiple slip systems, leading to quasi-homogeneous plastic flow.more » In contrast, highly irradiated specimens exhibit pronounced shear localization on parallel slip planes, which we ascribe to the onset of defect free channels normally seen in bulk irradiated materials. Our in situ testing system and approach allows for a quantitative study of the energy release and dynamics associated with defect free channel formation and subsequent localization. As a result, this study provides fundamental insight to the nature of interactions between mobile dislocations and irradiation-mediated and damage-dependent defect structures.« less

Effect of Gamma Ray Irradiation on Interlaminar Shear Strength of Glass Fiber Reinforced Plastics at 77 K

NASA Astrophysics Data System (ADS)

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

2008-03-01

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.

International Fusion Materials Irradiation Facility injector acceptance tests at CEA/Saclay: 140 mA/100 keV deuteron beam characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gobin, R., E-mail: rjgobin@cea.fr; Bogard, D.; Chauvin, N.

In the framework of the ITER broader approach, the International Fusion Materials Irradiation Facility (IFMIF) deuteron accelerator (2 × 125 mA at 40 MeV) is an irradiation tool dedicated to high neutron flux production for future nuclear plant material studies. During the validation phase, the Linear IFMIF Prototype Accelerator (LIPAc) machine will be tested on the Rokkasho site in Japan. This demonstrator aims to produce 125 mA/9 MeV deuteron beam. Involved in the LIPAc project for several years, specialists from CEA/Saclay designed the injector based on a SILHI type ECR source operating at 2.45 GHz and a 2 solenoid lowmore » energy beam line to produce such high intensity beam. The whole injector, equipped with its dedicated diagnostics, has been then installed and tested on the Saclay site. Before shipment from Europe to Japan, acceptance tests have been performed in November 2012 with 100 keV deuteron beam and intensity as high as 140 mA in continuous and pulsed mode. In this paper, the emittance measurements done for different duty cycles and different beam intensities will be presented as well as beam species fraction analysis. Then the reinstallation in Japan and commissioning plan on site will be reported.« less

Progress towards developing neutron tolerant magnetostrictive and piezoelectric transducers

NASA Astrophysics Data System (ADS)

Reinhardt, Brian; Tittmann, Bernhard; Rempe, Joy; Daw, Joshua; Kohse, Gordon; Carpenter, David; Ames, Michael; Ostrovsky, Yakov; Ramuhalli, Pradeep; Montgomery, Robert; Chien, Hualte; Wernsman, Bernard

2015-03-01

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is not feasible to obtain high resolution/microscale data with the current state of instrumentation technology. However, ultrasound-based sensors offer the ability to obtain such data if it is demonstrated that these sensors and their associated transducers are resistant to high neutron flux, high gamma radiation, and high temperature. To address this need, the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) is funding an irradiation, led by PSU, at the Massachusetts Institute of Technology Research Reactor to test the survivability of ultrasound transducers. As part of this effort, PSU and collaborators have designed, fabricated, and provided piezoelectric and magnetostrictive transducers that are optimized to perform in harsh, high flux, environments. Four piezoelectric transducers were fabricated with either aluminum nitride, zinc oxide, or bismuth titanate as the active element that were coupled to either Kovar or aluminum waveguides and two magnetostrictive transducers were fabricated with Remendur or Galfenol as the active elements. Pulse-echo ultrasonic measurements of these transducers are made in-situ. This paper will present an overview of the test design including selection criteria for candidate materials and optimization of test assembly parameters, data obtained from both out-of-pile and in-pile testing at elevated temperatures, and an assessment based on initial data of the expected performance of ultrasonic devices in irradiation conditions.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, James J.; Grandy, Christopher

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

Space environmental effects on graphite-epoxy compressive properties and epoxy tensile properties

NASA Technical Reports Server (NTRS)

Fox, Derek J.; Sykes, George F., Jr.; Herakovich, Carl T.

1987-01-01

This study characterizes the effects of electron radiation and temperature on a graphite-epoxy composite material. Compressive properties of the T300/934 material system were obtained at -250 F (-157 C), room temperature, and 250 F (121 C). Tensile specimens of the Fiberite 934 epoxy resin were fabricated and tested at room temperature and 250 F (121 C). Testing was conducted in the baseline (nonirradiated) and irradiated conditions. The radiation exposure was designed to simulate 30 year, worst-case exposure in geosynchronous Earth orbit. Mechanical properties tended to degrade at elevated temperature and improve at cryogenic temperature. Irradiation generally degraded properties at all temperatures.

The effects of gamma-ray irradiation on organic materials of different conjugation lengths

NASA Astrophysics Data System (ADS)

Zhang, Cheng; Taylor, Edward W.

2009-08-01

The radiation resistance of organic electro-optic and optoelectronic materials of different conjugation lengths for space applications is receiving increased attention. Earlier investigation reported that guest-host (G-H) poled polymer EO modulator devices composed of a phenyltetraene bridge-type chromophore in amorphous polycarbonate (CLD/APC) did not exhibit a decrease in EO response (i.e., an increase in modulation-switching voltage- Vπ) following irradiation by low dose [10-160 krad(Si)] 60Co gamma-rays. In this work, the post-irradiation responses of 60Co gamma-rays on CLD1/APC thin films are examined by various chemical and spectroscopic methods including: a solubility test, thin-layer chromatography, proton nuclear magnetic resonance spectroscopy, UV-vis absorption, and infra-red absorption. The results indicate that CLD1 and APC did not decompose under gamma-ray irradiation at dose levels ranging from 66-274 krad(Si) and from 61-154 krad(Si), respectively which support the previously reported data. A comparison with an in situ proton irradiated DRI/PMMA material is also presented.

The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Jiang, Hao

2016-01-12

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (ORNL) and is described fully in US Patent Application 20060070455, “Expanded plug method for developing circumferential mechanical properties of tubular materials.” This method is designed for testing fuel rod cladding ductility in a hot cell using an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are themore » simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.« less

Microhardness of dual-polymerizing resin cements and foundation composite resins for luting fiber-reinforced posts.

PubMed

Yoshida, Keiichi; Meng, Xiangfeng

2014-06-01

The optimal luting material for fiber-reinforced posts to ensure the longevity of foundation restorations remains undetermined. The purpose of this study was to evaluate the suitability of 3 dual-polymerizing resin cements and 2 dual-polymerizing foundation composite resins for luting fiber-reinforced posts by assessing their Knoop hardness number. Five specimens of dual-polymerizing resin cements (SA Cement Automix, G-Cem LincAce, and Panavia F2.0) and 5 specimens of dual-polymerizing foundation composite resins (Clearfil DC Core Plus and Unifil Core EM) were polymerized from the top by irradiation for 40 seconds. Knoop hardness numbers were measured at depths of 0.5, 2.0, 4.0, 6.0, 8.0, and 10.0 mm at 0.5 hours and 7 days after irradiation. Data were statistically analyzed by repeated measures ANOVA, 1-way ANOVA, and the Tukey compromise post hoc test (α=.05). At both times after irradiation, the 5 resins materials showed the highest Knoop hardness numbers at the 0.5-mm depth. At 7 days after irradiation, the Knoop hardness numbers of the resin materials did not differ significantly between the 8.0-mm and 10.0-mm depths (P>.05). For all materials, the Knoop hardness numbers at 7 days after irradiation were significantly higher than those at 0.5 hours after irradiation at all depths (P<.05). At 7 days after irradiation, the Knoop hardness numbers of the 5 resin materials were found to decrease in the following order: DC Core Plus, Unifil Core EM, Panavia F2.0, SA Cement Automix, and G-Cem LincAce (P<.05). The Knoop hardness number depends on the depth of the cavity, the length of time after irradiation, and the material brand. Although the Knoop hardness numbers of the 2 dual-polymerizing foundation composite resins were higher than those of the 3 dual-polymerizing resin cements, notable differences were seen among the 5 materials at all depths and at both times after irradiation. Copyright © 2014 Editorial Council for the Journal of Prosthetic Dentistry. Published by Elsevier Inc. All rights reserved.

Comparison between the Strength Levels of Baseline Nuclear-Grade Graphite and Graphite Irradiated in AGC-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carroll, Mark Christopher

2015-07-01

This report details the initial comparison of mechanical strength properties between the cylindrical nuclear-grade graphite specimens irradiated in the second Advanced Graphite Creep (AGC-2) experiment with the established baseline, or unirradiated, mechanical properties compiled in the Baseline Graphite Characterization program. The overall comparative analysis will describe the development of an appropriate test protocol for irradiated specimens, the execution of the mechanical tests on the AGC-2 sample population, and will further discuss the data in terms of developing an accurate irradiated property distribution in the limited amount of irradiated data by leveraging the considerably larger property datasets being captured in themore » Baseline Graphite Characterization program. Integrating information on the inherent variability in nuclear-grade graphite with more complete datasets is one of the goals of the VHTR Graphite Materials program. Between “sister” specimens, or specimens with the same geometry machined from the same sub-block of graphite from which the irradiated AGC specimens were extracted, and the Baseline datasets, a comprehensive body of data will exist that can provide both a direct and indirect indication of the full irradiated property distributions that can be expected of irradiated nuclear-grade graphite while in service in a VHTR system. While the most critical data will remain the actual irradiated property measurements, expansion of this data into accurate distributions based on the inherent variability in graphite properties will be a crucial step in qualifying graphite for nuclear use as a structural material in a VHTR environment.« less

Investigation of temperature dependence of fracture toughness in high-dose HT9 steel using small-specimen reuse technique

NASA Astrophysics Data System (ADS)

Baek, Jong-Hyuk; Byun, Thak Sang; Maloy, Start A.; Toloczko, Mychailo B.

2014-01-01

The temperature dependence of fracture toughness in HT9 steel irradiated to 3-145 dpa at 380-503 °C was investigated using miniature three-point bend (TPB) fracture specimens. A miniature-specimen reuse technique has been established: the tested halves of subsize Charpy impact specimens with dimensions of 27 mm × 3 mm × 4 mm were reused for this fracture test campaign by cutting a notch with a diamond-saw in the middle of each half, and by fatigue-precracking to generate a sharp crack tip. It was confirmed that the fracture toughness of HT9 steel in the dose range depends more strongly on the irradiation temperature than the irradiation dose. At an irradiation temperature <430 °C, the fracture toughness of irradiated HT9 increased with the test temperature, reached an upper shelf of 180-200 MPa √{m} at 350-450 °C, and then decreased with the test temperature. At an irradiation temperature ⩾430 °C, the fracture toughness was nearly unchanged up to about 450 °C and decreased slowly with test temperatures in a higher temperature range. Such a rather monotonic test temperature dependence after high-temperature irradiation is similar to that observed for an archive material generally showing a higher degree of toughness. A brittle fracture without stable crack growth occurred in only a few specimens with relatively lower irradiation and test temperatures. In this discussion, these TPB fracture toughness data are compared with previously published data from 12.7 mm diameter disc compact tension (DCT) specimens.

Investigation of temperature dependence of fracture toughness in high-dose HT9 steel using small-specimen reuse technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek, Jong-Hyuk; Byun, Thak Sang; Maloy, S

2014-01-01

The temperature dependence of fracture toughness in HT9 steel irradiated to 3 145 dpa at 380 503 C was investigated using miniature three-point bend (TPB) fracture specimens. A miniature-specimen reuse technique has been established: the tested halves of subsize Charpy impact specimens with dimensions of 27 mm 3mm 4 mm were reused for this fracture test campaign by cutting a notch with a diamond-saw in the middle of each half, and by fatigue-precracking to generate a sharp crack tip. It was confirmed that the fracture toughness of HT9 steel in the dose range depends more strongly on the irradiation temperaturemore » than the irradiation dose. At an irradiation temperature <430 C, the fracture toughness of irradiated HT9 increased with the test temperature, reached an upper shelf of 180 200 MPa ffiffiffiffiffi m p at 350 450 C, and then decreased with the test temperature. At an irradiation temperatureP430 C, the fracture toughness was nearly unchanged up to about 450 C and decreased slowly with test temperatures in a higher temperature range. Such a rather monotonic test temperature dependence after high-temperature irradiation is similar to that observed for an archive material generally showing a higher degree of toughness. A brittle fracture without stable crack growth occurred in only a few specimens with relatively lower irradiation and test temperatures. In this discussion, these TPB fracture toughness data are compared with previously published data from 12.7 mm diameter disc compact tension (DCT) specimens.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frederick, Greg; Sutton, Benjamin J.; Tatman, Jonathan K.

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

Materials and fabrication technology of modules intended for irradiation tests of blanket tritium-breeding zones in Russian fusion reactor projects

NASA Astrophysics Data System (ADS)

Kapychev, V.; Davydov, D.; Gorokhov, V.; Ioltukhovskiy, A.; Kazennov, Yu; Tebus, V.; Frolov, V.; Shikov, A.; Shishkov, N.; Kovalenko, V.; Shishkin, N.; Strebkov, Yu

2000-12-01

This paper surveys the modules and materials of blanket tritium-breeding zones developed in the Russian Federation for fusion reactors. Synthesis of lithium orthosilicate, metasilicate and aluminate, fabrication of ceramic pellets and pebbles and experimental reactor units are described. Results of tritium extraction kinetics under irradiation in a water-graphite reactor at a thermal neutron flux of 5×10 13 neutron/(s cm2) are considered. At the present time, development and fabrication of lithium orthosilicate-beryllium modules of the tritium-breeding zone (TBZ), have been carried out within the framework of the ITER and DEMO projects. Two modules containing orthosilicate pellets, porous beryllium and beryllium pebbles are suggested for irradiation tests in the temperature range of 350-700°C. Technical problems associated with manufacturing of the modules are discussed.

Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Shack, W. J.

2008-01-21

In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1more » MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.« less

Investigation of Basic Mechanisms of Radiation Effects in Carbon-Based Electronic Materials

DTIC Science & Technology

2017-06-01

materials characterization, and carbon nanotube diodes, FET, and PZT-memory test device structures for electrical measurements. Pre - and post -irradiation...definition (Radiation exposure) Task 2) The grantee shall perform testing to include: - Radiation testing . May be multiple types. - Pre and post -rad...technologies for electronic devices. Experiential radiation testing has included exposure to 10 keV X-rays, 4 MeV protons, heavy ions, and Ultra

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, J.A.

1996-04-01

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

Microstructural and mechanical investigation of aluminium alloy (Al 1050) melted by microwave hybrid heating

NASA Astrophysics Data System (ADS)

Shashank Lingappa, M.; Srinath, M. S.; Amarendra, H. J.

2017-07-01

Microwave processing of metals is an emerging area. Melting of bulk metallic materials through microwave irradiation is still immature. In view of this, the present paper discusses the melting of bulk Al 1050 metallic material through microwave irradiation. The melting process is carried out successfully in a domestic microwave oven with 900 W power at 2450 MHz frequency. Metallurgical and mechanical characterization of the processed and as-received material is carried out. Aluminium phase is found to be dominant in processed material when tested through x-ray diffraction (XRD). Microstructure study of as-cast metal through scanning electron microscopy (SEM) reveals the formation of uniform hexagonal grain structure free from pores and cavities. The average tensile strength of the cast material is found to be around 21% higher, when compared to as-received material. Vickers’ microhardness of the as-cast metal is measured and is 10% higher than that of the as-received metal. Radiography on as-cast metal shows no significant defects. Al 1050 material melted through microwave irradiation has exhibited superior properties than the as-received Al 1050.

Biostimulation effects of low-energy laser radiation on yeast cell suspensions

NASA Astrophysics Data System (ADS)

Anghel, Sorin; Stanescu, Constantin S.; Giosanu, Dana; Neagu, Ionica; Savulescu, Geta; Iorga-Siman, Ion

2000-02-01

This paper presents work to determine the effects produced by low energy laser radiation on the metabolism and growth of a yeast cell suspension. As experimental material, we used young yeast culture in liquid medium, then distributed on a solid medium, to obtain isolated colonies. As laser source, we used a He-Ne laser, and the irradiation was made with different exposure times. Form each irradiated material, a sample of white grape sterile must was sowed, that has fermented at 18 divided by 20 degrees C for 10 divided by 15 days, after that some properties was tested. Some microscopic studies were also made. The results prove some influence of low energy laser irradiation, which can induce mutations, with new properties of the irradiated material. These mutations can be obtained in a positive sense, with new and important perspectives in wine industry. Also, we observed an inhibitory effect of the laser radiation on the yeast cell growth, due, probably to the too high values of the exposure.

Neutron-Irradiated Samples as Test Materials for MPEX

DOE PAGES

Ellis, Ronald James; Rapp, Juergen

2015-10-09

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

10 CFR 170.31 - Schedule of fees for materials licenses and other regulatory services, including inspections, and...

Code of Federal Regulations, 2013 CFR

2013-01-01

... use of byproduct material in sealed sources for irradiation of materials in which the source is not... irradiation of materials in which the source is exposed for irradiation purposes. This category also includes underwater irradiators for irradiation of materials where the source is not exposed for irradiation purposes...

10 CFR 170.31 - Schedule of fees for materials licenses and other regulatory services, including inspections, and...

Code of Federal Regulations, 2012 CFR

2012-01-01

... use of byproduct material in sealed sources for irradiation of materials in which the source is not... irradiation of materials in which the source is exposed for irradiation purposes. This category also includes underwater irradiators for irradiation of materials where the source is not exposed for irradiation purposes...

Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sokolov, Mikhail A.

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of smallmore » specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy specimens of Midland beltline weld have been machined and just arrived to ORNL as part of this international collaboration. The ORNL will initiate tests of the irradiated Linde 80 weld in FY2017 and results of this international program will be reported in FY2018.« less

Thermal-stress analysis of IFMIF target back-wall made of reduced-activation ferritic steel and austenitic stainless steel

NASA Astrophysics Data System (ADS)

Ida, Mizuho; Chida, Teruo; Furuya, Kazuyuki; Wakai, Eiichi; Nakamura, Hiroo; Sugimoto, Masayoshi

2009-04-01

For long time operation of a liquid lithium target of the International Fusion Materials Irradiation Facility, annual replacement of a back-wall, a part of the flow channel, is planned, since the target suffers neutron damage of more than 50 dpa/fpy. Considering irradiation/activation conditions, remote weld on stainless steel 316L between a back-wall and a target assembly was employed. Furthermore, dissimilar weld between the 316L and a reduced-activation ferritic/martensitic steel F82H in the back-wall was employed. The objective of this study is to clarify structures and materials of the back-wall with acceptable thermal-stress under nuclear heating. Thermal-stress analysis was done using a code ABAQUS and data of the nuclear heating. As a result, thermal-stress in the back-wall is acceptable level, if thickness of the stress-mitigation part is more than 5 mm. With results of the analysis, necessity of material data for F82H and 316L under conditions of irradiation tests and mechanical tests are clarified.

Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit

DOE Office of Scientific and Technical Information (OSTI.GOV)

John H. Strumpell

2004-12-31

Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and/or disposal and enable plants to operate with longer more economical cycle lengths and/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding material and have irradiated several test fuel rods through four cycles. The U.S. Department of Energy (DOE) joined FANP and the BWOG in supporting this project during its final phase of collecting and evaluating high burnup data through post irradiationmore » examination (PIE).« less

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

Smith, Dale L.; Greenwood, Lawrence R.; Loomis, Benny A.

1989-01-01

An apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

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

1988-05-20

This paper discusses an apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

Smith, Dale L.; Greenwood, Lawrence R.; Loomis, Benny A.

1989-03-07

An apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Development of Continuum-Atomistic Approach for Modeling Metal Irradiation by Heavy Ions

NASA Astrophysics Data System (ADS)

Batgerel, Balt; Dimova, Stefka; Puzynin, Igor; Puzynina, Taisia; Hristov, Ivan; Hristova, Radoslava; Tukhliev, Zafar; Sharipov, Zarif

2018-02-01

Over the last several decades active research in the field of materials irradiation by high-energy heavy ions has been worked out. The experiments in this area are labor-consuming and expensive. Therefore the improvement of the existing mathematical models and the development of new ones based on the experimental data of interaction of high-energy heavy ions with materials are of interest. Presently, two approaches are used for studying these processes: a thermal spike model and molecular dynamics methods. The combination of these two approaches - the continuous-atomistic model - will give the opportunity to investigate more thoroughly the processes of irradiation of materials by high-energy heavy ions. To solve the equations of the continuous-atomistic model, a software package was developed and the block of molecular dynamics software was tested on the heterogeneous cluster HybriLIT.

Microstructural development under irradiation in European ODS ferritic/martensitic steels

NASA Astrophysics Data System (ADS)

Schäublin, R.; Ramar, A.; Baluc, N.; de Castro, V.; Monge, M. A.; Leguey, T.; Schmid, N.; Bonjour, C.

2006-06-01

Oxide dispersion strengthened steels based on the ferritic/martensitic steel EUROFER97 are promising candidates for a fusion reactor because of their improved high temperature mechanical properties and their potential higher radiation resistance relative to the base material. Several EUROFER97 based ODS F/M steels are investigated in this study. There are the Plansee ODS steels containing 0.3 wt% yttria, and the CRPP ODS steels, whose production route is described in detail. The reinforcing particles represent 0.3-0.5% weight and are composed of yttria. The effect of 0.3 wt% Ti addition is studied. ODS steel samples have been irradiated with 590 MeV protons to 0.3 and 1.0 dpa at room temperature and 350 °C. Microstructure is investigated by transmission electron microscopy and mechanical properties are assessed by tensile and Charpy tests. While the Plansee ODS presents a ferritic structure, the CRPP ODS material presents a tempered martensitic microstructure and a uniform distribution of the yttria particles. Both materials provide a yield stress higher than the base material, but with reduced elongation and brittle behaviour. Ti additions improve elongation at high temperatures. After irradiation, mechanical properties of the material are only slightly altered with an increase in the yield strength, but without significant decrease in the total elongation, relative to the base material. Samples irradiated at room temperature present radiation induced defects in the form of blacks dots with a size range from 2 to 3 nm, while after irradiation at 350 °C irradiation induced a0<1 0 0>{1 0 0} dislocation loops are clearly visible along with nanocavities. The dispersed yttria particles with an average size of 6-8 nm are found to be stable for all irradiation conditions. The density of the defects and the dispersoid are measured and found to be about 2.3 × 10 22 m -3 and 6.2 × 10 22 m -3, respectively. The weak impact of irradiation on mechanical properties of ODS F/M steel is thus explained by a lower density of irradiation induced defects relative to the density of reinforcing particles.

Test of the wire ageing induced by radiation for the CMS barrel muon chambers

NASA Astrophysics Data System (ADS)

Conti, E.; Gasparini, F.

2001-06-01

We have carried out laboratory tests to measure the ageing of a wire tube due to pollutants outgassed by various materials. The tested materials are those used in the barrel muon drift tubes of the CMS experiment at LHC. An X-ray gun irradiated the test tube to accelerate the ageing process. No ageing effect has been measured for a period equivalent to 10 years of operation at LHC.

Preliminary Analysis of the General Performance and Mechanical Behavior of Irradiated FeCrAl Base Alloys and Weldments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gussev, Maxim N.; Field, Kevin G.; Briggs, Samuel A.

The iron-based, iron-chromium-aluminum (FeCrAl) alloys are promising, robust materials for deployment in current and future nuclear power plants. This class of alloys demonstrates excellent performance in a range of environments and conditions, including high-temperature steam (>1000°C). Furthermore, these alloys have the potential to have prolonged survival under loss-of-coolant accident (LOCA) conditions compared to the more traditional cladding materials that are either Zr-based alloys or austenitic steels. However, one of the issues associated with FeCrAl alloys is cracking during welding. The present project investigates the possibility of mitigating welding-induced cracking via alloying and precise structure control of the weldments; in themore » frame work of the project, several advanced alloys were developed and are being investigated prior to and after neutron irradiation to provide insight into the radiation tolerance and mechanical performance of the weldments. The present report provides preliminary results on the post-irradiation characterization and mechanical tests performed during United States Fiscal Year (FY) 2016. Chapter 1 provides a general introduction, and Chapter 2 describes the alloy compositions, welding procedure, specimen geometry and manufacturing parameters. Also, a brief discussion of the irradiation at the High Flux Isotope Reactor (HFIR) is provided. Chapter 3 is devoted to the analysis of mechanical tests performed at the hot cell facility; tensile curves and mechanical properties are discussed in detail focusing on the irradiation temperature. Limited fractography results are also presented and analyzed. The discussion highlights the limitations of the testing within a hot cell. Chapter 4 underlines the advantages of in-situ testing and discusses the preliminary results obtained with newly developed miniature specimens. Specimens were moved to the Low Activation Materials Development and Analysis (LAMDA) laboratory and prepared for mechanical tests. Tensile tests were conducted at LAMDA using a modern digital image correlation approach allowing for strain distribution analysis. Plastic strain initiation and necking are discussed in detail; a concept of strain rate maps is also introduced and discussed. Follow-on SEM-EBSD and FIBTEM analysis is planned.« less

Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

NASA Astrophysics Data System (ADS)

Gussev, M. N.; Cakmak, E.; Field, K. G.

2018-06-01

Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. Here, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8-1.9 dpa in a temperature range of 195-559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

Investigation of temperature dependence of fracture toughness in high-dose HT9 steel using small-specimen reuse technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek, Jong-Hyuk; Byun, Thak Sang; Maloy, Stuart A.

2014-01-01

The temperature dependence of fracture toughness in HT9 steel irradiated to 3–145 dpa at 380–503 degrees*C was investigated using miniature three-point bend (TPB) fracture specimens. A miniature-specimen reuse technique has been established: the tested halves of subsize Charpy impact specimens with dimensions of 27 mm *3mm* 4 mm were reused for this fracture test campaign by cutting a notch with a diamond-saw in the middle of each half, and by fatigue-precracking to generate a sharp crack tip. It was confirmed that the fracture toughness of HT9 steel in the dose range depends more strongly on the irradiation temperature than themore » irradiation dose. At an irradiation temperature <430 *degreesC, the fracture toughness of irradiated HT9 increased with the test temperature, reached an upper shelf of 180—200 MPa*m^.5 at 350–450 degrees*C, and then decreased with the test temperature. At an irradiation temperature >430 degrees*C, the fracture toughness was nearly unchanged up to about 450 *degreesC and decreased slowly with test temperatures in a higher temperature range. Such a rather monotonic test temperature dependence after high-temperature irradiation is similar to that observed for an archive material generally showing a higher degree of toughness. A brittle fracture without stable crack growth occurred in only a few specimens with relatively lower irradiation and test temperatures. In this discussion, these TPB fracture toughness data are compared with previously published data from 12.7 mm diameter disc compact tension (DCT) specimens.« less

Efficiency of dual-cured resin cement polymerization induced by high-intensity LED curing units through ceramic material.

PubMed

Watanabe, H; Kazama, Re; Asai, T; Kanaya, F; Ishizaki, H; Fukushima, M; Okiji, T

2015-01-01

This study aimed to evaluate the ability of high-intensity light-emitting diode (LED) and other curing units to cure dual-cured resin cement through ceramic material. A halogen curing unit (Jetlite 3000, Morita), a second-generation LED curing unit (Demi, Kerr), and two high-intensity LED curing units (PenCure 2000, Morita; Valo, Ultradent) were tested. Feldspathic ceramic plates (VITABLOCS Mark II, A3; Vita Zahnfabrik) with thicknesses of 1.0, 2.0, and 3.0 mm were prepared. Dual-cured resin cement samples (Clearfil Esthetic Cement, Kuraray Noritake Dental) were irradiated directly or through one of the ceramic plates for different periods (5, 10, 15, or 20 seconds for the high-intensity LED units and 20, 40, 60, or 80 seconds for the others). The Knoop hardness test was used to determine the level of photopolymerization that had been induced in the resin cement. Data were analyzed by one-way analysis of variance and Dunnett's post-hoc test to identify test-control (maximum irradiation without a ceramic plate) differences for each curing unit (p<0.05). For all curing units, the curing conditions had a statistically significant effect on the Knoop hardness numbers (KHNs) of the irradiated cement samples (p<0.001). In general, the KHN decreased with increasing plate thickness and increased as the irradiation period was extended. Jetlite 3000 achieved control-level KHN values only when the plate thickness was 1.0 mm. At a plate thickness ≥2.0 mm, the LED units (except for PenCure 2000 at 3.0 mm) were able to achieve control-level KHN values when the irradiation time was extended. At a plate thickness of 3.0 mm, irradiation for 20 seconds with the Valo or for 80 seconds with the Demi were the only methods that produced KHN values equivalent to those produced by direct irradiation. Regardless of the type of curing unit used, indirect irradiation of dual-cured resin cement through a ceramic plate resulted in decreased KHN values compared with direct irradiation. When the irradiation period was extended, only the LED units were able to achieve similar KHN values to those observed under direct irradiation in the presence of plates ≥2.0-mm thick. High-intensity LED units require a shorter irradiation period than halogen and second-generation LED curing units to obtain KHN values similar to those observed during direct irradiation.

He implantation induced microstructure- and hardness-modification of the intermetallic γ-TiAl

NASA Astrophysics Data System (ADS)

Pouchon, Manuel A.; Chen, Jiachao; Hoffelner, Wolfgang

2009-05-01

TiAl is a well known high temperature material with good creep properties. It is investigated as a potential structural material for Generation IV high temperature gas cooled nuclear reactors. The tests are performed with the ABB-2 (Ti-rich TiAl with 2 at.% W) developed by ASEA Brown Boveri Ltd. (ABB). Thin samples are irradiated throughout with 24 MeV 4He2+ ions; the irradiated material is then investigated towards its microstructure and its hardness. The microstructure is studied by transmission electron microscopy and the hardness is investigated using a micro-hardness tester and a nano-indenter. Different effects can be identified. From room to moderate irradiation temperatures, the radiation induced hardening of the material slowly vanishes until the material completely recovers at about 943 K. Beyond this temperature, He-bubble formation seems to harden the material again, until beyond 1200 K a steep increase in hardening is detected. This effect can be correlated with bubbles being identified in the micrographs. The results are consistent and give strong indications to a microstructural development as a function of temperature.

Irradiation qualification testing of SNAP-10A components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chesavage, A. J.

1964-02-04

Selected SNAP 10A components were irradiated to about 10{sup14} nvt and 5{times} 10{sup 7} r at an average temperature of 136{degrees}F in a nominal vacuum of 2 {times} 10{sup {minus}5} torr. The components were operated periodically and the electrical characteristics recorded. Pre-irradiationand post-irradiation tests were conducted. Catastropic degradation occurred only in the low-level neutron detection system and about 1.5 {times} 10{sup 13} nvt and in the high-level neutron power supply at about 6{times} 10{sup 12} nvt. Marginal degradation occurred in the fusistors and in the silicone rubber insert material in connectors. The relays, low-voltage trip devices, expansion compensator position demodulator,more » resistance thermometer sensor and bridge, and the gamma detection system opearted within their respective specifications during and after irradiation. The insulation resistance of all components was adeqauate during and after irradiation.« less

10 CFR 171.16 - Annual fees: Materials licensees, holders of certificates of compliance, holders of sealed source...

Code of Federal Regulations, 2013 CFR

2013-01-01

.... Licenses for possession and use of byproduct material in sealed sources for irradiation of materials in... sources for irradiation of materials in which the source is exposed for irradiation purposes. This category also includes underwater irradiators for irradiation of materials in which the source is not...

Polarization of electron-beam irradiated LDPE films: contribution to charge generation and transport

NASA Astrophysics Data System (ADS)

Banda, M. E.; Griseri, V.; Teyssèdre, G.; Le Roy, S.

2018-04-01

Electron-beam irradiation is an alternative way to generate charges in insulating materials, at controlled position and quantity, in order to monitor their behaviour in regard to transport phenomena under the space charge induced electric field or external field applied. In this study, low density polyethylene (LDPE) films were irradiated by a 80 keV electron-beam with a flux of 1 nA cm‑2 during 10 min in an irradiation chamber under vacuum conditions, and were then characterized outside the chamber using three experimental methods. The electrical behaviour of the irradiated material was assessed by space charge measurements using the pulsed electro-acoustic (PEA) method under dc stress. The influence of the applied electric field polarity and amplitude has been tested in order to better understand the charge behaviour after electron-beam irradiation. Fourier transform infra-red spectroscopy (FTIR) and photoluminescence (PL) measurements were performed to evaluate the impact of the electron beam irradiation, i.e. deposited charges and energy, on the chemical structure of the irradiated samples. The present results show that the electrical behaviour in LDPE after irradiation is mostly driven by charges, i.e. by physical process functions of the electric field, and that changes in the chemical structure seems to be mild.

The effect of high energy concentration source irradiation on structure and properties of Fe-based bulk metallic glass

NASA Astrophysics Data System (ADS)

Pilarczyk, Wirginia

2016-06-01

Metallic glasses exhibit metastable structure and maintain this relatively stable amorphous state within certain temperature range. High intensity laser beam was used for the surface irradiation of Fe-Co-B-Si-Nb bulk metallic glasses. The variable parameter was laser beam pulse energy. For the analysis of structure and properties of bulk metallic glasses and their surface after laser remelting the X-ray analysis, microscopic observation and test of mechanical properties were carried out. Examination of the nanostructure of amorphous materials obtained by high pressure copper mold casting method and the irradiated with the use of TITAN 80-300 HRTEM was carried out. Nanohardness and reduced Young's modulus of particular amorphous and amorphous-crystalline material zone of the laser beam were examined with the use of Hysitron TI950 Triboindenter nanoindenter and with the use of Berkovich's indenter. The XRD and microscopic analysis showed that the test material is amorphous in its structure before irradiation. Microstructure observation with electron transmission microscopy gave information about alloy crystallization in the irradiated process. Identification of given crystal phases allows to determine the kind of crystal phases created in the first place and also further changes of phase composition of alloy. The main value of the nanohardness of the surface prepared by laser beam has the order of magnitude similar to bulk metallic glasses formed by casting process irrespective of the laser beam energy used. Research results analysis showed that the area between parent material and fusion zone is characterized by extraordinarily interesting structure which is and will be the subject of further analysis in the scope of bulk metallic glasses amorphous structure and high energy concentration source. The main goal of this work is the results' presentation of structure and chosen properties of the selected bulk metallic glasses after casting process and after irradiation process employing the high energy concentration sources.

Effects of long duration exposure to simulated space environment on nonmetallic materials properties

NASA Technical Reports Server (NTRS)

Peacock, C. L., Jr.; Whitaker, A. F.

1983-01-01

Nonmetallic materials specimens from the Viking program were tested in situ invacuo after continuous thermal vacuum exposure from 1971/1972 to the present. Eleven tests were done on appropriate specimens of 30 materials; however, no single material received all the tests. Some specimens also were exposed to 1 or 2.5 MeV electrons at differing fluences before testing. Baseline exposure data is reported for graphite/epoxy specimens that were exposed to vacuum since 1974. These materials were transferred to the thermal vacuum storage facility for future in situ testing and irradiation. Thin G/E specimens were tensile tested after thermal-vacuum cycling exposure. Photomicrographic examinations and SEM analyses were done on the failed specimens.

10 CFR 36.69 - Irradiation of explosive or flammable materials.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Irradiation of explosive or flammable materials. 36.69... IRRADIATORS Operation of Irradiators § 36.69 Irradiation of explosive or flammable materials. (a) Irradiation... cause radiation overexposures of personnel. (b) Irradiation of more than small quantities of flammable...

10 CFR 36.69 - Irradiation of explosive or flammable materials.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Irradiation of explosive or flammable materials. 36.69... IRRADIATORS Operation of Irradiators § 36.69 Irradiation of explosive or flammable materials. (a) Irradiation... cause radiation overexposures of personnel. (b) Irradiation of more than small quantities of flammable...

10 CFR 36.69 - Irradiation of explosive or flammable materials.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Irradiation of explosive or flammable materials. 36.69... IRRADIATORS Operation of Irradiators § 36.69 Irradiation of explosive or flammable materials. (a) Irradiation... cause radiation overexposures of personnel. (b) Irradiation of more than small quantities of flammable...

10 CFR 36.69 - Irradiation of explosive or flammable materials.

Code of Federal Regulations, 2010 CFR

2010-01-01

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

10 CFR 36.69 - Irradiation of explosive or flammable materials.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Irradiation of explosive or flammable materials. 36.69... IRRADIATORS Operation of Irradiators § 36.69 Irradiation of explosive or flammable materials. (a) Irradiation... cause radiation overexposures of personnel. (b) Irradiation of more than small quantities of flammable...

JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bignan, G.; Gonnier, C.; Lyoussi, A.

2015-07-01

Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and Dmore » support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under construction at CEA Cadarache research centre in the south of France will represent a major Research Infrastructure for scientific studies regarding material and fuel behavior under irradiation. It will also be devoted to medical isotopes production. Hence JHR will offer a real opportunity to perform R and D programs regarding needs above and hence will crucially contribute to the selection, optimization and qualification of these innovative materials and fuels. The JHR reactor objectives, principles and main characteristics associated to specific experimental devices associated to measurement techniques and methodology, their performances, their limitations and field of applications will be presented and discussed. (authors)« less

Low Activation Joining of SiC/SiC Composites for Fusion Applications: Modeling Miniature Torsion Tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.

2014-06-30

The use of SiC and SiC-composites in fission or fusion environments appears to require joining methods for assembling systems. The international fusion community has designed miniature torsion specimens for joint testing and for irradiation in HFIR. Therefore, miniature torsion joints were fabricated using displacement reactions between Si and TiC to produce Ti3SiC2 + SiC joints with CVD-SiC that were tested in shear prior to and after HFIR irradiation. However, these torsion specimens fail out-of-plane, which causes difficulties in determining a shear strength for the joints or for comparing unirradiated and irradiated joints. A finite element damage model has been developedmore » that indicates fracture is likely to occur within the joined pieces to cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. The implications for torsion shear joint data based on this sample design are discussed.« less

Irradiation creep-fatigue interaction of type 316L stainless steel

NASA Astrophysics Data System (ADS)

Scholz, R.; Mueller, R.

1996-10-01

Type 316L stainless steel samples in both, 20% cold-worked (cw) and recrystallised (rc) conditions were exposed to strain controlled fatigue cycling in torsion at 400°C during an irradiation with 19 MeV deuterons. The effect of irradiation creep induced stress relaxation on the fatigue life was studied by imposing a hold time at the minimum strain value in the loading cycle. For the cw material at strain ranges of 1.13% and 1.3%, the absolute stress values, τ H, maintained during the hold time decreased with the number of cycles due to the irradiation creep induced stress relaxation. A mean stress was built up. The number of cycles to failure was considerably reduced in comparison to continuous cycling tests under thermal conditions. For the rc material at strain ranges of 1.03% and 1.4%, the values of τ H increased with the number of cycles, despite the hold time imposed, due to irradiation and/or cyclic hardening.

Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

DOEpatents

Johnson, Jr., A. Burtron; Levy, Ira S.; Trimble, Dennis J.; Lanning, Donald D.; Gerber, Franna S.

1990-01-01

An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-bsed materials is disclosed. Samples of zirconium-based materials having different composition and/or fabrication are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280.degree. to 316.degree. C.). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hyriding for the same materials when subject to in-reactor (irradiated) corrision.

Neural network analysis of Charpy transition temperature of irradiated low-activation martensitic steels

NASA Astrophysics Data System (ADS)

Cottrell, G. A.; Kemp, R.; Bhadeshia, H. K. D. H.; Odette, G. R.; Yamamoto, T.

2007-08-01

We have constructed a Bayesian neural network model that predicts the change, due to neutron irradiation, of the Charpy ductile-brittle transition temperature (ΔDBTT) of low-activation martensitic steels given a set of multi-dimensional published data with doses <100 displacements per atom (dpa). Results show the high significance of irradiation temperature and (dpa) 1/2 in determining ΔDBTT. Sparse data regions were identified by the size of the modelling uncertainties, indicating areas where further experimental data are needed. The method has promise for selecting and ranking experiments on future irradiation materials test facilities.

International strategy for fusion materials development

NASA Astrophysics Data System (ADS)

Ehrlich, Karl; Bloom, E. E.; Kondo, T.

2000-12-01

In this paper, the results of an IEA-Workshop on Strategy and Planning of Fusion Materials Research and Development (R&D), held in October 1998 in Risø Denmark are summarised and further developed. Essential performance targets for materials to be used in first wall/breeding blanket components have been defined for the major materials groups under discussion: ferritic-martensitic steels, vanadium alloys and ceramic composites of the SiC/SiC-type. R&D strategies are proposed for their further development and qualification as reactor-relevant materials. The important role of existing irradiation facilities (mainly fission reactors) for materials testing within the next decade is described, and the limits for the transfer of results from such simulation experiments to fusion-relevant conditions are addressed. The importance of a fusion-relevant high-intensity neutron source for the development of structural as well as breeding and special purpose materials is elaborated and the reasons for the selection of an accelerator-driven D-Li-neutron source - the International Fusion Materials Irradiation Facility (IFMIF) - as an appropriate test bed are explained. Finally the necessity to execute the materials programme for fusion in close international collaboration, presently promoted by the International Energy Agency, IEA is emphasised.

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Effect of γ-irradiation on the optical and electrical properties of fiber reinforced composites

NASA Astrophysics Data System (ADS)

Anwar, Ahmad; Elfiky, Dalia; Ramadan, Ahmed M.; Hassan, G. M.

2017-05-01

The effect of gamma irradiation on the optical and electrical properties of the reinforced fiber polymeric based materials became an important issue. Fiberglass/epoxy and Kevlar fiber/epoxy were selected as investigated samples manufactured with hand lay-up without autoclave curing technique. The selected technique is simple and low cost while being rarely used in space materials production. The electric conductivity and dielectric constant for those samples were measured with increasing the gamma radiation dose. Moreover, the absorptivity, band gap and color change were determined. Fourier transform infrared (FTIR) was performed to each of the material's constituent to evaluate the change in the investigated materials due to radiation exposure dose. In this study, the change of electrical properties for both investigated materials showed a slight variation of the test parameters with respect to the gamma dose increase; this variation is placed in the insulators rang. The tested samples showed an insulator stable behavior during the test period. The change of optical properties for both composite specimens showed the maximum absorptivity at the gamma dose 750 kGy. These materials are suitable for structure materials and thermal control for orbital life less than 7 years. In addition, the transparency of epoxy matrix was degraded. However, there is no color change for either Kevlar fiber or fiberglass.

Materials science education: ion beam modification and analysis of materials

NASA Astrophysics Data System (ADS)

Zimmerman, Robert; Muntele, Claudiu; Ila, Daryush

2012-08-01

The Center for Irradiation of Materials (CIM) at Alabama A&M University (http://cim.aamu.edu) was established in 1990 to serve the University in its research, education and services to the need of the local community and industry. CIM irradiation capabilities are oriented around two tandem-type ion accelerators with seven beam lines providing high-resolution Rutherford backscattering spectrometry, MeV focus ion beam, high-energy ion implantation and irradiation damage studies, particle-induced X-ray emission, particle-induced gamma emission and ion-induced nuclear reaction analysis in addition to fully automated ion channeling. One of the two tandem ion accelerators is designed to produce high-flux ion beam for MeV ion implantation and ion irradiation damage studies. The facility is well equipped with a variety of surface analysis systems, such as SEM, ESCA, as well as scanning micro-Raman analysis, UV-VIS Spectrometry, luminescence spectroscopy, thermal conductivity, electrical conductivity, IV/CV systems, mechanical test systems, AFM, FTIR, voltammetry analysis as well as low-energy implanters, ion beam-assisted deposition and MBE systems. In this presentation, we will demonstrate how the facility is used in material science education, as well as providing services to university, government and industry researches.

PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

NASA Astrophysics Data System (ADS)

Leclercq, Sylvain; Lidbury, David; Van Dyck, Steven; Moinereau, Dominique; Alamo, Ana; Mazouzi, Abdou Al

2010-11-01

In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations… from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach young researchers in the field of materials' degradation. PERFORM 60 has officially started on March 1st, 2009 with 20 European organizations and Universities involved in the nuclear field.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maloy, Stuart Andrew; Pestovich, Kimberly Shay; Anderoglu, Osman

The Fuel Cycle Research and Development program is investigating methods of transmuting minor actinides in various fuel cycle options. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Recent results from testing numerous ferritic/martensitic steels at low temperatures suggest that improvements inmore » low temperature radiation tolerance can be achieved through carefully controlling the nitrogen content in these alloys. Thus, four new heats of HT-9 were produced with controlled nitrogen content: two by Metalwerks and two by Sophisticated Alloys. Initial results on these new alloys are presented including microstructural analysis and hardness testing. Future testing will include irradiation testing with ions and in reactor.« less

Ductility recovery in structural materials for spallation targets by post-irradiation annealing

NASA Astrophysics Data System (ADS)

Chen, J.; Jung, P.; Rödig, M.; Ullmaier, H.; Bauer, G. S.

2005-08-01

Low temperature irradiation embrittlement is one of the major criteria to determine the lifetime of spallation targets. Embrittlement is especially high at low service temperatures, e.g. 250 °C in liquid-mercury sources. It was the aim of the present study to investigate the effect of post-irradiation annealing on the mechanical properties of irradiated structural materials. The specimens used were obtained from spent target components of operating spallation facilities (Los Alamos Neutron Science Center, LANSCE, and the Spallation Neutron Source at Rutherford-Appleton Laboratory, ISIS). The investigated materials include a nickel-based alloy (IN718), an austenitic stainless steel (AISI 304L), a martensitic stainless steel (DIN 1.4926) and a refractory metal (Ta) which experienced 800 MeV proton irradiation to fluences of several 10 25 p/m 2. The specimens were annealed from 300 °C to 700 °C for 1 to 10 h, respectively, and their mechanical property changes were subsequently investigated at room temperature and 250 °C by tensile testing and fracture surface analysis conducted by scanning electron microscopy (SEM). The results showed that the ductility recovered to a large degree in 304L and DIN 1.4926 materials while their strength remained almost unchanged. Especially for DIN 1.4926, the ductility recovery is remarkable already at 400 °C. Together with its favorable thermo-mechanical properties, this makes martensitic steel a candidate for structural materials of spallation targets.

Crystal MD: The massively parallel molecular dynamics software for metal with BCC structure

NASA Astrophysics Data System (ADS)

Hu, Changjun; Bai, He; He, Xinfu; Zhang, Boyao; Nie, Ningming; Wang, Xianmeng; Ren, Yingwen

2017-02-01

Material irradiation effect is one of the most important keys to use nuclear power. However, the lack of high-throughput irradiation facility and knowledge of evolution process, lead to little understanding of the addressed issues. With the help of high-performance computing, we could make a further understanding of micro-level-material. In this paper, a new data structure is proposed for the massively parallel simulation of the evolution of metal materials under irradiation environment. Based on the proposed data structure, we developed the new molecular dynamics software named Crystal MD. The simulation with Crystal MD achieved over 90% parallel efficiency in test cases, and it takes more than 25% less memory on multi-core clusters than LAMMPS and IMD, which are two popular molecular dynamics simulation software. Using Crystal MD, a two trillion particles simulation has been performed on Tianhe-2 cluster.

Research and development on materials for the SPES target

NASA Astrophysics Data System (ADS)

Corradetti, Stefano; Andrighetto, Alberto; Manzolaro, Mattia; Scarpa, Daniele; Vasquez, Jesus; Rossignoli, Massimo; Monetti, Alberto; Calderolla, Michele; Prete, Gianfranco

2014-03-01

The SPES project at INFN-LNL (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro) is focused on the production of radioactive ion beams. The core of the SPES facility is constituted by the target, which will be irradiated with a 40 MeV, 200 µA proton beam in order to produce radioactive species. In order to efficiently produce and release isotopes, the material constituting the target should be able to work under extreme conditions (high vacuum and temperatures up to 2000 °C). Both neutron-rich and proton-rich isotopes will be produced; in the first case, carbon dispersed uranium carbide (UCx) will be used as a target, whereas to produce p-rich isotopes, several types of targets will have to be irradiated. The synthesis and characterization of different types of material will be reported. Moreover, the results of irradiation and isotopes release tests on different uranium carbide target prototypes will be discussed.

Meso-scale modeling of irradiated concrete in test reactor

DOE PAGES

Giorla, Alain B.; Vaitová, M.; Le Pape, Yann; ...

2015-10-18

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

10 CFR 170.31 - Schedule of fees for materials licenses and other regulatory services, including inspections, and...

Code of Federal Regulations, 2014 CFR

2014-01-01

... for irradiation of materials in which the source is not removed from its shield (self-shielded units... curies of byproduct material in sealed sources for irradiation of materials in which the source is exposed for irradiation purposes. This category also includes underwater irradiators for irradiation of...

Irradiation Induced Microstructure Evolution in Nanostructured Materials: A Review

PubMed Central

Liu, Wenbo; Ji, Yanzhou; Tan, Pengkang; Zang, Hang; He, Chaohui; Yun, Di; Zhang, Chi; Yang, Zhigang

2016-01-01

Nanostructured (NS) materials may have different irradiation resistance from their coarse-grained (CG) counterparts. In this review, we focus on the effect of grain boundaries (GBs)/interfaces on irradiation induced microstructure evolution and the irradiation tolerance of NS materials under irradiation. The features of void denuded zones (VDZs) and the unusual behavior of void formation near GBs/interfaces in metals due to the interactions between GBs/interfaces and irradiation-produced point defects are systematically reviewed. Some experimental results and calculation results show that NS materials have enhanced irradiation resistance, due to their extremely small grain sizes and large volume fractions of GBs/interfaces, which could absorb and annihilate the mobile defects produced during irradiation. However, there is also literature reporting reduced irradiation resistance or even amorphization of NS materials at a lower irradiation dose compared with their bulk counterparts, since the GBs are also characterized by excess energy (compared to that of single crystal materials) which could provide a shift in the total free energy that will lead to the amorphization process. The competition of these two effects leads to the different irradiation tolerance of NS materials. The irradiation-induced grain growth is dominated by irradiation temperature, dose, ion flux, character of GBs/interface and nanoprecipitates, although the decrease of grain sizes under irradiation is also observed in some experiments. PMID:28787902

HTGR fuels and core development program. Quarterly progress report for the period ending August 31, 1975

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1975-09-30

Studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies are described. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and, where appropriate, the data are presented in tables, graphs, and photographs. (auth)

Gamma irradiation induced effects of butyl rubber based damping material

NASA Astrophysics Data System (ADS)

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

2018-04-01

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

Irradiation-induced grain growth and defect evolution in nanocrystalline zirconia with doped grain boundaries

DOE PAGES

Dey, Sanchita; Mardinly, John; Wang, Yongqiang; ...

2016-05-27

Grain boundaries are effective sinks for radiation-induced defects, ultimately impacting the radiation tolerance of nanocrystalline materials (dense materials with nanosized grains) against net defect accumulation. However, irradiation-induced grain growth leads to grain boundary area decrease, shortening potential benefits of nanostructures. A possible approach to mitigate this is the introduction of dopants to target a decrease in grain boundary mobility or a reduction in grain boundary energy to eliminate driving forces for grain growth (using similar strategies as to control thermal growth). Here, in this study, we tested this concept in nanocrystalline zirconia doped with lanthanum. Although the dopant is observedmore » to segregate to the grain boundaries, causing grain boundary energy decrease and promoting dragging forces for thermally activated boundary movement, irradiation induced grain growth could not be avoided under heavy ion irradiation, suggesting a different growth mechanism as compared to thermal growth. Furthermore, it is apparent that reducing the grain boundary energy reduced the effectiveness of the grain boundary as sinks, and the number of defects in the doped material is higher than in undoped (La-free) YSZ.« less

Mechanical properties evaluation of extruded wood polymer composites

NASA Astrophysics Data System (ADS)

Zaini, A. S. Syah M.; Rus, Anika Zafiah M.; Rahman, Norherman Abdul; Jais, Farhana Hazwanee M.; Fauzan, M. Zarif; Sufian, N. Afiqah

2017-09-01

The rapidly expanding of interest in the manufacture of composite materials from waste industrial and agricultural materials is due to high demand for environmentally friendly materials. Wood polymer composite (WPC) are being used in many type of applications such as in the automobile, electronic, aerospace industry and construction. Therefore, this research study is to determine the mechanical properties behaviour of WPC after an extended Ultra Violet (UV) irradiation exposure. The fabricated sample has been used and to be compared in this research is consists of rice husk, waste fibre and polypropylene (PP) with 4 different types of WPC which are wood block waste (WBW), wood block virgin (WBV), wood sheet (WS) and wood sheet waste (WSW). The extruded specimens were tested for mechanical properties such as strength under compression, puncture strength and impact resistance, and density. In addition, the specimen has been irradiated with the UV exposure at 5000 hours, 10000 hours and 15000 hours. Generally, the mechanical properties the WPC which made from the recycled material were lower than the WPC from virgin material but the density was comparable between the two products after UV irradiation exposure.

Proton Irradiation Induced Effects in Titanium Carbide and Titanium Nitride: An Evaluation of Microstructures and Mechanical Properties

NASA Astrophysics Data System (ADS)

Dickerson, Clayton A.

The materials TiC and TiN have been identified as potential candidate materials for advanced coated nuclear fuel components for the gas-cooled fast reactor (GFR). While a number of their thermal and mechanical properties have been studied, little is known about how these ceramics respond to particle irradiation. The goal of this study was to investigate the radiation effects in TiC and TiN by analyzing the irradiated microstructures and mechanical properties. Irradiations of TiC and TiN were conducted with 2.6 MeV protons at the University of Wisconsin -- Madison to simulate proposed conditions expected in a reactor. Each material was subjected to three incident proton fluences resulting in doses of ˜0.2 dpa to ˜1 dpa at three temperatures, 600°C, 800°C, and 900°C. Post irradiation examination included microstructural analysis via TEM, lattice parameter determinations with XRD, and mechanical property measurements with micro indentation hardness and fracture toughness tests. The predominant irradiation induced aggregate defects found by high resolution TEM and diffraction contrast TEM in both irradiated TiC and TiN were interstitial faulted dislocation loops. Only circular loops were identified in TiC while both circular and triangular loops were present in TiN. The influences on the microstructural evolution from a high inherent density of dislocations and high porosity were also determined. The strains resulting from the development of the defective microstructures were measured with XRD and shown to be highly dependent on the density of dislocation loops. Maximum strains for the irradiated samples were on the order of 0.5%. Measurements of the fracture toughness of Tic samples were made by ion milling the surface of the samples to create micro cantilever beams which were subsequently fractured by nano indentation. The formation of high densities of dislocation loops in the irradiated samples was found to significantly decrease the material's fracture toughness.

The Charpy impact properties of martensitic 10.6% Cr steel (MANET-1) before and after neutron exposure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rieth, M.; Dafferner, B.; Rohrig, H.D.

1994-12-31

The MANET-I martensitic 10.6% Cr type of steel was developed as a potential structural material for the first wall and the blanket of a fusion device within the framework of the Nuclear Fusion Project. An extensive irradiation program (FRUST/SIENA) was elaborated to study the influence of radiation upon the Charpy impact characteristics. In addition to unirradiated reference specimens, 87 irradiated subsize Charpy specimens (3 x 4 x 27 mm{sup 3}) were examined under eight different heat treatments at irradiation temperatures between 287{degrees}C and 475{degrees}C and exposure doses of 5 dpa to 15 dpa. On the basis of the numerous testmore » results and their interpretation it is possible to describe radiation induced material embrittlement, and, consequently, the deterioration of the Charpy impact properties. The description is limited, on the one hand, by the variations in the test results and, on the other hand, by the gaps in the test matrix. Therefore, additional investigations, especially in the low irradiation temperature and low dose regimes will be the subject of further ongoing work.« less

Advances in tribological testing of artificial joint biomaterials using multidirectional pin-on-disk testers

PubMed Central

Baykal, D.; Siskey, R.S.; Haider, H.; Saikko, V.; Ahlroos, T.; Kurtz, S.M.

2013-01-01

The introduction of numerous formulations of Ultra-high molecular weight polyethylene (UHMWPE), which is widely used as a bearing material in orthopedic implants, necessitated screening of bearing couples to identify promising iterations for expensive joint simulations. Pin-on-disk (POD) testers capable of multidirectional sliding can correctly rank formulations of UHMWPE with respect to their predictive in vivo wear behavior. However, there are still uncertainties regarding POD test parameters for facilitating clinically relevant wear mechanisms of UHMWPE. Studies on the development of POD testing were briefly summarized. We systematically reviewed wear rate data of UHMWPE generated by POD testers. To determine if POD testing was capable of correctly ranking bearings and if test parameters outlined in ASTM F732 enabled differentiation between wear behavior of various formulations, mean wear rates of non-irradiated, conventional (25–50 kGy) and highly crosslinked (≥90 kGy) UHMWPE were grouped and compared. The mean wear rates of non-irradiated, conventional and highly crosslinked UHMWPEs were 7.03, 5.39 and 0.67 mm3/MC. Based on studies that complied with the guidelines of ASTM F732, the mean wear rates of non-irradiated, conventional and highly crosslinked UHMWPEs were 0.32, 0.21 and 0.04 mm3/km, respectively. In both sets of results, the mean wear rate of highly crosslinked UHMPWE was smaller than both conventional and non-irradiated UHMWPEs (p<0.05). Thus, POD testers can compare highly crosslinked and conventional UHMWPEs despite different test parameters. Narrowing the allowable range for standardized test parameters could improve sensitivity of multi-axial testers in correctly ranking materials. PMID:23831149

Testing of SRS and RFETS Nylon Bag Material

DOE Office of Scientific and Technical Information (OSTI.GOV)

Laurinat, J.E.

1998-11-03

This report compares the effects of radiation and heating on nylon bagout materials used at the Savannah River Site (SRS) and the Rocky Flats Environmental Technology Site (RFETS). Recently, to simplify the processing of sand, slag, and crucible (SS and C), FB-Line has replaced the low-density polyethylene (LDPE) and polyvinyl chloride (PVC) bags normally used to package cans of plutonium-bearing material with nylon bags. LDPE and PVC are not soluble in the nitric acid dissolver solution used in F-Canyon, so cans bagged using these materials had to be repackaged before they were added to the dissolver. Because nylon dissolves inmore » nitric acid, cans bagged in nylon can be charged to the F-Canyon dissolvers without repackaging, thereby reducing handling requirements and personnel exposure. As part of a program to process RFETS SS and C at SRS, RFETS has also begun to use a nylon bagout material. The RFETS bag materials is made from a copolymer of nylon 6 and nylon 6.9, while the SRS material is made from a nylon 6 monomer. In addition, the SRS nylon has an anti-static agent added. The RFETS nylon is slightly softer than the SRS nylon, but does not appear to be as resistant to flex cracks initiated by contact with sharp corners of the inner can containing the SS and C.2 FB-Line Operations has asked for measurement of the effects of radiation and heating on these materials. Specifically, they have requested a comparison of the material properties of the plastics before and after irradiation, a measurement of the amount of outgassing when the plastics are heated, and a calculation of the amount of radiolytic gas generation. Testing was performed on samples taken from material that is currently used in FB-Line (color coded orange) and at RFETS. The requested tests are the same tests previously performed on the original and replacement nylon and LDPE bag materials.3,4,5. To evaluate the effect of irradiation on material properties, tensile stresses and elongations to break w ere compared for unirradiated and irradiated samples. A standard ASTM method for the measurement of tensile plastic properties6 was used. Properties were measured both parallel to the direction of machining (MD) and transverse to the direction of machining (TD). Tensile strength measurements showed that the ultimate strengths of the SRS replacement bag material decreased by 22 percent in the MD orientation and 17 percent in the TD orientation after irradiation with 5 x 106 rad, a dose equivalent to about 8-9 months exposure in a plutonium can. For the RFETS material, the decreases were 23 percent in the MD orientation and 56 percent in the TD orientation. Although the 5 x 106 dose significantly degraded the properties of both materials, their strengths remained superior to those previously measured for LDPE,4 even after irradiation. Elongations to break also decreased, especially for the SRS material. The decrease for the SRS material were 86 percent in the MD orientation and 95 percent in the TD orientation. For the RFETS material, elongations to break decreased at least 18 percent in the MD orientation and 29 percent in the TD orientation. When samples of both the SRS and RFETS materials were heated in a sealed container to the maximum expected storage can temperature of about 95 C, they outgassed at pressures ranging from 16 to 22 psig. These pressure increases would not cause a can to fail. Using a representative G value of 1.6 molecules/100 ev, the amount of outgassing due to radiolysis was calculated to be negligible. In conclusion, it may be stated that the results of the strength tests and the outgassing measurements and calculations demonstrate that the SRS and RFETS replacement bag materials are acceptable substitutes for LDPE with respect to mechanical properties.« less

NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, Troy; Reichenberger, Michael; Stevenson, Sarah

2016-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report,more » the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating. • KSU delivered fissile HT MPFD parts to INL for final construction of HT MPFD prototype. • A prototype HT MPFD was constructed and analyzed at INL. • The HT MPFD has been modeled in MCNP to optimize the amount of fissile material deposition. • The HT MPFD has been modeled in MCNP to optimize the sensor location in the irradiation test. • The fissile material deposition is undergoing independent verifications. • Detector amplifier electronics have been revised and tested by KSU. • Several project meetings were held at INL and KSU to discuss the roles and responsibilities between INL, KSU, and CEA for development and deployment of the HT MPFDs. As documented in this report, FY16 funding has allowed the project to meet year two planned accomplishments to develop a HT MPFD. In addition, the accomplishments of this project have attracted independent funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for MTR irradiations of the MPFD technology. These are significant opportunities for this NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors project because the irradiation expense of these experiments could not be included in the original project scope.« less

Environmentally assisted cracking in light water reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Chung, H. M.; Clark, R. W.

2007-11-06

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the currentmore » choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature indicate that IASCC in 289 C water is dominated by a crack-tip grain-boundary process that involves S. An initial IASCC model has been proposed. A crack growth test was completed on mill annealed Alloy 600 in high-purity water at 289 C and 320 C under various environmental and loading conditions. The results from this test are compared with data obtained earlier on several other heats of Alloy 600.« less

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daw, J.E.; Knudson, D.L.; Villard, J.F.

2015-07-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physicalmore » property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO{sub 2} and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)« less

Predictive characterization of aging and degradation of reactor materials in extreme environments. Final report, December 20, 2013 - September 20, 2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qu, Jianmin

Understanding of reactor material behavior in extreme environments is vital not only to the development of new materials for the next generation nuclear reactors, but also to the extension of the operating lifetimes of the current fleet of nuclear reactors. To this end, this project conducted a suite of unique experimental techniques, augmented by a mesoscale computational framework, to understand and predict the long-term effects of irradiation, temperature, and stress on material microstructures and their macroscopic behavior. The experimental techniques and computational tools were demonstrated on two distinctive types of reactor materials, namely, Zr alloys and high-Cr martensitic steels. Thesemore » materials are chosen as the test beds because they are the archetypes of high-performance reactor materials (cladding, wrappers, ducts, pressure vessel, piping, etc.). To fill the knowledge gaps, and to meet the technology needs, a suite of innovative in situ transmission electron microscopy (TEM) characterization techniques (heating, heavy ion irradiation, He implantation, quantitative small-scale mechanical testing, and various combinations thereof) were developed and used to elucidate and map the fundamental mechanisms of microstructure evolution in both Zr and Cr alloys for a wide range environmental boundary conditions in the thermal-mechanical-irradiation input space. Knowledge gained from the experimental observations of the active mechanisms and the role of local microstructural defects on the response of the material has been incorporated into a mathematically rigorous and comprehensive three-dimensional mesoscale framework capable of accounting for the compositional variation, microstructural evolution and localized deformation (radiation damage) to predict aging and degradation of key reactor materials operating in extreme environments. Predictions from this mesoscale framework were compared with the in situ TEM observations to validate the model.« less

Mechanical strength of an ITER coil insulation system under static and dynamic load after reactor irradiation

NASA Astrophysics Data System (ADS)

Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.; Hamada, K.; Sugimoto, M.; Okuno, K.

2002-12-01

The insulation system proposed by the Japanese Home Team for the ITER Toroidal Field coil (TF coil) is a T-glass-fiber/Kapton reinforced epoxy prepreg system. In order to assess the material performance under the actual operating conditions of the coils, the insulation system was irradiated in the TRIGA reactor (Vienna) to a fast neutron fluence of 2×10 22 m -2 ( E>0.1 MeV). After measurements of swelling, all mechanical tests were carried out at 77 K. Tensile and short-beam-shear (SBS) tests were performed under static loading conditions. In addition, tension-tension fatigue experiments up to about 10 6 cycles were made. The laminate swells in the through-thickness direction by 0.86% at the highest dose level. The fatigue tests as well as the static tests do not show significant influences of the irradiation on the mechanical behavior of this composite.

Evaluation and prediction of long-term environmental effects of nonmetallic materials, second phase

NASA Technical Reports Server (NTRS)

1983-01-01

Changes in the functional properties of a number of nonmetallic materials were evaluated experimentally as a function of simulated space environments and to use such data to develop models for accelerated test methods useful for predicting such behavioral changes. The effects of changed particle irradiations on candidate space materials are evaluated.

Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

DOEpatents

Johnson, A.B. Jr.; Levy, I.S.; Trimble, D.J.; Lanning, D.D.; Gerber, F.S.

1990-04-10

An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-based materials is disclosed. Samples of zirconium-based materials having different compositions and/or fabrication methods are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280 to 316 C). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hydriding for the same materials when subject to in-reactor (irradiated) corrosion. 1 figure.

Effect of helium to dpa ratio on fatigue behavior of austenitic stainless steel irradiated to 2 dpa

NASA Astrophysics Data System (ADS)

Ioka, I.; Yonekawa, M.; Miwa, Y.; Mimura, H.; Tsuji, H.; Hoshiya, T.

2000-12-01

The effect of helium due to nuclear transmutation reactions during neutron irradiation on low cycle fatigue life of type 304 stainless steel was investigated. The specimens were irradiated in spectrally tailored capsules in the Japan Materials Testing Reactor (JMTR) at a temperature of 823 K to a neutron fluence of approximately 1×1025 n/m2 (E>1 MeV) and helium levels of 0.8, 2.5 and 8.1 appm. The low cycle fatigue tests were performed in total axial strain ranges of 0.8-1.6% at 823 K. A laser extensometer was used for controlling the axial strain of a specimen under cyclic testing. The difference between unirradiated and irradiated specimens is quite clear and appears to be a reduction by a factor of 2-5 in fatigue life. The helium concentration of the specimen is not the main factor to shorten fatigue life in the present experimental condition.

Charpy impact toughness of martensitic steels irradiated in FFTF: Effect of heat treatment

NASA Astrophysics Data System (ADS)

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

Charpy tests were made on plates of 9Cr-1MoVNb and 12Cr-1MoVW steels given four different normalizing-and-tempering treatments. One-third-size Charpy specimens from each steel were irradiated to 7.4 - 8 (times) 10(sup 26) n/m(sup 2) (about 34 - 37 dpa) at 420 C in the Materials Open Test Assembly of the Fast Flux Test Facility. Specimens were also thermally aged to 20000 h at 400 C to determine the effect of aging during irradiation. Previous work on these steels irradiated to 4 - 5 dpa at 365 C in MOTA were reexamined in light of the new results. The tests indicated that prior austenite grain size, which was varied by different normalizing treatments, had an effect on impact behavior of the 9Cr-1MoVNb but not on the 12Cr-1MoVW. Tempering treatment had relatively little effect on the shift in DBTT for both steels. Conclusions are presented on how heat treatment can be used to optimize properties.

Irradiation Testing of Ultrasonic Transducers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daw, Joshua; Tittmann, Bernhard; Reinhardt, Brian

2014-07-30

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphologymore » changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. For this reason, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 (E> 0.1 MeV). The goal of this research is to characterize magnetostrictive and piezoelectric transducer survivability during irradiation, enabling the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). As such, this test will be an instrumented lead test and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers.« less

10 CFR 171.16 - Annual fees: Materials licensees, holders of certificates of compliance, holders of sealed source...

Code of Federal Regulations, 2014 CFR

2014-01-01

... material in sealed sources for irradiation of materials in which the source is not removed from its shield... than 10,000 curies of byproduct material in sealed sources for irradiation of materials in which the source is exposed for irradiation purposes. This category also includes underwater irradiators for...

Influence of heat treatment on structural, mechanical and wear properties of crosslinked UHMWPE.

PubMed

Chiesa, R; Moscatelli, M; Giordano, C; Siccardi, F; Cigada, A

2004-01-01

New crosslinked ultra high molecular weight polyethylenes (UHMWPEs) have recently been developed, characterized and introduced in clinical applications. UHMWPE cross-linking treatments are very promising for reducing osteolysis induced by wear debris. The irradiation type, gamma or beta, the dosage and the thermal treatment performed during or following the irradiation process are all factors affecting polyethylene wear resistance. Thermal stabilization treatments performed after or during the irradiation process at a temperature above melting point (i.e. >130 degrees C) have been proven to effectively remove the free radicals generated during irradiation from UHMWPE, but their effect on the mechanical properties of UHMWPE are not completely clear. In addition to wear rate reduction, maintaining good mechanical properties is fundamental aspect in designing the new generation of crosslinked UHMWPE for artificial load bearing materials, especially considering the application in total knee replacements. In this study, we investigated the influence of different stabilization treatments, performed after gamma irradiation, on structural, wear and mechanical properties of UHMWPE. We performed four different stabilization treatments, with different temperatures and cooling rates, on 100 kGy gamma irradiated UHMWPE. Structural properties of UHMWPE were assessed by differential scanning calorimetry (DSC). To assess the mechanical performance of the materials, uni-axial tensile tests were performed according to the ASTM D638 standard, bi-axial tension performance was evaluated by small punch tests (ASTM F2183-02), toughness resistance was evaluated by the Izod method (ASTM F648), and cold flow resistance was analysed by a dynamic compressive test. Evaluation of wear resistance was by a multidirectional pin-on-disk screening machine. Materials considered were in "aged" and "non-aged" conditions. Results confirmed that cross-linking greatly enhances UHMWPE wear resistance, but introduces some detrimental effects on the mechanical properties. In this study, we found that the negative ef-fects on the mechanical properties of crosslinked UHMWPE can be modulated, to some extent, by choosing a thermal stabiliza-tion treatment at a correct temperature and cooling rate. (Journal of Applied Biomaterials & Biomechanics 2004; 2: 20-8).

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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.

Low Activation Joining of SiC/SiC Composites for Fusion Applications: Modeling Miniature Torsion Tests with Elastic and Elastic-Plastic Models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.

2015-06-30

The international fusion community designed miniature torsion specimens for joint testing and irradiation in test reactors with limited irradiation volumes since SiC and SiC-composites used in fission or fusion environments require joining methods for assembling systems. Torsion specimens fail out-of-plane when joints are strong and when elastic moduli are comparable to SiC, which causes difficulties in determining shear strengths for many joints or for comparing unirradiated and irradiated joints. A finite element damage model was developed to treat elastic joints such as SiC/Ti3SiC2+SiC and elastic-plastic joints such as SiC/epoxy and steel/epoxy. The model uses constitutive shear data and is validatedmore » using epoxy joint data. The elastic model indicates fracture is likely to occur within the joined pieces to cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. Lower modulus epoxy joints always fail in plane and provide good model validation.« less

Modernization of existing VVER-1000 surveillance programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kochkin, V.; Erak, D.; Makhotin, D.

2011-07-01

According to generally accepted world practice, evaluation of the reactor pressure vessel (RPV) material behavior during operation is carried out using tests of surveillance specimens. The main objective of the surveillance program consists in insurance of safe RPV operation during the design lifetime and lifetime-extension period. At present, the approaches of pressure vessels residual life validation based on the test results of their surveillance specimens have been developed and introduced in Russia and are under consideration in other countries where vodo-vodyanoi energetichesky reactors- (VVER-) 1000 are in operation. In this case, it is necessary to ensure leading irradiation of surveillancemore » specimens (as compared to the pressure vessel wall) and to provide uniformly irradiated specimen groups for mechanical testing. Standard surveillance program of VVER-1000 has several significant shortcomings and does not meet these requirements. Taking into account program of lifetime extension of VVER-1000 operating in Russia, it is necessary to carry out upgrading of the VVER-1000 surveillance program. This paper studies the conditions of a surveillance specimen's irradiation and upgrading of existing sets to provide monitoring and prognosis of RPV material properties for extension of the reactor's lifetime up to 60 years or more. (authors)« less

Space charge dynamic of irradiated cyanate ester/epoxy at cryogenic temperatures

NASA Astrophysics Data System (ADS)

Wang, Shaohe; Tu, Youping; Fan, Linzhen; Yi, Chengqian; Wu, Zhixiong; Li, Laifeng

2018-03-01

Glass fibre reinforced polymers (GFRPs) have been widely used as one of the main electrical insulating structures for superconducting magnets. A new type of GFRP insulation material using cyanate ester/epoxy resin as a matrix was developed in this study, and the samples were irradiated by Co-60 for 1 MGy and 5 MGy dose. Space charge distributed within the sample were tested using the pulsed electroacoustic method, and charge concentration was found at the interfaces between glass fibre and epoxy resin. Thermally stimulated current (TSC) and dc conduction current were also tested to evaluate the irradiation effect. It was supposed that charge mobility and density were suppressed at the beginning due to the crosslinking reaction, and for a higher irradiation dose, molecular chain degradation dominated and led to more sever space charge accumulation at interfaces which enhance the internal electric field higher than the external field, and transition field for conduction current was also decreased by irradiation. Space charge dynamic at cryogenic temperature was revealed by conduction current and TSC, and space charge injection was observed for the irradiated samples at 225 K, which was more obvious for the irradiated samples.

In-pile Hydrothermal Corrosion Evaluation of Coated SiC Ceramics and Composites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carpenter, David; Ang, Caen; Katoh, Yutai

2017-09-01

Hydrothermal corrosion accelerated by water radiolysis during normal operation is among the most critical technical feasibility issues remaining for silicon carbide (SiC) composite-based cladding that could provide enhanced accident-tolerance fuel technology for light water reactors. An integrated in-pile test was developed and performed to determine the synergistic effects of neutron irradiation, radiolysis, and pressurized water flow, all of which are relevant to a typical pressurized water reactor (PWR). The test specimens were chosen to cover a range of SiC materials and a variety of potential options for environmental barrier coatings. This document provides a summary of the irradiation vehicle design,more » operations of the experiment, and the specimen loading into the irradiation vehicle.« less

Gamma irradiation assisted fungal degradation of the polypropylene/biomass composites

NASA Astrophysics Data System (ADS)

Butnaru, Elena; Darie-Niţă, Raluca Nicoleta; Zaharescu, Traian; Balaeş, Tiberius; Tănase, Cătălin; Hitruc, Gabriela; Doroftei, Florica; Vasile, Cornelia

2016-08-01

White-rot fungus Bjerkandera adusta has been tested for its ability to degrade some biocomposites materials based on polypropylene and biomass (Eucalyptus globulus, pine cones, and Brassica rapa). γ-irradiation was applied to initiate the degradation of relatively inert polypropylene matrix. The degradation process has been studied by scanning electron microscopy, atomic force microscopy, infrared spectroscopy, contact angle measurements, rheological and chemiluminescence tests. These analyses showed that the polypropylene/biomass composites properties are worsen under the action of the selected microorganism. The formation of cracks and scrap particles over the entire matrix surface and the decrease of the complex viscosity values, as well as the dynamic moduli of gamma irradiated PP/biomass composite and exposed to Bjerkandera adusta fungus, indicate fungal efficiency in composite degradation.

Corrosion resistance investigation of vanadium alloys in liquid lithium

NASA Astrophysics Data System (ADS)

Borovitskaya, I. V.; Lyublinskiy, I. E.; Bondarenko, G. G.; Paramonova, V. V.; Korshunov, S. N.; Mansurova, A. N.; Lyakhovitskiy, M. M.; Zharkov, M. Yu.

2016-12-01

A major concern in using vanadium alloys for first wall/blanket systems in fusion reactors is their activity with regard to nonmetallic impurities in the coolants. This paper presents the results of studying the corrosion resistance in high-purity liquid lithium (with the nitrogen and carbon content of less than 10-3 wt %) of vanadium and vanadium alloys (V-1.86Ga, V-3.4Ga-0.62Si, V-4.81Ti-4.82Cr) both in the initial state and preliminarily irradiated with Ar+ ions with energy of 20 keV to a dose of 1022 m-2 at an irradiation temperature of 400°C. The degree of corrosion was estimated by measuring the changes in the weight and microhardness. Corrosion tests were carried out under static isothermal conditions at a temperature of 600°C for 400 h. The identity of corrosion mechanisms of materials both irradiated with Ar ions and not irradiated, which consisted in an insignificant penetration of nitrogen into the materials and a substantial escape of oxygen from the materials, causing the formation of a zone with a reduced microhardness near the surface, was established. The influence of the corrosive action of lithium on the surface morphology of the materials under study was found, resulting in the manifestation of grain boundaries and slip lines on the sample surface, the latter being most clearly observed in the case of preliminary irradiation with Ar ions.

Corrosion resistance investigation of vanadium alloys in liquid lithium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borovitskaya, I. V., E-mail: symp@imet.ac.ru; Lyublinskiy, I. E.; Bondarenko, G. G.

A major concern in using vanadium alloys for first wall/blanket systems in fusion reactors is their activity with regard to nonmetallic impurities in the coolants. This paper presents the results of studying the corrosion resistance in high-purity liquid lithium (with the nitrogen and carbon content of less than 10{sup –3} wt %) of vanadium and vanadium alloys (V–1.86Ga, V–3.4Ga–0.62Si, V–4.81Ti–4.82Cr) both in the initial state and preliminarily irradiated with Ar+ ions with energy of 20 keV to a dose of 10{sup 22} m{sup –2} at an irradiation temperature of ~400°C. The degree of corrosion was estimated by measuring the changesmore » in the weight and microhardness. Corrosion tests were carried out under static isothermal conditions at a temperature of 600°C for 400 h. The identity of corrosion mechanisms of materials both irradiated with Ar ions and not irradiated, which consisted in an insignificant penetration of nitrogen into the materials and a substantial escape of oxygen from the materials, causing the formation of a zone with a reduced microhardness near the surface, was established. The influence of the corrosive action of lithium on the surface morphology of the materials under study was found, resulting in the manifestation of grain boundaries and slip lines on the sample surface, the latter being most clearly observed in the case of preliminary irradiation with Ar ions.« less

21 CFR 179.45 - Packaging materials for use during the irradiation of prepackaged foods.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 21 Food and Drugs 3 2012-04-01 2012-04-01 false Packaging materials for use during the irradiation... OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION (CONTINUED) IRRADIATION IN THE... materials for use during the irradiation of prepackaged foods. The packaging materials identified in this...

21 CFR 179.45 - Packaging materials for use during the irradiation of prepackaged foods.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 21 Food and Drugs 3 2010-04-01 2009-04-01 true Packaging materials for use during the irradiation... OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION (CONTINUED) IRRADIATION IN THE... materials for use during the irradiation of prepackaged foods. The packaging materials identified in this...

21 CFR 179.45 - Packaging materials for use during the irradiation of prepackaged foods.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 21 Food and Drugs 3 2011-04-01 2011-04-01 false Packaging materials for use during the irradiation... OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION (CONTINUED) IRRADIATION IN THE... materials for use during the irradiation of prepackaged foods. The packaging materials identified in this...

21 CFR 179.45 - Packaging materials for use during the irradiation of prepackaged foods.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 21 Food and Drugs 3 2013-04-01 2013-04-01 false Packaging materials for use during the irradiation... OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION (CONTINUED) IRRADIATION IN THE... materials for use during the irradiation of prepackaged foods. The packaging materials identified in this...

High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donna Post Guillen; Larry R. Greenwood; James R. Parry

2014-06-22

A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluencemore » monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.« less

Simulated Aging of Spacecraft External Materials on Orbit

NASA Astrophysics Data System (ADS)

Khatipov, S.

Moscow State Engineering Physics Institute (MIFI), in cooperation with Air Force Research Laboratory's Satellite Assessment Center (SatAC), the European Office of Aerospace Research and Development (EOARD), and the International Science and Technology Center (ISTC), has developed a database describing the changes in optical properties of materials used on the external surfaces of spacecraft due to space environmental factors. The database includes data acquired from tests completed under contract with the ISTC and EOARD, as well as from previous Russian materials studies conducted within the last 30 years. The space environmental factors studied are for those found in Low Earth Orbits (LEO) and Geosynchronous orbits (GEO), including electron irradiation at 50, 100, and 200 keV, proton irradiation at 50, 150, 300, and 500 keV, and ultraviolet irradiation equivalent to 1 sun-year. The material characteristics investigated were solar absorption (aS), spectral reflectance (rl), solar reflectance (rS), emissivity (e), spectral transmission coefficient (Tl), solar transmittance (TS), optical density (D), relative optical density (D/x), Bi-directional Reflectance Distribution Function (BRDF), and change of appearance and color in the visible wavelengths. The materials tested in the project were thermal control coatings (paints), multilayer insulation (films), and solar cells. The ability to predict changes in optical properties of spacecraft materials is important to increase the fidelity of space observation tools, better understand observation of space objects, and increase the longevity of spacecraft. The end goal of our project is to build semi-empirical mathematical models to predict the long-term effects of space aging as a function of time and orbit.

Installation and first operation of the International Fusion Materials Irradiation Facility injector at the Rokkasho site

NASA Astrophysics Data System (ADS)

Gobin, Raphael; Bogard, Daniel; Bolzon, Benoit; Bourdelle, Gilles; Chauvin, Nicolas; Chel, Stéphane; Girardot, Patrick; Gomes, Adelino; Guiho, Patrice; Harrault, Francis; Loiseau, Denis; Lussignol, Yves; Misiara, Nicolas; Roger, Arnaud; Senée, Franck; Valette, Matthieu; Cara, Philippe; Duglué, Daniel; Gex, Dominique; Okumura, Yoshikazu; Marcos Ayala, Juan; Knaster, Juan; Marqueta, Alvaro; Kasugai, Atsushi; O'Hira, Shigeru; Shinto, Katsuhiro; Takahashi, Hiroki

2016-02-01

The International Fusion Materials Irradiation Facility (IFMIF) linear IFMIF prototype accelerator injector dedicated to high intensity deuteron beam production has been designed, built, and tested at CEA/Saclay between 2008 and 2012. After the completion of the acceptance tests at Saclay, the injector has been fully sent to Japan. The re-assembly of the injector has been performed between March and May 2014. Then after the check-out phase, the production of the first proton beam occurred in November 2014. Hydrogen and deuteron beam commissioning is now in progress after having proceeded with the final tests on the entire injector equipment including high power diagnostics. This article reports the different phases of the injector installation pointing out the safety and security needs, as well as the first beam production results in Japan and chopper tests. Detailed operation and commissioning results (with H+ and D+ 100 keV beams) are reported in a second article.

TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR

DOEpatents

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17

Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

Development of optimum process for electron beam cross-linking of high density polyethylene thermal energy storage pellets, process scale-up and production of application qualities of material

NASA Technical Reports Server (NTRS)

Salyer, I. O.

1980-01-01

The electron irradiation conditions required to prepare thermally from stable high density polyethylene (HDPE) were defined. The conditions were defined by evaluating the heat of fusion and the melting temperature of several HDPE specimens. The performance tests conducted on the specimens, including the thermal cycling tests in the thermal energy storage unit are described. The electron beam irradiation tests performed on the specimens, in which the total radiation dose received by the pellets, the electron beam current, the accelerating potential, and the atmospheres were varied, are discussed.

The role of niobium carbide in radiation induced segregation behaviour of type 347 austenitic stainless steel

NASA Astrophysics Data System (ADS)

Ahmedabadi, Parag; Kain, Vivekanand; Gupta, Manu; Samajdar, I.; Sharma, S. C.; Bhagwat, P.; Chowdhury, R.

2011-08-01

The effect of niobium carbide precipitates on radiation induced segregation (RIS) behaviour in type 347 stainless steel was investigated. The material in the as-received condition was irradiated using double-loop 4.8 MeV protons at 300 °C for 0.43 dpa (displacement per atom). The RIS in the proton irradiated specimen was characterized using double-loop electrochemical potentiokinetic reactivation (DL-EPR) test followed by atomic force microscopic examination. The nature of variation of DL-EPR values with the depth matched with the variation of the calculated irradiation damage (dpa) with the depth. The attack on grain boundaries during EPR tests was negligible indicating absence of chromium depletion zones. The interface between niobium carbide and the matrix acts as a sink for point defects generated during irradiation and this had reduced point defect flux toward grain boundaries. The attack was noticed at a few large cluster of niobium carbide after the DL-EPR test at the depth of maximum attack for the irradiated specimen. Pit-like features were not observed within the matrix indicating the absence of chromium depletion regions within the matrix.

Influence of light and oxygen on the color stability of five calcium silicate-based materials.

PubMed

Vallés, Marta; Mercadé, Montse; Duran-Sindreu, Fernando; Bourdelande, Jose L; Roig, Miguel

2013-04-01

Difficult handling, long setting time, and potential discoloration are important drawbacks of white mineral trioxide aggregate (WMTA). The development of Biodentine, a recently developed calcium silicate-based material (CSM), has overcome some of these shortcomings; however, there are no available data on its color stability. A previous study showed that WMTA discolors under light irradiation in an oxygen-free environment. The present study evaluated the influence of light irradiation and oxygen on the color stability of 5 CSMs. Fifteen samples of 5 CSMs (ProRoot WMTA, Angelus WMTA, White Portland Cement [PC], PC with bismuth oxide, and Biodentine) were divided into 5 groups. Each group was exposed to different oxygen and light conditions. A spectrophotometer was used to determine the color of each specimen at 0, 120 seconds, and 5 days. Data were analyzed by using analysis of variance and Tukey honestly significant difference test. The materials PC with bismuth oxide, Angelus WMTA, and ProRoot WMTA showed dark discoloration after light irradiation in an oxygen-free environment, which was statistically significantly different from Biodentine and PC. In groups that were exposed to no light irradiation or to an oxygen atmosphere, all materials showed color stability over time, and no significant differences were observed among them. PC and Biodentine maintained color stability in all conditions over time and showed no significant differences. The combination of light and anaerobic conditions (similar to those in clinical situations) results in differences in color of the tested CSMs during a period of 5 days, of which Biodentine and PC demonstrated color stability. Copyright © 2013 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

Microscale mechanical characterization of materials for extreme environments

NASA Astrophysics Data System (ADS)

Ozerinc, Sezer

Nanocrystalline metals are promising materials for applications that require outstanding strength and stability in extreme environments. Further improvements in the desirable mechanical properties of these materials require a better understanding of the relationship between their microstructure and grain boundary deformation behavior. Previous molecular dynamics simulations suggested that solute additions to grain boundaries can enhance the strength of nanocrystalline metals, but there has been a lack of experimental studies investigating this prediction. This dissertation presents mechanical and microstructural characterization of nanocrystalline Cu alloys and demonstrate that addition of Nb solutes to grain boundaries greatly enhances the strength of Cu. The measured hardness of Cu90Nb10 alloy is 5.6 GPa which is more than double the hardness of nanocrystalline pure Cu. Microstructural characterization through transmission electron microscopy and energy-dispersive X-ray spectroscopy on these alloys indicates a strong correlation between the grain boundary composition and the hardness. Variation of measured hardness with measured grain boundary composition is in very good agreement with previous molecular dynamics simulation predictions. The results of this work provide experimental evidence that grain boundary doping enhances the strength of nanocrystalline Cu far beyond that predicted by classical Hall-Petch strengthening and decreasing grain boundary energy through solute additions is the key to reaching theoretical strength in nanocrystalline metals. Irradiation induced creep is a deformation mechanism that takes place under combined stress and particle bombardment. Effective characterization of this phenomenon on nanostructured materials is crucial for the assessment of their potential use in next generation nuclear power plants. Direct measurements of irradiation induced creep under MeV-heavy ion bombardment have not been feasible until recently due to the requirements of micron-sized specimens, muN-level force sensitivity, and nm-level displacement sensitivity. A recently developed mechanical characterization technique, micropillar compression, has enabled the testing of miniaturized specimens; however, there has been no demonstration of the application of this technique to irradiation induced creep measurements. This dissertation presents the development of an in situ measurement apparatus for compression testing of micron-sized cylindrical specimens under MeV-heavy ion bombardment. The apparatus has a force resolution of 1 muN and a displacement resolution of 1 nm. The apparatus measured irradiation induced creep in four different amorphous materials and the findings clarified the significance of different creep mechanisms in these materials. In amorphous metals and amorphous Si, the measured irradiation induced fluidity is ≈ 3 dpa-1GPa-1 (dpa: displacements per atom). The measured fluidity is in excellent agreement with previous molecular dynamics simulation predictions, providing experimental evidence for point defect mediated plastic flow under ion bombardment. For amorphous SiO2, stress relaxation through thermal spikes further contribute to the creep response, resulting in higher fluidities up to ≈ 83 dpa-1GPa -1. Finally, this dissertation presents the further development of the creep testing apparatus for high temperature measurements. The apparatus demonstrated good thermal and mechanical stability and measured irradiation induced creep of nanocrystalline Cu at 200°C. Resulting irradiation induced fluidity is ≈ 10% of the fluidity of the amorphous metals, in agreement with previous measurements on free-standing films. Understanding the creep behavior of nanostructured metals under heavy ion bombardment at elevated temperatures is important for identifying the governing creep mechanisms in these materials. The developed apparatus provides a new and effective method of accelerated mechanical characterization of such promising materials for their potential use in future nuclear applications.

Report on Status of Shipment of High Fluence Austenitic Steel Samples for Characterization and Stress Corrosion Crack Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clark, Scarlett R.; Leonard, Keith J.

The goal of the Mechanisms of Irradiation Assisted Stress Corrosion Cracking (IASCC) task in the LWRS Program is to conduct experimental research into understanding how multiple variables influence the crack initiation and crack growth in materials subjected to stress under corrosive conditions. This includes understanding the influences of alloy composition, radiation condition, water chemistry and metallurgical starting condition (i.e., previous cold work or heat treatments and the resulting microstructure) has on the behavior of materials. Testing involves crack initiation and growth testing on irradiated specimens of single-variable alloys in simulated Light Water Reactor (LWR) environments, tensile testing, hardness testing, microstructuralmore » and microchemical analysis, and detailed efforts to characterize localized deformation. Combined, these single-variable experiments will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long-term, design IASCC-resistant materials. In support of this research, efforts are currently underway to arrange shipment of “free” high fluence austenitic alloys available through Électricité de France (EDF) for post irradiation testing at the Oak Ridge National Laboratory (ORNL) and IASCC testing at the University of Michigan. These high fluence materials range in damage values from 45 to 125 displacements per atom (dpa). The samples identified for transport to the United States, which include nine, no-cost, 304, 308 and 316 tensile bars, were relocated from the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad, Ulyanovsk Oblast, Russia, and received at the Halden Reactor in Halden, Norway, on August 23, 2016. ORNL has been notified that a significant amount of work is required to prepare the samples for further shipment to Oak Ridge, Tennessee. The preliminary work for sample shipment between Halden and Oak Ridge includes fabrication of an inner cask sample container, decontamination and preparation of a Type A container, preparation of new activity calculations, all necessary paperwork, and handling. ORNL will continue to work to track progress of sample preparation and shipment status, and to work toward an agreement that covers material shipping costs between the Halden Reactor and the Oak Ridge National Laboratory.« less

Methodology for Mechanical Property Testing of Fuel Cladding Using a Expanded Plug Wedge Test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John

2014-01-01

An expanded plug method was developed earlier for determining the tensile properties of irradiated fuel cladding. This method tests fuel rod cladding ductility by utilizing an expandable plug to radially stretch a small ring of irradiated cladding material. The circumferential or hoop strain is determined from the measured diametrical expansion of the ring. A developed procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves, from which material properties of the cladding can be extracted. However, several deficiencies existed in this expanded-plug test that can impact the accuracy of test results, suchmore » as that the large axial compressive stress resulted from the expansion plug test can potentially induce the shear failure mode of the tested specimen. Moreover, highly nonuniform stress and strain distribution in the deformed clad gage section and significant compressive stresses, induced by bending deformation due to clad bulging effect, will further result in highly nonconservative estimates of the mechanical properties for both strength and ductility of the tested clad. To overcome the aforementioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. By optimizing the specific geometry designs, selecting the appropriate material for the expansion plug, and adding new components into the testing system, a modified expansion plug testing protocol has been developed. A general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor, -factor, was used to convert the ring load Fring into hoop stress , and is written as _ = F_ring/tl , where t is the clad thickness and l is the clad length. The generated stress-strain curve agrees well with the associated tensile test data in both elastic and plastic deformation regions.« less

Does laser diode irradiation improve the degree of conversion of simplified dentin bonding systems?

PubMed Central

BRIANEZZI, Leticia Ferreira de Freitas; MAENOSONO, Rafael Massunari; BIM, Odair; ZABEU, Giovanna Speranza; PALMA-DIBB, Regina Guenka; ISHIKIRIAMA, Sérgio Kiyoshi

2017-01-01

Abstract Simplified dentin-bonding systems are clinically employed for most adhesive procedures, and they are prone to hydrolytic degradation. Objective This study aimed to investigate the effect of laser diode irradiation on the degree of conversion (DC), water sorption (WS), and water solubility (WSB) of these bonding systems in an attempt to improve their physico-mechanical resistance. Material and Methods Two bonding agents were tested: a two-step total-etch system [Adper™ Single Bond 2, 3M ESPE (SB)] and a universal system [Adper™ Single Bond Universal, 3M ESPE (SU)]. Square-shaped specimens were prepared and assigned into 4 groups (n=5): SB and SU (control groups – no laser irradiation) and SB-L and SU-L [SB and SU laser (L) – irradiated groups]. DC was assessed using Fourier transform infrared spectroscopy with attenuated total reflectance. Additional uncured resin samples (≈3.0 µL, n=5) of each adhesive were also scanned for final DC calculation. For WS/WSB tests, similar specimens (n=10) were prepared and measured by monitoring the mass changes after dehydration/water storage cycles. For both tests, adhesive fluids were dropped into standardized Teflon molds (6.0×6.0×1.0 mm), irradiated with a 970-nm laser diode, and then polymerized with an LED-curing unit (1 W/cm2). Results Laser irradiation immediately before photopolymerization increased the DC (%) of the tested adhesives: SB-L>SB>SU-L>SU. For WS/WSB (μg/mm3), only the dentin bonding system (DBS) was a significant factor (p<0.05): SB>SU. Conclusion Irradiation with a laser diode improved the degree of conversion of all tested simplified dentin bonding systems, with no impact on water sorption and solubility. PMID:28877276

Energy determination in industrial X-ray processing facilities

NASA Astrophysics Data System (ADS)

Cleland, M. R.; Gregoire, O.; Stichelbaut, F.; Gomola, I.; Galloway, R. A.; Schlecht, J.

2005-12-01

In industrial irradiation facilities, the determination of maximum photon or electron energy is important for regulated processes, such as food irradiation, and for assurance of treatment reproducibility. With electron beam irradiators, this has been done by measuring the depth-dose distribution in a homogeneous material. For X-ray irradiators, an analogous method has not yet been recommended. This paper describes a procedure suitable for typical industrial irradiation processes, which is based on common practice in the field of therapeutic X-ray treatment. It utilizes a measurement of the slope of the exponential attenuation curve of X-rays in a thick stack of polyethylene plates. Monte Carlo simulations and experimental tests have been performed to verify the suitability and accuracy of the method between 3 MeV and 8 MeV.

Packaging food for radiation processing

NASA Astrophysics Data System (ADS)

Komolprasert, Vanee

2016-12-01

Irradiation can play an important role in reducing pathogens that cause food borne illness. Food processors and food safety experts prefer that food be irradiated after packaging to prevent post-irradiation contamination. Food irradiation has been studied for the last century. However, the implementation of irradiation on prepackaged food still faces challenges on how to assess the suitability and safety of these packaging materials used during irradiation. Irradiation is known to induce chemical changes to the food packaging materials resulting in the formation of breakdown products, so called radiolysis products (RP), which may migrate into foods and affect the safety of the irradiated foods. Therefore, the safety of the food packaging material (both polymers and adjuvants) must be determined to ensure safety of irradiated packaged food. Evaluating the safety of food packaging materials presents technical challenges because of the range of possible chemicals generated by ionizing radiation. These challenges and the U.S. regulations on food irradiation are discussed in this article.

Final Report on MEGAPIE Target Irradiation and Post-Irradiation Examination

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yong, Dai

2015-06-30

Megawatt pilot experiment (MEGAPIE) was successfully performed in 2006. One of the important goals of MEGAPIE is to understand the behaviour of structural materials of the target components exposed to high fluxes of high-energy protons and spallation neutrons in flowing LBE (liquid lead-bismuth eutectic) environment by conducting post-irradiation examination (PIE). The PIE includes four major parts: non-destructive test, radiochemical analysis of production and distribution of radionuclides produced by spallation reaction in LBE, analysis of LBE corrosion effects on structural materials, T91 and SS 316L steels, and mechanical testing of the T91 and SS 316L steels irradiated in the lower partmore » of the target. The non-destructive test (NDT) including visual inspection and ultrasonic measurement was performed in the proton beam window area of the T91 calotte of the LBE container, the most intensively irradiated part of the MEGAPIE target. The visual inspection showed no visible failure and the ultrasonic measurement demonstrated no detectable change in thickness in the beam window area. Gamma mapping was also performed in the proton beam window area of the AlMg 3 safety-container. The gamma mapping results were used to evaluate the accumulated proton fluence distribution profile, the input data for determining irradiation parameters. Radiochemical analysis of radionuclides produced by spallation reaction in LBE is to improve the understanding of the production and distribution of radionuclides in the target. The results demonstrate that the radionuclides of noble metals, 207Bi, 194Hg/Au are rather homogeneously distributed within the target, while radionuclides of electropositive elements are found to be deposited on the steel-LBE interface. The corrosion effect of LBE on the structural components under intensive irradiation was investigated by metallography. The results show that no evident corrosion damages. However, unexpected deep cracks were found in the EBW (electron beam weld) of the LBE container in the intensive irradiation zone of the target, which should be formed during irradiation. In the SS 316L steel of the flow guide tube, inclusions or precipitates enriched with O, Si, S, Ca, Ti and Mn were observed. Many of them are very long, up to a few mm, and located on grain boundaries along the extrusion direction of the tube. The degradation of the mechanical properties of the T91 and SS 316L steels has been investigated by conducting tensile tests on the specimens extracted from the T91 and SS 316L components in the intensive irradiation region. The results obtained from the proton beam window of the T91 calotte exhibit a good ductility of T91 steel after irradiation at 6-7 dpa (displacement per atom) in contact with flowing LBE.« less

Evaluation of Cooling Conditions for a High Heat Flux Testing Facility Based on Plasma-Arc Lamps

DOE PAGES

Charry, Carlos H.; Abdel-khalik, Said I.; Yoda, Minami; ...

2015-07-31

The new Irradiated Material Target Station (IMTS) facility for fusion materials at Oak Ridge National Laboratory (ORNL) uses an infrared plasma-arc lamp (PAL) to deliver incident heat fluxes as high as 27 MW/m 2. The facility is being used to test irradiated plasma-facing component materials as part of the joint US-Japan PHENIX program. The irradiated samples are to be mounted on molybdenum sample holders attached to a water-cooled copper rod. Depending on the size and geometry of samples, several sample holders and copper rod configurations have been fabricated and tested. As a part of the effort to design sample holdersmore » compatible with the high heat flux (HHF) testing to be conducted at the IMTS facility, numerical simulations have been performed for two different water-cooled sample holder designs using the ANSYS FLUENT 14.0 commercial computational fluid dynamics (CFD) software package. The primary objective of this work is to evaluate the cooling capability of different sample holder designs, i.e. to estimate their maximum allowable incident heat flux values. 2D axisymmetric numerical simulations are performed using the realizable k-ε turbulence model and the RPI nucleate boiling model within ANSYS FLUENT 14.0. The results of the numerical model were compared against the experimental data for two sample holder designs tested in the IMTS facility. The model has been used to parametrically evaluate the effect of various operational parameters on the predicted temperature distributions. The results were used to identify the limiting parameter for safe operation of the two sample holders and the associated peak heat flux limits. The results of this investigation will help guide the development of new sample holder designs.« less

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)« less

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

NASA Astrophysics Data System (ADS)

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

2007-07-01

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

Arcjet Testing of Micro-Meteoroid Impacted Thermal Protection Materials

NASA Technical Reports Server (NTRS)

Agrawal, Parul; Munk, Michelle M.; Glaab, Louis J.

2013-01-01

There are several harsh space environments that could affect thermal protection systems and in turn pose risks to the atmospheric entry vehicles. These environments include micrometeoroid impact, extreme cold temperatures, and ionizing radiation during deep space cruise, all followed by atmospheric entry heating. To mitigate these risks, different thermal protection material samples were subjected to multiple tests, including hyper velocity impact, cold soak, irradiation, and arcjet testing, at various NASA facilities that simulated these environments. The materials included a variety of honeycomb packed ablative materials as well as carbon-based non-ablative thermal protection systems. The present paper describes the results of the multiple test campaign with a focus on arcjet testing of thermal protection materials. The tests showed promising results for ablative materials. However, the carbon-based non-ablative system presented some concerns regarding the potential risks to an entry vehicle. This study provides valuable information regarding the capability of various thermal protection materials to withstand harsh space environments, which is critical to sample return and planetary entry missions.

Ignitability analysis using the cone calorimeter and lift apparatus

Treesearch

Mark A. Dietenberger

1996-01-01

The irradiance plotted as function of time to ignition for wood materials tested in the Cone Calorimeter (ASTM E1354) differs signiticantly from that tested in the Lateral Ignition and Flame spread Test (LIFT) apparatus (ASTM E1321). This difference in piloted ignitabilty is primarily due to the difference in forced convective cooling of the specimen tested in both...

Fading test using the SAAD-POSL method for retrospective accidental dosimetry of building materials

NASA Astrophysics Data System (ADS)

Kim, M. J.; Lee, Y. J.; Lee, J. I.; Kim, J. L.; Hong, D. G.

2015-11-01

Fading test using the single aliquot additive dose method with pulsed optically stimulated luminescence (SAAD-POSL method) was applied to core-disc samples extracted from heated red brick, tile, roof-tile, and toilet porcelain after X-ray and beta irradiation. From thermoluminescence measurements of each material, the optimal preheat condition of the SAAD-POSL method was first determined as 170 °C for 10 s. Fading characteristics of core-disc samples of heated red brick obtained using the SAAD-POSL method were similar to those of quartz grains (90-250 μm) obtained using the SAR-OSL method, regardless of the differences in the sample and radiation type. Fading evaluations of the core-disc samples of these building materials two weeks after irradiation showed that the equivalent dose (ED) decreased between 5% and 42%. The results indicate that the fading characteristics will be able to contribute to a more accurate estimation of the ED value using the SAAD-POSL method.

Non-destructive diagnostics of irradiated materials using neutron scattering from pulsed neutron sources

NASA Astrophysics Data System (ADS)

Korenev, Sergey; Sikolenko, Vadim

2004-09-01

The advantage of neutron-scattering studies as compared to the standard X-ray technique is the high penetration of neutrons that allow us to study volume effects. The high resolution of instrumentation on the basis neutron scattering allows measurement of the parameters of lattice structure with high precision. We suggest the use of neutron scattering from pulsed neutron sources for analysis of materials irradiated with pulsed high current electron and ion beams. The results of preliminary tests using this method for Ni foils that have been studied by neutron diffraction at the IBR-2 (Pulsed Fast Reactor at Joint Institute for Nuclear Research) are presented.

Determination of He and D permeability of neutron-irradiated SiC tubes to examine the potential for release due to micro-cracking

DOE Office of Scientific and Technical Information (OSTI.GOV)

Katoh, Yutai; Hu, Xunxiang; Koyanagi, Takaaki

Driven by the need to enlarge the safety margins of light water reactors in both design-basis and beyond-design-basis accident scenarios, the research and development of accident-tolerant fuel (ATF) has become an importance topic in the nuclear engineering and materials community. Continuous SiC fiber-reinforced SiC matrix ceramic composites are under consideration as a replacement for traditional zirconium alloy cladding owing to their high-temperature stability, chemical inertness, and exceptional irradiation resistance. Among the key technical feasibility issues, potential failure of the fission product containment due to probabilistic penetrating cracking has been identified as one of the two most critical feasibility issues, togethermore » with the radiolysisassisted hydrothermal corrosion of SiC. The experimental capability to evaluate the hermeticity of SiC-based claddings is an urgent need. In this report, we present the development of a comprehensive permeation testing station established in the Low Activation Materials Development and Analysis laboratory at Oak Ridge National Laboratory. Preliminary results for the hermeticity evaluation of un-irradiated monolithic SiC tubes, uncoated and coated SiC/SiC composite tubes, and neutron-irradiated monolithic SiC tubes at room temperature are exhibited. The results indicate that this new permeation testing station is capable of evaluating the hermeticity of SiC-based tubes by determining the helium and deuterium permeation flux as a function of gas pressure at a high resolution of 8.07 x 10 -12 atm-cc/s for helium and 2.83 x 10 -12 atm-cc/s for deuterium, respectively. The detection limit of this system is sufficient to evaluate the maximum allowable helium leakage rate of lab-scale tubular samples, which is linearly extrapolated from the evaluation standard used for a commercial as-manufactured light water reactor fuel rod at room temperature. The un-irradiated monolithic SiC tube is hermetic, as is manifested by the un-detectable deuterium permeation flux at various feeding gas pressures. A large helium leakage rate was detected for the uncoated SiC/SiC composite tube exposed to atmosphere, indicating it is inherently not hermetic. The hermeticity of coated SiC/SiC composite tubes is strongly dependent on the coating materials and the preparation of the substrate SiC/SiC composite samples. To simulate the practical application environment, monolithic CVD SiC tubes were exposed to neutron irradiation at the High Flux Isotope Reactor under high heat flux from the internal surface to the external surface. Although finite element analysis and resonant ultrasound spectroscopy measurement indicated that the combined neutron irradiation and high heat flux gave rise to a high probability of cracking within the sample, the hermeticity evaluation of the tested sample still exhibited gas tightness, emphasizing that SiC cracking is inherently a statistical phenomenon. The developed permeation testing station is capable of measuring the gas permeation flux in the range of interest with full confidence based on the presented results. It is considered a critical pre- /post-irradiation examination technique to characterize SiC-based cladding materials in asreceived and irradiated states to aid the research and development of ATF.« less

In situ micro-compression testing of He2+ ion irradiated titanium aluminide

NASA Astrophysics Data System (ADS)

Wei, Tao; Xu, Alan; Zhu, Hanliang; Ionescu, Mihail; Bhattacharyya, Dhriti

2017-10-01

A titanium aluminide (TiAl) alloy 45XD has been irradiated by a He ion beam with an energy of 5 MeV on a tandem accelerator at the Australian Nuclear Science and Technology Organization (ANSTO). The total fluence of He ions was 5 × 1017 ion cm-2. A 17 μm uniform damage region from the material surface with a helium concentration of about 5000 appm was achieved by using an energy degrading wheel in front of the TiAl target. The micro-size test specimens from the damage layer were fabricated using a focused ion beam & scanning electron microscope (FIB-SEM) system. The in situ SEM micromechanical compressive testing was carried out inside an SEM and the results indicated irradiation embrittlement in the helium affected region. Electron back scatter diffraction (EBSD) analysis has been applied to reveal the orientation of the lamellae in the TiAl specimens, and used to understand the deformation processes in the sample. The irradiation damage of gallium ion beam from FIB on the surface of TiAl sample was also investigated.

Gap Size Uncertainty Quantification in Advanced Gas Reactor TRISO Fuel Irradiation Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pham, Binh T.; Einerson, Jeffrey J.; Hawkes, Grant L.

The Advanced Gas Reactor (AGR)-3/4 experiment is the combination of the third and fourth tests conducted within the tristructural isotropic fuel development and qualification research program. The AGR-3/4 test consists of twelve independent capsules containing a fuel stack in the center surrounded by three graphite cylinders and shrouded by a stainless steel shell. This capsule design enables temperature control of both the fuel and the graphite rings by varying the neon/helium gas mixture flowing through the four resulting gaps. Knowledge of fuel and graphite temperatures is crucial for establishing the functional relationship between fission product release and irradiation thermal conditions.more » These temperatures are predicted for each capsule using the commercial finite-element heat transfer code ABAQUS. Uncertainty quantification reveals that the gap size uncertainties are among the dominant factors contributing to predicted temperature uncertainty due to high input sensitivity and uncertainty. Gap size uncertainty originates from the fact that all gap sizes vary with time due to dimensional changes of the fuel compacts and three graphite rings caused by extended exposure to high temperatures and fast neutron irradiation. Gap sizes are estimated using as-fabricated dimensional measurements at the start of irradiation and post irradiation examination dimensional measurements at the end of irradiation. Uncertainties in these measurements provide a basis for quantifying gap size uncertainty. However, lack of gap size measurements during irradiation and lack of knowledge about the dimension change rates lead to gap size modeling assumptions, which could increase gap size uncertainty. In addition, the dimensional measurements are performed at room temperature, and must be corrected to account for thermal expansion of the materials at high irradiation temperatures. Uncertainty in the thermal expansion coefficients for the graphite materials used in the AGR-3/4 capsules also increases gap size uncertainty. This study focuses on analysis of modeling assumptions and uncertainty sources to evaluate their impacts on the gap size uncertainty.« less

Irradiation Creep in Graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13

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

Irradiated recycled plastic as a concrete additive for improved chemo-mechanical properties and lower carbon footprint.

PubMed

Schaefer, Carolyn E; Kupwade-Patil, Kunal; Ortega, Michael; Soriano, Carmen; Büyüköztürk, Oral; White, Anne E; Short, Michael P

2018-01-01

Concrete production contributes heavily to greenhouse gas emissions, thus a need exists for the development of durable and sustainable concrete with a lower carbon footprint. This can be achieved when cement is partially replaced with another material, such as waste plastic, though normally with a tradeoff in compressive strength. This study discusses progress toward a high/medium strength concrete with a dense, cementitious matrix that contains an irradiated plastic additive, recovering the compressive strength while displacing concrete with waste materials to reduce greenhouse gas generation. Compressive strength tests showed that the addition of high dose (100kGy) irradiated plastic in multiple concretes resulted in increased compressive strength as compared to samples containing regular, non-irradiated plastic. This suggests that irradiating plastic at a high dose is a viable potential solution for regaining some of the strength that is lost when plastic is added to cement paste. X-ray Diffraction (XRD), Backscattered Electron Microscopy (BSE), and X-ray microtomography explain the mechanisms for strength retention when using irradiated plastic as a filler for cement paste. By partially replacing Portland cement with a recycled waste plastic, this design may have a potential to contribute to reduced carbon emissions when scaled to the level of mass concrete production. Copyright © 2017 Elsevier Ltd. All rights reserved.

Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy

DOE PAGES

Field, Kevin G.; Briggs, Samuel A.; Hu, Xunxiang; ...

2016-11-01

FeCrAl alloys are an attractive materials class for nuclear power applications due to their increased environmental compatibility over more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300-400 °C have shown post-irradiation microstructures containing dislocation loops and Cr-rich ' phase. Although these initial works established the post-irradiation microstructures, little to no focus was applied towards the influence of pre-irradiation microstructures on this response. Here, a well annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 dpa at 382 °C and then the role of random high angle grain boundariesmore » on the spatial distribution and size of dislocation loops, dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and an increased size of dislocation loops in the vicinity directly adjacent to the grain boundary. Lastly, these results suggest the importance of the pre-irradiation microstructure on the radiation tolerance of FeCrAl alloys.« less

Technical Letter Report on the Cracking of Irradiated Cast Stainless Steels with Low Ferrite Content

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Y.; Alexandreanu, B.; Natesan, K.

2014-11-01

Crack growth rate and fracture toughness J-R curve tests were performed on CF-3 and CF-8 cast austenite stainless steels (CASS) with 13-14% of ferrite. The tests were conducted at ~320°C in either high-purity water with low dissolved oxygen or in simulated PWR water. The cyclic crack growth rates of CF-8 were higher than that of CF-3, and the differences between the aged and unaged specimens were small. No elevated SCC susceptibility was observed among these samples, and the SCC CGRs of these materials were comparable to those of CASS alloys with >23% ferrite. The fracture toughness values of unirradiated CF-3more » were similar between unaged and aged specimens, and neutron irradiation decreased the fracture toughness significantly. The fracture toughness of CF-8 was reduced after thermal aging, and declined further after irradiation. It appears that while lowering ferrite content may help reduce the tendency of thermal aging embrittlement, it is not very effective to mitigate irradiation-induced embrittlement. Under a combined condition of thermal aging and irradiation, neutron irradiation plays a dominant role in causing embrittlement in CASS alloys.« less

Progress In Developing Laser Based Post Irradiation Examination Infrastructure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, James A.; Scott, Clark L.; Benefiel, Brad C.

To be able to understand the performance of reactor fuels and materials, irradiated materials must be characterized effectively and efficiently in a high rad environment. The characterization work must be performed remotely and in an environment hostile to instrumentation. Laser based characterization techniques provide the ability to be remote and robust in a hot-cell environment. Laser based instrumentation also can provide high spatial resolution suitable for scanning and imaging large areas. The INL is currently developing three laser based Post Irradiation Examination (PIE) stations for the Hot Fuel Examination Facility at the INL. These laser based systems will characterize irradiatedmore » materials and fuels. The characterization systems are the following: Laser Shock Laser based ultrasonic C-scan system Gas Assay, Sample, and Recharge system (GASR, up-grade to an existing system). The laser shock technique will characterize material properties and failure loads/mechanisms in various materials such as LWR fuel, plate fuel, and next generation fuel forms, for PIE in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad bond. The shock wave travels as a compression wave through the material to the free (unconfined) back surface and reflects back through the material under test as a rarefaction (tensile) wave. This rarefaction wave is the physical mechanism that produces internal de-lamination failure. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize debonding caused by the laser shock technique. The laser ultrasonic system will be fully capable of performing classical non-destructive evaluation testing and imaging functions such as microstructure characterization, flaw detection and dimensional metrology in complex components. The purpose of the GASR is to measure the pressure/volume of the plenum of an irradiated fuel element and obtain fission gas samples for analysis. The study of pressure and volume in the plenum of an irradiated fuel element and the analysis of fission gases released from the fuel is important to understanding the performance of reactor fuels and materials. This system may also be used to measure the pressure/volume of other components (such as control blades) and obtain gas samples from these components for analysis. The main function of the laser in this application is to puncture the fuel element to allow the fission gas to escape and if necessary to weld the spot close. The GASR station will have the inherent capability to perform cutting welding and joining functions within a hot-cell.« less

Embrittlement and Flow Localization in Reactor Structural Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xianglin Wu; Xiao Pan; James Stubbins

2006-10-06

Many reactor components and structural members are made from metal alloys due, in large part, to their strength and ability to resist brittle fracture by plastic deformation. However, brittle fracture can occur when structural material cannot undergo extensive, or even limited, plastic deformation due to irradiation exposure. Certain irradiation conditions lead to the development of a damage microstructure where plastic flow is limited to very small volumes or regions of material, as opposed to the general plastic flow in unexposed materials. This process is referred to as flow localization or plastic instability. The true stress at the onset of neckingmore » is a constant regardless of the irradiation level. It is called 'critical stress' and this critical stress has strong temperature dependence. Interrupted tensile testes of 316L SS have been performed to investigate the microstructure evolution and competing mechanism between mechanic twinning and planar slip which are believed to be the controlling mechanism for flow localization. Deformation twinning is the major contribution of strain hardening and good ductility for low temperatures, and the activation of twinning system is determined by the critical twinning stress. Phases transform and texture analyses are also discussed in this study. Finite element analysis is carried out to complement the microstructural analysis and for the prediction of materaials performance with and without stress concentration and irradiation.« less

An evaluation of two flat-black silicone paints for space application

NASA Technical Reports Server (NTRS)

Clatterbuck, Carroll H.; Scialdone, John J.

1990-01-01

Tests were conducted on two flat-black silicone paints suggested for space applications to determine their optical, electrical, and mechanical properties. Three different types of substrate materials were chosen for these paint tests; the application of the paints onto the primed substrates was carried out by spray coating. The adhesion properties were verified by thermal shock and sudden immersion into liquid nitrogen. A controlled thermal vacuum tests was also carried out by varying the temperature of the paint from -100 to 225 C. The measured optical properties included normal and hemispherical emittance, and solar absorption/reflectance. A simultaneous exposure to low-energy proton/UV irradiation in vacuum, and high-energy proton/electron irradiation was carried out. Additional tests of the paints are described.

Characterization of irradiated AISI 316L stainless steel disks removed from the Spallation Neutron Source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vevera, Bradley J; Hyres, James W; McClintock, David A

2014-01-01

Irradiated AISI 316L stainless steel disks were removed from the Spallation Neutron Source (SNS) for post-irradiation examination (PIE) to assess mechanical property changes due to radiation damage and erosion of the target vessel. Topics reviewed include high-resolution photography of the disk specimens, cleaning to remove mercury (Hg) residue and surface oxides, profile mapping of cavitation pits using high frequency ultrasonic testing (UT), high-resolution surface replication, and machining of test specimens using wire electrical discharge machining (EDM), tensile testing, Rockwell Superficial hardness testing, Vickers microhardness testing, scanning electron microscopy (SEM), and energy dispersive spectroscopy (EDS). The effectiveness of the cleaning proceduremore » was evident in the pre- and post-cleaning photography and permitted accurate placement of the test specimens on the disks. Due to the limited amount of material available and the unique geometry of the disks, machine fixturing and test specimen design were critical aspects of this work. Multiple designs were considered and refined during mock-up test runs on unirradiated disks. The techniques used to successfully machine and test the various specimens will be presented along with a summary of important findings from the laboratory examinations.« less

Results of crack-arrest tests on irradiated a 508 class 3 steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iskander, S.K.; Milella, P.P.; Pini, M.A.

1998-02-01

Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K{sub la} of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10{degrees}C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTMmore » A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280{degrees}C, and to a fluence varying from 1.7 to 2.7 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284{degrees}C to a fluence of 3.2 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K{sub la} curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements measured were accounted for. 24 refs., 13 figs., 12 tabs.« less

New fixed-point mini-cell to investigate thermocouple drift in a high-temperature environment under neutron irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Laurie, M.; Vlahovic, L.; Rondinella, V.V.

Temperature measurements in the nuclear field require a high degree of reliability and accuracy. Despite their sheathed form, thermocouples subjected to nuclear radiations undergo changes due to radiation damage and transmutation that lead to significant EMF drift during long-term fuel irradiation experiment. For the purpose of a High Temperature Reactor fuel irradiation to take place in the High Flux Reactor Petten, a dedicated fixed-point cell was jointly developed by LNE-Cnam and JRC-IET. The developed cell to be housed in the irradiation rig was tailor made to quantify the thermocouple drift during the irradiation (about two year duration) and withstand highmore » temperature (in the range 950 deg. C - 1100 deg. C) in the presence of contaminated helium in a graphite environment. Considering the different levels of temperature achieved in the irradiation facility and the large palette of thermocouple types aimed at surveying the HTR fuel pebble during the qualification test both copper (1084.62 deg. C) and gold (1064.18 deg. C) fixed-point materials were considered. The aim of this paper is to first describe the fixed-point mini-cell designed to be embedded in the reactor rig and to discuss the preliminary results achieved during some out of pile tests as much as some robustness tests representative of the reactor scram scenarios. (authors)« less

Comparison of linear and square superposition hardening models for the surface nanoindentation of ion-irradiated materials

NASA Astrophysics Data System (ADS)

Xiao, Xiazi; Yu, Long

2018-05-01

Linear and square superposition hardening models are compared for the surface nanoindentation of ion-irradiated materials. Hardening mechanisms of both dislocations and defects within the plasticity affected region (PAR) are considered. Four sets of experimental data for ion-irradiated materials are adopted to compare with theoretical results of the two hardening models. It is indicated that both models describe experimental data equally well when the PAR is within the irradiated layer; whereas, when the PAR is beyond the irradiated region, the square superposition hardening model performs better. Therefore, the square superposition model is recommended to characterize the hardening behavior of ion-irradiated materials.

21 CFR 179.45 - Packaging materials for use during the irradiation of prepackaged foods.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 21 Food and Drugs 3 2014-04-01 2014-04-01 false Packaging materials for use during the irradiation... OF HEALTH AND HUMAN SERVICES (CONTINUED) IRRADIATION IN THE PRODUCTION, PROCESSING AND HANDLING OF... irradiation of prepackaged foods. The packaging materials identified in this section may be safely subjected...

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

NASA Astrophysics Data System (ADS)

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

2018-05-01

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

Long-Term Effects of {sup 56}Fe Irradiation on Spatial Memory of Mice: Role of Sex and Apolipoprotein E Isoform

DOE Office of Scientific and Technical Information (OSTI.GOV)

Villasana, Laura E.; Benice, Theodore S.; Raber, Jacob, E-mail: raberj@ohsu.ed

Purpose: To assess whether the effects of cranial {sup 56}Fe irradiation on the spatial memory of mice in the water maze are sex and apolipoprotein E (apoE) isoform dependent and whether radiation-induced changes in spatial memory are associated with changes in the dendritic marker microtubule-associated protein 2 (MAP-2) and the presynaptic marker synaptophysin. Methods and Materials: Two-month-old male and female mice expressing human apoE3 or apoE4 received either a 3-Gy dose of cranial {sup 56}Fe irradiation (600 MeV/amu) or sham irradiation. Mice were tested in a water maze task 13 months later to assess effects of irradiation on spatial memorymore » retention. After behavioral testing, the brain tissues of these mice were analyzed for synaptophysin and MAP-2 immunoreactivity. Results: After irradiation, spatial memory retention of apoE3 female, but not male, mice was impaired. A general genotype deficit in spatial memory was observed in sham-irradiated apoE4 mice. Strikingly, irradiation prevented this genotype deficit in apoE4 male mice. A similar but nonsignificant trend was observed in apoE4 female mice. Although there was no change in MAP-2 immunoreactivity after irradiation, synaptophysin immunoreactivity was increased in irradiated female mice, independent of genotype. Conclusions: The effects of {sup 56}Fe irradiation on the spatial memory retention of mice are critically influenced by sex, and the direction of these effects is influenced by apoE isoform. Although in female mice synaptophysin immunoreactivity provides a sensitive marker for effects of irradiation, it cannot explain the apoE genotype-dependent effects of irradiation on the spatial memory retention of the mice.« less

Radiochemical purity of Mo and Tc solution obtained after irradiation and dissolution of Mo-100-enriched and ultra-high-purity natural Mo disks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tkac, Peter; Gromov, Roman; Chemerisov, Sergey D.

2016-09-01

Four irradiations of ultra-high-purity natural Mo targets and one irradiation using 97.4% Mo-100-enriched material were performed. The purpose of these irradiations was to determine whether the presence of Sn stabilizer in the H 2O 2 used for the dissolution of sintered Mo disks can affect the radiochemical purity of the final K 2MoO 4 in 5M KOH solution. Results from radiochemical purity tests performed using thin-layer paper chromatography show that even 2– 3× excess of Sn-stabilized H 2O 2 typically used for dissolution of sintered Mo disks did not affect the radiochemical purity of the final product.

Setup for irradiation and characterization of materials and Si particle detectors at cryogenic temperatures

NASA Astrophysics Data System (ADS)

Väyrynen, S.; Pusa, P.; Sane, P.; Tikkanen, P.; Räisänen, J.; Kuitunen, K.; Tuomisto, F.; Härkönen, J.; Kassamakov, I.; Tuominen, E.; Tuovinen, E.

2007-03-01

A novel facility for proton irradiation with sample cryocooling has been developed at the Accelerator Laboratory of Helsinki University (equipped with a 5 MV tandem accelerator). The setup enables unique experiments to be carried out within the temperature range of 10-300 K. The setup has been constructed for "on-line" studies of vacancies with positron annihilation spectroscopy (PAS) including the option for optical ionization of the vacancies, and for current-voltage ( IV) measurements of irradiated silicon particle detectors. The setup is described in detail and typical performance characteristics are provided. The facility functionality was tested by performing PAS experiments with high-resistivity silicon and by IV measurements for two types of irradiated silicon particle detectors.

Avirulent Bacillus anthracis Strain with Molecular Assay Targets as Surrogate for Irradiation-Inactivated Virulent Spores.

PubMed

Plaut, Roger D; Staab, Andrea B; Munson, Mark A; Gebhardt, Joan S; Klimko, Christopher P; Quirk, Avery V; Cote, Christopher K; Buhr, Tony L; Rossmaier, Rebecca D; Bernhards, Robert C; Love, Courtney E; Berk, Kimberly L; Abshire, Teresa G; Rozak, David A; Beck, Linda C; Stibitz, Scott; Goodwin, Bruce G; Smith, Michael A; Sozhamannan, Shanmuga

2018-04-01

The revelation in May 2015 of the shipment of γ irradiation-inactivated wild-type Bacillus anthracis spore preparations containing a small number of live spores raised concern about the safety and security of these materials. The finding also raised doubts about the validity of the protocols and procedures used to prepare them. Such inactivated reference materials were used as positive controls in assays to detect suspected B. anthracis in samples because live agent cannot be shipped for use in field settings, in improvement of currently deployed detection methods or development of new methods, or for quality assurance and training activities. Hence, risk-mitigated B. anthracis strains are needed to fulfill these requirements. We constructed a genetically inactivated or attenuated strain containing relevant molecular assay targets and tested to compare assay performance using this strain to the historical data obtained using irradiation-inactivated virulent spores.

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

PubMed

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 < 0.05). The microhybrid composite resin exhibited significantly greater DC and gap width than the nanofilled material (P < 0.05). Irradiance had a mostly material-dependent influence on the hardness and DC, but not the polymerization shrinkage, of composite resins.

Gamma radiation-induced thermoluminescence emission of minerals adhered to Mexican sesame seeds

NASA Astrophysics Data System (ADS)

Rodríguez-Lazcano, Y.; Correcher, V.; Garcia-Guinea, J.; Cruz-Zaragoza, E.

2013-02-01

The thermoluminescence (TL) emission of minerals isolated from Mexican sesame seeds appear as a good tool to discern between irradiated and non-irradiated samples. According to the X-ray diffraction (XRD) and environmental scanning electron microscope (ESEM) data, the adhered dust in both samples is mainly composed of different amounts of quartz and feldspars. These mineral phases exhibit (i) enough sensitivity to ionizing radiation inducing good TL intensity, (ii) high stability of the TL signal during the storage of the material, i.e. low fading, and (iii) are thermally and chemically stable. Blind tests were performed under laboratory conditions, but simulating industrial preservation processes, allow us to distinguish between 1 kGy gamma-irradiated and non-irradiated samples even 15 months after irradiation processing followed the EN 1788 European Standard protocol in sesame samples.

Microstructural, mechanical and optical properties research of a carbon ion-irradiated Y2SiO5 crystal

NASA Astrophysics Data System (ADS)

Song, Hong-Lian; Yu, Xiao-Fei; Huang, Qing; Qiao, Mei; Wang, Tie-Jun; Zhang, Jing; Liu, Yong; Liu, Peng; Zhu, Zi-Hua; Wang, Xue-Lin

2017-09-01

Ion irradiation has been a popular method to modify properties of different kinds of materials. Ion-irradiated crystals have been studied for years, but the effects on microstructure and optical properties during irradiation process are still controversial. In this paper, we used 6 MeV C ions with a fluence of 1 × 1015 ion/cm2 irradiated Y2SiO5 (YSO) crystal at room temperature, and discussed the influence of C ion irradiation on the microstructure, mechanical and optical properties of YSO crystal by Rutherford backscattering/channeling analyzes (RBS/C), X-ray diffraction patterns (XRD), Raman, nano-indentation test, transmission and absorption spectroscopy, the prism coupling and the end-facet coupling experiments. We also used the secondary ion mass spectrometry (SIMS) to analyze the elements distribution along sputtering depth. 6 MeV C ions with a fluence of 1 × 1015 ion/cm2 irradiated caused the deformation of YSO structure and also influenced the spectral properties and lattice vibrations.

A Standard Method To Inactivate Bacillus anthracis Spores to Sterility via Gamma Irradiation

PubMed Central

Cote, Christopher K.; Buhr, Tony; Bernhards, Casey B.; Bohmke, Matthew D.; Calm, Alena M.; Esteban-Trexler, Josephine S.; Hunter, Melissa; Katoski, Sarah E.; Kennihan, Neil; Klimko, Christopher P.; Miller, Jeremy A.; Minter, Zachary A.; Pfarr, Jerry W.; Prugh, Amber M.; Quirk, Avery V.; Rivers, Bryan A.; Shea, April A.; Shoe, Jennifer L.; Sickler, Todd M.; Young, Alice A.; Fetterer, David P.; Welkos, Susan L.; McPherson, Derrell; Fountain, Augustus W.

2018-01-01

ABSTRACT In 2015, a laboratory of the United States Department of Defense (DoD) inadvertently shipped preparations of gamma-irradiated spores of Bacillus anthracis that contained live spores. In response, a systematic evidence-based method for preparing, concentrating, irradiating, and verifying the inactivation of spore materials was developed. We demonstrate the consistency of spore preparations across multiple biological replicates and show that two different DoD institutions independently obtained comparable dose-inactivation curves for a monodisperse suspension of B. anthracis spores containing 3 × 1010 CFU. Spore preparations from three different institutions and three strain backgrounds yielded similar decimal reduction (D10) values and irradiation doses required to ensure sterility (DSAL) to the point at which the probability of detecting a viable spore is 10−6. Furthermore, spores of a genetically tagged strain of B. anthracis strain Sterne were used to show that high densities of dead spores suppress the recovery of viable spores. Together, we present an integrated method for preparing, irradiating, and verifying the inactivation of spores of B. anthracis for use as standard reagents for testing and evaluating detection and diagnostic devices and techniques. IMPORTANCE The inadvertent shipment by a U.S. Department of Defense (DoD) laboratory of live Bacillus anthracis (anthrax) spores to U.S. and international destinations revealed the need to standardize inactivation methods for materials derived from biological select agents and toxins (BSAT) and for the development of evidence-based methods to prevent the recurrence of such an event. Following a retrospective analysis of the procedures previously employed to generate inactivated B. anthracis spores, a study was commissioned by the DoD to provide data required to support the production of inactivated spores for the biodefense community. The results of this work are presented in this publication, which details the method by which spores can be prepared, irradiated, and tested, such that the chance of finding residual living spores in any given preparation is 1/1,000,000. These irradiated spores are used to test equipment and methods for the detection of agents of biological warfare and bioterrorism. PMID:29654186

Modified glycogen as construction material for functional biomimetic microfibers.

PubMed

Rabyk, Mariia; Hruby, Martin; Vetrik, Miroslav; Kucka, Jan; Proks, Vladimir; Parizek, Martin; Konefal, Rafal; Krist, Pavel; Chvatil, David; Bacakova, Lucie; Slouf, Miroslav; Stepanek, Petr

2016-11-05

We describe a conceptually new, microfibrous, biodegradable functional material prepared from a modified storage polysaccharide also present in humans (glycogen) showing strong potential as direct-contact dressing/interface material for wound healing. Double bonds were introduced into glycogen via allylation and were further exploited for crosslinking of the microfibers. Triple bonds were introduced by propargylation and served for further click functionalization of the microfibers with bioactive peptide. A simple solvent-free method allowing the preparation of thick layers was used to produce microfibers (diameter ca 2μm) from allylated and/or propargylated glycogen. Crosslinking of the samples was performed by microtron beta-irradiation, and the irradiation dose was optimized to 2kGy. The results from biological testing showed that these highly porous, hydrophilic, readily functionalizable materials were completely nontoxic to cells growing in their presence. The fibers were gradually degraded in the presence of cells. Copyright © 2016 Elsevier Ltd. All rights reserved.

Increased Tensile Strength of Carbon Nanotube Yarns and Sheets through Chemical Modification and Electron Beam Irradiation

NASA Technical Reports Server (NTRS)

Miller, Sandi G.; Williams, Tiffany S.; Baker, James S.; Sola, Francisco; Lebron-Colon, Marisabel; McCorkle, Linda S.; Wilmoth, Nathan G.; Gaier, James; Chen, Michelle; Meador, Michael A.

2014-01-01

The inherent strength of individual carbon nanotubes offers considerable opportunity for the development of advanced, lightweight composite structures. Recent work in the fabrication and application of carbon nanotube (CNT) forms such as yarns and sheets has addressed early nanocomposite limitations with respect to nanotube dispersion and loading; and has pushed the technology toward structural composite applications. However, the high tensile strength of an individual CNT has not directly translated to macro-scale CNT forms where bulk material strength is limited by inter-tube electrostatic attraction and slippage. The focus of this work was to assess post processing of CNT sheet and yarn to improve the macro-scale strength of these material forms. Both small molecule functionalization and e-beam irradiation was evaluated as a means to enhance tensile strength and Youngs modulus of the bulk CNT material. Mechanical testing results revealed a tensile strength increase in CNT sheets by 57 when functionalized, while an additional 48 increase in tensile strength was observed when functionalized sheets were irradiated; compared to unfunctionalized sheets. Similarly, small molecule functionalization increased yarn tensile strength up to 25, whereas irradiation of the functionalized yarns pushed the tensile strength to 88 beyond that of the baseline yarn.

Chemical Synthesis and Oxide Dispersion Properties of Strengthened Tungsten via Spark Plasma Sintering

PubMed Central

Ding, Xiao-Yu; Luo, Lai-Ma; Chen, Hong-Yu; Zhu, Xiao-Yong; Zan, Xiang; Cheng, Ji-Gui; Wu, Yu-Cheng

2016-01-01

Highly uniform oxide dispersion-strengthened materials W–1 wt % Nd2O3 and W–1 wt % CeO2 were successfully fabricated via a novel wet chemical method followed by hydrogen reduction. The powders were consolidated by spark plasma sintering at 1700 °C to suppress grain growth. The samples were characterized by performing field emission scanning electron microscopy and transmission electron microscopy analyses, Vickers microhardness measurements, thermal conductivity, and tensile testing. The oxide particles were dispersed at the tungsten grain boundaries and within the grains. The thermal conductivity of the samples at room temperature exceeded 140 W/m·K. The tensile tests indicated that W–1 wt % CeO2 exhibited a ductile–brittle transition temperature between 500 °C and 550 °C, which was a lower range than that for W–1 wt % Nd2O3. Surface topography and Vickers microhardness analyses were conducted before and after irradiations with 50 eV He ions at a fluence of 1 × 1022 m−2 for 1 h in the large-powder material irradiation experiment system. The grain boundaries of the irradiated area became more evident than that of the unirradiated area for both samples. Irradiation hardening was recognized for the W–1 wt % Nd2O3 and W–1 wt % CeO2 samples. PMID:28773999

Low temperature embrittlement behaviour of different ferritic-martensitic alloys for fusion applications

NASA Astrophysics Data System (ADS)

Rieth, M.; Dafferner, B.

1996-10-01

In the last few years a lot of different low activation CrWVTa steels have been developed world-wide. Without irradiation some of these alloys show clearly a better low temperature embrittlement behaviour than commercial CrNiMoV(Nb) alloys. Within the MANITU project a study was carried out to compare, prior to the irradiation program, the embrittlement behaviour of different alloys in the unirradiated condition performing instrumented Charpy impact bending tests with sub-size specimens. The low activation materials (LAM) considered were different OPTIFER alloys (Forschungszentrum Karlsruhe), F82H (JAERI), 9Cr2WVTa (ORNL), and GA3X (PNL). The modified commercial 10-11% CrNiMoVNb steels were MANET and OPTIMAR. A meaningful comparison between these alloys could be drawn, since the specimens of all materials were manufactured and tested under the same conditions.

Development of a pepper-pot emittance meter for diagnostics of low-energy multiply charged heavy ion beams extracted from an ECR ion source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nagatomo, T., E-mail: nagatomo@riken.jp; Kase, M.; Kamigaito, O.

2016-02-15

Several fluorescent materials were tested for use in the imaging screen of a pepper-pot emittance meter that is suitable for investigating the beam dynamics of multiply charged heavy ions extracted from an ECR ion source. SiO{sub 2} (quartz), KBr, Eu-doped CaF{sub 2}, and Tl-doped CsI crystals were first irradiated with 6.52-keV protons to determine the effects of radiation damage on their fluorescence emission properties. For such a low-energy proton beam, only the quartz was found to be a suitable fluorescent material, since the other materials suffered a decay in fluorescence intensity with irradiation time. Subsequently, quartz was irradiated with heavymore » {sup 12}C{sup 4+}, {sup 16}O{sup 4+}, and {sup 40}Ar{sup 11+} ions, but it was found that the fluorescence intensity decreased too rapidly to measure the emittance of these heavy-ion beams. These results suggest that a different energy loss mechanism occurs for heavier ions and for protons.« less

The effects of gamma irradiation on diclofenac sodium, liposome and niosome ingredients for rheumatoid arthritis

PubMed Central

Turker, Selcan; Çolak, Seyda; Korkmaz, Mustafa; Kiliç, Ekrem; Özalp, Meral

2013-01-01

The use of gamma rays for the sterilization of pharmaceutical raw materials and dosage forms is an alternative method for sterilization. However, one of the major problems of the radiosterilization is the production of new radiolytic products during the irradiation process. Therefore, the principal problem in radiosterilization is to determine and to characterize these physical and chemical changes originating from high-energy radiation. Parenteral drug delivery systems were prepared and in vitro characterization, biodistribution and treatment studies were done in our previous studies. Drug delivery systems (liposomes, niosomes, lipogelosomes and niogelosomes) encapsulating diclofenac sodium (DFNa) were prepared for the treatment of rheumatoid arthritis (RA). This work complies information about the studies developed in order to find out if gamma radiation could be applied as a sterilization method to DFNa, and the raw materials as dimyristoyl phosphatidylcholine (DMPC), surfactant I [polyglyceryl-3-cethyl ether (SUR I)], dicethyl phosphate (DCP) and cholesterol (CHOL) that are used to prepare those systems. The raw materials were irradiated with different radiation doses (5, 10, 25 and 50 kGy) and physicochemical changes (organoleptic properties pH, UV and melting point), microbiological evaluation [sterility assurance level (SAL), sterility and pyrogen test] and electron spin resonance (ESR) characteristics were studied at normal (25 °C, 60% relative humidity) and accelerated (40 °C, 75% relative humidity) stability test conditions. PMID:24265902

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

NASA Astrophysics Data System (ADS)

Gelles, D. S.

1990-05-01

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

Emulation of reactor irradiation damage using ion beams

DOE PAGES

Was, G. S.; Jiao, Z.; Getto, E.; ...

2014-06-14

The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

Radiation resistance of elastomeric O-rings in mixed neutron and gamma fields: Testing methodology and experimental results

NASA Astrophysics Data System (ADS)

Zenoni, A.; Bignotti, F.; Donzella, A.; Donzella, G.; Ferrari, M.; Pandini, S.; Andrighetto, A.; Ballan, M.; Corradetti, S.; Manzolaro, M.; Monetti, A.; Rossignoli, M.; Scarpa, D.; Alloni, D.; Prata, M.; Salvini, A.; Zelaschi, F.

2017-11-01

Materials and components employed in the presence of intense neutron and gamma fields are expected to absorb high dose levels that may induce deep modifications of their physical and mechanical properties, possibly causing loss of their function. A protocol for irradiating elastomeric materials in reactor mixed neutron and gamma fields and for testing the evolution of their main mechanical and physical properties with absorbed dose has been developed. Four elastomeric compounds used for vacuum O-rings, one fluoroelastomer polymer (FPM) based and three ethylene propylene diene monomer rubber (EPDM) based, presently available on the market have been selected for the test. One EPDM is rated as radiation resistant in gamma fields, while the other elastomers are general purpose products. Particular care has been devoted to dosimetry calculations, since absorbed dose in neutron fields, unlike pure gamma fields, is strongly dependent on the material composition and, in particular, on the hydrogen content. The products have been tested up to about 2 MGy absorbed dose. The FPM based elastomer, in spite of its lower dose absorption in fast neutron fields, features the largest variations of properties, with a dramatic increase in stiffness and brittleness. Out of the three EPDM based compounds, one shows large and rapid changes in the main mechanical properties, whereas the other two feature more stable behaviors. The performance of the EPDM rated as radiation resistant in pure gamma fields does not appear significantly better than that of the standard product. The predictive capability of the accelerated irradiation tests performed as well as the applicable concepts of threshold of radiation damage is discussed in view of the use of the examined products in the selective production of exotic species facility, now under construction at the Legnaro National Laboratories of the Italian Istituto Nazionale di Fisica Nucleare. It results that a careful account of dose rate effects and oxygen penetration in the material, both during test irradiations and in operating conditions, is needed to obtain reliable predictions.

Radiation resistance of elastomeric O-rings in mixed neutron and gamma fields: Testing methodology and experimental results.

PubMed

Zenoni, A; Bignotti, F; Donzella, A; Donzella, G; Ferrari, M; Pandini, S; Andrighetto, A; Ballan, M; Corradetti, S; Manzolaro, M; Monetti, A; Rossignoli, M; Scarpa, D; Alloni, D; Prata, M; Salvini, A; Zelaschi, F

2017-11-01

Materials and components employed in the presence of intense neutron and gamma fields are expected to absorb high dose levels that may induce deep modifications of their physical and mechanical properties, possibly causing loss of their function. A protocol for irradiating elastomeric materials in reactor mixed neutron and gamma fields and for testing the evolution of their main mechanical and physical properties with absorbed dose has been developed. Four elastomeric compounds used for vacuum O-rings, one fluoroelastomer polymer (FPM) based and three ethylene propylene diene monomer rubber (EPDM) based, presently available on the market have been selected for the test. One EPDM is rated as radiation resistant in gamma fields, while the other elastomers are general purpose products. Particular care has been devoted to dosimetry calculations, since absorbed dose in neutron fields, unlike pure gamma fields, is strongly dependent on the material composition and, in particular, on the hydrogen content. The products have been tested up to about 2 MGy absorbed dose. The FPM based elastomer, in spite of its lower dose absorption in fast neutron fields, features the largest variations of properties, with a dramatic increase in stiffness and brittleness. Out of the three EPDM based compounds, one shows large and rapid changes in the main mechanical properties, whereas the other two feature more stable behaviors. The performance of the EPDM rated as radiation resistant in pure gamma fields does not appear significantly better than that of the standard product. The predictive capability of the accelerated irradiation tests performed as well as the applicable concepts of threshold of radiation damage is discussed in view of the use of the examined products in the selective production of exotic species facility, now under construction at the Legnaro National Laboratories of the Italian Istituto Nazionale di Fisica Nucleare. It results that a careful account of dose rate effects and oxygen penetration in the material, both during test irradiations and in operating conditions, is needed to obtain reliable predictions.

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

2017-05-01

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

DEVICE FOR TREATING MATERIALS

DOEpatents

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17

Test-hole construction in a reactor to facilitate inserting and removing test specimens from the reactor for irradiation therein is discussed. An elongated chamber extends from the outer face of the reactor shield into the reactor. A shield box, having an open end, is sealed to thc outer face of the reactor shield by its open end surrounding the outer end of the chamber. A removable door is provided in the side wall of the shield box for inscrtion and removal of test specimens. A means operable from thc exterior of the shield box is provided for transferring test specimens between the shield box and the irradiation position within the chamber and consists of an elongated rod having a specimen tray engaging member on its inner end, which may be manipulated by the operator.

Evaluation of cooling concepts and specimen geometries for high heat flux tests on neutron irradiated divertor elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linke, J.; Bolt. H.; Breitbach, G.

1994-12-31

To assess the lifetime and the long term heat removal capabilities of plasma facing components in future thermonuclear fusion reactors such as ITER, neutron irradiation and subsequent high heat flux tests will be most essential. The effect of neutron damage will be simulated in material test reactors (such as the HFR-Petten) in a fission neutron environment. To investigate the heat loads during normal and off-normal operation scenarios a 60 kW electron beam test stand (Juelich Divertor Test Facility in Hot Cells, JUDITH) has been installed in a hot cell which can be operated by remote handling techniques. In this facilitymore » inertially cooled test coupons can be handled as well as small actively cooled divertor mock-ups. A special clamping mechanism for small test coupons (25 mm x 25 mm x 35 mm) with an integrated coolant channel within a copper or TZM heat sink has been developed and tested in an electron beam test bed. This method is an attractive alternative to costly large scale tests on complete divertor modules. The temperature and stress fields in individual CFC or beryllium tiles brazed to metallic heat sink (e.g. copper or TZM) can be investigated before and after neutron irradiation with moderate efforts.« less

Material Compatibility for Historic Items Decontaminated with ...

EPA Pesticide Factsheets

Report This project continued research of the effects of decontamination methods for biological agents on materials identified as representative of types of irreplaceable objects or works of art found in museums and/or archive settings. In the previous research, surrogate materials were checked for compatibility with four decontamination methods: chlorine dioxide, hydrogen peroxide vapor, methyl bromide, and ethylene oxide gas. This project investigated the effects of gamma irradiation, which has also been shown to be an effective decontamination method for biological agents, on the surrogate test materials.

Accelerated radiation damage test facility using a 5 MV tandem ion accelerator

NASA Astrophysics Data System (ADS)

Wady, P. T.; Draude, A.; Shubeita, S. M.; Smith, A. D.; Mason, N.; Pimblott, S. M.; Jimenez-Melero, E.

2016-01-01

We have developed a new irradiation facility that allows to perform accelerated damage tests of nuclear reactor materials at temperatures up to 400 °C using the intense proton (<100 μA) and heavy ion (≈10 μA) beams produced by a 5 MV tandem ion accelerator. The dedicated beam line for radiation damage studies comprises: (1) beam diagnosis and focusing optical components, (2) a scanning and slit system that allows uniform irradiation of a sample area of 0.5-6 cm2, and (3) a sample stage designed to be able to monitor in-situ the sample temperature, current deposited on the sample, and the gamma spectrum of potential radio-active nuclides produced during the sample irradiation. The beam line capabilities have been tested by irradiating a 20Cr-25Ni-Nb stabilised stainless steel with a 3 MeV proton beam to a dose level of 3 dpa. The irradiation temperature was 356 °C, with a maximum range in temperature values of ±6 °C within the first 24 h of continuous irradiation. The sample stage is connected to ground through an electrometer to measure accurately the charge deposited on the sample. The charge can be integrated in hardware during irradiation, and this methodology removes uncertainties due to fluctuations in beam current. The measured gamma spectrum allowed the identification of the main radioactive nuclides produced during the proton bombardment from the lifetimes and gamma emissions. This dedicated radiation damage beam line is hosted by the Dalton Cumbrian Facility of the University of Manchester.

Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy

NASA Astrophysics Data System (ADS)

Field, Kevin G.; Briggs, Samuel A.; Hu, Xunxiang; Yamamoto, Yukinori; Howard, Richard H.; Sridharan, Kumar

2017-01-01

FeCrAl alloys are an attractive class of materials for nuclear power applications because of their increased environmental compatibility compared with more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300 and 400 °C have shown post-irradiation microstructures containing dislocation loops and a Cr-rich α‧ phase. Although these initial studies established the post-irradiation microstructures, there was little to no focus on understanding the influence of pre-irradiation microstructures on this response. In this study, a well-annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 displacements per atom (dpa) at 382 °C and then the effect of random high-angle grain boundaries on the spatial distribution and size of a〈100〉 dislocation loops, a/2〈111〉 dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with a/2〈111〉 dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and a〈100〉 dislocation loops exhibiting an increased size in the vicinity of the grain boundary. These results suggest the importance of the pre-irradiation microstructure and, specifically, defect sink density spacing to the radiation tolerance of FeCrAl alloys.

Heat Bonding of Irradiated Ethylene Vinyl Acetate

NASA Technical Reports Server (NTRS)

Slack, D. H.

1986-01-01

Reliable method now available for joining parts of this difficult-tobond material. Heating fixture encircles ethylene vinyl acetate multiplesocket part, providing heat to it and to tubes inserted in it. Fixtures specially designed to match parts to be bonded. Tube-and-socket bonds made with this technique subjected to tensile tests. Bond strengths of 50 percent that of base material obtained consistently.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« less

10 CFR 36.21 - Performance criteria for sealed sources.

Code of Federal Regulations, 2010 CFR

2010-01-01

... resistance if the sources are for use in irradiator pools; and (5) In prototype testing of the sealed source... under 10 CFR 32.210; (2) Must be doubly encapsulated; (3) Must use radioactive material that is as...

Radiation Characterization of Commercial GaN Devices

NASA Technical Reports Server (NTRS)

Harris, Richard D.; Scheick, Leif Z.; Hoffman, James P.; Thrivikraman, Tushar; Jenabi, Masud; Gim, Yonggyu; Miyahira, Tetsuo

2011-01-01

Radiative feedback from primordial protostars and final mass of the first star Commercially available devices fabricated from GaN are beginning to appear from a number of different suppliers. Based on previous materials and prototype device studies, it is expected that these commercial devices will be quite tolerant to the types of radiation encountered in space. This expectation needs to be verified and the study described herein was undertaken for that purpose. All of the parts discussed in this report are readily available commercially. The parts chosen for study are all targeted for RF applications. Three different studies were performed: 1) a preliminary DDD/TID test of a variety of part types was performed by irradiating with 50 MeV protons, 2) a detailed DDD/TID study of one particular part type was performed by irradiating with 50 MeV protons, and 3) a SEB/SEGR test was performed on a variety of part types by irradiating with heavy ions. No significant degradation was observed in the tests performed in this study.

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

NASA Astrophysics Data System (ADS)

Kobayashi, Koji; Akiyama, Yoko; Nishijima, Shigehiro

2017-09-01

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

Effect of electron beam irradiation and microencapsulation on the flame retardancy of ethylene-vinyl acetate copolymer materials during hot water ageing test

NASA Astrophysics Data System (ADS)

Sheng, Haibo; Zhang, Yan; Wang, Bibo; Yu, Bin; Shi, Yongqian; Song, Lei; Kundu, Chanchal Kumar; Tao, Youji; Jie, Ganxin; Feng, Hao; Hu, Yuan

2017-04-01

Microencapsulated ammonium polyphosphate (MCAPP) in combination with polyester polyurethane (TPU) was used to flame retardant ethylene-vinyl acetate copolymer (EVA). The EVA composites with different irradiation doses were immersed in hot water (80 °C) to accelerate ageing process. The microencapsulation and irradiation dose ensured positive impacts on the properties of the EVA composites in terms of better dimensional stability and flame retardant performance. The microencapsulation of APP could lower its solubility in water and the higher irradiation dose led to the more MCAPP immobilized in three dimensional crosslinked structure of the EVA matrix which could jointly enhance the flame retardant and electrical insulation properties of the EVA composites. So, the EVA composites with 180 kGy irradiation dose exhibited better dimensional stability than the EVA composites with 120 kGy due to the higher crosslinking degree. Moreover, the higher irradiation dose lead to the more MCAPP immobilizated in crosslinked three-dimensional structure of EVA, enhancing the flame retardancy and electrical insulation properties of the EVA composites. After ageing test in hot water at 80 °C for 2 weeks, the EVA/TPU/MCAPP composite with 180 kGy could still maintain the UL-94 V-0 rating and the limiting oxygen index (LOI) value was as high as 30%. This investigation indicated the flame retardant EVA cable containing MCAPP could achieve stable properties and lower electrical fire hazard risk during long-term hot water ageing test.

Effect of Nd: YAG laser irradiation on surface properties and bond strength of zirconia ceramics.

PubMed

Liu, Li; Liu, Suogang; Song, Xiaomeng; Zhu, Qingping; Zhang, Wei

2015-02-01

This study investigated the effect of neodymium-doped yttrium aluminum garnet (Nd: YAG) laser irradiation on surface properties and bond strength of zirconia ceramics. Specimens of zirconia ceramic pieces were divided into 11 groups according to surface treatments as follows: one control group (no treatment), one air abrasion group, and nine laser groups (Nd: YAG irradiation). The laser groups were divided by applying with different output power (1, 2, or 3 W) and irradiation time (30, 60, or 90 s). Following surface treatments, the morphological characteristics of ceramic pieces was observed, and the surface roughness was measured. All specimens were bonded to resin cement. After, stored in water for 24 h and additionally aged by thermocycling, the shear bond strength was measured. Dunnett's t test and one-way ANOVA were performed as the statistical analyses for the surface roughness and the shear bond strength, respectively, with α = .05. Rougher surface of the ceramics could be obtained by laser irradiation with higher output power (2 and 3 W). However, cracks and defects were also found on material surface. The shear bond strength of laser groups was not obviously increased, and it was significantly lower than that of air abrasion group. No significant differences of the shear bond strength were found among laser groups treated with different output power or irradiation time. Nd: YAG laser irradiation cannot improve the surface properties of zirconia ceramics and cannot increase the bond strength of the ceramics. Enhancing irradiation power and extending irradiation time cannot induce higher bond strength of the ceramics and may cause material defect.

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

NASA Astrophysics Data System (ADS)

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

2015-12-01

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

Microstructural stability and mechanical behavior of FeNiMnCr high entropy alloy under ion irradiation

DOE PAGES

Leonard, Keith J.; Bei, Hongbin; Zinkle, Steven J.; ...

2016-05-13

In recent years, high entropy alloys (HEAs) have attracted significant attention due to their excellent mechanical properties and good corrosion resistance, making them potential candidates for high temperature fission and fusion structural applications. However there is very little known about their radiation resistance, particularly at elevated temperatures relevant for energy applications. In the present study, a single phase (face centered cubic) concentrated solid solution alloy of composition 27%Fe-28%Ni-27%Mn-18%Cr was irradiated with 3 or 5.8 MeV Ni ions at temperatures ranging from room temperature to 700 °C and midrange doses from 0.03 to 10 displacements per atom (dpa). Transmission electron microscopymore » (TEM), scanning transmission electron microscopy with energy dispersive x-ray spectrometry (STEM/EDS) and X-ray diffraction (XRD) were used to characterize the radiation defects and microstructural changes. Irradiation at higher temperatures showed evidence of relatively sluggish solute diffusion with limited solute depletion or enrichment at grain boundaries. The main microstructural feature at all temperatures was high-density small dislocation loops. Voids were not observed at any irradiation condition. Nano-indentation tests on specimens irradiated at room temperature showed a rapid increase in hardness ~35% and ~80% higher than the unirradiated value at 0.03 and 0.3 dpa midrange doses, respectively. The irradiation-induced hardening was less pronounced for 500 °C irradiations (<20% increase after 3 dpa). Overall, the examined HEA material exhibits superior radiation resistance compared to conventional single phase Fe-Cr-Ni austenitic alloys such as stainless steels. Furthermore, the present study provides insight on the fundamental irradiation behavior of a single phase HEA material over a broad range of irradiation temperatures.« less

Design and performance evaluation of a cryogenic condenser for an in-pile experiment

NASA Technical Reports Server (NTRS)

Graham, R. W.; Crum, R. J.; Hsu, Y.

1972-01-01

An apparatus was designed to enable in-pile irradiation of materials in liquid hydrogen at cryogenic temperatures. One of the principal components of this apparatus was a horizontal tube condenser. The performance of the condenser was evaluated by running a liquid-nitrogen prototype of the apparatus at heat loads comparable to or greater than those expected during the irradiation. The test showed that the condenser was capable of handling the design heat load and that the design procedure was sound.

Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, andmore » up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six of them are equipped with instrumentation for condensed matter physics research: - H3 - spectrometer and diffractometer with double monochromator; - H4 - small angle scattering spectrometer; - H5 - polarized neutrons spectrometer; - H6, H7 - two 3-axial crystal neutron spectrometers; - H8 - neutron radiography stand. For two horizontal channels are ongoing exploitation programs: - H2 - station with epithermal neutron beam produced in uranium converter is being developed. Intelligent converter will be installed on the periphery of reactor core. The intensity of the beam will be at the level 2x10{sup 9} n cm{sup -2}s{sup -1} what makes the beam unique in the Europe. - H1 - special pneumatic horizontal mail is being developed for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. The number of neutron irradiation facilities in MARIA reactor is increasing every year. Numerous of thermal neutron irradiation channels including fast hydraulic rabbit system and large size channels for fast neutron irradiation are used routinely. Recently new in-pile facility with ITER-like neutron energy spectrum for 14 MeV neutron irradiation has been constructed. Taking into account its performance and ability of almost incessant operation the facility appears as one of the most powerful 14 MeV neutron sources. The facility shall be used for material research connected with thermonuclear devices (ITER) and 4. generation nuclear reactors. The system of independent fuels channels used in MARIA reactor appear to be very flexible and very convenient to be used as irradiation channels for uranium targets for {sup 99}Mo production. Currently, MARIA reactor supplies ca. 18% world production of {sup 99}Mo. The MARIA reactor research activities are still extended. The current scientific projects are connected e.g. with silicon neutron transmutation doping, in-pile gamma heating measurements, French calculation codes implementation (TRIPOLI4, APOLLO2). The horizontal neutron beams utilization is also developed. The MARIA reactor, due to its primary application connected with loop and fuel testing, is very convenient for testing the nuclear instrumentation, control and measurement systems.« less

Effects of γ-ray irradiation on optical absorption and laser damage performance of KDP crystals containing arsenic impurities.

PubMed

Guo, D C; Jiang, X D; Huang, J; Wang, F R; Liu, H J; Xiang, X; Yang, G X; Zheng, W G; Zu, X T

2014-11-17

The effects of γ-irradiation on potassium dihydrogen phosphate crystals containing arsenic impurities are investigated with different optical diagnostics, including UV-VIS absorption spectroscopy, photo-thermal common-path interferometer and photoluminescence spectroscopy. The optical absorption spectra indicate that a new broad absorption band near 260 nm appears after γ-irradiation. It is found that the intensity of absorption band increases with the increasing irradiation dose and arsenic impurity concentration. The simulation of radiation defects show that this absorption is assigned to the formation of AsO₄⁴⁻ centers due to arsenic ions substituting for phosphorus ions. Laser-induced damage threshold test is conducted by using 355 nm nanosecond laser pulses. The correlations between arsenic impurity concentration and laser induced damage threshold are presented. The results indicate that the damage performance of the material decreases with the increasing arsenic impurity concentration. Possible mechanisms of the irradiation-induced defects formation under γ-irradiation of KDP crystals are discussed.

Properties of radiation stable insulation composites for fusion magnet

NASA Astrophysics Data System (ADS)

Wu, Zhixiong; Huang, Rongjin; Huang, Chuanjun; Li, Laifeng

2017-09-01

High field superconducting magnets made of Nb3Al will be a suitable candidate for future fusion device which can provide magnetic field over 15T without critical current degradation caused by strain. The higher magnetic field and the larger current will produce a huge electromagnetic force. Therefore, it is necessary to develop high strength cryogenic structural materials and electrical insulation materials with excellent performance. On the other hand, superconducting magnets in fusion devices will experience significant nuclear radiation exposure during service. While typical structural materials like stainless steel and titanium have proven their ability to withstand these conditions, electrical insulation materials used in these coils have not fared as well. In fact, recent investigations have shown that electrical insulation breakdown is a limiting factor in the performance of high field magnets. The insulation materials used in the high field fusion magnets should be characterized by excellent mechanical properties, high radiation resistivity and good thermal conductivity. To meet these objectives, we designed various insulation materials based on epoxy resins and cyanate ester resins and investigated their processing characteristic and mechanical properties before and after irradiation at low temperature. In this paper, the recent progress of the radiation stable insulation composites for high field fusion magnet is presented. The materials have been irradiated by 60Co γ-ray irradiation in air at ambient temperature with a dose rate of 300 Gy/min. The total doses of 1 MGy, 5 MGy and 10 MGy were selected to the test specimens.

Application of subsize specimens in nuclear plant life extension

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosinski, S.T.; Kumar, A.S.; Cannon, S.C.

1991-01-01

The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missour-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) willmore » be tested. This paper will provide details of the program and present 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.« less

Application of subsize specimens in nuclear plant life extension

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosinski, S.T.; Kumar, A.S.; Cannon, S.C.

1991-12-31

The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missour-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) willmore » be tested. This paper will provide details of the program and present 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.« less

Stratospheric experiments on curing of composite materials

NASA Astrophysics Data System (ADS)

Chudinov, Viacheslav; Kondyurin, Alexey; Svistkov, Alexander L.; Efremov, Denis; Demin, Anton; Terpugov, Viktor; Rusakov, Sergey

2016-07-01

Future space exploration requires a large light-weight structure for habitats, greenhouses, space bases, space factories and other constructions. A new approach enabling large-size constructions in space relies on the use of the technology of polymerization of fiber-filled composites with a curable polymer matrix applied in the free space environment on Erath orbit. In orbit, the material is exposed to high vacuum, dramatic temperature changes, plasma of free space due to cosmic rays, sun irradiation and atomic oxygen (in low Earth orbit), micrometeorite fluence, electric charging and microgravitation. The development of appropriate polymer matrix composites requires an understanding of the chemical processes of polymer matrix curing under the specific free space conditions to be encountered. The goal of the stratospheric flight experiment is an investigation of the effect of the stratospheric conditions on the uncured polymer matrix of the composite material. The unique combination of low residual pressure, high intensity UV radiation including short-wave UV component, cosmic rays and other aspects associated with solar irradiation strongly influences the chemical processes in polymeric materials. We have done the stratospheric flight experiments with uncured composites (prepreg). A balloon with payload equipped with heater, temperature/pressure/irradiation sensors, microprocessor, carrying the samples of uncured prepreg has been launched to stratosphere of 25-30 km altitude. After the flight, the samples have been tested with FTIR, gel-fraction, tensile test and DMA. The effect of cosmic radiation has been observed. The composite was successfully cured during the stratospheric flight. The study was supported by RFBR grants 12-08-00970 and 14-08-96011.

Results of Simulated Galactic Cosmic Radiation (GCR) and Solar Particle Events (SPE) on Spectra Restraint Fabric

NASA Technical Reports Server (NTRS)

Peters, Benjamin; Hussain, Sarosh; Waller, Jess

2017-01-01

Spectra or similar Ultra-high-molecular-weight polyethylene (UHMWPE) fabric is the likely choice for future structural space suit restraint materials due to its high strength-to-weight ratio, abrasion resistance, and dimensional stability. During long duration space missions, space suits will be subjected to significant amounts of high-energy radiation from several different sources. To insure that pressure garment designs properly account for effects of radiation, it is important to characterize the mechanical changes to structural materials after they have been irradiated. White Sands Test Facility (WSFTF) collaborated with the Crew and Thermal Systems Division at the Johnson Space Center (JSC) to irradiate and test various space suit materials by examining their tensile properties through blunt probe puncture testing and single fiber tensile testing after the materials had been dosed at various levels of simulated GCR and SPE Iron and Proton beams at Brookhaven National Laboratories. The dosages were chosen based on a simulation developed by the Structural Engineering Division at JSC for the expected radiation dosages seen by space suit softgoods seen on a Mars reference mission. Spectra fabric tested in the effort saw equivalent dosages at 2x, 10x, and 20x the predicted dose as well as a simulated 50 year exposure to examine the range of effects on the material and examine whether any degradation due to GCR would be present if the suit softgoods were stored in deep space for a long period of time. This paper presents the results of this work and outlines the impact on space suit pressure garment design for long duration deep space missions.

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

DOE PAGES

Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian; ...

2017-02-27

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

The Study of Simulated Space Radiation Environment Effect on Conductive Properties of ITO Thermal Control Materials

NASA Astrophysics Data System (ADS)

Wei-Quan, Feng; Chun-Qing, Zhao; Zi-Cai, Shen; Yi-Gang, Ding; Fan, Zhang; Yu-Ming, Liu; Hui-Qi, Zheng; Xue, Zhao

In order to prevent detrimental effects of ESD caused by differential surface charging of spacecraft under space environments, an ITO transparent conductive coating is often deposited on the thermal control materials outside spacecraft. Since the ITO coating is exposed in space environment, the environment effects on electrical property of ITO coatings concern designers of spacecraft deeply. This paper introduces ground tests to simulate space radiation environmental effects on conductive property of ITO coating. Samples are made of ITO/OSR, ITO/Kapton/Al and ITO/FEP/Ag thermal control coatings. Simulated space radiation environment conditions are NUV of 500ESH, 40 keV electron of 2 × 1016 е/cm2, 40 keV proton of 2.5 × 1015 p/cm2. Conductive property is surface resistivity measured in-situ in vacuum. Test results proved that the surface resistivity for all ITO coatings have a sudden decrease in the beginning of environment test. The reasons for it may be the oxygen vacancies caused by vacuum and decayed RIC caused by radiation. Degradation in conductive properties caused by irradiation were found. ITO/FEP/Ag exhibits more degradation than other two kinds. The conductive property of ITO/kapton/Al is stable for vacuum irradiation. The analysis of SEM and XPS found more crackers and less Sn and In concentration after irradiation which may be the reason for conductive property degradation.

Laser-based irradiation apparatus and methods for monitoring the dose-rate response of semiconductor devices

DOEpatents

Horn, Kevin M [Albuquerque, NM

2006-03-28

A scanned, pulsed, focused laser irradiation apparatus can measure and image the photocurrent collection resulting from a dose-rate equivalent exposure to infrared laser light across an entire silicon die. Comparisons of dose-rate response images or time-delay images from before, during, and after accelerated aging of a device, or from periodic sampling of devices from fielded operational systems allows precise identification of those specific age-affected circuit structures within a device that merit further quantitative analysis with targeted materials or electrical testing techniques. Another embodiment of the invention comprises a broad-beam, dose rate-equivalent exposure apparatus. The broad-beam laser irradiation apparatus can determine if aging has affected the device's overall functionality. This embodiment can be combined with the synchronized introduction of external electrical transients into a device under test to simulate the electrical effects of the surrounding circuitry's response to a radiation exposure.

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

NASA Astrophysics Data System (ADS)

Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

2014-06-01

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Ultra-accelerated natural sunlight exposure testing

DOEpatents

Jorgensen, Gary J.; Bingham, Carl; Goggin, Rita; Lewandowski, Allan A.; Netter, Judy C.

2000-06-13

Process and apparatus for providing ultra accelerated natural sunlight exposure testing of samples under controlled weathering without introducing unrealistic failure mechanisms in exposed materials and without breaking reciprocity relationships between flux exposure levels and cumulative dose that includes multiple concurrent levels of temperature and relative humidity at high levels of natural sunlight comprising: a) concentrating solar flux uniformly; b) directing the controlled uniform sunlight onto sample materials in a chamber enclosing multiple concurrent levels of temperature and relative humidity to allow the sample materials to be subjected to accelerated irradiance exposure factors for a sufficient period of time in days to provide a corresponding time of about at least a years worth of representative weathering of the sample materials.

High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Was, Gary; Wirth, Brian; Motta, Athur

The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiatedmore » microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations.« less

Environmental stress cracking in gamma-irradiated polycarbonate - A diffusion approach

NASA Astrophysics Data System (ADS)

Silva, Pietro Paolo J. C. de O.; Araújo, Patricia L. B.; da Silveira, Leopoldo B. B.; Araújo, Elmo S.

2017-01-01

Polycarbonate (PC) is an engineering polymer which presents interesting properties. This material has been also used in medical devices, which is frequently exposed to gamma radiosterilization and to chemical agents. This may produce significant changes in polymer structure, leading to failure in service. The present work brings about a new approach on environmental stress cracking (ESC) processes elucidation in 100 kGy gamma-irradiated PC, by evaluating the diffusion process of methanol or 2-propanol in test specimens and determining the diffusion parameters on solvent-irradiated polymer systems. A comparison of diffusion parameters for both solvents indicated that methanol has a considerable ESC action on PC, with diffusion parameter of 7.5×10-14±1% m2 s-1 for non-irradiated PC and 7.8×10-14±2.8% m2 s-1 for PC irradiated at 100 kGy. In contrast, 2-propanol did not act as an ESC agent, as it did promote neither swelling nor cracks in the test specimens. These results were confirmed by visual analysis and optical microscopy. Unexpectedly, structural damages evidenced in tensile strength tests suggested that 2-propanol is as aggressive as methanol chemical for PC. Moreover, although some manufacturers indicate the use of 2-propanol as a cleaning product for PC artifacts, such use should be avoided in parts under mechanical stress.

The Gottingen minipig is a model of the hematopoietic acute radiation syndrome: G-CSF stimulates hematopoiesis and enhances survival from lethal total-body gamma-irradiation

PubMed Central

Moroni, Maria; Ngudiankama, Barbara F.; Christensen, Christine; Olsen, Cara H.; Owens, Rossitsa; Lombardini, Eric D.; Holt, Rebecca K.; Whitnall, Mark H.

2013-01-01

Purpose We are characterizing the Gottingen minipig as an additional large animal model for advanced drug testing for the Acute Radiation Syndrome (ARS), to enhance discovery and development of novel radiation countermeasures. Among the advantages provided by this model, the similarities to human hematological parameters and dynamics of cell loss/recovery following irradiation provide a convenient means to compare efficacy of drugs known to affect bone marrow cellularity and hematopoiesis. Methods and Materials Male Gottingen minipigs, 4–5 months old and weighing 9–11 kg were used for this study. We tested the standard off-label treatment for ARS, rhG-CSF (Neupogen®, 10 μg/kg/day for 17 days), at the estimated LD70/30 total-body gamma-irradiation (TBI) radiation dose for the hematopoietic syndrome, starting 24 hours after irradiation. Results Results indicate G-CSF enhanced survival, stimulated recovery from neutropenia, and induced mobilization of hematopoietic progenitor cells. In addition, administration of G-CSF resulted in maturation of monocytes/macrophages. Conclusion These results support continuing efforts toward validation of the minipig as a large animal model for advanced testing of radiation countermeasures and characterization of the pathophysiology of ARS, and suggest that the efficacy of G-CSF in improving survival after total body irradiation may involve mechanisms other than increasing numbers of circulating granulocytes. PMID:23845847

Microstructural, mechanical and optical properties research of a carbon ion-irradiated Y 2SiO 5 crystal

DOE PAGES

Song, Hong-Lian; Yu, Xiao-Fei; Huang, Qing; ...

2017-01-28

Ion irradiation has been a popular method to modify properties of different kinds of materials. Ion-irradiated crystals have been studied for years, but the effects on microstructure and optical properties during irradiation process are still controversial. In this study, we used 6 MeV C ions with a fluence of 1 × 10 15 ion/cm 2 irradiated Y 2SiO 5 (YSO) crystal at room temperature, and discussed the influence of C ion irradiation on the microstructure, mechanical and optical properties of YSO crystal by Rutherford backscattering/channeling analyzes (RBS/C), X-ray diffraction patterns (XRD), Raman, nano-indentation test, transmission and absorption spectroscopy, the prismmore » coupling and the end-facet coupling experiments. We also used the secondary ion mass spectrometry (SIMS) to analyze the elements distribution along sputtering depth. Finally, 6 MeV C ions with a fluence of 1 × 10 15 ion/cm 2 irradiated caused the deformation of YSO structure and also influenced the spectral properties and lattice vibrations.« less

Irradiation-induced sensitization and stress corrosion cracking of Type 304 stainless steel core-internal components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, H.M.; Ruther, W.E.; Sanecki, J.E.

1991-08-01

High- and commercial-purity heats of Type 304 stainless steel, obtained from neutron absorber tubes after irradiation to fluence levels of up to 2 {times} 10{sup 21} n{center dot}cm{sup {minus}2} (E > 1 MeV) in two boiling water reactors, were examined by Auger electron spectroscopy to characterize irradiation-induced grain- boundary segregation and depletion of alloying and impurity elements. Segregation of Si, P, Ni, and an unidentified element or compound that gives rise to an Auger energy peak at 59 eV was observed in the commercial-purity heat. Such segregation was negligible in high-purity material, except for Ni. No evidence of S segregationmore » was observed in either material. Cr depletion was more pronounced in the high-purity material than in the commercial-purity material. These observations suggest a synergism between the significant level of impurities and Cr depletion in the commercial-purity heat. In the absence of such synergism, Cr depletion appears more pronounced in the high-purity heat. Initial results of constant-extension-rate tests conducted on the two heats in air an in simulated BWR water were correlated with the results from analysis by Auger electron spectroscopy. 15 refs., 10 figs.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alfaro, R.; Sandoval, A.; Cruz, E.

We have performed radiation tolerance tests on the BCF-99-29MC wavelength shifting fibers and the BC404 plastic scintillator from Bicron as well as on silicon rubber optical couplers. We used the 60Co gamma source at the Instituto de Ciencias Nucleares facility to irradiate 30-cm fiber samples with doses from 50 Krad to 1 Mrad. We also irradiated a 10x10 cm2 scintillator detector with the WLS fibers embedded on it with a 200 krad dose and the optical conectors between the scintillator and the PMT with doses from 100 to 300 krad. We measured the radiation damage on the materials by comparingmore » the pre- and post-irradiation optical transparency as a function of time.« less

Dosimetry in radiobiological studies with the heavy ion beam of the Warsaw cyclotron

NASA Astrophysics Data System (ADS)

Kaźmierczak, U.; Banaś, D.; Braziewicz, J.; Czub, J.; Jaskóła, M.; Korman, A.; Kruszewski, M.; Lankoff, A.; Lisowska, H.; Malinowska, A.; Stępkowski, T.; Szefliński, Z.; Wojewódzka, M.

2015-12-01

The aim of this study was to verify various dosimetry methods in the irradiation of biological materials with a 12C ion beam at the Heavy Ion Laboratory of the University of Warsaw. To this end the number of ions hitting the cell nucleus, calculated on the basis of the Si-detector system used in the set-up, was compared with the number of ion tracks counted in irradiated Solid State Nuclear Track Detectors and with the number of ion tracks detected in irradiated Chinese Hamster Ovary cells processed for the γ-H2AX assay. Tests results were self-consistent and confirmed that the system serves its dosimetric purpose.

Interfacial adhesion improvement in carbon fiber/carbon nanotube reinforced hybrid composites by the application of a reactive hybrid resin initiated by gamma irradiation

NASA Astrophysics Data System (ADS)

Szebényi, G.; Faragó, D.; Lámfalusi, Cs.; Göbl, R.

2018-04-01

Interfacial adhesion is a key factor in composite materials. The effective co-working of the reinforcing materials and matrix is essential for the proper load transfer between them, and to achieve the desired reinforcing effect. In case of nanocomposites, especially carbon nanotube (CNT) reinforced nanocomposites the adhesion between the CNTs and the polymer matrix is poor. To improve the interfacial adhesion and exploit the reinforcing effect of these nanoparticles a two step curable epoxy (EP)/vinylester (VE) hybrid resin system was developed where the EP is cured using hardener in the first step, during the composite production, and in the second step the curing of the VE is initiated by gamma irradiation, which also activates the reinforcing materials and the cured matrix component. A total of six carbon fiber reinforced composite systems were compared with neat epoxy and EP/VE hybrid matrices with and without chemical initiator and MWCNT nano-reinforcement. The effect of gamma irradiation was investigated at four absorbed dose levels. According to our three point bending and interlaminar shear test results the adhesion has improved between all constituents of the composite system. It was demonstrated that gamma irradiation has beneficial effect on the static mechanical, especially interlaminar properties of both micro- and nanocomposites in terms of modulus, strength and interlaminar shear strength.

Development of medicine-intended isotope production technologies at Yerevan Physics Institute

NASA Astrophysics Data System (ADS)

Avetisyan, Albert; Avagyan, Robert; Kerobyan, Ivetta; Dallakyan, Ruben; Harutyunyan, Gevorg; Melkonyan, Aleksandr

2015-05-01

Accelerator-based 99mTc and 123I isotopes production technologies were created and developed at A.Alikhanyan National Science Laboratory (former Yerevan Physics Institute - YerPhI). The method involves the irradiation of natural molybdenum (for 99mTc production) and natural xenon (for 123I production) using high-intensity bremsstrahlung photons from the electron beam of the LUE50 linear electron accelerator located at the YerPhI. We have developed and tested the extraction of 99mTc and 123I from the irradiated natural MoO3 and natural Xe, respectively. The production method has been developed and shown to be successful. The current activity is devoted to creation and development of the technology of direct production 99mTc on the 100Mo as target materials using the proton beam from an IBA C18/18 cyclotron. The proton cyclotron C18/18 (producer - IBA, Belgium) was purchased and will be installed nearby AANL (YerPhI) till end 2014. The 18 MeV protons will be used to investigate accelerator-based schemes for the direct production of 99mTc. Main topics of studies will include experimental measurement of 99mTc production yield for different energies of protons, irradiation times, intensities, development of new methods of 99mTc extraction from irradiated materials, development of target preparation technology, development of target material recovery methods for multiple use and others.

Hot Cell Installation and Demonstration of the Severe Accident Test Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linton, Kory D.; Burns, Zachary M.; Terrani, Kurt A.

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examinemore » postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.« less

Simulated Space Environmental Testing on Thin Films

NASA Technical Reports Server (NTRS)

Russell, Dennis A.; Fogdall, Larry B.; Bohnhoff-Hlavacek, Gail; Connell, John W. (Technical Monitor)

2000-01-01

An exploratory program has been conducted, to irradiate some mature commercial and some experimental polymer films with radiation simulating certain Earth orbits, and to obtain data about the response of each test film's reflective and tensile properties. Protocols to conduct optimized tests were considered and developed to a "prototype" level during this program. Fifteen polymer film specimens were arranged on a specially designed test fixture. The fixture featured controlled exposure areas, and protected the ends of the samples for later gripping in tensile tests. The fixture featured controlled exposure areas, and protected the ends of the samples for later gripping in tensile tests. The fixture containing the films was installed in a clean vacuum chamber where protons, electrons and solar ultraviolet (UV) radiation could simultaneously irradiate the specimens. Near realtime UV rates were used, whereas proton and electron rates were accelerated appreciably to simulate 5 years in orbit during a two month test. Periodically, the spectral reflectance of each film was measured in situ. After the end of the irradiation, final reflectance measurements were made in situ, and solar absorptance values were derived for each specimen. These samples were then measured in air for thermal emittance and for tensile strength. Most specimens withstood the irradiation intact, but with reduced reflectance (increased solar absorptance). Thermal emittance changed slightly in several materials, as did their tensile strength and elongation at break. Conclusions are drawn about the performance of the films. Simulated testing to an expected 5 year dose of electrons and protons consistent with those expected at L2 and 0.98 AU orbits and 100 equivalent solar hours exposure.

Recent Advances in Understanding Radiation Damage in Reactor Cavity Concrete

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosseel, Thomas M; Field, Kevin G; Le Pape, Yann

License renewal up to 60 years and the possibility of subsequent license renewal to 80 years has resulted in a renewed focus on long-term aging of materials at nuclear power plants (NPPs) including concrete. Large irreplaceable sections of most nuclear generating stations include concrete. The Expanded Materials Degradation Analysis, jointly performed by the Department of Energy, the Nuclear Regulatory Commission and Nuclear Industry, identified the urgent need to develop a consistent knowledge base on irradiation effects in concrete (Graves et al., (2014)). Much of the historical mechanical performance data of irradiated concrete (Hilsdorf et al., (1978)) does not accurately reflectmore » typical radiation conditions in NPPs or conditions out to 60 or 80 years of radiation exposure (Kontani et al., (2011)). To address these potential gaps in the knowledge base, the Electric Power Research Institute and Oak Ridge National Laboratory, are working to better understand radiation damage as a degradation mechanism. This paper outlines recent progress toward: 1) assessing the radiation environment in concrete biological shields and defining the upper bound of the neutron and gamma dose levels expected in the biological shield for extended operation, and estimating adsorbed dose, 2) evaluating opportunities to harvest and test irradiated concrete from international NPPs, 3) evaluating opportunities to irradiate prototypical concrete and its components under accelerated neutron and gamma dose levels to establish conservative bounds and inform damage models, 4) developing improved models to enhance the understanding of the effects of radiation on concrete and 5) establishing an international collaborative research and information exchange effort to leverage capabilities and knowledge including developing cooperative test programs to improve confidence in data obtained from various concretes and from accelerated irradiation experiments.« less

Radiation Damage In Reactor Cavity Concrete

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G; Le Pape, Yann; Naus, Dan J

License renewal up to 60 years and the possibility of subsequent license renewal to 80 years has established a renewed focus on long-term aging of nuclear generating stations materials, and recently, on concrete. Large irreplaceable sections of most nuclear generating stations include concrete. The Expanded Materials Degradation Analysis (EMDA), jointly performed by the Department of Energy, the Nuclear Regulatory Commission and Industry, identified the urgent need to develop a consistent knowledge base on irradiation effects in concrete. Much of the historical mechanical performance data of irradiated concrete does not accurately reflect typical radiation conditions in NPPs or conditions out tomore » 60 or 80 years of radiation exposure. To address these potential gaps in the knowledge base, The Electric Power Research Institute and Oak Ridge National Laboratory are working to disposition radiation damage as a degradation mechanism. This paper outlines the research program within this pathway including: (i) defining the upper bound of the neutron and gamma dose levels expected in the biological shield concrete for extended operation (80 years of operation and beyond), (ii) determining the effects of neutron and gamma irradiation as well as extended time at temperature on concrete, (iii) evaluating opportunities to irradiate prototypical concrete under accelerated neutron and gamma dose levels to establish a conservative bound and share data obtained from different flux, temperature, and fluence levels, (iv) evaluating opportunities to harvest and test irradiated concrete from international NPPs, (v) developing cooperative test programs to improve confidence in the results from the various concretes and research reactors, (vi) furthering the understanding of the effects of radiation on concrete (see companion paper) and (vii) establishing an international collaborative research and information exchange effort to leverage capabilities and knowledge.« less

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

Radiation and temperature effects on the time-dependent response of T300/934 graphite/epoxy

NASA Technical Reports Server (NTRS)

Yancey, Robert N.; Pindera, Marek-Jerzy

1988-01-01

A time-dependent characterization study was performed on T300/934 graphite/epoxy in a simulated space environment. Creep tests on irradiated and nonirradiated graphite/epoxy and bulk resin specimens were carried out at temperatures of 72 and 250 F. Irradiated specimens were exposed to dosages of penetrating electron radiation equal to 30 years exposure at GEO-synchronous orbit. Radiation was shown to have little effect on the creep response of both the composite and bulk resin specimens at 72 F while radiation had a significant effect at 250 F. A healing process was shown to be present in the irradiated specimens where broken bonds in the epoxy due to radiation recombined over time to form cross-links in the 934 resin structure. An analytical micromechanical model was also developed to predict the viscoelastic response of fiber reinforced composite materials. The model was shown to correlate well with experimental results for linearly viscoelastic materials with relatively small creep strains.

Light Transmission of Novel CAD/CAM Materials and Their Influence on the Degree of Conversion of a Dual-curing Resin Cement.

PubMed

Egilmez, Ferhan; Ergun, Gulfem; Cekic-Nagas, Isil; Vallittu, Pekka K; Lassila, Lippo V J

To evaluate the light transmission characteristics of different types, shades, and thicknesses of novel CAD/CAM materials and their effect on the degree of conversion (DC) of a dual-curing resin cement. Square specimens (12 × 12 mm2) of three CAD/CAM materials - GC Cerasmart, Lava Ultimate, Vita Enamic - of different thicknesses (1.00, 1.50, and 2.00 mm, n = 5 per thickness) were irradiated with an LED unit. The amount of transmitted light was quantified. Thereafter, the DC% of the dual-curing resin cement (RelyX Ultimate) was recorded after 15 min using Fourier transform infrared spectroscopy. Statistical analysis was performed using two-way ANOVA followed by the Tukey's HSD post-hoc test at a significance level of p < 0.05. Regression analysis was performed to investigate the correlation between the DC and radiant energy, and the DC and thickness. Although the type and shade of CAD/CAM material significantly affect transmitted light irradiation (p < 0.0001), degrees of conversion are similar when the CAD/CAM material or material shade were taken into consideration (p > 0.05). Conversely, material thickness significantly affected light transmission (p < 0.0001) and DC (p < 0.0001). Multiple effects of material, shade, and thickness did not significantly affect the evaluated parameters (p = 0.638 for light irradiation; p = 0.637 for DC). Linear regression analysis showed a correlation between delivered energy and DC% results of the Vita Enamic (R² = 0.4169, p < 0.0001). Reduced light transmission in 2-mm-thick specimens of all CAD/CAM materials indicates that proper curing of the cement beneath CAD/CAM materials should be ensured.

RERTR-12 Post-irradiation Examination Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, Francine; Williams, Walter; Robinson, Adam

2015-02-01

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-propertymore » changes of the fuel and cladding.« less

The natural aging of austenitic stainless steels irradiated with fast neutrons

NASA Astrophysics Data System (ADS)

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.

Synthesis and characterization of zinc chloride containing poly(acrylic acid) hydrogel by gamma irradiation

NASA Astrophysics Data System (ADS)

Park, Jong-Seok; Kuang, Jia; Gwon, Hui-Jeong; Lim, Youn-Mook; Jeong, Sung-In; Shin, Young-Min; Seob Khil, Myung; Nho, Young-Chang

2013-07-01

In this study, the characterization of zinc chloride incorporated into a poly(acrylic acid) (PAAc) hydrogel prepared by gamma-ray irradiation was investigated. Zinc chloride powder with different concentrations was dissolved in the PAAc solution, and it was crosslinked with gamma-ray irradiation. The effects of various parameters such as zinc ion concentration and irradiation doses on characteristics of the hydrogel formed were investigated in detail for obtaining an antibacterial wound dressing. In addition, the gel content, pH-sensitive (pH 4 or 7) swelling ratio, and UV-vis absorption spectra of the zinc particles in the hydrogels were characterized. Moreover, antibacterial properties of these new materials against Staphylococcus aureus and Escherichia coli strains were observed on solid growth media. The antibacterial tests indicated that the zinc chloride containing PAAc hydrogels have good antibacterial activity.

Effect of γ-IRRADIATION on the Mechanical Properties of Al-Cu Alloy

NASA Astrophysics Data System (ADS)

Abo-Elsoud, M.; Ismail, H.; Sobhy, Maged S.

SEM observations and Vickers hardness tests were performed to identify the irradiation effects. γ-irradiation effect during the aging hardening process can be explained depending on the composition of the alloy and is used to derive quantitative information on the kinetics of the transformation precipitates. Increasing the Cu content of an Al-Cu alloy can improve the aging hardness. The present results of the hardness behavior, with SEM observations of surveillance specimens at different doses, suggest that the radiation-induced defects are probably complex valence-solute clusters. These clusters act as nuclei for the precipitation of θ-Al2Cu type. This can be effectively utilized to study the systematics of nucleation of precipitates at vacancy-type defects. γ-irradiation probably plays the key role in defects responsible for material strengthening and embrittlement.

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

NASA Astrophysics Data System (ADS)

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

2002-12-01

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dilli, S.; Garnett, J.L.

In a study of the mechanism of radiation-induced reactions with cellulose, the radiation chemistry of a number of simple crystalline sugars and polysaccharides was investigated. All solid sugars were irradiated in both air and vacuum to total doses of 10/sup 7/ and 10/sup 8/ rad in a Co/sup 60/ source at 1.25 megarad/hr. Examination of the gaseous products showed that irradiated cellobiose yields a relatively high hydrogen content, while amylose and amylopectin (amorphous) at doses of 10/sup 8/ rads show the presence of no water vapor. The same gases were also reported to result from the irradiation of crystalline glucose.more » In the solid state, the majority of the saccharides showed marked color changes following irradiation. The se colors, which were unchanged after 2 yr, varied from bright yellow with amylose, amylopectin, and glucose to dark brown with sucrose. Melibiose, lyxose, and fucose showed no change. Aqueous solutions of the irradiated materials were distinctly acid (pH 3-5). Paper chromatographic examination of the aqueous carbohydrate solutions showed no differences for the carbohydrates irradiated in air or vacuum. In marked contrast to the monosaccharides, the radiation stability of disaccharides was relatively poor. Each of the disaccharides tested yielded a large number of degradation products of which the component monosaccharides predominated. With the irradiated polysaccharides (amylose, amylopectin, and cellulose) characteristic chromatographic behavior in all solvents was a trail of reducing material often running to the end of the sheet. In the chromatography of all three compounds, only a faint spot corresponding to glucose was observed. Data are tabulated for the gas yields (H/sub 2/, CH/sub 4/, H/sub 2/O, CO, CO/sub 2/) and Rf values of the products from the irradiation of amylose, amylopectin, cellulose, trehalose, cellobiose, maltose, sucrose, lactose, and melibiose. (BBB)« less

Setup for in situ x-ray diffraction study of swift heavy ion irradiated materials.

PubMed

Kulriya, P K; Singh, F; Tripathi, A; Ahuja, R; Kothari, A; Dutt, R N; Mishra, Y K; Kumar, Amit; Avasthi, D K

2007-11-01

An in situ x-ray diffraction (XRD) setup is designed and installed in the materials science beam line of the Pelletron accelerator at the Inter-University Accelerator Centre for in situ studies of phase change in swift heavy ion irradiated materials. A high vacuum chamber with suitable windows for incident and diffracted X-rays is integrated with the goniometer and the beamline. Indigenously made liquid nitrogen (LN2) temperature sample cooling unit is installed. The snapshots of growth of particles with fluence of 90 MeV Ni ions were recorded using in situ XRD experiment, illustrating the potential of this in situ facility. A thin film of C60 was used to test the sample cooling unit. It shows that the phase of the C60 film transforms from a cubic lattice (at room temperature) to a fcc lattice at around T=255 K.

Setup for in situ x-ray diffraction study of swift heavy ion irradiated materials

NASA Astrophysics Data System (ADS)

Kulriya, P. K.; Singh, F.; Tripathi, A.; Ahuja, R.; Kothari, A.; Dutt, R. N.; Mishra, Y. K.; Kumar, Amit; Avasthi, D. K.

2007-11-01

An in situ x-ray diffraction (XRD) setup is designed and installed in the materials science beam line of the Pelletron accelerator at the Inter-University Accelerator Centre for in situ studies of phase change in swift heavy ion irradiated materials. A high vacuum chamber with suitable windows for incident and diffracted X-rays is integrated with the goniometer and the beamline. Indigenously made liquid nitrogen (LN2) temperature sample cooling unit is installed. The snapshots of growth of particles with fluence of 90MeV Ni ions were recorded using in situ XRD experiment, illustrating the potential of this in situ facility. A thin film of C60 was used to test the sample cooling unit. It shows that the phase of the C60 film transforms from a cubic lattice (at room temperature) to a fcc lattice at around T =255K.

Silver-Teflon contamination UV radiation study

NASA Technical Reports Server (NTRS)

Muscari, J. A.

1978-01-01

Silver-Teflon (Ag/FEP) is planned to be used as the thermal control material covering the radiator surfaces on the shuttle orbiter payload bay doors. These radiators require the use of materials that have a very low solar absorptance and a high emittance for heat rejection. However, operationally, materials used on these critical radiator surfaces, such as silver-Teflon, will be exposed to a variety of conditions which include both the natural as well as the induced environments from the Shuttle Orbiter. A complete test facility was assembled, and detailed test procedures and a test matrix were developed. Measurements of low solar absorptance were taken before and after contamination, at intervals during irradiation, and after sample cleaning to fulfill all the requirements.

Testing piezoelectric sensors in a nuclear reactor environment

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

2017-02-01

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

Structural responses of metallic glasses under neutron irradiation.

PubMed

Yang, L; Li, H Y; Wang, P W; Wu, S Y; Guo, G Q; Liao, B; Guo, Q L; Fan, X Q; Huang, P; Lou, H B; Guo, F M; Zeng, Q S; Sun, T; Ren, Y; Chen, L Y

2017-12-01

Seeking nuclear materials that possess a high resistance to particle irradiation damage is a long-standing issue. Permanent defects, induced by irradiation, are primary structural changes, the accumulation of which will lead to structural damage and performance degradation in crystalline materials served in nuclear plants. In this work, structural responses of neutron irradiation in metallic glasses (MGs) have been investigated by making a series of experimental measurements, coupled with simulations in ZrCu amorphous alloys. It is found that, compared with crystalline alloys, MGs have some specific structural responses to neutron irradiation. Although neutron irradiation can induce transient vacancy-like defects in MGs, they are fully annihilated after structural relaxation by rearrangement of free volumes. In addition, the rearrangement of free volumes depends strongly on constituent elements. In particular, the change in free volumes occurs around the Zr atoms, rather than the Cu centers. This implies that there is a feasible strategy for identifying glassy materials with high structural stability against neutron irradiation by tailoring the microstructures, the systems, or the compositions in alloys. This work will shed light on the development of materials with high irradiation resistance.

Mechanical Characterization of the Iter Mock-Up Insulation after Reactor Irradiation

NASA Astrophysics Data System (ADS)

Prokopec, R.; Humer, K.; Fillunger, H.; Maix, R. K.; Weber, H. W.

2010-04-01

The ITER mock-up project was launched in order to demonstrate the feasibility of an industrial impregnation process using the new cyanate ester/epoxy blend. The mock-up simulates the TF winding pack cross section by a stainless steel structure with the same dimensions as the TF winding pack at a length of 1 m. It consists of 7 plates simulating the double pancakes, each of them is wrapped with glass fiber/Kapton sandwich tapes. After stacking the 7 plates, additional insulation layers are wrapped to simulate the ground insulation. This paper presents the results of the mechanical quality tests on the mock-up pancake insulation. Tensile and short beam shear specimens were cut from the plates extracted from the mock-up and tested at 77 K using a servo-hydraulic material testing device. All tests were repeated after reactor irradiation to a fast neutron fluence of 1×1022 m-2 (E>0.1 MeV). In order to simulate the pulsed operation of ITER, tension-tension fatigue measurements were performed in the load controlled mode. Initial results show a high mechanical strength as expected from the high number of thin glass fiber layers, and an excellent homogeneity of the material.

Vanadium—lithium in-pile loop for comprehensive tests of vanadium alloys and multipurpose coatings

NASA Astrophysics Data System (ADS)

Lyublinski, I. E.; Evtikhin, V. A.; Ivanov, V. B.; Kazakov, V. A.; Korjavin, V. M.; Markovchev, V. K.; Melder, R. R.; Revyakin, Y. L.; Shpolyanskiy, V. N.

1996-10-01

The reliable information on design and material properties of self-cooled Li sbnd Li blanket and liquid metal divertor under neutron radiation conditions can be obtained using the concept of combined technological and material in-pile tests in a vanadium—lithium loop. The method of in-pile loop tests includes studies of vanadium—base alloys resistance, weld resistance under mechanical stress, multipurpose coating formation processes and coatings' resistance under the following conditions: high temperature (600-700°C), lithium velocities up to 10 m/s, lithium with controlled concentration of impurities and technological additions, a neutron load of 0.4-0.5 MW/m 2 and level of irradiation doses up to 5 dpa. The design of such an in-pile loop is considered. The experimental data on corrosion and compatibility with lithium, mechanical properties and welding technology of the vanadium alloys, methods of coatings formation and its radiation tests in lithium environment in the BOR-60 reactor (fast neutron fluence up to 10 26 m -2, irradiation temperature range of 500-523°C) are presented and analyzed as a basis for such loop development.

Mechanical property changes induced in structural alloys by neutron irradiations with different helium to displacement ratios*1

NASA Astrophysics Data System (ADS)

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

1988-07-01

Effects of helium on mechanical properties of irradiated structural materials are reviewed. In particular, variations in response to the ratio of helium to displacement damage serve as the focus. Ductility in creep and tensile tests is emphasized. A variety of early work has led to the current concentration on helium effects for fusion reactor materials applications. A battery of techniques has been developed by which the helium to displacement ratio can be varied. Our main discussion is devoted to the techniques of spectral tailoring and isotopic alloying currently of interest for mixed-spectrum reactors. Theoretical models of physical mechanisms by which helium interacts with displacement damage have been developed in terms of hardening to dislocation motion and grain boundary cavitation. Austenitic stainless steels, ferritic/martensitic steels and vanadium alloys are considered. In each case, work at low strain rates, where the main problems may lie, at the helium to displacement ratios appropriate to fusion reactor materials is lacking. Recent experimental evidence suggests that both in-reactor and high helium results may differ substantially from post-irradiation or low helium results. It is suggested that work in these areas is especially needed.

Application of thin layer activation technique for monitoring corrosion of carbon steel in hydrocarbon processing environment.

PubMed

Saxena, R C; Biswal, Jayashree; Pant, H J; Samantray, J S; Sharma, S C; Gupta, A K; Ray, S S

2018-05-01

Acidic crude oil transportation and processing in petroleum refining and petrochemical operations cause corrosion in the pipelines and associated components. Corrosion monitoring is invariably required to test and prove operational reliability. Thin Layer Activation (TLA) technique is a nuclear technique used for measurement of corrosion and erosion of materials. The technique involves irradiation of material with high energy ion beam from an accelerator and measurement of loss of radioactivity after the material is subjected to corrosive environment. In the present study, TLA technique has been used to monitor corrosion of carbon steel (CS) in crude oil environment at high temperature. Different CS coupons were irradiated with a 13 MeV proton beam to produce Cobalt-56 radioisotope on the surface of the coupons. The corrosion studies were carried out by subjecting the irradiated coupons to a corrosive environment, i.e, uninhibited straight run gas oil (SRGO) containing known amount of naphthenic acid (NA) at high temperature. The effects of different parameters, such as, concentration of NA, temperature and fluid velocity (rpm) on corrosion behaviour of CS were studied. Copyright © 2018 Elsevier Ltd. All rights reserved.

Testing in Support of Fission Surface Power System Qualification

NASA Technical Reports Server (NTRS)

Houts, Mike; Bragg-Sitton, Shannon; Godfroy, Tom; Martin, Jim; Pearson, Boise; VanDyke, Melissa

2007-01-01

The strategy for qualifying a FSP system could have a significant programmatic impact. The US has not qualified a space fission power system since launch of the SNAP-10A in 1965. This paper explores cost-effective options for obtaining data that would be needed for flight qualification of a fission system. Qualification data could be obtained from both nuclear and non-nuclear testing. The ability to perform highly realistic nonnuclear testing has advanced significantly throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modern FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.) and extensive data to be taken from the core region. For transient testing, pin power during a transient is calculated based on the reactivity feedback that would occur given measured values of test article temperature and/or dimensional changes. The reactivity feedback coefficients needed for the test are either calculated or measured using cold/warm zero-power criticals. In this way non-nuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. FSP fuels and materials are typically chosen to ensure very high confidence in operation at design burnups, fluences, and temperatures. However, facilities exist (e.g. ATR, HFIR) for affordably performing in-pile fuel and materials irradiations, if such testing is desired. Ex-core materials and components (such as alternator materials, control drum drives, etc.) could be irradiated in university or DOE reactors to ensure adequate radiation resistance. Facilities also exist for performing warm and cold zero-power criticals.

Advancements in internationally accepted standards for radiation processing

NASA Astrophysics Data System (ADS)

Farrar, Harry; Derr, Donald D.; Vehar, David W.

1993-10-01

Three subcommittees of the American Society for Testing and Materials (ASTM) are developing standards on various aspects of radiation processing. Subcommittee E10.01 "Dosimetry for Radiation Processing" has published 9 standards on how to select and calibrate dosimeters, where to put them, how many to use, and how to use individual types of dosimeter systems. The group is also developing standards on how to use gamma, electron beam, and x-ray facilities for radiation processing, and a standard on how to treat dose uncertainties. Efforts are underway to promote inclusion of these standards into procedures now being developed by government agencies and by international groups such as the United Nations' International Consultative Group on Food Irradiation (ICGFI) in order to harmonize regulations and help avoid trade barriers. Subcommittee F10.10 "Food Processing and Packaging" has completed standards on good irradiation practices for meat and poultry and for fresh fruits, and is developing similar standards for the irradiation of seafood and spices. These food-related standards are based on practices previously published by ICGFI. Subcommittee E10.07 on "Radiation Dosimetry for Radiation Effects on Materials and Devices" principally develops standards for determining doses for radiation hardness testing of electronics. Some, including their standards on the Fricke and TLD dosimetry systems are equally useful in other radiation processing applications.

Testing and optical modeling of novel concentrating solar receiver geometries to increase light trapping and effective solar absorptance

NASA Astrophysics Data System (ADS)

Yellowhair, Julius; Ho, Clifford K.; Ortega, Jesus D.; Christian, Joshua M.; Andraka, Charles E.

2015-09-01

Concentrating solar power receivers are comprised of panels of tubes arranged in a cylindrical or cubical shape on top of a tower. The tubes contain heat-transfer fluid that absorbs energy from the concentrated sunlight incident on the tubes. To increase the solar absorptance, black paint or a solar selective coating is applied to the surface of the tubes. However, these coatings degrade over time and must be reapplied, which reduces the system performance and increases costs. This paper presents an evaluation of novel receiver shapes and geometries that create a light-trapping effect, thereby increasing the effective solar absorptance and efficiency of the solar receiver. Several prototype shapes were fabricated from Inconel 718 and tested in Sandia's solar furnace at an irradiance of ~30 W/cm2. Photographic methods were used to capture the irradiance distribution on the receiver surfaces. The irradiance profiles were compared to results from raytracing models. The effective solar absorptance was also evaluated using the ray-tracing models. Results showed that relative to a flat plate, the new geometries could increase the effective solar absorptance from 86% to 92% for an intrinsic material absorptance of 86%, and from 60% to 73% for an intrinsic material absorptance of 60%.

Heavy-section steel technology and irradiation programs-retrospective and prospective views

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nanstad, Randy K; Bass, Bennett Richard; Rosseel, Thomas M

In 1965, the Atomic Energy Commission (AEC), at the advice of the Advisory Committee on Reactor Safeguards (ACRS), initiated the process that resulted in the establishment of the Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL). Dr. Spencer H. Bush of Battelle Northwest Laboratory, the man being honored by this symposium, representing the ACRS, was one of the Staff Advisors for the program and helped to guide its technical direction. In 1989, the Heavy-Section Steel Irradiation (HSSI) Program, formerly the HSST task on irradiation effects, was formed as a separate program, and this year the HSST/HSSImore » Programs, sponsored by the U.S. Nuclear Regulatory Commission (USNRC), celebrate 40 years of continuous research oriented toward the safety of light-water nuclear reactor pressure vessels. This paper presents a summary of results from those programs with a view to future activities. The HSST Program was established in 1967 and initially included extensive investigations of heavy-section low-alloy steel plates, forgings, and welds, including metallurgical studies, mechanical properties, fracture toughness (quasi-static and dynamic), fatigue crack-growth, and crack arrest toughness. Also included were irradiation effects studies, thermal shock analyses, testing of thick-section tensile and fracture specimens, and non-destructive testing. In the subsequent decades, the HSST Program conducted extensive large-scale experiments with intermediate-size vessels (with varying size flaws) pressurized to failure, similar experiments under conditions of thermal shock and even pressurized thermal shock (PTS), wide-plate crack arrest tests, and biaxial tests with cruciform-shaped specimens. Extensive analytical and numerical studies accompanied these experiments, including the development of computer codes such as the recent Fracture Analysis of Vessels Oak Ridge (FAVOR) code currently being used for PTS evaluations. In the absence of radiation damage to the RPV, fracture of the vessel is improbable. However, exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The HSSI Program has conducted a series of experiments to assess the effects of neutron irradiation on RPV material behavior, especially fracture toughness. These studies have included RPV plates and welds, varying chemical compositions, and fracture toughness specimens up to 4 in. thickness. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. Results from the HSST and HSSI Program are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety.« less

Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.

2012-06-01

Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

Neuropsychologic assessment of long-term survivors of childhood leukemia.

PubMed

Pfefferbaum-Levine, B; Copeland, D R; Fletcher, J M; Ried, H L; Jaffe, N; McKinnon, W R

1984-01-01

Thirty-two long-term survivors of childhood leukemia who were followed up at the University of Texas M. D. Anderson Hospital were evaluated with a battery of 17 neuropsychologic tests. These tests were selected to assess the development of cognitive skills and functions associated with brain impairment in children. Statistically significant differences were found between the group of children given CNS irradiation and the nonirradiated group on full-scale IQ and verbal IQ scores, mathematics skills, constructional skills, and memory for spatial material. Of particular interest was the absence of differences in language-based measures of verbal memory and the presence of group differences on measures of memory for spatial material. While the sample size was small, the findings delineate specific areas likely to be affected. These results indicate the need for caution when including cranial irradiation in CNS prophylaxis. When any CNS treatment is given, it seems appropriate that provisions be made for assessment and remediation of affected skills.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Xiang; Sokolov, Mikhail A.; Linton, Kory D.

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 neutronmore » 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.« less

Development and testing of a superconducting link for an IR detector

NASA Technical Reports Server (NTRS)

Caton, R.; Selim, R.

1991-01-01

The development and testing of a ceramic superconducting link for an infrared detector is summarized. Areas of study included the materials used, the electrical contacts, radiation and temperature cycling effects, aging, thermal conductivity, and computer models of an ideal link. Materials' samples were processed in a tube furnace at temperatures of 840 C to 865 C for periods up to 17 days and transition temperatures and critical current densities were recorded. The project achieved better quality high superconducting transition temperature material through improved processing and also achieved high quality electrical contacts. Studies on effects of electron irradiation, temperature cycling, and aging on superconducting properties indicate that the materials will be suitable for space applications. Various presentations and publications on the study's results are reported.

Detailed measurements of local thickness changes for U-7Mo dispersion fuel plates with Al-3.5Si matrix after irradiation at different powers in the RERTR-9B experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Williams, Walter; Robinson, Adam; Wachs, Dan; Moore, Glenn; Crawford, Doug

2017-10-01

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.

AGR-5/6/7 Irradiation Test Predictions using PARFUME

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skerjanc, William F.

PARFUME, (PARticle FUel ModEl) a fuel performance modeling code used for high temperature gas-cooled reactors (HTGRs), was used to model the Advanced Gas Reactor (AGR)-5/6/7 irradiation test using predicted physics and thermal hydraulics data. The AGR-5/6/7 test consists of the combined fifth, sixth, and seventh planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-5/6/7 test train is a multi-capsule, instrumented experiment that is designed for irradiation in the 133.4-mm diameter north east flux trap (NEFT) position of Advanced Test Reactor (ATR). Each capsule contains compacts filled with uranium oxycarbide (UCO) unaltered fuel particles. This report documents themore » calculations performed to predict the failure probability of tristructural isotropic (TRISO)-coated fuel particles during the AGR-5/6/7 experiment. In addition, this report documents the calculated source term from the driver fuel. The calculations include modeling of the AGR-5/6/7 irradiation that is scheduled to occur from October 2017 to April 2021 over a total of 13 ATR cycles, including nine normal cycles and four Power Axial Locator Mechanism (PALM) cycle for a total between 500 – 550 effective full power days (EFPD). The irradiation conditions and material properties of the AGR-5/6/7 test predicted zero fuel particle failures in Capsules 1, 2, and 4. Fuel particle failures were predicted in Capsule 3 due to internal particle pressure. These failures were predicted in the highest temperature compacts. Capsule 5 fuel particle failures were due to inner pyrolytic carbon (IPyC) cracking causing localized stresses concentrations in the SiC layer. This capsule predicted the highest particle failures due to the lower irradiation temperature. In addition, shrinkage of the buffer and IPyC layer during irradiation resulted in formation of a buffer-IPyC gap. The two capsules at the two ends of the test train, Capsules 1 and 5 experienced the smallest buffer-IPyC gap formation due to the lower irradiation fluences and temperatures. Capsule 3 experienced the largest buffer-IPyC gap formation of just under 24 µm. The release fraction of fission products Ag, Cs, and Sr silver (Ag), cesium (Cs), and strontium (Sr) vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 3, reaching up to 84.8% for the TRISO fuel particles. The release fraction of the other two fission products, Cs and Sr are much smaller and, in most cases, less than 1%. The notable exception is again in Capsule 3, where the release fraction for Cs and Sr reach up to 9.7% and 19.1%, respectively.« less

Glass transition temperature of hard chairside reline materials after post-polymerisation treatments.

PubMed

Urban, Vanessa M; Machado, Ana L; Alves, Marinês O; Maciel, Adeilton P; Vergani, Carlos E; Leite, Edson R

2010-09-01

This study evaluated the effect of post-polymerisation treatments on the glass transition temperature (T(g)) of five hard chairside reline materials (Duraliner II-D, Kooliner-K, New Truliner-N, Ufi Gel hard-U and Tokuso Rebase Fast-T). Specimens (10 x 10 x 1 mm) were made following the manufacturers' instructions and divided into three groups (n = 5). Control group specimens were left untreated. Specimens from the microwave group were irradiated with pre-determined power/time combinations, and specimens from the water-bath group were immersed in hot water at 55 degrees C for 10 min. Glass transition ( degrees C) was performed by differential scanning calorimetry. Data were analysed using anova, followed by post hoc Tukey's test (alpha = 0.05). Both post-polymerisation treatments promoted a significant (p < 0.05) increase in the T(g) of reline material K. Materials K, D and N showed the lowest T(g) (p < 0.05). No significant difference between T and U specimens was observed. Post-polymerisation treatments improved the glass transition of material Kooliner, with the effect being more pronounced for microwave irradiation.

Developmental status of thermionic materials.

NASA Technical Reports Server (NTRS)

Yang, L.; Chin, J.

1972-01-01

Description of the reference materials selected for the major components of the unit cell of a thermionic pile element (TFE), the out-of-pile and in-pile test results, and current efforts for improving the life and performance of thermionic fuel elements. The component materials are required to withstand the fuel burnup and fast neutron fluence dictated by the thermionic reactor system. Tungsten was selected as the cladding material because of its compatibility with both the carbide and the oxide fuel materials. Niobium was selected as the collector material because its thermal expansion coefficient matches closely with that of the thin aluminum oxide layer used to electrically insulate the collector from the TFE sheath. An unfueled converter has performed stably over 41,000 hr. Accelerated irradiation tests have attained burnups equivalent to that for 40,000 hr of the thermionic reactor under consideration.

The effect of the initial microstructure in terms of sink strength on the ion-irradiation-induced hardening of ODS alloys studied by nanoindentation

NASA Astrophysics Data System (ADS)

Duan, Binghuang; Heintze, Cornelia; Bergner, Frank; Ulbricht, Andreas; Akhmadaliev, Shavkat; Oñorbe, Elvira; de Carlan, Yann; Wang, Tieshan

2017-11-01

Oxide dispersion strengthened (ODS) Fe-Cr alloys are promising candidates for structural components in nuclear energy production. The small grain size, high dislocation density and the presence of particle matrix interfaces may contribute to the improved irradiation resistance of this class of alloys by providing sinks and/or traps for irradiation-induced point defects. The extent to which these effects impede hardening is still a matter of debate. To address this problem, a set of alloys of different grain size, dislocation density and oxide particle distribution were selected. In this study, three-step Fe-ion irradiation at both 300 °C and 500 °C up to 10 dpa was used to introduce damage in five different materials including three 9Cr-ODS alloys, one 14Cr-ODS alloy and one 14Cr-non-ODS alloy. Electron backscatter diffraction (EBSD), transmission electron microscopy (TEM), small angle neutron scattering (SANS), and nanoindentation testing were applied, the latter before and after irradiation. Significant hardening occurred for all materials and temperatures, but it is distinctly lower in the 14Cr alloys and also tends to be lower at the higher temperature. The possible contribution of Cr-rich α‧-phase particles is addressed. The impact of grain size, dislocation density and particle distribution is demonstrated in terms of an empirical trend between total sink strength and hardening.

Radiation sterilization of medical products in the Philippines

NASA Astrophysics Data System (ADS)

Singson, C.; Carmona, C.; de Guzman, Z.; Barrun, W.; Lanuza, L.

This paper presents the results of a comprehensive investigation of the biological, microbiological, physico-chemical, and dosimetry aspects of using gamma irradiation for the sterilization of locally manufactured medical products and pharmaceuticals. The objective of this study is to determine the technological feasibility of radiation sterilization for the said products in the Philippines. Hence, the materials used were directly obtained from local manufacturers. They are polyvinyl chloride or polyethylene based medical plastic disposables namely: absorbent cotton, surgical gauze, bandage, visceral packs, and some antibiotics and opthalmic ointments. The gamma facility of the Philippine Atomic Energy Commission was used for the irradiation. Result of biological studies indicate no signs of toxicity on experimental mice injected with extracts from irradiated samples. The contaminants are identified as Pseudomonas Sp. Staphyloccocus Aureus and Bacillus Subtilis. The D 10 values of survivors of higher doses ranged below 0.235 Megarad suggesting that these contaminants can be eliminated by the generally used sterilizing dose of 2.5 Mrads. The physico-chemical tests did not indicate any significant degradation of the irradiated products. Opthalmic and topical antibiotic ointments showed no marked decrease in potency. Fading tests on dosimeters used showed that red perspex is a more efficient dosimeter than clear perspex when irradiation time is prolonged. These studies indicated that radiation sterilization is technically feasible for locally manufactured medical products.

HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.

High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Testmore » Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two primary goals. First, the test was intended to assess the retention of fission products in loose kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material comprising the compact. Second, this test served as an evaluation of the FACS fission product condensation plate collection efficiency.« less

LWRS ATR Irradiation Testing Readiness Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett

2012-09-01

The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Testmore » Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics« less

Effects of neutron irradiation at 70-200 °C in beryllium

NASA Astrophysics Data System (ADS)

Chakin, V. P.; Kazakov, V. A.; Melder, R. R.; Goncharenko, Yu. D.; Kupriyanov, I. B.

2002-12-01

At present beryllium is considered one of the metals to be used as a plasma facing and blanket material. This paper presents the investigations of several Russian beryllium grades fabricated by HE and HIP technologies. Beryllium specimens were irradiated in the SM reactor at 70-200 °C up to a neutron fluence (0.6-3.9)×10 22 cm -2 ( E>0.1 MeV). It is shown that the relative mass decrease of beryllium specimens that were in contact with the water coolant during irradiation achieved the value >1.5% at the maximum dose. Swelling was in the range of 0.2-1.5% and monotonically increasing with the neutron dose. During mechanical tensile and compression tests one could observe the absolute brittle destruction of the irradiated specimens at the reduced strength level in comparison to the initial state. A comparatively higher level of brittle strength was observed on beryllium specimens irradiated at 200 °C. The basic type of destruction of the irradiated beryllium specimens is brittle and intergranular with some fraction of transgranular chip.

Qualification of heavy water based irradiation device in the JSI TRIGA reactor for irradiations of FT-TIMS samples for nuclear safeguards

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Kolšek, Aljaž; Fauré, Anne-Laure; Pottin, Anne-Claire; Pointurier, Fabien; Snoj, Luka

2018-03-01

The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications.

[Effect of sterilisation with formaldehyde, gamma irradiation and ethylene oxide on the properties of polyethylene joint replacement components].

PubMed

Fulín, P; Pokorný, D; Slouf, M; Vacková, T; Dybal, J; Sosna, A

2014-01-01

Each method of sterilisation has some effect on the structure and properties of UHMWPE and thus also on joint replacement longevity. This study was designed to compare, using objective methods of measurement, several kinds of sterilisation and to recommend the one which has the best prospect for making joint replacements last longer. Two groups of UHMWPE samples were tested. Group 1 included virgin GUR 1020 polyethylene, non-modified and non-sterilised (Meditech, Germany). Group 2 comprised of three sets of samples sterilised with formaldehyde, gamma irradiation and ethylene oxide, respectively. In both groups, physicochemical properties were assessed by infrared spectroscopy (IR), and the oxidation (OI) and trans-vinyl (VI) indices, which show the degree of oxidation of a material, were determined. Free-radical concentrations were measured by the method of electron spin resonance (ESR). The mechanical properties of each sample were studied using small punch tests (SPT) and testing microhardness (MH). Any change in mechanical properties can affect, to various degrees, the quality and longevity of a prosthetic joint. The samples sterilised by gamma irradiation showed higher values of both the OI (0.37) and the VI index (0.038) than the other samples (OI, 0.02 to 0.05 and VI, 0). Also, the free-radical concentration was detectable only in the gamma-sterilised sample. Values obtained for mechanical properties were as follows: peak load in the range of 58.48 N (gamma irradiation) to 59.60 N (ethylene oxide); ultimate load in the range of 46.69 N (gamma irradiation) to 57.50 N (ethylene oxide); ultimate displacement in the range of 4.29 mm (gamma irradiation) to 4.58 mm (virgin polyethylene and formaldehyde); and work to failure in the range of 185.18 mJ (gamma irradiation) to 205.89 mJ (virgin polyethylene). Microhardness values were obtained in the following ranges: 41.2 to 44.6 MPa (virgin polyethylene); 40.2 to 44.1 MPa (formaldehyde); 46.1 to 49.3 MPa (gamma irradiation); and 40.3 to 44.2 MPa (ethylene oxide). The samples sterilised with formaldehyde and ethylene oxide have mechanical properties very similar to virgin polyethylene, they are not damaged by oxidation and do not contain free radicals. Owing to these characteristics, the immediate and long-term oxidation stability of the three samples is higher. The sample sterilised by gamma irradiation showed the presence of free radicals and immediate and long-term oxidative degradation. This results in the deterioration of mechanical properties and the growth of crystallinity due to enhanced oxidation and leads to higher polyethylene microhardness. Sterilisation with gamma irradiation results in oxidative degradation and mechanical property deterioration, which is one of the potential risks of a shorter life span of joint replacements. The use of ethylene oxide or formaldehyde does not change polymer properties nor has any effect on oxidation of materials. Therefore, a longer life expectancy of the joint replacements sterilised with ethylene oxide can be expected. The life span of their joint replacements is a key issue for the patients. Prosthetic joint loosening is painful and the patient often requires re-implantation. A higher number of re-implantations is associated with higher costs for the institution involved and, consequently, for the whole health care system. Although this study basically deals with chemical issues, it informs the surgeon of the latest developments leading to the improvement of implanted materials, which can increase the life expectancy of joint replacements and patients' satisfaction.

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

NASA Astrophysics Data System (ADS)

Byun, T. S.; Farrell, K.; Lee, E. H.; Mansur, L. K.; Maloy, S. A.; James, M. R.; Johnson, W. R.

2002-05-01

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ammigan, Kavin; et al.

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

Shuttle filter study. Volume 2: Contaminant generation and sensitivity studies

NASA Technical Reports Server (NTRS)

1974-01-01

Contaminant generation studies were conducted at the component level using two different methods, radioactive tracer technique and gravimetric analysis test procedure. Both of these were reduced to practice during this program. In the first of these methods, radioactively tagged components typical of those used in spacecraft were studied to determine their contaminant generation characteristics under simulated operating conditions. Because the purpose of the work was: (1) to determine the types and quantities of contaminants generated; and (2) to evaluate improved monitoring and detection schemes, no attempt was made to evaluate or qualify specific components. The components used in this test program were therefore not flight hardware items. Some of them had been used in previous tests; some were obsolete; one was an experimental device. In addition to the component tests, various materials of interest to contaminant and filtration studies were irradiated and evaluated for use as autotracer materials. These included test dusts, plastics, valve seat materials, and bearing cage materials.

The Gottingen Minipig Is a Model of the Hematopoietic Acute Radiation Syndrome: G-Colony Stimulating Factor Stimulates Hematopoiesis and Enhances Survival From Lethal Total-Body γ-Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moroni, Maria, E-mail: maria.moroni@usuhs.edu; Ngudiankama, Barbara F.; Christensen, Christine

Purpose: We are characterizing the Gottingen minipig as an additional large animal model for advanced drug testing for the acute radiation syndrome (ARS) to enhance the discovery and development of novel radiation countermeasures. Among the advantages provided by this model, the similarities to human hematologic parameters and dynamics of cell loss/recovery after irradiation provide a convenient means to compare the efficacy of drugs known to affect bone marrow cellularity and hematopoiesis. Methods and Materials: Male Gottingen minipigs, 4 to 5 months old and weighing 9 to 11 kg, were used for this study. We tested the standard off-label treatment formore » ARS, rhG-CSF (Neupogen, 10 μg/kg/day for 17 days), at the estimated LD70/30 total-body γ-irradiation (TBI) radiation dose for the hematopoietic syndrome, starting 24 hours after irradiation. Results: The results indicated that granulocyte colony stimulating factor (G-CSF) enhanced survival, stimulated recovery from neutropenia, and induced mobilization of hematopoietic progenitor cells. In addition, the administration of G-CSF resulted in maturation of monocytes/macrophages. Conclusions: These results support continuing efforts toward validation of the minipig as a large animal model for advanced testing of radiation countermeasures and characterization of the pathophysiology of ARS, and they suggest that the efficacy of G-CSF in improving survival after total body irradiation may involve mechanisms other than increasing the numbers of circulating granulocytes.« less

Small-scale characterisation of irradiated nuclear materials: Part II nanoindentation and micro-cantilever testing of ion irradiated nuclear materials

NASA Astrophysics Data System (ADS)

Armstrong, D. E. J.; Hardie, C. D.; Gibson, J. S. K. L.; Bushby, A. J.; Edmondson, P. D.; Roberts, S. G.

2015-07-01

This paper demonstrates the ability of advanced micro-mechanical testing methods, based on FIB machined micro-cantilevers, to measure the mechanical properties of ion implanted layers without the influence of underlying unimplanted material. The first section describes a study of iron-12 wt% chromium alloy implanted with iron ions. It is shown that by careful cantilever design and finite element modelling that changes in yield stress after implantation can be measured even with the influence of a strong size effect. The second section describes a study of tungsten implanted with both tungsten ions and tungsten and helium ions using spherical and sharp nanoindentation, and micro-cantilevers. The spherical indentation allows yield properties and work hardening behaviour of the implanted layers to be measured. However the brittle nature of the implanted tungsten is only revealed when using micro-cantilevers. This demonstrates that when applying micro-mechanical methods to ion implanted layers care is needed to understand the nature of size effects, careful modelling of experimental procedure is required and multiple experimental techniques are needed to allow the maximum amount of mechanical behaviour information to be collected.

MTR BASEMENT. WORKERS (DON ALVORD AND CYRIL VAN ORDEN OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR BASEMENT. WORKERS (DON ALVORD AND CYRIL VAN ORDEN OF PHILLIPS PETROLEUM CO.) POSE FOR GAMMA IRRADIATION EXPERIMENT IN MTR CANAL. CANS OF FOOD WILL BE LOWERED TO CANAL BOTTOM, WHERE SPENT MTR FUEL ELEMENTS EMIT GAMMA RADIATION. INL NEGATIVE NO. 11746. Unknown Photographer, 8/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Evaluation of microbial loads, physical characteristics, chemical constituents and biological properties of radiation processed Fagonia arabica

NASA Astrophysics Data System (ADS)

Khattak, Khanzadi Fatima

2012-06-01

Whole plant of Fagonia arabica with 3 different particle sizes (30, 50 and 70 mesh) were exposed to gamma radiation doses of 1-10 kGy from a Cobalt 60 source. A series of tests was performed in order to check the feasibility of irradiation processing of the plant. The applied radiation doses did not affect (P<0.05) pH and antimicrobial activities of the plant. The total weight of the dry extracts in methanol as well as water was found increased with irradiation. The irradiated samples showed significant increase in phenolic content and free radical scavenging activity using DPPH. Shortly after irradiation (on the day of radiation treatment) high amounts of free radicals were detected in the irradiated plant samples and the chemiluminescence measurements were generally found to be dose dependent. Maximum luminescence intensity was observed in case of samples with mesh size of 30 for all the radiation doses applied. After a period of one month the chemiluminescence signals of the irradiated samples approximated those of the controls. The study suggests that gamma irradiation treatment is effective for quality improvement and enhances certain beneficial biological properties of the treated materials.

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

2012-08-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

Irradiation effect on mechanical properties in structural materials of fast breeder reactor plant

NASA Astrophysics Data System (ADS)

Nagae, Yuji; Takaya, Shigeru; Wakai, Eiichi; Aoto, Kazumi

2011-07-01

The effects of displacement per atom (dpa) level, helium content, and the ratio of helium content to dpa level on the tensile and creep properties have been investigated in the assumed irradiation damage range of FBR structural materials. The assumed irradiation damage range is up to about 1 dpa and about 30 appm for helium content. Austenitic stainless steel and high-chromium martensitic steel are considered as FBR structural materials. As a result, it is shown that the dpa level is a promising index for evaluating neutron irradiation damage.

Ultra-accelerated natural sunlight exposure testing facilities

DOEpatents

Lewandowski, Allan A.; Jorgensen, Gary J.

2003-08-12

A multi-faceted concentrator apparatus for providing ultra-accelerated natural sunlight exposure testing for sample materials under controlled weathering conditions comprising: facets that receive incident natural sunlight, transmits VIS/NIR and reflects UV/VIS to deliver a uniform flux of UV/VIS onto a sample exposure plane located near a center of a facet array in chamber means that provide concurrent levels of temperature and/or relative humidity at high levels of up to 100.times. of natural sunlight that allow sample materials to be subjected to accelerated irradiance exposure factors for a significant period of time of about 3 to 10 days to provide a corresponding time of about at least a years worth representative weathering of sample materials.

Studying Radiation Damage in Structural Materials by Using Ion Accelerators

NASA Astrophysics Data System (ADS)

Hosemann, Peter

2011-02-01

Radiation damage in structural materials is of major concern and a limiting factor for a wide range of engineering and scientific applications, including nuclear power production, medical applications, or components for scientific radiation sources. The usefulness of these applications is largely limited by the damage a material can sustain in the extreme environments of radiation, temperature, stress, and fatigue, over long periods of time. Although a wide range of materials has been extensively studied in nuclear reactors and neutron spallation sources since the beginning of the nuclear age, ion beam irradiations using particle accelerators are a more cost-effective alternative to study radiation damage in materials in a rather short period of time, allowing researchers to gain fundamental insights into the damage processes and to estimate the property changes due to irradiation. However, the comparison of results gained from ion beam irradiation, large-scale neutron irradiation, and a variety of experimental setups is not straightforward, and several effects have to be taken into account. It is the intention of this article to introduce the reader to the basic phenomena taking place and to point out the differences between classic reactor irradiations and ion irradiations. It will also provide an assessment of how accelerator-based ion beam irradiation is used today to gain insight into the damage in structural materials for large-scale engineering applications.

Three-Dimensional FIB/EBSD Characterization of Irradiated HfAl3-Al Composite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hua, Zilong; Guillen, Donna Post; Harris, William

2016-09-01

A thermal neutron absorbing material, comprised of 28.4 vol% HfAl3 in an Al matrix, was developed to serve as a conductively cooled thermal neutron filter to enable fast flux materials and fuels testing in a pressurized water reactor. In order to observe the microstructural change of the HfAl3-Al composite due to neutron irradiation, an EBSD-FIB characterization approach is developed and presented in this paper. Using the focused ion beam (FIB), the sample was fabricated to 25µm × 25µm × 20 µm and mounted on the grid. A series of operations were carried out repetitively on the sample top surface tomore » prepare it for scanning electron microscopy (SEM). First, a ~100-nm layer was removed by high voltage FIB milling. Then, several cleaning passes were performed on the newly exposed surface using low voltage FIB milling to improve the SEM image quality. Last, the surface was scanned by Electron Backscattering Diffraction (EBSD) to obtain the two-dimensional image. After 50 to 100 two-dimensional images were collected, the images were stacked to reconstruct a three-dimensional model using DREAM.3D software. Two such reconstructed three-dimensional models were obtained from samples of the original and post-irradiation HfAl3-Al composite respectively, from which the most significant microstructural change caused by neutron irradiation apparently is the size reduction of both HfAl3 and Al grains. The possible reason is the thermal expansion and related thermal strain from the thermal neutron absorption. This technique can be applied to three-dimensional microstructure characterization of irradiated materials.« less

Optical properties and light irradiance of monolithic zirconia at variable thicknesses.

PubMed

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 (p<0.05). Surface gloss 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=PRT

Tensile properties of vanadium-base alloys irradiated in the Fusion-1 low-temperature experiment in the BOR-60 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsai, H.; Gazda, J.; Nowicki, L.J.

The irradiation has been completed and the test specimens have been retrieved from the lithium-bonded capsule at the Research Institute of Atomic Reactors (RIAR) in Russia. During this reporting period, the Argonne National Laboratory (ANL) tensile specimens were received from RIAR and initial testing and examination of these specimens at ANL has been completed. The results, corroborating previous findings showed a significant loss of work hardening capability in the materials. There appears to be no significant difference in behavior among the various heats of vanadium-base alloys in the V-(4-5)Cr-(4-5)Ti composition range. The variations in the preirradiation annealing conditions also producedmore » no notable differences.« less

Status and improvement of CLAM for nuclear application

NASA Astrophysics Data System (ADS)

Huang, Qunying

2017-08-01

A program for China low activation martensitic steel (CLAM) development has been underway since 2001 to satisfy the material requirements of the test blanket module (TBM) for ITER, China fusion engineering test reactor and China fusion demonstration reactor. It has been undertaken by the Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences under wide domestic and international collaborations. Extensive work and efforts are being devoted to the R&D of CLAM, such as mechanical property evaluation before and after neutron irradiation, fabrication of scaled TBM by welding and additive manufacturing, improvement of its irradiation resistance as well as high temperature properties by precipitate strengthening to achieve its final successful application in fusion systems. The status and improvement of CLAM are introduced in this paper.

Investigation of radiosterilization feasibility of sulfamethoxazole by ESR spectroscopy

NASA Astrophysics Data System (ADS)

Çolak, Şeyda

2017-12-01

In the present study, the spectroscopic features of the radiolytic intermediates that were produced in gamma-irradiated (5, 10, 25 and 50 kGy) sulfamethoxazole (SMX) have been investigated by electron spin resonance (ESR) spectroscopy and the radiation sterilization feasibility of SMX by ionizing radiation was examined. Gamma-irradiated SMX exhibited a complex ESR spectrum consisting of 13 resonance lines where spectral parameters for the central resonance line were found to be g = 2.0062 and ΔHpp = 0.6 mT. The radiation yield of SMX was calculated to be relatively low (G = 0.1) by ESR spectroscopy and no meaningful difference was observed in the comparison of unirradiated and 50 kGy gamma irradiated SMX by the Fourier transform infrared (FT-IR) technique, confirming that SMX is a radioresistive material. Although SMX could not be accepted to be a good dosimetric material, the identification of irradiated SMX from the unirradiated sample was possible even for the low absorbed radiation doses and for a relatively long time (three months) after the irradiation process. Decay activation energy of the radical species, which is mostly responsible for the central intense resonance line, is calculated to be 45.15 kJ/mol by using the signal intensity decay data derived from annealing studies. Four radical species with different spectroscopic properties were accepted to be responsible for the ESR spectra of gamma-irradiated SMX, by simulation calculations. It is concluded that SMX and SMX-containing drugs can be sterilized by gamma radiation and ESR spectroscopy is an appropriate technique for the characterization of these induced radical intermediates during the gamma irradiation process of SMX. Toxicology tests should also be done for its safe usage.

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

Irradiation combined with SU5416: Microvascular changes and growth delay in a human xenograft glioblastoma tumor line

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schuuring, Janneke; Department of Neurology, Groene Hart Hospital, Gouda; Bussink, Johan

Purpose: The combination of irradiation and the antiangiogenic compound SU5416 was tested and compared with irradiation alone in a human glioblastoma tumor line xenografted in nude mice. The aim of this study was to monitor microenvironmental changes and growth delay. Methods and materials: A human glioblastoma xenograft tumor line was implanted in nude mice. Irradiations consisted of 10 Gy or 20 Gy with and without SU5416. Several microenvironmental parameters (tumor cell hypoxia, tumor blood perfusion, vascular volume, and microvascular density) were analyzed after imunohistochemical staining. Tumor growth delay was monitored for up to 200 days after treatment. Results: SU5416, whenmore » combined with irradiation, has an additive effect over treatment with irradiation alone. Analysis of the tumor microenvironment showed a decreased vascular density during treatment with SU5416. In tumors regrowing after reaching only a partial remission, vascular characteristics normalized shortly after cessation of SU5416. However, in tumors regrowing after reaching a complete remission, permanent microenvironmental changes and an increase of tumor necrosis with a subsequent slower tumor regrowth was found. Conclusions: Permanent vascular changes were seen after combined treatment resulting in complete remission. Antiangiogenic treatment with SU5416 when combined with irradiation has an additive effect over treatment with irradiation or antiangiogenic treatment alone.« less

Comparison of the properties of collagen-chitosan scaffolds after γ-ray irradiation and carbodiimide cross-linking.

PubMed

Chen, Zihao; Du, Tianming; Tang, Xiangyu; Liu, Changjun; Li, Ruixin; Xu, Cheng; Tian, Feng; Du, Zhenjie; Wu, Jimin

2016-07-01

The property of collagen-chitosan porous scaffold varies according to cross-linking density and scaffold composition. This study was designed to compare the properties of collagen-chitosan porous scaffolds cross-linked with γ-irradiation and carbodiimide (CAR) for the first time. Eleven sets of collagen-chitosan scaffolds containing different concentrations of chitosan at a 5% increasing gradient were fabricated. Fourier transform infrared spectroscopy was performed to confirm the success of cross-linking in the scaffolds. The scaffold morphology was evaluated under scanning electron microscope (SEM). SEM revealed that chitosan was an indispensable material for the fabrication of γ-ray irradiation scaffold. The microstructure of γ-ray irradiation scaffold was less stable than those of alternative scaffolds. Based upon swelling ratio, porosity factor, and collagenase degradation, γ-ray irradiation scaffold was less stable than CAR and 25% proportion of chitosan scaffolds. Mechanical property determines the orientation in γ-irradiation and CAR scaffold. In vitro degradation test indicated that γ-irradiation and CAR cross-linking can elevate the scaffold biocompatibility. Compared with γ-ray irradiation, CAR cross-linked scaffold containing 25% chitosan can more significantly enhance the bio-stability and biocompatibility of collagen-chitosan scaffolds. CAR cross-linked scaffold may be the best choice for future tissue engineering.

Investigation of high-energy ion-irradiated MA957 using synchrotron radiation under in-situ tension

DOE PAGES

Mo, Kun; Yun, Di; Miao, Yinbin; ...

2016-01-02

In this paper, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region similar to 7.5 μm in depth; the peak damage (similar to 40 dpa) was estimated to be at similar to 7 μm from the surface. Following the irradiation, in-situ high-energy X-ray diffraction measurements were performed at the Advanced Photon Source in order to study the dynamic deformation behavior of the specimens after ion irradiation damage. In-situ X-ray measurements taken during tensile testing ofmore » the ion-irradiated MA957 revealed a difference in loading behavior between the irradiated and un-irradiated regions of the specimen. At equivalent applied stresses, lower lattice strains were found in the radiation-damaged region than those in the un-irradiated region. This might be associated with a higher level of Type II stresses as a result of radiation hardening. The study has demonstrated the feasibility of combining high-energy ion radiation and high-energy synchrotron X-ray diffraction to study materials' radiation damage in a dynamic manner.« less

The Alliance of Advanced Process Control and Accountability – A Future Safeguards-By-Design Tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lumetta, Gregg J.; Bresee, James C.; Paviet, Patricia D.

For any chemical separation process producing a valuable product, a material balance is an important process control measurement. That is particularly true for the separation of actinides from irradiated nuclear fuel, not only for their intrinsic value but also because an incomplete material balance may indicate diversion for unauthorized use. The DOE Office of Nuclear Energy is currently carrying out at the Pacific Northwest National Laboratory an experimental measurement of how well and with what precision current technologies can implement near real-time actinide material balances. This measurement effort is called the CoDCon project. It involves the separation of a productmore » with a 70/30 uranium/plutonium mass ratio. Initial tests will use dissolved fuel simulants prepared with pure uranium and plutonium nitrates at the same input ratios as irradiated fuel. Subsequent testing with actual irradiated fuel would be done to verify the results obtained with simulants. The experiments will use advanced on-line instrumentation supported by dynamic process models. Since accountability uncertainties could mask diversions, the aim of the project is not only to measure present-day capabilities but also, through sensitivity analyses, to identify those measurements with the greatest potential for overall material-balance improvements. The latter results will help identify priorities for future fuel cycle R&D programs. Advanced separations process control and material accountability technologies thus have a common goal: to provide the best tools available for safeguards-by-design [defined by the International Atomic Energy Agency (IAEA) as the integration of the design of a new nuclear facility through planning, construction, operation and decommissioning]. Since the potential domestic use of CoDCon results may be later than their possible foreign applications, arrangements may be feasible for possible bilateral or multinational cooperation in the CoDCon project.« less

Test study of boron nitride as a new detector material for dosimetry in high-energy photon beams.

PubMed

Poppinga, D; Halbur, J; Lemmer, S; Delfs, B; Harder, D; Looe, H K; Poppe, B

2017-09-05

The aim of this test study is to check whether boron nitride (BN) might be applied as a detector material in high-energy photon-beam dosimetry. Boron nitride exists in various crystalline forms. Hexagonal boron nitride (h-BN) possesses high mobility of the electrons and holes as well as a high volume resistivity, so that ionizing radiation in the clinical range of the dose rate can be expected to produce a measurable electrical current at low background current. Due to the low atomic numbers of its constituents, its density (2.0 g cm -3 ) similar to silicon and its commercial availability, h-BN appears as possibly suitable for the dosimetry of ionizing radiation. Five h-BN plates were contacted to triaxial cables, and the detector current was measured in a solid-state ionization chamber circuit at an applied voltage of 50 V. Basic dosimetric properties such as formation by pre-irradiation, sensitivity, reproducibility, linearity and temporal resolution were measured with 6 MV photon irradiation. Depth dose curves at quadratic field sizes of 10 cm and 40 cm were measured and compared to ionization chamber measurements. After a pre-irradiation with 6 Gy, the devices show a stable current signal at a given dose rate. The current-voltage characteristic up to 400 V shows an increase in the collection efficiency with the voltage. The time-resolved detector current behavior during beam interrupts is comparable to diamond material, and the background current is negligible. The measured percentage depth dose curves at 10 cm  ×  10 cm field size agreed with the results of ionization chamber measurements within  ±2%. This is a first study of boron nitride as a detector material for high-energy photon radiation. By current measurements on solid ionization chambers made from boron nitride chips we could demonstrate that boron nitride is in principle suitable as a detector material for high-energy photon-beam dosimetry.

Test study of boron nitride as a new detector material for dosimetry in high-energy photon beams

NASA Astrophysics Data System (ADS)

Poppinga, D.; Halbur, J.; Lemmer, S.; Delfs, B.; Harder, D.; Looe, H. K.; Poppe, B.

2017-09-01

The aim of this test study is to check whether boron nitride (BN) might be applied as a detector material in high-energy photon-beam dosimetry. Boron nitride exists in various crystalline forms. Hexagonal boron nitride (h-BN) possesses high mobility of the electrons and holes as well as a high volume resistivity, so that ionizing radiation in the clinical range of the dose rate can be expected to produce a measurable electrical current at low background current. Due to the low atomic numbers of its constituents, its density (2.0 g cm-3) similar to silicon and its commercial availability, h-BN appears as possibly suitable for the dosimetry of ionizing radiation. Five h-BN plates were contacted to triaxial cables, and the detector current was measured in a solid-state ionization chamber circuit at an applied voltage of 50 V. Basic dosimetric properties such as formation by pre-irradiation, sensitivity, reproducibility, linearity and temporal resolution were measured with 6 MV photon irradiation. Depth dose curves at quadratic field sizes of 10 cm and 40 cm were measured and compared to ionization chamber measurements. After a pre-irradiation with 6 Gy, the devices show a stable current signal at a given dose rate. The current-voltage characteristic up to 400 V shows an increase in the collection efficiency with the voltage. The time-resolved detector current behavior during beam interrupts is comparable to diamond material, and the background current is negligible. The measured percentage depth dose curves at 10 cm  ×  10 cm field size agreed with the results of ionization chamber measurements within  ±2%. This is a first study of boron nitride as a detector material for high-energy photon radiation. By current measurements on solid ionization chambers made from boron nitride chips we could demonstrate that boron nitride is in principle suitable as a detector material for high-energy photon-beam dosimetry.

HEDL FACILITIES CATALOG 400 AREA

DOE Office of Scientific and Technical Information (OSTI.GOV)

MAYANCSIK BA

1987-03-01

The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

Summary report for the FY-2015 SACSESS Collaboration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterman, Dean Richard; Mincher, Bruce Jay

2015-09-01

During FY-2015, a collaborative research program was established by the Department of Energy-Nuclear Energy (DOE-NE) Material Recovery and Waste Form Development program and the European Union (EU) Safety of Actinide Separation Processes (SACSESS) program. One component of this collaboration was the evaluation of the radiolytic stability of a Selective ActiNide Extraction (SANEX) separation which utilized a TODGA-based organic solvent and an aqueous phase containing the hydrophilic complexing reagent, SO3-Ph-BTP. To best simulate process conditions, this experiment was irradiated in the radiolysis/hydrolysis test loop located at the Idaho National Laboratory. The effect of irradiation on a SACSESS program iSANEX formulation containingmore » a TODGA-based organic phase and a BTP-based aqueous phase was investigated using irradiations at INL in static and test loop modes. When irradiated in contact with only the acidic aqueous phase, the TODGA organic solution maintained excellent extraction performance of americium, cerium and europium to a maximum absorbed dose of nearly 0.9 MGy. When the aqueous phase was changed to that containing the aqueous soluble BTP, the irradiated aqueous phase showed a dramatic color change, but this does not appear to have adverse effects on solvent extraction performance. Only minor increases in distribution ratios for both the lanthanides and actinide were measured, and the separation factors were essentially unchanged to a maximum absorbed dose of 174 kGy. The determination of the americium, cerium, and europium distribution ratios for the remaining SACSESS test loop samples will be completed in the near future. The analysis of stable metals concentration in the the irradiated aqueous and organic phases will be completed shortly.« less

Mechanical properties and molecular structure analysis of subsurface dentin after Er:YAG laser irradiation.

PubMed

He, Zhengdi; Chen, Lingling; Hu, Xuejuan; Shimada, Yasushi; Otsuki, Masayuki; Tagami, Junji; Ruan, Shuangchen

2017-10-01

The purpose of this study was to evaluate the chemical and mechanical modifications in subsurface dentin layer after Er: YAG (Erbium-Yttrium Aluminium Garnet) laser irradiation, as the guidance of new dental restorative materials specific for laser irradiated dentin. Dentin disks obtained from extracted human molars were prepared and exposed to a single pulse Er:YAG laser irradiation at 80mJ/pulse. After laser irradiation the mechanical and chemical characteristics of intertubular dentin in subsurface layer were studied using nanoindentation tester and micro-Raman spectromy (μ-RS). The dentin 5-50µm depth beneath the lased surface was determined as testing area. Two-way analysis of variance (ANOVA) were used to compare the mechanical values between lased and untreated subsurface dentin (P = 0.05). A laser affected subsurface dentin layer after Er:YAG laser treatment is present. The laser irradiation is considered to decrease the mechanical properties in the superficial subsurface layer (<15µm deep). There was no significant difference in nanohardness and Young's modulus between lased subsurface dentin and untreated dentin (p > 0.05) under the depth of 15µm. However, the dentin at 5µm and 10µm depth beneath the lased surface exhibited significantly lower (~ 47.8% and ~ 33.6% respectively) hardness (p < 0.05). Er:YAG laser irradiation affected both mineral and organic components in subsurface dentin layer, a higher degree of crystallinity and reduced organic compounds occurred in the lased subsurface dentin. Under the tested laser parameters, Er:YAG laser irradiation causes lower mechanical values and reduction of organic components in subsurface dentin, which has deleterious effects on resin adhesion to this area. Copyright © 2017 The Authors. Published by Elsevier Ltd.. All rights reserved.

A comparison of the wear of cross-linked polyethylene against itself with the wear of ultra-high molecular weight polyethylene against itself.

PubMed

Joyce, T J; Unsworth, A

1996-01-01

Wear tests were carried out on reciprocating pin-on-plate machines which had pins loaded at 10 N and 40 N. The materials tested were irradiated cross-linked polyethylene sliding against itself, irradiated ultra-high molecular weight polyethylene sliding against itself and non-irradiated ultra-high molecular weight polyethylene sliding against itself. After 153.5 km of sliding, the non-irradiated ultra-high molecular weight polyethylene plates and pins showed mean wear factors under 10 N loads, or a nominal contact stress of 0.51 MPa, of 84.0 x 10(-6) mm3/N m for the plates and 81.3 x 10(-6) mm3/N m for the pins. Under 40 N loads, or a nominal contact stress of 2.04 MPa, the non-irradiated ultra-high molecular weight polyethylene pins sheared at 22.3 km. At the last measurement point prior to this failure, 19.1 km, wear factors of 158 x 10(-6) mm3/N m for the plates and 85.0 x 10(-6) mm3/N m for the pins had been measured. After 152.8 km. the irradiated ultra-high molecular weight polyethylene plates and pins showed mean wear factors under 10 N loads of 59.8 x 10(-6) mm3/N m for the plates and 31.1 x 10(-6) mm3/N m for the pins. In contrast, after 150.2 km, a mean wear factor of 0.72 x 10(-6) mm3/N m was found for the irradiated cross-linked polyethylene plates compared with 0.053 x 10(-6) mm3/N m for the irradiated cross-linked polyethylene pins.

Effect of Er:YAG laser irradiation on bonding property of zirconia ceramics to resin cement.

PubMed

Lin, Yihua; Song, Xiaomeng; Chen, Yaming; Zhu, Qingping; Zhang, Wei

2013-12-01

This study aimed to investigate whether or not an erbium: yttrium-aluminum-garnet (Er:YAG) laser could improve the bonding property of zirconia ceramics to resin cement. Surface treatments can improve the bonding properties of dental ceramics. However, little is known about the effect of Er:YAG laser irradiated on zirconia ceramics. Specimens of zirconia ceramic pieces were made, and randomly divided into 11 groups according to surface treatments, including one control group (no treatment), one air abrasion group, and nine Er:YAG laser groups. The laser groups were subdivided by applying different energy intensities (100, 200, or 300 mJ) and irradiation times (5, 10, or 15 sec). After surface treatments, ceramic pieces had their surface morphology observed, and their surface roughness was measured. All specimens were bonded to resin cement. Shear bond strength was measured after the bonded specimens were stored in water for 24 h, and additionally aged by thermocycling. Statistical analyses were performed using one way analysis of variance (ANOVA) and Tukey's test for shear bond strength, and Dunnett's t test for surface roughness, with α=0.05. Er:YAG laser irradiation changed the morphological characteristics of zirconia ceramics. Higher energy intensities (200, 300 mJ) could roughen the ceramics, but also caused surface cracks. There were no significant differences in the bond strength between the control group and the laser groups treated with different energy intensities or irradiation times. Air abrasion with alumina particles induced highest surface roughness and shear bond strength. Er:YAG laser irradiation cannot improve the bonding property of zirconia ceramics to resin cement. Enhancing irradiation intensities and extending irradiation time have no benefit on the bond of the ceramics, and might cause material defect.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rapp, Juergen; Aaron, A. M.; Bell, Gary L.

2015-10-20

Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panelmore » reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady-state heat fluxes of 5–20 MW/m 2 and ion fluxes up to 10 24 m -2s -1. Since PFCs will have to withstand neutron irradiation displacement damage up to 50 dpa, the target station design must accommodate radioactive specimens (materials to be irradiated in HFIR or at SNS) to enable investigations of the impact of neutron damage on materials. Therefore, the system will have to be able to install and extract irradiated specimens using equipment and methods to avoid sample modification, control contamination, and minimize worker dose. Included in the design considerations will be an assessment of all the steps between neutron irradiation and post-exposure materials examination/characterization, as well as an evaluation of the facility hazard categorization. In particular, the factors associated with the acquisition of radioactive specimens and their preparation, transportation, experimental configuration at the plasma-specimen interface, post-plasma-exposure sample handling, and specimen preparation will be evaluated. Neutronics calculations to determine the dose rates of the samples were carried out for a large number of potential plasma-facing materials.« less

Proton irradiation on materials

NASA Technical Reports Server (NTRS)

Chang, C. Ken

1993-01-01

A computer code is developed by utilizing a radiation transport code developed at NASA Langley Research Center to study the proton radiation effects on materials which have potential application in NASA's future space missions. The code covers the proton energy from 0.01 Mev to 100 Gev and is sufficient for energetic protons encountered in both low earth and geosynchronous orbits. With some modification, the code can be extended for particles heavier than proton as the radiation source. The code is capable of calculating the range, stopping power, exit energy, energy deposition coefficients, dose, and cumulative dose along the path of the proton in a target material. The target material can be any combination of the elements with atomic number ranging from 1 to 92, or any compound with known chemical composition. The generated cross section for a material is stored and is reused in future to save computer time. This information can be utilized to calculate the proton dose a material would receive in an orbit when the radiation environment is known. It can also be used to determine, in the laboratory, the parameters such as beam current of proton and irradiation time to attain the desired dosage for accelerated ground testing of any material. It is hoped that the present work be extended to include polymeric and composite materials which are prime candidates for use as coating, electronic components, and structure building. It is also desirable to determine, for ground testing these materials, the laboratory parameters in order to simulate the dose they would receive in space environments. A sample print-out for water subject to 1.5 Mev proton is included as a reference.

Decomposition Behavior of Curcumin during Solar Irradiation when Contact with Inorganic Particles

NASA Astrophysics Data System (ADS)

Nandiyanto, A. B. D.; Wiryani, A. S.; Rusli, A.; Purnamasari, A.; Abdullah, A. G.; Riza, L. S.

2017-03-01

Curcumin is one of materials which have been widely used in medicine, Asian cuisine, and traditional cosmetic. Therefore, understanding the stability of curcumin has been widely studied. The purpose of this study was to investigate the stability of curcumin solution against solar irradiation when making contact with inorganic material. As a model for the inorganic material, titanium dioxide (TiO2) was used. In the experimental method, the curcumin solution was irradiated using a solar irradiation. To confirm the stability of curcumin when contact with inorganic material, we added TiO2 micro particles with different concentrations. The results showed that the concentration of curcumin decreased during solar irradiation. The less concentration of curcumin affected the more decomposition rate obtained. The decomposition rate was increased greatly when TiO2 was added, in which the more TiO2 concentration added allowed the faster decomposition rate. Based on the result, we conclude that the curcumin is relatively stable as long as using higher concentration of curcumin and is no inorganic material existed. Then, the decomposition can be minimized by avoiding contact with inorganic material.

Dose rate effects in radiation degradation of polymer-based cable materials

NASA Astrophysics Data System (ADS)

Plaček, V.; Bartoníček, B.; Hnát, V.; Otáhal, B.

2003-08-01

Cable ageing under the nuclear power plant (NPP) conditions must be effectively managed to ensure that the required plant safety and reliability are maintained throughout the plant service life. Ionizing radiation is one of the main stressors causing age-related degradation of polymer-based cable materials in air. For a given absorbed dose, radiation-induced damage to a polymer in air environment usually depends on the dose rate of the exposure. In this work, the effect of dose rate on the degradation rate has been studied. Three types of NPP cables (with jacket/insulation combinations PVC/PVC, PVC/PE, XPE/XPE) were irradiated at room temperature using 60Co gamma ray source at average dose rates of 7, 30 and 100 Gy/h with the doses up to 590 kGy. The irradiated samples have been tested for their mechanical properties, thermo-oxidative stability (using differential scanning calorimetry, DSC), and density. In the case of PVC and PE samples, the tested properties have shown evident dose rate effects, while the XPE material has shown no noticeable ones. The values of elongation at break and the thermo-oxidative stability decrease with the advanced degradation, density tends to increase with the absorbed dose. For XPE samples this effect can be partially explained by the increase of crystallinity. It was tested by the DSC determination of the crystalline phase amount.

Graphene's Viability for Fusion Applications

NASA Astrophysics Data System (ADS)

Navarro, Marcos; Hall, Karla; Rojas, Richard; Santarius, John; Kulcinski, Gerald

2015-11-01

Graphene is a source of interest for multiple applications due to its unusual electronic and physical properties. As a coating material, it has reduced oxidation of the main substrate, though no effort has been reported of testing it under fusion conditions. A number of experimental studies have established that defect-free graphene is an excellent barrier material for gases. We explore its viability to maintain a significant pressure difference under ion irradiation. Deuterium is used as a projectile on graphene coated silicon over a range of 10-50 keV energies and various fluences. The vacancy yield (amount of damage) and natural resonance for graphene are found at around 1350 cm-1 and 1550 cm-1, respectively. Damage of each sample is quantified via Raman spectroscopy (RS) using the ratio of the intensities at these wavenumbers. Graphene is also tested here as a coating for some fusion components. Though tungsten is a very promising divertor and first wall candidate, after intense irradiation, it is prone to developing fuzz or grass structures, leading to a diminished lifetime. Graphene grown on tungsten is tested under reactor conditions with 30 keV He ions at several fluences, and the sputtering of both materials is studied via RS and Scanning Electron Microscopy. This work was supported by the Graduate Engineering Research Scholars and the TEAM-Science program at the University of Wisconsin-Madison.

Materials characterization study of conductive flexible second surface mirrors

NASA Technical Reports Server (NTRS)

Levadou, F.; Bosma, S. J.; Paillous, A.

1981-01-01

The status of prequalification and qualification work on conductive flexible second surface mirrors is described. The basic material is FEP Teflon witn either aluminium or silver vacuum deposited reflectors. The top layer has been made conductive by deposition of layer of a indium oxide. The results of a prequalification program comprised of decontamination, humidity, thermal cycling, thermal shock and vibration tests are presented. Thermo-optical and electrical properties. The results of a prequalification program comprised of decontamination, humidity, thermal cycling, thermal shock and vibration tests are presented. Thermo-optical and electrical properties, the electrostatic behavior of the materials under simulated substorm environment and electrical conductivity at low temperatures are characterized. The effects of simulated ultra violet and particles irradiation on electrical and thermo-optical properties of the materials are also presented.

Mass analysis of neutral particles and ions released during electrical breakdowns on spacecraft surfaces

NASA Technical Reports Server (NTRS)

Kendall, B. R. F.

1985-01-01

Charged-particle fluxes from breakdown events were studied. Methods to measure mass spectra and total emitted flux of neutral particles were developed. The design and construction of the specialized mass spectrometer was completed. Electrical breakdowns were initiated by a movable blunt contact touching the insulating surface. The contact discharge apparatus was used for final development of two different high-speed recording systems and for measurements of the composition of the materials given off by the discharge. It was shown that intense instantaneous fluxes of neutral particles were released from the sites of electrical breakdown events. A laser micropulse mass analyzer showed that visible discoloration at breakdown sites were correllated with the presence of iron on the polymer side of the film, presumably caused by punch-through to the Inconel backing. Kapton samples irradiated by an oxygen ion beam were tested. The irradiated samples were free of surface hydrocarbon contamination but otherwise behaved in the same way as the Kapton samples tested earlier. Only the two samples exposed to oxygen ion bombardment were relatively clean. This indicates an additional variable that should be considered when testing spacecraft materials in the laboratory.

Static and Dynamic Performance of Newly Developed ITER Relevant Insulation Systems after Neutron Irradiation

NASA Astrophysics Data System (ADS)

Prokopec, R.; Humer, K.; Fillunger, H.; Maix, R. K.; Weber, H. W.

2006-03-01

Fiber reinforced plastics will be used as insulation systems for the superconducting magnet coils of ITER. The fast neutron and gamma radiation environment present at the magnet location will lead to serious material degradation, particularly of the insulation. For this reason, advanced radiation-hard resin systems are of special interest. In this study various R-glass fiber / Kapton reinforced DGEBA epoxy and cyanate ester composites fabricated by the vacuum pressure impregnation method were investigated. All systems were irradiated at ambient temperature (340 K) in the TRIGA reactor (Vienna) to a fast neutron fluence of 1×1022 m-2 (E>0.1 MeV). Short-beam shear and static tensile tests were carried out at 77 K prior to and after irradiation. In addition, tension-tension fatigue measurements were used in order to assess the mechanical performance of the insulation systems under the pulsed operation conditions of ITER. For the cyanate ester based system the influence of interleaving Kapton layers on the static and dynamic material behavior was investigated as well.

Solar tests of aperture plate materials for solar thermal dish collectors

NASA Technical Reports Server (NTRS)

Jaffe, L. D.

1983-01-01

In parabolic dish solar collectors, walk-off of the spot of concentrated sunlight is a hazard if a malfunction causes the concentrator to stop following the Sun. Therefore, a test program was carried out to evaluate the behavior of various ceramics, metals, and polymers under solar irradiation of about 7000 kW/sq m. (peak) for 15 minutes. The only materials that did not slump or shatter were two grades of medium-grain extruded graphite. High purity, slip-cast silica might be satisfactory at somewhat lower flux. Oxidation of the graphite appeared acceptable during tests simulating walk-off, acquisition (2000 cycles on/off Sun), and spillage (continuous on-Sun operation).

Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corwin, W.R.; Broadhead, B.L.; Suzuki, M.

1997-02-01

There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

Nuclear and Physical Properties of Dielectrics under Neutron Irradiation in Fast (BN-600) and Fusion (DEMO-S) Reactors

NASA Astrophysics Data System (ADS)

Blokhin, D. A.; Chernov, V. M.; Blokhin, A. I.

2017-12-01

Nuclear and physical properties (activation and transmutation of elements) of BN and Al2O3 dielectric materials subjected to neutron irradiation for up to 5 years in Russian fast (BN-600) and fusion (DEMO-S) reactors were calculated using the ACDAM-2.0 software complex for different post-irradiation cooling times (up to 10 years). Analytical relations were derived for the calculated quantities. The results may be used in the analysis of properties of irradiated dielectric materials and may help establish the rules for safe handling of these materials.

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2010 CFR

2010-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2014 CFR

2014-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2012 CFR

2012-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2011 CFR

2011-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2013 CFR

2013-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

DOE PAGES

Jin, Ke; Guo, Wei; Lu, Chenyang; ...

2016-12-01

Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. In order to systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 10 13 to 3 × 10 16 cm -2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluencemore » regime, which is consistent to the results from molecular dynamic simulations. We did not observe any irradiation-induced compositional segregation in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.« less

Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin, Ke; Guo, Wei; Lu, Chenyang

Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. In order to systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 10 13 to 3 × 10 16 cm -2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluencemore » regime, which is consistent to the results from molecular dynamic simulations. We did not observe any irradiation-induced compositional segregation in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.« less

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

DOE PAGES

Fukuda, Makoto; Kiran Kumar, N. A. P.; Koyanagi, Takaaki; ...

2016-07-02

We performed a neutron irradiation to single crystal pure tungsten in the mixed spectrum High Flux Isotope Reactor (HFIR). In order to investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ~90–~800 °C and fast neutron fluences were 0.02–9.00 × 10 25 n/m 2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. Moreover, the hardness and microstructuremore » changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10 25 n/m 2 (E > 0.1 MeV). Finally, irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10 25 n/m 2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.« less

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fukuda, Makoto; Kiran Kumar, N. A. P.; Koyanagi, Takaaki

We performed a neutron irradiation to single crystal pure tungsten in the mixed spectrum High Flux Isotope Reactor (HFIR). In order to investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ~90–~800 °C and fast neutron fluences were 0.02–9.00 × 10 25 n/m 2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. Moreover, the hardness and microstructuremore » changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10 25 n/m 2 (E > 0.1 MeV). Finally, irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10 25 n/m 2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.« less

Environmentally assisted cracking in light water reactors : semiannual report, July 2000 - December 2000.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Chung, H. M.; Gruber, E. E.

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on themore » mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to {approx}2.0 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range on crack growth rates in air.« less

Advanced Numerical Model for Irradiated Concrete

DOE Office of Scientific and Technical Information (OSTI.GOV)

Giorla, Alain B.

In this report, we establish a numerical model for concrete exposed to irradiation to address these three critical points. The model accounts for creep in the cement paste and its coupling with damage, temperature and relative humidity. The shift in failure mode with the loading rate is also properly represented. The numerical model for creep has been validated and calibrated against different experiments in the literature [Wittmann, 1970, Le Roy, 1995]. Results from a simplified model are shown to showcase the ability of numerical homogenization to simulate irradiation effects in concrete. In future works, the complete model will be appliedmore » to the analysis of the irradiation experiments of Elleuch et al. [1972] and Kelly et al. [1969]. This requires a careful examination of the experimental environmental conditions as in both cases certain critical information are missing, including the relative humidity history. A sensitivity analysis will be conducted to provide lower and upper bounds of the concrete expansion under irradiation, and check if the scatter in the simulated results matches the one found in experiments. The numerical and experimental results will be compared in terms of expansion and loss of mechanical stiffness and strength. Both effects should be captured accordingly by the model to validate it. Once the model has been validated on these two experiments, it can be applied to simulate concrete from nuclear power plants. To do so, the materials used in these concrete must be as well characterized as possible. The main parameters required are the mechanical properties of each constituent in the concrete (aggregates, cement paste), namely the elastic modulus, the creep properties, the tensile and compressive strength, the thermal expansion coefficient, and the drying shrinkage. These can be either measured experimentally, estimated from the initial composition in the case of cement paste, or back-calculated from mechanical tests on concrete. If some are unknown, a sensitivity analysis must be carried out to provide lower and upper bounds of the material behaviour. Finally, the model can be used as a basis to formulate a macroscopic material model for concrete subject to irradiation, which later can be used in structural analyses to estimate the structural impact of irradiation on nuclear power plants.« less

Ceramic breeder research and development: progress and focus

NASA Astrophysics Data System (ADS)

van der Laan, J. G.; Kawamura, H.; Roux, N.; Yamaki, D.

2000-12-01

The world-wide efforts on ceramic breeder materials in the last two years concerned Li2O, Li4SiO4, Li2TiO3 and Li2ZrO3, with a clear emphasis on the development of Li2TiO3. Pebble-manufacturing processes have been developed up to a 10 kg scale. Characterisation of materials has advanced. A jump-wise progress is observed in the characterisation of pebble-beds, in particular of their thermo-mechanical behaviour. Thermal property data are still limited. A number of breeder materials have been or are being irradiated in material test reactors like HFR and JMTR. The EXOTIC-8 series of in-pile experiments is a major source of tritium release data. This paper discusses the technical advancements and proposes a focus for further research and development (R&D) : pebble-bed mechanical and thermal behaviour and its interactions with the blanket structure as a function of temperature, burn-up, irradiation dose and time; tritium release and retention properties; determination of the key factors limiting blanket life.

Bond strength of epoxy resin-based root canal sealer to human root dentin irradiated with Er,Cr:YSGG laser.

PubMed

Franceschini, Keila de Almeida; Silva-Sousa, Yara Teresinha Corrêa; Lopes, Fabiane Carneiro; Pereira, Rodrigo Dantas; Palma-Dibb, Regina Guenka; de Sousa-Neto, Manoel Damião

2016-12-01

The aim was to evaluate the influence of Er,Cr:YSGG laser irradiation associated with different final irrigation protocols on the bond strength of epoxy resin-based root canal sealer to root dentin, on the dentin/filling material interface and in the temperature variation during irradiation. Ninety-six maxillary canines were prepared with K3 rotary system up to #45/0.02 instrument, irrigating with distilled water between files. The specimens were randomly assigned to three groups-final irrigation (distilled water, 1% NaOCl, and 17% EDTAC) and four subgroups (n = 8)-laser parameters (non-irradiated, 2 W/20 Hz, 3 W/20 Hz, and 4 W/20 Hz). During irradiation, the temperatures were measured on the outer root dentin wall in the three thirds, and root apex. Canals were filled with lateral condensation of AHPlus sealer and gutta-percha cones. Two slices from each third were submitted to a push-out test in Instron machine and the failure mode was analyzed. One slice from each third was analyzed by confocal laser microscopy to evaluate the percentage of the perimeter of the root canal cross-section with sealer tags and depth of tags. Data were analyzed by ANOVA, Kruskal-Wallis, and Tukey's tests (P < 0.05). Er,Cr:YSGG laser irradiation increased sealer bond strength to dentin, regardless of the final irrigation. The highest values were obtained for 3 W (4.02 ± 1.32) and 4 W (4.18 ± 0.98) powers and different from the non-irradiated group (2.64 ± 0.58) (P < 0.05). The 2 W irradiation produced similar results to 3 W and 4 W when associated with 17% EDTA. Final irrigation with 17% EDTAC provided higher bond strength (4.01 ± 1.02) compared with distilled water (3.11 ± 1.09) and 1% NaOCl (3.47 ± 1.18) (P < 0.05). The cervical third (4.01 ± 1.21) presented significantly higher bond strength than the apical third (3.04 ± 0.89). There was a greater percentage of adhesive and mixed failure. In the groups irradiated with 3 W [21.1 (14.1-27.7)] and 4 W [17.8 (11.9-23.7)], a greater depth of filling material tags was observed compared with the non-irradiated group [12.9 (9.0-20.0)]. The greatest percentage of canal perimeter with sealer tags was observed in the irradiated groups, with no difference among them (P > 0.05). The temperature rise was proportional to the increase of laser power. Er,Cr:YSGG laser irradiation increased the bond strength of an epoxy resin-based sealer to root dentin, with greater formation of sealer tags for all tested powers, especially when combined with 17% EDTAC final irrigation; temperature rise during irradiation remained below the critical threshold biologically accepted. Lasers Surg. Med. 48:985-994, 2016. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

Effect of annealing high-dose heavy-ion irradiated high-temperature superconductor wires

NASA Astrophysics Data System (ADS)

Strickland, N. M.; Wimbush, S. C.; Kluth, P.; Mota-Santiago, P.; Ridgway, M. C.; Kennedy, J. V.; Long, N. J.

2017-10-01

Heavy-ion irradiation of high-temperature superconducting thin films has long been known to generate damage tracks of amorphized material that are of close-to-ideal dimension to effectively contribute to pinning of magnetic flux lines and thereby enhance the in-field critical current. At the same time, though, the presence of these tracks reduces the superconducting volume fraction available to transport current while the irradiation process itself generates oxygen depletion and disorder in the remaining superconducting material. We have irradiated commercially available superconducting coated conductors consisting of a thick film of (Y,Dy)Ba2Cu3O7 deposited on a buffered metal tape substrate in a continuous reel-to-reel process. Irradiation was by 185 MeV 197Au ions. A high fluence of 3 × 1011 ions/cm2 was chosen to emphasize the detrimental effects. The critical current was reduced following this irradiation, but annealing at relatively low temperatures of 200 °C and 400 °C substantially restore the critical current of the irradiated material. At high fields and high temperatures there is a net benefit of critical current compared to the untreated material.

Setup for in situ deep level transient spectroscopy of semiconductors during swift heavy ion irradiation.

PubMed

Kumar, Sandeep; Kumar, Sugam; Katharria, Y S; Safvan, C P; Kanjilal, D

2008-05-01

A computerized system for in situ deep level characterization during irradiation in semiconductors has been set up and tested in the beam line for materials science studies of the 15 MV Pelletron accelerator at the Inter-University Accelerator Centre, New Delhi. This is a new facility for in situ irradiation-induced deep level studies, available in the beam line of an accelerator laboratory. It is based on the well-known deep level transient spectroscopy (DLTS) technique. High versatility for data manipulation is achieved through multifunction data acquisition card and LABVIEW. In situ DLTS studies of deep levels produced by impact of 100 MeV Si ions on Aun-Si(100) Schottky barrier diode are presented to illustrate performance of the automated DLTS facility in the beam line.

PIE preparation of the MEGAPIE target

NASA Astrophysics Data System (ADS)

Wohlmuther, Michael; Wagner, Werner

2012-12-01

The MEGAPIE target, after successfully operating for 4 months at a beam power of 0.77 MW, is now being prepared for post irradiation examination PIE. The lead-bismuth eutectic (LBE) target was irradiated from August until December 2006, and in this period received a beam charge of 2.8 A h of 575 MeV protons. After that, the target was stored in the target storage facility of PSI, waiting for its post irradiation examination. In the meantime several campaigns of tests have been conducted by PSI and ZWILAG, the interim storage facility of Swiss nuclear power plants. In these tests the feasibility of the conditioning of the target and the extraction of sample material for the PIE has been proven. After transport to the hot cell facility at ZWILAG in June 2009, the dismantling of the MEGAPIE target started. It finally was cut into 21 pieces. Ten of these pieces will be shipped to the Hot Laboratory of PSI ('PSI hotlab') to extract samples from the structural materials as well as from the LBE. Currently it is foreseen that the sample extraction will start in the first half of 2011. The remaining parts of the MEGAPIE target were conditioned as radioactive waste. The present paper will mainly focus on the dismantling and first visual inspection of the MEGAPIE target. In addition an outlook on the PIE phase of MEGAPIE is given.

Data Fitting to Study Ablated Hard Dental Tissues by Nanosecond Laser Irradiation.

PubMed

Al-Hadeethi, Y; Al-Jedani, S; Razvi, M A N; Saeed, A; Abdel-Daiem, A M; Ansari, M Shahnawaze; Babkair, Saeed S; Salah, Numan A; Al-Mujtaba, A

2016-01-01

Laser ablation of dental hard tissues is one of the most important laser applications in dentistry. Many works have reported the interaction of laser radiations with tooth material to optimize laser parameters such as wavelength, energy density, etc. This work has focused on determining the relationship between energy density and ablation thresholds using pulsed, 5 nanosecond, neodymium-doped yttrium aluminum garnet; Nd:Y3Al5O12 (Nd:YAG) laser at 1064 nanometer. For enamel and dentin tissues, the ablations have been performed using laser-induced breakdown spectroscopy (LIBS) technique. The ablation thresholds and relationship between energy densities and peak areas of calcium lines, which appeared in LIBS, were determined using data fitting. Furthermore, the morphological changes were studied using Scanning Electron Microscope (SEM). Moreover, the chemical stability of the tooth material after ablation has been studied using Energy-Dispersive X-Ray Spectroscopy (EDX). The differences between carbon atomic % of non-irradiated and irradiated samples were tested using statistical t-test. Results revealed that the best fitting between energy densities and peak areas of calcium lines were exponential and linear for enamel and dentin, respectively. In addition, the ablation threshold of Nd:YAG lasers in enamel was higher than that of dentin. The morphology of the surrounded ablated region of enamel showed thermal damages. For enamel, the EDX quantitative analysis showed that the atomic % of carbon increased significantly when laser energy density increased.

Development and Testing of Dispersion-Strengthened Tungsten Alloys via Spark Plasma Sinterin

NASA Astrophysics Data System (ADS)

Lang, Eric; Madden, Nathan; Smith, Charles; Krogstad, Jessica; Allain, Jean Paul

2017-10-01

Tungsten (W) is a common plasma-facing component (PFC) material in the divertor region of tokamak fusion devices due to its high melting point and high sputter threshold. However, W is intrinsically brittle and is further embrittled under neutron irradiation, and the low recrystallization temperature pose complications in fusion environments. More ductile W alloys, such as dispersion-strengthened tungsten are being developed. In this work, W samples are processed via spark plasma sintering (SPS) with TiC, ZrC, and TaC dispersoids alloyed from 0.5 to 10 weight %. SPS is a powder compaction technique that provides high pressure and heating rates via electrical current, allowing for a lower final temperature and hold time for compaction. Initial testing of material properties, smicrostructure, and composition of specimens will be presented. Deuterium and helium irradiations have been performed in IGNIS, a multi-functional, in-situ irradiation and characterization facility at the University of Illinois. High-flux, low-energy exposures at the Magnum-PSI facility at DIFFER exposed samples to a D fluence of 1×1026 cm-2 and He fluence of 1x1025-1x1026 cm-2 at temperatures of 300-1000 C. In-situ chemistry changes via XPS and ex-situ morphology changes via SEM will be studied. Work supported by US DOE Contract DE-SC0014267.

Mechanical stability of Ti6Al4V implant material after femtosecond laser irradiation

NASA Astrophysics Data System (ADS)

Symietz, Christian; Lehmann, Erhard; Gildenhaar, Renate; Hackbarth, Andreas; Berger, Georg; Krüger, Jörg

2012-07-01

The surface of a titanium alloy (Ti6Al4V) implant material was covered with a bioactive calcium alkali phosphate ceramic with the aim to accelerate the healing and to form a stronger bond to living bone tissue. To fix the ceramic powder we used a femtosecond laser, which causes a thin surface melting of the metal. It is a requirement to prove that the laser irradiation would not reduce the lifetime of implants. Here we present the results of mechanical stability tests, determined by the rotating bending fatigue strength of sample rods. After describing the sample surfaces and their modifications caused by the laser treatment we give evidence for an unchanged mechanical stability. This applies not only to the ceramic fixation but also to a comparatively strong laser ablation.

Detecting self-ion irradiation-induced void swelling in pure copper using transient grating spectroscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dennett, C. A.; So, K. P.; Kushima, A.

Irradiation-induced void swelling remains a major challenge to nuclear reactor operation. Swelling may take years to initiate and often results in rapid material property degradation once started. Alloy development for advanced nuclear systems will require rapid characterization of the swelling breakaway dose in new alloys, yet this capability does not yet exist. In this paper, we demonstrate that transient grating spectroscopy (TGS) can detect void swelling in single crystal copper via changes in surface acoustic wave (SAW) velocity. Scanning transmission electron microscopy (STEM) links the TGS-observed changes with void swelling-induced microstructural evolution. Finally, these results are considered in the contextmore » of previous work to suggest that in situ TGS will be able to rapidly determine when new bulk materials begin void swelling, shortening alloy development and testing times.« less

Detecting self-ion irradiation-induced void swelling in pure copper using transient grating spectroscopy

DOE PAGES

Dennett, C. A.; So, K. P.; Kushima, A.; ...

2017-12-20

Irradiation-induced void swelling remains a major challenge to nuclear reactor operation. Swelling may take years to initiate and often results in rapid material property degradation once started. Alloy development for advanced nuclear systems will require rapid characterization of the swelling breakaway dose in new alloys, yet this capability does not yet exist. In this paper, we demonstrate that transient grating spectroscopy (TGS) can detect void swelling in single crystal copper via changes in surface acoustic wave (SAW) velocity. Scanning transmission electron microscopy (STEM) links the TGS-observed changes with void swelling-induced microstructural evolution. Finally, these results are considered in the contextmore » of previous work to suggest that in situ TGS will be able to rapidly determine when new bulk materials begin void swelling, shortening alloy development and testing times.« less

Study of properties of the plastic scintillator EJ-260 under irradiation with 150 MeV protons and 1.2MeV gamma-rays

NASA Astrophysics Data System (ADS)

Dormenev, V.; Brinkmann, K.-T.; Korjik, M.; Novotny, R. W.

2017-11-01

One of the most critical aspects for the application of a scintillation material in high energy physics is the degradation of properties of the material in an environment of highly ionizing particles in particular due to hadrons. There are presently several detector concepts in consideration being based on organic scintillator material for fast timing of charged particles or sampling calorimeters. We have tested different samples of the organic plastic scintillator EJ-260 produced by the company Eljen Technology (Sweetwater, TX, USA). The ongoing activity has characterized the relevant parameters such as light output, kinetics and temperature dependence. The study has focused on the change of performance after irradiation with 150 MeV protons up to an integral fluence of 5·1013 protons/cm2 as well as with a strong 60Co γ-source accumulating an integral dose of 100 Gy. The paper will report on the obtained results.

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacobsen, George; Back, Christina

2015-10-30

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called themore » endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.« less

Chromatographic extraction with di(2-ethylhexyl)orthophosphoric acid for production and purification of promethium-147

DOEpatents

Boll, Rose A [Knoxville, TN; Mirzadeh, Saed [Knoxville, TN

2008-10-14

A method of producing and purifying promethium-147 including the steps of: irradiating a target material including neodymium-146 with neutrons to produce promethium-147 within the irradiated target material; dissolving the irradiated target material to form an acidic solution; loading the acidic solution onto a chromatographic separation apparatus containing HDEHP; and eluting the apparatus to chromatographically separate the promethium-147 from the neodymium-146.

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

NASA Astrophysics Data System (ADS)

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

2018-04-01

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

X-ray sterilization of insects and microorganisms for cultural heritage applications

NASA Astrophysics Data System (ADS)

Borgognoni, F.; Vadrucci, M.; Bazzano, G.; Ferrari, P.; Massa, S.; Moretti, R.; Calvitti, M.; Ronsivalle, C.; Moriani, A.; Picardi, L.

2017-09-01

The APAM (Development of Particle Accelerators and Medical Applications) Laboratory of the ENEA Frascati Research Center is engaged in the preservation of cultural heritage as part of the COBRA (Sviluppo e diffusione di metodi, tecnologie e strumenti avanzati per la COnservazione dei Beni culturali, basati sull'applicazione di Radiazioni e di tecnologie Abilitanti) project addressed to the transfer of innovative technologies and methodologies from research to small and medium enterprises involved in the restorative measures. This work aims to demonstrate the effectiveness of ionizing radiation on the disinfection of biodegraded art objects. The conventional methods for the disinfestation of works of art, using chemicals toxic to humans and environment, might cause some damage to the treated material even on micrometric scale (i. e. either cellulose degradation). Ionizing radiations interact with the infesting biological material causing an irreversible DNA degradation. For this reason, they are certainly suitable for removal treatments of both macro organisms and bacterial colonies. A 4.8 MeV electron linear accelerator, normally dedicated to the characterization of dose detectors and radiographies, has been employed to produce Bremsstrahlung X-rays through a lead converter. The spectral fluence of the radiation source has been calculated using the Monte Carlo MCNPX code. The dosimetric characterization of the radiation field has been made using radiochromic films sensitive in the dose range of our interest (from 50 to 500 Gy) calibrated with a Markus ionization chamber. The irradiation of the artifact prototypes are made within a lead shielded room at a variable distance from the X-rays source. Samples subjected to irradiation consist of a soil bacterium, Agrobacterium rhizogenes, and an insect, Stegobium paniceum, that are found as wall paintings invasive coloniser and as a pest of books, wood works and paintings, respectively. Tests of irradiation have been performed on pest organisms as well as on woods mock-ups to evaluate potential damage to the material during the sterilization. The growing capacity of the treated bacterial cells re-cultured at the end of the treatment was evaluated on the bacterial sample and resulted to strongly inhibit cell growth during post-irradiation incubation, so that after incubation periods at 28 °C, no significant cell growth was observed. The induced levels of insect mortality and sterility vs absorbed dose and operative conditions have been also evaluated, demonstrating the induction of full sterility since the lower dose and 40% mortality by two days after the higher dose treatment. The experiments proved the ability to efficaciously treat objects of cultural heritage with X-rays in order to prevent the increase of the biodeterioration without damaging the materials: in fact, mechanical tests on both irradiated and not irradiated woods have demonstrated the absence of any induced degradation after the radiation exposition.

Preliminary evaluation of glass resin materials for solar cell cover use. [on spacecraft

NASA Technical Reports Server (NTRS)

Marsik, S. J.; Swartz, C. K.; Baraona, C. R.

1978-01-01

Silicon solar cells and silicon wafers coated with a heat-curable resin consisting of alternating Si-O atoms were subjected to three tests to evaluate the potential utility of this coating in space environments. These included UV irradiation in vacuum at an intensity of 10 air mass zero UV energy-equivalent solar constants for 728 hours followed by a long thermal cycle; 15 thermal shock cycles between 100 C and minus 196 C; and high temperature and humidity (65 C at 90% relative humidity). The UV tests resulted in a 8 to 24% loss in short-circuit current and darkening of the covers. Modification of the resin to provide a better match between the coefficients of expansion of the resin and silicon improved resistance to thermal shock, but also increased the darkening effect under UV irradiation. Silicon wafers coated with the resin were not adversely affected by the temperature/humidity test.

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang

2006-10-01

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progressmore » toward element testing will be reviewed.« less

Effect of gamma irradiation on the wear behavior of human tooth dentin.

PubMed

Qing, Ping; Huang, Shengbin; Gao, ShanShan; Qian, LinMao; Yu, HaiYang

2016-12-01

The objective of this study was to evaluate the effect of gamma irradiation on the wear behavior of human tooth dentin in terms of possible alterations in crystallinity, grain size, and composition. Human premolars (n = 19) were collected to obtain the perpendicular or parallel to the direction of the dentin tubule specimens. Each specimen was subjected to 60 Gy of gamma irradiation, in daily increments of 2 Gy. The nanoscratch tests were conducted. The scratch traces were observed via scanning electron microscope (SEM) and surface profilometer. X-ray diffraction (XRD) and Fourier transform infrared spectroscopy (FTIR) were used to investigate the alteration of crystallography and chemical composition of dentin after irradiation. The change of surface microhardness (SMH) was also evaluated. The nanoscratch results showed that the friction coefficient of dentin after irradiation became higher, and the depths and widths of scratch were greater than that of dentin before irradiation. Additionally, irradiation decreased the crystallinity of dentin and induced the formation of bigger crystals. The carbonate/mineral ratio was increased. Furthermore, a significant reduction in microhardness after irradiation was observed. The main damage mechanisms consisted of the formation of delamination and crack in both the specimens cut perpendicular and parallel to tubule dentin after irradiation. Irradiation affected directly the wear behavior of tooth dentin, accompanied by the alterations in crystallography, chemical composition, and surface microhardness of dentin. This would help extend understanding the influence of irradiation on dentin and provide suggestions for selecting more suitable materials for irradiated tooth.

MASSIVE LEAKAGE IRRADIATOR

DOEpatents

Wigner, E.P.; Szilard, L.; Christy, R.F.; Friedman, F.L.

1961-05-30

An irradiator designed to utilize the neutrons that leak out of a reactor around its periphery is described. It avoids wasting neutron energy and reduces interference with the core flux to a minimum. This is done by surrounding all or most of the core with removable segments of the material to be irradiated within a matrix of reflecting material.

Ultra-Accelerated Natural Sunlight Exposure Testing Facilities

DOEpatents

Lewandowski, Allan A.; Jorgensen, Gary J.

2004-11-23

A multi-faceted concentrator apparatus for providing ultra-accelerated natural sunlight exposure testing for sample materials under controlled weathering conditions comprising: facets that receive incident natural sunlight, transmits VIS/NIR and reflects UV/VIS onto a secondary reflector that delivers a uniform flux of UV/VIS onto a sample exposure plane located near a center of a facet array in a chamber that provide concurrent levels of temperature and/or relative humidity at high levels of up to 100.times. of natural sunlight that allow sample materials to be subjected to accelerated irradiance exposure factors for a significant period of time of about 3 to 10 days to provide a corresponding time of about at least a years worth representative weathering of sample materials.

Fission neutron source in Rome

NASA Astrophysics Data System (ADS)

Coppola, Mario; Di Majo, V.; Ingrao, G.; Rebessi, S.; Testa, A.

1997-02-01

A fission neutron source is operating in Rome at the ENEA Casaccia Research Center since 1971, consisting of a low power fast reactor named RSV-Tapiro. it is employed for a variety of experiments, including dosimetry, material testing, radiation protection and biology. In particular, application to experimental radiobiology includes studies of the biological action of neutrons in the whole-body irradiated animal, or in specialized systems in vivo or in vitro. For his purpose a vertical irradiation facility was originally constructed. Recently, a new horizontal irradiation facility has been designed to allow the exposure of larger samples or larger sample batches at one time. Dosimetry at the sample irradiation positions is routinely carried out by the conventional method of using two ion chambers. This physical dosimetry has recently been compared with the results of biological dosimetry based on the detection of chromosomal aberrations in peripheral blood human lymphocytes irradiated in vitro. A characterization of the radiation quality in the two configurations has been carried out by tissue equivalent proportional counter microdosimetry measurements. Information about the main characteristics of the reactor and the two irradiation facilities is provided and relevant results of the various measurements are summarized. Radiobiological results obtained using this neutron source are also briefly outlined.

NASA Astrophysics Data System (ADS)

Nishimura, A.; Takeuchi, T.; Nishijima, S.; Ochiai, K.; Nishijima, G.; Watanabe, K.; Shikama, T.

2010-04-01

To investigate the effect of neutron irradiation on superconducting properties, a collaboration network was established among superconducting material engineering and neutronics fields. Within the framework, irradiation test of Nb3Sn and Nb3Al wires by 14 MeV fusion neutron was planned and carried out at Fusion Neutronics Source in Japan Atomic Energy Agency. After the irradiation, critical current and critical magnetic field were measured with 28 T hybrid magnet at Institute for Metals Research in Tohoku University. The irradiation to 3.52×1020 n/m2 showed a slight increase of the critical current of the Nb3Sn wire, and the irradiation to 1.78×1021 n/m2 made the critical current appreciably larger. Regarding the critical magnetic field, no clear change was observed. In the case of Nb3Al wire, a sample irradiated to 1.78×1021 n/m2 showed no increase of the critical current below 200 A which was the limit of the power supply. As for the critical magnetic field, there was no clear improvement similar to the Nb3Sn wire. The increase of the critical current would be caused by knock-on effect of the fast neutron.

Void growth and coalescence in irradiated copper under deformation

NASA Astrophysics Data System (ADS)

Barrioz, P. O.; Hure, J.; Tanguy, B.

2018-04-01

A decrease of fracture toughness of irradiated materials is usually observed, as reported for austenitic stainless steels in Light Water Reactors (LWRs) or copper alloys for fusion applications. For a wide range of applications (e.g. structural steels irradiated at low homologous temperature), void growth and coalescence fracture mechanism has been shown to be still predominant. As a consequence, a comprehensive study of the effects of irradiation-induced hardening mechanisms on void growth and coalescence in irradiated materials is required. The effects of irradiation on ductile fracture mechanisms - void growth to coalescence - are assessed in this study based on model experiments. Pure copper thin tensile samples have been irradiated with protons up to 0.01 dpa. Micron-scale holes drilled through the thickness of these samples subjected to uniaxial loading conditions allow a detailed description of void growth and coalescence. In this study, experimental data show that physical mechanisms of micron-scale void growth and coalescence are similar between the unirradiated and irradiated copper. However, an acceleration of void growth is observed in the later case, resulting in earlier coalescence, which is consistent with the decrease of fracture toughness reported in irradiated materials. These results are qualitatively reproduced with numerical simulations accounting for irradiation macroscopic hardening and decrease of strain-hardening capability.

Low-energy electron-beam treatment as alternative for on-site sterilization of highly functionalized medical products - A feasibility study

NASA Astrophysics Data System (ADS)

Gotzmann, G.; Portillo, J.; Wronski, S.; Kohl, Y.; Gorjup, E.; Schuck, H.; Rögner, F. H.; Müller, M.; Chaberny, I. F.; Schönfelder, J.; Wetzel, C.

2018-09-01

Over the last decades, the medical device industry has grown significantly. Complex and highly functionalized medical devices and implants are being developed to improve patient treatment and to enhance their health-related quality of life. However, medical devices from this new generation often cannot be sterilized by standard methods such as autoclaving or sterilizing gases, as they are temperature sensitive, containing electronic components like sensors and microchips, or consist of polymers. Gamma irradiation for sterilization of such products is also problematic due to long processing times under highly reactive conditions resulting in material degradation or loss of functionality. Low-energy electron-beam treatment could enable irradiation sterilization of medical surfaces within seconds. This method is very fast in comparison to gamma irradiation because of its high dose rate and therefore degradation processes of polymers can be reduced or even prevented. Additionally, electron penetration depth can be precisely controlled to prevent damage of sensitive components like electronics and semiconductors. The presented study focuses on two key aspects: 1.) Can new and highly functionalized medical products in future be sterilized using low-energy electron-beam irradiation; and 2.) Is the low-energy electron-beam technology suitable to be set up on-site to speed up sterilization processing or make it available "just-in-time". To address these questions, different test specimens were chosen with complex geometry or electronic functional parts to gather information about the limitations and chances for this new approach. The test specimens were inoculated with clinical relevant test organisms (Pseudomonas aeruginosa) as well as with approved radiation resistant organisms (Deinococcus radiodurans and Bacillus pumilus) to prove the suitability of low-energy electron-beam treatment for the above-mentioned medical products. The calculation of the D10 value for B. pumilus revealed equal efficacy when compared to standard high-energy irradiation sterilization. All of the above-mentioned germs were successfully inactivated by low-energy electron-beam treatment when test specimens were inoculated with a germ load > 10^6 CFU and treated with doses ≥ 10 kGy (for B. pumilus and P. aeruginosa) and > 300 kGy (for D. radiodurans) respectively. As an example, for specialized electronic components to be sterilized, an impedance sensor for cell culture applications was sterilized and unimpaired functionality was demonstrated even after five repeated sterilization cycles to a total dose of 50 kGy. To address the second aspect of on-site suitability of this technology, the product handling for low-energy electron-beam treatment had to be adapted to minimize the size of the electron-beam facility. Therefore, a mini electron-beam source was used and a specialized sample holder and 3D-handling regime were developed to allow reproducible surface treatment for complex product geometries. Inactivation of B. pumilus inoculated medical screws (> 10^6 CFU) was successful using the developed handling procedure. In addition, a packaging material (PET12/PE50) for medical products was investigated for its suitability for low-energy irradiation sterilization. Biocompatibility assessment revealed the material to be eligible for this application as even overdoses did not impair the biocompatibility of the material. With these results, the principal suitability of low-energy electron-beam treatment for sterilization of medical products containing electronics like sensors is demonstrated. The low-energy technology and the specialized 3D-handling regime allow the on-site setup of the technology in hospitals, medical practices or any other point of care.

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

NASA Astrophysics Data System (ADS)

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

2018-04-01

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

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

Effects of neutron irradiation on deformation behavior of nickel-base fastener alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bajaj, R.; Mills, W.J.; Kammenzind, B.F.

1999-07-01

This paper presents the effects of neutron irradiation on the fracture behavior and deformation microstructure of high-strength nickel-base alloy fastener materials, Alloy X-750 and Alloy 625. Alloy X-750 in the HTH condition, and Alloy 625 in the direct aged condition were irradiated to a fluence of 2.4x10{sup 20} n/cm{sup 2} at 264 C in the Advanced Test Reactor. Deformation structures at low strains were examined. It was previously shown that Alloy X-750 undergoes hardening, a significant degradation in ductility and an increase in intergranular fracture. In contrast, Alloy 625 had shown softening with a concomitant increase in ductility and transgranularmore » failure after irradiation. The deformation microstructures of the two alloys were also different. Alloy X-750 deformed by a planar slip mechanism with fine microcracks forming at the intersections of slip bands with grain boundaries. Alloy 625 showed much more homogeneous deformation with fine, closely spaced slip bands and an absence of microcracks. The mechanism(s) of irradiation assisted stress corrosion cracking (IASCC) are discussed.« less

Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

DOE PAGES

Field, Kevin G.; Hu, Xunxiang; Littrell, Kenneth C.; ...

2015-07-14

The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition.more » Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.« less

Capabilities Development for Transient Testing of Advanced Nuclear Fuels at TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woolstenhulme, N. E.; Baker, C. C.; Bess, J. D.

2016-09-01

The TREAT facility is a unique capability at the Idaho National Laboratory currently being prepared for resumption of nuclear transient testing. Accordingly, designs for several transient irradiation tests are being pursued to enable development of advanced nuclear fuels and materials. In addition to the reactor itself, the foundation for TREAT’s capabilities also include a suite of irradiation vehicles and supporting infrastructure to provide the desired specimen boundary conditions while supporting a variety of instrumentation needs. The challenge of creating these vehicles, especially since many of the modern data needs were not historically addressed in TREAT experiment vehicles, has necessitated amore » sizeable engineering effort. This effort is currently underway and maturing rapidly. This paper summarizes the status, future plans, and rationale for TREAT experiment vehicle capabilities. Much of the current progress is focused around understanding and demonstrating the behavior of fuel design with enhanced accident tolerance in water-cooled reactors. Additionally, several related efforts are underway to facilitate transient testing in liquid sodium, inert gas, and steam environments. This paper discusses why such a variety of capabilities are needed, outlines plans to accomplish them, and describes the relationship between transient data needs and the irradiation hardware that will support the gathering of this information.« less

A review: applications of the phase field method in predicting microstructure and property evolution of irradiated nuclear materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Yulan; Hu, Shenyang; Sun, Xin

Complex microstructure changes occur in nuclear fuel and structural materials due to the extreme environments of intense irradiation and high temperature. This paper evaluates the role of the phase field (PF) method in predicting the microstructure evolution of irradiated nuclear materials and the impact on their mechanical, thermal, and magnetic properties. The paper starts with an overview of the important physical mechanisms of defect evolution and the significant gaps in simulating microstructure evolution in irradiated nuclear materials. Then, the PF method is introduced as a powerful and predictive tool and its applications to microstructure and property evolution in irradiated nuclearmore » materials are reviewed. The review shows that 1) FP models can correctly describe important phenomena such as spatial dependent generation, migration, and recombination of defects, radiation-induced dissolution, the Soret effect, strong interfacial energy anisotropy, and elastic interaction; 2) The PF method can qualitatively and quantitatively simulate 2-D and 3-D microstructure evolution, including radiation-induced segregation, second phase nucleation, void migration, void and gas bubble superlattice formation, interstitial loop evolution, hydrate formation, and grain growth, and 3) The FP method correctly predicts the relationships between microstructures and properties. The final section is dedicated to a discussion of the strengths and limitations of the PF method, as applied to irradiation effects in nuclear materials.« less

Surface Flashover on Epoxy-Resin Printed Circuit Boards in Vacuum under Electron Irradiation

NASA Astrophysics Data System (ADS)

Fujii, Haruhisa; Hasegawa, Taketoshi; Osuga, Hiroyuki; Matsui, Katsuaki

This paper deals with the surface flashover characteristics of dielectric material in vacuum during electron beam irradiation in order to design adequately the conductive patterns on printed circuit boards used inside a spacecraft. The dielectric material, glass-fiber reinforced epoxy resin, and the electrodes printed on it were irradiated with electrons of the energy of 3-10 keV. DC high voltage was applied between the two electrodes during electron irradiation. The voltage was increased stepwise until the surface flashover occurred on the dielectric material. We obtained the results that the surface flashover voltage increased with the insulation distance between the electrodes but electron irradiation made the flashover voltage lower. The flashover voltage characteristics were obtained as parameters of the electrode distance and the energy of the electron beam.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vogel, Sven C.; Losko, Adrian Simon; Pokharel, Reeju

The goal of the Advanced Non-destructive Fuel Examination (ANDE) work package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels, ultimately also to irradiated fuels. The results of these characterizations provide complete pre- and post-irradiation on length scales ranging from mm to nm, guide destructive examination, and inform modelling efforts. Besides technique development and application to samples to be irradiated, the ANDE work package also examines possible technologies to provide these characterization techniques pool-side, e.g. at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) using laser-driven intense pulsed neutronmore » and gamma sources. Neutron tomography and neutron diffraction characterizations were performed on nine pellets; four UN/ U-Si composite formulations (two enrichment levels), three pure U 3Si 5 reference formulations (two enrichment levels), and two reject pellets with visible flaws (to qualify the technique). The 235U enrichments ranged from 0.2 to 8.8 wt. %. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U 3Si 5 material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. We have also proposed a data format to build a database for characterization results of individual pellets. Neutron data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. This report provides the results for the characterized samples and discussion in the context of ANDE and APIE. We quantified the gamma spectra of several samples in their received state as well as after neutron irradiation to ensure that the neutron irradiation does not add significant activation that would complicate shipment and handling. We demonstrated synchrotron-based 3D X-ray microscopy on the composite fuel materials, providing unparalleled level of detail on the 3D microstructure. Furthermore, we initiated development of shielding containers allowing the characterizations presented herein while allowing handling of irradiated samples.« less

Minimizing material damage using low temperature irradiation

NASA Astrophysics Data System (ADS)

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

2012-08-01

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

Post-irradiation hardness development, chemical softening, and thermal stability of bulk-fill and conventional resin-composites.

PubMed

Alshali, Ruwaida Z; Salim, Nesreen A; Satterthwaite, Julian D; Silikas, Nick

2015-02-01

To measure bottom/top hardness ratio of bulk-fill and conventional resin-composite materials, and to assess hardness changes after dry and ethanol storage. Filler content and kinetics of thermal decomposition were also tested using thermogravimetric analysis (TGA). Six bulk-fill (SureFil SDR, Venus bulk fill, X-tra base, Filtek bulk fill flowable, Sonic fill, and Tetric EvoCeram bulk-fill) and eight conventional resin-composite materials (Grandioso flow, Venus Diamond flow, X-flow, Filtek Supreme Ultra Flowable, Grandioso, Venus Diamond, TPH Spectrum, and Filtek Z250) were tested (n=5). Initial and 24h (post-cure dry storage) top and bottom microhardness values were measured. Microhardness was re-measured after the samples were stored in 75% ethanol/water solution. Thermal decomposition and filler content were assessed by TGA. Results were analysed using one-way ANOVA and paired sample t-test (α=0.05). All materials showed significant increase of microhardness after 24h of dry storage which ranged from 100.1% to 9.1%. Bottom/top microhardness ratio >0.9 was exhibited by all materials. All materials showed significant decrease of microhardness after 24h of storage in 75% ethanol/water which ranged from 14.5% to 74.2%. The extent of post-irradiation hardness development was positively correlated to the extent of ethanol softening (R(2)=0.89, p<0.001). Initial thermal decomposition temperature assessed by TGA was variable and was correlated to ethanol softening. Bulk-fill resin-composites exhibit comparable bottom/top hardness ratio to conventional materials at recommended manufacturer thickness. Hardness was affected to a variable extent by storage with variable inorganic filler content and initial thermal decomposition shown by TGA. The manufacturer recommended depth of cure of bulk-fill resin-composites can be reached based on the microhardness method. Characterization of the primary polymer network of a resin-composite material should be considered when evaluating its stability in the aqueous oral environment. Copyright © 2014 Elsevier Ltd. All rights reserved.

Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

NASA Astrophysics Data System (ADS)

Magnusson, Per; Chen, Jiachao; Hoffelner, Wolfgang

2009-12-01

Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10 × 2 × 0.2 mm3 were investigated in a temperature range of 300 °C to 500 °C under irradiation, and significant creep strains were detected. At temperatures above 500 °C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900 °C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900 °C, the expected 104 hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloys.

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

NASA Astrophysics Data System (ADS)

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.

Correlation of Clinical and Dosimetric Factors With Adverse Pulmonary Outcomes in Children After Lung Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Venkatramani, Rajkumar, E-mail: rvenkatramani@chla.usc.edu; Department of Pediatrics, Keck School of Medicine, University of Southern California, Los Angeles, California; Kamath, Sunil

Purpose: To identify the incidence and the risk factors for pulmonary toxicity in children treated for cancer with contemporary lung irradiation. Methods and Materials: We analyzed clinical features, radiographic findings, pulmonary function tests, and dosimetric parameters of children receiving irradiation to the lung fields over a 10-year period. Results: We identified 109 patients (75 male patients). The median age at irradiation was 13.8 years (range, 0.04-20.9 years). The median follow-up period was 3.4 years. The median prescribed radiation dose was 21 Gy (range, 0.4-64.8 Gy). Pulmonary toxic chemotherapy included bleomycin in 58.7% of patients and cyclophosphamide in 83.5%. The followingmore » pulmonary outcomes were identified and the 5-year cumulative incidence after irradiation was determined: pneumonitis, 6%; chronic cough, 10%; pneumonia, 35%; dyspnea, 11%; supplemental oxygen requirement, 2%; radiographic interstitial lung disease, 40%; and chest wall deformity, 12%. One patient died of progressive respiratory failure. Post-irradiation pulmonary function tests available from 44 patients showed evidence of obstructive lung disease (25%), restrictive disease (11%), hyperinflation (32%), and abnormal diffusion capacity (12%). Thoracic surgery, bleomycin, age, mean lung irradiation dose (MLD), maximum lung dose, prescribed dose, and dosimetric parameters between V{sub 22} (volume of lung exposed to a radiation dose ≥22 Gy) and V{sub 30} (volume of lung exposed to a radiation dose ≥30 Gy) were significant for the development of adverse pulmonary outcomes on univariate analysis. MLD, maximum lung dose, and V{sub dose} (percentage of volume of lung receiving the threshold dose or greater) were highly correlated. On multivariate analysis, MLD was the sole significant predictor of adverse pulmonary outcome (P=.01). Conclusions: Significant pulmonary dysfunction occurs in children receiving lung irradiation by contemporary techniques. MLD rather than prescribed dose should be used to perform risk stratification of patients receiving lung irradiation.« less

Effects of gamma irradiation on deteriorated paper

NASA Astrophysics Data System (ADS)

Bicchieri, Marina; Monti, Michela; Piantanida, Giovanna; Sodo, Armida

2016-08-01

Even though gamma radiation application, also at the minimum dosage required for disinfection, causes depolymerization and degradation of the paper substrate, recently published papers seemed, instead, to suggest that γ-rays application could be envisaged in some conditions for Cultural Heritage original documents and books. In some of the published papers, the possible application of γ-rays was evaluated mainly by using mechanical tests that scarcely reflect the chemical modifications induced in the cellulosic support. In the present article the effect of low dosage γ-irradiation on cellulosic substrates was studied and monitored applying different techniques: colorimetry, spectroscopic measurements, carbonyl content and average viscometric degree of polymerization. Two different papers were investigated, a non-sized, non-filled cotton paper, and a commercial permanent paper. To simulate a real deteriorated document, which could need γ-rays irradiation, some samples were submitted to a hydrolysis treatment. We developed a treatment based on the exposition of paper to hydrochloric acid vapors, avoiding any contact of the samples with water. This method induces a degradation similar to that observed on original documents. The samples were then irradiated with 3 kGy γ-rays at a 5258 Gy/h rate. The aforementioned analyses were performed on the samples just irradiated and after artificial ageing. All tests showed negative effects of gamma irradiation on paper. Non-irradiated paper preserves better its appearance and chemical properties both in the short term and after ageing, while the irradiated samples show appreciable color change and higher oxidation extent. Since the Istituto centrale restauro e conservazione patrimonio archivistico e librario is responsible for the choice of all restoration treatments that could be applied on library and archival materials under the protection of the Italian State (http://www.icpal.beniculturali.it/allegati/DM-7-10-2008-Istituto.pdf), it has been evaluated that the modifications induced by γ-rays irradiation are not acceptable as safe conservation treatment (http://www.icpal.beniculturali.it/allegati/Nota_uso_raggi_gamma.pdf).

Complete Report on the Development of Welding Parameters for Irradiated Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frederick, Greg; Sutton, Benjamin J.; Tatman, Jonathan K.

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

Effect of high fluence neutron irradiation on transport properties of thermoelectrics

NASA Astrophysics Data System (ADS)

Wang, H.; Leonard, K. J.

2017-07-01

Thermoelectric materials were subjected to high fluence neutron irradiation in order to understand the effect of radiation damage on transport properties. This study is relevant to the NASA Radioisotope Thermoelectric Generator (RTG) program in which thermoelectric elements are exposed to radiation over a long period of time in space missions. Selected n-type and p-type bismuth telluride materials were irradiated at the High Flux Isotope Reactor with a neutron fluence of 1.3 × 1018 n/cm2 (E > 0.1 MeV). The increase in the Seebeck coefficient in the n-type material was partially off-set by an increase in electrical resistivity, making the power factor higher at lower temperatures. For the p-type materials, although the Seebeck coefficient was not affected by irradiation, electrical resistivity decreased slightly. The figure of merit, zT, showed a clear drop in the 300-400 K range for the p-type material and an increase for the n-type material. Considering that the p-type and n-type materials are connected in series in a module, the overall irradiation damages at the device level were limited. These results, at neutron fluences exceeding a typical space mission, are significant to ensure that the radiation damage to thermoelectrics does not affect the performance of RTGs.

Compact RF ion source for industrial electrostatic ion accelerator

NASA Astrophysics Data System (ADS)

Kwon, Hyeok-Jung; Park, Sae-Hoon; Kim, Dae-Il; Cho, Yong-Sub

2016-02-01

Korea Multi-purpose Accelerator Complex is developing a single-ended electrostatic ion accelerator to irradiate gaseous ions, such as hydrogen and nitrogen, on materials for industrial applications. ELV type high voltage power supply has been selected. Because of the limited space, electrical power, and robust operation, a 200 MHz RF ion source has been developed. In this paper, the accelerator system, test stand of the ion source, and its test results are described.

Compact RF ion source for industrial electrostatic ion accelerator.

PubMed

Kwon, Hyeok-Jung; Park, Sae-Hoon; Kim, Dae-Il; Cho, Yong-Sub

2016-02-01

Korea Multi-purpose Accelerator Complex is developing a single-ended electrostatic ion accelerator to irradiate gaseous ions, such as hydrogen and nitrogen, on materials for industrial applications. ELV type high voltage power supply has been selected. Because of the limited space, electrical power, and robust operation, a 200 MHz RF ion source has been developed. In this paper, the accelerator system, test stand of the ion source, and its test results are described.

Nonlinear bleaching, absorption, and scattering of 532-nm-irradiated plasmonic nanoparticles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liberman, V.; Sworin, M.; Kingsborough, R. P.

2013-02-07

Single-pulse irradiation of Au and Ag suspensions of nanospheres and nanodisks with 532-nm 4-ns pulses has identified complex optical nonlinearities while minimizing material damage. For all materials tested, we observe competition between saturable absorption (SA) and reverse SA (RSA), with RSA behavior dominating for intensities above {approx}50 MW/cm{sup 2}. Due to reduced laser damage in single-pulse experiments, the observed intrinsic nonlinear absorption coefficients are the highest reported to date for Au nanoparticles. We find size dependence to the nonlinear absorption enhancement for Au nanoparticles, peaking in magnitude for 80-nm nanospheres and falling off at larger sizes. The nonlinear absorption coefficientsmore » for Au and Ag spheres are comparable in magnitude. On the other hand, the nonlinear absorption for Ag disks, when corrected for volume fraction, is several times higher. These trends in nonlinear absorption are correlated to local electric field enhancement through quasi-static mean-field theory. Through variable size aperture measurements, we also separate nonlinear scattering from nonlinear absorption. For all materials tested, we find that nonlinear scattering is highly directional and that its magnitude is comparable to that of nonlinear absorption. These results indicate methods to improve the efficacy of plasmonic nanoparticles as optical limiters in pulsed laser systems.« less

Tungsten as a plasma-facing material in fusion devices: impact of helium high-temperature irradiation on hydrogen retention and damages in the material

NASA Astrophysics Data System (ADS)

Bernard, E.; Sakamoto, R.; Kreter, A.; Barthe, M. F.; Autissier, E.; Desgardin, P.; Yamada, H.; Garcia-Argote, S.; Pieters, G.; Chêne, J.; Rousseau, B.; Grisolia, C.

2017-12-01

Plasma-facing materials for next generation fusion devices, like ITER and DEMO, have to withstand intense fluxes of light elements (notably helium and hydrogen isotopes). For tungsten (W), helium (He) irradiation leads to major changes in the material morphology, rising concerns about properties such as material structure conservation and hydrogen (H) retention. The impact of preceeding He irradiation conditions (temperature, flux and fluence) on H trapping were investigated on a set of W samples exposed to the linear plasma device PSI-2. Positron annihilation spectroscopy (PAS) was carried out to probe the free volume of defects created by the He exposure in the W structure at the atomic scale. In parallel, tritium (T) inventory after exposure was evaluated through T gas loading and desorption at the Saclay Tritium Lab. First, we observed that the material preparation prior to He irradiation was crucial, with a major reduction of the T trapping when W was annealed at 1773 K for 2 h compared to the as-received material. PAS study confirms the presence of He in the bubbles created in the material surface layer, whose dimensions were previously characterized by transmission electron microscopy and grazing-incidence small-angle x-ray scattering, and demonstrates that even below the minimal energy for displacement of He in W, defects are created in almost all He irradiation conditions. The T loading study highlights that increasing the He fluence leads to higher T inventory. Also, for a given fluence, increasing the He flux reduces the T trapping. The very first steps of a parametric study were set to understand the mechanisms at stake in those observed material modifications, confirming the need to pursue the study with a more complete set of surface and irradiation conditions.

Final Report on Developing Microstructure-Property Correlation in Reactor Materials using in situ High-Energy X-rays

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Meimei; Almer, Jonathan D.; Yang, Yong

2016-01-01

This report provides a summary of research activities on understanding microstructure – property correlation in reactor materials using in situ high-energy X-rays. The report is a Level 2 deliverable in FY16 (M2CA-13-IL-AN_-0403-0111), under the Work Package CA-13-IL-AN_- 0403-01, “Microstructure-Property Correlation in Reactor Materials using in situ High Energy Xrays”, as part of the DOE-NE NEET Program. The objective of this project is to demonstrate the application of in situ high energy X-ray measurements of nuclear reactor materials under thermal-mechanical loading, to understand their microstructure-property relationships. The gained knowledge is expected to enable accurate predictions of mechanical performance of these materialsmore » subjected to extreme environments, and to further facilitate development of advanced reactor materials. The report provides detailed description of the in situ X-ray Radiated Materials (iRadMat) apparatus designed to interface with a servo-hydraulic load frame at beamline 1-ID at the Advanced Photon Source. This new capability allows in situ studies of radioactive specimens subject to thermal-mechanical loading using a suite of high-energy X-ray scattering and imaging techniques. We conducted several case studies using the iRadMat to obtain a better understanding of deformation and fracture mechanisms of irradiated materials. In situ X-ray measurements on neutron-irradiated pure metal and model alloy and several representative reactor materials, e.g. pure Fe, Fe-9Cr model alloy, 316 SS, HT-UPS, and duplex cast austenitic stainless steels (CASS) CF-8 were performed under tensile loading at temperatures of 20-400°C in vacuum. A combination of wide-angle X-ray scattering (WAXS), small-angle X-ray scattering (SAXS), and imaging techniques were utilized to interrogate microstructure at different length scales in real time while the specimen was subject to thermal-mechanical loading. In addition, in situ X-ray studies were complemented and benchmarked by ex situ characterization using advanced electron microscopy, atom probe tomography (APT) and micro/nano-indentation. The report presented in situ tensile test results on neutron-irradiated pure Fe, Fe-9Cr model alloy, 316 SS and CASS CF-8. These in situ experiments demonstrate the broad applications of the new capability in understanding several outstanding issues related to irradiated materials.« less

Gamma irradiation induces acetylcholine-evoked, endothelium-independent relaxation and activatesk-channels of isolated pulmonary artery of rats

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eder, Veronique; Gautier, Mathieu; Boissiere, Julien

2004-12-01

Purpose: To test the effects of irradiation (R*) on the pulmonary artery (PA). Methods and materials: Isolated PA rings were submitted to gamma irradiation (cesium, 8 Gy/min{sup -1}) at doses of 20 Gy-140 Gy. Rings were placed in an organ chamber, contracted with serotonin (10{sup -4} M 5-hydroxytryptamine [5-HT]), then exposed to acetylcholine (ACh) in incremental concentrations. Smooth muscle cell (SMC) membrane potential was measured with microelectrodes. Results: A high dose of irradiation (60 Gy) increased 5HT contraction by 20%, whereas lower (20 Gy) doses slightly decreased it compared with control. In the absence of the endothelium, 5-HT precontracted ringsmore » exposed to 20 Gy irradiation developed a dose-dependent relaxation induced by acetylcholine (EI-ACh) with maximal relaxation of 60 {+-} 17% (n = 13). This was totally blocked by L-NAME (10{sup -4} M), partly by 7-nitro indazole; it was abolished by hypoxia and iberiotoxin, decreased by tetra-ethyl-ammonium, and not affected by free radical scavengers. In irradiated rings, hypoxia induced a slight contraction which was never observed in control rings. No differences in SMC membrane potential were observed between irradiated and nonirradiated PA rings. Conclusion: Irradiation mediates endothelium independent relaxation by a mechanism involving the nitric oxide pathway and K-channels.« less

Oxygen ion irradiation effect on corrosion behavior of titanium in nitric acid medium

NASA Astrophysics Data System (ADS)

Ningshen, S.; Kamachi Mudali, U.; Mukherjee, P.; Barat, P.; Raj, Baldev

2011-01-01

The corrosion assessment and surface layer properties after O 5+ ion irradiation of commercially pure titanium (CP-Ti) has been studied in 11.5 N HNO 3. CP-Ti specimen was irradiated at different fluences of 1 × 10 13, 1 × 10 14 and 1 × 10 15 ions/cm 2 below 313 K, using 116 MeV O 5+ ions source. The corrosion resistance and surface layer were evaluated by using potentiodynamic polarization, electrochemical impedance spectroscopy (EIS), scanning electron microscopy (SEM) and glancing-angle X-ray diffraction (GXRD) methods. The potentiodynamic anodic polarization results of CP-Ti revealed that increased in ion fluence (1 × 10 13-1 × 10 15 ions/cm 2) resulted in increased passive current density due to higher anodic dissolution. SEM micrographs and GXRD analysis corroborated these results showing irradiation damage after corrosion test and modified oxide layer by O 5+ ion irradiation was observed. The EIS studies revealed that the stability and passive film resistance varied depending on the fluence of ion irradiation. The GXRD patterns of O 5+ ion irradiated CP-Ti revealed the oxides formed are mostly TiO 2, Ti 2O 3 and TiO. In this paper, the effects of O 5+ ion irradiation on material integrity and corrosion behavior of CP-Ti in nitric acid are described.

Nonlinear Ultrasonic Measurements in Nuclear Reactor Environments

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this thesis, two ultrasonic characterization techniques will be explored. The first, finite amplitude wave propagation has been demonstrated to be sensitive to microstructural material property changes. It is a strong candidate to determine fuel evolution; however, it has not been demonstrated for in-situ reactor applications. In this thesis, finite amplitude wave propagation will be used to measure the microstructural evolution in Al-6061. This is the first demonstration of finite amplitude wave propagation at temperatures in excess of 200 °C and during an irradiation test. Second, a method based on contact nonlinear acoustic theory will be developed to identify compressed cracks. Compressed cracks are typically transparent to ultrasonic wave propagation; however, by measuring harmonic content developed during finite amplitude wave propagation, it is shown that even compressed cracks can be characterized. Lastly, piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65x10 20 n/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 x1020 n/cm2, Zinc Oxide is capable of transduction up to 6.27 x1020 n/cm 2, and Aluminum Nitride is capable of transduction up to 8.65x x10 20 n/cm2.

A high field and cryogenic test facility for neutron irradiated superconducting wire

NASA Astrophysics Data System (ADS)

Nishimura, A.; Miyata, H.; Yoshida, M.; Iio, M.; Suzuki, K.; Nakamoto, T.; Yamazaki, M.; Toyama, T.

2017-12-01

A 15.5 T superconducting magnet and a variable temperature insert (VTI) system were installed at a radiation control area in Oarai center in Tohoku University to investigate the superconducting properties of activated superconducting materials by fast neutron. The superconductivity was measured at cryogenic temperature and high magnetic field. During these tests, some inconvenient problems were observed and the additional investigation was carried out. The variable temperature insert was designed and assembled to perform the superconducting property tests. without the liquid helium. To remove the heat induced by radiation and joule heating, high purity aluminum rod was used in VTI. The thermal contact was checked by FEM analysis and an additional support was added to confirm the decreasing the stress concentration and the good thermal contact. After the work for improvement, it was affirmed that the test system works well and all troubles were resolved. In this report, the improved technical solution is described and the first data set on the irradiation effect on Nb3Sn wire is presented.

Preliminary Evaluation of Commercial Off the Shelf (COTS) Packing Materials for Flight Medication Dispenser (FMD) Technology Development

NASA Technical Reports Server (NTRS)

Du, Brian; Daniels, Vernie; Crady, Camille; Putcha, Lakshmi

2010-01-01

With the advent of longer duration space missions, pharmaceutical use in space has increased. During the first 33 space shuttle missions, crew members took more than 500 individual doses of 31 different medications . Anecdotal reports from crew members described medications as generally "well tolerated" and "effective". However, reported use of increased medication doses and discrepancies in ground vs. flight efficacy may result from reduced potency or altered bioavailability due to changes in chemical and/or physical parameters of pharmaceutical stability. Based on preliminary results from a ground-based irradiation and an inflight study on pharmaceutical stability, three susceptible medications, Amoxicillin/Clavulanate and Sulfamethoxazole/trimethoprim antibiotics tablets and promethazine (PMZ), an antihistamine were selected for testing using two types of Oliver-Tolas bags, TPC-1475(Clear) and TPF-0599B (Foil) for radiation Shielding effectiveness. The material composition of the bags included aluminum coated Mylar sheathing coated with multifunctional nanocomposities based on polyethylene with dispersed boron-rich nanophases. Two bags of each medication were irradiated for different time intervals with 14.6 rad/min to achieve 0.1 Gy, 1 Gy and 10 Gy of cumulative radiation dose. Active pharmaceutical content (API) in each medication was determined and results analyzed. No significant difference in API content was observed between control and irradiated samples for both antibiotic tablets suggesting both types of bags may offer protection against gamma radiation; results with PMZ were inconclusive. These preliminary results suggest that Oliver-Tolas TPL-1475 and TPF-0599B materials may possess characteristics suitable for protection against ionizing radiation and can be considered for designing and further testing of FMD technology.

Disposition of Chicago Pile 5 (CP-5) Converter Tubes in the 10-160B Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pancake, Daniel C.; Rock, Cynthia

This paper will focus on the unique characterization, packaging, and transportation issues associated with the disposition of the two CP-5 Converter Tube assemblies from Argonne National Laboratory. The converter tubes were constructed of combinations of HEU and alloys of zirconium, and were part of the original research facilities attached to the CP-5 reactor during operating evolutions. These assemblies were heavily irradiated during their operational lifetime, and were segregated from the balance of irradiated test specimens when the reactor was deactivated and slated for Decontamination and Demolition (D&D). In addition, the substantial contribution of fissile material to the assemblies’ inventory mademore » the potential disposition pathways extremely challenging. As a result, these items became part of Argonne’s legacy “nuclear footprint”, and were added to the Nuclear Footprint Reduction Project scope for disposition. The Project was responsible for the size reduction and characterization of these items, as well as the ultimate disposition. After negotiating a disposal pathway for these tubes, there were significant transportation issues that required a small team to overcome, in order to successfully ship these items to the Nevada National Security Site (NNSS). The Project team at Argonne, technical support from transportation specialists, licensing support from the 10-160B license owner, the Savanah River National Lab (SRNL) Packaging Certification Team (PCT, and the DOE EM-33 staff contributed to license and safety analysis report amendments that eventually authorized the shipment of the material. The paper will identify the organizations, and the specific actions, required to successfully make three “one of a kind” shipments of irradiated test specimen material. This will include the unique packaging configurations, contents modification for the cask license (via the Amendment process), criticality evaluations, and associated review and approval processes.« less

Evaluation of Computed Tomography of Mock Uranium Fuel Rods at the Advanced Photon Source

DOE PAGES

Hunter, James F.; Brown, Donald William; Okuniewski, Maria

2015-06-01

This study discusses a multi-year effort to evaluate the utility of computed tomography at the Advanced Photon Source (APS) as a tool for non-destructive evaluation of uranium based fuel rods. The majority of the data presented is on mock material made with depleted uranium which mimics the x-ray attenuation characteristics of fuel rods while allowing for simpler handling. A range of data is presented including full thickness (5mm diameter) fuel rodlets, reduced thickness (1.8mm) sintering test samples, and pre/post irradiation samples (< 1mm thick). These data were taken on both a white beam (bending magnet) beamline and a high energy,more » monochromatic beamline. This data shows the utility of a synchrotron type source in the evealuation of manufacturing defects (pre-irradiation) and lays out the case for in situ CT of fuel pellet sintering. Finally, in addition data is shown from small post-irradiation samples and a case is made for post-irradiation CT of larger samples.« less

Radiation resistance of a gamma-ray irradiated nonlinear optic chromophore

NASA Astrophysics Data System (ADS)

Zhang, Cheng; Taylor, Edward W.

2009-11-01

The radiation resistance of organic electro-optic and optoelectronic materials for space applications is receiving increased attention. An earlier investigation reported that guest-host poled polymer EO modulator devices composed of a phenyltetraene bridge-type chromophore in amorphous polycarbonate (CLD/APC) did not exhibit a decrease in EO response (i.e., an increase in modulation-switching voltage- Vπ) following irradiation by low dose [10-160 krad(Si)] 60Co gamma-rays. To provide further evidences to the observed radiation stability, the post-irradiation responses of 60Co gamma-rays on CLD1/APC thin films are examined by various chemical and spectroscopic methods including: a solubility test, thin-layer chromatography, proton nuclear magnetic resonance spectroscopy, UV-vis absorption, and infra-red absorption. The results indicate that CLD1 and APC did not decompose under gamma-ray irradiation at dose levels ranging from 66-274 krad(Si) and from 61-154 krad(Si), respectively which support the previously reported data.

The Role of Grain Size on Neutron Irradiation Response of Nanocrystalline Copper

PubMed Central

Mohamed, Walid; Miller, Brandon; Porter, Douglas; Murty, Korukonda

2016-01-01

The role of grain size on the developed microstructure and mechanical properties of neutron irradiated nanocrystalline copper was investigated by comparing the radiation response of material to the conventional micrograined counterpart. Nanocrystalline (nc) and micrograined (MG) copper samples were subjected to a range of neutron exposure levels from 0.0034 to 2 dpa. At all damage levels, the response of MG-copper was governed by radiation hardening manifested by an increase in strength with accompanying ductility loss. Conversely, the response of nc-copper to neutron irradiation exhibited a dependence on the damage level. At low damage levels, grain growth was the primary response, with radiation hardening and embrittlement becoming the dominant responses with increasing damage levels. Annealing experiments revealed that grain growth in nc-copper is composed of both thermally-activated and irradiation-induced components. Tensile tests revealed minimal change in the source hardening component of the yield stress in MG-copper, while the source hardening component was found to decrease with increasing radiation exposure in nc-copper. PMID:28773270

Modeling defect cluster evolution in irradiated structural materials: Focus on comparing to high-resolution experimental characterization studies

DOE PAGES

Wirth, Brian D.; Hu, Xunxiang; Kohnert, Aaron; ...

2015-03-02

Exposure of metallic structural materials to irradiation environments results in significant microstructural evolution, property changes, and performance degradation, which limits the extended operation of current generation light water reactors and restricts the design of advanced fission and fusion reactors. Further, it is well recognized that these irradiation effects are a classic example of inherently multiscale phenomena and that the mix of radiation-induced features formed and the corresponding property degradation depend on a wide range of material and irradiation variables. This inherently multiscale evolution emphasizes the importance of closely integrating models with high-resolution experimental characterization of the evolving radiation-damaged microstructure. Lastly,more » this article provides a review of recent models of the defect microstructure evolution in irradiated body-centered cubic materials, which provide good agreement with experimental measurements, and presents some outstanding challenges, which will require coordinated high-resolution characterization and modeling to resolve.« less

Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

2010-06-01

2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has beenmore » restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.« less

Direct measurements of irradiation-induced creep in micropillars of amorphous Cu{sub 56}Ti{sub 38}Ag{sub 6}, Zr{sub 52}Ni{sub 48}, Si, and SiO{sub 2}

DOE Office of Scientific and Technical Information (OSTI.GOV)

Özerinç, Sezer; Kim, Hoe Joon; Averback, Robert S.

2015-01-14

We report in situ measurements of irradiation-induced creep on amorphous (a-) Cu{sub 56}Ti{sub 38}Ag{sub 6}, Zr{sub 52}Ni{sub 48}, Si, and SiO{sub 2}. Micropillars 1 μm in diameter and 2 μm in height were irradiated with ∼2 MeV heavy ions during uniaxial compression at room temperature. The creep measurements were performed using a custom mechanical testing apparatus utilizing a nanopositioner, a silicon beam transducer, and an interferometric laser displacement sensor. We observed Newtonian flow in all tested materials. For a-Cu{sub 56}Ti{sub 38}Ag{sub 6}, a-Zr{sub 52}Ni{sub 48}, a-Si, and Kr{sup +} irradiated a-SiO{sub 2} irradiation-induced fluidities were found to be nearly the same, ≈3 GPa{sup −1}more » dpa{sup −1}, whereas for Ne{sup +} irradiated a-SiO{sub 2} the fluidity was much higher, 83 GPa{sup −1} dpa{sup −1}. A fluidity of 3 GPa{sup −1} dpa{sup −1} can be explained by point-defect mediated plastic flow induced by nuclear collisions. The fluidity of a-SiO{sub 2} can also be explained by this model when nuclear stopping dominates the energy loss, but when the electronic stopping exceeds 1 keV/nm, stress relaxation in thermal spikes also contributes to the fluidity.« less

Effect of heavy ion beam irradiation on germination of local Toraja rice seed (M1-M2) mutant generation

NASA Astrophysics Data System (ADS)

Sjahril, R.; Riadi, M.; Rafiuddin; Sato, T.; Toriyama, K.; Abe, T.; Trisnawaty, A. R.

2018-05-01

Local rice in general has several weaknesses among others, long life, high plant posture and low yield result. The character is a limiting factor that causes farmers low interest to grow local rice. It is feared this will cause the lack of local rice cultivars as germplasm materials. Therefore, there is an effort to create a diversity of morphological characters, as the character of selection, especially related to the age of harvest and plant posture. One method is through breeding mutation by irradiation using ion beam. The objective of this research is to evaluate seed germination resulted after irradiation using ion beam in two varieties of Toraja local rice. The study was prepared based on a randomized block design pattern consisting of six treatments by testing two local Toraja rice varieties namely Pare Ambok and Pare Lea treated with ion beam irradiation of Argon and Carbon ion and control plant as comparison. Each grain from one panicle was germinated in one line method on a Ø15 cm Petri dish and transplanted into small plastic bags. Each treatment was repeated as much as 20 times which was then considered as a strain. The results showed that irradiation using Argon ion in local rice seed of Pare Ambok variety and of Pare Lea varieties produce better seedlings sprouts than irradiation using Carbon ion. Further M2 seed germination shows uniqueness in some seedlings produced such as lighter leaf color, albinism, wrinkled leaf, etc. which could prove potential mutant lines in tested M2 lines seed.

10 CFR 36.2 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... radioactive sealed sources for the irradiation of objects or materials and in which radiation dose rates... source and the area subject to irradiation are contained within a device and are not accessible to... supervisor present. Panoramic dry-source-storage irradiator means an irradiator in which the irradiations...

10 CFR 36.2 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... radioactive sealed sources for the irradiation of objects or materials and in which radiation dose rates... source and the area subject to irradiation are contained within a device and are not accessible to... supervisor present. Panoramic dry-source-storage irradiator means an irradiator in which the irradiations...

10 CFR 36.2 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... radioactive sealed sources for the irradiation of objects or materials and in which radiation dose rates... source and the area subject to irradiation are contained within a device and are not accessible to... supervisor present. Panoramic dry-source-storage irradiator means an irradiator in which the irradiations...

Irradiation stability and thermo-mechanical properties of NITE-SiC irradiated to 10 dpa

DOE PAGES

Terrani, Kurt A.; Ang, Caen; Snead, Lance L.; ...

2017-11-24

In this study, five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of <10 wt%. The dense monolithic materials were subsequently irradiated to 2 and 10 dpa in a mixed spectrum fission reactor at nominally 400 and 700°C. The evolution in swelling, strength, and thermal conductivity of these materials were examined after irradiation, where in all cases properties saturated at < 2dpa, without appreciable change for further irradiation to 10 dpa. Swelling behavior appeared similar to high-purity chemical vapor deposition (CVD) SiC within measurement uncertainty. The strength roughly doubled after irradiation. Finally,more » thermal resistivity increase as a result of irradiation was ~20% higher when compared to CVD-SiC.« less

Irradiation stability and thermo-mechanical properties of NITE-SiC irradiated to 10 dpa

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrani, Kurt A.; Ang, Caen; Snead, Lance L.

In this study, five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of <10 wt%. The dense monolithic materials were subsequently irradiated to 2 and 10 dpa in a mixed spectrum fission reactor at nominally 400 and 700°C. The evolution in swelling, strength, and thermal conductivity of these materials were examined after irradiation, where in all cases properties saturated at < 2dpa, without appreciable change for further irradiation to 10 dpa. Swelling behavior appeared similar to high-purity chemical vapor deposition (CVD) SiC within measurement uncertainty. The strength roughly doubled after irradiation. Finally,more » thermal resistivity increase as a result of irradiation was ~20% higher when compared to CVD-SiC.« less

Recovery of 131I from alkaline solution of n-irradiated tellurium target using a tiny Dowex-1 column.

PubMed

Chattopadhyay, Sankha; Saha Das, Sujata

2010-10-01

A simple and inexpensive ion-exchange chromatography method for the separation of medically useful no-carrier-added (nca) iodine radionuclides from bulk amounts of irradiated tellurium dioxide (TeO(2)) target was developed and tested using (131)I. The radiochemical separation was performed using a very small Dowex-1x8 ion-exchange column. The overall radiochemical yield for the complete separation of (131)I was 92+/-1.8 (standard deviation) % (n=8). The separated nca (131)I was of high, approximately 99%, radionuclidic and radiochemical purity and did not contain detectable amounts of the target material. This method may be adopted for the radiochemical separation of other different iodine radionuclides produced from tellurium matrices through cyclotron as well as reactor irradiation. Copyright 2010 Elsevier Ltd. All rights reserved.

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Chung, H. M.; Gruber, E. E.

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vesselmore » and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.« less

A microwave applicator for uniform irradiation by circularly polarized waves in an anechoic chamber

NASA Astrophysics Data System (ADS)

Chiang, W. Y.; Wu, M. H.; Wu, K. L.; Lin, M. H.; Teng, H. H.; Tsai, Y. F.; Ko, C. C.; Yang, E. C.; Jiang, J. A.; Barnett, L. R.; Chu, K. R.

2014-08-01

Microwave applicators are widely employed for materials heating in scientific research and industrial applications, such as food processing, wood drying, ceramic sintering, chemical synthesis, waste treatment, and insect control. For the majority of microwave applicators, materials are heated in the standing waves of a resonant cavity, which can be highly efficient in energy consumption, but often lacks the field uniformity and controllability required for a scientific study. Here, we report a microwave applicator for rapid heating of small samples by highly uniform irradiation. It features an anechoic chamber, a 24-GHz microwave source, and a linear-to-circular polarization converter. With a rather low energy efficiency, such an applicator functions mainly as a research tool. This paper discusses the significance of its special features and describes the structure, in situ diagnostic tools, calculated and measured field patterns, and a preliminary heating test of the overall system.

Photoelectrolysis of water at high current density - Use of laser light excitation of semiconductor-based photoelectrochemical cells

NASA Technical Reports Server (NTRS)

Wrighton, M. S.; Bocarsley, A. B.; Bolts, J. M.

1978-01-01

In the present paper, some results are given for UV laser light irradiation of the photoanode (SnO2, SrTiO3, or TiO2) in a cell for the light-driven electrolysis of H2O, at radiation intensities of up to 380 W/sq cm. The properties of the anode material are found to be independent of light intensity. Conversion of UV light to stored chemical energy in the form of 2H2/O2 from H2O was driven at a rate of up to 30 W/sq cm. High O2 evolution rates at the irradiated anodes without changes in the current-voltage curves are attributed to the excess oxidizing power associated with photogenerated holes. A test for this sort of hypothesis for H2 evolution at p-type materials is proposed.

Development of Advanced Ods Ferritic Steels for Fast Reactor Fuel Cladding

NASA Astrophysics Data System (ADS)

Ukai, S.; Oono, N.; Ohtsuka, S.; Kaito, T.

Recent progress of the 9CrODS steel development is presented focusing on their microstructure control to improve sufficient high-temperature strength as well as cladding manufacturing capability. The martensitic 9CrODS steel is primarily candidate cladding materials for the Generation IV fast reactor fuel. They are the attractive composite-like materials consisting of the hard residual ferrite and soft tempered martensite, which are able to be easily controlled by α-γ phase transformation. The residual ferrite containing extremely nanosized oxide particles leads to significantly improved creep rupture strength in 9CrODS cladding. The creep strength stability at extended time of 60,000 h at 700 ºC is ascribed to the stable nanosized oxide particles. It was also reviewed that 9CrODS steel has well irradiation stability and fuel pin irradiation test was conducted up to 12 at% burnup and 51 dpa at the cladding temperature of 700ºC.

A microwave applicator for uniform irradiation by circularly polarized waves in an anechoic chamber.

PubMed

Chiang, W Y; Wu, M H; Wu, K L; Lin, M H; Teng, H H; Tsai, Y F; Ko, C C; Yang, E C; Jiang, J A; Barnett, L R; Chu, K R

2014-08-01

Microwave applicators are widely employed for materials heating in scientific research and industrial applications, such as food processing, wood drying, ceramic sintering, chemical synthesis, waste treatment, and insect control. For the majority of microwave applicators, materials are heated in the standing waves of a resonant cavity, which can be highly efficient in energy consumption, but often lacks the field uniformity and controllability required for a scientific study. Here, we report a microwave applicator for rapid heating of small samples by highly uniform irradiation. It features an anechoic chamber, a 24-GHz microwave source, and a linear-to-circular polarization converter. With a rather low energy efficiency, such an applicator functions mainly as a research tool. This paper discusses the significance of its special features and describes the structure, in situ diagnostic tools, calculated and measured field patterns, and a preliminary heating test of the overall system.

Radiation hardness of Ce-doped sol-gel silica fibers for high energy physics applications.

PubMed

Cova, Francesca; Moretti, Federico; Fasoli, Mauro; Chiodini, Norberto; Pauwels, Kristof; Auffray, Etiennette; Lucchini, Marco Toliman; Baccaro, Stefania; Cemmi, Alessia; Bártová, Hana; Vedda, Anna

2018-02-15

The results of irradiation tests on Ce-doped sol-gel silica using x- and γ-rays up to 10 kGy are reported in order to investigate the radiation hardness of this material for high-energy physics applications. Sol-gel silica fibers with Ce concentrations of 0.0125 and 0.05 mol. % are characterized by means of optical absorption and attenuation length measurements before and after irradiation. The two different techniques give comparable results, evidencing the formation of a main broad radiation-induced absorption band, peaking at about 2.2 eV, related to radiation-induced color centers. The results are compared with those obtained on bulk silica. This study reveals that an improvement of the radiation hardness of Ce-doped silica fibers can be achieved by reducing Ce content inside the fiber core, paving the way for further material development.

Formation of a quasi-hollow beam of high-energy heavy ions using a multicell resonance RF deflector

NASA Astrophysics Data System (ADS)

Minaev, S. A.; Sitnikov, A. L.; Golubev, A. A.; Kulevoy, T. V.

2012-09-01

The generation of matter in an extreme state with precisely measurable parameters is of great interest for contemporary physics. One way of obtaining such a state is to irradiate the end of a hollow cylindrical shell at the center of which a test material is kept at a temperature of several Kelvin by an annular beam of high-energy heavy ions. Under the action of the beam, the shell starts explosively expanding both outwards and inwards, compressing the material to an extremely high pressure without subjecting it to direct heating. A method of producing a hollow cylindrical beam of high-energy heavy ions using a resonance rf deflector is described. The deflection of the beam in two transverse directions by means of an rf electric field allows it to rotate about the longitudinal axis and irradiate an annular domain on the end face of the target.

SiC Composite for Fuel Structure Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yueh, Ken

Extensive evaluation was performed to determine the suitability of using SiC composite as a boiling water reactor (BWR) fuel channel material. A thin walled SiC composite box, 10 cm in dimension by approximately 1.5 mm wall thickness was fabricated using chemical vapor deposition (CVD) for testing. Mechanical test results and performance evaluations indicate the material could meet BWR channel mechanical design requirement. However, large mass loss of up to 21% was measured in in-pile corrosion test under BWR-like conditions in under 3 months of irradiation. A fresh sister sample irradiated in a follow-up cycle under PWR conditions showed no measureablemore » weight loss and thus supports the hypothesis that the oxidizing condition of the BWR-like coolant chemistry was responsible for the high corrosion rate. A thermodynamic evaluation showed SiC is not stable and the material may oxidize to form SiO 2 and CO 2. Silica has demonstrated stability in high temperature steam environment and form a protective oxide layer under severe accident conditions. However, it does not form a protective layer in water under normal BWR operational conditions due to its high solubility. Corrosion product stabilization by modifying the SiC CVD surface is an approach evaluated in this study to mitigate the high corrosion rate. Titanium and zirconium have been selected as stabilizing elements since both TiSiO 4 and ZrSiO 4 are insoluble in water. Corrosion test results in oxygenated water autoclave indicate TiSiO4 does not form a protective layer. However, zirconium doped test samples appear to form a stable continuous layer of ZrSiO 4 during the corrosion process. Additional process development is needed to produce a good ZrSiC coating to verify functionality of the mitigation concept.« less

Improved Sterilization of Sensitive Biomaterials with Supercritical Carbon Dioxide at Low Temperature

PubMed Central

Bernhardt, Anne; Wehrl, Markus; Paul, Birgit; Hochmuth, Thomas; Schumacher, Matthias; Schütz, Kathleen; Gelinsky, Michael

2015-01-01

The development of bio-resorbable implant materials is rapidly going on. Sterilization of those materials is inevitable to assure the hygienic requirements for critical medical devices according to the medical device directive (MDD, 93/42/EG). Biopolymer-containing biomaterials are often highly sensitive towards classical sterilization procedures like steam, ethylene oxide treatment or gamma irradiation. Supercritical CO2 (scCO2) treatment is a promising strategy for the terminal sterilization of sensitive biomaterials at low temperature. In combination with low amounts of additives scCO2 treatment effectively inactivates microorganisms including bacterial spores. We established a scCO2 sterilization procedure under addition of 0.25% water, 0.15% hydrogen peroxide and 0.5% acetic anhydride. The procedure was successfully tested for the inactivation of a wide panel of microorganisms including endospores of different bacterial species, vegetative cells of gram positive and negative bacteria including mycobacteria, fungi including yeast, and bacteriophages. For robust testing of the sterilization effect with regard to later application of implant materials sterilization all microorganisms were embedded in alginate/agarose cylinders that were used as Process Challenge Devices (PCD). These PCD served as surrogate models for bioresorbable 3D scaffolds. Furthermore, the impact of scCO2 sterilization on mechanical properties of polysaccharide-based hydrogels and collagen-based scaffolds was analyzed. The procedure was shown to be less compromising on mechanical and rheological properties compared to established low-temperature sterilization methods like gamma irradiation and ethylene oxide exposure as well as conventional steam sterilization. Cytocompatibility of alginate gels and scaffolds from mineralized collagen was compared after sterilization with ethylene oxide, gamma irradiation, steam sterilization and scCO2 treatment. Human mesenchymal stem cell viability and proliferation were not compromised by scCO2 treatment of these materials and scaffolds. We conclude that scCO2 sterilization under addition of water, hydrogen peroxide and acetic anhydride is a very effective, gentle, non-cytotoxic and thus a promising alternative sterilization method especially for biomaterials. PMID:26067982

Importance of the gas phase role to the prediction of energetic material behavior: An experimental study

NASA Astrophysics Data System (ADS)

Ali, A. N.; Son, S. F.; Asay, B. W.; Sander, R. K.

2005-03-01

Various thermal (radiative, conductive, and convective) initiation experiments are performed to demonstrate the importance of the gas phase role in combustion modeling of energetic materials (EM). A previously published condensed phase model that includes a predicted critical irradiance above which ignition is not possible is compared to experimental laser ignition results for octahydro-1,3,5,7-tetranitro-1,3,5,7-tetrazocine (HMX) and 2,4,6-trinitrotoluene (TNT). Experimental results conflict with the predicted critical irradiance concept. The failure of the model is believed to result from a misconception about the role of the gas phase in the ignition process of energetic materials. The model assumes that ignition occurs at the surface and that evolution of gases inhibits ignition. High speed video of laser ignition, oven cook-off and hot wire ignition experiments captures the ignition of HMX and TNT in the gas phase. A laser ignition gap test is performed to further evaluate the effect of gas phase laser absorption and gas phase disruption on the ignition process. Results indicate that gas phase absorption of the laser energy is probably not the primary factor governing the gas phase ignition observations. It is discovered that a critical gap between an HMX pellet and a salt window of 6mm±0.4mm exists below which ignition by CO2 laser is not possible at the tested irradiances of 29W /cm2 and 38W/cm2 for HMX ignition. These observations demonstrate that a significant disruption of the gas phase, in certain scenarios, will inhibit ignition, independent of any condensed phase processes. These results underscore the importance of gas phase processes and illustrate that conditions can exist where simple condensed phase models are inadequate to accurately predict the behavior of energetic materials.

Optimization of microwave-assisted extraction of hydrocarbons in marine sediments: comparison with the Soxhlet extraction method.

PubMed

Vázquez Blanco, E; López Mahía, P; Muniategui Lorenzo, S; Prada Rodríguez, D; Fernández Fernández, E

2000-02-01

Microwave energy was applied to extract polycyclic aromatic hydrocarbons (PAHs) and linear aliphatic hydrocarbons (LAHs) from marine sediments. The influence of experimental conditions, such as different extracting solvents and mixtures, microwave power, irradiation time and number of samples extracted per run has been tested using real marine sediment samples; volume of the solvent, sample quantity and matrix effects were also evaluated. The yield of extracted compounds obtained by microwave irradiation was compared with that obtained using the traditional Soxhlet extraction. The best results were achieved with a mixture of acetone and hexane (1:1), and recoveries ranged from 92 to 106%. The extraction time is dependent on the irradiation power and the number of samples extracted per run, so when the irradiation power was set to 500 W, the extraction times varied from 6 min for 1 sample to 18 min for 8 samples. Analytical determinations were carried out by high-performance liquid chromatography (HPLC) with an ultraviolet-visible photodiode-array detector for PAHs and gas chromatography (GC) using a FID detector for LAHs. To test the accuracy of the microwave-assisted extraction (MAE) technique, optimized methodology was applied to the analysis of standard reference material (SRM 1941), obtaining acceptable results.

Survival of thermophilic and hyperthermophilic microorganisms after exposure to UV-C, ionizing radiation and desiccation.

PubMed

Beblo, Kristina; Douki, Thierry; Schmalz, Gottfried; Rachel, Reinhard; Wirth, Reinhard; Huber, Harald; Reitz, Günther; Rettberg, Petra

2011-11-01

In this study, we investigated the ability of several (hyper-) thermophilic Archaea and phylogenetically deep-branching thermophilic Bacteria to survive high fluences of monochromatic UV-C (254 nm) and high doses of ionizing radiation, respectively. Nine out of fourteen tested microorganisms showed a surprisingly high tolerance against ionizing radiation, and two species (Aquifex pyrophilus and Ignicoccus hospitalis) were even able to survive 20 kGy. Therefore, these species had a comparable survivability after exposure to ionizing radiation such as Deinococcus radiodurans. In contrast, there was nearly no difference in survival of the tested strains after exposure to UV-C under anoxic conditions. If the cells had been dried in advance of UV-C irradiation, they were more sensitive to UV-C radiation compared with cells irradiated in liquid suspension; this effect could be reversed by the addition of protective material like sulfidic ores before irradiation. By exposure to UV-C, photoproducts were formed in the DNA of irradiated Archaea and Bacteria. The distribution of the main photoproducts was species specific, but the amount of the photoproducts was only partly dependent on the applied fluence. Overall, our results show that tolerance to radiation seems to be a common phenomenon among thermophilic and hyperthermophilic microorganisms.

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sindelar, R.; Louthan, M.; PNNL, B.

2015-05-29

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history,more » residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to identical conditions and the material responses to thermo-mechanical exposures will be different depending on the materials and systems used. The discussions at the workshop showed several gaps in the standardization of processes and techniques necessary to assess the long term performance of irradiated zirconium alloy cladding during dry storage and transport. The development of, and adherence to, standards to help bridge these gaps will strengthen the technical basis for long term storage and post-storage operations, provide consistency across the nuclear industry, maximize the value of most observations, and enhance the understanding of behavioral differences among alloys. The need for, and potential benefits of, developing the recommended standards are illustrated in the various sections of this report.« less

Single ion hit detection set-up for the Zagreb ion microprobe

NASA Astrophysics Data System (ADS)

Smith, R. W.; Karlušić, M.; Jakšić, M.

2012-04-01

Irradiation of materials by heavy ions accelerated in MV tandem accelerators may lead to the production of latent ion tracks in many insulators and semiconductors. If irradiation is performed in a high resolution microprobe facility, ion tracks can be ordered by submicrometer positioning precision. However, full control of the ion track positioning can only be achieved by a reliable ion hit detection system that should provide a trigger signal irrespectively of the type and thickness of the material being irradiated. The most useful process that can be utilised for this purpose is emission of secondary electrons from the sample surface that follows the ion impact. The status report of the set-up presented here is based on the use of a channel electron multiplier (CEM) detector mounted on an interchangable sample holder that is inserted into the chamber in a close geometry along with the sample to be irradiated. The set-up has been tested at the Zagreb ion microprobe for different ions and energies, as well as different geometrical arrangements. For energies of heavy ions below 1 MeV/amu, results show that efficient (100%) control of ion impact can be achieved only for ions heavier than silicon. The successful use of the set-up is demonstrated by production of ordered single ion tracks in a polycarbonate film and by monitoring fluence during ion microbeam patterning of Foturan glass.

New Elastomeric Materials Based on Natural Rubber Obtained by Electron Beam Irradiation for Food and Pharmaceutical Use

PubMed Central

Craciun, Gabriela; Manaila, Elena; Stelescu, Maria Daniela

2016-01-01

The efficiency of polyfunctional monomers as cross-linking co-agents on the chemical properties of natural rubber vulcanized by electron beam irradiation was studied. The following polyfunctional monomers were used: trimethylolpropane-trimethacrylate, zinc-diacrylate, ethylene glycol dimethacrylate, triallylcyanurate and triallylisocyanurate. The electron beam treatment was done using irradiation doses in the range of 75 kGy–300 kGy. The gel fraction, crosslink density and effects of different aqueous solutions, by absorption tests, have been investigated as a function of polyfunctional monomers type and absorbed dose. The samples gel fraction and crosslink density were determined on the basis of equilibrium solvent-swelling measurements by applying the modified Flory–Rehner equation for tetra functional networks. The absorption tests were done in accordance with the SR ISI 1817:2015 using distilled water, acetic acid (10%), sodium hydroxide (1%), ethylic alcohol (96%), physiological serum (sodium chloride 0.9%) and glucose (glucose monohydrate 10%). The samples structure and morphology were investigated by Fourier Transform Infrared Spectroscopy and Scanning Electron Microscopy techniques. PMID:28774150

New Elastomeric Materials Based on Natural Rubber Obtained by Electron Beam Irradiation for Food and Pharmaceutical Use.

PubMed

Craciun, Gabriela; Manaila, Elena; Stelescu, Maria Daniela

2016-12-21

The efficiency of polyfunctional monomers as cross-linking co-agents on the chemical properties of natural rubber vulcanized by electron beam irradiation was studied. The following polyfunctional monomers were used: trimethylolpropane-trimethacrylate, zinc-diacrylate, ethylene glycol dimethacrylate, triallylcyanurate and triallylisocyanurate. The electron beam treatment was done using irradiation doses in the range of 75 kGy-300 kGy. The gel fraction, crosslink density and effects of different aqueous solutions, by absorption tests, have been investigated as a function of polyfunctional monomers type and absorbed dose. The samples gel fraction and crosslink density were determined on the basis of equilibrium solvent-swelling measurements by applying the modified Flory-Rehner equation for tetra functional networks. The absorption tests were done in accordance with the SR ISI 1817:2015 using distilled water, acetic acid (10%), sodium hydroxide (1%), ethylic alcohol (96%), physiological serum (sodium chloride 0.9%) and glucose (glucose monohydrate 10%). The samples structure and morphology were investigated by Fourier Transform Infrared Spectroscopy and Scanning Electron Microscopy techniques.

Effects of proton irradiation on structural and electrochemical charge storage properties of TiO 2 nanotube electrodes for lithium-ion batteries

DOE PAGES

Smith, Kassiopeia A.; Savva, Andreas I.; Deng, Changjian; ...

2017-03-23

The effects of proton irradiation on nanostructured metal oxides have been investigated. Recent studies suggest that the presence of structural defects (e.g. vacancies and interstitials) in metal oxides may enhance the material's electrochemical charge storage capacity. A new approach to introduce defects in electrode materials is to use ion irradiation as it can produce a supersaturation of point defects in the target material. In this work we report the effect of low-energy proton irradiation on amorphous TiO 2 nanotube electrodes at both room temperature and high temperature (250 °C). Upon room temperature irradiation the nanotubes demonstrate an irradiation-induced phase transformationmore » to a mixture of amorphous, anatase, and rutile domains while showing a 35% reduction in capacity compared to anatase TiO 2. On the other hand, the high temperature proton irradiation induced a disordered rutile phase within the nanotubes as characterized by Raman spectroscopy and transmission electron microscopy, which displays an improved capacity by 20% at ~240 mA h g –1 as well as improved rate capability compared to an unirradiated anatase sample. Voltammetric sweep data were used to determine the contributions from diffusion-limited intercalation and capacitive processes and it was found that the electrodes after irradiation had more contributions from diffusion in lithium charge storage. Finally, our work suggests that tailoring the defect generation through ion irradiation within metal oxide electrodes could present a new avenue for designing advanced electrode materials.« less

Effects of proton irradiation on structural and electrochemical charge storage properties of TiO 2 nanotube electrodes for lithium-ion batteries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Kassiopeia A.; Savva, Andreas I.; Deng, Changjian

The effects of proton irradiation on nanostructured metal oxides have been investigated. Recent studies suggest that the presence of structural defects (e.g. vacancies and interstitials) in metal oxides may enhance the material's electrochemical charge storage capacity. A new approach to introduce defects in electrode materials is to use ion irradiation as it can produce a supersaturation of point defects in the target material. In this work we report the effect of low-energy proton irradiation on amorphous TiO 2 nanotube electrodes at both room temperature and high temperature (250 °C). Upon room temperature irradiation the nanotubes demonstrate an irradiation-induced phase transformationmore » to a mixture of amorphous, anatase, and rutile domains while showing a 35% reduction in capacity compared to anatase TiO 2. On the other hand, the high temperature proton irradiation induced a disordered rutile phase within the nanotubes as characterized by Raman spectroscopy and transmission electron microscopy, which displays an improved capacity by 20% at ~240 mA h g –1 as well as improved rate capability compared to an unirradiated anatase sample. Voltammetric sweep data were used to determine the contributions from diffusion-limited intercalation and capacitive processes and it was found that the electrodes after irradiation had more contributions from diffusion in lithium charge storage. Finally, our work suggests that tailoring the defect generation through ion irradiation within metal oxide electrodes could present a new avenue for designing advanced electrode materials.« less

Infrared analysis of polyethylene wear specimens using attenuated total reflection spectroscopy. [effects of radiation on the surface properties of materials for total joint protheses

NASA Technical Reports Server (NTRS)

Jones, W. R.; Lauer, J. L.

1979-01-01

Attenuated total reflection infrared spectroscopy was used to analyze ultrahigh molecular weight polyethylene wear test specimens. Three different specimens were analyzed. One specimen was gamma irradiated to a dose of 5.0 MRad, another to a dose of 2.5 MRad, and the final specimen was unirradiated. There was no conclusive evidence of chemical changes (i.e., unsaturation or oxidation) in the surface regions of any of the polyethylene samples. Therefore, it was concluded that the gamma irradiation sterilization procedure shoud not alter the boundary lubricating properties of the polyethylene.

10 CFR 70.20b - General license for carriers of transient shipments of formula quantities of strategic special...

Code of Federal Regulations, 2010 CFR

2010-01-01

... significance, special nuclear material of low strategic significance, and irradiated reactor fuel. 70.20b..., special nuclear material of low strategic significance, and irradiated reactor fuel. (a) A general license... requirements of § 73.67 of this chapter. (3) Irradiated reactor fuel of the type and quantity subject to the...

10 CFR 70.20b - General license for carriers of transient shipments of formula quantities of strategic special...

Code of Federal Regulations, 2011 CFR

2011-01-01

... significance, special nuclear material of low strategic significance, and irradiated reactor fuel. 70.20b..., special nuclear material of low strategic significance, and irradiated reactor fuel. (a) A general license... requirements of § 73.67 of this chapter. (3) Irradiated reactor fuel of the type and quantity subject to the...