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

International Nuclear Information System (INIS)

Ishino, Shiori; Terai, Takayuki; Oku, Tatsuo

1999-08-01

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

High-temperature annealing of proton irradiated beryllium – A dilatometry-based study

Energy Technology Data Exchange (ETDEWEB)

Simos, Nikolaos, E-mail: simos@bnl.gov [Brookhaven National Laboratory, Upton, NY, 11973 (United States); Elbakhshwan, Mohamed; Zhong, Zhong; Ghose, Sanjit [Brookhaven National Laboratory, Upton, NY, 11973 (United States); Savkliyildiz, Ilyas [Rutgers University (United States)

2016-08-15

S−200 F grade beryllium has been irradiated with 160 MeV protons up to 1.2 10{sup 20} cm{sup −2} peak fluence and irradiation temperatures in the range of 100–200 °C. To address the effect of proton irradiation on dimensional stability, an important parameter in its consideration in fusion reactor applications, and to simulate high temperature irradiation conditions, multi-stage annealing using high precision dilatometry to temperatures up to 740 °C were conducted in air. X-ray diffraction studies were also performed to compliment the macroscopic thermal study and offer a microscopic view of the irradiation effects on the crystal lattice. The primary objective was to qualify the competing dimensional change processes occurring at elevated temperatures namely manufacturing defect annealing, lattice parameter recovery, transmutation {sup 4}He and {sup 3}H diffusion and swelling and oxidation kinetics. Further, quantification of the effect of irradiation dose and annealing temperature and duration on dimensional changes is sought. The study revealed the presence of manufacturing porosity in the beryllium grade, the oxidation acceleration effect of irradiation including the discontinuous character of oxidation advancement, the effect of annealing duration on the recovery of lattice parameters recovery and the triggering temperature for transmutation gas diffusion leading to swelling.

High-fluence hyperthermal ion irradiation of gallium nitride surfaces at elevated temperatures

Energy Technology Data Exchange (ETDEWEB)

Finzel, A.; Gerlach, J.W., E-mail: juergen.gerlach@iom-leipzig.de; Lorbeer, J.; Frost, F.; Rauschenbach, B.

2014-10-30

Highlights: • Irradiation of gallium nitride films with hyperthermal nitrogen ions. • Surface roughening at elevated sample temperatures was observed. • No thermal decomposition of gallium nitride films during irradiation. • Asymmetric surface diffusion processes cause local roughening. - Abstract: Wurtzitic GaN films deposited on 6H-SiC(0001) substrates by ion-beam assisted molecular-beam epitaxy were irradiated with hyperthermal nitrogen ions with different fluences at different substrate temperatures. In situ observations with reflection high energy electron diffraction showed that during the irradiation process the surface structure of the GaN films changed from two dimensional to three dimensional at elevated temperatures, but not at room temperature. Atomic force microscopy revealed an enhancement of nanometric holes and canyons upon the ion irradiation at higher temperatures. The roughness of the irradiated and heated GaN films was clearly increased by the ion irradiation in accordance with x-ray reflectivity measurements. A sole thermal decomposition of the films at the chosen temperatures could be excluded. The results are discussed taking into account temperature dependent sputtering and surface uphill adatom diffusion as a function of temperature.

Release behavior of fission products from irradiated dispersion fuels at high temperatures

International Nuclear Information System (INIS)

Iwai, Takashi; Shimizu, Michio; Nakagawa, Tetsuya

1990-02-01

As a framework of reduced enrichment fuel program of JMTR Project, the measurements of fission products release rates at high temperatures (600degC - 1100degC) were performed in order to take the data to use for safety evaluation of LEU fuel. Three type miniplates of dispersion silicide and aluminide fuel, 20% enrichment LEU fuel with 4.8 gU/cc (U 3 Si 2 90 %, USi 10 % and U 3 Si 2 50 %, U 3 Si 50 % dispersed in aluminium) and 45 % enrichment MEU fuel with 1.6 gU/cc, were irradiated in JMTR. The burnups attained by one cycle (22 days) irradiation were within 21.6 % - 22.5 % of initial 235 U. The specimens cut down from miniplates were measured on fission products release rates by means of new apparatus specially designed for this experiment. The specimens were heated up within 600degC - 1100degC in dry air. Then fission products such as 85 Kr, 133 Xe, 131 I, 137 Cs, 103 Ru, 129m Te were collected at each temperature and measured on release rates. In the results of measurement, the release rates of 85 Kr, 133 Xe, 131 I, 129m Te from all specimens were slightly less than that of G.W. Parker's data on U-Al alloy fuel. For 137 Cs and 103 Ru from a silicide specimen (U 3 Si 2 90 %, USi 10 % dispersed in aluminium) and 137 Cs from an aluminide specimen, the release rates were slightly higher than that of G.W. Parker's. (author)

Fluence behavior of polyethylene films irradiated with high energy electrons

International Nuclear Information System (INIS)

Pino, Eddy Segura; Silva, Leonardo G. Andrade e

1999-01-01

Polymers are viscoelastic materials at all temperatures, so that mechanical loads induce time dependable deformations. The recovery of these deformations, on load release, take some time and it is not always recovered completely. The main objective of this work was to analyse the creep behavior of electron irradiated polyethylene films. From the experimental results, it was sated that polyethylene creeps less with an increase on irradiation dose and also that creep recovery in this material increases with doses but it is not complete. This behavior can be attributed to the crosslinking effect witch stabilize elements of the molecular structure of the polyethylene, thus reducing their mobility and so inhibiting the creep mechanism. The partial creep recovery could be also attributed to the reticulation effect and to the polyethylene plastic behavior. Additional information on the creep behavior was obtained by fitting the experimental data with exponential functions and evaluating the mathematical parameters with a modified Kelvin-Voigt mechanical model. (author)

Understanding the Irradiation Behavior of Zirconium Carbide

International Nuclear Information System (INIS)

Motta, Arthur; Sridharan, Kumar; Morgan, Dane; Szlufarska, Izabela

2013-01-01

-induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.

Thermal conductivity of beryllium under low temperature high dose neutron irradiation

International Nuclear Information System (INIS)

Chakin, V.P.; Latypov, R.N.; Suslov, D.N.; Kupriyanov, I.B.

2004-01-01

Thermal conductivity of compact beryllium of several Russian grades such as TE-400, TE-56, TE-30, TIP and DIP differing in the production technology, grain size and impurity content has been investigated. The thermal diffusivity of beryllium was measured on the disks in the initial and irradiated conditions using the pulse method in the range from room temperature to 200degC. The thermal conductivity was calculated using the table values for the beryllium thermal capacity. The specimens and beryllium neutron source fragments were irradiation in the SM reactor at 70degC and 200degC to a neutron fluence of (0.5-11.4)·10 22 cm -2 (E>0.1 MeV) and in the BOR-60 reactor at 400degC to 16·10 22 cm -2 (E>0.1MeV), respectively. The low-temperature irradiation leads to the drop decrease of the beryllium thermal conductivity and the effect depends on the irradiation parameters. The paper analyses the effect of irradiation parameters (temperature, neutron fluence), measurement temperature and structural factors on beryllium conductivity. The experiments have revealed that the short time post-irradiation annealing at high temperature results in partial reduction of the thermal conductivity of irradiated beryllium. (author)

Radiation damage in stainless steel under varying temperature neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Yoshida, Naoaki [Kyushu Univ., Kasuga, Fukuoka (Japan). Research Inst. for Applied Mechanics

1998-03-01

Microstructural evolution of model alloys of 316SS was examined by neutron irradiation at JMTR under cyclic temperature varying condition. In the case of Fe-16Cr-17Ni, formation of interstitial loops and voids are strongly suppressed by varying the temperature from 473K to 673K. By adding Ti as miner element (0.25wt%), however, abnormal accumulation of vacancies (void swelling of 11%dpa at 0.1dpa) was observed. Theoretical analysis standing on the rate theory of defect clustering and simulation irradiation experiments with heavy ions indicates that the vacancy-rich condition which appears temporally during and after changing the temperature from low to high brings these results. It was also shown that only 1 dpa pre-irradiation at low temperature changes swelling behavior at high temperature above several 10 dpa. The understanding of non-steady-state defect processes under temperature varying irradiation is very important to estimate the radiation damage under fusion environment where short-term and long-term temperature variation is expected. (author)

The corrosion behavior of hafnium in high-temperature-water environments

Energy Technology Data Exchange (ETDEWEB)

Rishel, D.M.; Smee, J.D.; Kammenzind, B.F.

1999-10-01

The high-temperature-water corrosion performance of hafnium is evaluated. Corrosion kinetic data are used to develop correlations that are a function of time and temperature. The evaluation is based on corrosion tests conducted in out-of-pile autoclaves and in out-of-flux locations of the Advanced Test Reactor (ATR) at temperatures ranging from 288 to 360 C. Similar to the corrosion behavior of unalloyed zirconium, the high-temperature-water corrosion response of hafnium exhibits three corrosion regimes: pretransition, posttransition, and spalling. In the pretransition regime, cubic corrosion kinetics are exhibited, whereas in the posttransition regime, linear corrosion kinetics are exhibited. Because of the scatter in the spalling regime data, it is not reasonable to use a best fit of the data to describe spalling regime corrosion. Data also show that neutron irradiation does not alter the corrosion performance of hafnium. Finally, the data illustrate that the corrosion rate of hafnium is significantly less than that of Zircaloy-2 and Zircaloy-4.

Property changes in graphite irradiated at changing irradiation temperature

International Nuclear Information System (INIS)

Price, R.J.; Haag, G.

1979-07-01

Design data for irradiated graphite are usually presented as families of isothermal curves showing the change in physical property as a function of fast neutron fluence. In this report, procedures for combining isothermal curves to predict behavior under changing irradiation temperatures are compared with experimental data on irradiation-induced changes in dimensions, Young's modulus, thermal conductivity, and thermal expansivity. The suggested procedure fits the data quite well and is physically realistic

Saturation behavior of irradiation hardening in F82H irradiated in the HFIR

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-01

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

Zircaloy behaviour in high temperature irradiated water

International Nuclear Information System (INIS)

Urbanic, V.F.

1982-04-01

The corrosion and hydriding of Zircaloy during irradiation in high temperature water is strongly dependent on the oxygen concentration of the water. Corrosion tests in the NRX and NRU research reactors using small samples have demonstrated the importance of water chemistry in maintaining Zircaloy corrosion and hydriding within acceptable limits. Zircaloy fuel cladding develops non-uniform, patch-type oxides during irradiation in hich temperature water containing dissolved oxygen. Results from examinations of prototype fuel cladding irradiated in the research reactors are presented to show how local variations in coolant flow, fast neutron flux, metallurgical structure and surface condition can influence the onset of non-uniform corrosion under these conditions. Destructive examinations of CANDU-PHW reactor fuel cladding have emphasized the importance of good chemistry control, especially the dissolved oxygen concentration of the water. When reactor coolants are maintained under normal reducing conditions at high pH (5 to 10 cm 3 D 2 /kg D 2 O; 2 /kg D 2 O; pH > 10 with LiOD), Zircaloy cladding develops non-uniform, patch-type oxides. These patch-type oxides tend to coalesce with time to form a thick, uniform oxide layer after extended exposure. Under reducing coolant conditions, Zircaloy cladding absorbs less than 200 mg D/kg Zr (approximately 2.5 mg/dm 2 equivalent hydrogen) in about 500 days. With oxygen in the coolant, deuterium absorption is considerably less despite the significant increase in corrosion under such conditions

Experimental study of pyrolytic boron nitride at high temperature with and without proton and VUV irradiations

International Nuclear Information System (INIS)

Balat-Pichelin, M.; Eck, J.; Heurtault, S.; Glénat, H.

2014-01-01

Highlights: • New results for the high temperature study of pBN in high vacuum for the heat shield of solar probes. • Physico-chemical behavior of pBN studied up to 1700 K with proton and VUV irradiations. • Rather low effect of synergistic aggressions on the microstructure of pBN material. • The α/ε ratio of pBN coating on C/C measured up to 2200 K is 20% lower than for the C/C itself. - Abstract: In the frame of future exploration missions such as Solar Probe Plus (NASA) and PHOIBOS (ESA), research was carried out to study pyrolytic BN material envisaged as coating for their heat shields. The physico-chemical behavior of CVD pBN at very high temperature with or without hydrogen ions and VUV (Vacuum Ultra-Violet) irradiations was studied in high vacuum together with the in situ measurement of the thermal radiative properties conditioning the thermal equilibrium of the heat shield. Experimental results obtained on massive pBN samples are presented through in situ mass spectrometry and mass loss rate, and post-test microstructural characterization by XRD, SEM, AFM and nano-indentation techniques, some of them leading to mechanical properties. It could be concluded that synergistic effect of high temperature, protons and VUV radiation has an impact on the emission of gaseous species, the mass loss rate and the mechanical properties of the material

The behavior of lattice defects produced in Al2O3 irradiated by neutrons at high temperatures

International Nuclear Information System (INIS)

Atobe, K.; Koizumi, T.; Okada, M.

2003-01-01

Single crystals of α-Al 2 O 3 were irradiated by the two reactors, KUR and JMTR, at three different temperatures. Lattice defects produced by irradiation were studied by esr (electron spin resonance). Three kinds of esr spectram, which are denoted as A, B and C spectram, are observed. The spectram A was observed at three different irradiation temperatures and was ascribed to oxygen vacancies. The spectram B showed no angular dependence for the rotation of external magnetic field to the crystal axis, and the defect density of this spectram decreased with an increase of annealing temperature. When the specimen was annealed at 400 degC after irradiation at 200 degC, the spectram C was observed and was presumed to be due to Al-colloids. (Y. Kazumata)

Production of nanodiamonds by high-energy ion irradiation of graphite at room temperature

International Nuclear Information System (INIS)

Daulton, T.L.; Kirk, M.A.; Lewis, R.S.; Rehn, L.E.

2001-01-01

It has previously been shown that graphite can be transformed into diamond by MeV electron and ion irradiation at temperatures above approximately 600 deg. C. However, there exists geological evidence suggesting that carbonaceous materials can be transformed to diamond by irradiation at substantially lower temperatures. For example, submicron-size diamond aggregates have been found in uranium-rich, Precambrian carbonaceous deposits that never experienced high temperature or pressure. To test if diamonds can be formed at lower irradiation temperatures, sheets of fine-grain polycrystalline graphite were bombarded at 20 deg. C with 350±50 MeV Kr ions to fluences of 6x10 12 cm -2 using the Argonne tandem linear accelerator system (ATLAS). Ion-irradiated (and unirradiated control) graphite specimens were then subjected to acid dissolution treatments to remove untransformed graphite and isolate diamonds that were produced; these acid residues were subsequently characterized by high-resolution and analytical electron microscopy. The acid residue of the ion-irradiated graphite was found to contain nanodiamonds, demonstrating that ion irradiation of graphite at ambient temperature can produce diamond. The diamond yield under our irradiation conditions is low, ∼0.01 diamonds/ion. An important observation that emerges from comparing the present result with previous observations of diamond formation during irradiation is that nanodiamonds form under a surprisingly wide range of irradiation conditions. This propensity may be related to the very small difference in the graphite and diamond free-energies coupled with surface-energy considerations that may alter the relative stability of diamond and graphite at nanometer sizes

Production behavior of irradiation defects in solid breeder materials

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

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

International Nuclear Information System (INIS)

Sacksteder, Irene

2011-01-01

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

Long duration performance of high temperature irradiation resistant thermocouples

International Nuclear Information System (INIS)

Rempe, J.; Knudson, D.; Condie, K.; Cole, J.; Wilkins, S.C.

2007-01-01

Many advanced nuclear reactor designs require new fuel, cladding, and structural materials. Data are needed to characterize the performance of these new materials in high temperature, radiation conditions. However, traditional methods for measuring temperature in-pile degrade at temperatures above 1100 C degrees. To address this instrumentation need, the Idaho National Laboratory (INL) developed and evaluated the performance of a high temperature irradiation-resistant thermocouple that contains alloys of molybdenum and niobium. To verify the performance of INL's recommended thermocouple design, a series of high temperature (from 1200 to 1800 C) long duration (up to six months) tests has been initiated. This paper summarizes results from the tests that have been completed. Data are presented from 4000 hour tests conducted at 1200 and 1400 C that demonstrate the stability of this thermocouple (less than 2% drift). In addition, post test metallographic examinations are discussed which confirm the compatibility of thermocouple materials throughout these long duration, high temperature tests. (authors)

Irradiation performance of AGR-1 high temperature reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

2014-10-01

The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

Observation on the irradiation behavior of U-Mo alloy dispersion fuel

International Nuclear Information System (INIS)

Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

2000-01-01

Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

Lipid and carotenoid production by Rhodotorula glutinis under irradiation/high-temperature and dark/low-temperature cultivation.

Science.gov (United States)

Zhang, Zhiping; Zhang, Xu; Tan, Tianwei

2014-04-01

The capacity of lipid and carotenoid production by Rhodotorula glutinis was investigated under different irradiation conditions, temperatures and C/N ratios. The results showed that dark/low-temperature could enhance lipid content, while irradiation/high-temperature increased the yields of biomass and carotenoid. The optimum C/N ratio for production was between 80 and 100. A two-stage cultivation strategy was used for lipid and carotenoid production in a 5L fermenter. In the first stage, the maximum biomass reached 28.1g/L under irradiation/high-temperature. Then, the cultivation condition was changed to dark/low-temperature, and C/N ratio was adjusted to 90. After the second stage, the biomass, lipid content and carotenoid reached 86.2g/L, 26.7% and 4.2mg/L, respectively. More significantly, the yields of biomass and lipid were 43.1% and 11.5%, respectively. Lipids contained 79.7% 18C and 16.8% 16C fatty acids by GC analysis. HPLC quantified the main carotenoids were β-carotene (68.4%), torularhodin (21.5%) and torulene (10.1%). Crown Copyright © 2014. Published by Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

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

1987-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1987-09-01

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

Irradiation behavior evaluation of oxide dispersion strengthened ferritic steel cladding tubes irradiated in JOYO

Energy Technology Data Exchange (ETDEWEB)

Yamashita, Shinichiro, E-mail: yamashita.shinichiro@jaea.go.jp; Yano, Yasuhide; Ohtsuka, Satoshi; Yoshitake, Tsunemitsu; Kaito, Takeji; Koyama, Shin-ichi; Tanaka, Kenya

2013-11-15

Irradiation behavior of ODS steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS steel. Two types of ODS (9Cr–ODS{sub F}/M, 12Cr–ODS{sub F}) steel cladding tubes with differences in basic compositions and matrix phases were irradiated in JOYO. Post-irradiation examination data concerning hardness, ring tensile property, and microstructure were obtained. Hardness measurement after irradiation showed that there was an apparent irradiation temperature dependence on hardness for 9Cr–ODS{sub F}/M steel whereas no distinct temperature dependence for 12Cr–ODS{sub F} steel. Also, there was no significant change in tensile strengths after irradiation below 923 K, but those above 1023 K up to 6.6 × 10{sup 26} n/m{sup 2} (E > 0.1 MeV) were decreased by about 20%. TEM observations showed that the radiation-induced defect cluster formation during irradiation was suppressed because of high density sink site for defect such as initially-existed dislocation, and precipitate interfaces. In addition, oxide particles were stable up to the maximum doses of this irradiation test.

Ablation Behavior of Plasma-Sprayed La1-xSrxTiO3+δ Coating Irradiated by High-Intensity Continuous Laser.

Science.gov (United States)

Zhu, Jinpeng; Ma, Zhuang; Gao, Yinjun; Gao, Lihong; Pervak, Vladimir; Wang, Lijun; Wei, Chenghua; Wang, Fuchi

2017-10-11

Laser protection for optical components, particularly those in high-power laser systems, has been a major concern. La 1-x Sr x TiO 3+δ with its good optical and thermal properties can be potentially applied as a high-temperature optical protective coating or high-reflectivity material for optical components. However, the high-power laser ablation behavior of plasma-sprayed La 1-x Sr x TiO 3+δ (x = 0.1) coatings has rarely been investigated. Thus, in this study, laser irradiation experiments were performed to study the effect of high-intensity continuous laser on the ablation behavior of the La 1-x Sr x TiO 3+δ coating. The results show that the La 1-x Sr x TiO 3+δ coating undergoes three ablation stages during laser irradiation: coating oxidation, formation and growth of new structures (columnar and dendritic crystals), and mechanical failure. A finite-element simulation was also conducted to explore the mechanism of the ablation damage to the La 1-x Sr x TiO 3+δ coating and provided a good understanding of the ablation behavior. The apparent ablation characteristics are attributed to the different temperature gradients determined by the reflectivity and thermal diffusivity of the La 1-x Sr x TiO 3+δ coating material, which are critical factors for improving the antilaser ablation property. Now, the stainless steel substrate deposited by it can effectively work as a protective shield layer against ablation by laser irradiation.

Structural behavior of reinforced concrete structures at high temperatures

International Nuclear Information System (INIS)

Yamazaki, N.; Yamazaki, M.; Mochida, T.; Mutoh, A.; Miyashita, T.; Ueda, M.; Hasegawa, T.; Sugiyama, K.; Hirakawa, K.; Kikuchi, R.; Hiramoto, M.; Saito, K.

1995-01-01

To establish a method to predict the behavior of reinforced concrete structures subjected simultaneously to high temperatures and external loads, this paper presents the results obtained in several series of tests carried out recently in Japan. This paper reports on the material properties of concrete and steel bars under high temperatures. It also considers the heat transfer properties of thick concrete walls under transient high temperatures, and the structural behavior of reinforced concrete beams subjected to high temperatures. In the tests, data up to 800 C were obtained for use in developing a computational method to estimate the non-linear behavior of reinforced concrete structures exposed to high temperatures. (orig.)

Irradiation-assisted stress corrosion cracking considerations at temperatures below 288 degree C

International Nuclear Information System (INIS)

Simonen, E.P.; Jones, R.H.; Bruemmer, S.M.

1995-03-01

Irradiation-assisted stress corrosion cracking (IASCC) occurs above a critical neutron fluence in light-water reactor (LWR) water environments at 288 C, but very little information exists to indicate susceptibility as temperatures are reduced. Potential low-temperature behavior is assessed based on the temperature dependencies of intergranular (IG) SCC in the absence of irradiation, radiation-induced segregation (RIS) at grain boundaries and micromechanical deformation mechanisms. IGSCC of sensitized SS in the absence of irradiation exhibits high growth rates at temperatures down to 200 C under conditions of anodic dissolution control, while analysis of hydrogen-induced cracking suggests a peak crack growth rate near 100 C. Hence from environmental considerations, IASCC susceptibility appears to remain likely as water temperatures are decreased. Irradiation experiments and model predictions indicate that RIS also persists to low temperatures. Chromium depletion may be significant at temperatures below 100C for irradiation doses greater than 10 displacements per atom (dpa). Macromechanical effects of irradiation on strength and ductility are not strongly dependent on temperature below 288 C. However, temperature does significantly affect radiation effects on SS microstructure and micromechanical deformation mechanisms. The critical conditions for material susceptibility to IASCC at low temperatures may be controlled by radiation-induced grain boundary microchemistry, strain localization due to irradiation microstructure and irradiation creep processes. 39 refs

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Directory of Open Access Journals (Sweden)

Mohd Idzat Idris

2016-12-01

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

High-temperature irradiation effects on mechnical properties of HTGR graphites

International Nuclear Information System (INIS)

Oku, Tatsuo; Eto, Motokuni; Fujisaki, Katsuo

1978-04-01

The irradiation effects on stress-strain relation, Young's modulus, tensile strength, bending strength and compressive strength of HTGR graphites were studied in irradiation temperature ranges of 200 - 300 0 C and 800 - 1400 0 C and in neutron fluences up to 7.4 x 10 20 n/cm 2 and 3 x 10 21 n/cm 2 (> 0.18 MeV). Fracture criteria and strain energy to fracture of the unirradiated and the irradiated graphites were also examined. (1) Neutron fluence dependences are similar in Young's modulus, tensile strength and bending strength. (2) The change of compressive strength and of tensile and bending strengths with neutron fluence differ; the former varies with graphite kind. (3) At lower irradiation temperatures the bending fracture strain energy decreases with increasing neutron fluence and at higher irradiation temperatures it increases. (4) The fracture criteria of graphites deviates from the constant strain energy theory (α = 0.5) and the constant strain theory (α = 1), shifting from α asymptotically equals 0.5 to α asymptotically equals 1 with increasing irradiation temperature. (auth.)

High temperature superconductors for fusion magnets -influence of neutron irradiation

International Nuclear Information System (INIS)

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

2010-01-01

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

The behavior of lattice defects produced in Al{sub 2}O{sub 3} irradiated by neutrons at high temperatures

Energy Technology Data Exchange (ETDEWEB)

Atobe, K.; Koizumi, T. [Naruto Univ. of Education, Tokushima (Japan); Okada, M. [Kyoto Univ., Research Reactor Inst., Kumatori, Osaka (Japan)

2003-01-01

Single crystals of {alpha}-Al{sub 2}O{sub 3} were irradiated by the two reactors, KUR and JMTR, at three different temperatures. Lattice defects produced by irradiation were studied by esr (electron spin resonance). Three kinds of esr spectram, which are denoted as A, B and C spectram, are observed. The spectram A was observed at three different irradiation temperatures and was ascribed to oxygen vacancies. The spectram B showed no angular dependence for the rotation of external magnetic field to the crystal axis, and the defect density of this spectram decreased with an increase of annealing temperature. When the specimen was annealed at 400 degC after irradiation at 200 degC, the spectram C was observed and was presumed to be due to Al-colloids. (Y. Kazumata)

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

Science.gov (United States)

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

1988-07-01

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

Status for development of a capsule and instruments for high-temperature irradiation in HANARO

Energy Technology Data Exchange (ETDEWEB)

Cho, Man Soon; Choo, Kee Nam; Lee, Chul Yong; Yang, Seong Woo; Shim, Kyue Taek; Chung, Hwan-Sung [Korea Atomic Energy Research Institute, Taejeon (Korea, Republic of)

2012-03-15

As the reactors planned in the Gen-IV program will be operated at high temperature and under high neutron flux, the requirements for irradiation of materials at high temperature are recently being gradually increased. The irradiation tests of materials in HANARO up to the present have been performed usually at temperatures below 300degC at which the RPV materials of the commercial reactors are being operated. To overcome the restriction for high-temperature use of Al thermal media of the existing standard capsule, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated as a more advanced capsule than the single thermal media capsule. (author)

Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

Energy Technology Data Exchange (ETDEWEB)

Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D. [Laboratoire PROMES-CNRS, Laboratoire Procedes, Materiaux et Energie Solaire, 7 rue du Four Solaire, 66120 Font-Romeu Odeillo (France)

2007-07-01

Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H{sup +} ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

International Nuclear Information System (INIS)

Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D.

2007-01-01

Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H + ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

Energy Technology Data Exchange (ETDEWEB)

Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

1997-08-01

Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

Temperature dependence of magnetoresistance in neutron-irradiated and unirradiated high resistivity p-type silicon

International Nuclear Information System (INIS)

Yildirim, M.; Efeoglu, H.; Abay, B.; Yogurtcu, Y.K.

1996-01-01

The temperature dependence of the transverse magnetoresistance in irradiated and unirradiated p-type Si is studied in the range from 120 to 290 K. The magnetoresistance coefficients for the unirradiated left angle 001 right angle and left angle 1 anti 10 right angle samples increases with decreasing sample temperature in the range from 160 to 290 K, however, this behavior is reversed below 160 K. It is proposed that this reversal is due to the double injection effect. The magnetoresistance coefficient for the irradiated left angle 001 right angle sample increases with decreasing sample temperature in the range of 120 to 290 K and is greater than that for the unirradiated left angle 001 right angle sample. This result can be explained by increased scattering due to the increased number of defects produced by irradiation. On the other hand, the magnetoresistance coefficient for the unirradiated left angle 1 anti 10 right angle sample is found to be greater than that of the unirradiated left angle 001 right angle sample. (orig.)

Use of TRIGA-pulsed irradiations for high-temperature Doppler measurements

Energy Technology Data Exchange (ETDEWEB)

Foell, W K; Cashwell, R J; Bhattacharyya, S K [Argonne National Laboratory, Argonne, IL (United States); Russell, G J [Los Alamos Scientific Laboratory, University of California, Los Alamos, NM (United States)

1974-07-01

Conventional activation and reactivity measurements of the nuclear Doppler Effect have been limited to temperatures of about 2000{sup o}K because of problems with furnace equipment. There is a need for Doppler data at higher temperatures for design of reactors and analysis of reactor accidents. To fill this need, a novel technique using pulsed-mode operation of a TRIGA reactor has been developed at the University of Wisconsin. This new method, the Pulsed Activation Doppler (PAD) technique, has been used successfully for high temperature Doppler measurements of UO{sub 2} fuel pellets. In the PAD technique, UO{sub 2} test pellets were doped with varying amounts of U-235, with fissile enrichments varying from 0.22% to 12% by weight. The pellets were encapsulated in individual irradiation cells and electrically preheated to predetermined temperatures. Pyrofoam-graphite heaters were used to give preheat temperatures of up to 1720 deg. K. The cells were then positioned in the University of Wisconsin TRIGA reactor core and pulse-irradiated. During the rapid irradiation, adiabatic fission energy deposition occurred in the pellets and very high temperatures (over 3115 deg, K) were attained. Corresponding resonance neutron captures occurred at the elevated temperatures. The Doppler Ratio was deduced from the gamma activities of the Np-239 in the heated and unheated reference pellets. UO{sub 2} pellets of two nominal diameters, 210 mils (a surface-to-mass ratio, s/m = 1.1 cm{sup 2} /gm) and 360 mils (s/m = 0.63 cm{sup 2}/gm), were used for the experiments. For the 210 mil diameter pellets there was very good agreement between experimental results and Doppler ratios predicted both from extrapolations of the Hellstrand low-temperature resonance integral correlations and from GAROL calculations. Significantly, the agreement was good even for those pellets which experienced extensive melting. For the 360 mil diameter pellets the theoretical predictions were 10-15% lower than

Recovery behavior of high purity cubic SiC polycrystals by post-irradiation annealing up to 1673 K after low temperature neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Idris, Mohd Idzat, E-mail: idzat.i.aa@m.titech.ac.jp [Department of Nuclear Engineering, Graduate School of Science and Engineering, Tokyo Institute of Technology, 2-12-1, Ookayama, Meguro-ku, Tokyo 152-8550 Japan (Japan); The National University of Malaysia, School of Applied Physics, Faculty of Science and Technology, 43600 Bangi Selangor (Malaysia); Yamazaki, Saishun; Yoshida, Katsumi; Yano, Toyohiko [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1, Ookayama, Meguro-ku, Tokyo 152-8550 Japan (Japan)

2015-10-15

Two kinds of high purity cubic (β) SiC polycrystals, PureBeta-SiC and CVD-SiC, were irradiated in the BR2 reactor (Belgium) up to a fluence of 2.0–2.5 × 10{sup 24} (E > 0.1 MeV) at 333–363 K. Changes in macroscopic lengths were examined by post-irradiation thermal annealing using a precision dilatometer up to 1673 K with a step-heating method. The specimen was held at each temperature step for 6 h and the change in length of the specimen was recorded during each isothermal annealing step from 373 K to 1673 K with 50 K increments. The recovery curves were analyzed with the first order model, and rate constants at each annealing step were obtained. Recovery of defects, induced by neutron irradiation in high purity β-SiC, has four stages of different activation energies. At 373–573 K, the activation energy of PureBeta-SiC and CVD-SiC was in the range of 0.17–0.24 eV and 0.12–0.14 eV; 0.002–0.04 eV and 0.006–0.04 eV at 723–923 K; 0.20–0.27 eV and 0.26–0.31 eV at 923–1223 K; and 1.37–1.38 eV and 1.26–1.29 eV at 1323–1523 K, respectively. Below ∼1223 K the recombination occurred possibly for closely positioned C and Si Frenkel pairs, and no long range migration is deemed essential. Nearly three-fourths of recovery, induced by neutron irradiation, occur by this mechanism. In addition, at 1323–1523 K, recombination of slightly separated C Frenkel pairs and more long-range migration of Si interstitials may have occurred for PureBeta-SiC and CVD-SiC specimens. Migration of both vacancies may be restricted up to ∼1523 K. Comparing to hexagonal α-SiC, high purity β-SiC recovered more quickly in the lower annealing temperature range of less than 873 K, in particular less than 573 K. - Highlights: • Two kinds of high purity cubic (β) SiC polycrystals were irradiated. • Macroscopic lengths were examined by post-irradiation thermal annealing. • The recovery curves were analyzed with first order model. �

Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

International Nuclear Information System (INIS)

Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

1995-01-01

This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

Survey report on high temperature irradiation experiment programs for new ceramic materials in the HTTR (High Temperature Engineering Test Reactor). 2

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-02-01

A survey research on status of research activities on new ceramic materials in Japan was carried out under contract between Japan Atomic Energy Research Institute and Atomic Energy Society of Japan. The purpose of the survey is to provide information to prioritize prospective experiments and tests in the HTTR. The HTTR as a high temperature gas cooled reactor has a unique and superior capability to irradiate large-volumed specimen at high temperature up to approximately 800degC. The survey was focused on mainly the activities of functional ceramics and heat resisting ceramics as a kind of structural ceramics. As the result, the report recommends that the irradiation experiment of functional ceramics is feasible to date. (K. Itami)

HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

Energy Technology Data Exchange (ETDEWEB)

Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.; Chrisensen, Cad L.

2016-11-01

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 Test 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

High temperature tensile testing of modified 9Cr-1Mo after irradiation with high energy protons

International Nuclear Information System (INIS)

Toloczko, M.B.; Hamilton, M.L.; Maloy, S.A.

2003-01-01

This study examines the effect of tensile test temperatures ranging from 50 to 600 deg. C on the tensile properties of a modified 9Cr-1Mo ferritic steel after high energy proton irradiation at about 35-67 deg. C to doses from 1 to 3 dpa and 9 dpa. For the specimens irradiated to doses between 1 and 3 dpa, it was observed that the yield strength and ultimate strength decreased monotonically as a function of tensile test temperature, whereas the uniform elongation (UE) remained at approximately 1% for tensile test temperatures up to 250 deg. C and then increased for tensile test temperatures up to and including 500 deg. C. At 600 deg. C, the UE was observed to be less than the values at 400 and 500 deg. C. UE of the irradiated material tensile tested at 400-600 deg. C was observed to be greater than the values for the unirradiated material at the same temperatures. Tensile tests on the 9 dpa specimens followed similar trends

Neutron and X-ray diffraction analysis of the effect of irradiation dose and temperature on microstructure of irradiated HT-9 steel

International Nuclear Information System (INIS)

Mosbrucker, P.L.; Brown, D.W.; Anderoglu, O.; Balogh, L.; Maloy, S.A.; Sisneros, T.A.; Almer, J.; Tulk, E.F.; Morgenroth, W.; Dippel, A.C.

2013-01-01

Material harvested from several positions within a nuclear fuel duct (the ACO-3 duct) used in a 6-year irradiation of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF) was examined using neutron and high-energy X-ray diffraction. Samples with a wide range of irradiation dose and irradiation temperature history, reaching doses of up to 147 dpa and temperatures of up to 777 K, were examined. The response of various microstructural characteristics such as the weight fraction of M 23 C 6 carbides, the dislocation density and character, and the crystallographic texture were determined using whole profile analysis of the diffraction data and related to the macroscopic mechanical behavior. For instance, the dislocation density was observed to be intimately linked with observed flow strength of the irradiated materials, following the Taylor law. In general, at the high doses studied in this work, the irradiation temperature is the predominant controlling factor of the dislocation density and, thus, the flow strength of the irradiated material. The results, representing some of the first diffraction work done on samples exposed to such a high received dose, demonstrate how non-destructive and stand-off diffraction techniques can be used to characterize irradiation induced microstructure and at least estimate mechanical properties in irradiated materials without exposing workers to radiation hazards

Thermal behavior of spatial structures under solar irradiation

International Nuclear Information System (INIS)

Liu, Hongbo; Liao, Xiangwei; Chen, Zhihua; Zhang, Qian

2015-01-01

The temperature, particularly the non-uniform temperature under solar irradiation, is the main load for large-span steel structures. Due the shortage of in-site temperature test in previous studies, an in-site test was conducted on the large-span steel structures under solar irradiation, which was covered by glass roof and light roof, to gain insight into the temperature distribution of steel members under glass roof or light roof. A numerical method also was presented and verified to forecast the temperature of steel member under glass roof or light roof. Based on the on-site measurement and numerical analyses conducted, the following conclusions were obtained: 1) a remarkable temperature difference exists between the steel member under glass roof and that under light roof, 2) solar irradiation has a significant effect on the temperature distribution and thermal behavior of large-span spatial structures, 3) negative thermal load is the controlling factor for member stress, and the positive thermal load is the controlling factor for nodal displacement. - Highlights: • Temperature was measured for a steel structures under glass roof and light roof. • Temperature simulation method was presented and verified. • The thermal behavior of steel structures under glass or light roof was presented

High burnup, high power irradiation behavior of helium-bonded mixed carbide fuel pins

International Nuclear Information System (INIS)

Levine, P.J.; Nayak, U.P.; Boltax, A.

1983-01-01

Large diameter (9.4 mm) helium-bonded mixed carbide fuel pins were successfully irradiated in EBR-II to high burnup (12%) at high power levels (100 kW/m) with peak cladding midwall temperatures of 550 0 C. The wire-wrapped pins were clad with 0.51-mm-thick, 20% cold-worked Type 316 stainless steel and contained hyperstoichiometric (Usub(0.8)Pusub(0.2))C fuel covering the smeared density range from 75-82% TD. Post-irradiation examinations revealed: extensive fuel-cladding mechanical interaction over the entire length of the fuel column, 35% fission gas release at 12% burnup, cladding carburization and fuel restructuring. (orig.)

Irradiation test HT-31: high-temperature irradiation behavior of LASL-made extruded fuel rods and LASL-made coated particles

International Nuclear Information System (INIS)

Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; Davidson, K.V.; Schell, D.H.

1977-04-01

Three LASL-made extruded graphite and coated particle fuel rods have been irradiated in the Oak Ridge National Laboratory High Fluence Isotope Reactor test HT-31. Test conditions were about 9 x 10 21 nvt(E > .18 MeV) at 1250 0 C. The graphite matrix showed little or no effect of the irradiation. LASL-made ZrC containing coated particles with ZrC coats and ZrC-doped pyrolytic carbon coats showed no observable effects of the irradiation

Behavioral response and cell morphology changes of caenorhabditis elegans under high power millimeter wave irradiation

International Nuclear Information System (INIS)

Ren Changhong; Gao Yan; Wu Yonghong; Xu Zhiwei; Zhang Chenggang; Yuan Guangjiang; Xu Shouxi; Su Yinong; Liu Pukun

2010-01-01

C. elegans were exposed to high power millimeter waves (MMWs) with different mean power densities, to investigate their behavioral response and cell morphology changes under MMW irradiation. The time-course photomicrography system was used to record the behavioral changes of C. elegans. The behavioral response and cell morphology changes were further observed by stereoscopic microscopes. The results show that freely moving C. elegans will escape from the MMW irradiation region quickly. After the exposure to MMWs with output mean power of 10 W and 12 W, the bending speed of C. elegans increases significantly at first, while the movement gradually slows down until the bodies get rigid. However, exposed to 5 W MMW, C. elegans show a distinctive tolerant reaction because of the thermal effect. In addition, cell morphological observations show that the nuclear structure of the eggs are abnormal after abnormal after MMW irradiation. High power MMW significantly affects the behaviors and cell morphology of C. elegans, which suggests the C. elegans could be used as a typical model species to study the biological effects of MMW irradiation. (authors)

Irradiation-induced creep in fuel compacts for high-temperature reactor applications

Energy Technology Data Exchange (ETDEWEB)

Veringa, H; Blackstone, R [Stichting Energieonderzoek Centrum Nederland, Petten; Loelgen, R

1977-01-01

Restrained shrinkage experiments at neutron fluences up to 3 x 10/sup 21/ n cm/sup -2/ DNE in the temperature range 600 to 1200/sup 0/C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material.

Irradiation-induced creep in fuel compacts for high-temperature reactor applications

International Nuclear Information System (INIS)

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

1977-01-01

Restrained shrinkage experiments at neutron fluences up to 3 x 10 21 n cm -2 DNE in the temperature range 600 to 1200 0 C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material. (author)

Temperature annealing of tracks induced by ion irradiation of graphite

International Nuclear Information System (INIS)

Liu, J.; Yao, H.J.; Sun, Y.M.; Duan, J.L.; Hou, M.D.; Mo, D.; Wang, Z.G.; Jin, Y.F.; Abe, H.; Li, Z.C.; Sekimura, N.

2006-01-01

Highly oriented pyrolytic graphite (HOPG) samples were irradiated by Xe ions of initial kinetic energy of 3 MeV/u. The irradiations were performed at temperatures of 500 and 800 K. Scanning tunneling microscopy (STM) images show that the tracks occasionally have elongated structures under high-temperature irradiation. The track creation yield at 800 K is by three orders of magnitude smaller compared to that obtained during room-temperature irradiation. STM and Raman spectra show that amorphization occurs in graphite samples irradiated at 500 K to higher fluences, but not at 800 K. The obtained experimental results clearly reveal that the irradiation under high temperature causes track annealing

Effect of heat treatment and irradiation temperature on impact behavior of irradiated reduced-activation ferritic steels

International Nuclear Information System (INIS)

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

1998-01-01

Charpy tests were conducted on eight normalized-and-tempered reduced-activation ferritic steels irradiated in two different normalized conditions. Irradiation was conducted in the Fast Flux Test Facility at 393 C to ∼14 dpa on steels with 2.25, 5, 9, and 12% Cr (0.1% C) with varying amounts of W, V, and Ta. The different normalization treatments involved changing the cooling rate after austenitization. The faster cooling rate produced 100% bainite in the 2.25 Cr steels, compared to duplex structures of bainite and polygonal ferrite for the slower cooling rate. For both cooling rates, martensite formed in the 5 and 9% Cr steels, and martensite with ∼25% δ-ferrite formed in the 12% Cr steel. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The difference in microstructure in the low-chromium steels due to the different heat treatments had little effect on properties. For the high-chromium martensitic steels, only the 5 Cr steel was affected by heat treatment. When the results at 393 C were compared with previous results at 365 C, all but a 5 Cr and a 9 Cr steel showed the expected decrease in the shift in DBTT with increasing temperature

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Damage behavior in helium-irradiated reduced-activation martensitic steels at elevated temperatures

Energy Technology Data Exchange (ETDEWEB)

Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Chen, Jihong; Li, Tiecheng; Zheng, Zhongcheng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON (Canada); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

2014-12-15

Dislocation loops induced by helium irradiation at elevated temperatures in reduced-activation martensitic steels were investigated using transmission electron microscopy. Steels were irradiated with 100 keV helium ions to 0.8 dpa between 300 K and 723 K. At irradiation temperatures T{sub irr} ⩽ 573 K, small defects with both Burger vectors b = 1/2〈1 1 1〉 and b = 〈1 0 0〉 were observed, while at T{sub irr} ⩾ 623 K, the microstructure was dominated by large convoluted interstitial dislocation loops with b = 〈1 0 0〉. Only small cavities were found in the steels irradiated at 723 K.

Irradiation damage behavior of low alloy steel wrought and weld materials

International Nuclear Information System (INIS)

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

1993-01-01

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

High temperature oxidation behavior of ODS steels

Science.gov (United States)

Kaito, T.; Narita, T.; Ukai, S.; Matsuda, Y.

2004-08-01

Oxide dispersion strengthened (ODS) steels are being developing for application as advanced fast reactor cladding and fusion blanket materials, in order to allow increased operation temperature. Oxidation testing of ODS steel was conducted under a controlled dry air atmosphere to evaluate the high temperature oxidation behavior. This showed that 9Cr-ODS martensitic steels and 12Cr-ODS ferritic steels have superior high temperature oxidation resistance compared to 11 mass% Cr PNC-FMS and 17 mass% Cr ferritic stainless steel. This high temperature resistance is attributed to earlier formation of the protective α-Cr 2O 3 on the outer surface of ODS steels.

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

Science.gov (United States)

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

2017-02-01

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

Strain acceleration of the low temperature irradiated zirconium

International Nuclear Information System (INIS)

Fortis, Ana M.; Coccoz, Guillermina D. H.

2003-01-01

The strain of a Zr-0,06 at.% 235 U specimen irradiated during 4800 h in the RA-3 at a temperature near 40 C degrees is presented. An equivalent neutron fluence of 3.1 x 10 26 n m -2 was achieved by means of the generation of fission fragment within the material. The experimental conditions are described and a sudden strain acceleration independent of the neutron flux variations occurred during irradiation is shown. This behavior is compared with previous data obtained at different temperatures. (author)

High-Temperature Corrosion Behavior of Alloy 617 in Helium Environment of Very High Temperature Gas Reactor

International Nuclear Information System (INIS)

Lee, Gyeong-Geun; Jung, Sujin; Kim, Daejong; Jeong, Yong-Whan; Kim, Dong-Jin

2012-01-01

Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at 850°C-950°C in a helium environment containing the impurity gases H_2, CO, and CH_4, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high temperature corrosion behavior of Alloy 617 for the VHTR application.

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

Science.gov (United States)

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

2017-09-01

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

A new disordering mechanism in A15 type compounds submitted to low temperature irradiation or to quenching from high temperatures

International Nuclear Information System (INIS)

Fluekiger, R.

1984-05-01

A new diffusion mechanism describing the changes of the long range order parameter in A15 type compounds after both quenching from high temperatures or low temperature irradiation with high energy particles is presented. It is based on the occupation of nonequilibrium or 'virtual' sites centered halfway between two neighbouring A atoms on 6c sites, arising from the instability of a single 6c vacancy recently found by Welch and coworkers by pair potential calculations. After low temperature irradiation, the occupation of this interstitial site creates the necessary conditions for A B site exchanges over several interatomic distances by focused replacement collision sequences. Due to the occupation of a certain concentration of virtual sites, atomic 'overlapping' is not only possible between A atoms on the chains or between A and B atoms (due to deviations from perfect ordering),but also between B atoms on BBB sequences. The latter are retained after low temperature irradiation only and are responsible for the observed lattice expansion and static displacement. (orig.) [de

Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles

International Nuclear Information System (INIS)

Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

1977-10-01

Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10 21 nvt (E greater than 0.18 MeV) at 1350 0 C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particles with inert kernels showed no irradiation effects. The graded ZrC-C coats on the fissile particles were cracked. It is postulated that the cracking is associated with the low LTI deposition rate and is not related to the ZrC

The High Temperature Tensile and Creep Behaviors of High Entropy Superalloy.

Science.gov (United States)

Tsao, Te-Kang; Yeh, An-Chou; Kuo, Chen-Ming; Kakehi, Koji; Murakami, Hideyuki; Yeh, Jien-Wei; Jian, Sheng-Rui

2017-10-04

This article presents the high temperature tensile and creep behaviors of a novel high entropy alloy (HEA). The microstructure of this HEA resembles that of advanced superalloys with a high entropy FCC matrix and L1 2 ordered precipitates, so it is also named as "high entropy superalloy (HESA)". The tensile yield strengths of HESA surpass those of the reported HEAs from room temperature to elevated temperatures; furthermore, its creep resistance at 982 °C can be compared to those of some Ni-based superalloys. Analysis on experimental results indicate that HESA could be strengthened by the low stacking-fault energy of the matrix, high anti-phase boundary energy of the strengthening precipitate, and thermally stable microstructure. Positive misfit between FCC matrix and precipitate has yielded parallel raft microstructure during creep at 982 °C, and the creep curves of HESA were dominated by tertiary creep behavior. To the best of authors' knowledge, this article is the first to present the elevated temperature tensile creep study on full scale specimens of a high entropy alloy, and the potential of HESA for high temperature structural application is discussed.

Creep behavior of materials for high-temperature reactor application

International Nuclear Information System (INIS)

Schneider, K.; Hartnagel, W.; Iischner, B.; Schepp, P.

1984-01-01

Materials for high-temperature gas-cooled reactor (HTGR) application are selected according to their creep behavior. For two alloys--Incoloy-800 used for the live steam tubing of the thorium high-temperature reactor and Inconel-617 evaluated for tubings in advanced HTGRs--creep curves are measured and described by equations. A microstructural interpretation is given. An essential result is that nonstable microstructures determine the creep behavior

A Study on the High Temperature Irradiation Test Possibility for the HANARO Outer Core Region

Energy Technology Data Exchange (ETDEWEB)

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

2008-01-15

1. Information on the neutron flux levels and the gamma heat of the concerned test holes, which have been produced from a series of nuclear analysis and tests performed at KAERI since 1993, were collected and analyzed to develop the nuclear data for the concerned test holes of HANARO and to develop the new design concepts of a capsule for the high temperature irradiation devices. 2. From the literature survey and analysis about the system design characteristics of the new concepts of irradiation devices in the ATR and MIT reactor, U.S. and the JHR reactor, France, which are helpful in understanding the key issues for the on-going R and D programmes related to a SFR and a VHTR, the most important parameters for the design of high temperature irradiation devices are identified as the neutron spectrum, the heat generation density, the fuel and cladding temperature, and the coolant chemistry. 3. From the thermal analysis of a capsule by using a finite element program ANSYS, high temperature test possibility at the OR and IP holes of HANARO was investigated based on the data collected from a literature survey. The OR holes are recommended for the tests of the SFR and VHTR nuclear materials. The IP holes could be applicable for an intermediate temperature irradiation of the SWR and LMR materials. 4. A thermal analysis for the development of a capsule with a new configuration was also performed. The size of the center hole, which is located at the thermal media of a capsule, did not cause specimen temperature changes. The temperature differences are found to be less than 2%. The introduction of an additional gap in the thermal media was able to contribute to an increase in the specimen temperature by up to 27-90 %.

Irradiation temperature dependence of production efficiency of lattice defects in some neutron-irradiated oxides

International Nuclear Information System (INIS)

Okada, Moritami; Atobe, Kozo; Nakagawa, Masuo

2004-01-01

Temperature dependence of production efficiency of irradiation-induced defects in neutron-irradiated oxides has been investigated. Some oxide single crystals, MgO, α-Al 2 O 3 (sapphire) and TiO 2 (rutile), were irradiated at several controlled temperatures, 10, 20, 50, 100, 150 and 200 K, using the low-temperature irradiation facility of Kyoto University Reactor (KUR-LTL), and at ambient temperature (∼370 K) in the same facility. Irradiation temperature dependence of production efficiency of a 1 μm band in TiO 2 differs greatly from that of anion vacancy (F-type centers) in MgO and α-Al 2 O 3 . Results for MgO and α-Al 2 O 3 show steep negative gradients from 10 to 370 K, whereas that for TiO 2 includes a valley between 40 and 60 K and a hump at about 130 K, and then disappear at about 200 K. In MgO and α-Al 2 O 3 , this behavior can be explained by the recombination of Frenkel pairs, which is activated at higher temperature. In TiO 2 , in addition to the recombination mechanism, a covalent bonding property is thought to be exerted strong influence, and it is suggested that a disappearance of the 1 μm band at above 200 K is due to the recombination process of Frenkel pairs which is caused by the irradiation-induced crystallization

Irradiation temperature dependence of production efficiency of lattice defects in some neutron-irradiated oxides

Energy Technology Data Exchange (ETDEWEB)

Okada, Moritami [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 5900494 (Japan)]. E-mail: okada@rri.kyoto-u.ac.jp; Atobe, Kozo [Faculty of Science, Naruto University of Education, Naruto, Tokushima 7728502 (Japan); Nakagawa, Masuo [Faculty of Education, Kagawa University, Takamatsu, Kagawa 7608522 (Japan)

2004-11-01

Temperature dependence of production efficiency of irradiation-induced defects in neutron-irradiated oxides has been investigated. Some oxide single crystals, MgO, {alpha}-Al{sub 2}O{sub 3} (sapphire) and TiO{sub 2} (rutile), were irradiated at several controlled temperatures, 10, 20, 50, 100, 150 and 200 K, using the low-temperature irradiation facility of Kyoto University Reactor (KUR-LTL), and at ambient temperature ({approx}370 K) in the same facility. Irradiation temperature dependence of production efficiency of a 1 {mu}m band in TiO{sub 2} differs greatly from that of anion vacancy (F-type centers) in MgO and {alpha}-Al{sub 2}O{sub 3}. Results for MgO and {alpha}-Al{sub 2}O{sub 3} show steep negative gradients from 10 to 370 K, whereas that for TiO{sub 2} includes a valley between 40 and 60 K and a hump at about 130 K, and then disappear at about 200 K. In MgO and {alpha}-Al{sub 2}O{sub 3}, this behavior can be explained by the recombination of Frenkel pairs, which is activated at higher temperature. In TiO{sub 2}, in addition to the recombination mechanism, a covalent bonding property is thought to be exerted strong influence, and it is suggested that a disappearance of the 1 {mu}m band at above 200 K is due to the recombination process of Frenkel pairs which is caused by the irradiation-induced crystallization.

Modeling high temperature materials behavior for structural analysis

CERN Document Server

Naumenko, Konstantin

2016-01-01

This monograph presents approaches to characterize inelastic behavior of materials and structures at high temperature. Starting from experimental observations, it discusses basic features of inelastic phenomena including creep, plasticity, relaxation, low cycle and thermal fatigue. The authors formulate constitutive equations to describe the inelastic response for the given states of stress and microstructure. They introduce evolution equations to capture hardening, recovery, softening, ageing and damage processes. Principles of continuum mechanics and thermodynamics are presented to provide a framework for the modeling materials behavior with the aim of structural analysis of high-temperature engineering components.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Post irradiation characterization of beryllium and beryllides after high temperature irradiation up to 3000 appm helium production in HIDOBE-01

Energy Technology Data Exchange (ETDEWEB)

Fedorov, A.V., E-mail: fedorov@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Til, S. van; Stijkel, M.P. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Nakamichi, M. [Japan Atomic Energy Agency, Rokkasho (Japan); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/ Josep Pla, n° 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

2016-01-15

Titanium beryllides are considered as advanced candidate material for neutron multiplier for the helium cooled pebble bed (HCPB) and/or the water cooled ceramic breeder (WCCB) breeder blankets. In the HIDOBE-01 (HIgh DOse irradiation of BEryllium) experiment, beryllium and beryllide pellets with 5 at% and 7 at% Ti are irradiated at four different target temperatures (T{sub irr}): 425 °C, 525 °C, 650 °C and 750 °C up to the dose corresponding to 3000 appm He production in beryllium. The pellets were supplied by JAEA. During post irradiation examinations the critical properties of volumetric swelling and tritium retention were studied. Both titanium beryllide grades show significantly less swelling than the beryllium grade, with the difference increasing with the irradiation temperature. The irradiation induced swelling was studied by using direct dimensions. Both beryllide grades showed much less swelling as compare to the reference beryllium grade. Densities of the grades were studied by Archimedean immersion and by He-pycnometry, giving indications of porosity formation. While both beryllide grades show no significant reduction in density at all irradiation temperatures, the beryllium density falls steeply at higher T{sub irr}. Finally, the tritium release and retention were studied by temperature programmed desorption (TPD). Beryllium shows the same strong tritium retention as earlier observed in studies on beryllium pebbles, while the tritium inventory of the beryllides is significantly less, already at the lowest T{sub irr} of 425 °C.

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

Energy Technology Data Exchange (ETDEWEB)

Laurie, M.; Vlahovic, L.; Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe, (Germany); Sadli, M.; Failleau, G. [Laboratoire Commun de Metrologie, LNE-Cnam, Saint-Denis, (France); Fuetterer, M.; Lapetite, J.M. [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, (Netherlands); Fourrez, S. [Thermocoax, 8 rue du pre neuf, F-61100 St Georges des Groseillers, (France)

2015-07-01

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 high 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)

Parametric study of irradiation effects on the ductile damage and flow stress behavior in ferritic-martensitic steels

International Nuclear Information System (INIS)

Chakraborty, Pritam; Biner, S.Bulent

2015-01-01

Ferritic-martensitic steels are currently being considered as structural materials in fusion and Gen-IV nuclear reactors. These materials are expected to experience high dose radiation, which can increase their ductile to brittle transition temperature and susceptibility to failure during operation. Hence, to estimate the safe operational life of the reactors, precise evaluation of the ductile to brittle transition temperatures of ferritic-martensitic steels is necessary. Owing to the scarcity of irradiated samples, particularly at high dose levels, micro-mechanistic models are being employed to predict the shifts in the ductile to brittle transition temperatures. These models consider the ductile damage evolution, in the form of nucleation, growth and coalescence of voids; and the brittle fracture, in the form of probabilistic cleavage initiation, to estimate the influence of irradiation on the ductile to brittle transition temperature. However, the assessment of irradiation dependent material parameters is challenging and influences the accuracy of these models. In the present study, the effects of irradiation on the overall flow stress and ductile damage behavior of two ferritic-martensitic steels is parametrically investigated. The results indicate that the ductile damage model parameters are mostly insensitive to irradiation levels at higher dose levels though the resulting flow stress behavior varies significantly.

Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

Energy Technology Data Exchange (ETDEWEB)

Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

2014-01-15

Highlights: • Environmental constant extension rate tensile tests were performed on neutron irradiated steel. • Percentage of intergranular cracking quantified the cracking susceptibility. • Cracking susceptibility varied with test environment, solute addition, and cold work. • No singular microstructural change could explain increases in cracking susceptibility with irradiation dose. • The increment of yield strength due to irradiation correlated well with cracking susceptibility. -- Abstract: The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320 °C to doses between 4.4 and 47.5 dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni + Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

Study on structural recovery of graphite irradiated with swift heavy ions at high temperature

Energy Technology Data Exchange (ETDEWEB)

Pellemoine, F., E-mail: pellemoi@frib.msu.edu [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Avilov, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Bender, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Ewing, R.C. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Fernandes, S. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Lang, M. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996-2300 (United States); Li, W.X. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Mittig, W. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Schein, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Severin, D. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Tomut, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Laboratory of Magnetism and Superconductivity, National Institute for Materials Physics NIMP, Bucharest (Romania); Trautmann, C. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Dept. of Materials Science, Technische Universität Darmstadt, Darmstadt (Germany); and others

2015-12-15

Thin graphite foils bombarded with an intense high-energy (8.6 MeV/u) gold beam reaching fluences up to 1 × 10{sup 15} ions/cm{sup 2} lead to swelling and electrical resistivity changes. As shown earlier, these effects are diminished with increasing irradiation temperature. The work reported here extends the investigation of beam induced changes of these samples by structural analysis using synchrotron X-ray diffraction and transmission electron microscope. A nearly complete recovery from swelling at irradiation temperatures above about 1500 °C is identified.

Estimation of irradiation temperature within the irradiation program Rheinsberg

CERN Document Server

Stephan, I; Prokert, F; Scholz, A

2003-01-01

The temperature monitoring within the irradiation programme Rheinsberg II was performed by diamond powder monitors. The method bases on the effect of temperature on the irradiation-induced increase of the diamond lattice constant. The method is described by a Russian code. In order to determine the irradiation temperature, the lattice constant is measured by means of a X-ray diffractometer after irradiation and subsequent isochronic annealing. The kink of the linearized temperature-lattice constant curves provides a value for the irradiation temperature. It has to be corrected according to the local neutron flux. The results of the lattice constant measurements show strong scatter. Furthermore there is a systematic error. The results of temperature monitoring by diamond powder are not satisfying. The most probable value lays within 255 C and 265 C and is near the value estimated from the thermal condition of the irradiation experiments.

Evaluation of the Centerline Temperature for the Irradiated DUPIC Pellet

International Nuclear Information System (INIS)

Park, Chang Je; Lee, Cheol Yong; Kang, Kweon Ho; Song, Kee Chan

2007-01-01

The DUPIC (Direct Use of spent PWR fuels In a CANDU reactor) fuel has a proliferation-resistant property and provides an efficient utilization of a spent fuel through a direct fabrication with the OREOX process in which most of the fission products remain and some volatile elements such as Xe, Kr, Cs, and I are reduced significantly. It is expected that the performance of the DUPIC fuel exhibits different behavior when compared with the fresh uranium oxide fuel. To evaluate the performance of the DUPIC fuel, total five irradiation tests have been performed in the HANARO reactor since May 2000. Recently, the fifth irradiation test of the DUPIC fuel was successfully completed for a total of three cycles from March 2006 to July 2006. The important characteristics of the first irradiation test are a high power test and a validation of a remote assembly of an irradiation rig. The second irradiation test was instrumented with a SPND (self-powered neutron detector) first for a typical CANDU burnup test. The third test was an extensive irradiation test of the second test and the total burnup was estimated as 6,700 MWd/tU. The forth test was a remote instrumented test of the pellet centerline temperature and the inlet and outlet coolant temperatures. The first remote instrumentation test was achieved with our own technology. The fifth test was a remote-instrumented test of the pellet centerline temperature by extending the technology of the forth irradiation test. In this paper, a DUPIC fuel performance code (KAOS, KAERI Advanced Oxide fuel performance code System) was used to compare the main simulation results of the irradiation tests in the HANARO

Determination of irradiation temperature using SiC temperature monitors

International Nuclear Information System (INIS)

Maruyama, Tadashi; Onose, Shoji

1999-01-01

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

Analytical and numerical study of graphite IG110 parts in advanced reactor under high temperature and irradiation

Energy Technology Data Exchange (ETDEWEB)

Gao, Jinling, E-mail: Jinling_Gao@yeah.net; Yao, Wenjuan, E-mail: wj_yao@yeah.net; Ma, Yudong

2016-08-15

Graphical abstract: An analytical model and a numerical procedure are developed to study the mechanical response of IG-110 graphite bricks in HTGR subjected to high temperature and irradiation. The calculation results show great accordance with each other. Rational suggestions on the calculation and design of the IG-110 graphite structure are proposed based on the sensitivity analyses including temperature, irradiation dimensional change, creep and Poisson’s ratio. - Highlights: • Analytical solution of stress and displacement of IG-110 graphite components in HTGR. • Finite element procedure developed for stress analysis of HTGR graphite component. • Parameters analysis of mechanical response of graphite components during the whole life of the reflector. - Abstract: Structural design of nuclear power plant project is an important sub-discipline of civil engineering. Especially after appearance of the fourth generation advanced high temperature gas cooled reactor, structural mechanics in reactor technology becomes a popular subject in structural engineering. As basic ingredients of reflector in reactor, graphite bricks are subjected to high temperature and irradiation and the stress field of graphite structures determines integrity of reflector and makes a great difference to safety of whole structure. In this paper, based on assumptions of elasticity, side reflector is regarded approximately as a straight cylinder structure and primary creep strain is ignored. An analytical study on stress of IG110 graphite parts is present. Meanwhile, a finite element procedure for calculating stresses in the IG110 graphite structure exposed in the high temperature and irradiation is developed. Subsequently, numerical solution of stress in IG110 graphite structure is obtained. Analytical solution agrees well with numerical solution, which indicates that analytical derivation is accurate. Finally, influence of temperature, irradiation dimensional change, creep and Poisson�

High temperature oxidation behavior of TiAl-based intermetallics

International Nuclear Information System (INIS)

Stroosnijder, M.F.; Sunderkoetter, J.D.; Haanappel, V.A.C.

1996-01-01

TiAl-based intermetallic compounds have attracted considerable interest as structural materials for high-temperature applications due to their low density and substantial mechanical strength at high temperatures. However, one major drawback hindering industrial application arises from the insufficient oxidation resistance at temperatures beyond 700 C. In the present contribution some general aspects of high temperature oxidation of TiAl-based intermetallics will be presented. This will be followed by a discussion of the influence of alloying elements, in particular niobium, and of the effect of nitrogen in the oxidizing environment on the high temperature oxidation behavior of such materials

Heat-to-heat variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400-600{degrees}C

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

Irradiation creep data on ferritic/martensitic steels are difficult and expensive to obtain, and are not available for fusion-relevant neutron spectra and displacement rates. Therefore, an extensive creep data rescue and analysis effort is in progress to characterize irradiation creep of ferritic/martensitic alloys in other reactors and to develop a methodology for applying it to fusion applications. In the current study, four tube sets constructed from three nominally similar heats of HT9 subjected to one of two heat treatments were constructed as helium-pressurized creep tubes and irradiated in FFTF-MOTA at four temperatures between 400 and 600{degrees}C. Each of the four heats exhibited a different stress-free swelling behavior at 400{degrees}C, with the creep rate following the swelling according to the familiar B{sub o} + DS creep law. No stress-free swelling was observed at the other three irradiation temperatures. Using a stress exponent of n = 1.0 as the defining criterion, {open_quotes}classic{close_quotes} irradiation creep was found at all temperatures, but, only over limited stress ranges that decreased with increasing temperature. The creep coefficient B{sub o} is a little lower ({approx}50%) than that observed for austenitic steel, but the swelling-creep coupling coefficient D is comparable to that of austenitic steels. Primary transient creep behavior was also observed at all temperatures except 400{degrees}C, and thermal creep behavior was found to dominate the deformation at high stress levels at 550 and 600{degrees}C.

An explanation of the irreversibility behavior in the highly- anisotropic high-temperature superconductors

International Nuclear Information System (INIS)

Gray, K.E.; Kim, D.H.

1991-01-01

The wide temperature range of the reversible, lossy state of the new high-temperature superconductors in a magnetic field was recognized soon after their discovery. This behavior, which had gone virtually undetected in conventional superconductors, has generated considerable interest, both for a fundamental understanding of the HTS and because it degrades the performance of HTS for finite-field applications. We show that recently proposed explanation of this behavior for the highly-anisotropic high-temperature superconductors, as a dimensional crossover of the magnetic vortices, is strongly supported by recent experiments on a Bi 2 Sr 2 CaCu 2 O x single crystal using the high-Q mechanical oscillator techniques

Tritium release from beryllium pebbles after high temperature irradiation up to 3000 appm He in the HIDOBE-01 experiment

Energy Technology Data Exchange (ETDEWEB)

Til, S. van, E-mail: vantil@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Fedorov, A.V.; Stijkel, M.P.; Cobussen, H.L.; Mutnuru, R.K.; Idsert, P. van der [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and The Development of Fusion Energy, c/ Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

2013-11-15

In the HIDOBE (HIgh DOse irradiation of BEryllium) irradiation program, various grades of constrained and unconstrained beryllium pebbles, beryllium pellets and titanium-beryllide samples are irradiated in the High Flux Reactor (HFR) in Petten at four different temperatures (between 698 K and 1023 K) for 649 days [1]. The first of two HIDOBE irradiation experiments, HIDOBE-01, was completed after achieving a DEMO relevant helium production level of 3000 appm and the samples are retrieved for postirradiation examination (PIE). This work shows preliminary results of the out-of-pile tritium release analysis performed on different grades of irradiated beryllium pebbles (different in size). Relationships between irradiation temperature, tritium inventory and microstructural evolution have been observed by light microscopy and scanning electron microscopy.

Microstructural Evolution and Mechanical Behavior of High Temperature Solders: Effects of High Temperature Aging

Science.gov (United States)

Hasnine, M.; Tolla, B.; Vahora, N.

2018-04-01

This paper explores the effects of aging on the mechanical behavior, microstructure evolution and IMC formation on different surface finishes of two high temperature solders, Sn-5 wt.% Ag and Sn-5 wt.% Sb. High temperature aging showed significant degradation of Sn-5 wt.% Ag solder hardness (34%) while aging has little effect on Sn-5 wt.% Sb solder. Sn-5 wt.% Ag experienced rapid grain growth as well as the coarsening of particles during aging. Sn-5 wt.% Sb showed a stable microstructure due to solid solution strengthening and the stable nature of SnSb precipitates. The increase of intermetallic compound (IMC) thickness during aging follows a parabolic relationship with time. Regression analysis (time exponent, n) indicated that IMC growth kinetics is controlled by a diffusion mechanism. The results have important implications in the selection of high temperature solders used in high temperature applications.

A technique of melting temperature measurement and its application for irradiated high-burnup MOX fuels

International Nuclear Information System (INIS)

Namekawa, Takashi; Hirosawa, Takashi

1999-01-01

A melting temperature measurement technique for irradiated oxide fuels is described. In this technique, the melting temperature was determined from a thermal arrest on a heating curve of the specimen which was enclosed in a tungsten capsule to maintain constant chemical composition of the specimen during measurement. The measurement apparatus was installed in an alpha-tight steel box within a gamma-shielding cell and operated by remote handling. The temperature of the specimen was measured with a two-color pyrometer sighted on a black-body well at the bottom of the tungsten capsule. The diameter of the black-body well was optimized so that the uncertainties of measurement were reduced. To calibrate the measured temperature, two reference melting temperature materials, tantalum and molybdenum, were encapsulated and run before and after every oxide fuel test. The melting temperature data on fast reactor mixed oxide fuels irradiated up to 124 GWd/t were obtained. In addition, simulated high-burnup mixed oxide fuel up to 250 GWd/t by adding non-radioactive soluble fission products was examined. These data shows that the melting temperature decrease with increasing burnup and saturated at high burnup region. (author)

Behavior of implanted hydrogen in ferritic/martensitic steels under irradiation

Science.gov (United States)

Wan, F.; Takahashi, H.; Ohnuki, S.; Nagasaki, R.

1988-07-01

The aim of this study was to clarify the behavior of hydrogen under irradiation in ferritic/martensitic stainless steel Fe-10Cr-2Mo-1Ni. Hydrogen was implanted into the specimens by ion accelerator or chemical cathodic charging method, followed by electron irradiation in a HVEM at temperatures from room temperature to 773 K. Streaks in the electron diffraction patterns were observed only during electron irradiation at 623-723 K. From these results it is suggested that the occurrence of the streak pattern is due to the formation of radiation-induced complexes of Ni or Cr with hydrogen along directions.

Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

Science.gov (United States)

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 temperature measurements in the surveillance program

International Nuclear Information System (INIS)

Pav, T.; Krhounek, V.

1991-01-01

Evaluation of the diamond monitor method for the determination of the irradiation temperature in the surveillance programme of WWER-440 reactors is discussed. One of the difficulties with the practical application of the method is that the measured values of irradiation temperature are unlikely high. Using a thermodynamical model of the processes in the annealing of the irradiated diamond crystals, it was shown that experimental difficulties came from the principles of the method used. An analysis was performed of the thermal field inside the capsule of the surveillance chain in operational conditions, using the finite element method. The diamond monitor method was suggested to be eliminated from the surveillance programme and the use was proposed of the value of 273+-3 degC (as the most likely value) for the irradiation temperature of surveillance samples in WWER-440 reactors. (Z.S.). 3 tabs., 6 figs., 4 refs

High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

Energy Technology Data Exchange (ETDEWEB)

Was, Gary; Wirth, Brian; Motta, Athur; Morgan, Dane; Kaoumi, Djamel; Hosemann, Peter; Odette, Robert

2018-04-30

Project Objective: 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 irradiated 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

Effect of ion irradiation on tensile ductility, strength and fictive temperature in metallic glass nanowires

International Nuclear Information System (INIS)

Magagnosc, D.J.; Kumar, G.; Schroers, J.; Felfer, P.; Cairney, J.M.; Gianola, D.S.

2014-01-01

Ion irradiation of thermoplastically molded Pt 57.5 Cu 14.3 Ni 5.7 P 22.5 metallic glass nanowires is used to study the relationship between glass structure and tensile behavior across a wide range of structural states. Starting with the as-molded state of the glass, ion fluence and irradiated volume fraction are systematically varied to rejuvenate the glass, and the resulting plastic behavior of the metallic glass nanowires probed by in situ mechanical testing in a scanning electron microscope. Whereas the as-molded nanowires exhibit high strength, brittle-like fracture and negligible inelastic deformation, ion-irradiated nanowires show tensile ductility and quasi-homogeneous plastic deformation. Signatures of changes to the glass structure owing to ion irradiation as obtained from electron diffraction are subtle, despite relatively large yield strength reductions of hundreds of megapascals relative to the as-molded condition. To reconcile changes in mechanical behavior with glass properties, we adapt previous models equating the released strain energy during shear banding to a transit through the glass transition temperature by incorporating the excess enthalpy associated with distinct structural states. Our model suggests that ion irradiation increases the fictive temperature of our glass by tens of degrees – the equivalent of many orders of magnitude change in cooling rate. We further show our analytical description of yield strength to quantitatively describe literature results showing a correlation between severe plastic deformation and hardness in a single glass system. Our results highlight not only the capacity for room temperature ductile plastic flow in nanoscaled metallic glasses, but also processing strategies capable of glass rejuvenation outside of the realm of traditional thermal treatments

Elemental redistribution behavior in tellurite glass induced by high repetition rate femtosecond laser irradiation

International Nuclear Information System (INIS)

Teng, Yu; Zhou, Jiajia; Khisro, Said Nasir; Zhou, Shifeng; Qiu, Jianrong

2014-01-01

Highlights: • Abnormal elements redistribution behavior was observed in tellurite glass. • The refractive index and Raman intensity distribution changed significantly. • The relative glass composition remained unchanged while the glass density changed. • First time report on the abnormal element redistribution behavior in glass. • The glass network structure determines the elemental redistribution behavior. - Abstract: The success in the fabrication of micro-structures in glassy materials using femtosecond laser irradiation has proved its potential applications in the construction of three-dimensional micro-optical components or devices. In this paper, we report the elemental redistribution behavior in tellurite glass after the irradiation of high repetition rate femtosecond laser pulses. The relative glass composition remained unchanged while the glass density changed significantly, which is quite different from previously reported results about the high repetition rate femtosecond laser induced elemental redistribution in silicate glasses. The involved mechanism is discussed with the conclusion that the glass network structure plays the key role to determine the elemental redistribution. This observation not only helps to understand the interaction process of femtosecond laser with glassy materials, but also has potential applications in the fabrication of micro-optical devices

Temperature-dependent surface porosity of Nb{sub 2}O{sub 5} under high-flux, low-energy He{sup +} ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Novakowski, T.J., E-mail: tnovakow@purdue.edu; Tripathi, J.K.; Hosinski, G.M.; Joseph, G.; Hassanein, A.

2016-01-30

Graphical abstract: - Highlights: • Nb{sub 2}O{sub 5} surfaces are nanostructured with a novel He{sup +} ion irradiation process. • High-flux, low energy He{sup +} ion irradiation generates highly porous surfaces. • Top-down approach guarantees good contact between different crystallites. • Sample annealing demonstrates temperature effect on surface morphology. • Surface pore diameter increases with increasing temperature. - Abstract: The present study reports on high-flux, low-energy He{sup +} ion irradiation as a novel method of enhancing the surface porosity and surface area of naturally oxidized niobium (Nb). Our study shows that ion-irradiation-induced Nb surface micro- and nano-structures are highly tunable by varying the target temperature during ion bombardment. Mirror-polished Nb samples were irradiated with 100 eV He{sup +} ions at a flux of 1.2 × 10{sup 21} ions m{sup −2} s{sup −1} to a total fluence of 4.3 × 10{sup 24} ions m{sup −2} with simultaneous sample annealing in the temperature range of 773–1223 K to demonstrate the influence of sample temperature on the resulting Nb surface morphology. This surface morphology was primarily characterized using field-emission scanning electron microscopy (FE-SEM) and atomic force microscopy (AFM). Below 923 K, Nb surfaces form nano-scale tendrils and exhibit significant increases in surface porosity. Above 923 K, homogeneously populated nano-pores with an average diameter of ∼60 nm are observed in addition to a smaller population of sub-micron sized pores (up to ∼230 nm in diameter). Our analysis shows a significant reduction in surface pore number density and surface porosity with increasing sample temperature. High-resolution ex situ X-ray photoelectron spectroscopy (XPS) shows Nb{sub 2}O{sub 5} phase in all of the ion-irradiated samples. To further demonstrate the length scales in which radiation-induced surface roughening occurs, optical reflectivity was performed over a spectrum of

Effects of High-Energy Proton-Beam Irradiation on the Magnetic Properties of ZnO Nanorods

Energy Technology Data Exchange (ETDEWEB)

Park, Jun Kue; Kwon, Hyeok-Jung; Cho, Yong Sub [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

There are still many problem for the application due to its unstable magnetism state and too small magnetization values. Here we investigate magnetic properties of ZnO nanorods after high-energy proton-beam irradiation. Electron spin resonance (ESR) measurement on temperature was made to identify intrinsic or extrinsic defects as well as to observe magnetic ordering after irradiation. Understanding the effects of proton beam irradiation on magnetic behavior may help to shed light on the mechanism responsible for the magnetic ordering in this material. We have investigated proton-beam irradiation effects on the magnetic properties of ZnO nanorods. After irradiation a broad ESR line is observed, indicating emergence of ferromagnetic ordering up to room temperature. In M-H curve, stronger coercive field is observed after irradiation.

Behaviors of SiC fibers at high temperature

International Nuclear Information System (INIS)

Colin, C.; Falanga, V.; Gelebart, L.

2010-01-01

On the one hand, considering the improvements of mechanical and thermal behaviours of the last generation of SiC fibers (Hi-Nicalon S, Tyranno SA3); on the other hand, regarding physical and chemical properties and stability under irradiation, SiC/SiC composites are potential candidates for nuclear applications in advanced fission and fusion reactors. CEA must characterize and optimize these composites before their uses in reactors. In order to study this material, CEA is developing a multi-scale approach by modelling from fibers to bulk composite specimen: fibres behaviours must be well known in first. Thus, CEA developed a specific tensile test device on single fibers at high temperature, named MecaSiC. Using this device, we have already characterized the thermoelastic and thermoelectric behaviours of SiC fibers. Additional results about the plastic properties at high temperatures were also obtained. Indeed, we performed tensile tests between 1200 degrees C up to 1700 degrees C to characterize this plastic behaviour. Some thermal annealing, up to 3 hours at 1700 degrees C, had been also performed. Furthermore, we compare the mechanical behaviours with the thermal evolution of the electric resistivity of these SiC fibers. Soon, MecaSiC will be coupled to a new charged particle accelerator. Thus, in this configuration, we will be able to study in-situ irradiation effects on fibre behaviours, as swelling or creep for example

High temperature graphite irradiation creep experiment in the Dragon Reactor. Dragon Project report

Energy Technology Data Exchange (ETDEWEB)

Manzel, R.; Everett, M. R.; Graham, L. W.

1971-05-15

The irradiation induced creep of pressed Gilsocarbon graphite under constant tensile stress has been investigated in an experiment carried out in FE 317 of the OECD High Temperature Gass Cooled Reactor ''Dragon'' at Winfrith (England). The experiment covered a temperature range of 850 dec C to 1240 deg C and reached a maximum fast neutron dose of 1.19 x 10²¹ n cm^-2 NDE (Nickel Dose DIDO Equivalent). Irradiation induced dimensional changes of a string of unrestrained graphite specimens are compared with the dimensional changes of three strings of restrained graphite specimens stressed to 40%, 58%, and 70% of the initial ultimate tensile strength of pressed Gilsocarbon graphite. Total creep strains ranging from 0.18% to 1.25% have been measured and a linear dependence of creep strain on applied stress was observed. Mechanical property measurements carried out before and after irradiation demonstrate that Gilsocarbon graphite can accommodate significant creep strains without failure or structural deterioration. Total creep strains are in excellent agreement with other data, however the results indicate a relatively large temperature dependent primary creep component which at 1200 deg C approaches a value which is three times larger than the normally assumed initial elastic strain. Secondary creep constants derived from the experiment show a temperature dependence and are in fair agreement with data reported elsewhere. A possible determination of the results is given.

Modelling property changes in graphite irradiated at changing irradiation temperature

CSIR Research Space (South Africa)

Kok, S

2011-01-01

Full Text Available A new method is proposed to predict the irradiation induced property changes in nuclear; graphite, including the effect of a change in irradiation temperature. The currently used method; to account for changes in irradiation temperature, the scaled...

Post-irradiation examinations and high-temperature tests on undoped large-grain UO{sub 2} discs

Energy Technology Data Exchange (ETDEWEB)

Noirot, J., E-mail: jean.noirot@cea.fr [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Pontillon, Y. [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Yagnik, S. [EPRI, P.O. Box 10412, Palo Alto, CA 94303-0813 (United States); Turnbull, J.A. [Independent Consultant (United Kingdom)

2015-07-15

Within the Nuclear Fuel Industry Research (NFIR) programme, several fuel variants –in the form of thin circular discs – were irradiated in the Halden Boiling Water Reactor (HBWR) at burn-ups up to ∼100 GWd/t{sub HM}. The design of the fuel assembly was similar to that used in other HBWR programmes: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature differences within each fuel disc. One such variant was made of large-grain UO{sub 2} discs (3D grain size = ∼45 μm) which were subjected to three burn-ups: 42, 72 and 96 GWd/t{sub HM}. Detailed characterizations of some of these irradiated large-grain UO{sub 2} discs were performed in the CEA Cadarache LECA-STAR hot laboratory. The techniques used included electron probe microanalysis (EPMA), scanning electron microscopy (SEM) and secondary ion mass spectrometry (SIMS). Comparisons were then carried out with more standard grain size UO{sub 2} discs irradiated under the same conditions. Examination of the high burn-up large-grain UO{sub 2} discs revealed the limited formation of a high burn-up structure (HBS) when compared with the standard-grain UO{sub 2} discs at similar burn-up. High burn-up discs were submitted to temperature transients up to 1200 °C in the heating test device called Merarg at a relatively low temperature ramp rate (0.2 °C/s). In addition to the total gas release during these tests, the release peaks throughout the temperature ramp were monitored. Tests at 1600 °C were also conducted on the 42 GWd/t{sub HM} discs. The fuels were then characterized with the same microanalysis techniques as those used before the tests, to investigate the effects of these tests on the fuel’s microstructure and on the fission gas behaviour. This paper outlines the high resistance of this fuel to gas precipitation at high temperature and to HBS formation at high burn-up. It also shows the similarity of the positions, within the grains, where HBS forms

Evaluation of Candidate Linear Variable Displacement Transducers for High Temperature Irradiations in the Advanced Test Reactor

International Nuclear Information System (INIS)

Knudson, D.L.; Rempe, J.L.; Daw, J.E.

2009-01-01

The United States (U.S.) Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to promote nuclear science and technology in the U.S. Given this designation, the ATR is supporting new users from universities, laboratories, and industry as they conduct basic and applied nuclear research and development to advance the nation's energy security needs. A fundamental component of the ATR NSUF program is to develop in-pile instrumentation capable of providing real-time measurements of key parameters during irradiation experiments. Dimensional change is a key parameter that must be monitored during irradiation of new materials being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can experience significant changes during high temperature irradiation. Currently, dimensional changes are determined by repeatedly irradiating a specimen for a defined period of time in the ATR and then removing it from the reactor for evaluation. The time and labor to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data (i.e., only characterizing the end state when samples are removed from the reactor) and may disturb the phenomena of interest. To address these issues, the Idaho National Laboratory (INL) recently initiated efforts to evaluate candidate linear variable displacement transducers (LVDTs) for use during high temperature irradiation experiments in typical ATR test locations. Two nuclear grade LVDT vendor designs were identified for consideration - a smaller diameter design qualified for temperatures up to 350 C and a larger design with capabilities to 500 C. Initial evaluation efforts include collecting calibration data as a function of temperature, long duration testing of LVDT response while held at high temperature, and the assessment of changes

The effect of helium generation and irradiation temperature on tritium release from neutron irradiated beryllium

International Nuclear Information System (INIS)

Kupriyanov, I.B.; Gorokhov, V.A.; Vlasov, V.V.; Kovalev, A.M.; Chakin, V.P.

2004-01-01

The effect of neutron irradiation condition on tritium release from beryllium is described in this paper. Beryllium samples were irradiated in the SM reactor with neutron fluence (E > 0.1 MeV) of (0.37-2.0) x 10 22 cm -2 at 70-100degC and 650-700degC. Mass-spectrometer technique was used in out of tritium release experiments during stepped-temperature anneal within a temperature range from 250 to 1300degC. The total amount of helium accumulated in irradiated beryllium samples varied from 521 appm to 3061 appm. The first signs of tritium release were detected at temperature of 406-553degC. It was shown that irradiation temperature and helium generation level significantly affect the tritium release. A fraction of 44 - 74 % of tritium content in samples irradiated at low temperature (70 - 100degC) is release from beryllium at an annealing temperature below 800degC, whereas for samples after high temperature irradiation (650 - 700 degC) tritium release did not exceed 14 %. Majority of tritium (∼68%) is released within a temperature range from 800 to 920 degC. The increase of helium generation from 521 appm to 3061 appm results in lowering the temperature of maximal tritium release rate and the upper temperature of tritium release from beryllium by 100-130degC and 200-240degC, correspondingly. On the basis of data obtained, the diffusion coefficients of tritium in beryllium were calculated. (author)

Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to High Temperature

International Nuclear Information System (INIS)

Blue, Thomas; Windl, Wolfgang; Dickerson, Bryan

2013-01-01

The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica'@@s optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and

FEMAXI-7 analysis on behavior of medium and high burnup BWR fuels during base-irradiation and power ramp

Energy Technology Data Exchange (ETDEWEB)

Ogiyanagi, Jin, E-mail: ohgiyanagi.jin@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Two power ramp experiments of BWR fuels were analyzed by FEMAXI-7 code. Black-Right-Pointing-Pointer Calculated FGR and cladding deformation showed reasonable agreement with PIE data. Black-Right-Pointing-Pointer High temperature FGR could be predicted by the enhanced Turnbull FG diffusion constant. Black-Right-Pointing-Pointer Local PCMI model in the code could reasonably predict cladding ridging deformation. - Abstract: Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and subsequent power ramp test is analyzed by a fuel performance code FEMAXI-7. The code has a 1.5-D cylindrical geometry (4 axial segments) to have a coupled solution of thermal analysis and FEM mechanical analysis. Two kinds of target fuels are selected; one was subjected to a power ramp test in the DR3 reactor at RISO after the base-irradiation in a commercial BWR, and the other was subjected to the power ramp test in the DR3 reactor after the base-irradiation in the Halden boiling water reactor. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured data. In addition, the calculated ridging deformation of the cladding before and after the ramp test, which is obtained by using a local pellet-cladding mechanical interaction (PCMI) analysis geometry in FEMAXI-7, is compared with the measured data, and it is found that the FEMAXI-7 code is applicable to the local PCMI analysis of medium and high burnup rods under normal operation and power ramp conditions.

Numerical analysis and simulation of behavior of high burn-up PWR fuel pulse-irradiated in reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Suzuki, M.; Sugiyama, T.; Udagawa, Y.; Nagase, F.; Fuketa, T.

2010-01-01

The four cases of the NSRR experiments, consisting of two room temperature tests and two high temperature tests, using high burn-up PWR fuel rods are analyzed by using the RANNS code to discuss the fuel behavior in hypothetical pulse-irradiation conditions, and the results are compared with metallography observations of ruptured claddings. The cladding rupture occurred by a shear sliding which starts from the tip of incipient crack generated in the hydride dense layer. The analyses reveal that the onset of shear sliding leading to cladding rupture can be closely associated with the stress intensity factor KI at the crack tip and local plastic strain evolution around the tip as well, and that these two factors depend also on the temperature of cladding. Simulation calculations on the basis of experimental conditions reveals that the cladding stress is dependent on the height and half-width of pulse power, and for the same integral enthalpy of pulse a larger half-width mitigates the severity of transient and decreases KI to allow plastic strain by temperature rise, thus failure possibility would be markedly decreased

Low-temperature formation of high-quality gate oxide by ultraviolet irradiation on spin-on-glass

International Nuclear Information System (INIS)

Usuda, R.; Uchida, K.; Nozaki, S.

2015-01-01

Although a UV cure was found to effectively convert a perhydropolysilazane (PHPS) spin-on-glass film into a dense SiO x film at low temperature, the electrical characteristics were never reported in order to recommend the use of PHPS as a gate-oxide material that can be formed at low temperature. We have formed a high-quality gate oxide by UV irradiation on the PHPS film, and obtained an interface midgap trap density of 3.4 × 10 11  cm −2 eV −1 by the UV wet oxidation and UV post-metallization annealing (PMA), at a temperature as low as 160 °C. In contrast to the UV irradiation using short-wavelength UV light, which is well known to enhance oxidation by the production of the excited states of oxygen, the UV irradiation was carried out using longer-wavelength UV light from a metal halide lamp. The UV irradiation during the wet oxidation of the PHPS film generates electron-hole pairs. The electrons ionize the H 2 O molecules and facilitate dissociation of the molecules into H and OH − . The OH − ions are highly reactive with Si and improve the stoichiometry of the oxide. The UV irradiation during the PMA excites the electrons from the accumulation layer, and the built-in electric field makes the electron injection into the oxide much easier. The electrons injected into the oxide recombine with the trapped holes, which have caused a large negative flat band voltage shift after the UV wet oxidation, and also ionize the H 2 O molecules. The ionization results in the electron stimulated dissociation of H 2 O molecules and the decreased interface trap density

Low-temperature formation of high-quality gate oxide by ultraviolet irradiation on spin-on-glass

Energy Technology Data Exchange (ETDEWEB)

Usuda, R.; Uchida, K.; Nozaki, S., E-mail: nozaki@ee.uec.ac.jp [Graduate School of Informatics and Engineering, The University of Electro-Communications, 1-5-1 Chofugaoka, Chofu-shi, Tokyo 182-1515 (Japan)

2015-11-02

Although a UV cure was found to effectively convert a perhydropolysilazane (PHPS) spin-on-glass film into a dense SiO{sub x} film at low temperature, the electrical characteristics were never reported in order to recommend the use of PHPS as a gate-oxide material that can be formed at low temperature. We have formed a high-quality gate oxide by UV irradiation on the PHPS film, and obtained an interface midgap trap density of 3.4 × 10{sup 11 }cm{sup −2} eV{sup −1} by the UV wet oxidation and UV post-metallization annealing (PMA), at a temperature as low as 160 °C. In contrast to the UV irradiation using short-wavelength UV light, which is well known to enhance oxidation by the production of the excited states of oxygen, the UV irradiation was carried out using longer-wavelength UV light from a metal halide lamp. The UV irradiation during the wet oxidation of the PHPS film generates electron-hole pairs. The electrons ionize the H{sub 2}O molecules and facilitate dissociation of the molecules into H and OH{sup −}. The OH{sup −} ions are highly reactive with Si and improve the stoichiometry of the oxide. The UV irradiation during the PMA excites the electrons from the accumulation layer, and the built-in electric field makes the electron injection into the oxide much easier. The electrons injected into the oxide recombine with the trapped holes, which have caused a large negative flat band voltage shift after the UV wet oxidation, and also ionize the H{sub 2}O molecules. The ionization results in the electron stimulated dissociation of H{sub 2}O molecules and the decreased interface trap density.

Stability of lithium niobate on irradiation at elevated temperature

International Nuclear Information System (INIS)

Primak, W.; Gavin, A.P.; Anderson, T.T.; Monahan, E.

1977-01-01

In contrast to results obtained for neutron irradiation in a thermal reactor near room temperature, lithium niobate plates irradiated in the Experimental Breeder Reactor II (EBR-II) did not become metamict. This is attributed to the elevated temperature of the EBR-II. Ion bombardment experiments indicate that to avoid disordering of lithium niobate on irradiation, its temperature should be maintained above 673 K. Evidence for ionic conductivity was found at 873 K, indicating that it would be inadvisable to permit the temperature to rise that high, particularly with voltage across the plate. In reactor application as a microphone transducer, it is tentatively recommended that the lithium niobate be maintained in the middle of this temperature range for a major portion of reactor operating time

A Status of Art-Report on the Fission Products Behavior Released from Spent Fuel at High Temperature Conditions

International Nuclear Information System (INIS)

Park, Geun Il; Kim, J. H.; Lee, J. W.

2003-04-01

The experiments on the fission products release behavior from spent fuel at high temperature assuming reactor accident conditions have been carried out at Oak Ridge Nation Laboratory of USA in HI/VI tests, CEA of France in HEVA/VERCOS tests, AEA of England and CRNL of Canada in HOX test. The VEGA program to study the fission product release behavior from LWR irradiated fuel was recently initiated at JAERI. The key parameter affecting the fission product(FP) release behavior is temperature. In addition, other parameters such as fuel oxidation, burnup, pre-transient conditions are found to affect the FP releases considerably in the earlier tests. The atmosphere conditions such as oxidizing atmosphere (steam or air) or reducing atmosphere (hydrogen) can cause significant change of FPs release and transport behavior due to chemical forms of the reactive FPs which is dependent on the oxidation potential. The effect of fuel burnup on the Kr-85 or Cs-137 release showed that the release rates of these radionuclides increased with the increase of burnup, meaning that release rates are dominated by the atomic diffusions in the grains and they are primarily a function of temperature. However, the data on FPs release behavior using higher burnups above 50,000 MWD/MTU are not so many reported up to now. This report summarizes the test results of FPs release behavior in reactor accident conditions produced from other countries mentioned above. This review and analysis on earlier studies would be useful for predicting the release characteristics of FPs from domestic spent fuel. The release rates of fission gas or FPs from spent fuel at high temperature conditions during fabrication process of dry recycling fuel were also analyzed using many data obtained from earlier tests

Irradiation Behavior and Post-Irradiation Examinations of an Acoustic Sensor Using a Piezoelectric Transducer

International Nuclear Information System (INIS)

Lambert, T.; Zacharie-Aubrun, I.; Hanifi, K.; Valot, Ch.; Fayette, L.; Rosenkantz, E.; Ferrandis, J.Y.; Tiratay, X.

2013-06-01

The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. In the framework of high burn-up fuel experiments under transient operating conditions, an innovative sensor based on acoustic method was developed by CEA and IES (Southern Electronic Institute).This sensor is used to determine the on-line composition of the gases located in fuel rodlet free volume and thus, allows calculating the molar fractions of fission gases and helium. The main principle of the composition determination by acoustic method consists in measuring the time of flight of an acoustic signal emitted and reflected in a specific cavity. A piezoelectric transducer, driven by a pulse generator, generates the acoustic wave in the cavity. The piezoelectric transducer is a PZT ceramic disk, mainly consisting of lead, zirconium and titanium. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. However, during the irradiation test, acoustic signal degradation was observed, mainly due to irradiation effect but also due to the increasing of the gas temperature. Despite this acoustic signal degradation, the time of flight measurements were carried out with good accuracy throughout the test, thanks to the development of a more efficient signal processing. After experiment, neutronic calculations were performed in order to determine neutron fluence at the level of the piezoelectric transducer. In order to have a better understanding of the acoustic sensor behavior under irradiation, Post Irradiation Examination program was done on piezoelectric transducer and on acoustic coupling material too. These examinations were also realized on a non-irradiated acoustic sensor built in the same conditions and with the same materials and the same

Modeling the influence of high dose irradiation on the deformation and damage behavior of RAFM steels under low cycle fatigue conditions

Energy Technology Data Exchange (ETDEWEB)

Aktaa, J. [Forschungszentrum Karlsruhe GmbH, Institute for Materials Research II, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)], E-mail: aktaa@imf.fzk.de; Petersen, C. [Forschungszentrum Karlsruhe GmbH, Institute for Materials Research II, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2009-06-01

A viscoplastic deformation damage model developed for RAFM steels in the reference un-irradiated state was modified taking into account the irradiation influence. The modification mainly consisted in adding an irradiation hardening variable with an appropriate evolution equation including irradiation dose driven terms as well as inelastic deformation and thermal recovery terms. With this approach, the majority of the material and temperature dependent model parameters are no longer dependent on the irradiation dose and only few parameters need to be determined by applying the model to RAFM steels in the irradiated state. The modified model was then applied to describe the behavior of EUROFER 97 observed in the post irradiation examinations of the irradiation programs ARBOR 1, ARBOR 2 and SPICE. The application results will be presented and discussed in addition.

Using the quantum yields of photosystem II and the rate of net photosynthesis to moniter high irradiance and temperature stress in chrysanthemum (Dendrantherma grandiflora)

DEFF Research Database (Denmark)

Janka, Eshetu; Körner, Oliver; Rosenqvist, Eva

2015-01-01

and quantum yield of PSII remaining low until the temperature reaches 28 °C and 2) the integration of online measurements to monitor photosynthesis and PSII operating efficiency may be used to optimise dynamic greenhouse control regimes by detecting plant stress caused by extreme microclimatic conditions.......Under a dynamic greenhouse climate control regime, temperature is adjusted to optimise plant physiological responses to prevailing irradiance levels; thus, both temperature and irradiance are used by the plant to maximise the rate of photosynthesis, assuming other factors are not limiting...... irradiance, the maximum Pn and ETR were reached at 24 °C. Increased irradiance decreased the PSII operating efficiency and increased NPQ, while both high irradiance and temperature had a significant effect on the PSII operating efficiency at temperatures >28 °C. Under high irradiance and temperature, changes...

Behavior of fluorine 18 in neutron irradiated zeolites

International Nuclear Information System (INIS)

Estevez Lopez, D.R.

1992-01-01

The transformation of Li-exchanged H-Y zeolite has been investigated at 300, 550, 850 and 1050 Centigrade degree, formation of quartz structure in addition to an amorphous phase, was nited. The Li-aluminosilicate obtained was neutron irradiated and the chemical behavior of 18 F produced by the reaction sequence 6 Li (n, α) 3 H, 16 O ( 3 H, n) 18 F, was studied. The neutron irradiated material was purged with argon-hydron gas streams. It was found that the amount of released 18 F depends on the temperature used (Author)

Determinations of the temperature of terminal solid solubility in dissolution and precipitation of hydrogen/deuterium in irradiated Zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Vizcaino, P [CNEA-CONICET, Centro Atomico Ezeiza (Argentina)

2012-07-01

The proposed plan is an approach to the metallurgical consequences of the high neutron fluencies (10''2''2 n/cm''2) on the hydrogen behavior in zirconium based alloys, based on the significance of the microstructural behavior of the high burn up fuel claddings during the dry storage period. The studies are focused on Zircaloy-4, concerning to two processes: Neutron irradiation damage; Hydrogen pick up. The Zircaloy-4 was taken from cooling channels of the PHWR Atucha 1. These components remained more than 10 years in service, reaching neutron fluencies up to 10''2''2 n/cm''2. In the last recent years, measurements of the hydride dissolution temperatures have shown that hydrogen solubility is affected by the neutron irradiation, increasing it respect to the unirradiated Zircaloy solubility. In addition, in this material the amorphization/dissolution of the second phase particles (SPPs) was observed, being proposed an interaction between the hydrogen atoms, the SPPs and the irradiation defects as a possible explanation of the observed behavior. For the present case, attention will be focused on the hydride precipitation process, since it is strongly related with delay hydrogen cracking initiation, a problem of direct concern for the dry storage. The goal of the present proposal is to make an approach to the source of the observed effect, applying several specific techniques as differential scanning calorimetry (DSC), high resolution x-ray diffraction and transmission electron microscopy. The objectives can be divided as follows: Determination of the temperatures of terminal solid solubility in dissolution (TTSSd) and in precipitation (TTSSp) in high fluency irradiated Zircaloy-4, reproducing the temperatures at which the Zircaloy fuel claddings remain during dry storage by an annealing program during the DSC experiments; Observations by optical and transmission electron microscopy of the hydride distribution before (as received material) and after high temperature

Irradiation Creep and Swelling of Russian Ferritic-Martensitic Steels Irradiated to Very High Exposures in the BN-350 Fast Reactor at 305-335 degrees C

International Nuclear Information System (INIS)

Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

2003-01-01

Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb-Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

Development of irradiation technique with satured temperature capsule in the JMTR

International Nuclear Information System (INIS)

Ohtaka, Kimihiro

1999-01-01

The irradiation assisted stress corrosion cracking (IASCC) of in-core structural materials caused by the simultaneous effects of neutron irradiation and high temperature water environments has been pointed out as one of the major concerns not only for the light water reactors (LWRs) but also for the water-cooled fusion reactor, i.e,. ITER. The IASCC of the austenitic stainless steels or nickel base alloys has been studied for more than ten years under international efforts in the various projects for the plant life assessment and extension of LWRs. However its mechanism has not been clarified yet in spite of the extensive post-irradiation examinations. Under this situation, it is desired to perform irradiation tests under specially controlled conditions so that the effect of irradiation and high temperature water can be separately evaluated. In the Japan Materials Testing Reactor (JMTR), irradiation technique with the saturation temperature capsule (SATCAP) was developed for irradiation of the materials in the water with high, but constant, temperature and applied to study the IASCC. The capability of the SATCAP was improved by enhancing the temperature controllability to irradiate materials even in a low gamma region in the JMTR core. The performance tests of the improved SATCAP carried out in the JMTR have proven its capabilities. Based on experiences of the SATCAP, preliminary design study for the upgraded in-pile test facility are now underway in the JMTR. The test facility has a new test loop to achieve irradiate test simulated water environment of LWRs. The design, test results of the SATCAP and the design study of upgraded in-pile test facility are described in this paper

Radiation annealing mechanisms of low-alloy reactor pressure vessel steels dependent on irradiation temperature and neutron fluence

International Nuclear Information System (INIS)

Pachur, D.

1982-01-01

Heat treatment after irradiation of reactor pressure vessel steels showed annealing of irradiation embrittlement. Depending on the irradiation temperature, the embrittlement started to anneal at about 220 0 C and was completely annealed at 500 0 C with 4 h of annealing time. The annealing behavior was normally measured in terms of the Vickers hardness increase produced by irradiation relative to the initial hardness as a function of the annealing temperature. Annealing results of other mechanical properties correspond to hardness results. During annealing, various recovery mechanisms occur in different temperature ranges. These are characterized by activation energies from 1.5 to 2.1 eV. The individual mechanisms were determined by the different time dependencies at various temperatures. The relative contributions of the mechanisms showed a neutron fluence dependence, with the lower activation energy mechanisms being predominant at low fluence and vice versa. In the temperature range where partial annealing of a mechanism took place during irradiation, an increase in activation energy was observed. Trend curves for the increase in transition temperature with irradiation, for the relative increase of Vickers hardness and yield strength, and for the relative decrease of Charpy-V upper shelf energy are interpreted by the behavior of different mechanisms

Design, Fabrication, Test Report of the Material Capsule(08M-10K) with Double Thermal Media for High-temperature Irradiation

International Nuclear Information System (INIS)

Cho, Man Soon; Choo, K. N.; Kang, Y. H.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Kim, B. G.; Oh, S. Y.

2010-01-01

To overcome the restriction of the irradiation test at a high temperature of the existing material capsule with Al thermal media, a capsule suitable for the irradiation at the high temperature was developed and the performance test was undertaken. The 08M-10K capsule was designed and fabricated as that with double thermal media to verify the structural and external integrity in the high-temperature irradiation higher than 500 .deg. C. The thermal performance test was undertaken at the out-pile test facility, and the soundness of the double thermal media was confirmed with the naked eye after disassembling the capsule. Though the temperatures of the specimens reach 500±20 .deg. C as a result maintaining the capsule during 5 hours after setting the specimens temperatures in the target range, the high-temperature thermal media with double structure was confirmed to maintain the soundness. And the specimens and the thermal media were heated to 600 .deg. C for about 3 minutes, but the thermal media were maintained sound. Especially, the Al thermal media were heated for 5 hours in range of 500±20 .deg. C and for 3 minutes at the temperature of 600 .deg. C. However, the thermal media were confirmed to maintain the soundness. Whether a capsule has only Al thermal media or the high-temperature thermal media with double structure, any capsule can be used in the range of 500±20 .deg. C as the result of this experiment maintaining the specimens high-temperature

High Total Ionizing Dose and Temperature Effects on Micro- and Nano-electronic Devices

International Nuclear Information System (INIS)

Gaillardin, M.; Martinez, M.; Paillet, P.; Leray, J.L.; Marcandella, C.; Duhamel, O.; Raine, M.; Richard, N.; Girard, S.; Ouerdane, Y.; Boukenter, A.; Goiffon, V.; Magnan, P.; Andrieu, F.; Barraud, S.; Faynot, O.

2013-06-01

This paper investigates the vulnerability of several micro- and nano-electronic technologies to a mixed harsh environment including high total ionizing dose at MGy levels and high temperature. Such operating conditions have been revealed recently for several applications like new security systems in existing or future nuclear power plants, fusion experiments, or deep space missions. In this work, the competing effects already reported in literature of ionizing radiations and temperature are characterized in elementary devices made of MOS transistors from several technologies. First, devices are irradiated using a radiation laboratory X-ray source up to MGy dose levels at room temperature. Devices are grounded during irradiation to simulate a circuit which waits for a wake up signal, representing most of the lifetime of an integrated circuit operating in a harsh environment. Devices are then annealed at several temperatures to discuss the post-irradiation behavior and to determine whether an elevated temperature is an issue or not for circuit function in mixed harsh environments. (authors)

Temperature-dependent surface modification of Ta due to high-flux, low-energy He+ ion irradiation

International Nuclear Information System (INIS)

Novakowski, T.J.; Tripathi, J.K.; Hassanein, A.

2015-01-01

This work examines the response of Tantalum (Ta) as a potential candidate for plasma-facing components (PFCs) in future nuclear fusion reactors. Tantalum samples were exposed to high-flux, low-energy He + ion irradiation at different temperatures in the range of 823–1223 K. The samples were irradiated at normal incidence with 100 eV He + ions at constant flux of 1.2 × 10 21 ions m −2  s −1 to a total fluence of 4.3 × 10 24 ions m −2 . An additional Ta sample was also irradiated at 1023 K using a higher ion fluence of 1.7 × 10 25 ions m −2 (at the same flux of 1.2 × 10 21 ions m −2  s −1 ), to confirm the possibility of fuzz formation at higher fluence. This higher fluence was chosen to roughly correspond to the lower fluence threshold of fuzz formation in Tungsten (W). Surface morphology was characterized with a combination of field-emission scanning electron microscopy (FE-SEM) and atomic force microscopy (AFM). These results demonstrate that the main mode of surface damage is pinholes with an average size of ∼70 nm 2 for all temperatures. However, significantly larger pinholes are observed at elevated temperatures (1123 and 1223 K) resulting from the agglomeration of smaller pinholes. Ex situ X-ray photoelectron spectroscopy (XPS) provides information about the oxidation characteristics of irradiated surfaces, showing minimal exfoliation of the irradiated Ta surface. Additionally, optical reflectivity measurements are performed to further characterize radiation damage on Ta samples, showing gradual reductions in the optical reflectivity as a function of temperature.

Effects of cryogenic irradiation on temperature sensors

International Nuclear Information System (INIS)

Courts, S.S.; Holmes, D.S.

1996-01-01

Several types of commercially available cryogenic temperature sensors were calibrated, irradiated at 4.2 K by a gamma or neutron source, and recalibrated in-situ to determine their suitability for thermometry in radiation environments. Comparisons were made between pre- and post-irradiation calibrations with the equivalent temperature shift calculated for each sensor at various temperature in the 4.2 K to 330 K range. Four post-irradiation calibrations were performed with annealing steps performed at 20 K, 80 K, and 330 K. Temperature sensors which were gamma irradiated were given a total dose of 10,000 Gy. Temperature sensors which were neutron irradiated were irradiated to a total fluence of 2 x 10 12 n/cm 2 . In general, for gamma radiation environments, diodes are unsuitable for use. Both carbon glass and germanium resistance sensors performed well at lower temperature, while platinum resistance sensors performed best above 30 K. Thin-film rhodium and Cernox trademark resistance sensors both performed well over the 4.2 K to 330 K range. Only thin-film rhodium and Cernox trademark resistance temperature sensors were neutron irradiated and they both performed well over the 4.2 K to 330 K range

Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to Very High Temperatures

Energy Technology Data Exchange (ETDEWEB)

Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Windl, Wolfgang [The Ohio State Univ., Columbus, OH (United States); Dickerson, Bryan [Luna Innovations, Inc. (United States)

2013-01-03

The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica's optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and

Effect of individual or combined treatment by γ-irradiation or temperature (high or low) on bacillus subtilis spores and its application for sterilization of ground beef

International Nuclear Information System (INIS)

El-Zawahry, Y.A.; Mostafa, S.A.; Awny, N.M.

1986-01-01

The combination of two lethal agents such as irradiation and temperature (high or sub zero) resulted in synergistic death or B. subtilis spores (as indicated by decrease in the thermal D-value). The extent of this synergism in killing a spore population depended mainly on the sequence on application of the two physical agents. Irradiation-temperature (high or sub zero) sequence killed more but injured less B. subtilis spores than temperature irradiation sequence or irradiation and temperature applied separately. Storage at -20 0 C killed more spores than storage at -2 0 C if carried after irradiation, while the reverse was true of storage was prior irradiation. An irradiation dose of 8 KGY followed by thermal exposure to 70 0 C for 1 hr is suggested for the sterilization of ground beef. Irradiation induced certain quantitative changes on the amino-N, protein-N, RNA and DNA of the first subcultures of irradiated spores with stimulatory effect at low irradiation doses and inhibitory effect at the high irradiation doses. This might explain the increased sensitivity of irradiated spores to subsequent exposure to unfavourable temperature (high or sub zero). Exposure of B. subtilis spore to 70 0 C induced a stimulation in the amino- and protein-N of the resulting cells while exposure to 80 0 C resulted in a significant decrease in the amino-N. The protein-N remained more or less the same

Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys

International Nuclear Information System (INIS)

Ribis, J.

2007-12-01

The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

High temperature annealing of minority carrier traps in irradiated MOCVD n(+)p InP solar cell junctions

Science.gov (United States)

Messenger, S. R.; Walters, R. J.; Summers, G. P.

1993-01-01

Deep level transient spectroscopy was used to monitor thermal annealing of trapping centers in electron irradiated n(+)p InP junctions grown by metalorganic chemical vapor deposition, at temperatures ranging from 500 up to 650K. Special emphasis is given to the behavior of the minority carrier (electron) traps EA (0.24 eV), EC (0.12 eV), and ED (0.31 eV) which have received considerably less attention than the majority carrier (hole) traps H3, H4, and H5, although this work does extend the annealing behavior of the hole traps to higher temperatures than previously reported. It is found that H5 begins to anneal above 500K and is completely removed by 630K. The electron traps begin to anneal above 540K and are reduced to about half intensity by 630K. Although they each have slightly different annealing temperatures, EA, EC, and ED are all removed by 650K. A new hole trap called H3'(0.33 eV) grows as the other traps anneal and is the only trap remaining at 650K. This annealing behavior is much different than that reported for diffused junctions.

Behavior of carbon readsorbed on tungsten during low energy Ar ion irradiation at elevated temperatures

International Nuclear Information System (INIS)

Pranevicius, L.; Pranevicius, L.L.; Milcius, D.; Templier, C.; Bobrovaite, B.

2008-01-01

A study of the behavior of carbon sputtered and readsorbed after scattering collisions with particles of surrounding gas on the tungsten surface affected by Ar ion irradiation with the flux equal to 2 x 10 16 cm -2 s -1 extracted from plasma under 300 V negative bias voltage in the temperature range 370-870 K was performed. The dependence of the W sample weight change on the working gas pressure in the range 0.1-10 Pa was registered and the information was deduced about prevailing sputtering-redeposition processes. The depth profiles of carbon at the tungsten surface were measured. We found that carbon distribution profiles in tungsten depend on the C redeposition rate for fixed ion irradiation parameters. Three regimes have been distinguished: (i) at working gas pressure equal to 5 Pa and more, the C redeposition rate prevails the sample surface erosion rate and the W surface is covered by continuous amorphous carbon film (the C film growth regime), (ii) at working gas pressure equal to about 1 Pa, the C redepostion rate is approximately equal to the erosion rate and the W surface is partially covered by redeposited carbon, and (iii) at working gas pressure less than 0.2 Pa, the erosion rate prevails the C redeposition rate (the W surface erosion regime). In the regime of balanced redeposition and erosion deep C penetration depth into nanocrystalline W was registered. It is suggested that under simultaneous C adsorption and ion irradiation at elevated temperature C adatoms are driven from the W surface into grain boundaries and into the bulk by the difference in chemical potentials between the activated W surface and grain boundaries. As the W surface is covered by amorphous C film, the grain boundaries are blocked and the efficiency of carbon transport decreases

The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

International Nuclear Information System (INIS)

Fang, Xiang; Yu, Suyuan; Wang, Haitao; Li, Chenfeng

2014-01-01

Highlights: • The mechanical behavior of graphite component in HTRs under high temperature and neutron irradiation conditions is simulated. • The computational process of mechanical analysis is introduced. • Deformation, stresses and failure probability of the graphite component are obtained and discussed. • Various temperature and neutron dose ranges are selected in order to investigate the effect of in-core conditions on the results. - Abstract: In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750 °C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1 × 10 22 n cm −2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400 °C, 500 °C, 600 °C and 750 °C are selected as the

Change in digestibility of gamma-irradiated starch by low temperature cooking

International Nuclear Information System (INIS)

Kume, T.; Ishigaki, I.; Rahman, S.

1988-01-01

Combination effect of irradiation and low temperature cooking on starch digestibility has been investigated as a basic research for application of radiosterilization on starch fermentation. The digestion of corn starch by glucoamylase after cooking at low temperature was enhanced by γ-irradiation and the required cooking temperature was decreased from 75-80 0 C to 65 0 C by 25 kGy. Gelatinization of starches except tapioca starch was enhanced by irradiation and it corresponds to the digestibility. The digestibility of potato starch which has a high viscosity was especially enhanced at low temperature cooking because the viscosity was markedly decreased by irradiation. These results show that the irradiation of starches is useful not only for the sterilization of fermentation broth but also for the enhancement of digestion. (orig.) [de

Change in digestibility of gamma-irradiated starch by low temperature cooking

Energy Technology Data Exchange (ETDEWEB)

Kume, T.; Ishigaki, I.; Rahman, S.

1988-04-01

Combination effect of irradiation and low temperature cooking on starch digestibility has been investigated as a basic research for application of radiosterilization on starch fermentation. The digestion of corn starch by glucoamylase after cooking at low temperature was enhanced by ..gamma..-irradiation and the required cooking temperature was decreased from 75-80/sup 0/C to 65/sup 0/C by 25 kGy. Gelatinization of starches except tapioca starch was enhanced by irradiation and it corresponds to the digestibility. The digestibility of potato starch which has a high viscosity was especially enhanced at low temperature cooking because the viscosity was markedly decreased by irradiation. These results show that the irradiation of starches is useful not only for the sterilization of fermentation broth but also for the enhancement of digestion.

Prediction of the brittle-ductile transition temperature shift, from irradiation experiments obtained in France

International Nuclear Information System (INIS)

Miannay, D.; Dussarte, D.; Soulat, P.

1988-10-01

In integrety - evaluation studies of the nuclear-reactor vessel, the toughness of component materials is given by a reference curve, which represents the lowest level of the toughness possible values, as a function of the temperature. Its temperature-scale position is given by the null ductility temperature. In non-irradiated conditions, the temperature of the vessel materials is experimentally obtained. To take into account the irradiation effect, a shift of this temperature, given by the correlations where chemical composition and neutronic dose are included, is considered. Up to date correlations, settled for materials brittler than those applied in the French program, are called in question again. A correlation between results obtained in France since 1973, for base metals and welded joints is given. The behavior of CHOOZ vessel, under irradiation conditions, is discussed [fr

Electrical properties of as-grown and proton-irradiated high purity silicon

Energy Technology Data Exchange (ETDEWEB)

Krupka, Jerzy, E-mail: krupka@imio.pw.edu.pl [Institute of Microelectronics and Optoelectronics, Warsaw University of Technology, Koszykowa 75, 00-662 Warsaw (Poland); Karcz, Waldemar [Joint Institute for Nuclear Research, Joliot-Curie 6, 141980 Dubna (Russian Federation); Kamiński, Paweł [Institute of Electronic Materials Technology, Wólczyńska 13, 301-919 Warsaw (Poland); Jensen, Leif [Topsil Semiconductor Materials A/S, Siliciumvej 1, DK-3600 Frederikssund (Denmark)

2016-08-01

The complex permittivity of as-grown and proton-irradiated samples of high purity silicon obtained by the floating zone method was measured as a function of temperature at a few frequencies in microwave spectrum by employing the quasi TE{sub 011} and whispering gallery modes excited in the samples under test. The resistivity of the samples was determined from the measured imaginary part of the permittivity. The resistivity was additionally measured at RF frequencies employing capacitive spectroscopy as well as in a standard direct current experiment. The sample of as-grown material had the resistivity of ∼85 kΩ cm at room temperature. The sample irradiated with 23-MeV protons had the resistivity of ∼500 kΩ cm at 295 K and its behavior was typical of the intrinsic material at room and at elevated temperatures. For the irradiated sample, the extrinsic conductivity region is missing and at temperatures below 250 K hopping conductivity occurs. Thermal cycle hysteresis of the resistivity for the sample of as-grown material is observed. After heating and subsequent cooling of the sample, its resistivity decreases and then slowly (∼50 h) returns to the initial value.

Electrochemical behavior of TIO{sub 2} deposited stainless steel in high temperature water

Energy Technology Data Exchange (ETDEWEB)

Okamura, M.; Yamamoto, S. [Toshiba Corp., Kawasaki-city, Kanagawa (Japan); Urata, H.; Takagi, J. [Toshiba Corp., Yokohama-city, Kanagawa (Japan)

2010-07-01

It has previously been confirmed that the electrochemical corrosion potential (ECP) of stainless steel (SS) shifts in the negative direction by deposition of TiO{sub 2}. Recently we showed that TiO{sub 2} could decrease the ECP of SS in the absence of UV irradiation. In this study we measured the anodic polarization curve in high temperature water under UV irradiation and none irradiation condition and considered the mechanism of the ECP shift by TiO{sub 2} deposition. The anodic current density of the specimen increased with increasing the UV irradiation intensity and with increasing the amount of TiO{sub 2} deposition under none UV irradiation. Furthermore the oxide film of the specimen affects on the anodic current density was clarified. It was verified the ECP shift is caused by the anodic current density increasing with TiO{sub 2} deposition under both conditions of UV and none UV irradiation. (author)

Molecular weight distribution of electron and γ-ray irradiated PEEK measured by very high temperature GPC

International Nuclear Information System (INIS)

Nakahara, H.

1996-01-01

Poly(ether ether ketone)(PEEK) films were irradiated with electron beam in air and in helium. Gel fractions of the PEEK samples were determined as the ratio of the weight of insoluble fraction/total weigh by extracting the samples with 1-chloronaphthalene (1-CN) at 260degC. While unirradiated PEEK samples were dissolved in 1-CN completely, PEEK samples highly (10 - 50 MGy) irradiated in air were almost insoluble in the solvent. The weight-average molecular weight M w of soluble fractions of the samples were measured by very high temperature gel permeation chromatography (VHTGPC): it was found that the M w decreases with increasing dose. On the other hand, PEEK samples irradiated in helium gave gel fractions at lower doses (0 - 5 MGy) than in air. The PEEK films were also irradiated with 60 Co γ-rays in the dose range, i.e. from 0 to 5 MGy. The γ-irradiated PEEK samples were completely dissolved in 1-CN at 260degC. Their M w measured by VHTGPC decreases with increasing dose. (author)

IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

International Nuclear Information System (INIS)

Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

2002-01-01

Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

Theoretical and practical implications of the effects of temperature during irradiation and during pre- and post-irradiation storage on the response of thermoluminescence dosimeters

International Nuclear Information System (INIS)

Gail de Planque, E.

1984-01-01

Experiments have been conducted to determine the applicability of the Randall-Wilkins theory for describing the behavior of CaF 2 :Mn thermoluminescence dosimeters (Harshaw TLD-400 chips). Results were obtained for four different conditions: irradiation followed by storage, irradiation after storage, irradiation both preceded and followed by storage, and continuous simultaneous irradiation and storage. The experiments were performed for storage intervals of approximately 1, 2, 3, 5, 6 and 7 days at five different storage temperatures: -25, +20, +65, +150 and +175 0 C. The results indicate fading that is described not by the Randall-Wilkins theory but rather as a linear function of the logarithm of the storage time. While the results suggest that the trapping efficiency is independent of temperature, they do demonstrate a small decrease in TL response with storage time prior to irradiation which is independent of temperature and time (>17 hours) and hence probably not dosimetric in origin but perhaps optically related. Glow curve analyses support the concept of a band of traps rather than a single trap. The overall results are compared to other data available in the literature most of which is for room-temperature storage. These data, for storage periods ranging from minutes to one year, can also be described as a linear function of the logarithm of the storage time and are remarkably consistent when uniformly normalized. Although peripheral experiments revealed problems associated with self-irradiation as well as a decline in sensitivity with use, the stability experiment results verify the highly favorable stability properties of CaF 2 :Mn for widespread application

Effects of electron beam irradiation on mechanical properties at low and high temperature of fiber reinforced composites using PEEK as matrix material

International Nuclear Information System (INIS)

Sasuga, Tsuneo; Seguchi, Tadao; Sakai, Hideo; Odajima, Toshikazu; Nakakura, Toshiyuki; Masutani, Masahiro.

1987-11-01

Carbon fiber reinforced composite (PEEK-CF) using polyarylether-ether-ketone (PEEK) as a matrix material was prepared and the electron beam radiation effects on the mechanical properties at low and high temperature and the effects of annealing after irradiation were studied. Cooling down to 77 K, the flexural strength of PEEK-CF increased to about 20 % than that at room temperature. The data of flexural strength for the irradiated specimens showed some scattering, but the strength and modulus at 77 K were changed scarcely up to 120 MGy. The flexural strength and modulus in the unirradiated specimen decreased with increasing of measurement temperature, and the strength at 140 deg C, which is the just below temperature of the glass transition of PEEK, was to 70 % of the value at room temperature. For the irradiated specimens, the strength and modulus increased with dose and the values at 140 deg C for the specimen irradiated with 120 MGy were nearly the same with the unirradiated specimen measured at room temperature. The improvement of mechanical properties at high temperature by irradiation was supported by a viscoelastic measurement in which the glass transition shifted to the higher temperature by the radiation-induced crosslinking. A glass fiber reinforced PEEK composite (PEEK-GF) was prepared and its irradiation effects by electron beam was studied. Unirradiated PEEK-GF showed the same performance with that for GFRP of epoxide resin as matrix material, but by irradiation the flexual strength and modulus decreased with dose. It was revealed that this composite was destroyed by delamination because inter laminar shear strength (ILSS) decreased with dose and analysis of the profile of S-S curve showed typical delamination. Fractoglaphy by electron microscopy supported the delamination which is caused by the lowering of adhesion on interface between the fiber and matrix with increase of dose. (author)

Effect of high ambient temperature on behavior of sheep under semi-arid tropical environment.

Science.gov (United States)

De, Kalyan; Kumar, Davendra; Saxena, Vijay Kumar; Thirumurugan, Palanisamy; Naqvi, Syed Mohammed Khursheed

2017-07-01

High environmental temperature is a major constraint in sheep production under semi-arid tropical environment. Behavior is the earliest indicator of animal's adaptation and responses to the environmental alteration. Therefore, the objective of this study was to assess the effects of high ambient temperature on the behavior of sheep under a semi-arid tropical environment. The experiment was conducted for 6 weeks on 16 Malpura cross (Garole × Malpura × Malpura (GMM)) rams. The rams were divided equally into two groups, designated as C and T. The rams of C were kept in comfortable environmental conditions served as control. The rams of T were exposed to a different temperature at different hours of the day in a climatic chamber, to simulate a high environmental temperature of summer in semi-arid tropic. The behavioral observations were taken by direct instantaneous observation at 15-min intervals for each animal individually. The feeding, ruminating, standing, and lying behaviors were recorded twice a week from morning (0800 hours) to afternoon (1700 hours) for 6 weeks. Exposure of rams to high temperature (T) significantly (P behavior of sheep which is directed to circumvent the effect of the stressor.

Effect of high ambient temperature on behavior of sheep under semi-arid tropical environment

Science.gov (United States)

De, Kalyan; Kumar, Davendra; Saxena, Vijay Kumar; Thirumurugan, Palanisamy; Naqvi, Syed Mohammed Khursheed

2017-07-01

High environmental temperature is a major constraint in sheep production under semi-arid tropical environment. Behavior is the earliest indicator of animal's adaptation and responses to the environmental alteration. Therefore, the objective of this study was to assess the effects of high ambient temperature on the behavior of sheep under a semi-arid tropical environment. The experiment was conducted for 6 weeks on 16 Malpura cross (Garole × Malpura × Malpura (GMM)) rams. The rams were divided equally into two groups, designated as C and T. The rams of C were kept in comfortable environmental conditions served as control. The rams of T were exposed to a different temperature at different hours of the day in a climatic chamber, to simulate a high environmental temperature of summer in semi-arid tropic. The behavioral observations were taken by direct instantaneous observation at 15-min intervals for each animal individually. The feeding, ruminating, standing, and lying behaviors were recorded twice a week from morning (0800 hours) to afternoon (1700 hours) for 6 weeks. Exposure of rams to high temperature (T) significantly ( P animals of T spent significantly ( P behavior of sheep which is directed to circumvent the effect of the stressor.

Effects of combined action of temperature and irradiation on growth and crypt forming in fungi rhizopus nigricans and penicillium italicum

International Nuclear Information System (INIS)

Lezhneva, M.L.; Petrash, I.P.; Koval'skaya, L.P.

1974-01-01

The purpose of this study was to assess the effect of incubation temperature on the vegetative growth and sporulation of non-irradiated and irradiated molds. The behavior of irradiated molds at various incubation temperatures was found to depend on their physiological condition. By combining irradiation with exposure to properly selected storage temperatures, the microbial damage to fruits may apparently be reduced, even in the case of fruits containing radioresistant fungi adapted to growth at low above-zero C temperatures. (E.T.)

Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.

1985-01-01

Onset of melting is an important performance limit for irradiated UO 2 and UO 2 -based nuclear reactor fuels. Melting (solidus) temperatures are reasonably well known for starting fuel materials such as UO 2 and (U,PU)O 2 , however the influence of burnup on oxide fuel melting behavior continues to represent an area of considerable uncertainty. In this paper we report the results of a variety of melting temperature measurements on pseudo-binary fuel-fissia mixtures such as UO 2 -PUO 2 , UO 2 -CeO 2 , UO 2 -BaO, UO 2 -SrO, UO 2 -BaZrO 3 and UO 2 -SrZrO 3 . These measurements were performed using the thermal arrest technique on tungsten-encapsulated specimens. Several low melting eutectics, the existence of which had previously been inferred from post-irradiation examinations of high burnup mixed oxide fuels, were characterized in the course of the investigation. Also, an assessment of melting temperature changes in irradiated oxide fuels due to the production and incorporation of soluble oxidic fission products was performed by application of solution theory to the available pseudo-binary phase diagram data. The results of this assessment suggest that depression of oxide fuel solidus temperatures by dissolved fission products is substantially less than that indicated by earlier experimental studies. (orig.)

Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

International Nuclear Information System (INIS)

Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

2007-01-01

Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

Electron irradiation effect on the reverse phase transformation temperatures in TiNi shape memory alloy thin films

International Nuclear Information System (INIS)

Wang, Z.G.; Zu, X.T.; Fu, Y.Q.; Zhu, S.; Wang, L.M.

2005-01-01

In this work, Ti-Ni shape memory alloy thin films were irradiated by 1.7 MeV electron with three types of fluences: 4 x 10 20 , 7 x 10 20 and 1 x 10 21 /m 2 . The influence of electron irradiation on the transformation behavior of the TiNi thin films were investigated by differential scanning calorimetry. The transformation temperatures A s and A f shifted to higher temperature after electron irradiation, the martensite was stabilized. The electron irradiation effect can be easily eliminated by one thermal cycle. The shifts of the transformation temperatures can be explained from the change of potential energy barrier and coherency energy between parent phase and martensite after irradiation

Tensile behavior of irradiated manganese-stabilized stainless steel

Energy Technology Data Exchange (ETDEWEB)

Klueh, R.L. [Oak Ridge National Lab., TN (United States)

1996-10-01

Tensile tests were conducted on seven experimental, high-manganese austenitic stainless steels after irradiation up to 44 dpa in the FFTF. An Fe-20Mn-12Cr-0.25C base composition was used, to which various combinations of Ti, W, V, B, and P were added to improve strength. Nominal amounts added were 0.1% Ti, 1% W, 0.1% V, 0.005% B, and 0.03% P. Irradiation was carried out at 420, 520, and 600{degrees}C on the steels in the solution-annealed and 20% cold-worked conditions. Tensile tests were conducted at the irradiation temperature. Results were compared with type 316 SS. Neutron irradiation hardened all of the solution-annealed steels at 420, 520, and 600{degrees}C, as measured by the increase in yield stress and ultimate tensile strength. The steel to which all five elements were added to the base composition showed the least amount of hardening. It also showed a smaller loss of ductility (uniform and total elongation) than the other steels. The total and uniform elongations of this steel after irradiation at 420{degrees}C was over four times that of the other manganese-stabilized steels and 316 SS. There was much less difference in strength and ductility at the two higher irradiation temperatures, where there was considerably less hardening, and thus, less loss of ductility. In the cold-worked condition, hardening occured only after irradiation at 420{degrees}C, and there was much less difference in the properties of the steels after irradiation. At the 420{degrees}C irradiation temperature, most of the manganese-stabilized steels maintained more ductility than the 316 SS. After irradiation at 420{degrees}C, the temperature of maximum hardening, the steel to which all five of the elements were added had the best uniform elongation.

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

Energy Technology Data Exchange (ETDEWEB)

Roy, R B; Solly, B

1966-12-15

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

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

Science.gov (United States)

Kobayashi, Koji; Akiyama, Yoko; Nishijima, Shigehiro

2017-09-01

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

Characterisation of high-temperature damage mechanisms of oxide dispersion strengthened (ODS) ferritic steels

International Nuclear Information System (INIS)

Salmon-Legagneur, Hubert

2017-01-01

The development of the fourth generation of nuclear power plants relies on the improvement of cladding materials, in order to achieve resistance to high temperature, stress and irradiation dose levels. Strengthening of ferritic steels through nano-oxide dispersion allows obtaining good mechanical strength at high temperature and good resistance to irradiation induced swelling. Nonetheless, studies available from open literature evidenced an unusual creep behavior of these materials: high anisotropy in time to rupture and flow behavior, low ductility and quasi-inexistent tertiary creep stage. These phenomena, and their still unclear origin are addressed in this study. Three 14Cr ODS steels rods have been studied. Their mechanical behavior is similar to those of other ODS steels from open literature. During creep tests, the specimens fractured by through crack nucleation and propagation from the lateral surfaces, followed by ductile tearing once the critical stress intensity factor was reached at the crack tip. Tensile and creep properties did not depend on the chemical environment of specimens. Crack propagation tests performed at 650 C showed a low value of the stress intensity factor necessary to start crack propagation. The cracks followed an intergranular path through the smaller-grained regions, which partly explains the anisotropy of high temperature strength. Notched specimens have been used to study the impact of the main loading parameters (deformation rate, temperature, stress triaxiality) on macroscopic crack initiation and stable propagation, from the central part of the specimens. These tests allowed revealing cavities created during high temperature loading, but unexposed to the external environment. These cavities showed a high chemical reactivity of the free surfaces in this material. The performed tests also evidenced different types of grain boundaries, which presented different damage development behaviors, probably due to differences in local

High-light damage in air-dry thalli of the old forest lichen Lobaria pulmonaria - interactions of irradiance, exposure duration and high temperature

International Nuclear Information System (INIS)

Gauslaa, Y.; Solhaug, K.A.

1999-01-01

High-light damage in air-dry thalli of Lobaria pulmonaria were measured in the laboratory as reductions in maximal PSII efficiency (FV/FM) after a 48 h recovery in a hydrated state at low light to account for permanent damage. Thalli treated with the lowest light dose (90 mol photons m −2 ) recovered normal FV/FM-values with increasing irradiances (400–700 nm) up to 1000 µmol photons m −2 s −1 . Doubling this dose lowered the threshold level for damage from 1000 to 320 µmol photons m −2 s −1 , and reduced FV/FM at 1000 µmol photons m −2 s −1 by more than 50%. A second doubling of the dose to 360 mol photons m −2 caused damage at 200 µmol photons m −2 s −1 , and a nearly complete cessation of PSII efficiency occurred at 1000 µmol photons m −2 s −1 . No reciprocity of irradiance and duration of illumination for PSII function was found. The measured time-dependent decrease in FV/FM was remarkably similar for the naturally coupled, but artificially separated, light and temperature factors. Therefore, the damage of high light on desiccated L. pulmonaria seemed to be an additive effect of high irradiance and high temperatures. Air-dry thalli were highly heat susceptible, being affected already at temperatures around 40 °C. Logging operations in forests are likely to raise the solar radiation at remaining lichen sites to destructive levels. (author)

Minimizing material damage using low temperature irradiation

International Nuclear Information System (INIS)

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

2012-01-01

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

Effect of periodic temperature variations on the microstructure of neutron-irradiated metals

DEFF Research Database (Denmark)

Zinkle, S.J.; Hashimoto, N.; Hoelzer, D.T.

2002-01-01

Specimens of pure copper, a high purity austenitic stainless steel, and V–4Cr–4Ti were exposed to eight cycles of either constant temperature or periodic temperature variations during neutron irradiation in the High Flux Isotopes Reactor to a cumulative damage level of 4–5 displacements per atom.......-induced microstructural features consisted of dislocation loops, stacking fault tetrahedra and voids in the stainless steel, Ti-rich precipitates in the V alloy, and voids (along with a low density of stacking fault tetrahedra) in copper.......Specimens of pure copper, a high purity austenitic stainless steel, and V–4Cr–4Ti were exposed to eight cycles of either constant temperature or periodic temperature variations during neutron irradiation in the High Flux Isotopes Reactor to a cumulative damage level of 4–5 displacements per atom....... Specimens exposed to periodic temperature variations experienced a low temperature (360 °C) during the initial 10% of accrued dose in each of the eight cycles, and a higher temperature (520 °C) during the remaining 90% of accrued dose in each cycle. The microstructures of the irradiated stainless steel...

Investigation of hardening behavior in Xe ion-irradiated Zr–1Nb

Energy Technology Data Exchange (ETDEWEB)

Yan, Chunguang [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China); China Institute of Atomic Energy, Beijing 102413 (China); Wang, Rongshan [Life Management Technology Center, Suzhou Nuclear Power Research Institute, Suzhou 215004 (China); Dai, Xianyuan [Fujian Fuqing Nuclear Power Co., Ltd., Fuqing 350318 (China); Wang, Yanli, E-mail: wangyl@ustb.edu.cn [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China); Wang, Xitao [Collaborative Innovation Center of Steel Technology, University of Science and Technology Beijing, Beijing 100083 (China); Bai, Guanghai; Zhang, Yanwei [Life Management Technology Center, Suzhou Nuclear Power Research Institute, Suzhou 215004 (China)

2016-05-15

Irradiation hardening behavior of Zr–1Nb was examined by nanoindentation, slow positron annihilation technique, transmission electron microscopy and coplanar extremely asymmetric X-ray diffraction technique. Samples were irradiated at a dose rate of 2.78 × 10{sup −4} dpa/s to peak doses of 0.15, 0.5, 1.5 and 2.5 dpa with 6.37 MeV Xe{sup 26+} ion beam at room temperature. The increase in hardness as a function of dose followed a power law expression with the exponent of 0.46. With increasing irradiation dose, more mono-, di- and trivacancies were induced, but their concentration remained constant once formed due to the equilibrium between the formation and recombination of vacancy type clusters during irradiation. Meanwhile, the dislocation loops were also introduced and their linear density increased with dose. The dislocation loops played an important role in the irradiation hardening behavior. But the exact contribution of each microstructural components to the overall hardness still needs further study.

Nonequilibrium segregation and phase instability in alloy films during elevated-temperature irradiation in a high-voltage electron microscope

Science.gov (United States)

Lam, N. Q.; Okamoto, P. R.

1984-05-01

The effects of defect-production rate gradients, caused by the radial nonuniformity in the electron flux distribution, on solute segregation and phase stability in alloy films undergoing high-voltage electron-microscope (HVEM) irradiation at high temperatures are assessed. Two-dimensional (axially symmetric) compositional redistributions were calculated, taking into account both axial and transverse radial defect fluxes. It was found that when highly focused beams were employed radiation-induced segregation consisted of two stages: dominant axial segregation at the film surfaces at short irradiation times and competitive radial segregation at longer times. The average alloy composition within the irradiated region could differ greatly from that irradiated with a uniform beam, because of the additional atom transport from or to the region surrounding the irradiated zone under the influence of radial fluxes. Damage-rate gradient effects must be taken into account when interpreting in-situ HVEM observations of segregation-induced phase instabilities. The theoretical predictions are compared with experimental observations of the temporal and spatial dependence of segregation-induced precipitation in thin films of Ni-Al, Ni-Ge and Ni-Si solid solutions.

Using education on irradiated foods to change behavior of Korean elementary, middle, and high school students

OpenAIRE

Han, Eunok; Kim, Jaerok; Choi, Yoonseok

2014-01-01

BACKGROUND/OBJECTIVES Educational interventions targeted food selection perception, knowledge, attitude, and behavior. Education regarding irradiated food was intended to change food selection behavior specific to it. SUBJECTS AND METHODS There were 43 elementary students (35.0%), 45 middle school students (36.6%), and 35 high school students (28.5%). The first step was research design. Educational targets were selected and informed consent was obtained in step two. An initial survey was cond...

Irradiation behaviors of coated fuel particles, (4)

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kashimura, Satoru; Ogawa, Toru; Ikawa, Katsuichi; Iwamoto, Kazumi; Ishimoto, Kiyoshi

1981-09-01

Loose coated fuel particles prepared in confirmity to a preliminary design for the multi-purpose VHTR in fiscal 1972 - 1974 were irradiated by 73F - 12A capsule in JMTR. Main purpose for this irradiation experiment was to examine irradiation stability of the candidate TRISO coated fuel particles for the VHTR. Also the coated particles possessing low-density kernel (90%TD), highly anisotropic OLTI-PyC and ZrC coating layer were loaded with the candidate particles in this capsule. The coated particles were irradiated up to 1.5 x 10 21 n/cm 2 of fast neutron fluence (E > 0.18 MeV) and 3.2% FIMA of burnup. In the post irradiation examination it was observed that among three kinds of TRISO particles exposed to irradiation corresponding to the normal operating condition of the VHTR ones possessing poor characteristics of the coating layers did not show a good stability. The particles irradiated under abnormally high temperature condition (> 1800 0 C) revealed 6.7% of max. EOL failure fraction (95% confidence limit). Most of these particles were failed by the ameoba effect. Furthermore, among four kinds of the TRISO particles exposed to irradiation corresponding to the transient condition of the VHTR (--1500 0 C) the two showed a good stability, while the particles possessing highly anisotropic OLTI-PyC or poorly characteristic coating layers were not so good. (author)

Irradiation behavior of uranium oxide - Aluminum dispersion fuel

International Nuclear Information System (INIS)

Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

1996-01-01

An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

Irradiation behavior of uranium oxide-aluminum dispersion fuel

International Nuclear Information System (INIS)

Hofman, G.L.; Rest, J.; Snelgrove, J.L.

1996-01-01

An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

Role of the irradiation temperature on the modifications of swift-heavy-ion irradiated polyethylene

International Nuclear Information System (INIS)

Melot, M.; Ngono-Ravache, Y.; Balanzat, E.

2003-01-01

The damage processes triggered by swift heavy ions, SHI, can be very different to those induced by classical low ionising particles. This is due to the very high electronic stopping power, (dE/dx) e , of SHI. This paper concerns the effects of SHI on polyethylene, PE. In PE, low (dE/dx) e irradiations induce crosslinking and in-chain double bond formation. At high (dE/dx) e , the creation yield of vinyl groups becomes significant. Above a (dE/dx) e threshold, alkyne and allene groups appear. We present results on low temperature irradiations that bring new enlightenment on the damage process by preventing the migration of radiation-induced radicals and molecules. Two SHI specific modifications are studied: vinyl groups and alkyne end groups. We have irradiated PE films with oxygen and sulphur beams at 13.6 and 11.2 MeV/amu, respectively. The modifications were followed by in situ infrared spectroscopy (FTIR). We have performed irradiations at 8 and 290 K. The samples irradiated at 8 K have been annealed up to 290 K for investigating the effect of radical migration. Lowering the irradiation temperature to 8 K increases the creation yield of vinyl groups and alkyne end groups. The enhancement factor between 290 and 8 K is around three. Consequently the formation of defects specific to SHI irradiations is sensitive to radical migration and hence requires some time. During annealing, the alkyne concentration remains stable indicating that the creation of this group cannot be induced by radical recombination. The annealing spectra of vinyl groups are more complex

Stability of uranium silicides during high energy ion irradiation

International Nuclear Information System (INIS)

Birtcher, R.C; Wang, L.M.

1991-11-01

Changes induced by 1.5 MeV Kr ion irradiation of both U 3 Si and U 3 Si 2 have been followed by in situ transmission electron microscopy. When irradiated at sufficiently low temperatures, both alloys transform from the crystalline to the amorphous state. When irradiated at temperatures above the temperature limit for ion beam amorphization, both compounds disorder with the Martensite twin structure in U 3 Si disappearing from view in TEM. Prolonged irradiation of the disordered crystalline phases results in nucleation of small crystallites within the initially large crystal grains. The new crystallites increase in number during continued irradiation until a fine grain structure is formed. Electron diffraction yields a powder-like diffraction pattern that indicates a random alignment of the small crystallites. During a second irradiation at lower temperatures, the small crystallizes retard amorphization. After 2 dpa at high temperatures, the amorphization dose is increased by over twenty times compared to that of initially unirradiated material

Application of accelerated evaluation method of alteration temperature and constant dose rate irradiation on bipolar linear regulator LM317

International Nuclear Information System (INIS)

Deng Wei; Wu Xue; Wang Xin; Zhang Jinxin; Zhang Xiaofu; Zheng Qiwen; Ma Wuying; Lu Wu; Guo Qi; He Chengfa

2014-01-01

With different irradiation methods including high dose rate irradiation, low dose rate irradiation, alteration temperature and constant dose rate irradiation, and US military standard constant high temperature and constant dose rate irradiation, the ionizing radiation responses of bipolar linear regulator LM317 from three different companies were investigated under the operating and zero biases. The results show that compared with constant high temperature and constant dose rate irradiation method, the alteration temperature and constant dose rate irradiation method can not only very rapidly and accurately evaluate the dose rate effect of three bipolar linear regulators, but also well simulate the damage of low dose rate irradiation. Experiment results make the alteration temperature and constant dose rate irradiation method successfully apply to bipolar linear regulator. (authors)

Effect of heat treatments and pre-irradiation on the corrosion at high pressure and temperature of pressure tube Zr-2.5Nb

International Nuclear Information System (INIS)

Olmedo, A.M; Bordoni, R

2006-01-01

Zr and its alloys are widely used in the nuclear industry. The pressure tubes (PT) of CANDU reactors are made using the alloy Zr-2.5Nb because of its high mechanical resistance and its good behavior under aqueous corrosion at high pressure and temperature. The resulting microstructure after the production process consists of elongated grains of Zr-α with hcp structure (≤ 1%Nb) , approximately 0.3-0.5 μ thick, surrounded by a metastable β Zr bcc phase rich in Nb (≅20%). Temperatures of less than 600 C produce an evolution of the β Zr phase towards an equilibrium structure consisting of β Nb (≅85% of Nb) and α Zr phases. The conditions for the evolution of this phase may involve a Nb-poor intermediate phase called ω and the enrichment of Nb for the rest. The radiation, particularly the flow of rapid neutrons (E ≥1Mev), modifies the microstructure of the material inducing β-Nb precipitates in the grains of the α-Zr phase. The temperature and radiation induced changes in the microstructure modify the resistance to the corrosion of the Zr-2.5Nb of the PT. This work studies the behavior to corrosion of Zr-2.5Nb coupons of (PT) with aging heat treatments (HT) of (PT) at 400 o C for 72 and 1000 hours and at 500 o C for 2.6 and 10 hours. The results indicated that the corrosion speed of the test pieces with HT is less than for those without HT and that the decrement of the corrosion speed is much more pronounced in the first 2.6 h of aging at 500 o C and in the first 72 h at 400 o C. Also the greater the duration of the HT, the better the behavior to corrosion. This decreased corrosion speed is associated with the modifications produced in the microstructure of the PT material due to the effect of the HT, which cause the decomposition of the βZr phase and lead to a microstructure that is much closer to one of equilibrium. That is, the closer the material's microstructure is to equilibrium the greater will be the resistance to its corrosion. This effect

The pupal body temperature and inner space temperature of cocoon under microwave irradiation

International Nuclear Information System (INIS)

Kagawa, T.

1996-01-01

The temperature of pupal surface,body and inner space of cocoon on cocoon drying of microwave irradiation was investigated to make clear the effect of temperature with pupa and cocoon shell. After pupal surface temperature and body temperature were risen rapidly in early irradiation and slowly thereafter, these were done fast again. Then these rising degrees fell. The variation of inner space temperature consists three terms: as the first stage of rapidly rising on early irradiation, the second stage of slowly doing and the third stage of fast doing again in temperature. In the first stage and the second stage, the higher the temperature of sending air during irradiation was, the shorter the term was and the higher the reached temperature was. The surface, pupal body and inner space have reached higher temperature than the sending air before cocoon drying was over

Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA

International Nuclear Information System (INIS)

Huang, F.H.

1992-02-01

Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken

Artificial neural networks in prediction of mechanical behavior of concrete at high temperature

International Nuclear Information System (INIS)

Mukherjee, A.; Nag Biswas, S.

1997-01-01

The behavior of concrete structures that are exposed to extreme thermo-mechanical loading is an issue of great importance in nuclear engineering. The mechanical behavior of concrete at high temperature is non-linear. The properties that regulate its response are highly temperature dependent and extremely complex. In addition, the constituent materials, e.g. aggregates, influence the response significantly. Attempts have been made to trace the stress-strain curve through mathematical models and rheological models. However, it has been difficult to include all the contributing factors in the mathematical model. This paper examines a new programming paradigm, artificial neural networks, for the problem. Implementing a feedforward network and backpropagation algorithm the stress-strain relationship of the material is captured. The neural networks for the prediction of uniaxial behavior of concrete at high temperature has been presented here. The results of the present investigation are very encouraging. (orig.)

High temperature creep strength of Advanced Radiation Resistant Oxide Dispersion Strengthened Steels

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling well occurred to 120dpa at high temperatures and this leads the decrease of the mechanical properties and dimensional stability. Compared to this, ferritic/martensitic steel is a good solution because of excellent thermal conductivity and good swelling resistance. Unfortunately, the available temperature range of ferritic/martensitic steel is limited up to 650 .deg. C. ODS steel is the most promising structural material because of excellent creep and irradiation resistance by uniformly distributed nano-oxide particles with a high density which is extremely stable at the high temperature in ferritic/martensitic matrix. In this study, high temperature strength of advanced radiation resistance ODS steel was investigated for the core structural material of next generation nuclear systems. ODS martensitic steel was designed to have high homogeneity, productivity and reproducibility. Mechanical alloying, hot isostactic pressing and hot rolling processes were employed to fabricate the ODS steels, and creep rupture test as well as tensile test were examined to investigate the behavior at high temperatures. ODS steels were fabricated by a mechanical alloying and hot consolidation processes. Mechanical properties at high temperatures were investigated. The creep resistance of advanced radiation resistant ODS steels was more superior than those of ferritic/ martensitic steel, austenitic stainless steel and even a conventional ODS steel.

Mechanical behavior of high strength ceramic fibers at high temperatures

Science.gov (United States)

Tressler, R. E.; Pysher, D. J.

1991-01-01

The mechanical behavior of commercially available and developmental ceramic fibers, both oxide and nonoxide, has been experimentally studied at expected use temperatures. In addition, these properties have been compared to results from the literature. Tensile strengths were measured for three SiC-based and three oxide ceramic fibers for temperatures from 25 C to 1400 C. The SiC-based fibers were stronger but less stiff than the oxide fibers at room temperature and retained more of both strength and stiffness to high temperatures. Extensive creep and creep-rupture experiments have been performed on those fibers from this group which had the best strengths above 1200 C in both single filament tests and tests of fiber bundles. The creep rates for the oxides are on the order of two orders of magnitude faster than the polymer derived nonoxide fibers. The most creep resistant filaments available are single crystal c-axis sapphire filaments. Large diameter CVD fabricated SiC fibers are the most creep and rupture resistant nonoxide polycrystalline fibers tested to date.

Influence of high energy electron irradiation and gamma irradiation on the osmotic resistance of human erythrocyte membranes

International Nuclear Information System (INIS)

Catana, D.; Hategan, Alina; Moraru, Rodica; Popescu, Alina; Morariu, V. V.

1998-01-01

The effects of 5 MeV electrons and of gamma irradiation at 0 deg. C on the osmotic fragility of human erythrocyte membranes are presented. Both electron and gamma radiation in the range 0-400 Gy induced no hemolysis indicating that the membrane modifications due to radiation interaction do not reach a critical point as to cause swelling of the cells and subsequent lysis. The osmotic stress experiments performed after irradiation showed that the gamma irradiated erythrocytes exhibited an almost similar sigmoidal behavior for all irradiation doses, whereas the electron irradiated samples showed a much larger increase in hemolysis degree and, in the case of a given electron dose (100 Gy), the hemolysis was found much smaller than for the control sample (a similar behavior of the erythrocytes was found in the case of microwave irradiation at temperatures under 0 deg. C). Our experimental data suggest that electron radiation and gamma radiation have different impacts on the erythrocyte membrane fluidity, involving, probably, the different rate of energy deposition in the samples and the direct interaction of electrons with the erythrocyte membranes. (authors)

Corrosion behavior of low energy, high temperature nitrogen ion ...

Indian Academy of Sciences (India)

Corrosion behavior of low energy, high temperature nitrogen ion-implanted AISI 304 stainless steel. M GHORANNEVISS1, A SHOKOUHY1,∗, M M LARIJANI1,2,. S H HAJI HOSSEINI 1, M YARI1, A ANVARI4, M GHOLIPUR SHAHRAKI1,3,. A H SARI1 and M R HANTEHZADEH1. 1Plasma Physics Research Center, Science ...

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

Energy Technology Data Exchange (ETDEWEB)

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 experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of 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

Irradiation behavior of experimental miniature uranium silicide fuel plates

International Nuclear Information System (INIS)

Hofman, Gerard L.; Neimark, L.A.; Mattas, R.F.

1983-01-01

Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 10 20 cm -3 , far short of he approximately 20 x 10 20 cm -3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U 3 Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 10 20 cm -3 . This fuel swelling will likely result in unacceptably large plate-thickness increases. The U 3 Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 10 21 fission/cm 3 . The interdiffusion between fuel and matrix

Some behavioral aspects of adult rats irradiated prenatally

International Nuclear Information System (INIS)

Vekovishcheva, O.Yu.; Blagova, O.E.; Borovitskaya, A.E.; Evtushenko, V.I.; Khanson, K.P.

1992-01-01

This is a study of the effects of prenatal irradiation on the behavior of rats. The experiments were performed on 42 eighteen month old rats of both sexes. Eight of the males and thirteen females had been irradiated prenatally. The results of this experiment indicated that in general, the activation of behavior, the appearance of aggression and the increase in chaos along with the presence of behavior poses were typical of the suppressed condition of the prenatal irradiated animal. Also, among prenatally irradiated animals, there was a greater degree of anxiety, a slow rate of adjustment to unfamiliar situations and unfriendly relationships between animals of the same sex. These results were compared with the results of behavioral experiments on irradiated adult rats

Low temperature modification of gamma-irradiation effect on peas. II.Low temperature effect on the radio-sensitivity and the chlorophyll mutations

International Nuclear Information System (INIS)

Najdenova, N.; Vasileva, M.

1976-01-01

Dry pea seeds of cv.Ramonskii 77 with 11-12% moisture were γ-irradiated by 60 Co in doses 5, 15, 20 and 30 krad. Low temperature (-78 deg C) was effected in the form of dry ice for a 24 h period prior to, at the time of and post irradiation. As control were used: (a) dry non-irradiated seeds, stored at room temperature; (b) non-irradiated seeds subjected to low temperature (-78 deg C) for a 24 h period. and (c) seeds irradiated by the named doses, stored at room temperature until the time of irradiation. Treated and control seeds were sown in the field. Germination, survival rate and sterility were recorded in M 1 , while in M 2 chlorophyll mutations were scored. Results obtained showed that low temperature modification effect on the various irradiation doses depended on the time of its application; low temperature (-78 deg C) treatment prior to seed irradiation with doses 15, 20 and 30 krad increased germination percentage, plant survival and yield components in M 1 . The post-irradiation treatment did not have a significant effect on gamma-rays; highest protection effect was obtained in case seeds were irradiated at low temperature and then received supplementary treatment at high temperature. In this way the damaging effect of radiation was reduced to a maximum degree; low temperature treatment prior to irradiation with doses of 15 and 20 krad or at the time of irradiation with doses of 15, 20 and 30 krad resulted in a considerably wider chlorophyll mutation spectrum. (author)

Nonequilibrium segregation and phase instability in alloy films during elevated-temperature irradiation in a high-voltage electron microscope

International Nuclear Information System (INIS)

Lam, N.Q.; Okamoto, P.R.

1984-05-01

The effects of defect-production rate gradients, caused by the radial nonuniformity in the electron flux distribution, on solute segregation and phase stability in alloy films undergoing high-voltage electron-microscope (HVEM) irradiation at high temperatures are assessed. Two-dimensional (axially symmetric) compositional redistributions were calculated, taking into account both axial and transverse radial defect fluxes. It was found that when highly focused beams were employed radiation-induced segregation consisted of two stages: dominant axial segregation at the film surfaces at short irradiation times and competitive radial segregation at longer times. The average alloy composition within the irradiated region could differ greatly from that irradiated with a uniform beam, because of the additional atom transport from or to the region surrounding the irradiated zone under the influence of radial fluxes. As a result, damage-rate gradient effects must be taken into account when interpreting in-situ HVEM observations of segregation-induced phase instabilities. The theoretical predictions are compared with experimental observations of the temporal and spatial dependence of segregation-induced precipitation in thin films of Ni-Al, Ni-Ge and Ni-Si solid solutions

Further study of the glassy low-temperature properties of irradiated crystalline quartz: neutron and electron irradiation

International Nuclear Information System (INIS)

Laermans, C.; Daudin, B.

1979-01-01

Recently it has been shown that a quartz crystal after light fast neutron irradiation shows low temperature hypersonic properties which are similar to those found in glasses although the sample was still crystalline. Additional measurements have been carried out in the neutron-irradiated sample and a sample irradiated with high energy electrons has also been investigated. (Fast neutron dose 6 x 10 18 n/cm 2 , 2 MeV electron dose 3 x 10 19 e/cm 2 ). A magnetic field up to 1.5 T was found to have no influence in the hypersonic saturation behaviour of the neutron-irradiated sample (9 GHz, 1.65 K) and thermal conductivity measurements are consistent with a number of two level systems (2 LS) an order of magnitude lower than in vitreous silica as found before. Low temperature hypersonic measurements as a function of acoustic intensity and temperature as well as thermal conductivity measurements give no evidence for the presence of 2 LS in the electron irradiated sample. Considering the damage created in both samples this indicates that 2 LS are probably not related to point defects

Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

International Nuclear Information System (INIS)

Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

1999-01-01

This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

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

International Nuclear Information System (INIS)

Blondel, Antoine

2013-01-01

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

Behavior of pre-irradiated fuel under a simulated RIA condition

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide

1994-07-01

This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-3, conducted in the Nuclear Safety Research Reactor (NSSR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and analyses, interpretations, and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 19.6MWd/kgU with average linear heat rate of 25.3 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a deposited energy of 174±6 cal/g·fuel and a peak fuel enthalpy of 130±5 cal/g·fuel under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The cladding surface temperature increased to only 150degC, and the test resulted in slight fuel deformation and no fuel failure. An estimated rod-average fission gas release during the transient was about 2.2%. Through the detailed fuel examinations, the information concerning microstructural change in the fuel pellets were also obtained. (author)

Behavior of optical fibers under heavy irradiation

International Nuclear Information System (INIS)

Kakuta, T.; Sagawa, T.

1998-01-01

Several kinds of optical diagnostics are planned in a fusion reactor. Complicated optical systems such as periscopes are thought to be primary candidates for optical measurements, especially for visible wavelengths. However, optical fibers have several advantages over such optical systems. Also, the optical fibers could be a far better transmission line for signals under a high electromagnetic field. However, they have been considered vulnerable to heavy irradiation. In this study, several kinds of optical fibers were irradiated in the JMTR fission reactor. The optical transmissivity in fibers was measured in situ during fast neutron and gamma irradiation, up to doses of 2 x 10 24 n m -2 and 5 x 10 9 Gy, respectively. The irradiation temperature ranged from 300 to 700 K. For pure ionizing irradiation environments, some methods for improving the radiation resistance of optical fibers were indicated. The results showed that effects of the irradiation associated with fast neutrons would be different from the effects of pure ionizing irradiation. Some fibers were found to withstand the heavy irradiation, especially in an infrared wavelength range. (orig.)

Irradiation performance of HTGR fertile fuel in HFIR target capsules HT-12 through HT-15. Part I. Experiment description and fission product behavior

International Nuclear Information System (INIS)

Kania, M.J.; Lindemer, T.B.; Morgan, M.T.; Robbins, J.M.

1977-02-01

Sixteen types of Biso-coated designs, on ThO 2 kernels, were irradiated in High Flux Isotope Reactor target capsules HT-12 through HT-15. The report addresses the description of the experiment and extensive postirradiation analyses and experiments to determine fertile-particle burnup, fuel coating failures, and fission product behavior. Several low-temperature isotropic (LTI) pyrocarbon coatings, which ''survived'' according to visual inspection, were shown to have developed permeability during irradiation. These particles were irradiated at temperatures approximately equal to 1250 0 C and to burnups equal to or greater than 8 percent fission per initial heavy-metal atom (FIMA). No evidence of permeability was found in similar particles irradiated at temperatures approximately equal to 1550 0 C and burnups approximately equal to 16 percent FIMA. Failures due to permeability were not detectable by visual inspection but required a more extensive investigation by the 1000 0 C gaseous chlorine leaching technique. Maximum particle surface operating temperatures were found to be approximately 300 0 C in excess of design limits of 900 0 C (low-temperature magazines) and 1250 0 C (high-temperature magazines). The extremes of high temperatures and fast neutron fluences up to 1.6 x 10 22 neutrons/cm 2 produced severe degradation and swelling of the Poco graphite magazines and sample holders

Irradiation creep and growth behavior of Zircaloy-4 inner shell of HANARO

Energy Technology Data Exchange (ETDEWEB)

Chung, Jong-Ha; Cho, Yeong-Garp; Kim, Jong-In [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

2012-03-15

The inner shell of the reflector vessel of HANARO was made of Zircaloy-4 rolled plate. Zircaloy-4 rolled plate shows highly anisotropic behavior by fast neutron irradiation. This paper describes the analysis method for the irradiation induced creep and growth of the inner shell of HANARO. The anisotropic irradiation creep behavior was modeled as uniaxial strain-hardening power law modified by Hill's stress potential and the anisotropic irradiation growth was modeled by using volumetric swelling with anisotropic strain rate. In this study, the irradiation induced creep and growth behavior of the inner shell of the HANARO reflector vessel was re-evaluated. The rolling direction, the fast neutron flux, and the boundary conditions were applied with the same conditions as the actual inner shell. Analysis results show that deformation of the inner shell due to irradiation does not raise any problem for the lifetime of HANARO. (author)

Thermoluminescence analysis of co-doped NaCl at low temperature irradiations

Energy Technology Data Exchange (ETDEWEB)

Cruz-Zaragoza, E., E-mail: ecruz@nucleares.unam.m [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, A.P. 70-543, 04510 Mexico D.F. (Mexico); Ortiz, A. [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, A.P. 70-543, 04510 Mexico D.F. (Mexico); Unidad Profesional Interdisciplinaria de Ingenieria y Tecnologias Avanzadas, IPN, Av. Instituto Politecnico Nacional 2580, Col. La Laguna Ticoman, 07340 Mexico D.F. (Mexico); Furetta, C. [Touro University Rome, Circne Gianicolense 15-17, 00153 Rome (Italy); Flores J, C.; Hernandez A, J.; Murrieta S, H. [Instituto de Fisica, Universidad Nacional Autonoma de Mexico, A.P. 20-364, 01000 Mexico D.F. (Mexico)

2011-02-15

The thermoluminescent response and kinetics parameters of NaCl, doubly activated by Ca-Mn and Cd-Mn ions, exposed to gamma radiation are analyzed. The doped NaCl samples were irradiated at relative low temperature, i.e. at the liquid nitrogen temperature (LNT) and at dry ice temperature (DIT), and the glow curves obtained after 2 Gy of gamma irradiation were analyzed using the computerized glow curve deconvolution (CGCD). An evident variation in the glow curve structure after LNT and DIT was observed. It seems that different kinds of trapping levels are activated at relative low temperature. The original two prominent peaks in compositions A (Ca,Mn) and B (Ca,Mn) have been changed in only one main peak with satellites in the low temperature side of the glow curves. In compositions C (Cd,Mn) and D (Cd,Mn), low temperature peaks become stronger and prominent than the high temperature peaks; this effect could be explained considering that the trapping probability for low temperature traps, the one very close to the conduction band, is enhanced by low temperatures during irradiation.

Temperature evaluation of dental implant surface irradiated with high-power diode laser.

Science.gov (United States)

Rios, F G; Viana, E R; Ribeiro, G M; González, J C; Abelenda, A; Peruzzo, D C

2016-09-01

The prevalence of peri-implantitis and the absence of a standard approach for decontamination of the dental implant surface have led to searches for effective therapies. Since the source of diode lasers is portable, has reduced cost, and does not cause damage to the titanium surface of the implant, high-power diode lasers have been used for this purpose. The effect of laser irradiation on the implants is the elevation of the temperature surface. If this elevation exceeds 47 °C, the bone tissue is irreversibly damaged, so for a safety therapy, the laser parameters should be controlled. In this study, a diode laser of GaAsAl was used to irradiate titanium dental implants, for powers 1.32 to 2.64 W (real) or 2.00 to 4.00 W (nominal), in continuous/pulsed mode DC/AC, with exposure time of 5/10 s, with/without air flow for cooling. The elevation of the temperature was monitored in real time in two positions: cervical and apical. The best results for decontamination using a 968-nm diode laser were obtained for a power of 1.65 and 1.98 W (real) for 10 s, in DC or AC mode, with an air flow of 2.5 l/min. In our perspective in this article, we determine a suggested approach for decontamination of the dental implant surface using a 968-nm diode laser.

Influence of the irradiation temperature on the free-radical response of alanine

International Nuclear Information System (INIS)

Wieser, A.; Siegele, R.; Regulla, D.F.

1989-01-01

GSF operates the only IAEA high-level dosimetry reference laboratory and, as a joint project with the IAEA, the International Dose Assurance Service (IDAS). Dosimetry is based on long-lived free radicals in organic alanine induced by ionizing radiation and readout by ESR spectroscopy. The thermal time response of the radical concentration in alanine is fairly constant after irradiation provided that the alanine samples are stored at temperatures below 50 0 C. By contrast, a positive temperature coefficient had earlier been found at GSF for the production rate of alanine radicals, for irradiation temperatures between 0 and 50 0 C. This effect has to be considered for reference dosimetry in radiation processing. Radiation processing is also of interest at irradiation temperatures below 0 0 C. The present study describes experiments on the influence of irradiation temperatures between +50 and -100 0 C. Comparison is made between the present and earlier results, in the overlapping temperature range. An empirical function is proposed for the temperature coefficient based on the experimental data. (author)

National Low-Temperature Neutron-Irradiation Facility

International Nuclear Information System (INIS)

Coltman, R.R. Jr.; Klabunde, C.E.; Young, F.W. Jr.

1983-08-01

The Materials Sciences Division of the United States Department of Energy will establish a National Low Temperature Neutron Irradiation Facility (NLTNIF) which will utilize the Bulk Shielding Reactor (BSR) located at Oak Ridge National Laboratory. The facility will provide high radiation intensities and special environmental and testing conditions for qualified experiments at no cost to users. This report describes the planned experimental capabilities of the new facility

Computer simulation of defect behavior under fusion irradiation environments

International Nuclear Information System (INIS)

Muroga, T.; Ishino, S.

1983-01-01

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

Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement en fluage des alliages de zircomium irradies

Energy Technology Data Exchange (ETDEWEB)

Ribis, J

2007-12-15

The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

Neutron Irradiation Tests of Calibrated Cryogenic Sensors at Low Temperatures

CERN Document Server

Junquera, T; Thermeau, J P; Casas-Cubillos, J

1998-01-01

This paper presents the advancement of a program being carried out in view of selecting the cryogenic temperature sensors to be used in the LHC accelerator. About 10,000 sensors will be installed around the 26.6 km LHC ring, and most of them will be exposed to high radiation doses during the accelerator lifetime. The following thermometric sensors : carbon resistors, thin films, and platinum resistors, have been exposed to high neutron fluences (>10$^15$ n/cm$^2$) at the ISN (Grenoble, France) Cryogenic Irradiation Test Facility. A cryostat is placed in a shielded irradiation vault where a 20 MeV deuteron beam hits a Be target, resulting in a well collimated and intense neutron beam. The cryostat, the on-line acquisition system, the temperature references and the main characteristics of the irradiation facility are described. The main interest of this set-up is its ability to monitor online the evolution of the sensors by comparing its readout with temperature references that are in principle insensitive to t...

Superconductivity degradation in Gd-containing high temperature superconductors (HTSC) under thermal neutron irradiation

International Nuclear Information System (INIS)

Petrov, A.; Kudrenitskis, I.; Makletsov, A.; Arhipov, A.; Karklin, N.

1999-01-01

The physical properties of ordered crystals are extremely sensitive to the degree of order in the distribution of the various kinds of atoms over the corresponding sites in the crystal lattice. An increasingly popular means of creating disordered states is to use nuclear radiation. The type of radiation defects which appear and the nature and degree of the structural changes in ordered crystals depend on the kind of radiation and the fluence level, the irradiation temperature, the type of crystal structure, the composition and initial disorder of the material, the character of the interatomic forces, etc. There are many such scientific publications where the effects of fast neutron irradiation on high temperature superconductors (HTSC) have been studied in both polycrystalline and single crystalline superconductors. It is known also that the role of thermal neutrons in structural defects forming is negligible in comparison with fast neutrons because of their small (∼0.025 eV) energy. But it is evident enough that in superconductors containing isotopes with large thermal neutron cross sections the important results concerning the role of point defects could be obtained. Such point defects are creating due to soft displacements of isotopes having interacted with thermal neutrons. Such the possibility of creating point defects in solids including HTSC is investigating by several groups (Austria, USA, China, Latvia) and these investigations have found the support in the person of IAEA. In this review the authors consider the changes brought about by thermal-neutron irradiation (E∼0.025 eV) in the structure, superconducting and magnetic properties of gadolinium containing ordered HTSC with the structure 123, whose extreme electric and magnetic properties continue to attract both research and practical interest. All of the studies reviewed have been done on bulk polycrystalline samples RBa 2 Cu 3 O 7-δ (where R - natural mixture of Gd isotopes, 155 Gd, 157 Gd, 160

Influence of High Temperature Treatment on Mechanical Behavior of a Coarse-grained Marble

Science.gov (United States)

Rong, G.; Peng, J.; Jiang, M.

2017-12-01

High temperature has a significant influence on the physical and mechanical behavior of rocks. With increasing geotechnical engineering structures concerning with high temperature problems such as boreholes for oil or gas production, underground caverns for storage of radioactive waste, and deep wells for injection of carbon dioxides, etc., it is important to study the influence of temperature on the physical and mechanical properties of rocks. This paper experimentally investigates the triaxial compressive properties of a coarse-grained marble after exposure to different high temperatures. The rock specimens were first heated to a predetermined temperature (200, 400, and 600 oC) and then cooled down to room temperature. Triaxial compression tests on these heat-treated specimens subjected to different confining pressures (i.e., 0, 5, 10, 15, 20, 25, 30, 35, and 40 MPa) were then conducted. Triaxial compression tests on rock specimens with no heat treatment were also conducted for comparison. The results show that the high temperature treatment has a significant influence on the microstructure, porosity, P-wave velocity, stress-strain relation, strength and deformation parameters, and failure mode of the tested rock. As the treatment temperature gradually increases, the porosity slightly increases and the P-wave velocity dramatically decreases. Microscopic observation on thin sections reveals that many micro-cracks will be generated inside the rock specimen after high temperature treatment. The rock strength and Young's modulus show a decreasing trend with increase of the treatment temperature. The ductility of the rock is generally enhanced as the treatment temperature increases. In general, the high temperature treatment weakens the performance of the tested rock. Finally, a degradation parameter is defined and a strength degradation model is proposed to characterize the strength behavior of heat-treated rocks. The results in this study provide useful data for

Evaluation of Ion Irradiation Behavior of ODS Alloys

Energy Technology Data Exchange (ETDEWEB)

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

2006-08-15

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

Evaluation of Ion Irradiation Behavior of ODS Alloys

International Nuclear Information System (INIS)

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

2006-08-01

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

Optical properties of CsI single crystals irradiated with neutrons at low temperature

International Nuclear Information System (INIS)

Okada, M.; Atobe, K.; Itatani, N.; Ozawa, K.

1998-01-01

Optical properties of the irradiation-induced-defects in neutron-irradiated CsI single crystals have been investigated. The nominally pure CsI crystals are irradiated by reactor fast neutrons (E>0.1 MeV) with a fluence of 1.4 x 10 15 n/cm 2 at 20 K and by γ-rays from 60 Co source to a dose of 1.5 x 10 4 Gy at liquid nitrogen temperature (LNT). After the irradiations, isochronal annealings are performed to investigate the thermal behavior of the defects. The glow peaks of the thermoluminescence (TL) in each sample irradiated with neutrons at 20 K and with γ-rays at LNT are observed at about 100, 160 and 220 K. In the neutron-irradiated samples at 20 K, the emission band at 338 nm is observed at LNT. It is supposed that this emission band occurs by an excitation of γ-rays from 134 Cs, which is radioactivated by thermal neutrons among the reactor radiations. It is confirmed that the temperature dependence of the 338 nm band is similar with that of the emission band due to the self-trapped exciton which is introduced into the non-irradiated samples illuminated by higher energy photons. (orig.)

Optical properties of CsI single crystals irradiated with neutrons at low temperature

Energy Technology Data Exchange (ETDEWEB)

Okada, M. [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.; Nakagawa, M. [Faculty of Education, Kagawa Univ., Takamatsu, Kagawa (Japan); Atobe, K. [Faculty of Science, Naruto Univ. of Education, Naruto, Tokushima (Japan); Itatani, N.; Ozawa, K. [Horiba Ltd., Minamiku, Kyoto (Japan)

1998-05-01

Optical properties of the irradiation-induced-defects in neutron-irradiated CsI single crystals have been investigated. The nominally pure CsI crystals are irradiated by reactor fast neutrons (E>0.1 MeV) with a fluence of 1.4 x 10{sup 15} n/cm{sup 2} at 20 K and by {gamma}-rays from {sup 60}Co source to a dose of 1.5 x 10{sup 4} Gy at liquid nitrogen temperature (LNT). After the irradiations, isochronal annealings are performed to investigate the thermal behavior of the defects. The glow peaks of the thermoluminescence (TL) in each sample irradiated with neutrons at 20 K and with {gamma}-rays at LNT are observed at about 100, 160 and 220 K. In the neutron-irradiated samples at 20 K, the emission band at 338 nm is observed at LNT. It is supposed that this emission band occurs by an excitation of {gamma}-rays from {sup 134}Cs, which is radioactivated by thermal neutrons among the reactor radiations. It is confirmed that the temperature dependence of the 338 nm band is similar with that of the emission band due to the self-trapped exciton which is introduced into the non-irradiated samples illuminated by higher energy photons. (orig.) 13 refs.

Minimizing material damage using low temperature irradiation

Science.gov (United States)

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

2012-08-01

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

Microstructure evolution by neutron irradiation during cyclic temperature variation

International Nuclear Information System (INIS)

Kiritani, M.; Yoshiie, T.; Iseki, M.; Kojima, S.; Hamada, K.; Horiki, M.; Kizuka, Y.; Inoue, H.; Tada, T.; Ogasawara, Y.

1994-01-01

Utilizing a technique to control the temperature which is not influenced by the operation mode of a reactor, an irradiation during which the temperature was alternatively changed several times between two temperatures (T-cycle) has been performed. Some defect structures are understood as combinations of the defect processes at lower and higher temperatures, and some others are understood if the defect processes during the transient between the two temperatures are taken into consideration. However, the most remarkable characteristic of defect processes associated with the temperature variation is the reaction of point defect clusters induced by lower-temperature irradiation at the higher temperature. During lower-temperature irradiation, there is a greater accumulation of vacancy clusters as stacking fault tetrahedra in fcc metals than that of interstitial clusters as dislocation loops. Vacancies evaporated from the vacancy clusters at higher temperature can eliminate interstitial clusters completely, and the repetition of these processes leads to unexpectedly slow defect structure development by T-cycle irradiation. ((orig.))

Burst annealing of high temperature GaAs solar cells

Science.gov (United States)

Brothers, P. R.; Horne, W. E.

1991-01-01

One of the major limitations of solar cells in space power systems is their vulnerability to radiation damage. One solution to this problem is to periodically heat the cells to anneal the radiation damage. Annealing was demonstrated with silicon cells. The obstacle to annealing of GaAs cells was their susceptibility to thermal damage at the temperatures required to completely anneal the radiation damage. GaAs cells with high temperature contacts and encapsulation were developed. The cells tested are designed for concentrator use at 30 suns AMO. The circular active area is 2.5 mm in diameter for an area of 0.05 sq cm. Typical one sun AMO efficiency of these cells is over 18 percent. The cells were demonstrated to be resistant to damage after thermal excursions in excess of 600 C. This high temperature tolerance should allow these cells to survive the annealing of radiation damage. A limited set of experiments were devised to investigate the feasibility of annealing these high temperature cells. The effect of repeated cycles of electron and proton irradiation was tested. The damage mechanisms were analyzed. Limitations in annealing recovery suggested improvements in cell design for more complete recovery. These preliminary experiments also indicate the need for further study to isolate damage mechanisms. The primary objective of the experiments was to demonstrate and quantify the annealing behavior of high temperature GaAs cells. Secondary objectives were to measure the radiation degradation and to determine the effect of repeated irradiation and anneal cycles.

Burst annealing of high temperature GaAs solar cells

International Nuclear Information System (INIS)

Brothers, P.R.; Horne, W.E.

1991-01-01

One of the major limitations of solar cells in space power systems is their vulnerability to radiation damage. One solution to this problem is to periodically heat the cells to anneal the radiation damage. Annealing was demonstrated with silicon cells. The obstacle to annealing of GaAs cells was their susceptibility to thermal damage at the temperatures required to completely anneal the radiation damage. GaAs cells with high temperature contacts and encapsulation were developed. The cells tested are designed for concentrator use at 30 suns AMO. The circular active area is 2.5 mm in diameter for an area of 0.05 sq cm. Typical one sun AMO efficiency of these cells is over 18 percent. The cells were demonstrated to be resistant to damage after thermal excursions in excess of 600 degree C. This high temperature tolerance should allow these cells to survive the annealing of radiation damage. A limited set of experiments were devised to investigate the feasibility of annealing these high temperature cells. The effect of repeated cycles of electron and proton irradiation was tested. The damage mechanisms were analyzed. Limitations in annealing recovery suggested improvements in cell design for more complete recovery. These preliminary experiments also indicate the need for further study to isolate damage mechanisms. The primary objective of the experiments was to demonstrate and quantify the annealing behavior of high temperature GaAs cells. Secondary objectives were to measure the radiation degradation and to determine the effect of repeated irradiation and anneal cycles

High-Temperature Photovoltaic Effect in La.Ca.MnO/SiO/Si Heterojunction

Directory of Open Access Journals (Sweden)

Hao Ni

2012-01-01

Full Text Available We fabricated a heterojunction of La0.4Ca0.6MnO3/SiO/n-Si and investigated its electronic transport and ultraviolet photovoltaic properties at higher temperature up to 673 K. The rectifying behaviors vanished with the energy-band structure evolvement from 300 to 673 K. Under irradiation of a 248 nm pulse laser, the peak values of open-circuit photovoltage and short-circuit photocurrent decreased drastically. This understanding of the temperature-related current-voltage behavior and ultraviolet photodetection of oxide heterostructures should open a route for devising future microelectronic devices working at high temperature. PACS: 73.40.Lq, 71.27.+ a, 73.50.Pz.

Fusion neutron irradiation induced ordering and defect production in Cu3Au at high temperatures

International Nuclear Information System (INIS)

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

1987-08-01

We irradiate three Cu 3 Au alloys different degrees of initial long-range order at temperatures between 300K and 434K. The resistivity of samples is monitored during irradiation and related to the long-term order parameter by the Muto relation. The results show that the ordering rate, which is proportional to the concentration of freely migrating vacancies, increases at the beginning and then decreases later with fluence. The decrease is a result of the continuous production of sinks in the form of dislocation loops. The effect of sinks on vacancy annihilation in some cases causes a reversed temperature dependence of ordering rate. The free vacancy production rate and the rate of sink production are determined using an ordering kinetics theory. The results of the 14 MeV neutron irradiations are compared to those obtained in other neutron spectra and particle irradiations. The estimated free vacancy production rate is also compared to the primary defect production rate measured at 4.2K in disordered samples

Stability of Y-Ti-O nanoparticles in ODS alloys during heat treatment and high temperature swift heavy ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Skuratov, V.A. [FLNR, JINR, Dubna (Russian Federation); National Research Nuclear University MEPhI, Moscow (Russian Federation); Dubna State University, Dubna (Russian Federation); Sohatsky, A.S.; Kornieieva, K. [FLNR, JINR, Dubna (Russian Federation); O' Connell, J.H.; Neethling, J.H. [CHRTEM, NMMU, Port Elizabeth (South Africa); Nikitina, A.A.; Ageev, V.S. [JSC VNIINM, Moscow (Russian Federation); Zdorovets, M. [Institute of Nuclear Physics, Astana (Kazakhstan); Ural Federal University, Yekaterinburg (Russian Federation); Volkov, A.D. [Nazarbayev University, Astana (Kazakhstan)

2016-12-15

Aim of this report is to compare the morphology of swift (167 and 220 MeV) Xe ion induced latent tracks in Y{sub 2}Ti{sub 2}O{sub 7} nanoparticles during post-irradiation heat treatment and after irradiation at different temperatures in pre-thinned TEM foils and TEM targets prepared from hundreds microns thick irradiated oxide dispersion strengthened (ODS) steel. No difference in track parameters was found in room temperature irradiated nanoparticles in pre-thinned and conventional samples. Microstructural data gathered from pre-thinned foils irradiated in the temperature range 350-650 C or annealed at similar temperatures demonstrate that amorphous latent tracks interact with the surrounding matrix, changing the track and nanoparticle morphology, while such effect is not observed in conventional ODS material treated at the same conditions. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

The morphology of radiation damage in copper irradiated with neutrons at elevated temperatures

International Nuclear Information System (INIS)

Kemm, K.R.

1977-01-01

This thesis is an investigation of the radiation damage morphology of high purity copper crystals irradiated with fast neutrons at temperatures in the range of 250 to 400 degrees C. At these high temperatures neutron damage is found to accumulate into large 3-dimensional rafts up to 100 μm in size, and the well known homogeneous distribution of black dot damage which is characteristic of irradiations at low temperatures is not observed. The characteristics and composition of the rafts of damage at different temperatures in the range 250 to 400 degrees C have been compared and found to differ to a large extent. It has also been shown that the background areas between rafts contain a rather low density of damage at all temperatures studied. It is therefore concluded that many of the interstitial atoms formed during irradiation migrate over large distances through the crystal lattice to precipitate at the sites of the dislocations forming the large rafts, and so denuded inter-rafts areas are left behind. It is proposed that these large rafts originate from grown-in dislocations present in the crystal before irradiation

Basic properties of fuel determining its behavior under irradiation

International Nuclear Information System (INIS)

Konovalov, I.I.

2000-01-01

The theoretical model describing a swelling of nuclear fuel at low irradiation temperatures is considered. The critical physical parameters of substances determining behavior of point defects, gas fission atoms, dislocation density, nucleation and growth of gas-contained pores are determined. The correlation between meanings of critical parameters and physical properties of substance is offered. The accounts of swelling of various dense fuels with reference to work in conditions of research reactors are given. (author)

Temperature dependent investigation on optically active process of higher-order bands in irradiated silicon

International Nuclear Information System (INIS)

Shi Yi; Nanjing Univ., JS; Wu Fengmei; Nanjing Univ., JS; Zheng Youdou; Nanjing Univ., JS; Suezawa, M.; Imai, M.; Sumino, K.

1996-01-01

Optically active processes of the higher-order bands (HOB) are investigated at different temperatures in fast neutron irradiated silicon using Fourier transform infrared absorption measurement. It is shown that the optically active process is nearly temperature independent below 80 K, the slow decay process remains up to a heating temperature of 180 K. The observations are analyzed in terms of the relaxation behavior of photoexcited carriers governed by fast neutron radiation induced defect clusters. (orig.)

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

NARCIS (Netherlands)

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

1991-01-01

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

Irradiation behavior of miniature experimental uranium silicide fuel plates

International Nuclear Information System (INIS)

Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

1983-01-01

Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10 20 cm -3 , far short of the approximately 20 x 10 20 cm -3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix

Mechanistic insights into the room temperature transitions of polytetrafluoroethylene during electron-beam irradiation

Science.gov (United States)

Fu, Congli; Yu, Xianwei; Zhao, Xiaofeng; Wang, Xiuli; Gu, Aiqun; Xie, Meiju; Chen, Chen; Yu, Zili

2017-11-01

In order to recognize the characteristic thermal transitions of polytetrafluoroethylene (PTFE) occurring at 19 °C and 30 °C, PTFE is irradiated on electron beam accelerator at room temperature and analyzed by differential scanning calorimetry (DSC), Fourier transform infrared spectroscopy (FTIR) and X-ray diffraction (XRD). The results suggest that the two transition temperatures decrease considerably with increasing irradiation doses. Based on the results of structural analysis, the decrease of the two transition temperatures is supposed to be highly relevant to the structural changes. In particular, the content and structure of the side groups generated in PTFE are responsible for the variations of the two thermal transitions after irradiation, offering fundamental insights into the reaction mechanisms of PTFE during irradiation.

Microscale interfacial behavior at vapor film collapse on high-temperature particle surface

International Nuclear Information System (INIS)

Abe, Yutaka; Tochio, Daisuke

2009-01-01

It has been pointed out that vapor film on a premixed high-temperature droplet surface should be collapsed to trigger vapor explosion. Thus, it is important to clarify the micromechanism of vapor film collapse behavior for the occurrence of vapor explosion. In the present study, microscale vapor-liquid interface behavior upon vapor film collapse caused by an external pressure pulse is experimentally observed and qualitatively analyzed. In the analytical investigation, interfacial temperature and interface movement were estimated with heat conduction analysis and visual data processing technique. Results show that condensation can possibly occur at the vapor-liquid interface when the pressure pulse arrived. That is, this result indicates that the vapor film collapse behavior is dominated not by fluid motion but by phase change. (author)

Behavior of the dynamic magnetic susceptibility in ybco bula ceramics irradiated with gamma rays

International Nuclear Information System (INIS)

Leyva Fabelo, A.; Bouza Dominguez, J.; Cruz Inclan, Carlos M.

2001-01-01

Using measurements of the ac susceptibility, the behavior with the irradiation dose of YBa2Cu3O7- bulk ceramics synthesized by the classic reaction method in solid state, was studied. A Co60 gamma chamber model MPX-G-25M and a Cs137 source were employed as gamma ray sources. The behavior of the beginning temperature of the normal - superconducting state transition with the exposition dose show, independently of the incident gamma energy, a monotonous growth until reaching a threshold dose, after which, observe a fall, more abrupt in the case of the Co60. This behavior can be explained using the model that postulates the ability of the gamma radiation, in certain dose intervals, to stimulate the structural reordering in the oxygen sublattice. When the irradiation process takes place in the Co60 gamma chamber, the behavior of the superconducting volume fraction of the sample characterizes by the initial sharp fall with the dose, followed with an attenuation of the decrement. In the case of Cs137 irradiation, the behavior of the superconducting volume fraction is similar to the behavior of the Ton with the dose

Effect of strain rate and temperature at high strains on fatigue behavior of SAP alloys

DEFF Research Database (Denmark)

Blucher, J.T.; Knudsen, Per; Grant, N.J.

1968-01-01

Fatigue behavior of three SAP alloys of two nominal compositions (7 and 13% Al2O3) was studied in terms of strain rate and temperature at high strains; strain rate had no effect on life at 80 F, but had increasingly greater effect with increasing temperature above 500 F; life decreased with decre......Fatigue behavior of three SAP alloys of two nominal compositions (7 and 13% Al2O3) was studied in terms of strain rate and temperature at high strains; strain rate had no effect on life at 80 F, but had increasingly greater effect with increasing temperature above 500 F; life decreased...

Release of fission products from irradiated aluminide fuel at high temperature

International Nuclear Information System (INIS)

Shibata, Toshikazu; Kanda, Keiji; Mishima, Kaichiro; Tamai, Tadaharu; Hayashi, Masatoshi; Snelgrove, James L.; Stahl, David; Matos, James E.; Travelli, Armando; Case, F. Neil; Posey, John C.

1983-01-01

Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel cladding material. The release of fission products from the fuel plate at temperature below 500 deg. C was found negligible. The first rapid release of fission products was observed with the occurrence of blistering at 561±1 deg. C on the plates. The next release at 585. C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 deg. C of U-Al x . The released material was mostly xenon, but small amounts of iodine and cesium were observed. (author)

Release of fission products from irradiated aluminide fuel at high temperature

International Nuclear Information System (INIS)

Shibata, T.; Kanda, K.; Mishima, K.

1982-01-01

Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel-cladding material. The release of fission products from the fuel plate at temperature below 500 0 C was found negligible. The firist rapid release of fission products was observed with the occurrence of blistering at 561 +- 1 0 C on the plates. The next release at 585 0 C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 0 C of U-Al/sub x/. The released material was mostly xenon, but small amounts of iodine and cesium were observed

Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50–400)°C

Energy Technology Data Exchange (ETDEWEB)

Kuleshova, E.A., E-mail: evgenia-orm@yandex.ru [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation); National Research Nuclear University “MEPhI” (Moscow Engineering Physics Institute), Kashirskoe Highway 31, Moscow 115409 (Russian Federation); Gurovich, B.A.; Bukina, Z.V.; Frolov, A.S.; Maltsev, D.A.; Krikun, E.V.; Zhurko, D.A.; Zhuchkov, G.M. [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation)

2017-07-15

This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50–400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔT{sub K}) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects – dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔT{sub K} shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔT{sub K} shift in the studied range of irradiation temperature and fluence. - Highlights: •Structural elements in RPV steel are studied at different irradiation temperatures. •Highest number density dislocation loops are

Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Kim, Woo Gon; Kim, D. W.; Ryu, W. S.; Han, C. H.; Yoon, J. H.; Chang, J.

2005-01-01

Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

In Situ Irradiation and Measurement of Triple Junction Solar Cells at Low Intensity, Low Temperature (LILT) Conditions

Science.gov (United States)

Harris, R.D.; Imaizumi, M.; Walters, R.J.; Lorentzen, J.R.; Messenger, S.R.; Tischler, J.G.; Ohshima, T.; Sato, S.; Sharps, P.R.; Fatemi, N.S.

2008-01-01

The performance of triple junction InGaP/(In)GaAs/Ge space solar cells was studied following high energy electron irradiation at low temperature. Cell characterization was carried out in situ at the irradiation temperature while using low intensity illumination, and, as such, these conditions reflect those found for deep space, solar powered missions that are far from the sun. Cell characterization consisted of I-V measurements and quantum efficiency measurements. The low temperature irradiations caused substantial degradation that differs in some ways from that seen after room temperature irradiations. The short circuit current degrades more at low temperature while the open circuit voltage degrades more at room temperature. A room temperature anneal after the low temperature irradiation produced a substantial recovery in the degradation. Following irradiation at both temperatures and an extended room temperature anneal, quantum efficiency measurement suggests that the bulk of the remaining damage is in the (In)GaAs sub-cell

HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

Energy Technology Data Exchange (ETDEWEB)

Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

1998-03-01

As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

Behavior and failure of fresh, hydrided and irradiated Zircaloy-4 fuel claddings under RIA conditions

International Nuclear Information System (INIS)

Le Saux, M.

2008-01-01

The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fresh, hydrided and irradiated (in pressurized water reactors) cold-worked stress relieved Zircaloy-4 fuel claddings under reactivity initiated accident conditions. A model is proposed to describe the anisotropic viscoplastic mechanical behavior of the material as a function of temperature (from 20 C up to 1100 C), strain rate (from 3.10 -4 s -1 up to 5 s -1 ), fluence (from 0 up to 1026 n.m -2 ) and irradiation conditions. Axial tensile, hoop tensile, expansion due to compression and hoop plane strain tensile tests are performed at 25 C, 350 C and 480 C in order to analyse the anisotropic plastic and failure properties of the non-irradiated material hydrided up to 1200 ppm. Material strength and strain hardening depend on temperature and hydrogen in solid solution and precipitated hydride contents. Plastic anisotropy is not significantly modified by hydrogen. The material is embrittled by hydrides at room temperature. The plastic strain that leads to hydride cracking decreases with increasing hydrogen content. The material ductility, which increases with increasing temperature, is not deteriorated by hydrogen at 350 C and 480 C. Macroscopic fracture modes and damage mechanisms depend on specimen geometry, temperature and hydrogen content. A Gurson type model is finally proposed to describe both the anisotropic viscoplastic behavior and the ductile fracture of the material as a function of temperature and hydrogen content. (author) [fr

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-15

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Thermal effects on tenebrio molitor and lawn irradiated by high power centimeter wave

International Nuclear Information System (INIS)

Zhang Jie; Han Lijun; Qi Hongxing; Chen Shude; Qiao Dengjiang

2008-01-01

A system of high power centimeter wave(HPCW) was built up. The temperature change of tenebrio molitor and lawn was sampled during HPCW irradiation. It is shown that the relationship between the temperature increase of tenebrio molitor and irradiation time is approximately linear, and the relationship between the temperature increase of lawn and irradiation time is nonlinear. It is also considered that the temperature of lawn increases faster than that of tenebrio molitor does during the earlier stage of irradiation. The death percentage of tenebrio molitor and injury rate of lawn irradiated by HPCW could be fitted by normal distribution. The fitting curves of relations between death rate and irradiation time and between death rate and temperature are presented. (authors)

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

International Nuclear Information System (INIS)

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

1980-06-01

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

Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

International Nuclear Information System (INIS)

Nickel, H.

1990-05-01

The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

In-situ field-ion microscope study of the recovery behavior of heavy metal ion-irradiated tungsten, tungsten (rhenium) alloys and molybdenum

International Nuclear Information System (INIS)

Nielsen, C.H.

1977-06-01

Three field ion microscope (FIM) experiments were carried out to study the annealing behavior of heavy ion irradiated tungsten, tungsten (rhenium) alloys and molybdenum. The first experiment dealt with the stage I long-range migration of tungsten self interstitial atoms (SIAs) in high purity tungsten of resistivity ratio, R = 24,000 (R = rho 300 /rho 4 . 2 , where rho 300 and rho 4 . 2 are the room temperature and 0 0 C resistivities). The FIM specimens were irradiated in situ at 18 K with 30 keV W + ions to an average dose of 5 x 10 12 ions cm -2 and subsequently examined by the pulsed-field evaporation technique. The second experiment dealt with the phenomenon of impurity atom trapping of SIAs during long-range migration. It was shown that rhenium atoms in a tungsten matrix tend to capture tungsten SIAs and remain bound up to temperatures as high as 390 K. The final experiment was concerned with the low temperature annealing kinetics of irradiated molybdenum. High purity molybdenum of resistivity ratio R = 5700 was irradiated at 10 K with 30 keV Mo + ions to a dose of approximately 5 x 10 12 ions cm -2 . The results indicated that the electric field has only a minimal effect on the SIA annealing kinetics. This tends to strengthen the contention that the molybdenum SIA becomes mobile at 32 K

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

Science.gov (United States)

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

2017-10-01

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

Quantitative analysis of the gas evolved from high polymers in γ-irradiation

International Nuclear Information System (INIS)

Arakawa, Kazuo; Hayakawa, Naohiro; Kuriyama, Isamu

1977-09-01

Polymers are used as insulator of cables in nuclear-reactor radiation field. To estimate the evolution of gases when irradiated, total gas yield and composition were measured for variety of polymers. Samples were irradiated at room temperature in vacuo with 60 Co-γ rays. For ethylene propylene rubber (EPR), irradiation in high-temperature steam was also made. Composition of the gas was determined with a mass spectrometer. G-value of the total gaseous product was 3.2 to 3.4 for low-density polyethylene (LDPE) and 2.5 to 2.7 for high-density polyethylene (HDPE). In both polyethylene, hydrogen gas predominated. When an anti-radiation oil was added to LDPE, gas evolution was reduced drastically. For chloro-sulfonated PE (Hypalon), SO 2 gas was one of the major products even when the polymer contained only about 1% of sulfonyl groups. G-value of the total gas for EPR irradiated in high-temperature steam was 3.1, regardless of the temperature. (auth.)

Ultra high vacuum fracture and transfer device for AES analysis of irradiated austenitic stainless steel

International Nuclear Information System (INIS)

Urie, M.W.; Panayotou, N.F.; Robinson, J.E.

1980-01-01

An ultrahigh vacuum fracture and transfer device for analysis of irradiated and non-irradiated SS 316 fuel cladding is described. Mechanical property tests used to study the behavior of cladding during reactor transient over-power conditions are reported. The stress vs temperature curves show minimal differences between unirradiated cladding and unfueled cladding. The fueled cladding fails at a lower temperature. All fueled specimens failed in an intergranular mode

High temperature deformation behavior and microstructural evolutions of a high Zr containing WE magnesium alloy

Energy Technology Data Exchange (ETDEWEB)

Asqardoust, Sh.; Zarei-Hanzaki, A. [School of Metallurgical & Materials Engineering, University of Tehran, Tehran (Iran, Islamic Republic of); Fatemi, S.M., E-mail: mfatemi@ut.ac.ir [Shahid Rajaee Teacher Training University, Tehran (Iran, Islamic Republic of); Moradjoy-Hamedani, M. [School of Metallurgical & Materials Engineering, University of Tehran, Tehran (Iran, Islamic Republic of)

2016-06-05

Magnesium alloys containing RE elements (WE grade) are considered as potential materials for high temperature structural applications. To this end, it is crucial to study the flow behavior and the microstructural evolution of these alloys at high temperatures. In present work, the hot compression testing was employed to investigate the deformation behavior of a rolled WE54 magnesium alloy at elevated temperatures. The experimental material failed to deform to target strain of 0.6 at 250 and 300 °C, while the straining was successfully performed at 350 °C. A flow softening was observed at 350 °C, which was related to the depletion of RE strengthener elements, particularly Y atoms, from the solid solution and dynamic precipitation of β phases. It was suggested that the Zener pinning effect of the latter precipitates might retard the occurrence of dynamic recrystallization. As the temperature increased to 450 and 500 °C, the RE elements dissolved in the matrix and thus dynamic recrystallization could considerably progress in the microstructure. The comparative study of specimens cut along transverse ad normal direction (TD and ND specimens) implied that the presence of RE elements might effectively reduce the yield anisotropy in WE54 rolled alloy. Microstructural observations indicated a higher fraction of dynamically-recrystallized grains for the ND specimens. This was discussed relying on the different shares of deformation mechanism during compressing the TD and ND specimens. - Highlights: • Deformation behavior of a high Zr WE alloy was addressed at low strain rate. • Dynamic precipitation was realized at 350 °C. • The occurrence of DRX was retarded due to Zener pinning effect. • A higher DRX fraction was obtained in ND specimens comparing with TD ones.

Effects of temperature during the irradiation of calcium carbonate

Energy Technology Data Exchange (ETDEWEB)

Negron M, A.; Camargo R, C.; Ramos B, S. [UNAM, Instituto de Ciencias Nucleares, Circuito Exterior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Gomez V, V. [UNAM, Instituto de Quimica, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Uribe, R. M., E-mail: negron@nucleares.unam.mx [Kent State University, College of Technology, Kent 44240 Ohio (United States)

2015-10-15

The gamma irradiation of calcium carbonate at different doses (0 to 309 kGy) and temperature regimes (77 K to 298 K) was carried out to study the effects of irradiation temperature. The changes were followed by EPR spectroscopy. We observed the formation of a composite EPR spectrum even at low radiation doses and temperature. There is a strong effect on the evaluation of the radicals formed as a function of irradiation temperature, probably due to the diffusion in the frozen powder. Response curves show that this system tends to saturate at 10 MGy at 298 K. (Author)

Effects of temperature during the irradiation of calcium carbonate

International Nuclear Information System (INIS)

Negron M, A.; Camargo R, C.; Ramos B, S.; Gomez V, V.; Uribe, R. M.

2015-10-01

The gamma irradiation of calcium carbonate at different doses (0 to 309 kGy) and temperature regimes (77 K to 298 K) was carried out to study the effects of irradiation temperature. The changes were followed by EPR spectroscopy. We observed the formation of a composite EPR spectrum even at low radiation doses and temperature. There is a strong effect on the evaluation of the radicals formed as a function of irradiation temperature, probably due to the diffusion in the frozen powder. Response curves show that this system tends to saturate at 10 MGy at 298 K. (Author)

Numerical examination of temperature control in helium-cooled high flux test module of IFMIF

International Nuclear Information System (INIS)

Ebara, Shinji; Yokomine, Takehiko; Shimizu, Akihiko

2007-01-01

For long term irradiation of the International Fusion Materials Irradiation Facility (IFMIF), test specimens are needed to retain constant temperature to avoid change of its irradiation characteristics. The constant temperatures control is one of the most challenging issues for the IFMIF test facilities. We have proposed a new concept of test module which is capable of precisely measuring temperature, keeping uniform temperature with enhanced cooling performance. In the system according to the new design, cooling performances and temperature distributions of specimens were examined numerically under diverse conditions. Some transient behaviors corresponding to the prescribed temperature control mode were perseveringly simulated. It was confirmed that the thermal characteristics of the new design satisfied the severe requirement of IFMIF

Design practice and operational experience of highly irradiated, high-performance normal magnets

International Nuclear Information System (INIS)

Schultz, J.H.

1982-09-01

The limitations of high performance magnets are discussed in terms of mechanical, temperature, and electrical limits. The limitations of magnets that are highly irradiated by neutrons, gamma radiation, or x radiation are discussed

Analysis of irradiation-induced stresses in coating layers of coated fuel particles for the High Temperature Engineering Test Reactor (HTTR)

International Nuclear Information System (INIS)

Hayashi, Kimio; Kikuchi, Teruo; Fukuda, Kousaku; Sato, Sadao; Toyota, Junji; Shiozawa, Shusaku; Sawa, Kazuhiro; Kashimura, Satoru.

1991-07-01

Irradiation-induced stresses in coating layers of coated fuel particles were analyzed by the MICROS-2 code for the fuels of the High Temperature Engineering Test Reactor (HTTR) under its operating conditions. The analyses were made on the standard core fuel (A-type) and the test fuels comprising the advanced SiC-coated particle fuel (B-1 type) and the ZrC-coated particle fuel (B-2 type). For the B-1 type fuel, the stresses were relieved due to the thicker buffer and SiC layers than for the A type fuel. The slightly decreased thickness of the fourth layer for the B-1 type than for the A type fuel had no significant effect on the stresses. As for the B-2 type fuel, almost the same results as for the B-1 type were obtained under an assumption that the ZrC layer as well as the SiC layer undergoes negligible dimension change within the analysis conditions. The obtained results indicated that the B-1 and B-2 type fuels are better than the A type fuel in terms of integrity against the irradiation-induced stresses. Finally, research subjects for development of the analysis code on the fuel behavior are discussed. (author)

Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

Energy Technology Data Exchange (ETDEWEB)

Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi [Musashi Inst. of Tech., Tokyo (Japan); Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao

1996-04-01

SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young`s modulus lowers. In addition, its composition elements have the large (n,{alpha}) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

International Nuclear Information System (INIS)

Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi; Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao.

1996-01-01

SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young's modulus lowers. In addition, its composition elements have the large (n,α) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

Temperature, stress, and annealing effects on the luminescence from electron-irradiated silicon

Science.gov (United States)

Jones, C. E.; Johnson, E. S.; Compton, W. D.; Noonan, J. R.; Streetman, B. G.

1973-01-01

Low-temperature photoluminescence spectra are presented for Si crystals which have been irradiated with high-energy electrons. Studies of isochronal annealing, stress effects, and the temperature dependences of the luminescence are used to discuss the nature of the luminescent transitions and the properties of defects. Two dominant bands present after room-temperature anneal of irradiated material are discussed, and correlations of the properties of these bands are made with known Si defects. A band between 0.8 and 1.0 eV has properties which are related to those of the divacancy, and a band between 0.6 and 0.8 eV has properties related to those of the Si-G15(K) center. Additional peaks appear in the luminescence after high-temperature anneal; the influence of impurities and the effects of annealing of these lines are discussed.

Resistivity changes in superconducting-cavity-grade Nb following high-energy proton irradiation

International Nuclear Information System (INIS)

Snead, C.L. Jr.; Hanson, A.; Greene, G.A.

1997-01-01

Niobium superconducting rf cavities are proposed for use in the proton LINAC accelerators for spallation-neutron applications. Because of accidental beam loss and continual halo losses along the accelerator path, concern for the degradation of the superconducting properties of the cavities with accumulating damage arises. Residual-resistivity-ratio (RRR) specimens of Nb, with a range of initial RRR's were irradiated at room temperature with protons at energies from 200 to 2000 MeV. Four-probe resistance measurements were made at room temperature and at 4.2 K both prior to and after irradiation. Nonlinear increases in resistivity simulate expected behavior in cavity material after extended irradiation, followed by periodic anneals to room temperature: For RRR = 316 material, irradiations to (2 - 3) x 10 15 p/cm 2 produce degradations up to the 10% level, a change that is deemed operationally acceptable. Without. periodic warming to room temperature, the accumulated damage energy would be up to a factor of ten greater, resulting in unacceptable degradations. Likewise, should higher-RRR material be used, for the same damage energy imparted, relatively larger percentage changes in the RRR will result

Tensile behavior of EUROFER ODS steel after neutron irradiation up to 16.3 dpa between 250 and 450 °C

International Nuclear Information System (INIS)

Materna-Morris, Edeltraud; Lindau, Rainer; Schneider, Hans-Christian; Möslang, Anton

2015-01-01

Highlights: • The first 9%CrWVTa steel (0.5% Y_2O_3), EUROFER ODS HIP, have been neutron irradiated up to 16.3 dpa, between 250 and 450 °C, in the High Flux Reactor (HFR). • After post-irradiation tensile tests, there was not any increase of the upper yield strength or strain localization after irradiation which is typical of RAFM steels. • Initially higher yield strength, R_p_0_._2, and distinctive tensile strength, R_m, of EUROFER ODS HIP compared to EUROFER97 steel. • These values increased due to the neutron irradiation at lower irradiation temperatures. - Abstract: During the development of structural material for future fusion reactors, a 50 kg heat of reduced-activation ferritic-martensitic 9%CrWVTa steel with nanoscaled Y_2O_3-particles, EUROFER97 ODS HIP, was produced using powder metallurgy fabrication technology. This first batch of EUROFER97 ODS HIP and, for comparison, the steel EUROFER97 were prepared for a post-irradiation tensile test program. During neutron irradiation in the HFR (High Flux Reactor, The Netherlands), an accumulated dose of up to 16.3 dpa was reached for 771 days at full power, with the irradiation temperature ranging between 250 and 450 °C. During the post-examinations, all specimens showed the highest tensile strength at lower irradiation temperatures between 250 and 350 °C. However, ODS-alloy and steel were found to clearly differ in the mechanical behavior, which could be documented by fully instrumented tensile tests. In the un-irradiated state, tensile strength of the ODS-alloy already was increased considerably by about 60% compared to the steel. Strengthening was further increased by another 20% after neutron irradiation, but with a much better ductility than observed in the steel. The typical irradiation-induced strain localization of EUROFER97 or RAFM steels could not be observed in the EUROFER97 ODS HIP alloy.

Nanomechanical Characterization of Temperature-Dependent Mechanical Properties of Ion-Irradiated Zirconium with Consideration of Microstructure and Surface Damage

Science.gov (United States)

Marsh, Jonathan; Zhang, Yang; Verma, Devendra; Biswas, Sudipta; Haque, Aman; Tomar, Vikas

2015-12-01

Zirconium alloys for nuclear applications with different microstructures were produced by manufacturing processes such as chipping, rolling and annealing. The two Zr samples, rolled and rolled-annealed were subjected to different levels of irradiation, 1 keV and 100 eV, to study the effect of irradiation dosages. The effect of microstructure and irradiation on the mechanical properties (reduced modulus, hardness, indentation yield strength) was analyzed with nanoindentation experiments, which were carried out in the temperature range of 25°C to 450°C to investigate temperature dependence. An indentation size effect analysis was performed and the mechanical properties were also corrected for the oxidation effects at high temperatures. The irradiation-induced hardness was observed, with rolled samples exhibiting higher increase compared to rolled and annealed samples. The relevant material parameters of the Anand viscoplastic model were determined for Zr samples containing different level of irradiation to account for viscoplasticity at high temperatures. The effect of the microstructure and irradiation on the stress-strain curve along with the influence of temperature on the mechanisms of irradiation creep such as formation of vacancies and interstitials is presented. The yield strength of irradiated samples was found to be higher than the unirradiated samples which also showed a decreasing trend with the temperature.

Degradation of glass-fiber reinforced plastics by low temperature irradiation

International Nuclear Information System (INIS)

Nishijima, S.; Nishiura, T.; Ueno, S.; Tsukazaki, Y.; Okada, T.; Okada, T.M.; Miyata, K.; Kodaka, H.

1998-01-01

Low-temperature irradiation effects of glass-fiber reinforced plastics (GFRP) have been investigated in terms of mechanical properties such as interlaminar shear strength and creep, in order to obtain the selection standard of insulating materials of superconducting magnets used for fusion reactor. It was revealed that the degradation of interlaminar shear strength was strongly dependent of characteristics of matrix and/or glass/epoxy interface. Especially, the research has been carried out towards the creep behaviour of epoxy which is the matrix of GFRP, by both experimental and simulation method. It was suggested that the synergistic effects was observed in creep test. From the molecular dynamics simulation it was found that the cage effects was the one of the main reason of the stress effects of creep behavior under irradiation. (author)

Tensile behavior of RAFM alloys after neutron irradiation of up to 16.3 dpa between 250 and 450 °C

International Nuclear Information System (INIS)

Materna-Morris, E.; Schneider, H.-C.; Möslang, A.

2014-01-01

Tensile specimen of steel EUROFER97 and other alloys on the basis of RAFM steels such, as OPTIFER and F82H alloys, and Ga3X were irradiated and post-examined during a neutron irradiation program of up to 16.3 dpa between 250 and 450 °C in the HFR (High Flux Reactor) in the Netherlands. These tensile results were compared with former irradiation programs, with lower neutron doses of up to 0.8 and 2.4 dpa to quantify the difference in tensile strengthening. The average increase of tensile strength was in a range of 300 MPa between 0.8 and 16.3 dpa at temperatures of 250–300 °C. This behavior can be correlated with irradiation-induced changes in the microstructure. Most of the hardening can be attributed to dislocation loops, point defects or small precipitates as observed in boron-free alloys as F82H mod. and EUROFER97. Whereas the hardening in boron-containing alloys OPTIFER alloys and Ga3X can be correlated in addition with the combination of helium bubbles. At the highest irradiation and test temperature at 450 °C, all tensile data of all investigated materials were in the range of those of non-irradiated and irradiated material due to thermal aging effects

Tensile behavior of RAFM alloys after neutron irradiation of up to 16.3 dpa between 250 and 450 °C

Energy Technology Data Exchange (ETDEWEB)

Materna-Morris, E., E-mail: edeltraud.materna-morris@kit.edu; Schneider, H.-C., E-mail: hans-christian.schneider@kit.edu; Möslang, A., E-mail: anton.moeslang@kit.edu

2014-12-15

Tensile specimen of steel EUROFER97 and other alloys on the basis of RAFM steels such, as OPTIFER and F82H alloys, and Ga3X were irradiated and post-examined during a neutron irradiation program of up to 16.3 dpa between 250 and 450 °C in the HFR (High Flux Reactor) in the Netherlands. These tensile results were compared with former irradiation programs, with lower neutron doses of up to 0.8 and 2.4 dpa to quantify the difference in tensile strengthening. The average increase of tensile strength was in a range of 300 MPa between 0.8 and 16.3 dpa at temperatures of 250–300 °C. This behavior can be correlated with irradiation-induced changes in the microstructure. Most of the hardening can be attributed to dislocation loops, point defects or small precipitates as observed in boron-free alloys as F82H mod. and EUROFER97. Whereas the hardening in boron-containing alloys OPTIFER alloys and Ga3X can be correlated in addition with the combination of helium bubbles. At the highest irradiation and test temperature at 450 °C, all tensile data of all investigated materials were in the range of those of non-irradiated and irradiated material due to thermal aging effects.

Creep Behavior of High-Strength Concrete Subjected to Elevated Temperatures.

Science.gov (United States)

Yoon, Minho; Kim, Gyuyong; Kim, Youngsun; Lee, Taegyu; Choe, Gyeongcheol; Hwang, Euichul; Nam, Jeongsoo

2017-07-11

Strain is generated in concrete subjected to elevated temperatures owing to the influence of factors such as thermal expansion and design load. Such strains resulting from elevated temperatures and load can significantly influence the stability of a structure during and after a fire. In addition, the lower the water-to-binder (W-B) ratio and the smaller the quantity of aggregates in high-strength concrete, the more likely it is for unstable strain to occur. Hence, in this study, the compressive strength, elastic modulus, and creep behavior were evaluated at target temperatures of 100, 200, 300, 500, and 800 °C for high-strength concretes with W-B ratios of 30%, 26%, and 23%. The loading conditions were set as non-loading and 0.33f cu . It was found that as the compressive strength of the concrete increased, the mechanical characteristics deteriorated and transient creep increased. Furthermore, when the point at which creep strain occurred at elevated temperatures after the occurrence of transient creep was considered, greater shrinkage strain occurred as the compressive strength of the concrete increased. At a heating temperature of 800 °C, the 80 and 100 MPa test specimens showed creep failure within a shrinkage strain range similar to the strain at the maximum load.

Creep Behavior of High-Strength Concrete Subjected to Elevated Temperatures

Directory of Open Access Journals (Sweden)

Minho Yoon

2017-07-01

Full Text Available Strain is generated in concrete subjected to elevated temperatures owing to the influence of factors such as thermal expansion and design load. Such strains resulting from elevated temperatures and load can significantly influence the stability of a structure during and after a fire. In addition, the lower the water-to-binder (W–B ratio and the smaller the quantity of aggregates in high-strength concrete, the more likely it is for unstable strain to occur. Hence, in this study, the compressive strength, elastic modulus, and creep behavior were evaluated at target temperatures of 100, 200, 300, 500, and 800 °C for high-strength concretes with W–B ratios of 30%, 26%, and 23%. The loading conditions were set as non-loading and 0.33fcu. It was found that as the compressive strength of the concrete increased, the mechanical characteristics deteriorated and transient creep increased. Furthermore, when the point at which creep strain occurred at elevated temperatures after the occurrence of transient creep was considered, greater shrinkage strain occurred as the compressive strength of the concrete increased. At a heating temperature of 800 °C, the 80 and 100 MPa test specimens showed creep failure within a shrinkage strain range similar to the strain at the maximum load.

Repair behavior of He+-irradiated W-Y2O3 composites after different temperature-isochronal annealing experiments

Science.gov (United States)

Yao, Gang; Tan, Xiao-Yue; Luo, Lai-Ma; Zan, Xiang; Liu, Jia-Qin; Xu, Qiu; Zhu, Xifao-Yong; Wu, Yu-Cheng

2018-01-01

W-2%Y2O3 composites were prepared by wet chemical and powder metallurgy. Commercial roll tungsten was selected as a comparative sample in the He+ irradiation experiment. The experiment was conducted under He+ beam energy of 50 eV, irradiation dose of approximately 9.9 × 1024 ions/m2, and temperature of 1503-1553 K. The samples were annealed at 1173, 1373, and 1573 K for 1 h. The irradiation surface was observed in situ. The W-2%Y2O3 composites and pure tungsten displayed different grain orientation damage morphologies. In addition, the fuzzy structure was more likely to converge densely at the phase interface. Annealing repairs material surface irradiation damage, whereas the phase interface acts as a He+ migration channel.

Temperature detectors on irradiated silicon base

International Nuclear Information System (INIS)

Karimov, M.; Dzhalelov, M.A.; Kurbanov, A.O.

2005-01-01

It is well known, that the most suitable for thermal resistors production is compensated silicon with impurities forming deep lying in forbidden zone, having big negative resistance temperature coefficients (RTC). In the capacity of initial materials for thermal resistors with negative RTC the n-type monocrystalline silicon with specific resistance ∼30 Ω·cm at 300 K is applied. Before the irradiation the phosphorus diffusion is realizing at temperature ∼1000 deg. C for 10 min. Irradiation is putting into practise by WWR-SM reactor fast neutrons within the range (7-10)·10 13 cm -2 . The produced resistors have nominal resistance range (8-20)·10 3 Ω·cm, coefficient of the thermal sensitivity B=4000-6000 deg. C., RTC α 300K =4-6.6 %/grad. It is shown, that offered method allows to obtain same type resistors characteristics on the base of neutron-irradiated material

Modeling of mechanical behavior of quenched zirconium-based nuclear fuel claddings after a high temperature oxidation

International Nuclear Information System (INIS)

Cabrera-Salcedo, A.

2012-01-01

During the second stage of Loss Of Coolant Accident (LOCA) in Pressurized Water Reactors (PWR) zirconium-based fuel claddings undergo a high temperature oxidation (up to 1200 C), then a water quench. After a single-side steam oxidation followed by a direct quench, the cladding is composed of three layers: an oxide (Zirconia) outer layer (formed at HT), always brittle at Room Temperature (RT), an intermediate oxygen stabilized alpha layer, always brittle at RT, called alpha(O), and an inner 'prior-beta' layer, which is the only layer able to keep some significant Post Quench (PQ) ductility at RT. However, hydrogen absorbed because of service exposure or during the LOCA transient, concentrates in this layer and may leads to its embrittlement. To estimate the PQ mechanical properties of these materials, Ring Compression Tests (RCT) are widely used because of their simplicity. Small sample size makes RCTs advantageous when a comparison with irradiated samples is required. Despite their good reproducibility, these tests are difficult to interpret as they often present two or more load drops on the engineering load-displacement curve. Laboratories disagree about their interpretation. This study proposes an original fracture scenario for a stratified PQ cladding tested by RCT, and its associated FE model. Strong oxygen content gradient effect on layers mechanical properties is taken into account in the model. PQ thermal stresses resulting from water quench of HT oxidized cladding are investigated, as well as progressive damage of three layers during an RCT. The proposed scenario is based on interrupted RCT analysis, post- RCT sample's outer layers observation for damage evaluation, RCTs of prior-beta single-layer rings, and mechanical behavior of especially chemically adjusted samples. The force displacement curves appearance is correctly reproduced using the obtained FE model. The proposed fracture scenario elucidates RCTs of quenched zirconium-based nuclear fuel

Primary radiation damage characterization of α-iron under irradiation temperature for various PKA energies

Science.gov (United States)

Sahi, Qurat-ul-ain; Kim, Yong-Soo

2018-04-01

The understanding of radiation-induced microstructural defects in body-centered cubic (BCC) iron is of major interest to those using advanced steel under extreme conditions in nuclear reactors. In this study, molecular dynamics (MD) simulations were implemented to examine the primary radiation damage in BCC iron with displacement cascades of energy 1, 5, 10, 20, and 30 keV at temperatures ranging from 100 to 1000 K. Statistical analysis of eight MD simulations of collision cascades were carried out along each [110], [112], [111] and a high index [135] direction and the temperature dependence of the surviving number of point defects and the in-cascade clustering of vacancies and interstitials were studied. The peak time and the corresponding number of defects increase with increasing irradiation temperature and primary knock-on atom (PKA) energy. However, the final number of surviving point defects decreases with increasing lattice temperature. This is associated with the increase of thermal spike at high PKA energy and its long timespan at higher temperatures. Defect production efficiency (i.e., surviving MD defects, per Norgett-Robinson-Torrens displacements) also showed a continuous decrease with the increasing irradiation temperature and PKA energy. The number of interstitial clusters increases with both irradiation temperature and PKA energy. However, the increase in the number of vacancy clusters with PKA energy is minimal-to-constant and decreases as the irradiation temperature increases. Similarly, the probability and cluster size distribution for larger interstitials increase with temperature, whereas only smaller size vacancy clusters were observed at higher temperatures.

High-temperature absorbed dose measurements in the megagray range

International Nuclear Information System (INIS)

Balian, P.; Ardonceau, J.; Zuppiroli, L.

1988-01-01

Organic conductors of the tetraselenotetracene family have been tested as ''high-temperature'' absorbed dose dosimeters. They were heated up to 120 0 C and irradiated at this temperature with 1-MeV electrons in order to simulate, in a short time, a much longer γ-ray irradiation. The electric resistance increase of the crystal can be considered a good measurement of the absorbed dose in the range 10 6 Gy to a few 10 8 Gy and presumably one order of magnitude more. This dosimeter also permits on-line (in-situ) measurements of the absorbed dose without removing the sensor from the irradiation site. The respective advantages of organic and inorganic dosimeters at these temperature and dose ranges are also discussed. In this connection, we outline new, but negative, results concerning the possible use of silica as a high-temperature, high-dose dosimeter. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-07-01

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

Mechanical energy losses in plastically deformed and electron plus neutron irradiated high purity single crystalline molybdenum at elevated temperatures

Energy Technology Data Exchange (ETDEWEB)

Zelada, Griselda I. [Laboratorio de Materiales, Escuela de Ingenieria Electrica, Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Avda. Pellegrini 250, 2000 Rosario (Argentina); Lambri, Osvaldo Agustin [Laboratorio de Materiales, Escuela de Ingenieria Electrica, Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Avda. Pellegrini 250, 2000 Rosario (Argentina); Instituto de Fisica Rosario - CONICET, Member of the CONICET& #x27; s Research Staff, Avda. Pellegrini 250, 2000 Rosario (Argentina); Bozzano, Patricia B. [Laboratorio de Microscopia Electronica, Unidad de Actividad Materiales, Centro Atomico Constituyentes, Comision Nacional de Energia Atomica, Avda. Gral. Paz 1499, 1650 San Martin (Argentina); Garcia, Jose Angel [Departamento de Fisica Aplicada II, Facultad de Ciencias y Tecnologia, Universidad del Pais Vasco, Apdo. 644, 48080 Bilbao, Pais Vasco (Spain)

2012-10-15

Mechanical spectroscopy (MS) and transmission electron microscopy (TEM) studies have been performed in plastically deformed and electron plus neutron irradiated high purity single crystalline molybdenum, oriented for single slip, in order to study the dislocation dynamics in the temperature range within one third of the melting temperature. A damping peak related to the interaction of dislocation lines with both prismatic loops and tangles of dislocations was found. The peak temperature ranges between 900 and 1050 K, for an oscillating frequency of about 1 Hz. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Verification of thermal-irradiation stress analytical code VIENUS of graphite block

International Nuclear Information System (INIS)

Iyoku, Tatsuo; Ishihara, Masahiro; Shiozawa, Shusaku; Shirai, Hiroshi; Minato, Kazuo.

1992-02-01

The core graphite components of the High Temperature Engineering Test Reactor (HTTR) show both the dimensional change (irradiation shrinkage) and creep behavior due to fast neutron irradiation under the temperature and the fast neutron irradiation conditions of the HTTR. Therefore, thermal/irradiation stress analytical code, VIENUS, which treats these graphite irradiation behavior, is to be employed in order to design the core components such as fuel block etc. of the HTTR. The VIENUS is a two dimensional finite element viscoelastic stress analytical code to take account of changes in mechanical properties, thermal strain, irradiation-induced dimensional change and creep in the fast neutron irradiation environment. Verification analyses were carried out in order to prove the validity of this code based on the irradiation tests of the 8th OGL-1 fuel assembly and the fuel element of the Peach Bottom reactor. This report describes the outline of the VIENUS code and its verification analyses. (author)

Results and future plans for the innovative basic research on high temperature engineering

International Nuclear Information System (INIS)

2001-05-01

The High Temperature Engineering Test Reactor (HTTR) is under the rise-to-power stage at the Oarai Research Establishment of Japan Atomic Energy Research Institute (JAERI). This reactor is aimed not only at establishment of the infrastructural technology on high temperature gas-cooled reactor and its upgrading, but also at promotion of the innovative basic research on high temperature engineering. The research is a series of innovative high-temperature irradiation studies, making the best use of the characteristic of the HTTR that it provides a very wide irradiation space at high temperatures. The JAERI has been conducting preliminary tests of the innovative research since 1994, in collaboration with universities and other research institutes, in the fields of 1) new materials development, 2) high temperature radiation chemistry and fusion-related research, and 3) high temperature irradiation techniques and other nuclear research. The HTTR Utilization Research Committee has been examining the results and methodology of the preliminary tests and the future plans, as well as examining the preparatory arrangements of facilities for the HTTR irradiation and post-irradiation examinations. This report presents a summary of results of the preliminary tests and preparatory arrangements for about seven years, together with an outline of the future plans. (author)

High Temperature Degradation Behavior and its Mechanical Properties of Inconel 617 alloy for Intermediate Heat Exchanger of VHTR

International Nuclear Information System (INIS)

Jo, Tae Sun; Kim, Se Hoon; Kim, Young Do; Park, Ji Yeon

2008-01-01

Inconel 617 alloy is a candidate material of intermediate heat exchanger (IHX) and hot gas duct (HGD) for very high temperature reactor (VHTR) because of its excellent strength, creep-rupture strength, stability and oxidation resistance at high temperature. Among the alloying elements in Inconel 617, chromium (Cr) and aluminum (Al) can form dense oxide that act as a protective surface layer against degradation. This alloy supports severe operating conditions of pressure over 8 MPa and 950 .deg. C in He gas with some impurities. Thus, high temperature stability of Inconel 617 is very important. In this work, the oxidation behavior of Inconel 617 alloy was studied by exposure at high temperature and was discussed the high temperature degradation behavior with microstructural changes during the surface oxidation

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Long-term creep behavior of high-temperature gas turbine materials under constant and variable stress

International Nuclear Information System (INIS)

Granacher, J.; Preussler, T.

1987-01-01

Within the framework of the documented research project, extensive creep rupture tests were carried out with characteristic, high-temperature gas turbine materials for establishment of improved design data. In the range of the main application temperatures and in stress ranges down to application-relevant values the tests extended over a period of about 40,000 hours. In addition, long-term annealing tests were carried out in the most important temperature ranges for the measurement of the density-dependent straim, which almost always manifested itself as a material contraction. Furthermore, hot tensile tests were carried out for the description of the elastoplastic short-term behavior. Several creep curves were derived from the results of the different tests with a differentiated evaluation method. On the basis of these creep curves, creep equations were set up for a series of materials which are valid in the entire examined temperature range and stress range and up to the end of the secondary creep range. Also, equations for the time-temperature-dependent description of the material contraction behavior were derived. With these equations, the high-temperature deformation behavior of the examined materials under constant creep stress can be described simply and application-oriented. (orig.) With 109 figs., 19 tabs., 77 refs [de

Temperature Effects on the Mechanical Properties of Candidate SNS Target Container Materials after Proton and Neutron Irradiation; TOPICAL

International Nuclear Information System (INIS)

Byun, T.S.

2001-01-01

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 to 2.53 dpa. Irradiation temperatures were in the range 30 to 100 C. Tensile testing was performed at room temperature (20 C) and 164 C to study the effects of test temperature on the tensile properties. Test materials displayed significant radiation-induced hardening and loss of ductility due to irradiation. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative strain hardening. In the EC316LN stainless steel, increasing the test temperature from 20 C to 164 C decreased the strength by 13 to 18% and the ductility by 8 to 36%. The tensile data for the EC316LN stainless steel irradiated in spallation conditions were in line with the values in a database for 316 stainless steels for doses up to 1 dpa irradiated in fission reactors at temperatures below 200 C. However, extra strengthening induced by helium and hydrogen contents is evident in some specimens irradiated to above about 1 dpa. 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. It was estimated that the 316 stainless steels would retain more than 1% true stains to necking at 164 C after irradiation to 5 dpa. A calculation using reduction of area (RA) measurements and stress-strain data predicted positive strain hardening during plastic instability

Analysis of high school students' perception and attitude toward irradiated food

Energy Technology Data Exchange (ETDEWEB)

Choi, Yoon Seok; Han, Eun Ok [Dept. of Education and Research, Korea Academy of Nuclear Safety, Seoul (Korea, Republic of)

2014-04-15

We chose high school students, who are expected to show significant response to education, to try to understand perception and behavior toward irradiated food, and derive evidential materials for education. High school students who had ever acquired information or received education on irradiated food, and students who had ever purchased or eaten irradiated food, tended to have aproper attitude regarding its necessity, safety, and purchase without prejudice. It is necessary to provide information and education to high school students. Additionally, exposure to the irradiated food could be helpful in changing perception and attitude toward irradiated food.

Analysis of high school students' perception and attitude toward irradiated food

International Nuclear Information System (INIS)

Choi, Yoon Seok; Han, Eun Ok

2014-01-01

We chose high school students, who are expected to show significant response to education, to try to understand perception and behavior toward irradiated food, and derive evidential materials for education. High school students who had ever acquired information or received education on irradiated food, and students who had ever purchased or eaten irradiated food, tended to have aproper attitude regarding its necessity, safety, and purchase without prejudice. It is necessary to provide information and education to high school students. Additionally, exposure to the irradiated food could be helpful in changing perception and attitude toward irradiated food

Spin trapping of radicals formed in gamma-irradiated methanol: effect of the irradiation temperature from 77K to 300K

International Nuclear Information System (INIS)

Schlick, S.; Kevan, L.

1976-01-01

The neutral radicals formed in gamma-irradiated methanol were studied by spin trapping with phenyl-t-butylnitrone (PBN) in an attempt to probe the primary neutral radicals formed. In the temperature range from approximately 157 K to 300 K both CH 2 OH and CH 3 O spin adducts are observed and their limiting ratio at high PBN concentrations is CH 2 OH/CH 3 O=1.5 over this temperature range. Below approximately 157 K this ratio increases exponentially with decreasing temperature with an apparent activation energy of 5.8 kJ/mole (1.4 kcal/mole); this is consistent with the finding that only CH 2 OH radicals are formed by gamma radiolysis at 77 K. Several possible models for the primary neutral radicals formed in gamma-irradiated methanol and their subsequent reactions as a function of irradiation temperature are discussed. It is suggested that the primary radical formation mechanisms are similar in the gas and liquid phases and become temperature dependent when molecular motion is arrested in the solid. (Auth.)

The irradiation creep characteristics of graphite to high fluences

International Nuclear Information System (INIS)

Kennedy, C.R.; Cundy, M.; Kleist, G.

1988-01-01

High-temperature gas-cooled reactors (HTGR) have massive blocks of graphite with thermal and neutron-flux gradients causing high internal stresses. Thermal stresses are transient; however, stresses generated by differential growth due to neutron damage continue to increase with time. Fortunately, graphite also experiences creep under irradiation allowing relaxation of stresses to nominally safe levels. Because of complexity of irradiation creep experiments, data demonstrating this phenomenon are generally limited to fairly low fluences compared to the overall fluences expected in most reactors. Notable exceptions have been experiments at 300/degree/C and 500/degree/C run at Petten under tension and compression creep stresses to fluences greater than 4 /times/ 10 26 (E > 50 keV) neutrons/m 2 . This study complements the previous results by extending the irradiation temperature to 900/degree/C. 2 refs., 3 figs

Electrical treeing behaviors in silicone rubber under an impulse voltage considering high temperature

Science.gov (United States)

Yunxiao, ZHANG; Yuanxiang, ZHOU; Ling, ZHANG; Zhen, LIN; Jie, LIU; Zhongliu, ZHOU

2018-05-01

In this paper, work was conducted to reveal electrical tree behaviors (initiation and propagation) of silicone rubber (SIR) under an impulse voltage with high temperature. Impulse frequencies ranging from 10 Hz to 1 kHz were applied and the temperature was controlled between 30 °C and 90 °C. Experimental results show that tree initiation voltage decreases with increasing pulse frequency, and the descending amplitude is different in different frequency bands. As the pulse frequency increases, more frequent partial discharges occur in the channel, increasing the tree growth rate and the final shape intensity. As for temperature, the initiation voltage decreases and the tree shape becomes denser as the temperature gets higher. Based on differential scanning calorimetry results, we believe that partial segment relaxation of SIR at high temperature leads to a decrease in the initiation voltage. However, the tree growth rate decreases with increasing temperature. Carbonization deposition in the channel under high temperature was observed under microscope and proven by Raman analysis. Different tree growth models considering tree channel characteristics are proposed. It is believed that increasing the conductivity in the tree channel restrains the partial discharge, holding back the tree growth at high temperature.

Damage accumulation in MgO irradiated with MeV Au ions at elevated temperatures

Energy Technology Data Exchange (ETDEWEB)

Bachiller-Perea, Diana, E-mail: dianabachillerperea@gmail.com [Centre de Sciences Nucléaires et de Sciences de la Matière (CSNSM), Univ. Paris-Sud, CNRS-IN2P3, Université Paris-Saclay, 91405, Orsay Cedex (France); Centro de Micro-Análisis de Materiales, Universidad Autónoma de Madrid, C/Faraday 3, 28049, Madrid (Spain); Dpto. de Física Aplicada, Universidad Autónoma de Madrid, Ciudad Universitaria de Cantoblanco, 28049, Madrid (Spain); Debelle, Aurélien, E-mail: aurelien.debelle@u-psud.fr [Centre de Sciences Nucléaires et de Sciences de la Matière (CSNSM), Univ. Paris-Sud, CNRS-IN2P3, Université Paris-Saclay, 91405, Orsay Cedex (France); Thomé, Lionel [Centre de Sciences Nucléaires et de Sciences de la Matière (CSNSM), Univ. Paris-Sud, CNRS-IN2P3, Université Paris-Saclay, 91405, Orsay Cedex (France); Behar, Moni [Instituto de Física, Universidade Federal do Rio Grande do Sul, C.P. 15051, 91501-970, Porto Alegre, RS (Brazil)

2016-09-15

The damage accumulation process in MgO single crystals under medium-energy heavy ion irradiation (1.2 MeV Au) at fluences up to 4 × 10{sup 14} cm{sup −2} has been studied at three different temperatures: 573, 773, and 1073 K. Disorder depth profiles have been determined through the use of the Rutherford backscattering spectrometry in channeling configuration (RBS/C). The analysis of the RBS/C data reveals two steps in the MgO damage process, irrespective of the temperature. However, we find that for increasing irradiation temperature, the damage level decreases and the fluence at which the second step takes place increases. A shift of the damage peak at increasing fluence is observed for the three temperatures, although the position of the peak depends on the temperature. These results can be explained by an enhanced defect mobility which facilitates defect migration and may favor defect annealing. X-ray diffraction reciprocal space maps confirm the results obtained with the RBS/C technique. - Highlights: • High-temperature MeV-ion irradiated MgO exhibits a two-step damage process. • The occurrence of the second step is delayed with increasing temperature. • The damage level decreases with increasing temperature. • A shift of the damage peak is observed with increasing fluence. • A high defect mobility at high temperatures in MgO is clearly evidenced.

Temperature of loose coated particles in irradiation tests

International Nuclear Information System (INIS)

Conlin, J.A.

1975-04-01

An analysis is presented of the temperature of a monolayer bed of loose High-Temperature Gas-Cooled Reactor (HTGR) type fissioning fuel particles in an annular cavity. Both conduction and radiant heat transfer are taken into account, and the effect of particle contact with the annular cavity surfaces is evaluated. Charts are included for the determination of the maximum surface temperature of the particle coating for any size particle or power generation rate in a fuel bed of this type. The charts are intended for the design and evaluation of irradiation experiments on loose beds of coated fuel particles of the type used in HTGRs. Included in an Appendix is a method for estimating the temperature of a particle in circular hole. (U.S.)

A master curve-mechanism based approach to modeling the effects of constraint, loading rate and irradiation on the toughness-temperature behavior of a V-4Cr-4Ti alloy

International Nuclear Information System (INIS)

Odette, G.R.; Donahue, E.; Lucas, G.E.; Sheckherd, J.W.

1996-01-01

The influence of loading rate and constraint on the effective fracture toughness as a function of temperature [K e (T)] of the fusion program heat of V-4Cr-4Ti was measured using subsized, three point bend specimens. The constitutive behavior was characterized as a function of temperature and strain rate using small tensile specimens. Data in the literature on this alloy was also analysed to determine the effect of irradiation on K e (T) and the energy temperature (E-T) curves measured in subsized Charpy V-notch tests. It was found that V-4Cr-4Ti undergoes open-quotes normalclose quotes stress-controlled cleavage fracture below a temperature marking a sharp ductile-to-brittle transition. The transition temperature is increased by higher loading rates, irradiation hardening and triaxial constraint. Shifts in a reference transition temperature due to higher loading rates and irradiation can be reasonably predicted by a simple equivalent yield stress model. These results also suggest that size and geometry effects, which mediate constraint, can be modeled by combining local critical stressed area σ*/A* fracture criteria with finite element method simulations of crack tip stress fields. The fundamental understanding reflected in these models will be needed to develop K e (T) curves for a range of loading rates, irradiation conditions, structural size scales and geometries relying (in large part) on small specimen tests. Indeed, it may be possible to develop a master K e (T) curve-shift method to account for these variables. Such reliable and flexible failure assessment methods are critical to the design and safe operation of defect tolerant vanadium structures

Analysis of the behavior under irradiation of high burnup nuclear fuels with the computer programs FRAPCON and FRAPTRAN

Energy Technology Data Exchange (ETDEWEB)

Reis, Regis; Silva, Antonio Teixeira e, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

The objective of this paper is to verify the validity and accuracy of the results provided by computer programs FRAPCON-3.4a and FRAPTRAN-1.4, used in the simulation process of the irradiation behavior of Pressurized Water Reactors (PWR) fuel rods, in steady-state and transient operational conditions at high burnup. To achieve this goal, the results provided by these computer simulations are compared with experimental data available in the database FUMEX III. Through the results, it was found that the computer programs used have a good ability to predict the operational behavior of PWR fuel rods in high burnup steady-state conditions and under the influence of Reactivity Initiated Accident (RIA). (author)

The proceedings of 1993-workshop on 'development and application of facilities for low temperature irradiation as well as controlled irradiation'

International Nuclear Information System (INIS)

Kuramoto, Eiichi; Okada, Moritami

1993-03-01

This is the proceedings of 1992-workshop of the working group on 'Development and Application of Facilities for Low Temperature Irradiation as well as Controlled Irradiation' held at the Research Reactor Institute of Kyoto University on February 23 and 24, 1993. In this workshop until now, studies on irradiation effects in many materials irradiated at lower and higher temperatures have been reported. It has been clearly defined that a careful choice of irradiation conditions is most important. At the present time, a setting plan of exactly controlled irradiation facility, which is able to irradiate with higher temperatures, is in progress. On the other hand, a plan of vertical low temperature irradiation facility has not yet been performed for lack of funds. In last year, a middle scale plan of low temperature irradiation facility, which is possible to irradiate a fast-neutron dose above 10 17 n/cm 2 at about 5K, was proposed in this workshop. In this proceedings, the advanced facility is required to construct to the KUR as soon as possible by many of the workshop members. (author)

Characterization of mechanical properties and microstructure of highly irradiated SS 316

Energy Technology Data Exchange (ETDEWEB)

Karthik, V., E-mail: karthik@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Kumar, RanVijay; Vijayaragavan, A.; Venkiteswaran, C.N.; Anandaraj, V.; Parameswaran, P.; Saroja, S.; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.; Jayakumar, T.; Raj, Baldev [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

2013-08-15

Cold worked austenitic stainless steel type AISI 316 is used as the material for fuel cladding and wrapper of the Fast Breeder Test Reactor (FBTR), India. The evaluation of mechanical properties of these core structurals is very essential to assess its integrity and ensure safe and productive operation of FBTR to very high burn-ups. The changes in the mechanical properties of these core structurals are associated with microstructural changes caused by high fluence neutron irradiation and temperatures of 673–823 K. Remote tensile testing has been used for evaluating the tensile properties of irradiated clad tubes and shear punch test using small disk specimens for evaluating the properties of irradiated hexagonal wrapper. This paper will highlight the methods employed for evaluating the mechanical properties of the irradiated cladding and wrapper and discuss the trends in properties as a function of dpa (displacement per atom) and irradiation temperature.

Effects of irradiation at low temperature on V-4Cr-4Ti

International Nuclear Information System (INIS)

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

1996-01-01

Irradiation at low temperatures (100 to 275 degrees C) to 0.5 dpa causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile flow properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform ductility. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275 degrees C. This embrittlement is not due to pickup of O or other interstitial solutes during the irradiation

Effects of irradiation at low temperature on V-4Cr-4Ti

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

Irradiation at low temperatures (100 to 275{degrees}C) to 0.5 dpa causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile flow properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform ductility. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275{degrees}C. This embrittlement is not due to pickup of O or other interstitial solutes during the irradiation.

Damage structures in fission-neutron irradiated Ni-based alloys at high temperatures

Science.gov (United States)

Yamakawa, K.; Shimomura, Y.

1999-01-01

The defects formed in Ni based (Ni-Si, Ni-Cu and Ni-Fe) alloys which were irradiated with fission-neutrons were examined by electron microscopy. Irradiations were carried out at 473 K and 573 K. In the 473 K irradiated specimens, a high density of large interstitial loops and small vacancy clusters with stacking fault tetrahedra (SFT) were observed. The number densities of these two types of defects did not strongly depend on the amount of solute atoms in each alloy. The density of the loops in Ni-Si alloys was much higher than those in Ni-Cu and Ni-Fe alloys, while the density of SFT only slightly depended on the kind of solute. Also, the size of the loops depended on the kinds and amounts of solute. In 573 K irradiated Ni-Cu specimens, a high density of dislocation lines developed during the growth of interstitial loops. In Ni-Si alloys, the number density and size of the interstitial loops changed as a function of the amount of solute. Voids were formed in Ni-Cu alloys but scarcely formed in Ni-Si alloys. The number density of voids was one hundredth of that of SFT observed in 473 K irradiated Ni-Cu alloys. Possible formation processes of interstitial loops, SFT dislocation lines and voids are discussed.

Damage structures in fission-neutron irradiated Ni-based alloys at high temperatures

Energy Technology Data Exchange (ETDEWEB)

Yamakawa, K.; Shimomura, Y. [Hiroshima Univ. (Japan). Faculty of Engineering

1999-01-01

The defects formed in Ni based (Ni-Si, Ni-Cu and Ni-Fe) alloys which were irradiated with fission-neutrons were examined by electron microscopy. Irradiations were carried out at 473 K and 573 K. In the 473 K irradiated specimens, a high density of large interstitial loops and small vacancy clusters with stacking fault tetrahedra (SFT) were observed. The number densities of these two types of defects did not strongly depend on the amount of solute atoms in each alloy. The density of the loops in Ni-Si alloys was much higher than those in Ni-Cu and Ni-Fe alloys, while the density of SFT only slightly depended on the kind of solute. Also, the size of the loops depended on the kinds and amounts of solute. In 573 K irradiated Ni-Cu specimens, a high density of dislocation lines developed during the growth of interstitial loops. In Ni-Si alloys, the number density and size of the interstitial loops changed as a function of the amount of solute. Voids were formed in Ni-Cu alloys but scarcely formed in Ni-Si alloys. The number density of voids was one hundredth of that of SFT observed in 473 K irradiated Ni-Cu alloys. Possible formation processes of interstitial loops, SFT, dislocation lines and voids are discussed. (orig.) 8 refs.

Irradiation behavior of graphite shielding materials for FBR

International Nuclear Information System (INIS)

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

1994-01-01

Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor 'JOYO' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597degC. Postirradiation examination was carried out on dimensional change, elastic modulus, and the thermal conductivity. The result of measurement of dimensional change indicated that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased to two to three times of unirradiated values. A large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependency on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, but the change in specific heat was negligibly small. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (author)

High temperature oxidation behavior of SiC coating in TRISO coated particles

International Nuclear Information System (INIS)

Liu, Rongzheng; Liu, Bing; Zhang, Kaihong; Liu, Malin; Shao, Youlin; Tang, Chunhe

2014-01-01

Highlights: • High temperature oxidation tests of SiC coating in TRISO particles were carried out. • The dynamic oxidation process was established. • Oxidation mechanisms were proposed. • The existence of silicon oxycarbides at the SiO 2 /SiC interface was demonstrated. • Carbon was detected at the interface at high temperatures and long oxidation time. - Abstract: High temperature oxidation behavior of SiC coatings in tristructural-isotropic (TRISO) coated particles is crucial to the in-pile safety of fuel particles for a high temperature gas cooled reactor (HTGR). The postulated accident condition of air ingress was taken into account in evaluating the reliability of the SiC layer. Oxidation tests of SiC coatings were carried out in the ranges of temperature between 800 and 1600 °C and time between 1 and 48 h in air atmosphere. Based on the microstructure evolution of the oxide layer, the mechanisms and kinetics of the oxidation process were proposed. The existence of silicon oxycarbides (SiO x C y ) at the SiO 2 /SiC interface was demonstrated by X-ray photospectroscopy (XPS) analysis. Carbon was detected by Raman spectroscopy at the interface under conditions of very high temperatures and long oxidation time. From oxidation kinetics calculation, activation energies were 145 kJ/mol and 352 kJ/mol for the temperature ranges of 1200–1500 °C and 1550–1600 °C, respectively

Evaluation of rheological behavior and sensory properties of ready to eat salad dressings submitted to irradiation

International Nuclear Information System (INIS)

Gallo, Juliana Maria Altavista Sagretti

2013-01-01

This study evaluated the rheological behavior of 11 salad dressings ready to eat, processed by gamma irradiation, at doses: 3 kGy and 5 kGy. This assessment was made by rheograms traced by measurements of viscosity and shear stress by shear rate. Mathematical parameters obtained from these measures also contributed to the conclusion of the behavior exhibited and for choosing the best mathematical model applicable to them. The measurements were performed on a Brookfield viscometer Model LVDV - III. The Neslab thermal bath was used to maintain ambient temperature during the whole analysis. This study protocol was done in two stages. In step 1, the evaluation was performed soon after the samples have been irradiated when these were close to their dates of manufacture. In phase 2 the samples were reassessed after a period of storage, near the expiration date of their validity. Simultaneously pH measurement was performed to evaluate their stability in the face of treatment and sensory analyzes of two salad dressings were studied, in order to verify the acceptance of these when submitted to irradiation. The shear-thinning behavior was confirmed for all dressings through the mathematical model, the power law, that best model apply to this, in both phases. The irradiation at the absorbed doses studied did not influence this behavior. The results of sensory evaluation indicated good acceptance of irradiated sauces for the tasters. (author)

A study of silver behavior in Gas-turbine High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Sawa, Kazuhiro; Tanaka, Toshiyuki

1995-11-01

A Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR) is one of the promising reactor systems of future HTGRs. In the design of GT-HTGR, behavior of fission products, especially of silver, is considered to be important from the view point of maintenance of gas-turbine. A study of silver behavior in the GT-HTGR was carried out based on current knowledge. The purposes of this study were to determine an importance of the silver problem quantitatively, countermeasures to the problem and items of future research and development which will be needed. In this study, inventory, fractional release from fuel, plateout in the primary circuit and radiation dose were evaluated, respectively. Based on this study, it is predicted that gamma-ray from plateout silver in gas-turbine system contributes about a half of total radiation dose after reactor shutdown. In future, more detail data for silver release from fuel, plateout behavior, etc. using the High Temperature Engineering Test Reactor (HTTR), for example, will be needed to carry out reasonable design. (author)

Irradiation effects in tungsten-copper laminate composite

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-01

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

Corrosion of High Chromium Ferritic/Martensitic Steels in High Temperature Water. a Literature Review

Energy Technology Data Exchange (ETDEWEB)

Fernandez, P.; Lapena, J.; Blazquez, F. [Ciemat, Madrid (Spain)

2000-07-01

Available literature concerning corrosion of high-chromium ferritic/martensitic steels in high temperature water has been reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS) are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. The data on general corrosion indicate moderate corrosion rates in high temperature water up to 350 degree centigree. Considerably larger corrosion rates were observed under neutron irradiation. The works concerning to the behaviour of these alloys to stress corrosion cracking seem to conclude that in these materials is necessary to optimize the temper temperature and to carry out the post-weld heat treatments properly in order to avoid stress corrosion cracking. (Author) 40 refs.

Corrosion of High Chromium Ferritic/Martensitic Steels in High Temperature Water. a Literature Review

International Nuclear Information System (INIS)

Fernandez, P.; Lapena, J.; Blazquez, F.

2000-01-01

Available literature concerning corrosion of high-chromium ferritic/martensitic steels in high temperature water has been reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS) are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. The data on general corrosion indicate moderate corrosion rates in high temperature water up to 350 degree centigrade. Considerably larger corrosion rates were observed under neutron irradiation. The works concerning to the behaviour of these alloys to stress corrosion cracking seem to conclude that in these materials is necessary to optimize the temper temperature and to carry out the post-weld heat treatments properly in order to avoid stress corrosion cracking. (Author) 40 refs

Influence of nickel and beryllium content on swelling behavior of copper irradiated with fast neutrons

Energy Technology Data Exchange (ETDEWEB)

Singh, B.N.; Garner, F.A.; Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Evans, J.H.

1996-10-01

In the 1970`s, the effects of nickel content on the evolution of dislocation microstructures and the formation and growth of voids in Cu-Ni alloys were studied using 1 MeV electrons in a high voltage electron microscope. The swelling rate was found to decrease rapidly with increasing nickel content. The decrease in the swelling rate was associated with a decreasing void growth rate with increasing nickel content at irradiation temperatures up to 450{degrees}C. At 500{degrees}C, both void size and swelling rate were found to peak at 1 and 2% Ni, respectively, and then to decrease rapidly with increasing nickel content. However, recent work has demonstrated that the swelling behavior of Cu-5%Ni irradiated with fission neutrons is very similar for that of pure copper. The present experiments were designed to investigate this apparent discrepancy.

Low-temperature annealing of radiation defects in electron-irradiated gallium phosphide

International Nuclear Information System (INIS)

Kolb, A.A.; Megela, I.G.; Buturlakin, A.P.; Goyer, D.B.

1990-01-01

The isochronal annealing of radiation defects in high-energy electron irradiated n-GaP monocrystals within the 77 to 300 K range has been investigated by optical and electrical techniques. The changes in conductance and charge carrier mobility as functions of annealing temperature as well as the variation of optical absorption spectra of GaP under irradiation and annealing provide evidence that most of radiation defects are likely secondary complexes of defects

Void nucleation at elevated temperatures under cascade-damage irradiation

International Nuclear Information System (INIS)

Semenov, A.A.; Woo, C.H.

2002-01-01

The effects on void nucleation of fluctuations respectively due to the randomness of point-defect migratory jumps, the random generation of free point defects in discrete packages, and the fluctuating rate of vacancy emission from voids are considered. It was found that effects of the cascade-induced fluctuations are significant only at sufficiently high total sink strength. At lower sink strengths and elevated temperatures, the fluctuation in the rate of vacancy emission is the dominant factor. Application of the present theory to the void nucleation in annealed pure copper neutron-irradiated at elevated temperatures with doses of 10 -4 -10 -2 NRT dpa showed reasonable agreement between theory and experiment. This application also predicts correctly the temporal development of large-scale spatial heterogeneous microstructure during the void nucleation stage. Comparison between calculated and experimental void nucleation rates in neutron-irradiated molybdenum at temperatures where vacancy emission from voids is negligible showed reasonable agreement as well. It was clearly demonstrated that the athermal shrinkage of relatively large voids experimentally observable in molybdenum at such temperatures may be easily explained in the framework of the present theory

The tensile behavior of Ti36Ni49Hf15 high temperature shape memory alloy

International Nuclear Information System (INIS)

Wang, Y.Q.; Zheng, Y.F.; Cai, W.; Zhao, L.C.

1999-01-01

Recently, ternary Ti-Ni-Hf alloys have attracted great interest in the field of high temperature shape memory materials research and development. Extensive studies have been made on its manufacture process, constitutional phases, phase transformation behavior, the structure, substructure and interface structure of martensite and the precipitation behavior during ageing. Yet up to date there is no report about the fundamental mechanical properties of Ti-Ni-Hf alloys, such as the stress-strain data, the variation laws of the yield strength and elongation with the temperature. In the present study, tensile tests at various temperatures are employed to investigate the mechanical behavior of Ti-Ni-Hf alloy with different matrix structures, from full martensite to full parent phase structure, with the corresponding deformation mechanism discussed

Modeling of High Temperature Oxidation Behavior of FeCrAl Alloy by using Artificial Neural Network

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Joon; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Refractory alloys are candidate materials for replacing current zirconium-base cladding of light water reactors and they retain significant creep resistance and mechanical strength at high temperatures up to 1500 ℃ due to their high melting temperature. Thermal neutron cross sections of refractory metals are higher than that of zirconium, however the loss of neutron can be overcome by reducing cladding thickness which can be facilitated with enhanced mechanical properties. However, most refractory metals show the poor oxidation resistance at a high temperature. Oxidation behaviors of the various compositions of FeCrAl alloys in high temperature conditions were modeled by using Bayesian neural network. The automatic relevance determination (ARD) technique represented the influence of the composition of alloying elements on the oxidation resistance of FeCrAl alloys. This model can be utilized to understand the tendency of oxidation behavior along the composition of each element and prove the applicability of neural network modeling for the development of new cladding material of light water reactors.

low temperature irradiation effects in iron-alloys and ceramics

International Nuclear Information System (INIS)

Kuramoto, Eiichi; Abe, Hironobu; Tanaka, Minoru; Nishi, Kazuya; Tomiyama, Noriyuki.

1991-01-01

Electron beam irradiation at 77K and neutron irradiation at 20K were carried out on Fe-Cr and Fe-Cr-Ni alloys and ZnO and graphite system ceramics, and by measuring positron annihilation lifetime, the micro-information about irradiation-introduced defects was obtained. The temperature of the movement of atomic vacancies in pure iron is about 200K, but it was clarified that by the addition of Cr, it was not much affected. However, in the case of high concentration Cr alloys, the number of atomic vacancies which take part in the formation of micro-voids decreased as compared with the case of pure iron. It is considered that among the irradiation defects of ZnO, O-vac. restored below 300degC. It is considered that in the samples without irradiation, the stage of restoration exists around 550degC, which copes with structural defects. By the measurement of graphite without irradiation, the positron annihilation lifetime corresponding with the interface of matrix and crystal grains, grain boundaries and internal surfaces was almost determined. The materials taken up most actively in the research and development of nuclear fusion reactor materials are austenitic and ferritic stainless steels, and their irradiation defects have been studied. (K.I.)

Simulation of the irradiation-induced micro-thermo-mechanical behaviors evolution in ADS nuclear fuel pellets

Science.gov (United States)

Ding, Shurong; Zhao, Yunmei; Wan, Jibo; Gong, Xin; Wang, Canglong; Yang, Lei; Huo, Yongzhong

2013-11-01

An Accelerator Driven System (ADS) is dedicated to Minor Actinides (MA) transmutation. The fuels for ADS are highly innovative, which are composite fuel pellets with the fuel particles containing MA phases dispersed in a MgO or Mo matrix. Assuming that the fuel particles are distributed periodically in the MgO matrix, a three-dimensional finite element model is developed. The three-dimensional incremental large-deformation constitutive relations for the fuel particles and matrix are separately built, and a method is accordingly constructed to implement simulation of the micro-thermo-mechanical behaviors evolution. Evolutions of the temperature and mechanical fields are given and discussed. With irradiation creep included in the MgO matrix constitutive relation, the conclusions can be drawn as that (1) irradiation creep has a remarkable effect on the mechanical behaviors evolution in the matrix; (2) irradiation creep plays an important role in the damage mechanism interpretation of ceramic matrix fuel pellets. Thermal conductivity The thermal conductivity model is adopted as KUO2 = K0·FD·FP·FM·FR, which was proposed by Lucuta et al. [10] to adapt to the high burnup conditions with consideration of the effects of temperature, burnup, porosity and fission products. K0 is the thermal conductivity of fully dense un-irradiated UO2, as Eq. (1) in W/m K; FD, FP are the adjust factors reflecting the effects of dissolved and precipitated fission products; FM and FR are factors due to porosity and irradiation effects. The adopted thermal conductivity varies with temperature and burnup, which expresses its degradation with burnup, with the terms as k0={1}/{0.0375+2.165×10-4T}+{4.715×109}/{T2}exp-{16361}/{T} FD={1.09}/{B3.265}+{0.0643}/{√{B}}√{T}artan{1}/{1.09/B3.265}+{0.0643}/{√{B}}√{T} FP=1+0.019B/3-0.019B{1}/{1+exp(1200-T100)} FM={1-P}/{1+(s-1)P} FR=1-{0.2}/{1+expT-90080} Thermal expansion The engineering strain of thermal expansion [11] is given as {ΔL}/{L0

High temperature strength and aging behavior of 12%Cr-15%Mn austenitic steels

International Nuclear Information System (INIS)

Miyahara, Kazuya; Bae, Dong-Su; Sakai, Hidenori; Hosoi, Yuzo

1993-01-01

High Mn-Cr austenitic steels are still considered to be an important high temperature structural material from the point of view of reduced radio-activation. The objective of the present study is to make a fundamental research of mechanical properties and microstructure of 12%Cr-15%Mn austenitic steels. Especially the effects of alloying elements of V and Ti on the mechanical properties and microstructure evolution of high Mn-Cr steels were studied. Precipitation behaviors of carbides, nitrides and σ phase are investigated and their remarkable effects on the high temperature strength are found. The addition of V was very effective for strengthening the materials with the precipitation of fine VN. Ti was also found to be beneficial for the improvement of high temperature strength properties. The results of high temperature strengths of the 12Cr-15Mn austenitic steels were compared with those of the other candidate and/or reference materials, for example, JFMS (modified 9Cr-2Mo ferritic stainless steel) and JPCAs (modified 316 austenitic stainless steels). (author)

X-ray imaging for studying behavior of liquids at high pressures and high temperatures using Paris-Edinburgh press

International Nuclear Information System (INIS)

Kono, Yoshio; Kenney-Benson, Curtis; Park, Changyong; Shen, Guoyin; Shibazaki, Yuki; Wang, Yanbin

2015-01-01

Several X-ray techniques for studying structure, elastic properties, viscosity, and immiscibility of liquids at high pressures have been integrated using a Paris-Edinburgh press at the 16-BM-B beamline of the Advanced Photon Source. Here, we report the development of X-ray imaging techniques suitable for studying behavior of liquids at high pressures and high temperatures. White X-ray radiography allows for imaging phase separation and immiscibility of melts at high pressures, identified not only by density contrast but also by phase contrast imaging in particular for low density contrast liquids such as silicate and carbonate melts. In addition, ultrafast X-ray imaging, at frame rates up to ∼10 5 frames/second (fps) in air and up to ∼10 4 fps in Paris-Edinburgh press, enables us to investigate dynamics of liquids at high pressures. Very low viscosities of melts similar to that of water can be reliably measured. These high-pressure X-ray imaging techniques provide useful tools for understanding behavior of liquids or melts at high pressures and high temperatures

X-ray imaging for studying behavior of liquids at high pressures and high temperatures using Paris-Edinburgh press

Energy Technology Data Exchange (ETDEWEB)

Kono, Yoshio; Kenney-Benson, Curtis; Park, Changyong; Shen, Guoyin [HPCAT, Geophysical Laboratory, Carnegie Institution of Washington, 9700 S. Cass Ave., Argonne, Illinois 60439 (United States); Shibazaki, Yuki [Frontier Research Institute for Interdisciplinary Sciences, Tohoku University, Aramaki aza Aoba 6-3, Aoba-ku, Sendai 980-8578 (Japan); Wang, Yanbin [GeoSoilEnviroCARS, Center for Advanced Radiation Sources, The University of Chicago, 5640 S. Ellis Avenue, Chicago, Illinois 60637 (United States)

2015-07-15

Several X-ray techniques for studying structure, elastic properties, viscosity, and immiscibility of liquids at high pressures have been integrated using a Paris-Edinburgh press at the 16-BM-B beamline of the Advanced Photon Source. Here, we report the development of X-ray imaging techniques suitable for studying behavior of liquids at high pressures and high temperatures. White X-ray radiography allows for imaging phase separation and immiscibility of melts at high pressures, identified not only by density contrast but also by phase contrast imaging in particular for low density contrast liquids such as silicate and carbonate melts. In addition, ultrafast X-ray imaging, at frame rates up to ∼10{sup 5} frames/second (fps) in air and up to ∼10{sup 4} fps in Paris-Edinburgh press, enables us to investigate dynamics of liquids at high pressures. Very low viscosities of melts similar to that of water can be reliably measured. These high-pressure X-ray imaging techniques provide useful tools for understanding behavior of liquids or melts at high pressures and high temperatures.

Post-irradiation effects in CMOS integrated circuits

International Nuclear Information System (INIS)

Zietlow, T.C.; Barnes, C.E.; Morse, T.C.; Grusynski, J.S.; Nakamura, K.; Amram, A.; Wilson, K.T.

1988-01-01

The post-irradiation response of CMOS integrated circuits from three vendors has been measured as a function of temperature and irradiation bias. The author's have found that a worst-case anneal temperature for rebound testing is highly process dependent. At an anneal temperature of 80 0 C, the timing parameters of a 16K SRAM from vendor A quickly saturate at maximum values, and display no further changes at this temperature. At higher temperature, evidence for the anneal of interface state charge is observed. Dynamic bias during irradiation results in the same saturation value for the timing parameters, but the anneal time required to reach this value is longer. CMOS/SOS integrated circuits (vendor B) were also examined, and showed similar behavior, except that the saturation value for the timing parameters was stable up to 105 0 C. After irradiation to 10 Mrad(Si), a 16K SRAM (vendor C) was annealed at 80 0 C. In contrast to the results from the vendor A SRAM, the access time decreased toward prerad values during the anneal. Another part irradiated in the same manner but annealed at room temperature showed a slight increase during the anneal

Deformation and Fracture Properties in Neutron Irradiated Pure Mo and Mo Alloys

International Nuclear Information System (INIS)

Byun, T.S.; Snead, L.; Li, M.; Cockeram, B.V.

2007-01-01

Full text of publication follows: The evolution in microstructural and mechanical properties was investigated for molybdenum and molybdenum alloys after high temperature neutron irradiation. Test materials include oxide dispersion-strengthened (ODS) molybdenum alloy, molybdenum- 0.5% titanium-0.1% zirconium (TZM) alloy, and low carbon arc-cast (LCAC) molybdenum. Tensile specimens were irradiated in high flux isotope reactor (HFIR) at temperatures in the range ∼300 - 1000 deg. C to neutron fluences of 2.28 - 24.7 x 10 25 n/m 2 (E>0.1 MeV) or 1.2-13.1 dpa. Tensile tests were performed at temperatures ranging from -150 deg. C to 1000 deg. C. To evaluate irradiation effects, true stress parameters (yield stress, plastic instability stress, and true fracture stress) and ductility parameters (uniform strain, fracture strain, and reduction area) were compared for both irradiated and non-irradiated materials. Fracture toughness was also evaluated from the fracture stress and fracture strain data using a fracture strain model. The fracture strain was used to determine the ductile-to-brittle transition temperature (DBTT). Results indicate that irradiation in the temperature range of 600 - 800 deg. C hardened the materials by up to 70%, while the irradiation hardening outside this temperature range was much lower (<40%). The plastic instability stress was strongly dependent on test temperature; however, it was nearly independent of irradiation dose and temperature. It was also found that the true fracture stress was dependent on test temperature. The true fracture stress was not significantly influenced by irradiation at elevated and high test temperatures; however, it was decreased significantly at sub-zero temperatures after irradiation due to material embrittlement. The DBTT for 600 deg. C irradiated ODS molybdenum alloy was found to be about room temperature or lower, and among the test materials the ODS alloy showed the highest resistance to irradiation embrittlement

Overall viscoplastic behavior of non-irradiated porous nuclear ceramics

International Nuclear Information System (INIS)

Monerie, Yann; Gatt, Jean-Marie

2006-01-01

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

Development of crack growth and crack initiation test units for stress corrosion cracking examinations in high-temperature water environments under neutron irradiation (1) (Contract research)

International Nuclear Information System (INIS)

Izumo, Hironobu; Ishida, Takuya; Kawamata, Kazuo; Inoue, Shuichi; Ide, Hiroshi; Saito, Takashi; Ishitsuka, Etsuo; Chimi, Yasuhiro; Ise, Hideo; Miwa, Yukio; Ugachi, Hirokazu; Nakano, Junichi; Kaji, Yoshiyuki; Tsukada, Takashi

2009-04-01

To evaluate integrity of irradiation-assisted stress corrosion cracking (IASCC) on in-core structural materials used in light water reactors (LWRs), useful knowledge regarding IASCC has been obtained mainly by post-irradiation examinations (PIEs). In the core of commercial LWRs, however, the actual IASCC occurs under the effects of irradiation on both materials and high-temperature water environment. Therefore, it is necessary to confirm the suitability of the knowledge by PIE with comparison to IASCC behaviors during in-core SCC tests. Fundamental techniques for in-core crack growth and crack initiation tests have been developed already at the Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency (JAEA). For the in-core crack growth test technique, to evaluate the effects of neutron irradiation on stainless steels irradiated to low neutron fluences, it is indispensable to develop new loading technique which is applicable to compact tension (CT) specimens with thickness of 0.5 inch (0.5T), from the viewpoint of validity based on the fracture mechanics. Based on the present technical investigation for the in-core loading technique, it is expected that a target load of 7.6 kN approximately can apply to a 0.5T-CT specimen by adopting a loading unit of a lever type instead of the previous uni-axial tension type. For the in-core crack initiation test technique, moreover, construction of a loading unit adopting linear variable differential transformers (LVDTs) has been investigated and technical issues have examined. (author)

Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

International Nuclear Information System (INIS)

Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

2011-01-01

The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

Incoloy 800 anodic behavior in sulfate and chloride solutions at high temperature

International Nuclear Information System (INIS)

Lafont, C.; Alvarez, M.G.

1992-01-01

The anodic behavior and pitting corrosion resistance of Incoloy 800 in concentrated aqueous chloride and sulphate solutions has been studied by means of electrochemical techniques. The effect of different environmental variables, such as temperature (in the 100 0 C to 280 0 C range) and sulphate ion concentration (0.02 M to 2 M), was evaluated. In another set of experiments, the influence of sulphate ions additions on the pitting resistance and pitting morphology of Incoloy 800 in chloride solutions at high temperature was also examined. (author)

Microstructural evolution and hardening of GH3535 alloy under energetic Xe ion irradiation at room temperature and 650 °C

Science.gov (United States)

Huang, Hefei; Gao, Jie; Radiguet, Bertrand; Liu, Renduo; Li, Jianjian; Lei, Guanhong; Huang, Qing; Liu, Min; Xie, Ruobing

2018-02-01

The GH3535 alloy was irradiated with 7 MeV Xe26+ ions to a dose of 10 dpa at room temperature (RT) and 650 °C, and subsequently examined using Transmission Electron Microscopy (TEM) and nanoindentation. High numbers of nano-sized black dots, identified as dislocation loops were observed in both irradiated samples. The dislocation loops detected at the high temperature irradiated sample (size/number density: 9.5 nm/1.9 × 1021 m-3) were found to be larger in size but less in amount as compared to that of the case of RT irradiation (6.9 nm/18.7 × 1021 m-3). In addition, the large-sized Mo-Cr rich precipitates (16.4 nm/3.7 × 1021 m-3) were observed in the sample irradiated at 650 °C. Moreover, the Xe bubbles, with smaller size (2.9 nm) but higher number density (77.8 × 1021 m-3) among the irradiated induced defects, were also detected in the case of high temperature irradiated sample via the diffusion and aggregation of Xe atoms. Nanoindentaion measurements showed a hardening phenomenon for the irradiated sample, and the hardness increment is higher in the case of high temperature irradiated sample. Dispersed barrier-hardening (DBH) model was applied to predict the hardening produced from the irradiation induced defects. The yield strength increment calculated based on TEM observations and the nanohardness increment measured using nanoindentation are in excellent agreement.

High temperature brazing of reactor materials

International Nuclear Information System (INIS)

Orlov, A.V.; Nechaev, V.A.; Rybkin, B.V.; Ponimash, I.D.

1990-01-01

Application of high-temperature brazing for joining products of such materials as molybdenum, tungsten, zirconium, beryllium, magnesium, nickel and aluminium alloys, graphite ceramics etc. is described. Brazing materials composition and brazed joints properties are presented. A satisfactory strength of brazed joints is detected under reactor operation temperatures and coolant and irradiation effect

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

International Nuclear Information System (INIS)

Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

2002-10-01

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

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

Energy Technology Data Exchange (ETDEWEB)

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

2002-10-01

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

Aspects of Low Temperature Irradiation in Neutron Activation Analysis

International Nuclear Information System (INIS)

Brune, D.

1968-08-01

Neutron irradiation of the sample while frozen in a cooling device inserted in a reactor channel has been carried out in the analysis of iodine in aqueous samples as well as of mercury in biological tissue and water. For the simultaneous irradiation of a large number of aqueous solutions the samples were arranged in a suitable geometry in order to avoid mutual flux perturbation effects. The influence of the neutron temperature on the activation process has been discussed. Potential applications of the low temperature irradiation technique are outlined

Aspects of Low Temperature Irradiation in Neutron Activation Analysis

Energy Technology Data Exchange (ETDEWEB)

Brune, D

1968-08-15

Neutron irradiation of the sample while frozen in a cooling device inserted in a reactor channel has been carried out in the analysis of iodine in aqueous samples as well as of mercury in biological tissue and water. For the simultaneous irradiation of a large number of aqueous solutions the samples were arranged in a suitable geometry in order to avoid mutual flux perturbation effects. The influence of the neutron temperature on the activation process has been discussed. Potential applications of the low temperature irradiation technique are outlined.

Metallographic analysis of irradiated RERTR-3 fuel test specimens

International Nuclear Information System (INIS)

Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

2000-01-01

The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

Nitrogen compounds behavior under irradiation environment

International Nuclear Information System (INIS)

Ichikawa, Nagayoshi; Takagi, Junichi; Yotsuyanagi, Tadasu

1991-01-01

Laboratory experiments were performed to evaluate nitrogen compounds behavior in liquid phase under irradiation environments. Nitrogen compounds take a chemical form of ammonium ion under reducing condition by gamma irradiation, whereas ammonium ions are rather stable even under oxidizing conditions. Key reactions were pointed out and their reaction rate constants and activation energies were estimated through computer code simulation. A reaction scheme for nitrogen compounds including protonate reaction was proposed. (author)

Internal friction measurements of Mo after low-temperature proton irradiation

International Nuclear Information System (INIS)

Tanimoto, H.; Mizubayashi, H.; Masuda, R.; Okuda, S.; Tagishi, Y.

1992-01-01

Internal friction measurements are performed in Mo after 20 MeV proton irradiation in order to clarify the behavior of self-interstitial atoms (SIA's) in Mo. In the low dose range, strong dislocation pinning suggesting the free migration of defects is observed at about 40 K and weak pinning at about 25 K. The features are very similar to those reported after neutron irradiation except that the 25 K pinning is much smaller after proton irradiation. The result suggests that the migration of free SIA's is responsible for the 40 K pinning and that of SIA-defect clusters, probably di-SIA's, formed during irradiation for the 25 K pinning. In the high dose range, the relaxation peaks are observed at about 13 and 41 K, where the close similarities are found between the present peaks and the corresponding peaks reported after neutron irradiation except that the peak height of the 41 K peak per unit concentration of Frenkel pairs (FP) tends to increase strongly with decreasing dose here. The latter fact suggests the strong interaction between SIA's. Then the smallness of the 41 K peak reported after electron irradiation with very high dose could be explained by an increased interaction between SIA's, but not by the two-dimensional migration of SIA's as proposed by Jacques and Robrock. Deformation given prior to irradiation causes a drastic decrease in the modulus defects associated with FP's (so-called bulk effect) and in the 13 K peak height. After neutron irradiation, no such effect of deformation was reported. A possible origin for this difference is discussed. (orig.)

Gas diffusion and temperature dependence of bubble nucleation during irradiation

DEFF Research Database (Denmark)

Foreman, A. J. E.; Singh, Bachu Narain

1986-01-01

The continuous production of gases at relatively high rates under fusion irradiation conditions may enhance the nucleation of cavities. This can cause dimensional changes and could induce embrittlement arising from gas accumulation on grain boundaries. Computer calculations have been made...... of the diatomic nucleation of helium bubbles, assuming helium to diffuse substitutionally, with radiation-enhanced diffusion at lower temperatures. The calculated temperature dependence of the bubble density shows excellent agreement with that observed in 600 MeV proton irradiations, including a reduction...... in activation energy below Tm/2. The coalescence of diatomic nuclei due to Brownian motion markedly improves the agreement and also provides a well-defined terminal density. Bubble nucleation by this mechanism is sufficiently fast to inhibit any appreciable initial loss of gas to grain boundaries during...

Low temperature irradiation facility at Kyoto University Reactor (KUR)

International Nuclear Information System (INIS)

Atobe, Kozo; Okada, Moritami; Yoshida, Hiroyuki; Kodaka, Hisao; Miyata, Kiyomi.

1977-01-01

A new refrigeration system has been substituted to the low temperature irradiation facility at KUR instead of the previous one, since April in 1975. The model 1204 CTi He liquifier was designed to be modified for the refrigerator with the capacity of 30 watts at 10 K. The refrigeration capacity of 38 watts at 10 K was defined using a special cryostat and transfer-tubes, and the lowest temperature of about 18 K was measured using the irradiation loop without reactor operation. The reconstructed facility enables us to hold the many specimens simultaneously in the sample chamber of the irradiation loop at about 25 K during reactor operation of 5 MW. The irradiation dose has been reached about 6.6 x 10 16 n sub(f)/cm 2 and 6.1 x 10 17 n sub(th)/cm 2 with the normal reactor operation cycle of up to 77 hours. The stable operation condition of the machine and the special safety system for the refrigeration system enable us to maintain easily the facility with a constant operation condition for such a long time irradiation. Many kinds of low temperature neutron irradiation experiments are carried out using the facility, which techniques are partially reported. (auth.)

The effect of helium, radiation damage and irradiation temperature on the mechanical properties of beryllium

Energy Technology Data Exchange (ETDEWEB)

Fabritsiev, S.A. [D.V. Efremov Scientific Research Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S.

1998-01-01

In this work different RF beryllium grades were irradiated in the BOR-60 reactor to a dose of {approx}5-10 dpa at irradiation temperatures 350, 420, 500, 800degC. Irradiation at temperatures of 350-400degC is shown to result in Be hardening due to the accumulation of radiation defect complexes. Hardening is accompanied with a sharp drop in plasticity at T{sub test} {<=} 300degC. A strong anisotropy in plasticity has been found at a mechanical testing temperature of 400degC and this parameter may be preferable when the samples are cut crosswise to the pressing direction. High-temperature irradiation (T{sub irr} = 780degC) gives rise to large helium pores over the grain boundaries and smaller pores in the grain body. Fracture is brittle and intercrystallite at T{sub test} {>=} 600degC. Helium embrittlement is accompanied as well with a drop in the Be strength properties. (author)

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

International Nuclear Information System (INIS)

Miyashita, Makoto; Sugimoto, Masayoshi; Yutani, Toshiaki

2007-03-01

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

Effect of the irradiation temperature and relative humidity on PVG dosifilm

International Nuclear Information System (INIS)

Jia Haishun; Chen Wenxiu; Shen Yuxin

1999-01-01

The effect of environmental factors, such as irradiation temperature and relative humidity, on the PVG dosifilm irradiated by EB was tested. Experiments show that the temperature coefficient of irradiated PVG dosifilm was 0.008 deg. C -1 from 20 deg. C to 55 deg. C, and the humidity coefficient was 0.006 per r.h. (%) from r.h. 0% to 76%. The PVG dosifilm can be used as a routine dosimeter for dose measurement for low-energy EB processing. The absorbed dose values for various irradiation temperature and humidity can be corrected based on experimental data. (author)

Can prenatal low-dose irradiation affect behavior of adult rats?

International Nuclear Information System (INIS)

Smajda, B.; Tomasova, L.; Kokocova, N.

2011-01-01

The aim of our study was to determine whether exposure of very low dose gamma-rays during the critical phase of brain development affects some selected behavioral parameters in adult rats. Pregnant female Wistar rats were irradiated with 1 Gy gamma-rays from a cobalt source at 17 th day of pregnancy. The progeniture of irradiated as well as non-irradiated females have undergone behavioral tests at the age of 3 months. Irradiated animals exhibited lower locomotor and exploratory activity in the open field test. (authors)

Erosion and mass transfer of Mo, W and Nb under neutron irradiation of high temperature materials

International Nuclear Information System (INIS)

Berzhatyj, V.I.; Luk'yanov, A.N.; Zavalishin, A.A.; Tkach, V.N.; Fedorenko, A.I.

1980-01-01

Studies have been made of the medium composition in thermionic fuel elements of two types during reactor tests; erosion and mass transfer of electrode materials have been investigated in the after-reactor analysis of the tested fuel elements. The studies of electrode material evaporation at the conditions approaching (in environment temperature and composition) those of reactor tests of thermionic fuel elements have shown that the process proceeds in the form of metal oxides. Evaporation rates are determined, the mechanism of evaporation is discussed, and the analytical dependences are obtained for calculating the evaporation rates of Mo and W at certain temperature and gaseous medium composition. It is found that the main contribution to the material transfer off the Mo and Nb surfaces under a high-temperature reactor irradiation comes through the thermal evaporation; in the case of tungsten at the same experimental conditions the rates of mass transfer due to thermal evaporation and neutron sputtering are nearly the same [ru

Influence of hydride on the mechanical behavior under irradiation of zirconium and Zircaloy-4

International Nuclear Information System (INIS)

Vazquez, Carolina A.

2006-01-01

In the present work the radiation hardening law for zirconium with different thermal treatments and with two hydrogen contents, and Zircaloy-4 with low hydrogen content was studied. The irradiations were made at room temperature to a neutron fluences lower than those where the dislocation channeling and deformation inhomogeneities are expected. For zirconium with hydrogen contents lower than 20 ppm and with 100 ppm, irradiated to a neutron dose of 0.0008 dpa (3.6 x 10 17 n/cm 2 ), a behavior according to the Seeger's model was observed. The model predicts the lineal dependence of the yield stress with the square root of the neutron fluence. At this dose range and at room temperature, the hardening due the hydrogen content and the irradiation hardening are additive. Isochronal annealings to determine the hardening recovery temperature were carried out. Values for this temperature above 425 C degrees were found, which are circa 100 C degrees higher than those corresponding to Stage V of radiation damage in Zr and Zry without hydrogen. This suggests that there is a synergistic effect of radiation damage and hydrogen uptake on the mechanical properties; it is possible to think in a probable and severe ductility reduction, even at low neutron fluences. (author) [es

Behavior of beryllium pebbles under irradiation

Energy Technology Data Exchange (ETDEWEB)

Dalle-Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik; Baldwin, D.L.; Gelles, D.S.; Greenwood, L.R.; Kawamura, H.; Oliver, B.M.

1998-01-01

Beryllium pebbles are being considered in fusion reactor blanket designs as neutron multiplier. An example is the European `Helium Cooled Pebble Bed Blanket.` Several forms of beryllium pebbles are commercially available but little is known about these forms in response to fast neutron irradiation. Commercially available beryllium pebbles have been irradiated to approximately 1.3 x 10{sup 22} n/cm{sup 2} (E>1 MeV) at 390degC. Pebbles 1-mm in diameter manufactured by Brush Wellman, USA and by Nippon Gaishi Company, Japan, and 3-mm pebbles manufactured by Brush Wellman were included. All were irradiated in the below-core area of the Experimental Breeder Reactor-II in Idaho Falls, USA, in molybdenum alloy capsules containing helium. Post-irradiation results are presented on density change measurements, tritium release by assay, stepped-temperature anneal, and thermal ramp desorption tests, and helium release by assay and stepped-temperature anneal measurements, for Be pebbles from two manufacturing methods, and with two specimen diameters. The experimental results on density change and tritium and helium release are compared with the predictions of the code ANFIBE. (author)

DLTS spectra of silicon diodes with p+-n-junction irradiated with high energy krypton ions

Directory of Open Access Journals (Sweden)

Nikolai A. Poklonski

2016-06-01

Full Text Available p+-n-Diodes have been studied. The diodes were manufactured on wafers (thickness 460 μm, (111 plane of uniformly phosphorus doped float-zone-grown single-crystal silicon. The resistivity of silicon was 90 Ω cm and the phosphorus concentration was 5×1013 cm−3. The diodes were irradiated with 250 MeV krypton ions. The irradiation fluence was 108 cm−2. Deep-level transient spectroscopy (DLTS was used to examine the defects induced by high energy krypton ion implantation. The DLTS spectra were recorded at a frequency of 1 MHz in the 78–290 K temperature range. The capacity-voltage characteristics have been measured at a reverse bias voltage from 0 to −19 V at a frequency of 1 MHz. We show that the main irradiation-induced defects are A-centers and divacancies. The behavior of DLTS spectra in the 150–260 K temperature range depends essentially on the emission voltage Ue. The variation of Ue allows us to separate the contributions of different defects into the DLTS spectrum in the 150–260 K temperature range. We show that, in addition to A-centers and divacancies, irradiation produces multivacancy complexes with the energy level Et = Ec−(0.5±0.02 eV and an electron capture cross section of ~4×10–13 cm2.

Radiation behavior of superconductors

International Nuclear Information System (INIS)

Scanlan, R.M.; Raymond, E.L.

1979-01-01

High energy neutron irradiations have been performed on Nb 3 Sn superconductors to assess their behavior in a fusion reactor environment. Irradiations were performed at 4.2 K and property measurements were made without warming the samples. The critical current I/sub c/ increased with irradiation to a level about 50% above the unirradiated value at the highest fluences reached in our experiments. These results are compared with the results of other low temperature irradiations of Nb 3 Sn

Charge collection efficiency recovery in heavily irradiated silicon detectors operated at cryogenic temperatures

CERN Document Server

Da Vià, C; Berglund, P; Borchi, E; Borer, K; Bruzzi, Mara; Buontempo, S; Casagrande, L; Chapuy, S; Cindro, V; Dimcovski, Zlatomir; D'Ambrosio, N; de Boer, Wim; Dezillie, B; Esposito, A P; Granat, V; Grigoriev, E; Heijne, Erik H M; Heising, S; Janos, S; Koivuniemi, J H; Konotov, I; Li, Z; Lourenço, C; Mikuz, M; Niinikoski, T O; Pagano, S; Palmieri, V G; Paul, S; Pirollo, S; Pretzl, Klaus P; Ropotar, I; Ruggiero, G; Salmi, J; Seppä, H; Suni, I; Smith, K; Sonderegger, P; Valtonen, M J; Zavrtanik, M

1998-01-01

The charge collection efficiency (CCE) of high resistivity silicon detectors, previously neutron irradiated up to 2*10/sup 15/ n/cm/sup 2/, was measured at different cryogenic temperatures and different bias voltages. In order to $9 study reverse annealing (RA) effects, a few samples were heated to 80 degrees C and kept at room temperature for several months after irradiation. For comparison other samples (NRA) where kept at -10 C after irradiation. The RA and $9 NRA samples, measured at 250 V forward and reverse bias voltage, present a common temperature threshold at 150 K. Below 120 K the CCE is constant and ranges between 55and 65 0.000000or the RA and NRA sample respectively. Similar CCE $9 was measured for a device processed with low resistivity contacts (OHMIC), opening the prospect for a consistent reduction of the cost of large area particle tracking. (7 refs).

Irradiation effects on fuels for space reactors

International Nuclear Information System (INIS)

Ranken, W.A.; Cronenberg, A.W.

1984-01-01

A review of irradiation-induced swelling and gas release experience is presented here for the three principal fuels UO 2 , UC, and UN. The primary advantage of UC and UN over UO 2 is higher thermal conductivity and attendant lower fuel temperature at equivalent pellet diameter and power density, while UO 2 offers the distinct benefit of well-known irradiation performance. Irradiation test results indicate that at equivalent burnup, temperature, and porosity conditions, UC experiences higher swelling than UO 2 or UN. Fission gas swelling becomes important at fuel temperatures above 1320 K for UC, and at somewhat higher temperatures for UO 2 and UN. Evidence exists that at equivalent fuel temperatures and burnups, high density UO 2 and UN experience comparable swelling behavior; however, differences in thermal conductivity influence overall irradiation performance. The low conductivity of UO 2 results in higher thermal gradients which contribute to fuel microcracking and gas release. As a result UO 2 exhibits higher fractional gas release than UN, at least or burnups up to about 3%

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

International Nuclear Information System (INIS)

Zheng Yongnan; Xu Yongjun; Yuan Daqing

2014-01-01

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

Correlation between Mechanical Behavior and Actuator-type Performance of Ni-Ti-Pd High-temperature Shape Memory Alloys

Science.gov (United States)

Bigelow, Glen S.; Padula, Santo A., II; Garg, Anita; Noebe, Ronald D.

2007-01-01

High-temperature shape memory alloys in the NiTiPd system are being investigated as lower cost alternatives to NiTiPt alloys for use in compact solid-state actuators for the aerospace, automotive, and power generation industries. A range of ternary NiTiPd alloys containing 15 to 46 at.% Pd has been processed and actuator mimicking tests (thermal cycling under load) were used to measure transformation temperatures, work behavior, and dimensional stability. With increasing Pd content, the work output of the material decreased, while the amount of permanent strain resulting from each load-biased thermal cycle increased. Monotonic isothermal tension testing of the high-temperature austenite and low temperature martensite phases was used to partially explain these behaviors, where a mismatch in yield strength between the austenite and martensite phases was observed at high Pd levels. Moreover, to further understand the source of the permanent strain at lower Pd levels, strain recovery tests were conducted to determine the onset of plastic deformation in the martensite phase. Consequently, the work behavior and dimensional stability during thermal cycling under load of the various NiTiPd alloys is discussed in relation to the deformation behavior of the materials as revealed by the strain recovery and monotonic tension tests.

Beneficial effects of fruit extracts on neuronal function and behavior following 56Fe irradiation

Science.gov (United States)

Joseph, J. A.; Shukitt-Hale, B.; Carey, A. N.; Jenkins, D.; Rabin, B. M.

Exposing young rats to particles of high energy and charge HZE particles enhances indices of oxidative stress and inflammation and disrupts the functioning of the dopaminergic system and behaviors mediated by this system in a manner similar to that seen in aged animals Previous research has shown that diets supplemented with 2 blueberry or strawberry extracts have the ability to retard and even reverse age-related deficits in behavior and signal transduction in rats perhaps due to their antioxidant and anti-inflammatory properties This study evaluated the efficacy of these diets on irradiation-induced deficits in these parameters by maintaining rats on these diets or a control diet for 8 weeks prior to being exposed to whole-body irradiation with 1 5 Gy of 1 GeV n high-energy 56 Fe particles Irradiation impaired performance in the Morris water maze and measures of dopamine release one month following radiation these deficits were protected by the antioxidant diets The strawberry diet offered better protection against spatial deficits in the maze because strawberry-fed animals were better able to retain place information a hippocampally-mediated behavior compared to controls The blueberry diet on the other hand seemed to improve reversal learning a behavior more dependent on intact striatal function These data suggest that 56 Fe particle irradiation causes deficits in behavior and signaling in rats which were ameliorated by an antioxidant diet and that the polyphenols in these fruits might be acting in different brain regions

Beam-induced temperature changes in HVEM irradiations

International Nuclear Information System (INIS)

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

1975-01-01

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

Effect of high intensity vs. soft-start halogen irradiation on light-cured resin-based composites. Part I. Temperature rise and polymerization shrinkage.

Science.gov (United States)

Hofmann, Norbert; Markert, Tanja; Hugo, Burkard; Klaiber, Bernd

2003-12-01

To determine polymerization shrinkage kinetics and temperature rise of light-cured resin-based composites after high intensity vs. soft-start quartz tungsten halogen irradiation. Shrinkage kinetics was evaluated using the "deflecting disk technique", modified for simultaneous measurement of temperature within the resin-based composite using a thermocouple. Additional irradiations after 60 and 65 minutes allowed the determination of temperature rises caused by radiation or by reaction heat. Four hybrids (Filtek Z250, Herculite, Solitaire 2, Tetric Ceram), an inhomogeneously filled hybrid (InTen-S) and a microfill (Filtek A110, formerly Silux Plus) were cured using the quartz tungsten halogen units Astralis 10 and Optilux 501 in the high intensity (A10 HiPo: 10 seconds at 1300 mW/cm2; OL Boost: 10 seconds at 1140 mW/cm2) or soft-start modes (A10 Pulse: increase to 700 mW/cm2 within 10 seconds, three periods of 2 seconds at 1300 mW/cm2 alternating with two periods of 2 seconds at 700 mW/cm2; OL Ramp: exponential increase within 10 seconds, followed by 10 seconds at 1140 mW/cm2). The soft-start protocols produced less contraction, and polymerization shrinkage started later and progressed slower (or: more slowly), compared to high intensity irradiation [correction]. The lowest shrinkage was observed for InTen-S, followed by Filtek Z250 and A110, whereas Solitaire 2, Herculite and Tetric Ceram scored highest for this parameter. Temperature rise was caused more or less equally by radiation and by reaction heat and reached values of up to 28.9 degrees C relative to a baseline of 37 degrees C. For some combinations of curing modes and resin-based composites, less heat was generated by the soft-start protocols and by Optilux 501.