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JAEA Reports

Evaluation of the minimum critical amount for heterogeneous lattice systems composed of fuel rods utilized in low-power water-moderated research and test reactors by using continuous-energy Monte Carlo code MVP with JENDL-4.0

Yanagisawa, Hiroshi

JAEA-Technology 2021-023, 190 Pages, 2021/11


Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in low-power research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.

Journal Articles

Kinetic and Fourier transform infrared studies on the thermal decomposition of sodium hydride

Kawaguchi, Munemichi

Journal of Physical Chemistry C, 125(22), p.11813 - 11819, 2021/06

 Times Cited Count:0 Percentile:0(Chemistry, Physical)

Isothermal and constant heating thermogravimetry-differential thermal analysis (TG-DTA) and Fourier transform infrared spectrometer (FTIR) measurements have been performed for pre- and post-fired sodium hydride (NaH) in the temperature range of 500-700 K, respectively. Temperature dependence of NaH thermal decomposition rates obtained by the isothermal TGs showed an inflection point at around 620 K, which was caused by two kinds of hydrogen states (rapid diffusing and immobile hydrogen). In the FTIR spectra for the NaH and sodium (Na), the specific signals were observed at around 873.4, 1010.4, 1049.5 and 1125.7 cm$$^{-1}$$, and the integrated values of FTIR signals for post-fired NaH at below 550K and at above 698 K were comparable to those for pre-fired NaH and Na, respectively. Those for post-fired NaH at 602-667 K were the intermediate values of the pre-fired NaH and Na, which denoted that the Na-Na bonds haven't grown sufficiently and the hydrogen coexisted in metallic Na. In order to predict the practical kinetics of NaH thermal decomposition reaction, we suggested the simple kinetics model which assumed two kinds of rapidly diffusing and immobile hydrogen states. The simulation results revealed the inflection point in temperature dependence of the thermal decomposition rates accordingly because the transition from immobile hydrogen to rapid diffusing hydrogen crosses over at around 620 K.

Journal Articles

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06

 Times Cited Count:1 Percentile:23.17(Nuclear Science & Technology)

Journal Articles

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

 Times Cited Count:1 Percentile:23.17(Nuclear Science & Technology)

Journal Articles

Hydrogenation of $$L$$1$$_{2}$$-type AlNi$$_{3}$$ alloy at high pressure and temperature

Endo, Naruki*; Saito, Hiroyuki; Machida, Akihiko; Katayama, Yoshinori

Journal of Alloys and Compounds, 645(Suppl.1), p.S61 - S63, 2015/10

 Times Cited Count:0 Percentile:0.01(Chemistry, Physical)

Journal Articles

Hydrogenation of a TiFe-based alloy at high pressures and temperatures

Endo, Naruki*; Saita, Itoko*; Nakamura, Yumiko*; Saito, Hiroyuki; Machida, Akihiko

International Journal of Hydrogen Energy, 40(8), p.3283 - 3287, 2015/03

 Times Cited Count:11 Percentile:32.22(Chemistry, Physical)

Journal Articles

Phase diagram of the Eu-H system at high temperatures and high hydrogen pressures

Saito, Hiroyuki; Machida, Akihiko; Matsuoka, Takehiro*; Aoki, Katsutoshi*

Solid State Communications, 205, p.24 - 27, 2015/03

 Times Cited Count:8 Percentile:39.31(Physics, Condensed Matter)

JAEA Reports

The Evaluation of the influence of hydride rim and biaxial stress condition on the cladding failure under a reactivity-initiated-accident by using EDC test method

Shinozaki, Takashi; Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

JAEA-Research 2014-025, 34 Pages, 2014/12


EDC test is a test method on the mechanical property of fuel cladding tube, and it focuses on the stress condition generated by PCMI under a RIA. We conducted EDC tests which simulate the mechanical conditions during a RIA by using the unirradiated cladding tubes which simulate hydride rim. Circumferential residual strains observed in post-test specimens tended to decrease with increasing the hydrogen concentration in the test cladding tubes and the thickness of the hydride rim. We also prepared RAG tube and performed EDC tests on it. It was observed that circumferential total strains at failure tended to decrease with increasing pre-crack depth on the outer surface of RAG tube specimen. We conducted biaxial stress tests by applying longitudinal tensile load onto RAG tube specimens. It was observed that circumferential total strains at failure under biaxial stress conditions tended to decrease compared to the results under uniaxial tensile condition.

Journal Articles

Li$$_{4}$$FeH$$_{6}$$; Iron-containing complex hydride with high gravimetric hydrogen density

Saito, Hiroyuki; Takagi, Shigeyuki*; Matsuo, Motoaki*; Iijima, Yuki*; Endo, Naruki*; Aoki, Katsutoshi*; Orimo, Shinichi*

APL Materials (Internet), 2(7), p.076103_1 - 076103_7, 2014/07

 Times Cited Count:20 Percentile:67.66(Nanoscience & Nanotechnology)

Journal Articles

Neutronics assessment of advanced shield materials using metal hydride and borohydride for fusion reactors

Hayashi, Takao; Tobita, Kenji; Nishio, Satoshi; Ikeda, Kazuki*; Nakamori, Yuko*; Orimo, Shinichi*; DEMO Plant Design Team

Fusion Engineering and Design, 81(8-14), p.1285 - 1290, 2006/02

 Times Cited Count:20 Percentile:80.39(Nuclear Science & Technology)

Neutron transport calculations were carried out to evaluate the capability of metal hydrides and borohydrides as an advanced shielding material. Some hydrides indicated considerably higher hydrogen content than polyethylene and solid hydrogen. The hydrogen-rich hydrides show superior neutron shielding capability to the conventional materials. From the temperature dependence of dissociation pressure, ZrH$$_{2}$$ and TiH$$_{2}$$ can be used without releasing hydrogen at the temperature of less than 640 $$^{circ}$$C at 1 atm. ZrH$$_{2}$$ and Mg(BH$$_{4}$$)$$_{2}$$ can reduce the thickness of the shield by 30% and 20% compared to a combination of steel and water, respectively. Mixing some hydrides with F82H produces considerable effects in $$gamma$$-ray shielding. The neutron and $$gamma$$-ray shielding capabilities decrease in order of ZrH$$_{2}$$ $$>$$ Mg(BH$$_{4}$$)$$_{2}$$ and F82H $$>$$ TiH$$_{2}$$ and F82H $$>$$ water and F82H.

JAEA Reports

Driving force of PCMI failure under reactivity initiated accident conditions and influence of hydrogen embrittlement on failure limit

Tomiyasu, Kunihiko; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi

JAERI-Research 2005-022, 128 Pages, 2005/09


In order to clarify the driving force of PCMI failure on high burnup fuels and the influence of hydrogen embrittlement on failure limit under RIA conditions, simulated-RIA experiments were performed on fresh fuel rods in the NSRR. The driving force was restricted only to thermal expansion of pellet by using fresh pellets, and fresh claddings were pre-hydrided to simulate hydrogen absorption of high burnup fuels. In seven experiments, test rods resulted in PCMI failure, which was observed on high burnup fuels, in terms of transient behavior and fracture configuration. It indicates that the driving force is sufficiently explained with thermal expansion of pellet and a contribution of fission gas is small. Many incipient cracks were generated in the outer surface of the cladding, and they stopped at the boundary between hydride rim and metallic layer. It suggests that a toughness of metallic region except hydride rim has particular importantance for failure limit. Fuel enthalpy at failure correlates with the thickness of hydride rim, and tends to decrease with thicker hydride rim.

Journal Articles

Investigation of hydride rim effect on failure of Zircaloy-4 cladding with tube burst test

Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 42(1), p.58 - 65, 2005/01

 Times Cited Count:38 Percentile:91.94(Nuclear Science & Technology)

Tube burst tests have been performed with artificially hydrided Zircaloy-4 specimens at room temperature and at 620 K. Pressurization rate was increased to a maximum of 3.4 GPa/s in order to simulate rapid PCMI that occurs in high burnup fuel rods during a pulse-irradiation in the NSRR. Hydrogen content in the specimens ranged from 150 to 1050 ppm. Hydrides were accumulated in the cladding periphery and formed "hydride rim" as observed in high burnup PWR fuel claddings. The hydrided cladding tubes failed with an axial crack at the room temperature tests. Brittle fracture appeared in the hydride rim, and failure morphology was similar to that observed in the NSRR experiments. The hydrides rim obviously reduced burst pressure and residual hoop strain at the tests. The residual hoop strain was very small even at 620 K when thickness of the hydride rim exceeded 18% of cladding thickness. The present result accordingly indicates an important role of the hydrides layer in high burnup fuel rod failure under RIA conditions.

Journal Articles

Behavior of uranium-zirconium hydride fuel under reactivity initiated accident conditions

Sasajima, Hideo; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi; Uetsuka, Hiroshi

Proceedings of 7th International Topical Meeting on Research Reactor Fuel Management (ENS RRFM2003), p.109 - 113, 2003/03

Uranium-zirconium hydride (U-ZrHx) fuel has been widely utilized in the world as TRIGA reactor fuel. In order to obtain the fuel performance data under accident conditions and to enhance accountability of the safety assessment of the reactors using the fuel, irradiation tests under power burst type accident conditions have been conducted in the NSRR. Five pulse irradiation tests have been performed at peak fuel enthalpies ranging from 187 J/g to 483 J/g. Cladding surface temperature increased rapidly at the pulse and DNB occurred in peak fuel enthalpy over 187 J/g in the tests. The DNB occurred at lower fuel enthalpy in the U-ZrH1.6 fuel than in the UO$$_{2}$$ fuel rods. The rod internal pressure rose up to as high as 1MPa in the transient heating tests, suggesting considerable release of the hydrogen decomposed from the fuel. The peak pressure was lower than equilibrium hydrogen pressure of 1.5MPa at the peak temperature, suggesting the transient effect. Considerable PCMI was observed in the tests, through cladding elongation up to 3.3 mm synchronized to the pellet stack deformation.

JAEA Reports

Mechanical properties changes of high burnup PWR fuel cladding by temperature transient

Nagase, Fumihisa; Uetsuka, Hiroshi

JAERI-Research 2002-023, 23 Pages, 2002/11


To obtain basic data to evaluate fuel rod integrity during abnormal transient and accident of LWRs, high burnup PWR fuel claddings were heated for 0 to 600s at temperatures of 673 through 1173K, and the mechanical property changes were examined by using ring tensile test at room temperature. As a result of the test, it was shown that strength and ductility of the cladding are changed depending on heating temperature and time. The mechanical property changes by temperature transients are considered to be correspondent mainly to recovery of irradiation defect, recovery and recrystallization of the Zircaloy, phase transformations, and associated change of the hydride distribution and morphology. Comparison with unirradiated claddings suggested that irradiation effects are not completely annealed out by the short-term annealing at high temepratures. Radial change of hydrogen concentration was measured for the high burnup PWR fuel cladding and very high hydrogen concentration of about 2400wtppm was detected at the cladding periphery.

Journal Articles

Design, fabrication and irradiation experience of actinide-hydride fuel capsule in the JMTR

Komori, Yoshihiro; Amezawa, Hiroo; Komukai, Bunsaku; Narui, Minoru*; Konashi, Kenji*

KAERI/GP-195/2002, p.3 - 10, 2002/00

The actinide-hydride(UTh$$_{4}$$Zr$$_{10}$$H$$_{20}$$) fuel has been studied for transmutation of long-lived actinide contained in the high level wastes and the first irradiation test was successfully carried out in the Japan Materials Testing Reactor (JMTR) of JAERI. Fuel pellets were fabricated by alloying and hydrogenation within an expected diameter error. The fuel pellets were designed to be irradiated below 873K on the fuel surface in consideration of hydrogen dissociation. Irradiation temperature was well agreed with designed value. Fuel burnup reached 0.2%FIMA for two JMTR operation cycles.

JAEA Reports

High-pressurization-rate burst test of hydrided zircaloy-4 fuel cladding at 620K

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi

JAERI-Research 2000-046, 31 Pages, 2000/12


no abstracts in English

JAEA Reports

High-pressurization-rate burst test of hydrided Zircaloy-4 fuel cladding at room temperature

Nagase, Fumihisa; ; Uetsuka, Hiroshi

JAERI-Research 98-064, 25 Pages, 1998/11


no abstracts in English

Journal Articles

Hydride morphology and hydrogen embrittlement of Zircaloy fuel cladding used in NSRR/HBO experiment

Nagase, Fumihisa; Uetsuka, Hiroshi

Proc. of Int. Topical Meeting on LWR Fuel Performance, 0, p.677 - 684, 1997/03

no abstracts in English

JAEA Reports

Status and task of the study on the hydrogen embrittlement of zirconium alloys

Nagase, Fumihisa; *; Komatsu, Kazushi*; Furuta, Teruo

JAERI-Review 95-012, 122 Pages, 1995/08


no abstracts in English

JAEA Reports

Analysis of energy deposition and evaluation of maximum load of irradiation capsule for NSRR experiment with uranium-zirconium hydride fuel

Fuketa, Toyoshi; Ishijima, Kiyomi; Tanzawa, Sadamitsu; Nakamura, Takehiko; Sasajima, Hideo; Kashima, Yoichi; ;

JAERI-Research 95-005, 53 Pages, 1995/01


no abstracts in English

38 (Records 1-20 displayed on this page)