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Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident

更田 豊志*; 永瀬 文久

Zirconium in the Nuclear Industry; 18th International Symposium (ASTM STP 1597), p.52 - 92, 2018/01



Transient response of LWR fuels (RIA)

更田 豊志

Comprehensive Nuclear Materials, 2, p.579 - 593, 2012/02

Comprehensive Nuclear Materials will provide broad ranging, validated summaries of all of the major topics in the field of nuclear material research for fission as well as fusion reactor systems. The four volume work will consist of an extensive collection of comprehensive review articles written by a team of leading experts. Attention will be given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The chapter aims to provide a general outline of fuel behavior during a reactivity-initiated accident (RIA) postulated in light water reactors (LWRs) and to show experimental data providing technical basis with the current RIA-related regulatory criteria in Japan.


Ring compression ductility of high-burnup fuel cladding after exposure to simulated LOCA conditions

永瀬 文久; 中頭 利則; 更田 豊志

Journal of Nuclear Science and Technology, 48(11), p.1369 - 1376, 2011/11

 被引用回数:10 パーセンタイル:30.73(Nuclear Science & Technology)



Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

天谷 政樹; 中村 仁一; 永瀬 文久; 更田 豊志

Journal of Nuclear Materials, 414(2), p.303 - 308, 2011/07

 被引用回数:9 パーセンタイル:33.97(Materials Science, Multidisciplinary)



Behavior of coated fuel particle of High-Temperature Gas-cooled Reactor under reactivity-initiated accident conditions

梅田 幹; 杉山 智之; 永瀬 文久; 更田 豊志; 植田 祥平; 沢 和弘

Journal of Nuclear Science and Technology, 47(11), p.991 - 997, 2010/11

 被引用回数:6 パーセンタイル:53.15(Nuclear Science & Technology)

In order to clarify the failure mechanism and determine the failure limit of the High Temperature Gas-cooled Reactor (HTGR) fuel under reactivity-initiated accident (RIA) conditions, pulse irradiations were performed with unirradiated coated fuel particles at the Nuclear Safety Research Reactor (NSRR). The energy deposition ranged from 580 to 1,870 J/gUO$$_{2}$$ in the pulse irradiations and the estimated peak temperature at the center of the fuel particle ranged from about 1,510 to 3,950 K. Detailed examinations after the pulse irradiations showed that the coated fuel particles failed above about 1,400 J/gUO$$_{2}$$ where the peak fuel temperature reached over the melting point of UO$$_{2}$$ fuel. It was also shown that the coated fuel particle was failed by the mechanical interaction between the melted and swelled fuel kernel and the coating layer under RIA conditions.


Behavior of high burnup LWR fuels during design-basis accidents; Key observations and an outline of the coming program

更田 豊志; 永瀬 文久; 杉山 智之; 天谷 政樹

Proceedings of 2010 LWR Fuel Performance Meeting/TopFuel/WRFPM (CD-ROM), p.244 - 253, 2010/09



Ab initio study on plane defects in zirconium-hydrogen solid solution and zirconium hydride

宇田川 豊; 山口 正剛; 阿部 弘亨*; 関村 直人*; 更田 豊志

Acta Materialia, 58(11), p.3927 - 3938, 2010/06

 被引用回数:63 パーセンタイル:4.73(Materials Science, Multidisciplinary)

In order to elucidate the origin of the hydrogen-induced embrittlement of zirconium alloys, we here evaluate the surface energy (SE) and unstable stacking energy (USE) of Zr-H systems by making ab initio calculations. For solid solutions we found decrease in SE and USE with increased H/Zr ratio. For the hydride, we found 25% smaller SE and 200 to 300% larger USE than pure zirconium. This indicates that zirconium hydride is extremely brittle, due to the synergistic effect of small SE relative to pure zirconium, indicating easy generation of fractures on the surface, and large relative USE, indicating difficulty in dislocation motion. Furthermore, Rice's parameter D of ductility/brittleness becomes 1.1-1.5 in hydride, indicating that brittle fracture occurs more readily than iridium. These results seem enough to attribute hydrogen embrittlement of zirconium alloys substantially to the fundamentally brittle nature of the hydride itself.


Evaluation of initial temperature effect on transient fuel behavior under simulated reactivity-initiated accident conditions

杉山 智之; 宇田川 豊; 更田 豊志

Journal of Nuclear Science and Technology, 47(5), p.439 - 448, 2010/05

 被引用回数:5 パーセンタイル:59.02(Nuclear Science & Technology)

To evaluate possible effect of initial temperature on the fuel behavior, such as cladding deformation and fission gas release, in a reactivity-initiated accident, two comparative tests were performed on identical high burnup PWR fuel rods at different temperatures in the NSRR. Two tests RH-1 and RH-2 were performed, respectively, at room temperature and $$sim$$280 $$^{circ}$$C which corresponds to the hot zero power of PWR. The test rods did not fail in the both tests against fuel enthalpy increases of 462 and 378 J/g, respectively. The results of the two tests were consistent to those of previous tests at room temperature, if data were plotted as a function of the peak fuel enthalpy, not of the maximum increase of enthalpy. Computer analysis with the RANNS code confirmed that the cladding residual strain in RH-2 was driven only by the pellet thermal expansion and that the gas-induced strain did not occur because the cladding temperature did not reach high enough to enhance creep deformation.


Identification of radical position of fission gas release in high-burnup fuel pellets under RIA conditions

笹島 栄夫; 杉山 智之; 中頭 利則; 永瀬 文久; 中村 武彦; 更田 豊志

Journal of Nuclear Science and Technology, 47(2), p.202 - 210, 2010/02

 被引用回数:3 パーセンタイル:71.83(Nuclear Science & Technology)



Relationship between changes in the crystal lattice strain and thermal conductivity of high burnup UO$$_2$$ pellets

天谷 政樹; 中村 仁一; 更田 豊志; 高阪 裕二*

Journal of Nuclear Materials, 396(1), p.32 - 42, 2010/01

 被引用回数:3 パーセンタイル:71.83(Materials Science, Multidisciplinary)



Fuel safety limits; Experimental results and pending questions

Vitanza, C.*; 更田 豊志

EUROSAFE Tribune (Internet), 16, p.13 - 17, 2009/11

Considerable experimental effort has been made in the last decade to produce experimental data in support of the definition of fuel safety limits for a variety of fuel designs and considering the effect of burn-up. In particular, tests have been performed in specialized laboratories to address the fuel safety limits at conditions representative of design basis accidents, i.e. Reactivity Initiated Accident (RIA) and Loss-of-Coolant Accidents. In addition to assessing the effect of burn-up, the main focus of these tests has been on the safety performance of different cladding types, especially for PWR fuels. The main outcome of these efforts is discussed in this article.


Stress intensity factor at the tip of cladding incipient crack in RIA-simulating experiments for high burnup PWR fuels

宇田川 豊; 鈴木 元衛; 杉山 智之; 更田 豊志

Journal of Nuclear Science and Technology, 46(10), p.1012 - 1021, 2009/10

 被引用回数:5 パーセンタイル:59.19(Nuclear Science & Technology)

RIA-simulating experiments for high burnup PWR fuels have been performed in the NSRR and stress intensity factor $$K$$$$_{rm I}$$ I at the tip of cladding incipient crack has been evaluated in order to investigate its validity as a PCMI failure threshold in RIA conditions. An incipient crack depth was determined by observation of the metallographs. Hoop stress in cladding periphery during the pulse power transient was calculated by the RANNS code. Failure in elastic deformation range has never occurred with $$K$$$$_{rm I}$$ of less than 17 MPa m$$^{1/2}$$.


Thermal conductivity change in high burnup MOX fuel pellet

中村 仁一; 天谷 政樹; 永瀬 文久; 更田 豊志

Journal of Nuclear Science and Technology, 46(9), p.944 - 952, 2009/09



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

鈴木 元衛; 杉山 智之; 宇田川 豊; 永瀬 文久; 更田 豊志

Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 11 Pages, 2009/09



Applicability of NSRR room/high temperature test results to fuel safety evaluation under power reactor conditions

杉山 智之; 梅田 幹; 宇田川 豊; 笹島 栄夫; 鈴木 元衛; 更田 豊志

Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 12 Pages, 2009/09

Pulse irradiation tests of high burnup light water reactor fuels were performed at the Nuclear Safety Research Reactor (NSRR) in order to investigate transient fuel behavior and fuel failure limit under the reactivity-initiated accident (RIA) conditions. This paper presents new data from the NSRR high temperature tests at 250 to 290 $$^{circ}$$C as well as data from the room temperature tests at around 20 $$^{circ}$$C, and discusses the applicability of these data to the fuel safety evaluation under power reactor conditions.


Microstructure and mechanical property changes in fuel cladding during RIA-type temperature transients

永瀬 文久; 杉山 智之; 更田 豊志

Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 8 Pages, 2009/09



Current RIA-related regulatory criteria in Japan and their technical basis

更田 豊志; 杉山 智之

Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 12 Pages, 2009/09

This paper will describe current Japanese regulatory criteria concerning fuel behavior during postulated reactivity-initiated accident (RIA) and experimental data providing technical basis with those criteria. The current safety evaluation guideline for the reactivity-initiated events in light water reactors was established by the Nuclear Safety Commission (NSC) of Japan in 1984 based mainly on the results of the NSRR experiments. In the guideline, an absolute limit of fuel enthalpy during an RIA is defined to avoid mechanical forces generation. The guideline also defines an allowable limit of fuel enthalpy for fuel design as a function of difference between rod internal pressure and system pressure. Because only limited number of data had been available, a series of experiments with pre-irradiated fuel rods were initiated in 1989 in the NSRR. After a series of experiments, the NSC issued a regulatory report regarding behavior of burnup fuels during a postulated RIA in 1998.


Comparative analysis on behavior of high burnup PWR fuels pulse-irradiated in reactivity-initiated accident conditions

鈴木 元衛; 杉山 智之; 宇田川 豊; 永瀬 文久; 更田 豊志

Proceedings of Top Fuel 2009 (DVD-ROM), p.473 - 479, 2009/09



Effect of initial coolant temperature on mechanical fuel failure under reactivity-initiated accident conditions

杉山 智之; 梅田 幹; 笹島 栄夫; 鈴木 元衛; 更田 豊志

Proceedings of Top Fuel 2009 (DVD-ROM), p.489 - 496, 2009/09

Pulse irradiation tests, simulating reactivity-initiated accidents (RIAs), were performed on high burnup fuels at high temperature (HT) in the Nuclear Safety Research Reactor (NSRR). The NSRR tests have provided data of fuel failure limit against the pellet-cladding mechanical interaction (PCMI) at RIAs, but the coolant temperature in the previous tests was limited to room temperature (RT) of around 20 $$^{circ}$$C. Therefore, the obtained failure limits could be very conservative for RIAs at hot zero power or at operation. The possible effect of initial coolant temperature on the PCMI failure limit was investigated using a newly developed test capsule which can achieve 290 $$^{circ}$$C. PWR and BWR fuel rods were tested both at RT and HT conditions. Comparison of the test results indicated that the increased cladding ductility at HT raised the failure limit. Hence, the PCMI failure criterion based on the NSRR RT data has more than adequate safety margin for RIAs at HT condition.


Cladding embrittlement under LOCA conditions, examined by two test methodologies

永瀬 文久; 中頭 利則; 更田 豊志

Proceedings of Top Fuel 2009 (DVD-ROM), p.527 - 537, 2009/09


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