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論文

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

宇田川 豊; 三原 武; 谷口 良徳; 垣内 一雄; 天谷 政樹

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

 被引用回数:0 パーセンタイル:100(Nuclear Science & Technology)

This paper reports a computer-code analysis on the base-irradiation behavior of the chromia-and-alumina-doped BWR rod irradiated to 64 GWd/t in Oskarshamn-3, Sweden, and subjected to the reactivity-initiated-accident (RIA) test OS-1, which resulted in a fuel failure due to pellet-cladding mechanical interaction (PCMI) at the lowest fuel-enthalpy increase in all the BWR tests ever performed. The inverse calculation which utilized post-irradiation examination data as its constraint conditions revealed that the OS-1 rod had very likely experienced more intense PCMI loading due to higher swelling rate during base irradiation than other BWR rods subjected to previous RIA tests and thus had been prone to experience enhanced radial-hydride formation. The significant difference in the cladding hoop-stress more than 50 MPa discriminates the OS-1 rod from other BWR rods and supports the interpretation that enhanced radial-hydrides formation differentiated the PCMI-failure behavior observed in the test OS-1 from the previous BWR-fuel tests.

論文

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

天谷 政樹; 垣内 一雄; 三原 武

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

New fuel cladding alloys of which composition was changed from conventional ones have been developed by nuclear fuel vendors and utilities. Since the irradiation growth of fuel cladding is one of the most important parameters which determine the dimensional stability of fuel rod and/or fuel assembly during normal operation, the irradiation growth behavior of the improved Zr-based alloys for light-water reactor fuel cladding was investigated. The coupon specimens were prepared from fuel cladding tubes with various kinds of improved Zr-based alloys. The specimens were loaded into test rigs and had been irradiated in the Halden reactor in Norway under several coolant temperature conditions up to a fast-neutron fluence of $$sim$$7.8$$times$$10$$^{21}$$ (n/cm$$^{2}$$, E $$>$$ 1 MeV). Irradiation conditions such as specimen temperatures had been continuously monitored during the irradiation. During and after the irradiation, the amount of irradiation growth of each specimen was evaluated as a part of the interim and final inspections. The effect of the difference in alloy composition on the amount of irradiation growth seemed insignificant if the other conditions e.g. the final heat treatment condition at fabrication and the irradiation temperature were the same.

論文

Behavior of LWR fuels with additives under reactivity-initiated accident conditions

三原 武; 宇田川 豊; 天谷 政樹; 谷口 良徳; 垣内 一雄

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.544 - 550, 2019/09

In order to assess effects of additives for fuel pellet on the fuel behavior during a reactivity-initiated accident (RIA), fuels with additives irradiated in commercial light water reactors (LWRs) in Europe up to high burnup were subjected to pulse-irradiation experiments in Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Agency (JAEA). Two tests were performed: test LS-4 with chromia-doped UO$$_{2}$$ and Zry-2 cladding with liner and test OS-1 with ADOPT$$^{rm TM}$$ (chromia-and-alumina-doped UO$$_{2}$$) pellet and Zry-2 cladding with liner. The test fuel rod of LS-4 did not fail. The test fuel rod of OS-1 was considered to be failed by hydride-assisted pellet-cladding mechanical interaction (PCMI). The fuel failure limit in OS-1 was the lowest among the test results ever obtained at the NSRR in similar burnup range. The morphology of the hydrides precipitated in the fuel cladding of OS-1 was investigated by metallography and compared with previous results obtained in JAEA in connection focusing fuel failure limit. It was suggested that the observed lower limit of fuel failure was related to the amount and length of the hydride precipitated along the radial direction of cladding.

論文

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

谷口 良徳; 宇田川 豊; 三原 武; 天谷 政樹; 垣内 一雄

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

A pulse-irradiation test CN-1 on a high-burnup MOX fuel with M5$$^{TM}$$ cladding was conducted at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Agency (JAEA). Although the transient signals obtained during the pulse-irradiation test did not show any signs of the occurrence of PCMI failure, the failure of the test fuel rod was confirmed from the visual inspection carried out after test CN-1. Analyses using fuel performance codes FEMAXI-8 and RANNS were also performed in order to investigate the fuel behavior during normal operation and pulse-irradiation regarding the test fuel rod of CN-1, and the results were consistent with this observation result. These experimental and calculation results suggested that the failure of test fuel rod of CN-1 was not caused by hydride-assisted PCMI but high-temperature rupture following the increase in rod internal pressure. The occurrence of this failure mode might be related to the ductility remained in the M5$$^{TM}$$ cladding owing to its low content of the hydrogen absorbed during normal operation.

論文

Fracture behavior of recrystallized and stress-relieved Zircaloy-4 cladding under biaxial stress conditions

三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(8), p.724 - 730, 2019/08

 被引用回数:0 パーセンタイル:100(Nuclear Science & Technology)

Pellet-cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to failure of high-burnup fuel rods. Zircaloy cladding tubes are subjected to biaxial stress states induced by PCMI loading. This type of stress state, specific to PCMI, presumably makes the tubes more susceptible to failure. To clarify the influence of the anisotropic mechanical properties of Zircaloy cladding tubes on their fracture behavior under biaxial stress conditions, biaxial tensile tests were performed, and the measured stresses and strains were converted to reduced parameters such as equivalent strain, equivalent stress, and stress triaxiality by using the anisotropic constants of the Hill yield function derived in our previous study. The minimum fracture strains for recrystallized (RX) and stress-relieved (SR) specimens were located where the stress ratio of axial to circumferential is 0.75 in the measured range. The equivalent plastic fracture strains tended to decrease monotonously with increasing stress triaxiality, which is a typical trend observed in ductile fracture, in the range of 0.65-0.78 for both specimens. In the case of SR specimens, however, the analysis with stress triaxiality did not reduce the fracture strains well to a single trend curve, showing that the anisotropic constants used in the present work or Hill yield function itself is not enough to describe the whole anisotropy involved in the fracture process of SR material.

論文

The effect of hydride morphology on the failure strain of stress-relieved Zircaloy-4 cladding with an outer surface pre-crack under biaxial stress states

Li F.; 三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(5), p.432 - 439, 2019/05

 被引用回数:3 パーセンタイル:12.61(Nuclear Science & Technology)

Hydride precipitates are considered to affect cladding integrity adversely during pellet-cladding mechanical interaction (PCMI) in a reactivity-initiated accident (RIA). This study aims to clarify the role of hydride precipitates in cladding failure under the biaxial stress condition. The amount and distribution of hydride precipitates (hydride morphology) were evaluated quantitatively and hydrogen content was measured to assess its effect on the decrease in outer surface hoop strain at failure (failure strain) of the samples. The decrease in failure strain of the hydrided samples was found to be more significant under lower strain ratios in the samples with shallower pre-crack. The failure strain of sample tended to be more sensitive to hydrogen content under the strain ratio with a higher axial component in the case of samples with hydrogen contents higher than ~150 wppm.

論文

Behaviors of high-burnup LWR fuels with improved materials under design-basis accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

Fuels for light water reactors (LWRs) which consist of improved cladding materials and pellets have been developed by utilities and fuel vendors to acquire better fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate adequacy of the present regulatory criteria in Japan and safety margins regarding the fuel with improved materials, Japan Atomic Energy Agency (JAEA) has conducted ALPS-II program sponsored by Nuclear Regulation Authority (NRA), Japan. In this program, the tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) have been performed on the high burnup advanced fuels irradiated in commercial PWR or BWR in Europe. This paper presents recent results obtained in this program with respect to RIA, and main results of LOCA experiments, which have been obtained in the ALPS-II program, are summarized.

論文

Deformation behavior of recrystallized and stress-relieved Zircaloy-4 fuel cladding under biaxial stress conditions

三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 55(2), p.151 - 159, 2018/02

 被引用回数:4 パーセンタイル:23.21(Nuclear Science & Technology)

Pellet-cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to the failure of high-burnup fuel rods. Biaxial stress states generated by PCMI in Zircaloy cladding may make the cladding more susceptible to failure. In this study, we investigated the deformation behavior of Zircaloy cladding under biaxial stress conditions based on the concept of contours of equal plastic work. The major axis angles of the initial work contours of recrystallized (RX) and stress-relieved (SR) specimens were investigated and it was found that the shapes of the initial work contours of these kinds of specimens were almost symmetric across the direction where the ratio of axial stress to circumferential stress is 1. The shapes of subsequent work contours tended to change for the RX specimen while be the same as the initial for the SR specimen, as deformation proceeded. It was suggested that the textures and slip systems in the RX and SR specimens affect their initial work contours while the slip system in the RX specimens and the residual strain in the SR specimens influence the subsequent work contours.

論文

Behavior of high-burnup advanced LWR fuels under design-basis accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

JAEA has conducted a research program called ALPS-II program for advanced fuels of LWRs. In this program, the tests simulating a RIA and a LOCA have been performed on the high burnup advanced fuels irradiated in European commercial reactors. The failure limits of the high-burnup advanced fuels under RIA conditions have been obtained by the pulse irradiation tests at the NSRR in JAEA. The information about pellet fragmentation etc. during the pulse irradiations was also obtained from post-test examinations on the test rods after the pulse irradiation tests. As for the simulated LOCA test, integral thermal shock tests and high-temperature oxidation tests have been performed at the RFEF in JAEA. The fracture limits under LOCA and post-LOCA conditions etc. of the high-burnup advanced fuel cladding have been investigated, and it was found that in terms of these materials the fracture boundaries do not decrease and the oxidation does not significantly accelerate in the burnup level examined.

論文

Biaxial-EDC test attempts with pre-cracked zircaloy-4 cladding tubes

Li F.; 三原 武; 宇田川 豊; 天谷 政樹

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07

The failure behavior of cladding tube was investigated by using the improved EDC test apparatus. Cold-worked, stress-relieved and recrystallized Zircaloy-4 tubes with a pre-crack were used as test specimens: this pre-crack simulated the crack which is considered to form in the hydride rim of high-burnup fuel cladding at the beginning of PCMI failure. In the EDC test, a tensile stress in axial direction was applied and displacement-controlled loading was performed to keep the strain ratio of axial/hoop as a constant. The data of cladding deformation had been achieved in the range of strain ratio of 0, 0.25, 0.5 and 0.75 and pre-crack depth of 41-87 micrometers. Failures in hoop direction were observed in all the tested samples, and a general trend that higher strain ratio and deeper crack depth lead to lower failure limit in hoop direction could be seen. Different crack propagation mode was observed between recrystallized and stress relieved and cold worked samples.

論文

Improved-EDC tests on the Zircaloy-4 cladding tube with an outer surface pre-crack

篠崎 崇*; 宇田川 豊; 三原 武; 杉山 智之; 天谷 政樹

Journal of Nuclear Science and Technology, 53(9), p.1426 - 1434, 2016/09

 被引用回数:8 パーセンタイル:19.39(Nuclear Science & Technology)

In order to investigate the failure behavior of fuel cladding under a reactivity-initiated accident (RIA) condition, biaxial stress tests on unirradiated Zircaloy-4 cladding tube with an outer surface pre-crack were carried out under room temperature conditions by using an improved Expansion-Due-to-Compression (improved-EDC) test method which was developed by Japan Atomic Energy Agency (JAEA). The specimens with an outer surface pre-crack were prepared by using RAG (Rolling After Grooving) method. In each test, a constant longitudinal tensile load of 0, 5.0 or 10.0 kN was applied along the axial direction of specimen, respectively. All specimens failed during the tests, and the morphology at the failure opening of the specimens was similar to that observed in the result of post-irradiation examinations of high burnup fuel which failed during a pulse irradiation experiment. The longitudinal strain ($$varepsilon$$$$_{tz}$$) at failure clearly increased with increasing longitudinal tensile loads and the circumferential strain ($$varepsilon$$$$_{ttheta}$$) at failure significantly decreased in the case of 5.0 and 10.0 kN tests, compared with the case of 0 kN tests. It is considered that the data obtained in this study can be used as a fundamental basis for quantifying the failure criteria of fuel cladding under a biaxial stress state.

論文

Behavior of high-burnup advanced LWR fuels under accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.53 - 62, 2016/09

軽水炉用改良型燃料について、現行の安全基準の妥当性及び安全余裕を評価するため、また今後の規制のためのデータベースを提供するため、原子力機構ではALPS-IIと呼ばれる原子力規制庁からの委託事業を開始した。この事業は、商用PWR及びBWRで照射された高燃焼度改良型燃料を対象として、主として反応度投入事故及び冷却材喪失事故を模擬した試験から構成されている。最近、高燃焼度改良型燃料のRIA時破損限界がNSRRにて調べられ、パルス照射試験後の燃料を対象とした照射後試験が行われている。LCOA模擬試験に関しては、インテグラル熱衝撃試験及び高温酸化試験が燃料試験施設で行われ、高燃焼度改良型燃料被覆管の破断限界、高温酸化速度等が調べられた。本論文では、この事業で取得された最近のRIA及びLOCA模擬試験結果について主に述べる。

論文

Behavior of high burnup advanced fuels for LWR during design-basis accidents

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 杉山 智之

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09

高燃焼度領域での燃料性能を向上させるとともに既設の原子炉の安全性を向上させるため、高耐食性被覆管や核分裂生成ガス放出を抑えたペレットで構成された改良型燃料が事業者や燃料メーカによって開発されてきた。このような改良型燃料の現行の規制基準や安全裕度の妥当性を評価するため、またこれらに係る将来の規制のためのデータベースを提供するため、原子力機構はALPS-IIと呼ばれる新しい研究プロクラムを開始した。このプログラムは、欧州から輸送された高燃焼度改良型燃料を対象とした反応度事故(RIA)模擬試験及び冷却材喪失事故(LOCA)模擬試験から主に構成されている。本論文では、このプログラムの概要及び現在までに得られているRIA及びLOCA模擬試験結果について述べる。

報告書

EDC試験手法による反応度事故時の燃料被覆管破損に及ぼす水素化物偏在及び2軸応力状態の影響の評価

篠崎 崇; 三原 武; 宇田川 豊; 杉山 智之; 天谷 政樹

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

JAEA-Research-2014-025.pdf:6.05MB

EDC(Expansion-Due-to-Compression)試験は、燃料被覆管の機械特性試験の一手法であり、反応度事故(RIA)時におけるペレット-被覆管機械的相互作用(PCMI)に着目した試験手法である。本研究では、高燃焼度燃料被覆管に見られる"水素化物リム"を模擬するために外周部に水素化物を偏析させた未照射被覆管を使用し、高燃焼度燃料のRIA時に被覆管に負荷される機械的条件を模擬したEDC試験を実施した。試料の水素濃度および偏析した水素化物の厚みが増加すると、試験後試料の周方向残留ひずみが低下する傾向が見られた。また、RIA時に被覆管外面の水素化物に発生するき裂を模擬するため、外面に予き裂を有する被覆管(RAG管)を作製し、この試料を対象としたEDC試験を行った結果、試料の予き裂深さが増加するにつれて破損時の周方向全ひずみが低下する傾向が見られた。さらに、RAG管試料に軸方向引張荷重を負荷することで2軸応力状態とし、EDC試験を実施した。このような2軸応力状態では、単軸引張条件である通常のEDC試験と比較して破損時の周方向全ひずみが低下する傾向が見られた。

論文

Simulation of the fracture behavior of zircaloy-4 cladding under reactivity-initiated accident conditions with a damage mechanics model combined with fuel performance codes FEMAXI-7 and RANNS

宇田川 豊; 三原 武; 杉山 智之; 鈴木 元衛; 天谷 政樹

Journal of Nuclear Science and Technology, 51(2), p.208 - 219, 2014/02

AA2013-0436.pdf:3.87MB

 被引用回数:6 パーセンタイル:43.72(Nuclear Science & Technology)

In order to investigate the detailed processes of pellet cladding mechanical interaction (PCMI) failure, a continuum damage mechanics model using FEM calculations was proposed to be applied to analyses of the RIA-simulated NSRR tests with unirradiated and pre-hydrided claddings. The simulation made reasonable prediction regarding with cladding fracture strain in hoop direction and reproduced the typical fracture behaviour under PCMI loading characterized by a ductile shear zone. The effect of a local temperature rise in the cladding inner region on the failure strain was found to be less than 5%. Failure strains predicted under a plane strain loading were smaller by 20-30% than those predicted under equi-biaxial tensions between the hoop and the axial directions.

論文

Tilted-foil technique for producing a spin-polarized radioactive isotope beam

平山 賀一*; 三原 基嗣*; 渡辺 裕*; Jeong, S. C.*; 宮武 宇也*; 百田 佐多夫*; 橋本 尚志*; 今井 伸明*; 松多 健策*; 石山 博恒*; et al.

European Physical Journal A, 48(5), p.54_1 - 54_10, 2012/05

 被引用回数:2 パーセンタイル:79.74(Physics, Nuclear)

The tilted-foil method for producing spin-polarized radioactive isotope beams has been studied for the application to nuclear physics and materials science, using the radioactive nucleus $$^8$$Li produced at the Tokai Radioactive Ion Accelerator Complex (TRIAC). We successfully produced polarization in a $$^8$$Li beam using 15 thin polystyrene foils fabricated especially for this purpose. A systematic study of the nuclear polarization as a function of the number of foils, beam energy, tilt angles and foil material has been performed, confirming the features of the tilted-foil technique experimentally. The contributions made to the nuclear polarization of $$^8$$Li nuclei by the atomic states was investigated.

口頭

Evaluation of the long-term mechanical behavior in the near field considering chemical transitions of barrier materials

佐原 史浩*; 村上 武志*; 伊藤 弘之*; 三原 守弘; 大井 貴夫

no journal, , 

TRU廃棄物処分システムの、より信頼性の高い性能評価を行うためには、バリア材料の変遷挙動を考慮したニアフィールド水理場の長期的変遷評価を可能とするシステムの構築が必要である。この評価システムの構成要素として力学挙動解析システム(MACBECE)を開発した。このMACBECEは、バリア材の化学的変遷状況や岩盤クリープ変形量をインプットとして、バリア材の変形量を算出し、また化学的変質状況と変形量を加味した透水係数を算出するシステムである。MACBECEを用いて、円形処分坑道断面を対象に、仮定した化学的変遷状況をインプットとした長期力学挙動解析を実施し、ニアフィールドにおける長期的な変形、及び水理場の変遷状況を評価した。その結果、飽和後の人工バリアの変形は、坑道壁面の変位に強く依存し、坑道壁面の変位が無視できる場合は緩衝材の膨潤圧による坑道内部の変形は微小であり、緩衝材も安定していることがわかった。逆に坑道壁面の変位量が有意な場合は、緩衝材の厚さが減少する一方で密度は高くなるので、物質移行への影響は相殺される可能性があることが示された。

口頭

Modelling for the long-term mechanical and hydraulic behavior of bentonite-based materials considering chemical transitions

佐原 史浩*; 村上 武志*; 小林 一三*; 三原 守弘; 大井 貴夫

no journal, , 

TRU廃棄物処分における人工バリアの長期的な力学及び水理学的な挙動を評価するための評価システムを開発した。評価システムには、セメント系材料の影響により、緩衝材(ベントナイト)の特性が変化することを考慮したモデルを組み込んだ。緩衝材の変形モデルとして、粘弾塑性モデルが組み込まれており、その変質に伴う膨潤挙動の変化も考慮した人工バリアの長期評価を実施した。緩衝材の変質を考慮しても透水係数が緩衝材の性能目標値10$$^{-11}$$m/sとなることが示された。

口頭

ジルコニウム水素固溶体及び水素化物中面欠陥の第一原理計算

宇田川 豊; 山口 正剛; 阿部 弘亨*; 浅利 圭亮*; 篠原 靖周*; 村上 健太*; 中園 祥央*; 三原 武*; 澤山 陽平*; 関村 直人*; et al.

no journal, , 

ジルコニウム水素脆化の微視的機構を明らかにするため、ジルコニウム水素固溶体及び水素化物の表面エネルギと$$gamma$$表面を第一原理計算により評価した。計算には密度汎関数理論に基づく第一原理計算コードVASP(GGA-PAWポテンシャル)を用い、ジルコニウム単体($$alpha$$-hcp)及び水素固溶体(Hを含む$$alpha$$-hcp格子)については底面及び柱面、水素化物については(111), (110)及び(100)面の計算を行った。水素化物では表面エネルギ($$gamma$$$$_{S}$$)低下に加え、すべり,転位運動に対する抵抗の指標となる$$gamma$$表面鞍点エネルギ($$gamma$$$$_{US}$$)が大幅に増加し、これにより水素化物自身が極めて脆性的にふるまうことがわかった。一方水素固溶体における水素は$$gamma$$$$_{S}$$$$gamma$$$$_{US}$$をそれぞれ減少させる傾向が見られる。水素固溶体の傾向との比較から、水素化物に特有な脆化の要因は$$gamma$$$$_{US}$$増加、すなわちすべりや転位運動に対する障壁エネルギの増大にあると考えられる。

口頭

Failure behavior of LWR fuel cladding under accident conditions; Key observations from fuel safety research program at JAEA

永瀬 文久; 杉山 智之; 天谷 政樹; 宇田川 豊; 福田 拓司; 三原 武

no journal, , 

原子力機構は、軽水炉におけるRIA及びLOCA時の高燃焼度燃料挙動に関する研究計画を進めている。RIA時にはペレットと機械的に相互作用した被覆管が破損し、破損限界は燃焼度とともに低下する。NSRRを用いて破損限界に関する知見を蓄積するとともに、機械特性試験や解析コードを用いて破損メカニズムに関する知見を得ている。また、LOCA時挙動については、酸化試験やLOCA条件を模擬した試験により酸化速度や酸化した被覆管の破断限界に関する知見の取得や高燃焼度化の影響に関する評価を行っている。事故時燃料破損メカニズム解明及び解析コードを用いた安全評価手法の高度化に今後も取り組んでいく。

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