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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
高燃焼度PWR燃料を用いて、高温高圧水中で行われた2ケースのNSRR実験をRANNSコードで解析した。一方のケースでは被覆管は破損し、他方のケースでは非破損であったがDNBが発生した。解析結果は破損した被覆管の金相観察と比較され、破損を決める条件が、初期クラック深さや温度,応力の観点から考察された。これらの評価に基づき、初期温度の効果に関する感度解析、及び商用炉での仮想的なパルス照射条件でのシミュレーションを行った。これらの結果を相互に比較し、被覆管の破損の可能性について考察した。
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 
C as well as data from the room temperature tests at around 20 C, and discusses the applicability of these data to the fuel safety evaluation under power reactor conditions.
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
温度遷移における燃料被覆管のミクロ組織及び機械特性の変化を調べ、さまざまなRIA条件下での広範な温度範囲における燃料挙動を評価するために実験を行った。被覆管の延性は温度上昇とともに基本的に増大する。水素化した被覆管の延性は300から473Kの間で明確に増大するが、473から573Kでの増大は小さい。RIAタイプの速い温度遷移では再結晶による延性増大は期待できない。一方、1100K以上で起こる相変態は急激に起こり機械特性に影響を及ぼす。870K以上では水素化物リム中の水素化物が固溶するため、水素化物リムが及ぼす影響はないと考えられる。
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.
Georgenthum, V.*; 杉山 智之
Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 9 Pages, 2009/09
In the frame of their research programs on high burnup fuel safety, the IRSN and JAEA performed a large set of tests devoted to the study of PWR fuel rod behaviour during Reactivity Initiated Accident (RIA) respectively in the CABRI and NSRR. The test conditions are different in terms of coolant nature, temperature and pressure. In the CABRI reactor, tests were performed so far with sodium coolant at 280 C and 3 bar. In the NSRR most of the tests were performed with stagnant water at 20 C and atmospheric pressure, but recently a new high temperature high pressure capsule has been developed which allows up to 280 C and 70 bar. This paper discusses the influence of test conditions on rod behaviour during boiling phase, based on test results and SCANAIR code calculations. The study shows that when the boiling crisis is reached, the initial inner and outer rod pressure have an essential impact on the clad strain and possible ballooning.