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Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

竹田 武司; 大津 巌

Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08

An experiment was conducted for OECD/NEA ROSA-2 Project using LSTF, which simulated 17% hot leg intermediate-break LOCA in PWR. Core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on upper core plate. Results of uncertainty analysis with RELAP5/MOD3.3 code clarified influences of combination of multiple uncertain parameters on peak cladding temperature within defined uncertain ranges. An experiment was performed for OECD/NEA PKL-3 Project with PKL. The LSTF test simulated PWR 1% hot leg small-break LOCA with steam generator secondary-side depressurization as accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for primary pressure, core collapsed liquid level, and cladding surface temperature probably due to effects of differences between LSTF and PKL in configuration, geometry, and volumetric size.


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.


The Effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01

 被引用回数:3 パーセンタイル:41.33(Nuclear Science & Technology)

In order to investigate the effect of oxidation and crystal phase condition on the ballooning and rupture behaviors of cladding tube under simulated loss-of-coolant-accident (LOCA) conditions, laboratory-scale experiments were performed in which internally pressurized non-irradiated Zircaloy-4 (Zry-4) cladding specimens were heated to burst in steam and argon gas conditions. Values of the maximum circumferential strain were normalized by dividing them by engineering hoop stress at the time of rupture. The dependence of the normalized value on burst temperature and the relationship between the normalized value and the length, width and area of rupture opening were evaluated. The correlation between the normalized value and the burst temperature suggested that the fraction of the $$beta$$ phase in Zry-4 cladding specimens affected the strain in the specimens and the oxidation of specimens suppressed the amount of ballooning of the specimens. The relationship between the normalized value and the length, width and area of rupture opening indicated that the length, width and area of rupture opening depended on the crystal phase condition in Zry-4 cladding specimens irrespective of atmosphere in the case of the heating rate of $$sim$$3 K/s.


BWR loss-of-coolant accident tests at ROSA-III with high temperature emergency core coolant injection

中村 秀夫; 久木田 豊; 田坂 完二

Journal of Nuclear Science and Technology, 25(2), p.169 - 179, 1988/02



Assessment of core thermo-hydrodynamic models of REFLA-1D code with CCTF data for reflood phase of PWR-LOCA

大久保 努; 村尾 良夫

Journal of Nuclear Science and Technology, 22(12), p.983 - 994, 1985/12

 被引用回数:17 パーセンタイル:12.34(Nuclear Science & Technology)




桜井 文雄; 小山田 六郎

JAERI-M 85-001, 33 Pages, 1985/02


JMTR炉心の高濃縮度燃料(濃縮度:93%)から中濃縮度燃料(濃縮度:45%)への転換に係る安全審査において、LOCAの再評価が求められた。そして、1次冷却系配管の両端破断によるLOCAの熱水力的検討の結果、以下の事項が確認された。(1)燃料は炉心の冠水が維持されれば焼損しない。(2)1次冷却系配管の両端破断によるLOCAにおいて炉心が空気中に露出するのを防ぐためには、1次冷却系配管の小破断を想定して設置されている現行のサイフォンブレーク弁($$phi$$25mm)を大口径のもの($$phi$$60mm以上)に交換する必要がある。 サイフォンブレーク弁のサイズを検討するために、計算コードSBAC(Siphon Breaker Analysis Code)を作成した。本コードの精度は5%以内であることが検証実験により確認できた。


Evaluation Report on CCTF Core-I Reflood Tests Cl-16(Run 25),Cl-21(Run 40)and Cl-22(Run 41); Comparison of Rresults Between FLECHT Coupling Tests and FLECHT-SET Test

村尾 良夫; 須藤 高史; 井口 正

JAERI-M 83-065, 113 Pages, 1983/05


円筒炉心試験装置(CCTF)内の現象が他の試験装置内の現象と類似のものであることを確証するため、FLECHT-SET実験3105B,2714B,3420Bを模擬した試験条件で3回のCCTF試験を行った。ダウンカマおよび、上部プレナムの蓄水、健全ループの圧力損失は、CCTFとFLECHT-SETとで同じであったが、破断ループの圧力損失、及びそれによって生じたシステム内の水力的振動、炉心内の熱水力挙動は両者で異なっていた。これらの違いは、主としてC C T Fにおける破断コールドレグでの圧力損失によってもたらされたことがわかった。FLECHT-SET実験においては、構造の違い、装置の運転法の違いによりこの圧力損失は現われなかった。従って、両試験装置の構造、運転法の違いを考慮に入れれば、両試験装置で観測された現象は互に類似のものであると結論できる。



村尾 良夫

JAERI-M 83-032, 176 Pages, 1983/03




Evaluation Report on CCTF Core-I Reflood Tests C1-17(Run 36)and C1-20(Run 39); Thermally-Multidimensional Effects on Core Thermo-Hydrodynamics

村尾 良夫; 井口 正

JAERI-M 83-028, 106 Pages, 1983/03




Analysis of direct contact condensation of flowing steam onto injected water with a multifluid model of two-phase flow

秋本 肇; 小澤 由行*; 井上 晃*; 青木 成文*

Journal of Nuclear Science and Technology, 20(12), p.1006 - 1022, 1983/00

 被引用回数:7 パーセンタイル:33.61(Nuclear Science & Technology)



CCTF Core I Test Results

村尾 良夫; 須藤 高史; 秋本 肇; 井口 正; 杉本 純; 藤木 和男; 平野 見明

JAERI-M 82-073, 31 Pages, 1982/07




大型再冠水円筒炉心試験・第1次炉心・54年度試験; 試験結果の解析

村尾 良夫; 秋本 肇; 大久保 努; 須藤 高史; 平野 見明

JAERI-M 82-002, 144 Pages, 1982/03




Experimental modeling of core hydrodynamics during reflood phase of LOCA

村尾 良夫; 井口 正

Journal of Nuclear Science and Technology, 19(8), p.613 - 627, 1982/00

 被引用回数:19 パーセンタイル:13.63(Nuclear Science & Technology)



Experimental study of system behaivior during reflood phase of PWR-LOCA using CCTF

村尾 良夫; 秋本 肇; 須藤 高史; 大久保 努

Journal of Nuclear Science and Technology, 19(9), p.705 - 719, 1982/00

 被引用回数:30 パーセンタイル:7.28(Nuclear Science & Technology)



Development of computer code PRECIP-II for calculation of Zr-steam reaction; Comparison of calculation with experiments in temperature transient

鈴木 元衛; 川崎 了

Journal of Nuclear Science and Technology, 17(4), p.291 - 300, 1980/00

 被引用回数:6 パーセンタイル:40.26(Nuclear Science & Technology)



An Analysis of transients in experiments on loss-of-coolant accidents

傍島 真

Nuclear Science and Engineering, 60(1), p.10 - 18, 1976/01




高速二相流に関する研究,4; 大口径薄刃オリフィスの二相流出係数

安達 公道

日本原子力学会誌, 16(6), p.322 - 329, 1974/06



Study on the effect of azimuthal temperature difference on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions

成川 隆文

no journal, , 

Laboratory-scale burst tests on non-irradiated Zircaloy-4 cladding tube specimens were performed under transient-heating conditions using an external heating method. Then the obtained data were compared to those from a previous study, where an internal heating method was used. This comparison suggests that the amount of deformation in the ballooned region may be estimated based on engineering hoop stress, azimuthal temperature difference in the cladding, and the maximum circumferential strain divided by engineering hoop stress, $$k$$, which was obtained in this study, irrespective of heating method. The size of rupture opening may be affected by both the azimuthal and axial temperature differences in the cladding.

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