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論文

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.

論文

ROSA/LSTF test and RELAP5 analyses on PWR cold leg small-break LOCA with accident management measure and PKL counterpart test

竹田 武司; 大津 巌

Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08

An experiment using PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with LSTF on a cold leg small-break loss-of-coolant accident with an accident management measure in a PWR. The rate of steam generator secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

報告書

Analysis of PKL Test K9 by THYDE-P Code; CSNI ISP No.10 and THYDE-P Sample Calculation Run 70

小杉 誠司; 佐々木 忍; 朝日 義郎

JAERI-M 82-115, 63 Pages, 1982/09

JAERI-M-82-115.pdf:1.57MB

PKLテストK9の解析をTHYDE-Pコードを用いて行なった。テストK9は、OECD-NEA-CSNIの国際標準問題No.10である。実験の目的は、重力注水による再浸水・再冠水過程を研究することにあり、両端ギロチン破断がコールドレグに起ったとして、緊急炉心冷却水をコールドレグに注入している。THYDE-Pは、加圧水型軽水炉の冷却材喪失事故におけるブローダウン及び再浸水・再冠水過程を解析するコードである。本報告では、THYDE-Pの検証及びモデル開発のために、計算結果と実験値を比較し、検討した。最適評価オプションを用いることによって、実験値との良い一致が得られた。

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