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論文

Experimental study on outer surface cooling of containment vessel by using CIGMA

柴本 泰照; 石垣 将宏; 安部 諭; 与能本 泰介

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 14 Pages, 2017/09

The present paper introduces the recent outcome from the CIGMA experiments regarding containment vessel cooling, in which an external side of a vessel upper head was flooded by water. The test vessel was initially pressurized by steam and noncondensable gas (air and/or helium), and was subsequently cooled by pouring water to the outside of the vessel top. Similar experiments were performed by authors using air-steam binary system in the previous study, which showed several characteristic phenomena such as inverse temperature stratification. The experimental conditions were extended systematically in this study to investigate the effects of initial gas composition and distribution in a vessel. The measurement results indicated that natural circulation was significantly affected by distributions of each gas species. In particular, it was enhanced when the gas density became heavier after condensation on the vessel inner wall, while it was suppressed when the gas density became lighter, creating density stratification with helium-rich gas in the upper region. The results are explained by the simplified model.

論文

Outcome of first containment cooling experiments using CIGMA

柴本 泰照; 与能本 泰介; 石垣 将宏; 安部 諭

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 10 Pages, 2016/10

The Japan Atomic Energy Agency (JAEA) initiated the ROSA-SA project in 2013 for the purpose of studying thermal hydraulics relevant to over-temperature containment damage, hydrogen risk, and fission product transport. For this purpose, the JAEA newly constructed the Containment InteGral Measurement Apparatus (CIGMA) in 2015 for the experiments addressing containment responses, separate effects, and accident managements. Recently, we successfully conducted first experiments using CIGMA to characterize the facility under typical experimental conditions. Among these experiments, the present paper focuses on the results of containment cooling tests, for which an upper part of the vessel outer surface was cooled by spray water. Several distinctive phenomena were observed in the tests, including inverse temperature stratification in the vessel due to the cooling in the upper region. The RELAP5 analysis result was also presented to roughly indicate the prediction capability of the best-estimate two-phase flow code in predicting the containment thermal hydraulics.

論文

Heat removal tests for PWR containment spray by large scale facility

元木 保男; 成冨 満夫; 田中 貢; 西尾 軍治; 橋本 和一郎; 木谷 進

Nuclear Technology, 63, p.316 - 329, 1983/00

 被引用回数:2 パーセンタイル:35.76(Nuclear Science & Technology)

PWR格納容器スプレイの格納容器熱除去・減圧効果を明らかにするため、JAERIモデル格納容器にPWR用スプレイノズルを用いた、格納容器スプレイ熱除去試験を実施した。この試験結果から、スプレイ水滴の熱吸収率に関しては、隣接ノズルからスプレイされる水滴の相互干渉作用による影響は小さいことが判った。また、水滴周囲条件(水蒸気と空気の存在比)と熱吸収率との関係を水滴落下距離をパラメータとして整理した。スプレイの格納容器全体の熱除去効率である総括スプレイ熱吸収率に関しては、スプレイ流量とノズル取付け高さの熱吸収率に及ぼす影響を格納容器内の気相条件(水蒸気と空気の存在比)で整理した。また、減圧効果に影響する格納容器内壁熱伝達係数については、壁面流下スプレイ流量をパラメータとして熱伝達係数と気相部条件との関係を示した。これ等の試験データを計算コードCONTEMPT-LT/022の計算と比較して、試験結果が計算コードの使用上に有効な知見である事も確認した。

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