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高速実験炉「常陽」運転経験報告書; 原子炉容器内ナトリウム液位異常低下時における液位監視方法の確立

Operation experience report of experimental fast reactor JOYO; A special level monitoring for reactor vessel in the occurrance of the abnormal 1evel incident

藤枝 清; 竹内 徹; 高津戸 裕司; 今井 勝友; 小澤 健二; 堀米 利元; 照沼 誠一

not registered; not registered; not registered; not registered; Ozawa, kenji; not registered; Terunuma, Seiichi

「常陽」の原子炉容器ナトリウム液面計は,安全保護系に組み込まれている3本の誘導式ナトリウム液面計で構成されている。その測定範囲は,1本が長尺型で通常液位に対して+350mm$$sim$$-1600mm,他の2本は,+ー350mmで,1次主配管の下部レベルまでカバー出来るが,-1600mm以下の原子炉容器内ナトリウム液位を監視する手段がなかった。 このため1次補助冷却系の吐出配管が原子炉容器最下部まで挿入されていることから,1次補助冷却系電磁ポンプの吐出圧力計の指示値を用いて原子炉容器内ナトリウム液位を求めることが出来ると判断し,原子炉容器内ナトリウム液位と1次補助冷却系電磁ポンプ吐出圧力の関係を求める試験を実施した。試験の結果(1)1次補助冷却系を用いて燃料集合体上部までの原子炉容器内ナトリウム液位を推定することは,十分可能である。また,原子炉容器内ナトリウムドレン中の試験によって,原子炉容器内ナトリウムが47.5m3/hの速度で低下している過渡時においても,1次補助冷却系を用いて原子炉容器内ナトリウム液位の推定が可能であることを確認した。(2)1次補助冷却系電磁ポンプ吐出圧力,ナトリウム循環流量およびナトリウム温度から,原子炉容器内ナトリウム液位を求める近似式を導出した。(3)測定データを基に多重回帰分析を行い,1次補助冷却系電磁ポンプ吐出圧力およびナトリウム循環流量から,原子炉容器内ナトリウム液位を推定出来るグラフを作成した。

A reactor vessel in JOYO provides three induction type level meters which is defined in the safety protection system. They have two kinds of measuring range and display the sodium level below to the discharge nozzle of the primary cooling system. One is from 350mm about the normal sodium level to 1,600mm below it and other two sets are from 350mm above to 350mm below it. This report describes a special monitoring method of sodium level in the occurrence of the abnormal sodium level incident which reaches it more than 1600㎜ below the normal sodium level in the reactor vessel. The special monitoring method uses the discharge sodium pressure of the primary auxiliary cooling pump. A discharge sodium pipe from the primary auxiliary cooling pump is located in the bottom of the reactor vessel and it's discharge pressure is correlated with the reactor vessel sodium level which works back pressure to the pump. Therefore, it was assumed that abnormal sodium level which reaches it more than 1600mm below the normal sodium level can be monitored using this discharge sodium pressure. A verification test was conducted to measure the correlation of the discharge sodium pressure and the reactor vessel sodium level. Main results obtained from this test were as follows. (1)Validity of this special level monitoring method was confirmed in the sodium level range from normal to 3,390㎜ below it and in case of sodium level changing which is decreased at the rate of 47.5m$$^{3}$$/h by this test during the system sodium drain work. (2)A correlation equation is obtained using parameters of discharge sodium pressure, flow and temperature of the primary auxiliary cooling system to gain sodium level of reactor vessel. (3)Parametor chart of the reactor vessel sodium level was made using multi regressive analysis.

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