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久保 光太郎
Science and Technology of Nuclear Installations, 2023, p.7402217_1 - 7402217_12, 2023/06
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)Probabilistic risk assessment (PRA) is an effective methodology that could be used to improve the safety of nuclear power plants in a reasonable manner. Dynamic PRA, as an advanced PRA allows for more realistic and detailed analyses by handling time-dependent information. However, the applications of this method to practical problems are limited because it remains in the research and development stage. This study aimed to investigate the possibility of utilizing dynamic PRA in risk-informed decision-making. Specifically, the author performed an accident sequence precursor (ASP) analysis on the failure of emergency diesel generators that occurred at Unit 1 of the Tomari Nuclear Power Plant in Japan using dynamic PRA. The results were evaluated by comparison with the results of simplified classical PRA. The findings indicated that dynamic PRA may estimate lower risks compared with those obtained from classical PRA by reasonable modeling of alternating current power recovery. The author also showed that dynamic PRA can provide detailed information that cannot be obtained with classical PRA, such as uncertainty distribution of core damage timing and importance measure considering the system failure timing.
孫 昊旻; 柴本 泰照; 岡垣 百合亜; 与能本 泰介
Science and Technology of Nuclear Installations, 2019, p.1743982_1 - 1743982_15, 2019/06
被引用回数：13 パーセンタイル：82.61(Nuclear Science & Technology)Because a pool scrubbing is important for reducing radioactive aerosols to the environment for a nuclear reactor in a severe accident situation, many researches have been performed. However, decontamination factor (DF) dependence on aerosol concentration was seldom considered. DF dependence in the pool scrubbing with 2.4 m water submergence was investigated by light scattering aerosol spectrometers. It was observed that DF increased monotonically as decreasing particle number concentration in a constant thermohydraulic condition. Two validation experiments were conducted to confirm the observed DF dependence. In addition, characteristics of the DF dependence in different water submergences were investigated experimentally. It was found the DF dependence became more significant in higher water submergence.
竹田 武司; 大津 巌
Science and Technology of Nuclear Installations, 2018, p.7635878_1 - 7635878_19, 2018/00
被引用回数：2 パーセンタイル：20.93(Nuclear Science & Technology)Three tests were carried out with LSTF, simulating accident management (AM) measures during PWR station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total-failure of high-pressure injection system. As AM measures, steam generator (SG) depressurization was done by fully opening relief valves, and auxiliary feedwater was injected into secondary-side simultaneously. Conditions for break size and onset timing of AM measures were different. Primary pressure decreased to below 1 MPa with or without primary depressurization by fully opening pressurizer relief valve. Nonuniform flow behaviors were observed in SG U-tubes with gas ingress depending on gas accumulation rate in two tests that gas accumulated remarkably. The RELAP5/MOD3.3 code indicated remaining problems in predictions of primary pressure, SG U-tube liquid levels, and natural circulation mass flow rates after gas inflow and accumulator flow rate through analyses for two tests.
竹田 武司; 大貫 晃*; 金森 大輔*; 大津 巌
Science and Technology of Nuclear Installations, 2016, p.7481793_1 - 7481793_15, 2016/00
被引用回数：1 パーセンタイル：10.71(Nuclear Science & Technology)Two tests related to a new safety system for PWR were performed with ROSA/LSTF. The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.
Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春
Science and Technology of Nuclear Installations, 2015, p.964327_1 - 964327_14, 2015/00
被引用回数：6 パーセンタイル：45.92(Nuclear Science & Technology)To clarify the mechanisms underlying local fuel-coolant interactions (FCI) in a molten pool, in recent years several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.
鈴木 美寿; 寺尾 憲親
Science and Technology of Nuclear Installations, 2013, p.590684_1 - 590684_10, 2013/00
被引用回数：1 パーセンタイル：10.69(Nuclear Science & Technology)溶液監視装置は、再処理プラントが申告された通りに運転されていることを確かなものとする補助的手段として用いられてきた。最近、安全,保障措置、及びセキュリティを設計段階から考慮するとした活動が、原子力エネルギーの効果的効率的発展のために必要であるとして推進されている。この活動において、核拡散リスク評価は保障措置におけるリスク生起及び拡散リスクの頻度を考慮するために用いることができる。本研究では、保障措置及びセキュリティに対するリスク評価手法について議論し、不確実な条件下での施設の誤使用に対する意図的な行為に対して適用可能なリスク概念について調べる。マルコフモデルを用いた核拡散リスク解析、ゲームモデルを用いたけん制効果、及びファジー最適化による設計の最適化について適用化検討を行い、保障措置におけるリスク及び不確実性解析の可能性について調べる。
竹田 武司; 浅香 英明*; 中村 秀夫
Science and Technology of Nuclear Installations, 2012, p.957285_1 - 957285_15, 2012/00
被引用回数：11 パーセンタイル：63.41(Nuclear Science & Technology)PWRでの高圧注入系全故障を伴う主給水喪失事象において、スクラム失敗による高出力自然循環を仮定したOECD/NEA ROSA-2プロジェクト実験をLSTFを用いて行った。このとき、高出力自然循環を長期間よく観察するために補助給水有りの条件とし、炉心出力はRELAP5コードを用いたPWR解析により決定した。実験では、加圧器(PZR)逃し弁と蒸気発生器(SG)逃し弁の周期的な開閉によって一次系圧力は約16Mpa、SG二次側圧力は約8MPaに各々維持され、二相自然循環は早期に開始した。その結果、SG伝熱管でゆっくりとしたfill & dump形の大きな振幅を伴う水位振動が生じるとともに、二相自然循環流量は振動を伴いつつ低下した。RELAP5コードによる実験後解析において、SG伝熱管を詳細メッシュによる9本の並行チャンネルで模擬しかつ、Wallis型の気液対向流制限モデルを伝熱管入口部に適用することで、二相自然循環時におけるSG伝熱管のランダムでかつ大きな振幅を伴う水位振動を定性的に再現した。しかし、依然として水位振動の頻度や振幅に差が生じたことや一次系ループ流量の振動振幅を過大評価するなど、コードの現象予測に課題が残存した。
山野 秀将; 久保 重信; 島川 佳郎*; 藤田 薫; 鈴木 徹; 栗坂 健一
Science and Technology of Nuclear Installations, 2012, p.614973_1 - 614973_14, 2012/00
被引用回数：14 パーセンタイル：71.22(Nuclear Science & Technology)本論文では、原子力機構における深層防護と一致したJSFRの安全要求が述べられる。代表的DBEとして主ポンプ軸固着事故及び長期外部電源喪失事故安全解析により、JSFRに取り込まれた安全設計の妥当性が確認された。また、ATWSに対しても受動的炉停止系と影響緩和対策の有効性が示された。
