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Experimental investigation of decontamination factor dependence on aerosol concentration in pool scrubbing

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

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.


ROSA/LSTF tests and posttest analyses by RELAP5 code for accident management measures during PWR station blackout transient with loss of primary coolant and gas inflow

竹田 武司; 大津 巌

Science and Technology of Nuclear Installations, 2018, p.7635878_1 - 7635878_19, 2018/00

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.


ROSA/LSTF tests and RELAP5 posttest analyses for PWR safety system using steam generator secondary-side depressurization against effects of release of nitrogen gas dissolved in accumulator water

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:76.09(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.


SIMMER-III analyses of local fuel-coolant interactions in a simulated molten fuel pool; Effect of coolant quantity

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:1 パーセンタイル:81.41(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.


Solution monitoring evaluated by proliferation risk assessment and fuzzy optimization analysis for safeguards in a reprocessing process

鈴木 美寿; 寺尾 憲親

Science and Technology of Nuclear Installations, 2013, p.590684_1 - 590684_10, 2013/00

 パーセンタイル:100(Nuclear Science & Technology)



RELAP5 analysis of OECD/NEA ROSA project experiment simulating a PWR loss-of-feedwater transient with high-power natural circulation

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:9 パーセンタイル:29.05(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。


Safety design and evaluation in a large-scale Japan sodium-cooled fast reactor

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:4 パーセンタイル:55.75(Nuclear Science & Technology)


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