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竹田 武司; 和田 裕貴; 柴本 泰照
World Journal of Nuclear Science and Technology, 11(1), p.17 - 42, 2021/01
Many experiments have been conducted on accidents and transients of pressurized water reactor (PWR) employing the rig of safety assessment/large-scale test facility (ROSA/LSTF). Major results of the related integral effect tests with the LSTF were reviewed to experimentally identify thermal-hydraulic phenomena involved, regarding the PWR accident sequences in accordance with the new regulatory requirements for the Japanese light-water nuclear power plants. Key results of the recent integral effect tests utilizing the LSTF and future plans were presented relevant to multiple steam generator tube rupture accident with recovery operation, small-break loss-of-coolant accident (LOCA) with accident management measure on core exit temperature reliability, and small-break LOCA with thermal stratification under cold water injection from emergency core cooling system into cold legs.
竹田 武司; 大津 巌
Mechanical Engineering Journal (Internet), 5(4), p.18-00077_1 - 18-00077_14, 2018/08
We conducted an experiment focusing on nitrogen gas behavior during reflux condensation in PWR with ROSA/LSTF. The primary pressure was lower than 1 MPa under constant core power of 0.7% of volumetric-scaled (1/48) PWR nominal power. Steam generator (SG) secondary-side collapsed liquid level was maintained at certain liquid level above SG U-tube height. Nitrogen gas was injected stepwise into each SG inlet plenum at certain constant amount. The primary pressure and degree of subcooling of SG U-tubes were largely dependent on amount of nitrogen gas accumulated in SG U-tubes. Nitrogen gas accumulated from outlet towards inlet of SG U-tubes. Non-uniform flow behavior was observed among SG U-tubes with nitrogen gas ingress. The RELAP5/MOD3.3 code indicated remaining problems in predictions of the primary pressure and degree of subcooling of SG U-tubes depending on number of nitrogen gas injection. We studied further non-uniform flow behavior through sensitivity analyses.
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; 大津 巌
Annals of Nuclear Energy, 109, p.9 - 21, 2017/11
被引用回数：8 パーセンタイル：61.27(Nuclear Science & Technology)An experiment was conducted for the OECD/NEA ROSA-2 Project using LSTF, which simulated a cold leg intermediate-break loss-of-coolant accident with 17% break in a PWR. Assumptions were made such as single-failure of high-pressure and low-pressure injection systems. In the LSTF test, core dryout took place because of rapid drop in the core liquid level. Liquid was accumulated in upper plenum, SG U-tube upflow-side and inlet plena because of counter-current flow limiting (CCFL). The post-test analysis by RELAP5/MOD3.3 code revealed that peak cladding temperature (PCT) was overpredicted because of underprediction of the core liquid level due to inadequate prediction of accumulator flow rate. We found the combination of multiple uncertain parameters including the Wallis CCFL correlation at the upper core plate, core decay power, and steam convective heat transfer coefficient in the core within the defined uncertain ranges largely affected the PCT.
竹田 武司; 大津 巌
Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08
被引用回数：4 パーセンタイル：37.06(Nuclear Science & Technology)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.
竹田 武司; 大津 巌
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 11 Pages, 2017/07
An experiment focusing on nitrogen gas behavior during reflux cooling in a PWR was performed with the LSTF. The test conditions were made such as the constant core power of 0.7% of the volumetric-scaled PWR nominal power and the primary pressure of lower than 1 MPa. The steam generator (SG) secondary-side collapsed liquid level was maintained at a certain liquid level above the SG tube height. Nitrogen gas was injected stepwise into each SG inlet plenum at a certain constant amount. The primary pressure and the SG U-tube fluid temperatures were greatly dependent on the amount of nitrogen gas accumulated in the SG U-tubes. Non-uniform flow behavior was observed among the SG U-tubes with nitrogen gas ingress. The RELAP5/MOD3.3 code indicated remaining problems in the predictions of the primary pressure and the SG U-tube fluid temperatures after nitrogen gas inflow.
竹田 武司; 大津 巌
Mechanical Engineering Journal (Internet), 2(5), p.15-00132_1 - 15-00132_15, 2015/10
An experiment on a PWR station blackout transient with accident management (AM) measures was conducted using the ROSA/LSTF under an assumption of non-condensable gas inflow to primary system from accumulator tanks. The AM measures considered are SG secondary-side depressurization by fully opening safety valves in both SGs with start of core uncovery and coolant injection into the secondary-side of both SGs at low pressures. The primary pressure started to decrease when the SG primary-to-secondary heat removal resumed soon after the SG secondary-side coolant injection. The primary depressurization worsened due to the gas accumulation in SG U-tubes after accumulator completion. The RELAP5 code indicated remaining problems in the predictions of the SG U-tube collapsed liquid level and primary mass flow rate after gas ingress. The SG coolant injection flow rate was found to significantly affect the peak cladding temperature and the ACC actuation time through RELAP5 sensitivity analyses.
与能本 泰介; 柴本 泰照; 竹田 武司; 佐藤 聡; 石垣 将宏; 安部 諭; 岡垣 百合亜; 孫 昊旻; 栃尾 大輔
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5341 - 5352, 2015/08
This paper summarizes thermal-hydraulic (T/H) safety studies being conducted at JAEA based on the consideration of research issues after the Fukushima Dai-Ichi Nuclear Power Station accident. New researches have been initiated after the accident, which are related to containment thermal hydraulics and accident management (AM) measures for the prevention of core damage under severe multiple failure conditions. They are conducted in parallel with those initiated before the accident such as a research on scaling and uncertainty of the T/H phenomena which are important for the code validation. Those experimental studies are to obtain better understandings on the phenomena and establish databases for the validation of both lumped parameter (LP) and computational fluid dynamics (CFD) codes. The research project on containment thermal hydraulics is called the ROSA-SA project and investigates phenomena related to over-temperature containment damage, hydrogen risk and fission product (FP) transport. For this project, we have designed a large-scale containment vessel test facility called CIGMA (Containment InteGral Measurement Apparatus), which is characterized by the capability of conducting high-temperature experiments as well as those on hydrogen risk with CFD-grade instrumentation of high space resolution. This paper describes the plans for those researches and results obtained so far.
竹田 武司; 大津 巌
Journal of Energy and Power Sources, 2(7), p.274 - 290, 2015/07
An experiment on accident management (AM) measures during a PWR station blackout transient with leakage from primary coolant pump seals was conducted using the ROSA/LSTF under an assumption of non-condensable gas inflow to the primary system from accumulator tanks. The AM measures are steam generator (SG) secondary-side depressurization by fully opening safety valves (SVs) in both SGs and primary-side depressurization by fully opening SV in pressurizer with the start of core uncovery and coolant injection into the SG secondary-side at low pressures. The decrease was accelerated in the primary pressure when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes after accumulator completion. Remaining problems in the RELAP5 code include the predictions of pressure difference between the primary and SG secondary sides after the gas inflow.
竹田 武司; 大貫 晃*; 西 弘昭*
Journal of Energy and Power Engineering, 9(5), p.426 - 442, 2015/05
RELAP5 code analyses were performed on two ROSA/LSTF validation tests for PWR safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using SG (steam generator) secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place. Core collapsed liquid level recovered after the initiation of accumulator coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection system. Adjustment of break discharge coefficient for two-phase discharge flow predicted the break flow rate reasonably well. The code overpredicted the peak cladding temperature because of underprediction of the core collapsed liquid level due to inadequate prediction of the accumulator flow rate in the 8-in. break case.
竹田 武司; 大津 巌
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05
An experiment on accident management (AM) measures during a PWR station blackout transient with the TMLB' scenario and leakage from primary coolant pump seals was conducted using the ROSA/LSTF under an assumption of non-condensable gas inflow to the primary system from accumulator tanks. The AM measures are steam generator (SG) secondary-side depressurization by fully opening safety valves (SVs) in both SGs and primary-side depressurization by fully opening SV in pressurizer with the start of core uncovery and coolant injection into the SG secondary-side at low pressures. The decrease was accelerated in the primary pressure when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes. The RELAP5 code indicated remaining problems in the predictions of the primary pressure and SG U-tube collapsed liquid level.
竹田 武司; 大貫 晃*; 西 弘昭*
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 13 Pages, 2014/12
RELAP5 code post-test analyses were performed on two ROSA/LSTF validation tests for PWR safety system that simulated cold leg small-break LOCAs using SG secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves a little after a safety injection signal. In the 8-in. break test, core uncovery and heatup took place by boil-off. Core collapsed liquid level recovered after accumulator coolant injection. In the 4-in. break test, no core uncovery and heatup happened. Adjustment of break discharge coefficient for two-phase discharge flow predicted the break flow rate reasonably well. The code overpredicted the peak cladding temperature (PCT) because of underprediction of the core collapsed liquid level in the 8-in. break case. Sensitivity analyses indicated that a time delay for SG depressurization start and break discharge coefficient for two-phase discharge flow affect the PCT significantly in the 8-in. break case.
竹田 武司
International Journal of Nuclear Energy, 2014, p.803470_1 - 803470_17, 2014/00
RELAP5 code analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break LOCAs with different AM measures under an assumption of non-condensable gas inflow. Depressurization of and auxiliary feedwater (AFW) injection into both steam generators (SGs) as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI) system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow.
Susyadi; 与能本 泰介
JAERI-Research 2005-011, 64 Pages, 2005/06
JAERI-Research-2005-011.pdf:2.57MB
ROSA/LSTF実験で観測された蒸気発生器(SG)U字管群での非一様流動に着目しPWRにおける定常自然循環を検討した。RELAP5/MOD3コードを用いた解析では、SG挙動を、一次系を1本、又は、5本、又は、9本の平行流路で表し、実験に基づく境界条件を使用するSGモデルを用いて解析した。その結果、5ないし9本の平行流路を用いる場合、逆流,流入と排水,二相成層のような重要な非一様流動現象や、実験と同様な安定な出口流動を再現できることがわかった。しかし二次系への伝熱量は最大15%過小評価された。さらに、特に低圧条件において注意深く入口流量を設定する場合のみ安定な自然循環挙動が得られるなど、定常状態を確立するための問題が見いだされた。
与能本 泰介; 大津 巌
Proceedings of 12th Pacific Basin Nuclear Conference (PBNC 2000), Vol.1, p.317 - 329, 2000/00
蒸気発生器(SG)二次側除熱により事故後の崩壊熱除去を行うPWRにおいては、大気圧近傍圧力での自然循環除熱挙動を明らかにすることが重要である。本論文では、ROSA/LSTF装置を用いて、これに関して行った二つの実験について述べる。いずれの実験でもSG伝熱管群で気液二相が上下に分離した伝熱管(停滞管)と凝縮を伴いつつ二相流が流れる伝熱管が共存する非一様な観測された。停滞管では熱伝達が生じないため、非一様流動は一次系からSG二次側へ実効的な伝熱面積を減少させる効果がある。現行解析コードでは、この効果をモデル化できないため、自然循環挙動を適切に予測することはできない。重力注入水中の溶存空気の影響を検討した実験では、伝熱管への溶存空気の蓄積が観測されたが、7時間に渡る実験期間中に伝熱劣化は見られなかった。これは、空気が、伝熱に寄与しない停滞管に選択的に蓄積したことによる。この結果は、SG二次側除熱と重力注入を用いて、冷却材喪失事故後の長期冷却を行うシステムの有望性を示すものである。
竹田 武司; 大津 巌
no journal, ,
福島事故を踏まえ、PWRの一次冷却材ポンプシール部の漏洩を考慮した全電源喪失におけるアクシデントマネジメント(AM)策の検討に資するため、LSTFによる模擬実験を実施した。RELAP5/MOD3.2コードを用いた解析を基に、高圧ボイルオフ時の燃料被覆管温度への炉心ボイド率や燃料被覆管表面熱伝達率の影響を調べた。さらに、AM策に関するRELAP5コードによる感度解析を行い、燃料被覆管上昇温度に連動する蒸気発生器(SG)二次側安全弁全開によるSG二次側の減圧開始時間が遅く、かつ給水流量が少なくなるほど燃料被覆管最高温度が上昇することを明らかにした。
竹田 武司; 大津 巌
no journal, ,
PWR事故時に蓄圧注入系の隔離失敗により窒素ガスが一次系へ流入する際に一次系の冷却材量等が減圧阻害現象に与える影響を調べるため、ROSA/LSTFを用いて、低圧・リフラックス冷却条件下での窒素ガス挙動に着目した個別効果実験を実施した。一次系圧力や蒸気発生器(SG)伝熱管群の流動挙動は、蓄積したガス量に依存することを明らかにした。また、RELAP5/MOD3.3コードによる事後解析を通じて、窒素ガス流入後の一次系圧力やSG伝熱管流体温度の予測に課題があることが分かった。
竹田 武司; 大津 巌
no journal, ,
PWR全電源喪失時に一次冷却材喪失を伴う事象におけるアクシデントマネジメント策の有効性等を確認するため、ROSA/LSTFを用いて、窒素ガスの一次系への流入を仮定した条件で模擬実験を実施した。窒素ガスの流入により、一次系圧力の低下率が小さくなるとともに、蒸気発生器(SG)伝熱管群において非一様な流動を示した。また、RELAP5/MOD3.3コードにより事後解析を通じて、窒素ガス流入後の一次系圧力やSG伝熱管の水位等の予測に課題があることが分かった。
竹田 武司; 大津 巌
no journal, ,
PWR蒸気発生器(SG)伝熱管破断を模擬したROSA/LSTF実験のRELAP5/MOD3.3コードによる事後解析により、ループ間で異なる自然循環等主要現象は予測できたが、健全ループSG減圧後の破断ループSG二次側圧力等の予測に課題があることが分かった。また、感度解析を通じて、SG伝熱管ギロチン破断の本数や非常用炉心冷却系の高圧注入系の作動条件等が一次・二次圧力等に与える影響を明らかにした。
