発表形式

Initialising ...

掲載資料名

Initialising ...

発表会議名

Initialising ...

筆頭著者

Initialising ...

キーワード

Initialising ...

使用言語

Initialising ...

発行年

Initialising ...

開催年

Initialising ...

1
ROSA/LSTF test and RELAP5 analyses on PWR cold leg small-break LOCA with accident management measure and PKL counterpart test
竹田 武司; 大津 巌
Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08
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.
2
ROSA/LSTF test on nitrogen gas behavior during reflux cooling in PWR and RELAP5 post-test analysis
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.
3
Synthesized research report in the second mid-term research phase, Mizunami Underground Research Laboratory Project, Horonobe Underground Research Laboratory Project and Geo-stability Project (Translated document)
濱 克宏; 笹尾 英嗣; 岩月 輝希; 尾上 博則; 佐藤 稔紀; 藤田 朝雄; 笹本 広; 松岡 稔幸; 武田 匡樹; 青柳 和平; et al.
JAEA-Review 2016-014, 274 Pages, 2016/08
日本原子力研究開発機構は、高レベル放射性廃棄物の地層処分の実現に向けた国の第2期中期目標(平成22$$sim$$26年度)に基づき中期計画を策定し、処分事業と国による安全規制の両面を支える技術基盤を整備するため、地層処分研究開発と深地層の科学的研究の2つの領域において研究開発を進めている。今般、本中期計画期間における深地層の科学的研究分野(超深地層研究所計画、幌延深地層研究計画、地質環境の長期安定性に関する研究)の成果を取りまとめるにあたり、処分事業におけるサイト選定から処分開始に関する意思決定ポイントまでに必要な技術情報を事業者・規制機関が活用しやすい形式で体系化し、所期の目標の精密調査(前半)の段階に必要となる技術基盤として整備した。
4
Data report of ROSA/LSTF experiment TR-LF-07; Loss-of-feedwater transient with primary feed-and-bleed operation
竹田 武司
JAEA-Data/Code 2016-004, 59 Pages, 2016/07
LSTFを用いた実験(実験番号: TR-LF-07)が1992年6月23日に行われた。TR-LF-07実験では、PWRの給水喪失事象を模擬した。このとき、一次系フィード・アンド・ブリード運転とともに、補助給水系の不作動を仮定した。また、蒸気発生器(SG)の二次側水位が3mまで低下した時点でSI信号を発信し、その後30分で加圧器(PZR)の逃し弁(PORV)開放による一次系減圧を開始した。さらに、SI信号発信後12秒でPZRの有るループの高圧注入系(HPI)の作動を開始し、一次系圧力が10.7MPaまで低下した時点でPZRの無いループのHPIの作動を開始した。一次系とSG二次側の圧力は、PZRのPORVとSGの逃し弁の周期的開閉によりほぼ一定に維持された。PORVの開放にしたがい、PZRの水位が大きく低下し始め、高温側配管では水位が形成した。HPIの作動により、PZRと高温側配管の水位は回復した。一次系圧力はSG二次側圧力を下回り、両ループの蓄圧注入系(ACC)が作動した。炉心露出が生じなかったことから、PORV, HPIおよびACCを用いた一次系フィード・アンド・ブリード運転は、炉心冷却に有効であった。本報告書は、TR-LF-07実験の手順、条件および実験で観察された主な結果をまとめたものである。
5
Data report of ROSA/LSTF experiment SB-HL-12; 1% Hot leg break LOCA with SG depressurization and gas inflow
竹田 武司
JAEA-Data/Code 2015-022, 58 Pages, 2016/01
LSTFを用いた実験(実験番号: SB-HL-12)が1998年2月24日に行われた。SB-HL-12実験では、PWRの1%高温側配管小破断冷却材喪失事故を模擬した。このとき、高圧注入系の全故障とともに、蓄圧注入系(ACC)タンクからの非凝縮性ガス(窒素ガス)の流入を仮定した。また、アクシデントマネジメント(AM)策として両ループの蒸気発生器(SG)逃し弁全開による減圧を燃料棒表面最高温度が600Kに到達直後に開始した。一回目のボイルオフによる炉心露出に起因したAM策開始後、一次系圧力は低下したため、炉心二相混合水位は上昇し、燃料棒表面温度は635Kまでの上昇にとどまった。低温側配管内でのACC水と蒸気の凝縮に誘発されたループシールクリアリング(LSC)前に、二回目のボイルオフによる炉心露出が生じた。LSC後速やかに炉心水位は回復し、燃料棒表面温度は696Kまでの上昇にとどまった。窒素ガスの流入開始後、一次系とSG二次側の圧力差が大きくなった。SG伝熱管でのリフラックス凝縮時に、三回目のボイルオフによる炉心露出が生じ、燃料棒表面最高温度が908Kを超えた。本報告書は、SB-HL-12実験の手順、条件および実験で観察された主な結果をまとめたものである。
6
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
 パーセンタイル:100(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.
7
ROSA/LSTF experiment on a PWR station blackout transient with accident management measures and RELAP5 analyses
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.
8
Thermal hydraulic safety research at JAEA after the Fukushima Dai-ichi Nuclear Power Station accident
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.
9
Novel electrothermodynamic power generation
Kim, Y.*; Kim, J.*; 山中 暁*; 中島 啓*; 小川 孝*; 芹沢 毅*; 田中 裕久*; 馬場 将亮*; 福田 竜生; 吉井 賢資; et al.
Advanced Energy Materials, 5(13), p.1401942_1 - 1401942_6, 2015/07
空中に廃棄されている自動車排ガスの廃熱の再利用は、現在社会のエネルギー問題の重要な位置を占めるが、その一つとして強誘電体(焦電体)の誘電・焦電効果を応用したエネルギー回生技術の研究が進められている。焦電体を自動車排ガス中に設置するとともに、エンジン運転に伴う熱振動に同期した電場を外部から加えることで、回生エネルギーは大幅に上昇する。本研究ではこの時用いる取り出し電気回路の改良を行うとともに、典型的な焦電体PbZr$$_x$$Ti$$_{1-x}$$O$$_3$$(PZT)を用いて実際に有効活用できる回生エネルギーが非常に小さいながらもプラスであることを初めて確認した。また回生運転と同時測定した時分割X線回折により、焦電体の結晶構造の変化やドメイン比といったミクロな機構に関する知見を得るとともに、さらに実機エンジンを用いた試験でも実際に有効活用できる回生エネルギーの取得を確認できた。
10
ROSA/LSTF experiment on accident management measures during a PWR station blackout transient with pump seal leakage and RELAP5 analyses
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.
11
RELAP5 code study of ROSA/LSTF experiments on PWR safety system using steam generator secondary-side depressurization
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.
12
ROSA/LSTF experiment on AM measures during a PWR station blackout transient with pump seal leakage and RELAP5 POST-TEST analysis
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.
13
RELAP5 code study of ROSA/LSTF validation tests for PWR safety system using SG secondary-side depressurization
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.
14
Data report of ROSA/LSTF experiment SB-CL-32; 1% cold leg break LOCA with SG depressurization and no gas inflow
竹田 武司
JAEA-Data/Code 2014-021, 59 Pages, 2014/11
LSTFを用いた実験(実験番号: SB-CL-32)が1996年5月28日に行われた。SB-CL-32実験では、PWRの1%低温側配管小破断冷却材喪失事故を模擬した。このとき、非常用炉心冷却系である高圧注入系の全故障とともに、蓄圧注入系(ACC)タンクから非凝縮性ガスが流入しないと仮定した。また、アクシデントマネジメント(AM)策として両ループの蒸気発生器(SG)二次側減圧を破断後10分に一次系減圧率200K/hを目標として開始した。AM策開始後、SG二次側圧力の低下にしたがって一次系圧力は低下した。クロスオーバーレグの下降流側水位の低下とともに、ボイルオフによる炉心露出が開始した。一回目のループシールクリアリング(LSC)後速やかに炉心水位は回復し、模擬燃料棒表面温度は669Kまで上昇した。一次系減圧にしたがい低温側配管内でのACC水上の蒸気凝縮に誘発された二回目のLSC前に、ボイルオフによる炉心露出が生じた。二回目のLSC後速やかに炉心水位は回復し、観測された燃料棒表面最高温度は772Kであった。ACC隔離後、低圧注入系の注水による継続的な炉心冷却を確認して実験を終了した。本報告書は、SB-CL-32実験の手順、条件および実験で観察された主な結果をまとめたものである。
15
RELAP5 analyses on the influence of multi-dimensional flow in the core on core cooling during LSTF cold-leg intermediate break LOCA experiments in the OECD/NEA ROSA-2 Project
安部 諭; 佐藤 聡; 竹田 武司; 中村 秀夫
Journal of Nuclear Science and Technology, 51(10), p.1164 - 1176, 2014/10
 被引用回数:1 パーセンタイル:75.69(Nuclear Science & Technology)
PWRを模擬する大型熱水力試験装置LSTFを用いてコールドレグ中口径破断LOCA実験を2回(Test-2, Test-7)行い、それらの比較よりシステム全体の挙動及び熱水力現象の調査をOECD/ROSA2プロジェクトで行った。実験条件として、Test-2では破断サイズはコールドレグの17%相当とし、ECCSは単一故障を仮定した。Test-7では破断サイズは13%相当とし、ECCSは全注入を仮定した。本論文では、炉心を単チャンネルおよび複数チャンネルでノーディングしたRELAP5コード実験後解析の結果を発表する。結果として、炉心複数チャンネルモデルを用いた解析では、炉心内の多次元的な流れを考慮することができ、炉心単チャンネルモデルでは充分に再現できなかった燃料棒表面温度の挙動を再現することに成功した。
16
RELAP5 code study of ROSA/LSTF experiment on a PWR station blackout (TMLB') transient
竹田 武司; 中村 秀夫
Mechanical Engineering Journal (Internet), 1(4), p.TEP0015_1 - TEP0015_13, 2014/08
RELAP5 code analysis was performed on one of abnormal transient tests conducted with the ROSA/LSTF simulating a PWR station blackout transient with TMLB' scenario in 1995. The LSTF test revealed core uncovery by core boil-off took place a little after hot leg became empty of liquid. The code indicated remaining problems in the predictions of reverse flow U-tubes in SG during long-term single-phase liquid natural circulation. Sensitivity analyses were performed to clarify effectiveness of depressurization of and coolant injection into SG secondary-side as accident management measures. SG secondary-side depressurization was initiated by fully opening the safety valve in one of two SGs with incipience of core uncovery. Coolant injection was done into the secondary-side of the same SG at low pressures considering availability of fire engines. The SG depressurization with the coolant injection was found to well contribute to maintain core cooling by the actuation of accumulator system.
17
ROSA/LSTF experiment on a PWR station blackout transient with AM measures and RELAP5 post-test analysis
竹田 武司; 大津 巌; 与能本 泰介
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
An experiment on a PWR station blackout transient with the TMLB' scenario and accident management (AM) measures was conducted using the ROSA/LSTF at Japan Atomic Energy Agency under an assumption of non-condensable gas inflow to the primary system from accumulator tanks. The AM measures proposed in this study are steam generator (SG) secondary-side depressurization by fully opening safety valves in both SGs with the incipience of core uncovery and coolant injection into the secondary-side of both SGs at low pressures. The LSTF test revealed the primary pressure started to decrease 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 the completion of accumulator coolant injection. The RELAP5 code indicated remaining problems in the predictions of SG U-tube liquid level and primary mass flow rate after the gas ingress.
18
Deformation-driven $$p$$-wave halos at the drip-line; $$^{31}$$Ne
中村 隆司*; 小林 信之*; 近藤 洋介*; 佐藤 義輝*; Tostevin, J. A.*; 宇都野 穣; 青井 考*; 馬場 秀忠*; 福田 直樹*; Gibelin, J.*; et al.
Physical Review Letters, 112(14), p.142501_1 - 142501_5, 2014/04
 被引用回数:14 パーセンタイル:17.91(Physics, Multidisciplinary)
理化学研究所RIBFを用いて中性子過剰核$$^{31}$$Neの1中性子分離反応実験を行い、理論計算との比較から、$$^{31}$$Neが$$p$$波ハロー(一部の中性子が核内に局在せず、空間的に極めて広がっていること)を持つことを明らかにした。この実験では、ターゲットとしてクーロン分離反応が優位な鉛と核力分離反応が優位な炭素の両方を用いるとともに、脱励起$$gamma$$線も測定することによって、包括的な断面積のみならず、$$^{30}$$Neの基底状態への直接遷移のクーロン分解断面積を決めることに成功した。その実験結果を殻模型計算と比較した結果、$$^{31}$$Neの基底状態は、$$^{30}$$Neの基底状態に$$p$$波の中性子が付加されている確率が大きく、その中性子はハローになるという特異な構造を持つことがわかった。それは、変形による$$p$$波と$$f$$波の配位混合と、$$^{31}$$Neが極めて弱く束縛されていることの両面によるものであると考えられる。
19
RELAP5 analyses of ROSA/LSTF experiments on AM measures during PWR vessel bottom small-break LOCAs with gas inflow
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.
20
RELAP5/MOD3.2 sensitivity analysis using OECD/NEA ROSA-2 project 17% cold leg intermediate-break LOCA test data
竹田 武司; 渡辺 正; 丸山 結; 中村 秀夫
NEA/CSNI/R(2013)8/PART2 (Internet), p.173 - 183, 2013/12
PWRの低温側配管で17%中口径破断冷却材喪失事故(IBLOCA)が生じたことを想定し、LSTFを用いたOECD/NEA ROSA-2プロジェクト実験を行い、ループシールクリアリングの前に高速の蒸気流に起因した気液対向流制限(CCFL)により上部プレナム、蒸気発生器の伝熱管上昇流側と入口プレナムに蓄水が見られ、炉心水位の急速な低下によりドライアウトが生じた。RELAP5/MOD3.2.1.2コードを用いた実験後解析では、炉心露出が実験より遅く生じたため、燃料被覆管表面温度を過小予測した。実験分析と実験後解析の結果に基づいて、炉心水位挙動や燃料被覆管表面温度に影響を及ぼす重要現象と関与するパラメータを抽出した。実験後解析の条件をBase Caseとして、熱水力最適評価手法の不確かさ要因を調べるための感度解析を行い、安全余裕の評価に際して考慮が必要なパラメータとして、炉心出口でのWallis型CCFL相関式の係数Cや炉心内気液相間摩擦は燃料被覆管表面温度に対する影響が大きいことを確認した。