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O近藤 洋介*; Achouri, N. L.*; Al Falou, H.*; Atar, L.*; Aumann, T.*; 馬場 秀忠*; Boretzky, K.*; Caesar, C.*; Calvet, D.*; Chae, H.*; et al.
Nature, 620(7976), p.965 - 970, 2023/08
被引用回数：6 パーセンタイル：93.26(Multidisciplinary Sciences)非常に中性子が過剰な原子核Oは、陽子、中性子ともに魔法数であることから古くからその性質に興味が持たれていたが、酸素の最後の束縛核
Oよりも中性子が4個も多いため、これまで観測されてこなかった。この論文では、理化学研究所RIBFにて
Fからの1陽子ノックアウト反応によってOを生成し、そこから放出される中性子を測定することによって初めてその観測に成功した。核構造の観点からは、Oでは二重閉殻が保たれているか興味が持たれていたが、実験で得られた分光学的因子が殻模型計算で予言されて程度の大きいことから、閉殻構造をもたない可能性が高いことがわかった。
Wang, H.*; 大津 秀暁*; 千賀 信幸*; 川瀬 頌一郎*; 武内 聡*; 炭竃 聡之*; 小山 俊平*; 櫻井 博儀*; 渡辺 幸信*; 中山 梓介; et al.
Communications Physics (Internet), 2(1), p.78_1 - 78_6, 2019/07
被引用回数：8 パーセンタイル：55.71(Physics, Multidisciplinary)陽子(あるいは中性子)過剰核の効率的な生成経路を探索することは、原子核反応研究の主な動機のひとつである。本研究では、
Pdに対する核子当たり50MeVの陽子および重陽子入射による残留核生成断面積を逆運動学法によって測定した。その結果、重陽子入射ではAgやPd同位体の生成断面積が大きくなることを実験的に示した。また、理論計算による解析から、この生成断面積の増大は重陽子の不完全融合反応に起因することを示した。これらの結果は、陽子過剰核の生成において重陽子のような弱束縛核の利用が有効であることを示すものである。
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
Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08
被引用回数：14 パーセンタイル：79.14(Nuclear Science & Technology)An experiment was conducted for OECD/NEA ROSA-2 Project using LSTF, which simulated 17% hot leg intermediate-break LOCA in PWR. Core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on upper core plate. Results of uncertainty analysis with RELAP5/MOD3.3 code clarified influences of combination of multiple uncertain parameters on peak cladding temperature within defined uncertain ranges. An experiment was performed for OECD/NEA PKL-3 Project with PKL. The LSTF test simulated PWR 1% hot leg small-break LOCA with steam generator secondary-side depressurization as accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for primary pressure, core collapsed liquid level, and cladding surface temperature probably due to effects of differences between LSTF and PKL in configuration, geometry, and volumetric size.
知見 康弘; 岩田 景子; 飛田 徹; 大津 拓与; 高見澤 悠; 吉本 賢太郎*; 村上 毅*; 塙 悟史; 西山 裕孝
JAEA-Research 2017-018, 122 Pages, 2018/03
JAEA-Research-2017-018.pdf:44.03MB
原子炉圧力容器の加圧熱衝撃(Pressurized Thermal Shock: PTS)事象に対する構造健全性評価に与える影響項目の一つである高温予荷重(Warm Pre-stress: WPS)効果は、高温時に予め荷重を受けた場合に、冷却中の荷重減少過程では破壊が生じず、低温での再負荷時の破壊靱性が見かけ上増加する現象である。WPS効果については、主として弾性データによって再負荷時の見かけの破壊靱性を予測するための工学的評価モデルが提案されているが、試験片の寸法効果や表面亀裂に対して必要となる弾塑性評価は考慮されていない。本研究では、実機におけるPTS時の過渡事象を模擬した荷重-温度履歴を与える試験(WPS効果確認試験)を行い、WPS効果に対する試験片寸法や荷重-温度履歴の影響を確認するとともに、工学的評価モデルの検証を行った。再負荷時の見かけの破壊靭性について、予荷重時の塑性の程度が高くなると試験結果は工学的評価モデルによる予測結果を下回る傾向が見られた。比較的高い予荷重条件に対しては、塑性成分等を考慮することにより工学的評価モデルの高精度化が可能となる見通しが得られた。
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：20.74(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：60.93(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：36.71(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：10.6(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.
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.
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.
-wave halos at the drip-line; 
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：62 パーセンタイル：91.04(Physics, Multidisciplinary)理化学研究所RIBFを用いて中性子過剰核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、Neが波ハロー(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
線も測定することによって、包括的な断面積のみならず、
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、Neの基底状態は、Neの基底状態に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、変形による波と
波の配位混合と、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。
石野 雅彦; Faenov, A.*; 田中 桃子; Pikuz, T.; 保 智己*; 長谷川 登; 錦野 将元; Starikov, S. V.*; Stegailov, V. V.*; Norman, G.*; et al.
X-Ray Lasers 2012; Springer Proceedings in Physics, Vol.147, p.121 - 124, 2014/00
被引用回数：0 パーセンタイル：0(Engineering, Electrical & Electronic)We show experimentally the possibility of the precise nano-meter size surface structuring of metal surfaces induced by ultra low fluencies of pico-second soft X-ray laser single pulse. After irradiation processes, we observed the modified surfaces to understand the interactions between the soft X-ray laser pulses and various materials by a scanning electron microscope. The formations of unique modified structures caused by irradiations of the soft X-ray laser pulses were seen. On Al surface, the formations of conical structures were observed in the shallow features. On Au surface, the ripple-like structures were observed. The atomistic model of ablation is developed that reveals the ultra-low threshold fluency values of this process. Calculated ablation depth as a function of irradiation fluency is in good agreement with the experimental data presented as well as with the existing data on optical ablation. Our results will open new opportunities for nano-meter size processing for metal surfaces.
石野 雅彦; Faenov, A.*; 田中 桃子; 保 智己*; Pikuz, T.; 長谷川 登; 錦野 将元; Inogamov, N.*; Skobelev, I.*; Fortov, V.*; et al.
Proceedings of SPIE, Vol.8849, p.88490F_1 - 88490F_8, 2013/09
被引用回数：2 パーセンタイル：73.05(Optics)To study the interactions between soft X-ray laser (SXRL) beam and material surfaces, we irradiated the SXRL beam pulses having a wavelength of 13.9 nm and duration of 7 ps to Al, Au, Cu, and Si. Irradiated surfaces were observed using SEM and AFM. With single pulse irradiation, the formation of conical structures was observed on Al, and ripple-like structures were formed on Au and Cu. The conical structures on Al surface were destroyed under the multiple SXRL pulse exposures, but it was confirmed that the development of modified structures was observed after multiple pulse exposures on the Au and Cu surfaces. On the Si surface, deep holes that seemed to be melted structures induced by the accumulation of multiple pulses of irradiations were found. It was concluded that SXRL beam irradiation of various material surfaces causes different types of surface modifications, and the changes in the surface behaviors are attributed to the differences in the elemental properties, such as the melting points and the attenuation length of X-ray photons.
竹田 武司; 大津 巌; 中村 秀夫
Journal of Energy and Power Engineering, 7(6), p.1009 - 1022, 2013/06
Separate-effect experiment simulating steam direct-contact condensation on emergency core cooling system (ECCS) water in PWR cold legs during reflood phase of large-break LOCA was conducted in OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). A new test section was furnished in the downstream of the LSTF break unit horizontally attached to the cold leg. Significant condensation of steam appeared in a short distance from the simulated ECCS injection point, and the steam temperature in the test section decreased immediately after the initiation of the ECCS water injection. Total steam condensation rate estimated from the difference between steam flow rates at the test section inlet and outlet was in proportion to the simulated ECCS water mass flux until the complete condensation of steam. Clear images of high-speed video camera were successfully obtained on droplet behaviors through the viewer of the test section, especially for annular mist flow.
竹田 武司; 大津 巌; 中村 秀夫
Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 12 Pages, 2013/05
An OECD/NEA ROSA-2 Project experiment was conducted with the LSTF simulating a PWR hot leg small-break LOCA with a break size equivalent to 1.5% cold leg break under an assumption of total failure of HPI system as a counterpart to PKL-2 Project test. Major test objectives are to clarify responses of CETs versus cladding surface temperature at both of high- and low-pressure conditions corresponding to the pressure range of LSTF and PKL. Core uncovery took place in both phases with no reflux condensate. The observed peak temperature in the core was higher in the low-pressure phase because of longer core uncovery duration though core power and primary pressure were lower than in the high-pressure phase. One-dimensional representation of the core by RELAP5/MOD3.2.1.2 code indicated a limitation in the accuracy of CET responses. The lack in the multi-dimensional steam flow representation had a difficulty in the correct prediction of the peak steam temperature at the core exit.
石野 雅彦; Faenov, A. Ya.*; 田中 桃子; 保 智己*; 長谷川 登; 錦野 将元; Pikuz, T.; 海堀 岳史*; 河内 哲哉
Applied Physics A, 110(1), p.179 - 188, 2013/01
被引用回数：26 パーセンタイル：71.18(Materials Science, Multidisciplinary)To study the interactions between the picosecond soft X-ray laser (SXRL) beams and material surfaces, we irradiated the SXRL pulses having a wavelength of 13.9 nm, a duration of 7 ps to Au, Cu, and Si surfaces. Under a single pulse irradiation, the ripple-like structures were formed on the Au and Cu surfaces. These structures were different from the previously investigated conical structures formed on Al surface. And we could confirm that the developments of modified structures, i.e. growth of hillocks, on Au and Cu surfaces were observed under the multiple SXRL pulse exposures. On the Si surface, the deep holes, seemed to be melting structures, induced by the accumulation of multiple pulse irradiations were found. The SXRL beam irradiation on various material surfaces causes the different behaviors of surface modifications, and the changes of surface behaviors are attributed to the differences in elemental properties such as the attenuation lengths of X-ray photons.
