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下平 昌樹; 飛田 徹; 高見澤 悠; 勝山 仁哉; 塙 悟史
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08
JEAC4206「原子炉圧力容器に対する供用期間中の破壊靭性の確認方法」では、加圧熱衝撃事象時の原子炉圧力容器(RPV)の健全性評価において、原子炉圧力容器内面のステンレスオーバーレイクラッド(クラッド)下亀裂(UCC)を想定し、亀裂先端の応力拡大係数がRPV鋼の破壊靭性値を上回らないことを定めている。本研究では、クラッドの存在が破壊靭性値に与える影響を評価することを目的に、UCCまたは表面亀裂を有する試験体を用いた3点曲げ破壊靭性試験と有限要素解析(FEA)を行い、UCCに対する塑性拘束効果の影響を調べた。その結果、UCCの破壊靭性値が表面亀裂に比べて高いことを実験的に示した。また、有限要素解析により、クラッドの存在によりUCCの塑性拘束効果が弱められることを示した。
高見澤 悠; 勝山 仁哉; 河 侑成; 飛田 徹; 西山 裕孝; 鬼沢 邦雄
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 8 Pages, 2019/07
原子炉圧力容器鋼の溶接熱影響部(HAZ)について、実際の原子炉圧力容器を模擬した継手溶接材(継手HAZ)及びHAZの代表的な組織を再現した熱処理材を作製し、JRR-3を用いた中性子照射試験及び照射後試験を実施し、照射前後の微細組織変化及び機械的特性変化を調べた。未照射材において、継手HAZ及び細粒HAZの破壊靭性が母材よりも低く、その要因が島状マルテンサイトやフェライト相の存在に因ることを明らかにした。また、粗粒HAZの中性子照射脆化感受性は母材よりも小さい値を示し、継手HAZ及び細粒HAZは母材と同等であることを明らかにした。
飛田 徹; 西山 裕孝; 鬼沢 邦雄
JAEA-Data/Code 2018-013, 60 Pages, 2018/11
JAEA-Data-Code-2018-013.pdf:1.67MB
原子炉圧力容器の健全性を判断する上で、破壊靱性をはじめとする材料の機械的特性は重要な情報となる。本レポートは、日本原子力研究開発機構が取得した中性子照射材を含む原子炉圧力容器鋼材の機械的特性、具体的には引張試験, シャルピー衝撃試験, 落重試験及び破壊靱性試験の公開データをまとめたものである。対象とした材料は、初期プラントから最新プラント相当の不純物含有量及び靱性レベルで製造されたJIS SQV2A(ASTM A533B Class1)相当の5種類の原子炉圧力容器鋼である。また母材に加え、原子炉圧力容器の内張りとして用いられている2種類のステンレスオーバーレイクラッド材の機械的特性データについても記載した。これらの機械的特性データは、材料ごとにグラフで整理するとともに今後のデータの活用しやすさを考慮して表形式でリスト化した。
河 侑成; 飛田 徹; 大津 拓与; 高見澤 悠; 西山 裕孝
Journal of Pressure Vessel Technology, 140(5), p.051402_1 - 051402_6, 2018/10
被引用回数：1 パーセンタイル：82.95(Engineering, Mechanical)The applicability of miniature compact tension (Mini-C(T)) specimens to fracture toughness evaluation of neutron-irradiated reactor pressure vessel (RPV) steels was investigated. Three types of RPV steels neutron-irradiated to a high-fluence region were prepared and manufactured as Mini-C(T) specimens according to Japan Electric Association Code (JEAC) 4216-2015. Through careful selection of the test temperature by considering previously obtained mechanical properties data, valid fracture toughness, and reference temperature T
was obtained with a relatively small number of specimens. Comparing the fracture toughness and T values determined using other larger specimens with those determined using the Mini-C(T) specimens, T values of both unirradiated and irradiated Mini-C(T) specimens were found to be the acceptable margin of error. The scatter of 1T-equivalent fracture toughness values of both unirradiated and irradiated materials obtained using Mini-C(T) specimens did not differ significantly from the values obtained using larger specimens. The correlation between the Charpy 41J transition temperature (T
) and the T values agreed very well with that of the data in the literature, regardless of specimen size and fracture toughness of the materials before irradiation. Based on these findings, it was concluded that Mini-C(T) specimens can be applied to fracture toughness evaluation of neutron-irradiated materials without significant specimen size dependence.
河 侑成; 飛田 徹; 高見澤 悠; 塙 悟史; 西山 裕孝
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 6 Pages, 2018/07
An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature (T) values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40
C
60C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.
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。
Shamsuzzaman, M.*; 堀江 達郎*; 浮池 亮太*; 神山 基紀*; 森岡 徹*; 松元 達也*; 守田 幸路*; 田上 浩孝; 鈴木 徹*; 飛田 吉春
Annals of Nuclear Energy, 111, p.474 - 486, 2018/01
被引用回数：6 パーセンタイル：21.94(Nuclear Science & Technology)Particle bed characteristics are experimentally investigated for the sedimentation and subsequent bed formation of solid particles, related to the coolability aspects in core-disruptive accidents. Presently a series of experiments with gravity driven discharge of solid particles into a quiescent water pool was performed to evaluate bed formation characteristic in the course of particle sedimentation. We evaluated the effects of the crucial factors: nozzle diameter, particle density, particle diameter and nozzle height on four key quantitative parameters of bed shape: mound dimple area, mound dimple volume, repose angle and mound height to illustrate the role of the crucial factors on forming the particle bed shape. The investigated crucial factors exhibit a significant role that diversifies the particle bed formation process. Based on the data obtained in the experimental observations, we developed an empirical correlation to compare the predicted results with the experimental bed heights. The proposed empirical correlation can reasonably demonstrate the general trend of the experimental bed height. This correlation could be useful to assess the particle bed elevation, and to identify the governing parameters.
河 侑成; 飛田 徹; 高見澤 悠; 西山 裕孝
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 5 Pages, 2017/07
圧力容器鋼の破壊靱性評価へのミニチュアC(T)試験片の適用性を調べるため、中性子照射された圧力容器鋼材のシャルピー試験片からミニチュアC(T)試験片を加工するとともに破壊靱性試験に供し、参照温度
を評価した。その結果、ミニチュアC(T)試験片で得られるは疲労予亀裂入りシャルピー型破壊靭性試験片から得られる値とよく一致すること、ミニチュアC(T)試験片から得られる1T-C(T)相当の破壊靱性値のばらつきは疲労予亀裂入りシャルピー型破壊靭性試験片等から得られるものと大差が無いこと、参照温度とシャルピー吸収エネルギー41Jレベルの延性脆性遷移温度の関係は、米国データのばらつきの範囲内にあることが明らかになった。
小野田 雄一; 松場 賢一; 飛田 吉春; 鈴木 徹
Mechanical Engineering Journal (Internet), 4(3), p.16-00597_1 - 16-00597_14, 2017/06
For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow accident (ULOF) were preliminarily analyzed. The objective of this study is to demonstrate the integrity of the reactor vessel against the mechanical load induced by the energetics. Conservative energy production was assumed in order to confirm the robustness of the safety design of MONJU. Mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented. The mechanical energy, which were obtained by analyzing the expanding behavior of core materials after energetics, were about one order of magnitude below the thermodynamic work potential calculated by assuming isentropic expansion of the fuel vapor to one atmosphere, which was often used as an indicator to express the severity of the energetics. Structural integrity was then evaluated with coupled fluid-structure dynamics code using the obtained mechanical energy. No or very small circumferential residual strain of the reactor vessel was evaluated in most analytical cases, and even in the most conservative energy production case, the residual strain was only 0.008 % so that the integrity of the reactor vessel is maintained. The result obtained in the present study shows that MONJU has enough robustness against the mechanical load under energetics.
曽我部 丞司; 鈴木 徹; 和田 雄作; 飛田 吉春
日本機械学会論文集(インターネット), 83(848), p.16-00393_1 - 16-00393_10, 2017/04
高速炉の代表的な炉停止失敗事象である冷却材流量喪失時炉停止失敗事象(ULOF: Unprotected Loss of Flow)の熱的影響を評価するためには、事故が核的に終息した後の再配置/冷却過程において、損傷炉心物質が炉容器内のどこに再配置し、それぞれの場所で長期にわたって安定冷却できるかを示す必要がある。本報では、IVR(In-Vessel Retention)成立性に関する見通しを得るために実施した低圧プレナム移行燃料及び炉心残留燃料の安定冷却性評価について報告する。
鈴木 徹; 曽我部 丞司; 飛田 吉春; 堺 公明*; 中井 良大
日本機械学会論文集(インターネット), 83(848), p.16-00395_1 - 16-00395_9, 2017/04
高速炉の炉停止失敗事象(ATWS: Anticipated Transient without Scram)に対して、原子炉容器内での事象終息(IVR: In-Vessel Retention)の成立性を検討した。検討においては、確率論的評価に基づいて冷却材流量喪失時炉停止失敗事象(ULOF: Unprotected Loss of Flow)をATWSの代表事象に選定した上で、総合的安全解析コードや個別物理モデルを活用して炉心損傷時の事象進展を解析し、事故の機械的影響と熱的影響を評価した。本検討の結果から、原子炉容器は機械的にも熱的にも損傷することはなく、IVRが成立する見通しを得ることができた。
曽我部 丞司; 鈴木 徹; 和田 雄作; 飛田 吉春
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
The achievement of in-vessel retention (IVR) of accident consequences in an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, is an effective and rational approach to enhancing the safety characteristics of the sodium-cooled fast reactor. The objective of the present study is to show that the decay heat generated from the relocated fuels would be stably removed in post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phase, where the relocated fuels mean fuel discharged from the core into low-pressure plenum through control-rod guide tubes, and fuel remnant in the disrupted core region (non-discharged fuel). As a result of the assessment, it can be concluded that the stable cooling of the relocated fuels was confirmed and the prospect of IVR was obtained.
豊岡 淳一; 神山 健司; 飛田 吉春; 鈴木 徹
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
In this paper, for the purpose establishing more generalized models for the SIMMER code to reproduce the effect of steel component on mixture-to-wall heat transfer in the EAGLE-1 program, the authors performed a model improvement for the SIMMER code to treat the direct contact of the molten steel in a more mechanistic manner. By this model improvement, evaluations with unifying agreement on a result of the EAGLE-1 program using the SIMMER code could be possible.
松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 8 Pages, 2016/10
ナトリウム冷却高速炉の炉心損傷時に原子炉容器下部プレナムへ流出した溶融炉心物質がデブリ化するまでの距離の評価を目的として、溶融炉心模擬物質を冷却材中へ放出させる試験を行い、デブリ化距離と流出条件の関係を実験相関式として整理した。実験相関式による予測は実験結果とよく一致した。本研究により、冷却材の沸騰・膨張によるデブリ化促進効果を考慮することで、ナトリウム中におけるデブリ化距離を適切に評価可能であることがわかった。
岩田 景子; 飛田 徹; 高見澤 悠; 知見 康弘; 吉本 賢太郎*; 西山 裕孝
Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 6 Pages, 2016/07
原子炉圧力容器の破壊靱性に対する高温予荷重効果について評価を行った。1T-CT試験片を用いて様々な熱力学的荷重負荷条件でWPS試験を実施した。得られた結果は4つのWPS工学モデルによる予測値と比較した。また1T-, 0.4T-, 0.16T-CT試験片を用いて予荷重負荷-除荷-冷却-低温再負荷条件で試験片寸法依存性について調査した。試験片寸法による塑性域分布や残留応力分布の違いを解析により明らかにした。
曽我部 丞司; 鈴木 徹; 和田 雄作; 飛田 吉春
第21回動力・エネルギー技術シンポジウム講演論文集(USB Flash Drive), 3 Pages, 2016/06
高速炉の代表的な炉停止失敗事象(ATWS)である冷却材流量喪失時炉停止失敗事象(ULOF: Unprotected Loss of Flow)の再配置/冷却過程における事象推移を評価・検討し、IVR成立の見通しを得た。
鈴木 徹; 曽我部 丞司; 飛田 吉春; 堺 公明*; 中井 良大
第21回動力・エネルギー技術シンポジウム講演論文集(USB Flash Drive), 4 Pages, 2016/06
The achievement of In-Vessel Retention (IVR) against Anticipated Transient without Scram (ATWS) is an effective and rational approach in enhancing the safety characteristics of sodium-cooled fast reactors. Based on the Probabilistic Risk Assessment (Level 1 PRA) for a prototype fast-breeder reactor, Unprotected Loss of Flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, can be selected as a representative event of ATWS. The objective of the present study is to show that no significant mechanical energy release would take place during core disruption caused by ULOF, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. As a result of the present evaluation with computational codes and physical models reflecting the knowledge on relevant experimental studies, the prospect of IVR against ULOF was obtained.
小野田 雄一; 松場 賢一; 飛田 吉春; 鈴木 徹
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 10 Pages, 2016/06
For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow accident were preliminarily analyzed. The objective of this study is to demonstrate the integrity of the reactor vessel against the mechanical load induced by the energetics. Conservative energy production was assumed in order to confirm the robustness of the safety design of MONJU. Mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented. The mechanical energy, which were obtained by analyzing the expanding behavior of core materials after energetics, were about one order of magnitude below the thermodynamic work potential calculated by assuming isentropic expansion of the fuel vapor to one atmosphere, which was often used as an indicator to express the severity of the energetics. Structural integrity was then evaluated with coupled fluid-structure dynamics code using the obtained mechanical energy. No or very small circumferential residual strain of the reactor vessel was evaluated in most analytical cases, and even in the most conservative energy production case, the residual strain was only 0.008 % so that the integrity of the reactor vessel is maintained. The result obtained in the present study shows that MONJU has enough robustness against the mechanical load under energetics.
飛田 吉春; 神山 健司; 田上 浩孝; 松場 賢一; 鈴木 徹; 磯崎 三喜男; 山野 秀将; 守田 幸路*; Guo, L.*; Zhang, B.*
Journal of Nuclear Science and Technology, 53(5), p.698 - 706, 2016/05
被引用回数：6 パーセンタイル：31.55(Nuclear Science & Technology)炉心損傷事故(CDA)の炉内格納(IVR)はナトリウム冷却高速炉(SFR)の安全特性向上において極めて重要である。SFRのCDAにおいては、溶融炉心物質が炉容器の下部プレナムへ再配置し、構造物へ重大な熱的影響を及ぼし、炉容器の溶融貫通に至る可能性がある。この再配置過程の評価を可能とし、SFRのCDAではIVRで終息することが最も確からしいことを示すため、SFRのCDAにおける物質再配置挙動の評価手法を開発する研究計画が実施された。この計画では、炉心領域からの溶融物質流出挙動の解析手法、溶融炉心物質のナトリウムプール中への侵入挙動、デブリベッド挙動のシミュレーション手法を開発した。
Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春
Annals of Nuclear Energy, 85, p.740 - 752, 2015/11
被引用回数：19 パーセンタイル：7.56(Nuclear Science & Technology)Studies on local fuel-coolant interactions (FCI) in a molten pool are crucial to the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs). To clarify the characteristics of this interaction, in recent years a series of simulated experiments, which covers a variety of conditions including much difference in water volume, melt temperature, water subcooling and water release site (pool surface or bottom), was 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, motivated by acquiring further evidence for understanding its mechanisms, 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 confirmed that, similar to experiments, the water volume, melt temperature and water release site are observable to have remarkable impact on the interaction, while the role of water subcooling seems to be less prominent. The performed analyses also suggest that the most probable reason leading to the limited pressurization and resultant mechanical energy release for a given melt and water temperature within the non-film boiling range, even under a condition of much larger volume of water entrapped within the pool, should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.
