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telier effect in an Al-Mg-Zn-based crossover aluminum alloyZhang, X.*; Li, Y.*; Wei, S.*; Guo, H.*; He, Z.*; Yang, C.*; Gong, W.; Harjo, S.; Zhou, D.*; Li, Z.*; et al.
Acta Materialia, 308, p.121990_1 - 121990_18, 2026/04
被引用回数：7 パーセンタイル：98.38(Materials Science, Multidisciplinary)Dynamic strain aging (DSA) causes serrated flow, known as the Portevin-Le Chtelier (PLC) effect, through interactions between solute atoms and dislocations. In this study, tensile tests on an Al-Mg-Zn crossover solid-solution alloy revealed that DSA and strain-induced dynamic precipitation jointly control serration behavior. At low strain rates, DSA-assisted precipitation produced Type C serrations, whereas high strain rates suppressed precipitation and promoted Type A serrations. High-energy X-ray diffraction and TEM analyses showed that screw dislocations dominated early deformation and governed plastic flow localization through cross-slip. Frequent cross-slip at high strain rates generated dislocation microbands responsible for Type A serrations, highlighting the primary role of screw dislocations rather than edge dislocations. A constitutive model incorporating DSA-assisted strengthening successfully predicted flow behavior over a wide strain-rate range.
neutron diffraction mapping measurement山下 享介*; 柳樂 知也*; Gong, W.; 川崎 卓郎; Harjo, S.; 潮田 浩作*; 藤井 英俊*
ISIJ International, 66(5), p.673 - 684, 2026/04
被引用回数：0 パーセンタイル：0.00(Metallurgy & Metallurgical Engineering)In this study, neutron diffraction mapping was performed on linear friction welded (LFW) joints of 12 mm thick high-phosphorus weathering steel (SPA-H) to evaluate residual stress, dislocation density, and crystallographic orientation. Welding was conducted under applied pressures of 100 and 250 MPa. The weld interface mainly consisted of refined ferrite with minor retained austenite and martensite, indicating reverse transformation to austenite during welding. The 250 MPa condition resulted in a lower welding temperature. Elongated grains were observed near the surface along the oscillation direction, while equiaxed grains appeared at the center. Both joints showed high tensile residual stresses at the weld center and compressive stresses near the surface. Higher applied pressure increased dislocation density because of suppressed dynamic recovery. Strong texture formation due to plastic flow was observed, while the effect of applied pressure on texture development was limited.
山下 享介*; 柳樂 知也*; Gong, W.; 川崎 卓郎; Harjo, S.; 潮田 浩作*; 藤井 英俊*
鉄と鋼, 111(17), p.1057 - 1071, 2025/12
Neutron diffraction mapping was conducted on linear friction welded joints of 12 mm thick high-phosphorus weathering steel (SPA-H) to evaluate residual stress, dislocation density, and crystallographic orientation. Welding was performed at 100 MPa and 250 MPa. The weld interface mainly comprised refined ferrite with retained austenite and martensite, indicating temperatures exceeded A1 and induced reverse transformation. The 250 MPa joint showed a lower welding temperature. Elongated grains formed near the surface along the oscillation direction (OD), while equiaxed grains appeared at the center. High tensile residual stresses were found at the weld center, with compressive stress near the surface perpendicular to the weld. Pressure had little effect on overall stress trends. Dislocation density increased with pressure due to suppressed dynamic recovery. Strong texture developed at the interface, with limited pressure dependence.
Jeong, S. G.*; Kwon, J.*; Kim, E. S.*; Prasad, K.*; Harjo, S.; Gong, W.; 川崎 卓郎; Estrin, Y.*; Bouaziz, O.*; Hong, S. I.*; et al.
Materials Science & Engineering A, 942, p.148712_1 - 148712_11, 2025/10
被引用回数：2 パーセンタイル：46.97(Nanoscience & Nanotechnology)The cellular structure plays a key role in determining the mechanical properties of metal additive manufacturing (MAM) components. This study presents in situ neutron diffraction and dislocation density-based modeling for a CoCrFeMnNi high-entropy alloy (HEA) made via directed energy deposition (DED). A constitutive model based on the Kocks-Mecking-Estrin framework was used to represent the cellular structure. Parametric analysis showed lower dislocation accumulation and annihilation rates in the as-built sample (with cellular structure) than in the heat-treated one. These differences are linked to dislocation forest networks and local stacking fault energy variations. Dislocation density across cell interiors and walls was also compared with deformation-induced dislocation cells.
neutron diffractionLiu, Y.*; Yan, Z.*; Gao, Y.*; Li, Y.*; Gan, B.*; Harjo, S.; Gong, W.; 川崎 卓郎; Li, S.*; Wang, Y.-D.*
Microstructures (Internet), 5(4), p.2025096_1 - 2025096_15, 2025/10
The micromechanical behaviors and dislocation evolution in a polycrystalline Ni-Co-based superalloy were systematically investigated by 
neutron diffraction tensile testing combined with line profile analysis. The results reveal the sequential activation of 
' shearing and Orowan looping mechanisms, with interphase load partitioning governed by strain-dependent interactions of dislocation and precipitate. During the initial plastic deformation, the and ' phases undergo co-deformation through dislocation shearing without load transfer, while the Orowan looping facilitates the load transfer from to ' phase at a higher strain level. Furthermore, the low stacking fault energy leads to a rising fraction of screw dislocations by suppressing cross-slip. Crucially, the pinning effect of ' precipitates hinders the rearrangement of these dislocations into low-energy structures, resulting in the formation of high-energy, weakly screened dislocation configurations. These findings provide new evidence for the planar slip dominance in Ni-Co-based superalloys, enabling quantitative assessment of microstructural evolution and micromechanical responses.
垣内 一雄; 成川 隆文*; 宇田川 豊; 勝山 仁哉; 三原 武; 天谷 政樹
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1440 - 1449, 2025/10
Phenomena of fuel fragmentation, relocation, and dispersal (FFRD) of high burnup light water reactor fuels have been observed under simulated loss of coolant accident (LOCA) experiments. If the fuel fragments accumulate densely in the ballooned cladding during LOCA, the power of fuel rod may increase locally, which may increase the peak cladding temperature. Furthermore, if a large number of fuel fragments were dispersed from fuel rod to reactor core, the coolability of reactor core during and after the accident may be influenced. While recent studies suggest large impact of rod internal gas state on fuel fragmentation and dispersal, there have been few experimental data that enable to evaluate such impact. We thus performed three LOCA-simulated burst tests (Test no. MMDA3 / MMDA4 / LZRT5) using irradiated PWR and BWR UO
fuel rods whose plenum volumes were designed to be 1 cc and 5 cc, respectively, as the main test parameter, at the Reactor Fuel Examination Facility (RFEF) of Japan Atomic Energy Agency (JAEA). The tests highlighted the crucial role of plenum volume in fuel rod in FFRD: the burst appearance changed from a pin hole of MMDA3 with the 1cc plenum to a rupture opening of MMDA4 with 5 cc plenum, entailing increase in probable more substantial fragmentation and fuel fragments dispersal. Based on results from MMDA3 and MMDA4, the gas state, which was influenced by both the plenum volume and the gas communication, may significantly affect the amount of fuel fragment dispersion.
Mohamad, A. B.; 山下 真一郎; 根本 義之; 阿部 陽介; Pham, V. H.; 井岡 郁夫; 相馬 康孝; 石島 暖大; 佐藤 智徳; Rizaal, M.; et al.
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), 8 Pages, 2025/10
In parallel with the Advanced Technology Fuel (ATF) development program in Japan, Japan Atomic Energy Agency has established an R&D program on the ATF to provide more scientific support to the ATF fuel vendors in Japan. Based on the identified phenomenological issues, the R&D program covers the issues mainly on the light water reactors conditions such as normal operation, loss of coolant accident, beyond design basis accident, and severe accident. The R&D program such as irradiation test, corrosion test, LOCA test, and etc. are proposed. By acquiring the data from the experiment, the final goal is to implement the experimental data and model into the simulation code in order to predict the fuel performance behaviour at high burn-up. In 2024 TOPFUEL, the introduction of the R&D has been presented. In 2025, the update on the Cr-coated research to simulate normal operation and accident conditions will be given. For example, tests are currently underway to understand the in-reactor environment, such as ion irradiation, gamma irradiation and proton irradiation tests. In addition, a loss-of-coolant accident is being simulated for accident conditions, and the chemical effects of fission products on the cladding are being studied.
高梨 美咲*; 日高 僚太*; 大久保 亘太*; 増村 拓朗*; 土山 聡宏*; 諸岡 聡; 前田 拓也*; 中村 修一*; 植森 龍治*
ISIJ International, 65(9), p.1384 - 1393, 2025/08
被引用回数：1 パーセンタイル：26.96(Metallurgy & Metallurgical Engineering)This study investigates the strengthening mechanism of ausforming in martensitic steels, focusing on the role of dislocation inheritance from austenite. By analyzing Fe-5%Mn-C alloys, the researchers quantified dislocation densities before and after ausforming and examined how carbon content affects hardening. Results show that both hardness and dislocation density of ausformed martensite increase with greater deformation in austenite and higher carbon content. The findings support that ausforming strengthens martensite through dislocation accumulation, consistent with the Bailey-Hirsch relationship. 
neutron diffraction revealed that dislocation inheritance occurs only in steels with sufficient carbon, while additional dislocations are introduced during martensitic transformation.
/KNbO nanocomposite particles米田 安宏; 小林 徹; 辻 卓也; 芝田 悟朗; 竹田 幸治*; 斎藤 祐児; Khanal, G. P.*; 藤井 一郎*; 上野 慎太郎*; 佐藤 幸生*; et al.
Japanese Journal of Applied Physics, 64(8), p.08SP07_1 - 08SP07_8, 2025/08
被引用回数：1 パーセンタイル：28.15(Physics, Applied)ナノ粒子結晶の表面に別の異なる物質を成長させた複合ナノ粒子が高品位で合成できるようになってきた。このようなナノ構造体の評価には種々のスケールでの異なる評価方法を組み合わせることが効果的である。BaTiO/KNbOナノ複合粒子の種々の構造評価と電子状態評価を放射光を利用して行った。構造評価からナノ複合粒子はコア部分が100nmのBaTiOで20nmのKNbOで覆われていることが確認できた。最表面のKNbOのO-K吸収端スペクトルはバルクやナノ粒子のそれとは異なっており、複合粒子化することで化学結合状態が変化することがわかった。
竹田 武司
JAEA-Data/Code 2025-005, 106 Pages, 2025/06
JAEA-Data-Code-2025-005.pdf:2.93MB
JAEAでは、加圧水型軽水炉(PWR)解析のためのRELAP5/MOD3.3コードの入力データを、主に大型非定常実験装置(LSTF)の参照4ループPWRである敦賀発電所2号機の設計情報を基に作成してきた。PWR解析に関する代表的なOECD/NEAの活動として、BEMUSEプログラムの枠組みにおける低温側配管大破断冷却材喪失事故(LBLOCA)の計算が挙げられる。また、わが国の新規制基準に係るPWRの炉心損傷防止対策の有効性評価事象には、低温側配管LBLOCA時の非常用炉心冷却系(ECCS)の再循環機能喪失事象が含まれる。本検討において、PWRの安全設計上想定すべき設計基準事故の一つであるLBLOCAを解析するための入力データを整備した。本報告書では、PWRLBLOCA解析の入力データの主な特徴を示す。PWRの原子炉容器、加圧器(PZR)、高温側配管、蒸気発生器(SG)、SG二次系、クロスオーバーレグ、低温側配管、ECCSなどをモデル化し、参照4ループPWRを2ループで模擬した。その際、PZRは3ループ分を模擬するループAに接続し、破断口は1ループ分を模擬するループBに設置した。PWRのコンポーネントのノード分割は、LSTFのコンポーネントのノード分割を参照した。また、PWRLBLOCA解析の主な入力データに対して、解釈を加えるとともに、設定根拠などの付加情報を提供した。さらに、整備した入力データを用いて、ECCS再循環機能喪失事象を対象とした過渡解析を実施した。RELAP5/MOD3.3コードによる既往研究の計算と比較することにより、過渡解析は概ね妥当であることを確認した。加えて、RELAP5/MOD3.3コードを用いて感度解析を実施し、破断口の流出係数や代替再循環注水流量が燃料棒被覆管表面温度に及ぼす影響を明らかにした。本報告書では、設定した条件の範囲内での感度解析結果について示し、ECCS再循環機能喪失事象に対する既往研究の計算内容の一部を補完する。
Kim, Y. S.*; Kang, T.*; Hong, S.-K.*; Brechtl, J.*; Lebyodkin, M.*; Cheng, Y.-H.*; Huang, E.-W.*; Liaw, P. K.*; Harjo, S.; Gong, W.; et al.
Acta Materialia, 292, p.120970_1 - 120970_16, 2025/06
被引用回数：15 パーセンタイル：96.30(Materials Science, Multidisciplinary)Metallic materials can exhibit low-temperature serrated deformation (LTSD) at cryogenic temperatures, potentially causing sudden failures. Understanding LTSD is thus crucial for ensuring material reliability in such environments. LTSD has been explained by two main mechanisms: (i) dislocation-based mechanical instability and (ii) thermomechanical instability, but each has limitations when considered alone. To address this, we propose a new LTSD mechanism, a thermally induced dislocation dynamics model, based on cryogenic experimental evidence. This model accounts for dislocation avalanches and localized heating, leading to hierarchical dislocation networks and transitions in deformation modes. A modified deformation-mechanism map for SS316L is also presented. Our findings highlight the rate-dependent nature of LTSD and negative strain-rate sensitivity, including the first observation of links between small stress fluctuations and large serrations.
高梨 美咲*; 日高 僚太*; 大久保 亘太*; 増村 拓朗*; 土山 聡宏*; 諸岡 聡; 前田 拓也*; 中村 修一*; 植森 龍治*
鉄と鋼, 111(9), p.503 - 513, 2025/06
被引用回数：0 パーセンタイル：0.00(Metallurgy & Metallurgical Engineering)The strengthening mechanism of ausforming in martensitic steels is believed to be due to the inheritance of dislocations in austenite by the subsequently transformed martensite. However, no studies to date have quantified the dislocation density before and after ausforming. In this study, the dislocation densities of Fe-5%Mn-C alloys were analyzed, and the relationship between hardening by ausforming and dislocation accumulation, as well as the effect of carbon on this relationship, were investigated. The hardness of ausformed martensite increased with the ausforming reduction in austenite, and the strengthening effect of ausforming increased with the addition of carbon. Similarly, the dislocation density of ausformed martensite increased with the ausforming reduction in austenite, and the dislocation accumulation by ausforming increased with the addition of carbon. Because the hardness of the ausformed martensite follows the Bailey-Hirsch relationship, the strengthening mechanism owing to ausforming could be explained by dislocation strengthening. To understand the dislocation accumulation process during ausforming, the dislocation density of austenite immediately after ausforming was measured by in-situ heating neutron diffraction. Consequently, the dislocation density of the ausformed austenite was not dependent on the carbon content, indicating that dislocations are not inherited in carbon-free steels. By contrast, in steels with sufficient carbon content, not only are dislocations inherited but additional dislocations are introduced during martensitic transformation.
Naeem, M.*; Ma, Y.*; Tian, J.*; Kong, H.*; Romero-Resendiz, L.*; Fan, Z.*; Jiang, F.*; Gong, W.; Harjo, S.; Wu, Z.*; et al.
Materials Science & Engineering A, 924, p.147819_1 - 147819_10, 2025/02
被引用回数：7 パーセンタイル：86.13(Nanoscience & Nanotechnology)Face-centered cubic (fcc) medium-/high-entropy alloys (M/HEAs) typically enhance strength and ductility at cryogenic temperatures via stacking faults, twinning, or martensitic transformation. However, in-situ neutron diffraction on VCoNi MEA at 15 K reveals that strain hardening is driven solely by rapid dislocation accumulation, without these mechanisms. This results in increased yield strength, strain hardening, and fracture strain. The behavior, explained by the Orowan equation, challenges conventional views on cryogenic strengthening in fcc M/HEAs and highlights the role of dislocation-mediated plasticity at low temperatures.
佐藤 聡; 日引 俊*; 池田 遼; 柴本 泰照
Progress in Nuclear Energy, 180, p.105593_1 - 105593_11, 2025/02
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)加圧水型原子炉(PWR)の冷却材喪失事故では、ダウンカマーに流入するコールドレグに注入された緊急炉心冷却(ECC)水の流れにより、原子炉圧力容器(RPV)内壁に加圧熱衝撃(PTS)が発生するリスクがある。PTSは、ECC水によるダウンカマー壁の急冷によって発生し、ECC水の温度、壁面へのジェットの衝突位置と速度、壁面上の液膜の速度、液膜の厚さ、下降流の広がりなどに強く影響される。したがって、コールドレグからダウンカマーに流出するECC水の流れは、PTS事象に強く影響する可能性がある。この流動現象を理解するために、円管からの自由流出に関する研究をレビューした。流動条件の分類、流動条件間の遷移条件、端部深さ比、円管内の流れの自由表面形状、管から流出するナッペの形状に関する実験結果は、ほぼ一致した形で得られている。これに対し、コールドレグからダウンカマーへの流れを考慮する場合、自由空間ではなく狭い隙間への流れ、円管出口の角の丸みの存在、炉心からコールドレグへ流れる蒸気流の影響など、特殊な状況での流れ場を扱う必要がある。しかし、これらの要因を考慮した先行研究は少ないため、今後蓄積すべき知見としてまとめた。
岡垣 百合亜; 竹田 武司; 和田 裕貴; 安部 諭; 市原 京子*; 塩谷 仁*
Proceedings of 10th Workshop on Computational Fluid Dynamics for Nuclear Reactor Safety (CFD4NRS-10) (Internet), 12 Pages, 2025/00
Core Exit Temperature (CET), measured by core exit thermocouples (T/Cs), is utilized worldwide as a crucial parameter to start Accident Management (AM) operator action by detecting core temperature excursion during accidents in Pressurized Water Reactors (PWRs). The CET is used to switch accident response procedures from preventing core damage to preventing containment failure. Various thermal-hydraulic phenomena in the core influence the behavior of CET during an accident. Previous studies have indicated that CET may rise more slowly than the fuel cladding temperature. This study aimed to deepen the understanding of the relationship between CET and fuel cladding temperature during representative accident progressions by employing Computational Fluid Dynamics (CFD) simulations. It sought to complement experimental findings by evaluating the effects of three-dimensional core thermo-hydraulic behaviors, such as secondary flows in the upper core region. The CFD simulation was performed for a 1% vessel upper head small-break Loss-Of-Coolant Accident (LOCA) experiment, which was conducted at the Large-Scale Test Facility (LSTF) at Japan Atomic Energy Agency (JAEA) in 2023. The LSTF experiment assumed that the high-pressure injection system of the emergency core cooling system had totally failed. This study focused on the representative period when a significant rise in core temperature appeared during core uncovering. The transient solver for compressible fluids in OpenFOAM was employed for the CFD simulation. Boundary conditions, such as mass flow rate, temperature, and pressure at the core's top position, were applied. The turbulence model used was the Shear Stress Transport (SST) 
model. The CET distributions were compared with the experimental data, which had a total of 20 points. The velocity and temperature distributions in the mainstream and cross-sectional directions were visualized to elucidate thermal-hydraulic phenomena. This study provided valuable insights into CET behavior and related thermo-fluid dynamics.
竹田 武司
JAEA-Data/Code 2024-014, 76 Pages, 2024/12
JAEA-Data-Code-2024-014.pdf:4.0MB
ROSA-V計画において、大型非定常実験装置(LSTF)を用いた実験(実験番号:SB-PV-03)が2002年11月19日に行われた。ROSA/LSTFSB-PV-03実験では、加圧水型原子炉(PWR)の0.2%圧力容器底部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(ECCS)である高圧注入系の全故障とともに、蓄圧注入(ACC)タンクから一次系への非凝縮性ガス(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)策として両蒸気発生器(SG)二次側減圧を安全注入設備信号発信後10分に一次系減圧率55K/hを目標として開始し、その後継続した。さらに、AM策から少し遅れて両SG二次側への30分間の補助給水を開始した。ACCタンクから一次系への窒素ガスの流入開始まで、AM策は一次系減圧に対して有効であった。ACC系から両低温側配管への間欠的な冷却材注入により、炉心水位は振動しながら回復した。このため、炉心水位は小さな低下にとどまった。窒素ガスの流入後、一次系とSG二次側の圧力差が大きくなった。窒素ガス流入下におけるSG伝熱管でのリフラックス凝縮時に、ボイルオフによる炉心露出が生じた。模擬燃料棒の被覆管表面最高温度がLSTFの炉心保護のために予め決定した値(908K)を超えたとき、炉心出力は自動的に低下した。炉心出力の自動低下後、ECCSである低圧注入(LPI)系から両低温側配管への冷却材注入により、全炉心はクエンチした。LPI系の作動を通じた継続的な炉心冷却を確認後、実験を終了した。本報告書は、ROSA/LSTFSB-PV-03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。
高松 邦吉; 舩谷 俊平*
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 11 Pages, 2024/11
本研究は、外的ハザードに対する安全性向上に向けて、水や空気等の流体の駆動に期待するのではなく、事故時の崩壊熱及び残留熱を受動的に冷却可能な、放射冷却を利用した新たな冷却システム(炉容器冷却システム)の開発を目的とする。本発表では、提案する炉容器冷却システムを原子炉建家と一体化した構造概念を提示するとともに、これまでの実験及び解析検討結果に基づく性能評価結果を報告する。
neutron diffraction study to elucidate hydrogen effect on the deformation mechanism in Type 310S austenitic stainless steel伊東 達矢; 小川 祐平*; Gong, W.; Mao, W.*; 川崎 卓郎; 岡田 和歩*; 柴田 曉伸*; Harjo, S.
Proceedings of the 7th International Symposium on Steel Science (ISSS 2024), p.237 - 240, 2024/11
Hydrogen embrittlement has long been an obstacle to the development of safe infrastructure. However, in contrast to hydrogen's embrittling effect, recent research has revealed that the addition of hydrogen improves both the strength and uniform elongation of AISI Type 310S austenitic stainless steel. A detailed understanding of how hydrogen affects the deformation mechanism of this steel could pave the way for the development of more advanced materials with superior properties. In the present study, neutron diffraction experiments were conducted on Type 310S steel with and without hydrogen-charged to investigate the effect of hydrogen on the deformation mechanism. In addition to the effect of solid-solution strengthening by hydrogen, the q-value, a parameter representing the proportion of edge and screw dislocations in the accumulated dislocations, was quantitatively evaluated using CMWP analysis on neutron diffraction patterns. The comparison of q-values between the hydrogen-charged and non-charged samples reveals that hydrogen has minimal effect on dislocation character in Type 310S steel.
田崎 雄大; 成川 隆文; 宇田川 豊
Journal of Nuclear Science and Technology, 61(10), p.1349 - 1359, 2024/10
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)This study developed a probabilistic determination model with respect to cladding high-temperature burst conditions based on the Bayesian statistical method to reasonably evaluate fuel behaviors under loss-of-coolant accident conditions, including fuel fragmentation, relocation, and dispersal. The candidate models were based on the widely accepted empirical model established based on nonirradiated fuel cladding data. Explanatory variables were added to improve the applicability of these models with respect to irradiated materials and generalization performance. The posterior predictive distribution of each candidate model was evaluated using Bayesian estimation comprising 238 sets of high-temperature burst test data. The generalization performance was evaluated using information criteria. The results of model evaluation showed improved predictive performance by considering the effect of hydrogen content. A comparison with burnup as an alternative explanatory variable confirmed that hydrogen content was the better parameter and other burnup-associated effects, such as irradiation hardening of the metal matrix and oxide growth (reduction of the metal matrix), were less dominant under burst conditions.
曽我部 丞司; 石田 真也; 田上 浩孝; 岡野 靖; 神山 健司; 小野田 雄一; 松場 賢一; 山野 秀将; 久保 重信; 久保田 龍三朗*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
日仏協力の枠組みにおいて、タンク型ナトリウム冷却高速を対象とした過酷事故の評価手法を定義し、解析評価を実施した。
