Analysis of fracture conditions of Cr-coated Zr alloy claddings under LOCA conditions calculated using FEMAXI fuel performance code

Luu, V. N.; 谷口 良徳; 宇田川 豊; 田崎 雄大; 勝山 仁哉

Annals of Nuclear Energy, 230, p.112114_1 - 112114_14, 2026/06

https://doi.org/10.1016/j.anucene.2026.112114

Fracture behavior of chromium (Cr) coated cladding under loss of coolant accident (LOCA) conditions was investigated utilizing the FEMAXI fuel performance code. Cr coating degradation models were introduced to FEMAXI to calculate oxygen diffusion behavior within the cladding tube. The FEMAXI code reasonably simulated the observed evolution of cladding metallic and oxide layers under the simulated LOCA conditions, accounting for factors such as wall thinning due to cladding high temperature creep, Cr layer thinning by CrO formation and Cr/Zr interdiffusion, weight increase by oxygen absorption, associated oxide growth, and increased oxygen concentration in -Zr phase. According to sensitivity analyses of the cladding oxygen concentration, where the effects of wall thickness change and eutectic reactions were taken into account, the fracture condition of the Cr-coated cladding samples can be reasonably modelled by the fracture criteria based on the remaining -Zr thickness with an oxygen concentration of 0.9 wt%.

High-temperature oxidation failure in reactivity-initiated accidents; An Evaluation of failure criteria based on oxygen concentration from the previous NSRR experiments

Luu, V. N.; 谷口 良徳; 宇田川 豊; 勝山 仁哉

Nuclear Engineering and Design, 442, p.114222_1 - 114222_15, 2025/10

https://doi.org/10.1016/j.nucengdes.2025.114222

For near-term application, coated-Zr alloy claddings show potential for enhancing safety by providing better oxidation resistance and minimizing hydrogen absorption under design-basis accidents (DBA). This benefit could extend the burnup and operational cycles of fuel rods. In assessing safety, reactivity-initiated accidents (RIA) are considered as one of the DBA conditions. The current safety criteria for high-temperature oxidation failure, one of the failure modes linked to RIA, are defined by peak fuel enthalpy values that range from 205 to 270 cal/g. This wide variability presents challenges when attempting to generalize criteria for modified-Zr alloy claddings with superior oxidation resistance. Therefore, it may be more relevant to apply failure criteria based on embrittlement mechanisms, such as oxygen concentration in the -Zr phase. This study aimed to assess the failure based on both peak fuel enthalpy and cladding embrittlement by analyzing previous NSRR experiments conducted with conventional materials using the RANNS fuel performance code. The findings suggest that the failure criteria associated with cladding embrittlement can provide a rational evaluation of failure behavior compared to the existing criterion based on peak fuel enthalpy. The local failure criterion leading to the formation of through-wall cracks during quenching is consistent with Chung's proposal (NUREG/CR-1344): -Zr thickness of 0.9 wt% oxygen is less than 0.1 mm, and this corresponds to approximately 35% BJ-ECR.

Effect of rod internal gas state on FFRD behavior of high burnup fuel during LOCA conditions

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.

Improvement of PFM analysis code PASCAL-SP2 for evaluating the failure probability of stainless steel pipe with a stress corrosion crack in PWRs

山口 義仁; Li, S.; 勝山 仁哉

Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 8 Pages, 2025/07

https://doi.org/10.1115/PVP2025-154778

Recently, in Japan and France, the intergranular stress corrosion cracking (SCC) has been detected in welded joints of stainless steel pipes in primary piping systems in Pressurized Water Reactors (PWRs). It was observed that hardness in the welded joints with SCC was higher than that of the base metal. It is shown that hardness is an important parameter in the SCC propagation in the PWR primary system water environment. In our previous studies, a numerical method was proposed for the prediction of hardness and weld residual stress in Type 316 stainless steel welded joints. The proposed numerical method was validated through the comparisons between predicted results and measurements for welded stainless steel specimens. And then, the effect of welding condition on hardness and residual stresses in stainless steel pipe welds was investigated by using above method. In this study, probabilistic fracture mechanics analysis code PASCAL-SP Ver.2 developed by Japan Atomic Energy Agency is improved to be able to evaluate the failure probability of stainless steel pipe with SCC in PWRs by considering the effect of hardness, and the verification is carried out for an improved function. Additionally, failure probabilities are calculated considering crack growth due to SCC and fatigue in PWR primary water environment. The effect of welding condition such as heat input at welding is quantitatively evaluated based on the failure probabilities.

Effects of welding and constraint conditions on the welding residual stress and hardness of Type 316 stainless steel pipe

Li, S.; 山口 義仁; 勝山 仁哉; Li, Y.

Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 8 Pages, 2024/07

https://doi.org/10.1115/PVP2024-122731

Flaws due to stress corrosion cracking (SCC) were recently detected in welded joints of austenitic stainless steel piping lines in pressurized water reactors. Welding-induced high hardness and tensile residual stress are known as one of the main factors affecting SCC. In this work, thermal-elastic-plastic coupled three-dimensional finite element analyses were performed to investigate the distributions of welding residual stress and hardness in butt-welded joints of Type 316 stainless steel. Different heat inputs were applied to the pipe welds, including normal heat input, high heat input and very high heat input. Two different constraint conditions were considered for the welded joints, i.e., both ends free or clamped, the latter indicating that the welded joint is constrained by the surrounding piping system. Simulation results were compared with experimental data such as welding thermal cycle, axial shrinkage and residual stress for validation. The effects of heat input and constraint condition on the welding residual stress and hardness at different sections of the welded joints including the weld start/end location were discussed in detail.

Fracture toughness evaluation of weld-HAZ in RPV steel using Mini-C(T) specimens

河 侑成; 下平 昌樹; 勝山 仁哉

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 7 Pages, 2024/03

国内の原子炉圧力容器(RPV)の溶接継手熱影響部(継手HAZ)は監視試験対象部位である。継手HAZ内には非均質に金属組織が分布し、破壊靭性のばらつきが大きいと考えられるものの、HAZ試験片の採取位置における金属組織の違いが機械的特性及び照射脆化感受性に及ぼす影響は詳細に調べられていなかった。本研究では継手HAZの非均質な組織に着目し、まず未照射材について溶接金属と母材の境界からの距離に応じて3箇所(0.5mm, 1mm及び2mm)の継手HAZ、及び母材から採取した微小試験片(Mini-C(T)試験片)を用いて破壊靭性を評価した。その結果、HAZの破壊靭性は母材と比べてばらつきが大きく、特に溶接金属と母材の境界から0.5mmのHAZにおいて破壊靭性のばらつきが最も大きい傾向を示した。HAZに対する破壊靭性参照温度は採取した3箇所の中で有意な差が得られたものの、いずれの破壊靭性参照温度も母材より低く、破壊靭性値が高いことを確認した。本研究結果により、未照射状態では母材が継手HAZを代表して破壊靭性を保守的に評価できることを確認した。

Modeling of hardness and welding residual stress in Type 316 stainless steel components for the assessment of stress corrosion cracking

Li, S.; 山口 義仁; 勝山 仁哉; Li, Y.; Deng, D.*

Proceedings of the ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 7 Pages, 2023/07

https://doi.org/10.1115/PVP2023-103449

In this work, a framework was proposed on the comprehensive assessment of hardness and welding residual stress in Type 316 austenitic stainless steel welded joints. Firstly, an 8-pass butt-welded joint made of Type 316 stainless steel was fabricated. Finite element analysis of the welded joint was performed to investigate hardness and welding residual stress distributions. The grain growth model was developed for the hardness prediction. The Chaboche combined isotropic-kinematic strain hardening model and time-temperature dependent annealing model were adopted. The relationships between the Vickers hardness and the uniaxial plastic strain as well as grain size were collected from published literatures. The simulation results of the grain size and accumulated equivalent plastic strain were used for the hardness prediction of the welded joint. The predicted hardness was compared with the experimental data of hardness mapping. The distribution of welding residual stress on the outer surface of the welded pipe was measured by using the X-ray diffraction method and strain gauge method, respectively. The predicted welding residual stresses were compared with the measurements. The results obtained showed that the developed numerical approach can predict the hardness and welding residual stress of Type 316 stainless steel welded joints with satisfactory accuracy. The effects of structural constraint and heat input on the hardness and welding residual stress will be investigated as further works, as described in the proposed framework.

Fracture toughness evaluation of the heat-affected zone under the weld overlay cladding in reactor pressure vessel steel

河 侑成; 飛田 徹; 高見澤 悠; 勝山 仁哉

Journal of Pressure Vessel Technology, 145(2), p.021501_1 - 021501_9, 2023/04

https://doi.org/10.1115/1.4055626

The fracture toughness in the heat-affected zone (HAZ), which is located under the weld overlay cladding on the inner surface of the reactor pressure vessel (RPV), was evaluated by considering the inhomogeneous microstructures of the HAZ. In this regard, 0.4T-C(T) specimens and Mini-C(T) specimens were manufactured to clarify the effect of the HAZ on fracture toughness. First, the fracture toughness was evaluated using 0.4T-C(T) specimens targeted on microstructural distribution due to the cladding bead location. The reference temperature () of HAZ at middle region of cladding bead was 12C higher than that at overlap cladding beads. Second, the fracture toughness was evaluated using Mini-C(T) specimens focused on the microstructural difference with distance from the boundary between the cladding and the base metal. The value of the central position located at 3 mm to 7 mm from the boundary between the cladding and the base metal was higher than those of the other positions in HAZ. However, the 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 40C60C. Compared with the literature data that indicated fracture toughness at the inner surface without overlay cladding and the base metal at a quarter thickness, this study concluded that the HAZ under the overlay cladding showed no detrimental effect on the conservatism of the structural integrity assessment of RPV steel.

原子炉圧力容器用確率論的破壊力学解析コードPASCAL5の使用手引き及び解析手法

高見澤 悠; Lu, K.; 勝山 仁哉; 眞崎 浩一*; 宮本 裕平*; Li, Y.

JAEA-Data/Code 2022-006, 221 Pages, 2023/02

JAEA-Data-Code-2022-006.pdf:4.79MB

原子炉圧力容器(RPV: Reactor Pressure Vessel)は原子炉冷却材圧力バウンダリを構成する重要機器の1つであり、中性子照射等に伴う高経年化を考慮した構造健全性確保が極めて重要である。国立研究開発法人日本原子力研究開発機構(JAEA: Japan Atomic Energy Agency)では、RPVの構造健全性評価に関する研究の一環として、確率論的破壊力学(PFM: Probabilistic Fracture Mechanics)解析コードPASCAL(PFM Analysis of Structural Components in Aging LWR)の開発を進めている。本コードは、加圧水型軽水炉(PWR: Pressurized Water Reactor)及び沸騰水型軽水炉(BWR: Boiling Water Reactor)を対象に、影響因子が持つ不確実さを考慮し、加圧熱衝撃(PTS: Pressurized Thermal Shock)事象や低温過圧事象(LTOP: Low-Temperature Over Pressure)等の過渡によるRPVの炉心領域部の破損確率や破損頻度を求めるものである。破壊力学や確率論的計算手法等に関する最新知見や国内RPVに適した評価手法・評価モデルを踏まえ、新規解析機能の導入を進めるとともに、系統的なコード検証活動を通じて信頼性向上を図ってきた。平成12年度に公開したPASCALでは、PWRのPTS事象を対象に、RPVの破損確率を解析する基本的な枠組みを整備した。平成18年度に公開したPASCAL2では、内部亀裂の評価手法や様々な非破壊検査による亀裂の検出性に関する評価モデル等を導入し、過渡事象データベースを整備した。平成22年度に公開したPASCAL3では、肉盛溶接クラッド部に着目して、亀裂の評価機能等を改良した。平成29年度に公開したPASCAL4では、応力拡大係数解や破壊靭性の不確実さを考慮した評価モデル等の改良により解析機能の高度化を図るとともに、影響因子の不確実さを認識論的不確実さと偶然的不確実さに分類し、不確実さを考慮した信頼度評価機能等を整備した。平成30年度以降は、これまでPWRのPTS事象を対象としたRPV内面側亀裂の評価機能に加えて、BWRの起動事象、LTOP事象等を想定したRPV外面側亀裂の評価機能等の整備を進めてきた。これらの機能整備を踏まえ、国内PWR及びBWRのRPVを対象とした確率論的健全性評価に資する解析コードとして、PASCAL5へとバージョンアップした。PASCAL5はPFM解析モジュールであるPASCAL-RV、PASCAL-RVの入力データの生成やRPV炉心領域部を対象とした破損頻度の算出を行うモジュールであるPASCAL-Manager、付録として附属する簡易的な熱応力解析を

Improvement of the return mapping algorithm based on the implicit function theorem with application to ductile fracture analysis using the GTN model

真野 晃宏; 今井 隆太*; 宮本 裕平*; Lu, K.; 勝山 仁哉; Li, Y.

International Journal of Pressure Vessels and Piping, 199, p.104700_1 - 104700_13, 2022/10

https://doi.org/10.1016/j.ijpvp.2022.104700

有限要素法に基づく弾塑性解析は、金属材料の非線形挙動の再現に広く用いられる。この弾塑性解析を陰解法により実施する場合、応力値は通常リターンマッピングアルゴリズムを通じて時間増分ごとに更新される。一方、延性破壊等の複雑な挙動の解析では、リターンマッピングアルゴリズムに係る連立方程式の数が多くなり、収束解を得ることが困難となる場合がある。そのため、本論文では、連立方程式の数の削減を通じて陰解法による安定的な非線形解析を実現することを目的として、陰関数定理に基づき連立方程式の数を削減する一般的な手法を提案した。また、提案手法を、延性破壊解析に用いられるGurson-Tvergaard-Needleman(GTN)モデルに適用し、弾塑性解析を実施した。その結果、解析により、延性破壊に係る実験結果が再現できることを確認した。以上を踏まえ、提案手法は、金属材料の延性破壊解析に有用であると結論付けた。