Standard guideline for the seismic response analysis method using three-dimensional finite element model of reactor buildings (Contract research) (Translated document)

Choi, B.; Nishida, Akemi; Kawata, Manabu; Shiomi, Tadahiko; Li, Y.

JAEA-Research 2024-001, 206 Pages, 2024/03

JAEA-Research-2024-001.pdf:9.12MB

In the assessment of seismic safety and the design of building structures in nuclear facilities, lumped mass models have been used as standard methods. Recent advances in computer capabilities allow the use of three-dimensional finite element (3D FE) models to account for the 3D behavior of buildings, material nonlinearity, and the nonlinear soil-structure interaction effect. While 3D analysis method has many advantages, it is necessary to ensure its reliability as a new approach. The International Atomic Energy Agency performed an international benchmark study using the 3D FE analysis model for reactor building of Unit 7 at TEPCO's Kashiwazaki-Kariwa Nuclear Power Station based on recordings from the Niigataken Chuetsu-oki Earthquake in 2007. Multiple organizations from different countries participated in this study and the variation in their analytical results was significant, indicating an urgent need to improve the reliability of the analytical results by standardization of the analytical methods using 3D FE models. Additionally, it has been pointed out that it is necessary to understand the 3D behavior in the seismic fragility assessment of buildings and equipment, using realistic seismic response analysis method based on 3D FE models. In view of these considerations, a guideline for the seismic response analysis method using a 3D FE model was developed by incorporating the latest knowledge and findings in this area. The purpose of the guideline is to improve the reliability of the seismic response analysis method using 3D FE model of reactor buildings. The guideline consists of a main body, commentaries, and appendixes. The standard procedures, recommendations, key points to note, and technological bases for conducting seismic response analysis on reactor buildings using 3D FE models are provided in the guideline. In addition, the guideline will be revised reflecting the latest knowledge.

Numerical analyses on perforation damage using test results of reinforced concrete panel subjected to oblique impact

Kang, Z.; Okuda, Yukihiko; Nishida, Akemi; Tsubota, Haruji; Li, Y.

Transactions of 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03

Most studies conducted till now on local damage of reinforced concrete (RC) slab structures subjected to projectile impact are about normal impact, while few research related to oblique impact can be found. The objective of this study is to carry out impact tests under different impact conditions including oblique impacts, to confirm the different impact behaviors of the RC slab structure, to develop an analysis method by investigating the test results and analytical conditions, and to validate the analysis method through comparison with the test results. This study focuses on the scabbing damage which is one of the local damage modes of RC slab. Based on oblique impact test results due to soft projectile with hemispherical nose shape, we investigate the relationship between the criterion related to the concrete fracture and the occurrence of scabbing damage.

Failure probability evaluation for steam generator tubes with wall-thinning

Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.

Transactions of 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03

The steam generator (SG) is an important component of a pressurized water reactor. In addition, local wall-thinning has been reported in SG tubes. The burst differential pressure, considering both the internal and external pressures from the primary and secondary coolant systems, should be predicted for the failure probability evaluation or structural integrity assessment of SG tubes. In this study, based on the results of burst tests performed in Japan and the United States, we improved the existing burst pressure estimation method for SG tubes with wall-thinning. In addition, as an example of the utilization of the improved burst pressure estimation method, the conditional failure probabilities for SG tubes with local wall-thinning, which is necessary for probabilistic risk assessment and risk-informed decision making, are calculated considering the dimensions of the wall-thinning.

Microscopic origin of the spin-reorientation transition in the kagome topological magnet TbMnSn

Huang, Z.*; Wang, W.*; Ye, H.*; Bao, S.*; Shangguan, Y.*; Liao, J.*; Cao, S.*; Kajimoto, Ryoichi; Ikeuchi, Kazuhiko*; Deng, G.*; et al.

Physical Review B, 109(1), p.014434_1 - 014434_9, 2024/01

https://doi.org/10.1103/PhysRevB.109.014434

Investigation on damage evaluation index with ductility factor based on simulation analysis for loading test of piping support structure

Okuda, Yukihiko; Takito, Kiyotaka; Nishida, Akemi; Li, Y.

Mechanical Engineering Journal (Internet), 12 Pages, 2024/00

https://doi.org/10.1299/mej.23-00405

After the Great East Japan earthquake and the accident at the TEPCO's Fukushima Daiichi Nuclear Power Stations in March 2011, the regulation for nuclear power plants (NPPs) has been enhanced to take countermeasures against beyond-design-basis events. To improve the seismic safety of nuclear facilities against earthquakes that exceed the design input ground motion, the importance of seismic probabilistic risk assessment (PRA) has drawn much attention. It is essential to evaluate the realistic seismic response of the equipment and piping in NPPs for fragility assessment in seismic PRA. In particular, since piping systems have plant-specific complex route geometries, it is known that the arrangement and stiffness of piping support structures have a significant impact on seismic response characteristics of the entire piping system. To construct a realistic seismic response analysis method for excessive input ground motion exceeding the elastic response, it is desired to develop an elastic-plastic response analysis method that can estimate the realistic response of piping systems including pipe support structures. In this study, the applicability of the method is confirmed by the simulation analysis of the elasto-plastic response for the piping support structure loading test previously reported. Moreover, based on the good correlation between the ductility factor and the damage status obtained from the test results and simulation analysis results, it is shown that the ductility factor is effective as a damage evaluation index for piping support structures.

Chiral Dirac fermion in a collinear antiferromagnet

Zhang, A.*; Deng, K.*; Sheng, J.*; Liu, P.*; Kumar, S.*; Shimada, Kenya*; Jiang, Z.*; Liu, Z.*; Shen, D.*; Li, J.*; et al.

Chinese Physics Letters, 40(12), p.126101_1 - 126101_8, 2023/12

https://doi.org/10.1088/0256-307X/40/12/126101

PANDORA Project for the study of photonuclear reactions below

Tamii, Atsushi*; Pellegri, L.*; Sderstrm, P.-A.*; Allard, D.*; Goriely, S.*; Inakura, Tsunenori*; Khan, E.*; Kido, Eiji*; Kimura, Masaaki*; Litvinova, E.*; et al.

European Physical Journal A, 59(9), p.208_1 - 208_21, 2023/09

https://doi.org/10.1140/epja/s10050-023-01081-w

no abstracts in English

Characteristics of allowable axial cracks for pressurized pipes governed by limit load criteria

Hasegawa, Kunio; Li, Y.; Udyawar, A.*; Lacroix, V.*

International Journal of Pressure Vessels and Piping, 204, p.104952_1 - 104952_7, 2023/08

https://doi.org/10.1016/j.ijpvp.2023.104952

When axial cracks were detected in pipes, failure stresses for high toughness pipes are estimated using the Limit Load Criteria. The allowable stresses for the cracked pipes are derived from the combination of the failure stresses and safety factors. The allowable sizes of crack depths and lengths are determined from the allowable stresses. From the comparison of the allowable and failure stresses for through-wall cracks, the allowable cracks are not uniform. They can be separated into three different characteristics, i) leak-before-break (LBB) and crack growth stability, ii) non-LBB and crack growth stability and iii) non-LBB and crack growth instability. Inspectors and users should pay special attention to allowable cracks with the third characteristic to prevent unexpected failure, particularly for thin-wall pipes. The allowable crack depths and lengths that require special attention can be expressed by appropriate equations.

Verification of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessel

Lu, K.; Takamizawa, Hisashi; Li, Y.; Masaki, Koichi*; Takagoshi, Daiki*; Nagai, Masaki*; Nannichi, Takashi*; Murakami, Kenta*; Kanto, Yasuhiro*; Yashirodai, Kenji*; et al.

Mechanical Engineering Journal (Internet), 10(4), p.22-00484_1 - 22-00484_13, 2023/08

https://doi.org/10.1299/mej.22-00484

Level structures of Ca cast doubt on a doubly magic Ca

Chen, S.*; Browne, F.*; Doornenbal, P.*; Lee, J.*; Obertelli, A.*; Tsunoda, Yusuke*; Otsuka, Takaharu*; Chazono, Yoshiki*; Hagen, G.*; Holt, J. D.*; et al.

Physics Letters B, 843, p.138025_1 - 138025_7, 2023/08

https://doi.org/10.1016/j.physletb.2023.138025

Gamma decays were observed in Ca and Ca following quasi-free one-proton knockout reactions from Sc. For Ca, a ray transition was measured to be 1456(12) keV, while for Ca an indication for a transition was observed at 1115(34) keV. Both transitions were tentatively assigned as the decays. A shell-model calculation in a wide model space with a marginally modified effective nucleon-nucleon interaction depicts excellent agreement with experiment for level energies, two-neutron separation energies, and reaction cross sections, corroborating the formation of a new nuclear shell above the N = 34 shell. Its constituents, the and orbitals, are almost degenerate. This degeneracy precludes the possibility for a doubly magic Ca and potentially drives the dripline of Ca isotopes to Ca or even beyond.