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.

Construction of large-scale observation system for improvement of three-dimensional seismic analysis method for nuclear buildings

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

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 10 Pages, 2023/04

Our research and development are aimed at improving the accuracy of the three-dimensional seismic evaluation analysis method for nuclear buildings, which contributes to the probabilistic risk assessment caused by earthquakes (seismic PRA), using actual seismic observation records in collaboration with the Nuclear Regulation Authority since 2019. In this study, we constructed a large-scale observation system that enabled simultaneous measurements at multiple positions during earthquakes or against artificial waves. The accelerometers of the observation system were installed on/in the soil, floors, and on the walls of the nuclear building. This paper presents an outline of the large-scale observation system and the findings obtained from the analysis of the seismic observation records acquired using the system.

Analytical study for low ground contact ratio of buildings due to the basemat uplift using a three-dimensional finite element model

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

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08

In the seismic evaluation of nuclear facility buildings, basemat uplift-the phenomenon during which the bottom of the basemat of a building partially rises from the ground owing to overturning moments during earthquakes-is a very important aspect because it affects not only structural strength and integrity, but also the response of equipment installed in the building. However, there are not enough analytical studies on the behavior of buildings with a low ground contact ratio due to basemat uplift during earthquakes. In this study, we conducted a simulation using a three-dimensional finite element model from past experiments on basemat uplift; further, we confirmed the validity of this approach. In order to confirm the difference in the analytical results depending on the analysis code, the simulation was performed under the same analytical conditions using the three analysis codes, which are E-FrontISTR, FINAS/STAR and TDAPIII, and the obtained analysis results were compared. Accordingly, we investigated the influence of the difference in adhesion on the structural response at low ground contact ratio. In addition, we confirmed the effects of significant analysis parameters on the structural response via sensitivity analysis. In this paper, we report the analytical results and insights obtained from these investigations.

A Study on the improvement of accuracy of three-dimensional seismic evaluation analysis method for nuclear buildings using a large-scale observation system

Nishida, Akemi; Kawata, Manabu; Choi, B.; Iigaki, Kazuhiko; Li, Y.

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 10 Pages, 2022/07

We have conducted research and development with the aim of improving the accuracy of three-dimensional seismic evaluation analysis method for nuclear buildings that contributes to probabilistic risk assessment caused by earthquakes (seismic PRA). In 2019, we started our research on improving the accuracy and validating the three-dimensional seismic analysis method used for nuclear buildings using actual seismic observation records in collaboration with the Nuclear Regulation Authority. In this research, we constructed a large-scale observation system that enabled simultaneous observation at multiple positions during natural earthquakes or artificial waves by installing accelerometers not only on/in the soil and on the floors of the building but also on the walls of the building, targeting the High Temperature engineering Test Reactor, which is one of nuclear facilities of JAEA. In this paper, we report the outline of the large-scale observation system and the knowledge obtained from the analysis results of the seismic observation records acquired using this system.

Assessment of seismic fragility using a three-dimensional structural model of a reactor building

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

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 10 Pages, 2021/08

https://doi.org/10.1115/ICONE28-64300

The new regulatory requirements in Japan have strengthened the mitigation of damage caused by natural disasters, such as earthquakes, and the operational guide for safety improvement evaluation recommends the probabilistic risk assessment (PRA) as the evaluation method in Japan. In the PRA of an earthquake, also known as the seismic PRA, the realistic assessment of the structural seismic response and the damage probability (fragility) assessment using the realistic response assessment of the nuclear buildings and equipment is one of the most important issues. Accordingly, the authors have conducted this study on the realistic seismic response analysis methods and seismic fragility assessment methods to ensure the seismic safety of the nuclear buildings and equipment. In this study, a nonlinear seismic response analysis is conducted for input ground motions beyond the ground motions assumed in the design by using a three-dimensional (3D) structural model of a reactor building. In addition, the damage mode of the structural components of the reactor building associated with the equipment is identified, and the seismic fragility is assessed based on the 3D behavior of the reactor building. The local response and detailed damage process of the reactor building that have been obtained through seismic response analysis, are reported in this study, along with the results of the seismic fragility assessment.

The Working group on the analysis and management of accidents (WGAMA); A Historical review of major contributions

Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo

Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09

https://doi.org/10.1016/j.pnucene.2020.103432

Computation speeds and memory requirements of mesh-type ICRP reference computational phantoms in Geant4, MCNP6, and PHITS

Yeom, Y. S.*; Han, M. C.*; Choi, C.*; Han, H.*; Shin, B.*; Furuta, Takuya; Kim, C. H.*

Health Physics, 116(5), p.664 - 676, 2019/05

https://doi.org/10.1097/HP.0000000000000999

Recently, Task Group 103 of the ICRP developed the mesh-type reference computational phantoms (MCRPs), which are planned for use in future ICRP dose coefficient calculation. Performance of major Monte Carlo particle transport codes (Geant4, MCNP6, and PHITS) were tested with MCRP. External and internal exposure of various particles and energies were calculated and the computational times and required memories were compared. Additionally calculation for voxel-mesh phantom was also conducted so that the influence of different mesh-representation in each code was studied. Memory usage of MRCP was as large as 10 GB with Geant4 and MCNP6 while it is much less with PHITS (1.2 GB). In addition, the computational time required for MRCP tends to increase compared to voxel-mesh phantoms with Geant4 and MCNP6 while it is equal or tends to decrease with PHITS.

Soft, skin-interfaced microfluidic systems with wireless, battery-free electronics for digital, real-time tracking of sweat loss and electrolyte composition

Kim, S. B.*; Lee, K.-H.*; Raj, M. S.*; Reeder, J. T.*; Koo, J.*; Hourlier-Fargette, A.*; Bandodkar, A. J.*; Won, S. M.*; Sekine, Yurina; Choi, J.*; et al.

Small, 14(45), p.1802876_1 - 1802876_9, 2018/11

https://doi.org/10.1002/smll.201802876

Excretion of sweat from eccrine glands is a dynamic physiological process that varies with body position, activity level, and health status. Information content embodied in sweat rate and chemistry can be used to assess health status and athletic performance. This paper presents a thin, miniaturized, skin-interfaced microfluidic technology that includes a reusable, battery-free electronics module for measuring sweat conductivity and rate in real-time using wireless power from and data communication with capabilities in near field communications (NFC). Systematic studies of these combined microfluidic/electronic systems, accurate correlations of measurements performed with them to those of laboratory standard instrumentation, and field tests on human subjects establish the key operational features and their utility in sweat analytics.

Epistemic Uncertainty Quantification of Floor Responses for a Nuclear Reactor Building

Choi, B.; Nishida, Akemi; Li, Y.; Muramatsu, Ken*; Takada, Tsuyoshi*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

After the 2011 Fukushima accident, nuclear power plants are required to take countermeasures against accidents beyond design basis conditions. In seismic probabilistic risk assessment (SPRA), uncertainty can be classified as either aleatory uncertainty, which cannot be reduced, or epistemic uncertainty, which can be reduced with additional knowledge and/or information. To improve the reliability of SPRA, efforts should be made to identify and reduce the epistemic uncertainty caused by the lack of knowledge. In this study, we focused on the difference in seismic response by modeling methods, which is related epistemic uncertainty. We conducted a seismic response analysis with two kinds of modeling methods; a three-dimensional finite-element model and a conventional sway-rocking stick model, by using simulated various input ground motions, which is related to aleatory uncertainty. And then we quantified the seismic floor response results of the various input ground motions of each modeling methods. For the uncertainty quantification related to different modeling methods, we further perform a statistical analysis of the floor response results of the nuclear reactor building. Finally, we discussed how to utilize the results from these calculations for the quantification of uncertainty in fragility analysis for SPRA.

EBR-II passive safety demonstration tests benchmark analyses; Phase 2

Briggs, L.*; Monti, S.*; Hu, W.*; Sui, D.*; Su, G. H.*; Maas, L.*; Vezzoni, B.*; Partha Sarathy, U.*; Del Nevo, A.*; Petruzzi, A.*; et al.

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.3030 - 3043, 2015/08

The International Atomic Energy Agency Coordinated Research Project, "Benchmark Analyses of an EBR-II Shutdown Heat Removal Test" is in the third year of its four-year term. Nineteen participants representing eleven countries have simulated two of the most severe transients performed during the Shutdown Heat Removal Tests program conducted at Argonne's Experimental Breeder Reactor II. Benchmark specifications were created for these two transients, enabling project participants to develop computer models of the core and primary heat transport system, and simulate both transients. In phase 1 of the project, blind simulations were performed and then evaluated against recorded data. During phase 2, participants have refined their models to address areas where the phase 1 simulations did not predict as well as desired the experimental data. This paper describes the progress that has been made to date in phase 2 in improving on the earlier simulations and presents the direction of planned work for the remainder of the project.