All Issue

2026 Vol.39, Issue 5 Preview Page

Research Paper

Three-Dimensional Seismic Analysis and Seismic Capacity Evaluation of a Safety-Related Steel Support Structure of PHWR

중수형 원전 안전관련 철골 지지구조물의 지진취약도 평가를 위한 3차원 내진해석 및 내진성능 평가

Hong-Pyo Lee1
,
Jae-Wook Jung2*
,
In-Kil Choi3
이 홍표1
,
정 재욱2*
,
최 인길3
1Principal Researcher, Central Research Institute, KHNP, Daejeon, 34101, Korea
2Senior Researcher, Structural and Seismic Safety Research Division, KAERI, Daejeon, 34057, Korea
3Principal Researcher, Structural and Seismic Safety Research Division, KAERI, Daejeon, 34057, Korea
1한국수력원자력 중앙연구원 책임연구원
2한국원자력연구원 구조･지진안전연구부 선임연구원
3한국원자력연구원 구조･지진안전연구부 책임연구원
*Corresponding Author
31 October 2026. pp. 301-309
PDF XML Citation
Abstract

In this study, a seismic fragility assessment was performed on a safety-related steel support structure installed in a representative pressurized heavy water reactor (PHWR) to collect the input data for a seismic probabilistic safety assessment (SPSA). The target structure supports the equipment and appurtenances associated with major safety-related piping and valve systems. Therefore, its load-transfer capacity through the upper steel frame, connections, and anchorage system must be maintained during an earthquake. For this purpose, a three-dimensional seismic analysis model consisting of the foundation, lower supporting structure, and upper steel frame was developed based on design documents and seismic walkdown observations. Dead loads, equipment loads, platform and walkway loads, cladding loads, and other relevant loads were considered in the analysis model. In addition, modal analysis was conducted to compare the dynamic characteristics of the developed model with those of the original design information. The seismic fragility assessment was conducted based on the Conservative Deterministic Failure Margin (CDFM) method presented in EPRI NP-6041-SL and EPRI 3002012994. The major evaluation components were classified into steel members, connections, and column anchorage systems. The results showed that the seismic fragility of the structure was governed by a local failure mode of the anchorage system, which serves as the load-transfer region between the upper steel frame and lower supporting structure, rather than by the strength of the steel members. In particular, concrete edge breakout in shear was identified as the governing failure mode. Nevertheless, even this failure mode, which exhibited the lowest seismic capacity among the evaluated failure modes, was confirmed to have sufficient seismic performance when compared with the design-basis requirement. The results of this study can be used to support the development of SPSA input data and to prioritize seismic performance improvement measures for structures in PHWR.

Keywords
PHWR
safety-related steel structure
seismic fragility
CDFM
HCLPF
seismic PSA

본 연구에서는 중수형 원전에 설치된 안전관련 철골 지지구조물을 대상으로 확률론적 지진 안전성평가(Seismic Probabilistic Safety Assessment, SPSA)의 입력자료 생산을 위한 지진취약도 평가를 수행하였다. 대상 구조물은 주요 안전관련 배관･밸브 계통과 관련된 기기 및 부속설비를 지지하는 구조물로서, 지진 시 상부 철골 구조, 접합부 및 정착부를 통한 하중전달 성능 확보가 요구된다. 이를 위해 설계문서와 내진 현장답사 결과를 반영하여 기초부, 하부 지지부 및 상부 철골 프레임으로 구성된 3차원 내진해석 모델을 구축하였다. 해석모델에는 구조물 자중, 기기하중, 플랫폼 및 보행로 하중, 외장재 하중 등을 반영하였으며, 고유치 해석을 통해 기존 설계자료와 동특성 비교를 수행하였다. 지진취약도 평가는 EPRI NP-6041-SL 및 EPRI 3002012994에서 제시하는 보수적 결정론적 파손여유도 방법(CDFM)을 기반으로 수행하였다. 주요 평가 대상은 철골 부재, 접합부 및 기둥 정착부로 구분하였다. 평가 결과 구조물의 지진취약도는 부재 자체의 강도보다는 상부 철골 구조와 하부 구조 사이의 하중전달부분인 정착부의 국부 파손모드에 의해 지배되는 것으로 나타났다. 특히 전단에 의한 콘크리트 연단 파열이 지배 파손모드로 확인되었으며, 가장 낮은 내진성능을 가지는 파손모드일지라도 설계기준과 비교하여 충분한 내진성능을 확보하고 있음을 확인하였다. 본 연구 결과는 향후 중수형 원전 구조물의 지진 PSA 입력자료 생산 및 내진성능 개선 우선순위 도출에 활용될 수 있을 것으로 판단된다.

키워드
중수형 원전
안전관련 철골 구조물
지진취약도
CDFM
HCLPF
지진 PSA
References
1

ACI (2013) Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, ACI 349-13, American Concrete Institute, Farmington Hills, MI.

2

AISC (2016) Specification for Structural Steel Buildings, ANSI/AISC 360-16, American Institute of Steel Construction, Chicago, IL.

3

ASCE/SEI (2016) Seismic Analysis of Safety-Related Nuclear Structures, ASCE/SEI 4-16, American Society of Civil Engineers, Reston, VA.

4

ASCE/SEI (2019) Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, ASCE/SEI 43-19, American Society of Civil Engineers, Reston, VA.

5

ASME/ANS (2009) Addenda to ASME/ANS RA-S-2008: Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009, American Society of Mechanical Engineers, New York, NY.

6

EPRI (1991) A Methodology for Assessment of Nuclear Plant Seismic Margin, EPRI NP-6041-SL, Revision 1, Electric Power Research Institute, Palo Alto, CA.

7

EPRI (1994) Methodology for Developing Seismic Fragilities, EPRI TR-103959, Electric Power Research Institute, Palo Alto, CA.

8

EPRI (2002) Seismic Fragility Application Guide, EPRI 1002988, Electric Power Research Institute, Palo Alto, CA.

9

EPRI (2009) Seismic Fragility Applications Guide Update, EPRI 1019200, Electric Power Research Institute, Palo Alto, CA.

10

EPRI (2018) Seismic Fragility and Seismic Margin Guidance for Seismic Probabilistic Risk Assessments, EPRI 3002012994, Electric Power Research Institute, Palo Alto, CA.

11

IAEA (2013) Periodic Safety Review for Nuclear Power Plants, IAEA Safety Standards Series No. SSG-25, International Atomic Energy Agency, Vienna.

12

Republic of Korea (2015) Nuclear Safety Act, Act No. 13389, Republic of Korea.

13

SQUG (1992) Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 2, Seismic Qualification Utility Group.

14

U.S. NRC (1975) Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, WASH-1400, U.S. Nuclear Regulatory Commission, Washington, DC.

15

U.S. NRC (1978) Development of Criteria for Seismic Review of Selected Nuclear Power Plants, NUREG/CR-0098, U.S. Nuclear Regulatory Commission, Washington, DC.

16

U.S. NRC (1991a) Generic Letter 88-20, Supplement 4: Individual Plant Examination of External Events for Severe Accident Vulnerabilities, U.S. Nuclear Regulatory Commission, Washington, DC.

17

U.S. NRC (1991b) Procedural and Submittal Guidance for the Individual Plant Examination of External Events for Severe Accident Vulnerabilities, NUREG-1407, U.S. Nuclear Regulatory Commission, Washington, DC.

18

U.S. NRC (2020) Regulatory Guide 1.200, Revision 3: Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, U.S. Nuclear Regulatory Commission, Washington, DC.

19

U.S. NRC (2024) Backgrounder on Probabilistic Risk Assessment, U.S. Nuclear Regulatory Commission, Washington, DC.

Information
  • Publisher :Computational Structural Engineering Institute of Korea
  • Publisher(Ko) :한국전산구조공학회
  • Journal Title :Journal of the Computational Structural Engineering Institute of Korea
  • Journal Title(Ko) :한국전산구조공학회 논문집
  • Volume : 39
  • No :5
  • Pages :301-309
  • Received Date : 2026-05-29
  • Revised Date : 2026-06-17
  • Accepted Date : 2026-07-01
  • DOI :https://doi.org/10.7734/COSEIK.2026.39.5.301
Journal Informaiton Journal of the Computational Structural Engineering Institute of Korea Journal of the Computational Structural Engineering Institute of Korea
Online Submission
  • scopus
  • NRF
  • KOFST
  • crossref crossmark
  • crossref cited-by
  • crosscheck
  • orcid
  • open access
  • ccl
Journal Informaiton Journal Informaiton - close