Fire Safety Assessment of Accelerated-Aged Nuclear Power Plant Cables: Application of Heat Growth Index and Regression Analysis Techniques

doi:10.15683/kosdi.2025.9.30.536

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2025 Vol.21, Issue 3 Preview Page Next Page

Original Article

Fire Safety Assessment of Accelerated-Aged Nuclear Power Plant Cables: Application of Heat Growth Index and Regression Analysis Techniques 가속 열화된 원전 케이블의 화재 안전성 평가: 열성장지수와 회귀분석 기법의 적용: 박계원¹ㆍ곽지현²*ㆍ최두찬³ㆍ양현혁⁴
Kye-Won Park¹, Jihyun Kwark²*, Doo-Chan Choi³, Hyeon-Hyeok Yang⁴; ¹Principal Researcher, Fire Insurers Laboratories of Korea, Yeoju, Republic of Korea
²Principal Researcher, Fire Insurers Laboratories of Korea, Yeoju, Republic of Korea
³Chief Executive Officer, KF UBIS Co., Ltd., Seoul, Republic of Korea
⁴Researcher, KF UBIS Co., Ltd., Seoul, Republic of Korea

30 September 2025. pp. 536-544

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Abstract

Purpose: This study aims to establish a database of heat release rates of electrical cabinets that can be utilized for severity estimation for NPP fire risk assessment. Methods: For this purpose, fire scenarios were derived based on the specificity of the cabinet shape (volume, opening, combustibles, etc.), and for the design of the electrical cabinet fire test, a cone calorimeter test for combustibles selection and a cable vertical fire spread test for fire source calorific value selection were performed. Results: Based on the developed test method and selected fire scenarios, a Heat Release Rate (HRR) database for domestic nuclear power plant electrical cabinets was established, and its reliability was verified through FDS. Conclusion: The established electrical cabinet heat release rate DBs are applied as severity estimation factors for NPP fire risk assessment. In the future, a database covering all electrical cabinet fire scenarios is planned, and a follow-up study will be conducted to develop a technique to quantitatively estimate the fire severity of domestic electrical cabinets. These studies are expected to contribute to improving the accuracy and reliability of fire safety assessments for nuclear power plants.

Keywords

원전 케이블

화재 안전성

가속 열화

아레니우스 모델

IEEE 383

난연성

열성장지수

다중회귀분석

연구목적: IEEE 383과 같은 원자력발전소(원전) 케이블의 표준 화재 안전성 시험은 신품을 대상으로 하므로, 장기 운영에 따른 열화가 난연성능에 미치는 영향에 대한 지식 공백이 존재한다. 본 연구는 가속 열화를 통해 장기 사용된 원전 케이블의 화재 안전성능 변화를 정량적으로 분석하고, 이를 평가할 새로운 지표를 개발하는 것을 목적으로 한다. 연구방법: 안전등급 케이블 2종(A: XLPE, B: EPR/CSPE)에 대해 아레니우스 모델을 적용하여 최대 40년 등가 수명에 해당하는 가속 열화를 수행하였다. 이후 IEEE 383 수직 화염 시험으로 난연성능을 평가하였으며, 화재 거동을 정량화하기 위해 화원으로부터 특정 거리에서의 최대 온도상승률(°C/s)로 정의되는 새로운 열성장지수(TEGRA)를 제안하였다. 연구 결과: 가속 열화는 케이블의 난연성능에 유의미한 영향을 미쳤으며, 그 영향은 케이블 종류에 따라 다르게 나타났다. 제안된 열성장지수는 평균 탄화 길이를 예측하는 유의한 변수임이 확인되었다. A 케이블의 회귀 모델은 가속열화온도에 의해 높은 설명력(R2>0.6)을 보인 반면, B 케이블 모델의 설명력은 상대적으로 낮아(R2=0.313) 후자의 열화 거동이 더 복잡함을 시사했다. 결론: 본 연구는 기존 화재 안전성 표준이 경년 열화된 원전 케이블의 위험성을 충분히 반영하지 못할 수 있음을 실험적으로 입증하였다. 제안된 열성장지수는 성능 기반의 화재 안전성 평가를 위한 새로운 정량적 지표로서의 가능성을 제시하며, 원전의 경년 열화 케이블에 대한 보다 현실적인 화재 안전성 평가 기준 수립과 상태 기반 유지보수(CBM) 전략 개발의 기초 자료로 활용될 수 있을 것이다.

키워드

Nuclear Power Plant

Electric Cabinet Fire

Heat Release Rate

Probabilistic Safety Assessment(PSA)

Severity Factor

References

Arrhenius, S. (1889). “Über die Reaktionsgeschwindigkeit bei der Inversion von Rohrzucker durch Säuren.” Zeitschrift für physikalische Chemie, Vol. 4, No. 1, pp. 226-248. 10.1515/zpch-1889-0416
Frost, J. (2023). Interpreting Regression Output with Low R-squared Values. Statistics by Jim
IAEA (2005). Application of condition based maintenance in nuclear power plants. IAEA-TECDOC-1451.
IEC (2011). Test on Gases Evolved during Combustion of Materials from Cables - Part 1: Determination of the halogen acid gas content. IEC 60754-1.
IEEE (1974). IEEE Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations. IEEE Std 383-1974.
ISO (2019). Plastics - Smoke generation - Part 2: Determination of optical density by a single-chamber test. ISO 5659-2.
Kim, D.-C., Kwag, S.-Y., Kim, J.-T., Eem, S.-H. (2022). “Research trends on external event identification and screening methods for safety assessment of nuclear power plant.” Journal of the Korea Society of Disaster Information, Vol. 18, No. 2, pp. 252-260.
Kim, K.-M., Kim, S.-I., Son, E.-S., Ko, M.-H., Yang, H.-H., Choi, D.C. (2025) “A study on the analysis of key factors for reducing fire CDF in nuclear power plants.” Journal of the Korea Society of Disaster Information, Vol. 21, No. 2, pp. 342-347.
Laerd Statistics (2024). Independent t-test for Two Samples.
Salley, D. (2004). Browns Ferry Nuclear Plant, Unit 1, Fire of March 22, 1975. NUREG/IA-0204.
Seguchi, T. (2013). “Degradation mechanisms of cable insulation materials during radiation-thermal ageing in radiation environment.” Polymer Degradation and Stability, Vol. 98, No. 1, pp. 263-270.
U.S. Nuclear Regulatory Commission (2016). Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE-FIRE) Vol.. NUREG-2178.
U.S. Nuclear Regulatory Commission (2020). Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants. Regulatory Guide 1.89.

Information

Publisher :The Korean Society of Disaster Information
Publisher(Ko) :한국재난정보학회
Journal Title :Journal of the Society of Disaster Information
Journal Title(Ko) :한국재난정보학회논문집
Volume : 21
No :3
Pages :536-544
DOI :https://doi.org/10.15683/kosdi.2025.9.30.536

[1] Arrhenius, S. (1889). “Über die Reaktionsgeschwindigkeit bei der Inversion von Rohrzucker durch Säuren.” Zeitschrift für physikalische Chemie, Vol. 4, No. 1, pp. 226-248. 10.1515/zpch-1889-0416

[2] Frost, J. (2023). Interpreting Regression Output with Low R-squared Values. Statistics by Jim

[3] IAEA (2005). Application of condition based maintenance in nuclear power plants. IAEA-TECDOC-1451.

[4] IEC (2011). Test on Gases Evolved during Combustion of Materials from Cables - Part 1: Determination of the halogen acid gas content. IEC 60754-1.

[5] IEEE (1974). IEEE Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations. IEEE Std 383-1974.

[6] ISO (2019). Plastics - Smoke generation - Part 2: Determination of optical density by a single-chamber test. ISO 5659-2.

[7] Kim, D.-C., Kwag, S.-Y., Kim, J.-T., Eem, S.-H. (2022). “Research trends on external event identification and screening methods for safety assessment of nuclear power plant.” Journal of the Korea Society of Disaster Information, Vol. 18, No. 2, pp. 252-260.

[8] Kim, K.-M., Kim, S.-I., Son, E.-S., Ko, M.-H., Yang, H.-H., Choi, D.C. (2025) “A study on the analysis of key factors for reducing fire CDF in nuclear power plants.” Journal of the Korea Society of Disaster Information, Vol. 21, No. 2, pp. 342-347.

[9] Laerd Statistics (2024). Independent t-test for Two Samples.

[10] Salley, D. (2004). Browns Ferry Nuclear Plant, Unit 1, Fire of March 22, 1975. NUREG/IA-0204.

[11] Seguchi, T. (2013). “Degradation mechanisms of cable insulation materials during radiation-thermal ageing in radiation environment.” Polymer Degradation and Stability, Vol. 98, No. 1, pp. 263-270.

[12] U.S. Nuclear Regulatory Commission (2016). Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE-FIRE) Vol.. NUREG-2178.

[13] U.S. Nuclear Regulatory Commission (2020). Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants. Regulatory Guide 1.89.

Journal of the Society of Disaster Information ISSN:1976-2208(Print) 2671-5287(Online) 한국재난정보학회논문집

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