Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock

Ru-Feng Liu

doi:doi:10.11648/j.ijmea.20221002.11

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Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock

Ru-Feng Liu

Published in International Journal of Mechanical Engineering and Applications (Volume 10, Issue 2)

Received: 14 March 2022 Accepted: 1 April 2022 Published: 14 April 2022

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Abstract

This investigation is to present a verification study of probabilistic fracture mechanics (PFM) analysis code for a reactor pressure vessel (RPV) during pressurized thermal shock (PTS). The probabilistic fracture mechanics code FAVOR, developed by Oak Ridge National Laboratory, is used to calculate the conditional probabilities of crack initiation and penetration for welds that are located in the RPV beltline region. The procedure includes deterministic analyses of the temperature and stress distributions through the vessel wall at the PTS, and probabilistic analyses on the vessel failure probability as a result of PTS transients. The RPV geometries, material properties, and properties related to embrittlement are those in taken from previous studies. Two previously suggested hypothetical transients, which may seriously affect RPV integrity, are also taken into account. To verify the results of PFM round robin analysis of RPV during PTS events, several models and Monte Carlo methods for determining PFM performance are used and they agree on the accuracy of the failure assessment is obtained. The present work can be regarded as various important factors about performing PFM that affect in evaluating the structural safety and operational stability of RPVs. The comparisons of the paper also support the finding that the FAVOR code is very practically useful in assessing failure probability.

Published in	International Journal of Mechanical Engineering and Applications (Volume 10, Issue 2)
DOI	10.11648/j.ijmea.20221002.11
Page(s)	17-24
Creative Commons	This is an Open Access article, distributed under the terms of the Creative Commons Attribution 4.0 International License (http://creativecommons.org/licenses/by/4.0/), which permits unrestricted use, distribution and reproduction in any medium or format, provided the original work is properly cited.
Copyright	Copyright © The Author(s), 2022. Published by Science Publishing Group

Keywords

Pressurized Water Reactor, Reactor Pressure Vessel, Probabilistic Fracture Mechanics

References

[1]	10 CFR 50.61. (1984). Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission.
[2]	NUREG/CR-3770 (ORNL/TM-9176). (1986). Preliminary Development of an Integrated Approach to the Evaluation of Pressurized Thermal Shock as Applied to the Oconee Unit 1 Nuclear Power Plant, U.S. Nuclear Regulatory Commission.
[3]	NUREG/CR-4022 (ORNL/TM-9408). (1985). Pressurized Thermal Shock Evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant, U.S. Nuclear Regulatory Commission.
[4]	NUREG/CR-4183 (ORNL/TM-9567). (1985). Pressurized Thermal Shock Evaluation of the H. B. Robinson Nuclear Power Plant, U.S. Nuclear Regulatory Commission.
[5]	R. G. 1.154. (1987). Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors, U.S. Nuclear Regulatory Commission, U.S. Nuclear Regulatory Commission.
[6]	NEA/CSNI/R (2007) 18. (2008). Proceedings of the CSNI Workshop on Structural Reliability Evaluation and Mechanical Probabilistic Approaches of NPP Components, Nuclear Energy Agency.
[7]	Kanto, Y., Jhung, M.-J., Ting, K., & Yoshimura S. (2010). Summary of international PFM round robin analyses among asian countries on reactor pressure vessel integrity during pressurized thermal shock, The 8th International Workshop on the Integrity of Nuclear Components, Hyogo (Japan).
[8]	Soneda, N., Onchi, T. (1996). Benchmarking studies of probabilistic fracture mechanics analysis code, PROFMAC-11, for assessing pressurized thermal shock events of reactor pressure vessel integrity issues. J. Nucl. Sci. Technol 33 (1): 87-98. doi: 10.1080/18811248.1996.9731866.
[9]	Jhung, M. J., Kim, S. H., Choi, Y. H., Chang, Y. S., Xu, X., Kim, J. M., Kim, J. W., Jang, C. (2010). Probabilistic fracture mechanics round robin analysis of reactor pressure vessels during pressurized thermal shock. J. Nucl. Sci. Technol 47 (12): 1131-1139. doi: 10.1080/18811248.2010.9720980.
[10]	Fracture Analysis of Vessels Oak Ridge FAVOR, v09.1, Computer Code: Theory and Implementation of Algorithms, Methods and Correlations, Oak Ridge National Laboratory, 2010.
[11]	EPRI TR-105001. (1995). Documentation of Probabilistic Fracture Mechanics Codes Used for Reactor Pressure Vessels Subjected to Pressurized Thermal Shock Loading, Electric Power Research Institute.
[12]	Huang, C. C., Chou, H. W., Chen, B. Y., Liu, R. F., Lin, H. C. (2012). Probabilistic fracture analysis for boiling water reactor pressure vessels subjected to low temperature over-pressure event. Ann. Nucl 43: 61-67. doi: 10.1016/j.anucene.2011.12.028.
[13]	Park, J. S., Choi, Y. H., Jhung, M. J. (2016). Probabilistic fracture mechanics analysis of boling water reactor vessel for cool-down and low temperature over-pressurization transients. Nucl. Eng. Technol 48 (2): 543-553. doi: 10.1016/j.net.2015.11.006.
[14]	The Finite-Element Method in Heat Transfer Analysis. (1996). New York: John Wiley & Sons Press.
[15]	EPRI NP-719-SR. (1978). Flaw Evaluation Procedures: ASME Section XI, Electric Power Research Institute.

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APA Style

Ru-Feng Liu. (2022). Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock. International Journal of Mechanical Engineering and Applications, 10(2), 17-24. https://doi.org/10.11648/j.ijmea.20221002.11

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ACS Style

Ru-Feng Liu. Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock. Int. J. Mech. Eng. Appl. 2022, 10(2), 17-24. doi: 10.11648/j.ijmea.20221002.11

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AMA Style

Ru-Feng Liu. Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock. Int J Mech Eng Appl. 2022;10(2):17-24. doi: 10.11648/j.ijmea.20221002.11

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@article{10.11648/j.ijmea.20221002.11,
  author = {Ru-Feng Liu},
  title = {Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock},
  journal = {International Journal of Mechanical Engineering and Applications},
  volume = {10},
  number = {2},
  pages = {17-24},
  doi = {10.11648/j.ijmea.20221002.11},
  url = {https://doi.org/10.11648/j.ijmea.20221002.11},
  eprint = {https://article.sciencepublishinggroup.com/pdf/10.11648.j.ijmea.20221002.11},
  abstract = {This investigation is to present a verification study of probabilistic fracture mechanics (PFM) analysis code for a reactor pressure vessel (RPV) during pressurized thermal shock (PTS). The probabilistic fracture mechanics code FAVOR, developed by Oak Ridge National Laboratory, is used to calculate the conditional probabilities of crack initiation and penetration for welds that are located in the RPV beltline region. The procedure includes deterministic analyses of the temperature and stress distributions through the vessel wall at the PTS, and probabilistic analyses on the vessel failure probability as a result of PTS transients. The RPV geometries, material properties, and properties related to embrittlement are those in taken from previous studies. Two previously suggested hypothetical transients, which may seriously affect RPV integrity, are also taken into account. To verify the results of PFM round robin analysis of RPV during PTS events, several models and Monte Carlo methods for determining PFM performance are used and they agree on the accuracy of the failure assessment is obtained. The present work can be regarded as various important factors about performing PFM that affect in evaluating the structural safety and operational stability of RPVs. The comparisons of the paper also support the finding that the FAVOR code is very practically useful in assessing failure probability.},
 year = {2022}
}

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TY  - JOUR
T1  - Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock
AU  - Ru-Feng Liu
Y1  - 2022/04/14
PY  - 2022
N1  - https://doi.org/10.11648/j.ijmea.20221002.11
DO  - 10.11648/j.ijmea.20221002.11
T2  - International Journal of Mechanical Engineering and Applications
JF  - International Journal of Mechanical Engineering and Applications
JO  - International Journal of Mechanical Engineering and Applications
SP  - 17
EP  - 24
PB  - Science Publishing Group
SN  - 2330-0248
UR  - https://doi.org/10.11648/j.ijmea.20221002.11
AB  - This investigation is to present a verification study of probabilistic fracture mechanics (PFM) analysis code for a reactor pressure vessel (RPV) during pressurized thermal shock (PTS). The probabilistic fracture mechanics code FAVOR, developed by Oak Ridge National Laboratory, is used to calculate the conditional probabilities of crack initiation and penetration for welds that are located in the RPV beltline region. The procedure includes deterministic analyses of the temperature and stress distributions through the vessel wall at the PTS, and probabilistic analyses on the vessel failure probability as a result of PTS transients. The RPV geometries, material properties, and properties related to embrittlement are those in taken from previous studies. Two previously suggested hypothetical transients, which may seriously affect RPV integrity, are also taken into account. To verify the results of PFM round robin analysis of RPV during PTS events, several models and Monte Carlo methods for determining PFM performance are used and they agree on the accuracy of the failure assessment is obtained. The present work can be regarded as various important factors about performing PFM that affect in evaluating the structural safety and operational stability of RPVs. The comparisons of the paper also support the finding that the FAVOR code is very practically useful in assessing failure probability.
VL  - 10
IS  - 2
ER  -

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Author Information

Ru-Feng Liu

Engineering Technology and Facilities Operation Division, Institute of Nuclear Energy Research, Taoyuan, Taiwan (ROC)

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Plain Text BibTeX RIS

APA Style

Ru-Feng Liu. (2022). Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock. International Journal of Mechanical Engineering and Applications, 10(2), 17-24. https://doi.org/10.11648/j.ijmea.20221002.11

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ACS Style

Ru-Feng Liu. Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock. Int. J. Mech. Eng. Appl. 2022, 10(2), 17-24. doi: 10.11648/j.ijmea.20221002.11

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AMA Style

Ru-Feng Liu. Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock. Int J Mech Eng Appl. 2022;10(2):17-24. doi: 10.11648/j.ijmea.20221002.11

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@article{10.11648/j.ijmea.20221002.11,
  author = {Ru-Feng Liu},
  title = {Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock},
  journal = {International Journal of Mechanical Engineering and Applications},
  volume = {10},
  number = {2},
  pages = {17-24},
  doi = {10.11648/j.ijmea.20221002.11},
  url = {https://doi.org/10.11648/j.ijmea.20221002.11},
  eprint = {https://article.sciencepublishinggroup.com/pdf/10.11648.j.ijmea.20221002.11},
  abstract = {This investigation is to present a verification study of probabilistic fracture mechanics (PFM) analysis code for a reactor pressure vessel (RPV) during pressurized thermal shock (PTS). The probabilistic fracture mechanics code FAVOR, developed by Oak Ridge National Laboratory, is used to calculate the conditional probabilities of crack initiation and penetration for welds that are located in the RPV beltline region. The procedure includes deterministic analyses of the temperature and stress distributions through the vessel wall at the PTS, and probabilistic analyses on the vessel failure probability as a result of PTS transients. The RPV geometries, material properties, and properties related to embrittlement are those in taken from previous studies. Two previously suggested hypothetical transients, which may seriously affect RPV integrity, are also taken into account. To verify the results of PFM round robin analysis of RPV during PTS events, several models and Monte Carlo methods for determining PFM performance are used and they agree on the accuracy of the failure assessment is obtained. The present work can be regarded as various important factors about performing PFM that affect in evaluating the structural safety and operational stability of RPVs. The comparisons of the paper also support the finding that the FAVOR code is very practically useful in assessing failure probability.},
 year = {2022}
}

Copy | Download

TY  - JOUR
T1  - Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock
AU  - Ru-Feng Liu
Y1  - 2022/04/14
PY  - 2022
N1  - https://doi.org/10.11648/j.ijmea.20221002.11
DO  - 10.11648/j.ijmea.20221002.11
T2  - International Journal of Mechanical Engineering and Applications
JF  - International Journal of Mechanical Engineering and Applications
JO  - International Journal of Mechanical Engineering and Applications
SP  - 17
EP  - 24
PB  - Science Publishing Group
SN  - 2330-0248
UR  - https://doi.org/10.11648/j.ijmea.20221002.11
AB  - This investigation is to present a verification study of probabilistic fracture mechanics (PFM) analysis code for a reactor pressure vessel (RPV) during pressurized thermal shock (PTS). The probabilistic fracture mechanics code FAVOR, developed by Oak Ridge National Laboratory, is used to calculate the conditional probabilities of crack initiation and penetration for welds that are located in the RPV beltline region. The procedure includes deterministic analyses of the temperature and stress distributions through the vessel wall at the PTS, and probabilistic analyses on the vessel failure probability as a result of PTS transients. The RPV geometries, material properties, and properties related to embrittlement are those in taken from previous studies. Two previously suggested hypothetical transients, which may seriously affect RPV integrity, are also taken into account. To verify the results of PFM round robin analysis of RPV during PTS events, several models and Monte Carlo methods for determining PFM performance are used and they agree on the accuracy of the failure assessment is obtained. The present work can be regarded as various important factors about performing PFM that affect in evaluating the structural safety and operational stability of RPVs. The comparisons of the paper also support the finding that the FAVOR code is very practically useful in assessing failure probability.
VL  - 10
IS  - 2
ER  -

Copy | Download