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[1] Qian, G. & Niffenegger, M., Deterministic and probabilistic analysis of a reactor pressure vessel subjected to pressurized thermal shocks. Nuclear Engineering and Design, 273, pp. 381–395, 2014. [Crossref]
[2] Sonnenburg, H.G., Phänomenologische Versuchsauswertung des Versuchs UPTFTRAM C1 Thermisches Mischen im Kaltstrang. GRS-A-2434, 1997.
[3] Williams, P.T., Dickson, T.L. & Yin, S., Fracture analysis of vessels-Oak Ridge FAVOR, v 04.1, computer code: theory and implementation of algorithms, methods, and correlations. NUREG/CR -6854, 2004.
[4] Mahaffy, J., Chung, B., Dubois, F., Ducros, F., Graffard, E., Heitsch, M., Henriksson, M., Komen, E., Moretti, F., Morii, T., Muhlbauer, P., Rohde, U., Scheuerer, M., Smith, B.L., Song, C., Watanabe, T. & Zigh, G., Best practice guidelines for the use of CFD in nuclear reactors safety applications, NES/CSNI/R (2007)5.
[5] Niffenegger, M. & Reichlin, K., The proper use of thermal expansion coefficients in finite element calculations. Nuclear Engineering and Design, 243, pp. 356–359, 2012. [Crossref]
[6] Verordnung des UVEK über die Methodik und die Randbedingungen zur Überprüfung der Kriterien für die vorläufige Ausserbetriebnahme von Kernkraftwerken, (SR 732.114.5), 16.4.2008.
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Open Access
Research article

Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks

M. Niffenegger1,
G. Qian1,
V.F. Gonzalez-Albuixech1,
M. Sharabi2,
N. Lafferty2
1
Paul Scherrer Institut, Laboratory for Nuclear Materials
2
Laboratory for Thermal-hydraulics, CH-5232 Villigen PSI, Switzerland
International Journal of Computational Methods and Experimental Measurements
|
Volume 4, Issue 3, 2016
|
Pages 288-300
Received: N/A,
Revised: N/A,
Accepted: N/A,
Available online: N/A
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Abstract:

The integrity of reactor pressure vessels (RPVs) of nuclear power plants is one of the most important topics in the field of nuclear energy production. Therefore, the integrity of RPVs has to be assessed for normal operation as well as for emergency transients. A critical transient concerning the RPV integrity is the emergency cooling of a pressurized water reactor, initiated by a leak in the hot leg. Such shock-like cooling in combination with the pressure, the so-called pressurized thermal shock (PTS), causes high thermal stresses in the RPV wall and stress intensities of pre-existing cracks which could exceed the remaining fracture toughness of the material, which is additionally embrittled due to neutron irradiation. This may result in a cleavage fracture of the most safety relevant reactor component.

We present a PTS study of a reference reactor, starting with the calculation of the thermal-hydraulic system behaviour, followed by the simulation of the cold water temperature injection and mixing by means of computational fluid dynamics (CFD) method and the subsequent structural and fracture mechanics calculation. In the safety assessment, we compare the evolution of the stress intensity factors (SIF) during an emergency cooling transient with the fracture toughness at the tip of postulated cracks. Results and open questions will be discussed in the light of a realistic estimation of safety margins.

Keywords: Computational fluid dynamics, Finite element method, Fracture mechanics, Pressurized thermal shock, Reactor pressure vessel, RELAP5
Data Availability

The data used to support the findings of this study are available from the corresponding author upon request.

Conflicts of Interest

The authors declare that they have no conflicts of interest.

References
[1] Qian, G. & Niffenegger, M., Deterministic and probabilistic analysis of a reactor pressure vessel subjected to pressurized thermal shocks. Nuclear Engineering and Design, 273, pp. 381–395, 2014. [Crossref]
[2] Sonnenburg, H.G., Phänomenologische Versuchsauswertung des Versuchs UPTFTRAM C1 Thermisches Mischen im Kaltstrang. GRS-A-2434, 1997.
[3] Williams, P.T., Dickson, T.L. & Yin, S., Fracture analysis of vessels-Oak Ridge FAVOR, v 04.1, computer code: theory and implementation of algorithms, methods, and correlations. NUREG/CR -6854, 2004.
[4] Mahaffy, J., Chung, B., Dubois, F., Ducros, F., Graffard, E., Heitsch, M., Henriksson, M., Komen, E., Moretti, F., Morii, T., Muhlbauer, P., Rohde, U., Scheuerer, M., Smith, B.L., Song, C., Watanabe, T. & Zigh, G., Best practice guidelines for the use of CFD in nuclear reactors safety applications, NES/CSNI/R (2007)5.
[5] Niffenegger, M. & Reichlin, K., The proper use of thermal expansion coefficients in finite element calculations. Nuclear Engineering and Design, 243, pp. 356–359, 2012. [Crossref]
[6] Verordnung des UVEK über die Methodik und die Randbedingungen zur Überprüfung der Kriterien für die vorläufige Ausserbetriebnahme von Kernkraftwerken, (SR 732.114.5), 16.4.2008.
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Niffenegger, M., Qian, G., Gonzalez-Albuixech, V. F., Sharabi, M., & Lafferty, N. (2016). Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks. Int. J. Comput. Methods Exp. Meas., 4(3), 288-300. https://doi.org/10.2495/CMEM-V4-N3-288-300
M. Niffenegger, G. Qian, V. F. Gonzalez-Albuixech, M. Sharabi, and N. Lafferty, "Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks," Int. J. Comput. Methods Exp. Meas., vol. 4, no. 3, pp. 288-300, 2016. https://doi.org/10.2495/CMEM-V4-N3-288-300
@research-article{Niffenegger2016AnalysisOA,
title={Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks},
author={M. Niffenegger and G. Qian and V.F. Gonzalez-Albuixech and M. Sharabi and N. Lafferty},
journal={International Journal of Computational Methods and Experimental Measurements},
year={2016},
page={288-300},
doi={https://doi.org/10.2495/CMEM-V4-N3-288-300}
}
M. Niffenegger, et al. "Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks." International Journal of Computational Methods and Experimental Measurements, v 4, pp 288-300. doi: https://doi.org/10.2495/CMEM-V4-N3-288-300
M. Niffenegger, G. Qian, V.F. Gonzalez-Albuixech, M. Sharabi and N. Lafferty. "Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks." International Journal of Computational Methods and Experimental Measurements, 4, (2016): 288-300. doi: https://doi.org/10.2495/CMEM-V4-N3-288-300
NIFFENEGGER M， QIAN G， GONZALEZ-ALBUIXECH V F， et al. Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks[J]. International Journal of Computational Methods and Experimental Measurements， 2016， 4(3)： 288-300. https://doi.org/10.2495/CMEM-V4-N3-288-300