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http://dx.doi.org/10.3795/KSME-A.2004.28.6.693

The Effect of Analysis Variables on the Failure Probability of the Reactor Pressure Vessel by Pressurized Thermal Shock  

Jang, Chang-Heui (한국원자력안전기술원 금속재료실)
Jhung, Myung-Jo (한국원자력안전기술원 원자력안전연구실)
Kang, Suk-Chull (한국원자력안전기술원 금속재료)
Choi, Young-Hwan (한국원자력안전기술원 원자력안전연구)
Publication Information
Transactions of the Korean Society of Mechanical Engineers A / v.28, no.6, 2004 , pp. 693-700 More about this Journal
Abstract
The probabilistic fracture mechanics(PFM) is a useful analytical tool to assess the integrity of reactor pressure vessel(RPV) at the event of pressurized thermal shock(PTS). In PFM, the probabilities of flaw initiation and propagation are estimated by comparing the applied stress intensity factor with the fracture toughness calculated by the simulation of various stochastic variables. It is known that the results of PFM analyses are dependent on the choice of the stochastic parameters and assumptions. Of the various variables and assumptions, we investigated the effects of the RT$_{NDT}$ shift equations, fracture toughness curves, and flaw distributions on the PFM results for the three PTS transients. The results showed that the combined effects of the RT$_{NDT}$ shift equations and fracture toughness curves are complicated and dependent on the characteristics of the transients, the chemistry of the materials, the fast neutron fluence, and so on.
Keywords
Pressurized Thermal Shock; Reactor Pressure Vessel; Probabilistic Analysis; RT$_{NDT}$; Flaw Distribution; Fracture Toughness;
Citations & Related Records
Times Cited By KSCI : 1  (Citation Analysis)
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1 USNRC, 1987, Format And Content of Plant-Specific Pressurized Thermanl Shock Safety Analysis Reports For Pressurized Water Reactors, Regulatory Guide 1.154, US Nucelar Regulatory Commission
2 Dickson, T.L., 1994, FAVOR : A Fracture Anlaysis Code for Nuclear Reactor Pressure Vessels, Release 9401, ORNL/NRC/LTR/94/1, ORNL
3 KEPRI, 1999, PTS Evaluation of Kori 1 Reactor Pressure Vessel, TR. 96NJ12.J1999.81, Korea Electric Power Research Institute
4 OECD/NEA, 1999, Final Report on the International Comparative Assessment Study of Pressurized Thermal Shock in Reactor Pressure Vessels, NEA/CSNI/R(99)3, OECD
5 OECD/NEA, 2003, PROSIR - Probabilistic Structural Integrity of a PWR Reactor Pressure Vessel Round Robin Analysis, OECD
6 Jang, C. H., et al., 2001, 'Developement of the Improved Probabilistic Fracture Mechanics Analysis Code: VINTIN,' Proc. 2001 Spring Meeting of KNS, Jeju, Korea
7 USNRC, 1998, Technical Basis for an ASTM Standard on Determining the Reference Temperature, $T_o$, for Ferritic Steels in Transition Range, NUREG/CR-5505, US Nuclear Regulatory Commission
8 EPRI, 1998, Application for Master Curve Fracture Toughness Methodology for Ferritic Steels, TR-108390, Electric Power Research Institute
9 KINS, 2000, Round Robin Analysis of Pressurized Thermal Shock for Reactor Pressure Vessel, KINS/RR-029, Korea Institute of Nuclear Safety
10 Pierre Petrequin, 1996, A Review of Formulas for Predicting Irradiation Embrittlement of Reactor Pressure Vessel Materials, AMES Report No.6
11 USNRC, 1988, Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99 Rev. 2, US Nuclear Regulatory Commission