Browse > Article
http://dx.doi.org/10.5516/NET.07.2013.704

DEVELOPMENT OF THE ALTERNATE PRESSURIZED THERMAL SHOCK RULE (10 CFR 50.61a) IN THE UNITED STATES  

Kirk, Mark (Component Integrity Branch, Office of Nuclear Regulatory Research, United States Nuclear Regulatory Commission)
Publication Information
Nuclear Engineering and Technology / v.45, no.3, 2013 , pp. 277-294 More about this Journal
Abstract
In the early 1980s, attention focused on the possibility that pressurized thermal shock (PTS) events could challenge the integrity of a nuclear reactor pressure vessel (RPV) because operational experience suggested that overcooling events, while not common, did occur, and because the results of in-reactor materials surveillance programs showed that RPV steels and welds, particularly those having high copper content, experience a loss of toughness with time due to neutron irradiation embrittlement. These recognitions motivated analysis of PTS and the development of toughness limits for safe operation. It is now widely recognized that state of knowledge and data limitations from this time necessitated conservative treatment of several key parameters and models used in the probabilistic calculations that provided the technical of the PTS Rule, 10 CFR 50.61. To remove the unnecessary burden imposed by these conservatisms, and to improve the NRC's efficiency in processing exemption and license exemption requests, the NRC undertook the PTS re-evaluation project. This paper provides a synopsis of the results of that project, and the resulting Alternate PTS rule, 10 CFR 50.61a.
Keywords
Pressurized Thermal Shock; Nuclear Reactor Pressure Vessel; Embrittlement;
Citations & Related Records
연도 인용수 순위
  • Reference
1 N. Sui, "Uncertainty Analysis and Pressurized Thermal Shock: An Opinion," U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML992710066 (1999).
2 R.D. Cheverton, D.G Ball, S.E. Bolt, S.K. Iskander, and R.K. Nanstad, "Pressure Vessel Fracture Studies Pertaining to the PWR Thermal-Shock Issue: Experiments TSE-5, TSE-5A, and TSE-6," NUREG/CR-4249 (ORNL-6163), Oak Ridge National Laboratory (1985).
3 M.T. EricksonKirk, T. Dickson, T. Mintz, F. Simonen, "Sensitivity Studies of the Probabilistic Fracture Mechanics Model Used in FAVOR Version 03.1," NUREG-1808, U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML051780325 (2004).
4 M. EricksonKirk, B. Elliott, L. Abramson, ''Statistical Procedures for Assessing Surveillance Data for 10 CFR Part 50.61a,'' U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML081290654 (2008).
5 T.L. Dickson and P.T. Williams, "Fracture Analysis of Vessels Oak Ridge, FAVOR v04.1, Computer Code: User's Guide," NUREG/CR-6855, U.S. Nuclear Regulatory Commission (2004), http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6855/.
6 P.T. Williams and T.L. Dickson, "Fracture Analysis of Vessels Oak Ridge, FAVOR v04.1: Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations," NUREG/CR-6854, U.S. Nuclear Regulatory Commission (2004), http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6854/.
7 M.T. EricksonKirk, "Probabilistic Fracture Mechanics: Models, Parameters, and Uncertainty Treatment Used in FAVOR Version 04.1," NUREG-1807, U.S. Nuclear Regulatory Commission (2007), http://www.nrc.gov/readingrm/doc-collections/nuregs/staff/sr1807/.
8 F.A. Simonen, S.R. Doctor, G.J. Schuster, and P.G. Heasler, "A Generalized Procedure for Generating Flaw Related Inputs for the FAVOR Code," NUREG/CR-6817, Rev. 1, U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML051790410 (2003).
9 D.W. Marshall, "An Assessment of the Integrity of PWR Vessels," United Kingdom Atomic Energy Authority (1982).
10 T.L. Dickson and F.A. Simonen, "The Sensitivity of Pressurized Thermal Shock Results to Alternative Models for Weld Flaw Distributions," Proc. of the ASME Pressure Vessel and Piping, Vancouver, Canada, Aug. 4-8, 2002.
11 Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission (2001).
12 D. Moinereau, "The demonstration of warm pre-stress effect in RPV assessment: some experimental results and their interpretation by fracture mechanics," 7th Int. ASTM/ESIS Symposium on Fatigue and Fracture Mechanics, 36th ASTM National Symposium on Fatigue and Fracture Mechanics, Tampa, Florida, USA, Nov. 14-16, 2007.
13 Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission (1988).
14 C. English and W. Server, "Attenuation in US RPV Steels - MRP-56," Electric Power Research Institute (EPRI) (2002).
15 A.J. Brothers and S. Yukawa, "The Effect of Warm Prestressing on Notch Fracture Strength," J. of Basic Eng., vol. 85, pp. 97 (1963).   DOI
16 T.L. Dickson and M.T. EricksonKirk, "The Sensitivity of Risk-Informed Reactor Structural Integrity Analysis Results to Various Interpretations of Warm Pre-stress," Proc. of ASME Conference on Pressure Vessels and Piping, Prague, Czech Republic, July 26-30, 2009.
17 M. Kirk and M.E. Natishan, "Shift in Toughness Transition Temperature Due to Irradiation: ${\Delta}T_o$ vs. ${\Delta}T_{41J}$, A Comparison and Rationalization of Differences," Proc. of the IAEA Specialists Meeting on Master Curve Technology, Prague, Czech Republic, Sep. 17-19, 2001.
18 M. EricksonKirk and M. EricksonKirk, "The Relationship between the Transition and Upper Shelf Fracture Toughness of Ferritic Steels," Fatigue and Fracture of Engineering Materials and Structures, vol. 29, pp. 672-684 (2006).   DOI   ScienceOn
19 Title 10, Section 50.61a, of the Code of Federal Regulations, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," 10 CFR 50.61a, (2010).
20 WASH-740, "Theoretical Possibilities and Consequences of Major Accidents in Large Nuclear Power Plants," United States Atomic Energy Commission Report (1957), http://www.dissident-media.org/infonucleaire/wash740.pdf.
21 WASH-1400, "The Reactor Safety Study," United States Atomic Energy Commission Report (1975).
22 Report of the President's Commission on the Accident at Three Mile Island, http://www.pddoc.com/tmi2/kemeny/index.html.
23 51 FR 28044, August 4, 1986, and 51 FR 30028, August 21, 1986 (republication of 51 FR 28044 in its entirety at the Commission's request), "Safety Goals for the Operations of Nuclear Power Plants: Policy Statement."
24 STAFF REQUIREMENTS - SECY-00-0077, "Modifications to the Reactor Safety Goal Policy Statement," U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML00372 706 (2000).
25 Regulatory Guide 1.154, "Format and Content of Plant- Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," U.S. Nuclear Regulatory Commission (1987).
26 T.J. Burns, "Preliminary Development of an Integrated Approach to the Evaluation of Pressurized Thermal Shock as Applied to the Oconee Unit 1 Nuclear Power Plant," NUREG/CR-3770 (ORNL/TM-9176), U.S. Nuclear Regulatory Commission (1986).
27 D.L. Selby, "Pressurized Thermal Shock Evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant," NUREG/CR-4022 (ORNL/TM-9408), U.S. Nuclear Regulatory Commission (1985).
28 D.L. Selby, "Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant," NUREG/CR-4183 (ORNL/TM-9567), U.S. Nuclear Regulatory Commission (1985).
29 WCAP-15156, "WOG Pilot-Plant Application of the EPRI Alternative Method for Reactor Vessel PTS," Westinghouse Electric Company, LLC, Nuclear Services Division (1999).
30 NUREG-1624, Rev. 1, "Technical Basis and Implementation Guidelines for A Technique for Human Event Analysis (ATHEANA)," U.S. Nuclear Regulatory Commission (2000).
31 SAPHIRE, "Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 7.0," Idaho National Laboratory.
32 RELAP5/MOD3 Code Manual, Volume IV: Models and Correlations (1999).
33 C.D. Fletcher, D.A. Prelewicz, and W.C. Arcieri, "RELAP5/MOD3.2.2_ Assessment for Pressurized Thermal Shock Applications," NUREG/CR-6857, U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML043570394 (2004).
34 J.N. Reyes, et al., "Final Report for the OSU APEX-CE Integral Test Facility," NUREG/CR-6856, U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML043570405 (2004).
35 D.E. Bessette, "Thermal-Hydraulic Evaluations of Pressurized Thermal Shock," NUREG-1809, U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML050390012 (2005).
36 SECY-82-465, "Pressurized Thermal Shock," U.S. Nuclear Regulatory Commission (1982).
37 Title 10, Section 50.61, of the Code of Federal Regulations, "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," 10 CFR 50.61 (1984).
38 M. EricksonKirk, M. Junge, W. Arcieri, B.R. Bass, R. Beaton, D. Bessettle, T.H.J. Chang, T. Dickson, C.D. Fletcher, A. Kolaczkowski, S. Malik, T. Mintz, C. Pugh, F. Simonen, N. Siu, D. Whitehead, P. Williams, R. Woods, and S. Yin, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limits in the PTS Rule (10 CFR 50.61): Summary Report," NUREG-1806, U.S. Nuclear Regulatory Commission (2007), http://www.nrc.gov/ readingrm/doc-collections/nuregs/staff/sr1806/.
39 M. EricksonKirk and T.L. Dickson, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," NUREG-1874, U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML070860156 (2007).
40 SECY-06-0124, "Rulemaking Plan to Amend Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events (10 CFR 50.61)," U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML060530624 (2006).
41 STAFF REQUIREMENTS - SECY-06-0124, "Staff Requirements - SECY-06-0124 - Rulemaking Plan to Amend Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events (10 CFR 50.61)," U.S. Nuclear Regulatory Commission, ADAMS, Accession # ML061810148 (2006).