Status of the United States nuclear regulatory commission pressurized thermal shock rule re-evaluation project

International Conference on Nuclear Engineering, Proceedings, ICONE

Volumn 1, Issue , 2002, Pages 389-397

  Dickson, Terry L ^a   Malik, Shah N ^a   Kirk, Mark T ^a   Jackson, Deborah A ^a  

^a OAK RIDGE NATIONAL LABORATORY   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

HUMAN RELIABILITY ANALYSIS; MATERIALS EMBRITTLEMENT EFFECTS; NUCLEAR REACTOR PRESSURE VESSELS; PRESSURIZED THERMAL SHOCK; PROBABILISTIC RISK ASSESSMENT; THERMAL HYDRAULICS;

ACCIDENT PREVENTION; COMPUTATIONAL METHODS; CRACK INITIATION; FRACTURE MECHANICS; FRACTURE TOUGHNESS; PRESSURE VESSELS; RISK ASSESSMENT; TRANSIENTS;

PRESSURIZED WATER REACTORS;

EID: 0036392903     PISSN: None     EISSN: None     Source Type: Conference Proceeding    
DOI: 10.1115/ICONE10-22656     Document Type: Conference Paper

References (25)

1. U.S. Nuclear Regulatory Policy Issue, 1982, NRC Staff Evaluation of Pressurized Thermal Shock, SECY 82-465.
2. Code of Federal Regulations, Title 10 (Energy), Part 50, Section 50.61 and Appendix G, 2001.
3. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.154, Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors, 1987.
4. Selby, D.L., et al., Pressurized-Thermal-Shock Evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant, NUREG/CR-4022 (ORNL/TM-9408), Oak Ridge National Laboratory, Oak Ridge, TN, September 1985.
5. Selby, D.L., et al., Pressurized-Thermal-Shock Evaluation of the H.B. Robinson Nuclear Power Plant, NUREG/CR-4183 (ORNL/TM-9567), Oak Ridge National Laboratory, Oak Ridge, TN, September 1985.
6. Burns, T.J., et al., 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), Oak Ridge National Laboratory, Oak Ridge, TN, May 1986.
7. SECY 92-283, "Action Plan to Implement Lessons Learned from the Yankee Rowe Rector Vessel Embrittlement Issue," USNRC, August 14, 1992.
8. Marshall Committee, "An Assessment of the Integrity of PWR Pressure Vessels," Second Report, Chairman: D. W. Marshall, Published by: U.K. Atomic Energy Authority, March 1982.
9. T.L. Dickson, S. N. M. Malik, J. W. Bryson, and F. A. Simonen, "Revisiting the Integrated Pressurized Thermal Shock Studies of an Aging Pressurized Water Reactor," ASME PVP Volume 388, Fracture, Design Analysis of Pressure Vessels, Heat Exchangers, Piping Components, and Fitness for Service, ASME Pressure Vessels and Piping Conference, August, 1999.
10. U.S. Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Regulatory Guide 1.190, March 2001.
11. "DORT, "Two-Dimensional Discrete Ordinates Transport Code," RSIC Computer Code Collection, CCC-484, Oak Ridge National Laboratory, 1988.
12. D. T. Ingersoll, J. E. White, R. Q. Wright, H. T. Hunter, C. O. Slater, N. M. Greene, R. E. MacFarlane, R. W. Roussin, "Production and Testing of the VITAMIN-B6 Fine-Group and the BUGLE-93 Broad-Group Neutron/Photon Cross-section Libraries Derived from ENDF/B-VI Nuclear Data," ORNL-6795, NUREG/CR-6214, January 1995.
13. Arcieri W.C., Beaton, R., Lee, T. Bessette, D.E., "RELAP5 Thermal Hydraulic Analysis to Support PTS Evaluations for the Plant X Nuclear Power Plant, Draft NUREG/CR Report, January 2001.
14. Smith, C. L., et al., Testing, Verifying and Validating SAPHIRE Versions 6.0 and 7.0, NUREG/CR-6688, October 2000.
15. Technical Basis and Implementation Guidelines for A Technique for Human Event Analysis (ATHEANA), USNRC - Division of Risk Analysis and Applications, Office of Nuclear Regulatory Research, NUREG-1624, Revision 1, May 2000.
16. Jackson, D.A., and Abramson, L., 1999, Report on the Results of the Expert Judgment Process for the Generalized Flaw Size and Density Distribution for Domestic Reactor Pressure Vessels, U.S. Nuclear Regulatory Commission Office of Research, FY 2000-2001 Operating Milestone 1A1ACE.
17. Schuster, G.J., Doctor, S.R., Crawford, S.L., and Pardini, A.F., 1998, Characterization of Flaws in U.S. Reactor Pressure Vessels: Density and Distribution of Flaw Indications in PVRUF, USNRC Report NUREG/CR-6471, Vol. 1, U.S. Nuclear Regulatory Commission, Washington, D.C.,
18. Schuster, G.J., Doctor, S.R., and Heasler, P.G., 2000, Characterization of Flaws in U.S. Reactor Pressure Vessels: Validation of Flaw Density and Distribution in the Weld Metal of the PVRUF Vessel, USNRC Report NUREG/CR-6471, Vol. 2, U.S. Nuclear Regulatory Commission, Washington, D.C.
19. Schuster, G.J., Doctor, S.R., Crawford, S.L., and Pardini, A.F., 1999, Characterization of Flaws in U.S. Reactor Pressure Vessels: Density and Distribution of Flaw Indications in the Shoreham Vessel, USNRC Report NUREG/CR-6471, Vol. 3, U.S. Nuclear Regulatory Commission, Washington, D.C.
20. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," USNRC, Revision 2, May 1988.
21. E.D. Eason and J.E. Wright, "Updated Transition Temperature Shift Model," USNRC Contractor Draft Report, July 2000.
22. EPRI Special Report, 1978, Flaw Evaluation Procedures: ASME Section XI, EPRI NP-719-SR, Electric Power Research Institute, Palo Alto, CA.
23. Ia Fracture Toughness Databases for RPV Steels, ORNL/NRC/LTR-99/27, Oak Ridge National Laboratory, Oak Ridge, TN, February, 2000.
24. Dickson, T.L., Williams, P.T., and Bass, B.R., An Improved Probabilistic Fracture Mechanics Model for Pressurized Thermal Shock, 16th International Structural Mechanics in Reactor Technology (SMiRT) Conference, Washington, D.C., August, 2001.
25. Dickson, T.L. and Malik, S.M., An Updated Probabilistic Fracture Mechanics Methodology for Application to Pressurized Thermal Shock, International Journal of Pressure Vessels and Piping, 2001, pp. 155-163.