The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

References

References
1.
Boiler and Pressure Vessel Code
, Section XI, 1989 Edition With the 1989 Addenda up to and Including the 2004 Edition With the 2005 Addenda,
American Society of Mechanical Engineers
,
New York
.
2.
Title 10 of the Code of Federal Regulations
, 2011,
Part 50
,
U.S. Government Printing Office
,
Washington, DC
, April 11.
3.
Radiation Embrittlement of Reactor Vessel Materials
, 1988, Regulatory Guide 1.99, Revision 2,
U.S. Nuclear Regulatory Commission
,
Washington, DC
.
4.
Boiler and Pressure Vessel Code
, Section III, Paragraph NB-2300, 1989 Edition With the 1989 Addenda up to and Including the 2004 Edition With the 2005 Addenda,
American Society of Mechanical Engineers
,
New York
.
5.
Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis
, 2002, Regulatory Guide 1.174, Revision 1,
U.S. Nuclear Regulatory Commission
,
Washington, DC
.
6.
Recommended Screening Limits for Pressurized Thermal Shock (PTS), 2007, U.S. Nuclear Regulatory Commission, Washington, DC, NUREG-1874 (ADAMS Accession No. ML070860156).
7.
Fracture Analysis of Vessels—Oak Ridge FAVOR, v06.1, Rev. 2, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations, 2007, Oak Ridge National Laboratory, Oak Ridge, TN, Draft NUREG/CR (ORNL/TM-2007/0030).
8.
Fracture Analysis of Vessels—Oak Ridge-Heatup, FAVOR-HT, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations, 2006, Oak Ridge National Laboratory, Oak Ridge, TN.
9.
Materials Reliability Program: Validation and Verification of FAVOR v02.4: PFM Computational Algorithms and Associated Sampled Variables (MRP-90)
, 2003,
EPRI
,
Palo Alto, CA
,
1007826
.
10.
Materials Reliability Program: Validation and Verification of FAVOR v03.1: PFM Computational Algorithms and Associated Sampled Variables (MRP-125), 2004, EPRI, Palo Alto, CA, 1010953.
11.
Materials Reliability Program: Validation and Verification of FAVOR, v04.1: PFM Computational Algorithms and Associated Sampled Variables (MRP-171), 2005, EPRI, Palo Alto, CA, 1011795.
12.
Materials Reliability Program: Development of Alternate ASME Section XI Appendix G Methodology: Validation and Verification of FAVOR, v06.1 (MRP-226), 2007, EPRI, Palo Alto, CA, 1015012.
13.
Jackson
,
D. A.
,
Abramson
,
L.
,
Doctor
,
S. R.
,
Simonen
,
F. A.
, and
Shuster
,
G. J.
, 2001, “
Developing a Generalized Flaw Distribution for Reactor Pressure Vessels
,”
Nucl. Eng. Des.
,
208
, pp.
123
131
.
14.
Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10CFR50.61)
, 2007,
U.S. Nuclear Regulatory Commission
,
Washington, DC
, NUREG-1806.
You do not currently have access to this content.