ML062910417
| ML062910417 | |
| Person / Time | |
|---|---|
| Site: | Boiling Water Reactor Owners Group |
| Issue date: | 11/14/2006 |
| From: | Rosenberg S NRC/NRR/ADRA/DPR/PSPB |
| To: | Bunt R BWR Owners Group, Southern Nuclear Operating Co |
| Thompson J | |
| References | |
| SIR-05-044, TAC MC9694 | |
| Download: ML062910417 (16) | |
Text
November 14, 2006 Mr. Randy C. Bunt Chair, BWR Owners Group Southern Nuclear Operating Company 40 Inverness Center Parkway/Bin B057 Birmingham, AL 35242
SUBJECT:
DRAFT SAFETY EVALUATION FOR THE BOILING WATER REACTOR OWNERS GROUP (BWROG) STRUCTURAL INTEGRITY ASSOCIATES LICENSING TOPICAL REPORT (LTR) SIR-05-044, PRESSURE TEMPERATURE REPORT METHODOLOGY FOR BOILING WATER REACTORS (TAC NO. MC9694)
Dear Mr. Bunt:
By letter dated December 20, 2005, and supplement dated August 29, 2006, the BWROG submitted SIR-05-044, Pressure Temperature Report Methodology for Boiling Water Reactors, Revision 0 to the U.S. Nuclear Regulatory Commission (NRC) staff for review.
Enclosed for the BWROG review and comment is a copy of the NRC staff's draft safety evaluation (SE) for the LTR.
Twenty working days are provided to you to comment on any factual errors or clarity concerns contained in the SE. The final SE will be issued after making any necessary changes and will be made publicly available. The NRC staff's disposition of your comments on the draft SE will be discussed in the final SE.
To facilitate the NRC staff's review of your comments, please provide a marked-up copy of the draft SE showing proposed changes and provide a summary table of the proposed changes.
If you have any questions, please contact Michelle Honcharik at 301-415-1774.
Sincerely,
/RA/
Stacey L. Rosenberg, Chief Special Projects Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project No. 691
Enclosure:
Draft SE cc w/encl: See next page
BWR Owners Group Project No. 691 Mr. Doug Coleman Vice Chair, BWR Owners Group Energy Northwest Columbia Generating Station Mail Drp PE20 P.O. Box 968 Richland, WA 99352-0968 Mr. Amir Shahkarami Executive Chair, BWR Owners Group Exelon Generation Co., LLC Cornerstone II at Cantera 4300 Winfield Road Warrenville, IL 60555 Mr. Richard Libra Executive Vice Chair, BWR Owners Group DTE Energy - Fermi 2 M/C 280 OBA 6400 North Dixie Highway Newport, MI 48166 Mr. William A. Eaton Entergy Operations Inc.
P.O. Box 31995 Jackson, MS 39286 Mr. Richard Anderson First Energy Nuclear Operating Co Perry Nuclear Power Plant 10 Center Road Perry, OH 44081 Mr. Scott Oxenford Energy Northwest Columbia Generating Station Mail Drp PE04 P.O. Box 968 Richland, WA 99352-0968 Mr. James F. Klapproth GE Energy M/C A-16 3901 Castle Hayne Road Wilmington, NC 28401 Mr. Joseph E. Conen Regulatory Response Group Chair BWR Owners Group DTE Energy-Fermi 2 200 TAC 6400 N. Dixie Highway Newport, MI 48166 Mr. J. A. Gray, Jr.
Regulatory Response Group Vice-Chair BWR Owners Group Entergy Nuclear Northeast 440 Hamilton Avenue Mail Stop 12C White Plains, NY 10601-5029 Mr. Thomas G. Hurst GE Energy M/C A-16 3901 Castle Hayne Road Wilmington, NC 28401 Mr. Tim E. Abney GE Energy M/C A-16 3901 Castle Hayne Road Wilmington, NC 28401 BWR Owners Group
November 14, 2006 Mr. Randy C. Bunt Chair, BWR Owners Group Southern Nuclear Operating Company 40 Inverness Center Parkway/Bin B057 Birmingham, AL 35242
SUBJECT:
DRAFT SAFETY EVALUATION FOR THE BOILING WATER REACTOR OWNERS GROUP (BWROG) STRUCTURAL INTEGRITY ASSOCIATES LICENSING TOPICAL REPORT (LTR) SIR-05-044, PRESSURE TEMPERATURE REPORT METHODOLOGY FOR BOILING WATER REACTORS (TAC NO. MC9694)
Dear Mr. Bunt:
By letter dated December 20, 2005, and supplement dated August 29, 2006, the BWROG submitted SIR-05-044, Pressure Temperature Report Methodology for Boiling Water Reactors, Revision 0 to the U.S. Nuclear Regulatory Commission (NRC) staff for review.
Enclosed for the BWROG review and comment is a copy of the NRC staff's draft safety evaluation (SE) for the LTR.
Twenty working days are provided to you to comment on any factual errors or clarity concerns contained in the SE. The final SE will be issued after making any necessary changes and will be made publicly available. The NRC staff's disposition of your comments on the draft SE will be discussed in the final SE.
To facilitate the NRC staff's review of your comments, please provide a marked-up copy of the draft SE showing proposed changes and provide a summary table of the proposed changes.
If you have any questions, please contact Michelle Honcharik at 301-415-1774.
Sincerely,
/RA/
Stacey L. Rosenberg, Chief Special Projects Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project No. 691
Enclosure:
Draft SE cc w/encl: See next page DISTRIBUTION:
PUBLIC PSPB Reading File RidsNrrDpr RidsNrrDprPspb RidsNrrPMMHoncharik RidsNrrLADBaxley RidsOgcMailCenter RidsAcrsAcnwMailCenter BElliot MMitchell RidsNrrDci ADAMS ACCESSION NO.: ML062910417
- No major changes to SE input.
NRR-106 OFFICE PSPB/PM PSPB/LA Tech Branch*
PSPB/BC NAME MHoncharik DBaxley MMitchell SRosenberg DATE 11/9/06 11/9/06 10/10/06 11/14/06 OFFICIAL RECORD COPY
ENCLOSURE DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION LICENSING TOPICAL REPORT (LTR) SIR-05-044 PRESSURE TEMPERATURE REPORT METHODOLOGY FOR BOILING WATER REACTORS, REVISION 0 BOILING WATER REACTORS OWNERS GROUP (BWROG)
PROJECT NO. 691
1.0 INTRODUCTION
1 2
In a letter dated December 20, 2005, the Boiling Water Reactor Owners' Group (BWROG) 3 submitted LTR SIR-05-044, "Pressure Temperature Limits Report Methodology for Boiling 4
Water Reactors", Revision 0, dated December 2005 (Agencywide Documents Access and 5
Management System (ADAMS) Package Accession No. ML053560336) to the U.S. Nuclear 6
Regulatory Commission (NRC) for review and acceptance for referencing in subsequent 7
licensing actions. The BWROG provided this LTR to support the ability of boiling water reactor 8
(BWR) licensees to relocate their pressure-temperature (P/T) curves and associated numerical 9
values (such as heatup/cooldown rates) from facility Technical Specifications (TS) to a 10 Pressure Temperature Limits Report (PTLR), a licensee-controlled document, using the 11 guidelines provided in Generic Letter (GL) 96-03, Relocation of the Pressure Temperature 12 Limit Curves and Low Temperature Overpressure Protection System Limits, (Reference 1).
13 Proposed revisions to this LTR and responses to NRC staff requests for additional information 14 (RAIs) were provided in letter from the BWROG dated August 29, 2006 (ADAMS Accession No.
15 ML062440387).
16 17
2.0 REGULATORY EVALUATION
18 19 2.1 Requirements for Generating P/T Limits for Light-Water Reactors 20 21 The NRC has established requirements in Appendix G of Part 50 to Title 10 of the Code of 22 Federal Regulations (10 CFR Part 50, Appendix G; Reference 2), in order to protect the 23 integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. The 24 regulation at 10 CFR Part 50, Appendix G requires that the P/T limits for an operating 25 light-water nuclear reactor be at least as conservative as those that would be generated if the 26 methods of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) 27 Boiler and Pressure Vessel Code (Reference 3, ASME Code,Section XI, Appendix G) were 28 used to generate the P/T limits. The regulation at 10 CFR Part 50, Appendix G, also requires 29 that applicable surveillance data from reactor pressure vessel (RPV) material surveillance 30 programs be incorporated into the calculations of plant-specific P/T limits, and that the P/T 31 limits for operating reactors be generated using a method that accounts for the effects of 1
neutron irradiation on the material properties of the RPV beltline materials.
2 3
Table 1 to 10 CFR Part 50, Appendix G provides the NRC staffs criteria for meeting the P/T 4
limit requirements of ASME Code,Section XI, Appendix G, as well as the minimum temperature 5
requirements of the rule for bolting up the vessel during normal and pressure testing 6
operations. In addition, NRC staff regulatory guidance related to P/T limit curves is found in 7
Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, 8
(Reference 4), and Standard Review Plan Chapter 5.3.2, Pressure-Temperature Limits and 9
Pressurized Thermal Shock, (Reference 5).
10 11 The regulation at 10 CFR Part 50, Appendix H (Reference 6), provides the NRC staffs criteria 12 for the design and implementation of RPV material surveillance programs for operating light-13 water reactors.
14 15 In March 2001, the NRC staff issued RG 1.190, Calculational and Dosimetry Methods for 16 Determining Pressure Vessel Neutron Fluence (Reference 7). Fluence calculations are 17 acceptable if they are done with approved methodologies or with methods which are shown to 18 conform to the guidance in RG 1.190.
19 20 2.2 Technical Specification Requirements for P/T Limits 21 22 Section 182a of the Atomic Energy Act of 1954 requires applicants for nuclear power plant 23 operating licenses to include TS as part of the license. The Commission's regulatory 24 requirements related to the content of TS are set forth in 10 CFR 50.36 (Reference 8). That 25 regulation requires that the TS include items in five specific categories: (1) safety limits, limiting 26 safety system settings and limiting control settings; (2) limiting conditions for operation (LCOs);
27 (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.
28 29 The regulation at 10 CFR 50.36(c)(2)(ii) requires that LCOs be established for the P/T limits, 30 because the parameters fall within the scope of the Criterion 2 identified in the rule:
31 32 A process variable, design feature, or operating restriction that is an initial 33 condition of a design basis accident or transient analysis that either assumes the 34 failure of or presents a challenge to the integrity of a fission product barrier.
35 36 The P/T limits for BWR-designed light-water reactors fall within the scope of Criterion 2 of 37 10 CFR 50.36(c)(2)(ii) and are therefore ordinarily required to be included within the TS LCOs 38 for a plant-specific facility operating license. On January 31, 1996, the NRC staff issued 39 GL 96-03 to inform licensees that they may request a license amendment to relocate the P/T 40 limit curves and/or low temperature over-pressure protection (LTOP) limit setpoint values from 41 the TS LCOs into a PTLR or other licensee-controlled document that would be controlled 42 through the Administrative Controls Section of the TS. In GL 96-03, the NRC staff informed 43 licensees that in order to implement a PTLR, the P/T limits and LTOP limits for light-water 44 reactors would need to be generated in accordance with an NRC-approved methodology and 45 that the methodology to generate the P/T limits and LTOP limits would need to comply with the 46 requirements of 10 CFR Part 50, Appendices G and H; be documented in an NRC-approved 47 topical report or plant-specific submittal; and be incorporated by reference in the Administrative 48 Controls Section of the TS. The GL also mandated that the TS Administrative Controls Section 49 would need to reference the NRC staffs safety evaluation (SE) issued on the PTLR request 1
and that the PTLR be defined in Section 1.0 of the TS. Attachment 1 to GL 96-03 provided a 2
list of the criteria that the approved methodology and PTLR would be required to meet.
3 4
TS Task Force (TSTF) Traveler No. TSTF-419 (Reference 9) amended the Standard TS (STS) 5 (NUREGs-1430, -1431, -1432, -1433, and -1434) to: (1) delete references to the TS LCO 6
specifications for the P/T limits and LTOP system limits in the TS definition of the PTLR, and 7
(2) revised STS 5.6.6 to identify, by number and title, NRC-approved topical reports that 8
document PTLR methodologies, or the NRC safety evaluation for a plant-specific methodology 9
by NRC letter and date. A requirement was added to the reviewers note to specify the 10 complete citation of the PTLR methodology in the plant-specific PTLR, including the report 11 number, title, revision, date, and any supplements. Only the figures, values, and parameters 12 associated with the P/T limits and LTOP system limits are relocated to the PTLR. The 13 methodology for their development must be reviewed and approved by the NRC. TSTF-419 did 14 not change the requirements associated with the review and approval of the methodology or the 15 requirement to operate within the limits specified in the PTLR. Any changes to a methodology 16 that had not been approved by the NRC staff would continue to require NRC staff review and 17 approval pursuant to the license amendment request provisions and requirements of 18 10 CFR 50.90 (Reference 10).
19 20
3.0 TECHNICAL EVALUATION
21 22 As stated in Section 2.1 of this SE, the NRC staff has established a rule, 10 CFR Part 50, 23 Appendix G, that requires licensees to establish limits on the pressure and temperature of the 24 RCPB in order to protect the RCPB against brittle failure (i.e., against brittle fast-fracture).
25 These limits are defined by P/T limit curves for normal operations (including heatup and 26 cooldown operations of the reactor coolant system (RCS), normal operation of the RCS with the 27 reactor being in a critical condition, and transient operating conditions) and during pressure 28 testing conditions (i.e., either inservice leak rate testing and/or hydrostatic testing conditions).
29 30 The BWROG LTR SIR-05-44 was prepared by Structural Integrity Associates and has three 31 sections and two appendices. Section 1.0 describes the background and purpose for the LTR.
32 Section 2.0 provides the fracture mechanics methodology and its basis for developing P/T 33 limits. Section 3.0 provides a step-by-step procedure for calculating P/T limits. Appendix A 34 provides guidance for evaluating surveillance data. Appendix B provides a template PTLR.
35 36 3.1 Evaluation of Section 2.0 of the LTR 37 38 Section 2.0 provides the fracture mechanics methodology and its basis for developing P/T 39 limits. The NRC staff evaluation of this section is based on the criteria contained in 40 of GL 96-03. Attachment 1 of GL 96-03 contains seven technical criteria that the 41 contents of proposed methodology should conform to if license amendments requesting PTLRs 42 are to be approved by the NRC staff. The NRC staffs evaluations of the contents of BWROG 43 methodology against the seven criteria in Attachment 1 of GL 96-03 are given in the 44 subsections that follow.
45 46 GL 96-03, Attachment 1, Criterion 1:
1 2
Criterion 1 requires that the methodology describe the transport calculation methods including 3
computer codes and formulas used to calculate neutron fluences.
4 5
Table 1-1 in the BWROGs August 29, 2006, letter indicates this LTR does not describe the 6
transport calculation methods including computer codes and formulas used to calculate neutron 7
fluences. It indicates fluence methods and results must comply with RG 1.190 and have NRC 8
approval for use with this LTR. Table 1-1 will be included in the LTR. Therefore, this will be a 9
plant-specific action item to be addressed by licensees. Since Table 1-1 in the proposed LTR 10 methodology will indicate that the fluence methods and results must comply with RG 1.190 and 11 have NRC approval, this criterion has been satisfied.
12 13 GL 96-03, Attachment 1, Criterion 2:
14 15 Criterion 2 requires that the methodology describe the surveillance program and indicates that 16 the topical report should contain a place holder for the requested information.
17 18 Table 1-1 in the BWROGs August 29, 2006, letter indicates this information is in Appendix A of 19 the template PTLR, which is in Appendix B of the LTR. Therefore, this will be a plant-specific 20 action item to be addressed by licensees. Since Table 1-1 indicates the information will be 21 included in the PTLR, this criterion has been satisfied.
22 23 GL 96-03, Attachment 1, Criterion 3:
24 25 Criterion 3 requires that the methodology describe how the LTOP system limits are calculated 26 applying system/thermal hydraulics and fracture mechanics.
27 28 This LTR does not need to address this criterion since Criterion 3 only applies to pressurized 29 water reactors (PWRs) and this LTR applies to BWRs.
30 31 GL 96-03, Attachment 1, Criterion 4:
32 33 Criterion 4 requires that the methodology describe the method for calculating the Adjusted 34 Reference Temperature (ART) using RG 1.99, Revision 2.
35 36 Table 1-1 in the BWROGs August 29, 2006, letter indicates this information is in Section 2.3 of 37 the LTR. Section 2.3 of the LTR describes the methodology documented in RG 1.99, 38 Revision 2 for calculating ART. Therefore this criterion has been satisfied.
39 40 GL 96-03, Attachment 1, Criterion 5:
41 42 Criterion 5 requires that the methodology describe the application of fracture mechanics in the 43 construction of P/T curves based on ASME Code Section XI, Appendix G, and SRP, 44 Section 5.3.2.
45 46 Table 1-1 in the BWROGs August 29, 2006, letter indicates this information is in Sections 2.3 47 and 2.4 of the LTR. However, the information is in Sections 2.4 and 2.5 of the LTR (Table 1-1 48 needs to be revised to include Section 2.5). Section 2.4 describes the general fracture 49 mechanics methodology for calculating P/T limit curves. Section 2.5 describes the 1
methodology for calculating P/T limits for the RPV beltline, bottom head region, and non-beltline 2
region. The non-beltline region includes all regions outside the beltline, excluding the bottom 3
head.
4 5
Section 2.4 provides fracture mechanics criteria based on ASME Code,Section XI, 6
Appendix G, and ASME Code Cases N-640 and N-641. These code cases allow the use of 7
the reference stress intensity factor, KIC, for calculating P/T limit curves. NRC Regulatory Issue 8
Summary 2004-04 (Reference 11) indicates the use of NRC-approved ASME Code cases in 9
conjunction with earlier versions of the ASME Code endorsed in 10 CFR 50.55a may also be 10 used for development of P/T limit curves without the need for an exemption. NRC RG 1.147, 11 Revision 14 (Reference 12) approves these ASME Code cases. The use of the reference 12 stress intensity factor, KIC, for calculating P/T limit curves was first endorsed by the 1999 13 Addenda of the ASME Code. Therefore, licensees utilizing this methodology and versions of 14 ASME Code,Section XI, Appendix G that require P/T limit curves to be calculated using KIC do 15 not need to request an exemption.
16 17 Section 2.5 describes the methodology for calculating P/T limits for the RPV beltline, bottom 18 head region, and non-beltline region. For the beltline shell region, this section describes three 19 methods for calculating the thermal stress intensity factor, KIt: a) a stress linearizing technique 20 presented in ASME Code,Section XI, Nonmandatory Appendix A; b) a technique based on 21 Section XI, Appendix G; and c) a technique based on Welding Research Council (WRC) 22 Bulletin Number 175 (Reference 13). In response to NRC staff RAI No. 3, the BWROG 23 changed the stress linearizing technique to the method in ASME Code,Section XI, Appendix G.
24 The allowable pressure is calculated using the methodology and structural factors in ASME 25 Code,Section XI, Appendix G. Since these techniques are based on methodologies endorsed 26 by the NRC, they are acceptable. The NRC staff requires that this change be incorporated into 27 the -A version of the LTR.
28 29 The LTR indicates that the methodology for the calculating bottom head P/T limit curves should 30 follow the methodologies for the shell region except that a stress concentration factor is applied 31 to bottom head membrane stresses to account for the stress concentration resulting from 32 nozzles in the lower head. In addition, the pressure stress is considered entirely as a 33 membrane stress. Appendix 5 in WRC Bulletin Number 175 describes methods for calculating 34 the stress intensity factors at the inside corner of a nozzle. The stress concentration factors 35 described in these analyses are less than those utilized by the BWROG for the development of 36 bottom head P/T limits. The methodology proposed by the BWROG for the bottom head has 37 been previously reviewed by the NRC staff in a letter from D. S. Collins (NRC) to R. G. Byram 38 (Senior Vice President and Chief Nuclear Officer for Susquehanna Steam Electric Station, 39 Units 1 and 2) dated February 7, 2002 (ADAMS Accession No. ML013520605). The NRC staff 40 performed independent calculations and concluded that the method is consistent with the 41 methods in the 1995 Edition of Appendix G to Section XI of the ASME Code. Based on the use 42 of a conservative concentration factor and the NRC staffs previous evaluation of this 43 methodology, the NRC staff concludes that the methodology proposed by the BWROG for the 44 calculating bottom head P/T limit curves is acceptable.
45 46 The non-beltline region analysis method that was contained in Section 2.5 has been deleted 47 and replaced with a methodology that is described in the BWROG response to RAI No. 3. In 48 this methodology the location to be analyzed for determining the highest stresses in the 49 non-beltline region is the feedwater nozzle. The reference temperature, RTNDT, used in the 1
analysis is the limiting value for all non-beltline nozzles. The stress intensity factors for the 2
feedwater nozzle may be calculated from a detailed finite element model analysis of the nozzle.
3 The stress distribution from the finite element analysis is fit with a third order polynomial. The 4
stress intensity factors for the nozzle corner use the coefficients from the stress distribution 5
polynomial and a method proposed in General Electric (GE) Topical Report NEDE-21821-02 6
(Reference 14) for calculating stress intensity factors for nozzle corner cracks. The stress 7
intensity factor solutions documented in Reference 14 were verified by independent analysis 8
and experiment. Reference 14 was approved by the NRC staff in a letter from D. G. Eisenhut 9
(NRC) to R. Gridley (GE) dated January 14, 1980 (ADAMS Legacy Library Accession 10 No. 8002070141). The NRC staff concluded that each step in the GE analysis is acceptable, 11 but had specific comments. Since none of the comments were directed at the stress intensity 12 solutions for the nozzle corner crack, the stress intensity solutions proposed were considered 13 acceptable for evaluating nozzle corner cracks. The proposed methodology uses the stress 14 intensity factors from both thermal and pressure stress to develop P/T limits based on the 15 structural factors described in Appendix G to Section XI of the ASME Code. The NRC staff 16 finds the non-beltline methodology acceptable since it meets the requirements of ASME Code, 17 Section XI, Appendix G and the stress intensity factors are determined using a previously 18 reviewed methodology. However, the NRC staff requires that the information provided in 19 response to RAI No. 3 be incorporated into the -A version of the LTR.
20 21 Section 2.5 of the LTR and the methodology proposed in response to RAI No. 3 describe 22 methodologies for calculating bending and membrane stresses using computer code finite 23 element analyses. If these finite element analyses are to be utilized by licensees to develop 24 P/T limits, the NRC staff requested, in RAI No. 2, that the BWROG provide the following:
25 26 a)
Identify the computer codes that were used in the finite element stress analysis. How 27 were the codes benchmarked?
28 29 b)
Discuss briefly the assumptions [initial RTNDT] and the inputs to the stress analysis.
30 31 c)
Update the topical report methodology to require licensees to identify the finite element 32 codes used in the PTLR.
33 34 d)
Verify that this process for determining bending and membrane stresses will result in the 35 generation of P/T limits that are at least as conservative as those which would be 36 generated using the procedures of ASME Code,Section XI, Appendix G.
37 38 In response to the NRC staff request to items a), b), and c), the BWROG proposed to add the 39 following text to the Section 2.5 of the LTR:
40 41 In the subsections that follow, finite element analysis is discussed as a possible 42 approach for providing the necessary stress analysis for RPV regions. If finite element 43 analysis is utilized to develop P-T limits for any RPV region, the following information 44 shall be provided in the PTLR:
45 46 a.
Identify the computer code(s) that were used in the finite element stress 47 analysis.
48 49 b.
For any computer codes used, describe how the code(s) were verified or 1
benchmarked. Computer code verification shall be in accordance with a 2
qualified 10 CFR 50 Appendix B Quality Assurance Program. As a part of 3
computer code verification, benchmarking consistent with NRC GL 83-11, 4
Supplement 1 [17] shall be included.
5 6
c.
Identify the assumptions and the inputs to the finite element analysis.
7 Necessary inputs to the analysis include any or all of the following:
8 9
A description of plant operating conditions used (e.g., pressure and 10 temperature). The conditions used must represent current plant 11 operating conditions.
12 13 A description of the heat transfer coefficients used and the methodology 14 used to calculate them.
15 16 A description of the model developed, including materials, material 17 properties, finite element mesh pattern, and geometry.
18 19 New Reference 17 will be added to Section 4.0 of the LTR as follows:
20 21 17.
U. S. Nuclear Regulatory Commission, Generic Letter 88-11, 22 Supplement 1, "Licensee Qualification for Performing Safety Analyses,"
23 June 24, 1999.
24 25 Since the LTR will require licensees to provide the requested information in the PTLR, the 26 response is acceptable.
27 28 For item d), the BWROG proposed that the linearization techniques discussed in the LTR be 29 removed and replaced with polynomial fit techniques that are consistent with current ASME 30 Code,Section XI, Appendix G methodology. The proposed technique is described in the 31 BWROG response to RAI No. 3. Since the linearization technique will be replaced with a 32 technique which is consistent with the current ASME Code,Section XI, Appendix G 33 methodology, the change is acceptable. Since Sections 2.4 and 2.5 identify fracture mechanics 34 methods for the construction of P/T curves based on ASME Code,Section XI, Appendix G, this 35 criterion has been satisfied.
36 37 GL 96-03, Attachment 1, Criterion 6:
38 39 Criterion 6 requires that the methodology describe how the minimum temperature requirements 40 in Appendix G to 10 CFR Part 50 are applied to P/T curves for boltup temperature and 41 hydrotest temperature.
42 43 Table 1-1 in the BWROGs August 29, 2006, letter indicates this information is in Sections 2.7 44 and 2.8 of the LTR. Section 2.7 identifies the minimum metal temperature of the RPV closure 45 head flange and the RPV shell flange regions. This section also describes the minimum 46 requirements for hydrotest (hydrostatic pressure and leak tests). Section 2.8 identifies the 47 minimum boltup temperatures. Both of these sections identify minimum temperature 48 requirements that are contained in Appendix G to 10 CFR Part 50. Since the information in 49 these sections comply with the requirements in Appendix G to 10 CFR Part 50, this criterion 1
has been satisfied.
2 3
GL96-03, Attachment 1, Criterion 7:
4 5
Criterion 7 requires that the methodology describe how the data from multiple surveillance 6
capsules are used in the ART calculation.
7 8
Table 1-1 of the BWROGs August 29, 2006, letter indicates this information is in Sections 2.3 9
of the LTR. Criteria for evaluating surveillance data are contained in Appendix A to the LTR.
10 (Table 1-1 needs to be revised to include Appendix A). Appendix A documents two procedures 11 for calculating the ART. One procedure is applicable when RPV material and surveillance 12 material have identical heat numbers. This method follows the methodology documented in 13 Position 2.1 of RG 1.99, Revision 2 and the NRC staff guidance presented in an NRC/Industry 14 Workshop (Reference 15). Position 2.1 in RG 1.99, Revision 2 contains NRC staff guidance for 15 evaluating surveillance data when there are two or more credible surveillance data points.
16 Credibility is determined following the guidance in RG 1.99, Revision 2.
17 18 The second procedure is applicable when the heat number for the surveillance material does 19 not match the heat number for the RPV material. In this case the ART is determined using the 20 guidance in Position 1.1 of RG 1.99, Revision 2. Position 1.1 in RG 1.99, Revision 2 contains 21 NRC staff guidance for determining the ART from the chemical composition (weight-percent 22 copper and nickel) of the RPV material.
23 24 The NRC staff recommended changes to these procedures in RAIs sent to the BWROG.
25 These changes are discussed in the evaluation of Appendix A, which is discussed in Section 26 3.3 of this SE. The changes to Appendix A are acceptable, because they provide additional 27 guidance to the licensees and the guidance has been previously approved by the NRC staff.
28 Based on the changes documented in Section 3.3 and that the procedures follow guidance 29 recommended by the NRC staff, this criterion has been satisfied.
30 31 3.2 Evaluation of Section 3.0 of the LTR 32 33 Section 3.0 of the LTR provides a step-by-step procedure for calculating P/T limit curves. This 34 section indicates that P/T limits may also be developed for other RPV regions to provide 35 additional operating flexibility. In response to RAI No. 5, the BWROG indicated that a sentence 36 in the LTR will be revised to state:
37 38 P-T limit curves may also be developed for other RPV regions to provide 39 additional operating flexibility; however, for RPV regions other than those defined 40 in Section 2.0 of this report, licensees are required to submit methodologies to 41 the NRC for review and approval prior to use.
42 43 Since methods of evaluating other regions will be submitted to the NRC for review and approval 44 prior to use, the proposed change is acceptable. The NRC staff requires that this modification 45 be incorportated into the -A version of the LTR.
46 47 The guidance given in Section 3.0 does not indicate surveillance data is to be evaluated. In 48 response to RAI No. 6, the BWROG indicated a new Step (a) will be added to Section 3.0 of the 49 LTR and the previously defined steps will be re-labeled as Steps (b) through (i). The proposed 1
new Step (a) follows:
2 3
- a.
Evaluate surveillance data in accordance with Appendix A of this report.
4 5
Appendix A provides guidance for the use of the Boiling Water Reactor Vessel and Internals 6
Project (BWRVIP) Integrated Surveillance Program (ISP) surveillance data. The BWRVIP ISP 7
replaces individual plant RPV surveillance capsule programs with representative weld and base 8
materials data from host reactors. A representative material is a plate or weld material that is 9
selected from among all the existing plant surveillance programs or the Supplemental 10 Surveillance Program (SSP) to represent one or more limiting plate or weld materials in a plant.
11 The BWRVIP ISP is responsible to provide each BWR plant with surveillance data for the 12 materials assigned to represent that plant's limiting RPV weld and base materials. Plant 13 owners, in turn are responsible to evaluate the data using the methods in RG 1.99, Revision 2, 14 in accordance with 10 CFR Part 50, Appendix G, for determination of ART values.
15 16 Since the LTR will be revised to indicate surveillance data is to be evaluated in accordance with 17 Appendix A and Appendix A contains criteria for processing and reporting surveillance data, the 18 proposed change is acceptable. The NRC staff requires that this change be incorporated into 19 the -A version of the LTR.
20 21 3.3 Evaluation of Appendix A of the LTR 22 23 Appendix A provides guidance for evaluating surveillance data. In response to NRC staff RAI 24 No. 7, Appendix A will be revised to identify the source for the best estimate chemistries for the 25 BWR vessel and surveillance capsule materials and to identify that the best estimate 26 chemistries will be documented in the PTLR. The BWROG response adds the following note 27 and reference to Appendix A:
28 29 Note: Revised best estimate chemistries for selected BWR vessel and 30 surveillance capsule materials have been calculated by the BWRVIP, as 31 documented in BWRVIP-86-A [A-1]. Calculation of the best estimate chemistries 32 for all other vessel materials should be determined in accordance with the NRC 33 practice documented in Reference [A-7]. The suggested practice is documented 34 in guidelines contained in BWRVIP-135. This evaluation is the responsibility of 35 the plant, must be described in the PTLR, and must utilize NRC-approved 36 methods.
37 38 New Reference A-7 will be added to Section A.5 of the LTR as follows:
39 40 A-7.
"Generic Letter 92-01 and RPV Integrity Assessment - Status, Schedule, 41 and Issues," Presentation by K. Wichman, M. Mitchell, and A. Hiser at 42 NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.
43 44 In response to NRC staff RAI No. 8, Appendix A will be revised to describe the temperature 45 adjustment to the surveillance data if the temperature of the surveillance capsule is different 46 than that of the vessel. Appendix A, Procedure 1, Procedural Step 3(b) of the LTR will be 47 revised as follows:
48 49 b.
If the vessel wall temperature is an outlier, appropriate temperature 1
adjustments to the surveillance data may be required. An appropriate 2
temperature adjustment is a 1 oF degree increase in RTNDT per 1oF 3
decrease in irradiation temperature [A-7]. Alternatively, the temperature 4
adjustment can be determined using appropriate NRC guidance. Any 5
temperature adjustments shall be identified and described in the PTLR.
6 7
In response to NRC staff RAI No. 9, Appendix A will be revised to define the initial RTNDT, as 8
follows:
9 10 Initial RTNDT is the reference temperature for the unirradiated material as defined 11 in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel 12 Code. Some plants have measured values of initial RTNDT; other plants use 13 generic values. For generic values of weld metal, the following generic mean 14 values must be used: 0°F for welds made with Linde 80 flux, and -56°F for welds 15 made with Linde 0091, 1092, and 124 and ARCOS B-5 weld fluxes [A-6]. Other 16 generic mean values may be used, provided they are justified and have NRC 17 review and approval. The generic mean values used shall be identified in the 18 PTLR.
19 20 Reference A-6 is the Pressurized Thermal Shock rule, 10 CFR 50.61. The rule provides 21 generic initial RTNDT values for welds made with Linde 80, 0091, 1092, and 124 and ARCOS B-22 5 weld fluxes. These values have been reviewed and approved by the NRC staff. Therefore, 23 they are also applicable for BWR RPVs.
24 25 In response to NRC staff RAI No. 10, Appendix A will be revised to identify information that the 26 licensee should review to determine whether the data is credible or non-credible in 27 accordance with RG 1.99, Revision 2. The following two steps will be added to Appendix A, 28 Procedure 1, Procedural Step 3 of the LTR:
29 30 d.
Scatter in the plots of Charpy energy versus temperature for the irradiated and 31 unirradiated conditions should be small enough to permit the determination of 32 the 30 foot-pound temperature and the upper shelf energy unambiguously.
33 34
- e.
When there are two or more sets of surveillance data from one reactor, the 35 scatter of RTNDT values about a best-fit line drawn as described in Reg.
36 Guide 1.99 Rev. 2, Regulatory Position 2.1, normally should be less than 28oF 37 for welds and 17oF for base metal. Even if the fluence range is large (two or 38 more orders of magnitude), the scatter should not exceed twice those values.
39 Even if the data fail this criterion for use in shift calculations, they may be 40 credible for determining decrease in upper-shelf energy if the upper shelf can be 41 clearly determined, following the definition given in ASTM E185-82.
42 43 The changes to Appendix A are acceptable, because they provide additional guidance to the 44 licensees and the guidance has been previously approved by the NRC staff. The NRC staff 45 requires that these changes to Appendix A of the LTR be incorporated into the -A version of the 46 report.
47 48 3.4 Evaluation of Appendix B of the LTR 1
2 Appendix B provides a template PTLR. To ensure that the P/T limits were developed using the 3
LTR methodology, the NRC staff in RAI No. 11 requested that the following information be 4
included in the PTLR:
5 6
a)
The initial RTNDT [IRTNDT] for all reactor pressure vessel materials and the method of 7
determining the initial RTNDT (i.e., ASME Code, Generic Communication, Branch 8
Technical Position - MTEB 5-2 in SRP 5.3.2 in NUREG-0800, or other NRC-approved 9
methodologies),
10 11 b)
The chemistry (weight-percent copper and nickel) and ART at the 1/4 thickness location 12 for all beltline materials, and 13 14 c)
The computer codes used in the finite element analysis to determine for calculating 15 bending and membrane stresses from Section 2.5 of the methodology.
16 17 d)
Identify whether Procedure #1" or Procedure #2" was utilized to evaluate the 18 surveillance data. If surveillance data was utilized, provide the surveillance data and the 19 analysis of the surveillance data that was used to determine the ART. If surveillance 20 data was not utilized, state why it was not utilized.
21 22 In response to NRC staff RAI No. 11 items (a), (b), and (d), the BWROG proposed that the 23 following be added to Section 2.3 of the LTR:
24 25 The following information should be included in the PTLR with respect to the ART 26 calculations:
27 28 a.
The IRTNDT for all RPV materials and the method of determining the IRTNDT 29 (i.e., ASME Code, Generic Communication, Branch Technical Position MTEB 5-2 30 in Standard Review Plan 5.3.2 in NUREG-0800, or other NRC-approved 31 methodologies).
32 33 b.
The chemistry (weight-percent copper and nickel) and ART at the 1/4t location 34 for all beltline materials.
35 36 c.
Identify whether "Procedure 1" or "Procedure 2" from Appendix A was utilized to 37 evaluate the surveillance data. If surveillance data was utilized, provide the 38 surveillance [data] and the analysis of the surveillance data that was used to 39 determine the ART values. If surveillance data was not utilized, state why it was 40 not utilized.
41 42 The changes are acceptable, because they provide additional guidance for licensees and 43 provide information that the NRC staff needs to evaluate the PTLR. The NRC staff requires 44 that these changes be incorporated into the -A version of the report.
45 46 The response to item c) was discussed in the Section 3.1 of this SE (Evaluation of GL 96-03, 47, Criterion 5). Section 2.5 will be revised to request that the PTLR contain the 48 requested information.
49
4.0 CONCLUSION
1 2
The NRC staff concludes that BWROG LTR SIR-05-044 satisfies the criteria in Attachment 1 to 3
GL 96-03 and provides adequate methodology for BWR licensees to calculate P/T limit curves.
4 By using this methodology and following the PTLR guidance in GL 96-03, as amended by NRC 5
TSTF-419, BWR licensees will be able to relocate the P/T limit curves and the associated 6
heatup/cooldown rates from TS to a PTLR, a licensee-controlled document.
7 8
The NRC staff has recommended, as noted in this SE, additional changes to Table 1-1 of the 9
LTR. The BWROG must incorporate the NRC staff recommended changes and the changes 10 proposed by the BWROG in their letter dated August 29, 2006, into the -A version of the report.
11 12
5.0 REFERENCES
13 14 1.
NRC Generic Letter 96-03, Relocation of the Pressure-Temperature Limit Curves and 15 Low Temperature Overpressure Protection System Limits, January 31, 1996 (ADAMS 16 Legacy Library Accession No. 9601290350).
17 18 2.
10 CFR Part 50, Appendix G, Fracture Toughness Requirements, 2005 Edition.
19 20 3.
ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, Fracture Toughness 21 Criteria for Protection Against Failure, 2004 Edition.
22 23 4.
NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel 24 Materials, May 1988 (ADAMS Accession No. ML003740284).
25 26 5.
NUREG-0800, NRC Standard Review Plan, Section 5.3.2, Pressure-Temperature 27 Limits and Pressurized Thermal Shock, Draft Revision 2, June 1996.
28 29 6.
10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program 30 Requirements, 2005 Edition.
31 32 7.
NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining 33 Pressure Vessel Neutron Fluence, March 2001 (ADAMS Accession 34 No. ML010890301).
35 36 8.
10 CFR 50.36, Technical specifications, 2005 Edition.
37 38 9.
NRC Technical Specification Traveler Form TSTF-419, Revision 2, Pressure 39 Temperature Limits Report [PTLR], September 16, 2001 (ADAMS Accession 40 No. ML012690234).
41 42 10.
10 CFR 50.90, Application for amendment of license or construction permit, 43 2005 Edition.
44 45 11.
NRC Regulatory Issue Summary 2004-04, Use of Code Cases N-588, N-640 and 46 N-641 in Developing Pressure-Temperature Operating Limits, April 5, 2004 (ADAMS 47 Accession No. ML040920323).
48 49 12.
NRC Regulatory Guide 1.147, Revision 14, Inservice Inspection Code Case 1
Acceptability, ASME Section XI, Division 1, August 2005 (ADAMS Accession 2
No. ML052510117).
3 4
13.
WRC Bulletin No. 175, Pressure Vessel Research Committee (PVRC) 5 Recommendations on Fracture Toughness Requirements for Ferritic Materials, 6
August 1972.
7 8
14.
GE Topical Report NEDE-21821-02, BWR Feedwater Nozzle/Sparger Final Report, 9
Supplement 2, August 1979.
10 11 15.
NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, K. Wichman, 12 M. Mitchell, and A. Hiser, NRC/Industry Workshop on RPV Integrity Issues, 13 February 12, 1998.
14 15 Principle Contributor: B. Elliot 16 17 Date: November 14, 2006 18