ML090630457

Draft Safety Evaluation for Boiling Water Reactors Owners Group Licensing Topical Report (LTR) NEDC-33178P, General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves

ML090630457
Person / Time
Site:	Boiling Water Reactor Owners Group
Issue date:	03/23/2009
From:	NRC/NRR/DPR/PSPB
To:	Energy Northwest
honcharik, M C, NRR/DPR, 415-1774
Shared Package
ML090640127	List: ML090630457 ML090630461
References
NEDC-33178P, TAC MD2693
Download: ML090630457 (15)

Text

ENCLOSURE 1

DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 2

3 TOPICAL REPORT NEDC-33178P 4

5 "GENERAL ELECTRIC METHODOLOGY FOR DEVELOPMENT OF REACTOR PRESSURE 6

7 VESSEL PRESSURE-TEMPERATURE CURVES" 8

9 BOILING WATER REACTORS OWNERS GROUP 10 11 PROJECT NO. 691 12 13 14

1.0 INTRODUCTION AND BACKGROUND

15 16 By letter dated July 28, 2006, the Boiling Water Reactor Owners' Group (BWROG) submitted 17 Licensing Topical Report (LTR) NEDC-33178P, "General Electric Methodology for Development 18 of Reactor Pressure Vessel Pressure-Temperature Curves," Revision 0 (Reference 1), for the 19 Nuclear Regulatory Commission (NRC) review and acceptance for referencing in subsequent 20 licensing actions. The BWROG provided this LTR to support applications by BWR licensees to 21 relocate their pressure-temperature (P-T) curves from facility technical specifications (TS) to a 22 pressure temperature limits report (PTLR), a licensee-controlled document, using the guidelines 23 provided in Generic Letter (GL) 96-03, Relocation of the Pressure Temperature Limit Curves 24 and Low Temperature Overpressure Protection System Limits, (Reference 2). Responses to 25 NRC staffs requests for additional information (RAIs) were provided in a BWROG letter dated 26 July 31, 2007 (Reference 3), which was later superseded by a revised version of the LTR. LTR 27 NEDC-33178P, Revision 1, incorporating the proposed changes was provided to the NRC in a 28 letter dated January 19, 2009 (Reference 4).

29 30

2.0 REGULATORY EVALUATION

31 32 2.1 Requirements for Generating P-T Limits for Light-Water Reactors 33 34 The NRC has established requirements in Appendix G of Title 10, Code of Federal Regulations 35 Part 50 (10 CFR Part 50, Appendix G; Reference 5), to protect the integrity of the reactor 36 coolant pressure boundary (RCPB) in nuclear power plants. The regulation at 10 CFR Part 50, 37 Appendix G requires that the P-T limits for an operating light-water nuclear reactor be at least as 38 conservative as those that would be generated if the methods of Appendix G to Section XI of 39 the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code, 40 Section XI, Appendix G; Reference 6), were used to generate the P-T limits. The regulation at 41 10 CFR Part 50, Appendix G, also requires that applicable surveillance data from reactor 42 pressure vessel (RPV) material surveillance programs be incorporated into the calculations of 43 plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a 44 method that accounts for the effects of neutron irradiation on the material properties of the RPV 45 beltline materials.

46 47 Table 1 to 10 CFR Part 50, Appendix G provides the NRC staffs criteria for meeting the P-T 1

limit requirements of ASME Code,Section XI, Appendix G, as well as the minimum temperature 2

requirements of the rule for bolting up the vessel during normal and pressure testing operations.

3 In addition, NRC staff regulatory guidance related to P-T limit curves is found in Regulatory 4

Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, 5

(Reference 7), and NUREG-0800, Standard Review Plan (SRP), Section 5.3.2, 6

Pressure-Temperature Limits and Pressurized Thermal Shock (Reference 8).

7 8

The regulation at 10 CFR Part 50, Appendix H (Reference 9), provides the NRC staffs criteria 9

for the design and implementation of RPV material surveillance programs for operating 10 light-water reactors.

11 12 In March 2001, the NRC issued RG 1.190, Calculational and Dosimetry Methods for 13 Determining Pressure Vessel Neutron Fluence (Reference 10). Neutron fluence calculations 14 are acceptable if they are performed with approved methodologies or with methods which are 15 shown to conform to the guidance in RG 1.190.

16 17 2.2 Technical Specification Requirements for P-T Limits 18 19 Section 182a of the Atomic Energy Act of 1954 requires applicants for nuclear power plant 20 operating licenses to include TS as part of the license. The Commission's regulatory 21 requirements related to the content of TS are set forth in 10 CFR 50.36 (Reference 11). This 22 regulation requires that the TS include items in five specific categories: (1) safety limits, limiting 23 safety system settings and limiting control settings; (2) limiting conditions for operation (LCOs);

24 (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.

25 26 The regulation at 10 CFR 50.36(c)(2)(ii) requires that LCOs be established for the P-T limits 27 because the parameters fall within the scope of the Criterion 2 identified in the rule:

28 29 Criterion 2: A process variable, design feature, or operating restriction that is an 30 initial condition of a design basis accident or transient analysis that either 31 assumes the failure of or presents a challenge to the integrity of a fission product 32 barrier.

33 34 The P-T limits for BWRs fall within the scope of Criterion 2 of 10 CFR 50.36(c)(2)(ii) and were 35 therefore required to be included within the TS LCOs for a plant-specific facility operating 36 license. On January 31, 1996, the NRC staff issued GL 96-03 to inform licensees that they may 37 request a license amendment to relocate the P-T limit curves and/or low temperature 38 over-pressure protection (LTOP) limit setpoint values from the TS LCOs into a PTLR or other 39 licensee-controlled document that would be controlled through the Administrative Controls 40 Section of the TS. In GL 96-03, the NRC staff informed licensees that, in order to implement a 41 PTLR, the P-T limit curves and LTOP limits for U.S. licensed light-water reactors would need to 42 be generated in accordance with an NRC-approved methodology and that the methodology to 43 generate the P-T limit curves and LTOP limits would need to comply with the requirements of 44 10 CFR Part 50, Appendices G and H; be documented in an NRC-approved topical report or 45 plant-specific submittal; and be incorporated by reference in the Administrative Controls Section 46 of the TS. The GL also mandated that the TS Administrative Controls Section would need to 47 reference the NRC staffs safety evaluation (SE) issued on the PTLR request and that the PTLR 48 be defined in Section 1.0 of the TS. Attachment 1 to GL 96-03 provided a list of the criteria that 1

the approved methodology and PTLR would be required to meet.

2 3

Technical Specification Task Force (TSTF) Traveler No. TSTF-419, Revise PTLR Definition 4

and References in ISTS [Improved Standard Technical Specifications] 5.6.6, RCS PTLR 5

(Reference 12) amended the Standard Technical Specifications (STS) (NUREGs-1430, -1431, 6

-1432, -1433, and -1434) by: (1) deleting references to the TS LCO specifications for the P-T 7

limits and LTOP system limits in the TS definition of the PTLR, and (2) revising STS 5.6.6 to 8

identify, by number and title, NRC-approved topical reports that document PTLR methodologies, 9

or the NRC SE for a plant-specific methodology by NRC letter and date. A requirement was 10 added to the reviewers note to specify the complete citation of the PTLR methodology in the 11 plant-specific PTLR, including the report number, title, revision, date, and any supplements.

12 13 Only the figures, values, and parameters associated with the P-T limits and LTOP system limits 14 are relocated to the PTLR. The methodology for their development must be reviewed and 15 approved by the NRC. TSTF-419 did not change the requirements associated with the review 16 and approval of the methodology or the requirement to operate within the limits specified in the 17 PTLR. Any changes to a methodology that had not been approved by the NRC staff would 18 continue to require NRC staff review and approval pursuant to the license amendment request 19 provisions and requirements of 10 CFR 50.90 (Reference 13).

20 21

3.0 TECHNICAL EVALUATION

22 23 As stated in Section 2.1 of this SE, 10 CFR Part 50, Appendix G requires that licensees 24 establish limits on the pressure and temperature of the RCPB to protect it against brittle failure.

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) and system hydrostatic tests.

27 28 BWROG LTR NEDC-33178P, Revision 0 has six sections, nine appendices and two 29 attachments. Section 1.0 provides the introduction and the purpose for the LTR. Section 2.0 30 provides the scope of the analysis, and Section 3.0 refers to Attachment 1 for the assumptions 31 for the plant-specific P-T analysis. Section 4.0 describes the analysis methods for developing 32 P-T limits. Section 5.0 provides conclusions and recommendations and Section 6.0 provides 33 references. Attachment 1 provides an example of a P-T curve report template. Attachment 2 34 provides an example of a PTLR. Appendices A through H provide background information used 35 for performing the analyses described in Section 4.0 of the LTR. Appendix I provides guidance 36 for evaluating surveillance data.

37 38 3.1 Evaluation of Section 4.0 of the LTR 39 40 The NRC staffs evaluation of Sections 4.1 through 4.3 of the LTR is based on the criteria 41 contained in Attachment 1 of GL 96-03. Attachment 1 of GL 96-03 contains seven technical 42 criteria that the contents of proposed methodology should conform to for PTLRs acceptable to 43 the NRC staff. The NRC staffs evaluations of the contents of BWROG methodology against the 44 seven criteria in Attachment 1 of GL 96-03 are given below.

45 GL 96-03, Attachment 1 Methodology Criterion 1 1

2 Methodology Criterion 1 requires that the methodology describe the transport calculation 3

methods including computer codes and formulas used to calculate neutron fluence.

4 5

The GL 96-03 conformance table in the BWROGs November 15, 2007, letter indicates this LTR 6

does not describe the transport calculation methods including computer codes and formulas 7

used to calculate neutron fluence. However, Section 4.2.1.2 of the LTR indicates that the 8

neutron fluence will be determined using an approved methodology consistent with RG 1.190.

9 Further, this section indicates the neutron fluence is defined in Section 4.2.1.2 of Attachment 1 10 and Appendix B of the PTLR. Section 4.2.1.2 of Attachment 1 requires the licensee to identify 11 the report used to calculate the neutron fluence and to document that the plant-specific neutron 12 fluence calculation will be performed using an approved neutron fluence calculation 13 methodology. Therefore, this will be a plant-specific action item to be addressed by licensees.

14 Since the LTR methodology indicates that the neutron fluence calculation methodology must 15 comply with RG 1.190 and have been approved by the NRC, this criterion has been satisfied.

16 17 GL 96-03, Attachment 1 Methodology Criterion 2 18 19 Methodology Criterion 2 requires that the methodology describe the surveillance program and 20 indicates that the PTLR should contain a place holder for the requested information.

21 22 The GL 96-03 conformance table in the BWROGs November 15, 2007, letter indicates this 23 information is in Section 4.2.2 of the LTR. This section indicates that the BWR integrated 24 surveillance program is applicable to each BWR reactor vessel and is described in the 25 BWRVIP-102 report, BWR Vessel and Internals Project Integrated Surveillance Program 26 Implementation Guidelines, and the BWRVIP-135 report, BWR Vessels and Internals 27 Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. Since these 28 BWRVIP reports describe the BWR integrated surveillance program, this criterion has been 29 satisfied.

30 31 GL 96-03, Attachment 1 Methodology Criterion 3 32 33 Methodology Criterion 3 requires that the methodology describe how the LTOP system limits 34 are calculated applying system/thermal hydraulics and fracture mechanics.

35 36 This methodology does not need to address this criterion since it only applies to pressurized 37 water reactors (PWRs) and the methodology applies to BWRs.

38 39 GL 96-03, Attachment 1 Methodology Criterion 4 40 41 Methodology Criterion 4 requires that the methodology describe the method for calculating the 42 adjusted reference temperature (ART) using RG 1.99, Revision 2.

43 44 Sections 4.1 and 4.2 describe the method for determining the material properties for reactor 45 vessel beltline and non-beltline region materials. The unirradiated reference temperature (initial 46 RTNDT) is determined using the method described in ASME Code,Section III, Subsection 47 NB-2300, where sufficient data is available. If insufficient data is available, the initial RTNDT is 48 determined using the methodology described in GENE NEDC-32399-P (Reference 14). This 49 methodology was approved by the NRC on December 16, 1994 (Reference 15). The ART, an 1

indirect measure of the RPV material fracture toughness, is determined using the methodology 2

described in RG 1.99, Revision 2. The ART is defined in the RG as the sum of the initial RTNDT, 3

the shift in reference temperature caused by irradiation (RTNDT), and the margin term.

4 5

Section 4.1.2, Values of Initial RTNDT and Lowest Service Temperature (LST), indicates:

6 7

Where the lowest energy Charpy value is less than 50 ft-lb, it is adjusted by 8

adding 2 °F per ft-lb energy difference from 50 ft-lb. If the test specimens are 9

transverse and the lowest energy Charpy value is less than 50 ft-lb, it is adjusted 10 by adding 3 °F per ft-lb energy difference from 50 ft-lbs.

11 12 The NRC staff noted that the second sentence in the above statement is inconsistent with the 13 example that follows in this section.

14 15 The response to NRC staff RAI 10 in the BWROGs November 15, 2007, letter indicated the 16 following:

17 18 The example presented represents only the longitudinal specimen method. To 19 further clarify, a second example for the plate material will be added to this 20 section to demonstrate the methodology for a transverse specimen. The 21 methodology used for the transverse specimens is consistent with that 22 methodology defined in NEDC-32399P; this additional process was added to 23 account for older plants, where all of the ASME Code requirements were not met.

24 25 In the same response, the BWROG proposed to add the following to clarify this section 26 of the LTR:

27 28 A second example, for a plate material based upon transverse specimens, is 29 seen below.

30 31 The lowest Charpy energy and test temperature from the CMTRs [Certified 32 Material Test Reports] are 47 ft-lb and 13 °F. The estimated transverse 50 ft-lb 33 test temperature is:

34 35 T50T = 10 °F + [(50 - 47) ft-lb* 3 °F/ft-lb] = 19 °F 36 37 The initial RTNDT is the greater of NDT [nil-ductility temperature] or (T50T - 60 °F).

38 39 T50T - 60 °F = 19 °F - 60 °F = -41 °F 40 41 Dropweight testing to establish NDT for plate material is listed in the CMTR; the 42 NDT for this material is -20 °F. Therefore, the initial RTNDT for this plate heat is -

43 20 °F.

44 45 Since the proposed change to the LTR is consistent with the methodology approved by the NRC 46 staff, it is acceptable. The NRC staff verified that the change has been implemented in the 47 revised LTR dated December 2008.

48 The margin term that is defined in RG 1.99, Revision 2, is dependent upon the standard 1

deviation for the initial RTNDT (I) and the standard deviation for the RTNDT (). Section 4.2.1 2

of the LTR indicates:

3 4

The margin term as described above, is defined in RG 1.99: this methodology 5

is used except when Integrated Surveillance Program data from 6

BWRVIP-135is available, and BWRVIP-102methods are applied.

7 8

In response to NRC staff RAI 11, the BWROG indicated the following in its November 15, 2007, 9

letter:

10 11 This statement was intended to indicate that RG 1.99 is to be used to determine 12 the margin term, and that the procedures of BWRVIP-102 should be followed to 13 incorporate surveillance data from the ISP.

14 15 In the same response, the BWROG proposed to add the following to clarify this section 16 of the LTR:

17 18 The margin term, as described above, is defined in RG 1.99. When 19 Integrated Surveillance Program data from BWRVIP-135are available, 20 BWRVIP-102provides guidance with respect to applying the requirements of 21 RG 1.99 to this data. Appendix I of this report also contains guidance regarding 22 the application of surveillance data.

23 24 The proposed change and explanation provide the necessary clarification to ensure that 25 RG 1.99, Revision 2, is properly applied by licensees. The NRC staff verified that the change 26 has been implemented in the revised LTR dated December 2008.

27 28 Since the GL 96-03 conformance table in the BWROGs November 15, 2007, letter indicates the 29 information for calculating the ART is in Section 4.2 of the LTR, and this section describes the 30 methodology documented in RG 1.99, Revision 2, this criterion has been satisfied.

31 32 GL 96-03, Attachment 1 Methodology Criterion 5 33 34 Methodology Criterion 5 requires that the methodology describe the application of fracture 35 mechanics in the construction of P-T curves based on ASME Code Section XI, Appendix G, and 36 SRP Section 5.3.2.

37 38 The GL 96-03 conformance table in the BWROGs November 15, 2007, letter indicates this 39 information is in Section 4.3 of the LTR. This section of the report describes the methodology 40 for developing P-T curves for the lower vessel region, the upper vessel region, the core beltline 41 region, and the closure flange region of the RPV.

42 43 The lower vessel region analyses evaluate the materials in the bottom head, control rod drive 44 (CRD) penetrations and the nozzles, skirt, attachments to the bottom head (Tables 4-5a and 45 4-5b in the LTR). The bottom head analysis for pressure and leak test conditions is described in 46 Section 4.3.2.1.1 of the LTR. The limit for the coolant temperature rate change for the pressure 47 and leak test is 20 °F/hour. The bottom head analysis for core not critical heatup/cooldown 48 conditions is described in Section 4.3.2.1.2 of the LTR. The core not critical P-T limit curves 49 were developed based on assumed 100 °F/hour heatup/cooldown rates and bounding bottom 1

head transients defined on the plant-specific RPV thermal cycle and nozzle thermal cycle 2

diagrams.

3 4

The upper vessel region analyses evaluate materials in the upper shell, closure flange and the 5

nozzles, skirts and attachments to the upper shell (Tables 4-4a and 4-4b in the LTR). The 6

upper vessel region analysis for pressure and leak test conditions is described in Section 7

4.3.2.1.3 of the LTR. The upper shell region analysis for core not critical heatup/cooldown 8

conditions is described in Section 4.3.2.1.4 of the LTR. The core not critical P-T limit curves 9

were developed from the bounding feedwater transients defined on the plant-specific RPV 10 thermal cycle and nozzle thermal cycle diagrams.

11 12 The core beltline region analyses evaluate materials in the shell region of the RPV that are 13 adjacent to the active fuel, such that the neutron fluence is sufficient to cause a significant shift 14 of the RTNDT (i.e., at a neutron fluence exceeding 1.0E17 n/cm2 (E>1MeV)). The beltline region 15 analysis for pressure and leak test conditions is described in Sections 4.3.2.2.1 and 4.3.2.2.2 of 16 the LTR. The beltline region analysis for core not critical heatup/cooldown conditions is 17 described in Sections 4.3.2.2.3 and 4.3.2.2.4 of the LTR. Appendix E describes the method to 18 determine whether there are any RPV discontinuities be included in the beltline region.

19 20 The closure flange region analyses evaluate materials in the top head and shell closure flanges 21 in the RPV. The closure flange region analysis for pressure and leak test conditions and the 22 closure flange region analysis for core not critical heatup/cooldown conditions are described in 23 Section 4.3.2.3 of the LTR.

24 25 In addition to the fracture mechanics analyses documented in Section 4.3, Appendices F, G, 26 and H contain detailed fracture mechanics analyses for various RPV regions. Appendix F 27 describes the supplemental analysis performed for nozzles that are within the beltline region.

28 Appendix G provides supplemental analyses for thickness transition discontinuities between the 29 bottom head, lower torus and the upper torus and thickness transition discontinuities in the 30 beltline region. Appendix H provides a supplemental analysis for the bottom head CRD 31 penetrations.

32 33 The analyses described in Section 4.3 and Appendices F, G, and H of the LTR were performed 34 to satisfy the requirements in 10 CFR Part 50, Appendix G and Appendix G to Section XI of the 35 ASME Code. The Edition and Addenda of ASME Code,Section XI used in the plant-specific 36 evaluation will be specified in the plant-specific report provided to the licensee and the PTLR.

37 The methodology includes the following: 1) the use of KIc from Figure A-4200-1 of Appendix A 38 to ASME Code,Section XI and based on T-RTNDT, and 2) the use of the Mm calculation in ASME 39 Code,Section XI, Paragraph G-2214.1 for a postulated defect normal to the direction of 40 maximum stress. In its November 15, 2007, letter, the BWROG indicated in its response to 41 NRC staff RAI 1a that the methodology described in Section 4.3 and Appendix F has been 42 reported in P-T curve reports for Columbia Generating Station (Reference 16), Duane Arnold 43 Energy Center (Reference 17) and LaSalle County Station, Units 1 and 2 (References 18 44 and 19, respectively). At the time these reports were prepared, the beltline nozzle methodology 45 was not presented in a separate appendix; however, the methodology discussion was included 46 in the report text. The NRC staff approved the P-T curves for Columbia Generating Station, 47 Duane Arnold Energy Center and LaSalle County Station, Units 1 and 2 in SEs that are 48 documented in References 20, 21 and 22, respectively.

49 In the November 15, 2007, letter, the BWROG indicated that the methodology described in 1

Appendix G of the LTR has been reported in Appendix G of the P-T curve reports for Columbia 2

Generating Station, Fermi, Unit 2 (Reference 23), and LaSalle County Station, Unit 1. The NRC 3

staff approved the P-T curves for Fermi, Unit 2 in a SE that is documented in Reference 24.

4 5

In the November 15, 2007, letter, the BWROG indicated that the methodology described in 6

Appendix H of the LTR has been reported in Appendix F of Dresden Nuclear Power Station, 7

Units 2 and 3 (References 25 and 26, respectively) and Quad Cities Nuclear Power Station, 8

Units 1 and 2 (References 27 and 28, respectively) and Appendix G of the P-T limit curve report 9

for LaSalle County Station, Units 1 and 2. The NRC staff approved the P-T curves for Dresden 10 Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 in 11 an SE that is documented in Reference 29.

12 13 Since the methodologies contained in Section 4.3 and Appendices F, G, and H of the LTR have 14 been previously reviewed and approved by the NRC staff in the aforementioned plant-specific 15 reviews, the BWROGs fracture mechanics analyses are acceptable for utilization in calculating 16 P-T limit curves. Hence, this criterion has been satisfied.

17 18 GL 96-03, Attachment 1 Methodology Criterion 6 19 20 Methodology Criterion 6 requires that the methodology describe how the minimum temperature 21 requirements in Appendix G to 10 CFR Part 50 are applied to P-T limit curves.

22 23 The GL 96-03 conformance table in the BWROGs November 15, 2007, letter indicates this 24 information is in Section 4.3 of the LTR. Table 4-3 in the LTR identifies the 10 CFR Part 50, 25 Appendix G requirements for pressure and leak test, normal operation (heatup and cooldown, 26 including anticipated operational occurrences) with the core not critical, and operation with the 27 core critical conditions. As discussed under Criterion 5, the P-T limits for pressure and leak test 28 and heatup and cooldown, including anticipated operational occurrences with the core not 29 critical conditions have been previously evaluated and have satisfied the minimum temperature 30 requirements in Appendix G to 10 CFR Part 50.

31 32 The core critical operation condition evaluation is described in Section 4.3.2.4 of the LTR. This 33 evaluation satisfies the minimum temperature requirements of Appendix G to 10 CFR Part 50.

34 Hence, this criterion has been satisfied.

35 36 GL 96-03, Attachment 1 Methodology Criterion 7 37 38 Methodology Criterion 7 requires that the methodology describe how the data from multiple 39 surveillance capsules are used in the ART calculation.

40 41 The GL 96-03 conformance table in the BWROGs November 15, 2007, letter indicates this 42 information is in Section 4.2 and Appendix I of the LTR. Section 4.2 does not indicate that 43 surveillance data is to be evaluated in accordance with Appendix I. In response to NRC staff 44 RAI 4, the BWROG stated that Section 4.2 will be revised to indicate:

45 46 Surveillance material information, where applicable, shall be evaluated in 47 accordance with Section 4.2.2 and Appendix I.

48 49 This revision clarified the BWROGs evaluation of surveillance data. The NRC staff confirmed 1

that the revised LTR dated December 2008, has incorporated the stated change.

2 3

Appendix I contains the guidance for use of BWRVIP ISP surveillance data. This guidance 4

document indicates:

5 6

If there is new surveillance data for any heat that is located in the vessel beltline 7

(e.g., heat numbers match), then [Procedure 1] can be used as a guide for 8

evaluating the new information. A new Adjusted Reference Temperature (ART) 9 should be calculated for the vessel material to determine whether plant vessel 10 integrity evaluations are affected.

11 12 If there is new information but that same heat number is not contained in the 13 vessel beltline, then [Procedure 2] can be used as a guide for evaluating the new 14 information.

15 16 Procedure 1 follows the methodology documented in Position 2.1 of RG 1.99, Revision 2, and 17 the NRC staff guidance presented by the NRC staff in an NRC/Industry workshop 18 (Reference 30). Position 2.1 in RG 1.99, Revision 2, contains NRC staff guidance for evaluating 19 surveillance data when there are two or more credible surveillance data points. Credibility is 20 determined following the guidance in RG 1.99, Revision 2.

21 22 Procedure 2 is applicable when the heat number for the surveillance material does not match 23 the heat number for the RPV material. In this case the ART is determined using the guidance in 24 Position 1.1 of RG 1.99, Revision 2. Position 1.1 in RG 1.99, Revision 2, contains NRC staff 25 guidance for determining the ART based on the chemical composition (weight-percent copper 26 and nickel) of the RPV material.

27 28 The NRC staff identified issues with the procedures specified in Appendix I via RAIs sent to the 29 BWROG. The BWROG responded with proposed changes to Appendix I of the LTR. These 30 changes, as discussed below in the evaluation of Appendix I, are acceptable because they 31 provide additional guidance to the licensees and the guidance has been previously approved by 32 the NRC staff. Based on the changes documented in Section 3.2 of this SE and the fact that 33 the procedures follow guidance recommended by the NRC staff, this criterion has been 34 satisfied.

35 36 3.2 Evaluation of Appendix I of the LTR 37 38 Appendix I provided guidance for the use of the BWRVIP ISP surveillance data. The BWRVIP 39 ISP replaced individual plant RPV surveillance capsule programs with representative weld and 40 base materials data from host reactors. A representative material is a plate or weld material 41 that is selected from among all the existing plant surveillance programs or the Supplemental 42 Surveillance Program (SSP) to represent one or more limiting plate or weld materials in a plant.

43 The BWRVIP ISP is responsible for providing each BWR plant with surveillance data for the 44 materials assigned to represent that plant's limiting RPV weld and base materials. Plant 45 owners, in turn, are responsible for evaluating the data using the methods in RG 1.99, 46 Revision 2, in accordance with 10 CFR Part 50, Appendix G, for the determination of ART 47 values. Procedure 1 in the original LTR, as discussed above did not require that the 48 surveillance data meet all the criteria in RG 1.99, Revision 2, in determining the credibility of the 49 data. In response to NRC staff RAI 5, the BWROGs November 15, 2007, letter indicated that 1

the following two steps will be added to Step 3 in Procedure 1 of Appendix I:

2 3

d) Scatter in the plots of Charpy energy versus temperature for the irradiated 4

and unirradiated conditions should be small enough to permit the determination 5

of the 30 foot-pound temperature and the upper shelf energy unambiguously.

6 7

e) When there are two or more sets of surveillance data from one reactor, the 8

scatter of DRTNDT [(RTNDT)] values about a best-fit line drawn as described in 9

Regulatory Guide, Revision 2, Regulatory Position 2.1, normally should be less 10 than 28 °F for welds and 20 °F for base metal. Even if the fluence range is large 11 (two or more orders of magnitude), the scatter should not exceed twice those 12 values. Even if the data fail this criterion for use in shift calculations, they may be 13 credible for determining decrease in upper shelf energy if the upper shelf can be 14 clearly determined, following the definition given in ASTM E185-82.

15 16 Procedure 1 in the original LTR also did not contain an adequate description of the criteria to be 17 used if the vessel wall temperature is an outlier. In response to NRC staff RAIs 6 and 11, the 18 BWROG proposed to revise Procedure 1, Step 3(b), as follows, and add a reference to 19 available NRC staff guidance [Reference 60] to the LTR:

20 21 b) If the vessel wall temperature is an outlier, appropriate temperature 22 adjustments to the surveillance data may be required. An appropriate 23 temperature adjustment is a 1 °F increase in RTNDT per 1 °F decrease in 24 irradiation temperature [7]. Any temperature adjustments shall be identified and 25 described in the PTLR.

26 27 Note that Reference 30 to this SE is equivalent to the Reference 20 which was noted in the 28 revised text above and which the BWROG proposed to add to Section 6.0 of the LTR.

29 30 Procedures 1 and 2 from the original LTR did not provide an adequate description of the 31 determination of initial RTNDT. In response to NRC staff RAI 7, the BWROG proposed to revise 32 the Definitions and Background section of Procedures 1 and 2 as follows:

33 34 Initial RTNDT is the reference temperature for the unirradiated materials as defined 35 in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel 36 Code. Some plants have measured values of initial RTNDT; other plants use 37 generic values. For generic values of weld metal, the following generic mean 38 values must be used: 0 °F for welds made with Linde 80 flux, and - 56 °F for 39 welds made with Linde 0091, 1092, and 154 and ARCOS B-5 weld fluxes [6].

40 Other generic mean values may be used, provided they are justified and have 41 NRC review and approval. The generic mean values used shall be identified in 42 the PTLR.

43 44 Procedures 1 and 2 from the original LTR did not provide an adequate description of how to 45 determine the best estimate chemistry of a material. In response to NRC staff RAI 8, the 46 BWROG proposed to revise the note in Step 5 of Procedure 1 and Step 3 of Procedure 2 as 47 follows:

48 49 Note: Revised best estimate chemistries for selected BWR vessel and 1

surveillance capsule materials have been calculated by the BWRVIP, as 2

documented in BWRVIP-86-A [1]. Calculation of the best estimate chemistries 3

for all other vessel materials should be determined in accordance with the NRC 4

practice documented in [7]. The suggested practice is documented in guidelines 5

contained in BWRVIP-135. This evaluation is the responsibility of the plant, must 6

be described in the PTLR, and must utilize NRC-approved methods.

7 8

Based on the proposed changes to Procedures 1 and 2 in Appendix I of the LTR, the NRC staff 9

determined that Appendix I of the revised LTR, NEDC-33178, Revision 1, contains sufficient 10 information for licensees to evaluate surveillance data and the ART for the limiting beltline 11 material, in accordance with RG 1.99, Revision 2. The NRC staff verified that the change has 12 been implemented in the revised LTR dated December 2008.

13 14 3.3 Evaluation of Attachment 2 of the LTR 15 16 of the LTR provides a template PTLR. To ensure that the P-T limits were 17 developed using the LTR methodology, the NRC staff, in NRC RAI 9, requested that the 18 following information be included in the PTLR:

19 20 a)

The method of determining the initial RTNDT (i.e., ASME Code, Generic, Branch 21 Technical Position - MTEB 5-2 in SRP 5.3.2 in NUREG-0800, or other NRC 22 approved methodologies),

23 24 b)

The computer codes used in the finite element analysis to determine bending 25 and membrane stresses, 26 27 c)

Identify whether Procedure 1 or Procedure 2 was utilized to evaluate the 28 surveillance data. If surveillance data was utilized, provide the surveillance data 29 and the analysis of the surveillance data that was used to determine the ART. If 30 surveillance data was not utilized, state why it was not utilized, and 31 32 d)

Identify whether any of the P-T limit curves were adjusted to bound the analyses 33 documented in Section 4.3 of the LTP or in accordance with Attachment 1, 34 Appendix G. Identify the required adjustment in each P-T curve.

35 36 In its November 15, 2007, response to NRC staff RAI 9, the BWROG proposed that the 37 following be added to Section 5 of the template PTLR:

38 39 The method for determining the initial RTNDT for all vessel materials is that 40 defined in Section 4.1.2 of Reference 6.2. [Any deviations from this methodology 41 are discussed below.] Initial RTNDT values for all vessel materials considered are 42 presented in tables in this PTLR.

43 44 No new computer codes have been used in the development of the P-T curves.

45 46 OR 47 48 The following computer codes, which are not described in the topical report, have 1

been used in developing the P-T curves for [PLANT NAME].

2 3

For [PLANT NAME], the limiting material [HEAT #] considered Procedure [1]

4 defined in Appendix I of Reference 6.2. This procedure was used because [the 5

vessel material and the surveillance material are identical heats]. [If surveillance 6

data was utilized, provide the surveillance data and the analysis of the 7

surveillance data that was used to determine the adjusted reference temperature 8

(ART). If surveillance data was not utilized, state why it was not utilized.]

9 10 For [PLANT NAME], there is a thickness discontinuity in the vessel [between the 11 bottom head torus and dollar plate]. The P-T curves defined in Section 4.3 of 12 Reference 6.2 are based upon an RTNDT of [XXX] °F.

13 14 Based on the proposed changes to the template PTLR in Attachment 2, the NRC staff 15 determined that the template PTLR contains sufficient information for the NRC staff to perform 16 an independent evaluation of the P-T curves in accordance with RG 1.99, Revision 2 and the 17 fracture mechanics methodology described in Section 4.3 of the LTR. The NRC staff verified 18 that the change has been implemented in the revised LTR dated December 2008.

19 20 4.0 LIMITATIONS AND CONDITIONS 21 22 As documented in Section 3.1 of this SE, licensees who chose to implement NEDC-33178, 23 Revision 1 as their facilitys PTLR methodology must address one plant-specific action item:

24 25 The licensee must identify the report used to calculate the neutron fluence and 26 document that the plant-specific neutron fluence calculation will be performed 27 using an approved neutron fluence calculation methodology.

28 29 Information to address this licensee action item must be submitted with the licensees 30 requested license amendment to implement a PTLR for its facility.

31 32

5.0 CONCLUSION

33 34 The NRC staff concludes that BWROG LTR NEDC-33178P, Revision 1, satisfies the 35 criteria in Attachment 1 in GL 96-03 and provides adequate methodology for BWR 36 licensees to calculate P-T limit curves, given that licensees referencing this LTR comply 37 with the conditions listed in Section 4.0 of this SE. Using this methodology and following 38 the PTLR guidance in GL 96-03, as amended by NRC TSTF-419, BWR licensees will be 39 able to relocate the P-T limit curves from TS to a PTLR, a licensee-controlled document.

40 41

6.0 REFERENCES

42 43

1.

Boiling Water Reactor Owners' Group (BWROG) LTR NEDC-33178P, "General Electric 44 Methodology for Development of Reactor Pressure Vessel Pressure-Temperature 45 Curves", Revision 0, July 28, 2006. Agencywide Documents Access and Management 46 System (ADAMS) Accession No. ML062130323.

47 48

2.

NRC Generic Letter 96-03, Relocation of the Pressure-Temperature Limit Curves 1

and Low Temperature Overpressure Protection System Limits, January 31, 1996.

2 3

3.

Responses to Requests for Additional Information Regarding the BWROG 4

Submittal of General Electric Nuclear Energy (GENE) Licensing Topical Report 5

NEDC-33178P, Revision 0, "General Electric Methodology for Development of 6

Reactor Pressure-Temperature Curves," July 31, 2007. ADAMS Accession 7

No. ML072180598.

8 9

4.

Boiling Water Reactor Owners' Group (BWROG) LTR NEDC-33178P, "General 10 Electric Methodology for Development of Reactor Pressure Vessel Pressure-11 Temperature Curves", Revision 1, January 19, 2009. ADAMS Accession 12 No. ML090230247.

13 14

5.

Regulation 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, 15 2005 Edition.

16 17

6.

ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, Fracture 18 Toughness Criteria for Protection Against Failure, 2004 Edition.

19 20

7.

NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor 21 Vessel Materials, May 1988.

22 23

8.

NUREG-0800, NRC Standard Review Plan, Section 5.3.2, Pressure-Temperature 24 Limits and Pressurized Thermal Shock, Draft Revision 2, April 1996.

25 26

9.

Regulation 10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance 27 Program Requirements, 2005 Edition.

28 29

10.

NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining 30 Pressure Vessel Neutron Fluence, March 2001.

31 32

11.

Regulation 10 CFR 50.36, Technical Specifications, 2005 Edition.

33 34

12.

NRC Technical Specification Traveler Form TSTF-419, Revision 2, Revise PTLR 35 Definition and References in ISTS 5.6.6, RCS PTLR, September 16, 2001.

36 37

13.

Regulation 10 CFR 50.90, Application for Amendment of License or Construction 38 Permit, 2005 Edition.

39 40

14.

GE Nuclear Energy, NEDC-32399-P, Basis for GE RTNDT Estimation Method, Report 41 for BWR Owners, Group, San Jose, California, September 1994 (GE Proprietary).

42 43

15.

Letter from B. Sheron to R. A. Pinelli, Safety Assessment of Report NEDC-32399-P, 44 Basis for GE RTNDT Estimation Method, September 1994, USNRC, December 16, 1994.

45 46

16.

K. V. Norton and L. J. Tilly, Pressure-Temperature Curves for Energy Northwest 47 Columbia, GE Nuclear Energy, San Jose, CA, NEDC-33144P, Revision 0, April 2004 48 (GE Proprietary Information).

49 1

17.

M. C. OConnor, Pressure-Temperature Curves for Duane Arnold Energy Center, GE 2

Nuclear Energy, San Jose, CA, GE-NE-A22-00100-08-01-R2, Revision 2, August 2003 3

(GE Proprietary Information).

4 5

18.

L. J. Tilly, Pressure-Temperature Curves for Exelon LaSalle Unit 1, GE Nuclear 6

Energy, San Jose, CA, GE-NE-0000-0003-5526-02R1, Revision 1, May 2004 (GE 7

Proprietary Information).

8 9

19.

L. J. Tilly, Pressure-Temperature Curves for Exelon LaSalle Unit 2, GE Nuclear 10 Energy, San Jose, CA, GE-NE-0000-0003-5526-01R1, Revision 1, May 2004 (GE 11 Proprietary Information).

12 13

20.

Brian Benney (NRC) to J. V. Parrish (Energy Northwest), Columbia Generating Station 14

- Issuance of Amendment Re: Reactor Coolant System (RCS) Pressure and 15 Temperature Limits (TAC NO. MC3591), May 12, 2005, ADAMS Accession No.

16 ML051160277.

17 18

21.

Darl S. Hood (NRC) to Mark A. Peifer (DAEC), Duane Arnold Energy Center - Issuance 19 of Amendment Regarding Pressure and Temperature Limit Curves (TAC NO. MB8750),

20 August 25, 2003, ADAMS Accession No. ML032310536.

21 22

22.

Stephen P. Sands (NRC) to Christopher M. Crane (Exelon), LaSalle County Station, 23 Units 1 and 2, Issuance of Amendments RE: Pressure-Temperature Limits (TAC NOS.

24 MB7795 and MB7796), December 17, 2004, ADAMS Accession No. ML043240176.

25 26

23.

L. J. Tilly, Pressure-Temperature Curves for DTE Energy Fermi Unit 2, GE Nuclear 27 Energy, San Jose, CA, NEDC-33133P, Revision 0, February 2005 (GE Proprietary 28 Information).

29 30

24.

David H. Jaffe (NRC) to Donald K. Cobb (Detroit Edison Company), Fermi 2 -

31 Issuance of Amendment RE: Reactor Coolant System Pressure and Temperature 32 Curves (TAC NO. MC6468), January 25, 2006, ADAMS Accession No. ML053120186.

33 34

25.

A. R. Mehta and L. J. Tilly, Pressure-Temperature Curves for Exelon Dresden Unit 2, 35 GE Nuclear Energy, San Jose, CA, GE-NE-0000-0002-9629-01R1, Revision 1, May 36 2004 (GE Proprietary Information).

37 38

26.

L. J. Tilly, Pressure-Temperature Curves for Exelon Dresden Unit 3, GE Nuclear 39 Energy, San Jose, CA, GE-NE-0000-0002-9600-01R2, Revision 2, May 2004 (GE 40 Proprietary Information).

41 42

27.

L. J. Tilly, Pressure-Temperature Curves for Exelon Quad Cities Unit 1, GE Nuclear 43 Energy, San Jose, CA, GE-NE-0000-0002-9600-02R2, Revision 2, May 2004 (GE 44 Proprietary Information).

45 46

28.

L. J. Tilly, Pressure-Temperature Curves for Exelon Quad Cities Unit 2, GE Nuclear 47 Energy, San Jose, CA, GE-NE-0000-0002-9600-03R2, Revision 2, May 2004 (GE 48 Proprietary Information).

49 1

29.

Maitri Banerjee (NRC) to Christopher M. Crane (Exelon), Dresden Nuclear Power 2

Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance 3

of Amendments Regarding Pressure and Temperature Limits (TAC Nos. MC5160, 4

MC5161, MC5162, and MC5163), October 17, 2005, ADAMS Accession 5

No. ML052570761.

6 7

30.

NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, K. Wichman, M.

8 Mitchell, and A. Hiser, NRC/Industry Workshop on RPV Integrity Issues, 9

February 12, 1998.

10 11 12 Principle Contributor: C. F. Sheng 13 14 Date: March 23, 2009 15