NRC 2006-0054, Public Version, Request for Review of Reactor Vessel Toughness Fracture Mechanics Analysis

Public Version, Request for Review of Reactor Vessel Toughness Fracture Mechanics Analysis

ML062140522
Person / Time
Site:	Point Beach
Issue date:	06/06/2006
From:	Koehl D Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
F. Lyon LPLE X2296
Shared Package
ML062140272	List: NRC 2006-0054, Public Version, Request for Review of Reactor Vessel Toughness Fracture Mechanics Analysis
References
NRC 2006-0054, TAC MC2099, TAC MC2100
Download: ML062140522 (65)

Text

NMC,_

Committed to NuclearExcellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC June 6, 2006 NRC 2006-0054 10 CFR 50.90 10 CFR 50, App. G U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Point Beach Nuclear Plant Units I and 2 Dockets 50-266 and 50-301 License Nos. DPR-24 and DPR 27 Request for Review of Reactor Vessel Toughness Fracture Mechanics Analysis

Reference:

1) Letter from NMC to NRC dated October 25, 2004, NRC 2004-0111, "License Renewal Application Response to Request for Additional Information (TAC Nos. MC2099 and MC2100)"

Pursuant to 10 CFR 50.90, Nuclear Management Company, LLC (NMC), hereby requests a proposed amendment to the licenses for Point Beach Nuclear Plant (PBNP),

Units I and 2. The proposed amendment would support a change to the PBNP Final Safety Analysis Report (FSAR) by incorporating an updated analysis for satisfying the reactor vessel Charpy upper-shelf energy requirements of 10 CFR 50, Appendix G, Section IV.A.1.

Enclosed for Commission review and approval is Areva Document BAW-2467P, Revision 1, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Point Beach Units 1 and 2 for Extended Life through 53 Effective Full Power Years", dated October 2004. BAW-2467P incorporates the latest (2004) fluence projections to be consistent with the Pressurized Thermal Shock (PTS) evaluation. This analysis is being submitted in accordance with the requirements of 10 CFR 50, Appendix G, Section IV.A.1.c.

The non-proprietary version of this document (BAW-2467NP, Revision 1), had previously been submitted in Reference I as part of the license renewal process in accordance with i0 CFR54.

Enclosure I provides a description, justification, and a significant hazards determination for the reactor vessel toughness fracture mechanics analysis. Enclosure 2 submits BAW-2467NP, Revision I (non-proprietary). Enclosure 3 submits BAW-2467P, Revision 1, (proprietary). Enclosure 4 provides a Westinghouse authorization letter, accompanying affidavit, Proprietary Information Notice and Copyright Notice for the 6590 Nudear Road

• Tw Rivers, W&orin 54241 Telephae: 9.20155.2321

Document Control Desk Page 2 proprietary portion of the analysis. Also provided in Enclosure 4 are proprietary and non-proprietary versions of Westinghouse document WEP-06-33, which is the Westinghouse source document for Figure 5-1 of BAW-2467P.

Since the Areva document listed above as Proprietary contains information proprietary to Westinghouse, it is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity, for each, the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR 2.390.

Correspondence with respect to the copyright or proprietary aspects of the above documents, or the supporting Westinghouse affidavit, should reference the appropriate authorization letter (CAW-06-2141) and be addressed to B. F. Maurer, Acting Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

This letter contains no new commitments and no revision to existing commitments.

NMC requests approval of the proposed license amendment by May 2007, with the amendment being implemented within 60 days.

In accordance with 10 CFR 50.91, a copy of this submittal, with enclosures, is being provided to the designated Wisconsin Official.

I declare under penalty of peujury that the foregoing is true and correct. Executed on June 6, 2006.

Dennis L. Koehl Site Vice-President, Poi Beach Nuclear Plant Nuclear Management Company, LLC

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Regional Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE I REQUEST FOR REVIEW OF REACTOR VESSEL TOUGHNESS FRACTURE MECHANICS ANALYSIS

1.0 INTRODUCTION

As required by 10 CFR 50, Appendix G, Section IV.A.1.c, and in accordance with 10 CFR 50.90, Nuclear Management Company, LLC (NMC) requests review and approval of a revised reactor vessel toughness fracture mechanics analysis for the Point Beach Nuclear Plant (PBNP) Units I and 2. The proposed license amendment would support a change to the PBNP Final Safety Analysis Report (FSAR) by incorporating the updated analysis for satisfying the reactor vessel Charpy upper-shelf energy requirements of 10 CFR 50, Appendix G, Section IV.A.1.

2.0 DESCRIPTION

OF PROPOSED CHANGE NMC proposes changing the PBNP licensing basis to incorporate a revised equivalent margins assessment for the reactor pressure vessels (RPV), in PBNP Units 1 and 2, for material toughness when the upper-shelf Charpy energy level falls below 50 ft-lb. This assessment applied the 2004 Westinghouse fluence projection using full uprated power (1678 MWt), without crediting the presence of Hafnium power suppression inserts.

The Charpy upper-shelf value of reactor vessel beltline weld materials at Point Beach Units 1 and 2 may be less than 50 ft lb at 53 EFPY. In order to demonstrate that sufficient margins of safety against fracture remain to satisfy the requirements of Appendix G to 10 CFR Part 50, a low upper-shelf toughness fracture mechanics analysis has been performed. The limiting welds in the beltline region have been evaluated for ASME Levels A, B, C, and D Service Loadings based on the evaluation acceptance criteria of the ASME Code,Section XI, Appendix K.

The analysis demonstrates that the limiting reactor vessel beltline welds at Point Beach Units 1 and 2 satisfy the ASME Code requirements of Appendix K for ductile flaw extensions and tensile stability using projected low upper-shelf Charpy impact energy levels for the weld material at 53 EFPY.

3.0 BACKGROUND

The NlRC slafety evlUatinn renprt a*-sc-artd with li;.n.. ren-r'l of the Point Beach Nuclear Plant Units 1 and 2 (NUREG-1 839) references Areva analysis BAW-2467NP, dated July 2004. The July 2004 analysis was reissued October 2004 as Revision 1 to BAW-2467NP with updates to the fluence values used in the original analysis.

BAW-2467NP, Revision 1, dated October 2004, had previously been submitted in Reference I as part of the license renewal process in accordance with 10 CFR 54. In Page 1 of 4

accordance with 10 CFR 50, Appendix G, Section IV.A.1.c, NRC review and approval of this document is required for PBNP to incorporate it into the licensing basis.

4.0 TECHNICAL ANALYSIS

The technical justification for the revised reactor vessel toughness fracture mechanics analysis is contained in the enclosed Areva document. The revised reactor vessel toughness fracture mechanics analysis provides technical justification for satisfying the reactor vessel Charpy upper-shelf energy requirements of 10 CFR 50, Appendix G, Section IV.A.1.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Determination As required by 10 CFR 50, Appendix G, Section IV.A.1.c, and in accordance with 10 CFR 50.90, Nuclear Management Company, LLC (NMC) requests review and approval of a revised reactor vessel toughness fracture mechanics analysis for the Point Beach Nuclear Plant (PBNP) Units 1 and 2. The proposed license amendment would support a change to the PBNP Final Safety Analysis Report (FSAR) by incorporating the updated analysis for satisfying the reactor vessel Charpy upper-shelf energy requirements of 10 CFR 50, Appendix G, Section IV.A.1.

NMC has evaluated the proposed amendment in accordance with 10 CFR 50.91 against the standards in 10 CFR 50.92 and has determined

• that the operation of PBNP Units 1 and 2, in accordance with the proposed amendments, presents no significant hazards. The NMC evaluation against each of the criteria in 10 CFR 50.92 follows:

1. Would the proposed amendment involve a significant increase in the probability or consequences of any accident previously evaluated?

The proposed change incorporates the updated analysis for satisfying the reactor vessel Charpy upper-shelf energy requirements of 10 CFR 50, Appendix G, Section IV.A.1 into the FSAR. The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or the manner in which the plant is operated and maintained. The nronnonsrecrhannge dnos notf altfr nr prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the types or amounts of Page 2 of 4

radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed change is consistent with safety analysis assumptions and resultant consequences. Therefore, it is concluded that this change does not significantly increase the probability of occurrence of an accident previously evaluated.

2. Would the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change incorporates the updated analysis for satisfying the reactor vessel Charpy upper-shelf energy requirements of 10 CFR 50, Appendix G, Section IV.A.1 into the FSAR. The change does not impose any new or different requirements or eliminate any existing requirements. The change does not alter assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change would not create the possibility of a new or different kind of accident from any previously evaluated.

3. Would the proposed amendment result in a significant reduction in a margin of safety?

The proposed change incorporates the updated analysis for satisfying the reactor vessel Charpy upper-shelf energy requirements of 10 CFR 50, Appendix G, Section IV.A.1 into the FSAR. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The setpoints at which protective actions are initiated are not altered by the proposed change. Therefore, the proposed amendment does not result in a significant reduction in a margin of safety.

Conclusion Operation of PBNP in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of any accident previously analyzed; would not create the possibility of a new or different kind of accident from any accident previously analyzed; and, would not result in a significant reduction in any margin of safety. Therefore, operation of PBNP in accordance with the proposed amendment presents no significant hazards.

5.2 Applicable Regulatory Requirements 10 CFR 50, Appendix G, Section IV.A.1 promulgates reactor vessel Charpy upper-shelf energy requirements. 10 CFR 50, Appendix G, Section IV.A.I.c Page 3 of 4

requires that the analysis for satisfying the reactor vessel Charpy upper-shelf energy requirements of 10 CFR 50, Appendix G, Section IV.A.1 must be submitted for review and approval on an individual case basis at least three years prior to the date when the predicted Charpy upper-shelf energy will no longer satisfy the requirements of Section IVA.1.

10 CFR 50.71 (e) requires that licensees shall periodically update their final safety analysis report (FSAR), to assure that the information included in the report contains the latest information developed. This update shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement. The update shall also include the effects of all analyses of new safety issues performed by or on behalf of the licensee at Commission request.

Based upon the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in accordance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.3 Commitments There are no actions committed to by NMC in this document. Statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

6.0 ENVIRONMENTAL EVALUATION NMC has determined that the information for the proposed amendment does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of effluent release, or result in any significant increase in individual or cumulative occupational radiation exposure.

Accordingly, this proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with this proposed amendment.

Page 4 of 4

ENCLOSURE 2 TO REQUEST FOR REVIEW OF REACTOR VESSEL TOUGHNESS FRACTURE MECHANICS ANALYSIS AREVA DOCUMENT, BAW-2467NP, REVISION 1, "LOW UPPER-SHELF TOUGHNESS FRACTURE MECHANICS ANALYSIS OF REACTOR VESSEL OF POINT BEACH UNITS I AND 2 FOR EXTENDED LIFE THROUGH 53 EFFECTIVE FULL POWER YEARS", DATED OCTOBER 2004 (NON-PROPRIETARY)

(46 pages follow)

BAW-2467NP, Rev. I October 2004 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Point Beach Units I and 2 for Extended Life through 53 Effective Full Power Years AREVA Document No. 77-2467NP-01 (See Section 11 for document signatures.)

Prepared for Nuclear Management Company Prepared by Framatome ANPI Inc.

An AREVA and Siemens company 3315 Old Forest Road P. 0. Box 10935 Lynchburg, Virginia 24506-0935

BAW-2467NP, Rev. 1 EXECUTIVE

SUMMARY

Nuclear Management Company is considering plant life extension, power uprate to 1678 MWt and removal of hafnium power suppression assemblies from the core for Point Beach Units 1 and 2. As a result of these changes, operating conditions including vessel temperatures and projected fluence values at 53 effective full power years (EFPY) of plant operation have changed. It must be ensured that these changes do not affect the plant adversely from a regulatory compliance point of view. One of the compliance issues is Appendix G to 10 CFR Part 50 where low upper-shelf toughness is addressed. An equivalent margins assessment has to be made for material toughness when the.upper-shelf Charpy energy level falls below 50 ft-lb. This report addresses this particular compliance issue regarding low upper-shelf toughness only.

The Charpy upper-shelf value of reactor vessel beltline weld materials at Point Beach Units 1 and 2 may be less than 50 ft lb at 53 EFPY. In order to demonstrate that sufficient margins of safety against fracture remain to satisfy the requirements of Appendix G to 10 CFR Part 50, a low upper-shelf toughness fracture mechanics analysis has been performed. The limiting welds in the beltline region have been evaluated for ASME Levels A, B, C, and D Service Loadings based on the evaluation acceptance criteria of the ASME Code,Section XI, Appendix K.

The analysis presented in this report demonstrates that the limiting reactor vessel beltline weld at Point Beach Units 1 and 2 satisfies the ASME Code requirements of Appendix K for ductile flaw extensions and tensile stability using projected low upper-shelf Charpy impact energy levels for the weld material at 53 EFPY.

ii A A R EVA

BAW-2467NP, Rev. I RECORD OF REVISIONS Revision 0

1 Affected Pages All All Description Date Original release July 2004 Updated fluence values used for Evaluation October 2004 Condition I iii A

AREVA

BAW-2467NP, Rev. 1 TABLE OF CONTENTS 1.0 Introduction..................................................................................................................

1-1 2.0 Changes in Operating Condition Parameters...............................................................

2-1 3.0 Material Properties and Reactor Vessel Design Data.................................................

3-1 3.1 J-Integral Resistance Model for Mn-Mo-Ni/Linde 80 Welds.............................. 3-1 3.2 Reactor Vessel Design Data.............................................................................

3-1 3.3 Mechanical Properties for Weld Material..........................................................

3-1 3.3.1 Axial Weld SA-847........................................

3-2 3.3.2 Circumferential Weld SA-1 011.............................................................

3-3 3.3.3 Circumferential Weld SA-1484....................................................... 3-4 4.0 Analytical Methodology.........................................

4-1 4.1 Procedure for Evaluating Levels A and B Service Loadings.............

4-1 4.2 Procedure for Evaluating Levels C and D Service Loadings............................ 4-1 4.3 Temperature Range for Upper-Shelf Fracture Toughness Evaluations............ 4-3 4.4 Effect of Cladding Material................... :.......................................................... 4-3 5.0 A pplied Loads.......................

............................................................................. 5-1 5.1 Levels A and'B Service Loadings...................................................................

5-1 5.2 Levels C and D Service Loadings..............................

5-1 6.0 Evaluation for Levels A and B Service Loadings..........................................................

6-1 7.0 Evaluation for Levels C and D Service Loadings..........................................................

7-1 8.0 Summary of Results...........................................

8-1 9.0 Conclusion..................................................................................................................

9-1 10.0 References...................................................................................

........... 10-1 11.0 Certification................................................

11-1 12.0 AppendixA.................................................................

12-1 13.0 A ppendix B.................................................................................................................

13-1 iv A

AR EVA

BAW-2467NP, Rev. 1 LIST OF TABLES Table 2-1 Evaluation Conditions.............................................................................................

2-2 Table 3-1 Mechanical Properties for SA-847 Weld of Point Beach Unit 1................................

3-2 Table 3-2 Mechanical Properties for SA-1101 Weld of Point Beach Unit 1............................. 3-3 Table 3-3 Mechanical Properties for SA-1484 Weld of Point Beach Unit 2............................. 3-4 Table 6-1 Material J-Integral Resistance for Levels A and B Service Loadings - Evaluation Condition 1 - Uprated Power Conditions Without Hafnium Assemblies.................. 6-2 Table 6-2 Material J-Integral Resistance for Levels A and B Service Loadings - Evaluation Condition 2 - Current Power Conditions Without Hafnium Assemblies.................. 6-2 Table 6-3 Material J-Integral Resistance for Levels A and B Service Loadings - Evaluation Condition 3 - Current Power Conditions With Hafnium Assemblies....................... 6-2 Table 6-4 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 1 -

Uprated Power Conditions Without Hafnium Assemblies.......................................

6-3 Table 6-5 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 2 -

Current Power Conditions Without Hafnium Assemblies........................................

6-3 Table 6-6 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 3 -

Current Power Conditions With Hafnium Assemblies.............................................

6-3 Table 7-1 J-Integral vs. Flaw Extension for Evaluation Condition I - SA-847.......................... 7-6 Table 7-2 J-Integral vs. Flaw Extension for Evaluation Condition 1 - SA-1 101........................ 7-7 Table 7-3 J-Integral vs. Flaw Extension for Evaluation Condition I - SA-1484........................ 7-8 Table 7-4 Level D Service Loadings - Internal Pressure at Tensile Instability - SA-847.......... 7-9 VA A

AR EVA

BAW-2467NP, Rev. I 2-1 2-2 5-1 6-1 7-1 7-2 7-3 7-4 7-5 LIST OF FIGURES Reactor Vessel of Point Beach Unit I.................... ".................................................... 2-3 Reactor Vessel of Point Beach Unit 2....................................................

...... 2-4 Level D transients - Reactor Coolant Temperature and Pressure vs. Time................ 5-2 J-Integral vs. Flaw Extension for Levels A and B Service Loading - Evaluation Condition 2 - Current Power Conditions Without Hafnium Assemblies - Weld SA-847.............. 6-4 K, vs. Crack Tip Temperature for Evaluation Condition I - SA-847.............................

7-2 K, vs. Crack Tip Temperature for Evaluation Condition I - SA-1 101...........................

7-3 K, vs. Crack Tip Temperature for Evaluation Condition I - SA-1484...........................

7-4 J-Integral vs. Flaw Extension -All Welds...................................................................

7-10 J-Integral vs. Flaw Extension - SA-847......................................................................

7-11 vi A

AREVA

BAW-2467NP, Rev. 1 1.0 Introduction Nuclear Management Company is considering plant life extension, power uprate to 1678 MWt and removal of hafnium power suppression assemblies from the core for Point Beach Units 1 and 2. This document assesses the effect of these proposed changes on the upper-shelf fracture toughness of the reactor vessels.

The B&W Owners Group (B&WOG) fracture toughness model was used in the low upper-shelf toughness fracture mechanics analyses of the reactor vessels of the B&WOG Reactor Vessel Working Group (RVWG) which includes the Point Beach Units 1 and 2 reactor vessels. The low upper-shelf toughness analysis for all reactor vessels of the B&WOG RVWG for Levels A & B Service Loadings was documented in BAW-2192PA [1]. An additional fracture mechanics analysis for Levels C & D Service Loadings was carried out for all these reactor vessels and documented in BAW-2178PA [2]. Both these reports have been accepted by the NRC.

As a result of a subsequent power uprate, an additional low upper-shelf toughness analysis covering end-of-license and end-of-license renewal fluence values was performed for Point Beach Units 1 and 2 [3].

For the current planned changes, the effect on the reactor vessel materials upper-shelf toughness is assessed in this report.

Welds in the beltline region of all B&W Owners Group Reactor Vessel Working Group plants, including the Point Beach Units I and 2 vessels, have been analyzed [1, 2] for 32 effective full power years (EFPY) of operation to demonstrate that these low upper-shelf energy materials would continue to satisfy federal requirements for license renewal. In Reference 3, the Point Beach vessels were analyzed up to their forecasted end-of-license extension periods at a partially uprated power level of 165OMWt with hafnium power suppression assemblies, and both vessels were shown to be acceptable. The purpose of the present analysis is to perform a similar low upper-shelf toughness evaluation of the reactor vessel welds at the Point Beach plants for projected neutron fluences at 53 EFPY.

The present analysis addresses ASME Levels A, B, C, and D Service Loadings. For Levels A and B Service Loadings, the low upper-shelf toughness analysis is performed according to the acceptance criteria and evaluation procedures contained in Appendix K to Section Xl of the ASME Code [4]. The evaluation also utilizes the acceptance criteria and evaluation procedures prescribed in Appendix K for Levels C and D Service Loadings.

Levels C and D Service Loadings are evaluated using the one-dimensional, finite element, thermal and stress models and linear elastic fracture mechanics methodology of Framatome ANP's PCRIT computer code to determine stress intensity factors for a worst case pressurized thermal shock transient.

Revision 1 of this document utilizes the updated fluence values calculated in 2004 for the uprated power condition of 1678 MWt without the hafnium power suppression assemblies installed.

This input was provided by the Nuclear Management Company (NMC) and is included as Appendices A and B.

I A

AREVA

BAW-2467NP, Rev. 1 2.0 Changes in Operating Condition Parameters As a result of the planned updates to the Point Beach Units 1 and 2, there are increases in the projected end of life fluences for both the units. There are also changes in the plants' operating temperatures. These inputs were provided by the Nuclear Management Company and included as Appendices A and B and summarized in this section.

The analysis for current licensed rated power conditions (1540 MWt) gives a maximum cold leg temperature of 544.50F. As a result of the power uprate to 1678 MWt, the maximum cold leg temperature is reduced to 541.4 0F. The projected reactor vessel fluence values at 53 EFPY are provided in Table 2-1.

For this analysis, three cases, termed Evaluation Conditions, are studied -

uprated power conditions without hafnium assemblies, current power conditions without hafnium assemblies, and current power conditions with hafnium assemblies. Fluence values for these three cases are reported only for the controlling welds identified through review of the results reported in References 1, 2 and 3. Locations of the reactor vessel welds for Point Beach Units I and 2 are illustrated in Figures 2-1 and 2-2 respectively [1].

2-1 A

AR EVA

BAW-2467NP, Rev. 1 Table 2-1 Evaluation Conditions Fluence (n/cm2) at 53 EFPY Weld Location [1]

Weld Number

[1]

Cu (wt%)

[5]

Ni (wt%)

[5]

EVALUATION CONDITION I Uprated Power Conditions Without Hafnium Assemblies EVALUATION CONDITION 2 Current Power Conditions Without Hafnium Assemblies Cold Leg Temp:

544.50F EVALUATION CONDITION 3 Current Power Conditions With Hafnium Assemblies Cold Leg Temp:

544.50F Plant Cold Leg Temp:

541.4 0F PB-1 Lower Shell SA-847 0.23 0.52 3.25E+19 3.12E+19 2.67E+19 Long.

Inter.

Shell/Lower SA-1101 0.23 0.59 4.71E+19 4.52E+19 3.82E+19 Shell Circ.

Inter.

PB-2 Shell/Lower SA-1484 0.26 0.60 4.85E+19 4.65E+19 3.79E+19 Shell Circ.

2-2 A

AR EVA

BAW-2467NP, Rev. 1 Figure 2-1 Reactor Vessel of Point Beach Unit 1 [1]

L.

Weld SA-1426 75.

F l"Weld SA-812 Inside*27%

SA-775 Outside 73%

LIntermediate Shell (Plate) A981 1-1 I-a Weld SA-1101 Weld SA-847 Lower Shell (Plate) C1423-1

.39.87" 9 L8-*

Weld SA-1 101 2-3 A

AREVA

BAW-2467NP, Rev. 1 Figure 2-2 Reactor Vessel of Point Beach Unit 2 [1]

CE Weld Intermediate 123V500.VA1 Shell (Forging)

Weld SA-1484 Lower Shell (Forging) 122W195VA1 Kg 2-4 A

AREVA

BAW-2467NP, Rev. I 3.0 Material Properties and Reactor Vessel Design Data An upper-shelf fracture toughness material model is discussed below,, as well as mechanical properties for the weld material and reactor vessel design data.

3.1 J-Integral Resistance Model for Mn-Mo-Ni/Linde 80 Welds A model for the J-integral resistance versus crack extension curve (J-R curve) required to analyze low upper-shelf energy materials has been derived specifically for Mn-Mo-Ni/Linde 80 weld materials. A previous analysis of the reactor vessels of B&W Owners Group RVWG [1]

described the development of this toughness model from a large data base of fracture specimens.

A lower bound (-2Se) J-R curve is obtained by multiplying J-integrals from the mean J-R curve by 0.699 [1]. It was shown in a previous low upper-shelf toughness analysis performed for B&W Owners Group plants [6] that a typical lower bound J-R curve is a conservative representation of toughness values for reactor vessel beltline materials, as required by Appendix K (4] for Levels A, B, and C Service Loadings.

The best estimate representation of toughness required for Level D Service Loadings is provided by the mean J-R curve [7].

3.2 Reactor Vessel Design Data Pertinent design data for upper-shelf flaw evaluations in the beltline region of the reactor vessel are provided below for Point Beach Units 1 and 2.

Design Pressure, Pd

= 2485 psig [2] (use 2500 psig)

Inside radius, Ri

= 66 in. [2]

Vessel thickness, t

= 6.5 in. [2]

Nominal cladding thickness, tc

= 0.1875 in. [2]

3.3 Mechanical Properties for Weld Material Mechanical properties for the base and weld materials are presented in Tables 3-1 through 3-3.

The reactor vessel base metal at Point Beach Unit 1 is SA-302, Grade B low alloy steel, and at Point Beach Unit 2 is SA-508, Grade 2, Class 1 low alloy steel [8]. Base metal properties are found in the ASME Code [9].

Weld metal tensile properties are taken from appropriate surveillance capsule data of each weld material.

The ASME" transition region fracture toughness curve for Kc, used to define the beginning of the upper-shelf toughness region, is indexed by the initial RTNOT of the weld material. Also, Poisson's ratio, v, is taken to be 0.3.

3-1 A

AREVA

BAW-2467NP, Rev. I 3.3.1 Axial Weld SA-847 Table 3-1 Mechanical Properties for SA-847 Weld of Point Beach Unit 1 Temp.

E Yield Strength (cry)

Ultimate Strength (a,,)*

a Material:

Base Base Weld Base Weld Base Metal Metal SA-847 Metal SA-847 Metal S6urce:

Code Code Actual Code Actual Code

[Ref.]

[9]

[9]

[10]

[9]

[10]

[9]

(OF)

(ksi)

(ksi)

(ksi)

(ksi)

(ksi)

(in/in/°F) 100 29200 50.00 95.00 80 99.8 7.06E-06 200 28500 47.50 89.60 80 99.8 7.25E-06 300 28000 46.10 86.01 80

.99.8 7.43E-06 335 27790 45.74 85.10 80 97.6 7.48E-06 400 27400 45.10 84.77 80 99.8 7.58E-06 500 27000 44.50 84.26 80 99.8 7.70E-06 541.4 26751.6 44.16 84.04 80 99.8 7.75E-06 544.5 26733 44.14 84.03 80 99.8 7.76E-06 550 26700 44.11 84.00 80 99.8 7.77E-06 600 26400 43.80 83.74 80 99.8 7.83E-06

• Note: The ultimate strength values of the base and weld metals given here are not used in calculations Initial RTNDT = -5.0°F [5]

Margin = 48.30F [5]

3-2 A

AREVA

BAW-2467NP, Rev. 1 k

3.3.2 Circumferential Weld SA-1011 Table 3-2 Mechanical Properties for SA-1 101 Weld of Point Beach Unit 1 Temp.

E Yield Strength (a-y)

Ultimate Strength (au,)*

a Material:

Base Base Weld Base Weld Base Metal Metal SA-1101 Metal SA-1101 Metal Source:

Code Code Actual Code Actual Code

[Ref.]

[9]

[9]

[11]

[9]

[11]

[9]

(OF)

(ksi)

(ksi)

(ksi)

(ksi)

(ksi)

(in/in/°F) 100 29200 50.00 93.66 80 105.10 7.06E-06 200 28500 47.50 92.20 80 104.90 7.25E-06 300 28000 46.10 90.74 80 104.70 7.43E-06 400 27400 45.10 89.29 80 104.50 7.58E-06 500 27000 44.50 87.83 80 104.30 7.70E-06 541.4 26751.6 44.14 87.23 80 104.21 7.76E-06 544.5 26733 44.14 87.18 80 104.21 7.76E-06 550 26700 44.11 87.10 80 104.20 7.77E-06 600 26400 43.80 86.37 80 104.10 7.83E-06

• Note: The ultimate strength values of the base and weld metals given here are not used in calculations Initial RTNDT = 10.00F [5]

Margin = 56.0°F [5]

3-3 A

AREVA

BAW-2467NP, Rev. I 3.3.3 Circumferential Weld SA-1 484 Table 3-3 Mechanical Properties for SA-1484 Weld of Point Beach Unit 2 Temp.

E Yield Strength (oy)

Ultimate Strength (o-,)*

a Material:

Base Base Weld Base Weld Base Metal Metal SA-1484 Metal SA-1484 Metal Source:

Code Code Actual Code Actual Code

[Ref.]

[9]

[9]

[12]

[9]

[12]

[9]

("F)

(ksi)

(ksi)

(ksi)

(ksi)

(ksi)

(in/in/°F) 100 27800 50.00 82.10 80 96.90 6.50E-06 200 27100 47.50 79.57 80 92.98 6.67E-06 300 26700 46.10 78.00 80 90.40 6.87E-06 400 26100 45.10 77.17 80 89.41 7.07E-06 450 25900 44.76 76.80 80 89.60.

7.15E-06 500 25700 44.50 76.42 80 90.29 7.25E-06 541.4 25460 44.16 76.15 80 91.25 7.32E-06 544.5 25444 44.14 76.13 80 91.34 7.33E-06 580 25264 43.94 76.00 80 92.50 7.39E-06 600 25200 43.80 75.80 80 93.28 7.42E-06 Note: The ultimate strength values of the base and weld metals given here are not used in calculations Initial RTNDT = -5.00F [5]

Margin = 68.50F [5]

3-4 A

AR EVA

BAW-2467NP, Rev. 1 4.0 Analytical Methodology Upper-shelf toughness is evaluated through use of fracture mechanics analytical methods that utilize the acceptance criteria and evaluation procedures of Section Xl, Appendix K [4], where applicable.

4.1 Procedure for Evaluating Levels A and B Service Loadings The applied J-integral is calculated per Appendix K, paragraph K-4210 [4], using an effective flaw depth to account for small scale yielding at the crack tip, and evaluated per K-4220 for upper-shelf toughness and per K-4310 for flaw stability.

4.2 Procedure for Evaluating Levels C and D Service Loadings Levels C and D Service Loadings are evaluated using the one-dimensional, finite element, thermal and stress models and linear elastic fracture mechanics methodology of the PCRIT computer code to determine stress intensity factors. The beltline region welds identified in Section 3.3 are analyzed for all Level C and D transients. Two Level D transients are specified for the Point Beach Units. The original equipment specification includes a Steam Line Break (SLB) transient and a Reactor Coolant Line Break (LOCA) transient. The Point Beach FSAR contains a Steam Line Break (two loops in service) without Offsite Power transient [13].

The transients considered appear in Figure 5.1. Transients are assumed to hold steady at the end of their definitions, and are held constant until the thermal gradient through the shell has developed fully and begins to dissipate.

The evaluation is performed as follows:

(1)

For each transient described above, utilize PCRIT to calculate stress intensity factors for a semi-elliptical flaw of depth 1/10 of the base metal wall thickness, as a function of time, due to internal pressure and radial thermal gradients with a factor of safety of 1.0 on loading.

The applied stress intensity factor, KI, calculated by PCRIT for each of these transients is compared to the Kjc limit of the weld. The transient that most closely approaches the Kj, limit is chosen as the limiting transient, and the critical time in the limiting transient occurs at the point where K, most closely approaches the upper-shelf toughness curve.

(2)

At the critical transient time, develop a crack driving force diagram with the applied J-integral and J-R curves plotted as a function of flaw extension. The adequacy of the upper-shelf toughness is evaluated by comparing the applied J-integral with the J-R curve at a flaw extension of 0.10 in.

Flaw stability is assessed by examining the slopes of the applied J-integral and J-R curves at the points of intersection.

(3)

Verify that the extent of stable flaw extension is no greater than 75% of the vessel wall thickness by determining when the applied J-integral curve intersects the mean J-R curve.

4-1 A

AR EVA

BAW-2467NP, Rev. 1 (4)

Verify that the remainilng ligament is not subject to tensile instability. The internal pressure p shall be less than P,, where P, is the internal pressure at tensile instability of the remaining ligament. Equations for P, are given below for the axial and circumferential flaws [14]. These equations first appear in the 2001 Edition of the ASME Section Xl code that is cited.

(a) For an axial flaw, P, =1.07ao[(Ri1t-(A,1A)

[eqn. 1]

where cjy + au 0o 2 --

[eqn. 2]

2 A=t(e+t)

[eqn. 3]

Ac =n-

[eqn. 4]

4 and

= surface length of crack, six times the depth, a Rm = mean radius of vessel This equation for P, includes the effect of pressure on the flaw face.

(b) For a circumferential flaw, P, = 1.07ao[

1 -(Rrt)(Ac1A)]

[eqn. 5]

where uo, A, and Ac are given by equations 2, 3 and 4, respectively.

This equation for P, includes the effect of pressure on the flaw face. This equation is valid for internal pressures not exceeding the pressure at tensile instability caused by the applied hoop stress acting over the nominal wall thickness of the vessel. This validity limit on pressure for the circumferential flaw equation for P, is P,r 1.07ao-.]

[eqn. 6]

4-2 A

AREVA

BAW-2467NP, Rev. I 4.3 Temperature Range for Upper-Shelf Fracture Toughness Evaluations Upper-shelf fracture toughness is determined through use of Charpy V-notch impact energy versus temperature plots by noting the temperature above which the Charpy energy remains on a plateau, maintaining a relatively high constant energy level. Similarly, fracture toughness can be addressed in=' three different regions on the temperature scale, i.e. a lower-shelf toughness region, a transition region, and an upper-shelf toughness region: Fracture toughness of reactor vessel steel and associated weld metals are conservatively predicted by the ASME initiation toughness curve, Kl,, in the lower-shelf and transition regions. In the upper-shelf region, the upper-shelf toughness curve, Kj, is derived from the upper-shelf J-integral resistance model described in Section 3.1.

The upper-shelf toughness then becomes a function of fluence, copper content, temperature, and fracture specimen size. When upper-shelf toughness is plotted versus temperature, a plateau-like curve develops that decreases slightly with increasing temperature. Since the present analysis addresses the low upper-shelf toughness issue, only the upper-shelf temperature range, which begins at the intersection of Kc and the upper-shelf toughness curves, Kjc, is considered.

4.4 Effect of Cladding Material The PCRIT code utilized in the flaw evaluations for Levels C and D Service Loadings does not consider stresses in the cladding when calculating stress intensity factors for thermal loads. To account for this cladding effect, an additional stress intensity factor, Kjdad, is calculated separately and added to the total stress intensity factor computed by PCRIT.

The contribution of cladding stresses to stress intensity factor was examined previously [2]. In this low upper-shelf toughness analysis performed for B&W Owners Group Reactor Vessel Working Group plants, the Zion-1 WF-70 weld using thermal loads from the Turkey Point SLB was determined to be the bounding case. The Zion-1 vessel was as thick as or thicker than any other vessel. The thicknesses of the reactor vessels for the both Point Beach units are 6.5" whereas the Zion vessel is 8.44". The nominal cladding thickness is 3116" for both vessels.

From a thermal stress perspective, it is conservative to consider the thicker vessel. For the Zion vessel, the maximum value of Kidad, at any time during the transient and for any flaw depth, was determined to be 9.0 ksi4in. This bounding value is therefore used as the stress intensity factor for Kictad in this Point Beach low upper-shelf toughness analysis.

4-3 A

ARE VA

BAW-2467NP, Rev. I 5.0 Applied Loads The Levels A and B Service Loadings required by Appendix K are an accumulation pressure (internal pressure load) and a cooldown rate (thermal load). Since Levels C and D Service Loadings are not specified by the Code, Levels C and D pressurized thermal shock events are reviewed and a worst case transient is selected for use in flaw evaluations.

5.1 Levels A and B Service Loadings Per paragraph K-1300 of Appendix K [4], the accumulation pressure used for flaw evaluations should not exceed 1.1 times the design pressure. Using 2.5 ksi as the design pressure, the accumulation pressure is 2.75 ksi.

The cooldown rate is also taken to be the 'maximum required by Appendix K, 1 OO 0 F/hour.

5.2 Levels C and D Service Loadings As discussed in Section 4.2, the SLB and LOCA transients are evaluated using the computer code PCRIT. Pressure and temperature time histories for the two transients considered are shown in Figure 5-1.

5-1 A

AREVA

BAW-2467NP, Rev. I Figure 5-1 Level D transients - Reactor Coolant Temperature and Pressure vs. Time 5-2 A

ARE VA

BAW-2467NP, Rev. 1 6.0 Evaluation for Levels A and B Service Loadings The material mean and lower bounding J-R values for Evaluation Conditions 1, 2 and 3 detailed in Table 2-1 are given in Tables 6-1 through 6-3, respectively. Initial flaw depths equal to 1/4 of the vessel wall thickness are analyzed for Levels A and B Service Loadings following the procedure outlined in Section 4.1 and evaluated for acceptance based on values for the J-integral resistance of the materials from Section 3.3.

The results of the evaluation are presented in Table 6-4 through 6-6, where it is seen that the minimum ratio of material J-integral resistance (Jo.1) to applied J-integral (J1) is 1.87 for the SA-847 axial weld for Evaluation Condition 2, current power conditions without hafnium power suppression assemblies. This ratio is higher than the minimum acceptable value of 1.0. Also included in Table 6-4 through 6-6 is the applied J-integral at (Jo. 1) with a safety factor on pressure of 1.25.

The flaw evaluation for the controlling weld (SA-847) and controlling Evaluation Condition (2) is repeated by calculating applied J-integrals for various amounts of flaw extension with safety factors (on pressure) of 1.15 and 1.25. The results, along with mean and lower bound J-R curves, are plotted in Figure 6-1. The requirement for ductile and stable crack growth is also demonstrated by Figure 6-1 since the slope of the applied J-integral curve for a safety factor of 1.25 is considerably less than the slope of the lower bound J-R curve at the point where the two curves intersect.

6-1 A

AREVA

BAW-2467NP, Rev. I Table 6-1 Material J-Integral Resistance for Levels A and B Service Loadings - Evaluation Condition I - Uprated Power Conditions Without Hafnium Assemblies J-R at Aa = 0.1 in.

Cold Controlling Weld Fluence Lower Plant Leg Material Weld Cu x 1018 Mean Bound Temp.

ID Orientation Content (nlcm2) at -2Se

(°F)

(wt%)

at I.S.

at t/4 (lb/in)

(lb/in)

PB-1 541.4 SA-847 L

0.23 32.45 21.45 886 619 PB-I 541.4 SA-1101 C

0.23 47.10 31.13 871 609 PB-2 541.4 SA-1484 C

0.26 48.45 32.03 828 579 Table 6-2 Material J-Integral Resistance for Levels A and B Service Loadings - Evaluation Condition 2 - Current Power Conditions Without Hafnium Assemblies J-R at Aa = 0.1 in.

Cold Controlling Weld Fluence Lower Plant Leg Material Weld Cu x 1018 Mean Bound Temp.

ID Orientation Content (n/cm 2) at -2Se

('F)

(wt%)

at I.S.

at t/4 (lb/in)

(lb/in)

PB-1 544.5 SA-847 L

0.23 31.15 20.59 885 618 PB-I 544.5 SA-1101 C

0.23 45.20 29.88 870 608 PB-2 544.5 SA-1484 C

0.26 46.45 30.70 827 578 Table 6-3 Material J-Integral Resistance for Levels A and B Service Loadings - Evaluation Condition 3 - Current Power Conditions With Hafnium Assemblies J-R at Aa = 0.1 in.

Cold Controlling Weld Fluence Lower Plant Leg Material Weld Cu x 1018 Mean Bound Temp.

ID Orientation Content (n/cm2) at -2Se

('F)

(wt%)

at I.S.

at t/4 (lb/in)

(lb/in)

PB-I 544.5 SA-847 L

0.23 26.65 17.62 891 623 PRA-544.15

&A-I01 C

r0.21 3.2A 9 91r 877 613 PB-2 544.5 SA-1484 C

0.26 1

37.85 25.02 836 585 6-2 A

AREVA

BAW-2467NP, Rev. I Table 6-4 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 1 -

Uprated Power Conditions Without Hafnium Assemblies Lower Bounding SF = 1.15 SF 1.25 Plant Weld Weld Jc., at t/4 J

Jo.1 /Ji J1 Jo.1 /JI Number Orientation (lb/in)

(lb/in)

(lb/in)

PB-I SA-847 L

619 331 1.87 388 1.60 PB-I SA-1101 C

609 98 6.21 113 5.39 PB-2 SA-1484 C

579 104 5.57 119 4.87 Table 6-5 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 2 -

Current Power Conditions Without Hafnium Assemblies Lower Bounding SF = 1.15 SF 1.25 Plant Weld Weld Jo.1 at t/4 J

J0.1 /J1 J,

JO/J 1-Number Orientation (lb/in)

(Ib/in)

(lb/in)

PB-i SA-847 L

618 331 1.87 388 1.59 PB-I SA-1101 C

608 98 6.20 113 5.38 PB-2 SA-1484 C

578 104 5.56 119 4.86 Table 6-6 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 3 -

Current Power Conditions With.Hafnium Assemblies Lower Bounding SF = 1.15 SF = 1.25 Plant Weld Weld Jo.1 at t/4 J,

J* /J1 J0.1 /J1 Number Orientation (lb/in)

(lb/in)

(lb/in)

PB-1 SA-847 L

623

.331 1.88 388 1.61 PB-1 SA-1101 C

613 98 6.26 113 5.42 PB-2 SA-1484 C

585 104 5.63 119 4.92 6-3 A

AREVA

BAW-2467NP, Rev. 1 Figure 6-1 J-Integral vs. Flaw Extension for Levels A & B Service Loadings - Evaluation Condition 2 - Current Power Conditions Without Hafnium Assemblies - Weld SA-847 800 750 700 650 600 550 U,

C 500 450 400 350 300 '

0.00 0.05 0.10 0.15 Flaw Extension, Aa (in.)

0.20 0.25 6-4 A

AR EVA

BAW-2467NP, Rev. 1 7.0 Evaluation for Levels C and D Service Loadings A flaw depth of 1/10 of the base metal wall thickness, plus the cladding thickness, is used to evaluate the Level D Service Loadings. The stress intensity factor K, calculated by the PCRIT code is the sum of thermal, residual stress, deadweight, and pressure terms. PCRIT is run for each Level D transient.

RTNDT is also calculated by PCRIT. Transition region toughness is obtained from the ASME Section X1 equation for crack initiation [15].

K1,= 33.2 + 2.806 exp[0.02(T-RTNDT+ 100°F)]

[eqn. 7]

where:

= transition region toughness, ksi~'in T = crack tip temperature, OF Upper-shelf toughness is derived from the J-integral resistance model of Section 3.1 for a flaw depth of 1/10 of the wall thickness, a crack extension of 0.10 in., and fluence, as follows:

Kj=

[eqn. 8]

1000(1 - v2) where K= upper-shelf region toughness, ksi'Jin J=.1 J-integral resistance at Aa = 0.1 in.

Figure 7-1 through 7-3 shows the variation of applied stress intensity factor, K,, transition range toughness, Kc, and upper-shelf toughness, Kjc with temperature for the Evaluation Condition 1 described in Table 2-1 for the three welds. The markers on the K, curve indicate points in time at which PCRIT solutions are available. For all the three welds that were analyzed, the LOCA transient is limiting since it most closely approaches the Kjc limit of each weld. All subsequent analysis will pertain to this transient. In the upper-shelf toughness range, the K, curve is closest to the lower bound Kic curve at a particular time point into the transient for each weld, as listed below:

Weld Time (min)

SA-847 2.40 SA-1011 1.50 SA-1484 1.30 For each weld, the time specified above is selected as the critical time in the transient at which to perform the flaw evaluation for Level D Service Loadings.

7-1 A

AREVA

BAW-2467NP, Rev. 1 Figure 7-1 220 200 180 160 140 120 100

/

80

/

/

60 40 20 K, vs. Crack Tip Temperature for Evaluation Condition 1 - SA-847 C5:

0 L-275 325 375 425 475 Crack Tip Temperature (OF) 525 575 7-2 A

AR EVA

BAW-2467NP, Rev. I Figure 7-2 ' K, vs. Crack Tip Temperature for Evaluation Condition 1 - SA-1 101 220 200 180 160 140 5:

120 100 80 60 40 20 0 "

275 325 375 425 475 Crack Tip Temperature (OF) 525 575 7-3 A

AREVA

BAW-2467NP, Rev. 1 Figure 7-3 K, vs. Crack Tip Temperature for Evaluation Condition I - SA-1484 220 200 180 160 140 IlIi oI eI oI l

I I

Kic KJc Mean

-KJc Lower Bound Upper Shelf Limit KI at a=tIl 0 for 2003 FSAR SLB KI at a=t/1 0 for ESPEC LOCA -

KI at a=t/1 0 for ESPEC SLB I

I I

- ; I**~~*.-..

.1 It

'I I,

II I

Evlainpita 120

.Th 100

/

OU

.,,,ml.

Into trnient~lI 80 60 40 20 x

//

/

I oIi iI Upper-Shelf Toughness Range nl 275 325 375 425 475 Crack Tip Temperature (OF) 525 575 7-4 A

AR EVA

BAW-2467NP, Rev. I Applied J-integrals for the LOCA transient are calculated for each weld at the critical time points identified above for various flaw depths in Table 7-1, 7-2, and 7-3 using stress intensity factors from PCRIT and adding 9.0 ksiIin to account for cladding effects. Stress intensity factors are converted to J-integrals by the plain strain relationship, K,~~ (a)

V2 Japr,,ea (a)a)E

("-v2 )

[eqn. 9]

E Tables 7-1, 7-2, and 7-3 lists flaw extensions vs. applied J-integrals.

As the Point Beach vessels are 6.5 in. thick, the initial flaw depth of 1/10 of the wall thickness is 0.65 in.

Flaw extension from this flaw depth is calculated by subtracting 0.65 in. from the built-in PCRIT flaw depths in the base metal. The results, along with mean J-R curve, are plotted in Figure 7-4.

This figure indicates that Weld SA-847 is limiting as the ratio of the applied J-integral to the material J-R curve is less than the other two welds. Figure 7-5 is a plot of the applied J-integrals and the mean J-R curves for the three Evaluation Conditions from Table 2-1 for Weld SA-847. Evaluation Condition 1, uprated power conditions without hafnium power suppression assemblies, is the limiting case as the ratio of the mean J-R curves to applied J-integrals is the minimum of the three Evaluation Conditions. The requirements for ductile and stable crack growth are demonstrated by Figure 7-5 since the slopes of the applied J-integral curves are considerably less than the slopes mean J-R curves at the points of intersection. The Level D Service Loading requirement that the extent of stable flaw extension be no greater than 75% of the vessel wall thickness is easily satisfied since the applied J-integral curves intersects the mean J-R curves at flaw extensions that are only a small fraction of the wall thickness (less than 1%).

The last requirement is that the internal pressure p shall be less than PI, the internal pressure at tensile instability of the remaining ligament. Table 7-4 gives the results of the calculations for P/

for flaw depths up to 1.365 inches for Evaluation Condition 1. As the internal pressure p is less than PI, the remaining ligament is not subject to tensile instability.

7-5 A

AREVA

BAW-2467NP, Rev. 1 Table 7-1 J-Integral vs. Flaw Extension for Evaluation Condition I - SA-847 Time=

2.40 min E= 26751.6 ksi Crack tip at tV10

t.

6.5 in.

V =

0.3 (a*+/t)*40 a++

Aa Temp.

K1

ým I KIclad

K1tota, Japp (in.)

(in.)

(F)

(lb/in) 1 0.1625 246.40 62.06 9.0 71.1 172 2

0.3250 274.80 83.65 9.0 92.7 292 3

0.4875 302.10 94.64 9.0 103.6 365 4

0.6500 0.0000 328.00 100.97 9.0 110.0 411 5

0.8125 0.1625 352.70 104.24 9.0 113.2 436 6

0.9750 0.3250 375.90 105.82 9.0 114.8 448 7

1.1375 0.4875 397.70 106.12 9.0 115.1 451 8

1.3000 0.6500 417.90 105.76 9.0 114.8 448 9

1.4625 0.8125 436.50 104.86 9.0 113.9 441 10 1.6250 0.9750 453.60 103.22 9.0 112.2 428 12 1.9500 1.3000 483.10 98.74 9.0 107.7 395 14 2.2750 1.6250 507.00 93.05 9.0 102.1 354 16 2.6000 1.9500 525.80 88.28 9.0 97.3 322 18 2.9250 2.2750 540.10 82.87 9.0 91.9 287 20 3.2500 2.6000 550.70 77.27 9.0 86.3 253 22 3.5750 2.9250 558.40 71.71 9.0 80.7 222 24 3.9000 3.2500 563.90 66.53 9.0 75.5 194 26 4.2250 3.5750 567.60 61.81 9.0 70.8 171 28 4.5500 3.9000 570.00 57.20 9.0 66.2 149 30 4.8750 4.2250 571.60 52.58 9.0 61.6 129 32 5.2000 4.5500 572.60

• 48.13 9.0 57.1 111 Note:

a" is the flaw depth in the base metal 7-6 A

AR EVA

BAW-2467NP, Rev. 1 Table 7-2 J-Integral vs. Flaw Extension for Evaluation Condition I - SA-1 101 Time =

1.50 min E = 26751.6 ksi Crack tip at t/10 t=

6.5 in.

V =

0.3 (a+/t)*40 a

Aa Temp.

Klsum Kldad KItotai Japp (in.)

(in.)

(F)

(lb/in) 1 0.1625 280.80 59.65 9.0 68.7 160 2

0.3250 314.80 78.57 9.0 87.6 261 3

0.4875 346.70 86.65 9.0 95.7 311 4

0.6500 0.0000 376.30 90.22 9.0 99.2 335 5

0.8125 0.1625 403.60 91.26 9.0 100.3 342 6

0.9750 0.3250 428.40 90.74 9.0 99.7 338 7

1.1375 0.4875 450.60 89.06 9.0 98.1 327 8

1.3000 0.6500 470.50 86.71 9.0 95.7 312 9

1.4625 0.8125 488.00 83.66 9.0 92.7 292 10 1.6250 0.9750 503.10 80.42 9.0 89.4 272 12 1.9500 1.3000 527.20 72.98 9.0 82.0 229 14 2.2750 1.6250 544.30 65.06 9.0 74.1 187 16 2.6000 1.9500 555.90 57.27 9.0 66.3 149 18 2.9250 2.2750 563.40 49.24 9.0 58.2 115 20 3.2500 2.6000 568.10 41.31 9.0 50.3 86 22 3.5750 2.9250 570.90 34.09 9.0 43.1 63 24 3.9000 3.2500 572.40 27.47 9.0 36.5 45 26 4.2250 3.5750 573.30 21.94 9.0 30.9 33 28 4.5500 3.9000 573.70 17.63 9.0 26.6 24 30 4.8750 4.2250 573.90 14.36 9.0 23.4 19 32 5.2000 4.5500 574.00 11.59 9.0 20.6 14 Note:

a+ is the flaw depth in the base metal 7-7 A

AREVA

BAW-2467NP, Rev. 1 Table 7-3 J-Integral vs. Flaw Extension for Evaluation Condition I - SA-1484 Time =

1.30 min E = 25459.9 ksi Crack tip at t/10 t

6.5 in.

V=

0.3 (a++/t)*40 a++

Aa Temp.

KsUM Kiclad

Kliota, Japp (in.)

(in.)

(F)

(lb/in) 1 0.1625 292.60 51.19 9.0 60.2 129 2

0.3250 328.30 67.16 9.0 76.2 207 3

0.4875 361.60 73.97 9.0 83.0 246 4

0.6500 0.0000 392.10 76.91 9.0 85.9 264 5

0.8125 0.1625 419.80 77.72 9.0 86.7 269 6

0.9750 0.3250 444.70 77.16 9.0 86.2 265 7

1.1375 0.4875 466.60 75.59 9.0 84.6 256 8

1.3000 0.6500 485.80 73.43 9.0 82.4 243 9

1.4625 0.8125 502.50 70.67 9.0 79.7 227 10 1.6250 0.9750 516.40 67.71 9.0 76.7 210 12 1.9500 1.3000 538.10 61.07 9.0 70.1 175 14 2.2750 1.6250 552.60 54.04 9.0 63.0 142 16 2.6000 1.9500 561.80 47.18 9.0 56.2 113 18 2.9250 2.2750 567.40 40.21 9.0 49.2 87 20 3.2500 2.6000 570.60 33.42 9.0 42.4 64 22 3.5750 2.9250 572.40 27.38 9.0 36.4 47 24 3.9000 3.2500 573.30 21.99 9.0 31.0 34 26 4.2250 3.5750 573.80 17.69 9.0 26.7 25 28 4.5500 3.9000 574.00 14.53 9.0 23.5 20 30 4.8750 4.2250 574.00 12.34 9.0 21.3 16 32 5.2000 4.5500 574.10 10.58 9.0 19.6 14 Note:

a-is the flaw depth in the base metal 7-8 A

AR EVA

BAW-2467NP, Rev. 1 Table 7-4 Level D Service Loadings - Internal Pressure at Tensile Instability - SA-847 flaw depth a (in.)

P, (ksi) 0.0650 9.18 0.1300 9.16 0.1950 9.14 0.2600 9.12 0.3250 9.09 0.3900 9.06 0.4550 9.02 0.5200 8.98 0.5850 8,93 0.6500 8.88 0.7150 8.84 0.7800 8.78 0.8450 8.73 0.9100 8.68 0.9750 8.62 1.0400 8.56 1.1050 8.51 1.1700 8.45 1.2350 8.39 1.3000 8.32 1.3650 8.26 7-9 A

AR EVA

BAW-2467NP, Rev. 1 Figure 7-4. J-Integral vs. Flaw Extension - All Welds 1600 1400 1200 1000 C

I-0) ci)

C 800 600 400 200 04-0.00 0.05 0.10 0.15 0.20 0.25 0.30 Flaw Extension, Aa (in.)

0.35 0.40 0.45 0.50 7-10 A

AREVA

/

BAW-2467NP, Rev. 1 Figure 7-5. J-lntegral vs. Flaw Extension - Weld SA-847 1600 1400 1200 1000

-u--J-R Integral Uprated Without Hafnium J-R Integral Current Without Hafnium J-R Integral Current With Hafnium

-E-J applied Uprated Without Hafnium S800 -

-

• J applied Current Without Hafnium

-e-J applied Current With Hafnium 600 400 200 0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 Flaw Extension, Aa (in.)

7-11 A

AR EVA

BAW-2467NP, Rev. 1 8.0 Summary of Results A low upper-shelf toughness fracture mechanics analysis has been performed to evaluate the reactor vessel welds at Point Beach Units 1 and 2 for projected low upper-shelf energy levels at 53 EFPY, considering Levels A, B, C, and D Service Loadings of the ASME Code.

Evidence that the ASME Code, Section Xl, Appendix K [4] acceptance criteria have been satisfied for Levels A and B Service Loadings is provided by the following:

(1)

The limiting weld is the axial weld SA-847 of Point Beach Unit I in the current power condition without hafnium power suppression assemblies.

Figure 6-1 shows that with factors of safety of 1.15 on pressure and 1.0 on thermal loading,.

the applied J-integral (J1) is less.than the J-integral of the material at a ductile flaw extension of 0.10 in. (Jo.1). The ratio Jo.11J1 = 1.87 which is significantly greater than the required value of 1.0.

(2)

Figure 6-1 shows that with a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect.

Evidence that the ASME Code, Section Xl, Appendix K [4] acceptance criteria have been satisfied for Level D Service Loadings is provided by the following:

(1)

Figure 7-5 shows that with a factor of safety of 1.0 on loading, flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slopes of both the lower bound and mean J-R curves at the points of intersection.

(3)

Figure 7-5 shows that the flaw remains stable at much vessel wall thickness.

It has also been shown that the sufficient to preclude tensile instability by a large margin.

8-1 less than 75% of the remaining ligament is A

AREVA

BAW-2467NP, Rev. I 9.0 Conclusion The limiting Point Beach Units 1 and 2 reactor vessel beltline weld (axial weld SA-847 of Unit 1) satisfies the acceptance criteria of Appendix K to Section Xl of the ASME Code [4] for projected low upper-shelf Charpy impact energy levels at 53 effective full power years of plant operation for the three conditions evaluated: uprated power conditions (1678 MWt) without hafnium power suppression assemblies, current power conditions (1540 MWt) without hafnium power suppression assemblies, and current power conditions (1540 MWt) with hafnium power suppression assemblies.

9-1 A

AREVA

BAW-2467NP, Rev. 1 10.0 References

1. BAW-2192PA, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group For Level A & B Service Loads, April 1994.

2. BAW-2178PA, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group For Level C & D Service Loads, April 1994.

3. BAW-2255, Effect of Power Upgrade on Low Upper-Shelf Toughness Issue, May 1995.

4. ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition with Addenda through 2000.

5. USNRC Reactor Vessel Integrity Database Version 2.0.1 (RVID).

6. BAW-2275, Low Upper-Shelf Toughness Fracture Mechanics Analysis of B&W Designed Reactor Vessels for 48 EFPY, August 1996.

7. BAW-2312, Revision 1, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of Turkey Point Units 3 and 4 for Extended Life through 48 Effective Full Power Years, December 2000.

8. BAW-2150, Materials Information for Westinghouse-Designed Reactor Vessels Fabricated by B&W, December 1990.

9. ASME Boiler and Pressure Vessel Code,Section III, Appendices, 1989 Edition with no Addenda.

10. WCAP-1 3902, Analysis of Capsule S from the Rochester Gas and Electric Corporation R. E. Ginna Reactor Vessel Radiation Surveillance Program, December 1993.

11. WCAP-1 5916, Analysis of Capsule X from the Florida Power and Light Turkey Point 3 Reactor Vessel Radiation Surveillance Program, September 2002.

12. BAW-2254, Test Results of Capsule CR3-LG2: B&W Owners Group - Master Integrated Reactor Vessel Surveillance'Program, October 1995.

13. Point Beach Nuclear Plant Units I and 2 Final Safety Analysis Report, June 2003.

14. ASME Boiler and Pressure Vessel Code, Appendix K, Section Xl, 2001 Edition.

15. EPRI NP-719-SR, T.U. Marston, Flaw Evaluation Procedures: ASME Section Xl, Electric Power Research Institute, Palo Alto, California, August 1978.

10-1 A

AREVA

BAW-2467NP, Rev. 1 11.0 Certification This report is an accurate description of the low upper-shelf toughness fracture mechanics analysis performed for the reactor vessels-at Point Beach..

10 1II5-4 H. P. Gunawardane, Engineer III Materials and Structural Analysis Unit Date This report has been reviewed and found to be an accurate description of the low upper-shelf toughness fracture mechanics analysis performed for the reactor vessels at Point Beach.

A. D. Nana, Principal Engineer Date Materials and Structural Analysis Unit Verification of independent review.

1Co// 57o/0

. A. D. McKim, Manager Materials and Structural Analysis Unit Date This report is approved for release.

fO. Austin, Project Development Manager Date 11-1 A

AREVA

BAW-2467NP, Rev. I 12.0 Appendix A The following pages contain input information from Nuclear Management Company.

12-1 A

ARE VA

BAW-2467NP, Rev. 1 Committed to Nuclear Excellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC NPL 2004-0139 June 29, 2004 Heshan Gunawardane AREVA / Framatome ANP, Inc.

MS OF50 3315 Old Forest Road Lynchburg, VA 24501 Heshan:

This correspondence will serve to formally document the requested inputs for the PBNP Units 1 and 2 RPV Equivalent Margins Assessment that is being performed in accordance with AREVA.Proposal FANP-04-1067, April 2,2004.

Applicable ASME Section XI Code The PBNP ISI Program is in the fourth ten-year interval, which began on July 1, 2002 for both PBNP-1 and PBNP-2. The program is in accordance with the 1998 edition through 2000 addenda.(98A00) of ASME Section XI Code as modified by 10 CFR 50.55a and approved relief requests and code cases.

(Reference 1)

Fluence Projections For the case of full uprated power condition (1678 MWt), without hafnium absorber assemblies, for EOLE (53 EFPY) use the older talculated fluence projections contained in Section 2 of Reference 2.

This is requested for input consistency with the remaining RV embrittlement analyses.

For the cases of mini uprated power condition (1540 MWt), with and without hafnium absorber assemblies, for EOLE (53 EFPY) use the revised calculated fluence projections contained in Section 2 of Reference 3.

6590 Nuclear Road 9 Two Rivers, Wisconsin 54241 Telephone: 920.755.2321 12-2 R A AR EVA

BAW-2467NP, Rev. I NPL 2004-0139 June 29, 2004 Page 2 Normal Heatup and Cooldown Rates The PBNP RCS heatup and cooldown rates for normal operation are 100 degrees Fahrenheit per hour for both heatups and cooldowns. (Reference 4)

Predicted Operating Temperatures The analyses for current licensed rated power conditions (1540 MWt) include a range of full load T(avg)'s from 558.1 to 574 degrees Fahrenheit. The resulting T(hot) and T(cold) ranges are 588.1 to 603.5, and 528 to 544.5 degrees Fahrenheit, respectively (Reference 5). PBNP currently uses a T(avg) program of 547 to 570 degrees Fahrenheit (no load to full load) (Reference 6), resulting in a T(hot) and T(cold) of approximately 597 and 542 degrees Fahrenheit, respectively (Reference 7).

The analyses for the 10.5 percent uprated power condition (1678 MWt) include a range of T(avg) from 558.6 to 573.4 degrees Fahrenheit. The resulting T(hot) and T(cold) ranges are 591.2 to 605.5, and 526 to 541.4 degrees Fahrenheit, respectively (Reference 8).

Transient Information The original component transients are defined in each RPV design specification (References 9 and 10 for Units 1 and 2, respectively). A revised set of component design transients was generated to support steam generator replacement, a partial power uprate (8.7 percent), and license renewal (Reference 11).

The RPV transients weie evaluated and characterized for the partial power uprated condition in Reference 12. The RPV transients were further evaluated and characterized for full uprated conditions in Reference 13.

In addition, Chapter 14 of the PBNP FSAR (Reference 14) has been provided via previous correspondence. Chapter 14 containsthe PBNP safety analysis summaries. These transients should be reviewed for bounding conditions with respect to the component design transients.

Applicable ASME Section II and Ill Code ASME Boiler and Pressure Vessel Code,Section II, 1989, no Addenda.

ASME Boiler and Pressure Vessel Code, Section 111, 1989, no Addenda.

A 12-3 AREVA*

BAW-2467NP, Rev. 1 NPL 2004-0139 June 29,2004 Page 3 Sincerely, Brad Fromm PBNP License Renewal Nuclear Management Company James E. Knorr Manager of License Renewal PBNP Nuclear Management Company bins

References:

1. SER 2001-0010, "Point Beach Nuclear Plant, Units 1 and 2 - Relief Requests RR 1-24 (Unit 1)

And RR-2-30 (Unit 2) Re: Use Of ASME Code Section XI, 1998 Edition With Addenda Through 2000 (TAC Nos. MB2230 And MB223 1)", dated November 6, 2001.

2. Westinghouse Letter Report, LTR-REA-02-23, "Pressure Vessel Neutron Exposure Evaluations, Point Beach Units 1 and 2, S. L. Anderson, Radiation Engineering and Analysis, February 2002.

3. Westinghouse Letter Report, LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluations, Point Beach Units 1 and 2, S. L. Anderson, Radiation Engineering and Analysis, June 2004.

4. Point Beach Nuclear Plant Technical Requirements Manual Pressure Temperature Limits Report, Section 2.1, "RCS Pressure and Temperature Limits (LCO 3.4.3)", page 2.2-2, Revision 1, dated December 20, 2002.

5. NMC Letter, NRC 2002-0075, "Responses to Requests for Additional Information, License Amendment Request 226, Measurement Uncertainty Recapture Power Uprate", August 29,2002.

6. Setpoint Document, STPT 5.1, "Primary Control Systems Rod Speed Control", Revision 7.

A 12-4 AREVA

BAW-2467NP, Rev. 1 NPL 2004-0139 June 29, 2004 Page 4

7. Internal PBNP email, Steve Barkhahn to Brad Fromm, dated 4/17/04.

.8. Westinghouse, Power Uprate Project, PBNP Units 1 and 2, Volume 1 NSSS Engineering Report, and Volume 2 BOP Engine6ring Report, April 2002.

9. Section 4 of Westinghouse Equipment Specification G - 676243, "Reactor Coolant System -

Reactor Vessel", Revision 0, 05/0511966.

10. Section 4, and Figures 1 through 15 of Westinghouse Equipment Specification E-spec 677456, "Addendum to Equipment Specification 676413 Rev. 1, Reactor Coolant System - Reactor Vessel", Revision 2, 07/06/1971.

11. Appendix A of Westinghouse Design Specification, 414A83, "Point Beach Nuclear Plants Units 1 and 2, replacement Reactor Vessel Closure Head (RRVCH)", Revision 0.

12. Appendix B of WCAP-14448, "Addendum to the Stress Reports for the Point Beach Unit Nos. 1 and 2 Reactor Vessels (RSG/Uprating Evaluation), August 1995.

13. Section 5.1.4 of Westinghouse Report, "Tower Uprate Project, Point Beach Nuclear Plant, Units I and 2, NSSS Engineering Report", April 2002.

14. Chapter 14 of the PBNP Units I and 2 Final Safety Analysis Report, June, 2003.

Notes:

References 1, 4, 5, 6, 7, 8, and 14 document the sources of the information.

References 2, 3, 9, 10, 11, 12, and 13 are enclosed.

References 9, 10, 11, 12, and 13 are Westinghouse Proprietary and shall be treated in accordance with the associated Westinghouse Proprietary Agreement established between AREVA/Framatome-ANP, NMC, and Westinghouse in June 2004.

A 12-5 AREVA

BAW-2467NP, Rev. I 13.0 Appendix B The following page contaihs input information from Nuclear Management Company.

13-1 A

AR EVA

BAW-2467NP, Rev. 1 NMC-P Committed to Nuclear Excellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC NPL 2004-0236 October 14, 2004 Heshan Gunawardane AREVA / Framatome ANP, Inc.

MS OF50 3315 Old Forest Road Lynchburg, VA 24501 Heshan:

Subject:

PBNP Units 1 and 2 Equivalent Margins Assessment Revision, Framatome ANP, Inc. Proposal Number 416 0645, Addendum No. I This correspondence will serve to formally document NMC's request to revise the PBNP Units 1 and 2 RPV Equivalent Margins Assessment, Frametome ANP, Inc. Calculation Numbers 77-2647-00 and 77-2647NP-00, to use the 2004 Westinghouse fluence projection as the input to Evaluation Condition

1. Evaluation Condition 1 is full uprated power (1678 MWt), without the presence of Hafnium power suppression inserts.

Sincerely, Brad Fromm PBNP License Renewal Nuclear Management Company' ohn G. Thorger4sfor James E. Knorr Manager of License Renewal PBNP Nuclear Management Company bins 6590 Nuclear'Road - Two Rivers, Wisconsin 54241 Telephone: 920.755.2321 A

13-2 AREVA

ENCLOSURE 4 TO REQUEST FOR REVIEW OF REACTOR VESSEL TOUGHNESS FRACTURE MECHANICS ANALYSIS WESTINGHOUSE DOCUMENT, WEP-06-33 P ATTACHMENT, "TRANSMITTAL OF FIGURE 5-1 REACTOR COOLANT TEMPERATURE AND PRESSURE VS. TIME LEVEL D TRANSIENTS",

DATED MAY 10, 2006 (PROPRIETARY)

WESTINGHOUSE DOCUMENT, WEP-06-33 NP ATTACHMENT, "TRANSMITTAL OF FIGURE 5-1 REACTOR COOLANT TEMPERATURE AND PRESSURE VS. TIME LEVEL D TRANSIENTS",

DATED MAY 8,2006 (NON -PROPRIETARY)

WESTINGHOUSE AUTHORIZATION LETTER AFFIDAVIT PROPRIETARY INFORMATION NOTICE COPYRIGHT NOTICE (11 pages follow)

WWestinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA May 8, 2006 WEP-06-33 Mr. Jack Gadzala Nuclear Management Company Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, WI 54241 Nuclear Management Company Point Beach Units 1 and 2 Transmittal of Figure 5-1 Reactor Coolant Temperature and Pressure vs.

Time Level D. Transients

Dear Mr. Gadzala:

In response to your request attached please find proprietary and non-proprietary versions of the subject Figure 5-1 along with the appropriate documents to submit this information to the NRC.

If you have any questions regarding the attached, please call Mike Miller at 412-374-3353..

Very truly yours, WESTINGHOUSE ELECTRIC COMPANY Kerry B. Hanahan Customer Project Manager ElectronicallyApproved Records are Authenticated in the Electronic Document Management System A BNFL Group company

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WEP-06-33 NP Attachment Transmittal of Figure 5-1 Reactor Coolant Temperature and Pressure vs. Time Level D Transients May 8,2006 Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

© 2006 Westinghouse Electric Company LLC All Rights Reserved Page 1 of 2

Figure 5-1 Level D transients - Reactor Coolant Temperature and Pressure vs. Time b,c WEP-06-33-NP Attachment Page 2 of 2

Page 2 of 2 WEP-06-33 bcc: M. Miller R. Fagan

• Westinghouse Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA (412) 374-4419 (412) 374-4011 maurerbf@westinghouse.com U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Directtel:

Direct fax:

e-mail:

Our ref CAW-06-2141 May 12,2006 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

WEP-06-33 P-Attachment, "Transmittal of Figure 5-1 Reactor Coolant Temperature and Pressure vs. Time Level D Transients" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-06-2141 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (bX4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Nuclear Management Company.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-06-2141 and should be addressed to B. F. Maurer, Acting Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O.

Box 355, Pittsburgh, Pennsylvania 15230-0355.

Very truly yours, B. F. Maurer, Acting Manager Regulatory Compliance and Plant Licensing Enclosures A BNFL Group company

CAW-06-2141 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

s$

COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared B. F. Maurer, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

B. F. Maurer, Acting Manager Regulatory Compliance and Plant Licensing Sworn to and subscribed before me this

,=

"tay of 2006 Notary Public Notarial Seal Sharon L FRod. Notary Public Monreville Boro, Alegheny County My Commission Eires January 29,i20M Memtner, Pennsyvania Assnýtb-aJn Of No~taies

2 2CAW-06-2141 (1)

I am Acting Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2)

I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse "Application for Withholding" accompanying this Affidavit.

(3)

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii)

The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence.

The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as... iuws.

(a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's

3 CAW-06-2141 competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a)

The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b)

It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c)

Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

4 CAW-06-2141 (d)

Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e)

Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f)

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii)

The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv)

The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v)

The proprietary information sought to be withheld in this submittal is that which is appropriately marked in WEP-06-33 P-Attachment, "Transmittal of Figure 5-1 Reactor Coolant Temperature and Pressure vs. Time Level D Transients" (Proprietary) for submittal to the Commission, being transmitted by Nuclear Management Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with Nuclear Management Company's request for NRC approval of BAW-2467P, Revision 1 October 2004.

I 1

W h

AAjAl"CLWA~A &a

.Jkt V

L

&11"L hI1L11 vviII ralauiC,veLinV 0 ue to:

(a) Facilitate NMC in obtaining NRC approval of WEP-06-33 P-Attachment, "Transmittal of Figure 5-1 Reactor Coolant Temperature and Pressure vs. Time Level D Transients" (Proprietary).

5 5

~CAW-06-2 141 Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of this information to its customers for purposes of meeting NRC requirements for licensing documentation.

(b) Westinghouse can sell support and defense of the use of this information to its customers in the licensing process.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the abi~lity of competitors to provide similar calculations and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

Proprietary Information Notice Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the aff idavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.