View source for RS-04-075, Request for a License Amendment to Incorporate Approved Pressure and Temperature Limits Report (PTLR) Methodology Into Technical Specifications

{{Adams
| number = ML041470212
| issue date = 05/21/2004
| title = Request for a License Amendment to Incorporate Approved Pressure and Temperature Limits Report (PTLR) Methodology Into Technical Specifications
| author name = Ainger K
| author affiliation = Exelon Generation Co, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000454, 05000455, 05000456, 05000457
| license number = NPF-037, NPF-066, NPF-072, NPF-077
| contact person = 
| case reference number = FOIA/PA-2005-0108, RS-04-075
| document type = Letter type:RS, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Topical Report
| page count = 282
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Exelon Generation                    www.exeloncorp.com Exelkni 4300 Winfield Road                                                                          Nuclear Warrenville, IL 60555 10 CFR 50.90 RS-04-075 May 21, 2004 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

==Subject:==
Request for a License Amendment to Incorporate Approved Pressure and Temperature Limits Report (PTLR) Methodology into Technical Specifications

==Reference:==
Letter from U. S. NRC to O. D. Kingsley (Exelon Generation Company, LLC),
Issuance of Exemption from the Requirements of 10 CFR 50 Part 60 and Appendix G for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2," dated August 8, 2001 In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) is requesting an amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively.
The proposed amendment would revise TS 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," by adding the referenced letter as an acceptable method for determining reactor pressure vessel (RPV) pressure temperature (P-T) limits. The NRC approved the use of Code Cases N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section Xl, Division 1," and N-588, "Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels Section Xl, Division 1," for Braidwood Station, Units 1 and 2 and Byron Stations, Units 1 and 2 in the referenced letter.
The attached amendment request is subdivided as shown below.
Attachment 1 provides an evaluation of the proposed changes.
Attachments 2-A and 2-B include the marked-up TS pages with the proposed changes indicated for Braidwood Station and Byron Station, respectively.

U. S. Nuclear Regulatory Commission May 21, 2004 Page 2 Attachments 3-A and 3-B include the associated typed TS pages with the proposed changes incorporated for Braidwood Station and Byron Station, respectively.
Attachments 4-A through 4-D include reports developed by Westinghouse Electric Company (Westinghouse) that contain revised P-T limit curves for Braidwood Station and Byron Station. These reports are provided to demonstrate the results of applying Code Cases N-588 and N-640 to the currently approved methodology for developing the P-T limits. These reports also apply an additional allowance regarding the RPV closure flange, as discussed in Attachment 1, that has not been approved by the NRC. NRC approval of the reports is not being requested.
Attachment 5 provides a comparison of the 22 effective full power years (EFPY) data points in Attachments 4-A through 4-D to the data points of the existing P-T limit curves for Braidwood and Byron Stations. The 22 EFPY data points have been modified to restore the minimum flange temperature requirement of 10 CFR Appendix G Table 1 for each of the Braidwood and Byron Units.
Attachments 6-A and 6-B include P-T limit curves with revised applicability dates for Braidwood and Byron Stations, respectively. NRC approval of these curves is not being requested, since P-T limit curves are not part of the Braidwood and Byron Station TS.
The current Byron Station, Unit 1 P-T limit curves will reach the limiting reactor vessel neutron fluence on approximately November 3, 2004. Thus, we request approval of the proposed amendment by October 22, 2004. Once approved, the amendment will be implemented in a timeframe to accommodate Byron Station, Unit 1.
The proposed amendment has been reviewed by the Braidwood Station and the Byron Station Plant Operations Review Committees and approved by their respective Nuclear Safety Review Boards in accordance with the requirements of the EGC Quality Assurance Program.
EGC is notifying the State of Illinois of this application for a change to the TS by sending a copy of this letter and its attachments to the designated State Official.
Should you have any questions concerning this letter, please contact D. J. Chrzanowski at (630) 657-2816.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on            ;      2/.20&deg;YGV4 Kenneth A. Ainger Manager, Licensing : Evaluation of Proposed Changes -A: Markup of Proposed Technical Specifications Page Changes for Braidwood Station -B: Markup of Proposed Technical Specifications Page Changes for Byron Station -A: Typed Pages for Technical Specification Changes for Braidwood Station -B: Typed Pages for Technical Specification Changes for Byron Station

U. $. Nuclear Regulatory Commission May 21, 2004 Page 3 -A: WCAP-1 5364, Revision 2, Braidwood Unit 1 Heatup and Cooldown Limit Curves for Normal Operation -B: WCAP-15373, Revision 2, Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation -C: WCAP-1 5391, Revision 1, Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation -D: WCAP-1 5392, Revision 2, Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation : Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points -A: Revised P-T Limit Curves for Braidwood Station, Units 1 and 2 -B: Revised P-T Limit Curves for Byron Station, Units I and 2

I ATTACHMENT ATTACHMENT 1 Evaluation of Proposed Changes INDEX

==1.0 DESCRIPTION==

==2.0 PROPOSED CHANGE==

==3.0 BACKGROUND==

==4.0 TECHNICAL ANALYSIS==

==5.0 REGULATORY ANALYSIS==

5.1  No Significant Hazards Consideration.
5.2  Applicable Regulatory Requirements/Criteria

==6.0 ENVIRONMENTAL CONSIDERATION==

==7.0 REFERENCES==

Page 1 of 7

ATTACHMENT I Evaluation of Proposed Changes

==1.0     DESCRIPTION==

In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) is requesting an amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively. The proposed amendment would revise TS 5.6.6, "Pressure Temperature Limits Report," by adding the referenced letter to the list of approved methods for determining reactor pressure vessel (RPV) pressure temperature (P-T) limits. The referenced letter documents NRC approval of the use of American Society of Mechanical Engineers (ASME) Code Cases N-640, "Alternative Reference Fracture Toughness for Development of P-.
T Limit Curves for Section Xl, Division 1," and N-588, 'Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels Section Xl, Division 1," for Braidwood and Byron Stations.
By adding this acknowledgement to our approved methodology, EGC would have the basis to perform a change to the Braidwood and Byron Station PTLRs under 10 CFR 50.59, "Changes, tests, and experiments," to extend the applicability time of the current Braidwood and Byron heat up and cool down curves by two effective full power years (EFPY). This would provide a reasonable limit that would allow continued safe operation while accommodating a review and approval schedule for a future licensing action. The future submittal request would provide a completely updated set of heat-up and cooldown curves and low temperature overpressure protection setpoints based on the application of ASME Code Case N-640, ASME Code Case N-588 and, an exemption to 10 CFR 50, Appendix G, "Fracture Toughness Requirements,"
paragraph IV.A.2.c (minimum flange temperature requirements) as contained in WCAP-16143-P, "ReactorVessel Closure HeadNessel Flange Requirement Evaluation for Byron/Braidwood Units 1 and 2. This revised methodology would provide for Braidwood and Byron Station Pressure and Temperature Limit Reports that would be effective for 32 EFPY.
We request approval of the proposed amendment by October.22, 2004. Once approved, the amendment will be implemented in a timeframe necessary to accommodate the current Byron Station, Unit 1 Pressure and Temperature Limit Report expiration date.

==2.0      PROPOSED CHANGE==

TS Section 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," part b, states the following.
The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those' described in NRC letter dated January 21, 1998, "Byron Station Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report"; and The proposed change revises TS 5.6.6.b to state the following.
The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in NRC letters dated January 21, 1998, "Byron Station Units 1 and 2, and Page 2 of 7

ATTACHMENT I Evaluation of Proposed Changes Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report," and August 8, 2001, "Issuance of Exemption from the Requirements of 10 CFR Part 50 and Appendix G for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2"; and

==3.0      BACKGROUND==

Braidwood Station and Byron Station TS have adopted the provisions of NRC Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," to include the pressure-temperature (P-T) limit curves in a PTLR controlled by the licensee. TS 5.6.6 requires that the P-T limits be developed using the specific NRC-approved methodology described in TS 5.6.6.b.
The current Braidwood and Byron Station P-T limit curves were developed using the methodology described in the January 21, 1998, letter referenced in TS 5.6.6.b, which involves the use of the ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G, 1996 Addenda, the use of Code Case N-514, and use of the methodology in the 1996 ASME Code Section Xl, Appendix G, Article G-2000.
The current Braidwood and Byron Station P-T limit curves are applicable for RPV exposures less than or equal to 14 effective full power years (EFPY) for Braidwood Statioln Units 1 and 2, 15.6 EFPY for Byron Station, Unit 1, and 15.5 EFPY for Byron Station, Unit 2. ,The first unit to reach the limiting exposure for the P-T curves is Byron Station, Unit 1, which is' expected to reach 15.6 EFPY on approximately November 3, 2004.
In the referenced letter, the NRC granted Braidwood and Byron Stations an exe~ption from 10 CFR 50, Appendix G, "Fracture Toughness Requirements," to allow application of Code Cases N-640 and N-588 to determine the P-T limits. Code Case N-588 allows postula ion of a circumferentially-oriented reference flaw as the limiting flaw in an RPV circumferential weld in lieu of an axially-oriented flaw required by the 1995 Edition (1996 Addenda) of ASME Section Xl, Appendix G. Code Case N-640 allows use of the plane strain fracture toughness (K1,)curve in lieu of the crack arrest fracture toughness (Ku) curve as the lower bound for fracture toughness.
Braidwood and Byron Station TS Section 5.6.6 requires that the methods used to develop the P-T limit curves be those specifically listed in TS 5.6.6.b. Thus, the TS require revision to include the approval granted in the referenced letter.

==4.0      TECHNICAL ANALYSIS==

Use of Code Cases N-588 and N-640 has been determined to be acceptable for Braidwood and Byron Stations, as documented in the referenced letter. Thus, the addition of these methods to TS 5.6.6 represents an essentially administrative change to the TS.
Westinghouse Electric Company (Westinghouse) developed P-T limit curves for Braidwood and Byron Stations using the methods described in the referenced letter and, additionally, applying Code Cases N-588 and N-640. Westinghouse developed the curves using updated RPV fluence values developed from RPV material surveillance data. These curves are provided for Page 3 of 7

ATTACHMENT I Evaluation of Proposed Changes information as Westinghouse reports for each unit at Braidwood and Byron Station in Attachments 4-A through 4-D. These reports are provided to demonstrate the results of applying Code Cases N-588 and N-640 to the currently approved methodology for developing the P-T limits. NRC approval of the reports is not being requested, since P-T limit curves are not part of the Braidwood and Byron Station TS.
To support future licensing actions, the P-T limit curves in Attachments 4-A though 4-D were developed by eliminating the requirement of 10 CFR 50 Appendix G (Table 1, Note 2) that the minimum temperature of the highly stressed region of the closure flange exceed the unirradiated reference nil-ductility transition temperature (RTNDT) by 1201F for normal operation when the RPV pressure exceeds 20 percent of the pre-service hydrostatic pressure. The NRC has not approved the elimination of the requirement of Table 1, Note 2 for Braidwood and Byron Stations, and EGC is not requesting approval of the elimination of this requirement in these proposed changes. Attachment 5 provides a comparison of the 22 Effective Full Power Years (EFPY) data points in Attachments 4-A through 4-D to the data points of the existing P-T limit curves for Braidwood and Byron Stations. The 22 EFPY data points have been modified to restore the minimum flange temperature requirement of 10 CFR Appendix G Table I for each of the Braidwood and Byron Units.
Upon NRC approval of the proposed change to add the referenced letter to TS 5.6.6.b, Braidwood and Byron Stations will revise the current P-T limit curves by extending the current P-T limit curves an additional two EFPY, as provided in Attachments 6-A and 6-B, respectively.
The P-T limit curves in Attachments 6-A and 6-B are applicable until RPV exposures of 16 effective EFPY for Braidwood Station Units 1 and 2, 17.6 EFPY for Byron Station, Unit 1, and 17.5 EFPY for Byron Station, Unit 2. These curves provide more conservative P-T limits (i.e.,
higher required minimum temperature for a given pressure) than the curves in Attachments 4-A through 4-D for 22 EFPY. Thus, the P-T limit curves provided in Attachments 6-A and 6-B are acceptable for an additional two EFPY. These curves are being provided for information and NRC approval of these P-T limit curves is not being requested. Braidwood and Byron Stations will submit the revised P-T limit curves as part of the PTLR in accordance with the requirements of TS 5.6.6.c.
In a future licensing action, EGC plans to request NRC approval to eliminate the requirement of 10 CFR 50 Appendix G Table 1, Note 2, for Braidwood and Byron Stations. This licensing action, once approved, would then allow use of the P-T limit curves contained in Attachments 4-A through 4-D.

==5.0      REGULATORY ANALYSIS==

5.1      NO SIGNIFICANT HAZARDS CONSIDERATION Overview In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) is requesting an amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units I and 2, respectively.
The proposed amendment would revise TS 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," by adding a reference to the use of Code Cases N-640 Page 4 of 7

ATTACHMENT 1 Evaluation of Proposed Changes and N-588 as acceptable methods for determining reactor pressure vessel (RPV) pressure temperature (P-T) limits. The NRC has previously approved the use of Code Cases N-640 and N-588 for Braidwood and Byron Stations.
Criteria According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:
(1)    Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)    Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)    Involve a significant reduction in a margin of safety.
In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below regarding the proposed license amendment.
: 1.      The proposed TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The use of Code Cases N-588 and N-640 has been approved for Braidwood and Byron Stations. The use of P-T limits based on these Code Cases will continu to ensure that the RPV integrity is maintained under all conditions. Thus there is no in rease in the probability or consequences of an accident previously evaluated.
: 2.      The proposed TS change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change does not involve the use or installation of new equipment. No equipment will be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed change will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases.
Based on this evaluation, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3.      The proposed TS change does not involve a significant reduction in a margin of safety.
The P-T limits provide assurance that RPV integrity is maintained. The use of Code Cases N-588 and N-640 has been previously approved by the NRC for Braidwood and Byron Stations and will continue to ensure that RPV integrity is maintained. Thus, there is no reduction in the margin of safety.
Page 5 of 7

ATTACHMENT 1 Evaluation of Proposed Changes Based on the above discussions, it has been determined that the requested TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of a new or different kind of accident from any accident previously evaluated; or involve a significant reduction in a margin of safety. Therefore, the requested license amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).
5.2      APPLICABLE REGULATORY REQUIREMENTSICRITERIA 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," part (a) requires that the fracture toughness and material surveillance program for the reactor coolant pressure boundary must meet the requirements of 10 CFR 50, Appendix G. 10 CFR 50.60(b) allows alternatives to the requirements as specified in 10 CFR 50.12, "Specific exemptions." In the referenced letter, the NRC granted an exemption to Braidwood and Byron Stations allowing the use of Code Cases N-588 and N-640 in lieu of the requirements of 10 CFR 50 Appendix G. Thus, inclusion of the NRC approval in the Braidwood and Byron Stations TS continues to meet the requirements of 10 CFR 50.60 and 10 CFR 50.12.

==6.0      ENVIRONMENTAL CONSIDERATION==

EGC has evaluated this proposed operating license amendment consistent with the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." EGC has determined that these proposed changes meet the criteria for a categorical exclusion set forth in paragraph (c)(9) of 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," and as such, has determined that no irreversible consequences exist in accordance with paragraph (b) of 10 CFR 50.92, "Issuance of amendment." This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation," or which changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:
(i)      The amendment involves no significant hazards consideration.
As demonstrated in Section 5.1, "No Significant Hazards Consideration," the proposed change does not involve any significant hazards consideration.
(ii)     There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.
The proposed change does not result in an increase in power level, does not increase the production nor alter the flow path or method of disposal, of radioactive waste or byproducts; thus, there will be no change in the amounts of radiological effluents released offsite.
Page 6 of 7

ATTACHMENT 1 Evaluation of Proposed Changes Based on the above evaluation, the proposed change will not result in a significant change in the types or significant increase in the amounts of any effluent released offsite.
(iii)   There is no significant increase in individual or cumulative occupational radiation exposure.
The proposed change will not result in any changes to the configuration of the facility.
The proposed change will not cause a change in the level of controls or methodology used for the processing of radioactive effluents or handling of solid radioactive waste, nor will the proposed amendment result in any change in the normal radiation levels in the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.
7.0      REFERENCE Letter from U. S. NRC to 0. D. Kingsley (Exelon Generation Company, LLC), "Issuance of Exemption from the Requirements of 10 CFR 50 Part 60 and Appendix G for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2," dated August 8, 2001 Page 7 of 7

ATTACHMENT 2-A Markup of Proposed Technical Specifications Page Changes
* BRAIDWOOD STATION REVISED TS PAGE 5.6-5

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6       Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
: a.         RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and Power Operated Relief Valve (PORV) lift settings shall be established and documented in the PTLR for the following:
LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)
System";
: b.         The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in NRC.
letter,,dated January 21, 1998, "Byron Station Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report
: c.         The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7       Post Accident Monitoring Report When a report is required by Condition C or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
          .   "I+,1                        CI GA;".             *C4 S   l        Icsc   2T 4     RA.A           tree i o-              oA 2.a."               C BRAIDWOOD - UNITS 1 & 2                   5.6 - 5                      Amendmen   w

ATTACHMENT 2-B Markup of Proposed Technical Specification Page Changes BYRON STATION REVISED TS PAGE 5.6-5

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6       Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PT[R)
: a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and Power Operated Relief Valve (PORY) lift settings shall be established and documented in the PTLR for the following:
LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)
System";
: b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in NRC S            lettertdated January 21, 1998, "Byron Station Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance f Referencing of Pressure Temperature Limits Report ;an
: c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7       Post Accident Monitoring Report When a report is required by Condition C or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
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                                 .ry )W         k)V  Ys Z o> vvA 2. ckv-c   -3MtLAJ   4 ZO1A St,,-r ioo           O   His- I  cA      2 "     sced BYRON - UNITS 1 & 2                       5.6 - 5                         Amendment~21 '

ATTACHMENT 3-A Typed Pages for Technical Specification Changes BRAIDWOOD STATION REVISED TS PAGES 5.6-5 I

rements Reporting Requi
                                                       -Reporting    Requirements 5.6 5.6 Reporting Requirements 5.6.6       Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
: a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and Power Operated Relief Valve (PORV) lift settings shall be established and documented in the PTLR for the following:
LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)
System";
: b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in NRC letters dated January 21, 1998, "Byron Station Units 1 and 2,1 and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report," and August 8, 2001, "Issuance of Exemption from the requirements of 10 CFR Part 50 and Appendix G for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2"; and
: c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.                             i 5.6.7       Post Accident MonitorinQ Report When a report is required by Condition C or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inopera bility, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
BRAIDWOOD - UNITS 1 & 2              5.6 - 5                           Amendment

ATTACHMENT 3-B Typed Pages for Technical Specification Changes BYRON STATION REVISED TS PAGES 5.6-5

Requi rements Reporting Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6       Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
: a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and Power Operated Relief Valve (PORV) lift settings shall be established and documented in the PTLR for the following:
LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)
System";
: b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in NRC letters dated January 21, 1998, "Byron Station Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report," and August 8, 2001, "Issuance of Exemption from the requirements of 10CFR Part 50 and Appendix G for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2"; and
: c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7       Post Accident MonitorinQ Report When a report is required by Condition C or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
BYRON - UNITS 1 & 2                  5.6 - 5                         Amendment

ATTACHMENT 4-A WCAP-15364, Revision 2 Braidwood Unit I Heatup and Cooldown Limit Curves for Normal Operation

I                                                  EEMMW Westinghouse Non-Proprietary Class 3 WCAP-15364                                          November 2003 Revision 2 Braidwood Unit 1 Heatup and Cooldown Limit Curves for Normal Operation e Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15364, Revision 2 Braidwood Unit 1 Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubbam November 2003 Approved 2               jf'4   '~
R. A.      Cresham, Manager MReactor     Component Design & Analysis Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02003 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST OF TABLES ............................................                                iv LIST OF FIGURES ............................................                               vi PREFACE ..............................................                                    vii EXECUTIVE 

==SUMMARY==
 ............................................                           viii 1      INTRODUCTION.                                                                      1-1 2      PURPOSE .2-3      CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .3-1 4      CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .4-5      HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .5-1 6      REFERENCES .6-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS ............................................ A-1 Braidwood Unit 1 Pressure-Temperature Curves

iv LIST OF TABLES Table 4-1     Irradiation Temperatures of the Braidwood Unit I & 2 Reactor Vessels ........................ 4-2 Table 4-2     Summary of the Peak Pressure Vessel Neutron Fluence Values at 16 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV) .4-4 Table 4-3     Summary of the Peak Pressure Vessel Neutron Fluence Values at 22 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV) .4-5 Table 4-4     Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV) .4-5 Table 4-5     Calculated Integrated Neutron Exposure of the Braidwood Unit I Surveillance Capsules Tested to Date .4-6 Table 4-6     Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program .4-7 Table 4-7     Calculation of Best Estimate Cu and Ni Weight Percent for the Braidwood Unit I Forging Materials .4-8 Table 4-8     Calculation of the Average Cu and Ni Weight Percent for the Braidwood Unit I Surveillance Weld Material Only (Heat # 442011) .4-9 Table 4-9     Calculation of Best Estimate Cu and Ni Weight Percent Values for the Braidwood Units I & 2 Weld Material (Using Braidwood I & 2 Chemistry Test Results) ............. 4-10 Table 4-10    Reactor Vessel Beltline Material Unirradiated Toughness Properties ........................... 4-11 Table 4-11    Calculation of Chemistry Factors for Braidwood Unit I using Surveillance Capsule Data .                                                                                  4-12 Table 4-12    Summary of the Braidwood Unit I Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 .                              4-13 Table 4-13    Calculation of the '/4T and %T Fluence Factors Values used for the Generation of the 16 EFPY Heatup / Cooldown Curves .                                                       4-13 Table 4-14    Calculation of the 1/4AT and %4T Fluence Factors Values used for the Generation of the 22 EFPY Heatup / Cooldown Curves .                                                       4-14 Table 4-15    Calculation of the 'AT and %T Fluence Factors Values used for the Generation of the 32 EFPY Heatup / Cooldown Curves ............................................            4-14 Table 4-16    Calculation of the ART Values for the '/4T Location @ 16 EFPY ................................. 4-15 Braidwood Unit 1 Pressure-Temperature Curves

v LIST OF TABLES (Continued)
Table 4-17    Calculation of the ART Values for the %T Location @ 16 EFPY ................................. 4-16 Table 4-18    Calculation of the ART Values for the V4T Location @ 22 EFPY ................................. 4-17 Table 4-19    Calculation of the ART Values for the %T Location @22 EFPY ................................. 4-18 Table 4-20    Calculation of the ART Values for the '/4T Location @ 32 EFPY ................................. 4-19 Table 4-21    Calculation of the ART Values for the %T Location @ 32 EFPY ................................. 4-20 Table 4-22    Summary of the Limiting ART Values used in Generation of the Braidwood Unit I Reactor Vessel Heatup and Cooldown Curves at 1/4T and 3/4T Location .                    4-21 Table 5-1     Braidwood Unit I Heatup and Cooldown Curve Data Points for the 16 EFPY Curves Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) .5-5 Table 5-2     Braidwood Unit I Heatup and Cooldown Curve Data Points for the 22 EFPY Curves Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) .5-8 Table 5-3     Braidwood Unit I Heatup and Cooldown Curve Data Points for the 32 EFPY Curves Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) .5-11 Braidwood Unit 1 Pressure-Temperature Curves

vi LIST OF FIGURES Figure 5-1    Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of I 00 0 F/hr) Applicable to 16 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) .......................................... 5-3 Figure 5-2    Braidwood Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 1000 F/hr) Applicable to 16 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) ...................... 5-4 Figure 5-3    Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 100 0 F/hr) Applicable to 22 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) ........................................ 5-6 Figure 5-4    Braidwood Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100 &deg;F/hr) Applicable to 22 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) ...................... 5-7 Figure 5-5    Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of I 000 F/hr) Applicable to 32 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) ........................................ 5-9 Figure 5-6    Braidwood Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 andlOO0 F/hr) Applicable to 32 EFPY Using vs.1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors) .................... 5-10 Braidwood Unit 1 Pressure-Temperature Curves

vii PREFACE This report has been technically reviewed and verified by:
Reviewer:                                                             J. H. Ledger        AS'S Revision 1:
An error was detected in the "OPERLIM" Computer Program that Westinghouse uses to generate pressure-temperature (PT) limit curves. This error potentially affects the heatup curves when the 1996 Appendix G Methodology is used in generating the PT curves. It has been determined that WCAP-1 5364 Rev. 0 was impacted by this error. Thus, this revision provides corrected curves from WCAP-15364 Rev.
0.
Note that only the heatup curves and associated data point tables in section 5 have changed. The cooldown curves and data points remain valid and were not changed.
Revision 2:
This revision was completed to incorporate an updated reference for the flange elimination from the PT limit curves. In the previous revision, WCAP-15315 was provided as justification for the flange notch elimination and it was changed to WCAP-16143-P in this revision. In addition to this change, the thermal stress intensity factors for the highest heatup and cooldown rates were added to this report in Appendix A.
Braidwood Unit 1 Pressure-Temperature Curves

viii EXECUTIVE 

==SUMMARY==

The purpose of this report is to document the generation of pressure-temperature limit curves for Braidwood Unit I for normal operation at 16, 22 and 32 EFPY using the latest methodologies (i.e. WCAP 14040-NP-A, the 1996 ASME Boiler and Pressure Vessel Code, Section XI, Appendix 4 ASME Code Case N-588, ASME Code Case N-640 and WCAP-16143-P). Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values. The 1/4T and 3/4T values are summarized in Tables 4-16 through 4-21 and were calculated using the circumferential weld WF-562 (Heat 442011) (limiting material for circumferentially oriented flaws, Code Case N-588) and nozzle shell forging 5P-7016 (limiting material for axial flaws). The pressure-temperature limit curves were generated for a heatup rate of I 000 F/hr and cooldown rates of 0, 25, 50 and I00 0 F/hr. The axial oriented flaw cases are limiting for all EFPY values evaluated. Hence, only the axial oriented flaw curves are presented in this report and they can be found in Figures 5-1 through 5-6.
Braidwood Unit 1 Pressure-Temperature Curves

1-I I         INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 601F.
RTNDT   increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT   at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT(IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials"'.11 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves for normal operation.
NOTE: For the reactor vessel radiation surveillance program, Babcock and Wilcox Co. supplied Westinghouse with sections of SA508 Class 3 forging material used in the core region of the Braidwood Station Unit No. I reactor pressure vessel (Specifically from lower shell forging [49D867/49C813]-1-l).
Also supplied was a weldment made with non-copper coated weld wire using the automatic sub-arc welding process (Weld wire heat # 442011 Linde 80 flux, lot number 8061, which is identical to that used in the actual fabrication of the intermediate to lower shell girth weld of the pressure vessel).
Braidwood Unit 1 Pressure-Temperature Curves

2-1 2        PURPOSE The Commonwealth Edison Company (now known as Exelon Nuclear) contracted Westinghouse to analyze surveillance capsule W from the Braidwood Unit 1 reactor vessel. As a part of this analysis and the current uprating program, Westinghouse is to generated new heatup and cooldown curves for 16, 22 and 32 EFPY. These new Pressure-Temperature Curves are to be developed utilizing the following methodologies:
* ASME Code Case N-640 121,
* Elimination of the flange requirement of Appendix G to 10 CFR Part 50 13J (WCAP-16143-P[4 ], "Reactor Vessel Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2,"
* ASME Code Case N-58815] (where applicable), and
* Methodology of the 1996 version ofASME Section XI,Appendix G161.
* The P-T Curves will be developed WITHOUT margins or instrumentation errors.
Circumferential Flaw methodology (ASME Code Case N-588) in combination with 1996 Appendix G to ASME Section XI and ASME Code Case N-640 (K1 , methodology) will be used to develop the P-T Curves for the limiting circumferential weld material. Axial Flaw methodology, 1996 Appendix G to ASME Section XI and ASME Code Case N-640 (K1 , methodology) will be used to develop P-T Curves for the limiting forging/base metal material. In addition, the methodology to eliminate the 10 CFR Part 50Appendix G flange requirements (WCAP-16143-P) will be employed. The final P-TCurves will be developed by generating a P-T Curve based on the most limiting circumferential weld ART value P-T Curve and/or the P-T Curve based on the most limiting axial flaw material ART value. Hence the final P-T Curves will be a composite curve based of the limiting data from the two curves (ie. Circ flaw or axial flaw case).
The purpose of this report is to present the calculations and the development of the Braidwood Unit I heatup and cooldown curves for 16, 22 and 32 EFPY. This report documents the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2['1, for all the beltline materials and the development of the heatup and cooldown pressure-temperature limit curves for normal operation.
Per the request of Exelon, the surveillance weld data from the Braidwood Unit I and Braidwood Unit 2 surveillance programs has been integrated. The Braidwood Unit I surveillance weld metal was fabricated from the same weld wire heat as the Braidwood Unit 2 surveillance weld metal (Heat No.
442011). This weld metal is in the upper to lower shell gird weld seam WF-562 of both Units. Per WCAP-15366 71 , all the surveillance data has been determined to be credible.
Braidwood Unit 1 Pressure-Temperature Curves

3-1 3         CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 Overall Approach Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements"13 ] specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements. Section XI, Division 1, "Rule for Inservice Inspection of Nuclear Power Plant Components", Appendix G61, contains the conservative methods of analysis.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1c, for the metal temperature at that time. K1 , is obtained from the reference fracture toughness curve, defined in Code Case N-640 of Appendix G of the ASME Code, Section XI. The K1 , curve is given by the following equation:
Kic = 33.2 + 20.734
* e(0)02(T-Rr.Dr)J
: where, KI, = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K1, curve is based on the lower bound of static K, values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steels.
Braidwood Unit I Pressure-Temperature Curves

3-2 3.2 Methodology for Pressure-Temperature Limit Curve Developmeiit The governing equation for the heatup-cooldown analysis is defined in Code Case N-640 of Appendix G of the ASME Code as follows:
C*K Im+ Kit <Kic                                                      (2) where, Klm=   stress intensity factor caused by membrane (pressure) stress KII = stress intensity factor caused by the thermal gradients K,= = function of temperature relative to the RTNDT of the material C      = 2.0 for Level A and Level B service limits C      = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical For membrane tension, the K, corresponding to membrane tension for the postulated defect is:
Kim = Mm * (pR;
* t)                                         (3)
Where Mm for an inside surface is given by:
Mm = 1.85 for it < 2, Mm = 0.926 At for 2 *<tA     3.464, and Mm = 3.21 for t > 3.464.
Similarly, Mm for an outside surface flaw is given by:
Mm = 1.77 for t < 2, Mm = 0.893 4tt for 2
* lJt
* 3.464, and Mm = 3.09 for 4t > 3.464.
Where:
Ri = vessel inner radius, t   vessel wall thickness, and p = internal pressure, Braidwood Unit I Pressure-Temperature Curves

3-3 For Bending Stress, the K, corresponding to bending stress for the postulated defect is:
KMb = Mb
* maximum bending stress, where Mb is two-thirds of Mm For the Radial Thermal Gradient, the maximum K1 produced by radial thermal gradient for the postulated inside surface defect is:
K = 0.953x10 3 x CR xt25                                         (4) where:
CR= the cooldown rate in OF/hr.
For the Radial Thermal Gradient, the maximum K1 produced by radial thermal gradient for the postulated outside surface defect is:
K,= 0.753x10 3 x HU xt5                                         (5) where:
HU = the heatup rate in IF/hr.
The through-wall temperature difference associated with the maximum thermal K1 can be determined from ASME Section XI, Appendix G.Figure G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Section XI, Appendix G Figure G-2214-2 for the maximum thermal K1 .
(a) The maximum thermal K1 relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2) of Appendix G to ASME Section XI.
Braidwood Unit I Pressure-Temperature Curves

34 (b) Alternatively, the K1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a l4-thickness inside surface defect using the relationship:
Ku = (1.0359Co + 0.6322Ci + 0.4753C2 + 0.3855C3) */r                                        (6) or similarly, Krr during heatup for a l4-thickness outside surface defect using the relationship:
K,, = (1.043Co + 0.630Ci + 0.481C2 + 0.4O1C3)
* 4;                                         (7) where the coefficients C0 , C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
ci(x) = Co+ Ci(x/a)+ C2(x/a)2 + C3(x/a)3                                                (8) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.
Note, that equations 3 through 8 were added to the OPERLIM computer program, which is the Westinghouse computer program used to generate pressure-temperature limit curves. No other changes were made to the OPERLIM computer program with regard to the pressure-temperature curve calculation methodology. Hence, the pressure-temperature curve methodology described in WCAP-14040s1 Section 2.6 (equations 2.6.2-4 and 2.6.3-1) remains valid for the generation of the pressure-temperature curves documented in this report with the exceptions described above.
At any time during the heatup or cooldown transient, Kic is determined by the metal temperature at the tip of a postulated flaw at the '/4T and %T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kl,, for the Braidwood Unit I Pressure-Temperature Curves

3-5 reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.
Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the '/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of Klc at the A4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KIc exceeds K1 ,, the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the A/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a '/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIc for the Y4T crack during heatup is lower than the Kic for the 'AT crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIc values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the A/4T flaw is considered.
Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
Braidwood Unit I Pressure-Temperature Curves

3-6 The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a '4T flaw located at the '4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
Code Case N-588 for Circumferential Weld Flaws:
In 1997, ASME Section XI, Appendix G was revised to add methodology for the use of circumferential flaws when considering circumferential welds in developing pressure-temperature limit curves. This change was also implemented in a separate Code Case, N-588.
The earlier ASME Section XI, Appendix G approach mandated the postulation of an axial flaw in circumferential welds for the purposes of calculating pressure-temperature limits. Postulating the Appendix G reference flaw in a circumferential weld is physically unrealistic because the length of the reference flaw is 1.5 times the vessel thickness and is much longer than the width of the vessel girth welds. In addition, historical experience, with repair weld indications found during pre-service inspection and data taken from destructive examination of actual vessel welds, confirms that any flaws are small, laminar in nature and are not oriented transverse to the weld bead orientation. Because of this, any defects potentially introduced during fabrication process (and not detected during subsequent non-destructive examinations) should only be oriented along the direction of the weld fabrication. Thus, for circumferential welds, any postulated defect should be in the circumferential orientation.
The revision to Appendix G now eliminates additional conservatism in the assumed flaw orientation for circumferential welds. The following revisions were made to ASME Section XI, Appendix G:
G-2214.1 Membrane Tension...
The K, corresponding to membrane tension for the postulated circumferential defect of G-2120 is Ki m = Mmx(pRi/t)
Braidwood Unit I Pressure-Temperature Curves

3-7 where, Mm for an inside surface flaw is given by:
Mm     =      0.89 for 4   < 2, Mm     =      0.4434t for 24t        < 3.464, Mm     =      1.53 for 4t >3.464 Similarly, Mm for an outside surface flaw is given by:
Mm     =      0.89 for 4   < 2, Mm     =      0.4434t for 24t        < 3.464, Mm     =      1.53 for 4   >3.464 Note, that the only change relative to the OPERLIM computer code was the addition of the constants for Mm in a circ. weld limited condition. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. As stated previously, the P-T curve methodology is unchanged from that described in WCAP-14040!S] Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.
Braidwood Unit I Pressure-Temperature Curves

3-8 3.3 Closure HeadNessel Flange Requirements 10 CFR Part 50, Appendix G contains the requirements for the metal temperature of the closure head flange and vessel flange regions; This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure (3106 psig), which is 621 psig for the Braidwood Unit I reactor vessel.
This requirement was originally based on concerns about the fracture margin in the closure flange region.
During the boltup process, stresses in this region typically reach over 70 percent of the steady-state stress, without being at steady-state temperature. The margin of 1207F and the pressure limitation of 20 percent of hydrotest pressure were developed using the Ki, fracture toughness, in the mid 1970s.
Improved knowledge of fracture toughness and other issues which affect the integrity of the reactor vessel have led to the recent change to allow the use of Ki, in the development of pressure-temperature curves, as contained in Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division I".
The discussion given in WCAP-I 6143-P, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2," concluded that the integrity of the closure head/vessel flange region is not a concern for Byron or Braidwood, therefore the requirement of 10 CFR Part 50, Appendix G.can be eliminated from the Pressure-Temperature Curves contained in this report.
Braidwood Unit I Pressure-Temperature Curves

4-1 4        CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART = InitialRTXDT + A RTNDT + Margin                                 (9)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-233 1 of Section III of the ASME Boiler and Pressure Vessel Codel 91. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTNDT    is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows:
A RTD      =CF *fo     28
                                                            .101ogi                         (10)
To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
f(depth)   f .. f.ce
* e       )0(114) where x inches (vessel inner radius and beltline thickness is 86.625 inches and 8.5 inches, respectively)1'01 is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 10 to calculate the ARTNDT at the specific depth.
The Westinghouse Radiation Engineering and Analysis group evaluated the vessel fluence projections and the results are presented in Section 6 of WCAP-15316"111. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with the methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"181. Table 4-2 through 4-4, herein, contains the calculated vessel surface fluence values along with the Regulatory Guide 1.99, Revision 2, 1/4T and 3/4T calculated fluence values used to calculate the ART values for all beltline materials in the Braidwood Unit I reactor vessel for 16, 22 and 32 EFPY. Additionally, the calculated surveillance capsule fluence values are presented in Table 4-5.
Ratio Procedure and Temperature Adiustment:
The ratio procedure, as documented in Regulatory Guide 1.99, Revision 2 Position 2.1, was used, where applicable, to adjust the measured values of ARTNDT of the weld materials for differences in copper/nickel content. This adjustment is performed by multiplying the ARTNDT by the ratio of the vessel chemistry factor to the surveillance material chemistry factor. The adjusted ARTNDT values are then used to calculate the chemistry factor for the vessel materials.
Braidwood Unit I Pressure-Temperature Curves

4-2 From NRC Industry Meetings on November 12, 1997 and February 12th & 13wh of 1998, procedural guidelines were presented to adjust the ARTNDT for temperature differences when using surveillance data from one vessel applied to another vessel. The following guidance was presented at these industry meetings:
Irradiation temperature and fluence (or fluence factor) are first order environmental variables in assessing irradiation damage. To account for differences in temperature between surveillance specimens and vessel, an adjustment to the data must be performed. Studies have shown that for temperatures near 5507F, a 1IF decrease in irradiation temperature will result in approximately a IIF increase in ARTNDT-For capsules with irradiation temperature of Tcapsuic and a plant with an irradiation temperature of Tplant, an adjustment to normalize ARTNDT.mCasedto Tplan is made as follows:
Temp. Adjusted ARTNDT      = ARTNDT. measured + I.0*( Tcapsuie - Tpiant)           (12)
The irradiation temperatures of the Braidwood Unit I & 2 reactor vessels are as follows:
TABLE 4-1 Irradiation Temperatures of the Braidwood Unit I & 2 Reactor Vessels Fuel Cycle          Braidwood 1         Capsule           Braidwood 2           Capsule I                557 0 F               U               557 0 F             U 2                 551 0F                                5510F 3                 551 0F                                5510F 4                 551 0 F              X                5510 F              X 5              551-554 0 F           --               5510F
* 6                554 0F                               551 0F
* 7               554 0F
* W              5500 F**              W Average               5530 F                                5520 F Temp.         I     _I                              I                     I
* These values are from ComEd and are documented in BRW-DIT-97-3211'5 .
        ** This value is from ComEd and is documented in BRW-DIT-2000-00101 6 ].
Capsules U and X were exposed to the same average operating temperature of 5530F. Capsule W was exposed to the same average operating temperature of 553&deg;F for 5 of 7 cycles. For cycles 6 and 7, the Braidwood Unit I capsule W saw only a I &deg;F increase in average operating temperature. Thus, the Braidwood Unit I Pressure-Temperature Curves

4-3 average inlet operating temperature of both Braidwood Unit I and Unit 2 are essentially the same for the time of interest. Hence, no temperature adjustments are required.
Chemistry Factor:
The chemistry factor is obtained from the tables in Regulatory Guide 1.99, Revision 2 using the best estimate average copper and nickel content as reported in Tables 4-6 through 4-9. The chemistry factors were also calculated using Position 2.1 from the Regulatory Guide 1.99, Revision 2 using all available surveillance data. Per Reference 7, all available surveillance data for Braidwood Unit I is credible.
This assessment also calculated the vessel weld (including temperature and chemistry adjustment) chemistry factor using Braidwood Unit I stand alone data and it was determined to be 33.51F. Position 2.1 chemistry factors are calculated in Table 4-15.
Explanation of Margin Term:
When there are "two or more credible surveillance data sets"111 available for Braidwood Unit 1, Regulatory Guide 1.99 Rev. 2 (RGI.99R2) Position 2.1 states "To calculate the Margin in this case, use Equation 4; the values given there for qs may be cut in half". Equation 4 from RGI.99R2 is as follows:
M = 21,T 2.
Standard Deviation for Initial RTNDT Margin Term, a, If the initial RTNDT values are measured values, which they are in the case of Braidwood Unit 1, then o, is equal to 00F. On the other hand, if the initial RTNDT values were not measured, then a generic value of 17'F (base metal and weld metal) would have been required to be used for CY,.
Standard Deviation for ARTNDT Margin Term, 0c Per RG I.99R2 Position 1.1, the values of i,&  are referred to as "281F for welds and 170 F for base metal, except that as need not exceed 0.50 times the mean value of ARTNDT-" The mean value of ARTNDT is defined in RGI.99R2 by Equation 2 and defined herein by Equation 8.
Per RGI.99R2 Position 2.1, when there is credible surveillance data, 0c is taken to be the lesser of 12 ARTNDT or 14'F (280 F/2) for welds, or 8.5 0 F (17/F12) for base metal. Where ARTNDT again is defined herein by Equation 10.
Summary of the Margin Term Braidwood Unit I Pressure-Temperature Curves

4-4 Since a, is taken to be zero when a heat-specific measured value of initial RTNDT are available (as they are in this case), the total margin term, based on Equation 4 of RGI.99R2, will be as follows:
* Position 1.1:        Lesser of ARTNDT or 560 F for Welds Lesser of ARTNDT or 340 F for Base Metal
* Position 2.1:        Lesser of ARTNDT or 280 F for Welds Lesser of ARTNDT or 17'F for Base Metal TABLE 4-2 Summary of the Peak Pressure Vessel Neutron Fluence Values at 16 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV)
Material                          Surface                1/4T                 3/4T (n/cm2 ,E > 1.0 MeV) (n/cm2 ,E> 1.0 MeV)  (n/cm 2 ,E> 1.0 MeV)
Intermediate Shell Forging               1.03 x 10"9          6.19 x 10"           2.23 x 10"
[49D383/49C344]--1 I Lower Shell Forging                   1.03 x 10   9        6.19 x 10"1          2.23 x 10"
[49D687/49C813]-1-I Nozzle Shell Forging SP-7016               3.07 x 1O"           1.84 x 10"           6.65 x 10'7 Intermediate to Lower Shell Forging Circ.        1.00 x 10i9          6.00 x 10"           2.17 x 10" Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Inter. Shell Forging Circ.       3.07 x 10"           1.84 x 10"           6.65 x 1017
      'Weld Seam WF-645 (Heat H4498)
Note: All remaining vessel materials are below I x 101 n/cm 2, E > 1.0 MeV Braidwood Unit I Pressure-Temperature Curves

4-5 TABLE 4-3 Summary of the Peak Pressure Vessel Neutron Fluence Values at 22 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV)
Material                              Surface                 1/4T                 314T (n/cm 2 ,E  >1.0 MeV) (n/cm2  ,E> 1.0 MeV) (n/cm 2 ,E >  1.0 MeV)
Intermediate Shell Forging                    1.41 x 1o'9           8.47 x i0"'          3.05 x 10"
[49D383/49C344]-i -I Lower Shell Forging                       1.41 x 1019           8.47 x 10"           3.05 x 10"
[49D687/49C813]-1-I Nozzle Shell Forging 5P-7016                   4.20 x 10"            2.52 x 10"           9.09 x 1017 Intermediate to Lower Shell Forging Circ.            1.37 x 1lO'9          8.23 x 10"           2.97 x IO" Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Inter. Shell Forging Circ.           4.20 x 1O"            2.52 x 10"           9.09 x 1017 Weld Searn WF-645 (Heat H4498)
Note: All remaining vessel materials are below I x 107 n/cm 2 , E > 1.0 MeV TABLE 4-4 Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV)
Material                              Surface                 1/4T                 3/4T (n/cm 2 ,E > 1.0 MeV) (n/cm 2 ,E> 1.0 MeV) (n/cm 2 ,E> 1.0 MeV)
Intermediate Shell Forging                    2.05 x 1019           1.23 x 10'9          4.44 x 10" 149D383/49C344]-1-I               ._.-
Lower Shell Forging                       2.05 x I0'9           1.23 x 10'9          4.44 x 10"
[49D687/49C813]-I-I Nozzle Shell Forging 5P-7016                   6.08 x 10i"           3.65 x 10"           1.32 x IO" Intermediate to Lower Shell Forging Circ.            1.99 x 1o'9           1.19 x io'9          4.31 x 10"s Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Inter. Shell Forging Circ.          6.08 x 10"            3.65 x 10"           1.32 x 10" Weld Seam WF-645 (Heat H4498)
Note: All remaining vessel materials are below I x 1O"7 n/cm2 , E > 1.0 MeV Braidwood Unit I Pressure-Temperature Curves

4-6 TABLE 4-5 Calculated Integrated Neutron Exposure of the Braidwood Unit I Surveillance Capsules Tested to Date Capsule                             Fluence U                 3.87 x 1018 n/cm2 , (E > 1.0 MeV)
X                 1.24 x I0 9 n/cm 2 , (E > 1.0 meV)
W                 2.09 x 10 1 9 ncm2, (E > 1.0 MeV)
Contained in Table 4-6 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline materials. These measured shift values were obtained using CVGRAPH, Version 4.l1I], which is a symmetric hyperbolic tangent curve-fitting program.
Braidwood Unit I Pressure-Temperature Curves

4-7 TABLE 4-6 Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program Material                        Capsule           Measured 30 ft-lb Transition Temperature Shift(')
Intermediate Shell Forging                  U                         5.780 F
[49D867/49C813]-l-1                      X                        38.23 0 F (Tangential Orientation)                   W                        24.14 0 F Intermediate Shell Forging                  U                          0.OOF(b)
[49D867/49C813]--1                       X                        28.750 F (Axial Orientation)                     W                        37.11 OF Surveillance Program                     U                         17.060 F Weld Metal                          X                        30.15 0 F (Heat # 442011)                                                49.680 F Heat Affected Zone                      U                         56.31 OF X                         92.8 0 F W                         84.860 F Notes:
(a) From capsule W analysis results Calculated using the measured Charpy data plotted using CVGRAPH, Verision 4.1111'.
(b) Actual value for ARTNDT is -16.07. This physically should not occur, therefore for conservatism (i.e.
higher Chemistry Factor) a value of zero will be reported.
Braidwood Unit I Pressure-Temperature Curves

4-8 Table 4-7 contains the calculation of the best estimate weight percent copper and nickel for the Braidwood Unit 1 base materials in the beltline region. Table 4-8 contains the calculation of the best estimate weight percent copper and nickel for the Braidwood Unit I surveillance weld material, while Table 4-9 presents the overall best estimate average for that heat of weld. Table 4-10 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RTNDT of the beltline materials and vessel flanges. The weight percent values of Cu and Ni given in Table 4-10 were used to generate the calculated chemistry factor (CF) values based on Tables 1 and 2 of Regulatory Guide 1.99, Revision 2, and presented in Table 4-12. Table 4-11 provides the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Position 2.1, which are also summarized in Table 4-12.
TABLE 4-7 Calculation of the Best Estimate Cu and Ni Weight Percent for the Braidwood Unit I Forging Materials Intermediate Shell Forging             Lower Shell Forging
[49D383/49C3441-1-1 MK 24-2         149D867/49C8131-1-1 MK 24-3 Reference                  Cu %              Ni %              Cu %                Ni %
13                       0.05             0.73               ---                ---
11                       ---              ---                0.03               0.73 11 (Charpy EL-6)                 ---              ---               0.052              0.746 11 (Charpy ET-57)                  ---              ---               0.046              0.736 Charpy ET-31(t)                 ---              ---               0.045              0.687 Charpy EL-38(a)                 ---              ---               0.049              0.725 Charpy ET40(')                  ---                                0.047              0.721 Charpy EL-42(a)                 ---              ---               0.053              0.809 Best Estimate Average(b)               0.05             0.73               0.05               0.74 Notes:
(a)      Charpy Specimen From Capsule W of Braidwood Unit I (Ref 11).
(b) The best estimate average was rounded per ASTM E29, using the "Rounding Method".
Braidwood Unit I Pressure-Temperature Curves

4-9 TABLE 4-8 Calculation of the Average Cu and Ni Weight Percent for the Braidwood Unit I Surveillance Weld Material Only (Heat # 442011)
Reference                      Weight % Copper                 Weight % Nickel WCAP-14824, Rev.2(a)                        0.032                         0.671 Charpy EW-32(b)                           0.032                         0.674 Charpy EW-33(b)                           0.032                         0.660 Charpy EW-34()                            0.030                         0.661 Charpy EW-42(b)                           0.033                         0.690 Surveillance Weld Average                       0.03                          0.67 Notes:
(a)   This is the average of 29 data points. Therefore, to average with the new Charpy data, multiply the average copper and nickel values from Ref. 17 by 29, add four new points and divide by 33.
(b)    Charpy specimen from Capsule W of Braidwood Unit I (Ref. 11).
Braidwood Unit I Pressure-Temperature Curves

4-10 TABLE 4-9 Calculation of Best Estimate Cu and Ni Weight Percent Values for the Braidwood Units I & 2 Weld Material (Using Braidwood I & 2 Chemistry Test Results)
Chemistry hype                  Reference              Weight %            Weight %
Copper             Nickel B&W Weld Qualification             WCAP-14824,                  0.028               0.63 BANV-2261                       Rev. 2 B&W Weld Qualification             WCAP-14824,                   0.03               0.65 BAW-2261                       Rev. 2 B&W Weld Qualification              WCAP-14824,                  0.04               0.67 BAW-2261                       Rev. 2 Braidwood Unit I Surv.               Table 4-8                 0.032              0.671 Data Ave.(')
Braidwood Unit 2 Surv.             WCAP-15369                  0.034              0.710 Data Ave.(b)
BEST ESTIMATE                                                0.03               0.67 AVERAGE(C Notes:
(a)    The weld material in the Braidwood Unit I surveillance program was made of the same wire and flux as the reactor vessel inter. to lower shell girth seam weld (Weld seam WF-562, Wire Heat No. 442011, Flux Type Linde 80, Flux Lot No. 8061). The weld wire is type Linde MnMoNi (Low Cu-P).
(b)    The Braidwood Unit 2 surveillance weld is representative of that used in the Braidwood Unit I reactor vessel core region girth seam (WF-562) heat number 442011, with a Linde 80 type flux, lot number 8061 (i.e. The Braidwood Unit 2 surveillance weld was fabricated with Linde 80 type flux, Lot number 0344).
(c)    The best estimate chemistry values were obtained using the "average of averages" approach. In addition the best estimate average was rounded per ASTM E29, using the "Rounding Method".
Braidwood Unit I Pressure-Temperature Curves

4-11 TABLE 4-10 Reactor Vessel Beitline Material Unirradiated Toughness Properties Material Description                      Cu (%)        Ni (%)      Initial RTNDT Closure Head Flange 2030-V-1                    0.11          0.67             -20 Vessel Flange 122N357VAI                       -            0.77             -10 Nozzle Shell Forging 5P7016                    0.04          0.73               10 Intermediate Shell Forging                    0.05          0.73             -30
[49D383/49C344]-l -I
[Table 4-7]
Lower Shell Forging                       0.05          0.74             -20
[49D867/49C813]-i -I
[Table 4-7]
Intermediate to Lower Shell Forging                0.03          0.67              40 Circ. Weld Seam WF-562 (Heat 442011)
[Table 4-9]
Nozzle Shell to Intermediate Shell Forging            0.04          0.46              -25 Circ. Weld Seam WF-645 (Heat H4498)
Surveillance Weld Braidwood Unit 1                 0.03          0.67             ---
(Heat 442011)
[Table 4-91 Surveillance Weld Braidwood Unit 2                 0.03          0.71 (Heat 442011) rTable 4-9]
Notes:
(a)   The Initial RTNDT  values for the forgings and welds are based on measured data.
Braidwood Unit I Pressure-Temperature Curves

4-12 TABLE 4-11 Calculation of Chemistry Factors for Braidwood Unit 1 using Surveillance Capsule Data Material             Capsule       Capsule 0(1        FF(b)             ARTNDT(c)          FF*ARTNDT              FF2
_       (_F)                 (')
Lower Shell Forging           U             0.387           0.737                5.78                 4.26            0.543
[49D867/49C813]-1-I            X              1.24           1.060                38.23                40.52            1.124 (Tangential)              W              2.09           1.201                24.14                28.99            1.442 Lower Shell Forging           U             0.387           0.737                0.04)                 0.0            0.543
[49D867/49C813]-1-1            X              1.24           1.060                28.75                30.48            1.124 (Axial)               W              2.09           1.201                37.11               44.57             1.442 SUM:        148.82           6.218 l                  CF = V(FF
* RTNTT) +      X( FF2) = (148.820 F) + (6.218) = 23.9'F Braidwood Unit I            U             0.387           0.737               17.00)                12.57           0.543 Surveillance              X              1.24           1.060                  .
30 15   (d)            31.96            1.124 Weld Metal               W              2.09           1.201              49 .68 (d)             59.67            1.442 Braidwood Unit 2            U              0.40           0.746               0.0(d0                 0.0            0.557 Surveillance              X              1.23           1.058               26 .3 (d)            27.83             1.119 Weld Metal               W              2.25           1.220               23.9(d)              29.16             1.488 SUM:        161.19           6.273 CF = X(FF *RTNDT)      + X( FF )= (161.191F) + (6.273) = 25.71F 2
Notes:
1 1 (a)     f = Calculated fluence from the Braidwood Unit I capsule W dosimetry analysis results             Il and the Calulated t 41                 2 fluence from the Braidwood Unit 2 capsule W analysisl , (x l0'9 n/cm , E> 1.0 MeV).
(b)     FF = fluence factor = e(O.2-.Ologf)
(c)     ARTNDT values are the measured 30 fl-lb shift values for Braidwood Unit I taken from Ref. 11 and for Braidwood Unit 2 taken from Ref. 14.
(d)     The surveillance weld metal ARTNDT values have not been adjusted. The chemistry factors of the surveillance welds and the vessel weld is 41 IF, hence, the ratio factor is 1.0. In addition, the average inlet operating temperature of both Braidwood Unit I and Unit 2 are essentially the same for the time of interest. Hence, no temperature adjustments are required.
(e)     Actual value of ARTNDT is -16.07. This physically should not occur, therefore for conservatism (i.e. higher chemistry factor) a value of zero will be used.
(f)     Measured value is -0.58. 0 is assumed for conservatism.
Contained in Table 4-12 is a summary of the calculated chemistry factors of the Braidwood Unit I beltline materials based on the Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 methodology.
Braidwood Unit I Pressure-Temperature Curves

4-13 TABLE 4-12 Summary of the Braidwood Unit I Beltline Material Chemistry Factor Values Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Material                                          Chemist1 y Factor Position 1.1                Position 2.1 Intermediate Shell Forging                         31.0WF                         ---
[49D383/49C3441-1 -1 Lower Shell Forging                            31.0WF                       23.9WF
[49D867/49C813]-1 -1 Nozzle Shell Forging 5P7016                         26.0F                          ---
Intermediate Shell to Lower Shell Forging Circ.              41.0WF                       25.70F Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell Forging Circ.              54.0F                          ---
Weld Seam WF-645 (Heat H4498)
Braidwood Unit 1 Surveillance Program Weld                  41.0F                          ---
Metal Braidwood Unit I & 2 Surveillance Program                  41.0WF                         --  -
Weld Metal Contained in Tables 4-13 through 4-15 is a summary of the fluence factors (FF) used in the calculation of adjusted reference temperatures for the Braidwood Unit 1 reactor vessel beltline materials for 16, 22 and 32 EFPY.
TABLE 4-13 Calculation of the 1/4/4T and 3/4T Fluence Factor Values used for the Generation of the 16 EPFY Heatup/Cooldown Curves Material                            %T F             AT FF                'AT f             'AT FF (n/cm 2 ,E > 1.0 MeV)                 (n/cm 2 ,E > 1.0 MeV)
Intermediate Shell Forging               6.19 x 10"           0.866            2.23 x 1018            0.596
[49D383/49C344]-1     -I Lower Shell Forging                   6.19 x 108           0.866            2.23 x 1018            0.596
[49D687/49C8131-1-1I Nozzle Shell Forging 5P-7016               1.84 x 10"8          0.550            6.65 x 1017            0.340 Intermediate to Lower Shell               6.00 x 10"           0.857            2.17 x 1018            0.589 Forging Circ. Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Inter. Shell             1.84 x 1018          0.550            6.65 x 10"             0.340 Forging Circ. Weld Seam WF-645 (Heat H4498)
Braidwood Unit I Pressure-Temperature Curves

4-14 TABLE 4-14 Calculation of the 'hT and %T Fluence Factor Values used for the Generation of the 22 EFPY Heatup/Cooldown Curves Material                          'hT F            %TFF                  3/T f          3%TFF (n/cm 2 ,E > 1.0 MeV)                  (n/cm2  ,E > 1.0 MeV)
Intermediate Shell Forging              8.47 x 10'"          0.953            3.05 x I             0.675
[49D383/49C344]-1 -1 Lower Shell Forging                 8.47 x 10S           0.953            3.05 x 10"           0.675
[49D687/49C8131-1i-I Nozzle Shell Forging 5P-7016              2.52 x IO1           0.626            9.09 x 1017          0.398 Intermediate to Lower Shell              8.23 x 1018          0.945            2.97 x 10S           0.668 Forging Circ. Weld Seam WF-562 (Heat 442011)                                                                 .
Nozzle Shell to Inter. Shell            2.52 x 1018          0.626            9.09 x 1017          0.398 Forging Circ. Weld Seam WF-645 (Heat H4498)
TABLE 4-15 Calculation of the 'AT and 3%T Fluence Factor Values used for the Generation of the 32 EFPY Heatup/Cooldown Curves Material                          %T F             %T FF                 3/T f           'hT FF (n/cm2  ,E > 1.0 MeV)                  (n/cm2  ,E > 1.0 MeV)
Intermediate Shell Forging              1.23 x 10' 9         1.058            4.44 x lOs           0.774
[49D383/49C3441-1    -1 Lower Shell Forging                 1.23 x 1019          1.058            4.44 x 10"           0.774
[49D687/49C8131-I-I_
Nozzle Shell Forging 5P-7016              3.65 x 10"           0.722            1.32 x 1018          0.475 Intermediate to Lower Shell              1.19 x 1019          1.049            4.31 x 1018          0.766 Forging Circ. Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Inter. Shell            3.65 x 10's          0.722            1.32 x 1018          0.475 Forging Circ. Weld Seam WF-645 (Heat H4498)
Contained in Tables 4-16 and 4-21 are the calculations of the ART values used for the generation of the 16, 22 and 32 EFPY heatup and cooldown curves.
Braidwood Unit I Pressure-Temperature Curves

4-15 TABLE 4-16 Calculation of the ART Values for the '/4T Location @ 16 EFPY Reactor Vessel Bitline              Material                                                   16b-                  --
Region Locatlon               Identiflcatlon       Cu%        Ni%           CF         EFPY         'A-t r    'A-t FF       l(.)   ARTNDT`d)      al     a     o         Al       ARTic)
(x10)       (x 10")
Intermediate Shell Forging        49D383/49C344-1-1          0.05      0.73         31.0         1.03       0.619       0.866        -30       26.8         0         13.4       26.8       24 Lower shell Forging             49D867/49C813-1-1          0.05      0.74         31.0         1.03       0.619       0.866        -20       26.8         0         13.4       26.8        34 Lower Shell Forging                                                                23.9         1.03       0.619       0.866        -20       20,7         0         10.4       20.7       21
-+ using S/C Data Inter. to Lower Shell Circ.              WF-562             0.03      0.67         41.0         1.00       0.600       0.857        40        35.1         0         17.0       34.0       109 Weld Metal(')
Inter. to Lower Shell Circ.                                                        25.7         1.00       0.600       0.857        40        22.0         0         11.0       22.0        84 Weld Metalt 0
 -+ using S/C Data Nozzle Shell Forging                     5P-7016            0.04      0.73         26.0        0.307       0.184       0.550        10        14.3         0          7.2       14.3       39 Nozzle Shell to Inter. Shell             WF-645                  0.04 0.46         54.0        0.307       0.1AQ       0.550       -25        29.7                   14.9       29.7       34 Cire. Weld Metal                                                                          .
NOTES:
(a) Initial RTNDT values are measured values.
(b) Fluence, f, is based upon f,,,(1019 n/cm 2 , E>1.0 MeV).
(c) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(d) ARTNT = CF
* FF (e) The CF for the Braidwood Unit I Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I Weld (WF-562, heat # 44201 1) and the Braidwood 2 Weld (WF-562, Heat
           # 44201 1).
(f) The Braidwood Iand 2 surveillance programs contain the same heat of weld metal (heat # 44201 1). Ifence, the surveillance program results have been integrated and they are credible.
Braidwood Unit I Pressure-Temperaturc Curves I

4-16 TABLE 4-17 Calculation of the ART Values for the 'AT Location           @ 16 EFPY Reactor Vessel Beltilne             Material                                             f1@ 16(b)                                                  .
Region Location               Identification       Cu%       Ni%           CF          EFPY        3t4-t f     %-t FF                  ARTNDT t    at C)                      1 Ma       ART")
(lo101)    (X 10"1)
Intermediate Shell Forging         49D383/49C344-1-1        0.05      0.73         31.0          1.03        0.223       0.596       -30          18.5        0          9.3      18.5        7 Lower shell Forging              49D867/49C813-1-1        0.05      0.74         31.0          1.03        0.223       0.596       -20          18.5        0          9.3      18.5       17 Lower Shell Forging                                                                23.9          1.03        0.223       0.596       -20          14.2        0          7.1      14.2        8
-+ using S/C Data Inter. to Lower Shell Circ.               WF-562            0.03      0.67         41.0          1.00        0.217       0.589        40          24.1        0          12.1     24 1       88 Weld Metal(')_                                              _ _ _ _                                       _  __ _ _  _             _ _  _ _ ___             _  _ _ _ _ _  _ _ _
Inter. to Lower Shell Circ.                                                       25.7          1.00        0.217       0.589        40          15.1        0          7.6      15.1       70 Weld Metal")
-. using SIC Data                                                                                                                                                          --
Nozzle Shell Forging                     5P-7016           0.04      0.73         26.0         0.307        0.067       0.340         10          8.8        0          4.4       8.8       28 Nozzle Shell to Inter. Shell              WF-645            0.04      0.46          54.0        0.307        0.067       0.340       -25          18.4        0          9.2      18.4        12 Circ. Weld Metal NOTES:
(a) Initial RTNDT values are measured values.
(b) Fluence, f, is based upon fC,, (109 n/cm2 , E> 1.0 MeV).
(c) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(d) ARTNDT = CF
* FF (c) The CF for the Braidwood Unit I Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I Weld (WF-562, heat # 442011) and the Braidwood 2 Weld (WF-562, 1-eat
          # 442011).
(I) The Braidwood I and 2 surveillance programs contain the same heat of weld metal (heat # 442011). I lence, the surveillance program results have been integrated and they are credible.
Braidwood Unit I Pressure-Temperature Curves

4-17 TABLE 4-18 Calculation of ART Values for the '/4T Location ( 22 EFPY Reatdor Vesstl Beitline             Material                                             f @ 22'bI Region Location              Identification       Cu%        Ni%           CF         EFPY         A-t r      %A-tFF        I(q)   ARTNDTT"d      (1          aA         M1       ART'c)
(xlO"9)     (xlO")
Intermediate Shell Forging          49D383/49C344-1-I        0.05      0.73         31.0         1.41       0.847        0.953       -30        29.5        0          14.S      29.5        29 Lower shell Forging               49D867/49C813-1-1        0.05      0.74         31.0         1.41       0.847        0.953       -20        29.5        0          14.8      29.5        39 Lower Shell Forging                                                                 23.9         1.41       0.847        0.953       -20        22.8        0          11.4      22.8         26 using
      & S/C Data Inter. ToLower Shell Cire.                WF-562             0.03      0.67         41.0         1.37       0.823        0.945        40       38.7         0          19.4      38.7        117 Weld Metal)"
Inter. To Lower Shell Circ.                                                        25.7         1.37       0.823        0.945        40       24.3         0          12.2      24.3         89 Weld Metal(t
-i  using S/C Data Nozzle Shell Forging                      5P-7016            0.04      0.73         26.0        0.420       0.252        0.626        10        16.3        0          8.2        16.3        43 Nozzle Shell to Inter. Shell             WF-645             0.04      0.46         54.0        0.420       0.252        0.626       -25       33.8         0          16.9      33.8        43 Circ. Weld Metal NOTFS:
Th..
(a)  Initial RTNDT values are measured values.
(b)  Fluence, f, is based upon fer(l109 n/cm 2 , 1>1.0 MeV).
(c)  ART= I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(d)  tRThNTT = CF
* FF (e) The CF for the Braidwood Unit I Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I Weld (WF-562, heat #44201 1) and the Braidwood 2 Weld (WF-562, Heat
           # 44201 1).
(f) The Braidwood I and 2 surveillance programs contain the same heat of weld metal (heat # 44201 1). Ifence, the surveillance program results have been integrated and they are credible Braidwood Unit I Pressure-Temperature Curves

4-18 TABLE 4-19 Calculation of ART Values for the 3/4T Location        @ 22 EFPY Reactor Vessel Beltline              Material                                            F     22lb, Region Location               Identification       Cu%        NI%           CF         EFPY       3/4-t f    Y%-t FF       I(&)  ARTNDT(d        cra                    M        ART1te (x to")     (x 10")
Intermediate Shell Forging         49D383/49C344-1-1        0.05      0.73         31.0         1.41       0.305      0.675       -30       20.9         0         10.5       20.9         12 Lower shell Forging               49D867/49C813-1-1        0.05      0.74         31.0         1.41       0.305      0.675       -20       20.9         0         10.5       20.9         22 Lower Shell Forging                                                                23.9         1.41       0.305      0.675       -20       16.1         0          8.1        16.1        12
-+ using S/C Data                _
Inter. To Lower Shell Circ.              WF-562             0.03      0.67         41.0         1.37       0.297      0.668       40        27.4         0         13.7       27.4        95 Weld Metall')                                                                                                                 II Inter. to Lower Shell Circ.                                                        25.7         1.37       0.297      0.668       40        17.2         0          8.6        17.2        74 Weld Metaltf
-+ using S/C Data Nozzle Shell Forging                     5P-7016            0.04      0.73         26.0        0.420       0.091      0.398        10       10.3         0          5.2        10.3       31 Nozzle Shell to Inter. Shell             WF-645             0.04      0.46         54.0        0.420       0.091      0.398       -25       21.5         0         10.8       21.5         18 Circ. Weld Metal                                                                                                                                                .
NOTES:
(a) Initial RTNDT values are measured values.
(b) Fluence, f, is based upon f,,, (I019 n/cm 2, E>1.0 MeV).
(c) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(d)  ARTNDT     = CF
* FF (e) The CF for the Braidwood Unit I Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I Weld (WF-562, heat # 442011) and the Braidwood 2 Weld (WF-562, Heat
          # 44201 1).
(f) The Braidwood I and 2 surveillance programs contain the same heat of weld metal (heat # 442011). Hence, the surveillance program results have been integrated and they are credible Braidwood Unit I Pressure-Temperature Curves

4-19 TABLE 4-20 Calculation of ART Values for the '/4T Location @ 32 EFPY Reactor Vessel Beftline             Material                                             f  3 2 lb)
Region Location               Identification      Cu%        Ni%            CF         EFPY        YA-t  r   '/-t  FF      j(')   ARTNDTtd     ayl                      M    aA  ARTIM (xl019)     (x lo,)
Intermediate Shell Forging       49D383/49C344-1-1         0.05      0.73          31.0         2.05        1.23       1.058       -30       32.8          0        16.4       32.8        36 Lower shell Forging             49D867/49C813-1-1         0.05      0.74          31.0         2.05        1.23       1.058       -20       32.S          0         16.4      32.8        46 Lower Shell Forging                                                                23.9         2.05        1.23       1.058       *20       25.3          0        12.7       25.3        31
-. using S/C Data Inter. to Lower Shell Circ.              Wr-562            0.03      0.67          41.0          1.99       1.19       1.049       40        43.0          0        21.5       43.0       126 Weld Metal(')
Inter. to Lower Shell Circ.                                                        25.7          1.99       1.19      1.049        40        27.0          0        13.5       27.0        94 Weld Metalt)
-+ using S/C Data Nozzle Shell Forging                     5P-7016           0.04      0.73          26.0        0.608       0.365      0.722         10       18.8          0         9.4       18.8        48 Nozzle Shell to Inter. Shell             Wr-645            0.04      0.46          54.0        0.608       0.365      0.722       -25        39.0          0        19.5       39.0        53 Circ. Weld Metal NOTES:
(a) Initial RTNDT values are measured values.
(b) Fluence, f, is based upon ft,,(1019 n/cm2 , E>1.0 MeV).
(c) ART= I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(d) ARTNDT = CF* FF (e) The CF for the Braidwood Unit I Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I Weld (WF-562, heat # 44201 1) and the Braidwood 2 Weld (WF-562, Heat
         # 44201 1).
(I) The Braidwood I and 2 surveillance programs contain the same heat of weld metal (heat # 442011). Hence, the surveillance program results have been integrated and they are credible Braidwood Unit I Pressure-Temperature Curves

4-20 TABLE 4-21 Calculation of ART Values for the 3/4T Location            @32 EFPY Reactor Vessel Beltllne            Material                                              f  32'b)
Region Location              Identificstion      Cu%         Ni%            CF         EFPY        314-t f   I/.t FF       It'    ARTNDTf d t
i        OA         NI       ARfdc (x l0ll)    (x 10"l)
Intermediate Shell Forging        49D383/49C344-1-1        0.05       0.73          31.0         2.05        0.444     0.774        -30       24.0         0         12.0       24.0        18 Lower shell Forging             49D867/49C813-1*-        0.05       0.74          31.0         2.05        0.444     0.774        -20       24.0         0         12.0       24.0        28 Lower Shell Forging                                                                23.9         2.05        0.444     0.774       -20         18.5        0         9.3        18.5        17
-+ using S/C Data Inter. to Lower Shell Cire.             WF-562            0.03       0.67          41.0         1.99        0.431     0.766        40        31.4         0         15.7       31.4       103 Weld Metale"_
Inter. to Lower Shell Cire.                                                        25.7         1.99        0.431     0.766        40        19.7         0         9.9        19.7       79 Weld Metal( 0
-. using SIC Data Nozzle Shell Forging                     5P-7016          0.04        0.73          26.0        0.608        0.132     0.475         10        12.4        0         6.2        12.4       35 Nozzle Shell to Inter. Shell             WF-645           0.01        0.46          54.0        0.608        0.132     0.475       -25        25.7                   12.9       25.7       26 Circ. Weld Metal                .                                                          .                                                          .
NOTES:
(a) Initial RTNDT values are measured values.
(b) Fluence, f, is based upon f,.e(10 9 n/cm2, E>1.0 MeV).
(c) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(d) ARTNDT = CF
* FF (e) The CF for the Braidwood Unit I Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I Weld (WF-562, heat # 442011) and the Braidwood 2 Weld (WF-562, IHeat
           # 44201 1).
(f) The Braidwood I and 2 surveillance programs contain the same heat of weld metal (heat # 442011). Hence, the surveillance program results have been integrated and they are credible Braidwood Unit I Pressure-Temperature Curves

4-21 The intermediate to lower shell circumferential weld and the nozzle shell forging are the limiting beltline material for all heatup and cooldown curves to be generated. The ART value associated with this material will be used in all three sets of curves. The circumferential weld ART will be used when generating curves for Code Case N-588 (ie. Circ. Flaw). The ART associated with the limiting axial material must also be considered to determine if this case would be more conservative or overlap the circumferential flaw curves. Contained in Table 4-22 is a summary of the limiting ARTs to be used in the generation of the Braidwood Unit I reactor vessel heatup and cooldown curves.
TABLE 4-22 Summary of the Limiting ART Values used in Generation of the Braidwood Unit I Reactor Vessel Heatup and Cooldown Curves EFPY                    1/4T Limiting ART          3     T Limiting ART Limiting Circumferential Material (Weld Seam WF-562) 16                            84                          70 22                            89                          74 32                            94                          79 Limiting Axial Material (Nozzle Shell Forging 5P-7016) 16                            39                          28 22                            43                          31 32                            48                          35 Braidwood Unit I Pressure-Temperature Curves

5-1 5         HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Section 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-NP-A 181 , dated January 1996.
Figures 5-1 through 5-6 present the 16, 22 and 32 EFPY heatup and cooldown curves (without margins for possible instrumentation errors) for a heatup rate of I 000 F/hr and cooldown rates of 0, 25, 50 and 1000 F/hr using the 1996 Appendix G methodology, Code Case N-640 and Code Case N-588. The heatup and cooldown curves generated for the limiting circumferential material (Weld Seam WF-562) utilizing Code Case N-588 and the 1996 Appendix G methodology are less conservative than the heatup and cooldown curves generated for the axial flaw case even though the ART values is higher. This is true throughout the entire temperature range, including the criticality curve. Hence, the heatup and cooldown curves presented in this report were generated utilizing the limiting axial material ART (Nozzle Shell Forging 5P-7016), Code Case N-640 and the 1996 ASME Appendix G methodology.
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 through 5-6. This is in addition to other criteria which must be met before the reactor is made critical, as discussed in the following paragraphs.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-1, 5-3 and 5-5 (for the specific heatup rate being utilized). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-640 and Appendix G to Section XI of the ASME Code161 as follows:
1.5K Im< Kic                               (13)
: where, Klm is the stress intensity factor covered by membrane (pressure) stress, K1 c= 33.2 + 20.734 e 10.02(T-RTNDT)l T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 3. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40&deg;F higher than the minimum permissible temperature in the corresponding pressure-temperature curve Braidwood Unit I Pressure-Temperature Limit Curves

5-2 for heatup and cooldown calculated as described in Section 3 of this report. The minimum temperature for the inservice hydrostatic leak test for the Braidwood Unit I reactor vessel at 16 EFPY is 100IF at 2485 psig using the 1996 App. G Methodology and Code Case N-640. The minimum temperature for the inservice hydrostatic leak test for the Braidwood Unit I reactor vessel at 22 EFPY is 1031F at 2485 psig using the 1996 App. G Methodology and Code Case N-640. The minimum temperature for the inservice hydrostatic leak test for the Braidwood Unit I reactor vessel at 32 EFPY is 1081F at 2485 psig using the 1996 App. G Methodology and Code Case N-640. The approximately vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 5-1 through 5-6 define all of the above limits for ensuring prevention of nonductile failure for the Braidwood Unit I reactor vessel. The data points for the heatup and cooldown pressure-temperature limit curves shown in Figures 5-1 through 5-6 are presented in Tables 5-1 through 5-3, respectively.
Additionally, Westinghouse Engineering has reviewed the minimum boltup temperature requirements for the Braidwood Unit I reactor pressure vessel. According to Paragraph G-2222 of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Appendix G.the reactor vessel may be bolted up and pressurized to 20 percent of the initial hydrostatic test pressure at the initial RTNDT of the material stressed by the boltup. Therefore, since the most limiting initial RTNDT value is -100 F (vessel flange), the reactor vessel can be bolted up at this temperature.
Braidwood Unit I Pressure-Temperature Limit Curves

5-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: NOZZLE SHELL FORGING 5P-7016 LIMITING ART VALUES AT 16 EFPY:                                '/AT, 39&deg;F
                                                               'AT, 28&deg;F 2500                                                                                     rtim Versron:5.1 Run:8261                I
                      *II.         4stLim                 lt                                   lerlim Verslon:5w1Rn8261i I
                                                                                     ,!      i                           i 2250 i                                  I
                                 ..I                 4 2000                                                                               -rAcceptable IOperation l-'---                    -
Unacceptable               /     I OertioI IOeaion I                                                     't ',      .
1750                                                                                                       _ .   ........ ... . -
    , 1500 Heatup Rat I 00 Deg. FHr            -
Ciical Limit Deg. FIHr I
2 1 1250          :-         --      -                 f--                L:--
i f
IL
  ,   1000 750 Criticality Limit based on
                       .-                         inservice hydrostatic test                                         _   i 500 rtemperature (100 F) for the service period up to 16 EFPY                             t._
250 Boltup I       I 0            *'l .           ... I                 i                 -1 I '   i,   '                                           1 0        50         100         150    200           250     300       350      400         450          500                 550 Moderator Temperature (Deg. F)
FIGURE 5- 1 Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of I 00 0F/hr)
Applicable to 16 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Braidwood Unit I Pressure-Temperature Limit Curves

54 MATERIAL PROPERTY BASIS LIMITING MATERIAL: NOZZLE SHELL FORGING SP-7016 LIMITING ART VALUES AT 16 EFPY:                        IAT, 39 0 F 3/T, 280 F 2500 Opertim Version:5.1 Run:8261 2250   -
UnacceptableI Operation                                                            A.
ceptable 2000   -
1750   -                                                    :_ 1_,,7 0
CD
(. 1500   -
EL 1 1250 IL                          Cooldown Rates (FIlr)
   = 1000    -                steady-state,               I   ..    ......
                              -25,
                              -50, and                                !                      I
                              -100                                    :I 750 500-                                                                    .
:-   ..  .I            . -
I 250                         ~-- lTemp
                                          . l.__                        2                          i I
i           IiI nI    V' ...             I  ., ...         I      I  ,,      I     t      I          , I  r       I  ,  I 0        50       100        150    200    250          300      350         400     450        500   550 Moderator Temperature (Deg. F)
FIGURE 5-2 Braidwood Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I 00&deg;F/hr) Applicable to 16 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Braidwood Unit I Pressure-Temperature Limit Curves

5-5 TABLE 5-1 Braidwood Unit I Heatup and Cooldown Data Points for the 16 EFPY Curves Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Run = 8261      l                          Cooldown Curves Steady State                  0 25 Deg. F/Hr            50 Deg. 0F/Hr       100 Deg. 0 F/Hr Temp.        Press. Temp.         Press. Temp.          Press. Temp.         Press.
(`F)        (psig)     (OF)         (psig)     (0 F)        (psig)  (OF)         (psig) 60            0         60            0        60             0     60             0 60          1177       60           1177       60           1177    60           1177 65          1237        65          1237       65           1237    65           1237 70          1304        70          1304       70           1304    70           1304 75          1377        75          1377       75           1377    75           1377 80          1459        80          1459       80           1459     80          1459 85          1549        85          1549       85           1549    85           1549 90          1648        90          1648       90           1648    90           1648 95          1758        95          1758       95           1758    95           1758 100         1880       100          1880       100          1880    100          1880 105         2014       105          2014       105          2014    105         2014 110         2162       110          2162       110          2162    110          2162 115         2326       115          2326       115          2326    115         2326 Run    8261                                  Heatup Curves 0
100 Deg. F/Hr             Crit. Limit      Leak Test  Limit Temp.         Press. Temp.          Press. Temp.        Press.
(OF)         (psig)     (OF)         (psig)  (`F)         (psig) 60            0        100            0     82           2000 60          1151       100          1151    99          2485 65          1205       105          1205 70          1263       110          1263 75          1263       115          1263 80          1271       120          1271 85          1286      125           1286 90          1310       130          1310 95          1342       135          1342 100          1382       140          1382 105          1431       145          1431 110          1488       150          1488 115          1555       155          1555 120          1631       160          1631 125          1717       165          1717 130          1814       170          1814 135          1924       175          1924 140          2046       180          2046 145          2183       185          2183 150          2335       190          2335 Braidwood Unit 1 Pressure-Temperature Limit Curves

5-6 MATERIAL PROPERTY BASIS LIMITING MATERIAL: NOZZLE SHELL FORGING 5P-7016 LIMITING ART VALUES AT 22 EFPY:                               VAT, 43&deg;F 3/AT, 31 &deg;F 2500
                                    ,      Leak Test Limit                                        [Operlim Version:5.1 Run:9109l 2250 Acceptable 2000                                                                                                  Operation Unacceptable Operation             I 1750 Heatup Rate 100 Deg. FlHr                                1,i                 t   I                 i I                 I
: a. 1500-      -- _                     -- I          .-
I     ,Ij 2
1 1250 1                                                  !
i i         i I           I
* I
  .MS   c 9 1 000 ii U
I I                          I
* I 750  -
Criticality Limit based on
                                                 -  Inservice hydrostatic test temperature (103 F) for the t    service period up to 22 EFPY .
500  -
i           I
_       Boltu 250  -              4-          _-         Tem p               --  !-         4--               '_-   _.
0      . .,. . I    ...   ,..II. I....1.                  I        I..
                                                                             ..  -.. I....I...,I....                   I.. ...
0          50      100         150 200 250 300 350 400                                              450    500      550 Moderator Temperature (Deg. F)
FIGURE 5-3 Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 100&deg;F/hr)
Applicable to 22 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Braidwood Unit I Pressure-Temperature Limit Curves t        I

5-7 MATERIAL PROPERTY BASIS LIMITING MATERIAL: NOZZLE SHELL FORGING 5P-7016 LIMITING ART VALUES AT 22 EFPY:                                          1/4AT, 430 F 3%T,31 &deg;F 2500 Operlim Verslon:5.1 Run:9109 2250       Unacceptable                              ,
Operation                       lAccptabl                                                                         cl Operation 2000 1750
* I                                _j 0~
1500
                                              /.           'I                            I            i CL 1250 Cooldown Rates
                                           .. r U  1000                                 steady-state,
                                           -26, 40, and
                                           -100 750 K
_  .,        . _4-                                  -- --
500
                                                .                     ,i Ii                I 250                  .I.
i     I                                                                            i I t.              j                     1,...-
0                   -i                    ,    . T , I,i rI  , .     - S, I   -T I,    I   i.    -1i,       i -  I  , --             I 0             50                 100       150        200 250 300 350 400                                       450        500     550 Moderator Temperature (Deg. F)
FIGURE 5-4 Braidwood Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100&deg;F/hr) Applicable to 22 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Braidwood Unit I Pressure-Temperature Limit Curves

5-8 TABLE 5-2 Braidwood Unit I Heatup and Cooldown Curve Data Points for the 22 EFPY Curves Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Run - 9109                                 Cooldown Curves Steady State            25 Deg. &deg;F/lr          50 Deg. &deg;F/Hr         100 Deg. &deg;F/1lr Temp.         Press. Temp.      1   Prcss. Temp.          Press. Temp.          Prcss.
(&deg;F)         (psig)    (OF)          (psig)    (OF)       l  (psig)  (IF)          (psig) 60             0        60             0       60              0      60            0 60           1133       60           1133      60            1133     60           1133 65           1188       65           1188      65            1188     65           1188 70           1250       70           1250      70            1250    70            1250 75           1318       75           1318      75            1318    75            1318 80           1393       80           1393       80           1393     80           1393 85           1476       85           1476       85           1476    85            1476 90           1568       90           1568      90            1568     90           1568 95           1669       95           1669      95            1669    95            1669 100           1781      100           1781      100           1781    100           1781 105          1905      105           1905      105           1905    105           1905 110         2042       110          2042       110           2042    110          2042 115          2194       115          2194       115          2194     115          2194 120          2361       120          2361       120           2361    120          2361 Run   9109                                   llcatup Curves 100 Deg. FIlir            Crit. Limit        Leak Test Limit Temp.      1   Press. Temp.          Press. Temp.          Press.
('F)     1   (psig)    (OF)          (psig)  (OF)          (psig) 60             0       103             0     86           2000 60           1110      103           1110    103          2485 65           1163      105           1163 70           1218      110           1218 75           1222      115           1222 80           1229      120           1229 85           1243      125           1243 90           1264      130           1264 95           1294      135           1294 100           1331      140           1331 105           1376      145           1376 110           1430      150           1430 115           1492      155           1492 120           1563      160           1563 125           1644      165           1644 130           1736      170           1736 135           1838      175           1838 140           1954      180           1954 145          2082       185          2082 150          2225       190          2225 155           2384      195          2384 Braidwood Unit I Pressure-Temperature Limit Curves

5-9 MATERIAL PROPERTY BASIS LIMITING MATERIAL: NOZZLE SHELL FORGING SP-7016 LIMITING ART VALUES AT 32 EFPY:                       /4T, 480 F
                                                     %T,350 F 2500             ,
2500       iLeak                      Tst Limit 2250      --                                        i      t i Ij       1                     _  _1 2000 Unacceptable                                                   I1 Operation 1750             '      1---         1l-V)       i           /  l         }         jCritical Li
  . 1500          Heatup Ra w            l100 Deg. FlHr LD
  &sect;    1250 1250
                --- -- J-------t--
U
     *1o000 760         __         ____             __
Criticality Limit ba Inservice hydrosta
                                !       I__      temperature (108 I So                    -service                     period upl 250                       -
                             .Boitup 0
0         50     100     150 200 250 300 350 400                450    500     550 Moderator Temperature (Deg. F)
FIGURE 5-5 Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of I 00&deg;F/hr)
Applicable to 32 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Braidwood Unit I Pressure-Temperature Limit Curves

5-10 MATERIAL PROPERTY BASIS LIMITING MATERIAL: NOZZLE SHELL FORGING SP-7016 LIMITING ART VALUES AT 32 EFPY:.                                 114T, 480 F 3/4T, 35 0 F 2500-Operlim Version:5.1         ~Ru29~844 i
I 2250-.
I           i Unacceptabe Operaioni 2000  -
I  Acceptable             --       ~-1__
Opeaton 1750 -
: a. 1500    -                                                                                                        I 2
1 1250 .                                                                                                             L Cooldown Rates 0   1000               -W~      FHr)O.
LI)
I                  steady-state,
                                 -25.                                                                                  i
                            *-50. and                                                                                            I 750-
* 100                       -
I I
i i        i I       I    !        i
                                        !     /               t i        I                                                        I
                             -        -1 . -      ..
i                     i i                     I I        i i         I            Boltup 250  -
77F I ; II
                                          -          I Temp                                                 ____              -    -
U       -I-,-,-       - I....i I                       __   II Jr-ri
                                                                        ,  -    II r-r-1
                                                                                     -,- I, i-i-ri I -
iIin I-i-i-I-fl
                                                                                                                - -r-0           50         100            150     200     250        300        350        400          450         500         550 Moderator Temperature (Deg. F)
FIGURE 5-6      Braidwood Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I 00 0F/hr) Applicable to 32 EFPY Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Braidwood Unit 1 Pressure-Temperature Limit Curves

5-11 TABLE 5-3 Braidwood Unit I Heatup and Cooldown Curve Data Points for the 32 EFPY Curves Using 1996 Appendix G and Code Case N-640 Methodology (Without Margins for Instrumentation Errors)
Run - 29844        1                      . Cooldown Curves Steady State            25 Deg. _F/Hr          50 Deg. &deg;F/llr          100 Deg. 0 F/lr Temp. l Press.           Temp. 1   Press. Temp.           Press. Tcmp.          Press.
(OF)          (psig)     (OF)     r   (psig)     (&deg;F)      ]    (psig)  (&deg;F)           (psig) 60            0         60             0        60              0       60             0 60           1082        60           1078      60             1079     60            1079 65           1133       65           1133       65            1133      65            1133 70           1188       70           1188       70            1188      70            1188 75           1250       75           1250       75            1250      75            1250 80           1318        80          1318       80            1318      80            1318 85           1393       85           1393       85            1393      85            1393 90           1476       90           1476       90            1476      90            1476 95           1568       95           1568       95            1568      95            1568 100           1669       100          1669      100            1669     100            1669 105           1781       105          1781      105            1781     105            1781 110           1905       110          1905      110            1905     110            1905 115          2042        115          2042      115            2042     115           2042 120          2194        120          2194      120            2194     120           2194 125          2361        125          2361      125            2361     125           2361 Run =29844ileatup                                       Curves 100 Deg._FIIlr            Crit. Limit         Leak Test Limit Temp.         Press. Temp-.          Press   Temp.          Press.
(&deg;F)         (psig)    (OF)           (psig)   (f)           (psig) 60             0       108              0      91            2000 60           1064      108            1114     108           2485 65           1114      110            1166 70           1166      115            1172 75           1172      120            1176 80           1176      125            1188 85           1188      130            1207 90           1207      135            1234 95           1234      140            1267 100           1267      145            1308 105           1308      150            1357 110           1357      155            1414 115           1414      160            1479 120           1479      165            1554 125           1554      170            1638 130           1638      175            1732 135           1732      180            1838 140           1838      185            1956 145           1956      190           2088 150          2088       195           2235 155          2235      200            2397 160          2397 Braidwood Unit I Pressure-Temperature Limit Curves

6-1 6      REFERENCES I      Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U.S.
Nuclear Regulatory Commission, May, 1988.
2      Cases of ASME Boiler and Pressure Vessel Code, Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", Approved March 1999.
3      10 CFR Part 50, Appendix G."Fracture Toughness Requirements", Federal Register, Volume 60, No. 243, dated December 19, 1995.
4      WCAP-1 6143-P, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2", K.R. Hsu, November 2003.
5      Cases of ASME Boiler and Pressure Vessel Code, Case N-588, "Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels Section XI, Division 1",
Approved December 12, 1997.
6      ASME Boiler and Pressure Vessel Code, Section XI, "Rule for Inservice Inspection of Nuclear Power Plant Components", Appendix G,"Fracture Toughness Criteria for Protection Against Failure", December 1995.
7      WCAP-1 5366, Revision 2, "Commonwealth Edison Company Braidwood Unit I Surveillance Program Credibility Evaluation," T.J. Laubbam, November 2003.
8      WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D. Andrachek, et al., January 1996.
9       1989 Section 111, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Material for Vessels".
10     Babcock & Wilcox drawing number 185297E, Revision 2; "Longitudinal Section".
II     WCAP-15316, Revision 1,Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit I Reactor Vessel Radiation Surveillance Program", E. Terek, et al., December 1999.
12     CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1996.
13     Intermediate Shell Forging, Certification No. JQA-74-13-MC, Rev. 1, P.O. Number 318092MT, Part Mark 24-2, Heat Number 49D383/49C344-1 -1 Material Test Report, Dated October 31, 1974.
References

6-2 14     WCAP-15369, "Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program", T. J. Laubham, et al., March 2000.
15     Nuclear Design Information Transmittal NDIT No. BRW-DIT-97-321, Rev. 0.
16     Nuclear Design Information Transmittal NDIT No. BRW-DIT-2000-0010.
17     WCAP-14824, Revision 2, Byron Unit I Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood", T. J. Luabham, et al., November 1997.
References

A-I APPENDIX A Thermal Stress Intensity Factors (Kit)
Thermal Stress Intensity Factors

A-2 TABLE Al K1t Values for I 00&deg;F/hr Heatup Curve (16 EFPY)
Vessel Temperature          1/4T Thermal       Vessel Temperature       3/4T Thermal Water         @ 1/4T Location for               Stress      ( 3/4T Location for           Stress Temp.           100'F/hr Heatup         Intensity Factor      100'F/hr Heatup        Intensity Factor
(`F)                  (OF)           (KSI SQ. RT. IN.)             (OF)           (KSI SQ. RT. IN.)
60                  56.01                   -0.9945               55.05                0.4767 65                  58.62                   -2.4413               55.31                1.4422 PT Curves are Limited by the % T Location up to 65 0 F and %/4  T Limited for the Remainder of the Curve 70                  61.70                   -3.6836               56.01                2.4209 75                  65.01                   -4.8600               57.20                3.3365 80                  68.58                   -5.8718               58.80                4.1525 85                  72.27                   -6.7937               60.80                4.8806 90                  76.14                   -7.5912               63.14                5.5230 95                  80.11                   -8.3181               65.77                6.0965 100                  84.21                   -8.9517               68.68                6.6042 105                  88.40                   -9.5267               71.81                7.0564 110                  92.68                  -10.0310               75.15                7.4589 115                  97.04                  -10.4906               78.67                7.8204 120                 101.46                  -10.8960               82.35                8.1440 125                 105.95                  -11.2667               86.16                8.4352 130                 110.48                  -11.5953               90.10                8.6969 135                 115.06                  -11.8972               94.14                8.9336 140                 119.68                  -12.1661               98.28                9.1474 145                 124.34                  -12.4147              102.49                9.3417 150                 129.03                  -12.6374              106.78                9.5183 Vessel Radius to the h/4T and 3/4T Locations are as follows:
* 1/4T Radius = 88.750"
* 3/4T Radius = 93.000" Thermal Stress Intensity Factors

A-3 TABLE A2 K1t Values for l 00F/hr Cooldown Curve (16 EFPY)
Vessel Temperature 100'F/hr Cooldown Water        @ 1/4T Location for 114T Thermal Stress Temp.        100 0 F/hr Cooldown    Intensity Factor (OF)                (OF)         (KSI SQ. RT. IN.)
115               139.61              15.2187 110               134.53              15.1525 105               129.45              15.0860 100               124.36              15.0200 95                119.28              14.9538 90                114.19              14.8881 85                109.11              14.8221 80                104.02              14.7566 75                 98.94              14.6909 70                 93.86              14.6257 65                 88.77              14.5603 60                 83.69              14.4946 NOTE: All Cooldown Rates are limited by the Steady State Condition.
Thernal Stress Intensity Factors

A-4 TABLE A3 Kit Values for 100&deg;F/hr Heatup Curve (22 EFPY)
Vessel Temperature           1/4T Thermal         Vessel Temperature       3/4T Thermal Water        @ 1/4T Location for                Stress          @ 3/4T Location for          Stress Temp.          100&deg;F/hr Heatup           Intensity Factor        100F/hr Heatup         Intensity Factor (OF)                   (OF)            (KSI SQ. RT. IN.)               (OF)           (KSI SQ. RT. IN.)
60                   56.01                   -0.9945                 55.05                 0.4767 65                   58.62                   -2.4413                 55.31                 1.4422 70                   61.70                   -3.6836                 56.01                 2.4209 PT Curves are Limited by the       4 T Location up to 700 F and 3AT Limited for the Remainder of the Curve 75                   65.01                   -4.8600                 57.20                 3.3365 80                   68.58                   -5.8718                 58.80                 4.1525 85                   72.27                   -6.7937                 60.80                 4.8806 90                   76.14                   -7.5912                 63.14                 5.5230 95                   80.11                   -8.3181                 65.77                 6.0965 100                   84.21                   -8.9517                 68.68                 6.6042 105                   88.40                   -9.5267                 71.81                 7.0564 110                  92.68                  -10.0310                 75.15                 7.4589 115                  97.04                  -10.4906                 78.67                 7.8204 120                  101.46                  -10.8960                 82.35                 8.1440 125                  105.95                  -11.2667                 86.16                 8.4352 130                  110.48                  -11.5953                 90.10                 8.6969 135                  115.06                  -11.8972                 94.14                 8.9336 140                 119.68                  -12.1661                 98.28                 9.1474 145                  124.34                  -12.4147                 102.49                9.3417 150                  129.03                  -12.6374                 106.78                9.5183 155                 133.75                  -12.8446                 111.14                9.6799 I                      I Thermal Stress Intensity Factors

A-S TABLE A4 Kit Values for I 000 F/hr  Cooldown Curve (22 EFPY)
Vessel Temperature 1000 F/hr Cooldown Water        @ 1/4T Location for 1/4T Thermal Stress Temp.       100'F/hr Cooldown         Intensity Factor (OF)                  (OF)          (KSI SQ. RT. IN.)
120                144.70                15.2854 115                139.61                15.2187 110                134.53                15.1525 105                129.45                15.0860 100                124.36                15.0200 95                 119.28                14.9538 90                 114.19                14.8881 85                 109.11                14.8221 80                 104.02                14.7566 75                  98.94                14.6909 70                  93.86                14.6257 65                  88.77                14.5603 60                  83.69                14.4946 NOTE: All Cooldown Rates are limited by the Steady State Condition.
Thermal Stress Intensity Factors

A-6 TABLE A5 KI, Values for I 00F/hr Heatup Curve (32 EFPY)
Vessel Temperature            1/4T Thermal         Vessel Temperature      3/4T Thermal Water          @ 1/4T Location for                Stress          @ 3/4T Location for         Stress Temp.           100'F/hrHeatup              Intensity Factor       1001F/hrHeatup        Intensity Factor
( 0 F)                 (OF)               (KSI SQ. RT. IN.)              (OF)          (KSI SQ. RT. IN.)
60                   56.01                     -0.9945                 55.05                0.4767 65                   58.62                     -2.4413                 55.31                1.4422 70                   61.70                     -3.6836                 56.01                2.4209 PT Curves are Limited by the       1/4 T Location up to 70 0F and % T Limited for the Remainder of the Curve 75                   65.01                     -4.8600                 57.20                3.3365 80                   68.58                     -5.8718                 58.80                4.1525 85                   72.27                     -6.7937                 60.80                4.8806 90                   76.14                     -7.5912                 63.14                5.5230 95                   80.11                     -8.3181                 65.77                6.0965 100                  84.21                     -8.9517                 68.68                6.6042 105                  88.40                     -9.5267                 71.81                7.0564 110                  92.68                    -10.0310                 75.15                7.4589 115                  97.04                    -10.4906                 78.67                7.8204 120                 101.46                    -10.8960                 82.35                8.1440 125                  105.95                    -11.2667                 86.16                8.4352 130                 110.48                    -11.5953                 90.10                8.6969 135                 115.06                    -11.8972                 94.14                8.9336 140                 119.68                    -12.1661                 98.28                9.1474 145                 124.34                    -12.4147                 102.49               9.3417 150                 129.03                    -12.6374                 106.78               9.5183 155                 133.75                    -12.8446                 111.14               9.6799 160                 138.48                    -13.0315                 115.55               9.8277
                .1                          .a                      a Thermal Stress Intensity Factors

A-7 TABLE A6 K1 , Values for I 00 &deg;F/hr Cooldown Curve (32 EFPY)
Vessel Temperature 1000 F/hr Cooldown Water        @ 1/4T Location for 1/4T Thermal Stress Temp.        100 0 F/hr Cooldown      Intensity Factor (OF)                  (OF)          (KSI SQ. RT. IN.)
125                 149.78               15.3518 120                 144.70               15.2854 115                 139.61               15.2187 110                 134.53               15.1525 105                 129.45               15.0860 100                 124.36               15.0200 95                  119.28               14.9538 90                  114.19               14.8881 85                  109.11               14.8221 80                  104.02               14.7566 75                  98.94                14.6909 70                  93.86                14.6257 65                  88.77                14.5603 60                  83.69                14.4946 NOTE: All Cooldown Rates are limited by the Steady State Condition except for 60'F, where the 25&deg;F/hr rate controls the higher rates.
Thermal Stress Intensity Factors

ATTACHMENT 4-B WCAP-15373, Revision 2 Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation

_.. I- ... _.__
101 I Westinghouse Non-Proprietary Class 3 WCAP-15373                                          November 2003 Revision 2 Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation e Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15373, Revision 2 Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation November 2003 Approved.
e.A. resham, Manager A/Reactor Component Design & Analysis Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02003 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST OF TABLES ..................                                                                iv LIST OF FIGURES .vii PREFACE ......
EXECUTIVE 

==SUMMARY==
 .ix I     INTRODUCTION ............................................................                 1-1 2      PURPOSE ............................................................                     2-1 3      CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .................3-1 4      CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .......................................... 4-1 5      HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ....................... 5-1 6      REFERENCES ............................................................                  6-1 APPENDIX A        THERMAL STRESS INTENSITY FACTORS ................................... A-I

viii PREFACE This report has been technically reviewed and verified by:
Reviewer:                                                             J.H. Ledger    9 d Revision 1:
An error was detected in the "OPERLIM" Computer Program that Westinghouse uses to generate pressure-temperature (PT) limit curves. This error potentially effects the heatup curves when the 1996 Appendix G Methodology is used in generating the PT curves. It has been determined that WCAP-15364 Rev. 0 was impacted by this error. Thus, this revision provides corrected curves from WCAP-15364 Rev.
0.
Note that only the heatup curves and associated data point tables in section 5 have changed. The cooldown curves and data points remain valid and were not changed.
Revision 2:
This revision was completed to incorporate an updated reference for the flange elimination from the PT limit curves. In the previous revision, WCAP-15315 was provided as justification for the flange notch elimination and it was changed to WCAP-16143-P in this revision. In addition to this change, the thermal stress intensity factors for the highest heatup and cooldown rates were added to this report in Appendix A.

ix EXECUTIVE 

==SUMMARY==

The purpose of this report is to generate pressure-temperature limit curves for Braidwood Unit 2 for normal operation at 12, 16,22 and 32 EFPY using the Following methodologies: WCAP 14040-NP-A, the 1996 ASME Boiler and Pressure Vessel Code, Section XI Appendix G.ASME Code Case N-588, ASME Code Case N-640 and WCAP-16143-P. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART). The 1/4T and 3/4T values are summarized in Tables 4-18 through 4-25 and were calculated using the circumferential weld WF-562, Heat 442011 (The limiting material for circumferentially oriented flaws, Code Case N-588) and nozzle shell forging SP-7056 (The limiting material for axial flaws). The pressure-temperature limit curves were generated for a heatup rate of I00&deg;F/hr and cooldown rates of 0, 25, 50 and 1000 F/hr. The axial oriented flaw cases are limiting for all curves at each EFPY value evaluated. Hence, only the axial oriented flaw curves are presented in this report and they can be found in Figures 5-1 through 5-8.

iv LIST OF TABLES Table 4-1  Irradiated Temperature of the Braidwood Unit I and 2 Reactor Vessels ......................... 4-2 Table 4-2  Summary of the Peak Pressure Vessel Neutron Fluence Values at 12 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV) .4-4 Table 4-3  Summary of the Peak Pressure Vessel Neutron Fluence Values at 16 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV) .4-5 Table 4-4  Summary of the Peak Pressure Vessel Neutron Fluence Values at 22 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV) .4-5 Table 4-5  Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm 2 , E> 1.0 MeV) .4-6 Table 4-6  Calculated Integrated Neutron Exposure of the Braidwood Unit 2 Surveillance Capsules Tested to Date .4-6 Table 4-7  Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program .4-7 Table 4-8  Calculation of Best Estimate Cu and Ni Weight Percent for the Braidwood Unit 2 Forging Materials .4-8 Table 4-9  Calculation of the Average Cu and Ni Weight Percent for the Braidwood Unit 2 Surveillance Weld Material Only (Heat # 442011) .4-8 Table 4-10 Calculation of Best Estimate Cu and Ni Weight Percent Values for the Braidwood Units I & 2 Weld Material (Using Braidwood I & 2 Chemistry Test Results) ............... 4-9 Table 4-11 Reactor Vessel Beltline Material Unirradiated Toughness Properties ........................... 4-10 Table 4-12 Calculation of Chemistry Factors for Braidwood Unit 2 using Surveillance Capsule Data .4-11 Table 4-13 Summary of the Braidwood Unit 2 Reactor Vessel Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 ...4-12 Table 4-14 Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 12 EFPY Heatup / Cooldown Curves .                                                      4-13 Table 4-15 Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 16 EFPY Heatup / Cooldown Curves .                                                      4-13

V Table 4-16 Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 22 EFPY Heatup / Cooldown Curves ...............................................         4-14 Table 4-17 Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 32 EFPY Heatup / Cooldown Curves .4-14 Table 4-18 Calculation of the ART Values for the 1/4T Location @ 12 EFPY .                                 4-15 Table 4-19 Calculation of the ART Values for the 3/4T Location ( 12 EFPY .                                 4-16 Table 4-20 Calculation of the ART Values for the 1/4T Location ( 16 EFPY .4-17 Table 4-21 Calculation of the ART Values for the 3/4T Location ( 16 EFPY .4-18 Table 4-22 Calculation of the ART Values for the 1/4T Location @ 22 EFPY .4-19 Table 4-23 Calculation of the ART Values for the 3/4T Location     @ 22 EFPY .4-20 Table 4-24 Calculation of the ART Values for the 1/4T Location @32 EFPY .4-21 Table 4-25 Calculation of the ART Values for the 3/4T Location @ 32 EFPY .4-22 Table 4-26 Summary of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T Locations for 12 EFPY ..............                                                            4-23 Table 4-27 Summary of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T Locations for 16 EFPY ..............                                                            4-24 Table 4-28 Summary of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T Locations for 22 EFPY ..............                                                            4-25 Table 4-29 Summary of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T Locations for 32 EFPY ..............                                                            4-26 Table 5-1  Braidwood Unit 2 Heatup Data at 12 EFPY Using Code Case N-588 vs. 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) ............................ 5-5 Table 5-2  Braidwood Unit 2 Cooldown Data at 12 EFPY Using Code Case N-588 vs. 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) ............................ 5-6 Table 5-3  Braidwood Unit 2 Heatup Data at 16 EFPY Using Code Case N-588 vs. 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) ............................ 5-9 Table 54   Braidwood Unit 2 Cooldown Data at 16 EFPY Using Code Case N-588 vs. 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) .......................... 5-10

vi LIST OF TABLES Table 5-5 Braidwood Unit 2 Heatup Data at 22 EFPY Using Code Case N-588 vs. 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) .......................... 5-13 Table 5-6 Braidwood Unit 2 Cooldown Data at 22 EFPY Using Code Case N-588 vs. 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) .......................... 5-14 Table 5-7 Braidwood Unit 2 Heatup Data at 32 EFPY Using Code Case N-588 vs. 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) .......................... 5-17 Table 5-8 Braidwood Unit 2 Cooldown Data at 32 EFPY Using Code Case N-588 vs. 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) .......................... 5-18

vii LIST OF FIGURES Figure 5-1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100&deg;F/hr) Applicable to 12 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors) ...................................... 5-3 Figure 5-2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100 0F/hr) Applicable to 12 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors) .................. 54 Figure 5-3 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 1000 F/hr) Applicable to 16 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors) ...................................... 5-7 Figure 54  Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100&deg;F/hr) Applicable to 16 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors) .................. 5-8 Figure 5-5 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 1000 F/hr) Applicable to 22 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors) ....................................   -IlI Figure 5-6 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 1000 F/hr) Applicable to 22 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors) ................ 5-12 Figure 5-7 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of I 00&deg;F/hr) Applicable to 32 EFPY Using Code Case N-588 vs. 1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors) .................................... 5-15 Figure 5-8 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100 0F/hr) Applicable to 32 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors) ................ 5 -16

1-I I        INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60 0F.
RTNDT  increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT(IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials"[11 .
Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves for normal operation.
NOTE: For the reactor vessel radiation surveillance program, Babcock and Wilcox Co. supplied Westinghouse with sections of SA508 Class 3 forging material used in the core region of the Braidwood Station Unit No. 2 reactor pressure vessel (Specifically from lower shell forging 50D102-1/50C97-1).
Also supplied was a weldment made with weld wire heat # 442011 Linde 80 flux, lot number 0344, which is identical to that used in the actual fabrication of the intermediate to lower shell girth weld of the pressure vessel).
Introduction

2-1 2        PURPOSE The Commonwealth Edison Company (now known as Exelon Nuclear) contracted Westinghouse to analyze surveillance capsule W from the Braidwood Unit 2 reactor vessel and perform an evaluation for a 5% Uprating. As a part of these analysis Westinghouse generated new heatup and cooldown curves for 12, 16, 22 and 32 EFPY. These new Pressure-Temperature Curves are to be developed utilizing the following methodologies:
* Regulatory Guide 1.99, Revision 2111,
* ASME Code Case N-640t2 ],
* Elimination of the flange requirement of Appendix G to 10 CFR Part 50131 per WCAP-16143-P, "Reactor Vessel Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2 "'l4],
* ASME Code Case N-588 15] (where applicable),
* Methodology of the 1996 ASME B&P Vessel Code, Section XI, Appendix GQ16 1, and
* The PT Curves will be developed WITHOUT margins or instrumentation errors.
Based on the above methodologies, two sets of PT Curves will be generated. Set one will consist of the circumferential flaw methodology (ASME Code Case N-588) in combination with 1996 Appendix G to ASME Section XI and the Kic methodology (ASME Code Case N-640) for the limiting circumferential weld material. Set two will consist of the 1996 Appendix G to ASME Section XI and the K1 ,
methodology (ASME Code Case N-640) for the limiting forging/base metal material. Both sets of curves will used the methodology to eliminate the 10 CFR Part 50 Appendix G flange requirements (from WCAP-16143-P). The final PT curves to be presented herein will be the most limiting set of curves. If the situation arises where portions of each set of curves are limiting, then composite curves will be generated that are based on the most limiting data (i.e. Circ. Flaw or Axial Flaw Case).
The purpose of this report is to present the calculations and the development of the Commonwealth Edison Company Braidwood Unit 2 heatup and cooldown curves for 12, 16, 22 and 32 EFPY. This report documents the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2111, for all the beltline materials and the development of the heatup and cooldown pressure-temperature limit curves for normal operation.
Per the request of the Commonwealth Edison Company, the surveillance weld data from the Braidwood Unit I and Unit 2 surveillance programs has been integrated. Note that Braidwood Unit I surveillance weld is identical to the surveillance weld (Heat No. 442011) at Braidwood Unit 2. Per WCAP-15368171 ,
all the surveillance data has been determined to be credible.
Purpose

3-1 3        CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1      Overall Approach Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements"13] specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements. Section Xl, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components", Appendix GI , contains the conservative methods of analysis.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1 , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1 ,, for the metal temperature at that time. K1, is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of PT Limit Curves for Section Xl"[2 6 ] of the ASME Appendix G to Section XI. The Kic curve is given by the following equation:
Kic =332 +20.734
* er002(T-RTNDT)]                             (1) where,     Kl=         reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K1 , curve is based on the lower bound of static critical K1 values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2 and SA-508-3 steel.
3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C* Kim + Kit < K1,                                     (2)
: where, Km =           stress intensity factor caused by membrane (pressure) stress Kt=            stress intensity factor caused by the thermal gradients K,=            function of temperature relative to the RTNDT of the material C      =       2.0 for Level A and Level B service limits C      =        1.5 for hydrostatic and leak test conditions during which the reactor core is not critical Criteria For Allowable Pressure-Temperature Relationships

3-2 For membrane tension, the corresponding K, for the postulated defect is:
Kim = Mm x (pRi/ t)                                                    (3) where, Mm for an inside surface flaw is given by:
Mm     =      1.85 for f    < 2, Mm     =      0.926 ft for 2 <*f3.464, Mm     =      3.21 for r > 3.464 Similarly, Mm for an outside surface flaw is given by:
Mm     =      1.77for    t <2, Mm     =      0.893 A\l for 2*<,*<3.464, Mm     =      3.09 for ft > 3.464 and p = internal pressure, Ri = vessel inner radius, and t vessel wall thickness.
For bending stress, the corresponding K1 for the postulated defect is:
Klb = Mb
* Maximum Stress, where Mb is two-thirds of Mm The maximum K1 produced by radial thermal gradient for the postulated inside surface defect of G-2120 is K1, = 0.953x10, 3 x CR x t2 5, where CR is the cooldown rate in IF/hr., or for a postulated outside surface defect, K1, = 0.753x10, 3 x HU x t25 , where HU is the heatup rate in 'F/hr.
The through-wall temperature difference associated with the maximum thermal K1 can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.
G-2214-2 for the maximum thermal K1 .
(a)      The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).
(b)      Alternatively, the K1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4/4-thickness inside surface defect using the relationship:
A. = (1.0359Co + 0.6322Ci + 0.4753C2 + 03855C3) * \                                   (4)
Criteria For Allowable Pressure-Temperature Relationships

3-3 or similarly, Krr during heatup for a l/4-thickness outside surface defect using the relationship:
Kit = (1.043Co + 0.630Ci + 0.48 IC2 + 0.40 1C3)
* A/;                                 (5) where the coefficients C0, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
a(x) = Co + Ci(x / a) + C2(x / a)2 + C3(x / a)3                                     (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.
Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldwon Limit Curves" 18 1Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.
At any time during the heatup or cooldown transient, K1, is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K1 t, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of KI, at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1 , exceeds K1,, the calculated allowable pressure during cooldown will be greater than the steady-state value.
Criteria For Allowable Pressure-Temperature Relationships

34 The above procedures are needed because there is no direct control on temperature at the I/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of1/4Ta       defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; 1/4T crack during heatup is lower than the cK1 for the therefore, the K1 , for the                                                     1/4T crack during steady-state conditions at the same coolant temperature. During heatup,     especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1, values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the l/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a I/4T flaw located at the1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
Code Case N-588: Circumferential Welds:
In 1997, ASME Section XI, Appendix G was revised to add methodology for the use of circumferential flaws when considering circumferential welds in developing pressure-temperature limit curves. This change was also implemented in a separate Code Case, N-588.
The earlier ASME Section XI, Appendix G approach mandated the postulation of an axial flaw in circumferential welds for the purposes of calculating pressure-temperature limits. Postulating the Appendix G reference flaw in a circumferential weld is physically unrealistic because the length of the reference flaw is 1.5 times the vessel thickness and is much longer than the width of the vessel girth welds.
Criteria For Allowable Pressure-Temperature Relationships

3-5 In addition, historical experience, with repair weld indications found during pre-service inspection and data taken from destructive examination of actual vessel welds, confirms that any flaws are small, laminar in nature and are not oriented transverse to the weld bead orientation. Because of this, any defects potentially introduced during fabrication process (and not detected during subsequent non-destructive examinations) should only be oriented along the direction of the weld fabrication. Thus, for circumferential welds, any postulated defect should be in the circumferential orientation.
The revision to Appendix G now eliminates additional conservatism in the assumed flaw orientation for circumferential welds. The following revisions were made to ASME Section XI, Appendix G:
G-22 14.1 Membrane Tension...
The K, corresponding to membrane tension for the postulated circumferential defect of -2120 is Kim= Mmx(pRi/t) where, Mm for an inside surface flaw is given by:
Mm    =      0.89 for 4f <2, Mm    =      0.443 4    for 2 < t < 3.464, Mm    =      1.53 for   t > 3.464 Similarly, Mm for an outside surface flaw is given by:
Mm    =      0.89 for   t < 2, Mm    =      0.443 rt for 2 < r < 3.464, Mm    =       1.53 for ft > 3.464 Note again, that the only change relative to the OPERLIM computer code was the addition of the constants for Mm in a circ. weld limited condition. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. As stated previously, the P-T curve methodology is unchanged from that described in WCAP-14040!S] Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.
Criteria For Allowable Pressure-Temperature Relationships

3-6 3.3 Closure Head/Vessel Flange Requirements 10 CFR Part 50, Appendix G[3 1 addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120 0 F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for Braidwood Unit 2 reactor vessel.
This requirement was originally based on concerns about the fracture margin in the closure flange region.
During the boltup process, stresses in this region typically reach over 70 percent of the steady-state stress, without being at steady-state temperature. The margin of 1201F and pressure limitation of 20 percent of the hydrotest pressure were developed using the KIA fracture toughness from the mid 1970's.
Improved knowledge of the fracture toughness and other issues which affect the integrity of the reactor vessel have led to the recent change to allow the use of Kic in development of pressure-temperature curves, as contained in Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1,,121.
The discussion given in WCAP-I 6143-P, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2," concluded that the integrity of the closure head/vessel flange region is not a concern for Byron or Braidwood, therefore the requirement of 10 CFR Part 50, Appendix G,can be eliminated from the Pressure-Temperature Curves contained in this report.
Criteria For Allowable Pressure-Temperature Relationships

4-1 4         CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART = InitialRTNDT+ A RTDT + Margin                                   (7)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-233 1 of Section III of the ASME Boiler and Pressure Vessel Codel 91 . If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTNDT    is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows:
A RTADT      = CF
* f(0.280.101ogfi                         (8)
To calculate ARTNDT at any depth (e.g., at I/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
f (dieptx) = fsface
* e(02 4)                             (9) where x inches (vessel beltline thickness is 8.5 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNDT at the specific depth.
The Westinghouse Radiation Engineering and Analysis group evaluated the vessel fluence projections and the results are presented in Section 6 of WCAP-15369r1b] and WCAP-15316, Rev. 1i"'. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with the methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"18 3. Tables 4-2 through 4-5, herein, contain the calculated peak vessel surface fluence values along with the Regulatory Guide 1.99, Revision 2, 1/4T and 3/4T calculated fluences used to calculate the ART values for all beltline materials in the Braidwood Unit 2 reactor vessel. Additionally, the Braidwood Unit 2 calculated surveillance capsule fluence values are presented in Table 4-6.
Ratio Procedure and Temperature Adjustment:
The ratio procedure, as documented in Regulatory Guide 1.99, Revision 2, Position 2.1, was used, where applicable, to adjust the measured values of ARTNDT of the weld materials for differences in copper/nickel content. This adjustment is performed by multiplying the measured ARTNDT by the ratio of the vessel chemistry factor to the surveillance material chemistry factor. The adjusted measured ARTNDT values are then used to calculate the chemistry factor for the vessel materials.
Calculation of Adjusted Reference Temperature

4-2 From NRC Industry Meetings on November 12, 1997 and February 12Ih and 13Ih of 1998, procedural guidelines were presented to adjust the ARTNDT for temperature differences when using surveillance data from one vessel applied to another vessel. The following guidance was presented at these industry meetings:
Irradiation temperature and fluence (or fluence factor) are first order environmental variables in assessing irradition damage... To account for differences in temperature between surveillance specimens and vessel, an adjustment to the data must be performed. Studies have shown that for temperatures near 550'F, a 10 F decrease in irradiation temperature will result in approximately a 1IF increase in ARTNDT.
For capsules with irradiation temperature of T~psue and a plant with an irradiation temperature of Tpl,,,,,, an adjustment to normalize ARTNDT.ieasurcd to Tpla, is made as follows:
Temp. Adjusted ARTNDT = ARTNDT, mneasured +    1.0*( Tcapsul  - Tplant)               (10)
Per Reference 12, page C-5, following are the average operating inlet temperatures of the Braidwood Unit I and 2 reactor vessels along with when the capsules were removed.
TABLE 4-1 Irradiation Temperature of the Braidwood Unit I and 2 Reactor Vessels Fuel Cycle             Braidwood 1           Capsule        Braidwood 2            Capsule I                 557 0 F              U                5570 F                U 2                 551 0F                                5510F 3                 551 0F                                5510 F 4                  5510 F              X                5510 F                X 5               551-5540 F                             551 0F*
6                 554&deg;F                                551 0F*
7                554 0F*               W             550 0 F**               W Average                 5530 F                               552OF T em p.         _ _  _   _    _ _  _  _  _  _  _  _  _   _    _   _     _ _ _   _   _ _
          **        Information provided by ComEd in NDIT No. BRW-DIT-97-321        13 1.
               **   Information provided by CornEd inNDIT No. BRW-DIT-2000-001 0t 14 ).
Capsules U and X were exposed to the same average operating temperature of 5531F. Capsule W was exposed to the same average operating temperature of 553IF for 5 of the 7 cycles. For cycles 6 and 7, capsule W saw only a difference of IOF in average operating temperature. Thus, the inlet operating temperature of both Braidwood Unit I and Unit 2 are essentially the same for the time of interest and no temperature adjustments are required.
Calculation of Adjusted Reference Temperature

4-3 Chemistry Factor:
The chemistry factor is obtained from the tables in Regulatory Guide 1.99, Revision 2 using the best estimate average copper and nickel content as reported in Tables 4-8 through 4-10. The chemistry factors were also calculated using Position 2.1 from the Regulatory Guide 1.99, Revision 2 using all available surveillance data. Per Reference 7, the surveillance weld data for Braidwood Unit 2 is credible while the surveillance forging material is non-credible. In addition, Reference 7 also shows that the Table chemistry factor is non-conservative and the surveillance chemistry factor should be used with a full margin term. This assessment also calculated the vessel weld (including temperature and chemistry adjustment) chemistry factor using Braidwood Unit 2 stand alone data and Braidwood Unit I and 2 combined data. The chemistry factor was determined to be 25.70 F. Position 2.1 chemistry factors are calculated in Table 4-12.
Explanation of Margin Term:
When there are "two or more credible surveillance data sets"1t1 available for Braidwood Unit 2, Regulatory Guide 1.99 Rev. 2 (RG1.99R2) Position 2.1 states "To calculate the Margin in this case, use Equation 4; the values given there for aA may be cut in half". Equation 4 from RGI.99R2 is as follows:
M=24 T,2.
Standard Deviation for Initial RTNDT Margin Term, a, If the initial RTNDT values are measured values, which they are in the case of Braidwood Unit 2, then a, is equal to 0WF. On the other hand, if the initial RTNDT values were not measured, then a generic value of 171F (base metal and weld metal) would have been required to be used for a,.
Standard Deviation for ARTNDT Margin Term, caA Per RGI.99R2 Position 1.1, the values of ca, are referred to as "281F for welds and 17'F for base metal, except that aA need not exceed 0.50 times the mean value of ARTNDT." The mean value of ARTNDT is defined in RGI.99R2 by Equation 2 and defined herein by Equation 8.
Per RGI.99R2 Position 2.1, when there is credible surveillance data, a, is taken to be the lesser of 12 ARTNDT or 14'F (280 F/2) for welds, or 8.51F (17/F12) for base metal. Where ARTNDT again is defined herein by Equation 8.
Calculation of Adjusted Reference Temperature

44 Summary of the Margin Term Since ca, is taken to be zero when a heat-specific measured value of initial RTNDT are available (as they are in this case), the total margin term, based on Equation 4 of RGI.99R2, will be as follows:
* Position 1.1:       Lesser of ARTNDT or 561F for Welds Lesser of ARTNDT or 340 F for Base Metal
* Position 2.1:       Lesser of ARTNDT or 281F for Welds Lesser of ARTNDT or 17'F for Base Metal TABLE 4-2 Summary of the Peak Pressure Vessel Neutron Fluence Values at 12 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV)
Material(')                          Surface(b)                   1/4 T                3/4 T (n/cm 2 , E>1.O MeV)    (n/cm 2 , E>1.O MeV) (n/cm 2 , E>1.O NMeV)
Intermediate Shell Forging                   7.43 x 1018             4.46 x 1018          1.61 x 1018 49D963-1/49C904-1 Lower Shell Forging                      7.43 x 1018             4.46 x 1018          1.61 x 10" 5OD102-1/50C97-1 Nozzle Shell Forging 5P-7056                   2.16 x 108              1.30 x 108           4.68 x 1017 Intermediate to Lower Shell Forging Circ.             7.20 x 10i 8            4.32 x 108           1.56 x 10 Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell Forging            2.16 x 1018             1.30 x 1018          4.68 x 1017 Circ. Weld Seam WF-645 Notes:
(a)    All remaining vessel materials are below 1 x 1017 n/cm 2 , E > 1.0 MeV.
(b) Calculated Surface fluence documented in Reference 10.
Calculation of Adjusted Reference Temperature

4-5 TABLE 4-3 Summary of the Peak Pressure Vessel Neutron Fluence Values at 16 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MWV)
Material(')                          Surface(b)                   %T                  3/4 T (n/cm2 , E>1.O MeV)         2 (n/cm , E>1.O MeV)         2 (n/cm , E>1.O MeV)
Intermediate Shell Forging                   9.87 x 1018            5.93 x 1018          2.14 x 1018 49D963-1/49C904-1 Lower Shell Forging                       9.87 x 10"             5.93 x 10's          2.14 x 1018 5OD102-1/50C97-1 Nozzle Shell Forging 5P-7056                    2.86 x 1018            1.72 x 1O08          6.19 x    1017 Intermediate to Lower Shell Forging Circ.              9.54 x 1018            5.73 x 10s           2.07 x 10s Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell Forging             2.86 x 1018            1.72 x 1018          6.19 x 1017 Circ. Weld Seam WF-645 Notes:
(a) All remaining vessel materials are below I x 1017 n/cm2 , E > 1.0 MeV.
(b) Calculated Surface fluence documented in Reference 10.
TABLE 4-4 Summary of the Peak Pressure Vessel Neutron Fluence Values at 22 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV)
Material(')                         Surface(b)                   %T                  /4 T (n/cm2 , E>1.0 MeV)    (n/cm 2 , E>1.0 MeV) (n/cm2 , E>1.0 MeV)
Intermediate Shell Forging                   1.35 x 1O09            8.11 x 108           2.92 x 1018 49D963-1/49C904-1 Lower Shell Forging                       1.35 x 1019            8.11 x 10's          2.92x 1018 5OD102-1/50C97-1 Nozzle Shell Forging 5P-7056                    3.91 x 1018            2.35 x lOS           8.47 x 1017 Intermediate to Lower Shell Forging Circ.             1.31 x 10"9            7.87 x 108           2.84 x 1018 lWeld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell Forging             3.91 x 1018            2.35 x 1018          8.47 x 1017 Circ. Weld Seam WF-645 Notes:
(a)     All remaining vessel materials are below I x 1017 n/cm2 , E > 1.0 MeV.
(b)     Calculated Surface fluence documented in Reference 10.
Calculation of Adjusted Reference Temperature

4-6 TABLE 4-5 Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV)
Material(')                           Surface(b)                %1T                3 T 2                     2                  2 (n/cm , E>.O MeV)      (n/cm , E>l.O MeV) (n/cm , E>1.O MeV)
Intermediate Shell Forging                   1.96 x 10"             1.18 x 10"         4.24 x 108 49D963-1/49C904-1 Lower Shell Forging                       1.96 x 109             1.18 x 10'9        4.24 x 101 50D102-1/50C97-1 Nozzle Shell Forging 5P-7056                    5.67 x 10"             3.40 x 1018        1.23 x 1018 Intermediate to Lower Shell Forging Circ.             1.89 x 10'9            1.13 x 10'9        4.09 x 1018 Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell Forging            5.67 x 1018            3.40 x 10"8        1.23 x 1018 Circ. Weld Seam WF-645 Notes:
(a)    All remaining vessel materials are below I x 1017 n/cm2 , E > 1.0 MeV.
(b)    Calculated Surface fluence documented in Reference 10.
TABLE 4-6 Calculated Integrated Neutron Exposure of the Braidwood Unit 2 Surveillance Capsules Tested to Date Capsule                               Fluence U                   4.00x 1018 n/cm 2 , (E > 1.0 MeV)
X                   1.23 x1019 n/cm 2 , (E > 1.0 MeV)
W                   2.25 x 10'9 n/cm 2 , (E > 1.0 MeV)
Notes:
(a)    Documented in WCAP-15369'E1].
Contained in Table 4-7 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline materials. These measured shift values were obtained using CVGRAPH, Version 4.11"), which is a symmetric hyperbolic tangent curve-fitting program.
Calculation of Adjusted Reference Temperature

4-7 TABLE 4-7 Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program Material                               Capsule              Measured 30 ft-lb Transition Temperature Shift(')
Lower Shell Forging                              U                             -9.73 50D102-1/5OC97-1                                X                             -9.42 (Tangential Orientation)                           W                              4.53 Lower Shell Forging                              U                             -0.13 50D102-1/5OC97-1                                X                             33.94 (Axial Orientation)                             W                              33.2 Surveillance Program                             U                             -0.58 Weld Metal                                  X                              26.3 (Heat # 442011)                               W                              23.9 Heat Affected Zone                              U                            -34.45 X                             -9.54 W                              4.03 Notes:
(a) From capsule W analysis resultsll0 l.
Table 4-8 contains the calculation of the best estimate weight percent copper and nickel for the Braidwood Unit 2 base materials in the beltline region. Table 4-9 contains the calculation of the best estimate weight percent copper and nickel for the Braidwood Unit 2 surveillance weld material, while Table 4-10 presents the overall best estimate average for that heat of weld. Table 4-11 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RTNDT of the beltline materials and vessel flanges. The weight percent values of Cu and Ni given in Table 4-11 were used to generate the calculated chemistry factor (CF) values based on Tables I and 2 of Regulatory Guide 1.99, Revision 2, and presented in Table 4-13. Table 4-12 provides the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Position 2.1, which are also summarized in Table 4-13.
Calculation of Adjusted Reference Temperature

4-8 TABLE 4-8 Calculation of the Best Estimate Cu and Ni Weight Percent for the Braidwood Unit 2 Forging Materials Intermediate Shell Forging               Lower Shell Forging 49D963-1/49C904-1                   50D102-1/50C97-1 Reference                     Cu %                Ni %              Cu %              Ni%
WCAP-I 1188 1 161                 0.03                0.71 WCAP-I     11881161                ---                ---               0.06              0.75 WCAP-I 1188 1161                   ---                ---              0.057              0.77 Charpy FL-6 1'71                  ---                ---              0.049             0.745 Ref. 18                                                          0.056             0.804 Charpy FL-331 01                  ---                ---              0.065             0.680 Charpy FL-42 1101                 ---                ---              0.065             0.690 Charpy FT-43 1101                 ---                                 0.060             0.795 Charpy FT-451 01                  - --               ---              0.066             0.840 Best Estimate Average(')                0.03               0.71              0.06              0.76 Notes:
(a)     The best estimate average was rounded per ASTM E29, using the "Rounding Method".
TABLE 4-9 Calculation of the Average Cu and Ni Weight Percent for the Braidwood Unit 2 Surveillance Weld Material Only (Heat # 442011)
Reference                      Weight % Copper           Weight % Nickel WCAP-14824, Rev.2(')                        0.033                    0.708 Charpy FW-38(b)                          0.044                    0.703 Charpy FW-34(b)                          0.036                    0.774 Charpy FW-41(b)                          0.037                    0.747 Charpy FW-43(b)                          0.042                    0.670 Surveillance Weld Average(c)                      0.03                     0.71 Notes:
(a)     This is the average of 32 data points.
(b)     Charpy Specimens From Capsule W of Braidwood Unit 2 (Ref. 10).
(c)     The Surveillance weld average was rounded per ASTM E29, using the "Rounding Method".
Calculation of Adjusted Reference Temperature

4-9 TABLE 4-10 Calculation of Best Estimate Cu and Ni Weight Percent Values for the Braidwood Units I & 2 Weld Material (Using Braidwood I & 2 Chemistry Test Results)
Chemistry Type                            Reference              Weight % Copper            Weight % Nickel B&W WQ: BAW-2261                               Ref. 12                     0.028                        0.63 B&W WQ: BAW-2261                               Ref. 12                      0.03                        0.65 B&W WQ: BAW-2261                               Ref. 12                      0.04                        0.67 Braidwood Unit 2 Surv. Data Ave.(')                 Table 4-5                    0.034                        0.710 Braidwood Unit I Surv. Data Ave. (b)                 Ref. 11                     0.032                        0.671 BEST ESTIMATE AVERAGE                                                           0.03(c)                     0.67(c)
Notes:
(a) The weld material in the Braidwood Unit 2 surveillance program was made of the same wire and flux type as the reactor vessel inter. to lower shell girth seam weld (Weld seam WF-562, Wire Heat No. 442011, Flux Type Linde 80, Flux Lot No.8061). The surveillance weld flux lot # is 0344.
(b) The Braidwood Unit I surveillance weld is identical to that used in the Braidwood Unit 2 reactor vessel core region girth seam (WF-562) heat number 442011, with a Linde 80 type flux, lot number 8061.
(c) The best estimate chemistry values were obtained using the "average of averages" approach. In addition the best estimate average was rounded per ASTM E29, using the "Rounding Method".
Calculation of Adjusted Reference Temperature

4-10 TABLE 4-11 Reactor Vessel Beltline Material Unirradiated Toughness Properties Material Description                           Cu (%)             Ni(%)     Initial RTNDT Closure Head Flange 3P6566/5P7547/4P6986                       -                0.75        20 Vessel Flange 124P455                            0.07              0.70        20 Nozzle Shell Forging 5P7056                         0.04              0.90       30 Intermediate Shell Forging 49D963-1/49C904-1                   0.03              0.71       -30 Lower Shell Forging 50D102-1 / 50C97-1                     0.06              0.76      -30 Inter. to Lower Shell Forging Circ. Weld Seam                 0.03              0.67       40 WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell Forging Circ.               0.04              0.46      -25 Weld Seam WF-645 (Heat H4498)
Braidwood Units I & 2 Surveillance Program                0.03, 0.03        0.67, 0.71 Weld Metals (Heat # 442011)
Notes:
(a) The initial RTNDT values for the plates and welds are based on measured data.
Calculation of Adjusted Reference Temperature

4-11 TABLE 4-12 Calculation of Chemistry Factors for Braidwood Unit 2 using Surveillance Capsule Data Material             Capsule        Capsule      j0a)     FF(b)            ARTNDT(c)         FF*ARTNDT          FF2 Lower Shell Forging            U               0.400             0.746               0.0()                0          0.557 5OD102-1/50C97-1               X                 1.23            1.058               0.0()                0          1.119 (Tangential)                               2.25             1.220               4.53               5.53         1.488 Lower Shell Forging            U               0.400             0.746               0.0')                0          0.557 5OD102-1/5OC97-1               X                 1.23            1.058               33.94              35.91        1.119 (Axial)                 W                2.25             1.220                33.2              40.50         1.488 SUM:      81.94        6.328 CF = E(FF
* RTNDT)     + Z( FF2) = (81.94) * (6.328) = 12.91F Braidwood Unit I             U               0.387             0.737              17.06)              12.57        0.543 Surveillance              X                 1.24            1.060             30.15(d)             31.96        1.124 Weld Metal                W                2.09             1.201             4 9. 68 (d)          59.67        1.442 Braidwood Unit 2             U               0.400             0.746              O.( Oe                0          0.557 Surveillance              X                 1.23            1.058              26.3(d)             27.83        1.119 Weld Metal                W                2.25             1.220              23.9(d)             29.16        1.488 SUM:     161.19        6.273 CF= :(FF*RTNDT)        +  X( FF 2)= (161.19)+ (6.273) = 25.7 0F Notes:
(a) f = Calculated fluence from the Braidwood Unit 2 capsule W dosimetry analysis results 110kthe Braidwood Unit I calculated fluences are from capsule W analysis 11, (x 1019 n/cm2 , E > 1.0 MeV).
(b) FF = fluence factor = 1 o2s0.I'Jog O (c) ARTNDT    values are the measured 30 ft-lb shift values for Braidwood Unit 2 taken from WCAP-15369C1"'. Braidwood Unit I values are from WCAP-15316 Rev 1 (Ref. 11).
(d) The surveillance weld metal ARTNDT values have not been adjusted (i.e. the ratio factor is 1.0).
(e) Actual values of ARTNDT are -9.73 (Cap U Tang.), -9.42 (Cap. X Tang.), -0.13 (Cap. U Axial), -0.58 (Cap U Weld). This physically should not occur; therefore for conservatism (i.e. higher chemistry factor) a value of zero will be used.
Calculation of Adjusted Reference Temperature

4-12 TABLE 4-13 Summary of the Braidwood Unit 2 Reactor Vessel Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Material                             Chemistry Factor Position 1.1         Position 2.1 Intermediate Shell Forging                20.0F 49D963-1/49C904-1 Lower Shell Forging 50D102-1/50C97-1             37.0F                12.9 0 F Nozzle Shell Forging 5P7056                26.0 0 F               ---
Intermediate Shell to Lower Shell Forging         41.0 0 F             25.7 0F Circ. Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell            54.0 0 F Forging Circ. Weld Seam WF-645 (Heat H4498)
Braidwood Unit I & 2 Surveillance             41.0 0 F               ---
Program Weld Metal Calculation of Adjusted Reference Temperature

4-13 Contained in Tables 4-14 through 4-17 is the summary of the fluence factors (FF) used in the calculation of adjusted reference temperatures for the Braidwood Unit 2 reactor vessel beltline materials for 12, 16, 22 and 32 EFPY.
TABLE 4-14 Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 12 EPFY Heatup/Cooldown Curves Material                           114 T F      1/4T FF            314T F        314 T FF 2                                2 (n/cm , E > 1.0 MeV)             (n/cm , E >1.0  MeV)
Intermediate Shell Forging              4.46 x Io0 8 0.775           1.61 x 108        0.5 19 49D963-1/49C904-1 Lower Shell Forging                 4.46 x 10"       0.775           1.61 x 10"        0.519 50D102-1/50C97-1 Nozzle Shell Forging 5P-7056              1.30 x 10t       0.471          4.68 x 1017        0.282 Intermediate to Lower Shell Forging Circ.       4.32 x IO8       0.767           1.56 x 1018       0.512 Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell          1.30 x 1018      0.471          4.68 x 1017        0.282 Forging Circ. Weld Seam WF-645 TABLE 4-15 Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 16 EPFY Heatup/Cooldown Curves Material                           1/4 T F      1/4T FF            3/4T F        3/4 T FF (nlcm2, E > 1.0 MeV)             (n/cm 2, E >1.0 MeV)
Intermediate Shell Forging              5.93 x 10"       0.854          2.14 x 1018        0.586 49D963-1/49C904-1 LowerShell Forging                  5.93 x 1018      0.854          2.14x 10's         0.586 50D102-1/50C97-1 Nozzle Shell Forging 5P-7056              1.72 x 1018      0.534          6.19 x 1017        0.328 Intermediate to Lower Shell Forging Circ.       5.73 x 1018      0.844          2.07 x IO"'        0.578 Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell          1.72 x I 018      0.534         6.19 x 1017        0.328 Forging Circ. Weld Seam WF-645 Calculation of Adjusted Reference Temperature

4-14 TABLE 4-16 Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 22 EPFY Heatup/Cooldown Curves Material                          1/4 T F          1/4T FF        3/4T F           3/4 T FF (n/cm 2, E > 1.0 MeV)           (ncm  2 , E >1.0 MeV)_
Intermediate Shell Forging               8.11 x loll         0.941      2.92 x 1018           0.663 49D963-1 /49C904-1 Lower Shell Forging                  8.11 x 1018         0.941      2.92 x 108            0.663 50D102-1/50C97-1 Nozzle Shell Forging SP-7056              2.35 x 10 8       i 0.609      8.47 x 1017           0.384 Intermediate to Lower Shell Forging Circ.        7.87 x 10"           0.933     2.84 x 10"            0.656 Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell          2.35 x 1018          0.609     8.47 x 1017           0.384 Forging Circ. Weld Seam WF-645 TABLE 4-17*
Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 32 EPFY Heatup/Cooldown Curves Material                          1/4 T F          1/4T FF        3/4T F           314 T FF (n/cm 2 , E > 1.0 MeV)          (n/cm2 , E >1.0 MeV)   _
Intermediate Shell Forging              1.18 x 1019          1.046     4.24 x 1018           0.762 49D963-l/49C904-l Lower Shell Forging                  1.18 x 10"           1.046     4.24 x 108            0.762 50D102-1/5OC97-1 Nozzle Shell Forging SP-7056              3.40 x 10"           0.703      1.23 x 1018          0.460 Intermediate to Lower Shell Forging Circ.       1.13 x IO9           1.034     4.09 x IO8            0.752 Weld Seam WF-562 (Heat 442011)
Nozzle Shell to Intermediate Shell          3.40 x 10            0.703      1.23 x 1018          0.460 Forging Circ. Weld Seam WF-645 Contained in Tables 4-18 through 4-25 are the calculations of the ART values used for the generation of the 12, 16, 22 and 32 EFPY heatup and cooldown curves.
Calculation of Adjusted Reference Temperature

4-15 TABLE 4-18 Calculation of the ART Values for the 1/4T Location @ 12 EFPY Reactor Vessel Beltline             Material         Cu%         NI%         CF 8d)    f@ 12(m)     /4 t r    Y4-t         I     ARTNDTO) at1  o     M   ART7b)
    . Region Location               Identification                                         EFPY      (x 10")     FF Intermediate Shell                  49D963-1/          0.03       0.71         20.0      0.743      0.446    0.775        -30       15.5   0   7.75  15.5 Forging                             49C904-1 Lower Shell Forging                 50D102-1/          0.06       0.76         37.0      0.743      0.446    0.775        -30       28.7   0   14.35 28.7   27 50C97-1 Lower shell Forging                                                            12.9      0.743      0.446    0.775        -30       10.0   0   17.0  34.0   14
  -+ using S/C Data                                                                                                               I Inter. to Lower Shell                WF-562            0.03       0.67        41.0       0.720      0.432    0.767         40       31.4   0   15.7  31.4  103 Circ. Weld Metal Inter. to Lower Shell                                                         25.7       0.720      0.432    0.767         40       19.7   0   9.85  19.7   79 Circ. Weld Metal
   - using S/C Data Nozzle Shell Forging                 5P-7056           0.04        0.90       26.0       0.216      0.130    0.471         30       12.2   0    6.1  12.2   54 Nozzle Shell to Inter.               WF-645            0.04        0.46        54.0      0.216      0.130    0.471        -25       25.4   0   12.7  25.4   26 Shell Circ. Weld Metal NOTES:
(a)    Fluence, f, is the calculated peak clad/base metal interface fluence (1019 n/cm2, E>l.0 MeV).
(b)    ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c)    ARTNDT - CF
* FF (d)    The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I and 2 Welds (WF-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-16 TABLE 4-19 Calculation of the ART Values for the 3/4T Location          @ 12 EFPY Reactor Vessel Beltline            Material         Cu%          Ni%       CFdd)      f    I 12(s)   %-t t      %-t           I   ARTNDT(t) al  a    M   ARTrb)
Region Location             Identification                                          EFPY     (x 10")       FF Intermediate Shell                49D963-1/         0.03         0.71      20.0         0.743      0.161     0.519        -30     10.4    0  5.2  10.4   -9 Forging                           49C904-1 Lower Shell Forging               5oD 102-1/         0.06        0.76      37.0         0.743      0.161     0.519        -30     19.2    0  9.6  19.2    8 50C97-1I Lower shell Forging                                                         12.9        0.743      0.161     0.519        -30      6.7    0  17.0 34.0   11
  -e using S/C Data Inter, to Lower Shell              WF-562            0.03        0.67      41.0         0.720      0.156     0.512         40     21.0    0  10.5 21.0   82 Circ. Weld Metal Inter. to Lower Shell                                                      25.7         0.720      0.156     0.512         40     13.2    0  6.6  13.2   66 Circ. Weld Metal e- using S/C Data Nozzle Shell Forging               5P-7056           0.04        0.90      26.0         0.216     0.0468     0.282         30      7.3    0  3.65  7.3   45 Nozzle Shell to Inter.             WF-645            0.04        0.46      54.0         0.216     0.0468     0.282        -25     15.2    0  7.6  15.2    5 Shell Circ. Weld Metal NOTES:
(a) Fluence, f, is the calculated peak clad/base metal interface fluence (1019 n/cm 2, E>1.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c)    ARTNDT = CF
* FF (d)    The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I and 2 Welds (WF-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-17 TABLE 4-20 Calculation of the ART Values for the 1/4T Location @ 16 EFPY Reactor Vessel Beltline           Material         Cu%          Ni%       CF(d)      f  16(')     '/-t f     'A-t         I   ARTNDT! a1  aa    M   ARTK)
Region Location            Identification                                          EFPY      (x 10' )      FF Intermediate Shell                49D963-1/         0.03         0.71      20.0         0.987      0.593     0.854        -30     17.1  0  8.55  17.1   4 Forging                           49C904-1                                                      I Lower Shell Forging               50D102-/1         0.06         0.76      37.0         0.987      0.593     0.854        -30     31.6   0 15.8  31.6  33 50C97-1 Lower shell Forging                                                         12.9        0.987      0.593     0.854        -30     11.0   0 17.0  34.0   15
  . using S/C Data                                                                                                                                            l Inter. to Lower Shell              WF-562           0.03         0.67      41.0         0.954      0.573     0.844         40     34.6   0 17.0  34.0  109  l Circ. Weld Metal Inter. to Lower Shell                                                      25.7         0.954      0.573     0.844         40     21.7  0  10.85 21.7  83 Circ. Weld Metal
   -+ using S/C Data Nozzle Shell Forging               5P-7056          0.04         0.90      26.0         0.286      0.172     0.534         30     13.9  0  6.69  13.9  58 Nozzle Shell to Inter.             WF-645           0.04         0.46      54.0         0.286      0.172     0.534        -25     28.8  0  14.4  28.8  33 Shell Circ. Weld Metal NOTES:
(a) Fluence, f, is the calculated peak clad/base metal interface fluencc (1019 n/cm 2, E>1.0 MeV).
(b) ART = I + tRTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c)    ARTNDT = CF
* FF (d)    The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I and 2 Welds (Wr-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-18 TABLE 4-21 Calculation of the ART Values for the 3/4T Location @ 16 EFPY Reactor Vessel Beltline          Material         Cu%         Ni%         CFd)       fr@ 16&deg;"     %-t f     %-t          I   ARTNDT (0 aT   aA   M   ARTb)
Region Location            Identification                                           EFPY     (x 10")     FF Intermediate Shell               49D963-1/         0.03        0.71        20.0         0.987      0.214   0.586        -30     11.7     0 5.85  11.7  -7 Forging                          49C904-1 Lower Shell Forging              50D102-1I          0.06       0.76        37.0         0.987      0.214   0.586        -30     21.7     0 10.85 21.7   13 50C97-1I Lower shell Forging                                                         12.9        0.987      0.214   0.586        -30      7.6     0  17.0 34.0   12
  -* using S/C Data Inter. to Lower Shell             WF-562           0.03        0.67        41.0         0.954      0.207   0.578         40     23.7     0 11.85 23.7  87 Circ. Weld Metal Inter. to Lower Shell                                                      25.7         0.954      0.207   0.578         40     14.9     0 7.45  14.9  70 Circ. Weld Metal
  -+ using S/C Data Nozzle Shell Forging              5P-7056           0.04        0.90       26.0         0.286     0.0619   0.328         30      8.5     0 4.25   8.5  47 Nozzle Shell to Inter.            WF-645           0.04        0.46        54.0         0.286     0.0619   0.328        -25     17.7     0 8.85  17.7  10 Shell Circ. Weld Metal NOTES:
(a) Fluence, f, is the calculated peak clad/base metal interface fluence (1019 n/cm2 , E>1.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c) ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Braidwvood I and 2 Welds (WF-562, heat i 442011).
Calculation of Adjusted Reference Temperature

4-19 TABLE 4-22 Calculation of the ART Values for the 1/4T Location @ 22 EFPY Reactor Vessel Beltline          Material         Cu%          Ni%        CF(d)     rO 22(s)    . -t r    V4-t         I    ARTNDT(C) al ,r        M   ART"b)
Region Location           Identification                                         EFPY      (x 10")     FF Intermediate Shell               49D963-1/         0.03         0.71       20.0        1.35       0.811   0.941        -30      18.8   0      9.4  18.8    8 Forging                          49C904-1                   IlI Lower Shell Forging              50D102-1/         0.06         0.76        37.0       1.35       0.811   0.941        -30     34.8    0      17.0 34.0  39 50C97-1 Lower shell Forging                                                         12.9       1.35       0.811   0.941        -30      12.1   0      17.0 34.0   16
  -e using S/C Data Inter. to Lower Shell             WF-562           0.03         0.67       41.0        1.31       0.787   0.933        40      38.3    0     19.15 38.3  117 Circ. Weld Metal Inter. to Lower Shell                                                      25.7        1.31       0.787   0.933        40      24.0    0      12.0 24.0  88 Circ. Weld Metal
  -l. using SIC Data                              .
Nozzle Shell Forging              5P-7056          0.04         0.90       26.0       0.391       0.235   0.609        30       15.8   0      7.9  15.8  62 Nozzle Shell to Inter.            WF-645           0.04         0.46        54.0      0.391       0.235   0.609        -25     32.9    0     16.45 32.9  41 Shell Circ. Weld Metal NOTES:
(a) Fluence, f, is the calculated peak clad/base metal interface fluence (1019 n/cm2, E>l.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c) ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I and 2 Welds (Wr-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-20 TABLE 4-23 Calculation of the ART Values for the 3/4T Location ( 22 EFPY Reactor Vessel Beltline           Material         Cu%          Ni%       CF(d)   -  r    22(')    /e-t f    %-t           I   ARTNDri) al  CA    M   ART(b)
Region Location             Jdentification                                         EFPY      (x 10')       FF Intermediate Shell                49D963-1/         0.03         0.71      20.0          1.35     0.292      0.663        -30     13.3   0  6.65  13.3  -3 Forging                           49C904-1                                                                                                                    l Lower Shell Forging               50D102-1/         0.06         0.76      37.0          1.35     0.292      0.663        -30     24.5   0  12.25 24.5   19 50C97-1 Lower shell Forging                                                         12.9         1.35      0.292     0.663        -30      8.6   0  17.0  34.0   13
  -e using S/C Data Inter. to Lower Shell              WF-562           0.03         0.67      41.0          1.31     0.284      0.656         40     26.9   0  13.45 26.9  94 Circ. Weld Metal Inter. to Lower Shell                                                      25.7          1.31     0.284      0.656         40     16.9   0  8.45  16.9  74 Circ. Weld Metal
  -* using S/C Data Nozzle Shell Forging               5P-7056          0.04         0.90      26.0         0.391     0.0847     0.384         30     10.0   0   5.0  10.0  50 Nozzle Shell to Inter.             WF-645           0.04         0.46      54.0         0.391     0.0847     0.384        -25     20.7   0  10.35 20.7   16 Shell Circ. Weld Metal NOTES:
2 (a) Fluence, f, is the calculated peak clad/base metal interface fluence (l109 n/cm , E>l.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c)    ARTMT =    CF
* FF (d)    The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I and 2 Welds (WF-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-21 TABLE 4-24 Calculation of the ART Valucs for the 1/4T Location @ 32 EFPY Reactor Vessel Beltline          Material         Cu%          Ni%        CF(d)     f i32(')     Yet f     Y4-t         I   ARTNDT(c) a  aa    M   ARTr Region Location            Identification                                         EFPY      (x 10")     FF Intermediate Shell               49D963-1/          0.03        0.71       20.0         1.96      1.18     1.046       -30     20.9    0 10.45 20.9   12 Forging                          49C904-1 Lower Shell Forging              50D102-1/          0.06        0.76       37.0         1.96      1.18     1.046       -30     38.7    0 17.0  34.0   43 50C97-1 Lower shell Forging                                                        12.9         1.96      1.18     1.046       -30      13.5   0 17.0  34.0   18
  -+ using S/C Data Inter. to Lower Shell             WF-562           0.03         0.67       41.0         1.89      1.13     1.034        40     42.4    0 21.2  42.4  125 Circ. Weld Metal Inter. to Lower Shell                                                      25.7         1.89      1.13    1.034         40     26.6    0 13.3  26.6  93 Circ. Weld Metal
  -+ using S/C Data Nozzle Shell Forging              5P-7056          0.04         0.90       26.0        0.567     0.340    0.703         30      18.3   0 9.15  18.3  67 Nozzle Shell to Inter.            WF-645           0.04         0.46       54.0        0.567     0.340    0.703        -25     38.0    0 19.0  38.0  51 Shell Circ. Weld Metal NOTES:
(a) Fluence, f, is the calculated peak clad/base metal interface fluence (1019 n/cm2, E>1.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c) ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I and 2 Welds (WF-562, heat f 442011).
Calculation of Adjusted Reference Temperature

4-22 TABLE 4-25 Calculation of the ART Values for the 3/4T Location @ 32 EFPY Reactor Vessel Beltline          Material         Cu%         Ni%         CF(d)      fr    32(') -/4-t        f  %-t    I   ARTNDT.O a    GA   M   ART1b)
Region Location            Identification                                           EFPY            (x 109)   FF Intermediate Shell               49D963-1/         0.03         0.71       20.0           1.96           0.424   0.762 -30      15.2    0  7.6 15.2    0 Forging                          49C904-1 Lower Shell Forging              50D102-1/         0.06         0.76       37.0           1.96           0.424   0.762 -30     28.2     0 14.1 28.2   26 50C97-1 Lower shell Forging                                                         12.9          1.96           0.424   0.762 -30      9.8     0 17.0 34.0   14 e- using S/C Data Inter. to Lower Shell             WF-562           0.03         0.67       41.0           1.89           0.409   0.752  40     30.8     0 15.4 30.8  102 Circ. Weld Metal Inter. to Lower Shell                                                      25.7           1.89           0.409   0.752  40     19.3     0 9.65 19.3   79 Circ. Weld Metal
  -* using S/C Data Nozzle Shell Forging              5P-7056          0.04         0.90       26.0          0.567           0.123   0.460  30     12.0     0 6.0  12.0   54 Nozzle Shell to Inter.            WF-645           0.04         0.46       54.0          0.567           0.123   0.460 -25     24.8     0 12.4 24.8   25 Shell Circ. Weld Metal NOTES:
(a) Fluence, f, is the calculated peak clad/base metal interface fluence (1019 n/cm2 , E>l.0 MeV).
(b) ART = I + ARTNsT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c) ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Braidwood I and 2 Welds (WF-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-23 The girth weld WF-562 and the nozzle shell forging 5P-7056 are the limiting beltline materials for all heatup and cooldown curves to be generated. The ART value associated with these materials will be used in all four sets of curves. The girth weld ART will be used when generating curves for Code Case N-588 (ie. Circ. Flaw). The ART associated with the limiting axial material must also be considered to determine if this case would be more conservative or overlap the circ. flaw curves. Contained in Tables 4-26 through 4-29 is a summary of the limiting ARTs to be used in the generation of the Braidwood Unit 2 reactor vessel heatup and cooldown curves.
TABLE 4-26 Summary of Adjusted Reference Temperature (ART) values at the l/4T and 3/4T Locations for 12 EFPY Material                                               12 EFPY 1/4T ART                         3/4T ART Intermediate Shell Forging                               1                             -9 49D963-1 /49C904-1 Lower Shell Forging                                     27                              8 50D1102-1/50C97-1
          - Using Surveillance Data                                14                            11 Circumferential Weld WF-562                            103                             82
          - Using Surveillance Data                             79(m)                           66(m)
Nozzle Shell Forging 5P-7056                          54Ub)                           45(b)
Circumferential Weld WF-645                             26                              5 NOTES:
(a) These ART values were used to generate the Braidwood Unit 2 heatup and cooldown curves in Figures 5-1 and 5-2. See note (b).
(b) These ART values, using the '96 App. G Methodology, produced a more conservative curve (heatup and cooldown) with no overlap versus those curves generated with the circ. flaw ART values and Code Case N-588 Methodology.
Calculation of Adjusted Reference Temperature

4-24 TABLE 4-27 Summary of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T Locations for 16 EFPY Material                                   16 EFPY 1/4T ART                  3/4T ART Intermediate Shell Forging                        4                        -7 49D963-1/49C904-1 Lower Shell Forging                              33                        13 50D1 02-1/50C97-1
         - Using Surveillance Data                         15                       12 Circumferential Weld WF-562                      109                       87
         - Using Surveillance Data                            ;3.                  70(m)
Nozzle Shell Forging 5P-7056                    58(b)                     47(b)
Circumferential Weld WF-645                      33                        10 NOTES:
(a) These ART values were used to generate the Braidwood Unit 2 heatup and cooldown curves in Figures 5-3 and 54. See note (b).
(b) These ART values, using the '96 App. G Methodology, produced a more conservative curve (heatup and cooldown) with no overlap versus those curves generated with the circ. flaw ART values and Code Case N-588 Methodology.
Calculation of Adjusted Reference Temperature

4-25 TABLE 4-28 Summary of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T Locations for 22 EFPY Material                                    22 EFPY 1/4T ART                   3/4T ART Intermediate Shell Forging                        8                        -3 49D963-1/49C904-1 Lower Shell Forging                              39                        19 50D102-1/50C97-1
         - Using Surveillance Data                        16                        13 Circumferential Weld WF-562                      117                       94
         - Using Surveillance Data                       88(s)                     74(a)
Nozzle Shell Forging 5P-7056                    62((b)                    50b)
Circumferential Weld WF-645                      41                        16 NOTES:
(a) These ART values were used to generate the Braidwood Unit 2 heatup and cooldown curves in Figures 5-5 and 5-6. See note (b).
(b) These ART values, using the '96 App. G Methodology, produced a more conservative curve (heatup and cooldown) with no overlap versus those curves generated with the circ. flaw ART values and Code Case N-588 Methodology.
Calculation of Adjusted Reference Temperature

4-26 TABLE 4-29 Summary of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T Locations for 32 EFPY Material                                    32 EFPY 1/4T ART                   3/4T ART Intermediate Shell Forging                       12                         0 49D963-1/49C904-1 Lower Shell Forging                              43                        26 50D102-1/50C97-1
         - Using Surveillance Data                        18                         14 Circumferential Weld WF-562                      125                       102
         - Using Surveillance Data                       93(;)                     79(;)
Nozzle Shell Forging 5P-7056                    67(b)                     54(b)
Circumferential Weld WF-645                      51                        25 NOTES:
(a) These ART values were used to generate the Braidwood Unit 2 heatup and cooldown curves in Figures 5-7 and 5-8. See note (b).
(b) These ART values, using the '96 App. G Methodology, produced a more conservative curve (heatup and cooldown) with no overlap versus those curves generated with the circ. flaw ART values and Code Case N-588 Methodology.
Calculation of Adjusted Reference Temperature

5-1 5         HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-NP-A[ 8 ], dated January 1996.
Figures 5-1 through 5-8 present the 12, 16, 22 and 32 EFPY heatup and cooldown curves (without margins for possible instrumentation errors) for a heatup rate of 100 0F/hr and cooldown rates of 0, 25, 50 and I 000 F/hr using the 1996 Appendix G methodology16 ] and Code Case N-588t 5 3,respectively. The heatup and cooldown curves that are presented herein are actually curves generated using the 1996 App.
G methodology with the lower axialflaw ART value. The reason is these curves are more conservative than the curves generated using Code Case N-588 methodology with the higher circ.flaw ART value.
This is true throughout the entire temperature range, including criticality.
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 through 5-8. This is in addition to other criteria which must be met before the reactor is made critical, as discussed in the following paragraphs.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-1, 5-3, 5-5 and 5-7 (for the specific heatup rate being utilized).
The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-640 2 ] (approved in February 1999) as follows:
1.5K im<Kc                                 (11)
: where, Klm is the stress intensity factor covered by membrane (pressure) stress, K1, = 33.2 + 20.734 e(o02 Cr - RTNDT)]
T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 3. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40&deg;F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3 of this report. The minimum temperatures for the inservice hydrostatic leak test for the Braidwood Unit 2 reactor vessel at 12, 16, 22 and 32 EFPY Heatup and Cooldown Pressure-Temperature Limit Curves

5-2 are 1147F, 11 80F, 1221F and 1271F at 2485 psig 1996 App. G Methodology. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 5-1 through 5-8 define all of the above limits for ensuring prevention of nonductile failure for the Braidwood Unit 2 reactor vessel. The data points for the heatup and cooldown pressure-temperature limit curves shown in Figures 5-1 through 5-8 are presented in Tables 5-1 through 5-8, respectively.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-562 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 12 EFPY:                  1/4T, 790 F (N-588) & 54 0 F ('96 App. G) 3/4T, 660 F (N-588) & 450 F ('96 App. G) 2500 2250 2000 1750 in 2   1500 1 1250 IL M
LI 1000 750 500 250 0
0      50     100    150      200     250      300    350   400       450    500 550 Moderator Temperature (Deg. F)
FIGURE 5-1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 1000 F/hr)
Applicable to 12 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

54 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-562 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 12 EFPY:                  1/4T, 790 F (N-588) & 540 F ('96 App. G) 3/4T, 660 F (N-588) & 45 0 F ('96 App. G) 2500 2250 2000 1750 X 1500 e
l  1250 0~
* 1000 co i;
C.
750 500 250 0
0      50     100     150     200    250     300     350   400      450    500 550 Moderator Temperature (Deg. F)
FIGURE 5-2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 1000 F/hr) Applicable to 12 EFPY Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-5 TABLE 5-1 Braidwood Unit 2 Heatup Data at 12 EFPY Using Code Case N-588 vs. 1996 App. G Methodology (Without Margins for Instrumentation Errors) 100 Heatup                Critical. Limit               Leak Test Limit T             P      I       T              P       I      T           P .
60            0             114             0 '           97          2000 60          1014           114            1061            114         2485 65          1061           115            1061 70          1061           120            1062 75          1061           125            1069 80          1062           130            1083 85          1069           135            1103 90          1083           140            1128 95          1103            145           1161 100          1128            150           1199 105          1161            155           1244 110          1199            160           1297 115          1244           165            1357 120          1297           170            1425 125          1357            175           1501 130          1425           180            1587 135          1501           185            1683 140          1587            190           1791 145          1683            195           1910 150          1791           200            2042 155          1910           205            2189 160          2042           210            2352 165         2189 170         2352
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall. Thus, only the '96 App. G Axial Flaw results are presented in this Report.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-6 TABLE 5-2 Braidwood Unit 2 Cooldown Data at 12 EFPY Using Code Case N-588 vs. 1996 App. G Methodology (Without Margins for Instrumentation Errors)
Steady State                  25F                          50F                    lOOF T             P            T            P             T              P         T             P 60            0           60            0             60             0         60            0 60           1028          60           1017           60           1011 65          1073          65           1067 70          1122 75          1177 80          1237 85          1304 90           1377 95           1459 100          1549 105          1648 110          1758 115          1880 120          2014 125          2162 130          2326
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall. . Thus, only the '96 App. G Axial Flaw results are presented in this Report.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-7 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-562 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 16 EFPY:                  1/4T, 83&deg;F (N-588) & 58&deg;F ('96 App. G) 3/4T, 70&deg;F (N-588) & 47&deg;F ('96 App. G) 2500*                    _  _   _   _    _   _   _    _    _   _      _    _      _  _   _    _    _
0    I, operlim version:5.1 Run:9173 aest Limit 2250            --
Accetable 2000                                                                      Operation2 2Unacceptable  l                                          1   E    atios     LJ Operation 1750 VHeatup Rate                            CriticL 1500      10                                           Dg.
0Deg.FH0  FIHr   _
IL 1250 1000 750 500 250 0
0      50      100    150       200     250      300      350        400       450      500      550 Moderator Temperature (Deg. F)
FIGURE 5-3 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100&deg;F/hr)
Applicable to 16 EFPY Using Code Case N-588 vs.I 996 Appendix G with Axial ART (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-8 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-562 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 16 EFPY:                  1/4T, 830 F (N-588) & 580 F ('96 App. G) 3/4T, 700 F (N-588) & 470 F ('96 App. G) 2500 2250 2000 1750 1500 a-Cn E  1250 3 1000 IU 750 500 250 0
0       50     100     150     200      250     300     350   400      450    500 550 Moderator Temperature (Deg. F)
FIGURE 5-4 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100&deg;F/hr) Applicable to 16 EFPY Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-9 TABLE 5-3 Braidwood Unit 2 Heatup Data at 16 EFPY Using Code Case N-588 vs. 1996 App. G Methodology (Without Margins for Instrumentation Errors) 100 Heatup                  Critical. Limit              Leak Test Limit T             P                T                P            T             P 60            0               118               0           101          2000 60           983              118             1029          118          2485 65          1029              118             1042 70          1042              120             1042 75          1042              125             1048 80          1042              130             1061 85          1048              135             1079 90          1061              140             1104 95          1079              145             1134 100          1104              150             1171 105          1134              155             1214 110          1171              160             1264 115          1214              165             1322 120          1264              170             1387 125          1322              175             1460 130          1387              180             1543 135          1460              185             1635 140          1543              190             1738 145          1635              195             1852 150          1738              200             1980 155          1852              205             2121 160          1980              210             2277 165          2121              215             2450 170          2277 175          2450
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall. . Thus, only the '96 App. G Axial Flaw results are presented in this Report.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-10 TABLE 54 Braidwood Unit 2 Cooldown Data at 16 EFPY Using Code Case N-588 vs. 1996 App. G Methodology (Without Margins for Instrumentation Errors)
Steady State                   25F                           50F                    lOOF T            P            T              P             T                P        T            P 60            0           60             0             60               0       60             0 60          995           60           980             60             970 65          1037          65           1026            65            1022 70          1082          70           1077 75          1133 80          1188 85          1250 90          1318 95          1393 100          1476 105          1568 110          1669 115          1781 120          1905 125          2042 130          2194 135          2361
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall. . Thus, only the '96 App. G Axial Flaw results are presented in this Report.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-Il MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-562 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 22 EFPY                              l/4T, 88&deg;F (N-588) & 62&deg;F ('96 App. G) 3/4T, 74&deg;F (N-588) & 50&deg;F ('96 App. G) 2500 -                        IOperlim                                             Version:5.1 Run:15858l 2250 -          1        1                       l       1 2000                                   -                          Acceptable I                               Oertionj Unacceptable l 1 Operation               j 1750                                                               l_      ---              --     l-CL1500-I tlHeatup Rate l rl
                                             /l}-,
I
                                                    '     _   1 I
l I11i I i i
r         ---
150
               ,'n1 D l    /Crtical              00 Deg. F/Hr iiti 2d.                       J       ,        I                          I     j      ,                 ,
31000                                                                                 _
750                                          4 Criticality Limit based on
                       ,                4           Inservice hydrostatic test 500 -                                          temperature (122 F) for the _-
service period up to 22 EFPY 250   -  __      , .                                     ,_____     J4    _     I                    -
0 0         50      100      150       200         250       300    350    400    450       500       550 Moderator Temperature (Deg. F)
FIGURE 5-5 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of I 00&deg;F/hr)
Applicable to 22 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-12 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-562 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 22 EFPY:                      1/4T, 88&deg;F (N-588) & 62&deg;F ('96 App. G) 3/4T, 74&deg;F (N-588) & 50&deg;F ('96 App. G) 2500        Operlim Version:5.1 Run:15858 l 2250-                 I_  _                            _   _  _    - _  _      I    _
Unacceptable    i              i           _      _         i  _    l 2000 200       ~toOperation j  I_               II
{
* Acceptable 1750
* _IJOperation 1500 2
1250 U   1000 750 500 250 0
0          50      100       150   200    250      300   350   400      450   500 550 Moderator Temperature (Deg. F)
FIGURE 5-6 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100&deg;F/hr) Applicable to 22 EFPY Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-13 TABLE 5-5 Braidwood Unit 2 Heatup Data at 22 EFPY Using Code Case N-588 vs. 1996 App. G Methodology (Without Margins for Instrumentation Errors) 100 Heatup                  Critical. Limit              Leak Test Limit T             P                T                P            T             P 60             0               122               0.          105          2000 60            956              122             1000          122          2485 65           1000              122             1013 70           1013              125             1018 75           1013              130             1030 80          1013              135             1046 85          1018              140             1069 90          1030              145             1097 95          1046              150             1131 100          1069              155             1171 105          1097              160             1218 110          1131              165             1272 115          1171              170             1333 120          1218              175             1402 125          1272              180             1479 130          1333              185             1566 135          1402              190             1663 140          1479              195             1770 145          1566              200             1890 150          1663              205             2023 155          1770              210             2170 160          1890              215             2332 165          2023 170          2170 175          2332
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall. . Thus, only the '96 App. G Axial Flaw results are presented in this Report.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-14 TABLE 5-6 Braidwood Unit 2 Cooldown Data at 22 EFPY Using Code Case N-588 vs. 1996 App. G Methodology (Without Margins for Instrumentation Errors)
Steady State                   25F                           50F                    1OOF T             P           T              P             T               P        T             P 60            0           60             0             60               0       60            0 60           965          60           946             60            932        60           921 65          1003          65           989             65            980 70          1045          70           1036            70            1033 75          1092          75           1088 80          1143 85          1200 90          1263 95          1332 100          1409 105          1494 110          1587 115          1691 120          1805 125          1932 130          2071 135          2226 140          2396
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall. . Thus, only the '96 App. G Axial Flaw results are presented in this Report.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-15 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-562 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 32 EFPY                             1/4T, 93&deg;F (N-588) & 67&deg;F ('96 App. G) 3/4T, 79&deg;F (N-588) & 54&deg;F ('96 App. G) 2500              -                                                                                       I 22500  -                                         kT      L                         loperlimVersion:5.1 Run:19017 2250-       :I             !-   I_ _ I      I_                 _ _     _       _  _                         _  _&#xa2; 2000- _Operatio        2000                       l                        1_____Acceptable Unacceptable                                                    I          l Operation                                                              ,
1750-                      I C                  Heatup Ratee                                                 l C. 1500                       Hr           --                   !                __/
I    i  i           1Critical                       Limit aI            /i[                      l00Deg..F/Hr               ,
1250     ---                     /T-_I_                            I                                  _
1000                          --                 - - '-                                  _
750- _          I                                 I                                                      ___!
Criticality Llrmit based on Inservice hydrostatic test 500                         -                      temperature (127 F)for the I service period up to 32 EFPY 250                      ; _                          I;                 I I 0
0        50       100        150 200 250 300 350 400                                  450       500       550 Moderator Temperature (Deg. F)
FIGURE 5-7 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100&deg;F/hr)
Applicable to 32 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors)
Heatup and Cooldoun Pressure-Temperature Limit Curves

5-16 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-562 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:                    1/4T, 930 F (N-588) & 670 F ('96 App. G) 3/4T, 790 F (N-588) & 540 F ('96 App. G) 2500 2250 2000 1750 Q
1500 Q
2 1250 3
U  1000 750 500 250 0
0       50     100       150    200      250     300     350   400      450    500  550 Moderator Temperature (Deg. F)
FIGURE 5-8 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100 0 F/hr) Applicable to 32 EFPY Code Case N-588 vs. 1996 Appendix G with Axial ART (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-17 TABLE 5-7 Braidwood Unit 2 Heatup Data at 32 EFPY Using Code Case N-588 vs. 1996 App. G Methodology (Without Margins for Instrumentation Errors) 100 Heatup                  Critical. Limit              i Leak Test Limit T             P                T                P              T            P 60            0               127               0             110         2000 60           924              127              965            127         2485 65           965              127              979 70           977              127              977 75           977              127              981 80           977              130              990 85           981              135             1005 90           990              140             1025 95          1005              145             1051 100          1025              150             1081 105          1051              155             1118 110          1081              160             1161 115          1118              165             1210 120          1161              170             1266 125          1210              175             1329 130          1266              180             1400 135          1329              185             1480 140          1400              190             1569 145          1480              195             1668 150          1569              200             1778 155          1668              205             1901 160          1778              210             2036 165          1901              215             2186 170          2036              220             2353 175          2186 180          2353
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall. . Thus, only the '96 App. G Axial Flaw results are presented in this Report.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-18 TABLE 5-8 Braidwood Unit 2 Cooldown Data at 32 EFPY Using Code Case N-588 vs. 1996 App. G Methodology (Without Margins for Instrumentation Errors)
Steady State                    25F                           50F                   1OOF T             P            T             P              T               P       T             P 60             0           60            0              60               0     60             0 60           931           60           908             60            889      60            866 65           965           65           946             65            932      65           921 70           1003          70           989             70            980 75           1045          75           1036            75            1033 80           1092          80           1088 85           1143 90           1200 95           1263 100           1332 105           1409 110           1494 115           1587 120           1691 125           1805 130           1932 135           2071 140       2226 145           2396
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall. Thus, only the '96 App. G Axial Flaw results are presented in this Report.
Heatup and Cooldown Pressure-Temperature Limit Curves

6-1 6      REFERENCES 1      Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U.S.
Nuclear Regulatory Commission, May, 1988.
2      ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February 26, 1999.
3      Code of Federal Regulations, 10 CFR Part 50, Appendix G."Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
4      WCAP-16143-P, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2", K.R. Hsu, November 2003.
5      ASME Code Case N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels", Section XI, Division 1,Approved December 12, 1997.
6      ASME Boiler and Pressure Vessel Code, Section XI, "Rule for Inservice Inspection of Nuclear Power Plant Components", Appendix G."Fracture Toughness Criteria for Protection Against Failure", 1996.
7      WCAP-15368, "Commonwealth Edison Company Braidwood Unit 2 Surveillance Program Credibility Evaluation", T.J. Laubham, March 2000.
8      WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D. Andrachek, et al., January 1996.
9       1989 Section III, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Material for Vessels".
10    WCAP-15369, "Analysis of Capsule W from the Commonwealth Edison Co. Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program", T. J. Laubham, et al., March 2000.
11    WCAP-15316, Revision 1, "Analysis of Capsule W from the Commonwealth Edison Co.
Braidwood Unit I Reactor Vessel Radiation Surveillance Program", E.Terek, et.al., December 1999.
12    WCAP-14824, Revision 2, "Byron Unit I Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration For Byron and Braidwood", T. J. Laubham, et al., November 1997. Ref. Errata letter CAE-97-233, CCE-97-316, "Transmittal of Updated Tables to WCAP-14824 Rev. 2" 13    Nuclear Design Information Transmittal NDIT No. BRW-DIT-97-321, Rev. 0, "Braidwood Station Historical Tcold Data for Units I & 2", Dated October 22, 1997.
References

6-2 14     Nuclear Design Information Transmittal NDIT No. BRW-DIT-2000-0010, Rev. 0, "Braidwood Station Historical Tcold Data for Units I & 2 Cycle 7", Dated January 20, 2000.
15     CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed byATI Consulting, March 1996.
16     WCAP-1 11 88, "Commonwealth Edison Company Braidwood Station Unit No. 2 Reactor Vessel Radiation Surveillance Program", L.R. Singer, December 1986.
17     WCAP-12845, "Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program", E. Terek, et al., March 1991.
18     Analytical Request #15482, "Braidwood Nuclear Plant, Unit 2, Irradiated Low Alloy Steel Reactor Surveillance Dosimetry", L. Kardos, dated October 27, 1994.
References

A-I APPENDIX A Thermal Stress Intensity Factors (K1 t)
Thermal Stress Intensity Factors

A-2 TABLE Al KIt Values for 100 0 F/hr Heatup Curve (12 EFPY)
Vessel Temperature - 1/4T Thermial . Vessel Temperature                 3/4T Thermal Water         c   1/4T Location for            Stress          3/4T Location for          Stress Temp.           100WF/hr Heatup          Intensity Factor       100WF/hr Heatup      Intensity Factor (OF)                  (OF)            (KSI SQ. RT. IN.)              (OF)         (KSI SQ. RT. IN.)
60                  56.01                  -0.9945                 55.05               0.4767 65                  58.62                  -2.4413                 55.31                1.4422 PT Curves are Limited by the 1/4 T Location up to 650 F and 34 T Limited for the Remainder of the Curve 70                  61.70                  -3.6836                 56.01               2.4209 75                  65.01                  -4.8600                 57.20               3.3365 80                  68.58                  -5.8718                 58.80               4.1525 85                  72.27                  -6.7937                 60.80               4.8806 90                  76.14                  -7.5912                 63.14               5.5230 95                   80.11                 -8.3181                 65.77               6.0965 100                   84.21                 -8.9517                 68.68               6.6042 105                   88.40                 -9.5267                 71.81               7.0564 110                  92.68                 -10.0310                 75.15               7.4589 115                  97.04                 -10.4906                 78.67               7.8204 120                  101.46                -10.8960                 82.35               8.1440 125                  105.95                -11.2667                 86.16               8.4352 130                  110.48                -11.5953                 90.10               8.6969 135                  115.06                -11.8972                 94.14               8.9336 140                  119.68                -12.1661                 98.28               9.1474 145                  124.34                -12.4147                102.49               9.3417 150                  129.03                -12.6374                106.78               9.5183 155                  133.75                -12.8446                111.14               9.6799 160                  138.48                -13.0315                115.55               9.8277 165                  143.25                -13.2066                120.01               9.9638 170                  148.02                -13.3657                124.52               10.0892 Vessel Radius to the    Y4T and 3%T Locations   are as follows:
* 1/4T Radius = 88.750"
* 3/4T Radius = 93.000" Thermal Stress Intensity Factors

A-3 TABLE A2 K1t Values for 1000 F/hr Cooldown Curve (12 EFPY)
Vessel Temperature 100&deg;F/hr Cooldow'n Water          @ 1/4T Location for 1/4T Thernal Stress Temp.           100'F/hr Cooldown        Intensity Factor (OF)(         F)       0              (KSI SQ. RT. IN.)
130                 154.87                 15.4187 125                 149.78                 15.3518 120                  144.70                 15.2854 115                 139.61                 15.2187 110                 134.53                 15.1525 105                 129.45                 15.0860 100                 124.36                 15.0200 95                  119.28                 14.9538 90                  114.19                 14.8881 85                  109.11                 14.8221 80                  104.02                 14.7566 75                  98.94                  14.6909 70                  93.86                  14.6257 65                   88.77                 14.5603 60                   83.69                 14.4946 NOTE:       The I00F/hr Cooldown is limited by the lower rates at T = 60 to 701F, and by the Steady State condition thereafter.
Thermal Stress Intensity Factors

A-4 TABLE A3 Kit Values for 1007F/hr Heatup Curve (16 EFPY)
Vessel Temperature          1/4T Thermal      Vessel Temperature        3/4T Thermal Water         @ 1/4T Location for              Stress       @ 3/4T Location for            Stress Temp.               0 100 F/hr Heatup         Intensity Factor      100WF/hr Heatup        Intensity Factor (OF)                  (0 F)          (KSI SQ. RT. IN.)             (OF)           (KSI SQ. RT. IN.)
60                  56.01                  -0.9945               55.05                  0.4767 65                  58.62                  -2.4413               55.31                  1.4422 PT Curves are Limited by the 1/4 T Location up to 650 F and 3/4 T Limited for the Remainder of the Curve 70                  61.70                  -3.6836               56.01                 2.4209 75                  65.01                  -4.8600               57.20                 3.3365 80                  68.58                  -5.8718               58.80                 4.1525 85                  72.27                  -6.7937               60.80                 4.8806 90                  76.14                  -7.5912               63.14                 5.5230 95                  80.11                  -8.3181               65.77                  6.0965 100                  84.21                  -8.9517               68.68                 6.6042 105                  88.40                  -9.5267               71.81                  7.0564 110                  92.68                 -10.0310               75.15                  7.4589 115                  97.04                 -10.4906               78.67                  7.8204 120                 101.46                 -10.8960               82.35                  8.1440 125                 105.95                 -11.2667               86.16                  8.4352 130                 110.48                 -11.5953               90.10                  8.6969 135                 115.06                 -11.8972               94.14                  8.9336 140                 119.68                 -12.1661               98.28                 9.1474 145                 124.34                 -12.4147              102.49                 9.3417 150                 129.03                 -12.6374              106.78                 9.5183 155                 133.75                 -12.8446              111.14                 9.6799 160                 138.48                 -13.0315               115.55                9.8277 165                 143.25                 -13.2066              120.01                 9.9638 170                 148.02                 -13.3657              124.52                 10.0892 175                 152.82                 -13.5158              129.06                 10.2056 Thermal Stress Intensity Factors

A-S TABLEA4 Kit Values for  I 000 F/hr Cooldown Curve (16 EFPY)
Vessel Temperature 1001F/hr Cooldown Water         @ 1/4T Location for 1/4T Thermal Stress Temp.          100&deg;F/hr Cooldown        Intensity Factor (OF)(                     )          (KSI SQ. RT. IN.)
135                 159.95                15.4852 130                 154.87                15.4187 125                 149.78                15.3518 120                 144.70                15.2854 115                 139.61                15.2187 110                 134.53                15.1525 105                 129.45                15.0860 100                 124.36                15.0200 95                 119.28                14.9538 90                 114.19                14.8881 85                 109.11                14.8221 80                 104.02                14.7566 75                  98.94                14.6909 70                  93.86                14.6257 65                  88.77                14.5603 60                  83.69                14.4946 NOTE:      The I 000 F/hr Cooldown is limited by the lower rates at T = 60 to 701F, and by the Steady State condition thereafter.
Thermal Stress Intensity Factors

A-6 TABLE A5 K1t Values for I 00&deg;F/hr Heatup Curve (22 EFPY)
Vessel Temperature        --1/4T Thermal       Vessel Temperature      3/4T Thermal Water         @ 14T Location for                Stress -     @ 3/4T Location for          Stress Temp.          100'F/hr Heatup           Intensity Factor'. 100'F/hr Heatup         Intensity Factor (0 F)(KSI)                                                 (KSI SQ. RT. IN.)
(OF)             ,S                     SQ R.     .M          (OF) ,
60                   56.01                   -0.9945               55.05               0.4767 65                   58.62                   -2.4413              55.31                 1.4422 PT Curves are Limited by the %14T Location up to 650 F and 3/4 T Limited for the Remainder of the Curve 70                   61.70                   -3.6836               56.01               2.4209 75                   65.01                   -4.8600               57.20               3.3365 80                   68.58                   -5.8718               58.80               4.1525 85                   72.27                   -6.7937               60.80               4.8806 90                   76.14                   -7.5912               63.14               5.5230 95                   80.11                   -8.3181               65.77               6.0965 100                  84.21                   -8.9517               68.68               6.6042 105                  88.40                   -9.5267               71.81               7.0564 110                  92.68                  -10.0310               75.15               7.4589 115                  97.04                  -10.4906               78.67               7.8204 120                 101.46                  -10.8960               82.35               8.1440 125                 105.95                  -11.2667               86.16               8.4352 130                 110.48                  -11.5953              90.10                8.6969 135                 115.06                  -11.8972               94.14               8.9336 140                 119.68                  -12.1661               98.28               9.1474 145                 124.34                  -12.4147              102.49               9.3417 150                 129.03                  -12.6374              106.78               9.5183 155                 133.75                  -12.8446              111.14               9.6799 160                 138.48                  -13.0315              115.55               9.8277 165                 143.25                  -13.2066              120.01               9.9638 170                 148.02                  -13.3657              124.52               10.0892 175                 152.82                  -13.5158              129.06               10.2056 Thermal Stress Intensity Factors

A-7 TABLE A6 Kit Values for 100&deg;F/hr Cooldown Curve (22 EFPY)
Vessel Temperature 100&deg;F/hr Cooldown Water        @ 1/4T Location for 1/4T Thermal Stress Temp.         100&deg;F/hr Cooldown         Intensity Factor (0F)                 (OF)             ((KSI SQ. RT. IN.)
140                165.04                  15.5523 135                159.95                  15.4852 130                154.87                  15.4187 125                149.78                  15.3518 120                144.70                  15.2854 115                139.61                  15.2187 110                134.53                  15.1525 105                129.45                  15.0860 100                124.36                  15.0200 95                119.28                  14.9538 90                114.19                  14.8881 85                109.11                  14.8221 80                104.02                  14.7566 75                98.94                   14.6909 70                93.86                   14.6257 65                88.77                   14.5603 60                83.69                   14.4946 NOTE:      The 100IF/hr Cooldown is limited by the lower rates at T = 65 to 750F, and by the Steady State condition thereafter.
Thermal Stress Intensity Factors

A-8 TABLE A7 KIt Values for I 00 0 F/hr Heatup Curve (32 EFPY)
Vessel Temperature         1/4T Thermal          Vessel Temperature       3/4T Thermal Water         @ 1/4T Location for              Stress         @ 3/4T Location for            Stress Temp.           100'F/hr Heatup         Intensity Factor        100&deg;F/hr Heatup        Intensity Factor (OF)                   (OF)           (KSI SQ. RT. IN.)                (OF)          (KSI SQ. RT. IN.)
60                   56.01                  -0.9945                 55.05                 0.4767 65                   58.62                  -2.4413                 55.31                 1.4422 PT Curves are Limited by the 'M4   T Location up to 650 F and 3/4 T Limited for the Remainder of the Curve 70                   61.70                  -3.6836                 56.01                 2.4209 75                   65.01                  -4.8600                 57.20                 3.3365 80                   68.58                  -5.8718                 58.80                 4.1525 85                   72.27                  -6.7937                 60.80                 4.8806 90                   76.14                  -7.5912                 63.14                 5.5230 95                   80.11                  -8.3181                  65.77                6.0965 100                  84.21                  -8.9517                 68.68                 6.6042 105                  88.40                  -9.5267                 71.81                 7.0564 110                  92.68                 -10.0310                 75.15                 7.4589 115                  97.04                 -10.4906                 78.67                 7.8204 120                 101.46                 -10.8960                 82.35                 8.1440 125                 105.95                 -11.2667                 86.16                 8.4352 130                 110.48                 -11.5953                 90.10                 8.6969 135                 115.06                 -11.8972                 94.14                 8.9336 140                 119.68                 -12.1661                 98.28                 9.1474 145                 124.34                 -12.4147                 102.49                9.3417 150                 129.03                 -12.6374                 106.78                9.5183 155                 133.75                 -12.8446                 111.14                9.6799 160                 138.48                 -13.0315                 115.55                9.8277 165                 143.25                 -13.2066                 120.01                9.9638 170                 148.02                 -13.3657                 124.52                10.0892 175                 152.82                 -13.5158                 129.06                10.2056 180                 157.63                 -13.6533                 133.64                10.3136 Thermal Stress Intensity Factors

ATTACHMENT 4-C WCAP-15391, Revision 1 Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation

Westinghouse Non-Proprietary Class 3 WCAP-1 5391                                                      November 2003 Revision 1 Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation
                @)Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15391, Revision 1 Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubham November 2003 Approve  /
J. A. Gresham, Manager Reactor Component Design & Analysis Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02003 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST OF TABLES ............................................                                                      iv LIST OF FIGURES ............................................                                                    vi PREFACE ..............................................                                                          vii EXECUTIVE 

==SUMMARY==
 ............................................                                               viii I     INTRODUCTION.                                                                                           1-1 2      PURPOSE .2-3      CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .3-1 4      CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .4-5      HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .5-1 6      REFERENCES .6-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS                            . ......................................
A-I Revision 0

iv LIST OF TABLES Table 4-1  Summary of the Peak Pressure Vessel Neutron Fluence Values at 22 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV) ............................................... 4-4 Table 4-2  Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV) ......................................          4-4 Table 4-3  Calculated Integrated Neutron Exposure of the Byron Unit I Surveillance Capsules Tested to Date ...............                                                                 4-5 Table 4-4  Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program .........................                                         4-5 Table 4-5  Calculation of Best Estimate Cu and Ni Weight Percent for the Byron Unit I Forging Materials ..........                                                                            4-6 Table 4-6  Calculation of the Average Cu and Ni Weight Percent for the Byron Unit I Surveillance Weld Material Only (Heat # 442002) .................................                       4-6 Table 4-7  Calculation of Best Estimate Cu and Ni Weight Percent Values for the Byron Units I & 2 Weld Material (Using Byron I & 2 Chemistry Test Results) .4-7 Table 4-8  Reactor Vessel Beltline Material Unirradiated Toughness Properties .4-8 Table 4-9  Calculation of Chemistry Factors for Byron Unit I using Surveillance Capsule Data .......                                                                                    4-9 Table 4-10 Calculation of Chemistry Factors Using Braidwood Units I & 2 Surveillance Capsule Data .......                                                                                   4-10 Table 4-l1 Summary of the Byron Unit I Reactor Vessel Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 ...4-11 Table 4-12 Calculation of the I/4T and 3/4T Fluence Factor Values used for the Generation of the 22 EFPY Heatup / Cooldown Curves ............................                                   4-12 Table 4-13 Calculation of the 1/4T and 3/4T Fluence Factor Values used for the Generation of the 32 EFPY Heatup / Cooldown Curves .4-12 Table 4-14 Calculation of the ART Values for the 1/4T Location @ 22 EFPY .4-13 Table 4-15 Calculation of the ART Values for the 3/4T Location            @ 22 EFPY .4-14 Table 4-16 Calculation of the ART Values for the I/4T Location @ 32 EFPY .4-15 Revision 0

v Table 4-17 Calculation of the ART Values for the 3/4T Location @ 32 EFPY ............................... 4-16 Table 4-18 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location for 22 EFPY .                                                                                 4-17 Table 4-19 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location for 32 EFPY .                                                                                 4-18 Table 5-1  Byron Unit I Heatup Data at 22 EFPY Using 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) .5-5 Table 5-2  Byron Unit I Cooldown Data at 22 EFPY Using 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) .5-6 Table 5-3  Byron Unit I Heatup Data at 32 EFPY Using 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) .5-9 Table 5-4  Byron Unit I Cooldown Data at 32 EFPY Using Appendix G Methodology (Without Margins for Instrumentation Errors) .                                                5-10 Revision 0

vi LIST OF FIGURES Figure 5-1 Byron Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of I 000 F/hr) Applicable to 22 EFPY Using 1996 Appendix G (Without Margins for Instrumentation Errors) ..............................                                   5-3 Figure 5-2 Byron Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100&deg;F/hr) Applicable to 22 EFPY Using 1996 Appendix G (Without Margins for Instrumentation Errors).................................................................5-4 Figure 5-3 Byron Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 100&deg;F/hr) Applicable to 32 EFPY Using 1996 Appendix G (Without Margins for Instrumentation Errors) ..............................                                   5-7 Figure 5-4 Byron Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I00 0 F/hr) Applicable to 32 EFPY Using 1996 Appendix G (Without Margins for Instrumentation Errors) ..............................                                   5-8 Revision 0

vii PREFACE This report has been technically reviewed and verified by:
Reviewer:                                                             J. H. Ledger At'Z4 Revision 1:
This revision was completed to incorporate an updated reference for the flange elimination from the PT limit curves. In the previous revision, WCAP-15315 was provided as justification for the flange notch elimination and it was changed to WCAP-16143-P in this revision. In addition to this change, the thermal stress intensity factors for the highest heatup and cooldown rates were added to this report in Appendix A.
Revision 0

yiii EXECUTIVE 

==SUMMARY==

The purpose of this report is to generate new pressure-temperature limit curves for Byron Unit I for normal operation at 22 and 32 EFPY based of revised Uprated Fluences. The new pressure-temperature limit curves were generated using the methodology from WCAP 14040-NP-A, the 1996 ASME Boiler and Pressure Vessel Code, Section XI Appendix 4 ASME Code Case N-588, ASME Code Case N-640 and WCAP-16143-P ("Flange Notch" Elimination Justification). Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART). The 1/4T and 3/4T ART values are summarized in Tables 4-18 and 4-19 and the limiting material is the intermediate shell forging 5P-5933.
The pressure-temperature limit curves were generated for a heatup rate of 1000 F/hr and cooldown rates of 0, 25, 50 and I 000 F/hr. The axial oriented flaw cases are limiting for all curves at each EFPY value evaluated. Hence, Code Case N-588 was not used and only the axial oriented flaw curves are presented in this report and they can be found in Figures 5-1 through 5-4.
Revision 0

1-1 1         INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.
RTNDT   increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT(IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials"'1 ].
Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves for normal operation.
NOTE: For the reactor vessel radiation surveillance program, Babcock and Wilcox Co. supplied Westinghouse with sections of SA508 Class 3 forging material used in the core region of the Byron Station Unit No. I reactor pressure vessel (Specifically from forging 5P-5933). Also supplied was a weldment made with weld wire heat # 442002 Linde 80 flux lot number 8873, which is identical to that used in the actual fabrication of the intermediate to lower shell girth weld of the pressure vessel).
Intrducton                                                                          evison_
Introduction                                                                                       Revision 0

2-1 2         PURPOSE The Commonwealth Edison Company (now known as Exelon Nuclear) contracted Westinghouse to generate new pressure-temperature limit curves for 22 and 32 EFPY based on the revised fluences from the 5% Uprating. These new Pressure-Temperature Curves are to be developed utilizing the following methodologies:
* Regulatory Guide 1.99, Revision 2['1,
* ASME Code Case N-640 2 ],
* Elimination of the flange requirement of Appendix G to IOCFR Part 5O3 per WCAP-1 6143-P, "Reactor Vessel Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2"[41
* Methodology of the 1996 ASME B&P Vessel Code, Section XI, Appendix GJ161 and
* The PT Curves will be developed WITHOUT margins or instrumentation errors.
Based on the above methodologies, PT Curves will be generated consisting of the 1996 Appendix G to ASME Section XI and the Kl, methodologies (ASME Code Case N-640) for the limiting forging/base metal material. All curves will use the methodology to eliminate the 10 CFR Part 50 Appendix G flange requirements (from WCAP-16143-P). The final PT curves to be presented herein will be the most limiting set of curves.
The purpose of this report is to present the calculations and the development of the Byron Unit I heatup and cooldown curves for 22 and 32 EFPY. This report documents the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2[1], for all the beltline materials and the development of the heatup and cooldown pressure-temperature limit curves for normal operation.
Per the request of the Exelon, the surveillance weld data from the Byron Unit I and Unit 2 surveillance programs has been integrated. Note that Byron Unit 2 surveillance weld is identical to the surveillance weld (Heat No. 442002) at Byron Unit 1. In addition, the Braidwood Units I and 2 surveillance weld is identical to the nozzle to intermediate shell circumferential weld (Heat No. 442011). Per WCAP-151831', WCAP-151801"II and WCAP-153681 171 all weld metal surveillance data has been determined to be credible.
Revision 0 Purpose                                                                                          Revision 0

3-1 3          CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1       Overall Approach Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements"[31 specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements. Section XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components", Appendix G16 1,contains the conservative methods of analysis.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1c, for the metal temperature at that time. Kic is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of PT Limit Curves for Section Xl"12 61 of the ASME Appendix G to Section XI. The Kic curve is given by the following equation:
Kic =33.2+20.734*e e0. 02(T-RTNDT)]                           (1) where,       K    =    reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K1, curve is based on the lower bound of static critical K, values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, SA-508-3 steel.
3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C* Kim + Kit < K,.                                    (2)
: where, Kim    =     stress intensity factor caused by membrane (pressure) stress Kit    =     stress intensity factor caused by the thermal gradients K.=          function of temperature relative to the RTNDT of the material C      =     2.0 for Level A and Level B service limits C      =     1.5 for hydrostatic and leak test conditions during which the reactor core is not critical Revision 0 Pressure-Temperature Relationships Criteria For Allowable Prcssure-Tempcrature  Relationships                                          Revision 0

3-2 For membrane tension, the corresponding K, for the postulated defect is:
Ki   = Mmx(pRi /t)                                                     (3) where, Mm for an inside surface flaw is given by:
Mm      =     1.85 for 4 < 2, Mm      =     0.92647 for 2*4<t3A64, Mm      =     3.21 for 4t > 3.464 Similarly, Mm for an outside surface flaw is given by:
Mm      =     1.77 for  4   < 2, Mm            0.89347 for 2<4t<3.464, Mm            3.09 for 47 > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.
For bending stress, the corresponding K, for the postulated defect is:
Klb = Mb
* Maximum Stress, where Mb is two-thirds of Mm The maximum K, produced by radial thermal gradient for the postulated inside surface defect of G-2120 is Kit = 0.953xl 0,3 x CR x t25, where CR is the cooldown rate in IF/hr., or for a postulated outside surface defect, KI, = 0.753x1 0, 3 x HU x t25, where HU is the heatup rate in IF/hr.
The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.
G-2214-2 for the maximum thermal K,.
(a)       The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).
(b)       Alternatively, the K, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a l/4-thickness inside surface defect using the relationship:
Dt = (1.0359Co + 0.6322CI + 0.4753C2 + 0.3855C3) * -                                  (4)
Revision 0 Criteria For            Pressure-Temperature Relationships Allowable Pressure-Temperature For Allowable                        Relationships                                          Revision 0

3-3 or similarly, KIT during  heatup for a l-thickness outside surface defect using the relationship:
Ki, = (1.043Co + 0.630Ci + 0.481C2 + 0.401C3)*                                         (5) where the coefficients C0, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
a(x) = Co + Ci(x / a) + C2(x /    a) 2
                                                            + C3(x / a) 3                                (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.
Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"E8 1Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.
At any time during the heatup or cooldown transient, K1 , is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIt, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of KI, at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KI, exceeds KI,, the calculated allowable pressure during cooldown will be greater than the steady-state value.
Relationships                                            Revision 0 Criteria For Allowable Pressure-Temperature For Allok2ble  Pressure-Temperature Relationships                                             Revision 0

3-4 The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1c for the 1/4T crack during heatup is lower than the Kic for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1 , values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the I14T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a I/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
Criteria For Allowable Pressure-Temperature Relationships                                            Revision 0

3-5 3.3 Closure Head/Vessel Flange Requirements 10 CFR Part 50, Appendix G13 J addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for Byron Unit I reactor vessel.
This requirement was originally based on concerns about the fracture margin in the closure flange region.
During the boltup process, stresses in this region typically reach over 70 percent of the steady-state stress, without being at steady-state temperature. The margin of 1200 F and pressure limitation of 20 percent of the hydro-test pressure were developed using the KIA fracture toughness from the mid 1970's.
Improved knowledge of the fracture toughness and other issues which affect the integrity of the reactor vessel have led to the recent change to allow the use of Kic in development of pressure-temperature curves, as contained in Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1,,[2I.
The discussion given in WCAP-I 61 43-P, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2,"l4) concluded that the integrity of the closure head/vessel flange region is not a concern for Byron or Braidwood, therefore the requirement of 10 CFR Part 50, Appendix 4, can be eliminated from the Pressure-Temperature Curves contained in this report.
Criteria For Allowable Pressure-Temperature Relationships                                          Revision 0

4-1 4        CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART = InitialRTNDT + A RTNDT + Margin                                  (7)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-233 1 of Section III of the ASME Boiler and Pressure Vessel Codel9l. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTNDT is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows:
A RTNDT = CF
* f(.28-01OlO1gf                               (8)
To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
                                            =      ,*                                        ((-0.2 (9) where x inches (vessel inner radius and beltline thickness is 86.625 inches and 8.5 inches, respectively) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNDT at the specific depth.
The Westinghouse Radiation Engineering and Analysis group evaluated the vessel fluence projections for the 5% Uprating and the results are presented in SAE-REA-00-546 11 "). The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with the methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"[ 8 3. Table 4-1, herein, contains the calculated vessel surface fluence values along with the Regulatory Guide 1.99, Revision 2, 1/4T and 3/4T calculated fluences used to calculate the ART values for all beltline materials in the Byron Unit I reactor vessel. Additionally, the calculated surveillance capsule fluence values are presented in Table 4-1 and 4-2.
Ratio Procedure and Temperature Adiustment:
The ratio procedure, as documented in Regulatory Guide 1.99, Revision 2 Position 2.1, was used, where applicable, to adjust the measured values of ARTNDT of the weld materials for differences in copper/nickel content. This adjustment is performed by multiplying the ARTNDT by the ratio of the vessel chemistry factor to the surveillance material chemistry factor. The adjusted ARTNDT values are then used to calculate the chemistry factor for the vessel materials.
Revision 0 Calculation of          Reference Temperature Adjusted Reference of Adjusted            Temperature                                                      Revision 0

4-2 From NRC Industry Meetings on November 12, 1997 and February 12, 13 of 1998, procedural guidelines were presented to adjust the ARTNDT for temperature differences when using surveillance data from one vessel applied to another vessel. The following guidance was presented at these industry meetings:
Irradiation temperature and fluence (or fluence factor) are first order environmental variables in assessing irradiation damage... To account for differences in temperature between surveillance specimens and vessel, an adjustment to the data must be performed. Studies have shown that for temperatures near 550'F, a I F decrease in irradiation temperature will result in approximately a 1IF increase in ARTNDT.
For capsules with irradiation temperature of Tcapsute and a plant with an irradiation temperature of Tplant, an adjustment to normalize ARTNDT. measured to Tplan, is made as follows:
Temp. Adjusted ARTNDT = ARTNDT, measured + l .0*(     Teapsuie - Tpiant)             (1 0)
The irradiation temperatures from Byron Units I & 2 are presented in WCAP-14824, Revision 2121 . The average irradiation temperature from each of the four Units and operating cycles in question is 5530 F.
Therefore, no temperature adjustment is required.
Chemistry Factor:
The chemistry factor is obtained from the tables in Regulatory Guide 1.99, Revision 2 using the best estimate average copper and nickel content as reported in Tables 4-5 through 4-8. The chemistry factors were also calculated using Position 2.1 from the Regulatory Guide 1.99, Revision 2 using all available surveillance data. Per Reference 7, the surveillance weld data for Byron Unit I is credible while the surveillance forging material is non-credible. In addition, Reference 7 also shows that the Table chemistry factor is conservative. However, because the chemistry factor predicted by the Regulatory Guide 1.99 Position 2.1 for the forging surveillance data was greater than the Position 1.1 chemistry factor, then the Position 2.1 chemistry factor will be used to determine the adjusted reference temperature of the forging material with a full margin term. Per Reference 11, the surveillance weld data for Byron Unit 2 is credible. Position 2.1 chemistry factors are calculated in Table 4-9 and 4-10.
Explanation of Margin Term:
When there are "two or more credible surveillance data sets'"t11 available for Byron Unit 1, Regulatory Guide 1.99 Rev. 2 (RG1.99R2) Position 2.1 states "To calculate the Margin in this case, use Equation 4; the values given there for a, may be cut in half'. Equation 4 from RGI.99R2 is as follows:
M =2 G +        2a .
Calculation of Adjusted Reference Temperature                                                     Revision 0

4-3 Standard Deviation for Initial RTNDT Margin Term, c, If the initial RTNDT values are measured values, which they are in the case of Byron Unit 1, then cy, is equal to 00F. On the other hand, if the initial RTNDT values were not measured, then a generic value of 171F (base metal and weld metal) would have been required to be used for a, .
Standard Deviation for ARTNDT Margin Term, a, Per RGI.99R2 Position 1.1, the values of cA are referred to as "281F for welds and 17'F for base metal, except that a, need not exceed 0.50 times the mean value of ARTNDT-" The mean value of ARTNDT is defined in RGI.99R2 by Equation 2 and defined herein by Equation 8.
Per RG I.99R2 Position 2.1, when there is credible surveillance data, c&is taken to be the lesser of 1/2 ARTNDT or 14'F (280 F/2) for welds, or 8.50 F (I 70F/2) for base metal. Where ARTNDT again is defined herein by Equation 8.
Summary of the Margin Term Since a, is taken to be zero when a heat-specific measured value of initial RTNDT are available (as they are in this case), the total margin term, based on Equation 4 of RG1.99R2, will be as follows:
* Position 1.1:      Lesser of ARTNDT or 56 0 F for Welds Lesser of ARTNDT or 34 0 F for Base Metal
* Position 2.1:      Lesser of ARTNDT or 28'F for Welds Lesser of ARTNDT or 17&deg;F for Base Metal Revision 0 Calculation of of Adjusted Reference Temperature Adjusted Reference Temperature                                                  Revision 0

4-4 TABLE 4-1 Summary of the Peak Pressure Vessel Neutron Fluence Values at 22 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV)
Material(')                          Surface(b)                  %T                  3/4 T (n/cm 2, E>1.O MeV)    (n/cm2 , E>1.O  MeV) (n/cm2, E>1.O NeW)
Intermediate Shell Forging 5P-5933                 1.39 x 10'9            8.35 x 1018          3.01 x 108       l Lower Shell Forging 5P-5951                     1.39 x 10t9            8.35 x 10"8          3.01 x 1018 Nozzle Shell Forging 123J218                    4.15 x 1018            2.49 x 1018          8.99 x 10i 7 Intermediate to Lower Shell Forging Circ.              1.34 x 1O' 9           8.05 x 10"           2.90 x 108 Weld Seam WF-336 (Heat 442002)
Nozzle Shell to Inter. Shell Forging Circ.            4.15 x 1018            2.49 x 10"           8.99 x I0O" Weld Seam WF-501 (Heat 442011)
Notes:
(a) All remaining vessel materials are below I x 1017 n/cm 2, E > 1.0 MeV.
(b) Calculated Surface fluence documented in Reference 10.
TABLE 4-2 Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV)
Material(')                           Surface(b)                 1/4 T               M4 T (n/cm 2 , E>1.O MeV)   (n/cm 2, E>1.O MWV)  (n/cm2, E>l.O MeV)
Intermediate Shell Forging                   2.02 x 1019            1.21 x 10O9          4.37 x 10" 5P-5933 Lower Shell Forging 5P-5951                     2.02 x 1019            1.21 x 1019          4.37 x 1018 Nozzle Shell Forging 123J218                    6.04 x 1018            3.63 x 1018          1.31 x 10" Intermediate to Lower Shell Forging Circ.             1.94 x 1019            1.16 x    1O19       4.20 x 1018 Weld Seam WF336 (Heat 442002)
Nozzle Shell to Intermediate Shell Forging             6.04 x 1018            3.63 x 10'8          1.31 x 1018 Circ. Weld Seam WF501 (Heat 442011)
Notes:
(a)     All remaining vessel materials are below I x 1017 n/cm2 , E > 1.0 MeV.
(b) Calculated Surface fluence documented in Reference 10.
Calculation of Adjusted Reference Temperature                                                            Revision 0

4-5 TABLE 4-3 Calculated Integrated Neutron Exposure of the Byron Unit I Surveillance Capsules Tested to Date Capsule                          Fluence(a)
U                4.04 x IO 8 n/cm 2, (E > 1.0 MeV)
X                1.57X 1019 n/cm 2 , (E > 1.0 MeV) 2.43 x I 0' n/cm 2, (E > 1.0 MeV)
Notes:
(a)    Documented in WCAP-151231 18 ].
Contained in Table 44 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline materials. These measured shift values were obtained using CVGRAPH, Version 4. 116], which is a symmetric hyperbolic tangent curve-fitting program.
TABLE 4-4 Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program Material                             Capsule                 Measured 30 ft-lb Transition Temperature Shift(')
Intermediate Shell Forging 5P-5933                     U                            28.55 0 F X                             9.820 F (Tangential Orientation)                                                      49.20 F Intermediate Shell Forging 5P-5933                     U                            18.52 0 F X                            53.030 F (Axial Orientation)                           V                            29.34 0 F Surveillance Program                           U                             5.61 F Weld Metal                                X                            40.11`F W                            51.340 F Heat Affected Zone                            U                            -60.2&deg;F X                            13.45WF W                            15.230 F Notes:
(a) Documented in WCAP-15123118].
Calculation of Adjusted Reference Temperature                                                   Revision 0

4-6 Table 4-5 contains the best estimate weight percent copper and nickel for the Byron Unit I base materials in the beltline region. Table 4-6 contains the best estimate weight percent copper and nickel for the Byron Unit I surveillance weld material, while Table 4-7 presents the overall best estimate average for that heat of weld. Table 4-8 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RTNDT of the beltline materials and vessel flanges. The weight percent values of Cu and Ni given in Table 4-9 were used to generate the calculated chemistry factor (CF) values based on Tables I and 2 of Regulatory Guide 1.99, Revision 2, and presented in Table 4-11. Tables 4-9 and 4-10 provide the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Position 2.1, which are also summarized in Table 4-11.
TABLE 4-5 Calculation of the Best Estimate Cu and Ni Weight Percent for the Byron Unit I Forging Materials Intermediate Shell Forging 5P-5933       Lower Shell Forging 5P-5951 Reference                     Cu%                     Ni %         Cu %           l   Ni %
Ref. 12                     0.364(a)               0.747(a)       0.04              0.64 Charpy AL-31(b)                    0.36         j         0.70          ---               ---
Charpy AL-36(b)                    0.35                   0.76          ---               ---
Best Estimate                      0.04                   0.74         0.04              0.64 Average(c)                                   l Notes:
(a)    This is the average of 5 data points.
(b)    Charpy Specimens from Capsule W of Byron Unit I (Ref. 18).
(c)    The best estimate average was rounded per ASTM E29, using the "Rounding Method". This calculation is also presented in Reference 14.
TABLE 4-6 Calculation of the Average Cu and Ni Weight Percent for the Byron Unit I Surveillance Weld Material Only (Heat # 442002)
Reference                        Weight % Copper           Weight % Nickel Ref. 12                                0.022(a)               0.69(a)
Charpy AW-36(b)                              0.026                  0.72 Charpy AIN-40')                              0.027                  0.72 Surveillance Weld Average(c)          J               0.02                  0.69 Notes:
(a) This is the average of 21 data points.
(b) Charpy Specimens from Capsule W of Byron Unit I (Ref. 18).
(c) The best estimate average was rounded per ASTM E29, using the "Rounding Method". This calculation is also presented in Reference 14.
Calculation of Adjusted Reference Temperature                                                    Revision 0

4-7 TABLE 4-7 Calculation of Best Estimate Cu and Ni Weight Percent Values for the Byron Units I & 2 Weld Material (Using Byron I & 2 Chemistry Test Results)
Chemistry Type                            Reference              Weight % Copper              Weight % Nickel B&W WQ: BAW-2261                                Ref. 12                      0.024                       0.70 B&W WQ: BAW-2261                                Ref. 12                      0.031                       0.46 B&W WQ: BAW-2261                                Ref. 12                       0.03                       0.72 B&I WQ: BAW-2261                                Ref. 12                      0.068                       0.48 B&W WQ: BAW-2261                                Ref. 12                      0.053                       0.62 B&W WQ: BAW-2261                                Ref. 12                      0.059                       0.62 B&W WQ (from NDIT No.                            Ref. 12                      0.029                       0.65 BYR97-346, Rev. 0)
Round Robin Data Ave. on Weld                       Ref. 12                      0.038                      0.658 WF-336 (from NDIT No. BRW-DIT-97-391, Rev. 0)
Byron I Surveillance Data Ave. (a)                  Table 4-6                     0.022                       0.69 Byron 2 Surveillance Ave.(b)                      Ref. 12                      0.023                      0.712 BEST ESTIMATE AVERAGE                                                          0.04(c. d)                 0.63(c' d)
Notes:
(a) The weld material in the Byron Unit I surveillance program was made of the same wire and flux as the reactor vessel inter. to lower shell girth seam weld. (Weld seam WF-336, Wire Heat # 442002, Flux Type Linde 80, Flux Lot # 8873).
(b) The Byron Unit 2 surveillance weld is identical to that used inthe reactor vessel core region girth seam (WF-447). The weold wire is type Linde MnMoNi (Low Cu-P), heat number 442002, with a Linde 80 type flux, lot number 8064.
(c)    The best estimate chemistry values were obtained using the "average of averages" approach. Inaddition the best estimate average was rounded per ASTM E29, using the "Rounding Method". This calc as presented in Reference 14.
(d)    This ave. Cu & Ni excluded two data points (Cu = 0.114, Ni = 0.54 and Cu = 0.148, Ni = 0.60), see Ref. 22.
Calculation of Adjusted Reference Temperature                                                                    Revision 0

4-8 TABLE 4-8(b)
Reactor Vessel Beltline Material Unirradiated Toughness Properties Material Description                         Cu (%)              Ni(%)     Initial RTNDT)
Closure Head Flange 124K358VAI                              ,             0.74        60 Vessel Flange 123J219VAI                           ---               0.73        10 Nozzle Shell Forging    1 2 3 J2 1 8 (b)              0.05              0.72       30 Intermediate Shell Forging 5P-5933                      0.04              0.74       40 Lower Shell Forging 5P-5951                          0.04              0.64        10 Intermediate to Lower Shell Forging Circ. Weld                 0.04              0.63       -30 Seam WF-336 (Heat # 442002)
Nozzle Shell to Intermediate Shell Forging Circ.               0.03              0.67        10 Weld Seam WF-501 (Heat # 442011)
Byron Unit I Surveillance Program                       0.02              0.69      ---
Weld Metal (Heat # 442002)
Byron Unit 2 Surveillance Program                       0.02              0.71      ---
Weld Metal (Heat # 442002)
Braidwood Units I & 2 Surveillance Program                    0.03           0.67, 0.71   ---
Weld Metals (Heat # 442011)
Notes:
(a) The initial RTNDT values for the plates and welds are based on measured data.
(b) Table duplicated from Table 4-7 of Reference 14.
Revision 0 Calculation of          Reference Temperature Adjusted Reference of Adjusted            Temperature                                                Revision 0

4-9 TABLE 4-9 Calculation of Chemistry Factors using Byron Unit 1 Surveillance Capsule Data Material               Capsule             Capsule           FF(b)   ARTNDT(C)       FF*ARTNDT            FF2 esa)
Intermediate Shell               U                0.404           0.748       28.55            21.36          0.560 Forging 5P-5933                  X                1.57           1.124        9.82            11.04          1.263 (Tangential)                 W                2.43            1.239       49.20            60.96          1.535 Intermediate Shell                U               0.404           0.748       18.52            13.85          0.560 Forging 5P-5933                  X                1.57           1.124       53.03            59.61          1.263 (Axial)                    W                2.43            1.239       29.34            36.35          1.535 SUM:        203.17          6.716 CF5 p.5933      (FF
* RTNDT) . X( FF2 ) = (203.17) * (6.716)    30.31F Byron Unit I Surv. Weld              U               0.404           0.749       11.22             8.40          0.561 Material                                                              (5.61)(d)
(Heat # 442002)                 X                1.57           1.125       80.22            90.25          1.266 (40.1 1)(d)
W                2.43            1.239      102.68           127.22          1.535 (51. 3 4)(d)
Byron Unit 2 Surv. Weld              U               0.405           0.749       16.88            12.64          0.561 Material                                                              (8.44)(d)
(Heat # 442002)                                  1.27           1.067       57.76            61.63          1.138 (2 8 .8 8 )(d)
X               2.30            1.225      108.02           132.32          1.500 (54.01 )(d)
SUM:        432.46          6.561 CFsw Weld 442002 = X(FF
* RTNDT)
* X( FF2 ) = (432.46)  * (6.561) = 65.9 0 F Notes:
(a) Calculated Capsule Fluence taken from References 18(or Table 4-3 herein) & Reference 19 for Byron Units I and 2, respectively.
(b)    FF = fluence factor = fO.28-0.1frogf)
(c)    ARTNDT values are the measured 30 ft-lb shift values (Byron Unit I from Ref. 18 or Table 4-4 herein, &
Byron Unit 2 from Ref. 19).
(d) The Byron I & 2 surveillance weld metal ARTNDT values have been adjusted by a ratio factor of 2.00.
No temperature adjustments are required.
Calculation of Adjusted Reference Temperature                                                                 Revision 0

4-10 TABLE 4-10 Calculation of Chemistry Factors using Braidwood Units I & 2 Surveillance Capsule Data Material                Capsule       Capsule           FF(b)      ARTNDT(c)        FF*ARTNDT           FF2 Braidwood Unit I               U           0.387           0.737        17.06(d)           12.57          0.543 Surveillance                 X             1.24          1.060        3 0 .15 (d)        31.96           1.124 Weld Heat 442011, WF-501             W            2.09           1.201        49 .68(d)          59.67           1.442 Braidwood Unit 2               U           0.400           0.746                              0.0(d)      0.557 Surveillance                 X             1.23          1.058         26.3(d)           27.83           1.119 Weld Heat 442011, WF-501             W            2.25           1.220         23.9(d)           29.16           1.488 SUM:         161.19          6.273 CF = X(FF
* RTNDT)     +  X(FF2 ) = (161.19)    . (6.273) = 25.7WF Notes:
(a)    Calculated Capsule Fluence values taken from References 20 and 21 for Braidwood Units I and 2, respectively.
(b) FF = fluence factor = f(.28 *0. log 0 (c)    ARTNDT values are the measured 30 ft-lb shift values values (Braidwood Unit I from Ref. 20 & Braidwood Unit 2 from Ref. 21).
(d) The Braidwood I & 2 surveillance weld metal ARTNDT values do not require a ratio factor or temperature adjustment.
Revision 0 Calculation of of Adjusted Reference Temperature Adjusted Reference Temperature                                                             Revision 0

4-11 TABLE 4-11 Summary of the Byron Unit I Reactor Vessel Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Material                             Chemistry Factor Position 1.1         Position 2.1 Nozzle Shell Forging 123J218                31.0 0 F               ---
Intermediate Shell Forging 5P-5933             26.00 F              30.30F Lower Shell Forging 5P-5951                 26.07F Intermediate to Lower Shell Forging Circ.         54.0F                65.9 0F Weld Seam WF-336 (Heat 442002)
Nozzle Shell to Intermediate Shell            41.0LF               25.7 0F Forging Circ. Weld Seam WF-501 (Heat 442011)
Byron Unit I & 2 Surveillance Weld             27.0F Metal Braidwood Unit I & 2 Surveillance              41.00 F Weld Metal Revision 0 Calculation of          Reference Temperature Adjusted Reference of Adjusted           Temperature                                                Revision 0

4-12 Contained in Tables 4-12 and 4-13 is the summary of the fluence factors (FF) used in the calculation of adjusted reference temperatures for the Byron Unit I reactor vessel beltline materials for 22 and 32 EFPY.
TABLE 4-12 Calculation of the 1I4T and 3/4 T Fluence Factor Values used for the Generation of the 22 EPFY Heatup/Cooldown Curves Azimuth                            1/4 T F        114T FF         3/4T F           3/4 T FF (n/cm 2, E > 1.0 MeV)           (ncm2, E >1.0 MCV)
Intermediate Shell Forging 5P-5933              8.35 x 1018       0.949        3.01 x 1018          0.671 Lower Shell Forging 5P-5951                 8.35 x 1018       0.949        3.01 x 10I           0.671 Nozzle Shell Forging 123J218                 2.49 x 1O18       0.623        8.99 x 10'7          0.396 Inter. to Lower Shell Forging Circ.            8.05 x lO"s       0.939       2.90 x 108            0.662 Weld Seam WF336 (Heat 442002)
Nozzle Shell to Inter. Shell Forging            2.49 x lO1S       0.623        8.99 x 1IO7          0.396 Circ. Weld Seam WF501 (Heat 442011)
TABLE 4-13 Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 32 EPFY Heatup/Cooldown Curves Azimuth                           1/4 T F        1/4T FF         3/4T F           3/4 T FF l                                            (n/cm", E > 1.0 MeV)            (ncm 2, E >1.0 MeV)
Intermediate Shell Forging 5P-5933              1.21 x 10'9       1.053       4.37 x 10'8           0.770 Lower Shell Forging 5P-5951                 1.21 x 1019       1.053       4.37 x 1018           0.770 Nozzle Shell Forging 123J218                3.63 x 108        0.720        1.31 x 10l8          0.473 Inter. to Lower Shell Forging Circ.            1.16 x 10'9       1.041        4.20 x I0            0.759 Weld Seam WF336 (Heat 442002)
Nozzle Shell to Inter. Shell Forging            3.63 x 10'8       0.720        1.31 x 1018          0.473 Circ. Weld Seam WF501 (Heat 442011)
Contained in Tables 4-14 through 4-17 are the calculations of the ART values used for the generation of the 22 and 32 EFPY heatup and cooldown curves.
Revision 0 Calculation ofof Adjusted Reference Temperature Adjusted Reference Temperature                                                   Revision 0

4-13 TABLE 4-14 Calculation of the ART Values for the 1/4T Location              @ 22 EFPY Reactor Vessel Beltline            Material        Cu%         Ni%        CFld)l    f22(')           Y.-t r       YA-t         I       ARTNDT(C)        aT        CA         M      ART<)
Region Location             Identirication                                         EFPY         (x 10")        FF (x 10")
Intermediate Shell                  SP-5933          0.04        0.74       26.0        1.39         0.835       0.949        40           24.7          0        12.3      24.7       89 Forging Intermediate Shell                                                          30.3        1.39         0.835       0.949        40           28.8          0        17.0      34.0      103 Forging
  -+ using SIC Data(')
Lower shell Forging                 SP-5951          0.04        0.64       26.0        1.39         0.835       0.949         10          24.7          0        12.3      24.7       59 Inter. to Lower Shell               WF-336           0.04        0.63       54.0        1.34         0.805       0.939        -30          50.7          0        25.4      50.7      71 Circ. Weld Metal Inter. to Lower Shell                                                       65.9        1.34         0.805       0.939        -30          61.9          0        14.0      28.0      60 Circ. Weld Metal
  -e using S/C Data Nozzle Shell Forging                123J218          0.05        0.72       31.0       0.415         0.249       0.623        30           19.3          0         9.7      19.3      69 Nozzle Shell to Inter.              WF-501           0.03        0.67       41.0       0.415         0.249       0.623         10          25.5          0        12.8      25.5      61 Shell Circ. Weld Metal Nozzle Shell to Inter.                                                      25.7       0.415         0.249       0.623         10          16.0          0         8.0      16.0      42 Shell Circ. Weld Metal
  -4 using S/C Data NOTES:
(a) Fluence, f, is based upon fif (I0O9 n/cm 2, E>1.0 MeV).
(b) ART = I + ARTNJDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c)    ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Byron I Weld (WF-336, heat # 442002) and the Byron 2 Weld (WF-447, Heat # 442002).
The CF for the Nozzle Shell to Inter. Shell Circ. Weld is integrated between Byron I Weld (WF-501, heat # 442011) and the Braidwood I & 2 Welds (WIF-562, heat # 442011).
(e) Surveillance data is considered not credible. In addition, the Table chemistry factor is conservative and would normally be used for calculating RTPTS. However, because the chemistry factor predicted by the Regulatory Guide 1.99 Position 2.1 for the forging surveillance data was greater than the Position 1.1 chemistry factor, then the Position 2.1 chemistry factor will be used to determine ART with a full margin term.
Revision 0 Calculation of Adjusted Reference       Temperature Reference Temperature                                                                                                                             Revision 0

4-14 TABLE 4-15 Calculation of the ART Values for the 3/4T Location @ 22 EFPY 3
Reactor Vessel Beitline           Material         Cu%         Ni%        CFad)     r221')         /-t f        %-t           I       ARTNDTI)         a 0j     . A         M      ART4b)
Region Location             Identiflcation                                        EFPY         (a 10'9)        FF (x lo,~)
Intermediate Shell                 5P-5933            0.04       0.74       26.0        1.39         0.301       0.671         40           17.4         0         8.7       17.4      75 Forging Intermediate Shell                                                          30.3        1.39         0.301       0.671         40          20.3          0        17.0      34.0       94 Forging
  -e using S/C Data(')
Lower shell Forging                5P-5951           0.04        0.64       26.0        1.39         0.301       0.671         10           17.4         0         8.7       17.4      45 Inter. to Lower Shell               WF-336           0.04        0.63       54.0        1.34         0.290       0.662        -30          35.7          0        17.9      35.7      41 Circ. Weld Metal                                             .                                                            .
Inter. to Lower Shell                                                       65.9        1.34         0.290       0.662        -30          43.6          0        14.0      28.0       42 Circ. Weld Metal
  -4 using S/C Data                                                                                                         .
Nozzle Shell Forging                1231218          0.05        0.72       31.0       0.415        0.0899       0.396         30          12.3          0         6.1      12.3       55 Nozzle Shell to Inter.             WF-501            0.03        0.67       41.0       0.415        0.0899       0.396         10          16.2          0         8.1      16.2      42 Shell Circ. Weld Metal Nozzle Shell to Inter.                                                      25.7       0.415        0.0899       0.396         10          10.2          0         5.1      10.2      30 Shell Circ. Weld Metal
  - using S/C Data                                 .                                                                                                                  _
NOTES:
(a) Fluence, f, is based upon fe (1019 nckm, E>1.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c) ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Byron I Weld (WF-336, heat # 442002) and the Byron 2 Weld (WF-447, Ileat #442002).
The CF for the Nozzle Shell to Inter. Shell Cire. Weld is integrated between Byron I Weld (WF-501, heat # 442011) and the Braidwood I & 2 Welds (WF-562, heat # 442011).
(e) Surveillance data is considered not credible. In addition, the Table chemistry factor is conservative and would normally be used for calculating RTrs. However, because the chemistry factor predicted by the Regulatory Guide 1.99 Position 2.1 for the forging surveillance data was greater than the Position 1.1 chemistry factor, then the Position 2.1 chemistry factor will be used to determine ART with a full margin term.
Revision 0 Reference Temperature Calculation of Adjusted Reference       Temperature                                                                                                                            Revision 0

4-15 TABLE 4-16 Calculation of the ART Values for the 1/4T Location             @ 32 EFPY Reactor Vessel Beltline             Material         Cu%        Ni%        CFd)       f @ 32(')       VA-t *      '-t          I      ARTNDT(c)        ax        aa         M      ARIO)
Region Location              Identification                                          EFPY        (x 10&deg;9)       FF Intermediate Shell                    5P-5933          0.04      0.74       26.0          2.02          1.21      1.053        40          27.4          0       13.7      27.4       95 Forging                                                                                                                    .
Intermediate Shell                                                          30.3         2.02           1.21      1.053        40          31.9          0       17.0      34.0      106 Forging
  -) using S/C Data(')
Lower shell Forging                   5P-5951          0.04      0.64       26.0         2.02           1.21      1.053        10          27.4          0       13.7      27.4       65 Inter. to Lower Shell                 WF-336           0.04      0.63       54.0          1.94          1.16      1.041       -30          56.2         0        28.0      56.0       82 Circ. Weld Metal Inter. to Lower Shell                                                       65.9          1.94          1.16      1.041       -30          68.6         0        14.0      28.0       67 Circ. Weld Metal
   - using S/C Data Nozzle Shell Forging                  123J218          0.05      0.72       31.0         0.604         0.363      0.720        30          22.3          0       11.2      22.3       75 Nozzle Shell to Inter.                WF-501           0.03      0.67       41.0         0.604         0.363      0.720        10          29.5         0        14.8      29.5       69 Shell Circ. Weld Metal Nozzle Shell to Inter.                                                      25.7         0.604         0.363      0.720        10          18.5         0         9.3       18.5     47 Shell Circ. Weld Metal
  -l using S/C Data NOTES:
(a)    Fluence, f, is based upon f,,,.(1019 n/cm 2, 1>1.0 MeV).
(b)    ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c)   ARTNDT = CF
* FF (d)   The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Byron I Weld (WF-336, heat # 442002) and the Byron 2 Weld (WF-447, Heat # 442002).
The CF forthe Nozzle Shell to Inter. Shell Circ. Weld is integrated between Byron I Weld (WF-501, heat# 44201 1) and the Braidwood I & 2 Welds (WF-562, heat # 442011).
(e)   Surveillance data is considered not credible. In addition, the Table chemistry factor is conservative and would normally be used for calculating RTrrs. However, because the chemistry factor predicted by the Regulatory Guide 1.99 Position 2.1 forthe forging surveillance data was greater than the Position 1.1 chemistry factor, then the Position 2.1 chemistry factor will be used to determine ART with a full margin term.
Calculation of Adjusted Reference Temperature                                                                                                                                 Rcvision 0

4-16 TABLE 4-17 Calculation of the ART Values for tlic 3/4T Location @ 32 EFPY Reactor Vessel Beltline           Material          Cu%         NI%        CFOd)     f @ 32"')       'Yt f        3/4-t         I       ARTNDT(c        Gr         GA         M      ART(b)
Region Location              Identification                                        EFPY          (x 101)        FF (x 10"9)
Intermediate Shell                  5P-5933           0.04       0.74        26.0        2.02          .437        .770         40          20.0          0        10.0      20.0       80 Forging                                                                                                                                                                   I Intermediate Shell                                                           30.3        2.02          .437        .770        40           23.3          0        17.0      34.0        97 Forging
  -* using S/C Data(')
Lower shell Forging                 5P-5951           0.04       0.64        26.0        2.02          .437        .770         10          20.0          0        10.0      20.0       50 Inter. to Lower Shell               WF-336            0.04       0.63        54.0        1.94          .420        .759        -30          41.0          0        20.5      41.0       52 Circ. Weld Metal              l                   l Inter. to Lower Shell                                                        65.9        1.94          .420        .759        -30          50.0          0        14.0      28.0       48 Circ. Weld Metal
  -) using S/C Data Nozzle Shell Forging                123J218           0.05       0.72        31.0       0.604         0.131       0.473        30            14.7         0         7.3      14.7       59 Nozzle Shell to Inter.              WF-501            0.03       0.67        41.0       0.604         0.131       0.473         10          19.4          0         9.7      19.4       49 Shell Circ. Weld Metal                                                                                                                               II Nozzle Shell to Inter.                                                       25.7       0.604         0.131       0.473         10          12.2          0         6.1      12.2       34 Shell Circ. Weld Metal e using S/C Data NOTES:
(a) Fluence, f, is based upon ff r(l019 n/cm 2, E>1.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c) ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Byron I Weld (WF-336, heat # 442002) and the Byron 2 Weld (WF447, Heat # 442002).
The CF for the Nozzle Shell to Inter. Shell Circ. Weld is integrated between Byron I Weld (Wr-501, heat # 442011) and the Braidwood I & 2 Welds (WF-562, heat # 442011).
(e) Surveillance data is considered not credible. In addition, the Table chemistry factor is conservative and would normally be used for calculating RTprs. I Iowever, because the chemistry factor predicted by the Regulatory Guide 1.99 Position 2.1 for the forging surveillance data was greater than the Position 1.1 chemistry factor, then the Position 2.1 chemistry factor will be used to determine ART with a full margin term.
Calculation of Adjusted Reference Temperature                                                                                                                                   Revision 0

4-17 The intermediate shell forging 5P-5933 is the limiting beltline material for all heatup and cooldown curves to be generated. The ART values associated with this material will be used in all sets of curves.
Contained in Tables 4-18 and 4-19 is a summary of the limiting ARTs to be used in the generation of the Byron Unit I reactor vessel heatup and cooldown curves.
TABLE 4-18 Summary of Adjusted Reference Temperature (ART) at I/4T and 3/4T Location for 22 EFPY Material                                      22 EFPY 114T ART                    3/4T ART Intermediate Shell Forging 5P-5933                   89                          75
        - Using Surveillance Data(a)                       103(a)                       94(J)
Lower Shell Forging 5P-5951                          59                          45 Circumferential Weldc WF-336                         71                          41
        - Using Credible Surveillance Data                   60                          42 Circumferential 'eld WF-501                          61                          42
        -  Using Credible Surveillance Data                  42                          30 form Braidwood I and 2 Nozzle Shell Forging 1231218                         69                          55 NOTES:
(a) These ART values were used to calculate the Heatup and cooldown curves in Figure 5-1 and 5-2.
Revision 0 Calculation of           Reference Temperature Adjusted Reference of Adjusted           Temperature                                                       Revision 0

4-18 TABLE 4-19 Summary of Adjusted Reference Temperature (ART) at 1/4T and 314T Location for 32 EFPY Material                                        32 EFPY 1/4T ART                    3/4T ART Intermediate Shell Forging 5P-5933                  95                          80
        - Using Surveillance Data                          106(a)                       97(a)
Lower Shell Forging 5P-5951                          65                          50 Circumferential Weld WF-336                          82                          52
        - Using Credible Surveillance Data                   67                          48 Circumferential Weld WF-501                          69                          49
        - Using Credible Surveillance Data                   47                          34 form Braidwood I and 2 Nozzle Shell Forging 123J218                         75                          59 NOTES:
(a) These ART values were used to calculate the Heatup and cooldown curves in Figure 5-3 and 54.
Revision 0 Reference Temperature Calculation of Adjusted Reference Temperature                                                        Revision 0

5-s 5         HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Section 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-NP-A[8 1,dated January 1996.
Figures 5-1 through 54 present the 22 and 32 EFPY heatup and cooldown curves (without margins for possible instrumentation errors) for a heatup rate of I00&deg;F/hr and cooldown rates of 0, 25, 50 and I00&deg;F/hr using the 1996 Appendix G methodology 16 ]. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 through 5-4. This is in addition to other criteria which must be met before the reactor is made critical, as discussed in the following paragraphs.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-1 and 5-3 (for the specific heatup rate being utilized). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-64012 ] (approved in February 1999) as follows:
1.5K im< Kilc(1l
: where, Kim is the stress intensity factor covered by membrane (pressure) stress, KI, = 33.2 + 20.734 el.02 Cr- RTNDT)],
T is the minimum permissible metal temperature, and RTNDT   is the metal reference nil-ductility temperature The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 3. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3 of this report. The minimum temperatures for the inservice hydrostatic leak test for the Byron Unit I reactor vessel at 22 and 32 EFPY are 1630 F, 1661F at 2485 psig 1996 App. G Methodology. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Revision 0 Heatup and Cooldovwn                           Limit Curves Pressure-Temperature Limit Cooldown Pressure-Temperature           Curves                                         Revision 0

5-2 Figures 5-1 through 5-4 define all of the above limits for ensuring prevention of nonductile failure for the Byron Unit I reactor vessel. The data points for the heatup and cooldown pressure-temperature limit curves shown in Figures 5-1 through 5-4 are presented in Tables 5-1 through 5-4, respectively.
Heatup and Cooldown Pressure-Temperature Limit Curves                                             Revision 0

5-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 22 EFPY:                    I/4T, 1030F 314T, 940 F 2500 2250 2000 1750 in R 1500 2
1 IL 1250 3 1000 750 500 b
250 0
0       50      100     150     200      250    300 350   400     450     500      550 Moderator Temperature (Deg. F)
FIGURE 5-1 Byron Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of I 00 0F/hr)
Applicable to 22 EFPY using 1996 Appendix G (Without Margins for Instrumentation Errors)
Revision 0 Heatup and Cooldown Pressure-Temperature   Limit Curves Pressure-Temperature Limit Curves                                    Revision 0

5-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 22 EFPY:                                                          1/4T, 103&deg;F 3/4T, 94&deg;F 2500 1641813766                i-                       i'                     :I'                                                                        !I I                  I f                                               l       l             lI
______I                                           _l_                 ,       i              i                                        I                     I 2250                                                                                                                                      i9_        -. .~! __._w ----- _-        ._.
I.1                                                             ! i
                                                                                             ,       .             ;-                         I
                             .               I                        I I
2000       _  ,    _  _   _ .......           _   .      ....       .. __  .
j fl 1-i I
Unacceptable                                                                                    I                         i                   I Operation 1750   -                                                                                                                                                 -I.  -
Acceptablc Operation                                                  i 1500            -                                                                                                              ._
                                                                                                                                                           ----I------      1 - -     .  . -
   ,2 e
                                                                                                                    '                          I I              I                         i                   i         !
1250                                                I
                                                          .,   .                                     I                                        I_-
I
  .5 1000
                                                /
i:
n Cooldown Rates (Fillr)
                              .                               stedy-state, 750
                              .*              ,                   0, and
                                                              -100 500 Boltup 250                                 --    i---Temp                                        4            .----                                                 -i I      I                          I 0
0           50             100                    150              200              250         300                  350                      400           450    500        550 Moderator Temperature (Deg. F)
FIGURE 5-2 Byron Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I 00 0F/hr) Applicable to 22 EFPY using. 1996 Appendix G (Without Margins for Instrumentation Errors)
Limit Curves                                                                                            Revision 0 Pressure-Temperature Limit Heatup and Cooldown Pressure-Temperature                                             Curves                                                                                          Revision 0

5-5 TABLE 5-1 Byron Unit I Heatup Data at 22 EFPY Using 1996 App. G Methodology (Without Margins for Instrumentation Errors) 100 Heatup                   Critical. Limit              Leak Test Limit T              P              T              P              T           P 60              0             163             0             146         2000 60             732            163            732            163         2485 65             732            163            732 70             732            163            732 75             732            163            732 80             732            163            732 85             732            163            732 90             732            163            733 95             733            163            737 100            737            163            743 105            743            163            753 110            753            163            766 115            766            163            782 120            782            165            801 125             801            170            823 130            823            175            849 135            849            180            879 140            879            185            913 145            913            190            951 150             951            195           994 155            994            200           1042 160           1042            205           1095 165           1095            210           1155 170           1155            215           1221 175           1221            220           1294 180           1294            225           1376 185           1376            230           1466 190            1466            235           1566 195            1566            240           1676 200            1676            245           1798 205            1798            250           1932 210            1932            255           2081 215            2081            260           2245 220            2245            265           2425 225            2425 Curves                                        Revision 0 Heatup and Heatup               Pressure-Temperature Limit Cooldown Pressure-Temperature and Cooldown                       Limit Curves                                         Revision 0

5-6 TABLE 5-2 Byron Unit I Cooldown Data at 22 EFPY Using 1996 App. G Methodology (Without Margins for Instrumentation Errors)
Steady State                  25F                          50F             lOOF T            P             T             P            T             P    T         P 60           0            60              0           60             0  60         0 60          763           60           719            60           677  60        596 65          779           65           738            65           698  65        622 70          798           70           759            70           721  70        651 75          818           75           781            75           746  75        684 80          841           80           807            80           774  80        720 85          866           85           834            85           806   85       760 90          894           90           865            90           840  90        804 95          924           95           900            95           879  95        853 100          958           100          937           100           921  100       908 105          995           105          979           105           969 110         1037           110          1026          110           1021 115         1082           115          1077 120         1133 125         1188 130         1250 135         1318 140         1393 145         1476 150         1568 155         1669 160         1781 165         1905 170         2042 175         2194 180         2361 Curves                                Revision 0 Heatup and Heatup                Pressure-Temperature Limit Cooldown Pressure-Temperature and Cooldown                       Limit Curves                                 Revision 0

5-7 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:                   1/4T, 106 0 F 3/4T, 97 0 F 2500 2250 2000 1750 0L  1500 3
1 1250 I1 M
U 1000 750 500 250 0 0        50      10 0       so      100    150 200 250 300 350 400                  450      500    550 Moderator Temperature (Deg. F)
FIGURE 5-3 Byron Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of I00&deg;F/hr)
Applicable to 32 EFPY Using 1996 Appendix G (Without Margins for Instrumentation Errors)
Revision 0 Heatup and Cooldown                       Limit Curves Pressure-Temperature Limit Cooldown Pressure-Temperaturc        Curves                                    Revision 0

5-8 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:                             I/4T, 106&deg;F 3/4T, 97&deg;F 2500      l_               l -_                                _302616034 2250' 2000-'           i Unacceptable Operation l             l                                      I 1750-                .          ~                    ~-~       ---------               ----         ;
lI           Acceptable j    ILOperation SP1500----------/---                                                  -           ---              --   L S    ~       ~I                  ,           I         l 1250                            __                                           _      L     _                  _
1000       -- aCooldown Rates (F/Hr)
I           steady-state, 750                               -5!                                                                         -and-500-   -    -    ~    _   _              _   _   j_                      _   _     _      _ r-I            Boltup    [     t         l           l        l            {
250       ----    t    :         T-0          50      100        150      200         250        300      350          400    450     500     550 Moderator Temperature (Deg. F)
FIGURE 54 Byron Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I00&deg;F/hr) Applicable to 32 EFPY Using 1996 Appendix G (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves                                                             Revision 0

5-9 TABLE 5-3 Byron Unit I Heatup Data at 32 EFPY Using 1996 App. G Methodology (Without Margins for Instrumentation Errors) 100 Heatup                    Critical. Limit              Leak Test Limit T               P       I       T                P             T            P 60               0             166                0            149         2000 60              720            166               720           166         2485 65              720            166               720 70              720            166               720 75              720            166               720 80              720            166               720 85              720            166               720 90              720            166               720 95              720            166               723 100             723            166               729 105             729            166               737 110             737            166               749 115             749            166               764 120             764            166               781 125             781            170               802 130             802            175               826 135             826            180               854 140             854            185               886 145             886            190               921 150             921            195               962 155             962            200              1007 160            1007            205              1057 165            1057            210              1113 170            1113            215              1175 175            1175            220              1244 180            1244            225              1321 185            1321            230              1406 190            1406            235              1499 195            1499            240              1603 200             1603            245              1718 205             1718            250              1844 210             1844            255              1984 215             1984            260              2138 220             2138            265              2308 225             2308 Revision 0 Heatup                                     Limit Curves Pressure-Tempcrature Limit Hcatup and Cooldown Pressure-Temperature         Curves                                         Revision 0

5-10 TABLE 5-4 Byron Unit I Cooldown Data at 32 EFPY Using 1996 App. G Methodology (Without Margins for Instrumentation Errors)
Steady State                    25F                           5TF T            P       IT                  P               T POOF P   LT,    P 60            0             60             0             60              0     60   0 60           753            60           709             60            665     60  581 65           769            65           726             65            685     65  606 70           787            70           746             70            706     70  633 75           806            75           767             75            730     75  663 80           827            80           791             80            757     80  697 85           851            85           817             85            786     85  735 90           877            90           846             90            819     90  777 95           906            95           879             95            855     95  823 100           937           100           914             100           895    100  874 105           973           105           954             105           940    105  931 110          1011            110          997             110           989 115          1054           115          1045             115           1043 120          1102           120          1099 125          1154 130          1212 135          1276 140          1347 145          1425 150          1512 155          1607 160          1713 165          1829 170          1958 175          2101 180          2258 185          2433 Curves                                Revision 0 Heatup and            Pressure-Temperature Limit Cooldown Pressure-Temperature and Cooldown                       Limit Curves                                Revision 0

6-1 6      REFERENCES I      Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U.S.
Nuclear Regulatory Commission, May, 1988.
2      ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February 26, 1999.
3      Code of Federal Regulations, 10 CFR Part 50, Appendix G."Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
4      WCAP-16143-P, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2", K.R. Hsu, November 2003.
5      ASME Code Case N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels", Section XI, Division 1,Approved December 12, 1997.
6      ASME Boiler and Pressure Vessel Code, Section XI, "Rule for Inservice Inspection of Nuclear Power Plant Components", Appendix G "Fracture Toughness Criteria for Protection Against Failure", December 1995.
7      WCAP-15183, Revision 1, "Commonwealth Edison Company Byron Unit I Surveillance Program Credibility Evaluation", T.J. Laubham, November 2003.
8      WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D. Andrachek, et al., January 1996.
9       1989 Section III, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Material for Vessels".
10      SAE-REA-00-546, "Reactor Vessel Neutron Exposure Projections for the Byron/Braidwood Uprating", J.D. Perock, January 12, 2000.
11     WCAP-1 5180, Revision 1, "Commonwealth Edison Company Byron Unit 2 Surveillance Program Credibility Evaluation", T.J. Laubham, November 2003.
12     WCAP-14824, Revision 2, "Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration For Byron and Braidwood", T. J. Laubbam, et al., November 1997. Ref. Errata letter CAE-97-233, CCE-97-316, "Transmittal of Updated Tables to WCAP-14824 Rev. 2" 13     WCAP-1 5178, "Byron Unit 2 Heatup and Cooldown for Normal Operation", T. J. Laubham, June 1999.
References                                                                                   Revision 0

6-2 14     WCAP-15124, "Byron Unit 1 Heatup and Cooldown for Normal Operation", T. J. Laubham, November 1998.
15     WCAP-1 5373, Revision 1, "Braidwood Unit 2 Heatup and Cooldown for Normal Operation", T.
J. Laubham, October 2003.
16     CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1996.
17     WCAP-1 5368, "Commonwealth Edison Company Braidwood Unit 2 Surveillance Program Credibility Evaluation", T.J. Laubham, March 2000.
18     WCAP-15123, Revision 1, "Analysis of Capsule W from Commonwealth Edison Company Byron Unit I Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et. al., January 1999.
19     WCAP-15 176, "Analysis of Capsule X from Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et. al., March 1999.
20     WCAP-1 5316, Revision 1,"Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit I Reactor Vessel Radiation Surveillance Program," Ed Terek, et. al., December 1999.
21     WCAP-15369, "Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et. al., March 2000.
22     NDIT No. MSD-98-044, "Best Estimate Chemistry Values for Reactor Pressure Vessel Beltline Weld Heat Number 442002", Dated December 1998.
Revision 0 References                                                                                     Revision 0

A-I APPENDIX A Thermal Stress Intensity Factors (K1 t)
Revision 0 Thermnal        Intensity Factors Stress Intensity Thermal Stress            Factors                                      Revision 0

A-2 TABLE Al Kit Values for 100 0F/hr Heatup Curve (22 EFPY)
Vessel Temperature          1/4T Thermal        Vessel Temperature      314T Thermal Water          @ 1/4T Location for               Stress        @ 3/4T Location for         Stress Temp.           100'F/hr Heatup          Intensity Factor      100F/hr Heatup        Intensity Factor (OF)                   (OF)            (KSI SQ. RT. IN.)             (OF)          (KSI SQ. RT. IN.)
PT Curves are Limited by the 3/4T Location for the Entire Curve 60                   56.01                   -0.9945                55.05               0.4767 65                   58.62                   -2.4413                55.31                1.4422 70                   61.70                   -3.6836                56.01               2.4209 75                   65.01                    4.8600                57.20               3.3365 80                   68.58                   -5.8718                58.80               4.1525 85                   72.27                   -6.7937                60.80               4.8806 90                  76.14                   -7.5912                63.14               5.5230 95                  80.11                   -8.3181                65.77               6.0965 100                  84.21                   -8.9517                68.68               6.6042 105                  88.40                   -9.5267                71.81               7.0564 110                  92.68                  -10.0310                75.15               7.4589 115                  97.04                  -10.4906                78.67               7.8204 120                  101.46                 -10.8960                82.35               8.1440 125                  105.95                 -11.2667                86.16               8.4352 130                  110.48                 -11.5953                90.10               8.6969 135                  115.06                 -11.8972                94.14               8.9336 140                  119.68                 -12.1661                98.28               9.1474 145                  124.34                 -12.4147               102.49               9.3417 150                  129.03                 -12.6374               106.78               9.5183 155                  133.75                 -12.8446                111.14              9.6799 160                  138.48                 -13.0315                115.55              9.8277 165                  143.25                 -13.2066                120.01              9.9638 170                  148.02                 -13.3657                124.52              10.0892 175                  152.82                 -13.5158                129.06              10.2056 180                  157.63                 -13.6533                133.64              10.3136 185                  162.46                 -13.7841               138.25               10.4146 190                  167.29                 -13.9046               142.88               10.5091 195                  172.14                 -14.0204               147.54               10.5982 200                  176.99                 -14.1279                152.22              10.6822 205                  181.85                 -14.2319                156.92              10.7619 210                  186.72                 -14.3292                161.63              10.8377 215                  191.59                 -14.4240                166.36              10.9102 220                  196.47                  -14.5135               171.09              10.9796 225                  201.36                 -14.6012                175.85              11.0465        I Vessel Radius to the 'AT and 3%TLocations are as follows:
* I/4T Radius = 88.750"
* 3/4T Radius = 93.000" tres Inensty actos Theral                                                                        Rvison Therrnal Stress Intensity Factors                                                                   Revision 0

A-3 TABLE A2 K1t Values for I 00 0F/hr  Cooldown Curve (22 EFPY)
Vessel Temperature 100 0 F/hr Cooldown Water         @ 1/4T Location for 1/4T Thermal Stress Temp.         100'F/hr Cooldown             Intensity Factor (OF)                  (OF)               (KSI SQ. RT. IN.)
180                205.72                      16.0887 175                200.64                      16.0214 170                195.55                      15.9546 165                190.47                      15.8872 160                185.38                      15.8204 155                180.30                      15.7531 150                175.21                      15.6862 145                170.13                      15.6190 140                165.04                      15.5523 135                159.95                      15.4852 130                154.87                      15.4187 125                149.78                      15.3518 120                144.70                      15.2854 115                139.61                      15.2187 110                134.53                      15.1525 105                129.45                      15.0860 100                124.36                      15.0200 95                 119.28                      14.9538 90                 114.19                      14.8881 85                 109.11                      14.8221 80                 104.02                      14.7566 75                  98.94                      14.6909 70                  93.86                      14.6257 65                  88.77                      14.5603 60                  83.69                      14.4946 NOTE: From T = 105'F to       180 0 F, the I 00 0 F/hr Cooldown Rate is limited by the lower rates and/or the Steady State Condition.
Factors                                                                 Revision 0 Thermnal       Intensity Factors Thermal Stress Intensity                                                                         Revision 0

A-4 TABLE A3 Kit Values for 100 0F/hr Heatup Curve (32 EFPY)
Vessel Temperature           1/4T Thermal      Vessel Temperature       3/4T Thermal Water        @ 1/4T Location for               Stress       @ 314T Location for          Stress Temp.          100&deg;F/hr Heatup           Intensity Factor     100 0 F/hr Heatup     Intensity Factor (0 F)                  (OF)            (KSI SQ. RT. IN.)             (OF)         (KSI SQ. RT. IN.)
PT Curves are Limited by the % T Location for the Entire Curve 60                   56.01                  -0.9945                55.05               0.4767 65                   58.62                  -2.4413                55.31               1.4422 70                   61.70                  -3.6836                56.01               2.4209 75                   65.01                  -4.8600                57.20               3.3365 80                   68.58                  -5.8718                58.80               4.1525 85                   72.27                  -6.7937                60.80               4.8806 90                   76.14                  -7.5912                63.14               5.5230 95                   80.11                  -8.3181                65.77               6.0965 100                  84.21                  -8.9517                68.68               6.6042 105                  88.40                  -9.5267                71.81               7.0564 110                  92.68                 -10.0310                75.15               7.4589 115                  97.04                 -10.4906                78.67               7.8204 120                 101.46                 -10.8960                82.35               8.1440 125                 105.95                 -11.2667                86.16               8.4352 130                 110.48                 -11.5953                90.10               8.6969 135                 115.06                 -11.8972                94.14               8.9336 140                 119.68                 -12.1661                98.28               9.1474 145                 124.34                 -12.4147                102.49              9.3417 150                  129.03                -12.6374                106.78              9.5183 155                  133.75                 -12.8446               111.14              9.6799 160                  138.48                -13.0315                115.55              9.8277 165                  143.25                -13.2066                120.01              9.9638 170                  148.02                 -13.3657               124.52              10.0892 175                  152.82                 -13.5158               129.06              10.2056 180                  157.63                -13.6533                133.64              10.3136 185                  162.46                -13.7841                138.25              10.4146 190                  167.29                 -13.9046               142.88              10.5091 195                  172.14                 -14.0204               147.54              10.5982 200                  176.99                -14.1279                152.22              10.6822 205                  181.85                -14.2319                156.92              10.7619 210                  186.72                 -14.3292               161.63              10.8377 215                  191.59                 -14.4240               166.36              10.9102 220                  196.47                -14.5135                171.09              10.9796 225                 201.36                 -14.6012                175.85              11.0465 Revision 0 Thermal Stress Intensity Factors Stress Intensity Factors                                                                  Revision 0

A-5 TABLE A4 K1, Values for I000 F/hr Cooldown Curve (32 EFPY)
Vessel Temperature 100F/hr Cooldown Water        @ 1/4T Location for I/4T Thermal Stress Temp.         100&deg;F/hr Cooldown        Intensity Factor (0 F)                 (OF)           (KSI SQ. RT. IN.)
185                210.81                 16.1555 180                205.72                 16.0887 175                200.64                 16.0214 170                195.55                 15.9546 165                190.47                 15.8872 160                185.38                 15.8204 155                180.30                 15.7531 150                175.21                 15.6862 145                170.13                 15.6190 140                165.04                 15.5523 135                159.95                 15.4852 130                154.87                 15.4187 125                149.78                 15.3518 120                144.70                 15.2854 115                139.61                 15.2187 110                134.53                 15.1525 105                129.45                 15.0860 100                124.36                 15.0200 95                 119.28                 14.9538 90                 114.19                 14.8881 85                 109.11                 14.8221 80                 104.02                 14.7566 75                  98.94                 14.6909 70                  93.86                 14.6257 65                  88.77                 14.5603 60                  83.69                 14.4946 0         0          0 NOTE: From T 110 F to 180 F, the I 00 F/hr Cooldown Rate is limited by the lower rates and/or the Steady State Condition, Revision 0 Thermnal        Intensity Factors Stress Intensity Thermal Stress            Factors                                                                  Revision 0

ATTACHMENT 4-D WCAP-15392, Revision 2 Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation

Westinghouse Non-Proprietary Class 3 WCAP-15392                                             November 2003 Revision 2 Byron Unit 2 Heatup and Cooldown Limit Curves for' Normal Operation
           *Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15392, Revision 2 Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubham November 2003 Approved.
R. A. Cresham, Manager
                   ,/Reactor Component Design & Analysis Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02003 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST OF TABLES ..................                                                                iv LIST OF FIGURES ..................                                                               vi PREFACE .....................                                                                   vii EXECUTIVE 

==SUMMARY==
 ..................                                                           viii I     INTRODUCTION.                                                                             1-1 2     PURPOSE .2-3     CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .3-1 4     CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .......................................... 4-1 5     HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .5-1 6     REFERENCES .6-1 APPENDIX A ...........                                                                        A-I

iv LIST OF TABLES Table 4-1  Summary of the Peak Pressure Vessel Neutron Fluence Values at 22 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV) ............................................... 4-4 Table 4-2  Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV) ......................................           4-4 Table 4-3  Calculated Integrated Neutron Exposure of the Byron Unit 2 Surveillance Capsules Tested to Date ..............                                                                   4-5 Table 4-4  Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program .4-5 Table 4-5  Best Estimate Cu and Ni Weight Percent for the Byron Unit 2 Forging Materials .                          4-6 Table 4-6  Average Cu and Ni Weight Percent for the Byron Unit 2 Surveillance Weld Material Only (Heat # 442002) ....................                                                       4-6 Table 4-7  Best Estimate Cu and Ni Weight Percent Values for the Byron Units I & 2 Weld Material (Using Byron I & 2 Chemistry Test Results) .4-7 Table 4-8  Reactor Vessel Beltline Material Unirradiated Toughness Properties .4-8 Table 4-9  Calculation of Chemistry Factors Using Byron Unit 2 Surveillance Capsule Data .......                                                                                    4-9 Table 4-10 Calculation of Chemistry Factors Using Braidwood Units I & 2 Surveillance Capsule Data .......                                                                                   4-10 Table 4-11  Summary of the Byron Unit 2 Reactor Vessel Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 ...4-11 Table 4-12  Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 22 EFPY Heatup / Cooldown Curves ............................                                   4-12 Table 4-13  Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 32 EFPY Heatup / Cooldown Curves .4-12 Table 4-14  Calculation of the ART Values for the 1/4T Location @ 22 EFPY .4-13 Table 4-15  Calculation of the ART Values for the 3/4T Location @ 22 EFPY.                                         4-14 Table 4-16  Calculation of the ART Values for the 1/4T Location @ 32 EFPY .4-15 Table 4-17  Calculation of the ART Values for the 3/4T Location @32 EFPY .4-16

V Table 4-18 Summary of Adjusied Reference Temperature (ART) at 1/4T and 3/4T Location for 22 EFPY ...........................                                   4-17 Table 4-19 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location for 32 EFPY ......                                                        4-18 Table 5-1  Byron Unit 2 Heatup Data at 22 EFPY Using 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) ........................... 5-5 Table 5-2  Byron Unit 2 Cooldown Data at 22 EFPY Using 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) ........................... 5-6 Table 5-3  Byron Unit 2 Heatup Data at 32 EFPY Using 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) ........................... 5-9 Table 5-4  Byron Unit 2 Cooldown Data at 32 EFPY Using Appendix G Methodology (Without Margins for Instrumentation Errors) ........................... 5-10

vi LIST OF FIGURES Figure 5-1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of I 00 0 F/hr) Applicable to 22 EFPY Using Code Case N-640 & 1996 Appendix G (Without Margins for Instrumentation Errors) .......................................            5-3 Figure 5-2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I 000 F/hr) Applicable to 22 EFPY Using Code Case N-640 & 1996 Appendix G (Without Margins for Instrumentation Errors) ....................................... 5-4 Figure 5-3 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100&deg;F/hr) Applicable to 32 EFPY Using Code Case N-640 & 1996 Appendix G (Without Margins for Instrumentation Errors) ..............................                     5-7 Figure 5-4 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100&deg;F/hr) Applicable to 32 EFPY Using Code Case N-640 & 1996 Appendix G (Without Margins for Instrumentation Errors) ....................................... 5-8

vii PREFACE This report has been technically reviewed and verified by:
Reviewer:                                                             J. H. Ledger       4217d Revision 1:
An error was detected in the "OPERLIM" Computer Program that Westinghouse uses to generate pressure-temperature (PT) limit curves. This error potentially effects the heatup curves when the 1996 Appendix G Methodology is used in generating the PT curves. It has been determined that WCAP-1 5392 Rev. 0 was impacted by this error. Thus, this revision provides corrected curves from WCAP-I 5392 Rev.
0.
Note that only the heatup curves and associated data point tables in section 5 have changed. The cooldown curves and data points remain valid and were not changed.
Revision 2:
This revision was completed to incorporate an updated reference for the flange elimination from the PT limit curves. In the previous revision, WCAP-15315 was provided as justification for the flange notch elimination and it was changed to WCAP-16143-P in this revision. In addition to this change, the thermal stress intensity factors for the highest heatup and cooldown rates were added to this report in Appendix A.

viii EXECUTIVE 

==SUMMARY==

The purpose of this report is to generate new pressure-temperature limit curves for Byron Unit 2 for normal operation at 22 and 32 EFPY based of revised Uprated Fluences. The new pressure-temperature limit curves were generated using the methodology from WCAP 14040-NP-A, the 1996 ASME Boiler and Pressure Vessel Code, Section XI Appendix G ASME Code Case N-588, ASME Code Case N-640 and WCAP-16143-P ("Flange Notch" Elimination Justification).. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values. The I /4T and 3/4T values are summarized in Tables 4-18 and 19 and were calculated using the circumferential weld WF-447, Heat 442002 (The limiting material for circumferentially oriented flaws, Code Case N-588) and nozzle shell forging 4P-6107 (The limiting material for axial flaws). The pressure-temperature limit curves were generated for a heatup rate of I 007F/hr and cooldown rates of 0, 25, 50 and I 000F/hr. The axial oriented flaw cases are limiting for all curves at each EFPY value evaluated. Hence, only the axial oriented flaw curves are presented in this report and they can be found in Figures 5-1 through 5-4.

I-I 1        INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 607F.
RTNDT   increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT(IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials"[' 1 .
Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves for normal operation.
NOTE: For the reactor vessel radiation surveillance program, Babcock and Wilcox Co. supplied Westinghouse with sections of SA508 Class 3 forging material used in the core region of the Byron Station Unit No. 2 reactor pressure vessel (Specifically from forging 49D330-1/49C298-1). Also supplied was a weldment made with weld wire heat # 442002 Linde 80 flux, lot number 8064, which is identical to that used in the actual fabrication of the intermediate to lower shell girth weld of the pressure vessel).
Introduction

2-1 2        PURPOSE The Commonwealth Edison Company (now known as Exelon Nuclear) contracted Westinghouse to generate new pressure-temperature limit curves for 22 and 32 EFPY based on the revised fluences from the 5% Uprating. These new Pressure-Temperature Curves are to be developed utilizing the following methodologies:
* Regulatory Guide 1.99, Revision 2111,
* ASME Code Case N-640       21,
* Elimination of the flange requirement of Appendix G to I OCFR Part 50131 per WCAP-16143-P, "Reactor Vessel Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2'A4,
* ASME Code Case N-588t 51 (where applicable),
* Methodology of the 1996 ASME B&P Vessel Code, Section XI, Appendix          G16 1, and
* The PT Curves will be developed WITHOUT margins or instrumentation errors.
Based on the above methodologies, two sets of PT Curves will be generated. Set one will consist of the circumferential flaw methodology (ASME Code Case N-588) in combination with 1996 Appendix G to ASME Section XI and the K1 , methodology (ASME Code Case N-640) for the limiting circumferential weld material. Set two will consist of the 1996 Appendix G to ASME Section XI and the Kic methodology (ASME Code Case N-640) for the limiting forging/base metal material. Both sets of curves will use the methodology to eliminate the 10 CFR Part 50 Appendix G flange requirements (from WCAP-16143-P). The final PT curves to be presented herein will be the most limiting set of curves. If the situation arises where portions of each set of curves are limiting, then composite curves will be generated that are based on the most limiting data (i.e. Circ. Flaw or Axial Flaw Case).
The purpose of this report is to present the calculations and the development of the Byron Unit 2 heatup and cooldown curves for 22 and 32 EFPY. This report documents the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2111, for all the beltline materials and the development of the heatup and cooldown pressure-temperature limit curves for normal operation.
Per the request of the Exelon, the surveillance weld data from the Byron Unit I and Unit 2 surveillance programs has been integrated. Note that Byron Unit I surveillance weld is identical to the surveillance weld (Heat No. 442002) at Byron Unit 2. In addition, the Braidwood Units I and 2 surveillance weld is identical to the nozzle to intermediate shell circumferential weld (Heat No. 442011). Per WCAP-15183[7J, WCAP-151801 l1 and WCAP-15368117] all the surveillance data has been determined to be credible.
Purpose

3-1 3         CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1      Overall Approach Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements" 13 1specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements. Section XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components", Appendix GI6 , contains the conservative methods of analysis.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1,, for the metal temperature at that time. K1 , is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of PT Limit Curves for Section Xl"l2 61 of the ASME Appendix G to Section XI. The K1 , curve is given by the following equation:
K1, =332+20.734*e      1002(T RTNDT)]                          (1) where,      K1, =     reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K1 , curve is based on the lower bound of static critical K1 values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, SA-508-3 steel.
3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C* Kim + Kit < KI,                                     (2)
: where, Kim =        stress intensity factor caused by membrane (pressure) stress Kit =         stress intensity factor caused by the thermal gradients Ki, =         function of temperature relative to the RTNDT of the material C      =     2.0 for Level A and Level B service limits C     =       1.5 for hydrostatic and leak test conditions during which the reactor core is not critical Criteria For Allowable Pressure-Temperature Relationships

3-2 For membrane tension, the corresponding K, for the postulated defect is:
Ki m = Mmx(pRi/t)                                                     (3) where, Mm for an inside surface flaw is given by:
Mm      =       1.85 for ft < 2, Mm      =       0.926A/; for 2<ft<3.464, Mm      =       3.21 for r/t > 3.464 Similarly, Mm for an outside surface flaw is given by:
Mm      =       1.77 for a/ < 2, Mm      =       0.89317 for 2*<1<3.464, Mm      =       3.09 for ft > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.
For bending stress, the corresponding K, for the postulated defect is:
Klb   =  Mb
* Maximum Stress, where Mb is two-thirds of Mm The maximum K, produced by radial thermal gradient for the postulated inside surface defect of G-2120 is Kt = 0.953x1 0-3 x CR x t25 , where CR is the cooldown rate in IF/hr., or for a postulated outside surface defect, Ki, = 0.753xl 0-3 x HU x t2*5, where HU is the heatup rate in 'F/hr.
The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.
G-2214-2 for the maximum thermal K1.
(a)      The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).
(b)      Alternatively, the K, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a l/4-thickness inside surface defect using the relationship:
Ki t = (1.0359Co+ 0.6322Ci + 0.4753C2 + 0.3855C3) *Ir.                               (4)
Criteria For Allowable Pressure-Temperature Relationships

3-3 or similarly, Krr during heatup for a l4-thickness outside surface defect using the relationship:
K1 = (1.043Co + 0.630CI + 0.48 1C2 + 0.401C3)
* G                                     (5) where the coefficients C0 , C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
C(x) = Co + Ci(x / a) + C2(x / a)2 + C3(x / a)3                                     (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.
Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"181 Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.
At any time during the heatup or cooldown transient, K1, is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of Kic at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1 , exceeds K1 t, the calculated allowable pressure during cooldown will be greater than the steady-state value.
Criteria For Allowable Pressure-Temperature Relationships

3-4 The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a I14T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1. for the 1/4T crack during heatup is lower than the K1, for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower Kic values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
Code Case N-588: Circumferential Welds:
In 1997, ASME Section XI, Appendix G was revised to add methodology for the use of circumferential flaws when considering circumferential welds in developing pressure-temperature limit curves. This change was also implemented in a separate Code Case, N-588.
The earlier ASME Section XI, Appendix G approach mandated the postulation of an axial flaw in circumferential welds for the purposes of calculating pressure-temperature limits. Postulating the Appendix G reference flaw in a circumferential weld is physically unrealistic because the length of the reference flaw is 1.5 times the vessel thickness and is much longer than the width of the vessel girth welds.
Criteria For Allowable Pressure-Temperature Relationships

3-5 In addition, historical experience, with repair weld indications found during pre-service inspection and data taken from destructive examination of actual vessel welds, confirms that any flaws are small, laminar in nature and are not oriented transverse to the weld bead orientation. Because of this, any defects potentially introduced during fabrication process (and not detected during subsequent non-destructive examinations) should only be oriented along the direction of the weld fabrication. Thus, for circumferential welds, any postulated defect should be in the circumferential orientation.
The revision to Appendix G now eliminates additional conservatism in the assumed flaw orientation for circumferential welds. The following revisions were made to ASME Section XI, Appendix G:
G-2214.1 Membrane Tension...
The K, corresponding to membrane tension for the postulated circumferential defect of -2120 is Ki m = Mmx(pRi/1) where, Mm for an inside surface flaw is given by:
Mm     =      0.89 for a/ < 2, Mm     =      0.443 1t for 2*<t<3.464, Mm     =      1.53 for .11 > 3.464 Similarly, Mm for an outside surface flaw is given by:
Mm    =      0.89 for 4   < 2, Mm    =      0.44311 for 2<t<          3.464, Mm           1.53 for -it > 3.464 Note again, that the only change relative to the OPERLIM computer code was the addition of the constants for Mm in a circ. weld limited condition. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. As stated previously, the P-T curve methodology is unchanged from that described in WCAP-140401r] Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.
Criteria For Allowable Pressure-Temperature Relationships

3-6 3.3 Closure Head/Vessel Flange Requirements 10 CFR Part 50, Appendix G131 addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for Byron Unit 2 reactor vessel.
This requirement was originally based on concerns about the fracture margin in the closure flange region.
During the boltup process, stresses in this region typically reach over 70 percent of the steady-state stress, without being at steady-state temperature. The margin of 120'F and pressure limitation of 20 percent of the hydrotest pressure were developed using the KIA fracture toughness from the mid 1970's.
Improved knowledge of the fracture toughness and other issues which affect the integrity of the reactor vessel have led to the recent change to allow the use of Kic in development of pressure-temperature curves, as contained in Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1I"[21.
The discussion given in WCAP-16143-P, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2,"141 concluded that the integrity of the closure head/vessel flange region is not a concern for Byron or Braidwood, therefore the requirement of 10 CFR Part 50, Appendix G.can be eliminated from the Pressure-Temperature Curves contained in this report.
Criteria For Allowable Pressure-Temperature Relationships

4-1 4         CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART = InitialRTNDT + ARTNDT + Margin                                  (7)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-233 1 of Section III of the ASME Boiler and Pressure Vessel Codel 91 . If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTNDT    is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows:
A RTNDT = CF *f (0.280101ogfi                              (8)
To calculate ARTNDT at any depth (e.g., at 1/4T or 3/471), the following formula must first be used to attenuate the fluence at the specific depth.
f(deplh)x) = f luacc
* e0                                 (9) where x inches (vessel inner radius and beltline thickness is 86.625 inches and 8.5 inches, respectively) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNDT at the specific depth.
The Westinghouse Radiation Engineering and Analysis group evaluated the vessel fluence projections for the 5% Uprating and the results are presented in SAE-REA-00-546 1"]. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with the methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" 18 1. Tables 4-1 and 4-2, herein, contain the calculated vessel surface fluence values along with the Regulatory Guide 1.99, Revision 2, 1/4T and 3/4T calculated fluences used to calculate the ART values for all beltline materials in the Byron Unit 2 reactor vessel. Additionally, the calculated surveillance capsule fluence values are presented in Table 4-3.
Ratio Procedure and Temperature Adjustment:
The ratio procedure, as documented in Regulatory Guide 1.99, Revision 2 Position 2. 1, was used, where applicable, to adjust the measured values of ARTNDT of the weld materials for differences in copper/nickel content. This adjustment is performed by multiplying the ARTNDT by the ratio of the vessel chemistry factor to the surveillance material chemistry factor. The adjusted ARTNDT values are then used to calculate the chemistry factor for the vessel materials.
Calculation of Adjusted Reference Temperature

4-2 From NRC Industry Meetings on November 12, 1997 and February 12, 13 of 1998, procedural guidelines were presented to adjust the ARTNDT for temperature differences when using surveillance data from one vessel applied to another vessel. The following guidance was presented at these industry meetings:
Irradiation temperature and fluence (or fluence factor) are first order environmental variables in assessing irradiation damage... To account for differences in temperature between surveillance specimens and vessel, an adjustment to the data must be performed. Studies have shown that for temperatures near 5501F, a IIF decrease in irradiation temperature will result in approximately a I IF increase in ARTNDT-For capsules with irradiation temperature of T,,psuie and a plant with an irradiation temperature of Tplant, an adjustment to normalize ARTNDT,nmeasred to TpIant is made as follows:
Temp. Adjusted ARTNDT     ARTNDT.measurd + 1.0*( Tcapsule - Tplant)                (10)
The irradiation temperatures from Byron Units I & 2 are presented in WCAP-14824, Revision 2112]. The average irradiation temperature from each of the four Units and operating cycles in question is 5530 F.
Therefore, no temperature adjustment is required.
Chemistry Factor:
The chemistry factor is obtained from the tables in Regulatory Guide 1.99, Revision 2 using the best estimate average copper and nickel content as reported in Tables 4-5 through 4-8. The chemistry factors were also calculated using Position 2.1 from the Regulatory Guide 1.99, Revision 2 using all available surveillance data. Per Reference 11, the surveillance weld data and the lower shell forging data for Byron Unit 2 is credible. In addition, per Reference 7, the surveillance weld data for Byron Unit I is credible. Position 2.1 chemistry factors are calculated in Tables 4-9 and 4-10. Referencel 7, the Braidwood I & 2 surveillance weld data is credible.
Explanation of Margin Term:
When there are "two or more credible surveillance data sets"1t1 available for Byron Unit 2, Regulatory Guide 1.99 Rev. 2 (RGI.99R2) Position 2.1 states "To calculate the Margin in this case, use Equation 4; the values given there for oa may be cut in half'. Equation 4 from RGI.99R2 is as follows:
M = 21crT       +
Calculation of Adjusted Reference Temperature

4-3 Standard Deviation for Initial RTNDT Margin Term, a, If the initial RTNDT values are measured values, which they are in the case of Byron Unit 2, then Acis equal to 00F. On the other hand, if the initial RTNDT values were not measured, then a generic value of 17'F (base metal and weld metal) would have been required to be used for a,.
Standard Deviation for    ARTNDT   Margin Term, CFa Per RGl.99R2 Position 1.1, the values of ac are referred to as "281F for welds and 171F for base metal, except that i,&  need not exceed 0.50 times the mean value of ARTNDT-" The mean value of ARTNDT is defined in RG1.99R2 by Equation 2 and defined herein by Equation 8.
Per RGI .99R2 Position 2.1, when there is credible surveillance data, CA is taken to be the lesser of 1/2 ARTNDT or 147F (28 0 F/2) for welds, or 8.50 F (17'F/2) for base metal. Where ARTNDT again is defined herein by Equation 8.
Summary of the Margin Term Since cr1 is taken to be zero when a heat-specific measured value of initial RTNDT are available (as they are in this case), the total margin term, based on Equation 4 of RGI.99R2, will be as follows:
* Position 1.1:      Lesser of ARTNDT or 561F for Welds Lesser of ARTNDT or 34 0 F for Base Metal
* Position 2.1:      Lesser of ARTNDT or 280 F for Welds Lesser of ARTNDT or 17'F for Base Metal Calculation of Adjusted Reference Temperature

4-4 TABLE 4-1 Summary of the Peak Pressure Vessel Neutron Fluence Values at 22 EFPY used for the Calculation of ART Values (n/cm 2 , E > 1.0 MeV)
Material(')                           Surface(b)                  %T                3/4 T (n/cm2, E>l.0 MeV)     (n/Cm 2, E>1. MeV) (n/cm2, E>1.0 MeV)
Inter. Shell Forging 49D329-1/49C297-1                1.41 x 1O09            8.47 x 10s         3.05 x 1018 Lower Shell Forging 49D330-1/49C298-1                   1.41 x 10'9              8.47x 1018       3.05 x 108 Nozzle Shell Forging4P-6107                     3.58 x 10"             2.15 x 10i8        7.75 x 10"'
Intermediate to Lower Shell Forging Circ.               1.39 x 10' 9           8.35 x 10's        3.01 x 10" Weld Seam WF-447 (Heat 442002)
Nozzle Shell to Inter. Shell Forging Circ.             3.58 x 1018            2.15 x 108         7.75 x 10" Weld Seam WF-562 (Heat 442011)
Notes:
(a)     All remaining vessel materials are below I x 1017 n/cm 2, E > 1.0 MeV.
(b) Calculated Surface fluence documented in Reference 10.
TABLE 4-2 Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV)
Material(')                            Surface(b)                 %T                3   T 2                      2 (n/cm , E>1.0 MeN)     (n/cm , E>l.0 WeV) (n/cm2, E>1.0 N1 eV)
Inter. Shell Forging 49D329-1/49C297-1                2.06 x 1019            1.24 x 1019        4.46 x IO"1 Lower Shell Forging 49D330-1/49C298-1                  2.06 x 10'9            1.24 x 10'9        4.46 x 1018 Nozzle Shell Forging 4P-6107                    5.22 x 10'"            3.13 x 1018        1.13 x 1018 Intermediate to Lower Shell Forging Circ.              2.03 x 10'9             1.22 x IO'9       4.40 x 1018 Weld Seam WF-447 (Heat 442002)
Nozzle Shell to Inter. Shell Forging Circ.             5.22 x 1O' 8           3.13 x 10          1.13 x 10'"
Weld Seam WF-562 (Heat 442011)
Notes:
(a)     All remaining vessel materials are below I x 10'' n/cm 2 , E > 1.0 MeV.
(b)     Calculated Surface fluence documented in Reference 10.
Calculation of Adjusted Reference Temperature

4-5 TABLE 4-3 Calculated Integrated Neutron Exposure of the Byron Unit 2 Surveillance Capsules Tested to Date Capsule                           Fluence(')
U                 4.05 x 1018 n/cm 2 , (E > 1.0 MeV) 11.27 x 1019 n1cm 2 , (E > 1.0 meV)
X                 2.30 x1019 n/cm 2 , (E > 1.0 MeV)
Notes:
(a) Documented in WCAP-151761 '91.
Contained in Table 4-4 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline materials. These measured shift values were obtained using CVGRAPH, Version 4.l1161, which is a symmetric hyperbolic tangent curve-fitting program.
TABLE 4-4 Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program Material                              Capsule                  Measured 30 ft-lb Transition Temperature Shift(2)
Intermediate Shell Forging                          U                              -3.80F 49D330-1/49C298-1                               W                               3.650 F (Tangential Orientation)                           X                             15.750 F Intermediate Shell Forging                         U                              19.760 F 49D330-1/49C298-1                              W                              31.88WF (Axial Orientation)                            X                             38.910 F Surveillance Program                             U                              8.44OF Weld Metal                                 W                              28.880 F (Heat # 442002)                               X                             54.01OF Heat Affected Zone                              U                              6.74OF W                              30.440 F X                             34.22WF Notes:
(a) Documented in WCAP-151761"91.
Calculation of Adjusted Reference Temperature

4-6 Table 4-5 contains the best estimate weight percent copper and nickel for the Byron Unit 2 base materials in the beitline region. Table 4-6 contains the calculation of the best estimate weight percent copper and nickel for the Byron Unit 2 surveillance weld material, while Table 4-7 presents the overall best estimate average for that heat of weld. Table 4-8 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RTNDT of the beltline materials and vessel flanges. The weight percent values of Cu and Ni given in Table 4-8 were used to generate the calculated chemistry factor (CF) values based on Tables I and 2 of Regulatory Guide 1.99, Revision 2, and presented in Table 4-11. Tables 4-9 and 4-1 0 provide the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Position 2.1, which are also summarized in Table 4-11.
TABLE 4-5 Best Estimate Cu and Ni Weight Percent for the Byron Unit 2 Forging Materials Intermediate Shell Forging             Lower Shell Forging 49D329-1/49C297-1                 49D330-1/49C298-1 Reference                     Cu %               Ni %            Cu %          J    Ni %
Ref. 22                       0.01              0.70            0.05(a)            0.71(a)
Charpy YL-48(b)                    ---                ...            0.075               0.78 Charpy YT-52(b)                    ---                ---            0.073               0.78 Best Estimate Average(c)               0.01        l     0.70             0.06               0.73 Notes:
(a) This is the average of 4 data points.
(b) Charpy Specimens from Capsule X of Byron Unit 2 (Ref. 19).
(c)    The best estimate average was rounded per ASTM E29, using the "Rounding Method". This calculation is also presented in Reference 13.
TABLE 4-6 Average Cu and Ni Weight Percent for the Byron Unit 2 Surveillance Weld Material Only (Heat # 442002)
Reference                        Weight % Copper          Weight % Nickel Ref. 12                             0.023(a)                 0.712(a)
CharpyYYW-51(b)                            0.031                    0.70 Charpy YW-52&deg;b                             0.033                    0.70 Surveillance Weld Average(c)          j            0.02                    0.71 Notes:
(a) This is the average of 31 data points.
(b) Charpy Specimens from Capsule X of Byron Unit 2 (Ref. 19).
(c) The best estimate average was rounded per ASTM E29, using the "Rounding Method". This calculation is also presented in Reference 13.
Calculation of Adjusted Reference Temperature

4-7 TABLE 4-7 Best Estimate Cu and Ni Weight Percent Values for the Byron Units I & 2 Weld Material (Using Byron I & 2 Chemistry Test Results)
Chemistry Type                           Reference              Weight % Copper                Weight % Nickel B&W WQ: BAW-2261                                Ref. 12                       0.024                       0.70 B&W WQ: BAW-2261                                Ref. 12                       0.031                       0.46 B&W WQ: BAW-2261                                Ref. 12                        0.03                       0.72 B&W WQ: BAW-2261                                Ref. 12                       0.068                       0.48 B&W WQ: BAW-2261                                Ref. 12                       0.053                       0.62 B&W WQ: BAW-2261                                Ref. 12                       0.059                       0.62 B&W WQ (from NDIT No.                             Ref. 12                       0.029                       0.65 BYR97-346, Rev. 0)
Round Robin Data Ave. on Weld                       Ref. 12                       0.038                      0.658 WF-336 (from NDIT No. BRW-DIT-97-391, Rev. 0)
Byron I Surveillance Data Ave. '                     Ref. 12                       0.022                       0.69 Byron 2 Surveillance Ave.(b)                    Table 4-6                       0.023                      0.712 BEST ESTIMATE AVERAGE                                                           0.04(C  d)                  0.63(C. d)
Notes:
(a) The weld material in the Byron Unit I surveillance program was made of the same wire and flux as the reactor vessel inter. to lower shell girth seam weld. (Weld seam WF-336, Wire Heat # 442002, Flux Type Linde 80, Flux Lot # 8873).
(b) The Byron Unit 2 surveillance weld is identical to that used in the reactor vessel core region girth seam (WF-447). The weld wire is type Linde MnMoNi (Low Cu-P), heat number 442002, with a Linde 80 type flux, lot number 8064.
(c) The best estimate chemistry values were obtained using the "average of averages" approach. In addition the best estimate average was rounded per ASTM E29, using the "Rounding Method". This calc as presented in Reference 13.
(d) This ave. Cu &Ni excluded two data points (Cu = 0.114, Ni = 0.54 and Cu = 0.148, Ni = 0.60), see Ref. 23.
Calculation of Adjusted Reference Temperature

4-8 TABLE 4 -8 (b)
Reactor Vessel Beitline Material Unirradiated Toughness Properties Material Description                           Cu (%)              Ni(%)     Initial RTNDT)
Closure Head Flange 5P7382 / 3P6407                                          0.71        0 Vessel Flange 124L556VAI                             ---               0.70       30 Nozzle Shell Forging 4P-6107()                         0.05               0.74        10 Intermediate Shell Forging 49D329-1/49C297-1                   0.01               0.70      -20 Lower Shell Forging 49D330-1/49C298-1                       0.06               0.73      -20 Intermediate to Lower Shell Forging Circ. Weld                 0.04               0.63        10 Seam WF-447 (Heat # 442002)
Nozzle Shell to Intermediate Shell Forging Circ.               0.03               0.67        40 Weld Seam WF-562 (Heat # 442011)
Byron Unit I Surveillance Program                       0.02               0.69 Weld Metal (Heat # 442002)
Byron Unit 2 Surveillance Program                       0.02               0.71      ---
Weld Metal (Heat # 442002)
Braidwood Units I & 2 Surveillance Program                    0.03            0.67, 0.71   ---
Weld Metals (Heat # 442011)
Notes:
(a) The initial RT NDT values for the plates and welds are based on measured data.
(b) Table duplicated from Table 4-7 of Reference 13.
Calculation of Adjusted Reference Temperature

4-9 TABLE 4-9 Calculation of Chemistry Factors using Byron Unit 2 Surveillance Capsule Data Material               Capsule           Capsule          FF(b)     ARTNDT(C)      FF*ARTNDT            FF2 Lower Shell Forging             U                0.405         0.749          0.0(e)          0              0.561 49D330-1/49C298-1               W                 1.27         1.067          3.65          3.89             1.138 (Tangential)                X                2.30          1.225          15.75         19.29            1.500 Lower Shell Forging             U                0.405         0.749          19.76         14.80            0.561 49D330-1/49C298-1                W                 1.27         1.067         31.88         34.02             1.138 (Axial)                  X                2.30          1.225         38.91         47.66             1.500 SUM:      119.66            6.398 CFFoi = (FF
* RTNDT) -   X( FF2) = (119.66) * (6.398)    18.70 F Byron Unit I Surv. Weld           U                0.404         0.749          11.22         8.40             0.561 Material                                                              (5.61)(d)
(Heat # 442002)               X                 1.57         1.125          80.22        90.25             1.266 (40.1 1)(d) 2.43          1.239         102.68        127.22            1.535 (51. 3 4)(d)
Byron Unit 2 Surv. Weld           U                0.405         0.749          16.88         12.64            0.561 Material                                                              (8 .44)(d)
(Heat # 442002)               W                 1.27         1.067          57.76        61.63             1.138 (28.88y)(
X                 2.30         1.225         108.02        132.32            1.500 (54.01 )(d)
SUM:      432.46            6.561 CFs-1 . Wld. 442002 = X(FF
* RTNDT)
* X( FF2) = (432.46) + (6.561)    = 65.91F Notes:
(a) Calculated Capsule Fluence taken from References 18 and 19 (or Table 4-3 herein) for Byron Units I and 2, respectively.
(b) FF = fluence factor = fO.28-o.log (c)    ARTNDT values are the measured 30 ft-lb shift values (Byron Unit I from Ref. 18 & Byron Unit 2 from Ref.
19 or Table 4-4 herein).
(d) The Byron I & 2 surveillance weld metal ARTNDT values have been adjusted by a ratio factor of 2.00.
No temperature adjustment are required.
(e)    Actual shift was a -3.81. Since this physically should not occur, a conservative value of 0.0 will be used.
Calculation of Adjusted Reference Temperature

4-10 TABLE 4-1 0 Calculation of Chemistry Factors using Braidwood Units I & 2 Surveillance Capsule Data Material                Capsule        Capsule          FF(b)      ARTNDT()       FF*ARTNDT             FF 2 f<a)
Braidwood Unit I                  U            0.387         0.737         17.06(d)         12.57           0.543 Surveillance                  X             1.24         1.060        30 .15 (d)        31.96            1.124 Weld Heat 442011, WF-501               W            2.09          1.201        4 9 . 68 (d)      59.67            1.442 Braidwood Unit 2                  U           0.400          0.746         O.O(& e)           0             0.557 Surveillance                  X             1.23         1.058         26   .3(d)       27.83            1.119 Weld Heat 442011, WF-501               W            2.25          1.220         23.9(d)          29.16            1.488 SUM:      161.19           6.273 CF = X(FF
* RTNDT) . ( FF2) = (161.19) . (6.273) = 25.7 0F Notes:
(a) Calculated Capsule Fluence values taken from References 20 and 21 for Braidwood Units I and 2, respectively.
(b)    FF = fluence factor = f(O.28-0.logf)
(c)    ARTNDT values are the measured 30 fl-lb shift values (Braidwood Unit I from Ref. 20 & Braidwood Unit 2 from Ref. 21).
(d) The Braidwood I & 2 surveillance weld metal ARTNDT values do not require a ratio factor or temperature adjustment.
(e)    Actual shift was a -0.58. Since this physically should not occur, a conservative value of 0.0 will be used.
Calculation of Adjusted Reference Temperature

~--                                                                  -
4-11 TABLE 4-11 Summary of the Byron Unit 2 Reactor Vessel Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Material                             Chemistry Factor Position 1.1          Position 2.1 Inter. Shell Forging 49D329-1/49C297-1           20.0F                               l Lower Shell Forging                   37.0F                 18.7 0 F 49D330-1 /49C298-1 Nozzle Shell Forging 4P-6107               31.0LF                  ---
Intermediate Shell to Lower Shell Forging         54.0 0 F              65.9 0 F Circ. Weld Seam WF-447 (Heat 442002)
Nozzle Shell to Intermediate Shell           41.0 0F               25.7 0 F Forging Circ. Weld Seam WF-562 (Heat 442011)
Byron Unit I Surveillance Program             27.00 F Weld Metal Braidwood Unit I & 2 Surveillance             41.00 F Weld Metal Contained in Tables 4-12 and 4-13 is the summary of the fluence factors (FF) used in the calculation of adjusted reference temperatures for the Byron Unit 2 reactor vessel beltline materials for 22 and 32 EFPY.
Calculation of Adjusted Reference Temperature

4-12 TABLE 4-12 Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 22 EPFY Heatup/Cooldown Curves Azimuth                             1/4 T F       1/4T FF          314T F       3/4 T FF 2                              2 (n/cm , E >  1.0 MeV)          (n/cm , E  >1.0 MeV)
Intermediate. Shell Forging                8.47 x 1018        0.953      3.05 x 1018        0.675 49D329-1 /49C297-1 Lower Shell Forging                    8.47 x 1018        0.953      3.05 x 1018        0.675 49D330-1/49C298-1 Nozzle Shell Forging 4P-6107                 2.15 x 1018        0.587      7.75 x i01 7       0.368 Intermediate to Lower Shell Forging             8.35 x 1018        0.949      3.01 x 1018        0.671 Circ. Weld Seam WF-447 (Heat 442002)
Nozzle Shell to Inter. Shell Forging           2.15 x 1018        0.587      7.75 x 1017        0.368 Circ. Weld Seam WF-562 (Heat 442011)
TABLE 4-13 Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 32 EPFY Heatup/Cooldown Curves Azimuth                             1/4 T F       1/4T FF          3/4T F       3/4 T FF (n/cm 2, E > 1.0 MeV)                2 (n/cm , E >1.0 MeV)
Intermediate. Shell Forging               1.24 x 10'9         1.06      4.46 x 1018        0.775 49D329-1/49C297-1 Lower Shell Forging                   1.24 x 1019         1.06      4.46 x IO1         0.775 49D330-1/49C298-1 Nozzle Shell Forging 4P-6107               3.13 x 1018        0.681       1.13 x 1018       0.442 Intermediate to Lower Shell Forging Circ.        1.22 x 1019         1.05      4.40 x 1018        0.772 Weld Seam WF-447 (Heat 442002)                _
Nozzle Shell to Inter. Shell Forging Circ.       3.13 x 1018        0.681       1.13 x 1018       0.442 Weld Seam WF-562 (Heat 442011)
Contained in Tables 4-14 through 4-17 are the calculations of the ART values used for the generation of the 22 and 32 EFPY heatup and cooldown curves.
Calculation of Adjusted Reference Temperature

4-13 TABLE 4-14 Calculation of the ART Values for the 1/4T Location            @ 22 EFPY Reactor Vessel Beltline          Material           Cu%         Ni%        CF(d)     f  22(')      'A-t       '4-t        I       ARTNDT(C)      7      a         M      ARTb)l Region Location           Identilfication                                          EFPY      f(x 10")       FF (x 10"9)
Intermediate Shell                [49D329/            0.01        0.70       20.0        1.41       0.847      0.953      -20          19.1       0       9.5      19.1        18 Forging                         49C297]-1-1                                                                                       I Lower Shell Forging               [49D330/            0.06        0.74       37.0        1.41       0.847      0.953      -20         35.3        0      17.0      34.0       49 49C298]1--                                                                                                                                         l Lower shell Forging                                                          18.7        1.41       0.847      0.953      -20          17.8       0       8.5      17.0        15
  -e using S/C Data                                                                  I Inter. to Lower Shell              WF-447             0.04        0.63       54.0        1.39       0.835      0.949       10         51.2        0      25.6      51.2       112 Circ. Weld Metal                                                                                                                                                                   l Inter. to Lower Shell                                                        65.9        1.39       0.835      0.949       10         62.5        0      14.0      28.0       101 Circ. Weld Metal
  - using S/C Data                                                                                                                                                                   l Nozzle Shell Forging               4P-6107            0.05        0.74       31.0       0.358       0.215      0.587       10          18.2       0      9.1       18.2       46 Nozzle Shell to Inter.             WF-562             0.03        0.67       41.0       0.358       0.215      0.587       40         24.1        0      12.0      24.1       88 Shell Circ. Weld Metal                                                                          .                                                                                  l Nozzle Shell to Inter.                                                       25.7       0.358       0.215      0.587       40          15.1       0       7.5      15.1       70 Shell Circ. Weld Metal e using S/C Data                            -
NOTES:
(a) Fluence, f, is based upon fi,, (10X9 n/cm2, E>1.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c) ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Byron I Weld (WF-336, heat #442002) and the Byron 2 Weld (WF-447, I leat # 442002).
The CF for the Nozzle Shell to Inter. Shell Circ. Weld is integrated betveen Byron I Weld (WF-501, heat # 442011) and the Braidwood I & 2 Welds (WF-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-14 TABLE 4-15 Calculation of the ART Values for the 3/4T Location @ 22 EFPY Reactor Vessel Beltline           Material           Cu%         NI%        CFd)     f    22&deg;       7/4-t       7/.-t       I       ARTNDTOc)     a       CA        M      ARTO)
Region Location           Identirication                                           EFPY      f(x 10')        FF (X 10'9)
Intermediate Shell                 [49D329/            0.01       0.70       20.0        1.41       0.305       0.675      -20          13.5       0       6.7      13.5        7 Forging                         49C297-1l-1 Lower Shell Forging               [49D330/            0.06       0.74       37.0        1.41       0.305       0.675      -20         25.0        0      12.5      25.0       30 49C2981-1-1 Lower shell Forging                                                          18.7       1.41       0.305       0.675      -20          12.6       0      6.3       12.6        5
  -+ using S/C Data                                              _
Inter. to Lower Shell              WF-447             0.04       0.63       54.0        1.39       0.301       0.671       10         36.2        0      18.1      36.2       82 Circ. Weld Metal Inter. to Lower Shell                                                       65.9        1.39       0.301       0.671       10          44.2       0      14.0      28.0       82 Circ. Weld Metal
  - using S/C Data Nozzle Shell Forging                4P-6107            0.05       0.74       31.0       0.358      0.0775       0.368       10          11.4       0       5.7      11.4       33 Nozzle Shell to Inter.             WF-562             0.03       0.67       41.0       0.358      0.0775       0.368       40          15.1       0      7.5       15.1       70 Shell Circ. Weld Metal       _
Nozzle Shell to Inter.                                                      25.7       0.358      0.0775       0.368       40          9.5        0      4.7        9.5       59 Shell Circ. Weld Metal
   - using S/C Data NOTES:
(a) Fluence, E,is based upon f , (10'9 n/cm 2, l>1.0 MeV).
(b) ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c) ARTNDT = CF
* FF (d) The CF for the Inter. to Lower Shell Circ. Weld is integrated betwcen the Byron I Weld (WF-336, heat # 442002) and thc Byron 2 Weld (WF-447, Heat # 442002).
The CF for the Nozzle Shell to Inter. Shell Circ. Weld is integrated between Byron I Weld (WF-501, heat # 442011) and the Braidwood 1 & 2 Welds (WF-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-15 TABLE 4-16 Calculation of the ART Values for the 1/4T Location @ 32 EFPY Reactor Vessel Beitline             Material         Cu%         Ni%        CFd)     f @ 32(s)      Y4-t          4-t       I      ARTNDT(C)     CT         A       M      ARTO)
Region Location              Identification                                        EFPY      f(x 10")        FF (x 10")
Intermediate Shell                  [49D329/          0.01       0.70       20.0        2.06        1.24       1.060      -20         21.2        0       10.6     21.2       22 Forging                           49C2971-1-1 Lower Shell Forging                 [49D330/          0.06       0.74        37.0       2.06        1.24       1.060      -20         39.2        0       17.0     34.0       53 49C2981-1-1I Lower shell Forging                                                          18.7       2.06        1.24       1.060      -20         19.8        0       8.5      17.0        17
  -+ using S/C Data Inter. to Lower Shell                WF-447           0.04       0.63       54.0        2.03        1.22        1.05       10         56.7        0      28.0      56.0       123 Circ. Weld Metal                _
Inter. to Lower Shell                                                        65.9       2.03        1.22        1.05       10         69.2        0       14.0     28.0       107 Circ. Weld Metal
  -+ using S/C Data Nozzle Shell Forging                 4P-6107          0.05       0.74       31.0       0.522       0.313       0.681       10         21.1        0       10.6     21.1       52 Nozzle Shell to Inter.               WF-562           0.03       0.67       41.0       0.522       0.313       0.681       40         27.9        0       14.0     27.9       96 Shell Circ. Weld Metal Nozzle Shell to Inter.                                                      25.7       0.522       0.313       0.681       40         17.5        0       8.8      17.5       75 Shell Circ. Weld Metal
   - using S/C Data                                                   .
NOTES:
(a)    Fluence, f, is based upon fw(l0"9 n/cm2, E>1.0 MeV).
(b)    ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c)    ARTNDT = Cr
* FF (d)    The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Byron I Weld (WF-336, heat # 442002) and the Byron 2 Weld (WF-447, Heat # 442002).
The CF for the Nozzle Shell to Inter. Shell Circ. Weld is integrated between Byron I Weld (WF-501, heat # 442011) and the Braidwood I & 2 Welds (WF-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-16 TABLE 4-17 Calculation of the ART Values for the 3/4T Location           @ 32 EFPY 3                                                    M     ART(b)
Reactor Vessel Beltline             Material           Cu%       Ni%        C        fFd)32(')      3/-t         /-t        I      ARTNDT(       CT[
Region Location             Identirication                                        EFPY       f(x 10")        FF (x 10"9)
Intermediate Shell                   [49D329/            0.01     0.70       20.0       2.06         .446       .775       -20         15.5        0         7.8     15.5       11 Forging                            49C2971-1 -1 Lower Shell Forging                  [49D330/            0.06     0.74       37.0       2.06         .446       .775       -20         28.7        0         14.3    28.7      37 49C2981-1-1                                                                                                                                      l Lower shell Forging                                                          18.7      2.06         .446        .775      -20         14.5        0         7.2     14.5       9 l
 -* using S/C Data Inter. to Lower Shell                WF-447             0.04     0.63       54.0       2.03         .440       .772        10         41.7        0        20.8     41.7      93 Circ. Weld Metal Inter. to Lower Shell                                                       65.9       2.03         .440       .772        10         50.9        0         14.0    28.0      89 Circ. Weld Metal
  -) using S/C Data Nozzle Shell Forging                 4P-6107            0.05     0.74       31.0       0.522       0.113       0.442       10         13.7        0         6.9     13.7      37 Nozzle Shell to Inter.               WF-562             0.03     0.67       41.0       0.522       0.113       0.442       40         18.1        0         9.1     18.1      76 Shell Circ. Weld Metal Nozzle Shell to Inter.                                                      25.7       0.522       0.113       0.442       40         11.4        0         5.7     11.4      63 Shell Circ. Weld Metal
   - using S/C Data NWF ES:
(a)    Fluence, f, is based upon f,,ur(1019 n/cm 2 , E>1.0 MeV).
(b)    ART = I + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)
(c)    ARTNOT = CF
* FF (d)    The CF for the Inter. to Lower Shell Circ. Weld is integrated between the Byron I Weld (WF-336, heat # 442002) and the Byron 2 Weld (WF-447, I feat # 442002).
The CF for the Nozzle Shell to Inter. Shell Circ. Weld is integrated between Byron I Weld (WF-501, heat # 442011) and the Braidwood I & 2 Welds (WF-562, heat # 442011).
Calculation of Adjusted Reference Temperature

4-17 The girth weld WF-447 and the nozzle shell forging 4P-6107 are the limiting beltline materials for all heatup and cooldown curves to be generated. The ART value associated with these materials will be used in all three sets of curves. The girth weld ART will be used when generating curves for Code Case N-588 (ie. Circ. Flaw), otherwise, the nozzle shell forging ART will be used. The ART associated with the limiting axial material is considered to determine if this case would be more conservative or overlap the circ. flaw curves. Contained in Tables 4-18 and 4-19 is a summary of the limiting ARTs to be used in the generation of the Byron Unit 2 reactor vessel heatup and cooldown curves.
TABLE 4-18 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location for 22 EFPY Material                                       22 EFPY 1/4T ART                 3/4T ART Intermediate Shell Forging                           18                       7
[49D329/49C297]-1-1 Lower Shell Forging                                 49                       30
[49D330/49C298]-1 -I
          - Using Surveillance Data(a)                         15                       5 Circumferential Weld WF-447                         112                      82
          - Using Surveillance Data                           101                      82 Circumferential Weld WF-562                          88                      70
          - Using Surveillance Data from                      70                       59 Braidwood I and 2 Nozzle Shell Forging 4P-6107                       46(m)33(a)
NOTES:
(a) These ART values were used to calculate the Heatup and cooldown curves in Figure 5-1 and 5-2. They were generated using the '96 App. G Methodology, and were more conservative than the curves generated using the limiting circumferential flaw ART values and Code Case N-588 Methodology.
Calculation of Adjusted Reference Temperature

4-18 TABLE 4-19 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location for 32 EFPY Material                                      32 EFPY 1/4T ART                  3/4T ART Intermediate Shell Forging                          22                        11
[49D329/49C297]-l-1 Lower Shell Forging                                 53                        37
[49D330/49C298]-l -I
        - Using Surveillance Data(a)                        17                        9 Circumferential Weld WF-447                         123                       93
        - Using Surveillance Data                          107                        89 Circumferential Weld WF-562                        96                        76
        - Using Surveillance Data from                      75                        63 Braidwood I and 2 Nozzle Shell Forging 4P-6107                       52(')                     37('
NOTES:
(a)     These ART values were used to calculate the Heatup and cooldown curves in Figure 5-3 and 5-4. They were generated using the '96 App. G Methodology, and were more conservative than the curves generated using the limiting circumferential flaw ART values and Code Case N-588 Methodology.
Calculation of Adjusted Reference Temperature

S-I 5         HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beitline region using the methods discussed in Section 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-NP-A'l 1 , dated January 1996.
Figures 5-1 through 5-4 present the 22 and 32 EFPY heatup and cooldown curves (without margins for possible instrumentation errors) for a heatup rate of I 000 F/hr and cooldown rates of 0, 25, 50 and 100 0 F/hr using the 1996 Appendix G methodologyt 61 and Code Case N-588 151 , respectively. The heatup and cooldown curves that are presented herein are actually are curves generated using the 1996 App. G methodology with the lower arialflavARTvalue with exception to the last two temperatures of the 22 EFPY I 000 F/hr. heatup curve and the last four temperatures of the 32 EFPY I00 0 F/hr heatup curve. The reason is, these curves are more conservative (with exception noted) than the curves generated using Code Case N-588 methodology with the higher circ.flawARTvalue. This is true throughout the entire temperature range, including criticality.
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 through 5-4. This is in addition to other criteria which must be met before the reactor is made critical, as discussed in the following paragraphs.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-land 5-3 (for the specific heatup rate being utilized). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-640 121 (approved in February 1999) as follows:
1.5K Im<Kc                                 (1)
: where, Klm is the stress intensity factor covered by membrane (pressure) stress, K1, = 33.2 + 20.734 e1002 (T-RTNDT)1 T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 3. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40&deg;F higher than the minimum permissible temperature in the corresponding pressure-temperature curve Heatup and Cooldown Pressure-Temperature Limit Curves

5-2 for heatup and cooldown calculated as described in Section 3 of this report. The minimum temperatures for the inservice hydrostatic leak test for the Byron Unit 2 reactor vessel at 22 and 32 EFPY are 1061F and 112'F at 2485 psig 1996 App. G Methodology. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40IF higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 5-1 through 5-4 define all of the above limits for ensuring prevention of non-ductile failure for the Byron Unit 2 reactor vessel. The data points for the heatup and cooldown pressure-temperature limit curves shown in Figures 5-1 through 5-4 are presented in Tables 5-1 through 5-4, respectively.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 22 EFPY:                   1/4T, 101 0 F (N-588) & 460 F ('96 App. G) 3/4T, 82 0 F (N-588) & 33 0 F ('96 App. G) 2500 2250 2000 1750 2   1500 II 1250 I   1000 750 500 250 0
0       50     100     150    200      250       300    350   400      450    500 550 Moderator Temperature (Deg. F)
FIGURE 5-1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of lOOF/hr)
Applicable to 22 EFPY Using Code Case N-640 & 1996 Appendix G (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 22 EFPY:                  1/4T, 101 0 F (N-588) & 460 F ('96 App. G) 3/4T, 82 0 F (N-588) & 331F ('96 App. G) 2500 2250 2000 1750 K 1500 2
1  1250
* 1000
  .U 750 500 250 0
0       50      100     150     200     250      300     350   400      450    500 550 Moderator Temperature (Deg. F)
FIGURE 5-2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I 000 F/hr) Applicable to 22 EFPY Using Code Case N-640 & 1996 Appendix G (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-5 TABLE 5-1 Byron Unit 2 Heatup Data at 22 EFPY Using Code Case N-640 & 1996 Appendix G Methodology*
(Without Margins for Instrumentation Errors) 100 Heatup                Critical. Limit               Leak Test Limit T             P             T               P              T            P 60            0            106              0             89          2000 60          1082           106            1133            106         2485 65           1133           110            1186 70          1186           115            1197 75          1197           120            1202 80          1202           125            1215 85          1215           130            1235 90           1235           135            1263 95           1263           140            1299 100          1299           145            1342 105          1342           150            1393 110          1393           155            1452 115          1452           160            1520 120          1520           165            1598 125          1598           170            1686 130          1686           175            1784 135          1784           180            1895 140          1895           185            2018 145          2018           190            2155 150          2155           195            2304 155          2304           200            2428 160          2428
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall with exception to the last two temperatures, which are limited by the Circ-Flaw ART with Code Case N-588.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-6 TABLE 5-2 Byron Unit 2 Cooldown Data at 22 EFPY Using Code Case N-640 & 1996 App. G Methodology*
(Without Margins for Instrumentation Errors)
Steady State                  25F                          50F                 10OF T            P           T             P             T              P      T            P 60            0           60            0             60             0      60           0 60          1102 65          1154 70          1212 75          1276 80          1347 85          1425 90          1512 95          1607 100         1713 105         1829 110         1958 115        2101 120        2258 125         2433
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-7 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:                  1/4T, 107 0 F (N-588) & 52'F ('96 App. G) 3/4T, 89 0 F (N-588) & 370 F ('96 App. G) 2500 2250 2000 1750 K 1500 3
1   1250 3 1000 U0 750 500 250 0
0      50      100     150     200      250      300     350   400     450    500 550 Moderator Temperature (Deg. F)
FIGURE 5-3 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of I00FT/r)
Applicable to 32 EFPY Using Code Case N-640 & 1996 Appendix G (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-8 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:                  I/4T, 107 0 F (N-588) & 521F ('96 App. G) 3/4T, 89 0 F (N-588) & 370 F ('96 App. G) 2500 2250 2000 1750 2     1500 2
1    1250 I-
  =C.) 1000 750 500 250 0
0       50      100    150      200    250      300    350   400     450     500 550 Moderator Temperature (Deg. F)
FIGURE 5-4 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I 00&deg;F/hr) Applicable to 32 EFPY Using Code Case N-640 & 1996 Appendix G (Without Margins for Instrumentation Errors)
Heatup and Cooldown Pressure-Temperature Limit Curves

5-9 TABLE 5-3 Byron Unit 2 Heatup Data at 32 EFPY Using Code Case N-640 & 1996 App. G Methodology*
(Without Margins for Instrumentation Errors) 100 Heatup                   Critical. Limit               Leak Test Limit T             P                T                P       I     T             P 60            1030             112               0            95           2000 65            1078             112             1078           112          2485 70            1128             112             1128 75            1148             115             1148 80            1152             120             1152 85            1162             125             1162 90            1180             130             1180 95            1205             135             1205 100            1237             140             1237 105            1276             145             1276 110            1323             150             1323 115           1377             155             1377 120           1440             160             1440 125            1511             165             1511 130           1592             170             1592 135           1682             175             1682 140            1784             180             1784 145           1897             185             1897 150           2023             190             2023 155           2120             195             2120 160           2227            200              2227 165           2347            205              2347 170           2480            210              2480
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall with exception to the last four temperatures, which are limited by the Circ-Flaw ART with Code Case N-588.
Heatup and Cooldown Pressure-Temperature Limit Curves

5-10 TABLE 5-4 Byron Unit 2 Cooldown Data at 32 EFPY Using Code Case N-640 & 1996 App. G Methodology*
(Without Margins for Instrumentation Errors)
Steady State            2SF                           5OF                  1OOF T           P             T              P              T               P    T            P 60           0            60             0              60              0    60           0 60         1045           60            1036            60           1033 65         1092           65            1088 70         1143 75         1200 80         1263 85         1332 90         1409 95         1494 100         1587 105         1691 110         1805 115         1932 120         2071 125         2226 130         2396
* Note: The computer run for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall.
Heatup and Cooldown Pressure-Temperature Limit Curves

6-1 6      REFERENCES I      Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U.S.
Nuclear Regulatory Commission, May, 1988.
2      ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February 26, 1999.
3      Code of Federal Regulations, 10 CFR Part 50, Appendix G."Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
4      WCAP-1 6143-P, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2", K.R. Hsu, November 2003.
5      ASME Code Case N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels", Section XI, Division 1,Approved December 12, 1997.
6      ASME Boiler and Pressure Vessel Code, Section XI, "Rule for Inservice Inspection of Nuclear Power Plant Components", Appendix G; "Fracture Toughness Criteria for Protection Against Failure", December 1995.
7      WCAP-15183, Revision 1, "Commonwealth Edison Company Byron Unit I Surveillance Program Credibility Evaluation", T.J. Laubham, November 2003.
8      WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D. Andrachek, et al., January 1996.
9       1989 Section 111, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Material for Vessels".
10      SAE-REA-00-546, "Reactor Vessel Neutron Exposure Projections for the Byron/Braidwood Uprating", J.D. Perock, January 12, 2000.
11      WCAP-1 5180, Revision 1, "Commonwealth Edison Company Byron Unit 2 Surveillance Program Credibility Evaluation", T.J. Laubham, November 2003.
12      WCAP-14824, Revision 2, "Byron Unit I Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration For Byron and Braidwood", T. J. Laubham, et al., November 1997. Ref. Errata letter CAE-97-233, CCE-97-316, "Transmittal of Updated Tables to WCAP-I 4824 Rev. 2" 13      WCAP-15178, "Byron Unit 2 Heatup and Cooldown for Normal Operation", T. J. Laubham, June 1999.
References

6-2 14     WCAP-15 124, "Byron Unit I Heatup and Cooldown for Normal Operation", T. J. Laubham, November 1998.
15     WCAP-15373, Revision 2, "Braidwood Unit 2 Heatup and Cooldown for Normal Operation", T.
J. Laubham, November 2003.
16     CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1996.
17     WCAP-15368, "Commonwealth Edison Company Braidwood Unit 2 Surveillance Program Credibility Evaluation", T.J. Laubham, March 2000.
18     WCAP-15123, Revision 1, "Analysis of Capsule W from Commonwealth Edison Company Byron Unit I Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et. al., January 1999.
19     WCAP-15176, "Analysis of Capsule X from Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et. al., March 1999.
20     WCAP-1 5316, Revision 1, "Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit I Reactor Vessel Radiation Surveillance Program," Ed Terek, et. al., December 1999.
21     WCAP-1 5369, "Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et. al., March 2000.
22     WCAP-14940, "Byron Unit 2 Heatup and Cooldown Limit Curves For Normal Operation", T. J.
Laubham, October 1997. Ref. Errata letters CAE-97-210 & 232, CCE-97-289 & 316, "Transmittal of Updated Tables to WCAP-14940 and WCAP-14970" 23     NDIT No. MSD-98-044, "Best Estimate Chemistry Values for Reactor Pressure Vessel Beltline Weld Heat Number 442002", Dated December 1998.
References

A-I APPENDIX A Thermal Stress Intensity Factors (K1 t)
Revision 0 Thermnal        Intensity Factors Stress Intensity Thermal Stress            Factors                                      Revision 0

A-2 TABLE Al K.. Valuies for I 00 0 F/hr Meatim Crurve (22 FFPY)
Vessel Temperature            1/4T Thermal        Vessel Temperature     3/4T Thermal Water         @ 114T Location for                 Stress     i     3/4T Location for       Stress Temp.           100&deg;F/hr Heatup            Intensity Factor       100'F/hr Heatup     Intensity Factor (OF)                  (OF)              (KSI SQ. RT. IN.)               ( 0 F)     (KSI SQ. RT. IN.)
PT Curves are Limfited by the 3/T Location for the Entire Curve 60                  56.01                     -0.9945                 55.05             0.4767 65                  58.62                     -2.4413                 55.31             1.4422 70                  61.70                     -3.6836                 56.01             2.4209 75                  65.01                     -4.8600                 57.20             3.3365 80                  68.58                     -5.8718                 58.80             4.1525 85                  72.27                     -6.7937                 60.80             4.8806 90                  76.14                     -7.5912                 63.14             5.5230 95                   80.11                    -8.3181                 65.77             6.0965 100                   84.21                    -8.9517                 68.68             6.6042 105                   88.40                    -9.5267                 71.81             7.0564 110                  92.68                   -10.0310                 75.15             7.4589 115                  97.04                    -10.4906                 78.67             7.8204 120                 101.46                   -10.8960                 82.35             8.1440 125                  105.95                   -11.2667                 86.16             8.4352 130                 110.48                   -11.5953                 90.10             8.6969 135                 115.06                   -11.8972                 94.14             8.9336 140                 119.68                   -12.1661                 98.28             9.1474 145                 124.34                   -12.4147                 102.49            9.3417 150                 129.03                   -12.6374                 106.78            9.5183 155                 133.75                   -12.8446                 111.14            9.6799 160                 138.48                   -13.0315                 115.55            9.8277 Vessel Radius to the 'AT and %T Locations are as follows:
* I/4T Radius = 88.750"
* 3/4T Radius = 93.000" Theral tres Inensty actos                                               Rvison Therrnal Stress Intensity Factors                                                                   Revision 0

A-3 TABLE A2 K21 Values for I 00WF/hr Cooldown Curve (22 EFPY)
Vessel Temperature 100WF/hr Cooldown Water         @ 1/4T Location for 1/4T Thermal Stress Temp.         100&deg;F/hr Cooldown        Intensity Factor (OF)                 (OF)           (KSI SQ. RT. IN.)
125                149.78                 15.3518 120                144.70                 15.2854 115                139.61                 15.2187 110                134.53                 15.1525 105                129.45                 15.0860 100                124.36                 15.0200 95                 119.28                 14.9538 90                 114.19                 14.8881 85                 109.11                 14.8221 80                 104.02                 14.7566 75                 98.94                  14.6909 70                 93.86                  14.6257 65                 88.77                  14.5603 60                 83.69                  14.4946 NOTE: From T 601F to 1251F, the I 00OF/hr Cooldown Rate is limited by the Steady State Condition.
Revision 0 Thermal Stress Intensity Thermnal                 Factors Intensity Factors                                                    Revision 0

A-4 TABLE A3 KI, Values for I 00&deg;F/hr Heatup Curve (32 EFPY)
Vessel Temperature           1/4T Thermal      Vessel Temperature      3/4T Thermal Water        @ 1/4T Location for                Stress      @ 314T Location for         Stress Temp.           100F/hr Heatup           Intensity Factor     100OF/hr Heatup      Intensity Factor (OF)                  (OF)             (KSI SQ. RT. IN.)            (OF)         (KSI SQ. RT. IN.)
PT Curves are Limited by the %T Location for the Entire Curve 60                   56.01                   -0.9945              55.05               0.4767 65                   58.62                   -2.4413              55.31               1.4422 70                   61.70                   -3.6836              56.01               2.4209 75                   65.01                   4.8600               57.20               3.3365 80                   68.58                   -5.8718              58.80               4.1525 85                   72.27                   -6.7937              60.80               4.8806 90                   76.14                   -7.5912              63.14               5.5230 95                   80.11                   -8.3181              65.77               6.0965 100                   84.21                   -8.9517              68.68               6.6042 105                   88.40                   -9.5267              71.81               7.0564 110                   92.68                  -10.0310              75.15               7.4589 115                   97.04                  -10.4906              78.67               7.8204 120                  101.46                  -10.8960              82.35               8.1440 125                  105.95                  -11.2667              86.16               8.4352 130                  110.48                  -11.5953              90.10               8.6969 135                  115.06                  -11.8972              94.14               8.9336 140                  119.68                  -12.1661              98.28               9.1474 145                 124.34                  -12.4147              102.49              9.3417 150                 129.03                  -12.6374              106.78              9.5183 155                 133.75                  -12.8446              111.14              9.6799 160                 138.48                  -13.0315              115.55              9.8277 165                 143.25                  -13.2066              120.01              9.9638 170                 148.02                  -13.3657              124.52              10.0892 Revision 0 Stress Intensity Thermal Stress           Factors Intensity Factors                                                                 Revision 0

A-5 TABLE A4 K1, Values for 1000 F/hr Cooldown Curve (32 EFPY)
Vessel Temperature 100lF/hr Cooldown Water        @ 1/4T Location for 1/4T Thermal Stress Temp.         100&deg;F/hr Cooldown        Intensity Factor (OF)                  OF)           (KSI SQ. RT. IN.)
130               154.87                 15.4187 125               149.78                 15.3518 120               144.70                 15.2854 115               139.61                 15.2187 110               134.53                 15.1525 105               129.45                 15.0860 100               124.36                 15.0200 95                119.28                 14.9538 90                114.19                 14.8881 85                109.11                 14.8221 80                104.02                 14.7566 75                 98.94                 14.6909 70                 93.86                 14.6257 65                 88.77                 14.5603 60                 83.69                 14.4946 NOTE: From T 60 0 F to 1301F, the 100 0F/hr Cooldown Rate is limited by the lower rates and/or the Steady State Condition.
Revision 0 Thermal Stress Intensity Factors Thermnal                 Factors                                                                 Revision 0

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points The WCAP reports 1 provide the necessary information to validate the proper application of previously approved Cases of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Nuclear Code Cases N-640 2 and N-588 3. However, as discussed in , these reports also apply the exemption from the flange temperature requirements of 10 CFR 50 Appendix G. This regulation states that the metal temperature at the closure flange regions must exceed the material unirradiated initial reference nil-ductility transition temperature (RTNDT) by at least 120&deg; F for normal operation when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure. For the Braidwood Station and Byron Station units, this 20 percent pressure value is 621 psig.
To facilitate review, a comparison of the current Braidwood and Byron Station heat up, cooldown, steady state, and criticality limit data points to those of the 22 effective full power year (EFPY) curves with the flange temperature requirement of Appendix G restored and Code Cases N-640 and N-588 applied is provided. The purpose of this comparison is to demonstrate that the current Braidwood and Byron Station curves are more restrictive than the 22 EFPY limits using the approved methodology of ASME Code Cases N-640 and N-588. Consequently, it is justifiable to extend the applicability time of the current Braidwood and Byron Station P-T curves an additional two EFPY for each unit as provided in Attachment 6.
To clarify, EGC does not propose to adopt P-T curves based on Code Cases N-640 or N-588 or modify the data points in any way, but rather EGC intends to extend the current P-T curves by an additional two EFPY. The current P-T curves based on previously approved methodology, i.e., 1996 ASME Section Xl, Appendix G methodology and ASME Code Case N-514, are more conservative and operationally more restrictive. The low temperature overpressure protection (LTOP) system setpoints, the LTOP enable temperature, the reactor vessel boltup temperature, and the Reactor Vessel Minimum Pressurization Temperature are unchanged fr'om the current Braidwood and Byron Stations PTLRs. Since the Braidwood Station and Byron Station P-T limit curves use Kla to determine allowable pressure in the vessel, the LTOP system will continue to limit the maximum vessel pressure to 110% of the allowable pressure.
1 WCAP-15365,    Revision 2, Braidwood Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, WCAP-1 5373, Revision 2. Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, WCAP-15391, Revision 1, Byron Unit I Heatup and Cooldown Limit Curves for Normal Operation, and WCAP-15392, Revision 1, Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation.
2 Code Case N-640, 'Altemative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1" 3 Code Case N-588, Altemative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels Section XI, Division 1" Page 1 of 9

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points Braidwood Unit I 100 Degree  Heat Up            .          Criticality Limit                                 Leak Test Limit 22 EFPY Current PTLR 22 EFPY 22 EFPY Current               Current       2 EFPY 22 EFPY    Current     Current Temperature Pressure Pressure Temps Pressur PTLR Tern PTLR Pressure Temps' Pressures PTLR Temps PTLR Pressures 60         0             0    103        0           21              .0           86    2000           188        2000 60 565.09          565.09     103     1110           210            611.83       103    2485           210        2485 65    565.09       565.09     105     1163           210            597.56 70    565.09       565.09     110     1218           21             585.60 75    565.09       565.09     115     1222           210            576.77 80    565.09       565.09     120     1229           210            570.35 85    565.09       565.09     125     1243           210            566.61 90    565.09       565.09     130     1264           210            565.09 95    565.09       565.09     135     1294           210            565.87 100    565.87       565.87     140     1331           210            568.69 105    568.69       568.69     145     1376           210            573.56 110    573.56       573.56     150     1430           210            580.30 115      1492       580.30     155     1492           210            588.84 120      1563       588.84     160     1563           210            599.36 125      1644       599.36     165     1644           210            611.78 130      1736       611.78     170   - 1736           210            626.07 135      1838       626.07     175     1838           210            642.16 140      1954       642.16     180     1954           210            660.36 145      2082       660.36     185     2082           210            680.59 150      2225       680.59     190     2225           210            702.80                          I 155      2384       702.80     195     2384           210            727.33 160                 727.33                            210            754.07            '             'I 165                 754.07                            210            783.17           .
170                 783.17                            215            814.98 175                 814.98                            220            849.37 180                 849.37                            22             886.54 185                 886.54                            23             926.73           .
190                 926.73                            235            970.11                             I 195                 970.11                            24            1016.91                             i 200                1016.91                            245           1067.33                              \
205                1067.33                            25            1121.63 210                1121.63                            255           1180.01 215                1180.01                            260           1242.62 220                1242.62                            265           1309.84 225                1309.84                            270           1382.03 230                1382.03                            275           1459.45 235                1459.45                            280           1542.27 240                1542.27                            285           1630.97 245                1630.97                            290           1726.05 250                1726.05                             295          1827.80 255                1827.80                             300          1936.51 260                1936.51                             305          2052.39 265                2052.39                             310          2176.33 270                2176.323                            315          2308.42 275                2308.42                             320          2449.09 2801               2449.09 _                               I I         l             l      l         l _           Il _             l Page 2 of 9

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points Braidwood Unit 1 Stead State                  25 Degree F Cooldown            50 Degree F Cooldown        100 Degree F Cooldown Current                         Current                        Current                      Current 22 EFPY PTLR                    22 EFPY PTLR                   22 EFPY PTLR                 22 EFPY PTLR Temperature     Pressure Pressure Temperature Pressure Pressure Temperature Pressure Pressure Temperature , Pressure Pressure 60        0     0.00            60        0        0           60        0        0         60        0        0 60     .620   620.27            60      577   577.45           60      534  534.28          60      447   446.98 65      621   621.00            65      591   590.68           65      549  548.52          65      464   463.79 70      621   621.00            70      605   605.03           70      564  563.98          70      482   481.93 75      621   621.00            75      621   620.51           75      581  580.67          75      501   501.49 80      621   621.00            80      621   621.00           80      599  598.51          80      523   522.68 85      621   621.00            85      621   621.00           85      618  617.90          85      546   545.50 90      621   621.00            90      621   621.00           90      621  621.00          90      570   570.23 95      621   621.00            95      621   621.00           95      621  621.00          95      597   596.83 100      621   621.00          100       621   621.00          100      621  621.00        100       621   621.00 105      621   621.00          105       621   621.00          105      621  621.00        105       621   621.00 110      621   621.00          110       621   621.00          110      621  621.00        110       621   621.00 110     2042   795.92          110      2042   766.92          110     2042  739.27        110      2042   690.04 115     2194   821.55           115     2194   794.59          115     2194  769.53        115      2194   726.24 120     2361   849.00           120     2361   824.45          120     2361  801.97        120      2361   765.12 125            878.42           125            856.54          125           836.87        125             807.07 130 .          910.25           130            890.97          130           874.41        130             852.23 135            944.34           135            928.00          135           915.03        135             900.91 140            980.89           140            967.79          140           958.57        140             953.33 145           1020.15           145 _        1010.84           145          1005.42        145            1009.81 150 .         1062.35           150          1056.88           150          1055.76        150 155           1107.92           155          1106.38           155                         155 160           1156.42           160                            160                         160 165           1208.78           165                            165                          165 170           1265.05           170                            170                          170 .
175           1325.37           175          _                 175                          175 180           1390.04           180                            180                          180 185           1459.41           185                            185                          185 190           1533.55           190                            190                          190 195           1613.49           195                            195                          195 200           1699.01          200                             200                         200 205           1790.55          205                             205                         205 210           1888.61          210                             210                         210 215           1993.61           215                            215  _                      215 220           2105.69           220                            220                         220 225           2225.77           225                            225                         225 230           2353.75           230     =                      230                         230 Page 3 of 9

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points Braidwood Unit 2 100 Degree Heat Up          l              Criticallty Limit              l                  Leak Test Limit 22 EFPY Current PTLR 22 EFPY 22 EFPY Current            Current       22 EFP 22 EFPY        Current      Current emperature Pressure Pressure        Temps Pressure PTLR Temr PTLR Pressure Temps Pressures PTLR Temps PTLR Pressure 60         0            0       122       0           207             0        105      2000           186          2000 60       617          617       122    1000           207           621        122      2485           207          2485 65       617          617       122    1013           207           696             .
70       617          617       125    1018           207           715      .
75       617          617       130    1030           207           736t 80       617          617       135    1046           207           760 85       617          617       140    1069           207           786                                    .
90       617          617       145    1097           20            815 95       617          617       150    1131           21            846 100       617          617       155    1171           215           880 105       619          619       160    1218           220           917 110       621          621       165    1272           225           957               .
115       621          621       170    1333           230         1000 120       621          621       175    1402           235         1047 125       621          621       180    1479           240         1097 130       621          621       185    1566           245         1152 135       621          621       190    1663           250         1210 140       621          621       195    1770           255         1273 140      1479          696       200    1890           260         1341            .
145      1566          715       205    2023           265         1415 150      1663          736       210    2170           270         1493 .
155      1770          760       215    2332           275         1578 160      1890          786                             280         1669 165      2023          815                             285         1766                        _
170      2170          846 _                           290         1871       =
175      2332          880            _                295         1984       _
180                    917     =        _              300         2105       =
185                    957                             305         2235                            _
190                   1000                             31          2374                _
195                   1047 200                   1097                                                       .
205                   1152                                                                 .
2101                  1210 215                   1273                                           .
220                   1341     .
225                   1415 230                   1493 235                   1578 240                   1669           1 245                   1766 250                   1871-255                   1984                                        _
260                   2105                                                            _
265                   2235 270                   2374           1                     1             1 Page 4 of 9

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points Braidwood Unit 2 Ste dy State                25 Degree F Cooldown                50 Deae F Cooldown                100 Deg      F Cooldown Current                             Current                           Current                            Current 22 EFPY PTLR                     22 EFPY      PTLR                   22 EFPY    PTLR                    22 EFPY PTLR Temperature Pressure Pressure    Temperature Pressur Pressure        Temperature Pressure Pressure      Temperature Pressure Pressure 60        0         0           6           0           0           60          0         0             60        0          0 60      621       621           60       602          602           60       601        554             60      455        455 65      621       621           65       6161         616           65       619        568             65      471        471 70      621       621           70       621          621           70       621        583             70      489        489 75      621       621           75       621          621           75      '621        599             75      508        508 8t      621       621           80       621          621           80       621        6171            80      529        529 85      621       621           85       621          621            85      621        621             85      552        552 90      621       621            90      621          621            90      621        621             90      576        576 95      621       621           95       621          621            95      621        621             95      603        603 10t       621       621          100       621          621          100       621        621           100       621        621 105       621       621          105       621          621          105       621        621           105       621        621 110       621       621          110       621          621          110       621        621           110       621        621 115       621       621          115       621          621          115       621        621           115       621        621 120       621       621          120       621          621          120       621        621           120       621        621 125       621       621          125       621          621          125       621        6211          125       621        621 130       621       621          130       621          621          130       621        621,          130       621        621 135       621       621          135       621          621          135       621        621           135       621        621 140       621       621          140       621          621          140       621        6211           140      621        621 140      2396      1010          140      2396          991          140      2396        975            140     2396        957 145               105           145                   1034          145                1022             145                1013 150               1092          150                   1080          150                1072    i        150                1074 155               1137          155                   1129          155                1126             155                1137 160               1186          160                   1183          160                1185             160                1186 165               1239          165                   1239          165                12391            165                1239 170               1295          170                   1295          170 .              1295             170                1295 175               1356          175                   1356          175                1356             175                1356 18t               1422          180                   1422          180                1422             180                1422 185               1492          185                   1492          185                1492             185                1492 190               1567          190                   1567          190                1567             190                1567 195               1649          195                   1649          195                1649             195  _             1649 200                1736          200                   1736          200                1736             200                1736 205                1830          205   _               1830          205                 1830            205                1830 210                1931          210                   1931          210                 1931            210                1931 215               2039          215            1      2039          215                2039             215                2039 220               2156          220                   2156          220                2156             220                2156 225               2281          225                   2281          225                2281             225                2281 230               2416          230                   2416          230               .2416             230                2416 Page 5 of 9

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points Byron Unit 1            .       __
100 Degree Heat Up                              Criticality Limit                                 Leak Test Limit Current                                  Current Current                               Current       Current 22 EFPY     PTLR                        22 EFPY        PTLR PTLR           22 EFPY     22 EFPY       PTLR          PTLR Temperature       Pressure    Pressure    22 EFPYTemps   Pressure       Temps Pressure'       Temps     Pressures     Temps       Pressures 6              0            0          163             0        225           0        146        2000         204          2000 60           587         587           163          732         225        587         163        2485         225          2485 6            587         587           163          732         225        588 70           587         587           163          732         225        591 75           587         587           163          732         225        596 81           587         587           163          732         225        60 85           587         587           163          732         225        611 9            587         587           163          732         225        622 95           587         587           163          733         225        634 100            587         587           163          737         225        648 105            587         587           163          743         225        665 110            587         587           163          753         225        683 115            587         587           163          766         225        703 120            588         588           163          782         225        725 125            591         591           165          801         225        750 130            596         596           170          823         225        777 135            602         60            175           849        230        8061 140            611         611           18t           879        235        8381 145           621         621           185           913        240        872 150           621         621           190           951        245        910 155           621         621           195           994        250        950 160           621         621           200         1042         255        994 165           621         621           205         1095         260       1041 170           621         621           210         1155         265       1092 175           621         621           215         1221         270       1147 180           621         621           220         1294          275      1206t 180          1294         750           225         1376          280      1269 185          1376         777           230         1466          285      1338 190          1466         806           235         1566          290      1411 195          1566         838           240         1676          295      1490 200           1676         872           245         1798          300      1575 205           1798         910           250         1932          305      1666 21            193          950           255         2081          310      1764 _
215           2081         994           260         2245          315      1869 220           2245       1041            265         2425          320      1982 225           2425       1092                                      325      2104 230                      1147                                      330      2234 235                      1206                                      335      2374 240                      1269 245                      1338 250                     1411 255                     1490 260                     1575 265                     16661 270                     1764 275                     1869 280                     1982                                    =       -__                                         _
285                     2104                  _   _ _   _  _  _ _   __             _  _  _    _  _ _  _ _   _ _  _     _  _ _  _
290                     2234, 295 _                   2374                1
                                                            - Page6 of 9

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points Byron Unit I Steady State               25 Degr ee F Cooldown            50 Degree F Cooldown             100 Deg ree F Cooldown Current                          Current                          Current                         Current 22 EFPY PTLR                  22 EFPY PTLR                     22 EFPY PTLR                      22 EFPY PTLR Temperature Pressure Pressure Temperature Pressure Pressure Temperature Pressure Pressure Temperature Pressure Pressure 60  _                0        60          0           0         60        0            0         60         0       10 60       613      613         60        561        561          60      509         509          60       402       402 65       621      621         65        572        572          65      520         520          65       414       414 70       621      621         7         582        582          70      531         531          7        427       427 75       621      621         75        594        594          75      544         544          75       442       442 80       621      621         80        607        607          80      557         557          80       458       458 85       621      621         85        620        620          85      572         572          85       475       475 90       621      621         90        621        621          90      588         588          9        494       494 95       621      621         95        621        621          95      605         605          95       514       514 100       621      621        100        621        621         100      621         621         100       535       53 105       621      621        105        621        621         105      621         621         105       559       559 110       621      621        110        621        621         110      621         621         11t       584       584 115       621      621        115        621        621         115      621         621         115       611       611 120       587      621        120        621        621         120      621         621         120       621       621 125       588      621        125        621        621         125      621         621         12'       621       621 130       621      621        130        621        621         130      621         621  .      130       621       621 135       621      621        135        621        621         135      621         6211        135       621       621 140       621      621        140        621        621         140      621         621         140       621       621 145       621      621        145        621        621         145      621         621         145       621       621 15        621      621        150        621        621        -150      621         621         150       621       621 155       621      621        155        621        621         15'      621         621         15'       621       621 16        621      621        16         621        621         16       621         621         160       621       621 165       621      621        165        621        621         165      621         621         165       621       621 170       621      621        170        621        621         170      621         621         170       621       621 175       621      621        175        621        621         175      621         621         175       621       621 18t       621      621        18t        621        621         180      621         621         18t       621       621 180      2361     1207        18        2361       1205         180     2361        1205         18       2361      1205 185               1261        185                               185           .                  18'             .   -
190               1319        190                               190           .                  190 195               1382        195                               195           .                  195 200               1449        200   _                           200                              20 205               1521        205                               205                              205            .    -
210               1599        210                               210                              210           . .
215               1683        215            .                  215                              215 220               1773        220                               220         .                    220            .    -
225               1869        225                               225                              225 230               1973        230                               23                               230 235               2085 240               2205 2451              2334                       .
25t               2473                       .
Page 7 of 9

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byron and Braidwood Station PTLR Data Points
                                                       . Byron Unit 2 100 Degree Heat Up           l               Criticality Limit                  lI        _         Leak Test Limit             '
Current                                Current    Current                                Current    Current 22 EFPY       PTLR       22 EFPY      22 EFPY        PTLR       PTLR        22 EFPY       22 EFPY      PTLR       PTLR Temperature     Pressure    Pressure      Temps      Pressures Temps         Pressure Temps             Pressures Temps         Pressures 6              0          0           106             0       219           0           89          2000         198       2000 60           621       621            106         1133        219         635          106          2485         219       2485 65           621       621            110         1186        219         6741 70           621       621            115         1197        219         660                                  .
75           621       621            120         120         219         650 8C           621       621            125         1215        219         643                          .
85           621       621            130         1235        219         638              ;
90           621       621            135         1263        219         637                        .
95           621       621            140       '1299         219         637 100            621       621            145         134         219         641 105            621       621            150         1393        219         646 110            621       621            155         1452        219         654 115            621       621            160         1520        219         6641 120            621       621            165         1598        219         676 125            621       621            170         1686        219         690 130            621       621            175         1784        219         707 135            621       621            180         1895        219         725 140           621       621            185         2018        219        .746 145      . 621       621            190         2155        219         7701 150           621       621            195         2304        219 .       796 150          2155       707           200          2428        219         824 155          2304       725                                    220         855 .
160           2428       746                                    225         889 .
165                      7701                                   230         926 170                      796                                    235         966 175                      824      '                             240       1010 18C                      855                                    245       1057 180                      889                                    250       1107 .
185                      9261                                   255       1162 190                      966                                    260       1221 195                     1010                                    265       1285 200                     1057                                    270       1353 205                     1107                                    275       1427 210                     1162                                    280       1506 _               _
215                     1221                                    285       1591 _               _           _
220                     1285                                    290       1683 225                     1353                                    295       1781 230                     1427                                    300       18871 235                     1506                                    305       2001 240                     15911                                   310       2123                 _
245                     1683                                    315       2254 250                     1781                                    320       2395 255                     1887 260                     2001 ;
265                     2123 270                     22541 275 _                   2395                                           . ,
Page 8 of 9

ATTACHMENT 5 Comparison of 22 EFPY Data Points and Current Byr6h and Braidwood Station PTLR Data Points Byron     Unit 2 Steady State                25 Degree F Cooldown               50 Degree F Cooldown           100 Degree F Cooldown Current                          Current                         Current                         Current 22 EFPY     PTLR                22 EFPY PTLR                      22 EFPY   PTLR                 22 EFPY    PTLR Temperature    Pressure Pressure Temperature Pressure Pressure Temperature Pressure Pressure Temperature Pressure Pressure 60          0 _                 60         0   _                  60        0          0I         60        0        .0 60       621        621         60       574        574           60       523      523           60      418       4181 65      621        621         65       585        585           65       534      534           65      431       431 70       621        621         70       597.       597           70       547      547           70      446       446 75       621        621         75       610        610           75       561      561           75      462       462 80       621        621         80       621        621           8C       576      576           80      480       480 85      621        621         85       621        621           85       592      59,           85      498       498 90      621        621         90       621        621           90       609      609           90      519       519 95      621        621         95       621        621           95       621      621           95      541       541 100       621        621        100       621        621          100       621      621          10-      564       564 105       621        621        105       621        621          105       621      621          105      590       590 110       621        621        110       621        621          110       621      621          110      618       618 115       621        621        115       621        621          115       621      621          115      621       621 120       621        621        120       621        621          120       621      621          120      621       621 125       621        621        125       621        621          125       621      621   1      125      621       621 130       621        621        130       621        621          130       621      621          130      621       621 135       621        621        135       621        621          135       621      621          135      621       621 140       621        621        140       621        621          140       621      6211         140      621       621 145       621        621        145       621        621          145       621      621          145      621       621 150       621        621        150       621        621          150       621      621          15C      621       621 150      2433        995        150      2433        974          15C      2433      957          150     2433       935 155                 1033        155                 1016          155               1002          15                 989 160                 1075        160                 1061          160               10511         160               1048 165                1119         165                 1109          165               1104          165               1112 170                1166         170                 1161          170               1161          170 175                1217         175                 1217          175                             175 180                1272         18t                               180                             18 185                1331         185                               185                             185 190                1395         190                               19                              190 195                 1463        195                               195                             195 200                1536         200                               200                             20 205                1615         205                               205                             205 210                1700         210                               210                             210 215                1791         215                               215                             215 220                1889         220            _                  2201                            220 _
225                 1995        225                               225                             225 230                2108         230                               230                             230 235                2230         235                               235                             235 240  _             2361         240t                              240                             240 Page 9 of 9

ATTACHMENT 6-A Revised P-T Limit Curves for Braidwood Station, Units I and 2

BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT a.
2) n in 2
C-0, 0
0     50    100  150   200      250    300    350 400 450 500 ModeratorTemperature (deg. F)
Figure 2.1 Braidwood Unit I Reactor Coolant System Heatup Limitations (heatup rate up to 100&deg;F/hr)
Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) 4

BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT la 0
a IV a.,
to la a.
            *0 to U
0    50  100  150   200   250   300    350  400 450 500 ModeratorTemperature (deg. F)
Figure 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 0, 25, 50 and 100 0F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors)
S

BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a (Page 1 of 2)
Braidwood Unit 1 Heatup* Data Points at 16 EFPY
('Without Margins for Instrumentation Errors)
Heatup Curve 100 F Heatup        Criticality      Leak Test Limit            Limit T         P       T         P     T          P 60         0     210         0    188       2000 60      565.09   210      611.83  210       2485 65      565.09   210      597.56 70      565.09   210      585.60 75      565.09   210      576.77 80      565.09   210      570.35 85      565.09   210      566.61 90      565.09   210      565.09   =
95      565.09   210      565.87 100     565.87   210      568.69 105     568.69   210      573.56   =
110     573.56   210      580.30   _
115     580.30   210      588.84 120     588.84   210      599.36 125     599.36   210      611.78   =
130      611.78   210      626.07 135     626.07   210      642.16   _
140     642.16   210      660.36   =
145      660.36   210      680.59 150     680.59   210      702.80   =
155     702.80   210      727.33        _
160     727.33   210      754.07   =
165     754.07   210      783.17 170     783.17   215      814.98   _
175     814.98   220      849.37   =
180      849.37   225      886.54 185     886.54   230      926.73   =
190      926.73   235      970.11 195      970.11   240      1016.91 200     1016.91   245      1067.33 205     1067.33   250      1121.63  _
210     1121.63   255      1180.01  _
215     1180.01   260      1242.62  =
220     1242.62   265      1309.84  =
225     1309.84   270      1382.03 230     1382.03   275      1459.45      _
235     1459.45   280      1542.27 240     1542.27   285     1630.97 245     1630.97   290     1726.05   =      .     -
250     1726.05   295     1827.80 6

BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a Page 2 of 2 Heatup Curve 100 F Heatup                 Criticality           Leak Test Limit                Limit T                P           T           P          T        P 255           1827.80        300       1936.51 260           1936.51        305       2052.39 265           2052.39        310       2176.33 270           2176.33        315       2308.42 275           2308.42        320       2449.09 280           2449.09 0
* Heatup and Cooldown data includes vessel flange requirements of 11 0 F and 621 psig per I OCFR50, Appendix G.
7

BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1b Page I of 1 Braidwood Unit 1 Cooldown* Data Points at 16 EFPY**
(eithout Margins for Instrumentation Errors)
Cooldown Curves Steady State                 25 &deg;F                   50 &deg;F           100 IF Cooldown Cooldown                Cooldown T            P           T           P          T           P           T            P 60            0          60            0         60          0          60            0 60        620.27         60        577.45        60      534.28         60        446.98 65        621.00         65        590.68        65      548.52         65        463.79 70        621.00         70        605.03        70      563.98         70        481.93 75        621.00         75        620.51        75      580.67         75        501.49
                                 .80        621.00         80        621.00        80      598.51         80        522.68 85        621.00         85        621.00        85       617.90        85        545.50 90        621.00         90        621.00        90      621.00         90        570.23 95        621.00         95        621.00        95      621.00         95        596.83 100        621.00         100       621.00       100      621.00         100       621.00 105        621.00         105       621.00       105      621.00         105       621.00 110        621.00         110       621.00       110      621.00         110       621.00 110        795.92         110       766.92       110      739.27         110       690.04 115        821.55         115       794.59       115      769.53         115       726.24 120        849.00         120       824.45       120       801.97        120       765.12 125        878.42         125       856.54       125      836.87         125       807.07 130        910.25         130       890.97       130       874.41        130       852.23 135        944.34         135       928.00       135       915.03        135       900.91 140        980.89         140       967.79       140      958.57         140       953.33 145        1020.15        145      1010.84       145      1005.42        145       1009.81 150        1062.35        150      1056.88       150      1055.76 155        1107.92        155      1106.38 160        1156.42 165        1208.78 170        1265.05                            _                        _       _
175        1325.37 180        1390.04        _                       =
185        1459.41                              _     _
190        1533.55                                =
195        1613.49        _                       _
200       1699.01 205       1790.55 210        1888.61 215        1993.61 220       2105.69 225       2225.77 230       2353.75         _         _  _   _ __             _  __  __                _  _  _
0
* Heatup and Cooldown data includes vessel flange requirements of I10 F and 621 psig per IOCFR50, Appendix G.
*
* For each cooldown rate, the steady-stale pressure values shall govern Ihe temperature where no allowable pressure values are provided.
8

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2500  r 2250 2000 1750 a
in afi 1500 2
C
 @. 12500 U) 3 ID 1250 a-351000 U
750 750 500 250 0
0      50      100   150      200    250     300     350   400  450    500  550 Moderator Temperature (Deg. F)
Figure 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100 0 F/hr) Applicable for the First 16 EFPY Using the 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) 4

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2250 2000 1750 0,
a: 1500 I.CD to n
et (0
cob1250 C.
70
*5 1000 0
750 500 250 0
0      50    100   150    200    250     300     350    400 450   500   550 Moderator Temperature (Deg. F)
Figure 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and I00&deg;F/hr) Applicable to the First 16 EFPY using 1996 Appendix G Methodology (Without Margins of Instrumentation Errors) 5

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a (Page 1 of 2)
Braidwood Unit 2 Heatup* Data Points at 16 EFPY Using the 1996 Appendix G Methodology (Without Margins for Instrumentation Errors)
Heatup Curve 100 F Heatup     Criticality    Leak Test Limit Limit T            P      T           P    T           P 60           0     207          0   186        2000 60         617     207        621  207         2485 65         617     207        621 70     _   617     207        621 75         617     207        621 80         617     207        621 85         617     207        621 90         617     207        621 95         617     207        621 100         617     207        621 105         619     207        621                    l 110         621     207        621 115         621     207        621 120         621     207        621 125         621     207        621 130         621     207        621 135         621     207        621 140         621     207        696 140         621     207        715        _   .     -
140         696     207        736 145         715     207        760 150         736     207        786 155         760     207        815        _
160         786'    210        846 165         815     215        880 170         846     220        917 175         880     225        957 180         917     230        1000 185         957     235       1047 190         1000    240       1097 195         1047    245       1152 200         1097    250       1210 205         1152    255       1273 6

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a Page 2 of 2 Heatup Curve 100 F Heatup                  Criticality           Leak Test Limit Limit T              P              T          P            T             P 210            1210           260        1341 215            1273            265       1415 220            1341           270        1493 225            1415            275       1578 230            1493            280       1669 235            1578            285       1766 240            1669            290       1871 245            1766            295       1984                         _
250            1871            300       2105 255            1984            305       2235 260            2105.           310       2374 265            2235 270            2374          _                                 _      _
e Heatup and Cooldown data includes the vessel flange requirements of 140 'F and 621 psig per IOCFR50, Appendix G..
7

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1b (Page 1 of 1)
Braidwood Unit 2 Cooldown* Data at 16 EFPY Using the 1996 Appendix G Methodology (Without Margins for Instrumentation Errors)
Cooldown Curves Steady State           25 &deg;F               50 &deg;F               100OF Cooldown            Cooldown            Cooldown T         P         T         P         T          P         T         P 60        0         60        0        60          0        60          0 60       621        60       602       60         554       60        455 65       621        65       616       65         568   . 65        471.
70      621         70       621       70         583       70        489 75      621         75       621       75         599       75        508 80       621        80       621       80         617       80        529 85      621         85       621       85         621       85        552 90      621         90       621       90         621       90        576 95      621         95       621       95         621       95        603 100      621        100       621       100        621       100       621 105      621        105       621       105        621       105       621 110      621        110       621       110       .621       110       621 115      621        115       621       115        621       115       621 120      621        120       621       120        621       120      *621 125      621        125       621       125        621       125       621 130      621        130       621       130        621       130       621 135      621        135       621       135        621       135       621 140      621        140       621       140        621       140       621 140      621        140       621       140        621       140       621 140      1010       140       991       140        975       140       957 145      1050       145       1034      145       1022       145       1013 150      1092       150       1080      150       1072       150       1074 155      1137       155       1129      155       1126       155       1137 160      1186       160      1183       160       1185       160       1186 165      1239       165       1239      165       1239       165       1239 170      1295       170       1295      170       1295       170       1295 175      1356       175      1356       175       1356       175       1356 180      1422       180      1422       180       1422       180       1422 185      1492       185      1492       185       1492       185       1492 190      1567       190      1567       190       1567       190       1567 195      1649       195      1649       195       1649       195       1649 200       1736       200      1736      200        1736      200        1736 205       1830      205       1830      205        1830      205        1830 210      1931        210      1931      210        1931      210        1931 215      2039        215      2039      215        2039      215        2039 220      2156       220       2156      220       2156       220       2156 225      2281       225       2281      225       2281        225      2281 230      2416       230       2416      230       2416       230       2416 Heatup and Cooldown data includes the vessel nange requirements of 140 IFand 621 psig per IOCFRSO, Appendix G.
8

ATTACHMENT 6-B Revised PT Limit Curves for Byron Station, Units I and 2

BYRON - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT XtX111tttS                                                                           t            tt1 11>
22s0                           TEST LIMT                                                        -
1750 _fiXXXmWAf                                                                                                                  lll                    lll
                        ++
                         -+e                +        l~UNACCEPTABLE-SS++dHI                                                                                     M
                                            .w444OPERtf 4(              N C-S FW I WLL W                                         J JI                               L1It .II1i IS P TOATUP0 TE U
                      .- _   UP TO 100 F/HR
                                                                                                     'I 1000                                                                                                                               ACCEPTABLE IN           I I                 OPERATION 750 I U0IT IVjFOROCRIT.
t00 FA I 500                                                                                                             I I              .~~~~~~~~~~ . .I. I                II          I .             I      I   I   I I  .I . I
                                                                                                                       .I        I I 4 I  I. Il    .  . . I.         I I         I..
dBOLT.UP                                                                                                   II LTEWF.        I ,,T -1  .,, .    .
7TTTTT     ,,.  .
j    IV NHSCRITiCA~llY                LULMT BASED ON INSERFMCE I ....
I     I                          . .     . .. i..
250j II                                       HYDROSTATIC TEST TEMAERATURE (225 F)                  II FOR THE SERMCE PERIOD UP TO 17.6 EFPY
                   ... I I.1 1 1 1 1 I 1                                              tT-r I I I I I I-- I -1I I I I I I I I I II I I 0    _-
I                   . I  I,      I   I        I III,        II_       I   I  I   I
                                                                                                                                                                .5-   I  I      I 0           50          100                150                  200              250                300)            350            400          450            500 ModeUor TenVqe~ur, Peg. F)
Figure 2.1:
Byron Unit I Reactor Coolant System Heatup Limitations (Heatup rates up to 100 0F/hr) Applicable for 17.6 EFPY (W'ithout margins for instrumentation errors and using 1996 Appendix G Methodology) 4

BYRON - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2250 2000                         UNACCEPTABL
                              - - - - - OPERATION 1750 ACCEPTABLE OPERATION
       -a 1500 In 01 c- 1250 or C
U
* 0 1000 COOLDOWN RATES F/Hr.
750 0
25 500        50 100 -
250 0
0         50       100     150      200        250         300       350 400 450 500 Moderator Temperature (Deg F)
Figure 2.2:
Byron Unit I Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100 0F/hr)
Applicable for 17.6 EFPY (Without margins for instrumentation errors and using 1996 Appendix G Methodology) 5

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 Byron Unit I Heatup and Cooldown Data Points at 17.6 EFPY (Without Margins for Instrumentation Errors and Using the 1996 Appendix G Methodology)
             . Heatup Curve             _               Cooldown  Curves 100 F       Cnticality   Leak Test    Steady        25 F        50 F       100 F Heatup         Limit        Limit       State     Cooldown    Cooldown    Cooldown T       P     T       P. T I P       T I P       T       P   T       P    T     P 60      0 225          0 2041 2000     601    0    60      0   60      0   60     0 60 587 225 587 225 2485                60 613      60 561      60    509   60 402 65 587 225 587                         65 621      65 572      65 520      65 414 70 587 225 587                         70 621      70 582      70 531      70 427 75 587 225 587                         75 621      75 594      75 544      75 442 80 587 225 587                         80 621      80 607      80 557      80 458 85 587 225 587                         85 621      85 620      85 572      85 475 90 587 225 587                         90 621      90 621      90 588      90 494 95 587 225 587                         95 621      95 621      95 605      95 514 100 587 225 587                        100 621     100 621     100 621     100 535 105 587 225 587                        105 621     105 621     105 621     105 559 110 587 225 587                       110 621      110 621     110 621     110 584 115 587 225 587                       115 621      115 621     115 621     115 611 120 588 225 587                       120 621      120 621     120 621     120 621 125 591      225 587                  125 621     125 621      125 621     125 621 130 596 225 587                       130 621      130 621     130 621     130 621 135 602 225 587                        135 621      135 621     135 621     135 621 140 611 225 587                        140 621     140 621     140 621     140 621 145 621 225 588                        145 621     145 621     145 621     1451 621 150 621 225 591                       150 621     150 621      150 621     150 621 155 621 225 596                        155 621      155 621     155 621     155 621 160 621 225 602                        160 621      160 621     160 621     160 621 165 621 225 611                        165 621     165 621      165 621     165 621 170 621 225 622                        170 621     170 621      170 621     170 621 175 621 225 634                        175 621     175 621      175 621          _
180 621 225 648 _                      180 621     180 621 180 750 225 665                        180 1207    180 1205 185 777 225 683                        185 1261 190 806 225 703                        190 1319 195 838 225 725                        195 1382 200 872 225 750                        200 1449    =
205     910   225     777              205 1521 210    950   230      806              210 1599 215    994   235      838              215 1683 220 1041     240      872              220 1773 225 1092     245      910              225 1869 230 1147      250     950               230 1973 235 1206      255    994                235 2085 6

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 (Continued)
Heatup Curve                                                       Cooldown Curves 100 F           Criticality Leak Test                      Steady              25 F               50 F               100 F Heatup                Limit                Limit              State         Cooldown Cooldown                      Cooldown T I P              T I P                T lP             _T l P                 T lP              T        p Pl      -T           P 240 1269 260 1041                                            2401 2205 245 1338 265 1092                                            245 2334 250      1411      270       1147                            250    2473      -
255 1490           275       1206 260 1575           280       1269 265 1666           285       1338 270 1764           290       1411 275 1869           295       1490 280 11982          300       1575 285 2104           305       1666 2901 2234          310       1764 295 2374           315       1869 320       1982 325       2104 330       2234 335       2374                           =
0 Note I: Heatup and Cooldown data includes the vessel flange requirements of 180 F and 621 psig per I OCFRSO, Appendix G.
Note 2: For each cooldown rate, the steady-state pressure values shall govem the temperature where no allowable pressure values are provided.
Note 3: Temperatures and pressures are given in &deg; F and psig, respectively.
7

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 0     EC      1D      15D      2M)    2D       300     aD      400     45D      HD MXbA7TGTmEr&LW(D Fe Figure 2.1:
Byron Unit 2 Reactor Coolant System   Heatup Limitations  (Heatup rates up to 100 'F/hr) Applicable for 17.5 EFPY (W'ithout margins for instrumentation errors and using 1996 Appendix G Methodology) 4

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT
_J                                      11X1ILlL                                                                                                               ~~~            1     X
_)                                                                                                    __ + __                         __ IIIlIIII11' D __      W _--I IwrX                                  !Ia 4                                                                                                                  t 179;    3---T)               -- -- ,,,I,m                           IlX_1                           I__
CPEPAMCNNI i~PR                            111_1111
            )                                                                                                                                      ~             1                                    -
0 M
0 0 120
             )    Itk                                                                1T                                                 1                        11If ,,I
  .21
   =IU 0
RT             RH.             -      I I-I I I F .1            III           1 II             I I I _ I I I I I i i i I II                           I - iI I I I I II I-1,-f14
               -I I  I -I I I -II II I.      I I
                                                                      . I.
I
                                                                                  .I 4-M 11 1 11f11 1h1 .ill L
LI I 7
3-1     0
_      I II      I I I                I  II I    I         I I
I  I I
I    I   I  I I I I
I   I  I I
I I I
I  I r-I 1(
II..-
I
                                   't1-I-- +-
I tX                     2I .            I I   I 11 1 1 II                                                              I I     I    - I   I    I    I         I    I     I    I          I . I   .  .  . I    I     I    I      . I  I    I      I    I  I,  I      I N I cri ri         I   II I
I I I I
I I X, I l
llB9P.
1 i:IzL I I                                ....
il                  II    IT I            I .      .I .    .l       .            I I I I II. I I .I        .   .     .          .  .  .  . I O.         I I             ;                  II     I   I i_                          .   . I .          ;         _         ;              _  ;     _          ;
0                    50                      100               15D              20               250               30                                                     450            MD rbTurFr            pe       D (Do      F)
Figure 2.2:
Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 &deg;F/hr)
Applicable for 17.5 EFPY (Without Margins for Instrumentation Errors and using 1996 Appendix G Methodology) 5

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1: Byron Unit 2 Heatup and Cooldown Data Points at 17.5 EFPY (without Margins for Instrumentation Errors and Using the 1996 Annendix G Methodniniv)                                             He     sffIbJ Heatup Curve                     I                Cooldown Curves 100 F        Criticality Leak Test            Steady          25 F          50 F          100 F Heatup           Umrit          Unt             State       Ccdodn Coodo-Au             Cco&cimn T               T      P      T          P    T       P      T      P      T       P      T         P 600           219S0 198              200i        0 0        60C 0 60                0     60 3 60 621 219 635 219 2485                       60 621         60 574 60 523               60 418 65 621 219 674                                65 621         65 585 65 534               65 431 85 621 219 660-                   -           70 621         70 597 70 547                70 446 90 621 219 650 1                              751 621        75 610 75 561                75 46 95 621 219 643 1                              80 621         80 621 80 576                80 480 100 621 219 638                                 85 621         85 621 85 592              85 498 105 621 219 637                         -       90 621         90 621 90 6091               90 519 110 621 219 637                                 95 621         95 621 95 621                95 541 115 621 219 641                         -      100 621        100 621 100 621             100 564 120 621 219 646                                1051 621       105 621 105 621             105 590 125 621 219 654                                110 621        110 621 110 621             110 618 130 621 219 664                                115 621        115 621 115 621             115 621 135 621 219 676                                120 621        120 62112016211             120 621 140 621 219 690                                125 621        125 621 125 621             125 621 145 621 219 707                                130 621        130 621 130 621             130 621 150 621 219 725                                135 621        135 621 135 621             135 621 150 707 219 746                                140 621        140 621 140 621             140 6211 155 725 219 770                          -     145 621        145 621 145 621             145 621 160 746 219 796 =                              150 621        150 621 150 62              150 621 165 770 219 824 =                      _=     150 995        150 974 150 957             150 9T5 170 796 220 855                      _        155 1033       15' 1016 155 1002           155 989 175 824 225 8891                              160 107        160 1061 160 1051           160 1048 13C L          230 926 =                =     165 1119       165 1109 165 1104           165 111 185 889 23; 966 =                              170 1166       170 1161 170 1161 190 926 240 1010 =                             1751217        175 1217 195 966 245 1057                               180 1272 200 1010 250 1107                              185 1331                             =            -
205 1057 255 1162                     _        190 1395 210 1107 260 1221 1                      =     195 1463 215 1162 265 1285 =                           2001536 220 1221 270 1353                              205 1615                                            -
225 1285 275 1427                              210 1700 230 1353- 280 1506                             215 1791           -
235 1427 2851591                               220 1889 ==                          =
240 1506 290 16831                             225 1995 245 1591 295 1781                              230 2108 250 1683 300 1887                        -     235223C0              =      =                    -
255 1781 305 2001                              240 2361 ==
260 1887 310 2123 265 2001 315 2254 ._
270 212 320 2395 275 2254 Note 1: Heatup and Cooldown data includes the vessel flange requirements of 150 IF and 621 psig per I OCFR50, Appendix G.
Note 2: For each cooldown rate, the steady-state pressure values shall govern the temperature where no allowable pressure values are provided.
Note 3: Temperatures and pressures are given in I Fand psig, respectively.
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