ML20141J978
| ML20141J978 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 05/21/1997 |
| From: | Hosmer J COMMONWEALTH EDISON CO. |
| To: | NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20141J982 | List: |
| References | |
| NUDOCS 9705280275 | |
| Download: ML20141J978 (12) | |
Text
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Commonweahh Edison Company 1400 Opus l'iv:e Downers Grove, u.60515V01 i
May 21,1997 Oflice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk
SUBJECT:
Application for Amendment to Appendix A, Technical Specifications, for Facility Operating Licenses Byron Station Units 1 and 2 Facility Operating Licenses NPF-37 and NPF-66 NRC Docket Nos. 50-454 and 50-455 Braidwood Station Units 1 and 2 Facility Operating Licenses NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457
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Relocation of Pressure and Temperature Limits j
REFERENCES:
- 1) Westinghouse Owner's Group WCAP-14040-NP-A, " Pressure and Temperature Limits Methodology for Heatup, Cooldown, and COMS Analysis," dated January 1996.
- 2) NRC Safety Evaluation Report for use of ASME Code Case N-514 for Byron Station, letter dated November 29,1996.
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- 3) NRC Safety Evaluation Report for use of ASME Code Case N-514 for Braidwood Station, letter dated July 13,1995.
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- 4) Letter from G. Stanley and K. Graesser (Comed) to NRC Document Control Desk, " Conversion to the Improved Technical Specifications,"
dated December 13,1996.
- 5) Letter from J. Hosmer (Comed) to NRC Document Control Desk,
" Request for Exemption from Requirements of 10 CFR 50.60 to Use the 1996 Addenda of ASME Section XI, Appendix G, Article G-2000,
' Vessels'," dated April 3,1997
- 6) Letter from J. Hosmer (Comed) to NRC Document Control Desk, Reactor VesselIntegrated Suneillance Program 10 CFR 50, Appendix H, Section III.C," dated May 6,1997.
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NRC Document Control May 21,1997 Pursuant to 10 CFR 50.90, Commonwealth Edison (Comed) proposes to amend Appendix A, Technical Specifications, of Facility Operating Licenses NPF-37, NPF-66, NPF-72, and NPF-77 to allow licensee control of the Reactor Coolant System (RCS) Pressure and Temperature Limits for heatup, cooldown, low temperature operation, and hydrostatic.esting.
This change is being submitted consistent with the guidance provided in NRC Generic Letter 96-03, " Relocation of the Pressure Temperature L.imit Curves and Low Temperature Overpressure Protection System Limits", NUREG-1431, " Standard Technical Specifications for Westinghouse Plants," Revision 1, and WCAP-14040-NP-A (Reference 1), with exceptions noted in Attachment E.
The methodology used to generate the revised heatup, cooldown, low temperature overpressure protection (LTOP), and hydrostatic testing limits is documented in WCAP-14040-NP-A, with exceptions noted in Attachment E, in combination with ASME Code Case N-514. The NRC approved an exemption to allow use of ASME Code Case N-514 in the generation of LTOP setpoints for Byron Station in Reference 2 and for Braidwood Station in Reference 3. Note that the LTOP system is the same as cold overpressure mitigation system (COMS).
Comed also proposes to relocate the reactor vessel capsule withdrawal schedule in accordance with Generic Letter 91-01, " Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications."
In Reference 4, Comed submitted an application for conversion to the Improved Standard Technical Specifications (ITS). Corresponding changes are provided on the proposed ITS pages, in addition to the current Technical Specifications pages, in order to '; corporate license amendments for requests submitted after the ITS application.
Comed will require new low temperature overpressure protection setpoints following steam generator replacement as a result of the impact ofincreasing reactor coolant system volume on the heat injection transient. The Byron Unit I steam generator replacement outage is scheduled during the eighth refueling outage (BIR08). Approval of this proposed change is requested by November 3,1997, to support the current outage schedule. Comed requests that the NRC grant a 60 day implementation period after the amendment is issued to provide for revision of station procedures and other appropriate documents.
NRC Document Control May 21,1997 The following Attachments have been developed in support of this proposed change:
Attachment A Description and Safety Analysis Attachment B Marked Up Copy of Current and Proposed Improved Technical Specifications Attachment C Evaluation of Significant Hazards Considerations Attachment D Environmental Assessment Statement Attachment E Exceptions to WCAP-14040-NP-A Methodology Attachment F Pressure Temperature Limits Report for Byron Unit 1 Attachment G Pressure Temperature Limits Report for Byron Unit 2 Attachment H Pressure Temperature Limits Report for Braidwood Unit 1 Attachment I Pressure Temperature Limits Report for Braidwood Unit 2 In Reference 5, Comed requested permission to use a later version of the ASME Boiler and Pressure Vessel Code than is listed in 10 CFR 50.55a. The curves provided in the attached Pressure Temperature Limits Reports were based on the ASME Code that is currently required by 10 CFR 50.55a; however, future curves will be based on the later Code.
In Reference 6, Comed requested permission to integrate the weld metal surveillance programs for Byron Units 1 and 2 and for Braidwood Units 1 and 2. Note that due to capsule integration of the data pursuant to 10 CFR 50.61 and 10 CFR 50, Appendix H, the Byron Unit 2 and Braidwood Unit 2 pressure temperature (P/T) and LTOP curves that are contained in Attachments G and I, respectively, are negatively impacted in terms of applicable effective full power year (EFPY). In addition, the Byron Unit 2 curves are also adversely impacted by use of the ratio procedure as described in 10CFR50.61(c)(2)(ii)(B). The impact is due to the chemistry factor adjustment of ARTwr due to differences between the best estimate values for copper and nickel content of the weld wire heat numbers associated with the vessel and surveillance welds. Because of the negative impact on both units, new P/T and LTOP curves j
are being generated for both Byron Unit 2 and Braidwood Unit 2 that will provide appropriate conservatisms in the operating curves. These new curves will be provided to the NRC as a i
supplement to this submittal prior to final approval of the amendment request. The current Technical Specifications for Byron and Braidwood Stations remain applicable, pending NRC approval of the new integration process and this amendment request.
This proposed amendment has been reviewed and approved by Comed Onsite and Offsite Review in accordance with Comed procedures. Comed has reviewed the proposed amendment in accordance with 10 CFR 50.92(c) and has determined that no significant hazards consideration exists.
To the best of my knowledge and belief, the statements contained above are true and correct.
NRC Document Control May 21,1997 Comed is notifying the State ofIllinois of this application for amendment by transmitting a copy of this letter and its attachments to the designated state official.
Please address any comments or questions regarding this matter to Marcia Lesniak, Nuclear Licensing Administrator, at (630) 663-6484.
Sincerely, j,:::: OFFICIAL SEAL g
l JACQUELINE T EVANS i
p;uv comuession emnts:is, isis 7
, norAny rusuc, srArt or aunois John B. Hosmer Engineering Vice President
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Signed on this h day of 6y
,1997 by Owr >NT &w otar/Public Attachments cc:
A.B. Beach, Regional Administrator -RIII G.F. Dick, Byron /Braidwood Project Manager - NRR S.D. Burgess, Senior Resident Inspector - Byron C. Phillips, Senior Resident Inspector - Braidwood Office of Nuclear Facility Safety - IDNS
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l ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES l
NPF-37, NPF-66, NPF-72, AND NPF-77 i
A.
DESCRIPTION OF PROPOSED CHANGES i
Comed propeses to amend Appendix A, Technical Specifications (TS), ofFacility Operating Licenses NPF-37, NPF-66, NPF-72, and NPF-77 to allow licensee control of the Reactor
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Coolant System (RCS) Pressure and Temperature (P/T) limits for heatup, cooldown, low temperature overpressure protection system (LTOP), and hydrostatic testing. This change is
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consistent with the guidance provided in Nuclear Regulatory Commission (NRC) Generic 1
Letter (GL) 96-03, " Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits"; NUREG-1431, " Standard Technical i
Specifications for Westinghouse Plants," Revision 1; and WCAP-14040-NP-A, " Pressure and Temperature Limits Methodology for Heatup, Cooldown, and COMS Analysis."
Additionally, Comed proposes to relocate the reactor vessel surveillance program capsule withdrawal schedules in accordance with GL 91-01, " Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications."
The proposed changes are described in detail in Sections E and F of this Attachment. The marked up TS and proposed Improved TS (ITS) pages are shown in Attachments B-la and B-lb, respectively, for Byron, and Attachments B-2a and B-2b, respectively, for Braidwood.
3 B.
DESCRIPTION OF THE CURRENT REQUIREMENTS The Technical Specifications listed below contain the Limiting Conditions for Operation (LCOs) that establish P/T and LTOP limits for the RCS. The LCOs also require per: odic update of these limits and other supporting information. The operational limits themselves provide the acceptable range of operating temperatures and pressures for heatup, cooldown, low temperature overpressure protection, criticality, and inservice hydrostatic pressure test conditions. These limits are required to preserve the integrity of the RCS as a fission product barrier. Note that the LTOP system is the same as cold overpressure mitigation system (COMS).
i LCO 3.4.9.1 contains references to the RCS P/T limits contained in the TS and provides RCS heatup, cooldown, criticality, and inservice pressure test temperature limits.
Figures 3.4-2a and 3.4-2b contain the RCS heatup P/T limits and Figures 3.4-3a and 3.4-3b 1
contain RCS cooldown P/T limits for Units 1 and 2, respectively.
i LCO 3.4.9.3.b requires that the pressurizer power-operated reliefvalve (PORV) lift setpoint be within the limit established in Figures 3.4-4a and 3.4-4b for Units I and 2, respectively.
Table 4.4-5 contains the Unit I and Unit 2 reactor vessel capsule withdrawal schedules.
Tables B 3/4.4-la and B 3/4.4-lb, respectively, contain the Unit I and Unit 2 reactor vessel toughness data. For both Unit I and Unit 2, Figure B 3/4.4-1 contains the fast neutron fluence (E > 1 Mev) as a function of full power service life, while Figure B 3/4.4-2 contains the effect of fluence and copper on shift of the nil-ductility reference temperature (RTam) for reactor vessel steels exposed to irradiation at 550 F.
C.
BASES FOR THE CURRENT REQUIREMENTS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Appendix G. These limits preserve the integrity of the RCS as a fission product barrier.Section II.B.3 of Appendix H to Title 10 to the Code of Federal Regulations, Part 50 (10 CFR 50) requires that the preposed reactor vessel capsule withdrawal schedule be submitted and approved prior to implementation. The results obtained from the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. Results of these examinations are used to update P/T limits.
D.
NEED FOR REVISION OF THE REQUIREMENT Comed is submitting this amendment request in response to an internal commitment to obtain licensee control of the applicable limits, to incorporate the most recent surveillance capsule information in the limits for both stations, and to use the latest approved methodology for developing these limits.
E.
DESCRIPTION AND B ASES OF CHANGES TO CURRENT TECHNICAL SPECIFICATIONS (CTS)
Relocation of RCS Pressure and Temperature Limits Comed proposes to relocate RCS P/T limits and other information into a licensee-controlled document in accordance with GL 96-03. In GL 96-03, the NRC advised licensees that they may request a license amendment to relocate the P/T limit curves from their plant TS to a pressure temperature limits report (PTLR). The LTOP system limits may also be relocated to the PTLR. Relocating the curves and setpoints to a licensee-controlled document allows licensees to maintain these limits efficiently and at a lower cost, since a license request would 2
no longer be required. An alternative approach for controlling these limits was proposed during the development of vendor specific ITS (NUREGs 1430-1434). This alternative approach, similar to the one used for the Core Operating Limits, relocates the RCS P/T and LTOP limits, along with other supportive information to a PTLR. The PTLR then serves as the reference basis document in the affected LCOs. This amendment request is consistent with the guidance provided in GL 96-03 and NUREG-1431, Revision 1.. Descriptions of the specific TS changes follow.
The Definitions section is revised to add a new definition for the PTLR. Comed is deviating from the model definition in GL 96-03 by adding the PORV lift settings associated with the.
LTOP system to the PTLR. Although the GL described this change, it was not explicitly written in the model TS definition.
LCO 3.4.9.1 is revised to delete specific limits and references to figures that provide limits.
The limits shall be maintained within the limits of the PTLR. The corresponding action statement is also revised to reflect the limits in the PTLR.
LCO 3.4.9.3.b is revised to replace the reference to Figure 3.4-4 with a reference to the PTLR. The PORV lift setpoint curves are relocated to the PTLR.
The following figures are deleted from the TS and relocated to the PTLR, in accordance with GL 96-03:
Figures 3.4-2a and b ~ RCS Heatup Limitations (Units 1 and 2, respectively)
Figures 3.4-3a and b RCS Cooldown Limitations (Units 1 and 2, respectively)
Figure 3.4-4a Nominal PORV Pressure Relief Setpoint (Unit 1)
Figure 3.4-4b Nominal PORV Pressure Relief Setpoint (Unit 2)
The following figures and tables are deleted from the TS Bases and pertinent information is included in the PTLR:
Table B 3/4.4-1a Reactor Vessel Toughness (Unit 1)
Table B 3/4.4-lb Reactor Vessel Toughness (Unit 2)
Figure B 3/4.4-1 Fast Neutron Fluence as a Function of Full Power Service Life Figure B 3/4.4-2 Effect of Fluence and Copper on Shift of RTm for Reactor Vessel Steels Exposed to Irradiation at 550 F.
The Bases for 3/4.4.9, " Pressure / Temperature Limits," is replaced with text from the ITS Bases for the corresponding section. Changes from the ITS Bases were made to reflect CTS
. numbering and format. The changes are consistent with NUREG-1431 and the proposed ITS, and they provide more detail than the existing CTS Bases. The revised Bases are more appropriate, since the P/T limits will no longer be part of the TS.
A new specification is added to Section 6, " Administrative Controls," consistent with GL 96-03. The proposed specification lists the individual LCOs that reference the PTLR for 3
I RCS P/r limits, and references the methodology approved for use at Comed (WCAP-14040-NP-A, Revision 2). It also contains a requirement that the PTLR be provided to the NRC upcm issuance for each reactor vessel fluence period and for any revision or supplement thereto. WCAP-14040-NP-A was approved generically by the NRC staffin a safety evaluation dated October 16,1995.
Allowances for instrument uncertainty applied to LTOP setpoints and the minimum pressurization temperature have been determined using a process consistent with ANSI /ISA-S67.04-1994. This is the edition specified in WCAP-14040-NP-A, Revision 2.
Instmment uncertainties determined using a process consistent with ANSI /ISA-S67.04-1994 are also consistent with ANSI /IS A-S67.04-1982, the edition widely accepted within the NRC.
There are no conflicts between the two editions of ANSI /ISA-S67.04.
Comed is taking exception to some items in WCAP-14040-NP-A. These exceptions are described in detailin Attachment E. In order to incorporate these into the licensing basis, Comed has added a reference to tl e NRC's Safety Evaluation Report (SER) that will be issued when this license amendment request is approved. The SER will include the approval for the exceptions described in Attachment E. This approach was used by Comed's Zion Station for a similar request, which was approved in Amendment 177 to DPR-39 and Amendment 164 to DPR-48. The exceptions are described in the SER dated December 20, 1996.
Relocation of Reactor Vesgl Capsule Withdrawal Schedule Comed also proposes to relocate the reactor vessel capsule withdrawal schedule in accordance with GL 91-01. In GL 91-01, the NRC advised licensees that they may remove the withdrawal schedule for the reactor vessel material specimens, since the TS requirement duplicates the controls on changes to this schedule that have been established in 10 CFR 50, Appendix H.
Specifically, Surveillance Requirement 4.4.9.1.2 is revised to delete the reference to the reactor vessel material surveillance program withdrawal schedule and to information that will be relocated. Table 4.4-5, " Reactor Vessel Material Surveillance Program Withdrawal Schedule," is deleted and relocated to the PTLR. This change is consistent with the guidance provided in GL 91-01. Removal of the schedule for the withdrawal of reactor vessel capsules will not result in any loss of regulatory control because changes to this schedule are controlled by the requirements of Appendix H to 10 CFR 50. The requirements in the PTLR indicate that the capsules shall be removed and the specimens examined to determine changes ia their material properties, as required by Appendix H. Comed is deviating from the Generic Letter by placing the withdrawal schedule in the PTLR, rather than in the UFSAR, since the PTLR contains all of the related data. This is an administrative change since the requirements of 10 CFR 50 continue to be met, and the PTLR is updated as needed to reflect changes to the NRC-approved schedule. Controls on the PTLR and UFSAR are similar; a safety evaluation is required for changes to either document under the provisions of 10 CFR 50.59.
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Other Changes i
The references in the Bases to the ASME Boiler and Pressure Vessel Code,Section III, Appendix G, are changed to Section XI, Appendix G, to reflect 10 CFR 50, Appendix G.
10 CFR 50, Appendix G, was revised in December 1995 and included this change.
Section Ill, which applies to nuclear power plant components, was used for design, fabrication, installation, examination, and test requirement for the systems and components.
The TS involve periodic inspecting and testing throughout the life of the systems and components designed to Section III, therefore Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components,"is more appropriate. The require nents of the Appendices j
are identical.
The Table of Contents is revised to show the new definition for PTLR.
F.
DESCRIPTION AND BASES OF THE PROPOSED CIIANGES TO THE PROPOSED ITS
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Comed submitted an application for conversion to the Improved Technical Specifications (ITS) in a letter from G. Stanley and K. Graesser to the NRC dated December 13,1996. In order to incorporate license amendments for requests submitted after the date into the ITS, markups have been provided on the Proposed Improved Technical Specifications in addition to the current Technical Specifications. The proposed changes to the proposed ITS are as follows:
Relocation of RCS Pressure and Temocrature Limits The proposed ITS already include the PTLR definition, and individual LCOs already contain the appropriate referencer to a PTLR. The figures and tables that are being deleted in the CTS were not included in the proposed ITS, since this amendment request is expected to be approved before the ITS conversion. Some changes are proposed to the proposed ITS based
- on a review of changes proposed to the CTS.
Specification 5.6.6.b.1 of the proposed ITS is revised to include the appropriate version of WCAP-14040. WCAP-14040-NP-A, Revision 2 was issued to include the NRC's SER dated October 16,1995 accepting Topical Report WCAP-14040, Revision 1. The reference to the NRC SER in Specification 5.6.6.b.2, which was identified as an open item, is revised for consistency with the proposed CTS change. The SER for this amendment request should be referenced since it will approve exceptions to certain requirements in WCAP-14040, as described in Attachment E. The initial PTLR uses the methodology in WCAP-14040-NP-A, which is already referenced in Specification 5.6.6 as an NP. ' coroved method. Revisions must be made using NRC-approved methodology listed in. 4 YS. As stated in proposed ITS 5.6.6.c, the PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any resision or supplement thereto. Therefore, the proposed change is consistent with GL 96-03 and NUREG-1431 and the CTS proposal.
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4 The proposed changes to the proposed ITS are consistent with GL 96-03, as desciibed above.
Relocation of Reactor Vessel Capsule Withdrawal Schedule There are no changes required, the proposed ITS already reflects relocation of the withdrawal schedule.
Other Changra Changes are proposed to the Bases for Section 3.4.3, "RCS Pressure and Temperature (P/T)
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Limits." References to ASME Code Section III are changed to ASME Code Section XI, as described in the CTS changes.Section III references are retained in descriptions involving system design. References to reduced temperature are deleted from the Bases for Actions B.1 and B.2, since they can be misleading. In reduced pressure conditions, the possibility of propagation with undetected flaws is decreased. However, the material becomes tougher as i
temperature increases; the possibility of propagation of undetected flaws is decreased as i
temperature increases. In the reference section, the reference to WCAP-14040 is revised to reflect the current version. This change corresponds with the requirement in 5.6.6.b of the ITS described above. Additionally, the phrase," brittle cracking,"is changed to "non-ductile failure," to more accurately reflect ASME Code,Section III, Appendix G, " Protection Against Non-Ductile Failure."
In the Bases for Section 3.4.12, " Low Temperature Overpressure Protection (LTOP)
System," the phrase, " brittle cracking," is changed to "non-ductile failure," to more accurately j
reflect ASME Code,Section III, Appendix G, " Protection Against Non-Ductile Failure." In i
the same section, editorial changes are provided to the description of pressurizer PORV operability and P/T curves. These changes are for clarity, and they do not affect the meaning.
G.
IMPACT OF TIIE PROPOSED CIIANGES Relocating parameters to the PTLR provides an acceptable means of establishing and maintaining detailed values of the P/T limit curves and pressurizer PORV lift settings associated with the LTOP system limits. Plant operation will be limited in accordance with the requirements of Appendix G to 10 CFR Part 50. P/T and LTOP system limits maintained in the PTLR are established using a methodology approved by the NRC; these changes will not impact plant safety.
Maintaining licensee control eliminates the need to request a license amendment at the end of the effective period for P/T limit curves. The NRC agreed that processing amendment requests for changes to TS that are developed using an accepted methodology places an unnecessary burden on the licensee and NRC resources. Limits contained in the PTLR can only be revised under the provision of 10 CFR 50.59 and in accordance with the approved methodologies listed in the CTS /ITS. The methodology used to determine the P/T and LTOP 6
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l system limit parameters complies with the specific requirements of Appendices G and H to l
10 CFR 50; it is documented in an NRC-approved topical report (WCAP-14040-NP-A), and is incorporated by reference into the TS. Subsequent changes in the methodology must be l
approved by a license amendment. The methodology meets the requirements described in
' of GL 96-03.
GL 96-03 requires licensees to submit a formal PTLR containing the figures, values, and l
parameters derived from the application of the methodology approved by the NRC. These PTLRs are provided in Attachments F, G, H, and I for Byron Unit 1, Byron Unit 2, Braidwood Unit 1, and Braidwood Unit 2, respectively. Comed's proposal provides an equivalent assurance of compliance with design specifications as currently exists, yet it removes the unnecessary burden on both Licensee and NRC staff of processing amendment requests.
1 Relocating the schedule for the withdrawal of reactor vessel capsules will not result in any loss of regulatory control because changes to this schedule are controlled by the requirements of Appendix H to 10 CFR 50. The guiddines in GL 91-01 for removing the withdrawal schedule from TS are satisfied. There are no changes to the surveillance program itself. The requirements continue to be met for capsule withdrawal sequence and timing contained in i
ASTM E 185, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," which is incorporated by reference in 10 CFR 50
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Appendix H.
i H.
SCHEDULE REQUIREMENTS Comed will require new low temperature overpressure protection setpoints following steam generator replacement as a result of the impact ofincreasing reactor coolant system volume on the heat injection transient. The Byron Unit I steam generator replacement outage is scheduled during the eighth refueling outage (B1R08). Approval of this change is requested by November 3,1997 to support the current outage schedule.
Comed requested an exemption from 10 CFR 50.60 in a letter from J. Hosmer to the NRC dated April 3,1997. The exemption would allow Comed to use a later version of the ASME Code than currently specified in the CFR, i.e., the 1996 Addenda of ASME Section XI, Appendix G, Article G-2000, in the development of future allowable P/T limits, rather than addenda through the 1988 Addenda and editions through the 1989 Edition. Future PTLR curves will be developed using the later version of the Code.
In a letter from J. Hosmer to the NRC dated May 6,1997, Comed submitted a request to integrate capsule data from similar units into P/T curves pursuant to 10 CFR 50.61 and 10 CFR 50, Appendix H. This ensures consistency in the application of surveillance data to pressurized thermal shock values, as well as adjusted reference temperatures as inputs to P/T limits and LTOP setpoints. Byron Unit I and Unit 2 share beltline and surveillance weld metal heat numbers; Braidwood Unit I and Unit 2 also share beltline and surveillance weld metal 7
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heat numbers. The Byron Unit 2 and Braidwood Unit 2 P/T and LTOP curves that are contained in Attachments G and I, respectively, are negatively impacted in terms of applicable i
effective full power year (EFPY) by the integration of applicable surveillance data. In addition, the Byron Unit 2 curves are also adversely impacted by use of the ratio procedure as described in 10CFR50.61(c)(2)(ii)(B). This impact is due to the chemistry factor adjustment of ARTmr due to differences between the best estimate values for copper and nickel content of the weld wire heat numbers associated with the vessel and the surveillance welds. Because of the negative impact on both units, new P/T and LTOP curves are being generated for both Byron Unit 2 and Braidwood Unit 2, which will provide appropriate conservatisms in the i
operating curves: These new curves will be provided to the NRC as a supplement to this submittal prior to final approval of the amendment request. The current Technical
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Specifications for Byron and Braidwood Stations remain applicable, pending NRC approval of l
l the new integration process and this amendment request.
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