ML20069M289

Suppl to 940330 Application for Amend to Licenses NPF-72 & NPF-77,revising TS Sections 3.4.9.1 & 3.4.9.3 to Replace Figure 3.4-4a, Normal PORV Pressure Relief Setpoint Vs RCS Temp for Cold Overpressure Protection Sys..

ML20069M289
Person / Time
Site:	Braidwood
Issue date:	06/14/1994
From:	Saccomando D COMMONWEALTH EDISON CO.
To:	Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20069M290	List: ML20069M289 ML20069M295
References
NUDOCS 9406210276
Download: ML20069M289 (4)

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Downers Grove. Illinois 60515 June 14,1994 Mr. William Russell, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk

Subject:

Supplement to Request to Amend Technical Specification Sections 3.4.9.1 and 3.4.9.3 Braidwood Station Units 1 and 2 NPF-72/77; NRC Docket Nos. 50-456/457

Reference:

D. Saccomando letter to W. Russell dated March 30,1994, transmitting request to amend Technical Specification Sections 3.4.9.1 and 3.4.9.3

Dear Mr. Russell,

The reference letter transmitted Commonwealth Edison Company's (Comed) request to amend Sections 3.4.9.1 and 3.4.9.3 of the Technical Specification for Braidwood Units 1 and 2.

During a review associated with, but not directly part of the Low Temperature Overpressure Protection System (LTOPS) curve /setpoint effort, a calculational j anomaly was discovered in the proposed Technical Specification figure 3.4-4b. As a result of this situation, the proposed Technical Specification figure 3.4-4b has been revised. The attached figure supersedes the figure previously submitted in Attachment B with the reference letter. Additionally, this curve supersedes Figure 3 in Westinghouse letter CCE-92-193 (included as supporting information I with the original submittal). Also attached is Westinghouse letter CCE-94-232 which supports revision to this figure. .

1 This supplemental amendment request contains the following:

Attachment A: Description and Safety Analysis of Proposed Supplement Attachment B: Proposed Revision to the Technical Specifications The evaluation of Significant Hazards Considerations and the Environmental Assessment remains unchanged from the previous submittal.

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Mr. W. Russell June 14,1994 The proposed changes have been reviewed and approved by the On-site and Off-site Review Committees in accordance with CECO procedures. CECO has reviewed this proposed amendment in accordance with 10 CFR 50.92(c) and has determined that no significant hazards consideration exists.

CECO is notifying the State ofIllinois of our application for these amendments by transmitting a copy of this letter and the associated attachments to the designated State Official.

To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects these statements are not based on my personal knowledge, but on information furnished by other CECO employees, contractor employees, and/or consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.

Please address any further comments or questions regarding this matter to this office.

Sincerely, nu , m w e) d Denise M. Saccomando Nuclear Licensing Administrator Attachments cc: R. R. Assa, Braidwood Project Manager - NRR S. G. Dupont, Senior Resident Inspector - Braidwood B. Claytc,n, Branch Chief- Region III v S=====<==u Office of Nuclear Facility Safety - IDNS p OFFICIAL SEAL MARYELLEN D LONG W ARY PtJBitC. 5f Af f Or :(t Nogs My COMMiss:0N (xPIAf S 0411598

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. ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGES TO APPENDIX A l TECHNICAL SPECIFICATIONS OF l FACILITY OPERATING LICENSES l NPF-72 and NPF-77 )

A. DESCRIPTION OF THE PROPOSED CHANGE I

Replace the current Figure 3.4-4a, " Nominal PORV Pressure Relief Setpoint l Versus RCS Temperature for the Cold Overpressure Protection System Applicable l

up to 32 EFPY (Unit 1)" Page 3/4-40 in Commonwealth Edison's (Comed) March 30,1994, (reference letter) Application for Amendment to Facility Operating l Licenses-Reactor Coolant System with the attached Figure 3.4-4a, " Nominal PORV Pressure Relief Setpoint Versus RCS Temperature for the Cold Overpressure Protection System Applicable up to 32 EFPY (Unit 1)."

B. DESCRIPTION OF THE CURRENT REQUIREMENT

. The current Figure 3.4-4a describes the nominal Pressurizer Power Operated l Relief Valve (PORV) setpoints for the Low Temperature Overpressure Protection System (LTOPS) as a function of Reactor Coolant System (RCS) temperature.

C. BASES FOR THE CURRENT REQUIREMENT The setpoints provided for the LTOPS are selected such that the pressure peaks resulting from design basis overpressure events are limited to values less than those specified by Appendix G of Title 10 Code of Federal Regulations Part 50 (10 CFR 50). Appendix G provides the fiacture toughness requirements for reactor vessels under specified operating conditions, and Regulatory Guide (RG) 1.99, Rev.

! 2, " Radiation Embrittlement of Reactor Vessel Materials" specifies the procedure acceptable to the Nuclear Regulatory Commission (NRC) staff for calculating the pressure limits required by Appendix G.

D. NEED FOR REVISION OF THE REQUIREMENT During a review associated with the LTOPS curve and setpoint effort, a calculational error was discovered in the proposed Technical Specification Figure 3.4-4a. The original curve neglected to account for the 50 F thermal transport effect from the postulated heat injection transient. As a result, Figure 3.4-4a has been revised.

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r E. DESCRIPTION OF THE REVISED REQUIREMENT Figure 3.4-4a, " Nominal PORV pressure relief Setpoint Versus RCS Temperature J for the Cold Overpressure Protection System Applicable Up to 32 EFPY (Unit 1)" l in Comed's March 30,1994, submittal will be replaced with the attached Figure l 3.4-4a, " Nominal PORV Pressure Relief Setpoint Versus RCS Temperature for the Cold Overpressure Protection System Applicable Up to 32 EFPY (Unit 1)."  ;

i F. BASES FOR THE REVISED REQUIREMENT  !

l The setpoints provided for the LTOPS are selected such that the pressure peaks )

resulting from design basis overpressure events are limited to values less than i those specified by Appendix G of 10 CFR 50. Appendix G provides the fracture I toughness requirements for reactor vessels under specified operating conditions, 2

and RG 1.99, Rev. 2, " Radiation Embrittlement of Reactor Vessel Materials" i specifies the procedure acceptable to the NRC staff for calculating the pressure limits required by Appendix G.

G. IMPACT OF THE PROPOSED CHANGE This change corrects a calculational error in the LTOPS curve. No new equipment i is being installed, no new system interfaces are being created, no existing system interfaces are being modified. Thus, this change has no negative impact on any i system or operating mode. l j

j H. SCHEDULE REQUIREMENTS

! The proposed amendments dated March 30,1994, to Technical Specifications

3.4.9.1 and 3.4.9.3 are required for Braidwood Unit 1 to exceed 4.5 Effective Full i Power Years (EFPY). Braidwood Unit 1 is scheduled to reach 4.5 EFPY during the month of July,1994. Therefore, the March 30,1994, submittal requested that

, the amendments be approved by June 30,1994. Since this change is designed to replace one page of that submittal, the schedule requirements of the March 30, 1994, request applies. ,

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