ML20070J421

Supplemental Application for Amend to Licenses NPF-72 & NPF-77,revising TS Sections 3.4.9.1 & 3.4.9.3

ML20070J421
Person / Time
Site:	Braidwood
Issue date:	07/21/1994
From:	Saccomando D COMMONWEALTH EDISON CO.
To:	Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20070J422	List: ML20070J421 ML20070J423
References
NUDOCS 9407250092
Download: ML20070J421 (5)

Text

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. - - Commonwealth Edison L 1400 Opus Place oowners Grove, minos 60515

, July 21,1994 Mr. William Russell, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission i Washington, D.C. 20555 i Attn: Document Control Desk i ..

L

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. 50-456/452 I Teleconference dated July 18,1994, between Commonwealth

References:

1.

Edison Company and the Nuclear Regulatory Commission

2. D. Saccomando letter to W. Russell ,

dated June 14,1994, transmitting request to amend Technical Specification Sections 3.4.9.1 and 3.4.9.3

3. 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 3 transmitted Commonwealth Edison Company's (Comed) request to amend Sections 3.4.9.1 and 3.4.9.3 of the Technical Specifications for Braidwood Units 1 and 2. This amendment request was subsequently supplemented in reference letter 2.

During the reference teleconference it was brought to Comed's attention that Technical Specification 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)," needed to be revised to account for a 60 pounds per square inch gauge pressure channel measurement uncertainty. To accomplish this, it was necessary to revise the curve submitted in reference 2 by reducing the duration of applicability of Figure 3.4-4a from 32 Effective Full Power Years (EFPY) to 5.37 EFPY. This revised curve is more conservative than what was initially submitted because it accounts for a 50 thermal transport effect and pressure channel measurement uncertainty.

1-94o7250092 940721 PDR ADOCK 05000456.

PDR P. {l

. Mr. 5V. Russell July 21,1994 This supplemental amendment request contains the following:

Attachme;it A: Description and Safety Analysis of Proposed Supplement Attachment B: Proposed Supplemental Revision to the Technical Specifications The evaluation of Significant Hazards Considerations and tho &.vironmental Assessment remains unchanged from that previously submitted in reference 3.

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 of Illinois of our application for this supplemental request by transmitting a copy of this letter and the associated attachments to the.

designated State Oflicial.

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 I 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 )

oflice.

Sincerely,

- )

/*v~y pp ~~fj Denise M. Saccomando Nuclear Licensing Administrator Attachments cc: R. R. Assa, Braidwood Project Manager - NRR S. G. Dupont, Senior Resident Inspector - Braidwood B. Clayton, Branch Chief- Region III Office of Nuclear Facility Safety - IDNS k nin\bntwthitepry2\2 I

ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED SUPPLEMENTAL CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 A. DESCRIPTION OF THE PROPOSED CHANGE The index page VIII, and 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)," of Technical Specification 3.4.9.3 in Comed's March 30, 1994, submittal will be replaced with updated pages.

IL DESCRIPTION OF THE CURRENT REQUIREMENT The current index page VIII states that Figure 3.4-4a is applicable to 10 Effective Full Power Years (EFPY) for Unit 1. Figure 3.4-4a describes the nominal Pressurizer Power Operated 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 l 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 fracture toughness requirements for reactor vessels under specified operating conditions, and Regulatory Guide (RG) 1.99,

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

D. NEED FOR REVISION OF THE REQUIREMENT In order to allow suflicient margin to 10 CFR 50 Appendix G limits over the duration of applicability of Figure 3.4-4a, while accounting for a 60 pounds per square inch gauge (psig) pressure channel measurement uncertainty, from WCAP-

. 10529 Revision 1," Cold Overpressure Mitigating System," it was necessary to reduce the duration of applicability of Figure 3.4-4a from 32 EFPY to 5.37 EFPY.

This curve was previously revised in a submittal dated June 14,1994, to account for a 50 F thermal transport ef1'ect for the postulated heat injection transient. The curve included with this request accounts for both the thermal transport effect a' nd the pressure channel measurement uncertainty.

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The change in duration of applicability for Figure 3.4-4a required a corresponding change in the duration of applicability for this curve stated in the appropriate index page.

E. DESCRIPTION OF THE REVISED REQUIREMENT 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),"

in Comed's March 30,1994, submittal will be replaced with the attached Figure 3.4-4a " Nominal PORV Pressure Relief Setpoint Versus RCS Temperature for the Cold Overpressure Protection System Applicable Up to 5.37 EFPY (Unit 1)."

Index page VIII in Comed's March 30,1994, submittal will be replaced with the

attached index page VIII.

F. BASES FOR THE REVISED 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 10 CFR 50. Appendix G provides the fracture toughness requirements for reactor vessels under specified operating conditions, and RG 1.99 specifies the procedure acceptable to the NRC staff for calculating the pressure limits required by Appendix G.

The duration of applicability of Figure 3.4-4a was reduced from 32 EFPY to 5.37

. EFPY in order to allow sufficient margin to 10 CFR 50 Appendix G limits while accounting for a 60 psig pressure channel measurement uncertainty.

t G. IMPACT OF THE PROPOSED CHANGE This change reduces the duration of applicability of the LTOPS curve in order to provide sufficient margin to Appendix G limits taking into account pressure channel measurement uncertainties. This revised curve is more conservative than what was initially submitted because it accounts for a 50 thermal transport effect and pressure channel measurement uncertainty. No new equipment is being installed, no new system interfaces are being created, no existing system interfaces are being modified; therefore, this change has no negative impact on i any system or operating mode.

H. SCHEDULE REQUIREMENTS The proposed amendment dated March 30,1994, to Technical Specification 3.4.9.3 is required for Braidwood Unit 1 to exceed 4.5 EFPY. Braidwood Unit 1 is currently predicted to reach 4.5 EFPY on July 30,1994. Comed requests that this amendment be approved prior to July 30,1994.

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ATTACHMENT B MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES

.NPF-72 AND NPF-77 4

4 REVISED PAGES:

VIII 3/4 4-40a (Figure 3.4-4a)(New 3/4 4-40 page) 1 1

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