View source for DCL-11-055, Supplement to License Amendment Request 11-02, Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs).

{{Adams
| number = ML111120056
| issue date = 04/21/2011
| title = Supplement to License Amendment Request 11-02, Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs).
| author name = Becker J R
| author affiliation = Pacific Gas &amp; Electric Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000275, 05000323
| license number = DPR-080, DPR-082
| contact person = 
| case reference number = DCL-11-055
| document type = Letter type:DCL, Technical Specifications
| page count = 5
| project = 
| stage = Supplement
}}

=Text=
{{#Wiki_filter:Pacific Gas and Electric Company April 21, 2011 PG&E Letter DCL-11-055 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No*. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Supplement to License Amendment Request 11-02 James R. Becker Site Vice President Diablo Canyon Power Plant Mail Code 104/5/601
: p. O. Box 56 Avila Beach, CA 93424 805.545: 3462 Internal:
691.3462 Fax: 805.545.6445 10 CFR 50.90 Revision to Technical Specification 3.7.1, "Main Steam Safety Valves (MSSVs)" Dear Commissioners and Staff: By Pacific Gas and Electric Company (PG&E) Letter DCL-11-018, "License Amendment Request 11-02 Revision to Technical Specification 3.7.1, 'Main Steam Safety Valves (MSSV),'" dated February 17, 2011, PG&E submitted a license amendment request (LAR) (ADAMS Accession No. ML 110480870) to receive NRC review and approval of a change in method of evaluation due to the use of a new analysis methodology for establishing the reduced power range (PR) neutron flux high setpoint for one inoperable MSSV. In addition, the LAR requests the removal of a one-time note listed in Technical Specification (TS) Table 3.7.1-1, specific to Unit 2 Cycle 15, which is no longer applicable or needed. TS Bases B 3.7.1. is revised to reflect the new analysis methodology.
This includes the use of the RETRAN-02W computer code for establishing the reduced PR neutron flux high setpoint (for one inoperable MSSV) listed in TS Table 3.7.1-1. There is no proposed change to the PR neutron flux high setpoint; the value of 87 percent rated thermal power (RTP) listed in TS Table 3.7.1-1 for one inoperable MSSV will remain. PG&E is clarifying that the revision to the TS Bases is a revision to the Final Safety Analysis Report Update (FSARU). The TS Bases are incorporated into the FSARU by reference based on FSARU Section 16.1 stating the following:
The TS Bases provide the bases or reasons for these technical specifications . other than those covering administrative controls.
In accordance with 10 CFR 50.36, the TS Bases are not part of the TS, and are included by reference in this section of the FSAR Update in accordance with 10 CFR 50.34 and 10 CFR 50.36. Changes to the TS Bases are processed A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway.
Comanche Peak. Diablo Canyon. Palo Verde. San Onofre. South Texas Project. Wolf Creek Document Control Desk April 21, 2011 Page 2 PG&E Letter DCL-11-055 in accordance with TS 5.5.14, "Technical Specifications (TS) Bases Control Program." PG&E has included a markup of FSARU Sections 15.2.7.3 and 15.2.16 as a supplement to LAR 11-02 in the Enclosure to this letter to provide clarity that analyses for off normal conditions for MSSVs are discussed in the TS Bases Section B3.7 .1. PG&E will implement the proposed FSARU revision within 90 days of the issuance of the license amendment associated with LAR 11-02. This supplement does not affect the conclusions or no significant hazards consideration of LAR 11-02. PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this letter. If you have any questions or require additional information, please contact Tom Baldwin at 805-545-4720.
I declare under penalty of perjury that the foregoing is true and correct. Executed on April 21 , 2011 . James R. Becker Site Vice President phs3/50381928 Enclosures cc: Diablo Distribution cc/enc: Gary W. Butner, Branch Chief, California Department of Public Health Elmo E. Collins, NRC Regional Administrator, Region IV Michael S. Peck, NRC, Senior Resident Inspector James T. Polickoski, NRR Project Manager Alan B. Wang, NRR Project Manager A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway.
Comanche Peak. Diablo Canyon. Palo Verde. San Onofre. South Texas Project. Wolf Creek Enclosure PG&E Letter DCL-11-055 Proposed Final Safety Analysis Report Update Markup DCPP UNITS 1 & 2 FSAR UPDATE transient and never drops below its initial value. The pressurizer safety valves are not actuated in these transients.
Figures 15.2.7-9 and 15.2.7-10 show the transient responses for the total loss of load at BOL for the overpressure concern. No credit is taken for the pressurizer spray, pressurizer power-operated relief valves, or steam dump. The pressurizer and main steam safety valves are modeled as described in assumptions 7 and 8. The initial pressurizer pressure includes the pressurizer pressure control uncertainty to maximize the peak pressure. The reactor is tripped on the high pressurizer pressure signal. This case results in the highest RCS peak pressure among all cases. The peak RCS pressure is below 110 percent of the design value. Figures 15.2.7-11 and 15.2.7-12 show the transient responses for the total loss of load at BOL for the overpressure concern, assuming full credit for the pressurizer spray and the pressurizer power-operated relief valves. No credit is taken for the steam dump. The models for the pressurizer and main steam safety valves and the initial pressurizer pressure are the same as those used in the above case. The reactor trip due to high neutron flux is not credited in order to maximize the peak steam generator pressure.
The reactor is tripped on the high pressurizer pressure signal. This case results in the highest steam generator peak pressure among all cases. The peak steam generator pressure is below 110 percent of the design value. Reference 8 presents additional results for a complete loss of heat sink including loss of main feedwater. This report shows the overpressure protection that is afforded by the pressurizer and steam generator safety valves. Technical Specification 3.7.1 establishes reduced plant operating power limits for off normal conditions when one or more MSSVs are inoperable to ensure a loss of load event does not result in overpressurization of the steam generators. When two or more MSSVs are inoperable per ReS loop, the reduced power limits are established using a conservative energy balance algorithm established in the Westinghouse Nuclear Safety Advisory Letter NSAL-94-001 as documented in Reference
: 21. In order to evaluate off normal plant operation with a single inoperable MSSV on one or more steam generator loops , an additional spectrum of loss of load analyses are performed as documented in Reference
: 22. These analyses use the RETRAN-02W code to analyze the BOL loss of load overpressure case as discussed in this section and which represents the limiting case for challenging the steam generator peak pressure limit. These analysis results , as summarized in the Technical Specification Bases 3.7.1 , credit the overtemperature delta temperature reactor trip to demonstrate that the specified reduced operating power limit ensures that the available relief capacity with one inoperable MSSV per loop maintains the peak steam generator pressure below 110 percent of the design value. 15.2.7.4 Conclusions 15.2-27 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 9. J. S. Shefcheck, Application of the THINC Program to PWR Design, WCAP-7359-L, August 1969 (Proprietary), and WCAP-7838, January 1972. 10. R. L. Haessler, et aI., Methodology for the Analysis of the Dropped Rod Event, WCAP-11394 (Proprietary) and WCAP-11395 (Non-Proprietary), April 1987. 11. Deleted in Revision 16. 12. Deleted in Revision 16. 13. Deleted in Revision 16. 14. Deleted in Revision 18. 15. Deleted in Revision 18. 16. Westinghouse letter PGE-98-503, Diablo Canyon Units 1 & 2 Inadvertent ECCS Actuation at Power Analysis -PSV Operability Issue, January 13, 1998. 17. Westinghouse Letter NSAL-02-11, Reactor Protection System Response Time Requirements, July 29, 2002 18. Westinghouse Letter PGE-02-072, Diablo Canyon Units 1 & 2 Evaluation of Reactor Trip Functions for Uncontrolled RCCA Withdrawal at Power, December 13, 2002. 19. RETRAN-02 Modeling and Qualification for Westinghouse Non-LOCA Safety Analyses, WCAP-14882-P-A (Proprietary), April 1999, and WCAP-15234-A (Non-Proprietary), May 1999. 20. Westinghouse Letter NSAL-07-1 0, Loss-of-Normal Feedwater/Loss-of-Offsite AC Power Analysis PORV Modeling Assumptions, November 7,2007. 21. PG&E Design Calculation N-115, "Reduced Power Levels for a Number of MSSVs Inoperable", dated 3/14/94. 22. Westinghouse Letter PGE-1 0-43 , "Diablo Canyon Units 1 and 2 Loss of Load / Turbine Trip Analysis with One Inoperable MSSV per Steam Generator", September 2, 2010. 15.2-50 Revision 19 May 2010}}