ML20236L526

Forwards List of Licensee Clarifications/Changes to NRC Staff'S SE Re FOL TS Amends 137 & 129 Which Allow Operation at Increased Reactor Core Power Level of 2775 Mwt

ML20236L526
Person / Time
Site:	Farley
Issue date:	07/06/1998
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M98120, TAC-M98121, NUDOCS 9807130050
Download: ML20236L526 (5)

Text

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~

Dave Morey S:uttern Nuclear y Vice President Operating C:mp:ny

• Farley Project P.O. Box 1295 Birmingham, Alabama 35201 Tel 205.992.5131 l

July 6, 1998 SOUTHERN h COMFANY Energy to Serve Your World*

Docket Nos.: 50-348 10 CFR 50.4 50-364 U. S. Nuclear Regulatory Commission A'ITN.:~ Document Control Desk Washington,DC 20555 Joseph M. Farley Nuclear Plant Review of NRC Staff Safety Evaluation Associated With Facility Operating Licenses and Technical Soccifications Amendment Nos.137 (Unit 1) and 129 (Unit 2)

Ladies and Gentlemen:

NRC letter dated April 29,1998 (TAC Nos. M98120 and M98121) amended the Joseph M. Farley Nuclear Plant (FNP) Facility Operating Licenses and Technical Specifications to allow operation at an increased reactor core power level of 2775 megawatts thermal (Mwt). Southern Nuclear Operating Company (SNC) has reviewed the NRC Staff Safety Evaluation associated with Arnendment 137 to Facility Operating License No. NPF-2 and Amendment 129 to Facdity Operating License No. NPF-8. As a result of this review, several clarifications of the NRC Staffs Safety Evaluation are necessary. The SNC proposed clarifications / changes are provided as an atachment to this letter. In addition, the Staff should note that Clarification No.16 revises the SNC commitment to inspect all FNP steam generator U-bends in rows 1 and 2 (refer to SNC letter dated August 5,1997). It is SNC's intent that this commitment apply to FNP Unit 2 only, since it j is currently planned to replace steam generators in Unit I during the refueling outage following power uprate implementation.  !

Should you have any questions, please advise.

Respectfully submitted, 4 ffCOf 0hlpu Dave Morey l RWS/MGE/ cit:pwrup43. doc Attachment , cc: Mr. L. A. Reyes, Region II Administrator L Mr. J. I. Zimmerman, NRR Project Manager Mr. T. M. Ross, Plant Sr. Resident Inspector 9007130050 900706 PDR ADOCK 05000340 p PDR

O ATTACHMENT

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Joseph M. Farley Nuclear Plant Review of NRC Staff Safety Evaluation Associated With Facility Operating Licenses and Technical Specifications Amendment Nos.137 (Unit 1) and 129 (Unit 2) l i l i i I l

1 l 1 , ATTACHMENT Joseph M. Farley Nuclear Plant Review of NRC Staff Safety Evaluation Associated With Facility Operating Licenses and Technical Specifications Amendment Nos.137 (Unit 1) and 129 (Unit 2) Number Proposed Clarification / Change Reason / Reference

1. SE (p.1 - Section 1.0) should state: 'were documented in Completeness Westinghouse WCAP-14723, 'Farley Nuclear Plant Units I and 2 Isce 1, Attachant Power Uprate Project NSSS Licensing Report,' dated January yg)

I997 and the 'Farley Nuclear Plant Units I and 2 Pour Uprate Projed BOP Licensing Report' submitted by SNC with the February 14,1997, request."

2. SE (p. 2 - Section 3.0) should state: 'The transient analyses Typographical error presented rely heavily on analysis performed in the past to support other NRC-approved licensing actions (overpressure overpowr delta temperature (OPAT). . . ."

3. SE (p. 3 - Section 3.1.3) should state: %NC stated that a Typographical error calculation has been performed to determine the new hot leg switchover (HLSO) time and minimum hot leg reealculauen recircu'ation flow. . . . The new hot leg reenleulauen recirculation minimum flow for the worst break and single failure is 89.1 lbm/sec. This hot leg reenleule6en recirculation ,

minimum flow. . . ."

4. SE (p. 4 - Section 3.2) should state: "A reduced core flow of Typographical error 86,000 gpm per loop was employed in the analysis, which is and clarification associated with :: -"' 2:= " " c' 97,2^^ e minimum measuredflow of 87,800 gpm and 20 percent steam generator (* yT

                                                                                                           '^

6l 0 6 (SG) tube plugging. . . . 65)

5. SE (p. 5 - Section 3.2.3) should state: 'The dropped RCCA ar.d Typographical error the -"' -- "y statically misaligned assembly were analyzed at the uprated conditions. . . ."

6. SE (p. 5 - Section 3.2.5) should state: 'The transient was Clarification analyaed evaluated for the power uprate. The event was analyzed at the uprated conditions dile transitioning to VANTAGE-3 (see Ref.1 Attachment V, Section 6.2.0, p.6-fuel using the RTDP. . . ."

7. SE (p. 7 - Section 3.2.12) should state: " Additionally, although Typographical error the automatic actuation of the PORV is not considered safety- and clarification related, the assumulauen aduation circuits are routed :;eparately; there are two separate C'-- !E p--d transmitters dica are (see Ref.15'

                    ,                  ,,         ,, g              ,,             ,j ,, gg,,, Attachment 1, p.14)

IE, powered from IE power supplies, with Class IE procured relays."

8. SE (p.13 - Section 3.4.3) should state: "Although the power Typograpidcal error J !

uprate resulted in a small electrical load increase of the reactor coolant pumps and eharging condensate pumps on non-Class IE l 4160 V buses. . ." l ' l 1

                                                              'A-1                                                       j

s. Attachment - Review of NRC Staff Safety Evaluation

9. SE (p.18 - Section 3.5.1.4 - 5th paragraph) should state: Correction

                                                                                                           " cooling of the reactor core must occur through the use of the safny atmospheric reliefvalves."                                     (see Ref.15'y         ,

6.2.19.1, p. 6-155)

10. SE (p. 25 - Section 3.5.1.9) should state: "In this mesecsment, Typographical enor SNP SNC assumed the very same assumptions. . . ."

l. I1. SE (p. 26 - Section 4.0, under F_ Ins) should state: "using the Typographical error j procedures of Paragraph (c) of the revised ITS rule,10 CFR (see SE, References l 50.61 (Ref. 9 22). Section, p. 3)

12. SE (p. 29 - Section 4.3) should state: "the limiting components at Typographical error the power uprate conditions are identified on page 45 of Reference (see SE, References Section, p.1)

, 13. SE (p. 30 - Section 4.3) should state: "the staff concludes that the Typographical error l reactor internal components at FNP Units I and 2 will vetain remain within the allowable limits. . . ."

14. SE (p. 30 - Section 4.4) should state: "The components resiewed Typographical error include the full length (F/L) L406 L-106A CRDMs. . . ."

g V. Section 5.4, p. 5-12)

15. SE (p. 31 - Section 4.5) should state: "The results indicate that Correction l the fatigue usage of the U-bend will may exceed the ==p==

limit of 1.0 in 13.7 years after the implementation of the power (see Ref. 6 Attachment 1 p. 55 and Ref.1

( uprate at FNP Units 1 and 2 in 1998 outages. SNC concluded ttachinent V Section that the U-bend needs may need to be monitored and would may 5.7.3' p. 5-15) require sorac type of corrective action at that time, as necessary "

16. SE (p. 32 - Section 4.5.1.1.3) should state: "To ensure stress Clarification i corrosion cracking does not increase in the small radius U-bends, SNC enspectsplans to inspect all U-bends in the rows 1 and 2 at (seeRef.6 Attachment the refueling outage after the power uprate."

I,p.20) SNC decided to replace Additionally, it is SNC's intent that this commitment apply to YN' "'** 8'"*'*" Unit 2 only, since it is currently planned to replace steam generators in Unit i during the refueling outage following *I

                                                                                                                   "                                                                      "' '*y 'l 14,1997.

17. SE (p. 35 - Section 4.7) should state: "The maximum CUFs at Correction the limiting locations are 0.94 for the surge spray nozzle and 0.78 (seeRef.6, Attachment for the pressurizer upper head and shell."

I, Table B, p. 58) A-2 l

l i Attachment - Review of NRC Staff Safety Evaluation

18. SE (p. 38 - Section 4.10.3): The last sentence of the second Operation of the paragraph ("With two* SFP cooling trains in operation the SFP SFPCS was clarified by can be maintained below 140*F.") and footnote 4 ("4 SNC RAI response Simultaneous operation of both SFPCS trains is not a normal dated January 23,1998.

practice at FNP.") should be deleted. (see Ref.15, SE (p. 39 - Section 4.10.3): The second sentence of the fifth Attachment I, p. 6 - 9) paragraph in this section ("The cleanup system will be able to withstand the temperatures of pool water for partial and full core off-loads, as long as the two cleanup trains are in operation.") should be deleted. While analyses wre performed assunung two train SFPCS operation, simultaneous operation ofboth trains is not a normal practice at FNP; therefore, all references to tw train operation wre eliminatedfrom the Farley uprate BOP licensing report and A41 responses.

19. SE (p. 39 - Section 4.10.3) should state: "When only one train is Clarification in operation, the temperature of bulk water will may exceed 140*F' * *" Isee Ref.15' , p. 8 - 9)

Attachments

20. SE (p. 40 - Section 4.13) sitould state: "The effects of all changes Clarification due to plant operations at the proposed uprated power level on the EQ 8Pp1ies oniy to design and-EQ of mechanical components as well as EQ of electrical equipment.

sqfety-related electrical egaipment were evaluated. . . . The existing parameters used for 7"$in;; evaluating mechanical components inside and outside containment remain bounding. . .

21. SE (p. 42 - Section 4.13) should state: " Based on its resiew, the Clarification staff concludes that plant operation at the proposed uprated power EQ applies only to level will have an insignificant or no impact on the EQ-ef safety-electrical equipment.

related mechanical components inside or outside the containment.

22. SE (p. 42 - Section 4.15) should state: "The first change to Trip Correction Setpoint in TS Table 3.3-4 will meresse decrease the Trip Setpoint for Functional Unit 5.a, Turbine Trip and Feed Water (seeRef.1' Attachment able 6.7-2, p. 6 -

Isolation from Steam Generator Water Level - High - High, from 6 79.2 percent to 784 78.5 percent of narrow range instrument span."

23. SE (p. 53 - Section 7.1.2) should state: "The staff will impose Typographical error four three new license conditions. . . ."

(see Appendix C of NPF-2 and NPF-8)

24. SE (p. 56 - Section 7.1.12) shodbate: "The new value reflects Typographical error i

the current calculated total water and steam volume of the RCS at (see Ref.1, Attachment

                                                                                     $67.2 F."

III, TS 5.4.2) l 25. SE (Table 4.1.1-1): The "SNC Projected 1/4T USE (ft-lb)" "% Typographical error Decrease" values for " Low. Shell B69191" and " Low. Shell B6919-2" should be reversed. 'g) A-3 - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _}}