ML20024F454
| ML20024F454 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/07/1983 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8309090396 | |
| Download: ML20024F454 (4) | |
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TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 37401 400 Chestnut Street Tower II September 7,1983 Director of Nuclear Reactor Regulation Attention:
Ms. E. Adensam, Chief Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Ms. Adensam:
In the Matter of
)
Docket No.
50-327 Tennessee Valley Authority
)
328 Enclosed are the responses to the NRC questions regarding the Sequoyah Nuclear Plant unit 2, cycle 2 reload license application submitted by my July 1, 1983 letter to you. These responses were discussed with Carl Stahle of your staff and Mike Tokar, NRC, iu a telephone conversation on September 2, 1983 If you have any questions concerning this matter, please get in touch with J. E. Wills at FTS 858-2683 Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills, Man ger Nuclear Licensing Sworn to d subscri
,d be ore me 1983 this /7 day of Notagf Public
'F My Comission Expires
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Enclosure cc:
U.S. Nuclear Regulatory Comission (Enclosure)
Region II Attn:
Mr. James P. O'Reilly Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 k
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'983-TVA SOTH ANNIVERSARY 8309090396 830907 PDR ADOCK 05000328 An Equal Opportunity Employer P
ENCLOSURE TVA RESPONSE TO NRC QUESTIONS ON SEQUOYAH UNIT 2 CYCLE 2 RELOAD ON SUBMITTAL DATED JULY 1, 1983 NRC Ouestion No. 1 Please complete Table 1 of the Reload Safety Evaluation Report (RSER) by listing the approximate expected,end-of-cycle-2 burnup for each of the fuel regions.
TVA Response Region 1 Region 2 Region 3 Region 4 24,100 27,400 23,100 11,500 Approximate EOL BU Cycle 2 (MWD /MTU)
NRC Ouestion No. 2 It is stated on page 2 of the RSER that Wet Annular Burnable Absorber (WABA) rods are to be used. It is our understanding that Westinghouse will implement a surveillance program, details of which are described in the NRC safety evaluation of the WABA topical report, for the first two plants to utilize WABAs. Please confirm whether Sequoyah unit 2 is subject to and will follow the provisions of that surveillance program.
TVA Response
- Yes, the surveillance program will be followed.
NRC Question No. 3 During a recent refueling of McGuire unit 1, several broken nonfuel bearing i
component (NFBC) holddown springs were discovered. Some of the springs had double fractures "such that a... section might move... into the flow of the coolant ' system" (according to Reportable Occurrence Report No.
369/83-11). It is our understanding that Sequoyah unit 2 has an NFBC holddown springs of the same design as McGuire's and which, therefore, mr.v be subject to the same failure mode. Aside from the potential effects of loose parts, broken holddown springs could have an impact on UHI flow, with resultant effects on the LOCA peak cladding temperatures.
t t-Therefore, with regard to the potential effects on cycle 2 operation, discuss the results of post-irradiation examinations of the MFBC holddown springs during the current refueling outage. How many broken springs were observed? How many had double fractures? Discuss the cause and remedy of i
such failures. For cycle 2 operation, either provide reasonable assurance that broken springs will not occur or show that the potential effects of loose parts, UHI flow restrictions, and increases LOCA peak cladding temperature are insignificant.
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grid ass;mblics were prec:nted by W2stinghouas to Massrs. Miko Tokar, John Vogisweda, and Lenny Olshtn of the NRC on May 16, 1981. The identified NRC participants appeared to be satisfied with the information presen tad.
The modified grids have been incorporated into the fuel assemblies for the same nuclear plants listed in the response above.
NRC Ouestion No. 5 It is stated in the RSER that Region 4 fuel has been designed according to the fuel performance model in WCAP-8785. That model (PAD 3 3) was approved subject to certain restrictions identified in the NRC Staff's SER on the WCAP report. In an addendum to the report, the restrictions were revised, but still apply in areas dealing with conductance, fuel / cladding surface roughness, and cladding creep. Please state whether these restrictions were adhered to in the analysis of Sequoyah Unit 2 Cycle 2 fuel.
TVA Response The Westinghouse fuel performance model, as documented in WCAP-8720 (Proprietry) and WCAP-8785 (Non-Proprietary) was used in Sequoyah Unit 2
- Region 4 fuel design. This model was approved by the NRC with restriction identified during the NRC review of Addendum 1 to WCAP-8720. These restrictions, as modified by the NRC in the SER for Addendum 1 to WCAP-8720 (letter from Harold Bernhard (NRC) to E. P. Rahe (Westinghoure) dated July 20, 2982) have been adhered to in the analysis of Sequoyah Unit 2 Cycle 2.
NRC Question No. 6 The RSER indicates that the Westinghouse fuel rod revised internal pressure design basis, as described in WCAP-9964, is satisfied for Cycle 2 operation for all fuel regions. As noted in the NRC Staff's safety evaluation of that report, the proposed revised criterion is not sufficient to assure acceptable consequences for transient and accident conditions, and an amended criterion was eventually approved by the staff. Please state whether the amended criterion as stated in a letter from J. Stolz (NRC) to T. M. Anderson (Westinghouse) is not satisfied, provide justification.
TVA Response The NRC approved Westinghouse rod internal pressure criteria is documented in WCAP-8963-P-A (WCAP-8964, Non-Proprietry), and includes the amendment made by the NRC in the SER for WCAP-8963 (letter from J. Stolz (NRC) to T. Anderson (Westinghouse) dated May 19, 1978). The rod internal pressure criteria with the NRC amendment is satisfied for Sequoyah Unit 2 Cycle 2 fuel operation.
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