ML15218A237
| ML15218A237 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/04/1985 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Tucker H DUKE POWER CO. |
| References | |
| NUDOCS 8503210025 | |
| Download: ML15218A237 (4) | |
Text
I MAR 4 1985 Docket No. 50-270 DISTRIBUTION EDotket&l!e RIngram NRC PDR HNicolaras L PDR Gray File Mr. H. B. Tucker ORB#4 Rdg EBlackwood Vice President - Nuclear Production HThompson HOrnstein Duke Power Company OELD P. 0. Box 33189 EJordan 422 South Church Street BGrimes Charlotte, North Carolina 28242 JPartlow ACRS-10
Dear Mr. Tucker:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON OCONEE UNIT 2 CYCLE 8 RELOAD We have reviewed your December 19, 1984 submittal and proposed changes to the Technical Specifications which are required to support the operation of Oconee Unit 2 at full rated power during Cycle 8. We have also reviewed your January 22, 1985 letter as revised on February 20, 1985 concerning changes to the Oconee Startup Physics Test Program. With assistance from our consultant, Brookhaven National Laboratory, we have evaluated your letters and have determined that we need additional information to complete our evaluation. On February 21 and 27, 1985, we held a conference call with your staff to discuss these concerns. Since these questions have already been discussed with your staff, we request that you respond to the enclosed list of questions at least six weeks before startup of Unit's 2 Cycle 8. This request for information affects fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, JOHIN S $POLZ' John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
Request for Additional Information cc w/enclosure:
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Duke Power Company cc w/enclosure(s):
Mr. William L. Porter Duke Power Company P. 0. Box 33189 422 South Church Street Office of Intergovernmental Relations Charlotte, North Carolina 28242 116 West Jones Street Ral-eigh, North Carolina 27603 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621 Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission, Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Heyward G. Shealy, Chief Bureau of Radiological Health Sou'th Carolina Department of Health and Environmental Control Regional Radiation RepresentativeBull Street EReg-ion RaIatoereettv Columbia, South Carolina 29201 EPA Region IV 345 Courtland Street, N.E.
Atlanta, Georgia 30308 Mr. J. C. Bryant Senior Resident Inspector U.S. Nuclear Regulatory Commission Route 2, Box 610 Seneca, South Carolina 29678 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Manager, LIS NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 J. Michael McGarry, III, Esq.
Bishop, Liberman, Cook, Purcell & Reynolds 1200 17th Street, N.W.
Washington, D. C. 20036
0 ENCLOSURE Questions Concerning the Oconee-2 Cycle Reload Submittal I. The safety evaluation of the Mark BZ fuel assembly design required that a licensee referencing the Mark BZ design must submit:
a) A plant specific analysis of combined seismic and LOCA loads according to Appendix A to SRP 4.2 and b) An analysis of the reduction of the maximum allowable peaking and FA h for transitional mixed cores having both Mark B and Mark BZ fuel assemblies.
Since Oconee-2 Cycle 8 is a transitional Mark B/Mark BZ core, please provide the required analysis.
- 2. What were the results from the visual examination of the nine Oconee-1 Mark BZ demonstration assemblies?
- 3. Have there been any additional holddown spring failures at the Oconee units since the Oconee-2 Batch-7 failures?
- 4. How does the new control rod assembly design differ from the present Oconee-2 Cycle 7 design? Has this design been approved and is there any operating experience with this design?
- 5. What was the revision to the creep collapse analysis of the beta-quenched fuel rods in the advanced cladding pathfinder assemblies?
- 6. In previous reload analyses (e.g., Oconee-3 Cycle 8) the TACO2 fuel melt limit analysis employed conservative lower tolerance limit values for the initial pellet diameter and density. This conservatism has been deleted from the thermal design section of the Oconee-2 Cycle 8 report. Has this conservatism been removed from the thermal design analysis?
- 7. Is the - 5% margin between the 30,000 EFPH estimated residence time and the 31,400 EFPH cladding collapse time (provided in Table 4-1) sufficient to account for uncertainties in the estimated fuel residence time?
- 8. Was a cladding stress and strain analysis performed for the gray APSRs?
- 9. Is the Batch-10 Mark BZ reduced prepressurization above the minimum specified HZP prepressure used in the cladding stress calculations?
- 10. What are the changes in maximum FQ and FXY for a core operating with gray APSRs relative to one operating with black APSRs?
-2
- 11.
Figures 3.5.2-2, 3.5.2-5 and 3.5.2-8 of the proposed technical specifica tion revision indicate that the rod position limits (RPLs) for cycle 8 operation of Oconee 2 apply for the entire cycle of operation, rather than the three different exposure windows used in earlier cycles.
Explain how these RPLs were determined?
- 12.
Three sets of LOCA linear heat rate limits are given in Tables 7-2 and 7-3 of the submittal. The limits apply during the periods 0-25 EFPD, 25-65 EFPD, and 65 EFPD to the end of the cycle. The values given in the tables are identical to those given in Table 7-2 of the Oconee-1 Cycle 9 reload submittal, however, the exposure windows are considerably differ ent; i.e., 0-30+10/-0 EFPD, 30+10/ 250+10 EFPD and 250+10 EFPD to the end of the cycle. In view of the consider-ble similarities between the Oconee-2 Cycle 8 and Oconee-1 Cycle 9 reloads, please explain the differences in these limits?
- 13.
The key parameters used for accident analyses presented in the FSAR are compared to predicted values for Cycle 8 in Table 7.1.
While virtually all of the items considered in the "Key Safety Parameter Checklist" (Table 8-1 of NFS-1001, Rev. 4) are addressed in Table 7.1 and other tables in the submittal, the Minimum Tripped Rod Worth (MTRW) available in the case of a steamline break is not given. What is the value of the MTRW and how does it compare to the value used in the reference analysis, and to the values presented in the calculation of the Shutdown Margin in Table 5.2?
- 14. The sign of the calculated temperature rate of chanoe,aa/,included in the Justification for the Change in the MTC TEST described in Attachment 2 of reference 1, is incorrect. Please change the tabulated values of AQ/AT and the sign of the coefficient ofLa/1T in the equation foro((532).
What error is introduced by assuming thatAa/AT is determined only by the temperature rate of change of the moderator density?
REFERENCE
- 1. "Oconee Nucle.ar Station, Unit 2 Docket No. 50-270," Letter, H.B. Tucker (Duke Power) to J.F. Stolz (NRC), dated January 22, 1985.