ML070950041

Request for Additional Information on the Proposed Amendment to Revise Fuel Storage Pool Boron Concentration

ML070950041
Person / Time
Site:	San Onofre
Issue date:	04/12/2007
From:	Kalyanam N NRC/NRR/ADRO/DORL/LPLIV
To:	Rosenblum R Southern California Edison Co
Kalyanam N, NRR/DORL/LP4, 415-1480
References
TAC MD1405, TAC MD1406
Download: ML070950041 (9)

Text

April 12, 2007 Mr. Richard M. Rosenblum Senior Vice President and Chief Nuclear Officer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128

SUBJECT:

SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 -

REQUEST FOR ADDITIONAL INFORMATION ON THE PROPOSED AMENDMENT TO REVISE FUEL STORAGE POOL BORON CONCENTRATION (TAC NOS. MD1405 AND MD1406)

Dear Mr. Rosenblum:

By letter dated April 28, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML061220701), and as supplemented by letters dated November 13 and December 22, 2006 (ADAMS Accession Nos. ML063210425 and ML063610042, respectively), Southern California Edison submitted an application to change the San Onofre Nuclear Generating Station, Units 2 and 3, technical specifications related to fuel storage pool boron concentration. The proposed change will increase the minimum allowed boron concentration of the spent fuel pool and allow credit for soluble boron, guide tube inserts made from borated stainless steel, and fuel storage patterns in place of Boraflex.

After reviewing your request, the Nuclear Regulatory Commission staff has determined that additional information outlined in the enclosure is needed to complete the review. We discussed this information with your staff by telephone and they agreed to provide the additional information requested by April 27, 2007.

If you have any questions, please contact me at (301) 415-1480.

Sincerely,

/RA/

N. Kalyanam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-361 and 50-362

Enclosure:

Request for Additional Information cc: See next page

ML063210425 and ML063610042, respectively), Southern California Edison submitted an application to change the San Onofre Nuclear Generating Station, Units 2 and 3, technical specifications related to fuel storage pool boron concentration. The proposed change will increase the minimum allowed boron concentration of the spent fuel pool and allow credit for soluble boron, guide tube inserts made from borated stainless steel, and fuel storage patterns in place of Boraflex.

After reviewing your request, the Nuclear Regulatory Commission staff has determined that additional information outlined in the enclosure is needed to complete the review. We discussed this information with your staff by telephone and they agreed to provide the additional information requested by April 27, 2007.

If you have any questions, please contact me at (301) 415-1480.

Sincerely,

/RA/

N. Kalyanam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-361 and 50-362

Enclosure:

Request for Additional Information cc: See next page DISTRIBUTION:

PUBLIC RidsNrrDorl (CHaney/JLubinski) RidsNrrLAJBurkhardt LPLIV r/f RidsNrrDorlDpr RidsNrrPMNKalyanam RidsAcrsAcnwMailCenter RidsNrrDorlLpl4 (THiltz) RidsRgn4MailCenter (TPruett)

RidsOgcRp RidsNrrDssSpwb SSun, DSS ADAMS Accession No.: ML070950041 *No major change from Staff provided RAI.

OFFICE NRR/LPL4/PM NRR/LPL4/LA NRR/DSS/SPWB* NRR/LPL4/BC NAME NKalyanam JBurkhardt GCranston THiltz DATE 4/10/07 4/11/07 3/22/07 4/12/07 San Onofre Nuclear Generating Station Units 2 and 3 cc:

Mr. Daniel P. Breig Director, Radiologic Health Branch Southern California Edison Company State Department of Health Services San Onofre Nuclear Generating Station P.O. Box 997414, MS 7610 P.O. Box 128 Sacramento, CA 95899-7414 San Clemente, CA 92674-0128 Resident Inspector/San Onofre NPS Mr. Douglas K. Porter, Esquire c/o U.S. Nuclear Regulatory Commission Southern California Edison Company Post Office Box 4329 2244 Walnut Grove Avenue San Clemente, CA 92674 Rosemead, CA 91770 Mayor Mr. David Spath, Chief City of San Clemente Division of Drinking Water and 100 Avenida Presidio Environmental Management San Clemente, CA 92672 P.O. Box 942732 Sacramento, CA 94234-7320 Mr. James T. Reilly Southern California Edison Company Chairman, Board of Supervisors San Onofre Nuclear Generating Station County of San Diego P.O. Box 128 1600 Pacific Highway, Room 335 San Clemente, CA 92674-0128 San Diego, CA 92101 Mr. James D. Boyd, Commissioner Mark L. Parsons California Energy Commission Deputy City Attorney 1516 Ninth Street (MS 31)

City of Riverside Sacramento, CA 95814 3900 Main Street Riverside, CA 92522 Mr. Ray Waldo, Vice President Southern California Edison Company Mr. Gary L. Nolff San Onofre Nuclear Generating Station Assistant Director - Resources P.O. Box 128 City of Riverside San Clemente, CA 92764-0128 3900 Main Street Riverside, CA 92522 Mr. Brian Katz Southern California Edison Company Regional Administrator, Region IV San Onofre Nuclear Generating Station U.S. Nuclear Regulatory Commission P.O. Box 128 611 Ryan Plaza Drive, Suite 400 San Clemente, CA 92764-0128 Arlington, TX 76011-8064 Mr. Steve Hsu Mr. Michael R. Olson Department of Health Services San Diego Gas & Electric Company Radiologic Health Branch 8315 Century Park Ct. CP21G MS 7610, P.O. Box 997414 San Diego, CA 92123-1548 Sacramento, CA 95899 March 2006

San Onofre Nuclear Generating Station Units 2 and 3 cc:

Mr. A. Edward Scherer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128 March 2006

REQUEST FOR ADDITIONAL INFORMATION SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 (SONGS 2 AND 3)

SOUTHERN CALIFORNIA EDISON DOCKET NOS. 50-361 AND 50-362 REQUEST TO REVISE FUEL STORAGE POOL BORON CONCENTRATION

1. Criticality Accident Analyses Page 13 of Enclosure 2 to Reference 1 indicated that specific accidents considered for criticality analyses included fuel assembly drop, fuel misloading in the racks and spent fuel pool (SFP) water temperature changes. Also, Section 3.2.9 of Attachment L to Reference 1 discussed criticality analyses for various postulated accidents including: (1) pool water temperature accident and (2) fuel assembly misplacement. For the fuel assembly misplacement, the following accidents were discussed: (1) fuel assembly dropped horizontally on top of the rack, (2) fuel assembly dropped vertically into a storage location already containing a fuel assembly, (3) fuel assembly dropped to the SFP, and (4) fuel misloading in either Region I or II.

Furthermore, Section 5.6 of Attachment L to Reference 1 provided the quantitative results of analyses for a pool heat-up accident and fuel mishandling accident. However, the respective Attachments G and H containing the proposed Bases pages for SONGS 2 and 3 did not include adequate and sufficient information regarding applicable safety analyses. Specifically, Page B 3.7-71 stated in the applicable safety analyses section that this accident was analyzed assuming the misloading of one fresh assembly with the maximum permissible enrichment.

The discussion of other accidents analyzed or considered as discussed in the above paragraph were either deleted (a fuel assembly dropped vertically into a storage location already containing a fuel assembly), or omitted (such as the pool water temperature accident).

Modify the applicable safety analyses section on page B 3.7-71 to appropriately reflect the analyses considered for determining the required boron concentration in the SFP.

2. Licensee-Controlled Specification (LCS) 4.0.100 vs. NUREG-1432 Page 17 of Enclosure 2 to Reference 1 stated that LCS 4.0.100 was consistent with NUREG-1432, Standard Technical Specifications Combustion Engineering Plants. It should be noted that the requirement of the LCS documentation was discussed in Section 4.3(f) of NUREG-1432, which stated that new or partially spent fuel assemblies (SFAs) with a discharge burnup in the unacceptable range of Figure [3.7.18-1] would be stored in compliance with the Nuclear Regulatory Commission (NRC)-approved plant-specific information such as specific document containing the analytical methods, title, date or specific configuration or figure.

The proposed LCS 4.0.100 for SONGS 2 and 3 was contained in Attachment I and J, respectively. Based on the review of the proposed LCSs, it appeared to the NRC staff that LCS 4.0.100 did not contain the analytical methods used to determine the associated storage

patterns for the SFAs and thus, was not fully consistent with Section 4.3(f) of NUREG-1432 that requires the inclusion of analytical methods in the LCS.

Clarify the inconsistency between LCS 4.0.100 and Section 4.3(f) of NUREG-1432 regarding inclusion of analytical methods.

Also, Reference (5) in Section 7, REFERENCES, had the title: NRC: Standard Technical Specifications Combustion Engineering Plants Bases (NUREG-1432, Vol. 2, Rev. 3). The NRC staff found that Reference (5) misquoted the title of the Standard Technical Specifications (TS) since Bases sections do not contain information addressing LCS documentation requirements for the SFA storage patterns. The documentation requirements are discussed in Standard TS Section 4.3(f), as referenced in the first paragraph above. The licensee should correct the error by deleting Bases and replacing Vol. 2" with Vol. 1" in the title for Reference (5).

3. Inconsistency in the Note for LCS 4.0.100 The note on page 4.0.100-1 of Attachments I and J to Reference 1 indicated that the LCS was listed by revision number and date in TS 4.3.1. The NRC staff found that the LCS was listed in TS 4.3.1.k and TS 4.3.1.l (of Attachments C and D to Reference 1); however, the TSs listed the LCS revision number without specifying the date.

Clarify the inconsistency between the LCS note and TS 4.3.1 regarding inclusion of the LCS date.

4. No SONGS 1 Fuel to Be Stored in Region I Racks LCS Subsection 4.0.100.3 (in Attachments I and J to Reference 1) requires that SONGS 1 fuel not be stored in Region I racks. This requirement is consistent with the criticality analysis discussed in Section 4.6 of Attachment L that indicates that Unit 1 fuel has not been analyzed to be stored in Region I. However, Table 2-2, Spent Fuel Data (Each Unit), of Attachment L listed and/or SONGS 1 14x14 fuel in the row of the Table specifying the fuel types for the SFAs in Region I.

Modify Table 2-2 to correct the information that is inconsistent with LCS 4.0.100.3 and Section 4.6 of the criticality analysis.

5. Omissions of the Cooling Times in Figures Specifying Minimum Burnup for Category I Fuel Figures I-1 through I-6 in Attachments I and J to Reference 1 specify the relationship of the minimum burnup and initial uranium enrichment. There are five lines in each figure to show the minimum required burnup at the fuel cooling times of 0, 5, 10, 15, and 20 years. In each of the lines, the symbols to represent the cooling times were omitted.

Modify the figures to correct the omissions.

6. Control Element Assembly Lifetime Analysis Section 2.2.1 of Attachment L to Reference 1 indicated that a control element assembly (CEA) lifetime analysis would ensure that the CEAs used were suitable for use in the SFP. Before using any CEA, a visual inspection would be performed.

Discuss the CEA lifetime analysis and the associated acceptance criteria for selection of CEA to be used in the SFP. Discuss the criteria of the visual inspection for the CEA that would pass the inspection.

7. CASMO-3 Benchmarking Section 3.1.1 of Attachment L to Reference 1 indicated that CASMO-3 was used to perform the fuel depletion analyses and criticality analyses for fuel assemblies in the SFP, and CASMO-3 was compared with industry critical experiments with good agreement.

List the values of the bias and 95/95 uncertainty in the bias of the critical benchmark calculation for CASMO-3, and discuss how the CASMO-3 code bias and uncertainty were used in the fuel depletion analyses and criticality analyses for fuel assemblies in the SFP.

8. Computer Codes (KENO-V.a and the Related Codes) Benchmarking Section 3.1.2 of Attachment L to Reference 1 indicated that KENO-V.a was benchmarked by Southern California Edison against industry critical experiments performed by Babcock and Wilcox (B&W). The bias and 95/95 uncertainty in the bias for CELLDAN, NITAWL-II and KENO-V.a, and 27 group cross-section library were 0.00814 and 0.00172, respectively. The NRC staff found that these bias and uncertainty were different from that on page 9.1-13 of the updated final safety analysis report (UFSAR), which indicated that the bias and 95/95 uncertainty in the bias for CELLDAN, NITAWL-II and KENO-V.a, and 27 group cross-section library were 0.00928 and 0.00148, respectively. The bias and uncertainty in the UFSAR were also determined by analyses of B&W critical experiments.

Discuss why the values of the bias and uncertainty predicted for the KENO-V.a-related codes are different as they are shown in Section 3.1.2 of Reference 1 and Section 9.1 of the UFSAR, and justify that the values for the bias and uncertainty discussed in Section 3.1.2 of Reference 1 are adequate for use in the criticality analysis.

9. Manufacturing Tolerances Section 3.2.2 of Attachment L to Reference 1 discussed the reactivity effects of the manufacturing tolerances for the components including enrichment, stainless steel thickness and minimum cell inner dimension. It specifically indicated that the reactivity effect of the manufacturing tolerance of the storage cell pitch was only considered for Region I.

Discuss the storage rack configurations in Regions I and II to justify that the storage cell pitch tolerance effect needs to be considered for Region I only.

10. Minimum Axial Burnup Section 3.2.4 of Attachment L to Reference 1 indicated that the CASMO-3 depletions were performed based on the following reactor operating conditions: reactor outlet temperature of 600 ºF; constant beginning of cycle fuel temperature of 1200 ºF; and constant soluble boron of 1,000 parts per million (ppm). It indicated that the operating conditions so selected were to eliminate axial burnup effects.

Discuss the effects of the reactor outlet temperature, fuel temperature, and soluble boron on the axial burnup, and justify that the operating conditions assumed in the CASMO-3 depletion calculations would eliminate the axial burnup effects.

11. SIMULATE-3 All-Rod-Out 2D Depletions Section 3.2.5 of Attachment L to Reference 1 discussed the calculation of the axial burnup bias for two SIMULATE-3 cases. For the all-rod-out 2D cases, it indicated that at 0, 10, 20, 30, 40, 50, and 60 gigawatt days per ton (GWD/T), the 2D was expanded to 3D.

Explain how the 2D SIMULATE-3 calculations were expanded from 2D to 3D calculations, and justify the adequacy of the calculations in determining the axial burnup effects.

12. CEA Bias Section 3.2.6 of Attachment L to Reference 1 discussed the calculations for CEA bias. It indicated that CASMO-3 (through SIMULATE-3) has accurately predicted SONGS 2 and 3 CEA bank worth measurements.

Explain how CASMO-3 was benchmarked through SIMULATE-3 to predict the CEA worth measurements.

13. Boron Concentration of 370 ppm Under Non-Accident Conditions Section 5.1 of Attachment L to Reference 1 indicated that the soluble boron concentration needed to maintain effective multiplication factor (Keff) less than or equal to 0.95, including biases and uncertainties, under non-accident conditions was 370 ppm.

Discuss the boron worths in terms of ppm/k used in determining the required boron concentration of 370 ppm, and justify the adequacy of the values of the boron worth used.

Identify the limiting storage patterns with the associated enrichment that would result in a maximum required boron concentration of 370 ppm.

14. Reactivity Equivalence Uncertainty Section 5.2 of Attachment L to Reference 1 indicated that the reactivity equivalence uncertainty was 0.00 k at 0 GWD/T and 0.01 k at 30 GWD/T, linear with burnup. It also indicated that this reactivity equivalence uncertainty was approved by the NRC staff. Furthermore, it indicated that the soluble boron needed to compensate for the reactivity equivalence uncertainty was 178 ppm.

Provide the name of the author, title, and date of the NRC document approving the reactivity equivalence uncertainty. Discuss how the boron concentration of 178 was calculated from the reactivity equivalence uncertainty that was expressed in terms of k and a linear relationship with burnup.

15. Discharge Burnup Uncertainty Section 5.3 of Attachment L to Reference 1 indicated that the soluble boron needed to compensate for the fuel assembly discharge burnup uncertainty was 218 ppm. This boron compensation was based on a discharge burnup uncertainty of 7 percent for SONGS 2 and 3 fuel assemblies.

Discuss information including plant-specific fuel discharge burnup data to demonstrate the adequacy of the uncertainty of 7 percent used in the criticality analysis. Discuss how the boron compensation of 218 ppm was calculated from the fuel discharge burnup uncertainty of 7 percent.

Reference

1. Letter from B. Katz (SCE) to NRC, Docket Nos. 50-361 and 50-362, Amendment Application Numbers 243 and 227, Proposed Change Number (PCN) 556, Request to Revise Fuel Storage Pool Boron Concentration, San Onofre Nuclear Generating Station Units 2 and 3, dated April 28, 2006.