Text

Issuance of Amendment Nos. 213 and 205 Request to Revise Fuel Storage Pool Boron Concentration
	ML072550175
Person / Time
Site:	San Onofre
Issue date:	09/27/2007
From:	Kalyanam N; NRC/NRR/ADRO/DORL/LPLIV
To:	Rosenblum R; Southern California Edison Co
	Kalyanam N, NRR/DORL/LP4, 415-1480
Shared Package
ML072550164	List: ML072550175; ML072550183;
References
	TAC MD1405, TAC MD1406
	Download: ML072550175 (45)
	v • d • e

September 27, 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 -

ISSUANCE OF AMENDMENTS RE: REQUEST TO REVISE FUEL STORAGE POOL BORON CONCENTRATION (TAC NOS. MD1405 AND MD1406)

Dear Mr. Rosenblum:

The Commission has issued the enclosed Amendment No. 213 to Facility Operating License No. NPF-10 and Amendment No. 205 to Facility Operating License No. NPF-15 for San Onofre Nuclear Generating Station, Units 2 and 3, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated April 28, 2006, and as supplemented by letters dated November 13 and December 22, 2006, May 7, June 15, July 27, and September 11, 2007.

The amendments revise TSs 3.7.17, "Fuel Storage Pool Boron Concentration," 3.7.18, "Spent Fuel Assembly Storage," and 4.3, "Fuel Storage." The changes 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.

A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

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

Enclosures:

1. Amendment No. 213 to NPF-10

2. Amendment No. 205 to NPF-15

3. Safety Evaluation cc w/encls: See next page

PKG ML072550164, AMD ML072550175, TS Pages ML072550183

Staff supplied SE with minor editorial changes only.

- See previous concurrence OFFICE NRR/LPL4/PM NRR/LPL4/LA NRR/DSS/SPWB NRR/DSS/SBPB NAME NKalyanam JBurkhardt **

GCranston**

JSegala*

DATE 9/26/07 9/17/07 9/20/07 3/1/07 OFFICE NRR/DE/EMCB NRR/DIRS/ITSB OGC - NLO NRR/LPL4/BC NAME KManoly*

TKobetz**

AHodgdon **

THiltz DATE 6/15/07 9/26/07 9/25/07 9/27/07

May 2007 San Onofre Nuclear Generating Station Units 2 and 3 cc:

Mr. Raymond W. Waldo, Vice President, Nuclear Generation Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128 Mr. Douglas K. Porter, Esquire Southern California Edison Company 2244 Walnut Grove Avenue Rosemead, CA 91770 Dr. David Spath, Chief Division of Drinking Water and Environmental Management California Dept. of Health Services 850 Marina Parkway, Bldg P, 2nd Floor Richmond, CA 94804 Chairman, Board of Supervisors County of San Diego 1600 Pacific Highway, Room 335 San Diego, CA 92101 Mark L. Parsons Deputy City Attorney City of Riverside 3900 Main Street Riverside, CA 92522 Mr. Gary L. Nolff Assistant Director - Resources City of Riverside 3900 Main Street, 4th Floor Riverside, CA 92522 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 Mr. Michael J. DeMarco San Diego Gas & Electric Company 8315 Century Park Ct. CP21G San Diego, CA 92123-1548 Director, Radiologic Health Branch State Department of Health Services P.O. Box 997414, MS 7610 Sacramento, CA 95899-7414 Resident Inspector San Onofre Nuclear Generating Station c/o U.S. Nuclear Regulatory Commission Post Office Box 4329 San Clemente, CA 92674 Mayor City of San Clemente 100 Avenida Presidio San Clemente, CA 92672 Mr. James T. Reilly Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128 Mr. James D. Boyd, Commissioner California Energy Commission 1516 Ninth Street (MS 31)

Sacramento, CA 95814 Brian Katz Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92764-0128 Mr. Steve Hsu Department of Health Services Radiologic Health Branch MS 7610, P.O. Box 997414 Sacramento, CA 95899-7414 Mr. A. Edward Scherer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128

SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA DOCKET NO. 50-361 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 213 License No. NPF-10 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Southern California Edison Company, et al.

(SCE or the licensee), dated April 28, 2006, and as supplemented by letters dated November 13 and December 22, 2006, May 7, June 15, July 27, and September 11, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C(2) of Facility Operating License No. NPF-10 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 213, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 180 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Thomas G. Hiltz, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: September 27, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 213 FACILITY OPERATING LICENSE NO. NPF-10 DOCKET NO. 50-361 Replace the following pages of the Facility Operating License No. NPF-10 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 3.7-30 3.7-30 3.7-32 3.7-32 3.7-33 3.7-33 3.7-34 3.7-34 3.7-34a 3.7-34b 4.0-4 4.0-4 4.0-4a (3)

SCE, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of San Onofre Nuclear Generating Station, Units 1 and 2 and by the decommissioning of San Onofre Nuclear Generating Station Unit 1.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Southern California Edison Company (SCE) is authorized to operate the facility at reactor core power levels not in excess of full power (3438 megawatts thermal).

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 213, are hereby incorporated in l

the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 213

SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA DOCKET NO. 50-362 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 205 License No. NPF-15 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Southern California Edison Company, et al.

(SCE or the licensee), dated April 28, 2006, and as supplemented by letters dated November 13 and December 22, 2006, May 7, June 15, July 27, and September 11, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C(2) of Facility Operating License No. NPF-15 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 205, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 180 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Thomas G. Hiltz, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: September 27, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 205 FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 Replace the following pages of the Facility Operating License No. NPF-15 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 3.7-30 3.7-30 3.7-32 3.7-32 3.7-33 3.7-33 3.7-34 3.7-34 3.7-34a 3.7-34b 4.0-4 4.0-4 4.0-4a (3)

SCE, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear materials as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of San Onofre Nuclear Generating Station, Units 1 and 3 and by the decommissioning of San Onofre Nuclear Generating Station Unit 1.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Southern California Edison Company (SCE) is authorized to operate the facility at reactor core power levels not in excess of full power (3438 megawatts thermal).

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 205, are hereby incorporated in the license.

l Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 205

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 213 TO FACILITY OPERATING LICENSE NO. NPF-10 AND AMENDMENT NO. 205 TO FACILITY OPERATING LICENSE NO. NPF-15 SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 DOCKET NOS. 50-361 AND 50-362

1.0 INTRODUCTION

By application dated April 28, 2006 (Reference 1), and as supplemented by letters dated November 13 (Reference 2), and December 22, 2006 (Reference 3), May 7 (Reference 4),

June 15 (Reference 5), July 27 (Reference 6), and September 11, 2007 (Reference 7),

Southern California Edison Company (SCE, the licensee) requested changes to the Technical Specifications (TSs) for San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS 2 and 3). The supplements dated November 13 and December 22, 2006, May 7, June 15, July 27, and September 11, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on June 6, 2006 (71 FR 32606).

The changes revise the TSs requirements for spent fuel storage to remove credit for use of Boraflex, introduce borated stainless steel guide tube inserts (GT-lnserts) into the stored fuel, take credit for soluble boron, increase the required concentration of soluble boron, and provide allowable storage patterns to be controlled by the Licensee Controlled Specifications (LCS).

The changes revise TS 3.7.17, Fuel Storage Pool Boron Concentration, TS 3.7.18, Spent Fuel Assembly Storage, TS 4.3, Fuel Storage, and LCS 4.0.100 Fuel Storage Patterns.

Specifically, the proposed changes revise the minimum allowed boron concentration of the spent fuel pool and implement a Fuel Storage Program to allow credit for soluble boron, GT-Inserts, and Fuel Storage Patterns in place of Boraflex.

2.0 EVALUATION The evaluation of the license amendment request (LAR) covers three specific areas:

1) criticality/boron dilution, 2) structural/seismic analysis, and 3) spent fuel pool criticality analysis.

2.1 Criticality/Boron Dilution 2.1.1 Regulatory Evaluation The Nuclear Regulatory Commissions (NRC's) regulatory requirements related to the content of TS are set forth in Appendix A of Part 50 of the Title 10 of the Code of Federal Regulations (10 CFR 50) (Reference 8), General Design Criterion (GDC) 62, Prevention of criticality in fuel storage and handling, states, Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. The NRC has established a five percent subcriticality margin (k-effective < 0.95) for nuclear power plant operators to comply with GDC 62.

Part 68 of 10 CFR 50, Criticality accident requirements, states in subpart 50.68(b)(4), If no credit for soluble boron is taken, the k-effective [Keff, the effective neutron multiplication factor]

of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

Generic Letter (GL) 96-04, Boraflex Degradation In Spent Fuel Pool Storage Racks, dated June 26, 1996, was issued over concerns related to: (1) gamma radiation-induced shrinkage of Boraflex and the potential to develop tears or gaps in the material; and (2) long-term Boraflex performance throughout the intended service life of the racks resulting from gamma irradiation and exposure to the wet pool environment. The NRC staff requested the licensees to assess the ability of Boraflex to maintain a five percent subcriticality margin, and to submit a plan describing proposed actions if the five percent subcriticality margin could not be maintained by the Boraflex material due to current or projected material degradation. This LAR will eliminate credit taken in the SFP criticality analysis for the use of Boraflex while permitting credit for the use of an increase in soluble boron.

Accordingly, the licensee is proposing changes to the SONGS 2 and 3 TS which will support the requirements of 10 CFR 50.68 for crediting the use of soluble boron in the SFP. This SE addresses the dilution aspects of the review for approval of a change to TS Section 3.7.17, Fuel Storage Pool Boron Concentration. LCO 3.7.17 is revised to increase the minimum boron concentration from 1850 to 2000 parts per million (ppm).

2.1.2 Technical Evaluation 2.1.2.1 Boron Dilution Analysis The licensee proposes to increase the TS minimum soluble boron concentration, per TS 3.7.17, Fuel Storage Pool Boron Concentration, in the SFP as stated above from 1850 to 2000 ppm.

The licensee selected this value to provide sufficient margin (in combination with the geometric configuration of the fuel stored in the pool) to protect the design-basis subcriticality criterion of Keff < 0.95. Based on the licensees criticality analysis, the actual minimum soluble boron concentration required to maintain the SFP Keff < 0.95 during normal operation (including uncertainties, burnup, with a 95 percent probability at a 95 percent confidence level) is 970 ppm. SCE conservatively established a limit for final boron concentration following a boron dilution accident of 1700 ppm. It is noted that the SFP soluble boron concentration will be maintained at a minimum of 2000 ppm to protect against a fuel handling accident (maintain k-effective < 0.95) but a fuel handling accident is not assumed concurrent with a dilution event.

The licensee performed a boron dilution analysis to determine plant systems that could potentially dilute the SFP to either of these values, starting with an initial soluble boron concentration of 2000 ppm. The following paragraphs review the boron dilution analysis to ascertain that the licensee has adequate time to mitigate postulated dilution events such that the concentration of soluble boron in the SFP does not drop below 1700 ppm. The licensee determined, through walkdowns and evaluation of required actions, that 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months was needed to identify a given dilution source upon receipt of a Fuel Handling Building (FHB) high-high sump level alarm. It was determined that another 1/2 hour was needed to isolate the inflow to the SFP. Therefore, 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is needed to detect and isolate inflow to the SFP upon receiving an FHB sump Hi-Hi level alarm upon which the dilution analysis is based.

The licensee used flow rates from postulated events with fixed dilution times as inputs and calculated the final boron concentration of the pool. The fixed times were based on level alarms. The time the dilution flow ends is the time assumed by the licensee to terminate the dilution flow. The event is shown to be mitigated if the calculated boron concentration in the pool at termination of dilution flow is at an acceptable value. Data was entered and results calculated using a spreadsheet program. Cases were analyzed where unborated makeup does not overflow the pool and when unborated makeup causes the pool to overflow. These cases were calculated as a feed-and-bleed process with the SFP volume remaining essentially constant. Basic parameters and assumptions used in the analysis are: neglect cask and transfer pool volumes (the gates to these are normally open, assuming them to be closed conservatively utilizes a smaller total volume, that of the SFP only); initial SFP boron concentration is set at the proposed minimum limit of 2000 ppm; analysis assumes thorough mixing of unborated water added to the SFP resulting from SFP cooling flow.

Analysis was performed postulating dilution flow from the following sources: Primary Water System, Nuclear Service Water (NSW) addition through a service hose, NSW pipe break near the SFP, rupture of one tube in the SFP Heat Exchanger, pipe break in the Fire Water Header, pipe break in the SFP Cooling Water Return Header during normal conditions, pipe break in the SFP Cooling Water Return Header during fuel shipment. Each SONGS Unit has a separate SFP. A discussion of eight specific cases analyzing credible dilution sources presented in the licensees analysis follows and applies to either Unit.

2.1.2.2 Credible Dilution Paths Spent Fuel Pool Makeup System Dilution from the SFP Makeup System was analyzed and presented as the bounding case. The event postulated is a hypothetical alignment of normal makeup that is permitted to proceed unchecked and overflows the SFP. The dilution analysis considers makeup to the SFP from three unborated water sources. Makeup to the SFP can also be accomplished using water from the refueling water storage tank but this would not result in dilution because this source is borated. Therefore only makeup water from unborated sources were analyzed. These sources are: (1) SFP makeup using primary makeup water (demineralized water) from Primary Plant Makeup Storage Tanks SA1415MT055 and -056 to the SFP cooling pump, (2) SFP makeup using potentially demineralized water from Radwaste Primary Tanks (RPTs) SA1901MT065,

-066, and (3) SFP makeup using potentially demineralized water from RPTs SA1901MT067 and

-068. Of these three sources, the first one presented is the bounding case. When correctly aligned, this source can provide unborated water at a rate of 160 gallons per minute (gpm) via one of two primary makeup tank pumps. One of the two pumps remains in standby while one is operating. Makeup from any of the RPTs is bounded by makeup from the Primary Plant Makeup Storage Tanks because the design flow of the RPT pump (140 gpm) is less than that of one of the primary makeup tank pumps (160 gpm). Therefore, the licensee analyzed the case of 160 gpm of unborated makeup water being supplied to the SFP for this analysis.

The analysis shows acceptable results for the bounding case of 160 gpm inflow of unborated water. That is, that criticality limits would not be exceeded before plant operators could terminate the dilution. The licensee stated that makeup can be isolated 60 minutes after an SFP high-level (HL) alarm or, in the unlikely event of not receiving this alarm, 90 minutes after an FHB sump high-high (HI-HI) alarm which is 94 minutes after SFP overflow to the sump.

Numerical results that follow are based on time after receipt of HI-HI level sump alarm: time to reach 970 ppm (required for Keff < 0.95 under normal conditions) is over 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months , time to reach 1700 ppm (required for Keff < 0.95 during a fuel handling accident) soluble boron is 4 1/2 hours.

This affords operators sufficient time to terminate the event before exceeding the criticality margin. The NRC staff verified the time to dilution through independent calculation. In the bounding case presented above for makeup to the SFP from unborated sources and cases bounded by it, the licensee described a rigorous administrative procedure that must be followed for system alignment. This includes the operation of manual valves, manual start of the Primary Makeup Water Pump, and boron concentration sampling by the plant Chemistry Division. The NRC staff agrees that due to the rigorous nature of the procedure that must be followed for adding unborated makeup to the SFP, the probability of SFP dilution is very low. If this event were to develop, however, it has been shown by analysis that there is adequate time for operators to terminate dilution flow before criticality limits are exceeded.

Primary Water System Makeup with 1 gpm Pipe Break This case represents normal makeup due to postulated minor system leakage. A leak rate of 1 gpm is presented in the analysis. This was selected arbitrarily as an example of minor system leakage from any liquid system interfacing with the SFP. The unborated makeup constitutes the dilution. This case is presented for comparison with cases that follow involving pipe breaks that analyze larger outflows with subsequent unborated makeup. Calculated results for dilution from this source based on time after receipt of HI-HI sump level alarm are as follows: time to reach 970 ppm (required for Keff < 0.95 under normal conditions) is over 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months , time to reach 1700 ppm (required for Keff < 0.95 during a fuel handling accident) soluble boron is approximately 4 1/2 hours. The NRC staff verified the time to dilution through independent calculations. Due to identical initial conditions at the start of this case and the case above, the time to dilution is identical in both cases. The NRC staff concludes that this is adequate time for operators to isolate the dilution inflow to the SFP before exceeding criticality limits.

Nuclear Service Water Addition through a Service Hose Nuclear Service Water (NSW) is demineralized water (unborated) and is used at both units water service stations (hose stations) and washdown stations. The NSW storage tank has a gross capacity of 26,550 gallons. The dilution analysis shows that it would take 263,400 gallons of unborated water to dilute the SFP from 2000 to 970 ppm boron. The capacity of the NSW storage tank by itself would not be enough to dilute the SFP. However, makeup to the NSW storage tank is made from a much larger source, which contains enough unborated water to dilute the SFP. Makeup to the NSW storage tank is supplied from the Makeup Demineralizer (MUD) tanks SA1417MT266, -267, and -268. Each MUD tank has a nominal capacity of 535,000 gallons with level controlled at a minimum of 75 percent. In this case, it is postulated that an operator inadvertently leaves one hose running at a hose station in the vicinity of the SFP. The estimated flow of unborated water from a NSW hose spilling into the SPF is 50 gpm. Results for dilution from this source based on time after receipt of HI-HI sump level alarm are as follows: time to reach 970 ppm (required for Keff < 0.95 under normal conditions) is over 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months , time to reach 1700 ppm (required for Keff < 0.95 during a fuel handling accident) soluble boron is approximately 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months . The NRC staff verified the time to dilution through independent calculations. The NRC staff concludes that there is adequate time for operators to isolate the dilution inflow to the SFP from the hose before reaching criticality limits.

Pipe Crack in the Nuclear Service Water Header in the Vicinity of the SFP The licensee postulated this event because it is conceivable that a crack could develop in this piping, which is in the vicinity of the SFP and flow unborated water into the SFP. However, the flow rate was calculated to be only 30 gpm from this source. It is therefore bounded by the case above for the NSW hose station.

Tube Break in the SFP Heat Exchanger The SFP heat exchangers are cooled by component cooling water (CCW). The CCW contains no boron. CCW operating pressure is greater than that of the SFP cooling system. A tube leak, therefore, would introduce unborated water into the SFP cooling system and from there into the SFP. The CCW system contains a surge tank that is designed to accommodate fluid volumetric changes and to maintain a static pressure head at each CCW pump suction. The licensee determined that leakage from a single failed tube would result in automatic fill of the surge tank from a demineralized source with makeup from NSW without triggering a low-level (LL) alarm in the control room. This presents a credible dilution scenario. Therefore, leakage of one failed SFP heat exchanger tube, calculated to be 90 gpm, was postulated. Numerical results that follow are based on time after receipt of HI-HI level FHB sump alarm: time to reach 970 ppm (required for Keff < 0.95 under normal conditions) is 46 hours5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months , time to reach 1700 ppm (required for Keff < 0.95 during a fuel handling accident) soluble boron is over 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The NRC staff verified these time to dilution results by independent calculations. The NRC staff concludes that adequate time is available for operators to mitigate the event.

Pipe Break in the Fire Water Header Two fire protection hose stations are located in the vicinity of the SFP as noted above. These hose stations are supplied by the fire water system through a 4-inch header that branches to 2.5-inch lines to each hose station. For conservatism, the licensee postulated a break in the main header. Since the fire water system has moderate pressure, a critical crack was postulated. The pipe is a 4-inch standard weight pipe. The critical crack flow was calculated to be 110 gpm. Numerical results that follow are based on time after receipt of HI-HI level FHB sump alarm: time to reach 970 ppm (required for Keff < 0.95 under normal conditions) is over 37 1/2 hours, time to reach 1700 ppm (required for Keff < 0.95 during a fuel handling accident) soluble boron is over 6 1/2 hours. The NRC staff verified these time to dilution results by independent calculations. The NRC staff concludes that adequate time is available for operators to mitigate the event.

Pipe Break in the SFP Cooling Water Return Header This 12-inch pipe header in the vicinity of the SFP branches off to other smaller diameter pipes.

For conservatism, the licensee postulated a pipe break in the main header. Since the SFP cooling water system has moderate pressure, a critical crack was assumed. The critical crack flow was calculated to be 130 gpm. This break flow results in SFP level reduction below the LL alarm setpoint. Sixty minutes after LL alarm, makeup is started at full makeup flow, of unborated water, of 160 gpm. Numerical results that follow are based on time after receipt of HI-HI level FHB sump alarm: time to reach 970 ppm (required for Keff < 0.95 under normal conditions) is 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months , time to reach 1700 ppm (required for Keff < 0.95 during a fuel handling accident) soluble boron is over 3 1/2 hours. The NRC staff verified these time to dilution results by independent calculations. The NRC staff concludes that adequate time is available for operators to mitigate the event.

Pipe Break in the SFP Cooling Water Return Header During Fuel Shipment The licensee has analyzed the above scenario to occur during fuel shipment out of the SFP.

This is an infrequent operation in which the SFP low level is lower than during normal operation.

This was accounted for in the SFP dilution analysis for this scenario. The licensee indicated that the likelihood of SFP overflow during this operation is extremely low because operations personnel are present. Due to the operators presence, the evaluation of this scenario assumes that makeup would be isolated within 60 minutes after the SFP HI level alarm instead of 90 minutes after Fuel Building sump HI-HI level alarm. Numerical results that follow are based on time after receipt of SFP HI level alarm: time to reach 970 ppm (required for Keff < 0.95 under normal conditions) is over 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months , time to reach 1700 ppm (required for Keff < 0.95 during a fuel handling accident) soluble boron is slightly more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The NRC staff verified these time to dilution results by independent calculation. The NRC staff concludes that adequate time is available for operators to mitigate the event.

Feed and Bleed Dilution of the SFP In a letter dated December 22, 2006, the licensee responded to the NRC staffs question regarding a feed and bleed dilution of the SFP. In the case of a small leak (1 to 2 gpm) from the SFP concurrent with a small leak into the SFP at the same rate so the level does not change, the licensee determined that the dilution event will result in a soluble boron concentration of 40-80 ppm lower than the value attained in the previous weekly surveillance (SR 3.7.17.1). This reduction in boron concentration is outside the tolerance of the surveillance measurement of +/- 9 ppm and will be detected. The 40-80 ppm reduction in boron concentration between the weekly surveillance interval will not challenge the 1700 ppm limit.

The NRC staff concludes that this scenario is acceptable, as it will be detectable during the weekly surveillance and there is adequate time to mitigate the event.

2.1.2.3 Bounded Events The following events were evaluated and determined either to be bounded by one or more of the above cases, to be not credible, or to result in unborated makeup flow equivalent to the bounding case.

Loss of Offsite Power (LOOP)

Of the dilution sources listed above, the fire water header pipe break and the SFP cooling heat exchanger tube leak scenarios are capable of providing non-borated water to the SFP during a LOOP. This is because the fire water system is equipped with a diesel-driven fire pump, and the CCW pumps (which provide the motive force for a SFP cooling heat exchanger tube leak) are automatically loaded on the emergency diesel generators. Plant annunciators for SFP level and temperature are powered by 125 Volt direct current circuitry which would be available for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following a LOOP. If it became necessary during a LOOP, SFP boron concentration could be increased by manual addition of dry boric acid. The effects of the fire water header pipe break and the SFP cooling heat exchanger tube leak scenarios would be the same during a LOOP and are evaluated above.

Back Flushing the Fuel Pool Filter The fuel pool filter is periodically back flushed. Upstream and downstream isolation valves are closed for this process and nitrogen is used as the motive fluid for back flushing. There is, therefore, no boron dilution path during this alignment.

Resin Flush/Fill for SFP Ion Exchanger Resin transfer operation is performed approximately once each fuel cycle to flush spent resin from the SFP ion exchanger. The procedure for resin charging and flushing requires isolating the exchanger using rigorous administrative controls. Unborated water from the nuclear condensate system is used for flushing. Since the ion exchanger is isolated during this process the dilution path with condensate would be limited to the volume of the ion exchanger and the associated process piping up to the isolation valves. This volume of water, calculated to be 2000 gallons, will transfer to the SFP upon resumption of operation of the ion exchanger.

Analysis shows that this volume of unborated water would provide a very minimal dilution and is, therefore, bounded by the cases presented above.

Deployment of Fire Hoses in the SFP Area The SFP room is served by two fire hose stations with one hose per station. The fire hoses are fed by an unborated water source. The flow rate from one fire hose is 150 gpm. With two hoses deployed, the total flow would be 300 gpm. An SFP dilution by either one or two fire hoses is an event controlled by fire fighting personnel and readily observable by plant operators. The fire loading in the SFP room is very low per SONGS 2 and 3's Fire Hazard Analysis. The quantity of fire water needed to suppress a fire in the SFP room is only 360 gallons. This is a very small volume of water and would not significantly dilute the SFP. This event is, therefore, bounded by other events quantified above.

Pipe Break in the SFP Cooling Water Suction Header This event was calculated to result in a 107 gpm leakage flow. This event is bounded by the event evaluated above for a pipe break in the SFP Cooling Water Return Header which results in a leakage flow of 130 gpm.

Pipe Break in the SFP Purification Pump Discharge Header Dilution flow from a critical crack in this nominal 3-inch piping was calculated to be 38 gpm.

This event is, therefore, bounded by events evaluated above.

Pipe Break in the Demineralized Water Makeup Line This nominal 3-inch makeup pipe is connected to the 12-inch SFP cooling line. The effects of a break in this line are, therefore, bounded by a break in the 12-inch SFP cooling water return header which is evaluated above.

Pipe Breaks Due to Tornado or Hurricane Events The licensee reported in the dilution analysis that the effects of tornado or hurricane events were reviewed and found not to result in any rupture of piping adjacent to or associated with the SFP or related systems. Dilution due to tornado or hurricane, therefore, is not a credible event.

Crack in the SFP Liner Plate The SFP liner plate is a 3/16-inch thick stainless steel welded plate. Leakage through the liner plate, which is possible, would result in unborated makeup water addition. Behind the watertight liner plates are multiple chases which are connected by drains to a leak detection sump, which drains to the FHB sump, which is equipped with a HL alarm. Observation of the leakage from the drains allows identification of the approximate location of the leak. The transient volume of the leakage to annunciate the Fuel Building HI-HI level alarm is 1000 gallons. The leakage rate associated with SFP liner plate damage has been evaluated in a previous SONGS 2 and 3 licensing submittal for re-racking (License Amendment Applications 146 and 130, dated July 29, 1996). The leak rate was identified to be 49 gpm. If a leak is detected, a makeup, depending on the source chosen by operators, at a rate equivalent to or less than that analyzed above for the bounding case is initiated. Results for the time to reach 1700 ppm therefore are either bounded by or equivalent to that of the 160 gpm makeup event.

Leakage through the SFP Gates The SFP is connected to two other cavities. They are the cask pool and the fuel transfer pool.

These cavities can be isolated from the SFP by bulkhead gates. The bulkhead gates are normally open. Leakage through the closed gates, however, would require subsequent makeup which represents a potential dilution scenario. The potential for SFP gate leakage was a subject of NRC Office of Inspection and Enforcement (IE)Bulletin 84-03 dated August 24, 1984. SCE responded to this IE Bulletin stating that due to the design of the seals and low-seal pressure alarms, it is not credible to postulate a leak in these pneumatic water seals that would exceed the normal makeup capacity to the SFP. Unchecked makeup flow equivalent to the flow provided by normal makeup (160 gpm) was evaluated above and shown to be a recoverable event.

2.1.2.4 Summary The NRC staff reviewed the licensees evaluation of the above-postulated dilution events. In each case, the dilution event would be identified, either by the SFP high-water level alarm, the Fuel Building sump HI-HI level alarm, or by plant personnel performing routine plant rounds, with sufficient time to mitigate the event.

Based on the review of the licensees boron dilution analysis described above, the NRC staff finds that adequate time is available for detection and mitigation of events capable of diluting the SFP from the new TS minimum soluble boron concentration of 2000 ppm to the minimum soluble boron concentration required to maintain Keff < 0.95 (1700 ppm). Therefore, the dilution aspects implicit within 10 CFR 50.68 and GDC 62 are met. Also, the dilution aspects of the proposed change to TS SR 3.7.17 to increase the concentration of soluble boron in the SFP from 1850 to 2000 ppm is acceptable.

2.2 Structural/Seismic Analysis 2.2.1 Regulatory Evaluation The NRC staffs acceptance criterion for design of the safety-related structures being capable to withstand effects of natural phenomena is based on the regulations Reference 8, Criterion 2, Design bases for protection against natural phenomena. Part 50 of 10 CFR, Appendix A, Criterion 2 states, Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed. The structural effects due to the proposed change as described in Reference 1 will not alter compliance with 10 CFR 50, Appendix A, Criterion 2, nor any Code requirements or Code acceptance criteria.

2.2.2 Technical Evaluation The submittal is intended to demonstrate that the effect of adding borated stainless steel GT-Inserts into the fuel assemblies will not alter the structural adequacy of the spent fuel racks and the SFP.

The submittal states that the GT-Inserts will add a maximum weight of 120 pounds (lbs) to the dry weight of a fuel assembly. The total dry weight of a fuel assembly with GT-Inserts will be 1660 lbs which is approximately 7.8 percent heavier than the estimated standard fuel assembly weight of 1540 lbs that was licenced for the SONGS spent fuel racks. The submittal also indicates that an analysis had been performed to demonstrate that the spent fuel racks can safely handle a weight of 2904 lbs for the possible future storage of consolidated fuel. The submittal concludes that the structural integrity of the fuel rack and the SFP are unaffected by the added weight of the GT-Inserts.

In a request for additional information (RAI), the staff requested the licensee to identify that the amendment is not seeking approval for use of 2904 lb fuel assemblies. In its response (Reference 4), the licensee confirmed that the fuel pool racks are currently licensed for a single fuel assembly based on an analysis using a weight of 1540 lbs per assembly, and that NRC approval of this LAR request will enable the use of 1660 lb fuel assemblies with GT-Inserts at SONGS 2 and 3. The staff also requested the licensee to clarify whether any detailed structural/seismic analysis was performed related to this amendment. In its response, the licensee answered that it did not perform any new analysis. With respect to the acceptance criterion for the SFP, the licensee was requested to clarify whether the current analysis method for the SFP is consistent with ACI (American Concrete Institute)-318 method, which was used to design the fuel pool structure. In the RAI, the staff further pointed out that in the table of Evaluation Results for the Spent Fuel Walls and Basement in the submittal, the flexural capacity of concrete sections appeared to not be reduced to account for the presence of axial loads. In its response, the licensee stated that the SFP was designed in accordance with ACI-318 and will continue to meet the acceptance criteria of ACI-318 with the GT-Inserts. The licensee provided a comparison of calculation results from the updated final safety analysis report (UFSAR) for the flexural capacity of critical sections, which included the axial load effect, to the original flexural capacity without the axial load effect. The staff noted that significant reductions occurred in the flexural capacity for some sections, as a result of axial loads.

However, the reduced flexural capacity in each case was still greater than the required flexural capacity based on revised fuel pool loading.

The licensees response to the staffs RAI has clarified the basis of the design load for the spent fuel racks, and the acceptance criteria used for design of the spent fuel racks and the SFP structure. The response also confirmed the structural adequacy of the racks and the SFP structure, as a result of the added weight of GT-Inserts.

2.2.3 Summary Based on its review of the original submittal and RAI responses, the staff concludes that the licensee has properly analyzed the structural effect of adding GT-inserts into fuel assemblies and staff finds it acceptable.

2.3 Spent Fuel Pool Criticality Analysis 2.3.1

Background

The LAR would modify the TS requirements for spent fuel storage to increase the minimum allowed boron concentration of the SFP and allow credit for soluble boron, GT-Inserts made from borated stainless steel, and burnup and fuel storage patterns in place of Boraflex, as follows:

(1) increase the minimum boron concentration in TS 3.7.17, Fuel Storage Pool Boron Concentration, from 1850 to 2000 ppm; (2) replace Figures 3.7.18-1 and 3.7.18-2 in TS 3.7.18, Spent Fuel Assembly Storage, with new Figures 3.7.18-1, 3.7.18-2, 3.7.18-3, and 3.7.18-4 to show minimum assembly burnup versus cooling time and enrichment for unrestricted and peripheral storage in the SONGS 2 and 3 SFP; (3) delete items b through l in TS 4.3.1.1, Criticality, and replace with items b through m to incorporate new spent fuel storage requirements; and (4) change LCS 4.0.100 to reflect allowable fuel storage patterns. This LCS provides storage patterns when the conditions of TS Figures 3.7.18-1 through 3.7.18-4 cannot be met.

The SONGS fuel storage facility is designed to store either new nuclear fuel assemblies, or burned fuel assemblies in a vertical configuration under water. The storage pool is sized to store 1542 fuel assemblies. Two types of spent fuel storage racks are used (Region I and Region II). The two Region I racks each contain 156 storage locations each spaced 10.40 inches on center in a 12x13 array. Each storage location consists of Type 304LN stainless steel square cell 8.64 inches in inside dimension, with 0.110-inch thick walls. The spent fuel assembly is located within the stainless steel cell. The cells in Region I are separated from each other by a minimum water gap of about 1.1 inches. The six Region II storage racks are similar to Region I except that for four of Region II racks each contains 210 storage locations in a 14x15 array. Each of the remaining two Region II racks contains 195 locations in a 13x15 array. All Region II locations are spaced 8.85 inches on center. The cells in Region II do not have a water gap. The Region II storage racks consist of stainless steel cells placed in a checkerboard pattern. Cells are located in every other location and welded together at the cell corners. This results in non-cell storage locations, each one formed by one side-wall of four checkerboard cells. Currently, the fuel racks use Boraflex, a neutron absorber, in the rack design for reactivity control. Because of the concerns with Boraflex dissolution in the spent fuel racks, the SONGS SFP storage rack criticality design basis is being changed to reflect a zero Boraflex credit. The removal of the Boraflex rack poison credit needs to be counter-balanced with the use of fuel storage pattern credit, soluble boron credit, borated stainless steel GT-Inserts credit, inserted control element assemblies (CEAs) and burnup credits in order to comply with the regulatory reactivity acceptance criteria for loading the storage racks.

The proposed TS changes are to provide fuel storage design (without Boraflex rack poison credit), and fuel storage pattern requirements (based on initial enrichment, burnup, and cooling time), to maintain Keff, less than 1.0 when flooded with unborated water, and maintain Keff less than or equal to 0.95 with soluble boron.

In Reference 5, the licensee resubmitted the LAR in its entirety to reflect its responses to the NRC staffs RAIs. Reference 5 included the following major changes to the original submittal (Reference 1):

TS SR 3.7.18.1, TS 4.3.2.k and TS 4.3.1.l - add Rev 2, dated 09/27/2007 to LCS 4.0.100. The changes reflect the revision of the LCS 4.0.100 and the date of NRC approval of the TS.

Enclosure - Add Enclosure 4 to include the responses to NRC staff questions regarding proposed LAR.

In Reference 6, the licensee provided additional clarification to the RAI responses in Reference 5.

Subsequently, the licensee updated in Reference 7 the following requirements to the applicable TSs:

Prior to using the storage criteria of LCO 3.7.18 and LCS 4.0.100, the following uncertainties will be applied:

(1)

The calculated discharge burnup of San Onofre Units 2 and 3 assemblies will be reduced by 6.6 percent.

(2)

The calculated burnup of San Onofre Unit 1 fuel assemblies will be reduced by 10.0 percent.

The requirements were added to TS 4.3.1.g. The originally proposed items g to I were renamed as items h to m. There were no other changes to the 4.3.1.

The same requirements were also added as a note to LCS 4.0.100.

2.3.2 Regulatory Evaluation GDC 62, Prevention of criticality in fuel storage and handling, in Reference 8, specifies that the licensee must limit the potential for criticality in the fuel handling and storage system by physical systems or process.

10 CFR 50.68, Criticality Accident Requirements, specifies the NRC regulatory requirements and acceptance criteria for maintaining subcritical conditions in SFPs. As part of its proposed changes to the design and operation of the SFP, the licensee is also changing the licensing basis of the SFP from an exemption to 10 CFR 70.24, Criticality Accident Requirements, to compliance with 10 CFR 50.68.

The acceptance criteria specified in 10 CFR 50.68(b)(4) for criticality prevention in the SFP are the following:

The Keff shall not exceed 0.95, if the SFP is fully flooded with borated water at a 95/95 level; and The Keff shall be less than 1.0, if the SFP is fully flooded with unborated water at a 95/95 level.

The NRC staff reviewed the LAR to ensure compliance with GDC 62 and 10 CFR 50.68.

2.3.3 Technical Evaluation The technical evaluation contains the following sections: Section 2.3.3.1, the criticality analyses used in determination of the allowable fuel storage patterns; Section 2.3.3.2, the proposed TS changes that include the allowable fuel storage pattens and required boron in the SFP; and Section 2.3.3.3, the proposed licensing basis changes for the SONGS SFP from a 10 CFR 70.24 exemption to a 10 CFR 50.68 compliance.

2.3.3.1 Criticality Analyses The evaluation of the criticality analysis is based on the information in the criticality analysis report (CAR) included in Attachment L to Reference 5. In determining the acceptability of the LAR, the NRC staff reviewed three aspects of the licensees analyses: (1) the computer codes employed for the analyses, (2) the methodology used to calculate the maximum Keff, and (3) the criticality analyses to demonstrate compliance with the regulatory reactivity limits. For each part of the review, the NRC staff evaluated whether the licensees analyses and methodologies demonstrated that adequate safety margins developed were in accordance with NRC regulations and could be maintained in the SONGS SFP.

2.3.3.1.1 Computer Codes The licensee performed the SONGS criticality analysis of the SFP racks with following codes:

CELLDAN, NITAWL-II, KENO-V.a, CASMO-3, and SIMULATE-3.

CELLDAN calculates the atoms/barn-cm of U235, U238, and oxygen in the UO2 fuel. It also calculates the atoms/barn-cm of hydrogen, oxygen, B10, and B11 in the water. In addition, CELLDAN calculates the Dancoff factor, and U235 and oxygen scattering cross-sections per U238 atom for NITAWL-II that generates a 27-group cross-section library for KENO-V.a.

KENO-V.a is a three-dimensional Monte Carlo criticality code used to calculate the Keff. The code is the nuclear industry standard code for criticality analyses.

CASMO-3, a multi-group, two-dimensional transport theory program, was used to (1) determine the reactivity variations due to the rack manufacturing tolerances and normal pool temperature variation, (2) generate the initial enrichment versus discharge burnup criteria, (3) analyze the pool heatup accident, and (4) determine the required soluble boron concentration for the fuel mishandling accident. CASMO-3 was benchmarked against measured experiments, measured fuel isotopics, and measured pinwise LA-140 distributions. The depletion calculation was validated by comparison with the Yankee Core-1 and Zion-measured uranium and plutonium isotopics. These comparisons were performed for a range of pin-cell spectra and indicated good agreement for the fuel isotopics versus burnup (Reference 13 and Reference 6, RAI 7).

SIMULATE-3 is a three-dimensional code used by the licensee to calculate the SONGS reactor cores power distribution, rod worths, etc. In the CAR, SIMULATE-3 was used to calculate the axial burnup bias.

The NRC staff reviewed the licensees application of the codes and determined that each code could be used reasonably to calculate the appropriate parameters necessary to support the maximum Keff analyses, since the NRC staff found that in Reference 11, CELLDAN, NITAWL-II, KENO-V.a, and CASMO-3 were NRC-approved codes for use in the analysis of fuel assemblies stored in the SFP, and that in Reference 13, CASMO-3 and SIMULATE-3 were NRC-approved codes for use in SONGS reactor physical analyses.

2.3.3.1.2 Methodology The licensee performed its SFP criticality analyses with inclusion of the method biases using 95/95 analysis techniques. The major components in the analyses were the nominal reference reactivity (Knominal) calculated using the KENO-V.a code based on fresh assemblies and nominal rack dimensions, KENO-V.a code and spent fuel pool temperature biases, CEA bias, axial burnup uncertainty, a statistical sum of 95/95 uncertainties and worst-case k manufacturing tolerances, and discharge burnup uncertainty. The effective neutron multiplication factor at 95/95 probability/confidence level, Keff, was calculated using the following equations:

Keff = Knominal + Bmethod + Btemp + BCEA + Baxial + Buncert + Bburnup (1) and Buncert = (C2 k 2 + bias, 95/95 2 + ktol 2)1/2 (2),

Where Keff

=

effective neutron multiplication factor at 95/95 probability/confidence level; Knominal

=

nominal reference reactivity - the KENO V.a calculated Keff result; Bmethod

=

the KENO-V.a method bias determined from benchmark critical experiments; Btemp

=

temperature bias (68 oF to 160 oF);

BCEA

=

CEA bias for rodded cases only; Baxial

=

axial burnup bias; Buncert

=

statistical summation of uncertainty components; Bburnup

=

discharge burnup uncertainty based on the statistical summation of a reactivity equivalencing uncertainty, an assembly power measurement uncertainty and a plant power uncertainty; C

=

confidence multiplier (1.763) based on 500 neutron generations; k

=

standard deviation of Knominal; bias, 95/95

=

method bias 95/95 uncertainty; and ktol

=

statistical combination of statistically independent k values due to manufacturing tolerances, e.g., fuel enrichment, cell pitch, etc, and eccentric placement of fuel assemblies.

The following NRC staffs review of the calculation of the value for each item in Equations 1 and 2 is based on the CAR in Attachment L of Reference 5 and the RAI responses in References 5, 6, and 7.

2.3.3.1.2.1 Nominal Reference Reactivity Section 3.2.1 of the CAR described the calculation of Knominal. KENO-V.a was used to calculate a Knominal using input parameters that were consistent with the NRC-approved criticality analysis (Reference 11) for the SONGS SFP. The input included: (1) the nominal spent fuel storage rack and fuel assembly dimensions; (2) the UO2 stack density of 96 percent of the theoretical value; (3) the temperature of all materials at 68 oF; and (4) the axial length of active fuel of 150 inches with a 30-centimeter water reflector above and below the active fuel region. Only the storage cell box wall and Boraflex wrapper were modeled. Fuel assembly grids and end fittings were conservatively not modeled. Also, no Boraflex rack poison was credited. In the KENO-V.a models, at least 503 neutron generations were run with at least 2000 neutrons per generation. The confidence multiplier, C, in Equation 2 is the single-sided tolerance limit factor.

The value of C (1.763) was obtained based on the KEN-V.a cases for 500 neutron generations (Appendix D to Reference 5). The standard deviation of Knominal is designated as k in Equation 2. Since the values of Knominal were calculated by using the previous NRC-approved KENO-V.a code and input parameters consistent with the previous NRC-approved analysis for the assemblies in the SONGS SFP, the calculated Knominal values are acceptable.

2.3.3.1.2.2 Methodology Bias Section 3.1.2 of the CAR described the calculation of the methodology bias which is related to the computer KENO-V.a code used for the criticality calculation. The licensee benchmarked the KENO-V.a code against critical experiments performed by Babcock and Wilcox (B&W)

(Reference 9), and determined that the KENO-V.a code calculation bias (Bmethod) is 0.00814 with a 95/95 bias uncertainty (bias, 95/95) of + 0.00172 using 27-group cross-section library. The boron concentration in the experiments ranges from 0 ppm to 1037 ppm which bounds the total soluble boron requirement of 970 ppm to maintain Keff less than or equal to 0.95. The B&W critical experiment data was previously used by the licensee for SFP criticality analyses that were approved by the NRC (Reference 11). Since KENO-V.a code bias and bias-related uncertainty were calculated based on the NRC-approved KENO-V.a code and B&W critical experimental data that are consistent with that used in the NRC-approved SFP analysis for the SONGS, the NRC staff determined that the calculated code bias and uncertainty are acceptable.

2.3.3.1.2.3 Fuel Pool Temperature Bias Sections 3.2.2 of the CAR described the calculation of the fuel pool temperature bias (Btemp) using CASMO-3. For both storage regions, the fuel pool temperature range considered is from 68 oF to 160 oF, which covers the expected lowest and highest non-accident fuel pool temperature ranges. The soluble boron range considered is from 0 ppm to 1,000 ppm, which bounds the total soluble boron requirement of 970 ppm to maintain Keff less than or equal to 0.95. For Region I, the enrichment range considered is from 1.85 to 5.1 weight percent (w/o),

which bounds the effective Region I fresh enrichment of 2.47 w/o. For Region II, the enrichment range is from 1.20 to 1.85 w/o, which bounds the final Region II fresh enrichment of 1.23 w/o. The results of the calculation were shown in Table 4-1 of the CAR. The licensee used the bounding values of 0.00914 k and 0.003 k for the Region I and Region II assemblies, respectively, in the criticality analysis. The pool temperature has no random variation; the pool temperature bias was added algebraically to the reference nominal reactivity.

The NRC staff found that the most adverse values for the whole ranges of the enrichment, fuel pool temperature and boron concentration were selected to account for the temperature bias effect. Therefore, the NRC staff determined the calculated temperature biases are acceptable.

2.3.3.1.2.4 GT-Insert Bias The licensee used CEAs, GT-Inserts, and erbium oxide (erbia) in fuel rods to control reactivity for assemblies in the SFP. The following subsections contain the NRC staffs evaluation of the licensees use of CEAs, GT-Inserts, and erbium rods in criticality analyses.

2.3.3.1.2.4.1 CEA Bias As indicated in Section 2.2.1 of the CAR, the licensee credited the presence of five-finger CEAs placed in SONGS fuel assemblies, except for SONGS-1 fuel assemblies, stored in Regions I and II. The presence of these CEAs provides additional negative reactivity. A CEA lifetime analysis and a visual inspection would be used to ensure that CEAs used are adequate for use in the SFP. In response to the NRC staffs request, the licensee provided information (RAIs 6 and 16 of Reference 5) describing its CEA lifetime analysis and visual inspection procedures.

The CEA lifetime analysis applied to the CEAs for service in the reactor core would be used to qualify the CEAs used for reactivity control service in the SPF. The measured reactivity of any CEA used in the SFP would be required to match the predicted reactivity effect in one of the previous two cycles of operations during low-power physics testing. The neutron flux and gamma flux level in the SFP would be well below 106 neutrons/centimeter2/second (n/cm2-sec),

which is orders magnitude below the flux level of 1014 n/cm2-sec in the reactor. In addition, the core service life limits would be based on allowable plastic strain with consideration of fast neutron fluence exposure history and susceptibility to irradiation assisted stress-corrosion cracking described in the SONGS UFSAR Section 4.2.3.4. In the visual inspection program, the licensee requires that any CEA with over 20 percent through-wall wear be assessed in detail prior to qualification for continued use. This wear threshold criterion for further evaluation for core service will also apply to the SFP reactivity control service. The licensee performed quantitative inspections on the SONGS 2 CEAs that had served in Cycles 1 through 7. The results showed that the highest observed wear measurement for a CEA that is still suitable for service was 5 percent through-wall. Based on its review of the RAI response, the NRC staff determined that the licensee provided reasonable assurance that its CEA lifetime analysis and visual inspection program are appropriate to preclude phenomena such as absorber depletion, irradiation assisted stress-corrosion cracking, and cladding wear from affecting the SFP criticality analyses.

The likelihood of an inadvertent withdrawal of a CEA is very low because specialized tooling is required for withdrawing a CEA from a fuel assembly. Existing SONGS procedures require that operators validate tool weight only on the spent fuel handling machines load cell readout after ungrappling from a fuel assembly and raising the hoist slightly. The procedures require the operator to report this information to the engineer directing fuel movement. Besides, the use of CEA for reactivity control in criticality analyses was previously approved for a Combustion Engineering (CE) plant (Reference 14). Therefore, the NRC staff determined that the use of CEAs in the SONGS criticality analyses is acceptable.

Sections 3.2.6 and 4.3 of the CAR described the calculation of the CEA bias (BCEA). The bias was evaluated at 68 0F and 0 ppm. These conditions maximize CEA worth and thus the bias.

The bias for full length, five-finger CEAs was determined by comparison of CASMO-3 and KENO-v.a for rodded and unrodded cases. CASMO-3 has accurately predicated SONGS 2 and 3 CEA bank worth measurements. The bias between CASMO-3 and the measured CEA worth data is 0.0 k. The result of the comparison between two computer codes (RAI 25(a)(v),

Reference 6) showed that KENO-V.a is conservative compared to CASMO-3 (i.e., KENO-V.a underpredicted CEA worth by 0.007 k, thus requiring a higher discharge burnup for storage).

In the CAR, a conservative CEA insertion bias of 0.007 k, instead of a calculated value of

-0.007 k, was selected. Therefore, the NRC staff determined that the value of the CEA bias is acceptable.

2.3.3.1.2.4.2 Borated Stainless Steel GT-Insert Credit As indicated in Sections 2.2.2 and 3.2.8 of the CAR, the licensee credited the presence of three or five borated stainless steel GT-Inserts placed in SONGS fuel assemblies stored in Region II.

The presence of these GT-Inserts provides additional negative reactivity. The mechanical design configuration of the GT-Inserts is similar to the shape, size, and weight of a CEA finger.

Each of the GT-Inserts is manufactured with an outside diameter of approximate 0.75 inches and a boron content of approximately 2 w/o. It is designed to cover the entire active fuel length of 150.0 inches. The thermal considerations of the fuel are unaffected by the presence of the GT-Inserts because the guide tube is designed for the presence of a CEA, and it is not a primary coolant flow area. In response to the staffs RAI 19 (Reference 5), the licensee indicated that the GT-Inserts are made from borated stainless steel, Type 304 B7, grade A, and are manufactured in accordance with the requirements of standard American Society for Testing and Material (ASTM) A 887 and ASTM A 484. The SONGS procurement process will include vender source inspection and qualification, and material certification and test results in compliance with the ASTM specifications. When three GT-Inserts are used, they are installed in an assemblys center guide tube, the guide tube associated with the serial number, and the diagonally opposite guide tube (Figure 3-3 of the CAR).

Although intergranular corrosion resistance of borated stainless steel exposed to acid conditions decreases with increased boron content, long-term tests with borated stainless steel indicated that in the SFP environment no measurable corrosion effects take place. In order to provide assurance that at all times there is enough poison material for reactivity control, the licensee committed (RAI 19, Reference 5) to institute a surveillance program where, at 5-year intervals, one percent of the GT-Inserts will be visually inspected for any material degradation.

Also, inadvertently withdrawing a GT-Insert and misloading of the GT-Inserts are unlikely to occur because the GT-Insert (RAI 16, Reference 5) will be contained in guide tubes of the designed assemblies and a specialized tool is required to remove it. In addition, the design of the installation equipment, procedure controls, and the double verification requirement that will be in place to ensure that the GT-Inserts are installed properly were considered in the NRC staffs review of the GT-Inserts.

The NRC staff determined that the materials used in GT-Inserts and the surveillance program for inspection of material degradation are consistent with those previously approved by the NRC for a CE plant (Reference 15). Therefore, the NRC staff concluded that the use of GT-Inserts in the SONGS criticality analysis is acceptable.

2.3.3.1.2.5 Erbium Rods Credit As indicated in Section 3.2.7 of the CAR and the RAI 20 response (Reference 5), the licensee credited erbium rods in fresh fuel assemblies for reactivity control. No reactivity credit was considered for the remaining erbia in the assembly returned from the core. Erbia is applied as an integral part of the fuel pellet. The NRC has issued a generic approval (Reference 10) for core designs containing erbium burnable absorber for CE reactors. In the criticality analysis, fresh fuel assemblies containing 40 and 80 erbium rods were considered and the fuel assembly was modeled at its most reactive point of life (beginning-of-cycle). The NRC-approved CASMO-3 code was used to calculate the erbia equivalent fresh enrichment. The burnable poison load for the erbium rods was assumed at 2.0 w/o erbia, which was reduced from the nominal burnable poison loading of 2.1 w/o by 5 percent to account for manufacturing tolerances. Since CASMO-3 is an NRC-approved code and the assumption of using the initial erbium loading reduced by 5 percent, resulting in a smaller neutron poison effect, is conservative, the NRC staff concluded that the use of erbium rods in the criticality analysis is acceptable.

2.3.3.1.2.6 Axial Burnup Bias As indicated in Section 3.2.4 of the CAR, the relationships of discharge burnup versus initial enrichment were generated with two-dimensional (2D) CASMO-3 with uniform axial burnup and isotopic distribution. The assumption of uniform axial distribution may result in nonconservative reactivity results. To offset the potential nonconservatism, the licensee performed all CASMO-3 assembly depletion calculations at moderator and fuel temperatures higher than the core average conditions. The moderator temperature was assumed to be 600 oF, which is significantly higher than the core average moderator temperature of about 580 oF. The fuel temperature was assumed to be 1200 oF, instead of the fuel temperature decreased with burnup. The boron concentration of 1000 ppm was also chosen to bound the cycle average hot-full power boron concentration. Performing calculations at higher temperatures and slightly higher boron concentration enhances the build-up of PU239. The higher PU239 concentration results in a higher, and thus more conservative, equivalent fresh fuel enrichment.

The axial burnup bias (Baxial) was defined as the k difference between three-dimensional (3D) axially dependent burnup distribution with a given average burnup and 2D uniform burnup distribution at the same average burnup. In response to the NRC staffs request, the licensee discussed the calculation of axial burnup bias using SIMULATE-3 in its RAI 11 response (Reference 5). For each enrichment, the 3D depletion was performed at the nominal reactor coolant system (RCS) inlet temperature of 553 oF and the fuel temperatures corresponding to various burnup points of interest, while the 2D depletion was performed at a constant RCS temperature of 600 oF and a fuel temperature of 1200 oF. The SFP temperature was assumed at 68 oF. The results of the analysis for cases with burnup values of 0, 10, 20, 30, 40, 50, and 60 GWD/T (Gigawatt-Day per Ton) and initial enrichment levels of 1.87 w/o and 4.45 w/o showed that the effective reactivities based on the 2D depletion with higher moderator and fuel temperatures were higher than that based on the 3D depletion with nominal moderator and fuel temperatures. Therefore, the licensee conservatively set the axial burnup bias to be 0.0 k. In response to RAI 25.b regarding a concern of use of the axial burnup bias to bound the applicable operating range, the licensee performed additional calculations. The calculations (Reference 7) were performed for enrichment values of 1.87 w/o, 4.45 w/o, and 5.0 w/o. For each enrichment, the 3D depletion was performed at RCS inlet temperatures ranging from 533 0F to 560 oF. This temperature range bounds the SONGS TS limit of 535 oF to 558 oF specified in TS 3.4.1(b).2. Also, 3D depletion was performed at 533 oF and at 30 percent power to bound the part power operating condition. The calculations showed that the 2D depletion with the high RCS inlet temperature of 600 oF resulted in the highest Keff values for all enrichment cases. In addition, the above calculations were performed at the SFP temperature of 160 oF to verify that the conclusion is valid for the entire SFP range of 68 oF to 160 oF. The calculation showed that the 2D depletion with the high RCS inlet temperature of 600 oF resulted in the highest Keff values at the SFP temperature of 160 oF.

In an NRC-approved submittal (Reference 11), an independent method was used to show that the SONGS fuel axial burnup bias is 0.0 k. The burnup distribution from a discharged SONGS assembly was converted to equivalent fresh enrichments. The equivalent fresh enrichments were input to a 3D KENO model. A second 3D KENO model with uniform enrichment corresponding to the assembly average burnup was set up. The comparison of the calculational results showed that the axial burnup bias is 0.0 k.

An Oak Ridge Report (NUREG/CR-6801) presented studies of the axial burnup effect using axial burnup profiles provided by pressurized-water reactor (PWR) plants of various designs.

The results of the analysis in Section 4.2.2 of the NUREG report showed that CE fuel types, which were used in SONGS, exhibit a smaller axial burnup effect: for most cases, k values for CE fuel are close to zero; and the maximum axial burn effect is less than 0.01 k as compared to a maximum axial burnup effect of up to 0.04 k for B&W fuel. This is consistent with Calvert Cliffs (a CE plant) results using plant-specific axial burnup profiles (Section 9.E.2.k of the Calvert Cliffs submittal dated September 30, 2003, ADAMS ML033140579). The Calvert Cliffs results showed that the worst-case axial burnup bias is a negative value.

The NRC staff noted that the results of analyses in NUREG/CR-6801 indicated that the number of nodes representing axial length of fuel assemblies modeled in the computer codes would significantly affect the results of the axial burnup credit. In light of the NUREG/CR-6801 results, the NRC staff requested the licensee to show that the number of axial-zones used in SIMULATE-3 is adequate and acceptable for calculations of the axial burnup. In response (RAI 24, Reference 5), the licensee stated that its SIMULATE-3 model contains 20 axial zones in the active fuel zone. As indicated in the NRC-approved report (Reference 13), SIMULATE-3 predicted axial power distribution agrees well with measurements. The NRC staff determined that the number of nodes (20) used in SIMULATE-3 is acceptable since it is consistent with the NUREG/CR-6801 results that showed a model with 18 or more axial zones is adequate to predict a reliable burnup bias.

Based on its review discussed above, the NRC staff found that: (1) the licensees calculations showed that the 2D depletion with the high RCS inlet temperature of 600 oF resulted in the highest Keff values for the applicable operating conditions; and (2) the results of analyses performed by the nuclear industry demonstrated that the axial burnup biases are very close to zero for CE fuel. Therefore, the NRC staff determined that the 2D depletion model with the high RCS inlet temperature of 600 oF and the associated axial burnup bias of 0.0 k are acceptable.

2.3.3.1.2.7 Reactivity Equivalencing Uncertainty and Fuel Assembly Burnup Uncertainty In considering the effect of the uncertainty of fuel depletion calculations, the licensee applied to CASMO-3 calculational results a reactivity uncertainty that was zero at zero burnup and increased linearly with burnup, passing through 0.01 k at 30 GWD/T. The same reactivity equivalencing uncertainties were previously approved by the NRC (Reference 16) for other PWR plants. In addition, the NRC previously approved the licensees use of a 5 percent uncertainty applied to the total reactivity decrement calculated by CASMO-3. The licensee showed (RAI 14, Reference 5) that the 5 percent uncertainty on the reactivity decrement is comparable to the 0.01 k value at 30 GWD/T and 0.02 k value at 60 GWD/T. These values were converted to the burnup equivalencing uncertainty of 3.98 percent (RAI 26(a),

Reference 6).

In response to RAI 26(a), the licensee indicated that its criticality analyses will account for the reactivity equivalencing and discharge burnup uncertainty by reducing the discharge burnup of the fuel assemblies to be placed in the SFP racks. After the discharge burnup was decreased by these uncertainties, the Region I and Region II Tables and Figures of proposed TS 4.3.1 and LCS 4.0.100 were used to determine the storage patterns in meeting the criterion of being less than 1.0 for unborated water conditions. For SONGS 2 and 3, a 6.6 percent reduction (Bburnup) will be applied to the calculation using the NRC-approved CECOR code (CENPD-153-P, Rev. 1-P-A) for determination of the fuel burnup for all fuel assemblies prior to determination of the allowable storage per TS 4.3.1 and LCS 4.0.100. The reduction of 6.6 percent was based on (RAI 26(a), Reference 6) the statistical summation of a 3.98 percent burnup equivalencing uncertainty, a 4.76 percent assembly power measurement uncertainty and a 2 percent plant power uncertainty. For conservatism, a reduction of 10 percent will be used to account for the reactivity equivalencing and discharge burnup uncertainty for SONGS 1.

The NRC staff found that the values of 6.6 percent and 10 percent are based on the NRC-approved reactivity equvalencing uncertainty and the power measurement uncertainty using the NRC-approved CECOR code. Therefore, the NRC staff determined that the values are acceptable.

2.3.3.1.2.8 Manufacturing Tolerances and Eccentric Placement Bias Sections 3.2.2 and 4.1 of the CAR described the calculation of the manufacturing tolerance biases applied to the Regions I and II racks that were used in the determination of Keff. The CASMO-3 code was used to determine the reactivity effects of dimensional tolerances for the fuel assemblies in the SFP and storage racks. The calculation was based on the dimensional tolerances for SONGS 16x16 and 14x14 design-basis fuel assemblies. There are tolerances for four components including rack storage cell-wall thickness, rack storage cell inside diameter, rack storage cell pitch and U235 enrichment that will have significant reactivity impact and contribute to the system bias effects on Keff. The licensee performed a sensitivity study (Table 4.1 of the CAR) for the tolerances over the expected range of soluble boron concentration and fuel enrichments. The soluble boron range considered is from 0 ppm to 1000 ppm, which bounds the total soluble boron requirement of 970 ppm to maintain Keff less than or equal to 0.95. For Region I, the enrichment range considered is from 1.85 to 5.1 w/o, which bounds the effective Region I fresh enrichment of 2.47 w/o. For Region II, the enrichment range is from 1.20 to 1.85 w/o, which bounds the effective Region II fresh enrichment of 1.23 w/o. The manufacturing dimensional tolerances were obtained from Westinghouse, the manufacturer of the racks, and the enrichment tolerance limit of 0.05 w/o was consistent with the licensees fuel manufacturing enrichment specifications (RAI 25(iv),

Reference 6). Since the licensee used the highest calculated tolerance biases for each boron concentration level and the fuel enrichment level, the NRC staff determined that the calculated tolerance biases are conservative and, thus, acceptable.

Sections 3.2.3 and 4.2 of the CAR described the calculation of eccentric placement bias. The bias due to eccentric placement of fuel assemblies in the cell was calculated using with KENO-V.a for SONGS SFP racks at 68 oF and 0 ppm. This eccentric placement bias is insensitive to the variation of enrichment and load pattern because the Keff of a system is a flux-and volume-weighted integral quantity and is not affected by small local variations (RAI 25(iii),

Reference 6). In the KENO-V.a models, groups of four assemblies were moved as close together as possible in the corner where four storage locations met. An SFP criticality analysis for a typical PWR showed that the eccentric placement bias k decreases as fewer than four assemblies are moved together. The NRC staff determined that the calculated bias is conservative since it was based on the KENO-V.a models that represent the worst fuel pattern of four assemblies moved together and, therefore, it is acceptable. This bias was statistically combined with the manufacturing tolerances.

Based on the review discussed in Section 2.3.3.1.2 of this evaluation, the NRC staff concluded that the methods and calculated numeric values of all parameters in Equations 1 and 2 are acceptable for use in the criticality analysis to show that Keff is less than or equal to 0.95 for fresh and spent fuel in SPF Regions I and II with borated water, and Keff is less than 1.0 for fuel in the SFPs with unborated water.

2.3.3.1.3 Analysis of Storage Racks in Regions I and II The SONGS SFP Region I contains CE, SONGS 2 and 3, Zircaloy-clad, 16x16 fuel assemblies with a maximum design enrichment of 4.8 w/o, while Region II contains both CE 16x16 fuel assemblies and Westinghouse, SONGS 1, stainless-steel-clad, 14x14 fuel assemblies with a maximum design enrichment of 4.0 w/o.

The SONGS SFP storage rack criticality analyses were performed to reflect a zero Boraflex credit. Removal of Boraflex poison credit needs to be counter-balanced with the use of fuel burnup credit (BUC), CEA credit, GT-Insert credit, erbium rod credit, and SFP soluble boron credit to provide safe storage of the fuel assemblies in the SFP and comply with the regulatory reactivity limits. There are many combinations of BUC, CEA credit, GT-Insert credit, erbium rod credit, and soluble boron credits possible to offset the reactivity increase due to the removal of Boraflex credit. The soluble boron credit should be limited to avoid deboration time requirements and to maintain the Keff below 1.0 for an unborated SFP, whereas the BUC loading restrictions should not be excessively demanding (i.e., requiring very high assembly burnup) in order to remain useful and applicable to the expected fuel assembly discharge burnups. The licensee performed the criticality analyses to determine the BUC loading restrictions and the pool boron concentration requirement to meet the reactivity limits for both normal loading and accident or upset conditions.

2.3.3.1.3.1 Burnup Loading Curve and Usage Requirements Section 3.2 of the CAR described the criticality analysis for fuel assemblies in Regions I and II.

The criticality analysis developed BUC loading curves for an allowable storage pattern that allows the discharged assemblies to be stored in the SFP racks with acceptable burnup-enrichment combinations for cooling times of 0, 5, 10, 15, and 20 years. To bound the existing fuel being stored in the SONGS SFP, the criticality analysis was performed first without boron credit to determine the storage patterns of the spent fuel racks such that the final KENO-V.a Keff, including all uncertainties, is less than 1.0. The determined storage pattern is then evaluated to set the pool soluble boron concentration requirements such that the regulatory criterion of Keff less than or equal to 0.95 is met for both storage and fuel handling, as well as accident/upset conditions. Keff includes Knominal and k calculated by KENO-V.a and all other allowances for bias and uncertainties for codes, methods, and manufacturing tolerances contained in Equations (1) and (2) of Section 3.1.2 of this report discussed above.

2.3.3.1.3.2 Fuel Storage Patterns To compensate for no Boraflex poison credit, the licensee considered in its criticality analyses the following storage patterns and GT-Inserts: (1) unrestricted storage, (2) SFP peripheral storage, (3) 2x2 storage patterns, (4) 3x3 storage patterns, (5) credit for inserted CEAs, (6) credit for Erbium rods, (7) credit for PU241 decay, (8) credit for GT-Inserts, and (9) credit of burnup effects.

The enrichment-burnup curve calculations for each storage pattern were performed for a discrete number of initial assembly enrichment up to a maximum of 5.0 w/o using the methods discussed in Section 2.3.3.1.2 of this report. The calculation determined the allowable minimum assembly average burnup for each of the initial assembly enrichment, as well as a determination of the allowable initial assembly enrichment with the zero burnup. For each fuel initial enrichment, a sensitivity study was performed to determine assembly burnup that results in Keff slightly below 1.0 with unborated water in the SFP, and not exceeding 0.95 with borated water in the SFP.

For SONGS 2 and 3, the results of the criticality analyses were provided in Sections 4.5, 4.6, Tables 4-3 through 4-25, Figures 4-1 through 4-21, and Figure 4-32 of the CAR. The results are: (1) Tables 4-3 through 4-10 and Figures 4-1 through 4-6 show the permissible Region I storage patterns with the associated allowable burnup-enrichment limits; (2) Tables 4-11 through 4-25 and Figures 4-7 through 4-21 show the permissible Region II storage patterns; and (3) Figure 4-32 summarizes the assembly boundary interface requirements to prevent an undesirable increase in reactivity.

For SONGS 1, the licensee did not analyze the fuel in Region I and, therefore, the SONGS 1 fuel was not allowed to be stored in Region I. For SONGS 1 fuel assemblies in Region II, the results of criticality analyses were provided in Section 4.6 of the CAR for the allowable storage requirements.

Since the fuel patterns and the interface requirements were determined based on the acceptable method discussed in Section 2.3.3.1.2 of this evaluation report, and the results showed that for unborated conditions, Keff is less than 1.0 including 95/95 uncertainty, satisfying the 10 CFR 50.68(b)(4) requirements, the NRC staff determined that they are acceptable.

2.3.3.1.3.3 Soluble Boron Credit Sections 3.2.10 and 5 of the CAR described the calculation of the soluble boron credit used to provide safety margin by maintaining Keff less than or equal to 0.95 including 95/95 uncertainty.

Based on the spent fuel storage configurations discussed in Sections 3.2 and 4 of the CAR for unborated water conditions, the licensee calculated the soluble boron concentration requirement to bring the Keff, including all uncertainties, to be less than or equal to 0.95 for the allowable storage patterns. A series of full-pool KENO-V.a cases with various boron concentrations were performed (RAI 13, Reference 5) to determine the Keff for each case. In each case, each cell was loaded with fresh fuel at the maximum allowed enrichment for the location. The required boron concentration was calculated by interpolating for Keff of 0.95. The results showed that the required soluble boron concentration for the SFP rack normal condition was 370 ppm. Since the NRC-approved KENO-V.a code and allowable storage patterns were used to determine the required boron concentration, the NRC staff concluded that the boron concentration of 370 ppm is acceptable.

As discussed in Section 2.3.3.1.2.7 of this report, the licensee applied to CASMO-3 fuel depletion calculational results a reactivity equivalencing uncertainty that was zero at zero burnup, and increased linearly with burnup, passing through 0.01 k at 30 GWD/T (i.e., 0.02 k at 60 GWD/T). In calculating the required boron concentration to account for the reactivity decrement, the licensee used (RAI 14, Reference 5) the highest reactivity value of 0.02 k that was based on the limiting peak pin burnup of 60 GWD/T in any SONGS assembly. The licensee used the calculated highest boron worth based on a full-pool KENO-V.a model at the expected range of the boron concentration in the SFP. Since the required boron concentration of 178 ppm is based on the highest reactivity equivalencing uncertainty and boron worth, the NRC staff determined that the boron concentration is conservative and acceptable.

The licensee calculated the required boron concentration of 218 ppm to compensate for the fuel assembly discharge burnup uncertainty of 7 percent. This value was conservative as compared to the 95/95 uncertainty of 4.76 percent (RAI 15, Reference 5) for the measurement of the fuel assembly power using an NRC-approved code (CENPD-153-P, Rev. 1-P-A, INCA/CECOR Power Peaking Uncertainty) with a conservative multiplier to account for the plant power uncertainty. In converting the discharge burnup uncertainty into the boron concentration, the licensee used the calculated highest boron worth at the expected range of the boron concentration in the SFP. Therefore, the NRC staff determined that the calculated boron requirements of 218 ppm is acceptable.

The licensee included in the criticality analyses the soluble boron measurement uncertainty of 50 ppm, which was previously used by the licensee for the SONGS SFP criticality analyses. In addition, the licensee included an allowance of 154 ppm reserved for future requirements.

2.3.3.1.3.4 Accident Conditions Evaluation The licensee also evaluated the reactivity accidents with consideration of compliance with the double contingency principle that at least two unlikely, independent accidents have to occur concurrently for a criticality event to be possible. The licensee evaluated the following events to determine the bounding reactivity accident that requires the most boron credit to maintain Keff less than or equal to 0.95:

(1) loss of cooling resulted in SFP water temperature rise to 248 oF, (2) an assembly dropped horizontally on top of the racks, (3) a fuel assembly dropped vertically into a storage location already containing a fuel assembly, (4) a fuel assembly dropped to the SFP floor, and (5) a misloading of a single 4.8 w/o fresh fuel assembly in either Region I or Region II.

Sections 3.2.9 and 5.6 of the CAR discussed calculations to determine the additional soluble boron credit needed to counter-balance the increased reactivity to maintain and satisfy the regulatory acceptance criteria of Keff less than or equal to 0.95 for each of these accident conditions. Among these accident conditions evaluated, the licensee identified that the misloading of a single 4.8 w/o fresh fuel assembly in Region II, 1.56 w/o x 0.94 w/o checkerboard pattern, is the limiting case, requiring the most boron credit of approximately 730 ppm. Since the boron concentration of 730 ppm was based on the NRC-approved CASMO-3 code and results of analysis for the limiting accident, the NRC staff determined that it is acceptable.

2.3.3.1.3.5 Deboration Accident Evaluation An analysis in Attachment K of the Reference 5 showed that a deboration accident that would result in the dilution of the SFP boron concentration below 1700 ppm from an initial concentration of 2000 ppm is not a credible event. The deboration accident evaluation is addressed in Section 2.1 of this SE.

Section 5 of the CAR showed that the total soluble boron required to maintain Keff to be less than or equal to 0.95, including all biases and uncertainties, under accident conditions, is 1700 ppm (370 ppm + 178 ppm + 218 ppm + 50 ppm + 154 ppm + 730 ppm), with the exception of boron dilution. With inclusion of the boron required to compensate for the boron dilution event, the SFP boron should be maintained at a concentration of 2000 ppm.

Based on the discussion in Section 2.3.3.1 of this evaluation, the NRC staff found that: (1) the acceptable methods (including computer codes and models) were used in performing criticality analyses; (2) the bounding values of biases and uncertainties were applied to the applicable conditions; and (3) the results showed that the calculated Keff with inclusion of the credible biases and uncertainties is less than 1.0 for unborated water conditions, and is not greater than 0.95 for borated water conditions, meeting the 10 CFR 50.68(b)(4) requirements. Therefore, the NRC staff concluded that the criticality analysis is acceptable.

2.3.3.2 Proposed TS Changes The LAR would (1) increase the minimum boron in TS 3.7.17, Fuel Storage Pool Boron Concentration, from 1850 to 2000 ppm, (2) replace Figures 3.7.18-1 and 3.7.18-2 in TS, Spent Fuel Assembly Storage, with new Figures 3.7.18-1, 3.7.18-2, 3.7.18-3, and 3.7.18-4 to show minimum assemblies burnup versus cooling time and enrichment for unrestricted and peripheral storage in the SONGS SFP, (3) delete items b through I in TS 4.3.1.1, Criticality, and replace with items b through m to incorporate new storage requirements; and (4) change LCS 4.0.100 to include allowable fuel storage patterns. The marked-up TSs are included in Attachments C and D, and the revised TSs are located in Attachments E and F of Reference 5 for SONGS 2 and 3, respectively. The NRC staff reviewed each of the changes against licensing regulations discussed in Section 2.0 of this evaluation and found them acceptable.

The basis for the NRC staffs acceptance and a description of the review it performed is discussed in Section 2.3.3.1 and subsections discussed as follows.

2.3.3.2.1 Revision to TS 3.7.17, Fuel Storage Pool Boron Concentration TS 3.7.17 would be revised to increase the minimum boron concentration from 1850 to 2000 ppm. The frequency of verification is not changed.

Currently, the soluble boron is not credited in determining fuel storage requirements that maintain Keff less than or equal to 0.95. With the anticipated loss of Boraflex due to erosion or dissolution, as has been experienced in the industry, a minimum concentration of soluble boron would be required to maintain Keff less than or equal to 0.95. The increase in the TS-required concentration from 1850 to 2000 ppm ensures that there is no credible boron dilution event that would cause Keff to exceed 0.95.

Current TS 3.7.17 is applicable to whenever any fuel assemblies are stored in the fuel storage pool, and a fuel storage pool verification has not been performed since the last movement of fuel assemblies in the fuel storage pool. If fuel storage pool boron concentration is not within the limit, Action A.2.2 requires the operator to verify immediately by administrative means that Region II fuel storage pool verification has been performed since the last movement of fuel assemblies in fuel storage pool.

In support of the storage patterns being implemented by proposed TS 3.7.18, the licensee proposed to expand the current applicability of TS 3.7.17 to whenever a fuel assembly is stored in the fuel storage pool, and as a result, Action A.2.2 is no longer applicable and can be deleted.

The acceptability of the SFP boron concentration of 2000 ppm was demonstrated by the criticality analysis and deboration analysis discussed in Section 2.3.3.1 of this SE. Therefore, the NRC staff determined that the proposed TS 3.7.17 changes are acceptable.

2.3.3.2.2 Revision to TS 3.7.18, Spent Fuel Assembly Storage TS Figures 3.7.18-1 and 3.7.18-2 would be replaced with new TS Figures 3.7.18-1, 3.7.18-2, 3.7.18-3, and 3.7.18-4. The new TS figures showed minimum burnup versus cooling time and enrichment for unrestricted and peripheral storage in the SONGS SFP. Additional storage patterns for SONGS 1, 2, and 3 fuel assemblies storage in the SFP would be contained in LCS 4.0.100 (Revision 2), which would be used to store fuel assemblies that did not meet the requirements of Figures 3.7.18-1 through 3.7.18-4. Currently, the Boraflex in the SFP racks limit Keff less than or equal to 0.95 with minimal limitations on fuel assembly initial enrichment and burnup storage location criteria. The anticipated future loss of Boraflex requires additional storage requirements that are more stringent than those currently in place.

The acceptability of the requirements specified in the proposed TS 3.7.18 and LCS 4.0.100 (Revision 2) for various fuel assembly storage patterns in the SFP was demonstrated by the criticality analysis and deboration analysis discussed in Section 2.3.3.1 of this SE. Therefore, the NRC staff determined that the proposed TS changes are acceptable.

2.3.3.2.3 Revision to TS 4.3.1, Criticality The LAR (References 5 and 7) proposed to revise TS 4.3.1.1 by deleting items b through I and replacing them with new items b through m as follows:

b.

Keff < 1.0 if flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; c.

Keff < 0.95 if fully flooded with water borated to 1700 ppm, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; d.

Three or five Borated stainless steel guide tube inserts (GT-Insert) may be used.

When three Borated stainless steel guide tube inserts are used, they will be installed in an assemblys center guide tube, the guide tube associated with the serial number, and the diagonally opposite guide tube. Fuel containing GT-Inserts may be placed in either Region I or Region II. However, credit for GT-Inserts is only taken for Region II storage; A five-finger CEA may be installed in an assembly. Fuel containing a five-finger CEA may be placed in either Region I or Region II. Credit for inserted 5-finger CEAs is taken for both Region I and Region II; e.

A nominal 8.85 inch center to center distance between fuel assemblies placed in Region II; f.

A nominal 10.40 inch center to center distance between fuel assemblies placed in Region I; g.

Prior to using the storage criteria of LCO 3.7.18 and LCS 4.0.100, the following uncertainties will be applied:

(1)

The calculated discharge burnup of San Onofre Units 2 and 3 assemblies will be reduced by 6.6%.

(2)

The calculated discharge burnup of San Onofre Unit 1 assemblies will be reduced by 10%.

h.

Units 2 and 3 fuel assemblies with a burnup in the acceptable range of Figure 3.7.18-1 are allowed unrestricted storage in Region I; i.

Units 2 and 3 fuel assemblies with a burnup in the acceptable range of Figure 3.7.18-2 are allowed unrestricted storage in the peripheral pool locations with 1 and 2 faces toward the spent fuel pool walls of Region I; j.

Units 2 and 3 fuel assemblies with a burnup in the acceptable range of Figure 3.7.18-3 are allowed unrestricted storage in Region II; k.

Units 2 and 3 fuel assemblies with a burnup in the acceptable range of Figure 3.7.18-4 are allowed unrestricted storage in the peripheral pool locations with 1 or 2 faces toward the spent fuel pool walls of Region II; l.

Units 2 and 3 fuel assemblies with a burnup in the unacceptable range of Figure 3.7.18-1, Figure 3.7.18-2, Figure 3.7.18-3, and Figure 3.7.18-4 will be stored in compliance with Licensee Controlled Specification 4.0.100 Rev. 2, dated 09/27/07; and m.

Each SONGS 1 uranium dioxide spent fuel assembly store in Region II shall be stored in accordance with Licensee Controlled Specification 4.0.100 Rev. 2, dated 09/27/07.

Items b, c, g, h, i, j, k, l, and m above incorporate new storage requirements to be implemented as a result of the anticipated loss of Boraflex. Items e and f above are identical to the current requirements of TS 4.3.1.1.c and 4.3.1.1.d. Items g through m provide a reference to TS 3.7.18 and LCS 4.0.100, Revision 2 (with a date of 09/27/07 representing the date of the NRC-approval of this LAR) that contains expanded fuel storage requirements including the use of GT-Inserts.

It should be noted that the term Keff in TS 4.3.1.1 is the effective neutron multiplication factor Keff that includes an allowance for all uncertainties evaluated at a 95 percent probability, 95 percent confidence level. The proposed TS 4.3.1.1.b and 4.3.1.1.c for the spent fuel racks in Regions I and II are consistent with the requirements specified in 10 CFR 50.68(b)(4). This revised TS 4.3.1 takes credit for the minimum soluble boron of 1700 ppm to maintain Keff not exceeding 0.95 during conditions without considering the boron dilution event. For conditions with consideration of the boron dilution event, the proposed revision of TS 3.7.17 with the minimum soluble boron of 2000 ppm applies to maintain a Keff less than or equal to 0.95.

TS 4.3.1.1.d specifies the requirements of using GT-Inserts and CEAs to compensate for no Boraflex poison credit. The requirements are consistent with the assumptions used in the criticality analyses discussed in Section 2.3.3.1.2.4 of this SE.

TS 4.3.1.1.e and TS 4.3.1.1.f simply include the storage racks located in Regions I and II and specify a center-to-center spacing (pitch) of 8.85 inches and 10.4 inches between fuel assemblies placed in Regions I and II.

TS 4.3.1.1.g includes the reactivity equivalencing and discharge burnup uncertainty by reducing the discharge burnup of the fuel assemblies to be placed in the SFP racks. The requirements are consistent with the assumptions used in the criticality analyses discussed in Section 2.3.3.1.2.7 of this SE.

TS 4.3.1.1.h and TS 4.3.1.1.j restrict assemblies to be loaded in Regions I and II to be within the acceptable range of the enrichment-burnup restrictions (EB loading curves) shown in Figures 3.7.18-1 and 3.7.18-3, respectively. TS 4.3.1.1.i and TS 4.3.1.1.k specify EB loading curves for the respective Regions I and II assemblies stored in peripheral locations.

TS 4.3.1.1.l and TS 4.3.1.1.m restrict assemblies not meeting the EB loading curves to be stored in compliance with LCS 4.0.100 Revision 2 for SONGS 2 and 3 fuel assemblies, and SONGS 1 fuel assemblies, respectively. These limitations are to assure that TS 4.3.1.1.b and 4.3.1.1.c are complied with as demonstrated by the criticality analyses.

The acceptability of the requirements specified in the proposed TS 4.3.1.1 for various fuel assembly storage patterns in the SFP was demonstrated by the criticality analysis and deboration analysis discussed in Section 2.3.3.1 of this SE report. Therefore, the NRC staff determined that the proposed TS 4.3.1 changes are acceptable.

2.3.3.2.4 Revision to LCS 4.0.100, Fuel Storage Patterns The LAR proposed to revise LCS 4.0.100 to provide the allowable fuel storage patterns, including the use of soluble boron, GT-Inserts, CEAs, erbia in fresh assemblies, and PU241 decay effect to control reactivity for assemblies in the SFP. This LCS provides storage patterns when the conditions of TS Figures 3.7.18-1 through 3.7.18-4 cannot be met. The licensee will place (RAI 2, Reference 5) the date of NRC approval on each page of LCS 4.0.100 when the NRC staff approves the LCS. Future revisions of LCS 4.0.100 (pages 1 through 61, i.e.,

excluding the Bases pages) would be made in conjunction with an LAR associated with TS 4.3.1. The NRC staff found that the licensees approach to revise the LCS was consistent with the requirements of TS 4.3.1.f of NUREG-1432, Standard Technical Specifications for CE Plants. Therefore, the NRC staff determined that the licensees approach is acceptable.

The LAR proposed to revise LCS 4.0.100 as follows:

LCS 4.0.100 Fuel Storage Patterns NOTE 1:

This Licensee Controlled Specification is listed by revision number and date in Technical Specification 4.3.1. All changes to pages 1 through 61, Rev. 2 dated 09/27/07 of this LCS (i.e., excluding the Bases pages) must be approved by the NRC via the amendment application process in conjunction with an associated change to Technical Specification 4.3.1.

NOTE 2:

Prior to using the storage criteria in 4.0.100.1, 4.0.100.2, and 4.0.100.4 below, the following uncertainties shall be applied:

(1)

The calculated discharge burnup of San Onofre Units 2 and 3 fuel assemblies will be reduced by 6.6%.

(2)

The calculated discharge burnup of San Onofre Unit 1 fuel assemblies will be reduced by 10.0%.

VALIDITY STATEMENT:

Rev. 2 effective upon NRC approval 09/27/07, to be implemented within 180 days.

4.0.100 New or burned fuel (which does not meet the criteria of LCO 3.7.18 for unrestricted storage or storage at the pool periphery) may be stored in Region I or Region II in accordance with the allowable storage patterns described in this LCS.

4.0.100.1 Region I =

Region I Storage Patterns are given in Tables I-1 through I-8 and Figures I-1 through I-9.

4.0.100.2 Region II =

Region II Storage Patterns are given in Tables II-1 through II-15 and Figures II-1 through II-22.

4.0.100.3 SONGS Unit 1 Fuel shall not be stored in Region I Racks.

4.0.100.4 The burnup of each SONGS Unit 1 uranium dioxide spent fuel assembly stored in Region II shall meet the following criteria:

4.0.100.4.1 SONGS Unit 1 nominal 3.40 w/o assemblies can be stored in the Region II Racks (unrestricted) if:

the burnup is greater than 25,000 MWD/T [Megawatt-Day per Ton], and the cooling time is greater than 5 years.

4.0.100.4.2 SONGS Unit 1 nominal 4.00 w/o assemblies can be stored in the Region II Racks (unrestricted) if:

the burnup is greater than 26,300 MWD/T, and the cooling time is greater than 20 years.

the burnup is greater than 27,100 MWD/T, and the cooling time is greater than 15 years.

the burnup is greater than 28,200 MWD/T, and the cooling time is greater than 10 years.

4.0.100.4.3 SONGS Unit 1 nominal 4.00 w/o assemblies can be stored in the Region II Racks (SFP perpherery) if:

the burnup is greater than 20,000 MWD/T, and the cooling time is greater than 0 years.

4.0.100.5 Design Requirements for Guide Tube Inserts (i)

GT-Inserts shall be 0.75 inches O.D. minimum, completely cover the active fuel region (150 inches), and have a minimum boron content of 0.02434 grams of B-10 per cm3.

(ii)

Three (3) or 5 GT-Inserts are allowed. The orientation of every fuel assembly with 3 guide tube inserts shall be the same (Figure II-23).

(iii)

A 5-finger, full length Control Element Assembly (CEA) may be used in place of GT-inserts.

4.0.100.6 Design requirements For Erbia Assemblies containing 40 or 80 erbia rods shall have the erbia rods distributed per Figures II-24 and II-25. The minimum initial nominal erbia loading shall be 2.0 w/o Er203.

4.0.100.7 The Failed Fuel Rod Storage Basket (FFRSB)

The Failed Fuel Rod Storage Basket (FFRSB) shall be treated as if it were an assembly with enrichment and burnup of the rod in the basket with the most limiting combination of enrichment and burnup.

4.0.100.8 Non-Fuel Components Neutron sources and non-fuel bearing assembly components (thimble plugs, CEAs, etc.) may be stored in the fuel assemblies without affecting the storage requirements of these assemblies. A storage basket containing no fissile material can be stored in any storage location, and can be used as a storage cell blocker for reactivity control.

4.0.100.9 Fuel Assembly Reconstitution Station A Fuel assembly reconstitution station is a special case of a checkerboard pattern. A reconstitution station is permitted anywhere in the Region I racks. The empty cells in the checkerboard pattern do not need to be blocked. A reconstitution station is permitted anywhere in the Region II racks provided that empty cells in the checkerboard patterns are blocked to make it impossible to misload a fuel assembly during reconstitution activities.

The NRC staff found that the requirements specified in the proposed LCS 4.0.100 (Revision 2) for various fuel assembly storage patterns in the SFP were consistent with those included in the CAR (Reference 5) and discussed in Sections 2.3.3.1.2.7 and 2.3.3.1.3.2 of this SE, and the acceptability of the requirements was demonstrated by the criticality analysis and deboration analysis discussed in Section 2.0 of this SE. Therefore, the NRC staff determined that the proposed LCS changes are acceptable.

2.3.3.3 Licensing Basis Change As discussed in Section 2.3.2 of this report, the licensee proposed to change the licensing basis for the SONGS SFP from a 10 CFR 70.24 exemption to a 10 CFR 50.68 compliance. As such, the NRC staff reviewed the information provided in References 1 and 5 to determine whether the SFP complied with the 10 CFR 50.68 requirements. The licensee provided a description of how it complied with the requirements in 10 CFR 50.68(b) in Enclosure 3 of Reference 5.

These requirements include the following: (1) using plant procedures to ensure subcriticality and safe handling of fuel assemblies; (2) ensuring new fuel storage racks are subcritical by defined margins under both unborated and optimum moderation conditions; (3) verifying SPF racks are subcritical by defined margins under both borated and unborated conditions; (4) ensuring the quantity of Special Nuclear Material stored onsite is less than the quantity necessary for a critical mass; (5) providing radiation monitor in fuel storage and handling areas; (6) maintaining the maximum U235 enrichment of fresh fuel assemblies less than or equal to five percent by weight; and (7) updating the Final Safety Analysis Report in a timely fashion after choosing to comply with 10 CFR 50.68.

The NRC staff reviewed each of the requirements that did not require a criticality analysis to verify that the licensee would meet the conditions. The NRC staff found that the licensees responses as presented in Enclosure 3 of Reference 5 provided reasonable assurance that it would meet each of these requirements.

For requirements that require a criticality analysis to demonstrate compliance, the NRC staff reviewed the information provided by the licensee in support of this LAR (Reference 5). In the NRC staffs review of this LAR, it used the regulatory limits of Keff that are described in 10 CFR 50.68 for spent fuel storage racks. The NRC staffs review found the licensees criticality analyses acceptable and in compliance with the regulatory limits. Therefore, the NRC staff determined that the licensee will comply with all of the requirements of 10 CFR 50.68 and that the change in the licensing basis for the SONGS SFP is adequate and acceptable.

2.3.4 Summary The NRC staff reviewed the effects of the LAR using the appropriate requirements of 10 CFR 50.68 and GDC 62. The NRC staff found that the proposed changes to TSs and LCS in the LAR correctly reflected the results of the acceptable criticality analysis, which provided reasonable assurance that under both normal and accident conditions, the licensee would be able to safely operate the plant and comply with NRC regulations. Therefore, the NRC staff determined that the proposed changes to TSs and LCS are acceptable.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendment. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published June 6, 2006 (71 FR 32606). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

6.0 REFERENCES

1.

Letter from B. Katz, SCE to NRC,

Subject:

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," April 28, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML061220701).

2.

Letter from A. Edward Scherer, SCE to NRC, "Docket Nos. 50-361 and 50-362, Additional Information in Support of Amendment Application Numbers 243 and 227 (TAC Nos. MD1405 and MD1406), San Onofre Nuclear Generating Station, Units 2 and 3," November 13, 2006 (ADAMS Accession No. ML063210425).

3.

Letter from A. Edward Scherer, SCE to NRC, "Docket Nos. 50-361 and 50-362, Additional Information in Support of Amendment Application Numbers 243 and 227 (TAC Nos. MD1405 and MD1406), San Onofre Nuclear Generating Station, Units 2 and 3," December 22, 2006 (ADAMS Accession No. ML063610042).

4.

Letter from James Reilly, SCE to NRC, "Docket Nos. 50-361 and 50-362, Additional Information in Support of Amendment Application, Numbers 243 and 227 (TAC Nos.

MD1405 and MD1406), Request to Revise Fuel Storage Pool Boron Concentration, San Onofre Nuclear Generating Station, Units 2 and 3," May 7, 2007 (ADAMS Accession No. ML071280703).

5.

Letter from A. Edward Scherer, SCE to NRC, "Docket Nos. 50-361 and 50-362, Additional Information in Support of Amendment Application, Numbers 243 and 227 (TAC Nos. MD1405 and MD1406), Request to Revise Fuel Storage Pool Boron Concentration, San Onofre Nuclear Generating Station, Units 2 and 3," June 15, 2007 (ADAMS Accession No. ML071700097).

6.

Letter from A. Edward Scherer, SCE to NRC, "Docket Nos. 50-361 and 50-362, Additional Information in Support of Amendment Application, Numbers 243 and 227 (TAC Nos. MD1405 and MD1406), Request to Revise Fuel Storage Pool Boron Concentration, San Onofre Nuclear Generating Station, Units 2 and 3," July 27, 2007 (ADAMS Accession No. ML072130057).

7.

Letter from A. Edward Scherer, SCE to NRC, "Docket Nos. 50-361 and 50-362, Additional Information in Support of Amendment Application, Numbers 243 and 227 (TAC Nos. MD1405 and MD1406), Request to Revise Fuel Storage Pool Boron Concentration, San Onofre Nuclear Generating Station, Units 2 and 3," September 11, 2007 (ADAMS Accession No. ML072550304).

8.

10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants.

9.

BAW-1484-7, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, July 1979.

10.

Letter from A. Thadani (NRC) to S. Toelle (CE), Acceptance for Referencing of Topical Report CENPD-382-P, Methodology for Core Designs Containing Eribum Burnable Absorbers (TAC Nos. M79061 and M82959), June 29, 1993.

11.

Letter from NRC to SCE, "Issuance of Amendment for San Onofre Nuclear Generating Station, Unit 2 (TAC No. M94624 and Unit 3 (TAC NO. M94625," October 3, 1996 (ADAMS Accession No. ML022000232).

12.

Memorandum from L. Kopp (NRC) to T. Collins (NRC), "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 19, 1998.

13.

SCE-9001-A, "Southern California Edison Company PWR Reactor Physics Methodology Using CASMO-3/SIMULATE-3," September 1992.

14.

Letter from B. Moroney (NRC) to J. A. Stall (FPL), St. Lucie, Unit 1, License Amendment, Permits Credit Soluble Boron, Fuel Loading Restrictions & Control Element Assemblies in Spent Fuel Pool Criticality Analyses & Eliminate Need to Credit Boraflex Neutron Absorbing Material for Reactivity Control (TAC No. MB68640), September 23, 2004 (ADAMS Accession No. ML072670562).

15.

Letter from G. S. Vissing (NRC) to J. F. Opeka (Northeast Nuclear Energy Company),

"Issuance of Amendment (TAC NO. M86361)," March 1, 1994.

16.

Letter from L. Olshan (NRC) to W. R. McCollum, Jr. (Duke Energy), Oconee Nuclear Station, Units 1, 2, and 3, RE: Issuance of Amendments (TAC NOS. MB0894, MB0895, and MB0896), April 22, 2002 (ADAMS Accession No. ML020930470).

Principal Contributors: S. Sun J. Ma J. Wilson G. Waig Date: September 27, 2007

v t e Letter Sequence Approval
TAC:MD1405, (Approved, Closed) TAC:MD1406, (Approved, Closed)
Initiation Request Acceptance... Supplement, Supplement Administration Questions 26 October 2006 16 March 2007 12 April 2007 7 May 2007 14 May 2007 18 April 2007 20 April 2007 Answers 27 July 2007 Results Approval Other: ML063210425, ML063610042, ML072550183
MONTHYEAR2007-04-20 [Table View]

ML072550175

Text

SUBJECT:

Dear Mr. Rosenblum:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

Background

3.0 STATE CONSULTATION

4.0 ENVIRONMENTAL CONSIDERATION

5.0 CONCLUSION

6.0 REFERENCES

Subject:

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