ML071700097

Amendment Application Numbers 243, Supplement 1 and 227, Supplement 1, Proposed Change Number (PCN) 556, Revision 1, Request to Revise Fuel Storage Pool Boron Concentration

ML071700097
Person / Time
Site:	San Onofre
Issue date:	06/15/2007
From:	Reilly J Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PCN 556
Download: ML071700097 (411)

Text

EDISON SOUTHERN CALIFORNIA James T. Rely Vice President An EDISON INTERNATIONAL Company June 15, 2007 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Docket Nos. 50-361 and 50-362 Amendment Application Numbers 243, Supplement I and 227, Supplement 1 Proposed Change Number (PCN) 556, Revision I Request to Revise Fuel Storage Pool Boron Concentration San Onofre Nuclear Generating Station Units 2 and 3

Reference:

1) Letter from N. Kalyanam (NRC) to Richard M. Rosenblum (SCE) dated May 7, 2007;

Subject:

San Onofre Nuclear Generating Station, Units 2 and 3 - Request for Additional Information on the Proposed Amendment to Revise Fuel Storage Pool Boron Concentration (TAC Nos. MD1405 and MD 1406)

2) Letter from Brian Katz (SCE) to the U. S. Nuclear Regulatory Commission dated April 28, 2006;

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

Dear Sir or Madam:

By letter dated May 7, 2007, and through subsequent teleconferences, the U.S. Nuclear Regulatory Commission issued a request for additional information (Reference 1) regarding Proposed Change Number (PCN) 556. Southern California Edison (SCE) requested approval of Amendment Application Numbers 243 and 227, which consist of PCN 556, by letter dated April 28, 2006 (Reference 2). PCN 556 proposes to revise Technical Specifications 3.7.17, "Fuel Storage Pool Boron Concentration," 3.7.18, "Spent Fuel Assembly Storage," and 4.3, "Fuel Storage." This proposed change will increase the minimum allowed boron concentration of the spent fuel pool and allow credit for soluble boron, guide tube inserts (GT-Inserts) made from borated stainless steel, and Fuel Storage Patterns in place of Boraflex. This letter provides Amendment Application Numbers 243 Supplement 1 and 227 Supplement 1, which consist of PCN 556 Revision 1 (Enclosures 1, 2, and 3); and the answers to questions 1 through 24 and 27 from the NRC Staff (Enclosure 4).

P.O. Box 128 San Clemente, CA 92674-0128 4o 949-368-1480 Fax 949-368-1490

Document Control Desk June 15, 2007 The responses to questions 25 and 26 are not yet complete. SCE will provide the responses to these questions on or before July 31, 2007.

SCE has evaluated the supplemental information under the standards set forth in 10CFR50.92(c) and determined that SCE's original finding of "no significant hazards consideration" is not changed.

If you have any questions or require additional information, please contact Ms. Linda T.

Conklin at (949) 368-9443.

Sincerely, Enclosures

1. Notarized Affidavits

2. Licensee's Evaluation Attachments A. Existing Technical Specification Pages, Unit 2 B. Existing Technical Specification Pages, Unit 3 C. Proposed Technical Specification Pages, Redline and Strikeout, Unit 2 D. Proposed Technical Specification Pages, Redline and Strikeout, Unit 3 E. Proposed Technical Specification Pages, Unit 2 F. Proposed Technical Specification Pages, Unit 3 G. Proposed Bases Pages (For Information Only), Unit 2 H. Proposed Bases Pages (For Information Only), Unit 3 I. Proposed LCS 4.0.100, Unit 2 J. Proposed LCS 4.0.100, Unit 3 K. Spent Fuel Pool Dilution Analysis L. Spent Fuel Pool Criticality Analysis

3. Declaration of Compliance with 1 0CFR50.68

4. Responses to NRC Staff Questions Regarding Proposed Change Notice (PCN) 556 cc: B. S. Mallett, Regional Administrator, NRC Region IV N. Kalyanam, NRC Project Manager, San Onofre Units 2 and 3 C. C. Osterholtz, NRC Senior Resident Inspector, San Onofre Units 2 and 3 S. Y. Hsu, California Department of Health Services, Radiologic Health Branch

Enclosure 1 Notarized Affidavits

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA )

EDISON COMPANY, ET AL. for a Class 103 ) Docket No. 50-361 License to Acquire, Possess, and Use )

a Utilization Facility as Part of ) Amendment Application Unit No. 2 of the San Onofre Nuclear ) No. 243, Supplement 1 Generating Station )

SOUTHERN CALIFORNIA EDISON COMPANY et al., pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 243, Supplement 1. This amendment application consists of proposed change No. NPF-10-556, Revision 1 to Facility Operating License NPF-10. Proposed change No. NPF-10-556, Revision 1 is a request to revise Technical Specification 3.7.17, "Fuel Storage Pool Boron Concentration," 3.7.18, "Spent Fuel Assembly Storage," 4.3, "Fuel Storage," and Licensee Controlled Specification 4.0.100, "Fuel Storage Patterns." This proposed change will revise the minimum allowed boron concentration of the spent fuel pool and implement a Fuel Storage Program to allow credit for soluble boron, guide tube inserts, and Fuel Storage Patterns in place of Boraflex.

State of California County of San Diego es T. Re ice Pres t of Engineering and Technical Services Subscribed and sworn to (er-afiame) before me this ________day of

%t--ae- ,12007.

by: jtn- e-e 5 T", U1 personally known to me or proved to me on the baz3 cf 9aticfacto.-' evidence to be the person who appeared before me. -_-__- __- __-_

Notary Pubfic a *"

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA ))

EDISON COMPANY, ET AL. for a Class 103 Docket No. 50-362 License to Acquire, Possess, and Use )

a Utilization Facility as Part of ) Amendment Application Unit No. 3 of the San Onofre Nuclear ) No. 227, Supplement 1 Generating Station )

SOUTHERN CALIFORNIA EDISON COMPANY et al., pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 227, Supplement 1. This amendment application consists of proposed change No. NPF-15-556, Revision 1 to Facility Operating License NPF-15. Proposed change No. NPF-15-556, Revision 1 is a request to revise Technical Specification 3.7.17, "Fuel Storage Pool Boron Concentration," 3.7.18, "Spent Fuel Assembly Storage," 4.3, "Fuel Storage," and Licensee Controlled Specification 4.0.100, "Fuel Storage Patterns." This proposed change will revise the minimum allowed boron concentration of the spent fuel pool and implement a Fuel Storage Program to allow credit for soluble boron, guide tube inserts, and Fuel Storage Patterns in place of Boraflex.

State of California County of San Diego Pnic S"T. Reilly, Vite PEKnt of Engineering and Technical Services Subscribed and sworn to (rf-affifi9ed) before me this JJ1b day of To Y) 4e, 2007.

by: Taaie1s) I. ýe I/LI personally known to me er-*,uved to on tho -bmci of satisfaGctor eviddnco to be the person who appeared before me.

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Enclosure 2 Licensee's Evaluation PCN 556, Revision 1

Enclosure 2 LICENSEE'S EVALUATION PCN 556, Revision 1 SPENT FUEL POOL STORAGE WITH NO BORAFLEX AND CREDIT FOR SOLUBLE BORON SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 SECTION PAGE 1.0 INTRO D UC T IO N ................................................. 2 2.0 PRO POSED CHANGE ............................................ 2 3.0 BAC KG RO UND .................................................. 5

4.0 TECHNICAL ANALYSIS

........................................... 5 5.0 REGULATORY SAFETY ANALYSIS ................................. 10 5.1 No Significant Hazards Consideration ........................... 10 5.2 Applicable Regulatory Requirements/Criteria ..................... 15 5.3 Adm inistrative Controls ...................................... 17 5.4 C onclusion ............................................... 17

6.0 ENVIRONMENTAL CONSIDERATION

............................... 17 7.0 R EFER ENC ES ................................................. 18 ATTACHM ENTS A. Existing Technical Specification Pages, Unit 2 B. Existing Technical Specification Pages, Unit 3 C. Proposed Technical Specification Pages, Redline and Strikeout, Unit 2 D. Proposed Technical Specification Pages, Redline and Strikeout, Unit 3 E. Proposed Technical Specification Pages, Unit 2 F. Proposed Technical Specification Pages, Unit 3 G. Proposed Bases Pages (For Information Only), Unit 2 H. Proposed Bases Pages (For Information Only), Unit 3 I. Proposed LCS 4.0.100, Unit 2 J. Proposed LCS 4.0.100, Unit 3 K. Spent Fuel Pool Dilution Analysis L. Spent Fuel Pool Criticality Analysis

Enclosure 2 Licensee Evaluation Page 2 of 18

1.0 INTRODUCTION

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

This is a request to revise the following San Onofre Nuclear Generating Station (SONGS)

Units 2 and 3 Technical Specifications:

3.7.17, "Fuel Storage Pool Boron Concentration" 3.7.18, "Spent Fuel Assembly Storage" 4.3, "Fuel Storage", and to revise Licensee Controlled Specification:

4.0.100, "Fuel Storage Patterns."

2.0 PROPOSED CHANGE

TS 3.7.17 Fuel Storage Pool Boron Concentration LCO 3.7.17 is revised to increase the minimum boron concentration from 1,850 to 2,000 parts per million (ppm). The frequency of verification (7 days) is not changed.

Currently, soluble boron is not credited in determining fuel storage requirements which maintain Keff _<0.95. However, with the anticipated loss of Boraflex due to erosion/dissolution, as has been experienced in the industry, a minimum concentration of soluble boron will be required to maintain Keff to _<0.95.

The increase in the TS required concentration from 1,850 to 2,000 ppm ensures that there is no credible boron dilution event that would cause Keff to exceed 0.95. A final boron concentration of 970 ppm following a boron dilution accident would result in acceptable consequences. Conservatively, SCE has calculated that the minimum possible boron concentration following a boron dilution accident is 1700 ppm, assuming an initial boron concentration of 2000 ppm. The difference between 970 ppm (required for Keff < 0.95, non-accident conditions) and 1,700 ppm (assumed in the boron dilution analysis) is discretionary margin. To support the stricter controls being implemented by proposed TS 3.7.18, the applicability of TS 3.7.17 is being expanded 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 is being deleted.

Enclosure 2 Licensee Evaluation Page 3 of 18 TS 3.7.18 Spent Fuel Assembly Storage LCO 3.7.18 is completely revised. Figures 3.7.18-1 and 3.7.18-2 are being replaced with new Figures 3.7.18-1, 3.7.18-2, 3.7.18-3, and 3.7.18-4. These new figures show minimum burnup versus cooling time and enrichment for unrestricted and peripheral storage in the SONGS spent fuel pools. Additional Storage Patterns for SONGS Units 1, 2, and 3 fuel assemblies stored in Units 2 and 3 Spent Fuel Pools (SFP) will be contained in LCS 4.0.100. LCS 4.0.100 will be used to store fuel assemblies which do not meet the requirements of Figures 3.7.18-1 through 3.7.18-4.

Currently the Boraflex in the SFP racks limits Keff 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, which are more stringent than those currently in place.

TS 4.3.1 Criticality Sub-section 4.3.1.1 is revised as follows.

Items (b) through (i) are deleted and replaced with new items (b) through (I).

(b) Keff < 1.0 if fully 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 assembly's 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 I1.

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 I1. Credit for inserted 5-finger CEAs is taken for both Region I and Region I1.

(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) 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;

Enclosure 2 Licensee Evaluation Page 4 of 18 (h) 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 or 2 faces toward the spent fuel pool walls of Region I; (i) 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; (j) 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; (k) 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; and (I) Each SONGS 1 uranium dioxide spent fuel assembly stored in Region II shall be stored in accordance with Licensee Controlled Specification 4.0.100 Rev. 2.

Items (b), (c), (g), (h), (i), (j), (k), and (I) above incorporate new storage requirements to be implemented as a result of the anticipated loss of Boraflex. Item (d) describes the use of GT-Inserts, and Control Element Assemblies (CEA). 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 (I) above provide a reference to TS 3.7.18 and LCS 4.0.100 which contains expanded fuel storage requirements including the use of GT-Inserts.

Changes to LCS 4.0.100 (included for review and approval)

LCS 4.0.100 provides allowable fuel storage patterns, including the use of GT-Inserts.

This LCS provides storage patterns when the conditions of LCO Figures 3.7.18-1 through 3.7.18-4 can not be met. Once approved by the NRC, the date of NRC approval will be placed on each page of LCS 4.0.100. Future revisions of LCS 4.0.100 (pages 1 through 61, i.e., excluding the Bases pages) will be made in conjunction with a license amendment request associated with Technical Specification 4.3.1. Upon approval of PCN 556, XX/XX/XX at the bottom of each page of LCS 4.0.100 shall be replaced with the date of NRC approval.

Changes to TS Bases (included for information only)

Changes to the Bases of TS 3.7.17 and TS 3.7.18 are made for consistency with the proposed changes described above. They are provided in Attachments G and H for information only.

Enclosure 2 Licensee Evaluation Page 5 of 18

3.0 BACKGROUND

The spent fuel storage racks at SONGS Units 2 and 3 consist of two storage regions.

Region I is generally reserved for temporary storage of new fuel or partially irradiated fuel which does not qualify for Region II storage. Region II is generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying burnup levels.

As originally installed and currently licensed, both regions use Boraflex, a neutron absorbing material. Boraflex consists of fine boron carbide particles distributed in a polymeric silicone encapsulant.

The spent fuel storage racks at SONGS Units 2 and 3 are licensed to store two fuel assembly types:

(1) 16x16, Zircaloy or ZIRLO TM cladding, SONGS Units 2 and 3 fuel assemblies with a maximum enrichment of up to 4.80 weight percent (w/o)

(2) 14x14, Stainless Steel cladding, SONGS Unit 1 fuel assemblies (transhipped from Unit 1) with a maximum nominal enrichment of up to 4.00 w/o Erosion/dissolution of Boraflex in Pressurized Water Reactor (PWR) spent fuel pool racks is an industry-wide problem. Silica levels (an indicator of Boraflex dissolution) have been increasing in the SONGS Units 2 and 3 spent fuel pools and are approximately 6,400 ppb in Unit 2 and approximately 8,600 ppb in Unit 3, as of March 14, 2006.

Although there is currently sufficient Boraflex, the spent fuel storage rack criticality analyses have been redone assuming no Boraflex. Assuming no Boraflex for SONGS Units 2 and 3 will totally eliminate any Boraflex related reactivity concerns in the future, and monitoring programs will not be needed to ensure that an adequate amount of Boraflex is always present.

Consistent with other applications approved by the NRC, Southern California Edison (SCE) proposes to take credit for soluble boron in the spent fuel pool water, GT-Inserts and fuel storage patterns.

4.0 TECHNICAL ANALYSIS

The results of criticality, boron dilution, radiological decay heat, and structural/seismic analyses show that SONGS Units 1, 2, and 3 fuel assemblies can be safely stored in the SONGS Units 2 and 3 spent fuel racks assuming no Boraflex is present by taking credit for soluble boron, using GT-Inserts, and/or storing the fuel in analyzed fuel storage patterns.

The maximum fresh enrichment of SONGS Units 2 and 3 fuel assemblies is 4.8 w/o. The SONGS Units 2 and 3 fuel assemblies are 16x16 and have Zircaloy or ZIRLO TM cladding.

The maximum fresh enrichment of SONGS Unit 1 fuel assemblies is 4.0 w/o with an uncertainty of 0.05 w/o. The SONGS Unit 1 fuel assemblies are 14x14 and have stainless steel cladding.

Enclosure 2 Licensee Evaluation Page 6 of 18 Assuming no Boraflex, fuel storage patterns, and the use of GT-Inserts have been identified and analyzed in the criticality report (Attachment L) such that:

(1) Under normal conditions, the neutron multiplication factor, Keff, is less than 1.0, including all uncertainties (95/95 probability/confidence level), if the storage racks are flooded with unborated water.

(2) Under normal conditions, the neutron multiplication factor, Keff, is less than 0.95, including all uncertainties (95/95 probability/confidence level), if the storage racks are flooded with borated water at 970 ppm.

(3) Under accident conditions, the neutron multiplication factor, Keff, is less than or equal to 0.95, including all uncertainties (95/95 probability/confidence level), if the storage racks are flooded with borated water at 1,700 ppm.

Assuming 2,000 ppm soluble boron in the spent fuel pool water, a boron dilution analysis (Attachment K) shows that it is not credible for the soluble boron concentration to fall below 1,700 ppm.

The minimum assumed soluble boron for the fuel handling accident is 1,700 ppm. One is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, a final boron concentration of 970 ppm following a boron dilution accident would be acceptable. Conservatively, SCE has established a lirnit for final boron concentration following a boron dilution accident of 1,700 ppm. SCE requests that the difference between 970 ppm (required for Keff _<0.95, non-accident conditions) and 1,700 ppm (assumed in the boron dilution analysis) is to be considered a discretionary margin for use by SCE.

Previous decay heat and radiological analyses are not changed by the presence or absence of Boraflex in the spent fuel racks or the concentration of the soluble boron.

The proposed use of GT-Inserts is discussed in detail in the Spent Fuel Pool Criticality Analysis, Attachment L.

The mechanical design configuration of the GT-Inserts is similar to the shape, size, and weight of a control element assembly (CEA) finger. Each of the GT-Inserts is approximately 0.78 inch outside diameter (OD) solid stainless steel, with a boron content of approximately 2 w/o. A small counterbore is machined at the top for handling and a rounded bottom is machined. The OD of these GT-Inserts is less than that of a CEA finger. The material (borated stainless steel) is approved by the American Society for Testing and Materials (ASTM) and has been licensed by the NRC for use in spent fuel storage technologies and spent fuel pools.

The structural effect of adding GT-Inserts into the fuel assemblies will be small and within the capability of the spent fuel racks and the spent fuel pool.

Enclosure 2 Licensee Evaluation Page 7 of 18 The GT-Inserts will add a maximum weight of 120 lbs to the dry weight of a fuel assembly.

The 120 lbs is based on 5 GT-Inserts, each GT-Insert weighs about 22 lbs. The total dry _

weight of a fuel assembly with GT-Inserts will be 1660 lbs which is approximately 7.8%

heavier than the estimated standard fuel assembly weight of 1540 lbs that was considered in the design of the SONGS spent fuel racks. The spent fuel racks were also analyzed using a weight of 2904 Ibs, for the possible future storage of consolidated fuel.

The design of the spent fuel racks considered two load cases for each storage cell location: a dry weight of 1540 lbs for the standard fuel assembly and also a load of 2904 lbs. Tables 1 and 2 provide the comparison of the minimum margins to allowables for the two load cases of 1540 lbs and 2904 lbs per cell. The Margin to Allowable (MA) is calculated by comparing the acceptance limit with the applied load (see equation form below). The acceptance limit is the limit load of the structural component (see NF 3340 of ASME Code Section III), and the applied load is the factored load obtained for the applicable load combinations which include dead loads, live loads, seismic loads, and thermal loads.

MA = Allowable Load -1 Limit Load -1 Applied Load Factored Load Since a fuel assembly with GT-Inserts (1660 Ibs) will weigh less than 2904 lb., the spent fuel rack stresses will still be within allowables.

The reinforced concrete walls and basemat of the spent fuel pool were also evaluated for the load cases of 1540 lbs and 2904 lbs stored in each storage cell of the racks. The utilization factors for the governing concrete elements are presented in Table 3 for the two load cases. The utilization factor is defined as the percentage of resistance of the reinforced concrete section that has been utilized relative to the zero curvature line. A utilization factor of 100% indicates that the section is fully utilized by the design load. The largest utilization factor for a concrete section is 94% (at least a 6 % margin remains against the section allowable). The results show that the spent fuel pool will be capable of maintaining its structural integrity for the increase in weight from the GT-Inserts.

The acceptability of a fuel assembly drop with GT-Inserts was evaluated by comparing the small weight increase to the results of the drop analysis for the single fuel assembly weighing 1540 lbs. The structural acceptance criterion used for the drop analysis was that perforation of the spent fuel pool liner does not occur. The fuel assembly with GT-Inserts will weigh approximately 7.8% more than the analyzed fuel assembly. The maximum plate ily stress of the analyzed fuel assembly is only 43% of the ASME Code allowable limit (0.7 Su= 51100 psi) for Faulted Conditions. Therefore, there is ample stress margin in the liner plate to prevent perforation for the postulated drop of a fuel assembly with GT-Inserts.

The maximum fuel impact force on the rack storage cell is 3671 lbs for the load case of a fuel assembly weighing 2904 lbs. The allowable impact load on a C-E fuel assembly at the spacer grid is 6300 lbs. Therefore, the increased weight of the GT inserts will not lead to fuel damage.

Enclosure 2 Licensee Evaluation Page 8 of 18 Table 1 MINIMUM MARGIN(a) TO ALLOWABLE REGION I SPENT FUEL RACKS 1540 lb. Load /Cell 2904 lb. Load/Cell OBE DBE OBE DBE Support Pads 2.04 1.21 1.02 0.46 Cells 1.64 0.96 0.84 0.45 Grids 2.13 1.19 1.65 0.91 Cell to Cell Clips 2.01 1.01 1.65 0.57 Welds Cell to Grid 0.46 0.21 0.46 0.21 Cell to Clip 1.85 0.55 1.21 0.21 Grid to Grid 3.62 1.66 2.90 1.33 Grid to Base Plate 1.53 0.60 0.95 0.31 Cell Seam 1.40 0.41 1.12 0.21 Cell to Wrapper 0.81 0.57 0.35 0.23 Table 2 MINIMUM MARGIN(a) TO ALLOWABLE REGION II SPENT FUEL RACKS 1540 lb. Load /Cell 2904 lb. Load /Cell OBE DBE OBE DBE Support Pads 1.65 0.84 0.86 0.27 Cells 1.14 0.55 0.51 0.19 Welds Cell to Base Plate 1.38 0.46 0.83 0.19 Cell to Cell 1.03 0.46 1.15 0.23 Cell Seam 1.66 0.49 1.01 0.20 Cell to Wrapper 0.63 0.43 0.35 0.23 (a) Margin is defined as [Allowable Load/Applied Load] - 1. So any value greater than zero indicates a margin.

Enclosure 2 Licensee Evaluation Page 9 of 18 Table 3 EVALUATION RESULTS FOR THE SPENT FUEL POOL WALLS AND BASEMAT 1540 lb Load/Cell 2904 Ib, Load/Cell Utilization Factor(b) Utilization Factor(b)

North and South Walls Horizontal Reinforcement 88.4 90.8 Vertical Reinforcement 37.4 44.9 East Wall Horizontal Reinforcement 23.1 22.9 Vertical Reinforcement 47.0 59.8 West Wall Horizontal Reinforcement 28.1 28.4 Vertical Reinforcement 79.5 94.0 Basemat North-South Reinforcement 51.7 54.6 East-West Reinforcement 81.4 80.6 (b) The Utilization Factor is defined as the percentage of resistance to the reinforced concrete section that has been utilized relative to the zero-curvature line. A utilization factor of 100% indicates that the section is fully utilized by the design load.

Enclosure 2 Licensee Evaluation Page 10 of 18 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; therefore, it is not a primary coolant flow area. The fuel rack normal thermal cooling and malfunctioned blocked cooling scenarios are unaffected by the presence of the GT-Inserts in the fuel assemblies.

The possibility of accidentally withdrawing a GT-Insert is minimized because special tooling is required to remove it, and it will be completely contained within the guide tubes of the designated assemblies. Potential misloading of the GT-Inserts is minimized due to the design of the installation equipment, procedural controls, and the double verification requirement that will be in place to ensure the GT-Inserts are installed properly.

The possibility of accidentally withdrawing a CEA is minimized because specialized tooling is required for withdrawing a CEA from a fuel assembly. It is physically possible for the spent fuel handling tool to bind on a CEA after ungrappling from a fuel assembly and raising the tool. However, existing SONGS procedures require that the operator validate "tool weight only" on the spent fuel handling machine's load cell read out after ungrappling from a fuel assembly and raising the hoist slightly, and to report this information to the engineer directing fuel movement.

Two detailed reports are attached in Attachments K and L:

(Attachment K) Spent Fuel Pool Dilution Analysis, September 2005 (Attachment L) Spent Fuel Pool Criticality Analysis (With No Borafex And Credit For Soluble Boron), Revision 1, November 2005 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

NO.

Dropped Fuel Assembly There is no significant increase in the probability of a fuel assembly drop accident in the spent fuel pool when assuming a complete loss of the Boraflex panels in the spent fuel pool racks and considering the presence of soluble boron in the spent fuel pool water for criticality control.

Enclosure 2 Licensee Evaluation Page 11 of 18 Neither the presence of soluble boron in the spent fuel pool water, nor the placement of borated stainless steel guide tube inserts (GT-Inserts) in the fuel assemblies for criticality control, will increase the probability of a fuel assembly drop accident. The handling of the fuel assemblies in the spent fuel pool has always been performed in borated water, and the quantity of Boraflex remaining in the racks or GT-Inserts placed in the fuel assemblies, has no affect on the probability of such a drop accident.

Southern California Edison (SCE) has performed a criticality analysis which shows that the consequences of a fuel assembly drop accident in the spent fuel pool are not affected when considering a complete loss of the Boraflex in the spent fuel racks and the presence of soluble boron. The rack Keff remains less than or equal to 0.95.

The fuel, the fuel rack, and the fuel pool qualifications have been evaluated and determined to be unaffected by the installation of the GT-Inserts. The mechanical design configuration of the GT-Inserts is similar to the shape, size, and weight of a control element assembly (CEA) finger. Each of the GT-Inserts are approximately 0.78 inch outside diameter (OD) solid stainless steel, with a boron content of approximately 2 weight percent (w/o). A small counterbore is machined at the top for handling and a rounded bottom is machined. The OD of these GT-Inserts is less than that of a CEA finger. The material (borated stainless steel) is American Society for Testing and Materials (ASTM) approved and has been licensed by the United States Nuclear Regulatory Commission (NRC) for use in spent fuel storage technologies and spent fuel pools. The structural effect of the weight of the GT-Inserts on the fuel, the fuel rack, and the fuel pool structural interfaces and drop qualifications are unaffected. This is because the addition of five GT-Inserts (which increases the dry weight of a fuel assembly by 120 lbs.)

brings the total weight to 1660 lbs. which is enveloped by the 2904 lbs.

assumed in the calculation for fuel rack design.

Fuel Misloading There is no significant increase in the probability of the accidental misloading of spent fuel assemblies into the spent fuel racks when assuming a complete loss of the Boraflex panels and considering the presence of soluble boron in the pool water for criticality control. Fuel assembly placement will continue to be controlled pursuant to approved fuel handling procedures and will be in accordance with Technical Specification (TS) 3.7.18 "Spent Fuel Assembly Storage" and Licensee Controlled Specification (LCS) 4.0.100, "Fuel Storage Patterns," which will specify spent fuel rack storage configuration limitations.

Enclosure 2 Licensee Evaluation Page 12 of 18 There is no increase in the consequences of the accidental misloading of a spent fuel assembly into the spent fuel racks. The criticality analysis, performed by SCE, demonstrates that the pool Keff will be maintained less than or equal to 0.95 following an accidental misloading by the boron concentration of the pool. The proposed TS 3.7.17 "Fuel Storage Pool Boron Concentration" will ensure that an adequate spent fuel pool boron concentration is maintained.

Change in Spent Fuel Pool Temperature There is no significant increase in the probability of either the loss of normal cooling to the spent fuel pool water or a decrease in pool water temperature from a large emergency makeup when assuming a complete loss of the Boraflex panels and considering the presence of soluble boron in the spent fuel pool water. A high proposed concentration (> 2000 parts per million (ppm)) of soluble boron is consistent with current operating practices maintained in the spent fuel pool water. The proposed minimum boron concentration of 2000 ppm in TS 3.7.17 will ensure that an adequate concentration is maintained in the spent fuel pools.

A loss of normal cooling to the spent fuel pool water causes an increase in the temperature of the water passing through the stored fuel assemblies. This causes a decrease in water density, and when coupled with the assumption of a complete loss of Boraflex, may result in a positive reactivity addition.

However, the additional negative reactivity provided by the boron concentration limit in the proposed TS 3.7.17 will compensate for the increased reactivity which could result from a loss of spent fuel pool cooling.

Because adequate soluble boron will be maintained in the spent fuel pool water to maintain Keft less than or equal to 0.95, the consequences of a loss of normal cooling to the spent fuel pool will not be increased.

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; therefore, it is not a primary coolant flow area. The fuel rack normal thermal cooling and malfunctioned blocked cooling scenarios are unaffected by the presence of the GT-Inserts in the fuel assemblies.

This proposed change does not involve an increase in the probability or consequences of an accident previously evaluated.

Enclosure 2 Licensee Evaluation Page 13 of 18

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

NO.

The consideration of criticality accidents in the spent fuel pool are not new or different. They have been analyzed in the Updated Final Safety Analysis Report (UFSAR) and in previous submittals to the NRC.

Specific accidents considered and evaluated include fuel assembly drop, fuel assembly misloading in the racks, and spent fuel pool water temperature changes.

The possibility for creating a new or different kind of accident is not credible. Neither Boraflex or soluble boron are accident initiators. The proposed change takes credit for soluble boron in the spent fuel pool while maintaining the necessary margin of safety. Because soluble boron has always been present in the spent fuel pool, a dilution of the spent fuel pool soluble boron has always been a possibility. However, a criticality accident resulting from a dilution accident was not considered credible. For this proposed amendment, SCE performed a spent fuel pool dilution analysis, which demonstrated that a dilution of the boron concentration in the spent fuel pool water which could increase the rack Keff to greater than 0.95 (constituting a reduction of the required margin to criticality) is not a credible event. The requirement to maintain boron concentration in the spent fuel pool water for reactivity control will have no effect on normal pool operations and maintenance. There are no changes in equipment design or in plant configuration.

The possibility of accidentally withdrawing a GT-Insert is minimized because special tooling is required to remove it, and itis completely contained within the guide tubes of the designated assemblies. Potential misloading of the GT-Inserts is minimized due to the design of the installation equipment, procedural controls, and double verification that will be in place to ensure the GT-Inserts are installed properly.

The possibility of accidentally withdrawing a CEA is minimized because specialized tooling is required for withdrawing a CEA from a fuel assembly. It is physically possible for the spent fuel handling tool to bind on a CEA after ungrappling from a fuel assembly and raising the tool.

However, existing SONGS procedures require that the operator validate "tool weight only" on the spent fuel handling machine's load cell read out after ungrappling from a fuel assembly and raising the hoist slightly, and to report this information to the engineer directing fuel movement.

Therefore, the proposed change will not result in the possibility of a new or different kind of accident from any accident previously evaluated.

Enclosure 2 Licensee Evaluation Page 14 of 18

3. Does the proposed change involve a significant reduction in a margin of safety?

NO.

The TS changes proposed by this license amendment request and the resulting spent fuel storage operation limits will provide adequate safety margin to ensure that the stored fuel assembly array will always remain subcritical. Those limits are based on a San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 plant specific analysis that was performed in accordance with a methodology previously approved by the NRC.

The proposed change takes partial credit for soluble boron in the spent fuel pool. SCE's analyses show that spent fuel storage requirements meet the following NRC acceptance criteria for preventing criticality outside the reactor:

(1) The neutron multiplication factor, Keff, including all uncertainties, shall be less than 1.0 when flooded with unborated water, and, (2) The neutron multiplication factor, Keff, including all uncertainties, shall be less than or equal to 0.95 when flooded with borated water.

The criticality analysis utilized credit for soluble boron to ensure Keff will be less than or equal to 0.95 under normal circumstances, and storage configurations have been defined using a 95/95 Keff calculation to ensure that the spent fuel rack will be less than 1.0 with no soluble boron.

Soluble boron credit is used to provide safety margin by maintaining Keff less than or equal to 0.95 including uncertainties, tolerances and accident conditions in the presence of spent fuel pool soluble boron.

SCE evaluated the loss of a substantial amount of soluble boron from the spent fuel pool water which could lead to Keff exceeding 0.95 and showed that it was not credible.

Also, the spent fuel rack Keff will remain less than 1.0 with the spent fuel pool flooded with unborated water.

Decay heat, radiological effects, and seismic loads are unchanged by the absence of Boraflex.

The mechanical properties and weight of the fuel assemblies remain essentially unchanged with the inclusion of the weight of five GT-Inserts per assembly. The original mechanical and thermal analysis of the fuel assembly/fuel rack and fuel pool building interfaces currently approved remain valid and conservative.

Therefore, the proposed change does not involve a significant reduction in the plant's margin of safety.

Enclosure 2 Licensee Evaluation Page 15 of 18 5.2 Applicable Regulatory Requirements/Criteria Code of Federal Regulations, 10 CFR 50, Appendix A Criterion 61 -- Fuel storage and handling and radioactivity control. The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

Criterion 62 -- Prevention of criticality in fuel storage and handling. Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

Criterion 63 -- Monitoring fuel and waste storage. Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.

The NRC initially reviewed the fuel handling and storage at San Onofre Units 2 and 3 and found it acceptable, as documented in NUREG 0712, Reference 1. On May 1, 1990, the NRC staff issued the safety evaluation report, Reference 2, prior to spent fuel pool rack replacement. As documented in Reference 2, the staff found that SCE met the requirements of 10 CFR 50, Appendix A, General Design Criterion 61, regarding the capability to permit appropriate periodic inspection and testing of fuel storage components, and General Design Criterion 62, regarding prevention of criticality by the structural integrity of components and of the neutron absorber. On September 19, 1996, a letter to SCE from the NRC, Reference 3, documented the completion of a detailed review of spent fuel storage pool safety issues. The results of the staff's review are documented in a report to the Commission enclosed with the September 19, 1996, letter. In the report, the staff concluded that existing structures, systems, and components related to the storage of irradiated fuel provide adequate protection of public health and safety.

This report addressed adherence to General Design Criteria 61, 62, and 63. By letter dated October 3, 1996, Reference 4, the NRC issued amendments which revised Technical Specification Section 4.3, "Fuel Storage," to allow fuel assemblies having a maximum U-235 enrichment of 4.8 weight percent (w/o) to be stored in the spent fuel racks. The safety evaluation report enclosed with the October 3, 1996, letter stated that General Design Criterion 62 continued to be satisfied.

Enclosure 2 Licensee Evaluation Page 16 of 18 SCE is anticipating that in the future, the Boraflex panels that line the fuel storage cells will cease to effectively absorb neutrons, and therefore, SCE Will not be able to credit Boraflex in maintaining Keff to less than 0.95. The purpose of the proposed technical specifications is to provide fuel storage design (without Boraflex), and fuel storage configuration requirements (based on initial enrichment, burnup, and cooling time), to maintain Keff less than 1, and to maintain Keff less than or equal to 0.95 with soluble boron. The anticipated loss of Boraflex and the crediting of soluble boron, already present in the spent fuel pool, does not affect the ability of the fuel storage and handling system to meet Criterion 61 and Criterion 63. Therefore, the previously reviewed findings that General Design Criterion 61 and 63 are met at San Onofre Units 2 and 3 are not affected by this proposed change.

LCS 4.0.100, "Fuel Storage Patterns," Attachments I and J, meets General Design Criterion 62 by establishing allowable storage configurations that will maintain Keff to less than 1.0. Soluble boron is credited to maintain Keff to less than or equal to 0.95.

Regulatory Guide Regulatory Guide 1.13 "Spent Fuel Storage Facility Design Basis" describes a method acceptable to the NRC staff for implementing Criterion 61. As stated above, the ability of the fuel storage and handling system to meet Criterion 61 is not affected by this proposed change.

Regulatory Correspondence Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, B. K.

Grimes, April 14, 1978, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," as amended by the NRC letter dated January 18, 1979 USNRC, Office Of Nuclear Reactor Regulation, Reactor Systems Branch, 1998, "Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light-Water Reactor Power Plants" NRC Regulatory Issue Summary 2001-12, May 18, 2001, "Nonconservatism in Pressurized Water Reactor Spent Fuel Storage Pool Reactivity Equivalencing Calculations" The criticality analysis, Attachment L, which supports TS 3.7.18 and LCS 4.0.100, Attachments I and J, addressed the guidance provided in the above correspondence.

Enclosure 2 Licensee Evaluation Page 17 of 18 10 CFR 50.36 (B)

The operating restrictions and design features that constitute an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, are contained in Technical Specifications 3.7.17, 3.7.18, and 4.3.

LCS 4.0.100 is consistent with NUREG 1432 (Reference 5). Technical Specification 4.3 directs that fuel be stored in compliance with the specific document (LCS 4.0.100 Rev. 2) containing specific configurations approved by the NRC.

5.3 Administrative Controls Storage of fuel assemblies in the SONGS Units 2 and 3 spent fuel storage racks will continue to be in accordance with the NRC approved administrative controls described in LCS 4.0.100.

LCS 4.0.100, is being submitted as part of this Amendment Application for NRC review and approval. Once approved by the NRC, SCE will issue LCS 4.0.100 Revision 2, with the date of the NRC approval. Future revisions of LCS 4.0.100 will made in conjunction with a license amendment request associated with Technical Specification 4.3.1.

5.4 Conclusion In conclusion, based on the considerations discussed above, (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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

The proposed Technical Specification amendment has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. SCE has determined that the proposed amendment does not involve a significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite,. nor increase individual or cumulative occupational radiation exposures. Therefore, the proposed amendment meets the criteria given in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Assessment.

Enclosure 2 Licensee Evaluation Page 18 of 18

7.0 REFERENCES

(1) NUREG-0712 Safety Evaluation Report Related to the Operation of San Onofre Nuclear Generating Station, Units 2 and 3, dated February 1981 (2) Letter from L. E. Kokajko (NRC) to H. B. Ray (SCE) dated May 1, 1990;

Subject:

Issuance of Amendment No. 87 to Facility Operating License No.

NPF-10 and Amendment No. 77 to Facility Operating License No. NPF-15, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, (TAC Nos. 68308 and 68309)

(3) Letter from M. B. Fields (NRC) to H. B. Ray (SCE) dated September 19, 1996;

Subject:

Resolution of Spent Fuel Storage Pool Safety Issues: Issuance of Final Report, San Onofre Nuclear Generating Station, Units. 2 and 3, (TAC No. M88094)

(4) Letter from M. B. Fields (NRC) to H. B. Ray (SCE) dated October 3, 1996, Issuance of Amendment for San Onofre Nuclear Generating Station Unit No. 2 (TAC NO. M94624) and Unit No. 3 (TAC NO. M94625)

(5) NRC: Standard Technical Specifications Combustion Engineering Plants (NUREG-1432, Vol. 1, Rev. 3)

PCN-556 Attachment A (Existing Technical Specification Pages, Unit 2)

Fuel Storage Pool Boron Concentration 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Fuel Storage Pool Boron Concentration LCO 3.7.17 The fuel storage pool boron concentration shall be

> 1850 ppm.

APPLICABILITY: When 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.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool ------------ NOTE-----------

boron concentration LCO 3.0.3 is not applicable.

not within limit.

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool.

AND A.2.1 Initiate action to Immediately restore fuel storage pool boron concentration to within limit.

OR A.2.2 Verify by Immediately administrative means Region II fuel storage pool verification has been performed since the last movement of fuel assemblies in the fuel storage pool.

SAN ONOFRE--UNIT 2 3.7-30 Amendment No. 127

Spent Fuel Assembly Storage 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Spent Fuel Assembly Storage LCO 3.7.18 The combination of initial enrichment and burnup of each SONGS 2 and 3 spent fuel assembly stored in Region II shall be within the acceptable burnup domain of Figure 3.7.18-1 or Figure 3.7.18-2, or the fuel assembly shall be stored in accordance with Licensee Controlled Specification 4.0.100.

The burnup of each SONGS I uranium dioxide spent fuel assembly stored in Region II shall be greater than or equal to 18.0 GWD/T for interior locations of 5.5 GWD/T for peripheral locations, or the fuel assembly shall be stored in accordance with Licensee Controlled Specification 4.0.100.

APPLICABILITY: Whenever any fuel assembly is stored in Region II of the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 -------- NOTE-------

LCO not met. LCO 3.0.3 is not applicable.

Initiate action to move Immediately the noncomplying fuel assembly from Region II.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify by administrative means the initial Prior to enrichment and burnup of the fuel assembly storing the is in accordance with LCO 3.7.18. fuel assembly in Region II SAN ONOFRE-UNIT 2 3.7-32 Amendment No. 7-, 131

Spent Fuel Assembly Storage 3.7.18 40 I i ,7- 4.-.

& ~-.--.-"-*------4

~35

.4- - -4. --. - - J-.---- .4. - . .~- 4

~30 c-25 SACCEPTABLE j, F I

• i , - * ,,,

.20

-I cmI0 z L-5 I_________ .7

, i,,r NOT ACCEPTAiB 7

0-1.

I,.

Zr'-._ i 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5 Initial U-235 Enrichment (w/o)

MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION II RACKS FIGURE 3.7.18-1 SAN ONOFRE-UNIT 2 3.7-33 Amendment No. +2-T, 131

Spent Fuel Assembly Storage 3.7.18 j40 035

~30 c-25

~20 0 15 CD 8 10 00 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION II PERIPHERAL POOL LOCATIONS FIGURE 3.7.18-2 SAN ONOFRE-UNIT 2 3.7-34 Amendment No. 4, 131

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.8 weight percent;

b. Keff

• 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;

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

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

e. Units 1, 2, and 3 fuel assemblies may be stored in Region I with no restrictions;

f. 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 II;

g. 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 or 2 faces toward the spent fuel pool walls of Region II;

h. Fuel assemblies with a burnup in the "unacceptable range" of Figure 3.7.18-1 and Figure 3.7.18-2 will be stored in compliance with the Licensee Controlled Specification 4.0.100; and

i. The burnup of each SONGS I uranium dioxide spent fuel assembly stored in Region II shall be greater than or equal to 18.0 GWD/T for interior locations or 5.5 GWD/T for peripheral locations, or the fuel assembly shall be stored in accordance with Licensee Controlled Specification 4.0.100.

(continued)

SAN ONOFRE-UNIT 2 4.0-4 Amendment No. 4, 131

PCN-556 Attachment B (Existing Technical Specification Pages, Unit 3)

Fuel Storage Pool Boron Concentration 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Fuel Storage Pool Boron Concentration LCO 3.7.17 The fuel storage pool boron concentration shall be

> 1850 ppm.

APPLICABILITY: When 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.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool -------------NOTE--------

boron concentration LCO 3.0.3 is not applicable.

not within limit.

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool.

AND A.2.1 Initiate action to restore fuel storage Immediately pool boron concentration to within limit.

OR A.2.2 Verify by administrative means Immediately Region II fuel storage pool verification has been performed since the last movement of fuel assemblies in the fuel storage pool.

SAN ONOFRE-UNIT 3 3.7-30 Amendment No. 116

Spent Fuel Assembly Storage 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Spent Fuel Assembly Storage LCO 3.7.18 The combination of initial enrichment and burnup of each SONGS 2 and 3 spent fuel assembly stored in Region II shall be within the acceptable burnup domain of Figure 3.7.18-1 or Figure 3.7.18-2, or the fuel assembly shall be stored in accordance with Licensee Controlled Specification 4.0.100.

The burnup of each SONGS I uranium dioxide spent fuel assembly stored in Region II shall be greater than or equal to 18.0 GWD/T for interior locations of 5.5 GWD/T for peripheral locations, or the fuel assembly shall be stored in accordance with Licensee Controlled Specification 4.0.100.

APPLICABILITY: Whenever any fuel assembly is stored in Region II of the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 -------- NOTE-------

LCO not met. LCO 3.0.3 is not applicable.

Initiate action to move Immediately the noncomplying fuel assembly from Region II.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify by administrative means the initial Prior to enrichment and burnup of the fuel assembly storing the is in accordance with LCO 3.7.18. fuel assembly in Region II SAN ONOFRE-UNIT 3 3.7-32 Amendment No. 41-+, 120

Spent Fuel Assembly Storage 3.7.18 40 Z35

~30 c-25

• ACCEPTABLE

" 20 a) 15 CU 10

- -i".-- - NOT ACCEPTABLE

.- 1--- --

• Ci~

1.5 2.0 2.5 3.0 3.5 4.0 4.5 5 Initial U-235 Enrichment (w/o)

MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION II RACKS FIGURE 3.7.18-1 SAN ONOFRE-UNIT 3 3.7-33 Amendment No. +-+-6, 120

Spent Fuel Assembly Storage 3.7.18

,--40 035 iý:30 cL 25

~20

~15 C"Z10 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION II PERIPHERAL POOL LOCATIONS FIGURE 3.7.18-2 SAN ONOFRE-UNIT 3 3.7-34 Amendment No. 1-i-E, 120

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.8 weight percent;

b. Keff

• 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;

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

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

e. Units 1, 2, and 3 fuel assemblies may be stored in Region I with no restrictions;

f. 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 II;

g. 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 or 2 faces toward the spent fuel pool walls of Region II;

h. Fuel assemblies with a burnup in the "unacceptable range" of Figure 3.7.18-1 and Figure 3.7.18-2 will be stored in compliance with the Licensee Controlled Specification 4.0.100; and

i. The burnup of each SONGS 1 uranium dioxide spent fuel assembly stored in Region II shall be greater than or equal to 18.0 GWD/T for interior locations or 5.5 GWD/T for peripheral locations, or the fuel assembly shall be stored in accordance with Licensee Controlled Specification 4.0.100.

(continued)

SAN ONOFRE--UNIT 3 4.0-4 Amendment No. --1-6, 120

PCN-556 Attachment C

(Proposed Technical Specification (Redline and Pages)

Strikeout, Unit 2)

Fuel Storage Pool Boron Concentration 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Fuel Storage Pool Boron Concentration LCO 3.7.17 The fuPi qtorage pool boron concentration shall be 85-02000 ppm.

APPLICABILITY: Whenever any fuel assembly+es-*e-eis stored in the fuel storage pool. and a fuel storage pool verifi.ation has not been performed since the last movement of fuel U a IIU I W-ttit113 ttI I t-III IUý-I aku I U~JK- JU . I.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool ------------ NOTE-----------

boron concentration LCO 3.0.3 is not applicable.

not within limit.

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool.

AND A.2--4 Initiate action to restore fuel storage Immediately pool boron concentration to within limit.

A.2.2 Verify by administrative means +/-mmedi-ate!y Region 11 fuel verification has beenf performed since the last movement of fuc, assemblies in the fuel storage pool.

SAN ONOFRE--UNIT 2 3.7-30 Amendment No. +24

Spent Fuel Assembly Storage 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Spent Fuel Assembly Storage LCO 3.7.18 The combination of initial enrichment and burnup of -ach SONGS 2 and 3 spent fuel assembly stored in Region I+ shall be within the acceptable burnup domain of F Figure 3.7.18-1 or Figure 3.7.18-2, or the fuel assembl" -h-11 _ stored in acriAnce 4 .3. 1I. with Liensee COntrlled Technical Specification I0--1-00.

The combination of initial enrichment and burnup of each,-,

SONGS 2 and 3 spent fuel assembly stored in Region II shall be within the acceptable burnup domain of Fiqure 3.7.18-3 or Fiqure 3.7.18-4, or the fuel assembly shall be stored in accordance with Technical Specification 4.3.1.1.,

The of burnup e Each SONGS 1 uranium dioxide s ent fuel Specification 4.3.4. 10-I-O.

APPLICABILITY: Whenever any fuel assembly is stored in ,Re*.i. 1II of-the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reouirements of the A.1 -------- NOTE-------

LCO not met. LCO 3.0.3 is not applicable.

Tnitilte action to moiee Immediately bring the noncomplying fuel asspmh1v 4- I Reg-i-oi9--linto compliance.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify by administrative means the initial Prior to st orimng enrichment., e-i'td-burnup, and cooling time of t4emoving a:fuel the fuel assembly -siare in accordance with assembly ."*

LCO 3.7.18-., or Desiqn Features 4.3.1.1, or Rec,,on4IT rII to LCS,4.0.100. Rev 2, dated xx/xx/xxxxi any spent fuel pool storage 1ocation.!

SAN ONOFRE--UNIT 2 3.7-32 Amendment No. 2-7, 131

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-1 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION I RACKS 25 iF20 0

L,

= 15 E10 al)

Cl) tLL5 0 .

2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

- OYears e 5 Years IF 10Years 0§ 15Years EE 20 Years SAN ONOFRE--UNIT 2 3.7-33 Amendment No. +/-27, 131

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-2 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN PERIPHERAL POOL LOCATIONS IN REGION I RACKS 15 10 c2o

-0 E

a)

C,,

C,, 5 I LL I

01 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

-- Ei- 0 Years -- Eý- 5SYears -- w- 10 Years- -- 15 Years--H- 20 Years SAN ONOFRE--UNIT 2 3.7-34 Amendment No. 127, 131

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-3 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION II RACKS 60- __

50 -- Acceptable Region -

0 --

• .2.40 " .... ...- '---. .. ...... - -"-'

C 00 30 E

~20 -

10 __ _ -jUnacceptable Rg 1 2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years -e 5 Years ---- 10Years -ý1 15Years F- 20Years SAN ONOFRE--UNIT 2 3.7-34ia Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-4 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN PERIPHERAL POOL LOCATIONS IN REGION II RACKS 40 r 030 E

M20~

U..

O .--

1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E 0 Years -e- 5 Years - 10 Years v 15 Years -- 20 Years SAN ONOFRE-UNIT 2 3.7-34b Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.8 weight percent;

b. Kef* n n< 1.0*if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; C. Kpff. 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 stainl-ess steel guide tlbe inse6rts are used, they will be installed in an assembly's center guide tube, the guide tube associated with the serial number, and the diagonally oppositeý 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.i A five-finger CEA may be installed in an assembly.

Fuel cn*'ta~ining a five-finger CEA, may bet,ýplaced in either Region I or Region II. Credit for inserted 5-finger CEAs is taken for both Region I and Region II.;

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

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

e. Units 1, 2, and 3 fuel assemblies may be stared in Region I mith6 restrictions.

(continued)

SAN ONOFRE--UNIT 2 4.0-4 Amendment No. 1'27 131

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1 Criticality (continued)

{g. 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; f h. 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 I or 2 face,, toward the spent fuel pool walls of in Region -I-I; i;. Un'its 2 and 3 fuel assemblies with a burnup in the "acceptable ranqe" of Fiqure 3.7.18-3 are allowed un~res~tri ctedstorage Regijon 'I I j.ý Units 2 and 3 fuel assemblies with a burnup in the "acceptable ranqe" of Fiqure 3.7.18-4 are allowed unrestricted storaqe in the peripheral pool, locations widtih 1 or 2 faces toward, the, spent fuel pool walls of Region II;!

hk. 'Un*itS 2 and 3 Ffuel assemblies with a burnup in the "unacceptable ranqe" of Figure 3.7.18-1!, Figure 3-:.,71.1i8-2,, Fig~ure 3.7.1873, and Figure 3.7.18--Z41 will be stored in compliance with, th-Licensee Controlled Specification 4.0.100 Rev. 2, dated xx/xx/xx; and

+-1. The burnu.p of eEach SONGS 1 uranium dioxide spent fuel assembly stored in Region II shall be-greeat-er than or equal to 18.0 GWD,'T for interior locations-or 5.5 GWD/T for peripheral locations, or the fuel

,ssemblyshal be stored in accordance with Licensee Controlled Specification 4.0.100 Rev. 2, dated xx/xx/xx.

(continued)

SAN ONOFRE--UNIT 2 4.0O-4,a Amendment No. 127,- 13

Attachment D (Proposed Technical Specification Pages)

(Redline and Strikeout, Unit 3)

Fuel Storage Pool Boron Concentration 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Fuel Storage Pool Boron Concentration LCO 3.7.17 The fuPl ctorage pool boron concentration shall be

> -- *02000 ppm.

APPLICABILITY: Whenever any fuel assembly-i-ex eis stored in the fuel storage pool. and a fuel storage pool verification has not been performed .snce the .. last.'.1 movement of fuel

..... kl'.1 "I .-r

• fi-th

, 1 *4 ..

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool ------------ NOTE-----------

boron concentration LCO 3.0.3 is not applicable.

not within limit.

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool.

AND A.2-+- Initiate action to restore fuel storage Immediately pool boron concentration to within limit.

OR A.2.2 Verify by administrative means immediately Regon H fupel veriicaionhas been rfome shftL uee-tk assemblies iný"the' fuel storage-pool SAN ONOFRE--UNIT 3 3.7-30 Amendment No. +46

Spent Fuel Assembly Storage 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Spent Fuel Assembly Storage LCO 3.7.18 The combination of initial enrichment and burnup of -ach SONGS 2 and 3 spent fuel assembly stored in Region I+ shall be within the acceptable burnup domain of Figure 3.7.18-1 or Figure 3.7.18-2, or the fuel assembly shall be stored in accordance with Licensee Controlled Technical Specification 4.3.1.1910*--.

The combination of initial enrichment and burnup of each SONGS 2 and3spent fuel assembly stored in Reqion II sha'll be within the acceptable burnup domain of Figure 3.7.18-3 or Fiqure 3.7.18-4, or the fuel assembly shall be stored in accordance with Technical Specification 4.3.1.1.

The burrup of e Each SONGS 1 uranium dioxide spent fuel assembly stored in Region II O,,u,, 6- rater, t........ , ual pe,,4 ph.e a,1,,

, ,r*,o a u l,,,lt

,eat, shall be stored in accordance with Licesee Controlled Technical, Specification 4.3.1.10.-1-0.

APPLICABILITY: Whenever any fuel assembly is stored iný'Regian I the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 -------- NOTE-------

LCO not met. LCO 3.0.3 is not applicable.

Tnit*B.te action to move bring the noncomplying Immediately fuel assepmhlvyrem Reg-i-n-l-4iJnto compliance.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify by administrative means the initial Prior to moving a enrichment., ftrd-burnup,, and cooling time of s-ta-ring-t-he fuel the fuel assembly i-iare' in accordance with assembly +/-ft LCO 3.7.18--, orý. Desiqn "Features. 4.3.1.1, or Reqion I or lto I LCS 4.0.100. Rev 2, dated xx/xx/xxxx any spent fuel pool storage location SAN ONOFRE--UNIT 3 3.7-32 Amendment No. 11-6*120

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-1 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION I RACKS 25

~20 0

=1 15 E10 CD, LL 0

2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

EE3 OYears e 5 Years 3-z 10Years-z-- 15Years 20 Years SAN ONOFRE--UNIT 3 3.7-33 Amendment No. 116,-12

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-2 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN PERIPHERAL POOL LOCATIONS IN REGION I RACKS 15

~10 E

Ui)

Ci) 5 LL 0

3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

-E 0 Years -- G- 5 Years -- Iv- 10 Years --A- 15 Years EE 20 Years SAN ONOFRE--UNIT 3 3.7-34 Amendment No. 116, 12-

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-3 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION II RACKS 60 SUbAcceptable RegionR

_10 .. . . . ... . ........

0-.- ---- I 2 3 45 Initial U-235 Enrichment (w/o)

EZI 0 Years e 5 Years -- 10 Years 1;F 15 Years -F- 20 Years SAN ONOFRE--UNIT 3 3.7 a Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-4 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN PERIPHERAL POOL LOCATIONS IN REGION II RACKS 40 --

Acceptable Region Al ]

o3Q- l (9

E cn20 -

E -

al)

Unacceptable Region 400 IL lI .. ... .7 0-- I I i

I , ,

1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E 0 Years E- 5 Years -* 10 Years s*z 15 Years --- 20 Years SAN ONOFRE--UNIT 3 3.7-34b Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.8 weight percent;

b. Kef *.95 < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties-as described in Section 9.1 of the UFSAR;

c. Kf* "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.ý Three o r five Borated' stainless steel guide tube inserts (GT-Inserts) may be used. When three.

Borated stainless steel guide tube inserts are

,used, they wi lIl be installed in an assembly's.

'center guide tube, the guide tube associated with'

,the serial number, and the diagonally opposi~te' guide tube. Fuel cont'a'ining GT-Inserts mayýýbe..

placed in either Region I or Region II. However,1

'credit for GT-Inserts is only taken for Region Hi1

storage.ý A five-finger CEA may be installed in an assembly.,l Fuel containing a fiVe-finger CEA may be placed in' either Region I or Region II. Credit for inserted.

35-finger CEAs is taken for both Region. I andRegiQn

.II.

e-.-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;

e. Uits 1, 2, and 3 fuel asIemblies may be stored in Reg, no rest-icotieons (conti nued)

SAN ONOFRE-UNIT 3 4.0-4 Amendment No. 16,-120

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1 Criticality (continued)

1. g. 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; gh. 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 or 2 faces toward the spent fuel pool walls of Region +-+I;

,i.ý Units 2 and 3 fuel assemblies with a burnup in the "acceptable range" of Fiqure 3.7.18-3 are allowed unrestricted storage in Region II;

?j. Units 2 and 3 fuel assemblies with a burnup in the

• "acceptable range" of Fiqure 3.7.18-4 are allowed unrestricted storaQe in the peripheral pool,ý locations with 1 or 2 faces-toward the spent ifueli pool walls of Region II;'

h-kUnwits 2 and 3 'Ffuel assemblies with a burnupin the "unacceptable ranqe" of Figure 3.7.18-l', Figure' 3.7.18-2, Figure 3.7.18-3, and Figure 3.7.18-Z4 will be stored in compliance with t4e-Licensee Controlled Specification 4.0.100 Rev. 2,' dated:

xx/xx/xx; and

+ ,T.burnup of e:Eiach SONGS 1 uranium dioxide spent The fuel assembly stored in Region II shall be-greater than or equal to 18.0 GWD/T for interior locations or 5.5 CWD/T for peripheral loations, or the fuel s-e...b.ly s,,,l be stored in accordance with Licensee Controlled Specification 4.0.100 Rev. 2, dated xx/xx/xx;. IA (continued)

SAN ONOFRE-UNIT 3 4.0O- 4',a Amendment No. 116, 120

PCN 556 Attachment E (Proposed Technical Specification Pages, Unit 2)

Fuel Storage Pool Boron Concentration 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Fuel Storage Pool Boron Concentration LCO 3.7.17 The fuel storage pool boron concentration shall be 2 2000 ppm.

APPLICABILITY: Whenever any fuel assembly is stored in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool ------------- NOTE-----------

boron concentration LCO 3.0.3 is not applicable.

not within limit.

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool.

AND A.2 Initiate action to Immediately restore fuel storage pool boron concentration to within limit.

SAN ONOFRE-UNIT 2 3.7-30 Amendment No.

Spent Fuel Assembly Storage 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Spent Fuel Assembly Storage LCO 3.7.18 The combination of initial enrichment and burnup of each SONGS 2 and 3 spent fuel assembly stored in Region I shall be within the acceptable burnup domain of Figure 3.7.18-1 or Figure 3.7.18-2, or the fuel assembly shall be stored in accordance with Technical Specification 4.3.1.1.

The combination of initial enrichment and burnup of each SONGS 2 and 3 spent fuel assembly stored in Region II shall be within the acceptable burnup domain of Figure 3.7.18-3 or Figure 3.7.18-4, or the fuel assembly shall be stored in accordance with Technical Specification 4.3.1.1.

Each SONGS I uranium dioxide spent fuel assembly stored in Region II shall be stored in accordance with Technical Specification 4.3.1.1.

APPLICABILITY: Whenever any fuel assembly is stored in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 -------- NOTE-------

LCO not met. LCO 3.0.3 is not applicable.

Initiate action to bring Immediately the noncomplying fuel assembly into compliance.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify by administrative means the initial Prior to moving a enrichment, burnup, and cooling time of the fuel assembly to fuel assembly are in accordance with LCO any spent fuel 3.7.18, or Design Features 4.3.1.1, or LCS pool storage 4.0.100. Rev 2, dated location.

SAN ONOFRE--UNIT 2 3.7-32 Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-1 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION I RACKS 25

~20 0

CL E10 al)

C,)

FD 0

2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

-- 0 Years e 5 Years -I-- 10 Years 2!- 15 Years ER 20 Years SAN ONOFRE--UNIT 2 3.7-33 Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-2 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN PERIPHERAL POOL LOCATIONS IN REGION I RACKS 15 10 E

W) 5 L-0L 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

-- E- 0OYears -- @o- 5 Years I-v-- 10 Years-~ 15 Years- -- 20 Years SAN ONOFRE--UNIT 2 3.7-34 Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-3 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION II RACKS 60 -

e:'40 _

C0 m30 -- ------

Accptbl Region.uacetoz

• - - elo ......

E 20 C-f' 10 Unacceptable Region 1 0 12 3 4 5 Initial U-235 Enrichment (w/o)

Eý 0 Years e 5 Years ---- 10 Years - 15 Years 9- 20 Years SAN ONOFRE--UNIT 2 3.7-34a Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-4 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN PERIPHERAL POOL LOCATIONS IN REGION II RACKS 40 030-(9 E

m20*

E ai)

Un

~310 0-1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E 0 Years -E 5 Years ----- 10 Years -I- 15 Years ---- 20 Years SAN ONOFRE-UNIT 2 3.7-34b Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.8 weight percent;

b. Kff < 1.0 if fully flooded with unborated water, wtich 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 assembly's 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. 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; (continued)

SAN ONOFRE--UNIT 2 4.0-4 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1 Criticality (continued)

h. 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 I or 2 faces toward the spent fuel pool walls of Region I;

i. 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;

j. 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;

k. 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 xx/xx/xx; and

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

(continued)

SAN ONOFRE--UNIT 2 4.0-4a Amendment No.

PCN-556 Attachment F (Proposed Technical Specification Pages, Unit 3)

Fuel Storage Pool Boron Concentration 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Fuel Storage Pool Boron Concentration LCO 3.7.17 The fuel storage pool boron concentration shall be

Ž 2000 ppm.

APPLICABILITY: Whenever any fuel assembly is stored in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool ------------- NOTE-----------

boron concentration LCO 3.0.3 is not applicable.

not within limit.

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool.

AND A.2 Initiate action to Immediately restore fuel storage pool boron concentration to within limit.

SAN ONOFRE--UNIT 3 3.7-30 Amendment No.

Spent Fuel Assembly Storage 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Spent Fuel Assembly Storage LCO 3.7.18 The combination of initial enrichment and burnup of each SONGS 2 and 3 spent fuel assembly stored in Region I shall be within the acceptable burnup domain of Figure 3.7.18-1 or Figure 3.7.18-2, or the fuel assembly shall be stored in accordance with Technical Specification 4.3.1.1.

The combination of initial enrichment and burnup of each SONGS 2 and 3 spent fuel assembly stored in Region II shall be within the acceptable burnup domain of Figure 3.7.18-3 or Figure 3.7.18-4, or the fuel assembly shall be stored in accordance with Technical Specification 4.3.1.1.

Each SONGS 1 uranium dioxide spent fuel assembly stored in Region II shall be stored in accordance with Technical Specification 4.3.1.1.

APPLICABILITY: Whenever any fuel assembly is stored in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 -------- NOTE-------

LCO not met. LCO 3.0.3 is not applicable.

Initiate action to bring Immediately the noncomplying fuel assembly into compliance.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify by administrative means the initial Prior to moving a enrichment, burnup, and cooling time of the fuel assembly to LCO fuel assembly are in accordance with 3.7.18, or Design Features 4.3.1.1, or LCS any spent fuel 4.3.100. pool storage location.

SAN ONOFRE--UNIT 3 3.7-32 Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-1 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION I RACKS 25 20 C!,

0.

E- 15 co

..0 E 10 U-LL) 0 L.

2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o) 0 Years ---- 5 Years -;, 10 Years 2ýv 15 Years ER 20 Years SAN ONOFRE--UNIT 3 3.7-33 Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-2 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN PERIPHERAL POOL LOCATIONS IN REGION I RACKS 15 10 CL E

a) un 5

3 L-0 L-3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

- 0 Years O- -- 6 5 Years - 10 Years 2ý- 15 Years H 20 Years SAN ONOFRE--UNIT 3 3.7-34 Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-3 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR UNRESTRICTED PLACEMENT OF SONGS 2 AND 3 FUEL IN REGION II RACKS 60

-50 40 ca 30 E

0 LL 1 I 2 3 4 5 Initial U-235 Enrichment (w/o)

SOYears -e- 5Years E 10Years IF 15Years -ý 20Years SAN ONOFRE--UNIT 3 3.7-34a Amendment No.

Spent Fuel Assembly Storage 3.7.18 FIGURE 3.7.18-4 MINIMUM BURNUP AND COOLING TIME VS. INITIAL ENRICHMENT FOR PLACEMENT OF SONGS 2 AND 3 FUEL IN PERIPHERAL POOL LOCATIONS IN REGION II RACKS 40 030 (9

m0 20-E a)

-610

=3L 0 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

---

0Years e 5 Years ---- 10Years  ;ý- 15Years --- 20Years SAN ONOFRE--UNIT 3 3.7-34b Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.8 weight percent;

b. Keff < 1.0 if fully 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-Inserts) may be used. When three Borated stainless steel guide tube inserts are used, they will be installed in an assembly's 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. 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; (continued)

SAN ONOFRE-UNIT 3 4.0-4 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1 Criticality (continued)

h. 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 or 2 faces toward the spent fuel pool walls of Region I;

i. 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;

j. 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;

k. 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 xx/xx/xx; and

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

(continued)

SAN ONOFRE-UNIT 3 4.0-4a Amendment No.

PCN-556 Attachment G (Proposed Bases Pages (For Information Only) Unit 2)

Fuel Storage Pool Boron Concentration B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Fuel Storage Pool Boron Concentration BASES BACKGROUND As described in LCO 3.7.18, "Spent Fuel Assembly Storage,"

fuel assemblies are stored in the spent fuel racks in accordance with criteria based on initial enrichment, &f-e4 discharge burnup, and cooling time (plutonium decay).

Although the water in the spent fuel pool is normally borated to >Ž 1850 2000 ppm, the criteria that limit the storage of a fuel assembly to specific rack locations is conservatively developed without takingcredit for boron while maintaining K*ff < 1.0. Credit for boron is taken to, maintain Kef,

• 0,95.9 APPLICABLE Soluble boron in the spent fuel pool is credited in SAFETY ANALYSES criticality, analyses for normal and acc~ident conditions.

Thie relevant ac'cidehts are 1) Fuel Assembly Dropped Horizontally On Top of the Racks, 2) Fuel Assembly Dropped Vertically Into a Storaqe Location Already Containinq a Fuel]

Assembly, 3),,Fuel,-Assembly Dropped to. the SFP Floor, and 4)ýý Fuel Misloadinq in either Reqion I or Reqion II. The limitinq accident is Fuel Misloading in either Region I or ~1 Region II.

A fuel assembly could be inadvertently loaded into a spent fuel rack location not allowed by LCO 3.7.18 (e.g., an un-irradiated fuel assembly or an insufficiently depleted fuel assembly). This accident is analyzed assuminq 'the misloadinq of one fresh assembly with the maximum permissible enrichment,:,oading t.. eei.. I..... pot racks 1it mic(x~r~ ie the wrst ae)u .*ledia assmbic o a crihmrtwhchbounds 48 w'.Aother typeof pstulated aeeident is assocatei;*d With a futl a's embflPy that is drafpped ont te uly oaded fucl peal storaerak. [ither incident -1ud, h-e a positiv reactUiit cfet derai4th agnt ciiaiy However t:Ihe 'neg'ativ'e re"actii~vi t'y eTff ct' of't'he' solTubl b"ýoro n compensates for the increased reactivity ca,-,sed by eit-e-ae'eof the two postulated accident scenario-s-.

(continued)

SAN ONOFRE-UNIT 2 B 3.7-71 Amendment No. *2*-04/2*o*3

Fuel Storage Pool Boron Concentration B 3.7.17 BASES (continued)

APPLICABLE Under normal, non-accident conditions, the soluble boron SAFETY ANALYSES needed to maintain Kf less than or equal to 0.95, including (continued) uncertainties, is 970 ppm. Under accident conditions, the soluble boron needed to maintain Kff less than or equal toý 0.95, including uncertainties, is 1700-ppm. A SFP boron dilution analysis shows that dilution from 2000 ppm to below 1700 is not credible. Therefore, the minimum required soluble boron concentration is 2000 ppm.F The concentration of dissolved boron in the fuel pool satisfies Criterion 2 of the NRC Policy Statement.

(conti nued)

SAN ONOFRE-UNIT 2 B 3.7-71a Amendment No. 127 n2/-nn

Fuel Storage Pool Boron Concentration B 3.7.17 BASES (continued)

LCO The specified concentration of 2000 ppm dissolved boron in the fuel pool preserves the assumptions used in the analyses of t p nial aeeident sce- . "

ri* described above. This concentration of dissolved boron is the minimum required conc'n÷t 'ion for fuel assembly storage and movement within the spent fuel pool.

APPLICABILITY This LCO appliec whenever .,1.14 fuel. 1--.. assemblies

-. .. 4- C. are

-I,1 -- stored

' in the spent fuel pool. u a e ,lf- the l Iot. a I 1.

f II g Ithe crif I V1. I In1e the v.r ifi cati oan VouI Uo1t thIre are me misoaded fIuel assemblies.

Wit. .. . It-^ t. e - ..

fue se*l

... 1m,*4l C.avme t  !*!' .. .. . , ..... r, I IIU . pL*U.JIIL 111

... IIi AII U for.

Iu  ; 1I UUU:U u~ I~UE I U as-,1ul y3;II UI -, U dU U op IpJJUe UfI-amb4--*

ACTIONS A.17.and A.2-,-&nd -A-.3 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentratJ"n vf h.oron in the spent fuel pool is less than required :2000 ppm, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. This does not preclude the movement of fuel assemblies to a safe position. In addition, action musho ... mmot .. n ," ni fiated to restore boron concentration to the required 2000 ppm. wi th in limit. Alternatel, an vcf i iA4dat A *n;A yIadmiinistratite means, of tHe, fucl I bll.. I iheI 3 Il IU as*I I.IU vIIII eL III embi i the fuel storage pool, can be performed.

If moving irradiated fuel assemblies while in MODE 5 or 6,

[CO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

(continued)

SAN ONOFRE-UNIT 2 B 3.7-72 Amendment No. 4-2'IT

Fuel Storage Pool Boron Concentration B 3.7.17 BASES (continued)

SURVEILLANCE SR 3.7.17.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed.

The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over a short period of time.

REFERENCES 1. UFSAR, Section 9.1.

SAN ONOFRE-UNIT 2 B 3.7-73 Amendment No. +2'4

Spent Fuel Assembly Storage B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Spent Fuel Assembly Storage BASES BACKGROUND Th ent fu 1 st rage The ionirraat?) nuc facility lear TUel isassemblies, design ed toor store burnedqither Tirra latea) He assemblies in a verticai ,confiyuratlon underwater. The storaqe pool js sized to store 1542 fuel ass m*Iie$. Two ty~es.sies of s pent fuel $torage r cks are use Reeign I and Region.HI). T e two Re ion I racks eacn contain. 15b ]orge locations eacn spaced ,O.4O inchns on center in 8 M1 array. Four.Regign H storage rgcs each contain u21 storage 9cations p arr e hn ex15 rem inang two a ixi*aray, Re n T ac

• /.regionil - ocati]o* are sp~cea ins 1g- 1 l

• in incses on cedter. a nee*

• WU~~~~~

b *. lU, 1 , J . L.~~

I3 i vy,* Il yG U,,II*..*r ,.IU*, ,I *,J. .* IIi b . 1 TomaiRI 15 ntain ac aimmercmn op, .8.5frsen5ulo o nnto inric (2) unPresriphera storage,minimum discharge burnupf and coolingl time requirements vs. initial enri chment,u (3)' 2x2 storaqe patterns, minimum discharqe burnup

'and coolinq time requirements vs.. initial:

enri chment ,j

(4), 3x3 storaqe patterns, minimum discharqe burnup and cooling time requirements vs. initial enrichment,!

(conti nued)

SAN ONOFRE--UNIT 2 B 3.7-74 Amendment No. 2-7 04/29/43

Spent Fuel Assembly Storage B 3.7.18 BASES (continued)

RACKGROUIIND (5) credit for inserted Control Element Assemblies (continued) (CEAs),

(6) credit for erbia in fresh assemblies, (7) credit for cooling time (Pu-241 decay), and,.

(8), credit for borated stainless steel guide tube inserts When snliihle hnron is credited, the following acceptance criteria apply:,

i(1)* Under normal conditions, the 95/95 neutron multiplication factor (Klff), including all uncertainties, shall be less than 1.0 when flooded with unborated water, and,'

,(2) Under normal and accident "conditions; the 95/95 neutron multiplication factor (Kjff), including

'all uncertainties, shall be less than or equal to 0.95 when flooded with borated water.

APPLICABLE The spent fuel storage facility is designed for SAFETY ANALYSES noncriticality by use of adequate spacing, and "flux trp" enn~trucI tin L hecb th- fu-l aggsemblies;1%Q Rr. ýi A*-

neutron ahsorhino stainleps steel ca.ns,- hnrated watpr withý a minimum solubjhle boron concentration of 970 nnm. and' storaae of fu*l assemblies in accordance with the.. .

administrative controls in LCO 3.7.,18 and LCS 4.0.100, "Fuel, Storage Patterns".,

The spent fuel assembly storage satisfies Criterion 2 of the NRC Policy Statement.

LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, in the accompanying LCO, ensure& that the K,,, of the spent fuel pool will always remain t 0.-95

< 1.00 under normal, non-accident conditions assuminq the pool to be ,flooded with unborated water. The Keff of the (continued)

SAN ONOFRE--UNIT 2 B 3.7-74'a Amendment No.

Spent Fuel Assembly Storage B 3.7.18 BASES (continued)

LCO spent fuel pool will always remain ! 0.95 under normal, (continued) non-accident conditions assuminq the pool to be flooded with tmborated water with a minimum soluble boron concentration of 970 ppm. The K.;, of the spent fuel pool will always*

remain

• 0.95 under accident conditions assuminq the pool to be flooded with borated water with a minimum soluble boron concentration of 1700 ppm. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool. aearding to the L**'-.

Fuel assemblies not meeting the LCO shall be stared in aeeardamee with Speeifieation 4.3.1.1.

(continued)

SAN ONOFRE--UNIT 2 B 3.7-74b Amendment No.

Spent Fuel Assembly Storage B 3.7.18 BASES (continued)

APPLICABILITY This Lrn Anliac whenever any fuel assembly is stored in Regions I and II of the spent fuel pool.

ACTIONS A.1 Re quired Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When tho rnnin,,ration of fuel assemblies stored in Region's Iand II of the spent fuel pool ic n-, 1n *ccordance with Figure 3. 1and F*i* u* 3...CO 3.7.18, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with the [CO.

If moving irradiated fuel assemblies while in MODE 5 or 6,

[CO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.18.1 REQUIREMENTS This SR verifies by administrative means that, ...... ,ilal enriehment and barig-r f 1-h fi-.l ncconmh1i/i,Iccftnradin, A rdinrwith LCO 3.7.18, or Design Features 4.3.1.1, or, C!LS4.0.10 in the a ,,flanjing [CO Pn I 1Ccmh 1i'6 c

• ha1-4-V"- 1r*,-e range of t1he LCO not stored in accordance with LCO 3.7.18, performance of this SR Will ensure compliahce with Specification 4.3.1.1. LCO 3.7.18.

Jhis surveillance is performed,,,prior to the initial storaqe

,of a fuel assembly in a spent fuel pool location and prio-r to each subsequent movement to a new location.!

REFERENCES UFSAR, Section 9.1.2.2.

SAN ONOFRE--UNIT 2 B 3.7-75 Amendment No. 127 04-/2n

'PCN-556 Attachment H (Proposed Bases Pages (For Information Only) Unit 3)

Fuel Storage Pool Boron Concentration B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Fuel Storage Pool Boron Concentration BASES BACKGROUND As described in LCO 3.7.18, "Spent Fuel Assembly Storage,"

fuel assemblies are stored in the spent fuel racks i-accordance with cri+-i+/- i h*cod nn init1l nnrirkhmanta-an4 discharge burnup, and cooling time (plutonium decay).

Although the water in t-he spent fuel pool is norma ly borated to Ž 185 -2000. ppm, the criteria that limit the storage of a fuel assembly to specific rack locations is conservatively developed without taking credit for boron while maint*aining K~ff <1T'.0. Credit for boron is taken to maintain Keff

• 0.95.ý APPLICABLE Soluble boron in the spent fuel pool is credited in SAFETY ANALYSES criticality analyses for normal andaccident conditions.,

The relevant accidents are 1) Fuel Assembly Dropped.

Horizontally On Top ofthe'Racks, 2) Fuel Assembly *;Droipped .......

Vertically Into a Storage Location Already Containinqga Fuel' Assembly. 3) Fuel Assembly Dropped to the SFP Floor, ýand 4)i Fuel Misloadinq in either Region I or Region II. The limitinq accident is Fuel Misloading in either Region I orI Region II.

A fuel assembly could be inadvertently loaded into a spent fuel rack location not allowed by LCO 3.7.18 (e g.-,-a fuel ra*k lacation m*t allowed by LCO 3.7.18 (e.g., an un-irradiated fuel assembly or an insufficiently depleted fuel a.s.sembly) . ThiJs-acc-ident is analyzed ass uminq the misloadinq Iof one fresh assembly with the maximum permissible enrichment"odng thegin II fue! pool racks th nie(33 r*rýa*y ii te.,

...... .o.rst a) un,, irradiated assembllics o an. enr ichment which bouinds 4-8 W/o Another t -typ fpostul ated accident is associatedwihafc assmbl that is droppedOto!.thefullyr loaded fuel pool stoErage rack. Eithericdetcud aca oii

-reactivit effect, decreasing the m-rgin tocii4iy However, the negative reactivity e'ýffect 'of ut~he sluble boron compensates for the increased reactivity ca--sed by-e-it+her~

,one--of the 'two-postulated accident scenario-s7.

(continued)

SAN ONOFRE--UNIT 3 B 3.7-71 Amendment No. 116 *4/2a943

Fuel Storage Pool Boron Concentration B 3.7.17 BASES (continued)

APPLICABLE Under normal, non-accident conditions, the soluble boron SAFETY ANALYSES needed to maintain Kff less than or equal to 0.95, including (continued) uncertainties, is 970 ppm. Under accident conditions, the soluble boron needed to maintain Kf less than or equal to.

0.95, including uncertainties, is 1700 ppm. A SFP boron dilution analysis shows that dilution from 2000 ppm to below 1700 is not credible. Therefore, the minimum required soluble boron concentration is 2000 ppm.

The concentration of dissolved boron in the fuel pool satisfies Criterion 2 of the NRC Policy Statement.

(conti nued)

SAN ONOFRE--UNIT 3 B 3.7-71;a" Amendment No. 116 04/29/n%

Fuel Storage Pool Boron Concentration B 3.7.17 BASES (continued)

LCO The specified concentration of 2000 ppm dissolved boron in the fuel pool preserves the assumptions used in the analyses of the potential accident scenarios described above. This concentration of dissolved boron is the minimum required concentrAtion for fuel assembly storage and movement within the spent fuel pool.

APPLICABILITY This LCO applies, whenever fuel assemblies are stored in the spent fuel pool.. until a complete spent fuel pol verifiation has been performed following the last lovement of fuel assemfblies in the spent fuel pool. This LCO does not apply following the verification since the verification wo~uld confirni that there are no ffisloadcd fuel asseniblies.

With no further fuel assemibly movements iprres, there is no potential for a mnisloaded fuel assembly or a dropped fuel assemblyf.-

ACTIONS A.17 ,and A.2-,-&ftid-4--3 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentrati~cnnnf moron in the spent fuel pool is less than required 2000 ppm, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. This does not preclude the movement of fuel assemblies to a safe position. In addition, action Must hp i.mmdi~+/-r1~i iPynitiated to restore boron concentration to :the required 2000,ppm Within limit. Alternately, an immeiateverif+66t46oqby administrative means, of the fuel st'orage pool fuel locations, to ensure proper locations of the 1fuel s'Jincc the last rnovenmcnt of fuel assemblies in the fuel storage pool, can be performed.

If moving irradiated fuel assemblies while in MODE 5 or 6,

[CO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

(contionued)

SAN ONOFRE--UNIT 3 B 3.7-72 Amendment No. ++6

Fuel Storage Pool Boron Concentration B 3.7.17 BASES (continued)

SURVEILLANCE SR 3.7.17.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed.

The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over a short period of time.

REFERENCES 1. UFSAR, Section 9.1.

SAN ONOFRE--UNIT 3 B 3.7-73 Amendment No. +4-6

Spent Fuel Assembly Storage B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Spent Fuel Assembly Storage BASES BACKGROUND The spent fuel storage facility is desi gned to store either new (nonirradiated) nuclear fuel assemblies, or burned (irradiated) fuel assemblies in a vertical configuration underwater. The storage pool is sized to store 1542 fuel assemblies. Two types~sizes 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. Four Region II storage racks each contain 210 storage locations in a 14x15 array. The remaining two Region,-'T "v:',- each contain 195 locations in a 13x15 array. All Region II locations are spaced 8.85 inches on center. e,sau^e 88 To mai4ntainK 0.95 for s1ent fuel of --maximum Penrichment to

-u 4.8 w/o. (1) soluble boron is credited and (2) followinc storaae patterns and borated stainless steel guide tube, ins~erts are used as needed:'

(1) unrestricted storaoe. minimum discharoe burnup and coolinc time requirements vs. initial:

en'richment,'

!(,2) SFP Peripheral storaae. minimum discharae burnup and icoolinq time requirements vs. initialI enrichment,:

(3) 2x? storace Datterns. minimum di~scharae burnup and coolinQ time requirements vs. initial, enrichment, (4) 3X3 storaoIe patterns. minimum discharae burnup and coolinq time requirements vs. initial enrichment, (conti nued)

SAN ONOFRE--UNIT 3 B 3.7-74 Amendment No. 116 29/0/3

Spent Fuel Assembly Storage B 3.7.18 BASES (continued)

RACKGROIIND (5) credit, for inserted Control Element Assemblies (continued) (CEAs),

(6) cf-edit, for erbia in fresh assemblies, (7), credit for cooling time (Pu-241 decay), and, (8) credit for borated stainless steel guide tube inserts:

When ,olihlp horon is credited, the following acceptance criteria apply:

() Unhder normal conditions, the 95/95 neutron multiplication factor (Kpff), including all uncertainties, shall be less than 1.0 when flooded with unborated~water, and,:

-(2)1 Under normal and acci~dent conditions, the 95/95!

neutron multiplication factor (Kpf,), including all uncertainties, shall be less than or equal to 0%95 when flooded with borated water.'

APPLICABLE The spent fuel storage facility is designed for SAFETY ANALYSES noncriticality by use of adequate spacing, amd ,,uxwrt--tia-p-A

-enmrtructicn whreby1 the fuel aseb 1-r~m in 2 iR 4 A* An+/-

npuitrnn ahsnrhina ctainless qtppl can,. hnratpd water with a minimum soluble hornn concentration of W7O nDm: and storaae of" fuel assembl ies in accordance with the......

administrative controls in LCO 3.7.18 and LCS 4.0.100, "Fuel Storage Patterns".:

The spent fuel assembly storage satisfies Criterion 2 of the NRC Policy Statement.

LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, in the accompanying LCO, ensures that the -K... of the sDent fuel Dool will always remain -t--95

< 1.00 und~e'r normal, non-accident conditions assumitn* the pool to be flooded with unborated water. The Keff of the (conti nued)

SAN ONOFRE--UNIT 3 B 3.7-74a: Amendment No. 16- 04/2/-9n3

Spent Fuel Assembly Storage B 3.7.18 BASES (continued)

LCO soent fuel oool will alwavs remain

• 0.95 under normal.

(continued non-accident conditions assumina the pool to be flooded with borated water with a minimum soluble boron concentration of 970 DDm. The K_ of the spent fuel pool will always remain 0.95 under accident conditions assuminc the oool to be flooded with tmborated water with a minimum soluble boron concentrationof i700 ppm. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool ..... rdig to the L"O.

Fuel assemblies not mieeting the CO shall be stored in ac..rdace. with Spe.ifi.ation 4.3.1.1.

(continued)

SAN ONOFRE--UNIT 3 B 3.7-74b Amendment No. 116 04/29/43

Spent Fuel Assembly Storage B 3.7.18 BASES (continued)

APPLICABILITY This Lrn rznnr1ic whenever any fuel assembly is stored in Regions I and II of the spent fuel pool.

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

Whan. +hn configuration of fuel assemblies stored in Regions

-Figure I and II: of the - spent

• fuel

.... Q..1I* is n+LC in,"'rrdance with ool Fig 37-81 1and , Figure 3.7.182- LCO 3.7.1a, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance-w-i-t-h the-Le-.

If moving irradiated fuel assemblies while in MODE 5 or 6,

[CO 3.0. would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.18.1 REQUIREMENTS This SR verifies by administrative means that +*6 iitial e mr ie ,he mt anIn ub i . . .. ' -h.. f - in....cc...c tr r in

-rrord.nnr. -,ith 'LCO-3.7.18, or Design Features 4.3.1.1, or LCS 4.0.100... the,,ee-,,in , i,,,, -H - . .. f1. 1l c.omh. i c +-

erae fte O not stored in accordance with LCO 3.7.,18, performance of this SR will ensure compliance with Specification 4.3.1.1.LGO 3.7.18.

This, urveilla~nce is nerformed nrior to the initial storaae,

,of a fuel assemblv in a snent fuel nool location and prior to each subse-quent movement to a new location.

REFERENCES UFSAR, Section 9.1.2.2.

SAN ONOFRE--UNIT 3 B 3.7-75 Amendment No. 116 04/29/-94

PCN-556 Attachment I (Proposed LCS 4.0.100 Unit 2)

(The date XX/XX/XX on each page of LCS 4.0.100 will be the date of NRC approval of License Amendment Application XXX)

Fuel Storage Patterns for Region,II Raks LCS 4.0.100 4.0 DESIGN FEATURES LCS 4.0.100 Fuel Storage Patterns for Region II Raks and Region- +

Raeks Reconstitution Statin NOTE:,

This Licensee Controlled Specification is listed by revision number and date in Technical Specification 4.3.1. All chanqes t~o paqes 1 throuqh 61, Rev. 2 dated XX/XX/XX of this LCS (i.e., excludinq 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.

VALIDITY STATEMENT: Rev. 2+ effective upon NRC approval'1-21/97, to be implemented within 3180 days.

4.0. 100 New or burned fuel (which does not meet the criteria of LCO 3.7.18 for unrestricted storaqe or storage at the pool periphery) may be stored in Reqion I or Reqion I i~+/-f-al"! the in accordance, with the allowable Storage Patterns described in this LCS.,follotvig conditions are ,,,t.

4.0. 100.1I Region I,, Region I Storaqe Patterns are qiven in Tables I-i through 1-8 and Figures I-1 through 1-9.

I u*I I pI - Aliaew or burnied fuil whi h does not mect the ,ritria of LCO 3.7.18 for unrstricted storae or storage at- tHe pool periphery.

4.0.100.2, Region II =Region II Storage Patterns are given in Tables 11-1 through 11-15 and Figures II-I through 11-22.

Fue!V Type 2 - Fuel 'GO 3.7. 'whieh ,o- _dbe§

,f _f mneet......

the the t.i'te4 cri teri a of sA N r..tor e -g in'fl. \

RegIo n II .

(A) Fue-l ....H 1 s qav intial enric hment t 4.8 (B) Fue I Tpe Js hall be I tored in ReI onr I1 in a chekcroar pttern The four (4R bsi eu i-rements 1 '1-,

J1.j - ,,' -4f 1I,,.. ^ -[4-4 -" -- ...

(1) Type 1 fuel assemblies can not be in adan lotn They can, however,

()Tp -1 l assembic. 1tordi ego sl aestwo(2) 4 w sids facing an empty lction.

(3 Te hepa ckerbd pater doe ronotneed t 6 no; (4) igure 1 prolIdes an illustration ofa -ceptabl fuel star-age pattern.

SAN ONOFRE-UNIT 2 4.0-100-1 Rev. 2'XX/XX/XXi

Fuel Storage Patterns for Region II Racks LCS 4.0.100 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:

(C) A reconstitution station is - sp-ecial ease of-a-checkerboard pattern. The ;Laberus for-checkerbOarding permit a reconstitution station anywhere in the Region 1I rocks. Single or multiple row reconstitution stations are permitted. Figure 4.0.100 2 provides ant

() Sam Onofre IUMi 1 fuel asse blies may be stored imn--egion Ii a 'Three if4the assemb- urntis out of four' patter least 1.7-GWB4.

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

the burnup is qreater than 25,000 MWD/T, and

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

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

the burnup is qreater 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 periphery) if:!

the burnup is qreater than 20,000 MWD/T. and.

the cooling time is greater than 0 years.

4.0. 100.5 Desig*n'Requirements For Guide Tube lnserts, (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-IO per cm'.

SAN ONOFRE-UNIT 2 4.0-100-2 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Ranks LCS 4.0.100 (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 11-23).

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

4.0.100.6 *Design requirements For Erbia Assemblies containinq 40 or 80 erbia rods shall have the erbia!

rod's' distributed per Fiqures 11-24 and :11-25. The mi~nimum!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 Storaqe 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

,burnupJ.

4.0.100.8 ýNon-Fuel Components Neut~ron sou;rces and non-fuelibearinq assembly c6mponents "(thimble pluqs, CEAs, etc) may be stored in fuel assemblies without!'-

affectinq the storaqe requirements of these assemblies. A

storage basket containing no fissile material can, be stored in'

,any storaqe'location, and can be used as a storage cel1 blo'cke r 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 permittedi anywhere in he Reqion I racks. The empty celIs in the .

.checkerboar"d Pattern do not need to be blocked.' A reconstitution

station is permitted anywhere in the Reqion II racks provided that empty cells in the checkerboard Pattern are blocked to make it .impossible to misload a fuel- assembly during recons-titution acctivities.;

SAN ONOFRE-UNIT 2 4.0-100-3 Rev. .2XX/XX/XX

ý T I'- -I.-

Fuel Storage Patterns for ,Re, LII I I4U.U0.

LCS 4.0.100

• x x * * * * *

• 0 __ 0 __ 0 0 *

• __ 0 __ 0 __ 0 __ *

• 0 __ 0 __ 0 0

• H~=~TYPE 2 FUE-L E -TYPE 1 FUEL FH7 -EMP TYlc4n.r I I Figu -tI re 4.U .100- J 0__ ~ 0 _ 0 0_0*

_ __ 0._ 0 _ 0 __ _

H-77 TYPE 2 Ftffb I-- I . . I /1 1 "

-- TYPE 4.E'II00I I ____Figure I 4.0.100-2 SAN ONOFRE-UNIT 2 4.0-100-4 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region 1" Raeks LCS 4.0.100 Table I-I REGION I Category I-i Fuel (Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 21.47 2.59 19.67 5.00 22.84. 20.04 a".50 18.761 16- . 8 9 17".'57 16.45 4.00 14.30 13. 58 13 .09 12.78 12.571 3.50 9.84 9.40 9.11 8.92 8.79 3.00 5o.:02 4.84 4 .79 2.47 0.00 0 .00 So.boo; 0.00 0.00 tategor) Category Categ-or Category goryI-Note: Category T-1 and fresh fuel with full-length 5-finger CEAsji(Tale 1-5) may. be ;tored together with no restrictions.

SAN ONOFRE-UNIT 2 4.0-100-5 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table 1-2 REGION I Category 1-2 Fuel (SFP Peripheral Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling

5. 00, 12.55 12.15. 11. 882 11 .61 1i.47!

[4.50 8 85, 8.49 8.40 9.09 8. 63ý

4. 00 5.58 5.43, 5.25 5.21 3 .50 2 .22 2.13 2.09 2.05 2.03 3.20 0.00 0.00 0.00 0.00 0.00 P P 0

pL 0

'0 0 L

o o 0 w w A 0 L 0 L

L 0 o o POOL WALL SAN ONOFRE-UNIT 2 4.0-100-6 Rev. .2'XX/XX/XX

Fuel Storage Patterns for Region 1. Rocks LCS 4.0.100 Table 1-3 REGION I Category 1-3 Fuel (Filler Assembly For l-out-of-4 Pattern)

Ihiti~al Minimum Burnup (GWD/T)'

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling 'Cooling 34.217 5.00 39.99 36.28 33.04 32.22 4.50 34.95 31.71 29.94 28.84 28.12 26 . 991 25.46 :23 .89:

4.00, 29-.71 :24'.511 3.50 24.22 22 .03 20.79. 20.02 19 .52 300, 18.37. 16.84 15.91 i5.34 14 .97

'2. 50 1$2!.2 l11.3oQ 10.72 10. 37, 10.13 2.00 5.28 5.05 4 . 85 4.72" 4.62-1.71' 0.00 0 .00 0. 00 0.00 0 .000 4.80 Category Fresh 1-,3 Category Ct~eg-or'y 1-31 1-3ý SAN ONOFRE-UNIT 2 4.0-100-7 Rev. 2 XX/Xk/XX

Fuel Storage Patterns for Region II, Raks LCS 4.0.100 Table 1-4 REGION I 4.80 w/o Fresh Fuel Checkerboard Pattern Initial Minimum Burnup (GWD/T)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 4.80 0. 00 0.00 :0. O0 0.00 .0.00o 4.80o Emptyý Fresh

• Empty *4.8 Fresh SAN ONOFRE-UNIT 24 4.0-100-8 Rev. 12XX/XX ,

/XX

Fuel Storage Patterns for Regi Ln 4IRa.ks LCS 4.0.100 Table 1-5 REGION I 4.8 w/o Fresh Fuel With Full-Length, 5-Finger CEA (Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o), Years Y 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 4.80 0-. 00 0.00 0.00 0.00 0.00 4.-80o 4.80 Fresh Fresh WithCEA With CEA

'4.80 4.80, Fresh Fresh With CEA lWith CEA Note: Category I-1 and fresh fuel with full-length 5-finger CEAs (Table 1-5) may be stored together with no restrictions, SAN ONOFRE-UNIT 2 4.0-100-9 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II RoIks LCS 4.0.100 Table 1-6 REGION I Category 1-4 Fuel (Filler Assembly For l-out-of-4 Pattern)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 26.57 24.71' 23.59 22.90. 22.44 4.50 22.12 20.62 19.73 19.17 18.80 4.00 1.7. '54 16.46 15.78 15.35 15.o07

3. 150 12. 84i 19' i2ý 11'. 66 11..37 11 .18 3.00 7.95 7.56 7.31 7.115; 7 . 05 2.50 2.76 2 . 64 2.56 2.50 2.46 2 .27 0.00 0.00 0.00 0 .o00 4 Category Fresh 1-4 80 Erbia Cateego r-y Category 1-4 1-4 SAN ONOFRE-UNIT 2 4.0-100-10 Rev. 2 X-X/XX/`XX

Fuel Storage Patterns for Region II Raks LCS 4.0.100 Table 1-7 REGION I Category 1-5 Fuel (Filler Assembly For l-out-of-4 Pattern)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/O), 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 30.81 28.40. 27.00 26.14 *25.57

'4.50 26.17 24-.17, 212.-99, 22.26 21.78 4.00 21 .32 19.771 18-.84 18.27 17.88 3.50 16.32 15.22 14.55 14.13 13 .85 3 ý 00 11 .11 10 .45 10.05 .9.61

.9.79 2.50 5.55 5.30 5.14 5.04 4.98 2.07 0.00 0.o0 0.00 0 00 0.00 4f Category Fresli I-5

,40 Erbia C e go9 ry Category 17S 1-5 SAN ONOFRE-UNIT 2 4.0-100-11 Rev. 2 XX/XX/0

Fuel Storage Patterns for Region 1< Raeks LCS 4.0.100 Table 1-8 REGION I Checkerboard Pattern Category 1-6 Fuel (4.80 w/o Assembly Depleted to 18.0 GWD/MTU)

Initial Minimum Burnup (GWD/MTU)

-Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 19.82 18. 841 i8.12]

1 17.67 17.37 4 1.50 15.83 15 .1.1 14.-58' 14.01 4-.00 11 .75 11.28 '10.92 10.69 10.54 3 .50 7.56 7.23 71.04 6.91 6.83

3. 00 3 .28 3.15 3 . 07: 3.03 2.99 2.65' 0.00 0.00, 0.00 0.00 0.00 Category 1-4 Fuel (Checkerboard Partner-For Category 1-6 Fuel) initial .Minimum Burnupo (GWD/MTU)ý Enrichment i(w/o):

0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 50~0 26.57 224, 7,1 23 .5-9 ;22.90g 19 . 17 22 .44-4A.50, 22 .12 20 .162 191. 73 '18.80 17.54 16.46 15.78, 15.35ý 15.07 3 . 501 12..84 12.12: 11. 66 11. 3 7i 11.18 3.00 7.95 7.56 7.31 7:.715, 7.05 2.76 2 '.64 2 .56' 50 2.46

2. 27 0.00 0 A00 0.00 ý0.00 0.00 Cate~gory Category Category Category 1-6 1-4 SAN ONOFRE-UNIT 2 4.0-100-12 Rev. 2-. XX/XXA,X,

Fuel Storage Patterns for Region II Raks LCS 4.0.100 Table II-i REGION II Category II-i Fuel (Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment

,(w/'o) 0 Years 5 Years. 10 Years 1,5i.Years 20 Years Cooling Cooling Cooling Cooling Cooling

5. 00 53.76 47.77 44.75. '43 .00. 41.86

,4.5 48.43 42.93 40.15 38.52' 37.47 4 .00 42.91 37.94 35.40 -33 .92 32.96 3.00 30.99 27.26 25.30 24.161 23.43

2. 00 17.05 14 .97 13 .90 13 .

25. 12.83 1.87 14.93, 131.23 12.2 61

!1.6 11.31 1 .23 0.00 0:.00 02.00 0,).o00 0.00 Category CategoIy it-i it-i Note: Category I-i1, 11-8, 11-9, and II-15'may be stored together with no restrictions.

SAN ONOFRE-UNIT 2 4.0-100-13 Rev. 2 XX/xx/xx:

Fuel Storage Patterris for Region 11 Raeks LCS 4.0.100 Table 11-2 REGION II Category 11-2 Fuel (SFP Peripheral Storage) initial Minimum Burnup GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling

.5.00 36.95 33.68 31 .89 30.81 30.10, 4.50 ,26.91 32.29 29.44 27.87. 26.2.8 4'.00 27.44 25. 04 23. .70o 22.88 22.35 3.00. 16.95 15 .62 14.83. 14.34 14 . 03 2.00 4 .93 4 .67 4 .52 4'.42 4.35 1.87 3 . 04' 2 .87 2 .76 2. 69. ,2,. 64

.1.70 0.00 0.00 0.00 0.00 0.00'

ýp P 0

0, o o 0 W W A 0 A 0 L L

0, 0 0 ~0 POOL WALL SAN ONOFRE-UNIT 2 4.0-100-14 Rev. 2,XX/XX/XX:

Fuel Storage Patterns f or Region II Ilocks LCS 4.0.100 Table 11-3 REGION II Checkerboard Storage Category 11-3 Fuel (Checkerboard Partner For Category 11-4 Fuel)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 41.18 37.27 35.18 33.93' 33 .12 4.50 36.34 32.87 31.01 29.88 29.15 4.00 31 .2.9 28.31 26.69 25.70 I25. 0 6 3 .00 20.3-2 ,18.50 ý17 47 16. 84i 2.00 7..81 7.25 1116.9i 6.71 1661.

o. 42 58, 1.87 5, 90 5. 53' 5.30 5 . 17 5.09, 1.56 0.*0 0.00 0.00 0..00 .0'.00 Category 11-4 Fuel (Checkerboard Partner'For Category 11-3 Fuel)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years ý5 Years 10 Years '15 Years '20 Yearsý Cooliing Cooling Cooling Cooling Co. 01 p

45. O0 75 .42! 61. 90, '56.85; 54.18 52. 6"0 4.50 68. 08 49 .25

56 . 12 51. 65 47.76 14.00 60 .74 .46.44 44. 119 42 78 13:8.80:

,0 8. 356

3. O0 46.06 35.41 :33 . 52 .32.31 2.00 31.38, '25.71 23.12 21 . 65 20 .71 1.87 29.19ý 23 . 83 21.34 19.9fi 19. 08 0700 10.94 0.00, 0.00 0. 00 0.00 Category CategQoy 11T- 4 ii-ý Category Category 11-3 11-4 SAN ONOFRE-UNIT 2 4.0-100-15 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Regi Arq T T fl~VzpI LCS 4.0.100 Table 11-4 REGION II Checkerboard Storage Category 11-5 Fuel (Checkerboard Partner For Category 11-6 Fuel)

Initial Minimum' Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling

5. 00' 47.50 40.03 38.53 37.55, 4.50 42.40 37.95, 35.64 34.26 33.37' 4.00 37.10 33. 16 31.10 29.86, 29 .06 3.00 25.64 22.89, 21.40 20.52 19 .95'

-2 9o 121.291 11. 10 10.4.2 10.01I 9.75.

1.87 10. 24 9 .35' 8.80 8.46, 8.25 1.38 0. 00 0 - 00 0.00 0.00 0.00 Category 11-6 Fuel (Checkerb6ard'Partner For Category 11-5 Fuel)

Initial Minimum Burnup (GWD/MTU);

Enrichment 0 Years 5 Year-s 10 Years 15 Years 2.0 Years Cooling Cooli-ng Cooling Cooling Cooling i5. 00 62.37. '53 . 95 50.33 *48.25 46.91, 56.21 488 90' 45.51 !43 .56 42.3-1 4.500 A.00 50.04' 4 3. 67, 40.54 13-7,. 57, 32.56, 29.97 :28. 52, ,27.58ý

.2. O0b 23.30 19._8.0, 18.13 17 .14' 16.50 1.87 21.11 18. 02 16.42 15.48 ,14.888 1.08 0.000 0...0 0.00 0.00 0.00 Category Category 111-6 11-5' Category Category I1-5 11-6 SAN ONOFRE-UNIT 2 4.0-100-16 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Regi LCS 4.0.100 Table 11-5 REGION II Checkerboard Storage Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 4.'80 0.00: 0.00 , 0.00 0".0o 0.00 4.80 Empty Fresh (Blocked),

EmptV 4.80 (Blo-cke-d') Fresh SAN ONOFRE-UNIT 2 4.0-100-17 Rev. 2, XýXAVXXI'

Fuel Storage Patterns for Region.. Ra.ks LCS 4.0.100 Table 11-6 REGION II Category 11-7 Fuel (3 Out Of 4 Pattern)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 34.20 31.35 29.74 28.76 28.12 4.50 29.67, 27.21 25.82 24*.97 24 .41 4.00 24.94' 22.92 21.17 5ý 21.05 20.59, 3.00 14.79, 13.76 13. 13 12 .73, 12 .47 2.00 3.16 3.00 2.90 2.83 2.79ý 1.87 1 .21 1.14 1.09 1.06 1.04 1.80 0.00 0.00 0.00 0.00 0.00 Category Empty 11-7* (BIocked)ý Category Category 11-7 I1-7 SAN ONOFRE-UNIT 2 4.0-100-18 Rev. Z,,XX/XX/XX

Fuel Storage Patterns for Re<gSnn TTI PA1I L,.

LCS 4.0.100 Table 11-7 REGION II Category 11-8 Fuel (Fuel With 5 GT-Inserts)

(Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o)Y 0 Years 5C Years 10 Years :15 Years 20 Years Cooling Coc)ling Cooling Cooling Cooling 5.00 37.68 34.53 32.72 31.61 30.89 4.50 32.61 29.90 28. 33 27.36 26.72 4.I00 27'.33, 25.10 23.78 22.97 22 . 43.

3 .00 15.86 14.76 14. 06 13.61 13 .32 2.00 2.04 1.97 1.89 1.84 1.81

1. 90 0.00 0.00 0.00 0.00 0.00 Category Category 11-8 at11-g8 Category 'Cate-gory 11-8 1-Note: Category 11-1, 11-8, 11-9, and 11-15_may be stored together with nb restrictions.

SAN ONOFRE-UNIT 2 4.0-100-19 Rev. 2,XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table 11-8 REGION II Category 11-9 Fuel (Fuel With 3 GT-Inserts)

(Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o). 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 44. 16 39.95 37.68 36.31 35.42 4.50 38.99 35.25 33.22 31.99 31.18 4.00 30.38 28.60Q ý7.52, 26.81 3.00 ý21. 92' 19.86 18.72, 18.01 17.56 2.00 8.28 7.72 7.34 7.11 6.96 1.87 6.18 5 . 83 5.-58 5.43 5.34 1.59 - 0 .00 0.00 0.00 0.00 '0.00 Category Category 11-9 11-9 Category Category I1=9 11-9 Note: catgory 11-,ii-8, 11-9, and I1715 may be stored together with no restriction-s.

SAN ONOFRE-UNIT 2 4.0-100-20 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Raes LCS 4.0.100 Table 11-9 REGION II Category II-10 Fuel (Filler Assembly With 5 GT-Inserts)

Initial Minimum Burnup (GWD/IMTU)

(w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cc oling Cooling Cooling 5.00 80 . 0.91 65 . 66 60 .12 57 .45 55 .68 4.50 72.13' 59.43 54 .55 52.02 50.33 4.00 64.18 53.19 483.98 46.58 44 . 98

3. 0o 488.27 40.72 :37 .

.1 6, 35 . 03 33.75,

2. 00 32.35 26.59 233.79 22.25 '211.-25 1.03. 0.00 0.00 '01 .00 0.00 0.00 Category Category Category II-10 II-10. II-10

( -Inserts) (GT-Inserts) (GT-Inserts) ategory 4 .80 Category Ii-10 II'10

(GT-Inserts)- Fresh (GT-Inserts)I Category Category Category GT-I-1nserts) T (GT-Inserts)i SAN ONOFRE-UNIT 2 4.0-100-21 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region Raeks n1 LCS 4.0.100 Table 11-10 REGION II Category II-ll Fuel (Filler Assembly With 5 GT-Inserts)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 47.04 42 .52 40.05 38.57 37.60 4.50 41.62 37.58 35.36 34.02 33 . 14' 35.97 30.50 29.32 28.54; 4.00

3 . 003 23.70 21.142ý 20.09 19.31 18.79Y 2.00 9.17 8.54- 8.09 7.81 7 f.62 1.59 0.00 0.00 0.00 0.00 0. o00 Caltegory Category Category ii-li iIT-ll il-~li (GT-Inserts) * (GT-Inserts), (GT-Inserts)

Cate'gogry 4.80. atg~ory II-ll Fresh II-ll (GT-Inserts)_ Inserts (GT-Inserts)i II-ll IIn-llt iI-5 (GT-Inserts) Q(T-Inserts) (GT-Inserts)--

SAN ONOFRE-UNIT 2 4.0-100-22 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table II-1i REGION II Category 11-12 Fuel (Filler Assembly With 3 GT-Inserts)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years io Years 15 Years 20 Years-Cooling Cooling Cooling Cooling Cooling

. 00 54.33 48.48 45.46 43.67 42.51, 4.50 48.181 43 .1ý 40.67, 39.02 37.9 5' 4.00 43.07 38.24 35.72 34.22 33 .26 3 . 00 30.65 27. 11 25.18 24.05 23.33

,2. 00 16.,01 ,14,.,2 3i 13 22: 12 . 62 121. 22!

1. 87; 130.082 12.35: 11.47 10.94 10. 60 1.32. 0.00 0.00, 0.00 0.00 0.00O Category Category Category 11-12 .I-12' 11-12 (3 inse'rts) (3 Inserts) (3 Inserts)I Category 8 80 Category 11-12: 'Fresh II-12Z (3 Inserts) 5 Tnserts (3 Inserts)'

Category Category Category 11-12' 11-12 I!-12'

.(3 Inserts) ý(3 Inserts) (3 Inserts):

SAN ONOFRE-UNIT 2 4.0-100-23 Rev. 2, XX/X,,jXX

Fuel Storage Patterns for Region II Ra"ks LCS 4.0.100 Table 11-12 REGION II Category 11-13 Fuel (Filler Assembly With No GT-Inserts)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 64 . 24 51,. 59 49.41 48. 03ý

57. 99ý '50.23 46 :V3, 44.67 43.38, 4.00 51.75 44.94, 41.71 39.79 38.59 3.00 39.25 33.75 31.051 29.48 28.50

'2.00 24.76 20.95 19. 07, 18.01 17 .33,

ý1.87 22.64 19.10 17.381 16.37 15.72 1;.05 0.00 0.00 0.00 0.00 0.00 Category Category Category 11-13 11-13 11-13 4.80 Category Category 11-13: Fresh 11-13 5 Inserts Category 'Category Category 11I-13 11-31I-13 SAN ONOFRE-UNIT 2 4.0-100-24 Rev. 2 XX/XX/X1Xý

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Table 11-13 REGION II Category 11-14 Fuel (4.80 w/o Assembly Depleted to 18.0 GWD/MTU)

Initial Minimum Burnup (GWD/MTU)

'Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5 . oo0 19.59, 18.61 17.96 17.54 17.27,

4.50 15.93 15,.17 14.68 14.36 14 . 15 4.00 12.18 11.64 11.29 11. 07 10.93 3 .00 4.28 4.12 4.05 4.00 3.98 2.51i 0. 00

'0.00 0-.,00 '0.00 0.00 Category 11-13 FueI (Filler Assembly For Category 11-14 Fuel) initial, Minimum, Burnup (GWD/MTU)I, Enrichment 0 Years 5 Years 10 Years! 15 Years: 20 Years Cooling Cooling Cooling, Cooling Cooling

.64.24 55.151; 51.59. 49.41 48.o03 14.5 0o 57.99 150 . 23, 46 .73' 44. 67 43.38

}39.79,

4. o0 51.75 '44.94! '41.71. 38.59

'3'. Ooo 39.25 J33<.75 31'. 05' ,29.48 28 .501 2.00, 24.76' 20.9 5 19 .07, ,18.01 177.33

1..87 22 . 64 19. 10, 17.38 16.37 15 .72 1,.05, 0.00 0.00 0.00 0.00 0.00 Category Category Category ii-13ý 11-13 1i-131 Category Category Category 11-13 Cto Ctg11-13 Category Category Category 11-13 *2iI13 I-3 SAN ONOFRE-UNIT 2 4.0-100-25 Rev. 2 H/U/Xý,

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table 11-14 REGION II Category 11-14 Fuel (4.80 w/o Assembly Depleted to 18.0 GWD/MTU)

Initial Minimum *Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5 .00 19.59 18.61 17.96 17. 54! 17.27

4. 50d 15.93 15. 171 14.68& 14.3-6ý '14.15' 4.0d 12.18 11.64 11.29 11.07' 10.93 3.00 4.28 4.12 4.05 4.00 3.98 2.51, 0.00 0.00 0.00 0.00 0.00 Category II-11 Fuel!

(Filler Assembly With 5 Guide Tube Inserts)'

Initial, Minimum Burnup (GWD/MTU)

Enrichment (w/o0) 0 Years ý5 Years 10 Years 15 Years' 20 Years Cooling Cooling Cooling Cooling Cooling

5. 0 47.04 42.52 40 .05 ý38.571 37.60 A41 . 62', 37.58 35..36 :3 4. 021 33.14

,4.00, 35.97 32.46 '30.50 29.32 28.54 3.00. 23.70, 21.42, 20.09 19.31 18.79 2.001 9.17 8.54 8.09 77.81 .7.62

1. 59. 0.00 0. O0 0.00 0.00 0.00 Category Category Category

(,5 Inserts)Y (5' Inserts)ý (5 InsertsYý Category Category Cate gory II-11 lIhi (5Inserts) 1114, (5 Inserts)

Category Category Category (5 Inserts)  :(5 Inserts) (5 Inserts)

SAN ONOFRE-UNIT 2 4.0-100-26 Rev. 2 H/H/Xý,

Fuel Storage Patterns for R-egion 11 Reks LCS 4.0.100 Table 11-15 REGION II Category 11-15 Fuel (Fuel With 5 Finger Full Length'CEA)

(Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling o

29. 24 27.24 26.00 25.22 24.70 4.50 24 44 22.84 21.81 21 .1 20.75 4.30 19.41, 18.216 17. 49 17.00 16.68 3.00 8.,83 8.47 8.19 8. 02 7. 90 2.30 0.00 0.00 0.00 0.00 0.00 Category Category 11-15 11-15 Category Category 11-15 11-15 Note:: Category 11-1, 11-8, 11-9, and 11-15 may be stored together with no restrictions.,

SAN ONOFRE-UNIT 2 4.0-100-27 Rev. 2 H/ýXI/XIX,

Fuel Storage Patterns for Region TT 'Ranks LCS 4.0.100 Figure I-I REGION I MINIMUM BURNUP FOR CATEGORY I-i FUEL 25 WAcceptable Region 0 15 E

Cn

(*o

< Unacceptable Region U- 5 ... ......

2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

-- 0 Years a 5 Years 1-v 10 Years eý 15 Years m 20 Years SAN ONOFRE-UNIT 2 4.0-100-28 Rev. 12XX/XX/XX

Fuel Storage Patterns for Region.. Racks LCS 4.0.100 Figure 1-2 REGION I MINIMUM BURNUP FOR CATEGORY 1-2 FUEL 12xj I SAN ONOFRE-UNIT 2 4.0-100-29 Rev. 2 ',X-XjXX/ýX

Fuel Storage Patterns for Regian . Raeks LCS 4.0.100 Figure 1-3 REGION I MINIMUM BURNUP FOR CATEGORY 1-3 FUEL 40 35 30.

25-20 E 15 U)

U) 10 5

0-2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

E9 0 Years --9 5 Years -v- 10 Years -*- 15 Years -,-- 20 Years SAN ONOFRE-UNIT 2 4.0-100-30 Rev. 2 XX-/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 1-4 REGION I MINIMUM BURNTUP FOR CATEGORY 1-4 FUEL 30______ _____

25 (D 20 1 Aceptable Region m 1 E

WU)' 10_ _

- Unacceptable Region 0 - - ------- - - ----- ----

0r 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

- 0 Years -e- 5 Years -- v-- 10 Years 2-c 15 Years 3R 20 Years SAN ONOFRE-UNIT 2 4.0-100-31 Rev. ,2 XX/XX/XX

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Figure 1-5 REGION I MINIMUM BURNUP FOR CATEGORY 1-5 FUEL 35

30. ]

C1

20 - _ _ _ _ _ _ _ _ _ _

a)

< 10

.: Unacceptable L 5. Region 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o) 0 Years - 5 Years -w- 10 Years --A,- 15 Years x 20 Years SAN ONOFRE-UNIT 2 4.0-100-32 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 1-6 REGION I MINIMUM BURNUP FOR CATEGORY 1-6 FUEL 20 ]

15 Acceptable Region -

F:

o 10 I

n-1 E

,5 ... Unacceptable Region..

=3 0

2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

- 0Years O-- -- E 5Years - l0Years - 15Years - 20 Years SAN ONOFRE-UNIT 2 4.0-100-33 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Reoks LCS 4.0.100 Figure 1-7 REGION I Boundary Between Unrestricted Storage And Checkerboard Storage I-i i-i I-i I-i I-ii I-l:

I-i I-i I-i I-i, I-!; I-i I-l EI-i I-i I-i I-i

!Empty 4.80 Empty I'1: I-I i-i 8 Empty I-i i I-i I-1 Interface i-I7 f-i I- i-i ii I i~r

-lI

!I-i I-il I-l, I-l, I-* Ii:

I- il I- i I -i i -i T-4 I_6 I -4 I-i I-i i-l)

I-6 ,1-4 I-1 I-i I--1 I -11 U- .L I 11 II Interface Note: (1) A row of empty cells can be used at the interface to separate the configurations.

(2) :it is acceptable to replace an as sembly with an empty_ cell.

SAN ONOFRE-UNIT 2 4.0-100-34 Rev. XX/,XX/XX

Fuel Storage Patterns for Region II Raks LCS 4.0.100 Figure 1-8 REGION I Boundary Between Unrestricted Storage And 1 Out Of 4 Storage 11-3 :1:

1-1 1-.-

I-1 4.80 i-3 1- 1 -I-

I-3 1-3 :i1-3 I-1 I -1 1-17 Interface Note:i (1)i A row of empty cells can be .used at the interface to separate the configurations,.

(2), It is acceptable to replace an assembly with an empty cell.

SAN ONOFRE-UNIT 2 4.0-100-35 Rev. 12 XX/,XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 1-9 REGION I Boundary Between Checkerboard Storage And 1 Out Of 4 Storage 4.800 Empty 4.80 Empty 4.80 Empty Empty 4.80' Empty 4.80 Empty 4.80

-4.80 Em 48

4. 0 Empty 4.80 Empty Epty Empty I-3 )Empty 4.80 Empty 4.80 1-3; A4.80! 1-3 Empty 4. 80b -Empty 3:

.- I-3* Empty 4.80 Empty .80 In-ter-face 1" 6 1-4, 1-6' I-4 1-6 1-4' 1-4 1-6 1-4 1-6 1-4 '1-6 1-6 1-4; -6 1-4 1-6 1-4 T-3 J -3: .-3 1- 4 6

-3J4.8jj14 -6 Y' I~6 1- 14 '16 II- 3: '-31 1- 3; 1- 6 1-4 *I - 6; I

_11 Interface Note: (1) A row of empty cells can be'used at the interface to separate the configurations.

(2) It iýs acceptable to-replac-e an assembly with an empty cell.

SAN ONOFRE-UNIT 2 4.0-100-36 Rev. 2 XX/XX/XX'i

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Figure II-i REGION II MINIMUM BURNUP'FOR CATEGORY Ii-i FUEL 60

_,50 (40 a-1 E

m3O E

a)

(020 U-10 0.-4 I 2 3 4 5 Initial U-235 Enrichment (w/o)

E3 OYears -e 5 Years - 10Years v, 15Years F- 20Years SAN ONOFRE-UNIT 2 4.0-100-37 Rev. 2 H/H/Xý,

Fuel Storage Patterns for Region . .. a.ks LCS 4.0.100 Figure 11-2 REGION II MINIMUM BURNUP FOR CATEGORY 11-2 FUEL 40 o 30-0 D.

n20 E

-30~

U-0-

1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o) 0 Years 5 Years - 10 Years IF 15 Years imE 20 Years SAN ONOFRE-UNIT 2 4.0-100-38 Rev. 2 XX/XX/XXI

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 11-3 REGION II MINIMUM BURNUP FOR CATEGORY II-3 FUEL 50 0'30 E

E 20

,=)

0-1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

ER 0 Years -e--5 Years -->-- 10Years -- 15Years --- 20Years SAN ONOFRE-UNIT 2 4.0-100-39 Rev. 2 XX/n/u

Fuel Storage Patterns far,,e- ^^RegioI5 Ra4ks LCiS 4.0.100 Figure 11-4 REGION II MINIMUM BURNUP FOR CATEGORY 11-4 FUEL 80.

060 co40 ZI, 020 IL 1 2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years O- e 5 Years - 10 Years -v- 15 Years -ýE 20 Years SAN ONOFRE-UNIT 2 4.0-100-40 Rev.

Fuel Storage Patterns for,,egi an 4IRa.ks LCS 4.0.100 Figure 11-5 REGION II MINIMUM BURNUP FOR CATEGORY II-5 FUEL 50 E40-0 93o (D

= 30 P

m

.a E 20-a)

U-m10-0 1 2 3 4 5 Initial U-235 Enrichment (w/o)

-s- OYears e 5Years l0Years 0 IF 15Years -p- 20 Years SAN ONOFRE-UNIT 2 4.0-100-41 Rev. 2 :,xx

--Y"J"

Fuel Storage Patterns fora Region ..RaIks LCS 4.0.100 Figure 11-6 REGION Ii MINIMUM BURNUP FOR CATEGORY 11-6 FUEL 70 60 50

• 40

-30 E

co

< 20 10 0

1 2 3 4 5 Initial U-235 Enrichment (w/o)

Fri0 Years e 5 Years - 10Years 1-- 53 15 Years iR 20 Years SAN ONOFRE-UNIT 2 4.0-100-42 Rev. 2 XX/Xxl I Ax'II

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Figure 11-7 REGION II MINIMUM BURNUP FOR CATEGORY 11-7 FUEL 35 30 0

.25

• 20 co "15 E

<10 LL 5

0 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

-i 0 Years E- 5 Years -> 10 Years  ;ý, 15 Years -- R 20 Years SAN ONOFRE-UNIT 2 4.0-100-43 Rev. 2 XX/XX/XXI

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Figure 11-8 REGION II MINIMUM BURNUP FOR CATEGORY 11-8 FUEL 40 CL

=30 20 m20 E

(n (01

-LL1 0-1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

-iR 0 Years -E- 5 Years --- 10 Years -'-i 15 Years i*i 20 Years SAN ONOFRE-UNIT 2 4.0-100-44 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Re"ks LCS 4.0.100 Figure 11-9 REGION II MINIMUM BURNUP FOR CATEGORY 11-9 FUEL 50

-'40 (9

E 20-Ci)

U, U,

LL 10-1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E 0 Years --e- 5 Years ->- 10 Years -ý1 15 Years 11 20 Years SAN ONOFRE-UNIT 2 4.0-100-45 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure II-10 REGION II MINIMUM BURNUP FOR CATEGORY II-10 FUEL 100 E9 0.

E

-0 E

0i a)

LL 1 2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years - 5 Years - 1OYears l- 1;F 15 Years -i 20 Years SAN ONOFRE-UNIT 2 4.0-100-46 Rev. 2"XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure II-l1 REGION II MINIMUM BURNUP FOR CATEGORY II-1l FUEL 50 40-0 m3 E 20 (D

U 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E-i 0 Years - -- 5 Years *-- 10 Years IF 15 Years --- 20 Years SAN ONOFRE-UNIT 2 4.0-100-47 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Raks LCS 4.0.100 Figure 11-12 REGION II MINIMUM BURNUP FOR CATEGORY 11-12 FUEL 60 50 (40

=1_

E m 30 E

a) 020 -

10 0-2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years e 5 Years -E 10 Years Fvz 15 Years --- 20 Years SAN ONOFRE-UNIT 2 4.0-100-48 Rev. 2 H/H/Hi,

Fuel Storage Patterns for Region IIiRaks LCS 4.0.100 Figure 11-13 REGION II MINIMUM BURNUP FOR CATEGORY 11-13 FUEL 70 60 50 5o E 40

-5 30 E

4)

(I,

<20-LL 10 0-2 3 4 5 Initial U-235 Enrichment (w/o)

E3 0 Years e 5 Years --- 10Years F 15Years EE 20Years SAN ONOFRE-UNIT 2 4.0-100-49 Rev. 2 H/XX/Hý

Fuel Storage Patterns for Region II RaIks LCS 4.0.100 Figure 11-14 REGION II MINIMUM BURNUP FOR CATEGORY 11-14 FUEL 20 015-10 E

ai) cn LL 0

2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E 0 Years E35 Years -)- 10Years -ý 15Years -E-- 20Years SAN ONOFRE-UNIT 2 4.0-100-50 Rev. 2_XX/XX/XX

Fuel Storage Patterns for Region II Raeks LCS 4.0.100 Figure 11-15 REGION II MINIMUM BURNUP FOR CATEGORY 11-15 FUEL 2.2 2.7 3.2 3.7 4.2 4.7 Initial U-235 Enrichment (w/o)

E3 OYears e 5Years - lOYears -, 15Years ER 20QYears SAN ONOFRE-UNIT 2 4.0-100-51 Rev. 2'XX/XX/XX,

Fuel Storage Patterns for Region 11 Racks LCS 4.0.100 Figure 11-16 REGION II Boundary Between Unrestricted Storage And Checkerboard Storage II-i Ii-i II-i li-i II-I II-i II-1 li-i IT-i II-i II-i TI-i IT-i iI-i II-il TI-i II-i IT-il II-i Ji~locked Ii-i' i-ii Tii-i I1-i Blocked 4.801 Blocke IT-i fi-i II-i 4.80 B-l*0ke] IT-i -iý -i- I-i

. II. II Interface 11-4 if-i i-- 11-1 11i- II-i II-i JIT-' II-i .II-i fl-i II-l 11-4 J 1'4 4il* 11i-1Z 11-1 1I-1 iI- :f73 *ii44'I-i1i' II-i 11-3 I1 -4 TII-1 ITIL ý!Iz 1 II-i IInterfac Interface Note: (1) A row of empty cells can be used at the interface to separate the configurations.

(2) It is acceptable to replace an assembly with an empty cell.

SAN ONOFRE-UNIT 2 4.0-100-52 Rev. 2 XX/XX/Xx

Fuel Storage Patterns for eq-i [m II Racks LCS 4.0.100 Figure 11-17 REGION II Boundary Between Unrestricted Storage And 3 Out Of 4 Storage iI-i II-I Ii-i TI-i II-i II-i ti-i II-1 **I 1 II-i~ TI-i II-i Blocked 1 iT~i Blocked ilgi iT-i TIii 11-27 TIl- iI-l1 Il-i! ITil! .- i Bl1ocked 11-7, IBlocfked :Ii*il I I.-i Interface Not Ie:

I(il. A'row of empty cells can be used at the interface to separate the configurations.

(2) It is acceptable to replace an assembly with an empty cell.

SAN ONOFRE-UNIT 2 4.0-100-53 Rev. 2 XX/XX/XK

Fuel Storage Patterns for Region II Ra'ks LCS 4.0.100 Figure 11-18 REGION II Boundary Between Checkerboard Storage And 3 Out Of 4 Storage 11-7. Blocked 11-7 Blocked 11-7 Blocked 11-7 11-7 11-7. 11-7 11-7 11-7 11-7 Blocked !I-7ý Blocked li-7 Blocked B4-0oked J 1-7k Blocke] Iice 11-7 11-7 4A.80 JBlocked~ 1HJ - Blo .cked I'I-7 Bloc-ked

,II-7 Blocked 4.80. IBockedce 11-7 11-7, II II Interface Note. (1) A row* 6empty cells can be used at the interface to separate the configurations (2) ýIt is acceptable to replace an assembly with an empty cell.

SAN ONOFRE-UNIT 2 4.0-100-54 Rev. 2 XX/ý,X,/ýX

Fuel Storage Patterns for Regio-LCS 4.0.100 Figure 11-19 REGION II Boundary Between l-Out-Of-9 Storage And Unrestricted Storage iI-i Il-i II-i Il-i Il-i II-i TI-i1 II-1 II-i II-i II-i- II-1 Filer' F'il er Fi-ler I-i II -i Ii-i

'Filler A Filler- 1l-i i-i 11-i Filler Filier Filler II 1I-i I-I II Interface Where: Tf~ A 4.80 w/o + 0 Inserts, Filler Category II-10 (2) If .4.80 W/o + 5 Inserts, Filler Category (3) If A 4.80 w/o + 5 Inserts, Filler Category (4) If A 4. 8.0 w/ o + 5 Inserts, Filler Category 11-13 (5) If A Category 11-14, Filler Category 11-13 A

(6) if Category 11-14, Filler Category SAN ONOFRE-UNIT 2 4.0-100-55 Rev. 2- XX/XX/X-X

Fuel Storanp Pattprnq LCS 4.0.100 Figure 11-20 REGION II Boundary Between 1-Out-Of-9 Storage And Checkerboard Storage 11-4 11-3 11-4 11-3 11-4 11-3 11-3 11-4 11-3 Ii-4 11-3 11-4 11-41 11-3 11-4 +/-11-3 11-4 11-3 Filler Filler Fille r z 11-4 11-3 11-4 Fi4l!,e r Al Fili ie'.r =1-3 11-4 11 -3ý Filler Filler' Filler II-4 11-3 11-4 Interface Where: (2) If 4.80 w/o + 0 Inserts; Filler Category (2) If A 4.80 w/o + 5 Inserts, Filler Category (3) ,I f A 4.80 w/o + 5 Inserts, Filler Category 11-'12 A

(-4). ,I f A 4.C80 w/o + 5 Inserts, Filler Category 11-.13

'If¸ Category II-14,, Filler Category 11-.13 (5) A -Catego-91ry II -14,1; Filler Category IIl-li SAN ONOFRE-UNIT 2 4.0-100-56 Rev. 2" H/U/Xý,

Fuel Storanp PAttprns LCS 4.0.100 Figure 11-21 REGION II Boundary Between l-Out-Of-9 Storage And Checkerboard Storage Blocked 4.80 Blocked 4.80 Blocked 4.80 4.80 Blocked 4.80 Blocked 4.80 Blocked Blocked 4.80 Blocked 4.80 Blocked 4.8Q FFiler Blocked Filler Biocked 4.80 Blocked Filler A Blocked 4. 80 Blocked 4.80 F,il.1er, B 10ocked4.80 Blocked4,8 iFiller Filler Filler ,Blocked 4.80 .Blocked II.

• Interface Where: If A -4.80 w/o + 0 Insertq, Filler Category (2) I f A 4.80 w/o + 5 Inserts, Filler Category 11-12

.If A 4.80 w/o

• 5 inserts, Filler Cat egory 11-13 If A 4.80 w/o

• 5 Inserts, Filler Category 11-13 i-li3 (5)t If A Category Filler Cat-egory I I-13!

S6-), If A. Category Filler Calteg~ory SAN ONOFRE-UNIT 2 4.0-100-57 Rev. 2,XX/XX/XX

Fuel Storanp Pattprnq LCS 4.0.100 Figure 11-22 REGION II Boundary Between l-Out-Of-9 Storage And 3-Out-Of-A Storage 11-7 Blocked 11-7 Blocked 11-7 Blocked 11-7 11-7 11-7 11-7 11-7 11-7 11-7 Blocked 117: Blocked 11-7, Blocked Filler Filler Ffie r I1 11-7, I-I7 I-7 Filler A 'Filler Blocked 11-7 Blocked Filler Filler Filler 11I-7; 11-7, iI1-7

. .... l .

Interface Where-:! (I) If A 4.80 w/o + 0 Inserts, ,Filler Category 11-10 (2)- If A 4.80 w/o + 5 Inserts, Filler Category il-il

_(3) If A 4.80 w/~o + 5 Inserts, Filler Category 11 -1,2

(4) If A 4.80 w/o + 5 Inserts, Category 11-13 Filler i(5) If A Category 11-14 ,! Filler Category 11-13 Category If A 'Category 'Filler 11-11 SAN ONOFRE-UNIT 2 4.0-100-58 Rev. 2 XX/XX/XX

Fuel Storage Patterrqs for Region 11 Racks LCS 4.0.100 Figure 11-23 Orientation Of 3 Guide Tube Inserts Fuel Assembly yyy <--- Serial y Number y

00 xx 00 xx XX XX Xx 00 xx 00 xx Guide Tube With Insert xx 00 = Empt*y Guide Tube 00 SAN ONOFRE-UNIT 2 4.0-100-59 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region 1I Racks LCS 4.0.100 Figure 11-24 SONGS UNITS 2 AND 3 FUEL ASSEMBLY WITH 40 ERBIA RODS 40 Erbia Rods!

D High Enriched Fuel Rod Low Enric~hed Fuel Rod I Erbia Fuel Rod SAN ONOFRE-UNIT 2 4.0- 100-60 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II fo"ks LCS 4.0.100 Figure 11-25 SONGS UNITS 2 AND 3 FUEL ASSEMBLY WITH 80 ERBIA RODS if . . . . . . , I I I I I * --

80 Erbia Rods D High Enriched'Fuel Rod Low Enriched Fuel Rod U Erbi a Fuel Rod SAN ONOFRE-UNIT 2 4.0-100-61 Rev. 2,-XX/XX/XX,

Fuel Storaae Patterns.

LCS 4.0.100 B LCS 4.0.100 Fuel Storage Patterns BASES BACKGROUND The spent fuel storaqe facility is desiqned to store either new (nonirradiated) nuclear fuel assemblies, orburned (irradiated) fuel assemblies in a vertical confiquration, underwater. The storaqe pool is sized to store 1542 fuel!

assemblies. Two types/sizes of spent fuel storage racks are used (Reqion I and Reqion II). The two Reqion I racks each"'

contain 156 storaqe locations each spaced 10.40 inches on' center in a 12x13 array. Four Reqion II storaqe racks each, contain *,210 storaqe locations in a 14x15 array'. Theý-,-

remaininq two Reqion II racks each contain 195 locations in' a 13x15 array. All Region II locations are spaced 8.85.

inch e s .n, ce nter.

As described in LCS 4.0.100, "Fuel Storage Patterns," fuell assembli~es are stored in the spent fuel racks in accordance' with criteria based on ilnitial enrichment, discharge burnup,[I and cooling time (plutonium decay).'

A'ddi-It'i-ona]l1-ly, LGS"CS 4.0. 100. al1owls` the f(oll owing storageý options: . .

2x2 storaqe patterns, 3x3,ý,storaqe: patterns,I credit for erbia in fresh assemblies, inserted CEAs,!

and borated stainless steel guide tube inserts (GT- Inserts)..

The effect of these storaqe options is to reduce the' required discharqe burnup of irradiated assemblies and add flexibility in storing fresh assembl'es'cntain'ing erbia.

Finally, credit,for boron is taken to maintain Ke`< 0.95.i The storaqe options allowed by LCS 4.0.100 and credit for soluble boron are needed because the boraflex is assumed to be completely absent. Boraflex erosion/dissolution is an industry problem' and SONGS Units 2 and 3 are affected.-.

• Taking no credit for boraflex for SONGS Units 2 and 3 (continued)

SAN ONOFRE-UNIT 2 4.0-100-62 Rev. 2 XX/XX/XX;

Fuel Storaae Patterns LCS 4.0.100 BASLS (continued)

BACKGROUND eliminates any boraflex concerns and monitorinq proqrams to (continued) ensure that an adequate amount of boraflex is present will not be needed.

APPLICABLE' When soluble bo6ro is credited, the following acceptance SAFETY ANALYSIS& criteria apply:,

(1) Under normal conditions, the 95/95 neutron multiplication factor (Kf,), including all:

uncertainties, shall be less than 1.0 when flooded with unborated water, and, (2) Under normal a-nd accident conditions., thee 95/95.

ýneutrron multiplilcation factor, (K,,) ,,/inc~lud'ing all,

,u~ncertainti'esý,'shall be less 'than or equal. to 0.95 when flooded with borated water.,

The storaae ootions allowed bv LCS 4.0.100 meet accentance

,criterion 1 ablove. Under normal. non-accident conditions,;

0 ppm.

Keff iS less than'.1.00 with The storage options allowed by LCS 4.0.100 meet acceptance,

,criterion 2 above. Under normal, non-accident conditions,;

Ithe TS. 3.7.171limit for, soluble boron concentration of 2000 ppm will ensure that, K*fremains less than or equal to 0.95.i The analytical methods used to demonstrate that the storage, options allowed by LCS 4.0.100 meet acceptance criteria 1 and 2 are described in the SONGS Spent Fuel Pool Criticality Analysis (Reference 1)..

A fuel assembly' couldbe inadvertently loaded into a spent,;

fuel rack location not allowed by LCS *4.0, 100 (e.g., an unirradiated fuel assembly or an insuffi'ciently depleted, fuel assembly). However, the negative reactivity effect of

,the soluble boron compensates for the increased reactivity, caused by the postulated accident scenario. For this accident condition, the TS 3.7.17 limit for soluble boron concentration of 2000 ppm will ensure that K'ff remains less than or equal to 0.95.

(continued)

SAN ONOFRE-UNIT 2 4.0-100-6.3 Rev. 2 XX/XX/ ,XXI

Fuel Storaae Patterns LCS 4.0.100 BASFL (continued)

LCS The placement of fuel assemblies within the spent fuel pool according to LCS 4.0.100 ensures that the Kff of the spent fuel pool will always remain < 1.00 under normal, non-accident conditions assuming the pool to be flooded withunborated water. The K~ff of the spent fuel pool will:

always remain

• 0.95 under normal and accident conditions:

assuming the pool to be flooded with borated water at 2,0.0,0, ppm.ri

'APPLICABILITY This LCS applies whenever any fuel assembly is stored in the spent fuel pool.!

ACTIONS See the Bases for LCO 3.7.18 A SURVEILLANCE See the Bases, for SR 3.7.18.1I REFERENCES 1. Spent Fuel . Pool Criticality'Analysis, (with no Boraflex and Credit for Soluble Boron),, Southern' California Edison, San Onofre Nuclear Generating.

~2~Of StationUnits 2 and 3, Revision 2, April 2007.ý (continued)

SAN ONOFRE-UNIT 2 4.0-100-64 Rev. 2 H/H/Hý i

PCN-556 Attachment J (Proposed LCS 4.0.100 Unit 3)

(The date XX/XX/XX on each page of LCS 4.0.100 will be the date of NRC approval of License Amendment Application XXX)

Fuel Storage Patterns for Region II Racks LCS 4.0.100 4.0 DESIGN FEATURES LCS 4.0.100 Fuel Storage Patterns for Region II Rack QAd PQRiir ITT Ralks ReIonstitution Statio NOTE::

This Licensee Controlled Specification is listed by revision number and data in Technical Specification 4.3.1'. All chanqes to paqes 1 through 61,Rev. 2 dated XX/XX/XX of this LCS (i.e., excludinq 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.'

VALIDITY STATEMENT: Rev. +2 effective upon NRC approval: !12-17, to be implemented within 49--180 days.

4.0.100 New or burned fuel (which does not meet the criteria of LCO 3.7.18 for unrestricted storaqe _or storage at the, Pool pe~riphery) may be stored in Reqion I or Region IL if a-l-I-the stored in accordance with the allowable Storage Patterns described in this, LCS.f.llo,,ing conditions are met.

'4.0.100.I1 Region I Reqion I Storaqe Patterns are qiven in Tables I-i through 1-8 and Figures I-1 through 1-9.. .

Fuel Type 1 - NeW or buirmed fuel which does mat meeta the criteria of LGO 3.7.18 for unrestricted storage or storage at-th pool perihe.ry.

4,0. 100.2 Region II Reqion II Storaqe Patterns are qgiven in Tables I1,,i-i through 11-15 and Figures I!-I through 11-22.,

Fuel Type 2 - Puc, '6-*ý4 3w;hich dbos :o^**F-_themeet

....... the criteria iete4 of chcc (B) s a!r .e. be stared in Rei I- in a pattn The four (

HguIremets f Rio 1I IIstorage a.J. U, (1) Type19 fue* l ass elies can.not b" in adjcen eloations. They can, however, bce~ in dl11I ago nal locaeatioans.

(2)Tpel fuel aCssemiblies stared in Refg~ion.

S,,shall have at least two (2) sides fa 1 I. _ t a,

• "1"I I Ul I -I".A . .L**/-

3J L **

-

• 1 IU 1; I. I H I U.-*l ., I U t.,I )I

() IguII ~re*.0~10 1 pIrovies an illustration of anaccetabl fue stoarage pattern.

SAN ONOFRE--UNIT 3 4.0-100-1 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Reks LCS 4.0.100 4.0.100.3 SONGS Unit 1 Fuel shall not be stored in Region I Racks.

4.0.100.4 The hurnun of each SONGS Unit 1 Uranium dioxide snent fuel.

assembly stored in Region II shall meet the following criteria:

(C)reonsituionstation is a speci.al c-ase of a eheckerboard pattern. The aboerle-o 111U L,IIn pI l't-It*J IIII l arIt permitted. Fgure

, .0.100 2 provides ant I I 1U

.9 b l# l.

(D) San Orofre Unit 1 fuel assemblis mybe stored in egonIIin a 'Three. otffurpattcrn iftes~semb!Y-burnuo is at least 1. GYM/T 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 qreater than 25,000 MWD/T, and:

the cooling time is greater than 5 years.:

4.0.100.4.2 SONGS Un-i't-l nominal 4.00 w/O assemblies can be stored in the" Region 11 Racksj(unrestricted) if:

the burnup is qreater than 26,300 MWD/T, and' the cooling time is greater than 20 years.

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

• the burnup is qreater 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 peri phery) i: f the burnuD is greater than 20,000 MWD/T. and' the cooling time is greater than 0 years.ý 4.0.100.5 Design Requirements For Guide Tube Inserts (i) GT-Inserts shall beO.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 cm .

SAN ONOFRE--UNIT 3 4.0-100-2 Rev. 2,,.XX/XX/XX

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 (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 11-23).

(iii) A 5-finqer, full lenqth Control Element Assembly (CEA) may be ,,used in place of borated stainless steel or.

aluminum guide tube inserts.

4.0.100.6 Design requirements For Erbia Assemblies containinq 40 or 80 erbia rods shall have the erbiai rods distributed per Fiqures 11-24 and 11-25. The minimum initial nominal erbia loading shall be 2.O w/o Er203.,

4.0.100.7 The Failed Fuel Rod Storage Basket-(FFRSB)

The Failed Fuel IRod Storaqe Basket (FFRSB) shall be treated aIs 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 bearinq assembly components,(thimble pluqs, CEAs, etc) may be stored in fuel assemblies without!

affectinq the storaqe reQuirements of these assemblies., ._

storaqe basket containinq-no fisisile material can be st6ded, in!

any storaqe location, and can be used as a storage cell blocker for reactivity control.;

4.0.100.9 Fuel Ass'e'mbly Reconstitution Station A fuel assembly reconstitution station is a special case of al chechckerbolard-, paatt%:e~rhn. ,A reconstitution station *Is permi'tted"

anywhere in the Reqion I racks. The empty cells in the..

checkerboard pattern do not need to be blocked. A reconstitution, istationis Permitted anywhere in the Reqion II racks, provided

,that empty cells in the checkerboard pattern are blocked0to make it impossible to misload a fuel assembly during reconstitution' activities..

SAN ONOFRE--UNIT 3 4.0-100-3 Rev. 2, XWX,/ýý

Fuel Storage Patterns for Regian II Racks LCS 4.0.100 x* x

-( X -x 0

x

• 0 __ 0 __ 0

. _ _ 0 _ _ 0

_0X H-TYFPE 2 FUE-6

-- "VTYPE" 1 4.0l.10 I -I . ,II Fi ur 4 . . 0 2.

SAN ONOFRE--UNIT 3 4.0-100-4 Rev. .2 XX/",Xx/xxl

Fuel Storage Patterns for Region 11 Ra'ks LCS 4.0.100 Table I-i REGION I Category I-i Fuel (Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling, Cooling Cooling 5.00 22.84 21 .47 2 0.5 9 20b. 04 ,19.67 4.5o0 18.61 17 .57 16.89 16.45 16.17 4.00 14.30 13.58 13. 09: 12.78 12.57 3.50 9.84 9.40 9.11 8 ; 92 8.179 3.00. 5.24 5.02 4.91 4.84 4.79 2.47 0.00 0.00 0. 00, 0.00 0.00 Category Category I-1to I-1' Category Category I-li. I-l Note: Category I-1 and fresh fuelwi th- full-length 5-finger CEAs (Table 1-5) may be stored together with no restrictions.

SAN ONOFRE--UNIT 3 4.0-100-5 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region 11 Racks LCS 4.0.100 Table 1-2 REGION I Category 1-2 Fuel (SFP Peripheral Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (W/I0): .0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling -Cooling Cooling Cooling 5.500! 12.55 12.15 i1_. 82 11.61 11.47 4.50 9.09, 8.85 8.63 8.49 8.40

'4.001 5.58 5.43 5.33 5. 25 5.21 3.50 2.22 2.13 2.09 2.05; 2.03 3.20' 0. oo0 0.00: ý0.00 0.00 0.00 P P 0 0 Q

L L w

A 0 A L 0 L L 0 L!

0 o 0 POOL WALL SAN ONOFRE--UNIT 3 4.0-100-6 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region ,.

Raeks LCS 4.0.100 Table 1-3 REGION I Category 1-3 Fuel (Filler Assembly For l-out-of-4 Pattern)

Initial Minimum Burnup (GWD/T)

Enrichment (w/o) 0 Years 5 Years ,10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 i39.: 99 36.28 34.27 33 . 04: ý32. 22 -

4. 50 34 . 95 31. 71 29.94 28.84 28.12 4.00. 29.71 26.99 25.46, 24.51. ,23.89 3.50 24 .22 22 . 03z 20.7 9ý 20.02 19.52 3 00 18.37 16. 84 15.91 15.34 14.97 2.50 12 .21 11.30 10.72 10.37 ý10.13 2:.00 5.28 5 . 05 S4.85. .4.72 4.62 0.00 0. 00 0.00

. 0.00 !0.00 4.80 Category Fresh 1-3, Category Category 1-3 1-3 SAN ONOFRE--UNIT 3 4.0-100-7 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II RaIks LCS 4.0.100 Table 1-4 REGION I 4.80 w/o Fresh Fuel Checkerboard Pattern Initial Minimum Burnup (GWD/T)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years' Cooling Cooling Cooling Cooling Cooling 4.80 0.06b 0.00 0.00 0.00 0.00 4.80 Empty Fresh Empty 4. 80ý Fresh SAN ONOFRE--UNIT 3 4.0-100-8 Rev. 2:XX/XX/XX

Fuel Storage Patterns for Region 11 Racks LCS 4.0.100 Table 1-5 REGION I 4.8 w/o Fresh Fuel With Full-Length, 5-Finger CEA (Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 years 15 Years 20 Years, Cooling Cooling Cooling Cooling Cooling 4.80, 0.00 0.00 0.00. 0.00' 0.00 4.80, 4.80 Fresh, Fresh Wi thý C EA With CEA 4.80 4.80, Fresh Fresh With CEA With CEA Note: Category I-1 and fresh fuel with full-length 5-finger CEAs (Table 1-5) may be stored together with no restrictions.

SAN ONOFRE--UNIT 3 4.0-100-9 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Regioam 1 Raeks LCS 4.0.100 Table 1-6 REGION I Category 1-4 Fuel (Filler Assembly For l-out-of-4 Pattern)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0.Years 5 Years' 10 Years 15 Years, 2,0, Years Cooling Cooling Cooling Cooling Cooling

5. O0 26.57 24.71 23.59 22.90, 22.44 4.50, 22.12 20.62, 19.73i 19.17, 18'. 80 4.00 17.54 16.46. 15.781 15.35' 15.07 3.50 12.84 12. 12 11.66 11.37 11.18 3.00 7.95 7 . 56' 7.31 7.15 7.05 2.50 2.76 2.64ý 2.56 2.50. 2.46

,,0.0:0

,2. 2 7 .ý0.-O00 0.. 010 0 .OQ 0.00 4.80, Category Fr es'h 1'-4 80 Erbia Category Category 1-41 1-4 SAN ONOFRE--UNIT 3 4.0-100-10 Rev. 2 XX/XX/H

Fuel Storage Patterns for Region Ra*k*

,1 LCS 4.0.100

.Table 1-7 REGION I Category 1-5 Fuel (Filler Assembly For 1-out-of-4 Pattern)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years' 20 Years Cooling Cooling Cooling Cooling Cooling

ý5. 00. 30.81 28.406 27.00 26.14ý 25.57 4.50' 26.17 24.17, 22.99 22.26 21.78 4.00, 21.32 19.77 18.84. 18.27ý 17.88

,3.do 16.32 15.'22 14. 55 14.. 13 13.85 3.0o 11 .11 10.45ý 10.05 9.79 9.61 2.50§ 5.55 5.30: 5.14, 5.04 4.98 2 . 07 0.00 0.00o 0.00 0.00 0.00 4.806 Category Fresh I-5, 40, Erbia Category Category I-5 I-5, SAN ONOFRE--UNIT 3 4.0-100-11 Rev. ,2-XX/X'X/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table 1-8 REGION I Checkerboard Pattern Category 1-6 Fuel (4.80 w/o Assembly Depleted to 18.0 GWD/MTU)

Initial Minimum Burnup (GWD/MTU)

Enrichment

((w/o): 0 Years 5 Years 10 Years 15 Years, 20 Years Cooling Cooling Cooling Cooling(, Cooling

5. 00 19.82 18.84 18.12 17.67 17.37 15 .11i 4.50 15.83 14.58 -14..24 14.01 4 . 00 11. 75 11.28 10.92 10.54 77b4" 3.50 7 .56 7.23, 3.07 6. 91, 6. 83 23.0 06 3.28 3.15; 3 .03: 2.99

. 0.0

.2. 6 5' 0.00 0.00 0.00 *0.00 0.00

,Catelgory 1-4 Fueli

'(Checkeýrbo'ard Partner For ICategory 1-16 -Fu .el)I Initial Minimum Burnup (GWD/MTU)j Enrichment 1w.*o~ o0 Years 5 Years, ,10 Years 15 Years' 20 Years Cooling Coo0ing Coolin C0oolin 'Cooling 5.00 26.57 23.59 ý22.90 .22.44 4.50, 22.12 20. 62 19.73 19 .17: :18.80

4. 00' 17.54 16.466 15 .788 -15. 3 5! 15.07 1284 12 .'-12. 11.66, 11.37 11i. 18 3.00 7.95 7 .561 7.31 7..05 2.76 2.64x 2.56 2.46 22.27 0.00 0.00 0.00 0.O0 0.00 Category Category 1a-4 1C-e6 Category Category 1-6 1-4 SAN ONOFRE--UNIT 3 4.0-100-12 Rev. 2 XX/XX/Xx11,ý

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table II-1 REGION II Category II-i Fuel (Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling, Cooling 5.oo 53.76 47 .77 44.75 43.00 41.86 4.50 48.43 42.93 40.15 38.52 37.47,

42 :.91 4.00, 3,7 . 94 35.40 33.92: 32.96 3 %.00 130 .99 25.130 '24-716 23 .43 2.00 17.05 14. 97" 13.90 13 .25: 12.83 1.87 14 .93 13.23 12.26 11i.68 il. 31 1.23;ý 0 .00. 0.00 0.00 0. O0 0.00 Category Category Cag-i II-ir SCategory Category II-1 II-1I Note: Category I.-i, 11-8, 11-9, and 11-15 may be stored together with no restrictions.,

SAN ONOFRE--UNIT 3 4.0-100-13 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Reg- AVn TT PAPVz LCS 4.0.100 Table 11-2 REGION II Category 11-2 Fuel (SFP Peripheral Storage)

Initial Minimum Burnup GWD/MTU)

Enrichment (w/o), o Years 5 Years 10 Years 15 Years 20 Years, Cooling Cooling Cooling Cooling Cooling 5.00 36.95, 3368 31.189: 301.81, 30.10 4.50 32. 29! 29.44 27. 87 26.91 26.28

.4.00 27.44- 25.04 23.70 22 .-88' 22.35 3.00 16. 95 15.62 14 . 83 14.34 14. 03 2.00 493: .- 4.67 4. 52, 4.42 4.35 1.87 3. 04 2.87 2.76 2. 69 2'.64, 1.70 0.00 0.00 0.00 0.00 0.00 P

' ' 0 L

0 0 C0 ,category Category Category 1---' 11-2 W

0 A

.. .. . .. . ... .. .. . .. ... .. ..<L 0 L L

0 o 0 0 POOL WALL SAN ONOFRE--UNIT 3 4.0-100-14 Rev. 2,,,,.XX/XX/`,ýX

Fuel Storage Patterns for Region 11 Raks LCS 4.0.100 Table 11-3 REGION II Checkerboard Storage Category 11-3 Fuel (Checkerboard Partner For Category 11-4 Fuel)

Initial Minimum Burnup (GWD/MTU)

Enrichment!

(w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 41.18 37 27 35.18 33 . 93 33.12ý 4.50 36.34 32.887 31.01 29. 88 29.15 4.00 31. 29 26.69 25.70 25-.06I 3.00 20.32; 18.50b 17. 47ý 16. 84 16 . 42' 2.00 7.81 7.25 6.910 6. 71 6.58 1.87 5.90, 5.53' 5.30, 5.17 5 .0 1.56' 0 .00 0. 0~0 '0.00 0.00 0. 00 Category II-4-Fuel (Checkerboard Partner For Category 11-3 Fuel)

Initial Minimum Burnup (GWD/MTU).

Enrichment (w/o)' ,0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 75 .42: 61.90] 5'6 .85}

53. 00 54.18 52' 60'

ý4. 5 0 :68. 381 56.12 51.65 49.25! 47.76

50. 35 46.44 44. 19 42 .78 46 06. ý38. 80 35.41 33.52 32 -. 31f 31 .38: 25.71, 23 .12 21.65 1'. 87 29.19 23.83: 21'34 19.91 19.08

°0.00 0.00o 0.94ý 0 .00o 0.00 0. 00 Category Category 11-4 11-31 Category Category 11-3 11-4 SAN ONOFRE--UNIT 3 4.0-100-15 Rev. 2 X1X/XX/X,1k

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table 11-4 REGION II Checkerboard Storage Category II-5 Fuel (Checkerboard Partner For Category 11-6 Fuel)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 47.50 42.58 40.03, 38.53 37.55 4.50 42..40 37.95, 35.64 34.26 33 .37, 4 . 00 37.10 33'.16I 312. 140 29.86 29.06, 25.64ý 22.89, 21.40! 20.52 19.95 3.00 2.00 12.29 11.10, 10.42 10.01 9.75, 1.87 10.24 9.35' 8.80 8.46 8.25 1.38 0.00 0.060 0.006 0.00 0.00, CategoryII-6 Fuel

,(Checkerboard Partner For Category 11-5 Fuel).

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) .0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cgooling 5.00 62 .37 53 95 50.33' 48.25 46.91.

,4.50 56.21, 48 .90 ,45 .51, 43356 42.31

43. 67, 40:. 54 38.73 37.57 3 .00 37.71, 32.56 ý29.97 28.52 27.58, 2.00 23. 30 19..80i 18 .13i 17 .14 116.50 1-.87 21. 116 16 .42 15.48 14 88

1. 08' 0 .00 0.00 0.00; o.00 0.00 Category Category 11-6 I1-5 Category Category 11-5 11-6 SAN ONOFRE--UNIT 3 4.0-100-16 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Regi Aiq IT P.;rlbc LCS 4.0.100 Table 11-5 REGION II Checkerboard Storage initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Yearst Cooling Cooling Cooling Cooling Cooling 4.::80' 0.00 0.00G o0.00 0. 00. *0.00 4.80 Empty Fresh, (Blocked),

Empty 4.80 (Blocked). Fresh SAN ONOFRE--UNIT 3 4.0-100-17 Rev. 2 XX/XXI/XX,

Fuel Storage Patterns forReg-i-AR TT Pnr*z LCS 4.0.100 Table 11-6 REGION. II Category 11-7 Fuel (3 Out Of 4 Pattern)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o)X 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Coqoing Cooling "5.00 34.20 31.35' 29.74 28.76 28.12

4. 50 29.67 27.21 25.82 24.97, 24.41 4.00 24.94 :22. 92; 21.75 21. 05 20.59' 3 .00 14.79 13.76, 13 . 13 12 .73 12. 47, 2.90:

2.00, 3 .16 3. o. 2 . 83' 2.79,

1. 87 1.21 1.14 1. 09ý 1. 06; 1.04 1.80 0.00 0.00 0. 00 0. o0 0.00 Category Empty 11-7 (Blocked):

Category Category 11-71 11-7 SAN ONOFRE--UNIT 3 4.0-100-18 Rev. 2 XX/,XX/XX

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Table 11-7 REGION II Category 11-8 Fuel (Fuel With 5 GT-Inserts)

(Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5 .00 37.68 34.53 32.72 31.61 30.88.

4.50 32.61. 29.90 28.33 27.36 26 ..72 4.00 27.33ý 25.10 23.78 22 .97; 22 . 431 015.86 14.76 14.06: 13.332 32.00 0.00 2~ 1.97 1. 89 o.o00 1.84 1.081 1.90 0 .00, 0.00o 0.00.

Category Category 11-8 II-8 Category Category

.. 8 S1- 8 Note: Category 11-1, 11-8, 11-9, and-.11-15 may be stored together with no restrictions.-

SAN ONOFRE--UNIT 3 4.0-100-19 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Reks LCS 4.0.100 Table 11-8 REGION II Category 11-9 Fuel (Fuel With 3 GT-Inserts)

(Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)'

Enrichment (w/o) 0 Yearsý 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.. 0o 44 .16 39.95 36.31 35.42

,4.50 38.99: 35.25 33 .22 31.99' 3 1. 18:

4.00 33.61, 30.38 28.60 27 .52 26.81:

3.00. 18.72* 17.:56' 21 92 19.86 18.01 '1 56:

8 .28 7.72, 7.34.

1.87. 6.18 5.83 5.58 5.43 S5.34

1. 59. 0.00o 0. 00' 0.00. 0.00G 0.00' Category Category 11-9 11-9 Category Category 11-9 11-9 Note:i Category II-1l,' 11-8, 11-9, and 11-7'15 may 'be stored togqeither- w'ith no' restrictions..

SAN ONOFRE--UNIT 3 4.0-100-20 Rev. 2,XX/XX/XX

Fuel Storage Patterns for Region 11 Racks LCS 4.0.100 Table 11-9 REGION II Category II-10 Fuel (Filler Assembly With 5 Guide Tube Inserts)

Initial Minimum Burnup (GWD/MTU)

(w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 80.09 65. 66 60.12 57.45 55.68 4.50 72.13 59 .43i 54.55 52 . 02 50.33:

4 6.5 8 4.00 64. 18ý ,53.19 48.98 44 .98 3.00 48.27 40.72ý 37.16t 35.03i 33. 75 2.00 3. 35 26.5 9; 23.79 22.25 21.25

1. 03! 0.0oo: 0.00 0.00 0.00 Category 'Category Category II-10 Ji-10 I-10 (GT-Inserts) )(GT-Inserts) (GT-Inserts)

Category qn Category I1-10: Fresh ii-10 i(GT-Inserts) (GT-Inserts)l Category Category Category It-10 TI-10 (GT-z

ý(GT'-Inserts) f(GT-Inserts), (CT-Inse -rts),

SAN ONOFRE--UNIT 3 4.0-100-21 Rev. 2 XX/XX/XX,

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Table II-10 REGION II Category II-11 Fuel (Filler Assembly With 5 Guide Tube Inserts)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o)Y 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 47.04 42.52ý 40.051 38.57 37.60 4.50' 41 .62 37 .58 35.36 34.021 33.14 4.00 35.97 32.46, 30.50 29.32 28.54 3.00 23.70 21.42: 20.09 19.31. 18.79 21.00 9.17 8'. 5 8.:09' 7.881' 7.62 1.59 0.00 O.o0 0..00 0.00 0.0 Category Category Category II-ll III-ll' II-Ii (GT-Inserts) ý(CT-Inserts), (GT-Inserts)

Category 4-80 Category ll-1l Fresh Ii-ii

ý(GT-.Inserts)ý 57nsertS (GT-Inserts)i Category Category Categoryý,

II-il -I1se! )11-11I (CT-Inserts). (GT-I!nserts), (GT-Inserts)i SAN ONOFRE--UNIT 3 4.0-100-22 Rev. 2 XX/XX/XX,,'

Fuel Storage Patterns for Region II Ranks LCS 4.0.100 Table II-11 REGION II Category 11-12 Fuel (Filler Assembly With 3 GT-Inserts)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/0), 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 54.33 48. 48 45'.46 43.67 42.51 4.50 48.81 43-. 45' 40.67 39.02. 37.95

4. 00: 43.07 38.24 35.72 34.22: 33.26 3.00, 30.65 27 .11 25.181 24. 05 23.33 16.01 14. 23: 13.22 121.62 12.22'

'1.87 1-3 .82 12<. 35' i1' 47 10. 94ý 10.60

,1.32ý 0 .00 0.00 0. 0,0b 0. 00, 0.00 Category Category Category 11-12, 11-12 11-121 (GT-Inserts)t (GT-Inserts), (GT-Inserts)

Category" 4 8O0! Category 11-12 Fresh II-12 (GT-Inserts) 5GInserts (CT-Inserts)

Category Category Category 11-12 ii-12 II-12 (GT-Inserts) (GT-Inserts)' (GT-Inserts)

SAN ONOFRE--UNIT 3 4.0-100-23 Rev. 2 XX/XX/XX,

Fuel Storage Patterrns for Region 11 Raeks LCS 4.0.100 Table 11-12 REGION II Category 11-13 Fuel (Filler Assembly With No GT-Inserts)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o), 0 Years 5 Years 10 Years 15 Years, 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 64.24 51.59 49.41ý 48. 03

ý4. 50, 57.99 50.23 46 . 73: 44 .67, 43.38 51.75 44.94 41.71. 39 .79i 38.59 32.00 39.25 33.75 31.05 29 .48 28. 50

,2. 00 24.76 20.95 :19. 07 18.01, 17 .33 1 - 87: 22.64 19.10 17. 38 16.37 15.72 1.05 6.00 0'.0'0 0.00 0.00o 0. 00 Category Category Ca-tegory 11-13 11-13 1i-13

,4.80 Category Category 13' ,....Fresh -i3

ý5 Inserts Category Category Category ii-i3, ;11-13 ii1 SAN ONOFRE--UNIT 3 4.0-100-24 Rev. 2 XX/XX/XX,

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table 11-13 REGION II Category 11-14 Fuel (4.80 w/o Assembly Depleted to 18.0 GWD/MTU)

Initial Minimum Burnup (GWD/MTU),

Enrichment, (w/o), 0 Years 5 Years 10 Years 15 Years' 20 Years Cooling Cooling Cooling Cooling Cooling

5. 00, 1,9 .59 18.61 17.96 17. 54' 17. 27, 4 . 5'0 15 .93ý 1,5,.17: ,14.68 14.36 14 ý.15,

4. 00 12.1 8] 11.64 11.29 11. 07 10.93!

23.-0.5 4.28 4 .12 4.05 4.00 3.98 2.51i, 0.00 0.00 0.00 0.o0 0.00' Category 11-13 Fuel' 11-14 Fuel)1

ý(Filler *Assembly For Category Initial Minimum Burnup (GWD/MTU),

,Enrichment (w,/ 0 ) 0 Years 5 Years 10 Years' 15 Years, 20 Years Cooling Cooling Cooling Cooling Cooling 500f 64.24- :55.51, ;51.59 :49 .41, 48.03, 4.50' 57.99 50.231 46.73, :44 .67 43.38

4. 0o0 i51..75 44.94 41.71, 39.79 ý38.59Y 3.00o ;39. 2 5 33j.5 31. 05' ,29.48, 28.50 2.00 ;24.76 20.95 019.07 18.01O ý17. 33 22.64 i9 .1.o :17 .38 16.37 1.87t 15.72 0.00 0.00 0.00 0.00 0.00 Category Category Category, 11-13 11-131 II-131 Category 11-13 Category 11-14, Category 11- 13!

Category Category Category 11-13 11-13 11-13 SAN ONOFRE--UNIT 3 4.0-100-25 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table 11-14 REGION II Category 11-14 Fuel (4.80 w/o Assembly Depleted to 18.0 GWD/MTU)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 19.59 18.61 17.96 17.54 17.27 4.50 15.93 15.17 14.68 14.36 14.151, 4.00, 12. 18 11.64 11 29 11. 07, 10. 93 3 .00 4.28, 4.12 ý4.o05, 4.00 3.98ý 2.51i 0.00 0.00 0.00 0.00o 0.00 Category II-11 Fuel (Filler Assembly With 5 Guide Tube Inserts),ý Initial Minimum Burnup (GWD/MTU)

Enrichment (w/O)ý 0 Years ý5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling C 'doingi

'5. 00. .47. 04' '42.52 40.05 38.57 37.60

4. 501 41. 62 37.58- 35.36. 33 .14 34 . 02

'28. 541

35. 9 7' 32.46 30.550 29.32

ý3.

42.00O00 23.70, '21. 42' 20.09 19. 31; 18.79; 9.17; 8.54 8 .09; 7. 81: 7 . 62ý 1.59' 0.00. 0.00 .0.00 0.00 0. 00 Category Category Category II-ll II-1l1 II-1

,(,5 Inserts), (5ý Inserts) (5 Insertzs)

Category

'I11-1 Category Category II'ii-1 (5 Inserts) 11-14 (5 Inserts)

Category Category Category (5 Inserts) (5 Inserts) (5 Inserts)

SAN ONOFRE--UNIT 3 4.0-100-26 Rev. 2 XX/XX/ýX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Table 11-15 REGION II Category 11-15 Fuel (Fuel With 5 Finger Full Length CEA)

(Unrestricted Storage)

Initial Minimum Burnup (GWD/MTU)

Enrichment (w0o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 29 .24" 2-7. 24 26.00 25.22, 24.70' 22.8s4 21. 81 20.75

4. 50' 24.44 21.17T 4.00 19.41 18."26;

8. 47ý 17.49 17.00, 16. 68 8.83 8.i9 8 . 02 7.90,

,2.30 .0.00 0. 00 0.00' 0.00 0.00 Category Category 11-15 11-15 Category Category 11-15 11-15 N:ot'e:' Category II-1, II-8, 1i-9, and II-15 may be stored together with no

.restrictions.:

SAN ONOFRE--UNIT 3 4.0-100-27 Rev. 2',XX,/,XX/XX

Fuel Storage Patterns for egi- LnCS 4.0.1 LCS 4.0.100 Figure I-1 REGION I MINIMUM BURNUP FOR CATEGORY I-i FUEL 25 20 CL 15 00 E 10 Ql)

Cl)

U- 5 0 L-2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

- -- 0 Years e 5 Years -;, 10 Years -ý 15 Years 3ý- 20 Years SAN ONOFRE--UNIT 3 4.0-100-28 Rev. ý XX/'XX'/XX:

Fuel Storage Patterns for Region II Raks LCS 4.0.100 Figure 1-2 REGION I MINIMUM BURNUP FOR CATEGORY 1-2 FUEL 15-10 (CL1

-0 E

a1)

U,5 LL 0-4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

-a 0 Years E) 5 Years V 10 Years - 15 Years ff 20 Years SAN ONOFRE--UNIT 3 4.0-100-29 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 1-3 REGION I MINIMUM BURNUP FOR CATEGORY 1-3 FUEL I

i SAN ONOFRE--UNIT 3 4.0- 100-30 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Regon II Racks LCS 4.0.100 Figure 1-4 REGION I MINIMUM BURNUP FOR CATEGORY 1-4 FUEL 3o 0 20 .

E

. Acceptable Region M15

< Unacceptable Region 0~

2.0 2.5 3.0 3.5 Initial U-235 Enrichment (w/o) 4.0 4.5 5.0 &w 0 Years OEI a 5 Years w 10 Years -*- 15 Years m 20 Years SAN ONOFRE--UNIT 3 4.0-100-31 Rev. 2'XX/XX/XX,

Fuel Storage Patterns for Region II R*-ks LCS 4.0.100 Figure 1-5 REGION I MINIMUM BURNUP FOR CATEGORY 1-5 FUEL 35 30 ____ ___

25___ __

Acceptable Region

_ 20 -

zn> 15 . . . ...... .. .....................-..-.... .. . . .. ..... ... ... . .. ... . .. .

E co,

< 10

-Unacceptable 0LI 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

- 0Years O- 5 Years --- lOYears -.- 15Years x 20 Years SAN ONOFRE--UNIT 3 4.0-100-32 Rev. ,

2 XX/XX/XX

Fuel Storage Patterns for Region , 1 Ra"ks LCS 4.0.100 Figure 1-6 REGION I MINIMUM BURNUP FOR CATEGORY 1-6 FUEL 20 0 15 - Acceptable Region

3 FD 5 ..................................... Unacceptable Region U_

0 L 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o) 4V 1

- 0Years 0- 5Years I-vt 10 Years - 15Years -N 1- 20 Years SAN ONOFRE--UNIT 3 4.0-100-33 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region 11 Racks LCS 4.0.100 Figure 1-7 REGION I Boundary Between Unrestricted Storage And Checkerboard Storage I-i I-i I-ll I-i I-i i-l; I-i Ii-i I-i I-i I-i *I-i I-1 Empty I-1 I-i i-i I-l Empty 4. 8E0 1Emp ty I-I 1i 7-1 4.80 EmptyJ I-11-,~-i

~II Interface I -i i- i :i-li I-ii I -i .I -i}

I -i i i -ii I-1 I- 1 i-i

-l 1-64I-i

,Il ili

• IiI 4 ii{ i-1 I-i Ili!

1-6 1-4

Ii- i T-1 I-i .I-ý

• , II Interface Note: (1) A row of empty cells can be used at the interface to separate the configurations.'

(2) It is acceptable to replace an assembly with an empty cell.

SAN ONOFRE--UNIT 3 4.0-100-34 Rev. 2 XX/XX/XX:

Fuel Storage Patterris for Region II Racks LCS 4.0.100 Figure I-8 REGION I Boundary Between Unrestricted Storage And 1 Out Of 4 Storage I- i I1-1 -i I -i I-i' I~ I-i I -I,1 I-i I-i 1-i I-i. I 1' ,I -. 'Iii I - i I-l]

• 2 2! 13-i I-i I-i' S-3 i - 3 I -1i I-1 I -1 Interface Note:, (I) A row o~f empty ce'lls_ canbe used at the interface to separate the configurations; (2) It is acceptable to replace -an assembly with an empty cell.

SAN ONOFRE--UNIT 3 4.0-100-35 Rev. 2,XX/XX/XX

Fuel Storage Patterr s for Region 11 Racks LCS 4.0.100 Figure 1-9 REGION I Boundary BetweenCheck'erboard Storage And 1 Out Of 4 Storage 4.80 EMptiy 4.8 Empty 4.80 Empty Empty 4.80O EmptyA480 Empty .4-.80 4.80 Enit .0 Empty

".~,Eri Empljty J1-3~ Empty 4.80 ~Empty 48 0 1-31 1-p3 4.8,0 4.80.8 1-3 4.80: Epy 4. Empty Interface 1-6 I4I6 1-4 T-6

?i--4 1-6

-4 I 6 1-4 1-6 T-~4, -

1-6 1-4 T-6 1:-4, 16 -

1- J 1-3 T-13 64-6, 1 7- T-3 1 ,-, 1-6 Iý - 4-3 1-3 :ig3 1-46 d:*6:

11 11 Interfa~ce Note: A row, of exnpty cells can be used at the interfaceto separate the 4-)

configurations.

C2, it is acceptable to, replace an assebly with an e*mpty cell SAN ONOFRE--UNIT 3 4.0-100-36 Rev . 2 XX/XX/X-X

Fuel Storage Pattern 1z for Regi A Iq IT n-P.Or LCS 4.0.100 Figure II-1 REGION II MINIMUM BURNUP FOR CATEGORY IIl FUEL 60

,- 50 440 CL I3-E

)

20o

" 10 -

0 I 2 3 4 5 Initial U-235 Enrichment (w/o)

-E- 0 Years E3 5 Years - 10 Years - 15 Years Eý 20 Years SAN ONOFRE--UNIT 3 4.0-100-37 Rev. 2 XX/xx/xýý,

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 11-2 REGION II MINIMUM BURNUP FOR CATEGORY 11-2 FUEL 40 030-(D E

E U)

-~10-0-

1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E 0 Years -e--- 5 Years -ý 10Years 'ýz 15Years 20 Years SAN ONOFRE--UNIT 3 4.0-100-38 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Reginonr Raeks LCS 4.0.100 Figure 11-3 REGION II MINIMUM BURNUP FOR CATEGORY 11-3 FUEL 50 E40 30

-o M

E 20 Cl)

U) 0 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

SOYears 5 Years - l0Years -v- 15Years -s- 20Years SAN ONOFRE--UNIT 3 4.0-100-39 Rev. 21,XOX,/ý,X

Fuel Storage Patterns for Regi LCS 4.0.100 Figure 11-4 REGION II MINI MUM BU1NUP FOR CATEGORY, I-4.FUEL 80-060 (9

D.

m 40 2i, S20

.L, 2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years -E 5 Years - 10 Years l- 15 Years -s 20 Years SAN ONOFRE--UNIT 3 4.0-100-40 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Raeks LCS 4.0.100 Figure 11-5 REGION II MINIMUM BURNUP FOR CATEGORY 11-5 FUEL 50 E40-CL

=330 E20 a)

L~0 2 3 4 5 Initial U-235 Enrichment (w/o)

E3 OYears E3 5Years -- lYears IF 15Years -i 20 Years SAN ONOFRE--UNIT 3 4.0-100-41 Rev. ý2-,XX/.XX/XX

Fuel Storage Patterns for,-Re,*i*, n If Racks CS 4.0.100 Figure 11-6 REGION II MINIMUM BURNUP FOR CATEGORY 11-6 FUEL 70 2 3 4 5 Initial U-235 Enrichment (w/o)

- 0 Years -e-5 Years -- i10 Years IF 15 Years ---- 20 Years SAN ONOFRE--UNIT 3 4.0-100-42 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Figure 11-7 REGION II MINIMUM BURNUP FOR CATEGORY 11-7 FUEL 35 0

25 E 20 x15 E

fl,

<10 U_

1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

- ~0 Years E- 5 Years -+- 10 Years -ý- 15 Years -- 20 Years SAN ONOFRE--UNIT 3 4.0-100-43 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 11-8 REGION II MINIMUM BURNUP FOR CATEGORY 11-8 FUEL 40 6 30 m 20 E

(I)

W10-

--,)

U-0 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

S--.- 0 Years -E- 5 Years - 10 Years - 15 Years iw- 20 Years SAN ONOFRE--UNIT 3 4.0-100-44 Rev. 2,XX/XX/XX

Fuel Storage Patterns for Region II Rac"s LCS 4.0.100 Figure 11-9 REGION II MINIMUM BURNUP FOR CATEGORY 11-9 FUEL 50

- 40 30 E 20-0-

1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

- 0 Years -e- 5 Years -ý 10 Years v, 15 Years Ei 20 Years SAN ONOFRE--UNIT 3 4.0-100-45 Rev. ?.XX/XX/XX

Fuel Storage Patterns for Region II R*oks LCS 4.0.100 Figure II-10 REGION II MINIMUM BURNUP FOR CATEGORY II-10 FUEL 100 80 CD

-60 E

co

-0 E 40 U(

LL20 0

12 3 4 5 Initial U-235 Enrichment (w/o) 0 Years e 5 Years ý 10Years  ;, 15Years -s 20Years SAN ONOFRE--UNIT 3 4.0-100-46 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Ra"ks LCS 4.0.100 Figure II-1i REGION II MINIMUM BURNUP FOR CATEGORY II-l1 FUEL 50

-~40 30 E 20 a-Cn EL20 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

Eý OYears e 5Years lOYears sz 155Years iR 20Years SAN ONOFRE--UNIT 3 4.0-100-47 Rev. 2 XX/XX/XX,

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Figure 11-12 REGION II MINIMUM BURNUP FOR CATEGORY 11-12 FUEL 60 50 (40 a

E m 30 E

c 20 LiL 10 0

1 2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years Ozi -e- 5 Years -- 10 Years  ;;& 15 Years is- 20 Years SAN ONOFRE--UNIT 3 4.0-100-48 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region 1- Raeks LCS 4.0.100 Figure 11-13 REGION II MINIMU!M BURNUP FOR CATEGORY 11-13 FUEL 70-60

(.950-c 40 In -

"30-E U)

< 20-10-0-

1 2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years 5 Years - 10 Years v 15 Years iý 20 Years SAN ONOFRE--UNIT 3 4.0-100-49 Rev. 2 XX/Xx/xxý

Fuel Storage Patterns for Regi LCS 4.0.100 Figure 11-14 REGION II MINIMUIM BURNUP FOR CATEGORY 11-14 FUEL 20 015 P

m 0-tnl E

a) 5 0

2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

-s-0 Years - 5Years 5- - 10Years - 155Years EE 20Years SAN ONOFRE--UNIT 3 4.0-100-50 Rev. 2 XX/,XX/XX

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Figure 11-15 REGION II MINIMUM BURNUP FOR CATEGORY 11-15 FUEL 2.2 2.7 3.2 3.7 4.2 4.7 Initial U-235 Enrichment (w/o) 0 Years O~- e 5 Years - 10 Years li- 15 Years 11 20 Years SAN ONOFRE--UNIT 3 4.0-100-51 Rev. 2 XX/XX/XX

Fuel Storage Patterns for ,Reg-LCS 4.0.100 Figure 11-16 REGION II Boundary Between Unrestricted Storage And Checkerboard Storage II-1 lok I-i 11-1 II-i II-1 IIi II-i fIIi II-i II-i iI-i*

TI-i ITi1 1I1-! TI-i IT-i il-1 11i- JBlocked 11~-1 TI-i TI-i Inli Blocked A4.80 Blocked II-I iI-1 II-i; 4.`80d Blacke] ..-. t ,I Il-i Ii-i Interface Ii- 1 TI7-i1 I i1-1 Il-i TiI -i TIi- i iI-i iIli-! Ii II-i IT-i Ti-i liI-i i l <ii.i II ~1 I- i .<IT- ii Ti-i

-II-4 11-il': [I4 1+/--i TI-i ITi-iI 11-3 I I -4 i I-- i II-i IILI* :IT-T intef Interface ie Note: (1) A row of empty cells can be used at the interface to separaie the configurations.

(2)! It is acceptable to replace an assembly with an empty cell.

SAN ONOFRE--UNIT 3 4.0-100-52 Rev. 2 XX/XX/XX

Fuel Storage Patterns for eR-i LCS 4.0.100 Figure 11-17 REGION II Boundary Between Unrestricted Storage And 3 Out Of 4 Storage Ii-i II-i II-1 II-i II-i II-i IIi-1 11-1 II-i 1II-i lI-i TI-i Ih-id II-f, 1 i -i II1-i I-1 11-I biock7d 1i 1 Bl cked - I -1 IT-i I-i i11-7 J II 7 1 IT-i I IilT-- 11-i:

TI-I II-i I I -I 'I Blocked TII-7, B.locked Inter ec Note: (1) A row of empty ce-lls can be used at the interface to separate the conficigurations.,

(2.) It is acceptable to rgep~ace an assembly with an empty cell.

SAN ONOFRE--UNIT 3 4.0-100-53 Rev. 2 XX/XX/X)(,

Fuel Storage Patterns for egi-am 1- Racks LCS 4.0.100 Figure 11-18 REGION II Boundary Between Checkerboard Storage And 3 Out Of 4 Storage 11-7 Blocked 11-7 Blocked 11-7 Blocked I1-7 11-:7 11-1. 117 II-7 11-7, 11-7 BlOcked 11-7 Blocked 11-7, Blocked Blocked i1-75 JBocked 11-7 I1-7 I1-7 4.80 Blocked ,i-7 Blocked 1i-7. Blocked iII=7 Bid 4.80 ti o c k e d 11-7 11  :.11 I n Ee'rf a-c e Note:, (1) A row 'of Empty cell-s' c-ain ,be used-at the interface to1 separate the configurations.'

(2) It is acceptable to replace .an assembly with an empty cell' SAN ONOFRE--UNIT 3 4.0-100-54 Rev. 2'XX/XX/,XX

Fuel Storage Patterns for Region .. Racks LCS 4.0.100 Figure 11-19 REGION II Boundary Between l-Out-Of-9 Storage And Unrestricted Storage II-i II-i IT-i Ii-i II-i Il II-1 TIi-1 Ii11 iI-i TI-lI II-i Filler Filler Fillert I-i ii-1 Ii-i Filler A Filler 11-1i lIf'i 11- I Filler Filler Filler II- i ii iI-i Interface Wherte,: (1) If A 4.80 w/o + 0 Inserts, Filler Category II-10 (2) 5 Inserts, If A + Filler Category 11-11 (3) If A 4.80 w/o + 5 Inserts, Filler Category 11-12 (4) If A 4.80 w/o + 5 Inserts, Filler Category 11-13 (5) If A Category 11-14, Filler Category 11-13 (6) If A Category 11-14, Filler Category SAN ONOFRE--UNIT 3 4.0-100-55 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 11-20 REGION II Boundary Between l-Out,-Of-9 Storage And Checkerboard Storage 11-4, 11-3 11-4 11-3 I1-4ý 11-3 II-3 11-4 11-3 11-4II1 11-3 11-4 11-4 11-3 I1-4 11-3 11-4 11-3 Filler Filler Filler 11-4: 11-3, ii-4 Filler A ,Filler -3 11- 4 *I-3' 1

Ftiller 1i1e r FileI P"74' I1-3,3 TI-4 Interface Where:! If A 4.80 w/o + 0 Inserts, Filler = Category II-10 i(2 ) If A 4.80. w/o + 5 Inserts, Filler Category II-11 if 4.80 w/,o + ,5 Inserts, Filler = Category TT-12 A

if 4.80 w~/o + Inserts, I Filler = Category II-13 A

(5.) If Category 1I-I 4, Filler = Category 11-13 A

(6) if Category 11-14, Filler = Category 11-11 SAN ONOFRE--UNIT 3 4.0-100-56 Rev. 2 H/H/Xý,,

Fuel Storage Patterns for Region 11 Raeks LCS 4.0.100 Figure 11-21 REGION II Boundary Between l-Out-Of-9 Storage And Checkerboard Storage Blocked 4 80 Blocked 4.80 Blocked 4.80 4.80 Blocked 4.80 Blocked 4.80 Blocked Blocked 4. 80 Blocked 4.80 Blocked 4.80 F ile-r Alo-cke'd Filler Blocked 4.80 bl!o'c'k'e-d Filler' A Blocked 4.80 Biocked 4.880

'Filler ililer Filler Blocked 4.80 blocked Interface Where: (1 ) A 4.80 w/o

• 0 Inserts, Filler Category II-10 (2). if 4.80 w/o + 5 Inserts, Fililer If A Category (3) If A 4.80 w/0o + 5 Inserts, Filler Category 11-12 (4) A 4.80 w/o + 5 Inserts, Filler Category II -11 If (5)" -Category 11-14, Filler Category ,I -1.3 A

(.6) Filler Category SAN ONOFRE--UNIT 3 4.0-100-57 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 11-22 REGION II Boundary Between l-Out-Of-9 Storage And 3-Out-Of-4 Storage 11-7 Blocked 11-7 Blocked 11-7 Blocked 11-7 TII-7, 11-7 11-7 11-7; 11-7 II-7 Bl6ocked Ii-7 Blocked II-7ý Blocked Filler Filler Filler 11-7: 11-7 i-7 Filer- Ak Filler Bl0c ke d I I- Blocked Filler Fille-r 'Filler 11-7, 11-7 iI'-7 II Interface Where: (1) :I f A -4.80-w-/op + 0 Inserts, Filler Category II-10 (2) If A 4.80 wT/o + 5 Inserts, !Fi'ller Category Filler I dI-1 (3) if A 4.-80-w/o + 5 Inserts, Filler Category.

(4) 4.80 w/o + 5 Inserts, Category 11-13 If A (5) Category 11-14, Filler Category II--13 If A Fille-r (6) Category II-14, Category SAN ONOFRE--UNIT 3 4.0-100-58 Rev. 2 XX/XX/XX,,,

Fuel Storage Patterns for Region II Reks LCS 4.0.100 Figure 11-23 Orientation Of 3 Guide Tube Inserts Fuel Assembly

<--- Serial

.yy Number yI 00 yI xx 00 xx xx xx 00 xx 00 XX : Guide Tube With Insert 00 = Empty Guide Tube SAN ONOFRE--UNIT 3 4.0-100-59 Rev. 2 XX/XX/ýXý

Fuel Storage Patterns for Region II Racks LCS 4.0.100 Figure 11-24 SONGS UNITS 2 AND 3 FUEL ASSEMBLY WITH 40 ERBIA RODS 40 Erbial Ro~ds n High Enriched Fuel Rod Low Enriched Fuel Rod U 'Erbia Fuel Rod SAN ONOFRE--UNIT 3 4.0-100-60 Rev. 2 XX/XX/XX

Fuel Storage Patterns for Region II Raks LCS 4.0.100 Figure 11-25 SONGS UNITS 2 AND 3 FUEL-ASSEMBLY WITH 80 ERBIA RODS I I

~-Vf 80 Erbia Rods' DHgi h En riched Fuel -Rod ZLow Enriched Fuel Rod EErbia Fue 1 Rod SAN ONOFRE--UNIT 3 4.0-100-61 Rev. 2 1H/u/n

Fuel Storanp Pattprnq LCS 4.0.100 B LCS 4.0.100 Fuel Storage Patterns BASES BACKGROUND The spent fuel storage facility is designed to store either new (nonirradiated) nuclear fuel assemblies, or burned (irradiated) fuel assemblies in a vertical configuration underwater. The storage pool is sized to store 1542 fuel assemblies. Two types/sizes 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. Four Region II storage racks each

,contain 210 storage locations in a 14x15 array. The

remaining two Region II racks each contain 195 locations in' a 13x15 array. All Region II locations are spaced 8.85 inches on center.

As des cribed in LCS 4.0.100, "Fuel Storage Patterns fuel assemblies are st~ored in the spent' fuelracks in accordance, width criteria based on initial enrichment', di scharge burnu~p, and cooling time (plutonium decay).

,Addi-tional I y, L'CS-4-.0.1I00 a1lows t*he fol-lowing storage options:

x storage patterns.,

3x3 storage patterns,!,

credit for erbia in fresh assemblies, inserted CEAs, and borated stainless steel guide tube inserts

.(GT-Inserts)..

The effect of these storage options is to reduce the

,required discharge burnup of irradiated assemblies and add flexibility in storing fresh assemblies containing erbia.

Finally, credit for boron is taken to maintain Keff _<0.95.

The storaqe options' allowed by LCS 4.0.100 and credit for.

soluble boron are needed because the boraflex is assumed to be completely absent. Boraflex erosion/dissolution is an

,industry problem' and SONGS Units 2 and 3 are affected.

Taking no cred'it for boraflex for SONGSý Units 2 and 3 (continued)

SAN ONOFRE--UNIT 3 4.0-100-62 Rev. 2 XX/XX/XX,

Fuel Storanm Patt.Prns LCS 4.0. 100 BASES (continued)

BACKGROUND eliminates any boraflex concerns and monitorinqg Droqrams to (continued) ensure that an adequate amount of boraflex is present will not be needed.

A*PPLICABLE When soluble boron is credited,, the foll*o Iwing, acceptance

-SAFETY ANALYSIS criteria apply:

(1) Under normal conditions, the 95/95 neutron multi plication factor (K.,,,), iincludinq all, uncertainties, shall be less than 1.0 wh en. flooded with unb~orated water, and, (2)~ Und~er normal and accide'nt conditions, the 95/95~ neutron' miul~tiplication atr(., includinq all uncertainties, shall be less than or eqgual to 0.95 when flooded with borated water.

The storaceoptions allowed by LCS 4.0.100 meet acceptance criterion I,above. Under, normal, non-accident conditions, K~ffis l~ess than. 1.00 ~with 0 ppm.

The storage options allowed by LCS 4.0.100 meet acceptance criterion 2 above. Under normal, non-accident condit&ions.,

theTS3.7.17 limit for soluble boron concentration of 2000 ppmywill ensure that Keff remains less than or equal to 0.95.

The analytical methods used to demonstrate that the stforage options allVowed by [CS 4.0.100 meet acceptance c~riteria 1 and 2 are described in the SONGS Sp entF*el Poo'l Criticality Analysis (Reference 1).

A fuel assembly could be inadvertently loaded into a s pent fuel rack location not allow4ed by LCS 4.0.100 (e.g., an unirrradi ated fuel assembly,or an i nsuffi ci entl y depl ete'd fuel assembly). However, the neqative reactivi.:ty effect of the soluble boron compensates' for the increased reactivity caused by the Dostul]ated accident scenario. >Fo;r this accident condition, the TS 3.7.17li'mit for soluble boron con centrat on of 2000 ppm will ensure that Kep remains less or equal to 0,.95.

(conti nued)

SAN ONOFRE--UNIT 3 4.0-100-63 Rev. 27:xXXlXX/kX

Fuel Storanp Paf~tprnq LCS 4.0.100 BASES (continued),

LCS :The placement of fuel assemblies within the spent fuel pool accordinq to LCS 4.0.100 ensures that the K,,, of the spent fuel pool will always remain < 1.00 under normal,'non-accident conditions assuminq the pool to be floodedwith' unborated water. The K,, of the spent fuel pool will always remain < 0.95 under normal and accident conditions assuming.

the pool to be flooded with borated water at 2000 ppm..

APPLICABILITY This LCS applies whenever any fuel,assembly is stored-in the

spent fuel

pool ACTIONS: 'See" the Bas'es "for [CO 3.7'.18 A SURVEILLANCýE See the*Bases for'SR 3.7.18.1 REQU IREMENTS REFERENCES 1.ý Spent Fuel Pool Criticality, Analysis, (with no Borafl ex and Credit for Soluble Boron), Southern California:

Edison, S&an Onofre Nuclear Generating Station-Units 2, and 3, Revision 2, April 2007.:

(continued)

SAN ONOFRE--UNIT 3 4.0-100-64 Rev. :2XX/xx/xx

PCN 556 Attachment K (Spent Fuel Pool Dilution Analysis)

SPENT FUEL POOL DILUTION ANALYSIS SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 SEPTEMBER, 2005 SOUTHERN CALIFORNIA EDISON COMPANY

S023 SFP BORON DILUTION ANALYSIS Page 2 of 35 TABLE OF CONTENTS SECTION PAGE

1.0 INTRODUCTION

...................................................... 3 2.0 SPENT FUEL POOL AND RELATED SYSTEM FEATURES ................. 3 2.1 Spent Fuel Pool ................................................... 3 2.2 Spent Fuel Pool Storage Racks ...................................... 5 2.3 Spent Fuel Pool Cooling System ..................................... 5 2.4 Spent Fuel Pool Purification System ................................. 6 2.5 Dilution Sources and Flowrates ..................................... 6 2.6 Boration Sources ................................................ 15 2.7 Loss of Offsite Power (LOOP) ..................................... 16 2.8 Piping ...... ................................................... 17 2.9 Spent Fuel Pool Instrumentation ................................... 17 2.10 Administrative Controls .......................................... 17 3.0 SPENT FUEL POOL DILUTION EVALUATION .......................... 20 3.1 Calculation of Boron Dilution Times and Volumes .................... 20 3.2 Evaluation of Boron Dilution Events ................................ 22 3.3 Summary of Boron Dilution Events ................................. 23

4.0 CONCLUSION

S ...................................................... 34

5.0 REFERENCES

........................................................ 35

S023 SFP BORON DILUTION ANALYSIS Page 3 of 35

1.0 INTRODUCTION

A boron dilution analysis has been completed for crediting soluble boron in the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 spent fuel rack criticality analysis. The boron dilution analysis includes an evaluation of the following plant specific features:

- Dilution Sources

- Dilution Flow Rates

- Boration Sources

- Instrumentation

- Administrative Controls

- Piping

- Loss of Offsite Power Impact

- Boron Dilution Initiating Events

- Boron Dilution Times and Volumes The boron dilution analysis was performed to ensure that sufficient time, administrative procedures, and instrumentation are available to detect and mitigate the boron dilution before the spent fuel racks criticality analysis 0.95 Keff design basis is exceeded.

This analysis demonstrates that the final minimum boron concentration following a boron dilution event is 1,700 ppm. This soluble boron concentration is more conservative than the minimum boron concentration of 970 ppm required to maintain the spent fuel rack Keff less than 0.95 assuming normal plant operations during the dilution event. No other accidents, such as misloading a fuel assembly, are assumed to occur during the dilution accident.

2.0 SPENT FUEL POOL AND RELATED SYSTEM FEATURES This section provides background information on the SONGS Units 2 and 3 spent fuel pool and its related systems and features.

2.1 Spent Fuel Pool The purpose of the Spent Fuel Pool (SFP) is to provide safe storage for irradiated fuel assemblies. The SFP is filled with borated water. The water functions as a sink for decay heat generated by the irradiated fuel, as a shield to reduce personnel radiation exposure and to ensure that the offsite dose consequences due to a postulated fuel handling accident are acceptable.

Evaporation of SFP water occurs on a continuous basis due to the decay heat from irradiated fuel, and periodic SFP makeup is required. Because the evaporation process does not remove boron, makeup may be from an unborated water source. If the SFP is filled with borated water, evaporation may increase the boron concentration in the pool.

S023 SFP BORON DILUTION ANALYSIS Page 4 of 35 Each SONGS Unit has a separate SFP. The SFP is a reinforced concrete structure lined with 3/16" thick stainless steel welded liner plate. Behind the watertight liner plates are multiple horizontal and vertical chases which are connected by their individual drains to a leak detection sump. Observation of the leakage from the drains allows identification of the approximate location of the leak. The fuel handling building (FHB) and the SFP are designed as seismic Class I structures.

The SFP operating deck is located above grade at the 63.5 feet elevation of the fuel handling building. The nominal SFP water depth is 43.5 feet, which corresponds to the plant elevation of 61 feet or the level of 28 feet on the SFP ruler. The high water level alarm is at an elevation 61'-

4", the low level alarm is at an elevation 59'-6" and the minimum water level is at an elevation 55'-11 15/16", which is 23 feet above the top of the stored fuel assemblies at an elevation 32'-11 15/16".

The plant elevations and water levels are summarized in Table 2-1, and illustrated in Figure 2-1.

Both of the SONGS Units 2 and 3 SFPs are divided into three areas which are thermally and hydraulically coupled, but are separated by gates. The larger area (the main SFP) is used primarily for the storage of fuel. One smaller area, the cask storage pool (CP) is used primarily for loading of fuel storage or transportation casks. Another smaller area is the fuel transfer pool (TP), used during refueling.

The refueling canal (cavity) lies adjacent to the fuel transfer pool, and is connected to it by a transfer tube. The transfer tube is normally closed by an installed blind flange and a manually operated gate valve. If draining the refueling canal is desired following the completion of refueling operations, the gate is closed to isolate the refueling canal and the fuel transfer pool from the main SFP. The gate may also be closed to isolate the cask loading area from the main SFP; these gates are normally open to ensure SFP water quality. The elevation of both keyway bottoms is 35'-5 1/1/2", which is above the top of the stored fuel assemblies at an elevation of 32'-11 15/16" (Table 2-1).

Considering the volume displaced by a full loading of irradiated fuel, the volume contained in the SFP (with the CP and TP isolated) corresponding to the low level of 59'-6" is 349,931 gallons.

2.1.1 Spent Fuel Pool Overflow The main SFP overflow line S2(3)1219ML041 is connected to the overflow line S2(3)1219ML021 from the fuel transfer pool which is connected to the overflow line S2(3)1219ML022 from the cask storage pool. All the above overflows are directed into the fuel handling building sump. The sump is provided with a HI-HI level switch 2(3)LSHH5827, which alarms in the control room (CR).

The Fuel Handling Building Sump is located at the elevation 17.5 ft, has an overall length of 4.8 ft by 3.9 ft wide by 6 ft deep. The sump houses two fuel handling building sump pumps, S2(3)2426MP328 and -329, with a nominal flowrate of 50 gpm. The sump pumps are operated by a hand-switch. The evaluation of line and sump volumes (Reference 1) identified the total volume up to the sump level switch setpoint to be 600 gallons.

S023 SFP BORON DILUTION ANALYSIS Page 5 of 35 The fuel building sump pumps are normally not in operation and are started manually after the HI-FR alarm. The response to the HI-HI alarm is per Alarm Response Instruction S023-15-56.C, item 56C56 (Fuel Hdlg Bldg Sump Level HI-HI), which describes the Operations activity after that alarm. Specifically, alarm response instruction S023-15-56.C requires inspection for leakage and performing a tour of the Fuel Handling Building to uncover the cause of the HI-HI level. Investigation showed that Operations personnel need about 30 minutes to perform this search. The time needed to uncover the real cause of the overflow has been identified by Operations to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, based on the walkdown of the area. Thus, SFP overflow would cause HI-HI level after a transient volume of 600 gallons is filled, and the Operations can isolate the inflow 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the alarm sounds in the control room.

2.1.2 Required hydraulic head at the SFP overflow pipe The top of the overflow elbow is at the elevation of 61'-5". To this elevation we will add a hydraulic head due to the flow into the elbow derived in Reference 1.

2.2 Spent Fuel Pool Storage Racks The spent fuel storage racks provide for storage of new and spent fuel assemblies in the spent fuel pool, while maintaining a coolable geometry, preventing criticality, and protecting the fuel assemblies from excess mechanical or thermal loadings. Storage is divided into two regions within the pool. Region I has 312 locations and is generally reserved for temporary storage of new fuel or partially irradiated fuel which would not qualify for Region 11 storage. Region 1t has 1230 locations and is generally used for long term storage of permanently discharged fuel that has achieved qualifying burnup levels.

2.3 Spent Fuel Pool Cooling System The SFP cooling system is designed to remove the decay heat generated by the irradiated fuel assemblies stored in the pool. The fuel pool cooling system consists of two pumps and two SFP heat exchangers (HXs) arranged in parallel with each pump capable of directing flow to either heat exchanger. Each SFP heat exchanger rejects heat to the component cooling water system which is cooled by the ultimate heat sink. Piping for the SFP cooling system is Seismic Category I and is arranged so that a piping failure will not drain the SFP below the top of the stored fuel assemblies. The SFP cooling system has piping ties with the safety injection system (shutdown cooling heat exchanger), which provides an alternate means of cooling the SFP.

The fuel pool cooling system suction line penetrates the pool liner at elevation 54'-10". The operating deck elevation is at 63.5 feet. The fuel pool cooling return line $2(3)-031-10"-D-LL0 enters the pool from above on the west side of the pool. The fuel pool cooling return line is equipped with an anti-siphon pipe S2(3)-080-1.5"-D-LL0, which has open ends at the elevation 58'-11". The top of the stored fuel assemblies is at 32'- 11 15/16".

The fuel pool cooling system is capable of removing the design basis heat loads as described in UIFSAR Section 9.1.3.

S023 SFP BORON DILUTION ANALYSIS Page 6 of 35 2.4 Spent Fuel Pool Purification System The SFP cleanup or purification system is designed to remove soluble and insoluble foreign matter from the SFP water and dust from the pool surface. This maintains the SFP water purity and clarity, permitting visual observation of underwater operations. The purification system interfaces with the SFP separate from the SFP cooling system and consists of a purification pump, a filter, an ion exchanger, an ion exchanger strainer, a surface debris skimmer, and various valves and instrumentation. Purification is conducted on an intermittent basis as required by SFP conditions. The fuel pool purification pump has a design flowrate of 150 gpm.

In addition to purifying the SFP water, the SFP makeup water from the refueling water tank(s) may be cleaned through connections to the purification loop.

2.5 Dilution Sources and Flowrates Table 2-2 summarizes the credible dilution sources and associated flowrates. The listed events will be used in Section 3 for numerical evaluation of SFP boron dilution times and SFP boron concentration values. The dilution sources are discussed below.

2.5.1 Spent Fuel Pool Makeup System The makeup which is evaluated here is the unborated water makeup, as that one has the potential for diluting the SFP. Makeup from the refueling water storage tank (RWST) would not result in dilution of the SFP, as RWST contains borated water with a minimum concentration of 2,350 ppm per Technical Specification 3.5.4.

The demineralized water makeup is proceduralized to assure that dilution will not occur.

Procedure S023-3-2.11.1 describes three ways of adding demineralized water to the SFP:

a) SFP makeup using primary makeup water (demineralized water) from SA1415MT055 and -056 to the SFP cooling pump.

b) SFP mhkeup (potentially demineralized water) from Radwaste Primary Tanks (RPTs) SA1901MT065 and -066.

c) SFP makeup (potentially demineralized water) from RPTs SA1901MT067 and

-068.

Procedure S023-3-2.11.1 uses a rigorous method for adding demineralized (unborated) makeup to the SFP. Prior to such makeup, the boron concentration in the SFP must be sampled, and an initial and final boron concentration determined by a calculation performed in the procedure, based on SFP initial and final water levels. Due to the rigorous nature of this administrative procedure, a potential operator error is eliminated. Thus the probability of transferring more than the required quantity of unborated makeup to the SFP is judged to be extremely low. However, for this evaluation, it is assumed that the operators will not isolate the primary makeup. This case is shown on Table 3-1, which shows that for an initial boron concentration of 2,000 ppm and when the SFP reaches the high level, and consequent high level alarm, the operators will have 280 minutes before the SFP boron concentration reaches the 1,700 ppm limit to verify that the possible makeup sources are closed. The Operations personnel will be able to isolate the makeup within this time.

S023 SFP BORON DILUTION ANALYSIS Page 7 of 35 2.5.1.1 Makeup from the Primary Plant Makeup Storage Tank(s)

Prior to the system alignment for makeup from the Primary Plant Makeup Storage Tank(s)

SA1415MT055 and -056, procedure S023-3-2.11.1 requires performing calculations of SFP volumes and boron concentration, as a result of the makeup. If these calculations show that SFP boron concentration following makeup would be Ž 2,650 ppm, plant procedures permit SFP makeup to originate from the Primary Plant Makeup Storage Tanks. In this evaluation, the primary makeup water is assumed to contain 0 ppm soluble boron.

The makeup requires a specific valve alignment (opening normally closed valves, etc.). The Primary Makeup Storage Tank Pumps SA1415MP200, -201, -202 and -203 draw suction from the 300,000 gallon capacity Primary Makeup Storage Tanks SA1415MT055 and -056 (S023-407-3-61). These tanks have high and low level alarms which annunciate locally and in the control room. The.Unit 2 and 3 primary plant makeup storage tanks are filled by a flow from the radwaste deborating ion exchangers SA1415MEO83 and -084.

Normally, one primary makeup tank pump is in service; the second pump remains in standby.

Either pump is started manually by a handswitch and the pumps are prevented from starting (or are stopped) when low-low level takes place in the Primary Plant Makeup Storage Tank (at either Units 2 or 3). Each pump has a design flowrate of 160 gpm and provides flow to the coolant polishing demineralizer, from where the condensate is distributed to a variety of plant equipment.

Thus, 160 gpm would be assumed to be directed to the SFP.

Prior to the makeup operation, the makeup mode selector hand switch is first selected to be in the manual mode. Then, for the demineralized makeup to be lined-up to the SFP, manual valves SA1901MU572 (Unit 2) or SA1901MU584 (Unit 3) need to be opened, in addition to manual valve $2(3)1901MU574. The primary makeup water pump is then manually started.

The contents of the Primary Plant Makeup Storage Tank(s) SA1415MT055(-056) can be transferred to the SFP through a single 3-inch branch line using the primary makeup tank pump(s) as motive force. The makeup line S2(3)1219ML071 connects to the SFP cooling pump(s) suction line S2(3)1219ML010 at approximately 38 foot elevation. Makeup to the SFP through this 3-inch line may be accomplished by opening the normally closed valves S2(3)1901MMU574 and S2(3)1219MU096, which is the procedurally specified makeup path.

Normal Operating Procedure S023-3-2.11.1 specifies that primary makeup flow is established by manually opening (2 - 4 turns) the 3" ASME Section III isolation valve $2(3)1219MU096 located by SFP HX at the elevation 30' of the FHB. The valve is then gradually opened. The primary makeup flow is indicated on CR flow indicator. The flow path is directed to the suction side of SFP Cooling Pumps.

Due to the rigorous administrative procedure used for adding makeup to the SFP, the probability of SFP dilution is very low.

S023 SFP BORON DILUTION ANALYSIS Page 8 of 35 2.5.1.2 Makeup from the Radwaste Primary Tank(s) SA1901MT065/066 Makeup to the SFP from the Radwaste Primary Tanks (RPTs) is done through the SFP Purification Ion Exchanger $2(3)1219ME071, SFP purification discharge line, and into the SFP Cooling system. The SFP Purification pump $2(3)1219MP014 will be stopped (due to lack of miniflow during the transfer) and then restarted after the transfer.

Prior to the makeup, the Chemistry division will obtain a sample from the RPTs, to verify the boron concentration. If the boron concentration of the water in the RPTs is higher than 2,650 ppm, no additional calculation of the SFP final boron concentration is required, and, the makeup operation can proceed. If, however, the boron concentration in the RPTs is < 2,650 ppm, the calculation of final (after makeup) boron concentration in the SFP is required per procedure S023-3-2.1 1.I. Currently the makeup is only allowed, if the final SFP boron concentration is higher than 1,900 ppm with the SFP not connected to the Refueling Cavity, or, is >_2,650 ppm with the SFP connected to the Refueling Cavity.

If the makeup to the SFP is allowed per the procedure, a crosstie path alignment is then established. The discharge valve S2(3)1219MU021 from the SFP purification pump S2(3)1219MP014 is closed, the pump miniflow isolation valve $2(3)1219MU098, the recirculation isolation valve SAI901MU410 to Radwaste Primary Tank SA1901MT065 and the recirculation isolation valve SA1901MU411 to Radwaste Primary Tank SA1901MT066 are closed. The RPT crosstie to Unit 2 (throttle valve S2(3)1219MU173) is opened. RPT crosstie to Unit 2 (isolation valve S2(3)1219MU172) is unlocked and opened.

Likewise, the valves in the RPT crosstie to Unit 3 are opened. To start the transfer of water, valve SA1901MU476 is opened and the RPT pump SA1901MP169 is started. The transfer flow is about 60 gpm. Communications must be established to communicate SFP level changes to the control room.

As the design flow of the RPT pump is 140 gpm, this case is bounded by the previous case in Section 2.5.1.1. Due to the rigorous administrative procedure for adding makeup to the SFP (plant procedure S023-3-2.11.1), the probability of SFP dilution using this path is very low.

2.5.1.3 Makeup from the Radwaste Primary Tank(s) SA1901MT067/068 The makeup from Primary Radwaste Tank(s) SA1901MT067/068 is similar to the makeup from tanks SA1901MT065/066 and the same rigor is used for the makeup alignment as outlined above. This case is bounded by the case described in Section 2.5.1. 1. Due to the rigorous administrative procedure of adding makeup to the SFP, the probability of SFP dilution using this path is very low.

S023 SFP BORON DILUTION ANALYSIS Page 9 of 35 2.5.2 Nuclear Service Water Systems Nuclear service water (NSW) is a demineralized water and it is used at both units for water service stations (hoses) and washdown stations. The NSW is delivered to the system by NSW pumps SA1415MP138 or -139, which take suction from the nuclear service water storage tank SAI415MT104. The NSW storage tank has a gross capacity of 26,550 gallons.

The makeup to the NSW storage tank is supplied from the demineralized water makeup, supplied by the Makeup Demineralizer (MUD) tanks SA1417MT266, -267, -268. Each MUD tank has a nominal capacity of 535,000 gallons, with a level controlled at a minimum of 75%. Procedure S023-11-6, specifies that High Flow MUD system is operated when the level drops to 75%, to replenish the tank. Based on the above data, sufficient capacity exists in these tanks to provide makeup in case of postulated outflow from the NSW system.

There are 2 NSW pumps of nominal 200 gpm capacity, one in operation and one in standby, actuated by a low pressure signal in the pump discharge header. This double-pump NSW system is common to both Units 2 and 3 and the NSW is supplied throughout both Units 2 and 3. There are dedicated 1.5 inch pipe headers leading into each unit's fuel handling building.

The NSW piping is routed in the vicinity of the SFPs, specifically 1.5 inch line S2(3)1415-223-1.5"-R-LL1, which serves several water stations and is finally reduced to a 1 inch size (similar at both units).

In this case it is postulated that an operator forgets to close one hose station, thus allowing the hose to continue flowing at an estimated flowrate of 50 gpm.

Also, the NSW header will be evaluated for a pipe break flow. As the NSW piping is routed in the direct vicinity of the SFPs, and, the system is normally pressurized, this system is considered to a be viable source of boron dilution.

2.5.3 Component Cooling Water Component cooling water (CCW) is the cooling medium for the SFP cooling heat exchangers.

The CCW contains zero (0) boron. The portion of CCW that interacts with the SFP heat exchangers is seismic Category I. There is no direct connection between the component cooling water system and the SFP cooling water system. However, if a leak were to develop in either of the SFP heat exchangers (S2(3)1219ME003 or S2(3)1219ME004) while in service, a connection between the two systems would be made. As the CCW system operating pressure (110 psig) is higher than the SFP cooling system operating pressure (35 psig), the flow would be from the CCW to the SFP.

S023 SFP BORON DILUTION ANALYSIS Page 10 of 35 The CCW system contains a surge tank (S2(3)1203MT003/004) which is designed to accommodate fluid volumetric changes and to maintain a static pressure head at each CCW pump suction. Any leakage path between the fuel pool heat exchanger shell and tube sides will result in a reduction in the surge tank level and will cause demineralized water to be automatically added to the CCW surge tank via an automatic water level control system, with makeup from the nuclear service water.

Although the CCW surge tank has an automatic level control, should a major leakage occur, beyond the capacity of the makeup (200 gpm), low level would trigger a low level alarm in the control room. A calculation performed in Reference I has shown that the upper bound leak rate from a single failed tube is approximately 90 gpm. As the leak rate associated with the rupture of one tube is smaller than 200 gpm, a CCW surge tank low level alarm may not sound in the control room as a result of one broken heat exchanger tube.

Based on the above, the credible dilution scenario for the CCW is the rupture of the SFP Heat Exchanger tube.

2.5.4 Back-flushing the Fuel Pool Filter Fuel Pool Filter S2(3)1219MF016 is backflushed periodically by using procedure S023-8-3.

The backflushing can be automatic or manual. Upstream/downstream process isolation valves must be closed prior to the backflushing. After that, nitrogen is used for backflushing the filter into the backflush filter crud tank T-073. The process of filter backflushing does not require the use of nuclear condensate, however, a connection is provided on the return line to the crud tank T-073 to allow flushing that line. The backflushing valves must be closed before the process valves are open, otherwise the system would not operate properly.

Specifically, during the back-flushing process, the process flow, i.e. the SFP purification flow, is isolated by closing the air-operated inlet/outlet valves 2(3)HV-7733A and 2(3)HV-7733B. The back-flush inlet/outlet valves 2(3)HIV-7733C and 2(3)HV-7733D are then opened to allow the back-flushing to take place.

Based on the fact that backflushing of the filter uses nitrogen as the motive fluid, there is no boron dilution path during this alignment.

2.5.5 Resin Flush Line/Resin Fill for Spent Fuel Pool Ion Exchanger E-071 The resin transfer operation is performed approximately once each fuel cycle (18-24 months), to flush spent resin from the ion exchanger.

SFP ion exchanger S2(3)1219ME071 is equipped with resin charge/flush lines. Nuclear condensate is used for flushing the ion exchanger during this flushing operation. The nuclear condensate system is supplied by the nuclear service water system, by line SA1415-703-2"-J-LLO. The resin transfer is performed in accordance with procedure S023-8-12. The procedure requires closing the process isolation valves upstream/downstream of the exchanger S2(3)1219MU028/032, thus the dilution with condensate would be limited only to the piping around the ion exchanger and the ion exchanger itself.

S023 SFP BORON DILUTION ANALYSIS Page 11 of 35 The opening as well as the closing of the nuclear condensate valve is done with administrative verification of valve alignment (an operator initial is required in the procedure to verify that the alignment has been made). In the case of this procedure there is a high degree of certainty that the alignment is performed, due to the structure of the procedure. Specifically, not only an alignment must be verified, but there are three valves ($2(3)1219MU056, S2(3)1219MU057 and

$2(3)1219MU058) that are aligned (opened or closed) at the same time at specific steps of the procedure, and those valve numbers are displayed prominently in a special text box in the procedure.

The volume of condensate is thus limited to the volume of the exchanger and the associated process piping up to the isolation valves. This volume has been derived in Reference I to be 2000 gallons. Calculations documented for other dilution paths (see for example Table 3-7) show a very minimal dilution with dilution volume of 7,800 gallons. Thus, this dilution path need not be evaluated separately as it is bounded by all other evaluated dilution events.

2.5.6 Fire Protection System In addition to the primary makeup water, the SFP room is served by two fire hose stations, fed by the plant fire protection system pumps, which take suction at the plant fire tanks. The quality of the fire tank water is similar to the domestic water, which contains zero boron. Two fire protection hose stations are located in the vicinity of the SFP (FHS 118 and 119 in Unit 2 and FHS 128 and 129 in Unit 3). 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. Normally closed isolation valves are used to control the delivery of water to the hoses, however in case of fire, it is assumed that both hoses will be deployed. Any discharge from the fire hose to the SFP would require its manual removal from the reel and operation by fire fighting personnel. A flow rate of 150 gpm is assumed to result from initiation of one fire hose flow. With 2 hoses, the total flow is 300 gpm.

A SFP dilution through a deployed fire hose is an evolution that is easily observable by operations personnel. As the fire loading in the SFP room is less than 1E7 Btu, the duration of the fire is only 0.02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (per SONGS 2 and 3 Fire Hazard Analysis, fire hazard area 2FH17-123) thus the quantity of fire water discharged into the SFP is only 360 gallons. Even if we assume more than twice the volume (1000 gallons), the dilution volume is too low to affect the boron concentration in the SFP significantly. The easy visibility combined with the operator involvement and low volume mean that fire water supplied to the SFP through the two hose stations is not a credible dilution path to dilute the SFP. Based on the above, the dilution source from fire fighting activities is bounded by all other evaluated dilution events, and will not be evaluated separately.

Another concern is a potential dilution stre~am due to leakage and/or crack in the fire protection header, which would divert unborated water into the SFP. As the fire water piping is routed in the direct vicinity of the SFPs, and, the fire protection water system is normally pressurized, this system is considered to be a viable source of boron dilution and is discussed in the next section.

S023 SFP BORON DILUTION ANALYSIS Page 12 of 35 2.5.7 Dilution from Pipe / Component Break Events The fuel handling building is a seismic category I reinforced concrete structure containing the SFP, spent fuel cask area, refueling canal, spent fuel cooling and purification pumps, heat exchangers, filters and ventilation equipment. The FHB exterior walls, floors and partitions are designed to protect the equipment inside from the effects of hurricane and tornado winds, external missiles and flooding. The SFP portion of the F-B, including the walls and roof directly above the pool, is designed to withstand, without penetration, the impact of external missiles that might occur during the passage of a tornado. The SFP is located above grade with a pool floor elevation of 17.5 feet and an operating deck elevation of 63.5 feet.

The below Sections evaluate the relevant pipe/component break flows at Unit 2 (U3 is similar) piping in the vicinity of the SFP.

2.5.7.1 Pipe Break in the Fire Water Header SA-2301-051-4"-W-LS I - Fire Water. This header branches off to other pipes/hose stations; however, for conservatism it is postulated that there is one pipe break in this main header. As the fire water has moderate pressure, a critical crack will be postulated. The pipe is 4 inch standard (std) weight (schedule 40), 4.026 inch internal diameter (ID), with 0.237 inch wall.

The critical crack flow has been determined in Reference 1 to be 110 gpm.

2.5.7.2 Pipe Break in the Nuclear Service Water Header S2-1415-223-1.5"-R-LL4 - Nuclear Service Water. This header branches off to other pipes/hose stations; however, for conservatism it is postulated that there is one pipe break in this main header. The pipe is stainless steel, schedule 40S, 1.61 inch ID, with 0.145 inch wall.

The critical crack flow has been determined in Reference 1 to be 30 gpm.

2.5.7.3 Pipe Break in the SFP Cooling Water Return Header S2-1219-025-12"-D-LLO - SFP Cooling water return header. This header branches off to other smaller diameter pipes; however, for conservatism it is postulated that there is one pipe break in this main header. As the SFP cooling water has moderate pressure, a critical crack will be assumed in this piping. The pipe is schedule 10S, with 12.39" ID, and wall thickness of 0.18".

The critical crack flow has been determined in Reference 1 to be 130 gpm.

2.5.7.4 Pipe Break in the SFP Cooling Water Suction Header S2-1219-010-14"-D-LLO - SFP Cooling water suction header. A critical crack will be assumed in this piping. The pipe is schedule 10S, with 13.624" ID, and wall thickness of 0.188". The critical crack flow has been determined in Reference 1 to be 107.2 gpm.

S023 SFP BORON DILUTION ANALYSIS Page 13 of 35 2.5.7.5 Pipe Break in the Fuel Pool Purification Pump P-014 Discharge Header S2-1219-018-3"-J-LL0, SFP Purification pump discharge header. A critical crack will be assumed in this piping. The pipe is schedule 10S, with 3.26" ID, and wall thickness of 0.12".

The critical crack flow has been determined in Reference I to be 38 gpm.

2.5.7.6 Tube Break in the SFP Heat Exchanger SFP Heat Exchangers $2(3)1219ME005, -ME006 tube bundle features 490 3/4" tubes 22 gage of 304 stainless steel (S/S) material. Rupture of only I tube will be assumed.

The tube outside diameter (OD)= 0.75", and the tube wall is 0.028". The design pressure of the SFP cooling piping is 50 psi, and the operating pressure is 35 psi. The flow through a ruptured tube has been derived in Reference 1 to be 90 gpm.

2.5.7.7 Pipe Break in the Demineralized Water Makeup This makeup pipe (S2-1219-029-3"-J-LLO) is connected to the SFP cooling line at an elevation 20.75'. The effects of rupture of demineralized water makeup pipe S2-1219-029-3"-J-LLO need not be considered, as they are bounded by the break in the 12" SFP cooling water return header line.

2.5.7.8 Pipe Breaks due to Tornado Events The effects of tornado or hurricane events have been reviewed and found to not result in any rupture of piping adjacent/associated with the SFP or related systems. As a result, dilution resulting from a tornado or hurricane is not a credible event.

2.5.7.9 Crack in the SFP Liner Plate The SFP liner plate is a 3/16" thick stainless steel welded plate. Behind the watertight liner plates are multiple horizontal and vertical chases which are connected by their individual drains to a leak detection sump. Observation of the leakage from the drains allows identification of the approximate location of the leak. Thus the liner plate leakage ends up directly in the Fuel Handling Building Sump, which is equipped with a 1I-1-HI alarm. The transient volume of the leakage to result in the sump HI-HI level is conservatively estimated at 1000 gallons.

The leakage rate associated with SFP liner plate damage has been evaluated before in the licensing submittal for the SONGS 2 and 3 SFP re-racking (Licensing Amendment Applications 146 and 130, dated July 29, 1996 ) and it was identified as 49 gpm.

S023 SFP BORON DILUTION ANALYSIS Page 14 of 35 2.5.7.10 Leakage through the SFP Gates As mentioned above, the SFP is connected to two other cavities, i.e. 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 in the SONGS Units 2 and 3 SFPs. However, this evaluation assumes that these two gates are closed (thus isolating the two cavities from the SFP), as that condition provides a smaller SFP initial volume. This is more conservative from the boron dilution standpoint, than the situation with the gates open. Leakage through the closed gates would require subsequent makeup, which represents a potential dilution scenario.

Each SFP bulkhead seal consists of 3-5" wide by 28'-7.5" high gate with two independent pressurized bladders which provide redundant means for sealing the surfaces between the bulkhead gates and the SFP walls. These pressurized seals are constructed from heavy radiation resistant reinforced fabric. Both seals on each gate are held in place by a total of 174 two inch wide clips. Eighty-seven of these clips are welded to the gate and 87 clips are held in place by 174 screws.

The normal operating pressure of the seal is 20 psig and the motive air is the service air, backed-up by compressed air bottles. Low pressure switches at the south gate inner seal (2(3)PSL-7777A) and outer seal (2(3)PSL-7777B) alarm at the local SFP panel. Likewise, low pressure switches at the north gate inner seal (2(3)PSL-7778A) and outer seal (2(3)PSL-7778B) alarm at the local SFP panel.

The potential for the SFP gate leakage was a subject of NRC IE Bulletin 84-03 dated August 24, 1984. SCE responded to this IE Bulletin in a letter from M.D. Medford to USNRC dated October 26, 1984, which concluded that due to the redundant nature of the motive gas, and the seals, as well as low pressure alarms, it is not credible to postulate a leak in these pneumatic water seals, which would exceed the normal makeup capacity to the SFP.

2.5.8 Evaluation of Infrequent Spent Fuel Pool Configurations 2.5.8.1 Dilution of SFP with Cask Storage Area Isolated Although unlikely, it is possible that the main SFP could be unintentionally isolated from the cask storage area. But as noted above, this evaluation already assumes the cask pool isolated, as this lineup is more conservative. Based on that, no additional calculations need to be done for this lineup.

S023 SFP BORON DILUTION ANALYSIS Page 15 of 35 2.5.8.2 Filling the Refueling Canal and Pool To prepare for refueling activities, the fuel transfer (refueling) canal must be filled with borated water with at least 2,650 ppm boron, per procedure S023-3-2.11.1. As noted previously, a bulkhead between the SFP and the transfer canal is normally open. However, prior to the refueling operations, the gate at the fuel transfer pool is closed, and, the borated water pumped out to a holding tank (such as Radwaste Primary Tank(s) SA1901MT067, 068), to facilitate the installation of the refueling equipment. The transfer pool is then refilled typically from the refueling pool side, which is filled from the RWST. Otherwise, plant procedures used for filling the transfer canal specify that makeup is taken directly from the RWST, which is a borated tank.

Also, per plant procedures, the bulkheads between the pools may not be open until the water level in the corresponding areas is approximately equal to the water level in the SFP. Opening the bulkheads when the adjacent area is empty is thus precluded by plant operation procedures.

Based on the fact that the boron concentration during refueling is higher than 2,600 ppm, this case is bounded by previous SFP boron dilution cases.

2.5.8.3 SFP Dilution During Fuel Shipment Fuel shipment out of the SFP (to dry cask storage or other destinations) is an infrequent operation, and during this operation the SFP low level is lower than during normal operation.

The water levels during fuel shipment will not be used except for the worst-case scenario (Case 7), which will be verified (Table 3-7A) with the SFP.LL setpoint lowered to 57.667 ft from the normal 59.5 ft.

It should be noted that the likelihood of SFP overflow during this operation is extremely low, as operation personnel are present during the fuel trans-shipment, and, this operation requires a specially lowered water level. Thus rising water level in the pool would not go un-noticed, and using the operator response times, as used in the other scenarios, is extremely conservative.

Due to the operators' presence, this evaluation assumes that the makeup would be isolated within 60 minutes after the SFP HI level alarm.

2.6 Boration Sources The normal source of borated water to the SFP is from the RWST. It is also possible to borate the SFP through the addition of dry boric acid directly to the SFP water. The boration sources are listed here for information only, and as such they will not be considered as boron dilution sources in this evaluation, because the boron concentration of these sources is higher than the boron concentration of the SFP.

S023 SFP BORON DILUTION ANALYSIS Page 16 of 35 2.6.1 Refueling Water Storage Tank(s)

There are two 245,000 gallon Refueling Water Storage Tanks (RWSTs) per unit,

$2(3)1204MT005, 006. Both tanks are cross-connected by a 24 inch cross-connecting pipe, resulting in the combined nominal volume of 490,000 gallons. The RWST(s) are connected to the SFP purification loop through a three inch feed line ($2(3)1219ML018) via the SFP Makeup Pump S2(3)1219MP011 and a four inch return line ($2(3)1219ML036). These connections are used as a flow path for makeup to the SFP from the RWST and also may be used to process the contents of the RWST through the purification filters and ion exchanger. Using the makeup flow path, the makeup pump can supply a makeup flow rate to the SFP of approximately 160 gpm.

Technical Specification 3.5.4 requires that the boron concentration in the RWST be maintained at least 2,350 ppm (in the range from 2,350 ppm to 2,800 ppm). The RWST boron is normally maintained Ž!2,650 ppm, for refueling purposes.

2.6.2 Boric Acid Makeup Tanks The contents of either BAMU tank ($2(3)1218MT071 or $2(3)1218MT72) can be directed to the RWST $2(3)1204MT006 by using a blending tee which mixes demineralized water with the borated water from the BAMU pumps S2(3)1218MP174 or S2(3)1218MP175 to a selected mix concentration. From the RWST, this fluid may be used to borate the SFP. To pass flow from the BAMU tanks to the RWST, a number of valves must be repositioned to utilize this non-standard lineup. To be in service (operable), LCS 3.1.104 requires the BAMU tanks to contain at least 4150 gallons of water with a concentration of greater than 4371 ppm boron.

2.6.3 Direct Addition of Boric Acid If necessary, the boron concentration of the SFP can be increased by emptying barrels of dry boric acid directly into the SFP. The dry boric acid will dissolve in the SFP water and will be mixed throughout the pool by the SFP cooling system flow and by the thermal convection created by the SFP decay heat.

2.7 Loss of Offsite Power (LOOP)

Of the dilution sources listed in Section 2.5, only the fire water and CCW piping are capable of providing non-borated water to the SFP during a loss of offsite power (LOOP), coincident with a postulated pipe break. This is because the fire water system is equipped with a diesel-driven fire pump (SA2301MP220), and, the CCW pumps are automatically loaded on the emergency diesel generators ($2(3)2420MG002/003). Plant annunciators, including the control room (such as for SFP level and temperature annunciation) are powered by 125 Volt DC non-lE power supply, so they should be operable following a LOOP.

A LOOP would also affect the ability to respond to a dilution event. The fuel pool purification pump $2(3)1219MP014 is not automatically loaded on the emergency diesel generators. Manual boron addition could be used if it became necessary to increase spent fuel boron concentration during a LOOP.

The SONGS Unit 2 and 3 SFP cooling pumps $2(3)1219MP009 and $2(3)1219MP010 are not automatically loaded onto the emergency diesel generators in the event of a LOOP.

S023 SFP .BORON DILUTION ANALYSIS Page 17 of 35 In conclusion, the only potential dilution sources after the LOOP will be the fire water and the CCW water, which are already addressed in Sections 2.5.6 and 2.5.3, and, the affects thereof are evaluated.

2.8 Piping There are no systems (other than those listed in Sections 2.5.6 and 2.5.7) identified which have piping in the vicinity of the SFP which could result in a dilution of the spent fuel pool if they were to fail.

Fire protection and nuclear service water, if damaged, could provide a source of SFP dilution.

However, the effects of these dilutions are bounded by the effects of primary water system makeup into the SFP, associated with postulated operator errors.

2.9 Spent Fuel Pool Instrumentation Instrumentation is available at SONGS Units 2 and 3 to monitor SFP water level and temperature. Additional instrumentation is available to monitor the status of each SFP cooling pump, the pump discharge line pressure and the upstream/downstream temperatures of the SFP heat exchangers. Local instrumentation is available to indicate the purification pump discharge pressure and temperature.

The instrumentation provided to monitor the SFP water level and temperature has a local indication and is annunciated in the control room. The SFP water level is maintained at a nominal elevation between 59'-6"(low) and 61'-4"(high). High level alarm (at 61'-4")

annunciates in the control room. Low level alarm (at 59'-6") is annunciated locally.

2.10 Administrative Controls The following administrative controls are in place to control and monitor the SFP boron concentration and water inventory:

A. The SFP boron concentration must be verified in accordance with LCS 3.7.116 at least every 30 days. This is done per chemistry procedure SO 123-111-1.1.23 on a weekly basis.

B. The SFP water level must be verified in accordance with LCS 3.7.117 at least every 7 days. This is done on a weekly basis by plant procedures per S023-3-3.27 and S023 3.27.1.

C. Plant procedure S023-3-2.11.1 requires sampling the SFP for boron concentration following makeup with a un-borated water source (e.g. step 2.4.7 in Attachment 1 of the procedure).

D. Administrative controls on the use of the primary makeup dilution paths are present.

Administrative controls are also present for the positioning of the valves in the lines connecting the RWST and SFP (see procedure S023-3-2.11.1).

S023 SFP BORON DILUTION ANALYSIS Page 18 of 35 Table 2-1: SFP Elevations Condition Water level Plant Elevation SFP operating deck N.A. 63.5 ft nominal level 43.5 ft 61 ft high level alarm 43.833 ft 61'-4" low level alarm 42 ft 59'-6" minimum level 38.4375 ft 55'-11 15/16" top of stored fuel 15.4375 ft 32'-11 15/16" assemblies Table 2-2: Dilution Summary Dilution Source Break Flow Max Dilution

1. Normal Primary Water System makeup from T-055/-056 NA 160 gpm

2. Primary Water System makeup from T-055/-056 I gpm 160 gpm

3. NSW Addition through (1) service hose NA 50 gpm

4. NSW pipe break near the SFP 30 gpm 30 gpm

5. Rupture of 1 tube in the SFP Heat Exchanger 90 gpm 90 gpm

6. Pipe Break in the Fire Water Header 110 gpm 160 gpm

7. Pipe Break in the SFP Cooling Water Return Header 130 gpm 160 gpm 7A. Same as case 7, but during fuel shipment 130 gpm 160 gpm

S023 SFP BORON DILUTION ANALYSIS Page 19 of 35 T/NEFLM 1, PPE Operating deck at EL M35 VFL 6I41

• LsELFl EL 61.3 EL39-5 1/2" (TYP.)

CASK (riot incd' in voume)

Bot. EL 16.6 Legedars+/--pladanteeaflon.

Note 1: The SFP vater level is initially assumed at the Low Level setpoint for exarrme calc and Table 3-1 Note 2 Cperations start 160 gprn deineralized water MUJ at the LL alarm level Note 3: DaTineralized vter MJ is assured to continue past -L setpint NWte 4: QOerator neglects to stop IVJ at I-L alarm Note 5: Bottom invert elevation of the overflow ppe Note 6:

• EL of er ater is a fution of flowrate (irdcated value is for example calc.& Table 3-1)

Note 7: Inclicated values are applicable for exaiple calculation and Table 3-1:

Ra.q 2-1: Schematic Diaqramn of the Spent Fuel Pool and Rant Bevations for exarrle cadc. & Table 3-1

S023 SFP BORON DILUTION ANALYSIS Page 20 of 35 3.0 SPENT FUEL POOL DILUTION EVALUATION 3.1 Calculation of Boron Dilution Times and Volumes This evaluation uses only the volume of the SFP (while the cask pool gate and transfer pool gate are normally open, they are assumed closed for conservatism), as this results in conservative boron concentration values after dilution as compared to the volumes with the cask pool and transfer pool connected with the SFP. The total pool volume at the start of the dilution is 349,931 gallons, which corresponds to the LL level in the SFP, when the cask pit is isolated. The low level represents the volume of the SFP filled to the low elevation of 59.5'. This value of pool volume is derived by using the procedure S023-3-2.11.1, which accounts for the water displaced by the fuel storage racks and the contained fuel assemblies.

For the purposes of identifying the dilution times and volumes, the initial SFP boron concentration is assumed to be at the proposed Technical Specification 3.7.17 limit of 2,000 ppm. This evaluation assumes thorough mixing of all non-borated water added to the SFP.

The time to dilute depends on the initial volume of the SFP and the postulated rate of dilution.

The dilution times and required volumes for the considered dilution events will be calculated based on the below methodology.

3.1.1 Boron Concentration Derivations When the Makeup Does Not Result in an Overflow For SFP water levels up to the SFP overflow, the boron concentration is calculated based on mix correlations, as follows:

Cfinal = (V1

• Cinit) / (V1 + dV) [ppm] ... (1) where:

VI ... Initial volume of SFP [gal]

dV ... Added volume of unborated makeup [gal]

Cinit ... Initial boron concentration in the SFP (relative to volume VI) [ppm]

Cfinal ... Final boron concentration after unborated makeup with volume dV [ppm]

3.1.2 Boron Concentration Derivations When the Makeup Results in an Overflow When the SFP overflows as a result of unborated makeup, this is considered to be a 'feed and bleed' process, with the SFP volume to remain essentially same. The water head required for the overflow is neglected in this evaluation, thus the volume at the overflow would be assumed to be equal to the SFP volume at the overflow level.

S023 SFP BORON DILUTION ANALYSIS Page 21 of 35 The rate of change of boron concentration in the SFP is then described by the following equation V

• dC/dt = -QC [ppm] ... (2) where:

V -SFP volume corresponding to overflow level (365,075 gallons for SFP only) [gal]

C -SFP boron concentration [ppm]

Q -Volumetric flowrate of unborated water [gpm] (assumed constant) t - Dilution time [min]

The solution of Equation 2 can be written as:

C(t) = Cinit

• e-t/tau [ppm] ... (3) where:

Cinit -Initial boron concentration [ppm]

C(t) -Boron concentration after t minutes [ppm]

tau = V/Q = boron dilution time constant [min]

In terms of the total added makeup, the Equation 3 can be expressed in terms of the initial SFP volume V, added makeup volume, say Vmakeup and initial and final concentrations, as follows:

Vmakeup = Q

• t thus, C(t) = Cinit

• e-(V/(vi/vakeup/t))

and, C(t) = Cinit

• e-(VmakeupN) [ppm] ... (4)

To calculate the time required to reach a specific boron value, the original Equation 3 can be re-written as:

t = ln(Cinit / C(t))

• V/Q or t =ln(Cinit / C(t))

• tau [min] ... (5) 3.1.3 Dilution Calculation Example The SFP elevations for this example (and for Table 3-1) are shown on Figure 2-1.

Determine SFP volumes in the range of concern (from LL to OVFL) per page 91 of S023-3-2.11.1:

In the range of LL-OVFL: Volume of 1 ft. of SFP = 7,572 gallons As an example, a dilution by unborated makeup at 160 gpm is evaluated. The operator is assumed to start makeup after the SFP low level alarm is annunciated in the control room. The low level setpoint is 59.5 ft. and the alarm is assumed to sound after a 0.25 inch band.

Thus the makeup is started when the SFP is at the following level:

59.5 - 0.25/12 = 59.479 ft.

S023 SFP BORON DILUTION ANALYSIS Page 22 of 35 The SFP volume corresponding to elevation 59.479 ft is 349,773 gallons.

The overflow elevation is 61'-5" ft. The hydraulic head associated with the overflow (per Reference 1) is:

Hofl = Q2 / 162,844 = 0.157 ft.

Thus the overflow elevation will be:

61.417 + .157 = 61.574 ft.

Adding 2.095 ft. of water from elevation of 59.479 ft. to 61.574 ft. represents adding 15,861 gallons of unborated water to the SFP. If the initial boron concentration was 2,000 ppm, the final concentration is calculated as follows:

VI = 349,773 gallons dV = 15,861 gallons Cinit = 2,000 ppm Cfinal = (VI

• Cinit) / (VI + dV) = 1,913 ppm The SFP volume corresponding to the elevation 61.574 ft is 365,634 gallons.

Now, adding 7,937 gallons of unborated water to the volume of 365,634 gallons with initial concentration of 1,913 ppm would result in following final concentration:

VI = 365,634 gallons dV = 7,937 gallons Cinit = 1,913 ppm Cfinal = Cinit

• e-(dV/Vl) = 1,872 ppm 3.2 Evaluation of Boron Dilution Events The SFP boron concentrations for bounding boron dilution events at SONGS Units 2 and 3 (Table 2-2), are derived by using the methodology described above in Section 3.1.

The dilution flowrates from Table 2-2 are used as an input in specific boron concentration calculations, which are documented in Tables 3-1 through 3-7A.

The minimum final boron concentration after dilution is 1700 ppm (Reference 2). Also, the gallons required to dilute to 970 ppm will be calculated to show the available discretionary margin.

The selected dilution flows are considered to be added (not simultaneously) until the time when they are isolated (either 60 minutes from SFP HI level alarm, or 90+ minutes from SFP overflow, based on Fuel Handling Building Sump HI-Hf level alarm).

S023 SFP BORON DILUTION ANALYSIS Page 23 of 35 The dilution calculations are performed in spreadsheet tables as summarized below:

Dilution Source (Table 2-2) Table

1. Normal Primary Water System makeup from T-055/-056 Table 3-1

2. Primary Water System makeup from T-055/-056 Table 3-2 after minor SFP outflow

3. NSW Addition through (1) service hose Table 3-3

4. NSW pipe break near the SFP Table 3-4

5. Rupture of I tube in the SFP Heat Exchanger Table 3-5

6. Pipe Break in the Fire Water Header Table 3-6

7. Pipe Break in the SFP Cooling Water Return Header Table 3-7 7A. Same as case 7, but during fuel shipment Table 3-7A 3.3 Summary of Boron Dilution Events The boron dilution results are summarized in Table 3-8.

It is concluded that an unplanned or inadvertent event which would result in the dilution of the SFP boron concentration below 1,700 ppm from an initial concentration of 2,000 ppm is a not a credible event.

The following discussion applies:

A. Boron dilution during normal primary make-up (flowrate = 160 gpm) would have to continue for 4.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the SFP HI level alarm is sounded in the control room for the boron concentration to be diluted from initial 2,000 ppm to 1,700 ppm (see Table 3-1). This would be adequate time for the operators to isolate the makeup.

B During normal primary makeup, in the unlikely event that the Operators neglect or do not get the SFP HI level alarm, the primary make-up flowrate of 160 gpm will continue and the water will reach an overflow level. After the SFP overflows, water will spill into the fuel handling building sump, which is equipped with a HI-HI alarm. This alarm will be triggered and annunciated in the control room about 4 minutes after the SFP starts to overflow. The alarm response procedure S023-15-56.C requires the operators to determine the cause of the sump HI-HI level. The operating staff have concluded that this activity can be accomplished within half an hour after the HI-HI alarm is sounded. Another I hour would be needed to isolate the inflow, thus the total time to isolate the in-leakage would be 94 minutes after the SFP starts to overflow. Based on the above, the inflow would be isolated when the boron concentration in the SFP reaches about 1,836 ppm.

S023 SFP BORON DILUTION ANALYSIS Page 24 of 35 C. The above discussions illustrate that there are two separate opportunities for the operators to isolate the SFP unborated inflow, based on two different alarms. The alarm associated with the SFP HI level is annunciated in the control room at panel 61, while the Fuel Handling Building Sump HI-HI level alarm is annunciated in the control room at panel 56. This alarm multiplicity prevents further long-term dilution of the SFP (longer than about 193 minutes during the above mentioned normal primary makeup per Table 3-1). This multiple alarm is applicable to all evaluated scenarios in this evaluation, because the SFP dilution means extended inflow of unborated water to the pool.

D. The normal makeup paths to the SFP from the primary makeup water system and the nuclear service water system are maintained closed.

E. In-place administrative controls on the primary letdown path from the SFP (return line to the RWST) ensure that any prolonged, inadvertent SFP makeup would result in pool overflow.

F. Besides documenting the required dilution volumes to dilute the SFP to 1,700 ppm boron, the following tables show for information the dilution volumes which would dilute the SFP to 970 ppm boron, to illustrate the dilution margin.

S023 SFP BORON DILUTION ANALYSIS Page 25 of 35 Table 3-1: SFP Boron Dilution Due To Primary Makeup Makeup Flow= 160 [gpm]

Makeup is started after SFP LL alarm sounds in the control room Makeup can be isolated 60 minutes after SFP HL alarm, or Makeup can be isolated 90 minutes after sump HI-HI alarm, which is, 94 [min] after SFP overflow SFP dilution (from LL level to >OFL) from Cinit= 2,000 [ppm]

Total Makeup Time after Time after Time after Final SFP Makeup Flow SFP SFP SFP boron Condition added LL alarm HL alarm overflow conc.

[Gal] [gpm] [min] [min] [min] [ppm]

SFP overflowing 264,100 160 1,711 1,562 1,551 970 SFP overflowing 58,980 160 429 280 269 1,700 SFP overflowing 30,929 160 253 105 94 1,836 SFP overflowing 23,858 160 209 60 50 1,872 SFP overflowing 15,861 160 159 10 0 1,913 OFL pipe level 14,671 160 152 3 1,919 HL alarm 14,198 160 149 0 1,922 HL setpoint 14,040 160 148 1,923 LL setpoint 158 160 61 1,999 LL alarm 0 160 60 2,000 OP starts MU 0 160 60 2,000 Note: It would take 264,100 gallons of unborated water to dilute the SFP to 970 ppm I

S023 SFP BORON DILUTION ANALYSIS Page 26 of 35 Table 3-2: SFP Boron Dilution Due To Primarv Makeup)

After Minor Leak Leak Flow= 1 [gpm]

Makeup Flow= 160 [gpm]

Leakage results in SFP level reduction below LL alarm, 60 minutes after LL alarm, makeup is started at a full makeup flow of 160 gpm.

Makeup can be isolated 60 minutes after SFP HL alarm, or Makeup can be isolated 90 minutes after sump HI-HI alarm, which is, 94 [min] after SFP overflow SFP dilution (from LL level to >OFLI from Cinit= 2.000 [onmi Total Makeup Time after Time after Time after Final SFP Makeup Flow SFP SFP SFP boron Condition added LL alarm HL alarm overflow conc.

[Gal] [gpm] [rin] [nin] [min] [ppm]

SFP overflowing 264,100 160 1,711 1,562 1,551 970 SFP overflowing 58,980 160 429 280 269 1,700 SFP overflowing 30,929 160 253 104 94 1,836 SFP overflowing 23,858 160 209 60 50 1,872 SFP overflowing 15,921 160 160 10 0 1,913 OFL pipe level 14,731 160 152 3 1,919 HL alarm 14,258 160 149 0 1,922 HL setpoint 14,100 160 148 1,922 LL setpoint 218 160 61 1,999 LL alarm 60 160 60 2,000 OP starts MU 0 160 60 2,000 Note: Itwould take 264,100 gallons of unborated waterto dilute the SFP to 970 ppm boron.

S023 SFP BORON DILUTION ANALYSIS Page 27 of 35 Table 3-3: SFP Boron Dilution Due To One Flowinc NSW Hose Hose Flow= 50 [gpm]

Inflow results in SFP level increase from LL Inflow can be isolated 60 minutes after SFP HL alarm, or Inflow can be isolated 90 minutes after sump HI-HI alarm, which is, 102 [min] after SFP overflow SFP dilution (from LL level to >OFL) from Cinit= 2,000 [ppm]

Total SFP Time after Time after Time after Final SFP Inflow Inflow SFP SFP SFP boron Condition added LL alarm HL alarm overflow conc.

[Gal] [gpm] [rin] [min] [nin] [ppm]

SFP overflowing 263,400 50 5,268 4,984 4,972 970 SFP overflowing 59,000 50 1,180 896 884 1,700 SFP overflowing 19,887 50 398 114 102 1,892 SFP overflowing 17,198 50 344 60 48 1,906 SFP overflowing 14,787 50 296 12 0 1,919 OFL pipe level 14,671 50 293 9 1,919 HL alarm 14,198 50 284 0 1,922 HL setpoint 14,040 50 281 1,923 LL setpoint 158 50 3 1,999 LL alarm 0 50 0 2,000 Inflow starts 0 50 0 2,000 Note: Itwould take 263,400 gallons of unborated water to dilute the SFP to 970 ppm boron.

S023 SFP BORON DILUTION ANALYSIS Page 28 of 35 Table 3-4: SFP Boron Dilution Due To Cracked NSW Header Break Flow= 30 [gpm]

Inflow results in SFP level increase from LL Inflow can be isolated 60 minutes after SFP HL alarm, or

'Inflow can be isolated 90 minutes after sump HI-HI alarm, which is, 110 [min] after SFP overflow SFP dilution lfrom LL level to >OFL* from Cinit= 2.000 [oomi

-F diuto fro LL lee t-LI >OF fro Ciit 2 000- LEE -

Total SFP Time after Time after Time after Final SFP Inflow Inflow SFP SFP SFP boron Condition added LL alarm HL alarm overflow conc.

[Gal] [gpm] [min] [min] [min] [ppm]

SFP overflowing 263,300 30 8,777 8,303 8,286 970 SFP overflowing 59,000 30 1,967 1,493 1,476 1,700 SFP overflowing 18,013 30 600 127 110 1,902 SFP overflowing 15,998 30 533 60 43 1,913 OPL* 14,713 30 490 17 0 1,919 ORL pipe level 14,671 30 489 16 1,919 HL alarm 14,198 30 473 0 1,922 HL setpoint 14,040 30 468 1,923 LL setpoint 158 30 5 1,999 LL alarm 0 30 0 2,000 Inflow starts 0 30 0 2,000 Note: Itwould take 263,300 gallons of unborated water to dilute the SFP to 970 ppm boron.

S023 SFP BORON DILUTION ANALYSIS Page 29 of 35 Table 3-5: SFP Boron Dilution Due To One Failed SFP H-X Tube Break Flow= 90 [gpm]

Inflow results in SFP level increase from LL Inflow can be isolated 60 minutes after SFP HL alarm, or Inflow can be isolated 90 minutes after sump HI-HI alarm, which is, 97 [min] after SFP overflow SFP dilution (from LL level to >OFL) from Cinit= 2,000 [ppm]

Total SFP Time after Time after Time after Final SFP Inflow Inflow SFP SFP SFP boron Condition added LL alarm HL alarm overflow conc.

[Gal] [gpm] [rin] [min] [min] [ppm]

SFP overflowing 263,500 90 2,928 2,770 2,761 970 SFP overflowing 59,000 90 656 498 488 1,700 SFP overflowing 23,750 90 264 106 97 1,872 SFP overflowing 19,598 90 218 60 51 1,894 SFP overflowing 15,047 90 167 9 0 1,918 OFL pipe level 14,671 90 163 5 1,919 HL alarm 14,198 90 158 0 1,922 HL setpoint 14,040 90 156 1,923 LL setpoint 158 90 2 1,999 LL alarm 0 90 0 2,000 Inflow starts 0 90 0 2,000 Note: It would take 263,500 gallons of unborated water to dilute the SFP to 970 ppm boron.

S023 SFP BORON DILUTION ANALYSIS Page 30 of 35 Table 3-6: SFP Boron Dilution Due To Cracked Fire Water Header Break Flow= 110 [gpm]

Inflow results in SFP level increase from LL Inflow can be isolated 60 minutes after SFP HL alarm, or Inflow can be isolated 90 minutes after sump HI-HI alarm, which is, 95 [min] after SFP overflow SFP dilution from LL level to >OFL) from Cinit= 2,000 [ppm]

Total SFP Time after Time after Time after Final SFP Inflow Inflow SFP SFP SFP boron Condition added LL alarm HL alarm overflow conc.

,[Gall [gpm] [min] [min] [min] [ppm]

SFP overflowing 263,700 110 2,397 2,268 2,259 970 SFP overflowing 59,000 110 536 407 398 1,700 SFP overflowing 25,738 110 234 105 95 1,862 SFP overflowing 20,797 110 189 60 51 1,888 SFP overflowing 15,233 110 138 9 0 1,917 OFR pipe level 14,671 110 133 4 1,919 HL alarm 14,198 110 129 0 1,922 HL setpoint 14,040 110 128 1,923 LL setpoint 158 110 1 1,999 LL alarm 0 110 0 2,000 Inflow starts 0 110 0 2,000 Note: Itwould take 263,700 gallons of unborated water to dilute the SFP to 970 ppm boron.

S023 SFP BORON DILUTION ANALYSIS Page 31 of 35 Table 3-7: SFP Boron Dilution Due To Cracked SFP Cooling Header Break Flow= 130 [gpm]

Makeup Flow= 160 [gpm]

Break flow results in SFP level reduction below LL alarm, 60 minutes after LL alarm, makeup is started at a full makeup flow of 160 gpm.

Makeup can be isolated 60 minutes after SFP HL alarm, or Makeup can be isolated 90 minutes after sump HI-HI alarm, which is, 94 [min] after SFP overflow SFP dilution (from LL level to >OFL) from Cinit= 2,000 [ppm]

Total Makeup Time after Time after Time after Final SFP Makeup Flow SFP SFP SFP boron Condition added LL alarm HL alarm overflow conc.

[Gal] [gpm] [min] [nin] [rin] [ppm]

SFP overflowing 263,600 160 1,708 1,510 1,500 970 SFP overflowing 58,700 160 427 229 219 1,700 SFP overflowing 38,669 160 302 104 94 1,795 SFP overflowing 31,598 160 257 60 50 1,830 SFP overflowing 23,661 160 208 10 0 1,871 OFL pipe level 22,471 160 200 3 1,877 HL alarm 21,998 160 197 0 1,879 HL setpoint 21,840 160 196 1,880 LL setpoint 7,958 160 110 1,955 LL alarm 7,800 160 109 1,955 OP starts MU 0 160 60 2,000 Note: Itwould take 263,600 gallons of unborated water to dilute the SFP to 970 ppm boron.

S023 SFP BORON DILUTION ANALYSIS Page 32 of 35 Table 3-7A: SFP Boron Dilution Due To Cracked SFP Cooling Header Durina Fuel Shipment Break Flow= 130 [gpm]

Makeup Flow= 160 [gpm]

Break flow results in SFP level reduction below LL alarm, (The LL alarm during fuel trans-shipment is lowered below normal setpoint) 60 minutes after LL alarm, makeup is started ata full makeup flow of 160 gpm.

Due to operators presence, M-U can be isolated 60 minutes after SFP HL alarm.

SFP dilution (from LL level to >OFL) from Cinit= 2,000 [ppm]

Total Makeup Time after Time after Time after Final SFP Makeup Flow SFP SFP SFP boron Condition added LL alarm HL alarm overflow conc.

[Gal] [gpm] [min] [rnin] [min] [ppm]

SFP overflowing (2) 262,400 160 1,700 1,416 1,405 970 SFP overflowing 57,300 160 418 134 123 1,700 SFP overflowing 52,600 160 389 105 (1) 94 1,722 SFP overflowing 45,480 160 344 60 50 1,756 SFP overflowing 37,541 160 295 10 0 1,795 OFL pipe level 36,350 160 287 3 1,801 HL alarrn 35,877 160 284 0 1,803 HL setpoint 35,719 160 283 1,804 LL setpoint 7,958 160 110 1,953 LL alarm 7,800 160 109 1,954 OP starts MU 0 160 60 2,000 Notes:

(1) Data for information only, if M-U isolated 90 minutes after sump HI-HI alarm, which is: 94 [min] after SFP overflow (2) It would take 262,400 gallons of unborated waterto dilute the SFP to 970 ppm boron.

S023 SFP BORON DILUTION ANALYSIS Page 33 of 35 Table 3-8: SFP Boron Dilution Summary Normal Primary SFP Makeup 3-1 N/A 160 60 94 1,872 1,836 Prrary SFP MU after rrinor leak 3-2 1 160 60 94 1,872 1,836 Row from 1 NSW Hose 3-3 50 50 60 102 1,906 1,892 Cracked NSW Header 3-4 30 30 60 110 1,913 1,902 Failed SFP H-X Tube 3-5 90 90 60 97 1,894 1,872 Cracked Fire Water Header 3-6 110 110 60 95 1,888 1,862 Cracked SFP Cooling Header 3-7 130 160 60 94 1,830 1,795 Case per 3-7 w/ fuel shipment 3-7A 130 160 60 94 1,756 1,722

S023 SFP BORON DILUTION ANALYSIS Page 34 of 35

4.0 CONCLUSION

S A boron dilution analysis has been completed for the spent fuel pool. As a result of this spent fuel pool boron dilution analysis, it is concluded that an unplanned or inadvertent event which would result in the dilution of the spent fuel pool boron concentration from 2,000 ppm to 1,700 ppm is not a credible event.

An operator would have to initiate dilution flow, then abandon monitoring of pool level, ignore tagged valves, violate administrative procedures, and ignore spent fuel pool and building sump level alarms.

A spent fuel pool dilution event would be readily detected by plant personnel via alarms, flooding in the fuel handling building, or by normal operator rounds through the spent fuel pool area.

It should be noted that this boron dilution analysis was conducted by evaluating the time and water volumes required to dilute the spent fuel pool from 2,000 ppm to 1,700 ppm. Under normal, non-accident conditions, only 970 ppm is required to keep Keff less than 0.95. This is a discretionary margin of 730 ppm. As shown in Tables 3-1 through 3-7A, a minimum of 262,400 gallons of unborated water would have to be added to dilute from 2,000 ppm to 970 ppm. Plant instrumentation and administrative procedures are in place to prevent the inadvertent dilution of this magnitude.

Finally, the criticality analyses show that on a 95/95 basis the spent fuel rack Keff remains less than 1.0 with non-borated water in the pool. Thus, even if the spent fuel pool were diluted to zero ppm, the spent fuel would remain subcritical and the health and safety of the public would be assured.

S023 SFP BORON DILUTION ANALYSIS Page 35 of 35

5.0 REFERENCES

1. M-0022-019, Rev. 0, SFP Boron Dilution Analysis

2. "Spent Fuel Pool Criticality Analysis (With No Boraflex And Credit For Soluble Boron)",

Southern California Edison Company, San Onofre Nuclear Generating Station, Units 2 And 3, Revision 0, November 2001.

PCN 556 Attachment L (Spent Fuel Pool Criticality Analysis)

SPENT FUEL POOL CRITICALITY ANALYSIS (WITH NO BORAFLEX AND CREDIT FOR SOLUBLE BORON)

SOUTHERN CALIFORNIA EDISON SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 REVISION 2 APRIL 2007

S023 SFP CRITICALITY ANALYSIS Page 2 of 121 TABLE OF CONTENTS Page EXECUTIVE

SUMMARY

.............................................. 3

1. INTR O D U CTIO N .......................................................... 4

2. FUEL STORAGE DESCRIPTION ............................................. 5 2.1 FUEL ASSEMBLY DESCRIPTIONS ..................................... 5 2.2 GUIDE TUBE INSERTS ................................................ 5 2.3 SPENT FUEL STORAGE RACK DESCRIPTION ........................... 6

3. COMPUTER PROGRAMS AND METHODOLOGY ............................. 13 3.1 COM PUTER PROGRAM S ............................................. 13 3.2 METHODOLOGY .................................................... 16

4. CRITICALITY SAFETY ANALYSES ......................................... 29 4.1 MANUFACTURING TOLERANCES AND POOL TEMPERATURE BIAS ...... 29 4.2 ECCENTRIC PLACEMENT BIAS ...................................... 29 4.3 CONTROL ELEMENT ASSEMBLY (CEA) BIAS .......................... 30 4.4 AXIAL BURNUP BIAS ............................................... 30 4.5 SONGS UNITS 2 AND 3 FUEL ASSEMBLIES ............................ 30 4.6 SONGS UNIT 1 FUEL ASSEMBLIES .................................... 32 4.7 INTER-MODULE SPACING ........................................... 32 4.8 RECONSTITUTION STATION ......................................... 33 4.9 FAILED FUEL ROD STORAGE BASKET ................................ 33 4.10 FUEL HANDLING EQUIPMENT ....................................... 33 4.11 NON-FUEL COMPONENTS ........................................... 33

5. SOLUBLE BORON REQUIREMENTS ........................................ 91 5.1 Keff LESS THAN OR EQUAL TO 0.95 ................................... 91 5.2 REACTIVITY EQUIVALENCING UNCERTAINTY ........................ 91 5.3 DISCHARGE BURNUP UNCERTAINTY ................................ 92 5.4 SOLUBLE BORON MEASUREMENT UNCERTAINTY .................... 92 5.5 MARGIN FOR FUTURE REQUIREMENTS .............................. 92 5.6 ACCIDENT CONDITIONS ............................................ 92

6. REFEREN CES ........................................................... 94 APPENDIX A (SPENT FUEL RACK DIAGRAMS) .............................. 96 APPENDIX B (REGION I KENO-V.a MODEL) ............................... 100 APPENDIX C (REGION II KENO-V.a MODEL) ............................... 109 APPENDIX D (

SUMMARY

OF TOLERANCES, BIASES, AND UNCERTAINTIES .. 119

S023 SFP CRITICALITY ANALYSIS Page 3 of 121 EXECUTIVE

SUMMARY

This report describes the criticality analyses performed for San Onofre Nuclear Generating Station (SONGS) Units 2 and 3, Facility Operating Licenses NPF-10 and NPF-15, respectively.

The results of the criticality analyses show that the existing spent fuel storage racks with no Boraflex, and supporting systems and components, have been adequately designed to accommodate the storage and handling of SONGS Units 2 and 3 fuel with a maximum fuel pin enrichment of 4.8 weight percent (w/o). For all normal and postulated accident conditions (with the exception of boron dilution) in the spent fuel pool, a minimum concentration of 1,700 ppm soluble boron is required. A soluble boron level of 2,000 ppm in the spent fuel pool is required for all postulated accident conditions and a concurrent boron dilution event.

To compensate for no Boraflex, SONGS Units 2 and 3 will use the following storage patterns and guide tube inserts as needed:

(1) unrestricted storage, minimum discharge burnup and cooling time requirements vs initial enrichment, (2) SFP Peripheral storage, minimum discharge burnup and cooling time requirements vs initial enrichment, (3) 2x2 storage patterns, minimum discharge burnup and cooling time requirements vs initial enrichment, (4) 3x3 storage patterns, minimum discharge burnup and cooling time requirements vs initial enrichment, (5) credit for inserted Control Element Assemblies (CEAs)

(6) credit for erbia in fresh assemblies, (7) credit for cooling time (Pu-241 decay), and, (8) credit for borated stainless steel guide tube inserts.

The criticality analyses also show that San Onofre Nuclear Generating Station (SONGS) Unit 1 fuel assemblies can be safely stored in the SONGS Units 2 and 3 spent fuel storage racks with no Boraflex. The maximum fresh enrichment of the SONGS Unit 1 assemblies is 4.0 w/o with an uncertainty of 0.05 w/o.

Edison is not asking NRC approval of a new methodology. The analyses documented herein use methodologies and computer programs previously reviewed and approved by the NRC. In particular, credit for soluble boron in the spent fuel pool water is taken. This means that the acceptance criteria for the spent fuel storage racks are Keff less than 1.0 with unborated water, and Keff less than or equal to 0.95 with borated water.

On the basis of the information and evaluations presented in this report, Edison concludes that the proposed changes (no Boraflex and credit for soluble boron) in fuel storage for the SONGS Units 2 and 3 spent fuel storage facilities will provide safe fuel storage and are in conformance with NRC requirements. The changes will have no significant impact on the health and safety of the general public.

S023 SFP CRITICALITY ANALYSIS Page 4 of 121

1. INTRODUCTION The criticality analyses documented in this report show that the SONGS Units 2 and 3 spent fuel storage racks meet the NRC's acceptance criteria for criticality (1) assuming no Boraflex, and (2) taking credit for 1,700 ppm soluble boron in the spent fuel pool water.

The NRC is not asked to approve a new methodology. The methodology and computer programs employed herein have been previously reviewed and approved by the NRC. (3,4,.6,7 )

When credit for soluble boron is taken, the acceptance criteria are:(5)

(1) Under normal conditions, the 95/95 neutron multiplication factor (Keff), including all uncertainties, shall be less than 1.0 when flooded with unborated water.

(2) Under normal and accident conditions, the 95/95 neutron multiplication factor (Keff),

including all uncertainties, shall be less than or equal to 0.95 when flooded with borated water.

To compensate for no Boraflex, SONGS Units 2 and 3 will use the following storage patterns and guide tube inserts as needed:

(1) unrestricted storage, minimum discharge burnup and cooling time requirements vs initial enrichment, (2) SFP Peripheral storage, minimum discharge burnup and cooling time requirements vs initial enrichment, (3) 2x2 storage patterns, minimum discharge burnup and cooling time requirements vs initial enrichment, (4) 3x3 storage patterns, minimum discharge burnup and cooling time requirements vs initial enrichment, (5) credit for inserted Control Element Assemblies (CEAs)

(6) credit for erbia in fresh assemblies, (7) credit for cooling time (Pu-241 decay), and, (8) credit for borated stainless steel guide tube inserts.

Boraflex erosion/dissolution is an industry problem, and SONGS Units 2 and 3 are affected. Silica levels in the SONGS Units 2 and 3 spent fuel pools are increasing, and this indicates the Boraflex is eroding/dissolving. Although there is currently sufficient Boraflex, it is prudent to plan for the long term. Taking no credit for Boraflex for SONGS Units 2 and 3 will totally eliminate any Boraflex concerns in the future, and monitoring programs will not be required to ensure that an adequate amount of Boraflex is present.

S023 SFP CRITICALITY ANALYSIS Page 5 of 121

2. FUEL STORAGE DESCRIPTION This section presents a description of the SONGS Units 2 and 3 spent fuel storage racks, the SONGS Units 1, 2, and 3 fuel assemblies licensed for storage in the SONGS Units 2 and 3 spent fuel storage racks, and guide tube inserts which will be used as needed.

2.1 FUEL ASSEMBLY DESCRIPTIONS Two fuel assembly designs are currently licensed for storage in the SONGS Units 2 and 3 fuel storage racks (UFSAR Sections 4.2 and 9.1.2):

(1) Westinghouse-Combustion Engineering (W/CE), Zircaloy-clad, 16x16 Fuel Assemblies, 4.8 w/o maximum enrichment Note: Westinghouse-Combustion Engineering was formerly ABB-Combustion Engineering was formerly Combustion Engineering.

(2) Westinghouse, Stainless-steel-clad, 14x14 Fuel Assemblies transhipped from Unit 1, 4.0 w/o maximum nominal enrichment The characteristics of the W/CE SONGS Units 2 and 3 and Westinghouse SONGS Unit 1 fuel assembly designs are given in Table 2-1.

2.2 GUIDE TUBE INSERTS 2.2.1 Control Element Assemblies (CEAs)

Full length, 5-finger CEAs may be used in W/CE fuel assemblies stored in Regions I and 11. Use of CEAs allows flexibility in fuel assembly placement and greater utilization of the spent fuel pool by lowering fuel assembly reactivity, which is expressed as a lower required discharge burnup. The characteristics of the CEAs assumed in the criticality analyses documented herein are described in Section 4.2 of the UFSAR. No CEA insert will be used in the Westinghouse SONGS Unit 1 fuel assemblies.

The control elements of a full- length CEA consist of an Inconel 625 tube loaded with a stack of cylindrical absorber pellets. The absorber material is boron carbide (B4C), with the exception of the lower portion which is silver-indium-cadmium (Ag-In-Cd) alloy cylinders and Inconel end plug.

A CEA lifetime analysis will ensure that the CEAs used are suitable for use in the spent fuel pools.

Before using any CEA, a visual inspection will be performed.

S023 SFP CRITICALITY ANALYSIS Page 6 of 121 2.2.2 Borated Stainless Steel (SS) Guide Tube Inserts Three or five borated SS guide tube inserts may be used in fuel assemblies stored in Region HI.Use of inserts allows flexibility in fuel assembly placement and greater utilization of the spent fuel pool by lowering fuel assembly reactivity, which is expressed as a lower required discharge burnup.

The characteristics assumed for the inserts in the criticality analyses documented herein are described below.

The inserts are 0.75 inches O.D. minimum, and must cover the entire active fuel length of 150.0 inches. The inserts must have a minimum boron loading of:

3 0.02434 grams of B-10 per cm When 3 guide tube inserts are used, the orientation shall be the same in every assembly in the spent fuel pool (Figure 3-3).

2.3 SPENT FUEL STORAGE RACK DESCRIPTION The spent fuel storage racks (12) provide for storage of new and spent fuel assemblies in the spent fuel pool, while maintaining a coolable geometry, preventing criticality, and protecting the fuel assemblies from excess mechanical or thermal loadings. SONGS Units 1, 2, and 3 fuel may be stored in the SONGS Units 2 and 3 racks, as well as miscellaneous storage items (e.g., trash baskets, dummy fuel assemblies, neutron sources), and the failed rod storage baskets.

Fuel is stored in two regions within each pool (Table 2-2, Figure 2-1):

(1) Region I (312 locations)

(2) Region HI(1230 locations)

As originally installed and currently licensed, both regions use Boraflex, a neutron absorbing material. Boraflex consists of fine boron carbide particles distributed in a polymeric silicone encapsulant.

The criticality analyses documented herein assume no Boraflex. Conservatively, it is assumed that the Boraflex has completely dissolved/eroded, and the pocket is filled with spent fuel pool water.

S023 SFP CRITICALITY ANALYSIS Page 7 of 121 2.3.1 Materials The Region I and Region II racks are constructed from Type 304LN stainless steel except the leveling screws which are SA-564 Type 630 stainless steel and some leveling pads which are either SA-182 Type F-304 stainless steel or SA-240 (or SA-479) Type 304 stainless steel. The floor plates under the rack support pads are made from SA-240 Type 304 stainless steel, which has the same corrosion resistance characteristics as the rack materials.

The Region I and Region HIracks are neither anchored to the floor nor braced to the pool walls or each other. Also, the pool floor plates are not attached to the pool floor.

2.3.2 Region I Spent Fuel Storage Rack Description Region 1 (312 locations) consists of two high density fuel racks, each with 12x13 cells. The nominal dimensions of each rack are 125.5 inches by 135.9 inches. The cells within a rack are interconnected by grid assemblies and stiffener clips to form an integral structure as shown in Figure 2-2. The cells in Region I are separated from each other by a minimum water gap of about 1.1 inches.

Region I is generally reserved for temporary storage of new fuel or partially irradiated fuel which would not qualify for Region II storage.

2.3.3 Region II Spent Fuel Storage Rack Description Region II (1230 locations) has six high density fuel racks, four with 14x15 cells and two with 13x15 cells and provides normal storage for spent fuel assemblies. The nominal dimensions of the 14x15 rack are 124.82 inches by 133.67 inches; the nominal dimensions of the 13x15 rack are 115.97 inches by 133.67 inches. The cells in Region II do not have a water gap.

The six Region 1I storage racks consist of stainless steel cells assembled in a checkerboard pattern, producing the structure shown in Figure 2-3. Cells are located in every other location and are welded together at the cell corners. This results in "non-cell" storage locations, each one formed by one outside wall of four checkerboard cells.

Region II is generally used for long term storage of permanently discharged fuel that has achieved qualifying burnup levels.

S023 SFP CRITICALITY ANALYSIS Page 8 of 121 Table 2-1 FUEL ASSEMBLY DATA FOR SONGS UNITS 1, 2, AND 3 SONGS 1 SONGS2&3 Maximum Fuel Pin Enrichment (w/o) 4.0 4.8 Cladding Type SS Zr Rod Array 14x14 16x16 Fuel Rod Pitch (in.)* 0.556 0.506 Number of Rods Per Assembly 180 236 Fuel Rod Outer Diameter (in.) 0.422 0.382 Fuel Pellet Diameter (in.) 0.3835 0.3250- 0.3255 Active Fuel Length (in.) 120.0 150.0 Cladding Thickness (in.) 0.0165 0.025 Number of Guide Tubes 16 5 Guide Tube Outer Diameter (in.) 0.535 0.980 Guide Tube Inner Diameter (in.) 0.511 0.900 Guide Tube Material SS Zr

• Fuel rod pitch is the spacing between fuel rods measured as the distance from centerline to centerline of the rod. Both assembly types are square pitch arrays.

S023 SFP CRITICALITY ANALYSIS Page 9 of 121 Table 2-2 SPENT FUEL RACK DATA (Each Unit)

Region I Region II Number of Storage 312 1230 Locations Number of Rack Two 12 x 13 Four 14 x 15 Anays Two 13 x 15 Center-to-Center 10.40 8.85 Spacing (inches)

Cell Inside Width 8.64 8.63 (inches)

Type of Fuel SONGS 2 and 3 16 x 16 SONGS 2 and 3 16 x 16 and/or and/or SONGS 1 14 x 14* SONGS 1 14 x 14 Rack Assembly Outline 126 x 136 x 198.5 125 x 134 x 198.5 Dimensions (inches) (14 x 15) 116 x 134 x 198.5 (13 x 15)

• SONGS 1 fuel is included under Region 1 for historical reasons. SONGS 1 fuel may only be stored in Region II. flV

S023 SFP CRITICALITY ANALYSIS Page 10 of 121 Figure 2-1 SAN ONOFRE UNITS 2 AND 3 SPENT FUEL POOL LAYOUT (NOMINAL DIMENSIONS in INCHES)

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S023 SFP CRITICALITY ANALYSIS Page II of 121 Figure 2-2 SAN ONOFRE UNITS 2 AND 3 REGION I SPENT FUEL STORAGE CELLS YAS CWA1)

S023 SFP CRITICALITY ANALYSIS Page 12 of 121 Figure 2-3 SAN ONOFRE UNITS 2 AND 3 REGION II SPENT FUEL STORAGE CELLS 8.630 RAL f. Square can i=nsij I

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S023 SFP CRITICALITY ANALYSIS Page 13 of 121

3. COMPUTER PROGRAMS AND METHODOLOGY This section describes the computer programs and methodology used for the criticality analyses of the SONGS Units 2 and 3 spent fuel storage racks.

The NRC is not asked to approve a new methodology.

3.1 COMPUTER PROGRAMS 3.1.1 Computer Program Descriptions CELLDAN, NITAWL-ll, KENO-V.a, CASMO-3, and SIMULATE-3 are the computer programs used in the analyses. (10.11.12)

CELLDAN, NITAWL-ll, KENO-V.a, and CASMO-3 have been used in previous SCE spent fuel pool criticality analyses approved by the NRC.(7 )

SCE has NRC approval('7 ) to use CASMO-3 and SIMULATE-3 for reactor physics analyses.

CELLDAN calculates the atoms/barn-cm of U235, U238, and Oxygen in the U0 2 fuel. CELLDAN also calculates the atoms/barn-cm of Hydrogen, Oxygen, B-10, and B-il in the water. Finally, CELLDAN calculates the Dancoff factor, and U235 and Oxygen scattering cross-sections per U238 atom for NITAWL-II.

NITAWL-II generates a binary cross-section library for KENO-V.a. The library contains 27 group cross-section data for every nuclide in the KENO-V.a problem. Using the U238 number density, Dancoff factor, and U235/Oxygen scattering cross-sections per U238 atom from CELLDAN, NITAWL-ll uses the Nordheim Method to do resonance shielding of the U238 cross-section.

KENO-V.a is the nuclear industry standard program for criticality analyses. KENO-V.a is a three-dimensional, multi-group, Monte Carlo program.

CASMO-3 is a multi-group two-dimensional transport theory program for calculations on BWR and PWR fuel assemblies. It has been extensively used by utilities in the U.S. In these analyses, CASMO-3 is used for three purposes. First, CASMO-3 is used to evaluate the reactivity variations (Delta-k) due to the rack manufacturing tolerances and normal pool temperature variation. Second, CASMO-3 is used to generate the initial enrichment versus discharge burnup criteria. Thirdly, CASMO-3 is used to evaluate the pool heat up accident, and determine the required soluble boron concentration for the fuel mishandling accident.

S023 SFP CRITICALITY ANALYSIS Page 14 of 121 SIMULATE-3 is the computer program used by SCE to model the San Onofre reactor cores to calculate power distribution, Boron letdown curve, MTC, rod worths, etc. In this analysis, SIMULATE-3 is used to evaluate the axial burnup bias.

3.1.2 Computer Program Benchmarking KENO-V.a has been benchmarked by SCE against industry standard critical experiments performed by B&W .13.14) The bias and 95/95 uncertainty in the bias for CELLDAN, NITAWL-II, and KENO-V.a and the 27 group cross-section library is 0.008.14 + 0.00172.

(Table 3-1)

The criticality experiments examined have similar nuclear characteristics to spent fuel storage and are applicable to conditions encountered during the handling of LWR fuel outside reactors. This B&W critical experiment case set has been previously used by SCE for spent fuel pool criticality analyses reviewed and approved by the NRC.(7 ) The number of benchmarking cases used by SCE (16 cases) compared to Reference 4 (32 cases) is conservative and results in a higher bias for SCE's analyses.

15 6 Since the enrichment in the B&W critical experiments was 2.46 w/o, additional comparisons( ' )

were done at 4.30 w/o and 4.75 w/o. The bias and uncertainty determined at 2.46 w/o was found to be applicable for these higher enrichments.

CASMO-3 and SIMULATE-3 have been validated by accurately predicting SONGS Units 2 and 3 startup physics test data and core follow results.(17) CASMO-3 U and Pu isotopic predictions agree quite well with measurements for all measured isotopes (Yankee Core) throughout the burnup range. CASMO-3 has also been compared to industry critical experiments and other measured data*

with good agreement.

• 28 Ro = Ratio of epithermal to thermal U2 38 capture rates 25 D*elt = Ratio of epithermal to thermal U 2 35 fission rates 28 De*lt, = Ratio of U 238 to U 235 fission rates CR = Ratio of U238 capture to U235 fission rates (Conversion Ratio)

S023 SFP CRITICALITY ANALYSIS Page 15 of 121 Table 3-1 KENO-V.a Analyses Of Critical Experiments For The Determination Of Calculational Bias And Uncertainty B&W Core Measured k-eff KENO-V.a k-eff Difference I 1 .0002 0.99037 0 .00983 ZI 1.0001 0.99278 0 .00732 ZII 1.0000 0.99346 0.00654 IX 1.0030 0.99044 0.01256 X 1.0001 0.99039 0.00971 XI 1.0000 0.99441 0 .00559 XII 1.0000 0.99362 0.00638 XIII 1.0000 0.99868 0 .00132 XIV 1.0001 0.99382 0.00628 XV 0.9998 0.98833 0 .01147 XVI 1. 0001 0. 98892 0 .01118 XVII 1.0000 0.99234 0.00766 XVIII 1. 0002 0.99119 0.00901 XIX 1 .0002 0.99146 0 .00874 XX 1. 0003 0.99085 0 .00945 XXI 0.9997 0.99256 0.00714 Mean = 0.00814 Standard Deviation = 0.00273 Bias + Uncertainty = Mean + k95/95

• Standard Deviation SQRT(Number Of Cases)

= 0.00814 + (2.524)(0.00273)/SQRT(16)

= 0.00814 + 0.00172

S023 SFP CRITICALITY ANALYSIS Page 16 of 121 3.2 METHODOLOGY The methodology used for this analysis is consistent with methods previously reviewed and approved by the NRC. O3 4' 56A7 The NRC is not asked to approve a new methodology.

The methodology of this analysis follows NRC guidance and the Westinghouse Owners Group (WOG) methodology for spent fuel storage rack criticality analyses. This methodology includes credit for soluble boron and burnable poison integral with the fuel rods. The WOG methodology has been found non-conservative for the axial burnup bias.(6) In this analysis, SCE evaluates a SONGS specific axial burnup bias.

Reference 7 is NRC approval of SCE's methodology for fuel rack criticality analyses using KENO-V.a and CASMO-3. The SCE methodology does not include soluble boron or burnable poison credit. Otherwise, the SCE and WOG methodologies are essentially the same (except for axial burnup bias) and follow well established industry guidance. 8

• Reference 9 documents that an equivalent fresh enrichment determined at 0 ppm is under-estimated if used in a KENO-V.a model which includes dissolved boron. SCE's analyses ensured that the conditions under which equivalent fresh enrichments were determined remained unchanged in the down-stream analyses which used the equivalent fresh enrichments. For example, fresh enrichment equivalence was determined at 500 ppm soluble boron for downstream cases which assumed 500 ppm soluble boron.

Reference 19 documents that an error has been discovered in the KENO-V.a computer code program. SCE has reviewed all analyses which use KENO V.a and verified that the identified error does not impact any analyses.

The following methodology elements are discussed below:

(1) Reference Reactivity (2) Manufacturing Tolerances And Pool Temperature Bias (3) Eccentric Placement in Storage Cells (4) Fuel Assembly Burnup Credit (Reactivity Equivalencing)

(5) Axial Burnup Bias (6) Control Element Assembly (CEA) Bias (7) Integral Fuel Burnable Absorber Credit (8) Borated Stainless Steel (SS) Inserts (9) Postulated Accidents (10) Soluble Boron Credit Methodology The relationship between these methodology elements is shown in Figure 3-1.

S023 SFP CRITICALITY ANALYSIS Page 17 of 121 3.2.1 Reference Reactivity KENO-V.a is used to establish a nominal reference reactivity, using fresh assemblies and nominal rack dimensions.

The following input parameters/assumptions are consistent with the WOG and SCE methodologies( 4.7) approved by the NRC.

(a) Nominal spent fuel storage rack and fuel assembly dimensions are used.

(b) The U0 2 stack density is 96 % of theoretical.

This bounds the small tolerances in fuel rod/assembly dimensions, including the two fuel pellet ODs of 0.325 inches and 0.3255 inches.

(c) The temperature of all materials is 68 degrees F.

(d) Axially, 150 inches of active fuel are modeled.

A 30 centimeter water reflector is used above and below the active fuel region.

(e) Storage box walls above the active fuel are conservatively not modeled.

(f) Only the storage cell box wall and Boraflex wrapper are modeled.

Storage rack structural materials, braces, or supports are not modeled.

Boraflex is not modeled, and is conservatively replaced with spent fuel pool water.

(g) Fuel assembly grids and end fittings are conservatively not modeled.

(h) In the KENO-V.a models, at least 503 neutron generations will be run with at least 2000 neutrons per generation. Kof will be taken after skipping the first 3 generations.

(i) The following formula is used to determine the final Keff:

Keff = knominal + Bmethod + Btemp + Buncert + BCEA + BAxial where: knominaI = KENO-V.a Keff Bmethod = method bias determined from benchmark critical experiments Btemp = temperature bias (68'F to 160'F)

Buncert = statistical summation of uncertainty components BCEA = CEA bias for rodded cases only BAxiaM = Axial Burnup Bias

S023 SFP CRITICALITY ANALYSIS Page 18 of 121 Representative Region I and Region II KENO-V.a models, and the whole pool KENO-V.a model used for this analysis are shown in Appendix A.

All steps for calculating the Region I and Region II zero burnup enrichment for unrestricted storage at 0 ppm (Keff < 1.0) are shown in Appendices B and C.

All tolerances, biases, and uncertainties are summarized in Appendix D.

3.2.2 Manufacturing Tolerances And Pool Temperature Bias The reactivity effects of possible variations in material characteristics and construction dimensions must be evaluated and included in the final neutron multiplication factor (KNff) of the spent fuel racks. The reactivity effects of the following tolerances are evaluated using CASMO-3:

Enrichment - The standard DOE enrichment tolerance of + 0.05 w/o U235 is used Stainless steel thickness (cell wall thickness and Boraflex wrapper thickness)

Minimum cell inner dimension Storage cell pitch (Region I only)

The delta Keff's due to these tolerances are calculated with CASMO-3 because the delta Keff's are small and can be lost in the statistical uncertainty in KENO-V.a results (Keff +/- sigma).

The variations in material characteristics due to manufacturing tolerances are random. These random variations both increase and decrease Keff, but on the average there is no net effect.

Therefore, the CASMO-3 tolerance delta-k results are conservatively combined statistically (Square root of the sum of the squares) with the methodology bias uncertainty, reference KENO-V.a Ketf uncertainty, and eccentric placement of fuel assemblies in the storage cells. Manufacturing tolerance delta-k's are conventionally reported as positive values, because the value is squared when combined statistically with other tolerances and uncertainties.

Rather than analyze tolerances on U0 2 stack density, the fuel is analyzed at a bounding value of 96% of theoretical density with no pellet dishing.

The normal fuel pool temperature range is 68'F to 160 'F. CASMO-3 is used to evaluate the fuel pool temperature bias. Since the pool temperature has no random variation, the pool temperature bias is added directly to the reference KENO-V.a result. The pool temperature bias may be either positive or negative depending on boron concentration and storage rack geometry (absence or presence of a water gap around the storage cells)

The manufacturing tolerance and pool temperature results are in Section 4. 1.

S023 SFP CRITICALITY ANALYSIS Page 19 of 121 3.2.3 Eccentric Placement in Storage Cells For eccentric placement of fuel assemblies in the storage cells, a KENO-V.a model is set up with four assemblies moved as close together as possible in the corner where four storage locations meet. The results may depend on spent fuel rack region and enrichment.

The results for eccentric placement of fuel assemblies is in Section 4.2.

3.2.4 Fuel Assembly Burnup Credit (Reactivity Equivalencing)

Spent fuel storage in the Region I and II spent fuel storage racks is achievable by means of "reactivity equivalencing". The concept of "reactivity equivalencing" is based on the fact that reactivity decreases with fuel assembly burnup. A series of reactivity calculations are performed to generate a set of "enrichment - fuel assembly discharge burnup" pairs which all give the equivalent Keff when the fuel is stored in the Region I and II racks.

The "enrichment - burnup pairs" were generated with CASMO-3. CASMO-3 allows a fuel assembly to be depleted at hot full power reactor conditions, and then placed into fuel storage rack geometry at 20 degrees C, 0 ppm soluble boron concentration, and no Xenon. The most reactive point in time for a fuel assembly after discharge is conservatively approximated by removing the Xenon. Samarium buildup after shutdown is conservatively not modeled.

To eliminate axial burnup effects, the CASMO-3 depletions are performed at the following extreme reactor operating conditions which enhance plutonium buildup:

(1) Reactor Outlet Temperature = 600°F (2) Constant BOC Fuel Temperature = 1200'F (3) Constant Soluble Boron = 1000 ppm Because the burnup history is not known exactly for the discharged fuel assemblies, the fuel assembly isotopic content (U, Pu, etc) and distribution is not known exactly. Therefore, a bounding uncertainty is applied to CASMO-3 calculational results which is zero at zero burnup and increases linearly with burnup, passing through 0.01 delta-k at 30,000 MWD/T. This uncertainty is covered by an amount of soluble boron.

As part of the reactivity equivalencing process, Pu-241 decay is also credited for up to 20 years of cooling.

S023 SFP CRITICALITY ANALYSIS Page 20 of 121 3.2.5 Axial Bumup Bias Curves of discharge burnup vs initial enrichment are generated with 2D axially infinite models (CASMO-3) which gives the modeled assembly a uniform axial burnup. However, physical fuel assemblies have a non-uniform axial burnup caused by neutron leakage from the ends of the finite length fuel assembly. Thus, the ends of the assembly have a lower burnup than the assembly average. The delta-k difference between the 3D axially dependent burnup distribution with a given average burnup and the 2D uniform burnup distribution at the same average burnup is the axial burnup bias. The bias may be either positive or negative.

The axial burnup bias is evaluated with two SIMULATE-3 Cases:

(1) All-rods-out (ARO) 2D depletion at constant Tmoa = 600 F, constant Tfuel = 1200 F, constant 1000 ppm, and burnup from 0 to 60 GWD/T.

At 0, 10, 20, 30, 40, 50, and 60 GWD/T, the 2D depleted assembly is expanded to 3D at 68 F, 0 ppm, no Xenon, and with top and bottom reflectors ( from the SIMULATE-3 models used for core follow, physics databook, and startup test predictions).

(2) ARO 3D depletion at constant Tinlt = 553 F, Tfuel - f(Tmod, Burnup), 1000 ppm, top and bottom reflectors, and burnup from 0 to 60 GWD/T. A Tinlet of 553 F bounds lower inlet temperatures.

At 0, 10, 20, 30, 40, 50, and 60 GWD/T, the 3D depleted assembly is restarted at 68 F, 0 ppm, no Xenon, and with top and bottom reflectors (from the SIMULATE-3 models used for core follow, physics databook, and startup test predictions).

If the 2D case has higher assembly k-inf than the 3D case, the bias is zero. Therefore CASMO-3 reactivity equivalencing cases run at constant Tmod = 600 F, constant Tfuel = 1200 F, and constant 1000 ppm are conservative. If the 3D case has higher assembly k-inf than the 2D case, an appropriate bias will be determined and included in the final calculational results. The axial burnup bias results are in Section 4.4.

3.2.6 Control Element Assembly (CEA) Bias Full length, 5-finger CEAs may be used in SONGS Units 2 and 3 fuel assemblies stored in Region I and It. Since CEAs will be modeled in KENO-V.a, the need for a CEA bias to apply to the KENO-V.a results will be investigated.

The potential KENO-V.a bias when CEAs are present will be determined by comparison of calculated CEA worth between KENO-V.a and CASMO-3. CASMO-3 (through SIMULATE-3) has accurately predicted SONGS Units 2 and 3 CEA bank worth measurements.(") The bias between CASMO-3 and measured CEA worth data is 0.0 delta-k.

S023 SFP CRITICALITY ANALYSIS Page 21 of 121 The CEA bias result is in Section 4.3.

CEA tip depletion is not a concern for the following reasons:

- SONGS Units 2 and 3 operation history is essentially unrodded.

- The bottom portion of the CEA finger is composed of non-depleting Silver-Indium-Cadmium.

- The CEA tip is in a low importance region. At shutdown, the flux shifts to the top of the assembly; the CEA tip is at the bottom of the assembly.

- During the lifetime of the CEAs , the rod worth is measured at the beginning of each cycle and no depletion effects are discernible.

- The W/CE CEA design has a much larger cross sectional area than the Westinghouse RCCA design, which significantly reduces CEA depletion.

3.2.7 Integral Fuel Burnable Absorber Credit Credit for burnable absorbers integral with the fuel (Erbia) includes:

- The fuel assembly is modeled at its most reactive point in life. (BOC)

- The nominal burnable poison loading is decreased by 5% to conservatively account for manufacturing tolerances.

In this analysis, fresh fuel assemblies containing 40 and 80 Erbium rods are considered.

Conservatively, enrichment zoning in the assembly is not modeled. Every fuel rod is at a nominal 4.80 weight percent (w/o), including the fuel rods with the erbia. Normally, the erbia containing fuel rods would be approximately 0.4 w/o less enriched. The erbia cutback region is modeled as 4.80 w/o instead of 4.40 w/o.

The presence of erbia is converted to an equivalent fresh enrichment:

(1) Run an assembly with erbia in CASMO-3 to determine a zero burnup enrichment rack k-inf. The fresh fuel assembly is modeled as follows:

4.80 w/o U-235 in all pins (No zoning) 40 or 80 erbia rods 2.0 w/o Erbia (Reduced from 2.1 wt%)

(2) Run an assembly containing only U0 2 fuel rods with a single U-235 enrichment.

There are NO erbia fuel rods, and there is no enrichment zoning. The U-235 enrichment shall be iterated until the rack k-inf of this case matches (1) above.

S023 SFP CRITICALITY ANALYSIS Page 22 of 121 3.2.8 Borated Stainless Steel (SS) Guide Tube Inserts Three or five borated SS guide tube inserts may be used in fuel assemblies stored in Region HI. Use of inserts allows flexibility in fuel assembly placement and greater utilization of the spent fuel pool by lowering fuel assembly reactivity, which is expressed as a lower required discharge bumup.

The inserts are 0.75 inches O.D. minimum, and must cover the entire active fuel length of 150.0 inches. The inserts must have a minimum boron loading of:

0.02434 grams of B-10 per cm 3 When 3 guide tube inserts are used, the orientation shall be the same in every assembly in the spent fuel pool (Figure 3-3).

3.2.9 Postulated Accidents Two accident conditions must be addressed: (4,18)

Pool Water Temperature Accident Fuel Assembly Misplacement 3.2.9.1 Pool Water Temperature Accident For the Pool Water Temperature Accident, CASMO-3 is used to determine the amount of reactivity associated with an increase or decrease in spent fuel pool water temperature.

The normal operating temperature range is 68 F to 160 F. This range is covered by a bias added to KENO-V.a results.

The accident range is 50 F to 248 F + 10% void. At the bottom of the racks where pressure is greater than atmospheric, 248 F (120 C) is the approximate boiling temperature. Ten percent voiding is an additional conservatism in the SCE methodology.

S023 SFP CRITICALITY ANALYSIS Page 23 of 121 3.2.9.2 Fuel Assembly Misplacement The following fuel assembly misplacement accidents are considered: 1 8

• Fuel Assembly Dropped Horizontally On Top Of The Racks Fuel Assembly Dropped Vertically Into A Storage Location Already Containing A Fuel Assembly Fuel Assembly Dropped To The SFP Floor Fuel Misloading in either Region I or Region HI 3.2.9.2.1 Fuel Assembly Dropped Horizontally On Top Of The Racks In a previous submittal,"1 8' SCE has shown that more than 12 inches of water separates the active fuel region of the dropped assembly lying on top of the racks from the active fuel region of assemblies in the storage racks. A SONGS Units 2 and 3 fuel assembly is 176.8 inches long. A storage cell is about 190 inches deep. Therefore the storage cell extends about 13 inches above the top of the upper end fitting of the fuel assembly in storage. The active fuel is about 21 inches below the top of the upper end fitting. Thus the active fuel regions of the dropped and stored fuel assemblies are neutronically isolated and reactivity does not increase.

A single un-irradiated, 5.1 w/o fuel assembly, with no burnable absorbers, in water at 68 degrees F and 0 ppm has Keff = 0.92,(") which is less than the acceptance criterion of 0.95.

3.2.9.2.2 Fuel Assembly Dropped Vertically Into A Storage Location Already Containing A Fuel Assembly In a previous submittal,(*)8 SCE has shown that more than 12 inches of water, steel, and zircaloy (fuel rod end cap and lower end fitting of dropped assembly; upper end fitting plus fuel rod end caps and plenum region of stored assembly) separates the active fuel region of the dropped assembly from the active fuel region of assemblies in the storage racks. Thus the active fuel regions of the dropped and stored fuel assemblies are neutronically isolated and reactivity does not increase.

3.2.9.2.3 Fuel Assembly Dropped To The SFP Floor A dropped fuel assembly can not fit between rack modules. However, a fuel assembly can fit between a Region I module and the pool wall.

Therefore, this case is analyzed in Section 3.2.9.2.4 below.

S023 SFP CRITICALITY ANALYSIS Page 24 of 121 3.2.9.2.4 Fuel Misloading in either Region I or Region II Misloading a single fresh 4.8 w/o fuel assembly in Regions I and II is analyzed.

The following misplacement locations were considered:

Region I: Center Of Module I Periphery Of Module I Periphery Of Module 2 Module 1-3 Interface (Center)

Module 1-3 Interface (Periphery)

Next to Module 1 (Outside Racks)

Checkerboard Patterns In Module I Region II: Center Of Module 5 Periphery Of Module 5 Periphery Of Module 6 Module 1-3 Interface (Center)

Module 1-3 Interface (Periphery)

Checkerboard Patterns In Module 5 3 Out of 4 Patterns In Module 5 I Out Of 9 patterns In Module 5 These locations are shown in Figure 3-2.

The amount of reactivity increase caused by each possible accident scenario is calculated using KENO-V.a.

For these accident conditions, the presence of soluble boron can be assumed as a realistic initial condition.

Using the results of KENO-V.a or CASMO-3 soluble boron worth calculations, the amount of soluble boron needed to offset the highest reactivity increase caused by all accident conditions and maintain Keff less than or equal to 0.95 is determined.

3.2.10 Soluble Boron Credit Methodology The soluble boron credit methodology has four steps which determine three soluble boron concentrations. The four steps are:

S023 SFP CRITICALITY ANALYSIS Page 25 of 121 (a) Determine the storage configuration of the spent fuel racks using no soluble boron 95/95 k-eff conditions such that the final KENO Keff, including all uncertainties, is less than 1.0.

(b) Using the configuration from Step (a), determine the soluble boron concentration which maintains Keff less than or equal to 0.95. This step is performed with either CASMO-3 or KENO-V.a.

(c) Since soluble boron is now credited, uncertainties in reactivity equivalencing and discharge burnup are now off-set with soluble boron. This step is performed with either CASMO-3 or KENO-V.a.

(d) Determine the increase in reactivity caused by postulated accidents and the corresponding additional amount of soluble boron needed to off-set these reactivity increases. The increase in reactivity is determined with KENO-V.a (fuel mishandling) and CASMO-3 (pool heat up).

The amount of soluble boron needed to off-set the reactivity increase for the pool heatup accident is calculated with CASMO-3. The amount of soluble boron needed to off-set the reactivity increase for the fuel mishandling accident is calculated with both CASMO-3 and KENO-V.a to compare results.

The final soluble boron requirement is the sum of the requirements determined in steps (b),

(c), and (d) above.

The total soluble boron credit requirement along with the storage configuration specified for no soluble boron shows that the spent fuel racks will always maintain Kaff less than or equal to 0.95. Further, the no soluble boron storage configuration will ensure that Keff remains less than 1.0 with no soluble boron in the spent fuel pool.

Finally, Reference 9 documents that an equivalent fresh enrichment determined at 0 ppm is under-estimated if used in a KENO-V.a model which includes dissolved boron. SCE's analyses ensured that the conditions under which equivalent fresh enrichments were determined remained unchanged in the down-stream analyses which used the equivalent fresh enrichments.

S023 SFP CRITICALITY ANALYSIS Page 26 of 121 Figure 3-1 SFP Criticality Methodology CELLDAN NITAWL-II I

KENO-V.a CASMO-3 (1) Reference Reactivity (1) Manufacturing Tolerances (2) Eccentric Placement In Pool Temp Bias Storage Cells (2) Reactivity Equivalencing (3) CEA Bias (Burnup/Cooling Time Tables)

(4) Integral Fuel Burnable (3) CEA Bias Absorber Credit (4) Integral Fuel Burnable (5) Borated SS Guide Tube Inserts Absorber Credit (6) Postulated Accidents (5) Borated SS Guide Tube (7) Soluble Boron Credit Inserts (6) Postulated Accidents (7) Soluble Boron Credit I CASMO-3 X-Sections SIMULATE-3 Axial Burnup Bias (2D vs 3D Comparison)

S023 SFP CRITICALITY ANALYSIS Page 27 of 121 Figure 3-2 Fuel Mishandling Analysis Locations RI-I RII-3 RII-5 RII-7 x

x X1 Ix X I

I X X1 Ix Ix 2/4 1 I 2/4 1/9 I

I I

I

• ,I-2 I I

I R1 I I x RH-4 RII-6 RII-8 Ry-z = Region y, Module z x = Misload Location

S023 SFP CRITICALITY ANALYSIS Page 28 of 121 Figure 3-3 Orientation Of 3 Guide Tube Inserts Fuel Assembly

<--- Serial yI Number 00 xx 00 xx xx xx xx 00 xx 00 xx = Guide Tube With Insert xx 00 - Empty Guide Tube 00

S023 SFP CRITICALITY ANALYSIS Page 29 of 121

4. CRITICALITY SAFETY ANALYSES This section summarizes the results of the criticality analyses preformed for the SONGS Units 2 and 3 spent fuel storage racks assuming no Boraflex. The analyses were performed at 0 ppm. The acceptance criteria is Keff < 1.0, including all uncertainties.

First, results are presented for : Manufacturing Tolerances Pool Temperature Bias Eccentric Placement Bias CEA Bias Axial Burnup Bias The delta-k's from these analyses (and the bias and uncertainty from Section 3.1.2) are needed to calculate a spent fuel storage rack Keff which includes all biases and uncertainties. (All tolerances, biases, and uncertainties are summarized in Appendix D.

Then, permissible storage patterns for both Region I and Region II are given. SONGS Unit 1 assembly results are given. Finally, results are given for: Inter-module Spacing Reconstitution Station Failed Fuel Rod Storage Basket Fuel Handling Equipment Non-fuel components 4.1 MANUFACTURING TOLERANCES AND POOL TEMPERATURE BIAS The manufacturing tolerance and normal pool temperature range results are shown in Table 4-1.

These results were calculated with CASMO-3. The manufacturing tolerances are combined statistically (square root of the sum of the squares). The pool temperature bias is added directly to the KENO-V.a result.

4.2 ECCENTRIC PLACEMENT BIAS Eccentric Placement of fuel assemblies in the storage cells has been evaluated with KENO-V.a.

The results are:

Region I Delta-k = 0.01383 (4.80 w/o) 0.00767 (2.47 w/o)

Region HIDelta-k = 0.0 (No enrichment dependence)

As discussed in Section 3.2.3, this result is combined statistically with the manufacturing tolerance results.

S023 SFP CRITICALITY ANALYSIS Page 30 of 121 4.3 CONTROL ELEMENT ASSEMBLY (CEA) BIAS The bias for 5-finger, full-length CEAs in SONGS Units 2 and 3 fuel assemblies is 0.007 delta-k.

This bias is independent of enrichment and was determined by inter-comparison of CASMO-3 and KENO-V.a for rodded and unrodded cases. The CEA bias is added directly to the KENO-V.a results.

4.4 AXIAL BURNUP BIAS The axial burnup bias for SONGS fuel assemblies is 0.0 delta-k at all burnups from 0 to 60 GWD/T. This bias is added directly to the KENO-V.a results.

SCE's analyses have determined that an assembly with a uniform axial burnup of X GWD/T (X = 0 to 60 GWD/T) has higher reactivity than an assembly with a 3D burnup profile with average bumup of X GWD/T provided:

(1) The mode of operation is ARO.

(2) The uniform axial bumup results from depletion at constant Tmod = 600 F, Tfue! = 1200 F.

(3) The 3D axial burnup distribution results from depletion at actual reactor conditions of T *nle 553 F, Tfe 1C = f(Tmod, Bumrup), and axial variation of these temperatures.

The SIMULATE-3 results are shown in Table 4-2. The axial burnup bias is 0.0 delta-k. In fact, there is a small credit which increases with burnup. Conservatively, this credit is not taken.

4.5 SONGS UNITS 2 AND 3 FUEL ASSEMBLIES 4.5.1 Region I The non-accident neutron multiplication factor (Keff) for the Region I spent fuel storage racks is less than 1.0, including all uncertainties, assuming a soluble boron concentration of 0 ppm.

The permissible Region I storage patterns are shown in Tables 4-3 through 4-10 and Figures 4-1 through 4-6.

S023 SFP CRITICALITY ANALYSIS Page 31 of 121 4.5.2 Region II The non-accident neutron multiplication factor (Keff) for the Region II spent fuel storage racks is less than 1.0, including all uncertainties, assuming a soluble boron concentration of 0 ppm.

The permissible Region II storage patterns are shown in Tables 4-11 through 4-25 and Figures 4-7 through 4-21. When 3 guide tube inserts are used, the orientation shall be the same in every assembly in the spent fuel pool (Figure 3-3). A 5-finger, full-length CEA may be used in place of 3 or 5 borated SS guide tube inserts.

4.5.3 Region I And Region H Checkerboard Pattern Interface requirements The boundary between checkerboard zones and the boundary between a checkerboard zone and all cell storage must be controlled to prevent an undesirable increase in reactivity. This is accomplished by examining each 2x2 assembly matrix interface and ensuring that each matrix conforms to restrictions for both regions.

For example, consider a fuel assembly location E in the following matrix of storage cells.

Storage Pattern 1 A B C Interface D E F Storage Pattern 2 G H I Four 2x2 matrices of storage cells which include cell E are created in the above figure. They include (A, B, D, E), (B, C, F, E), (E, F, I, H), and (D, E, H, G). Matrices (A,B,D,E) and (B,C,F,E) must meet the requirements of Storage Pattern 1. Matrices (E,F,I,H) and (D,E,G,H) must meet the requirements of Storage Pattern 2.

A row of empty storage cells can also be used at the interface to separate different storage patterns.

The interface requirements are shown in Figures 4-22 through 4-27, and are summarized in Figure 4-32.

4.5.4 Region II One Out Of Nine Pattern Interface requirements The boundary between One Out of Nine and the boundary between all cell storage, checkerboard storage, and 3-out-of-4 storage must be controlled to prevent an undesirable increase in reactivity.

Specific KENO-V.a cases were run to determine the interface requirements.

S023 SFP CRITICALITY ANALYSIS Page 32 of 121 A row of empty storage cells can also be used at the interface to separate different storage patterns.

The interface requirement are shown in Figure 4-28 through 4-31, and are summarized in Figure 4-32.

4.6 SONGS UNIT I FUEL ASSEMBLIES Unit I Fuel has not been analyzed to be stored in Region I.

4.6.1 Unrestricted Storage in Region II 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, and the cooling time is greater than 5 years.

SONGS Unit I 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.

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

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

4.6.2 SFP Peripheral Storage in Region II SONGS Unit 1 nominal 4.00 w/o assemblies can be stored in the Region II Racks (SFP periphery) if:

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

4.7 INTER-MODULE SPACING A full pool KENO-V.a model (Figure A-3 in Appendix A) was used to evaluate the inter-module spacing assuming no Boraflex. The inter-module spacing needed when Boraflex is present is conservatively unchanged when there is no Boraflex.

S023 SFP CRITICALITY ANALYSIS Page 33 of 121 4.8 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 pattern are blocked to make it impossible to misload a fuel assembly during reconstitution activities.

4.9 FAILED FUEL ROD STORAGE BASKET The failed fuel rod storage basket (FFRSB) is less reactive than an intact fuel assembly.

Therefore, for storage of the FFRSB in the SFP storage racks, the 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. Alternatively, explicit analyses using the methodology of Section 3.2 may be performed to determine storage requirements for the FFRSB.

4.10 FUEL HANDLING EQUIPMENT This equipment (Upenders, Transfer Baskets, Refueling Machine Mast) is not affected by the presence or absence of Boraflex.

4.11 NON-FUEL COMPONENTS Neutron sources and non-fuel bearing assembly components (thimble plugs, CEAs, etc) may be stored in fuel assemblies without affecting the storage requirements of these assemblies. The neutron source material is an absorber which reduces reactivity. Thus, a neutron source may be stored in an empty cell or in an assembly. 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.

S023 SFP CRITICALITY ANALYSIS Page 34 of 121 Table 4-1 Manufacturing Tolerance And Pool Temperature Results REGION I DELTA-Kinf 5.10 w/o 1.85 w/o Tolerance 0 ppm 500 ppm 1000 ppm 0 ppm 500 ppm 1000 ppm Enrichment 0.00179 0.00201 0.00216 0.00757 0.00772 0.00774 SS Thickness 0.00518 0.00370 0.00283 0.00460 0.00291 0.00215 Cell ID 0.00531 0.00518 0.00502 0.00360 0.00350 0.00335 Cell Pitch 0.00796 0.00807 0.00790 0.00620 0.00586 0.00560 40 C (104 F) 0.00383 0.00344 0.00314 0.00182 0.00152 0.00144 71 C (160 F) 0.00914 0.00862 0.00829 0.00389 0.00438 0.00454 REGION II DELTA-Kinf 1.85 w/o 1.20 w/o Tolerance 0 ppm 500 ppm 1000 ppm 0 ppm 500 ppm 1000 ppm Enrichment 0.00907 0.00959 0.00976 0.01543 0.01547 0.01510 SS Thickness 0.00174 0.00107 0.00062 0.00165 0.00100 0.00058 Cell ID 0.00244 0.00304 0.00331 0.00223 0.00267 0.00286 Cell Pitch N/A N/A N/A N/A N/A N/A 40 C (104 F) -0.00040 0.00031 0.00077 -0.00076 0.00000 0.00043 71 C (160 F) -0.00099 0.00147 0.00285 -0.00160 0.00083 0.00215 Note: Region HIdoes not have a water gap between storage cells. Therefore, cell pitch is not applicable in the Region H racks

S023 SFP CRITICALITY ANALYSIS Page 35 of 121 Table 4-2 Axial Burnup Bias 1.87 w/o 1.87 w/o Bumup 2D K__ff 3D Keff* Delta-k (2D - 3D) 0 1 .24454 1.24454 0.00000 10 1.11561 1.10809 0.00752 20 1.01562 1.00765 0.00797 30 0.94019 0.93183 0.00836 40 0.88513 0.87637 0.00876 50 0 .84777 0.83876 0 .00901 60 0 .82274 0.81284 0 .00990 4.45 w/o 4.45 w/o Bumup 2D K~e~f 3D K~ff* Delta-k (2D - 3D) 0 1.45672 1.45671 0. 00001 10 1.34135 1.33404 0.00731 20 1.25052 1 .24275 0.00777 30 1.16868 1. 16048 0.00820 40 1. 08984 1. 08169 0.00815 50 1. 01564 1.00736 0.00828 60 0.94916 0.93970 0.00946

• 3D Axial Burnup Profile

S023 SFP CRITICALITY ANALYSIS Page 36 of 121 Table 4-3 Region I Category I-i Unrestricted Storage Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling

5. 00 22 .84 21.47 20.59 20.04 19.67 4.50 18.61 17.57 16. 89 16 .45 16.17 4.00 14.30 13.58 13 .09 12.78 12.57 3.50 9.84 9.40 9.11 8.92 8.79 3.00 5.24 5.02 4.91 4.84 4.79 2 .47 0.00 0.00 0.00 0. 00 0.00 Note: Category I-i and fresh fuel with full-length 5-finger CEAs (Table 4-7) may be stored together with no restrictions.

S023 SFP CRITICALITY ANALYSIS Page 37 of 121 Table 4-4 Region I Category 1-2 SFP Peripheral Storage Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 12.55 12.15 11.82 11.61 11.47 4.50 9.09 8.85 8.63 8.49 8.40 4.00 5.58 5.43 5.33 5.25 5.21 3.50 2.22 2.13 2.09 2.05 2.03 3.20 0.00 0.00 0.00 0.00 0.00

S023 SFP CRITICALITY ANALYSIS Page 38 of 121 Table 4-5 Region I Category 1-3 Filler Assembly For l-out-of-4 Pattern Initial Minimum Burnup(GWD/T)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5 .00 39.99 36.28 34.27 33.04 32.22 4.50 34.95 31.71 29.94 28.84 28.12 4.00 29.71 26.99 25 .46 24.51 23 .89 3 .50 24.22 22 .03 20.79 20.02 19 .52 3 .00 18.37 16 .84 15 .91 15.34 14 .97 2.50 12.21 11.30 10.72 10.37 10.13 2.00 5.28 5. 05 4.85 4.72 4.62 1.71 0.00 0. 00 0. 00 0.00 0.00 4.80 w/o 1.71 w/o Fresh 1.71 w/o 1.71 w/o

S023 SFP CRITICALITY ANALYSIS Page 39 of 121 Table 4-6 Region I 4.80 w/o Fresh Fuel Checkerboard Initial Minimum Burnup(GWD/T)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 4.80 0.00 0.00 0.00 0.00 0.00 (Empty) 4.80 w/o 4.80 w/o (Empty)

S023 SFP CRITICALITY ANALYSIS Page 40 of 121 Table 4-7 Region I 4.80 w/o Fresh Fuel With Full-length, 5-finger CEA Initial Minimum Burnup(GWD/MTU)

Enrichment (wIo) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 4.80 0.00 0.00 0.00 0.00 0.00 4.80 w/o 4.80 w/o Fresh Fresh With CEA With CEA 4.80 w/o 4.80 w/o Fresh Fresh With CEA With CEA Note: Category I-i and fresh fuel with full-length 5-finger CEAs (Table 4-7) may be stored together with no restrictions.

S023 SFP CRITICALITY ANALYSIS Page 41 of 121 Table 4-8 Region I Category 1-4 Filler Assembly For l-out-of-4 Pattern Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 26.57 24.71 23 .59 22.90 22.44 4.50 22 .12 20.62 19 .73 19. 17 18.80 4.00 17 .54 16.46 15 .78 15.35 15 .07 3.50 12 .84 12 .12 11.66 11.37 11 18 3 .00 7.95 7.56 7.31 7.15 7.05 2.50 2.76 2 .64 2.56 2.50 2.46 2 .27 0.00 0. 00 0. 00 0.00 0.00 4.80 w/o 2.27 w/o Fresh 80 Erbia 2.27 w/o 2.27 w/o

S023 SFP CRITICALITY ANALYSIS Page 42 of 121 Table 4-9 Region I Category 1-5 Filler Assembly For 1-out-of-4 Pattern Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 30.81 28.40 27.00 26.14 25.57 4.50 26.17 24. 17 22 .99 22 .26 21.78 4.00 21.32 19.77 18.84 18.27 17.88 3.50 16.32 15.22 14 .55 14 .13 13 .85 3 .00 11.11 10.45 10.05 9.79 9.61 2.50 5.55 5.30 5 .14 5.04 4.98 2.07 0.00 0.00 0.00 0. 00 0.00 4.80 w/o 2.07 w/o Fresh 40 Erbia 2.07 w/o 2.07 w/o

S023 SFP CRITICALITY ANALYSIS Page 43 of 121 Table 4-10 Region I Category 1-6 4.80 w/o Assembly Depleted to 18.0 GWD/MTU Initial Minimum Burnup(GWD/MTU)

Enrichment (wlo) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 19.82 18.84 18 .12 17.:67 17.37 4.50 15.83 15.11 14.58 14.24 14.01 4.00 11.75 11.28 10 .92 10.69 10.54 3.50 7.56 7.23 7 .04 6.91 6.83 3.00 3.28 3.15 3 .07 3.03 2.99 2.65 0.00 0.00 0. 00 0.00 0.00 Category 1-4 Checkerboard Partner For Category 1-6 Fuel Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 26.57 24.71 23.59 22 .90 22.44 4.50 22.12 20.62 19.73 19 .17 18.80 4.00 17.54 16.46 15.78 15.35 15.07 3.50 12.84 12 .12 11.66 11.37 11.18 3.00 7.95 7.56 7.31 7.15 7.05 2.50 2.76 2.64 2.56 2.50 2.46 2.27 0.00 0.00 0.00 0.00 0.00 2.65 w/o 2.27 w/o 2.27 w/o 2.65 w/o

S023 SFP CRITICALITY ANALYSIS Page 44 of 121 Table 4-11 Region II Category II-1 Unrestricted Storage Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 53.76 47 .77 44.75 43.00 41.86 4.50 48 .43 42 .93 40.15 38.52 37 .47 4.00 42 .91 37.94 35.40 33.92 32.96 3 .00 30.99 27.26 25.30 24.16 23.43 2.00 17 .05 14 .97 13 .90 13.25 12 .83 1 .87 14 .93 13.23 12.26 11.68 11.31 1.23 0.00 0.00 0.00 0.00 0.00 Note: Category II-1, 11-8, 11-9, and 11-15 may be stored together with no restrictions.

S023 SFP CRITICALITY ANALYSIS Page 45 of 121 Table 4-12 Region II Category 11-2 SFP Peripheral Storage Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 36.95 33.68 31.89 30.81 30.10 4.50 32.29 29.44 27.87 26.91 26.28 4.00 27.44 25.04 23.70 22.88 22.35 3.00 16.95 15.62 14.83 14.34 14.03 2.00 4.93 4.67 4.52 4.42 4.35 1.87 3.04 2.87 2.76 2.69 2.64 1.70 0.00 0.00 0.00 0.00 0.00

S023 SFP CRITICALITY ANALYSIS Page 46 of 121 Table 4-13 Region II Category 11-3 (Checkerboard Pattern for Category 11-4 Fuel)

Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 41. 18 37.27 35.18 33.93 33 .12 4.50 36.34 32 .87 31.01 29.88 29.15 4.00 31.29 28.31 26.69 25.70 25.06 3.00 20.32 18.50 17 .47 16.84 16 .42 2.00 7.81 7.25 6.91 6.71 6.58

1. 87 5.90 5 .53 5.30 5.17 5.09 1.56 0.00 0.00 0.00 0.00 0.00 Category 11-4 (Checkerboard Partner For Category 11-3 Fuel)

Initia 1 Minimum Burnup(GWD/MTU)

Enrichme tnt (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 75.42 61.90 56.85 54. 18 52.60 4.50 68.08 56. 12 51.65 49.25 47.76 4.00 60.74 50.35 46.44 44. 19 42 .78 3.00 46.06 38.80 35.41 33 .52 32.31 2.00 31.38 25.71 23.12 21.65 20.71 1 .87 29. 19 23 .83 21.34 19. 91 19. 08 0.94 0.00 0.00 0.00 0.00 0.00 0.94 w/o 1.56 w/o 1.56 w/o 0.94 w/o

S023 SFP CRITICALITY ANALYSIS Page 47 of 121 Table 4-14 Region II Category 11-5 (Checkerboard Pattern for Category 11-6 Fuel)

Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 47 .50 42. 58 40.03 38.53 37.55 4.50 42 .40 37.95 35.64 34.26 33.37 4.00 37.10 33.16 31.10 29.86 29.06 3.00 25.64 22.89 21.40 20.52 19.95 2.00 12 .29 11.10 10 .42 10.01 9.75 1.87 10.24 9.35 8.80 8.46 8.25 1.38 0.00 0.00 0.00 0.00 0.00 Category 11-6 (Checkerboard Partner For Category 11-5 Fuel)

Initia 1 Minimum Burnup(GWD/MTU)

Enrichme nt (wlo) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 62 .37 53 .95 50.33 48.25 46.91 4.50 56.21 48.90 45.51 43 .56 42.31 4.00 50.04 43 .67 40.54 38.73 37 .57 3 .00 37.71 32.56 29 .97 28. 52 27.58 2.00 23.30 19.80 18.13 17 .14 16.50 1 .87 21.11 18. 02 16 .42 15.48 14 .88 1.08 0.00 0.00 0.00 0.00 0.00 1.08 w/o 1.38 w/o 1.38 w/o 1.08 w/o

S023 SFP CRITICALITY ANALYSIS Page 48 of 121 Table 4-15 Region II Checkerboard Storage Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 4.80 0.00 0.00 0.00 0.00 0.00 Empty (Blocked) 4.80 w/o Empty 4.80 w/o (Blocked)

S023 SFP CRITICALITY ANALYSIS Page 49 of 121 Table 4-16 Region II Category 11-7 3 Out of 4 Storage Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling

5. 00 34.20 31. 35 29.74 28 .76 28 12

4. 50 29. 67 27. 21 25.82 24 97 24 .41

4. 00 24. 94 22. 92 21.75 21 .05 20 .59

3. 00 14.79 13. 76 13 .13 12 .73 12 .47

2. 00 3.16 3. 00 2.90 2 .83 2 .79

1. 87 1.21 1. 14 1.09 1 .06 1 04

1. 80 0.00 0. 00 0.00 0 .00 0 .00 1.80 w/o 1.80 w/o Empty 1.80 w/o (Blocked)

S023 SFP CRITICALITY ANALYSIS Page 50 of 121 Table 4-17 Region II Category 11-8 Unrestricted Storage With 5 Borated SS Guide Tube Inserts In Every Assembly Initial Minimum Burnup (GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 37.68 34.53 32.72 31. 61 30 .88 4.50 32.61 29.90 28.33 27. 36 26 .72 4.00 27.33 25.10 23.78 22. 97 22 .43 3.00 15.86 14.76 14 .06 13. 61 13 .32 2.00 2.04 1.97 1.89 1. 84 1 .81 1.90 0.00 0.00 0.00 0. 00 0 .00 1.90 w/o 1.90 w/o 1.90 w/o 1.90 w/o 1.90 w/o 1.90 w/o 5 Inserts 5 Inserts 5 Inserts 5 Inserts 5 Inserts 5 Inserts Note: Category I-1, 11-8, 11-9, and 11-15 may be stored together with no restrictions.

S023 SFP CRITICALITY ANALYSIS Page 51 of 121 Table 4-18 Region II Category 11-9 Unrestricted Storage With 3 Borated SS Guide Tubes Inserts In Every Assembly Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 44.16 39.95 37.68 36.31 35.42 4.50 38.99 35.25 33.22 31.99 31.18 4.00 33.61 30.38 28.60 27.52 26.81 3.00 21.92 19.86 18.72 18.01 17 .56 2.00 8.28 7 .72 7.34 7.11 6.96 1 .87 6.18 5 .83 5.58 5.43 5.34 1.59 0.00 0.00 0.00 0.00 0.00 1.59 w/o 1.59 w/o 1.59 w/o 1.59 w/o 1.59 w/o 1.59 w/o 3 Inserts 3 Inserts 3 Inserts 3 Inserts 3 Inserts 3 Inserts Note: Category 11-1, 11-8, 11-9, and 11-15 may be stored together with no restrictions.

S023 SFP CRITICALITY ANALYSIS Page 52 of 121 Table 4-19 Region II Category II-10 Filler Assembly With 5 Borated SS Guide Tube Inserts Ini tial Minimum Burnup(GWD/MTU)

Enric=hment (w /0) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 80.09 65.66 60 .12 57 .45 55.68 4.50 72 .13 59.43 54 .55 52. 02 50.33 4.00 64 .18 53 .19 48.98 46. 58 44 .98 3 .00 48.27 40.72 37.16 35.03 33.75 2.00 32.35 26.59 23.79 22.25 21.25 1.03 0.00 0.00 0.00 0.00 0.00 1.03 w/o 1.03 w/o 1.03 w/o 5 Inserts 5 Inserts 5 Inserts 1.03 w/o 4.80 w/o 1.03 w/o 5 Inserts Fresh 5 Inserts 1.03 w/o 1.03 w/o 1.03 w/o 5 Inserts 5 Inserts 5 Inserts

S023 SFP CRITICALITY ANALYSIS Page 53 of 121 Table 4-20 Region II Category II-11 Filler Assembly With 5 Borated SS Guide Tube Inserts Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 47 .04 42 .52 40.05 38. 57 37.60 4.50 41.62 37.58 35.36 34. 02 33.14 4.00 35.97 32.46 30.50 29.32 28.54 3 .00 23.70 21.42 20.09 19.31 18.79 2.00 9 .17 8. 54 8.09 7.81 7 .62 1.59 0.00 0.00 0.00 0.00 0.00 1.59 w/o 1.59 w/o 1.59 w/o 5 Inserts 5 Inserts 5 Inserts 4.80 w/o 1.59 w/o Fresh 1.59 w/o 5 Inserts 5 Inserts 5 Inserts 1.59 w/o 1.59 w/o 1.59 w/o 5 Inserts 5 Inserts 5 Inserts

S023 SFP CRITICALITY ANALYSIS Page 54 of 121 Table 4-21 Region II Category 11-12 Filler Assembly With 3 Borated SS Guide Tube Inserts Initial Minimum Burnup(GWD/MTU)

Enrichment (wIo) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 54 .33 48.48 45.46 43 .67 42. 51 4.50 48.81 43.45 40. 67 39.02 37 .95 4.00 43 .07 38.24 35.72 34.22 33 .26 3 .00 30.65 27.11 25.18 24.05 23 .33 2.00 16.01 14 .23 13 .22 12 .62 12 .22 1 .87 13 .82 12.35 11.47 10 .94 10.60 1.32 0.00 0.00 0.00 0.00 0.00 1.32 w/o 1.32 w/o 1.32 w/o 3 Inserts 3 Inserts 3 Inserts 4.80 w/o 1.32 w/o Fresh 1.32 w/o 3 Inserts 5 Inserts 3 Inserts 1.32 w/o 1.32 w/o 1.32 w/o 3 Inserts 3 Inserts 3 Inserts

S023 SFP CRITICALITY ANALYSIS Page 55 of 121 Table 4-22 Region II Category 11-13 Filler Assembly With No Borated SS Guide Tube Inserts Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 64.24 55. 51 51.59 49. 41 48. 03 4.50 57 .99 50.23 46.73 44. 67 43. 38 4.00 51.75 44 .94 41.71 39. 79 38. 59 3 .00 39.25 33.75 31.05 29. 48 28. 50 2.00 24.76 20.95 19.07 18. 01 17. 33 1 .87 22 .64 19.10 17.38 16. 37 15. 72 1.05 0.00 0.00 0.00 0. 00 0. 00 1.05 w/o 1.05 w/o 1.05 w/o 4.80 w/o 1.05 w/o Fresh 1.05 w/o 5 Inserts 1.05 w/o 1.05 w/o 1.05 w/o

S023 SFP CRITICALITY ANALYSIS Page 56 of 121 Table 4-23 Region II Category 11-13 Filler Assembly For Category 11-14 Fuel Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 64 .24 55.51 51.59 49.41 48.03 4.50 57 .99 50.23 46.73 44. 67 43 .38 4.00 51.75 44. 94 41.71 39.79 38.59 3.00 39.25 33.75 31.05 29.48 28.50 2.00 24.76 20.95 19. 07 18.01 17 .33 1.87 22 .64 19. 10 17.38 16.37 15.72 1.05 0.00 0.00 0.00 0.00 0.00 Category 11-14 4.80 w/o Assembly Depleted to 18.0 GWD/MTU Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 19.59 18.61 17.96 17.54 17.27 4.50 15.93 15.17 14.68 14.36 14.15 4.00 12.18 11.64 11.29 11.07 10.93 3.00 4.28 4.12 4.05 4.00 3.98 2.51 0.00 0.00 0.00 0.00 0.00 1.05 w/o 1.05 w/o 1.05 w/o 1.05 w/o 2.51 w/o 1.05 w/o 1.05 w/o 1.05 w/o 1.05 w/o

S023 SFP CRITICALITY ANALYSIS Page 57 of 121 Table 4-24 Region II Category 11-14 4.80 w/o Assembly Depleted to 18.0 GWD/MTU Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 19.59 18.61 17.96 17.54 17.27 4.50 15.93 15.17 14.68 14.36 14.15 4.00 12.18 11.64 11.29 11.07 10.93 3.00 4.28 4.12 4.05 4.00 3.98 2.51 0.00 0.00 0.00 0.00 0.00 Category II-11 Filler Assembly With 5 Borated SS Guide Tube Inserts Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 0 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5.00 47. 04 42 .52 40.05 38.57 37.60 4 .50 41. 62 37.58 35.36 34 .02 33 .14 4.00 35. 97 32.46 30.50 29.32 28.54 3 .00 23. 70 21.42 20.09 19.31 18.79 2.00 9. 17 8.54 8.09 7.81 7 .62 1.59 0. 00 0.00 0.00 0.00 0.00 1.59 w/o 1.59 w/o 1.59 w/o 5 Inserts 5 Inserts 5 Inserts 1.59 w/o 2.51 w/o 1.59 w/o 5 Inserts 5 Inserts 1.59 w/o 1.59 w/o 1.59 w/o 5 Inserts 5 Inserts 5 Inserts

S023 SFP CRITICALITY ANALYSIS Page 58 of 121 Table 4-25 Region II Category 11-15 Unrestricted Storage With A Full-Length, 5-Finger CEA In Every Assembly Initial Minimum Burnup(GWD/MTU)

Enrichment (w/o) 0 Years 5 Years 10 Years 15 Years 20 Years Cooling Cooling Cooling Cooling Cooling 5 .00 29.24 27 .24 26 00 25.22 24.70 4 .50 24.44 22 .84 21 81 21.17 20.75 4 .00 19.41 18 .26 17 49 17.00 16.68 3 .00 8.83 8 .47 8 19 8.02 7.90 2 .30 0.00 0 .00 0 00 0.00 0.00 2.30 w/o 2.30 w/o j 2.30 w/o 2.30 w/o 2.30 w/o 2.30 w/o CEA CEA CEA CEA CEA CEA Note: Category II-1, 11-8, 11-9, and 11-15 may be stored together with no restrictions.

S023 SFP CRITICALITY ANALYSIS Page 59 of 121 Figure 4-1 REGION I MINIMUM BURNUP FOR CATEGORY I-1 FUEL (UNRESTRICTED STORAGE) 25

~20 0

z3E 15 E10 Cl)

Cl)

IL 5 01 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

E3 OYears -G-- 5 Years IF 10Years-z-- 15Years ER 20 Years (Figure data points are from Table 4-3)

S023 SFP CRITICALITY ANALYSIS Page 60 of 121 Figure 4-2 REGION I MINIMUM BURNUP FOR CATEGORY 1-2 FUEL (SFP PERIPHERAL STORAGE) 15 10 c1o E

a)

(n U) 5 IL 0L--

3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

-- E-- 0 Years -- E 5 Years --- 10 Years 2ýt 15 Years -R 20 Years (Figure data points are from Table 4-4)

S023 SFP CRITICALITY ANALYSIS Page 61 of 121 Figure 4-3 REGION I MINIMUM BURNUP FOR CATEGORY 1-3 FUEL (FILLER ASSEMBLY FOR I-OUT-OF-4 PATTERN) 40 35 S30 c- 25 mo 20 E

(D 15 g 10 U-5 0L-1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o) 0 Years -e- 5 Years rvz 10 Years t*i 15 Years 11 20 Years (Figure data points are from Table 4-5)

S023 SFP CRITICALITY ANALYSIS Page 62 of 121 Figure 4-4 REGION I MINIMUM BURNUP FOR CATEGORY 1-4 FUEL (FILLER ASSEMBLY FOR 1-OUT-OF-4 PATTERN) 30 25 20 E

"-I m 15 E

ct10 U-5 01 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o) 0 Years OE3 - 5 Years - 10 Years -ýr 15 Years 20 Years (Figure data points are from Table 4-8)

S023 SFP CRITICALITY ANALYSIS Page 63 of 121 Figure 4-5 REGION I MINIMUM BURNUP FOR CATEGORY 1-5 FUEL (FILLER ASSEMBLY FOR I-OUT-OF-4 PATTERN) 35 30 25 E 20 0,

U,fl15 E

10 5

0 A 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o) 0 Years 9 5 Years IF 10 Years---- 15 Years X 20 Years (Figure data points are from Table 4-9)

S023 SFP CRITICALITY ANALYSIS Page 64 of 121 Figure 4-6 REGION I MINIMUM BURNUP FOR CATEGORY 1-6 FUEL (4.80 W/O ASSEMBLY DEPLETED TO 18.0 GWD/T) 20 c) 15 C

3:

0.

E co 10 2,

U) 6 5 13 0L-2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)

-a-- 0Years -E? 5 Years v;X 10 Years z!: 15 Years -i- 20 Years (Figure data points are from Table 4-10)

S023 SFP CRITICALITY ANALYSIS Page 65 of 121 Figure 4-7 REGION 11 MINIMUM BURNUP FOR CATEGORY 11-I FUEL (UNRESTRICTED STORAGE) 60--

S:240 CL E

30-1 20 U -

0 I 2 3 4 5 Initial U-235 Enrichment (w/o)

E 0OYears - 5Years --- l0Years -vl 15Years F- 20Years (Figure data points are from Table 4-11)

S023 SFP CRITICALITY ANALYSIS Page 66 of 121 Figure 4-8 REGION II MINIMUM BURNUP FOR*CATEGORY 11-2 FUEL (SFP PERIPHERAL STORAGE) 40 030

(-9 m20 e.0 E

0-1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E 0 Years e 5 Years - 10Years 1-- I-v 15Years -s 20Years (Figure data points are from Table 4-12)

S023 SFP CRITICALITY ANALYSIS Page 67 of 121 Figure 4-9 REGION H1 MINIMUM BURNUP FOR CATEGORY 11-3 FUEL (CHECKERBOARD PARTNER FOR CATEGORY 11-4) 50

-40 30 E

E20 0) 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E3 OYears e 5Years - 10 Years -- 15 Years -s- 20 Years (Figure data points are from Table 4-13)

S023 SFP CRITICALITY ANALYSIS Page 68 of 121 Figure 4-10 REGION 11 MINIMUM BURNUP FOR CATEGORY 11-4 FUEL (CHECKERBOARD PARTNER FOR CATEGORY 11-3)80-060 C

I40~

m 40 E

-U, 20 ILJ 0

2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years 5 Years - 10 Years -- 15Years -i- 20 Years (Figure data points are from Table 4-13)

S023 SFP CRITICALITY ANALYSIS Page 69 of 121 Figure 4-11 REGION H MINIMUM BURNUP FOR CATEGORY 11-5 FUEL (CHECKERBOARD PARTNER FOR CATEGORY 11-6) 50 CF40-CD 5 30 E 20 L10 0

1 2 3 4 5 Initial U-235 Enrichment (w/o)

-~i 0 Years ---- 5 Years ---- 10 Years 1ýz 15 Years iý 20 Years (Figure data points are from Table 4-14)

S023 SFP CRITICALITY ANALYSIS Page 70 of 121 Figure 4-12 REGION H MINIMUM BURNUP FOR CATEGORY 11-6 FUEL (CHECKERBOARD PARTNER FOR CATEGORY 11-5) 70 60 20 5o--

0t.

• 40.-

I-

.* 30-10--

<20.

(It 10 1 2 3 4 5 Initial U-235 Enrichment (w/o)

ER 0 Years -e- 5 Years - 10 Years - 15 Years -- 20 Years (Figure data points are from Table 4-14)

S023 SFP CRITICALITY ANALYSIS Page 71 of 121 Figure 4-13 REGION 1H MINIMUM BURNUP FOR CATEGORY 11-7 FUEL (3-OUT-OF-4 STORAGE) 35 30-25-(9 a..

20-

"15-E (0

<10--

5-0-

1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

-i 0 Years -E 5 Years -- <-- 10 Years ';F 15 Years ii 20 Years (Figure data points are from Table 4-16)

S023 SFP CRITICALITY ANALYSIS Page 72 of 121 Figure 4-14 REGION 11 MINIMUM BURNUP FOR CATEGORY 11-8 FUEL (UNRESTRICTED STORAGE WITH 5 BORATED SS GUIDE TUBE INSERTS) 40 630 M 20 E

U-0 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

- OYears - 5Years - 10 Years -*- 15 Years i+E 20Years (Figure data points are from Table 4-17)

S023 SFP CRITICALITY ANALYSIS Page 73 of 121 Figure 4-15 REGION II MINIMUM BURNUP FOR CATEGORY 11-9 FUEL (UNRESTRICTED STORAGE WITH 3 BORATED SS GUIDE TUBE INSERTS) 50 i S40 jAcceptable Region .

E 0~

=03 30 -_-------- -

mn - J I cn 1o0- - Unacceptable Region"1

. . 3 4 ...

1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

E-- 0 Years -e- 5 Years -- )- 10 Years -,- 15Years R 20 Years (Figure data points are from Table 4-18)

S023 SFP CRITICALITY ANALYSIS Page 74 of 121 Figure 4-16 REGION 11 MINIMUM BURNUP FOR CATEGORY 11-10 FUEL (FILLER ASSEMBLY WITH 5 BORATED SS GUIDE TUBE INSERTS) 1 (D

E E

CD, Ii-2 3 4 5 Initial U-235 Enrichment (w/o) 0 Years -e- 5 Years -*- 10 Years ý, 15Years -- E 20 Years (Figure data points are from Table 4-19)

S023 SFP CRITICALITY ANALYSIS Page 75 of 121 Figure 4-17 REGION II MINIMUM BURNUP FOR CATEGORY I-1 FUEL (FILLER ASSEMBLY WITH 5 BORATED SS GUIDE TUBE INSERTS) 50 "40-(.9

=a30-E-3 E 20 U)

U- 10Q 1.5 2 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

-~ 0 Years -- 9 5 Years --- X 10 Years -vz 15 Years ---- 20 Years (Figure data points are from Table 4-20)

S023 SFP CRITICALITY ANALYSIS Page 76 of 121 Figure 4-18 REGION 11 MINIMUM BURNUP FOR CATEGORY 11-12 FUEL (FILLER ASSEMBLY WITH 3 BORATED SS GUIDE TUBE INSERTS) 60° ..

50-r .

0 4

• -4o CL m30 E

• 20 UL 10 0

1 2 3 4 5 Initial U-235 Enrichment (w/o)

SOYears -e- 5 Years E 10 Years - 15 Years - 20 Years (Figure data points are from Table 4-21)

S023 SFP CRITICALITY ANALYSIS Page 77 of 121 Figure 4-19 REGION H1 MINIMUM BURNUP FOR CATEGORY 11-13 FUEL (FILLER ASSEMBLY WITH NO INSERTS) 70 60 E 40 c30 E

(I,

< 20-75 10 0

1 2 3 4 5 Initial U-235 Enrichment (w/o)

--- 0 Years e 5 Years ---- 10 Years -*- 15 Years -s- 20 Years (Figure data points are from Table 4-22)

S023 SFP CRITICALITY ANALYSIS Page 78 of 121 Figure 4-20 REGION 11 MINIMUM BURNUP FOR CATEGORY 11-14 FUEL (4.80 W/O ASSEMBLY DEPLETED TO 18.0 GWD/T) 20 2.5 3 3.5 4 4.5 5 Initial U-235 Enrichment (w/o)

-ý OYears -e- 5Years - 10Years - 15Years Ei 20Years (Figure data points are from Table 4-23)

S023 SFP CRITICALITY ANALYSIS Page 79 of 121 Figure 4-21 REGION 11 MINIMUM BURNUP FOR CATEGORY H-15 FUEL (ASSEMBLY WITH FULL-LENGTH, 5-FINGER CEA) 2.2 2.7 3.2 3.7 4.2 4.7 Initial U-235 Enrichment (w/o)

E~- OYears e 5Years - lOYears -ý- 15Years 11 20Years (Figure data points are from Table 4-25)

S023 SFP CRITICALITY ANALYSIS Page 80 of 121 Figure 4-22 REGION I BOUNDARY BETWEEN ALL CELL STORAGE AND CHECKERBOARD STORAGE (VALUES ARE U-235 W/O) 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 Empty 2.47 2.47 2.47 2.47 Empty 4.80 Empty 2.47 2.47 2.47 4.80 Empty 2.47 2.47 2.47 2.47 II II Interface 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.27 2.47 2.47 2.47 2.47 2.27 2.65 2.27 2.47 2.47 2.47 2.65 2.27 2.47 2.47 2.47 2.47 1i i1 Interface Note: (1) A row of empty cells can be used at the interface to separate the configurations.

(2) It is acceptable to replace an assembly with an empty cell.

S023 SFP CRITICALITY ANALYSIS Page 81 of 121 Figure 4-23 REGION I BOUNDARY BETWEEN ALL CELL STORAGE AND I OUT OF 4 STORAGE (VALUES ARE U-235 W/O) 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 2.47 1.71 1.71 1.71 2.47 2.47 2.47 1.71 4.80 1.71 2.47 2.47 2.47 1.71 1.71 1.71 2.47 2.47 2.47 II 11 Interface Note: (1) A row of empty cells can be used at the interface to separate the configurations.

(2) It is acceptable to replace an assembly with an empty cell.

S023 SFP CRITICALITY ANALYSIS Page 82 of 121 Figure 4-24 REGION I BOUNDARY BETWEEN CHECKERBOARD STORAGE AND 1 OUT OF 4 STORAGE (VALUES ARE U-235 W/O) 4.80 Empty 4.80 Empty 4.80 Empty Empty 4.80 Empty 4.80 Empty 4.80 4.80 Empty 4.80 Empty 4.80 Empty Empty 1.71 Empty 4.80 Empty 4.80 1.71 4.80 1.71 Empty 4.80 Empty 1.71 1.71 Empty 4.80 Empty 4.80 II I Interface 2.65 2.27 2.65 2.27 2.65 2.27 2.27 2.65 2.27 2.65 2.27 2.65 2.65 2.27 2.65 2.27 2.65 2.27 1.71 1.71 1.71 2.65 2.27 2.65 1.71 4.80 1.71 2.27 2.65 2.27 1.71 1.71 2.65 2.27 2.65 II II Interface Note: (1) A row of empty cells can be used at the interface to separate the configurations.

(2) It is acceptable to replace an assembly with an empty cell.

S023 SFP CRITICALITY ANALYSIS Page 83 of 121 Figure 4-25 REGION II BOUNDARY BETWEEN ALL CELL STORAGE AND CHECKERBOARD STORAGE (VALUES ARE U-235 W/O) 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 Blocked 1.23 1.23 1.23 1.23 Blocked 4.80 Blocked 1.23 1.23 1.23 4.80 Blocked 1.23 1.23 1.23 1.23 II II Interface 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 0.94 1.23 1.23 1.23 1.23 0.94 1.56 0.94 1.23 1.23 1.23 1.56 0.94 1.23 1.23 1.23 2.47 II II Interface Note: (1) A row of empty cells can be used at the interface to separate the configurations.

(2) It is acceptable to replace an assembly with an empty cell.

S023 SFP CRITICALITY ANALYSIS Page 84 of 121 Figure 4-26 REGION H BOUNDARY BETWEEN ALL CELL STORAGE AND 3 OUT OF 4 STORAGE (VALUES ARE U-235 W/O) 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 Blocked 1.23 Blocked 1.23 1.23 1.23 1.80 1.80 1.23 1.23 1.23 1.23 Blocked 1.80 Blocked 1.23 1.23 1.23 II 11 Interface Note: (1) A row of empty cells can be used at the interface to separate the configurations.

(2) It is acceptable to replace an assembly with an empty cell.

S023 SFP CRITICALITY ANALYSIS Page 85 of 121 Figure 4-27 REGION II BOUNDARY BETWEEN CHECKERBOARD STORAGE AND 3 OUT OF 4 STORAGE (VALUES ARE U-235 W/O) 1.80 Blocked 1.80 Blocked 1.80 Blocked 1.80 1.80 1.80 1.80 1.80 1.80 1.80 Blocked 1.80 Blocked 1.80 Blocked Blocked 1.80 Blocked 1.80 1.80 1.80 4.80 Blocked 1.80 Blocked 1.80 Blocked Blocked 4.80 Blocked 1.80 1.80 1.80 Intr Interface II Note: (1) A row of empty cells can be used at the interface to separate the configurations.

(2) It is acceptable to replace an assembly with an empty cell.

S023 SFP CRITICALITY ANALYSIS Page 86 of 121 Figure 4-28 REGION HI BOUNDARY REQUIREMENTS FOR 1 OUT OF 9 STORAGE AND ALL CELL STORAGE (VALUES ARE U-235 W/O) 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 1.23 Filler Filler Filler 1.23 1.23 1.23 Filler A Filler 1.23 1.23 1.23 Filler Filler Filler 1.23 1.23 1.23 II Interface Where: (1) If A = 2.51 w/o, Filler= 1.05 w/o or 1.59 w/o + 5 Inserts.

(2) If A = 4.80 w/o, Filler = 1.03 w/o + 5 Inserts.

(3) If A = 4.80 w/o + 5 Inserts, Filler = 1.59 w/o + 5 Inserts.

(4) If A = 4.80 w/o + 5 Inserts, Filler = 1.32 w/o + 3 Inserts.

(5) If A = 4.80 w/o + 5 Inserts, Filler = 1.05 w/o.

S023 SFP CRITICALITY ANALYSIS Page 87 of 121 Figure 4-29 REGION II BOUNDARY REQUIREMENTS FOR I OUT OF 9 AND CHECKERBOARD STORAGE (VALUES ARE U-235 W/O) 0.94 1.56 0.94 1.56 0.94 1.56 1.56 0.94 1.56 0.94 1.56 0.94 0.94 1.56 0.94 1.56 0.94 1.56 Filler Filler Filler 0.94 1.56 0.94 Filler A Filler 1.56 0.94 1.56 Filler Filler Filler 0.94 1.56 0.94 Interface Where: (1) If A = 2.51 w/o, Filler = 1.05 w/o or 1.59 w/o + 5 Inserts.

(2) If A = 4.80 w/o, Filler = 1.03 w/o + 5 Inserts.

(3) If A = 4.80 w/o + 5 Inserts, Filler = 1.59 w/o + 5 Inserts.

(4) If A = 4.80 w/o + 5 Inserts, Filler = 1.32 w/o + 3 Inserts.

(5) If A = 4.80 w/o + 5 Inserts, Filler = 1.05 w/o.

S023 SFP CRITICALITY ANALYSIS Page 88 of 121 Figure 4-30 REGION II BOUNDARY REQUIREMENTS FOR I OUT OF 9 AND CHECKERBOARD STORAGE (VALUES ARE U-235 W/O)

Blocked 4.80 Blocked 4.80 Blocked 4.80 4.80 Blocked 4.80 Blocked 4.80 Blocked Blocked 4.80 Blocked 4.80 Blocked 4.80 Filler Blocked Filler Blocked 4.80 Blocked Filler A Blocked 4.80 Blocked 4.80 Filler Filler Filler Blocked 4.80 Blocked In Interf ace Where: (1) If A = 2.51 w/o, Filler = 1.05 w/o or 1.59 w/o + 5 Inserts.

(2) If A = 4.80 w/o, Filler = 1.03 w/o + 5 Inserts.

(3) If A = 4.80 w/o + 5 Inserts, Filler = 1.59 w/o + 5 Inserts.

(4) If A = 4.80 w/o + 5 Inserts, Filler = 1.32 w/o + 3 Inserts.

(5) If A = 4.80 w/o + 5 Inserts, Filler = 1.05 w/o.

S023 SFP CRITICALITY ANALYSIS Page 89 of 121 Figure 4-31 REGION II BOUNDARY REQUIREMENTS FOR I OUT OF 9 AND 3 OUT OF 4 STORAGE (VALUES ARE U-235 W/O) 1.80 Blocked 1.80 Blocked 1.80 Blocked 1.80 1.80 1.80 1.80 1.80 1.80 1.80 Blocked 1.80 Blocked 1.80 Blocked Filler Filler Filler 1.80 1.80 1.80 Filler A Filler Blocked 1.80 Blocked Filler Filler Filler 1.80 1.80 1.80 In Interface Where: (1) If A = 2.51 w/o, Filler= 1.05 w/o or 1.59 w/o + 5 Inserts.

(2) If A = 4.80 w/o, Filler = 1.03 w/o + 5 Inserts.

(3) If A = 4.80 w/o + 5 Inserts, Filler = 1.59 w/o + 5 Inserts.

(4) If A = 4.80 w/o + 5 Inserts, Filler = 1.32 w/o + 3 Inserts.

(5) If A = 4.80 w/o + 5 Inserts, Filler = 1.05 w/o.

S023 SFP CRITICALITY ANALYSIS Page 90 of 121 Figure 4-32

SUMMARY

OF REGION I BOUNDARY REQUIREMENTS All Cell Checkerboard 1 out of 4 All Cell N/A Figure 4-22 Figure 4-23 Checkerboard Figure 4-22 N/A Figure 4-24 1 out of 4 Figure 4-23 Figure 4-24 N/A

SUMMARY

OF REGION II BOUNDARY REQUIREMENTS All Cell Checkerboard 3 out of 4 1 out of 9 All Cell N/A Figure 4-25 Figure 4-26 Figure 4-28 Checkerboard Figure 4-25 N/A Figure 4-27 Figure 4-29 Figure 4-30 3 out of 4 Figure 4-26 Figure 4-27 N/A Figure 4-31 1 out of 9 Figure 4-28 Figure 4-29 Figure 4-31 N/A Figure 4-30

S023 SFP CRITICALITY ANALYSIS Page 91 of 121

5. SOLUBLE BORON REQUIREMENTS This analysis takes credit for soluble boron in the spent fuel pool for both normal and accident conditions.

The total soluble boron required to maintain Keff less than 0.95, including all biases and uncertainties, under non-accident conditions, is 970 ppm. This result has the following components:

(Section 5.1) (1) Keff less than or equal to 0.95 - 370 ppm (Section 5.2) (2) Reactivity equivalencing uncertainty - 178 ppm (Section 5.3) (3) Discharge burnup uncertainty - 218 ppm (Section 5.4) (4) Soluble boron measurement uncertainty - 50 ppm (Section 5.5) (5) Margin for future requirements - 154 ppm TOTAL 970 ppm (Section 5.6) The total soluble boron required to maintain Keff less than 0.95, including all biases and uncertainties, under accident conditions, is 1,700 ppm, with the exception of boron dilution. A spent fuel pool boron concentration of 2000 ppm is used to cover both accident conditions and a concurrent boron dilution.

5.1 Keff LESS THAN OR EQUAL TO 0.95 The soluble boron concentration needed to maintain Keff less than or equal to 0.95, including biases and uncertainties, under non-accident conditions is 370 ppm. This amount of soluble boron does not include uncertainties in reactivity equivalencing, discharge burnup, and soluble boron measurement. This value was determined from a whole-pool (8 modules) KENO-V.a model.

Region I and II unrestricted and SFP peripheral fresh enrichments from Section 4 above were used.

5.2 REACTIVITY EQUIVALENCING UNCERTAINTY The soluble boron needed to compensate for reactivity equivalencing uncertainties is 178 ppm. The reactivity equivalencing uncertainty is 0.00 Delta-k at 0 GWD/T and 0.01 Delta-k at 30 GWD/T, linear with burnup.

In previous SCE analyses, which did not credit soluble boron, the reactivity equivalencing uncertainty was used to adjust the reactivity acceptance criteria. Since this analysis credits soluble boron, the reactivity equivalencing uncertainty is expressed in ppm units. This reactivity equivalencing uncertainty is the industry standard practice submitted to and approved by the NRC.

S023 SFP CRITICALITY ANALYSIS Page 92 of 121 5.3 DISCHARGE BURNUP UNCERTAINTY The soluble boron needed to compensate for the fuel assembly discharge burnup uncertainty is 218 ppm.

This result is based on a discharge burnup uncertainty of 7% for SONGS Units 2 and 3 fuel assemblies.

For SONGS Unit I assemblies, the discharge burnup uncertainty is 10%. However, the higher discharge burnups of SONGS Units 2 and 3 assemblies make these the assemblies to use to evaluate the soluble boron requirement.

5.4 SOLUBLE BORON MEASUREMENT UNCERTAINTY Previous SONGS spent fuel pool criticality analyses have assumed the soluble boron measurement uncertainty is 50 ppm. This conservative assumption is used in this analysis also.

5.5 MARGIN FOR FUTURE REQUIREMENTS A conservative allowance of 154 ppm is reserved for future requirements.

5.6 ACCIDENT CONDITIONS The total soluble boron concentration needed to maintain Keff less than or equal to 0.95, including all biases and uncertainties, under accident conditions is 1,700 ppm. This is based on misloading a single 4.8 w/o fresh fuel assembly in Region II. The fuel misloading accident is more severe than the pool heat up accident.

5.6.1 Pool Heat-Up Accident The amount of soluble boron required for the pool heat up accident is 125 ppm.

The pool heat-up accident was evaluated using CASMO-3. Infinitely large Regions I and II were separately considered, and the larger of the two region results is the final result.

S023 SFP CRITICALITY ANALYSIS Page 93 of 121 The pool heat up accident considers the temperature range from 50 degrees F to 248 degrees F + 10% void.

5.6.2 Fuel Mishandling Accident The amount of soluble boron required for the fuel mishandling accident is 730 ppm.

The fuel mishandling accident was evaluated with a whole-pool model. A single fresh 4.8 w/o fuel assembly was placed in different rack locations and storage patterns. It was discovered that Region 11 storage patterns which include an empty storage location into which the misload might occur would have un-acceptably large soluble boron requirements.

Thus it was decided to only allow these patterns if the empty cell is blocked to preclude the misloading accident. (Region I empty cell patterns are acceptable.)

After excluding Region II empty cell patterns, the fuel misloading event which produces the largest increase in spent fuel pool Keff is misloading in Region 1, 1.56 w/o x 0.94 w/o checkerboard pattern (Table 4-13). Starting at Keff = 0.95 with 370 ppm, this accident requires an additional 730 ppm to maintain Keff less than or equal to 0.95, including all biases and uncertainties.

S023 SFP CRITICALITY ANALYSIS Page 94 of 121

6. REFERENCES

1. San Onofre Nuclear Generating Station Units 2 and 3 Updated Final Safety Analysis Report, Revision 27, Chapter 9, Docket Nos. 50-361 and 50-362.

2. Spent Fuel Pool Reracking Licensing Report, Revision 6, Southern California Edison, San Onofre Nuclear Generating Station Units 2 and 3, February 16, 1990.

3. (A) Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, B. K. Grimes, April 14, 1978, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," as amended by the NRC letter dated January 18, 1979 (B) USNRC, Office Of Nuclear Reactor Regulation, Reactor Systems Branch, 1998, "Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light-Water Reactor Power Plants"

4. WCAP-14416-NP-A, Rev 1, Westinghouse Electric Corporation, November 1996, "Westinghouse Spent Fuel Rack Criticality Analysis Methodology"

5. NRC Letter to Westinghouse Owners Group, October 25, 1996, "Acceptance For Referencing Of Licensing Topical Report WCAP-14416-P,

'Westinghouse Spent Fuel Rack Criticality Analysis Methodology (TAC No. M93254)'

6. NRC To Westinghouse Letter Dated July 27, 2001 "Non-Conservatisms In Axial Burnup Biases For Spent Fuel Rack Criticality Analysis Methodology"

7. NRC to SCE Letter Dated October 3, 1996, "Issuance Of Amendment For San Onofre Nuclear Generating Station, Unit No. 2 (TAC No.

M94624) and Unit No. 3 (TAC No. M94625)"

8. ANSI/ANS-57.2-1983, "American National Standard Design Requirements For Light Water Reactor Spent Fuel Storage Facilities At Nuclear Power Plants"

9. NUREG/CR-6683, ORNL/TM-2000/230, Oak Ridge National Laboratory, September 2000 "A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage"

SO23 SFP CRITICALITY ANALYSIS Page 95 of 121

10. CCC-545, RSIC Computer Code Collection, Oak Ridge National Laboratory "SCALE 4.3 Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation for Workstations and Personal Computers"

11. STUDS VIK/NFA-89/3, User's Manual, Studsvik AB, 1989 "CASMO-3 Fuel Assembly Burnup Program"

12. STUDSVIK/SOA-92/01, User's Manual, Studsvik AB, 1992 "SIMULATE-3 Advanced Three-Dimensional Two-Group Reactor Analysis Code"

13. BAW-1484-7, The Babcock & Wilcox Company, July, 1979 "Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel,"

14. SCR-607, Sandia Corporation, March 1963 "Factors For One-Sided Tolerance Limits And For Variables Sampling Plans"

15. Nuclear Technology, Volume 50, September 1980, "Dissolution And Storage Experiment With 4.75-wt% 2 35U-Enriched U0 2 Rods"

16. NUREG/CR-0073, Battelle Pacific Northwest Laboratories, 1978, "Critical Separation Between Subcritical Clusters of 4.29 wt% 235U Enriched U0 2 Rods In Water With Fixed Neutron Poisons"

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

18. SCE to NRC Letter dated December 6, 1995, "Docket Nos. 50-361 And 50-362 Amendment Application Nos. 153 and 137 Storing Nuclear Fuel, San Onofre Nuclear Generating Station, Units 2 And 3" Attachment E - "Evaluation Of The Handling And Storage Of 4.8 w/o Enriched Fuel",

October 18, 1995

19. NRC Information Notice 05-13, May 17, 2005, "Potential Non-Conservative Error in modeling in the KENO-V.a Criticality Code"

S023 SFP CRITICALITY ANALYSIS Page 96 of 121 APPENDIX A SPENT FUEL RACK DIAGRAMS

S023 SFP CRITICALITY ANALYSIS Page 97 of 122 Figure A- 1 REGION 1 ESFP Water SFP Structure UCladding / Guide Tube lFuel Pin

S023 SFP CRITICALITY ANALYSIS Page 98 of 122 Figure A-2 REGION 2 SFP Water

• SFP Structure Cladding / Guide Tube

- Fuel Pin

SEIDEEIDEDODDOEIDD DOOODOEIDDOODOON M11110EIDDOE100000D ODOODEIDEIDOE10 DOE MOODOEIDETIDOODEE EIDOET11:11:1DODOOE]EIN MDOEIDDETIODE11:10EIE, DODEDEIDDE]EIDEIDEIM NEIDEIDEIDDEEIE10E]EIE, ElElElOOODDEEIDOEIDE MOODODOODEIDOODE DEJEJEJEJOETIE100EIDE] U) 0)

MEJEJE]DEUEJOETIOE10 E' DEJODEJOE100000 NODDEEIE]EIDDOODEE EIDI:IEIDEIEI[:IEIEIEII:IEIEIM 001:10EIDE]DEDETIE]EID E]EIE1EIE1[JE1OI:1ElE3EIDOE (L CL MOODOETIEDEDOODD OEIII DO OEIDEll 1111 El 0 M 0 MOEIDOEIDDETIDEIDEll 01100001:1EIDEIDEIDOM C-4) MF]FJFFTJDEJOM_ PE-IE]Fj o) 0)

MI]EIDEIDDOODETIODE ElODEEIDE10EIDEIDOOM MODEIDEIDOOEEIDOEIE EDEDEJEJEODDOEIDEIM MEIE]EIE:IEIEIEIEIEIEIEI[:IEE] EIDDOODOEIDE11:1000M MOETIEDE111DOOODE111 DDOODE1001:1EIETIDOM ME__1ODEIE:IEQEI0QO1ý1EIE [:IEIEIEIEII:IEIEIEIEII]o1:1 MODEEDDI]EIDDOODE ODEI ETI EIE1OF_10[__1El [:][:] 0 NEE10DOODEIDDEDOE [:]Ell:] 1:111 ODDE10 00 OD 0 0 0 EXIDDEDDEEII]DOODE .2 .2 MODDEIDE Ell E100011 E DEIDEDOEIDEIDOODIIN 0) 0)

EDEIEDE:11:10DEIDEIDOE DE10001:11--lOOF-11:1DEIDE 0D 0)

ME]EIDDOI:IEDEIDEIDEIL'OEIDDEJOEIDEJE]DEIE]EIS ME1001:100000OF-IDDE ODD DO EIF-IEIEI--IEEI DOE NOEIDEIDE] Ell] [--][:]Ell:] 1:11:1 ElOO DO ODOEODO El[:1 M MFII-IFIFI[-]FIEII--ýFIFI[-]FýFýE EEII-I FIF] FIEDD-P []FIF] FIE 01:10DEIDOODF-IODEID11 Ell__IEIEIE]IJEJOEI0OO0[:1M

-MEI]1:11:11:1 El ODE] DOE El El EODOODOODODOO MEIE]E][:][:IEI:IEIEIF-11:11:11:]LI 01:10OODE1001100EIDE 0 LL ME]EIDEIEF-1011DEIDDEE Ell:][:] EEI Q [:][:][:] Q EIQ Q [IN o co 00 11 EIDE] D F-10 El ODD ED ODOODOO [:11:10 0 0 ME] El DEI[--]El 1:11:1 El 1:111E] E E EIE1O DO 0000EIF-10 Ell] M 01:111E]EEIETID 0DODE E DOODOODDE1001:1000 ME]EIDEIDDOODDEJOEE, DOOF-IODODOODOODE MOODDEDOF-10E]EIDEE, [:IF-]O DO 01=0000 [--ILI M MODE101:1001:111EIDEED 001:11:10001:1DE]EIDEIDE SEIE]EIIIE]EICEII:11:]Ell:llll:I 01:10 EQ EDOOE11:10 Ell] 0 0000110CEIDEIEDEID0 ED 1:11--][:] El 01:11] Q DO El El 0 MFI F1 FIFE-1 1:1 PFl I-IFIL-11-11 111-7 FlIF] FIFI 1-1 FIF-TI F] FID FI11M Flamm MONSEENS m:lm.mlmmm.mmmmmf-i onismommomm-om Damommommmom mmmmmmmmmmml--]

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S023 SFP CRITICALITY ANALYSIS Page 100 of 121 APPENDIX B REGION I KENO-V.a MODEL

S023 SFP CRITICALITY ANALYSIS Page 101 of 121 SONGS UNITS 2 AND 3 FUEL ASSEMBLY DATA Fuel Pellet O.D. = 0.325 inches Fuel Pellet Density = 10.5216 gms/cc (0.96

• 10.96 gms/cc)

Clad I.D. = 0.332 inches Clad O.D. = 0.382 inches Clad material = Zircaloy Guide Tube I.D. = 0.90 inches Guide Tube O.D. - 0.98 inches Guide Tube Material = Zircaloy Number Of Guide Tubes = 5 Fuel Rod Array = 16x16 Fuel Rod Pitch = 0.506 inches Iif il

S023 SFP CRITICALITY ANALYSIS Page 102 of 121 REGION I KENO-V.a MODEL

<-- 10.28192 cm ->

<-- 10.97280 cm ------ >

<---11.62558 cm --------------- >

<---13.20800 cm---------------------

< ---- 11.37285 cm >(Origin to center of Boraflex hole)

BORAFLEX REPLACED BY WATER I I

- I - I I I I I I I I I I I I I I I .1 I I I I I I I I I I I I I I I I I I I I I I I I I I I x I I I I I I I I I I FUEL ASSEMBLY I I I I I I I I I I I I I I I I I I I I I I I

- I - I I I I I I I I BORAFLEX REPLACED BY WATER I I I I (SS BOX WALL + POCKET + WRAPPER)

(WATER GAP) 1/2/2(Center-to-center) (2.54)(10.40)/2 3.20800 cm 1/2A(Cell+Wall+Pocket+Wrap) = (2.54)(8.64/2 + 0.11

+ 0.127 +0.02) = 11.62558 cm 1

1/2(Storage cell Id) = (2.54) (8.64)/2 = 10.97280 cm 1/2/(Fuel Assembly) = (2.54) (16)(0.506)/2 = 10.28192 cm 1

/2(BFLX thick) = (2.54)(0.095)/2 = 0.12065 cm 1/2/2(BFLX width) = (2.54)(7.522)(0.96)/2 = 9.17082 cm Origin to BFLX Hole = (2.54) (8.64/2 + 0.11 + 0.095/2) = 11.37285 cm

S023 SFP CRITICALITY ANALYSIS Page 103 of 121 CELLDAN PROGRAM - CELL CONSTANTS AND DANCOFF FACTOR SCE VERSION Ri 10-09-97 DATE OF RUN - 12 :50 :47 05/01/**

CE 16 X 16 FUEL FUEL OD ..... .3250 IN. .8255 CM RADIUS = .4127 CM CLAD ID ..... .3320 IN. .8433 CM RADIUS = .4216 CM CLAD OD ..... .3820 IN. .9703 CM RADIUS = .4851 CM

/ PITCH ROD PITCH .. .5060 IN. 1.2852 CM .6426 CM ZIRCONIUM CLAD - SQUARE GEOMETRY FUEL DENSITY ........... 10.522 G/CC U02 ENRICHMENT ............. 2.470 WT% U-235 MODERATOR TEMPERATURE 20 DEGREES C DENSITY FACTOR ......... 1.000 WATER DENSITY .......... .99823 G/CC HYDROGEN NUMBER DENSITY =* .066740 OXYGEN NUMBER DENSITY == .033370 IN MODERATOR U-235 NUMBER DENSITY = 5.87114E-04 U-238 NUMBER DENSITY = .022890 OXYGEN NUMBER DENSITY = .046954 IN FUEL NO OF FUEL RODS 236 NO THIM AND GUIDE TUBES 5 OVERALL CELL DIM, IN. 8.0960 THIMBLE ID, INCHES .9000 DANCOFF HOMOG FACTOR-i 3.5590 THIMBLE OD, INCHES .9800 DANCOFF HOMOG FACTOR-2 3.2200 ID/OD IN CM 1.1430/1.2446 HOMOGENIZED DENSITIES AT A WATER-TO-FUEL VOL RATIO OF 1.936 U235 = 1.7537E-04 U238 = 6.8371E-03 CLAD = 4.6736E-03 HYDROGEN = 3.8598E-02 OXYGEN = 3.3324E-02 U-235 SCAT/A-238 = .2693 MOD MEAN FREE PATH = .6689 OXYGEN SCAT/A-238 = 7.6924 CLAD MEAN FREE PATH = 3.7925 SIG-PO 17.4617 DANCOFF FACTORS:

FIRST ROW (PER ROD) .0490 SECOND ROW (PER ROD) .0133 CORR. - OUTER ROWS .013618 DANCOFF FACTOR .2229

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S023 SFP CRITICALITY ANALYSIS Page 108 of 121 CALCULATION OF REGION I FINAL Keff FOR UNRESTRICTED STORAGE KENO-V.a Result 0.96400 + 0.00063 (Keff + sigma)

Bias 0.00814 Delta-k Bias Uncertainty 0.00172 Delta-k K95/95 For 500 Cases = 1.763 Pool Temperature Bias 0.00914 Delta-k Tolerances: SS Thickness 0.00518 Delta-k Storage Cell ID 0.00531

.Storage Cell Pitch 0.00807 Enrichment 0.00774 Eccentric Loading 0.00767 Final Keff = 0.96400 + 0.00814 + 0.00914

+ SQRT[(1.763*0.00063)2 +(0.00172) 2

+(0.00518 )2

+(0.00531 )2

+(0.00807 )2

+(0.00774 )2

+(0.00767 )2

= 0.99687

S023 SFP CRITICALITY ANALYSIS Page 109 of 121 APPENDIX C REGION 1I KENO-V.a MODEL

S023 SFP CRITICALITY ANALYSIS Page 110 of 121 SONGS UNITS 2 AND 3 FUEL ASSEMBLY DATA Fuel Pellet O.D. = 0.325 inches Fuel Pellet Density = 10.5216 gms/cc (0.96

• 10.96 gms/cc)

Clad I.D. = 0.332 inches Clad O.D. = 0.382 inches Clad material = Zircaloy Guide Tube I.D. = 0.90 inches Guide Tube O.D. = 0.98 inches Guide Tube Material = Zircaloy Number Of Guide Tubes = 5 Fuel Rod Array = 16x16 Fuel Rod Pitch = 0.506 inches

S023 SFP CRITICALITY ANALYSIS Page III of 121 REGION II KENO-V.a MODEL

<-- 10.28192 cm ->

<-- 10.96010 cm--------

<---11.23950 cm --------------- >

< - 11.19913 cm-------- >

(Origin to center of Bflex Sheet Replaced By H2 0) 1/2 BFLEX REPLACED BY H2 0 9.17082 cm I - - -- - -- - -- -- -- - - I ---

I FUE ASSEMBL v

1ABFUEXELASSEDBLY H20

<--> = 1/2 (Boraflex thickness) = 0.07874 cm 1/2(Center-to-center) = (2.54) (8.85)/2 = 11.23950 cm 1/2/2(Storage cell Id) = (2.54) (8.63)/2 = 10.96010 cm 1/2(Fuel Assembly) = (2.54)(16)(0.506)/2 = 10.28192 cm 12(BFLX thick) = 2.54) (0.062)/2 = 0.07874 cm 1/2A(BFLX width) = 2.54)(7.522)(0.96)/2 = 9.17082 cm Origin to BFLX Hole = 11.23950 - 0.07874/2 - 0.001 = 11.19913 cm

S023 SFP CRITICALITY ANALYSIS Page 112 of 121 CELLDAN PROGRAM - CELL CONSTANTS AND DANCOFF FACTOR SCE VERSION Ri 10-09-97 DATE OF RUN - 08:03:26 04/27/**

CE 16 X 16 FUEL FUEL OD ..... 3250 IN. .8255 CM RADIUS = .4127 CM CLAD ID ..... 3320 IN. .8433 CM RADIUS = .4216 CM CLAD OD ..... 3820 IN. .9703 CM RADIUS = .4851 CM ROD PITCH .. .5060 IN. 1.2852 CM 1/2 PITCH = .6426 CM ZIRCONIUM CLAD - SQUARE GEOMETRY FUEL DENSITY ........... 10.522 G/CC U02 ENRICHMENT ............. 1.230 WT% U-235 MODERATOR TEMPERATURE .. 20 DEGREES C DENSITY FACTOR ......... 1.000 WATER DENSITY .......... .99823 G/CC HYDROGEN NUMBER DENSI TY = .066740 OXYGEN NUMBER DENSI TY = .033370 IN MODERATOR U-235 NUMBER DENSI TY = 2.92374E-04 U-238 NUMBER DENSI TY = .023181 OXYGEN NUMBER DENSI TY = .046947 IN FUEL NO OF FUEL RODS 236 NO THIM AND GUIDE TUBES 5 OVERALL CELL DIM, IN. 8.0960 THIMBLE ID, INCHES .9000 DANCOFF HOMOG FACTOR-1 3.5590 THIMBLE OD, INCHES .9800 DANCOFF HOMOG FACTOR-2 3.2200 ID/OD IN CM 1.1430/1.2446 HOMOGENIZED DENSITIES AT A WATER-TO-FUEL VOL RATIO OF 1.936 U235 = 8.7331E-05 U238 = 6.9241E-03 CLAD = 4.6736E-03 HYDROGEN = 3.8598E-02 OXYGEN = 3.3322E-02 U-235 SCAT/A-238 = .1324 MOD MEAN FREE PATH = .6689 OXYGEN SCAT/A-238 = 7.5946 CLAD MEAN FREE PATH = 3.7925 SIG-PO = 17.2270 DANCOFF FACTORS:

FIRST ROW (PER ROD) .0490 SECOND ROW (PER ROD) .0133 CORR. - OUTER ROWS .013618 DANCOFF FACTOR .2229

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S023 SFP CRITICALITY ANALYSIS Page 118 of 121 CALCULATION OF REGION HI FINAL Keff FOR UNRESTRICTED STORAGE KENO-V.a Result = 0.97086 + 0.00048 (Keff +. sigma)

Bias 0.00814 Delta-k Bias Uncertainty  : 0.00172 Delta-k K95/95 For 500 Cases : 1.763 Pool Temperature Bias : 0.00300 Delta-k Tolerances: SS Thickness - 0.00174 Delta-k Storage Cell ID 0.00331 Enrichment 0.01547 Eccentric Loading 0.00000 Final Keff

, 0.97086 + 0.00814 + 0.00300

+ SQRT[(1.763*0.00048)2 +(0.00172)2

+(0.00174)2

+(0.00331)2

+(0.01547)2

+(0.00000)2

= 0.99803

S023 SFP CRITICALITY ANALYSIS Page 119 of 121 APPENDIX D

SUMMARY

OF TOLERANCES, BIASES, AND UNCERTAINTIES

S023 SFP CRITICALITY ANALYSIS Page 120 of 121 The keff calculation process uses the following equation:

Final keff =: knomila 1 + Bmethod + Btemp + Buncert + BCEA + BAxial where: knominai = KENO-V.a keff Bmethod = method bias determined from benchmark critical experiments Btemp = temperature bias (68 F to 160 F)

Buncert = statistical summation of uncertainty components:

method bias 95/95 uncertainty 1.763

• KENO-V.a sigma SS thickness Storage Cell ID Storage Cell Pitch Enrichment Eccentric Loading BCEA = CEA bias for rodded cases only BAxiaI = Axial Burnup Bias The individual contributions to the final keff were obtained as follows:

Description Value Source KENO-V.a Method Bias: 0.00814 Ak Industry Critical Experiments (Reference 13)

KENO-V.a Method Bias 95/95 Uncertainty: 0.00172 Ak Industry Critical Experiments (Reference 13) knominai (Case Specific) 0.70 - 1.20

• KENO-V.a sigma (Case Specific) 0.0001 - 0.0030

• KENO-V.a k95/95 for 500 neutron generations: 1.763 Reference 14

• Range of expected values.

Normal Pool Temperature Bias Region I (68F - 160F) 0.00914 Ak CASMO-3 Region II (68F - 160F) 0.00300 Ak CASMO-3 CEA Insertion Bias CEAs Present 0.00700 Ak CASMO-3, KENO-V.a CEAs Not Present 0.00000 Ak

S023 SFP CRITICALITY ANALYSIS Page 121 of 121 Axial Burnup Distribution Bias 0.00000 Ak CASMO-3, SIMULATE-3 (0 to 60,000 MWD/T)

Eccentric Loading(4) (Region I)

(2.47 w/o) 0.00767 Ak KENO-V.a (4.80 w/o) 0.01383 Ak KENO-V.a Eccentric Loading(4) (Region II)

(All w/o) 0.00000 Ak KENO-V.a 123 Manufacturing Tolerances( ' ' )

Region I Rack Storage Cell Wall Thickness +0.004" 0.00518 Ak CASMO-3 Rack Storage Cell ID +0.025" 0.00531 Ak CASMO-3 Rack Storage Cell Pitch +0.06" 0.00807 Ak CASMO-3 U235 Enrichment +0.05 w/o 0.00774 Ak CASMO-3 Region 11 Rack Storage Cell Wall Thickness +0.004" 0.00174 Ak CASMO-3 Rack Storage Cell ID +0.025" 0.00331 Ak CASMO-3 Rack Storage Cell Pitch N/A U235 Enrichment +0.05 w/o 0.01547 Ak CASMO-3 Notes: (1) Fuel rod and fuel pellet manufacturing tolerances (dimensions and U0 2 density) are conservatively bounded by using a U0 2 density of 96% of theoretical density and no fuel pellet dishing.

(2) The manufacturing dimensional tolerances are from the Westinghouse criticality analysis report: Westinghouse to SCE Letter PA-89-0075/8924, May 05, 1989,

Subject:

Comprehensive Criticality Report. The U235 enrichment tolerance is the standard DOE value and is included in SCE's enrichment specification.

(3) The tolerances / variations were evaluated over the range of expected soluble boron concentrations.

(4) Eccentric loading means four assemblies moved as close together as possible in the corner where four storage locations meet.

Enclosure 3 DECLARATION OF COMPLIANCE WITH 10 CFR 50.68

Enclosure 3 Declaration of compliance with 10 CFR 50.68 Southern California Edison (SCE) has elected to comply with 10 CFR 50.68, "Criticality Accident Requirements," and will no longer rely on an exemption from the requirements of 10 CFR 70.24. As demonstrated below, upon NRC approval of the license amendment request, SCE will comply with 10 CFR 50.68(b), paragraphs (1) through (7). Within 180 days of NRC issuance of the license amendment, SCE will complete the UFSAR changes described below for compliance with 10 CFR 50.68(b),

paragraph 8.

REQUIREMENT (1): Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

STATEMENT OF Movement and storage is limited in procedures S023-X-7 COMPLIANCE: "Nuclear Fuel Movement for Refueling Cycles" and S023-X-7.2 "Nuclear Fuel Movement - Spent Fuel Pool" to analyzed conditions using approved methodology. Based on the physical design at the San Onofre Nuclear Generating Station (SONGS) described in UFSAR section 9.1.4, fuel is handled and stored in four areas in the fuel storage building:

Up-ender and new fuel elevator pool

2) Storage rack pool

3) Cask pool

4) New fuel storage room The pools are physically separated by walls with gates to allow fuel transfer using the spent fuel handling machine between them. The new fuel storage room is located in a separate room. A single rail connects the new storage room to the up-ender and new fuel elevator pool via the new fuel crane.

The restrictions on handling and storage for each area are as follows:

1) Up-ender and new fuel elevator pool:

Movement: One fuel assembly to be moved at one time with spent fuel handling machine or new fuel elevator or new fuel crane Storage: Two fuel assemblies may be in the pool at one time with:

a) two assemblies in the up-ender and none in motion or, b) one assembly in the up-ender and one or none in motion or, c) one assembly in the new fuel elevator and none in motion.

1 OF 6

Enclosure 3 Declaration of compliance with 10 CFR 50.68

2) Storage rack pool:

Movement: One fuel assembly to be moved at one time with spent fuel handling machine Storage: Assemblies may be stored in spent fuel storage racks in accordance with PCN 556.

3) Cask pool:

Movement: One fuel assembly with the spent fuel handling machine or approved cask with the cask crane to be moved at one time. Fuel loading into the cask is per approved procedure S023-X-9, "Dry Cask Storage Loading".

Storage: One Fuel assembly or approved cask configurations.

4) New Fuel Rack Room Movement: One fuel assembly or new fuel cask to be moved at one time with new fuel crane.

Storage: New Fuel rack or approved cask configurations.

SONGS may move assemblies simultaneously in each of the four areas as restricted above since motions are limited to separate and distinct physical locations by crane and pool designs. For example the Upenders could be loweredand transferred into containment while the spent fuel handling machine is retrieving a fuel assembly from the spent fuel racks.

REQUIREMENT (2): The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

2 OF 6

Enclosure 3 Declaration of compliance with 10 CFR 50.68 STATEMENT OF Per SCE calculation N-1 020-060, "SONGS 2/3 New Fuel COMPLIANCE: Racks - 5.1 w/o", k-effective of the fresh fuel in the fresh fuel storage racks is less than 0.95, at a 95 percent probability, 95 percent confidence level assuming the racks are loaded with fuel of the maximum fuel assembly reactivity allowed by T.S. and flooded with unborated water. In addition at SONGS Units 2 and 3, the maximum U-235 enrichment of the fresh fuel assemblies is limited to 4.8 weight percent (w/o) by Limiting Condition for Operation (LCO) 4.3.1.1 (a) -- Fuel Storage, Criticality. This limit is not changed by Proposed Change Notice (PCN) 556.

REQUIREMENT (3): If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.

STATEMENT OF Per SCE calculation N-1 020-060, "SONGS 2/3 New Fuel COMPLIANCE: Racks - 5.1 w/o", k-effective of the fresh fuel in the fresh fuel storage racks is less than 0.98, at a 95 percent probability, 95 percent confidence level assuming the racks are loaded with fuel of the maximum fuel assembly reactivity allowed by Technical Specification and assuming unborated water ranging in density from 0.0 grams/cc to 1.0 grams/cc (optimum moderation). In addition at SONGS Units 2 and 3, the maximum U-235 enrichment of the fresh fuel assemblies is limited to 4.8 percent by weight by LCO 4.3.1.1 (a) -- Fuel Storage, Criticality. This limit is not changed by PCN 556.

REQUIREMENT (4): If no credit for soluble boron is taken, 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 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.

3 OF 6

Enclosure 3 Declaration of compliance with 10 CFR 50.68 (1) Spent Fuel Racks STATEMENT OF SCE calculation N-1020-159, "SONGS 2 and 3 Spent Fuel COMPLIANCE: Racks - No Boraflex - Regions 1 & 2 Criticality Analyses",

assumes no accidental miss-loadings, PCN 556 loading of assemblies in the spent fuel pool racks, and a boron level greater than 970 ppm which is lower than LCO 3.7.17 which requires 2000 ppm. The result of the analysis is the k-effective of the spent fuel in the spent fuel storage racks is less than 0.95 at a 95 percent probability, 95 percent confidence level.

Also the k-effective will remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

(2) Dry Shielded Canister (24 PT4-OSC) in The Cask Pool Calculation S023-207-16-C15, "Criticality Evaluation of CE 16x1 6 Fuel Assemblies for the Advanced NUHOMS@ System 24-PT4 Day Shielded Canister", assumes no soluble boron in the spent fuel pool/cask pool water, and the cask fully loaded with fresh fuel assemblies (no burnup) with maximum permitted enrichment. The result of this analysis is k-effective less than 0.95 at a 95 percent probability, 95 percent confidence level.

REQUIREMENT (5): The quantity of Special Nuclear Material (SNM), other than nuclear fuel stored onsite, is less than the quantity necessary for a critical mass.

STATEMENT OF At SONGS Units 2 and 3, the quantity of SNM, other than COMPLIANCE: nuclear fuel stored onsite, is less than the quantity necessary for a critical mass. In incore detectors (fission chambers),

there are 34/32 grams of Total U/U235 at Unit 2 and 44/41 gms of Total U/U235 at Unit 3. In neutron sources, there are 1.8 grams of Pu238 at Unit 2 and 1.7 grams of Pu238 at Unit 3.

REQUIREMENT (6): Radiation monitors are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

STATEMENT OF The fuel storage and handling facilities at San Onofre Units 2 COMPLIANCE: and 3 are equipped with a monitoring system for detection of increases in direct gamma and gaseous radiation levels.

4 OF 6

Enclosure 3 Declaration of compliance with 10 CFR 50.68 Area Radiation Monitor The single area radiation monitor for direct gamma radiation in each of the Units 2 and 3 fuel handling buildings functions to: immediately alert personnel entering or working in the area of increasing or abnormally high radiation levels which, if unnoticed, could result in inadvertent overexposure; inform the control room operator of the abnormal radiation increase in the area; and, comply with the requirements of 10 CFR 50, Appendix A, General Design Criteria 63, for monitoring fuel and waste storage and handling areas.

This single area radiation monitor in each fuel handling area is of a non-saturating design so it can register full scale if exposed to radiation levels up to 100 times full-scale indication. The monitor consists of a gamma-sensitive radiation detector transmitting to an alarm-readout unit in the main control room. Indicating alarm units (both audible and visual) are also mounted near the detector location to alert personnel in the area. This monitor views the cask lay down area, the cask hatch area, and a portion of the spent fuel pool. It is located approximately 30 feet from the nearest corner and approximately 90 feet from the farthest corner of the spent fuel pool. The alarm setpoint is fully adjustable throughout the entire range of the rate meter. The setpoint is set high enough to avoid spurious alarms yet set low enough to provide an early warning detection of any significant increase in radiation such as from a fuel handling accident. The nominal setpoint is 2.5 mr/hr; however, this may be increased to avoid spurious alarms if the background increases such as when spent fuel is moved. In addition to a high-level alarm, the detector is provided with an instrument failure alarm to alert the loss of monitoring capability. The radiation level is trended on the radiation monitoring Data Acquisition System (DAS) which is available to the control room operators.

Airborne Monitors The fuel handling airborne monitoring system consists of two one-channel monitors that detect noble gases only. These monitors sample ventilation exhaust from each fuel handling area to the continuous exhaust plenum. The gaseous airborne radiation monitors function to alarm and initiate isolation of the fuel handling area from the normal ventilation system and actuate the fuel handling area clean-up unit if pre-established levels are exceeded. The radiation level is also trended on the radiation monitoring DAS which is available to the control room operators.

Procedures are followed to test instrument performance on a preset frequency based on instrument use, reliability and past experience.

REQUIREMENT (7): The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.

5 OF 6

I~

Enclosure 3 Declaration of compliance with 10 CFR 50.68 STATEMENT OF At SONGS Units 2 and 3, the maximum U-235 enrichment COMPLIANCE: of the fresh fuel assemblies is limited to 4.8 percent by weight by LCO 4.3.1.1 (a) -- Fuel Storage, Criticality. This limit is not changed by PCN 556.

REQUIREMENT (8): The UFSAR is amended no later than the next update which 10CFR50.71 (e) of this part requires, indicating that the licensee has chosen to comply with 10CFR50.68(b).

Upon approval of this License Amendment Request (PCN 556), a statement indicating compliance with §50.68(b) will be included in the next SONGS UFSAR update per procedures:

1) SO1 23-XXX-2.3, "Preparation of Amendment Applications for the SONGS 1 Possession Only Licensing and the SIGNS 2 and 3 Facility Operation Licenses"

2) S0123-XXX-5.2, "Control of Licensing Document Changes" 6 OF 6

Southern California Edison (SCE)

San Onofre Nuclear Generating Station (SONGS), Units 2 and 3 Docket Nos. 50-361 and 50-362 Enclosure 4 Responses to NRC Staff Questions Regarding Proposed Change Notice (PCN) 556

NRC Question 1: Criticality Accident Analyses Page 13 of Enclosure 2 to Reference 1 indicated that specific accidents considered for criticality analyses included fuel assembly drop, fuel misloading in the racks and spent fuel pool (SFP) water temperature changes. Also, Section 3.2.9 of Attachment L to Reference 1 discussed criticality analyses for various postulated accidents including: (1) pool water temperature accident; and (2) fuel assembly misplacement. For the fuel assembly misplacement, the following accidents were discussed: (1) fuel assembly dropped horizontally on top of the rack; (2) fuel assembly dropped vertically into a storage location already containing a fuel assembly; (3) fuel assembly dropped to the (SFP); and (4) fuel misloading in either region I or I1. Furthermore, Section 5.6 of Attachment L to Reference 1 provided the quantitative results of analyses for a pool heat-up accident and fuel mishandling accident. However, the respective Attachment G and H containing the proposed Bases pages for SONGS Units 2 and 3 did not include adequate and sufficient information regarding applicable safety analyses. Specifically, Page B 3.7-71 stated in the applicable safety analyses section that this accident was analyzed assuming the misloading of one fresh assembly with the maximum permissible enrichment. The discussion of other accidents analyzed or considered as discussed in the above paragraph were either deleted (a fuel assembly dropped vertically into a storage location already containing a fuel assembly), or omitted (such as the pool water temperature accident).

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

SCE Response:

Amendment Application Numbers 243 Supplement 1 and 227 Supplement 1, consisting of PCN 556 Revision 1, provide the modified applicable safety analyses section on page B.3.7-71.

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

The proposed LCS 4.0.100 for Units 2 and 3 was contained in Attachment I and J, respectively. Based on the review of the proposed LCSs, it appeared to the NRC staff that LCS.4.0.100 did not contain the analytical methods used to determine the associated storage patterns for the spent fuel assemblies (SFAs) and thus, was not Page 1 of 25

fully consistent with Section 4.3(f) of NUREG-1432 that requires to include analytical methods in the LCS.

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

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

The licensee should correct the error by deleting "Bases" and replacing "Vol. 2" with "Vol. 1" in the title for Reference (5).

SCE Response:

Amendment Application Numbers 243 Supplement 1 and 227 Supplement 1, which consist of PCN 556 Revision 1, provide requested clarification and correct the noted discrepancy.

NRC Question 3: Inconsistency in the Note for LCS 4.0.100 The note on page 4.0.100-1 of Attachment I and J to Reference 1 indicated that the LCS was listed by revision number and date in TS 4.3.1. The NRC staff found that the LCS was listed in TS 4.3.1.k and TS 4.3.1.1 (of Attachment C and D to Reference 1).

However, the TSs listed the LCS revision number without specifying the date.

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

SCE Response:

Amendment Application Numbers 243 Supplement 1 and 227 Supplement 1, which consist of PCN 556 Revision 1, clarify the inconsistency by adding the date to TS 4.3.1.

NRC Question 4: No SONGS 1 Fuel to be stored in Region I Racks LCS Subsection 4.0.100.3 (in Attachment I and J to Reference 1) requires that SONGS Unit 1 fuel not be stored in Region I racks. This requirement is consistent with the criticality analysis discussed in Section 4.6 of Attachment L that indicates that Unit 1 fuel has not been analyzed to be stored in Region I. However, Table 2-2, "Spent Fuel Data (Each Unit)," of Attachment L lists "and/or SONGS 1 14x14" fuel in the row of the Table specifying the fuel types for the SFAs in Region I Page 2 of 25

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

SCE Response:

Amendment Application Numbers 243 Supplement 1 and 227 Supplement 1, which consist of PCN 556 Revision 1, add a Note to Region I for SONGS 1 assemblies.

(SONGS 1 fuel included for historical reasons)

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

Modify the figures to correct the omissions.

SCE Response:

Amendment Application Numbers 243 Supplement 1 and 227 Supplement 1, which consist of PCN 556 Revision 1, provide better quality figures with symbols.

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

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

SCE Response:

CEA Lifetime Analysis: The CEAs qualified by the CEA lifetime analyses for service in the reactor core are also judged to be acceptable for reactivity control service in the spent fuel pool due to the less severe environment and lack of CEA insertion/scram time requirements. Any CEA used in the Spent Fuel Pool for reactivity control would have had to pass the CEA drop time surveillance (Reference 4) in the previous cycle of operation and the measured reactivity would have matched the predicted reactivity effect in one of the previous 2 cycles of operations during low power physics testing (Reference 5). The core service life limits are based on allowable plastic strain with consideration of the fast neutron fluence exposure history and susceptibility to Page 3 of 25

irradiation assisted stress corrosion cracking as described in the SONGS Updated Final Safety Analysis Report (UFSAR) Section 4.2.3.4.

Control Element Assembly CEA lifetime is based on estimates of neutron absorber burnup, allowable plastic strain of the Inconel 625 cladding, and the resultant dimensional clearances of the elements within the fuel assembly guide tubes. Consideration of these factors results in the following CEA lifetime limits: 1) 18.0 E21 neutrons/cm2 fast fluence in2 the silver-indium-cadmium (AglnCd) absorber region, and 2) 2.0 E21 neutrons/cm fast fluence in the B4C absorber region.

Visual Inspection: CEAs which are deemed not suitable for core operation may still be deemed suitable for reactivity control in the spent fuel pool; however, the CEA will have to pass a visual examination for defects. This visual examination is "qualitative"; but if abnormalities are noted, further evaluation will be required which can include quantitative examination of the CEA such as the extent of remaining clad wall thickness via eddy-current testing. SONGS Units 2 and 3 established the requirement that any CEA with over 20% through-wall wear would require a detailed assessment prior to qualification for continued service. This wear threshold criteria for further evaluation for core service will also apply to spent fuel pool reactivity control service.

In 2002, quantitative inspections were performed on the Unit 2 CEAs which had served in Cycles 1 through 7. The results were that minor vibrational wear, or scratching, was detected on most of the CEAs. One CEA indicated a wear/defect of over 50% through wall; however, this CEA was deemed not suitable for core service following Cycle 2.

The highest observed wear measurement for a CEA which is still suitable for service was 5% through-wall.

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

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

SCE Response:

CASMO-3 was approved by the NRC for SCE to perform steady state core design activities in Reference 1. CASMO-3 was benchmarked against measured critical experiments, measured fuel isotopics, and measured pinwise LA-140 distributions. The depletion calculation was validated by comparisons with the Yankee Core-1 and Zion Page 4 of 25

measured uranium and plutonium isotopics which were industry standard benchmark sources. These comparisons were performed for a range of pin-cell spectra and indicated good agreement for the fuel isotopics versus burnup. As further validation, a set of uniform critical measurements was also calculated. CASMO-3 reproduced 74 criticals to within 1% delta-k/k. The average multiplication factor for all 74 criticals is 0.9954 + 0.0088.

SCE benchmarked the CASMO-3 computer program capability for modeling the San Onofre spent fuel rack geometries:

Region I CASMO-3 bias and uncertainty = 0.00000 + 0.00139 Region II CASMO-3 bias and uncertainty = 0.00153 + 0.00152 These results were determined by comparing CASMO-3 against KENO-V.a after KENO-V.a had been benchmarked against Babcock and Wilcox (B&W) critical experiments.

In the calculations supporting this submittal, CASMO-3 was used to determine the following:

1. Depletion of assemblies to generate fuel isotopics for reactivity equivalencing

2. Reactivity variations due to manufacturing tolerances and normal pool temperature variations

3. Initial enrichment versus burnup criteria (reactivity equivalencing)

4. Reactivity variation due to fuel pool heat up event For the depletion calculation, uncertainty of 0.00 Ak at 0 GWD/T and 0.01 Ak at 30 GWD/T, linear with burnup, was used. As described in the response to RAI#14, this uncertainty is consistent with the 5% reactivity decrement approach previously approved in Reference 6.

For the criticality (reactivity) calculations, CASMO-3 was used to generate reactivity changes (Ak) with respect to variations (tolerances, temperature, fuel isotopics) of interest. For Ak calculations, bias and uncertainty in the bias are not needed since they cancel out. In summary, the basis for criticality analysis is still KENO-V.a and appropriate bias and uncertainty in the bias for KENO-V.a was applied. This approach is consistent with past SCE analysis approved in the same NRC letter (Reference 6).

NRC Question 8: Computer Codes (KENO-V.a and the Related Codes)

Benchmarking Section 3.1.2 of Attachment L to Reference 1 indicated that KENO-V.a was benchmarked by Southern California Edison (SCE) against industry critical experiments performed by Babcock and Wilcox (B&W). The bias and 95/95 uncertainty in the bias for CELLDAN, NITAWL-II and KENO-V.a, and 27 group cross section library were 0.00814 and 0.00172, respectively. The NRC staff found that these bias and Page 5 of 25

uncertainty were different from that on page 9.1-13 of the Updated Final Safety Analysis Report (UFSAR), which indicated that the bias and 95/95 uncertainty in the bias for CELLDAN, NITAWL-II and KENO-V.a, and 27 group cross-section library were 0.00928 and 0.00148, respectively. The bias and uncertainty in the UFSAR were also determined by analyses of B&W critical experiments.

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

SCE Response:

The same set of B&W critical experiments were used in the benchmarking. Small differences in the bias and 95/95 uncertainty in the bias for CELLDAN, NITAWL-11 and KENO-V.a are due to the different architecture and precision on different computer platforms. The KENO-V.a bias and uncertainty stated in the UFSAR were determined by executing the computer codes on an old IBM mainframe computer, which has since been retired. For the current submittal, all CELLDAN, NITAWL-11 and KENO-V.a computer cases were executed on a newer IBM AIX workstation. The KENO-V.a bias and uncertainty stated in Attachment L of this submittal are consistent with the use of the KENO-V.a computer code system on the IBM AIX workstation.

As shown in the table below, when appropriate bias and uncertainty are applied, the predicted Keff is generally more conservative than the measured values. This is consistent with the 95/95 probability/confidence.

Page 6 of 25

Comparison of Measured and Predicted Keff for B&W Critical Experiments (Computer Codes executed on IBM AIX computer)

B&W Core Measured Keff KENO-Va Keff Predicted Keff Pred - Meas.

I 1.0002 0.99037 1.00023 0.00003 II 1.0001 0.99278 1.00264 0.00254 III 1.0000 0.99346 1.00332 0.00332 IX 1.0030 0.99044 1.00030 -0.00270 X 1.0001 0.99039 1.00025 0.00015 Xl 1.0000 0.99441 1.00427 0.00427 Xll 1.0000 0.99362 1.00348 0.00348 XlII 1.0000 0.99868 1.00854 0.00854 XIV 1.0001 0.99382 1.00368 0.00358 XV 0.9998 0.98833 0.99819 -0.00161 XVI 1.0001 0.98892 0.99878 -0.00132 XVII 1.0000 0.99234 1.00220 0.00220 XVIII 1.0002 0.99119 1.00105 0.00085 XIX 1.0002 0.99146 1.00132 0.00112 XX 1.0003 0.99085 1.00071 0.00041 XXI 0.9997 0.99256 1.00242 0.00272 Source: Table 3-1, Attachment L of the submittal

• Predicted Keff = KENO-Va Keff + Bias (0.00814) + Uncertainty (0.00172)

The UFSAR values will be updated after this PCN is approved.

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

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

SCE Response:

As shown in Figure 2-2 of Attachment L of the submittal, the Region I rack is composed of individual storage boxes with a water gap between cells. The water gap acts as a flux trap and is a reactivity control feature of the Region I racks. The size of the gap affects the reactivity control of the racks. Manufacturing tolerance for the cell pitch affects the size of the water gap between storage boxes.

Page 7 of 25

The Region II rack is composed of storage boxes arranged in an alternating pattern with storage cells within each storage box and in between storage boxes (Figure 2-3 of Attachment L of the submittal). There are no water gaps among cells. Therefore, there is no need to evaluate the effect of the size of the water gap due to the cell pitch tolerance.

In addition to the cell pitch tolerance, a manufacturing tolerance on the storage cell inside dimension is evaluated for both Region I and Region II racks (see Section 3.2.2 of Attachment L of this submittal).

NRC Question 10: Minimum Axial Burnup Section 3.2.4 of Attachment L to Reference 1 indicated that the CASMO-3 depletions were performed based on the following reactor operating conditions: reactor outlet temperature of 600 OF; constant beginning of cycle fuel temperature of 1200 OF; and constant soluble boron of 1000 parts per million (ppm). It indicated that the operating conditions so selected were to eliminate axial burnup effects.

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

SCE Response:

"Axial Burnup Effect" refers to the non-conservative results due to the 2D approximation that uses the axially averaged assembly burnup values in the criticality calculation. The effect of axial burnup distribution was investigated by SCE and was bounded by performing all CASMO-3 isotopic calculations for reactivity equivalencing at a moderator temperature of 600 OF, which is significantly higher than the core average moderator temperature at about 580 OF. Performing calculations at 600 OF enhances the build-up of Plutonium-239, as compared to the 580 OF results at the same assembly burnup value. The higher Plutonium-239 concentration results in a higher (more conservative) equivalent fresh fuel enrichment.

Maintaining the fuel temperature at 1200 OF, instead of simulating the fuel temperature decrease with burnup, also has a similar effect.

Sections 3.2.5 and 4.4 of Attachment L of the SCE submittal describe the process for the determination of the axial burnup effect (bias) and concluded that, by performing 2D CASMO-3 assembly depletion calculations at a high moderator temperature of 600 F and at a constant fuel temperature of 1200 OF, the conservatism introduced due to enhanced Plutonium-239 build-up is sufficient to offset the axial burnup effect. More discussions will be presented in response to RAI #11.

Page 8 of 25

The boron concentration of 1000 ppm was chosen to bound the cycle average Hot-Full-Power (HFP) critical boron concentration. A slightly higher boron concentration hardens the neutron spectrum and enhances the build-up of Plutonium-239.

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

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

SCE Response:

As indicated in Section 3 of Attachment L of the SCE submittal, both KENO-V.a criticality and CASMO-3 reactivity equivalencing calculations were performed with uniform axial burnup and isotopic (e.g., fissile) distributions. However, physical assemblies have a non-uniform axial burnup and isotopic distribution. The assumption of uniform axial distribution may result in non-conservative reactivity results. To offset the potential non-conservatism, SCE 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 at about 580 OF. The fuel temperature was maintained at 1200 OF, instead of simulating the fuel temperature decrease with burnup. Performing calculations at higher temperatures enhances the build-up of Plutonium-239, as compared to the nominal temperature results at the same assembly burnup value. The higher Plutonium-239 concentration results in a higher (more conservative) equivalent fresh fuel enrichment.

Section 3.2.5 of Attachment L of the SCE submittal discussed the process used to determine whether the use of conservative temperatures can override the "axial burnup effect." The process is summarized below.

Step 1. Construct a single assembly SIMULATE-3 model in 3D geometry for the assembly of interest. Deplete the assembly with the nominal moderator and fuel temperatures to various burnup points of interest. These temperatures produce a realistic, axially-varying isotopic inventory.

Step 2. Construct a single assembly SIMULATE-3 model in 2D geometry for the assembly in Step 1 above. Instead of the nominal moderator and fuel temperatures, the 2D depletion was performed at a constant moderator temperature of 600 OF and a Page 9 of 25

constant fuel temperature of 1200 OF. These constant temperatures produce a more conservative isotopic make-up due to the harder neutron spectrum. The burnup and temperature history were expanded to the 3D geometry consistent with the geometry in Step 1 above. All axial elevations of the fuel will have the same burnup and temperature history. This is consistent with the 2D modeling employed in CASMO-3.

Step 3. Using the results from Steps 1 and 2 above, perform SIMULATE-3 restarts at burnup values of 0, 10, 20, 30, 40, 50, and 60 GWD/T; at 0 ppm; 68 OF; and no xenon to simulate the spent fuel rack conditions. Keff values from the case with the conservative moderator and fuel temperatures (Step 2) and the case with nominal temperatures (Step 1) were compared to verify that the conservative temperature approach was sufficient to override the axial burnup effect.

The SIMULATE-3 results for two different assembly enrichments are summarized below:

1.87 w/o 1.87 w/o Burnup 2D Keff 3D Keff Delta-k (2D - 3D) 0 1.24454 1.24454 0.00000 10 1.11561 1.10809 0.00752 20 1.01562 1.00765 0.00797 30 0.94019 0.93183 0.00836 40 0.88513 0.87637 0.00876 50 0.84777 0.83876 0.00901 60 0.82274 0.81284 0.00990 4.45 w/o 4.45 w/o Burnup 2D Keff 3D Keff Delta-k (2D - 3D) 0 1.45672 1.45671 0.00001 10 1.34135 1.33404 0.00731 20 1.25052 1.24275 0.00777 30 1.16868 1.16048 0.00820 40 1.08984 1.08169 0.00815 50 1.01564 1.00736 0.00828 60 0.94916 0.93970 0.00946 As shown above, the conservative temperature approach (2D) was more conservative.

Therefore, the axial burnup bias is conservatively set at 0.0 delta-k.

NRC Question 12: CEA Bias Section 3.2.6 of Attachment L to Reference 1 discussed the calculations for CEA bias.

It indicated that CASMO-3 (through SIMULATE-3) has accurately predicted SONGS Units 2 and 3 CEA bank worth measurements.

Page 10 of 25

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

SCE Response:

The San Onofre reactors are modeled with CASMO-3 (cross-sections) and SIMULATE-3 (3D two-group nodal code for both PWRs and BWRs).

SCE's CASMO-3/SIMULATE-3 topical report (Reference 1) documents benchmarking (comparison of measured startup test results with SIMULATE-3 predictions) for six cycles from SONGS Unit 1, five cycles from SONGS Unit 2, four cycles from SONGS Unit 3, and the initial cycle of ANO-2.

From Table 1.2 of SCE's NRC approved topical report, the CASMO-3/SIMULATE-3 bias and 95/95 uncertainty for CEA bank worth are 1.2% + 8.2%.

Benchmarking has continued to be performed (Reference 3) by comparing measured startup data for CEA bank worth to predictions. Recent results for SONGS 2&3 cycles 10 - 14 have confirmed that the bias and uncertainty are closer to 0% + 8.2%.

Thus CASMO-3 is benchmarked against measured San Onofre data by way of providing the cross-section data to the computer program (SIMULATE-3) which models the reactor cores and is used to predict the startup test measurements and perform core-follow.

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

Discuss the boron worths in terms of ppm/Ak used in determining the required boron concentration of 370 pmm, and justify the adequacy of the values of the boron worth used. Identify the limiting storage patterns with the associated enrichment that would result in a maximum required boron concentration of 370 ppm.

SCE Response:

As briefly described in Section 5.1 of Attachment L of the submittal, the non-accident boron concentration (370 ppm) was determined with the entire spent fuel pool (8 storage modules) modeled in KENO-V.a, precluding the need or use of a boron worth value. The process is further described below.

A series of full-pool KENO-V.a cases with various soluble boron concentrations were executed to determine the Keff for each case. In each case, each cell is loaded with fresh fuel at the maximum allowed enrichment for the location (e.g., 2.47 w/o for the Page 11 of 25

interior cells in Region I). The cases with Keff close to 0.95 were selected. As shown below, after all biases and uncertainties are added, the final results are:

300 PPM Final 95/95 Keff = 0.96211 350 0.95293 400 0.94554 The required boron concentration is calculated by interpolating for Keff = 0.95, precluding the need to calculate and use a boron worth value.

NUREG/CR-6683 and NRC Regulatory Issue Summary 2001-12 discuss the issue that the presence of soluble boron may result in a significantly higher equivalent fresh fuel enrichment from the burnup equivalencing. To ensure conservative results, equivalent enrichments were determined at each soluble boron concentration used for each full-pool case.

NRC Question 14: Reactivity Equivalence Uncertainty Section 5.2 of Attachment L to Reference 1 indicated that the reactivity equivalence uncertainty was 0.0 Ak at 0 GWD/IT and 0.01 Ak at 30 GWD/T, linear with burnup. It also indicated that this reactivity equivalence uncertainty was approved by the NRC.

Furthermore, it indicated that the soluble boron needed to compensate for the reactivity equivalence uncertainty was 178 ppm.

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

SCE Response:

The reactivity equivalence uncertainty accounts for the uncertainty in the prediction of the fuel assembly isotopic content (U, Pu, etc.) and distribution within the assembly.

The reactivity equivalencing uncertainty of 0.01 Ak at 30 GWD/T was described in Westinghouse topical report WCAP-14416-NP-A, "Westinghouse Spent Fuel Rack Criticality Analysis Methodology." This report was approved in Reference 7. The same uncertainty was used in Reference 8 and was approved by the NRC in Reference 9.

Prior to this SCE submittal, the NRC approved the SCE use of a 5% uncertainty applied to the total reactivity decrement calculated by CASMO-3. The NRC approval was provided in Reference 6.

The 5% uncertainty on the reactivity decrement is comparable to the 0.01 Ak value at 30 GWDFT. For example, for a 5.0 w/o fuel assembly, the reactivity decrement between burnups of 0 GWD/T and 30 GWD/T is calculated as follows.

Page 12 of 25

Calculation of Reactivity Decrement for a 5.0 w/o SONGS 2/3 assembly Burnup (GWD/T) CASMO-3 Kinf Reactivity Decrement (Ak/k) 0 1.40058 N/A 30 1.14909 0.15626 55 0.97176 0.31507 60 N/A 0.34682 (extrapolated)

A 5% uncertainty on the reactivity decrement is 0.0078 (0.15626 x 0.05) at 30 GWD, which is close to 0.01 Ak/k. Applying the reactivity (Ak/k) in terms of Keff change is acceptable for the criticality analysis since the base Keff of interest is around 1.00. As will be shown below, a value of 0.02 Ak, corresponding to a maximum burnup of 60 GWD/T, was applied to this analysis. Applying a 0.02 Ak/k at 60 GWD/T bounds the 5% reactivity decrement of 0.0173 (0.34682 x 0.05).

This 0.01 Ak at 30 GWD/T (and linear with burnup) was equivalenced to 178 ppm as follows.

From the full pool KENO-Va model at various boron concentrations:

300 PPM Keff = 0.96211 (No miss-load) 400 0.94554 (No miss-load) 800 0.95586 (miss-loaded fuel assembly study) 900 0.94465 (miss-loaded fuel assembly study)

Thus the boron worth is: (900-800)/(0.94465-0.95586) = -8,921 PPM/Ak (400-300)/(0.94554-0.96211) = -6,035 PPM/Ak For non-accident conditions, the required boron is in the 300-400 PPM range and the boron worth is -6,035 PPM/Ak.

However, we conservatively choose -8,921 PPM/Ak for the boron worth.

The peak pin burnup in any SONGS assembly is limited to 60 GWD/T.

The average assembly burnup at discharge is less than 60 GWD/T.

We conservatively assume the whole assembly (every pin) is 60 GWD/T.

0.01 Ak at 30 GWD/T is 0.02 Ak at 60 GWD/T.

(0.02 Ak) * (8,921 PPM/Ak) = 178 PPM Applying an uncertainty based on the maximum burnup of 60 GWD/T is conservative since the spent fuel pool Keff is mainly driven by fuel with low burnups.

Section 5.0 identifies a discretionary boron concentration margin of 154 ppm, which is equivalent to an additional conservatism of 0.017 Ak.

Page 13 of 25

In addition to the conservatism above, conservative bias/uncertainty terms were developed for the Keff formula (Section 3.2.1(i) of Attachment L of the submittal).

Conservatism exists in the treatment of the manufacturing tolerances, eccentric placement, and pool temperature bias. The tolerance evaluation assumes that every cell in the storage rack has the dimension (or enrichment or eccentric placement or temparature) which produces the highest Keff. In the actual racks, the variations are expected to average out to no impact. Thus one measure of the conservatism in the KENO final Keff is to set the manufacturing/enrichment tolerances, eccentric loading, and pool temperarture delta-k's to 0.0:

Region 1 conservatism = 0.99687 - 0.98333 = 0.01354 Region 2 conservatism = 0.99803 - 0.98092 = 0.01711 NRC Question 15: Discharge Burnup Uncertainty Section 5.3 of Attachment L to Reference 1 indicated that the soluble boron needed to compensate for the fuel assembly discharge burnup uncertainty was 218 ppm. This boron compensation was based on a discharge burnup uncertainty of 7 percent for SONGS Units 2 and 3 fuel assemblies.

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

SCE Response:

The discharge burnup uncertainty of 7% is a conservative estimate based on the 95/95 uncertainty of 4.76% for the measurement of the fuel assembly power using the CECOR computer code (CENPD-153-P, Rev. 1-P-A, "INCA/CECOR Power Peaking uncertainty," Reference 2) with a conservative multiplier (1.02) to account for the plant power uncertainty. This approach is very conservative since the burnup is a time-integral of the assembly power. As a measure of conservatism, EPRI Report TR-112054, ("Determination of the Accuracy of Utility Spent Fuel Burnup Records," July 1999), concludes that measured burnup uncertainties on the order of 2% are justifiable.

The 7% uncertainty was converted to 218 ppm as follows.

(1) Convert the 7% burnup uncertainty to burnup in GWD/T for the range of burnup of interest.

(7.0%)

• 60 GWD/T = 4.2 GWD/T (7.0%)

• 40 GWD/T = 2.8 GWD/T (7.0%)

• 30 GWD/T = 2.1 GWD/T (2) From CASMO-3 results, determine a conversion factor from GWD/IT to reactivity.

The conversion factor was found to be: 0.00583 Ak/GWD/T, 0.00633 Ak/GWD/T, and Page 14 of 25

0.00686 Ak/GWD/T for burnup values of 60 GWD/T, 40 GWD/T, and 30 GWD/T, respectively. Convert the GWD/T values from step (1) above to reactivity as follows.

(4.2 GWD/T) * (0.00583 Ak/GWD/T) = 0.02447 Ak (2.8 GWD/T) * (0.00633 Ak/GWD/T) = 0.01772 Ak (2.1 GWD/T) * (0.00686 Ak/GWD/T) = 0.01441 Ak (3) We conservatively choose the highest Ak from (2) above and the conservative boron worth of -8,921 PPM/Ak determined from a full-rack study at a conservatively high soluble boron concentration range of 800 - 900 ppm. A high soluble boron concentration increases the conversion factor from PPM to reactivity, and the boron requirement. The boron requirement is calculated below.

(0.02447 Ak) * (8,921 PPM/Ak) = 218 PPM NRC Question 16: CEA Life Time Analysis (Follow up to Question 6)

Control Element Assemblies (CEAs) were credited in the criticality analysis. How long do you have to take this credit? How do you know that the neutron absorber (AglnCd) inventory in the CEAs is sufficient as this credit is needed?

The questions are also applied to the guide tube (GT) inserts with the neutron absorber of B-1 0. In addition, please confirm that the credit of GT inserts was previously approved by Nuclear Regulatory Commission (NRC) staff for the criticality analysis SCE Response:

The CEAs and GT-inserts will be credited for reactivity holddown as necessary for the remainder of the plant life. Applying the reactivity credit on a long term basis is acceptable because of the very low neutron flux and gamma flux in the spent fuel pool as compared to the reactor core. The flux level in SFP is well below 106 n/cm2 -sec, which is orders of magnitude below the flux level of 1014 n/cm2 -sec in the reactor. It is expected21 that 2the fluence will not exceed the CEA lifetime limits of 18.Oxl 021 n/cm2 and 2.Oxl1 n/cm for AglnCd and B4C, respectively.

The mechanical design configuration of the GT-Inserts is similar to the shape, size, and weight of a control element assembly (CEA) finger. Each of the GT-Inserts is approximately 0.78 inch outside diameter (OD) solid stainless steel, with a boron content of approximately 2 w/o. A small counterbore is machined at the top for handling and a rounded bottom is machined. The OD of these GT-Inserts is less than that of a CEA finger. The material (borated stainless steel) is approved by the American Society for Testing and Materials (ASTM) and has been licensed by the NRC for use in spent fuel storage technologies and spent fuel pools. Credit of GT-Inserts was previously approved for Millstone Unit 2 in NRC letter dated March 1, 1994 (TAC No. M86361, Adams Accession No. ML012850287).

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Borated stainless steel is a two phase alloy, composed of a complex boride phase in an austenitic chromium-nickel-iron matrix. In addition to the U.S., this material is also manufactured in Austria, Great Britain, and Japan with boron content ranging from 0.2 to 2.25 w/o. It is extensively used by foreign nuclear industry for making spent fuel storage racks and spent fuel transportation casks. In the U.S., it was used in the Indian Point Nuclear Station as a poison plate material in their spent fuel pool. In general, the physical properties of the material resemble those of 304 austenitic stainless steel.

However, the yield strength, ultimate tensile strength and hardness increase with increasing levels of boron and ductility and impact strength are decreased. These properties also vary with the exposure to neutron fluency, but no significant changes occur for neutron fluences below 1017 n/cm 2 . This value is much higher than the typical spent fuel pool lifetime neutron fluence of 1012 n/cm 2 . Since poison rodlets do not carry any loads when inserted in the guide tubes, mechanical properties of the material are not of primary importance.

Although intergranular corrosion resistance of borated stainless steel exposed to acidic conditions decreases with increased boron content, long term tests with borated stainless steel have indicated that in the spent fuel pool environment no measurable corrosion effects take place. It is not expected, therefore, that any meaningful corrosion degradation of poison rodlets will occur during their service life.

The possibility of accidentally withdrawing a GT-Insert is minimized because special tooling is required to remove it, and it will be completely contained within the guide tubes of the designated assemblies. Potential misloading of the GT-Inserts is minimized due to the design of the installation equipment, procedural controls, and the double verification requirement that will be in place to ensure the GT-Inserts are installed properly.

The possibility of accidentally withdrawing a CEA is minimized because specialized tooling is required for withdrawing a CEA from a fuel assembly. It is physically possible for the spent fuel handling tool to bind on a CEA after ungrappling from a fuel assembly and raising the tool. However, existing SONGS procedures require that the operator validate "tool weight only" on the spent fuel handling machine's load cell read out after ungrappling from a fuel assembly and raising the hoist slightly, and to report this information to the engineer directing fuel movement.

NRC Question 17: Criticality Analysis; Design Basis Event Section 4.5 of the criticality analysis report specifies the limitations for locations of fuel assemblies in the spent fuel pools (SFPs) to meet the reactivity requirements used in the criticality analysis. Discuss the control programs and plant procedures to assure that the worst fuel misloading case is limited to the single assembly misloading event, which is the worst design-basis event considered in the criticality analysis.

SCE Response:

Preplanned moves of fuel and fuel inserts are reviewed and approved by a minimum of two Page 16 of 25

qualified engineers. Existing SONGS procedures require a peer-check of direction given by the engineer in charge of fuel movement, to ensure the engineer does not read the wrong step of the preplanned move sheet, nor misread the correct step (e.g., mistakes an "F" for an "E"). The spent fuel handling machine operator is not permitted to lower the spent fuel handling tool over a location without first receiving confirmation from a peer-checker that the machine is over the correct location. Engineers in charge of fuel movement, and personnel operating the spent fuel handling machine, receive formal training, documented in a required personnel qualification standard (PQS). The actual (physical) movement of fuel and CEAs will not be anymore difficult with implementation of the PCN. SONGS has already organized the Unit 2 and 3 spent fuel pools to meet the requirements of PCN 556. This was accomplished in 2002; since this time, SONGS has reorganized the spent fuel pools each cycle to maintain compliance with PCN 556.

NRC Question 18: Use of CEAs In accordance with the Safety Evaluation (SE) for St Lucie-1 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML042670562), the bases for the NRC staff to accept the use of CEAs are:

(a) the licensee has demonstrated that it had appropriate controls, procedures, and analyses to both understand and preclude the phenomena including 1) cladding wear, 2) unrecoverable cladding strain,

3) irradiation assisted stress-corrosion cracking, and 4) absorber depletion, from affecting the SFP criticality analyses, and (b) the licensee has demonstrated that design and use of the removable CEAs proposed in St Lucie 1 satisfied the guidance in Section 5.1 (a and c only) of Regulatory Guide 1.13, "Spent Fuel Storage Facility Design Basis."

SONGS 2 and 3 is requested to provide information to demonstrate that it satisfies the SE conditions (a) and (b) listed above for St Lucie -1, and justify for any deviations.

SCE Response:

As described in response to questions 6, 16, and 17, appropriate controls, procedures, and analyses have been implemented to both understand and preclude the phenomena including 1) cladding wear, 2) unrecoverable cladding strain, 3) irradiation assisted stress corrosion cracking, and absorber depletion, from affecting the spent fuel pool criticality analyses.

In addition, controls and procedures will be implemented to preclude the inadvertent removal of CEA or GT-Inserts. Assemblies loaded into locations for which the loading pattern specifies CEA or GT-inserts are moved into and out of their storage locations with the CEA or GT-Inserts already inserted. The CEA handling tool's physical design limits the potential to inadvertently mistake it for the spent fuel handling tool. The spent fuel handling tool is neither designed nor used to handle CEA or GT-Inserts.

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Procedures are in place to control the use of each piece of equipment, and do not provide for such use of the spent fuel handling tool. Finally, load cell circuitry enables the operator to monitor the weight of the fuel assemblies during hoisting and placement operation, thereby providing positive means to prevent inadvertent snagging and removal of a CEA during the grappling and ungrappling process.

NRC Question 19: Use of GT-Inserts In accordance with the SE for Millstone 2 (ADAMS Accession No. ML012850287), the bases for the NRC staff to accept the use of rodlets are:

(a) the rodlets were made from borated stainless steel, Type 304 B7, grade A with 2 weight percent of boron, manufactured in accordance with the requirements of standard ASTM A 887-89 and ASTM A 484-91; and (b) the licensee committed to implement a surveillance program where, at 5-year intervals, 1 percent of the rodlets will be visually inspected for any material degradation to assure that at all times there is enough poison material for reactivity control.

Please describe the material used in the SONGS 2 and 3 GT inserts and surveillance programs, and demonstrate that it satisfies the SE conditions (a) and (b) listed above for Millstone-2. Justification should be provided if the material of GT inserts and surveillance programs used at SONGS 2 and 3 are different.

SCE Response:

a) The rodlets (GT-Inserts) are made from borated stainless steel, type 304 B7, grade A in accordance with ASTM specification A 887 and A 484. The SONGS procurement process includes vendor source inspection and qualification and material certification and tests results in conformance with the ASTM specifications.

b) A surveillance program will be in place prior to installation of the rodlets. This program will be drafted from the Millstone surveillance program which requires visual inspection of 1% of the rodlets at 5 year intervals and includes rodlet visual surface examination for cracks, blisters, missing pieces of material, corrosion and pitting.

NRC Question 20: Use of Erbium Rods Provide the ADAMS Accession No. for the SE approving the use of Erbium rods for pressurized-water reactor (PWR) licensing applications, and address how the applicable range and conditions specified in the SE are met for the application of Erbium rods in SONGS 2 and 3.

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SCE Response:

Erbium is the burnable absorber used in San Onofre reactor core designs. Erbium oxide (erbia) is mixed with uranium oxide in the fuel pellet. Reference 10 provides the generic approval for core designs containing erbium burnable absorber for Combustion Engineering reactors. Conditions and requirements for the application of erbium are met through the SCE reload process described in Reference 11. There are no specific requirements for spent fuel storage.

In Attachment L to the submittal, the presence of erbia was credited for placement of fresh (new) fuel in Region I1. Crediting erbia in fresh fuel enables the storage of new fuel prior to loading the new assemblies in the core. No reactivity credit is considered for the remaining erbia in the assembly returned from the reactor core.

The characteristic of erbia is similar to gadolinia, which is used more extensively in the nuclear industry. Erbia is applied as an integral part of the fuel pellet, similar to the Westinghouse ZrB2 Integral fuel burnable absorber. Reference 7 provides the approval for credit of ZrB2 Integral Fuel Burnable Absorber in the spent fuel application.

Consistent with the treatment of B10 loading in Reference 7, the initial erbium loading is reduced by 5% for conservatism.

NRC Question 21: Interface Requirements Please provide the explanation on the checkerboard pattern interface requirements described in Section 4.5.3 of Attachment L to your application dated April 28, 2006.

SCE Response:

The boundary between checkerboard zones and the boundary between a checkerboard zone and all cell storage must be controlled to prevent an increase in reactivity. This is accomplished by examining each 2x2 assembly matrix interface and ensuring that each matrix confirms to restrictions for both regions. For example, consider a fuel assembly located in E in the following matrix of storage cells, Storage Pattern 1 A B C Interface Row D E F Storage Pattern 2 G H I Four 2x2 matrices of storage cells which include cell E are created in the above figure.

They include (A,B,D,E), (B,C,F,E), (E,F,I,H), and (D,E,H,G). (A,B,D,E) and (B,C,F,E) must satisfy the requirements of storage pattern 1. (E,F,I,H) and (D,E,H,G) must satisfy the requirements of storage pattern 2.

The figure below shows the interface requirement for placing a 2x2 matrix of fresh 4.8 w/o fuel with empty cells. For Region II, the maximum enrichment for unrestricted storage is 1.23 w/o. Placing the 2x2 matrix in Region II requires that the interface meet Page 19 of 25

both the unrestricted storage requirement and the checkerboard requirement.

A B C E F G 1 1.23 1.23 1.23 1.23 1.23 1.23 2 1.23 1.23 1.23 1.23 1.23 1.23 3 1.23 1.23 1.23 1.23 1.23 1.23 4 1.23 Blocked 1.23 1.23 1.23 1.23 5 Blocked 4.80 Blocked 1.23 1.23 1.23 6 4.80 [Blocked 1.23 1.23 1.23 1.23 The checkerboard pattern is in cells A5 through B6. The interface between the checkerboard pattern and the unrestricted storage pattern is cells A4, B4, C4, C5, and C6. These interface cells must meet the requirements of both unrestricted storage and checkerboard storage. Therefore, B4 and C5 must be empty, as if A4, C4, and C6 were 4.8 w/o fuel. The interface assemblies must also satisfy requirements for assemblies surrounding the 2x2 matrix.

NRC Question 22: Credit for Pu-241 Decay Based on the information on page 3 (item 7) and Figures 4-1 through 4-21 of Appendix L to the application dated April 28, 2006, it is clear that the effect of cooling time (Pu-241 decay) was included in the criticality analysis to determine the requirements for various assembly storage patterns proposed for storage in SFP.

Discuss the cooling effect, and describe how the effect was accounted for determining the burnup-enrichment requirements shown in Figures 4-1 through 4-21.

SCE Response:

The cooling effect refers to the decay of Pu-241 and other longer-life fissile and fissionable nuclides. The reactivity of an irradiated fuel assembly will decrease following its discharge from the reactor due to the decay of fissile and fissionable nuclides. The reactivity effect is calculated using the CASMO-3 computer code, consistent with the Reactivity Equivalencing description in Section 3.2.4 of Attachment L of the SCE submittal.

NRC Question 23: Boron Concentration Requirements Proposed TS 3.7.17 requires that whenever any fuel assembly fuel is stored in the fuel storage pool, the fuel storage pool boron concentration shall be greater than or equal to 2,000 parts per million (ppm).

Item c of TS 4.3.1.1 requires that K-eff shall be less than or equal to 0.95 if full flooded with water borated to 1,700 ppm, which includes an allowance for Page 20 of 25

uncertainty as described in Section 9.1 of the Updated Final Safety Analyses Report (UFSAR).

Clarify: (1) the contradictory requirements for boron concentration (2,000 ppm versus 1700 ppm) in the fuel storage pool; and (2) the specific pages in UFSAR Section 9.1 that contain the information related to an allowance indicated in item c of TS 4.3.1.1.21.

SCE Response:

The 2000 ppm boron concentration requirement in the proposed TS 3.7.17 is based on the boron dilution analysis (Attachment K of the SCE submittal). With an initial boron concentration of 2000 ppm, there is no credible boron dilution event that would reduce the boron concentration to a level required to maintain Keff below 0.95.

The proposed TS 4.3.1.1 (Design Features) describes the minimum boron concentration required to maintain Keff to below 0.95. As described in Section 5 of Attachment L of the SCE submittal, the total soluble boron required to maintain Keff less than 0.95, including all biases and uncertainties, under accident conditions, is 1,700 ppm.

The SCE proposed TS 3.7.17 and TS 4.3.1.1 are consistent with the Tech. Spec.

approved for St. Lucie-1 (ML042670562, ML042710211).

UFSAR Section 9.1.2 describes the spent fuel pool design basis, facilities description, and safety evaluation consistent with the current licensing basis approved in Reference

6. This section will be updated after this PCN is approved. Uncertainties described in Section 9.1.2.3 will be revised to be consistent with Appendix D of the criticality report (Attachment L of the SCE submittal).

NRC Question 24: Adequacy of the Axial Model This question is regarding the draft RAI 11 response which provided a discussion of calculations of the axial burnup bias using SIMULATE-3.

The results of analyses in NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses," indicated that the number of the nodes representing axial length of the fuel assemblies modeled in the computer codes would significantly affect the results of the axial burnup credit.

Specifically, item 2 on page A-8 of NUREG/CR-6801 indicated that the 10 axial-zone model was inadequate for the axial burnup bias calculation. The results showed that a model with 18 or more axial zones could produce a reliable axial burnup bias.

In light of the NUREG/CR-6801 results, provide a discussion to show that the number of the axial-zone used in SIMULATE-3 is adequate and acceptable for the calculation of the axial burnup bias.

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SCE Response:

The SCE SIMULATE-3 model contains 20 axial zones in the active fuel region. As described in Reference 1, SIMULATE-3 predicted axial power distribution agrees well with measurements.

Page A-8 of NUREG/CR-6801 concludes that a model with 18 or more axial zones are adequate to produce a reliable axial burnup bias.

NRC Question 25: Bounding Conditions for Storage Patterns 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 to compensate for no Boraflex poison credit. Based on its results of the criticality analyses, the licensee proposed the allowable storage patterns and the associated enrichment-burnup limits at discrete numbers of cooling years in Tables 4-3 through 4-34, and Figures 4-1 through 4-32 of the criticality analysis report (CAR) for the NRC staff to review. The Tables and Figures include 23 individual storage patterns and various combined patterns with associated enrichment ranging from 0.94 to 5.0 w/o, the burnup of up to 55.68 GWD/MTU, and the cooling times of up to 20 years).

Discuss the conditions in terms of the storage pattern, enrichment, burnup, cooling time, and boron concentration used to determine the values of each (underlined for emphasis) of the following parameters, and demonstrate that these values are the bounding values and applicable to the proposed allowable storage patterns for the assemblies in the SFP.

(a) the parameters included in Appendix D to the CAR. These parameters include:

kENO-V.a method bias, KENO-V.a method bias 95/95 uncertainty, k-nominal and its associated sigma, normal pool temperature bias, CEA insertion bias, eccentric loadings, and manufacturing tolerances including tolerances of rack storage cell wall thickness, rack storage cell ID, rack storage cell pitch and U235 enrichment.

(b) the axial burnup bias discussed in the draft RAI 11 response.

(c) the reactivity equivalencing uncertainty and discharge burnup uncertainties discussed in the draft RAI 14, and RAI 15 response, respectively.

(d) the boron concentration of 370 ppm to maintain k-eff less than 0.95 discussed in Section 5.1 of the CAR.

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SCE Response:

SCE is evaluating the response to this question and SCE will provide the response to this question on or before July 31, 2007.

NRC Question 26: Available Margins in the Results of the Criticality Analyses As indicated in the CAR, the licensee determined the allowable storage patterns and the associated enrichment limits based on calculations of following sets of conditions:

I. For cases with unborated water in the SFP, the k-eff is maintained to be less than 1.0 at a 95/95 level, and II. For Cases with borated water in the SFP, the k-eff is maintained to be less than or equal to 0.95 at a 95/95 level.

Provide information to demonstrate that the available margins in the results of the criticality analyses for all the allowable storage patterns are sufficient to bound the uncertainties for the following parameters that are not included in the criticality analyses. If the available margins are not sufficient, the licensee should perform the criticality analyses including all uncertainties, and present the results of analyses and the allowable storage patterns and the associated enrichment-burnup limits for the NRC staff to review.

(a) the reactivity equivalencing uncertainty and discharge burnup uncertainty (discussed in RAI 1(c)) that are not included in the criticality analyses for cases with unborated water (discussed in RAI 2.1), and (b) the biases and uncertainties for the parameters included in Appendix D to the CAR (discussed in RAI 1(a)), and the axial burnup bias discussed in RAI 1(b) that are not included in the criticality analyses for cases with borated water (discussed in RAI 2.11).

SCE Response:

SCE is evaluating the response to this question and SCE will provide the response to this question on or before July 31, 2007.

NRC Question 27: SR 3.8.18.1 reference to LCS 4.0.100 SR 3.8.18.1 references LCS 4.0.100. The LCS should be added with the revision number and the date to be consistent with TS.4.3.1 and the note on page 4.0.100-1 discussed in the draft RAI 3 response.

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SCE Response:

Amendment Application Numbers 243 Supplement 1 and 227 Supplement 1, which consist of PCN 556 Revision 1, clarify the inconsistency by adding the revision number and date to SR 3.8.18.1..

References:

1) NRC letter from M. B. Fields to H. B. Ray, SCE,

Subject:

Acceptance of Topical Report SCE-9001, PWR Reactor Physics Methodology Using CASMO-3/SIMULATE-3" For use at San Onofre Nuclear Generating Station, Units 1, 2, and 3, dated August 10, 1992.

2) NRC letter from James R. Miller to A. E. Scherer (Combustion Engineering),

Subject:

Review of Combustion Engineering Topical Report CENPD-153-P/CENPD-153-P (Revision 1), "Evaluation of Uncertainty In the Nuclear Power Peaking Measured by the Self-Powered, Fixed In-Core Detector System",

July 02, 1980.

3) San Onofre Nuclear Generating Station, Units 2 and 3, Engineering Procedure SO23- XXXVI-2.9, "Evaluation of Reactor Physics Biases and Uncertainties".

4) San Onofre Nuclear Generating Station, Units 2 and 3 Technical Specification Surveillance Requirement 3.1.5.5, "Verify each full length CEA drop time and the arithmetic average of all full length drop times is within at least one of the limit sets."

5) NRC letter from L. Raghavan to H. B. Ray,

Subject:

San Onofre Nuclear Generating Station, Units 2 and 3 Issuance of Amendments Related to Low Power Physics Testing Methodology, dated October 10, 2000.

6) NRC letter (M. B. Fields) to SCE (H. B. Ray), "Issuance of Amendment for San Onofre Nuclear Generating Station, Unit No. 2 (TAC No. M94624) and Unit No. 3 (TAC No.M94625)," October 3, 1996.

7) Timothy E. Collins (NRC) to Tom Greene (WOG), "Acceptance for Referencing of licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Criticality Analysis Methodology (TAC NO. M93254)," October 25, 1996.

8) "Duke Energy Corporation Oconee Nuclear Station, Units 1, 2, and 3, Docket Numbers 50-269, 50-270, and 50-287, Proposed Technical Specification Amendment Generic Letter 96 Spent Fuel Storage Racks (TSR 2000-01),"

December 28, 2000.

9) Leonard N. Olshan (NRC) to William 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.

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10) Ashok C. Thadani (NRC) to S. A. Toelle (Combustion Engineering), "Acceptance for Referencing of Topical Report CENPD-382-P, Methodology for Core Designs Containing Erbium Burnable Absorbers, (TAC NOS. M79067 and M82959),"

June 29, 1993.

11) NRC Letter (Stephen Dembek) to H. B. Ray (SCE), "San Onofre Nuclear Generating Station, Units 2 and 3 -" Evaluation of Reload Analysis Methodology Technology Transfer*(TAC Nos. MA4289 and MA4290)," June 2, 1999.

List of Commitments Contained in SCE Responses:

1) CEA Visual Inspection for SFP use: The CEA will have to pass a visual examination for defects. This visual examination is "qualitative"; but if abnormalities are noted, further evaluation will be required which can include quantitative examination of the CEA such as the extent of remaining clad wall thickness via eddy-current testing. SONGS Units 2 and 3 established the requirement that any CEA with over 20% through-wall wear would require a detailed assessment prior to qualification for continued service. This wear threshold criteria for further evaluation for core service will also apply to spent fuel pool reactivity control service.

2) The affected UFSAR sections will be updated after this PCN is approved to reflect design changes, design requirements, and the results of the supporting calculations.

3) Prior to the use of GT-Inserts, procedures will be implemented to preclude the inadvertent removal of CEAs and GT-Inserts and to ensure they are placed into the correct locations. These procedures will require:

a. Special tooling for installation of CEAs and GT-Inserts,

b. GT-Inserts to be completely contained within the guide tubes of the designated assemblies.

c. Special tooling for removal of CEAs and GT-Inserts, and

d. double verification requirement.

4) Procedures will require that assemblies loaded into locations for which the loading pattern specifies CEA or GT-inserts be moved into and out of their storage locations with the CEA or GT-Inserts already inserted.

5). The GT-Insert surveillance program will require visual inspection of 1% of the rodlets at 5 year intervals and includes rodlet visual surface examination for cracks, blisters, missing pieces of material, corrosion and pitting.

6) SCE will provide a timeline to the NRC Project Manager for responding to questions 25 and 26 as soon as it is developed.

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