ML20116C674

Application for Amends 146 & 130 to Licenses NPF-10 & NPF-15 Consisting of Change Request 443,permitting Increase in Licensed Storage Capacity of Spent Fuel Pools,By Allowing Fuel to Be Consolidated After Required Time in SFP

ML20116C674
Person / Time
Site:	San Onofre
Issue date:	07/29/1996
From:	Nunn D SOUTHERN CALIFORNIA EDISON CO.
To:
Shared Package
ML20116C666	List: ML20116C663 ML20116C674 ML20116C680 ML20116C684
References
NUDOCS 9608010067
Download: ML20116C674 (39)

Text

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA ) Docket No. 50-361 EDIS0N COMPANY, H .A_L. for a C1 ass 103 )

License to Acquire, Possess, and Use )

a Utilization Facility as Part of ) Amendment Application Unit No. 2 of the San Onofre Nuclear ) No. 146 Generating Station- )

SOUTHERN CALIFORNIA EDISON COMPANY, H AL. pursuant to 10 CFR 50.90, hereby submit Amendment Application No.146.

This amendment application consists of Proposed Change Number (PCN) NPF-10-443 to Facility Operating License No. NPF-10. PCN NPF-10-443 requests changes to the Unit 2 Amendment No. 127 (as revised by Amendment No. 128) approved Technical Specification (TS) Sections 3.7 (Plant Systems) and 4.3 (Fuel ,

i Storage). These changes will permit an increase in the licensed storage l

(

capacity of the spent fuel pool of Unit No. 2 by allowing spent fuel to be '

consolidated after a minimum residence time in the spent fuel pool.

9608010067 960729 PDR ADOCK 05000361 P PDR

1 i

2 Subscribed on this day of v [U , 1996 Respectfully submitted, I

l SOUTHERN CALIFORNIA EDIS0N (QMPANY By: b I .L 9

' Dwight}E" Nunn Vice Pr sident State of California b n ( fQre lkAb I O ,

persor. ally appeared d it h W . M tt ii n - , personally known to me to be the person whose Wame is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and that by his signature on the instrument the person, or the entity upon behalf of hich the person acted, executed the instrument.

WITNESSmyhndandofficialseal Signature J J .

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a MARANE,SANCHEZ C=. lau m f 2

j Notone pttsc - CoCfornia ORANGE COUNiy f

L- W %. Eg*ee OCT 14

________,imf, , ,

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UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION

Application of SOUTHERN CALIFORNIA ) Docket No. 50-362 l

EDIS0N COMPANY, H R . for a Class 103 )

License to Acquire, Possess, and Use )

a Utilization Facility as Part of ) Amendment Application Unit No. 3 of the San Onofre Nuclear ) No. 130 Generating Station )

SOUTHERN CALIFORNIA EDIS0N COMPANY, H R . pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 130. )

l This amendment application consists of Proposed Change Number (PCN) NPF-15-443 to Facility Operating License No. NPF-15. PCN NPF-15-443 requests changes to the Unit 3 Amendment No. 116 (as revised by Amendment No. 117) approved Technical Specification (TS) Sections 3.7 (Plant Systems) and 4.3 (Fuel Storage). These changes will permit an increase in the licensed storage I capacity of the spent fuel pool of Unit No. 3 by allowing spent fuel to be j consolidated after a minimum residence time in the spent fuel pool.

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2-Subscribed on this __ ' day of w b ,1996 v

Respectfully submitted, 3

SOUTHERN CALIF 0 IA EDIS0N COM 'ANY yk -

" Dwight . Nunn Vice Pr ident l

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State of California 'q '

A fore m 016 fL b,{ B , IVbh) 1 hhb&,

h R n R , personally known to pers6rfally ' appeared I

me to be the person whbse Yiame is subscr'ibed to the within instrument and 1 acknowledged to me that he executed the same in his authorized capacity, l and that, by his signature on the instrument the person, or the entity upon behalf o which the person acted, executed the instrument.

WITNESS mI ' d nd official se .

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SignaturfI b j MARIANE SANCHEZ COMM. # 10337M f

v / z Mary Pubile - Colifornia h

] ORANGE COUNTY s 1 Comrn. Expirm OCT R Ign ,

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Enclosure 1 DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-10/15-443 Proposed Change Number 443 (PCN-443) is a request to revise Section 3.7,

" Plant Systems", and Section 4.3, " Fuel Storage," of the San Onofre Nuclear Generating Station (SONGS) Unit 2 Amendment Nos. 127/128 and Unit 3 Amendment Nos. 116/117 approved Technical Specifications (TS). These revisions will allow SONGS 2 and 3 spent fuel and SONGS 1 uranium dioxide spent fuel to be consolidated and stored in the existing spent fuel pools of SONGS 2 ad 3.

The revisions are independent of and in addition to those requested for TS Sections 3.7 and 4.3 through PCN-449. PCN-449, dated December 6, 1995, requested an increase in the maximum U-235 enrichment of spent fuel that may be stored in the SONGS 2 and 3 spent fuel pools from 4.1 weight percent (w/o) l to 4.8 w/o.

NRC action on PCN-449 is currently pending.

Unit 2 Amendment Nos. 127/128 and Unit 3 Amendment Nos. 116/117 Approved Technical Specifications Unit 2: See Attachment A Unit 3: See Attachment B Unit 2 Amendment Nos. 127/128 and Unit 3 Amendment Nos. 116/117 Approved Technical Specifications as Revised by PCN-443 Unit 2: See Attachment C Unit 3: See Attachment D Unit 2 Amendment Nos. 127/128 and Unit 3 Amendment Nos. 116/117 Approved Technical Specifications as Revised by PCN-443 and PCN-449 Unit 2: See Attachment E Unit 3: See Attachment F Description of Changes PCN-443 requests changes to TS Sections 3.7 and 4.3 to permit an increase in the licensed storage capacity of the spent fuel racks in the spent fuel pools (SFPs) of SONGS 2 and 3. The increased capacity will be provided by consolidating spent fuel after a minimum required residence time in the SFPs.

The increased capacity will enable SONGS 2 and 3 to operate until the end of their licensed life (2013) and provide room for uranium dioxide spent fuel remaining in the SONGS 1 SFP to be consolidated and stored in the SONGS 2 and 3 SFPs. SONGS 2 and 3 are presently licensed to store fuel assemblies having a maximum U-235 enrichment of 4.1 w/o. Upon NRC approval of PCN-449, the maximum licensed enrichment for fuel stored at SONGS 2 and 3 would increase to 4.8 w/o. The criticality analyses for PCN-443 have been performed conservatively assuming 5.1 w/o maximum fuel enrichment for SONGS 2 and 3.

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The details of our proposal, the reasons and justification for increased SFP storage capacity, and the supporting analyses are presented in Enclosure 2, entitled " Fuel Consolidation and Storage Report." The most significant information is presented in the section " Discussion of Changes" below.

The consolidation and storage of spent fuel will take place entirely within the fuel handling buildings (FHBs) of SONGS 2 and 3. Section 3.7, Plant Systems, and Section 4.3, Fuel Storage, of the Technical Specifications (TS) contain requirements and restrictions relating to the handling and storage of spent fuel in the FHBs. These requirements and restrictions must be revised to support fuel consolidation and storage inside the FHBs. The sections below include discussion of the required TS revisions and the associated safety analyses.

Discussion of Changes Need for Increased Storage Capacity:

PCN-443 requests NRC approval to increase the SFP storage capacity. SONGS 2 and 3 SFPs were reracked in 1990 and 1991, respectively, increasing the storage capacity of each SFP from 800 to 1542 fuel assemblies. This design change was authorized by SONGS 2 License Amendment No. 87 and SONGS 3 License Amendment No. 77. It resulted in 1542 high density fuel storage locations (or cells) for each SFP, with each location capable of storing one fuel assembly.

The same license amendments also allowed SONGS 1 uranium dioxide assemblies to be stored at SONGS 2 and 3. SONGS 1 was permanently shut down in November 1992, and spent fuel is no longer being generated there. At present, SONGS 1 uranium dioxide fuel is stored in all three SONGS units, while SONGS 1 mixed oxide fuel (four assemblies) that was used for demonstration purposes is stored at SONGS 1. If the present SONGS 2 and 3 production factors and power / refueling cycle durations (based on 4.1 w/o fuel enrichment) continue in the future, then SONGS 3 will lose full core offload reserve (FCOR) capability in 2003 and SONGS 2 will lose FCOR capability in 2005. The different dates f are mainly due to SONGS 3 having more SONGS 1 fuel in storage than is stored j at SONGS 2. Contingent on NRC approval of 4.8 w/o enrichment fuel starting j with Cycle 9 operation, as proposed in PCN-449, loss of FCOR for both units is ,

estimated to occur in 2006 (end of Cycle 13 operation). Considering that l

! SONGS 2 and 3 are licensed to operate until the year 2013, additional storage j l capacity must be provided to support operation for the full term of their ~

i L licenses. Fuel consolidation will increase the storage capacity of each SFP

! from 1542 fuel assemblies to approximately 2867 fuel assemblies with minimal

( impact on existing plant structures, systems, and equipment. For more information concerning the need for increased storage, refer to Enclosure 2.

Fuel Consolidation:

l PCN-443 requests NRC approval to perform onsite spent fuel consolidation.

Edison has not selected a consolidation vendor or a specific consolidation j process at this time. A generic description of the consolidation process is provided in Section 5 of Enclosure 2.

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Typically, consolidation will involve pulling fuel rods out of two spent fuel assemblies that have similar physical charactertstics (one SONGS 2 or 3 assembly with another SONGS 2 or 3 assembly, each containing 236 fuel rods).

The undamaged rods from both assemblies will be densely packed in a metal canister sized for storage in the space occupied by one assembly prior to consolidation. Damaged fuel rods will be separated and stored in either a rod storage basket or a consolidated fuel canister with other damaged rods.

Criteria for separating damaged rods from undamaged rods will be developed prior to implementing consolidation. After all the fuel rods are removed, the non-fuel bearing components (NFBC) of the assembly, such as grid spacers, end fittings, guide tubes, etc., will be compacted and stored in separate containers. The maximum achievable fuel rod consolidation ratio is 2:1. The maximum achievable NFBC hardware compaction ratio is estimated to be 10:1.

The physical and structural characteristics of the various canisters and containers used to store the products of fuel consolidation will ensure that these radioactive products can be handled and stored safely in the SFPs. The design criteria for the consolidated fuel canisters are presented in section 4.3 of Enclosure 2. The SONGS 1 consolidated fuel canisters will be smaller, I weigh less, and hold fewer fuel rods than the SONGS 2 and 3 consolidated fuel canisters.

At SONGS 2 and 3 fuel consolidation will take place in borated water in the cask pool area, which is adjacent to the SFP. The consolidation process and its major mechanical components are described in Section 5 of Enclosure 2.

The same equipment presently used to move fuel assemblies inside the FHB will be used to transfer candidate fuel assemblies for consolidation from the SFP to the cask pool area and the products of consolidation back to the SFP. This process simplifies consolidated fuel handling operations and the analytical work needed to support fuel consolidation, and minimizes the probability of accidents. Based on analysis, a minimum decay time of six (6) months must be met before a fuel assembly is ready for consolidation. This time is sufficient to allow thermal cooling of the irradiated fuel and decay of more than 99.9 percent (%) of the radioactivity present in the volatile radioisotopes (noble gases and iodine), so that safety parameters, such as the bulk SFP water temperature and the fuel cladding temperature, are not adversely affected. It also ensures that the consequences of postulated radiological accidents associated with fuel consolidation and storage would remain within regulatory limits.

In the safety analyses for fuel consolidation and storage, credit has been taken for the operability of the Control Room Emergency Air Cleanup System (CREACUS) to maintain the radiological dose to the control room operators below regulatory limits. No credit was needed nor taken for the operability of the FHB post-accident cleanup system. In the criticality analyses relating to passive storage of consolidated fuel in the SFP, the boron concentration in the pool water has been conservatively assumed to be zero. For postulated accidents where an increase in the effective neutron multiplication factor (k-eff) would result from the accident, the boron concentration has been assumed in criticality calculations to be 1800 parts per million or ppm (the Technical Specification limit of 1850 ppm includes 50 ppm measurement uncertainty). This assumption is in agreement with the " double contingency

principle" of ANSI N16.1-1975 (ANSI /ANS-8.1-1983), which in effect states that it is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.

In the analyses of accidents involving consolidated fuel, credit has been taken for maintaining a water depth of at least 23 feet over the. top cf the irradiated fuel assemblies or consolidated fuel canisters seated in the storage racks. While fuel consolidation is in progress, the gate separating

the SFP and the cask pool will be kept open. This will ensure that the minimum boron concentration limit and the minimum water level requirement are satisfied for the cask pool and the SFP simultaneously.

1 j Proposed Increase in SFP Capacity:

PCN-443 requests an increase in SFP storage capacity from 1542 to 2867 fuel assemblies. The SONGS 2 and 3 SFPs consist of two regions. Region I (312 cell locations) is used for full core offload (217 fuel assemblies). The remaining 95 cell locations in Region I may be used for either fresh fuel or cpent fuel that has not achieved the minimum required burnup for storage in Region II. Region II (1230 cell locations) provides normal storage for spent fuel assemblies that have achieved the minimum required burnup. Fuel that does not meet the burnup criterion may be placed in Region II in accordance with existing Technical Specification (TS) 4.3.1.1.g (proposed TS 4.3.1.1.h) provisions.

Regions I and II together have 1542 cell locations. Each cell location can store one standard fuel assembly. If 217 cell locations are reserved for the ,

full core offload, the renaining 1325 cell locations can each accommodate one )

consolidated fuel canister. A canister designed for SONGS 2 or SONGS 3 fuel can hold up to 472 fuel rods, and a canister designed for SONGS 1 fuel can hold up to 360 fuel rods. The number of rods in each case is equivalent to two full spent fuel assemblies. Assuming for purposes of licensed capacity determination, a full pool of consolidated fuel except for 217 assemblies from one full core offload, each SFP can store up to 217 + (2 x 1325) = 2867 spent fuel assemblies. This estimate conservatively assumes that fresh fuel assemblies, compacted NFBC, damaged fuel rod baskets, and other permitted storage items will not be stored in the SFP with the consolidated fuel.

(However, this assumption does not preclude permitted storage items to be stored in the SFP along with consolidated fuel.)

Decay Heat Loads and SFP Bulk Temperature:

PCN-443 requests an increase in the acceptance criterion for the " maximum normal heat load" case SFP bulk temperature from 140 F to 145 F. This request is the result of analyses performed to determine the " maximum normal" and

maximum abnormal" case heat loads on the SFP cooling system. These analyses were performed consistent with the conditions delineated in the NRC's "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April,1978, as amended by NRC letter dated January 18, 1979, and the guidance in Standard Review Plan (SRP) 9.1.3, and are discussed below. Although not considered in the OT Position or the SRP, the maximum

anticipated heat load during refueling operations with consolidated fuel stored in the SFPs was also analyzed and is discussed below.

l The first two heat loads were modeled as follows:

- Maximum normal heat load: one refueling load discharge load after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decay and one refueling load discharge load after one year of 1

decay in a full SFP Maximum abnormal heat load: one full core offload after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decay and one refueling load discharge load after 36 days of decay in a full SFP These two selected models are consistent with SRP 9.1.3 guidelines. The anticipated consolidated fuel heat loads are 26.9 MBTU/hr (refueling discharge J load of 104 fuel assemblies plus 2546 previously used fuel assemblies consolidated and stored in the SFP) and 52.7 MBTU/hr (full core offload plus 2650 previously used fuel assemblies consolidated and stored in the SFP). In the first case, the number of fuel assemblies selected was the maximum number needed to fill the SFP leaving enough room for a full core offload. In the second case, the number of fuel assemblies selected was the maximum number needed to fill the SFP. In each case, the anticipated consolidated fuel heat i load was based on 635 effective full power day (EFPD) operating cycles and a refueling discharge load of 104 fuel assemblies. The anticipated consolidated fuel heat loads were found to be less than the design basis consolidated fuel heat loads assumed in the SFP cooling system evaluation performed during the SFP reracking project: 27.3 MBTV/hr (maximum normal heat load) and 53.3 4

MBTU/hr (maximum abnormal heat load). These two design basis consolidated fuel heat loads are based on 570 EFPD and a refueling discharge. load of 108 fuel assemblies. Since fuel consolidation has not previously been implemented at SONGS, the SONGS 2 and 3 Updated Final Safety Analysis Report (UFSAR) does

not discuss fuel consolidation. As a result, the two anticipated consolidated

fuel heat loads and the two design basis consolidated fuel heat loads are not mentioned in the UFSAR. The anticipated consolidated fuel heat loads are

smaller than the corresponding design basis consolidated fuel heat loads since reducing the discharge batch size from 108 fuel assemblies to 104 fuel

! assemblies more than offsets increasing the irradiation time from 570 EFPD to 635 EFPD. Therefore, no change is needed to the existing SFP cooling system hardware to accommodate the anticipated consolidated fuel heat loads.

4 The existing SFP cooling system is adequate to maintain the SFP bulk temperature below 145*F for the maximum normal (design basis) heat load with the consolidated fuel, assuming a single active failure (one pump inoperable) as specified in SRP 9.1.3. This temperature limit is higher than the existing bulk temperature limit, which is 140 F, consistent with SRP 9.1.3.III.1.d.

However, the new limit of 145 F has been conservatively calculated, and the existing cooling system components are qualified for the higher temperature.

For the maximum SFP temperature of 145*F, during the maximum normal heat load condition (normal fuel handling building ventilation and exhaust subsystem operating), the SFP area air temperature will be below 104 F, which is within the fuel handling building maximum acceptable normal air temperature of 109*F.

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The 109 F air temperature is based on the design maximum SFP water temperature of 160*F and the design maximum site ambient air temperature of 85*F during normal operation of the building ventilation and cooling system. The 109*F air temperature is acceptable since the design operating air temperature for the fuel handling equipment used to move fuel between the reactor core and the SFP is 120 F. The SFP purification system uses ion exchange resins for water purification. The purification pump is stopped at 140*F by a temperature switch to protect the resins. The SFP liner plate is analyzed for temperatures up to 216*F. Therefore, the higher SFP temperature limit does not significantly decrease the plant safety margins and is acceptable.

Consistent with SRP 9.1.3, a single active failure need not be considered for the maximum abnormal heat load condition. With both SFP cooling system heat exchangers and pumps in operation, the SFP temperature will be 160 F or less assuming a maximum abnormal decay heat load of 53.3 MBTU/hr. During the reracking licensing process the maximum SFP temperature associated with the maximum abnormal heat load was 156*F. The higher temperature of 160 F is still well within the SRP value (less than 212 F) and is reflected in the UFSAR. The calculated heat load to the environment will only increase by about 0.03%. For the maximum SFP temperature of 160*F, during the maximum abnormal heat load condition (fuel handling accident conditions and the post accident cleanup subsystem operating), the SFP area air temperature will be approximately 140*F, which is well within the 150*F design air temperature for the fuel handling building post accident cleanup subsystem. This analysis assumes that, upon fuel handling isolation signal (FHIS) actuation, the fuel handling equipment has ceased operation and refueling personnel have evacuated the SFP area. For this condition, with the spent fuel pool pump room emergency cooling subsystem operating, the spent fuel pool pump room temperature will not exceed 104*F. Fuel consolidation equipment will not be operated if the air temperature near the consolidation equipment exceeds 104 F (see Attachment G for the administrative controls established for consolidated fuel storage).

SONGS 2 and 3 conduct refueling by offloading either half the core (108 fuel assemblies) or the full core (217 fuel assemblies). The full core offload refueling provides the greater of the two heat loads. Therefore, in addition to the " maximum normal" and " maximum abnormal" heat loads specified in SRP Section 9.1.3, the maximum heat load during refueling operations was also evaluated. For this case, the heat load was evaluated assuming a two year refueling cycle, the spent fuel pool completely filled with spent fuel (except the full core offload at the end), and the full core offloaded at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decay. In addition, it was conservatively assumed that plant operating cycles have 570 EFPDs and refueling discharge loads consist of 108 fuel assemblies (except last full core offload, which is 217 assemblies). With these assumptions, the maximum consolidated fuel refueling heat load was 44.8 MBTU/hr. This is slightly above the maximum refueling heat load without fuel consolidation, which is 43.0 MBTV/hr. With either heat load, a single SFP cooling pump with two heat exchangers will maintain the SFP temperature below 160 F, assuming the component cooling water temperature is 88 F and the ocean water temperature is 76 F. The corresponding SFP area air temperature

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will be within the maximum normal fuel handling building air temperature of l 109*F with the nnrmal fuel handling building ventilation and exhaust subsystem operating. Fuel consolidation equipment will not be operated if the air l temperature near the consolidation equipment exceeds 104*F, as stated in the I administrative controls in Attachment G. Therefore, the impact of the maximum  :

SFP temperature of 160 F on the air temperature, the ventilation equipment, I and the fuel consolidation equipment is acceptable.

The shutdown cooling system (SDCS), if available, can be used as an alternate heat dissipation path for cooling the SFP. The SDCS has been evaluated for the maximum normal and maximum abnormal heat loads and it has been determined that the system and interconnecting ties are adequate to maintain the SFP temperature below 145 F for the maximum normal heat load and below 160 F for the maximum abnormal heat load. Since the maximum abnormal heat load bounds the maximum refueling heat load, there is no need to evaluate the SDCS for the maximum refueling heat load. For the maximum refueling heat load, the SDCS does not meet the single failure criterion for SFP cooling; however, the use of the SDCS for SFP cooling during Modes 5 and 6 of plant operation was evaluated and considered acceptable by the NRC in the SDC cross-connection safety evaluation report dated June 4,1993.

For further details concerning the consolidated fuel decay heat analysis, refer to Section 3 of Enclosure 2.

Identification of Changes A number of changes to the SONGS 2 and 3 Technical Specifications are required before the fuel consolidation and storage plan discussed above and described more fully in Enclosure 2 can be implemented. These changes are identified below. These changes will be supplemented with administrative controls, which are presented for information in Attachment G and which will be incorporated i into the Licensee Controlled Specifications (LCS) prior to any fuel (

consolidation activities.

Changes to Technical Specification 3.7.11: J Specification 3.7.11, Control Room Emergency Air Cleanup System (CREACUS),

provides operability requirements for the two CREACUS trains in plant MODES 1 through 6 and during movement of irradiated fuel assemblies. It specifies required actions and required completion times for these actions should one or both trains become inoperable. The primary objective is to provide clean, filtered air to the control room operators after a postulated uncontrolled release of radioactivity. Certain surveillance requirements are also stated.

The radiological consequences of potential accidents related to fuel consolidation and consolidated fuel handling and storage have been analyzed.

These accidents are described in Chapter 3 and evaluated in Chapter 6 of Enclosure 2. They are also discussed below under the sections titled

" Discussion of Accidents" and " Safety Analysis." The results indicate that consolidated fuel storage will not increase the radiological consequences of fuel handling accidents previously evaluated in the UFSAR. Specifically, l

_8-l doses to the control room personnel and offsite personnel will not be increased. Additionally, fuel handling accidents for SONGS 1 fuel which is

( stored in the SONGS 2 and 3 SFPs would be bounded by fuel handling accidents j involving SONGS 2 and 3 fuel.

The new analyses take credit for operation of the CREACUS after a postulated release of radioactivity from the consolidated fuel. Consequently, both CREACUS trains must be operable during fuel consolidation and movement of i consolidated fuel in the same way they are presently required to be operable

! while moving spent fuel assemblies. Accordingly, the following changes are proposed to Specification 3.7.11:

l a. The Applicability statement will be modified to replace the phrase "During movement of irradiated fuel assemblies." with "During movement of irradiated fuel."

b. The CONDITION column of ACTIONS C and E will be modified to replace "during movement of irradiated fuel assemblies" with "during movement of irradiated fuel ." The REQUIRED ACTION column of ACTIONS C and E will be modified to replace " Suspend movement of irradiated fuel assemblies" with " Suspend movement of irradiated fuel."

The term " movement," as applied to irradiated fuel in Specification 3.7.11, has been defined in the Bases (see B 3.7.11 in Attachment H). The term

" movement" in this application refers to the movement of irradiated fuel l pellet (s), fuel rod (s), a partial fuel assembly or assemblies, a complete fuel l assembly or assemblies, or consolidated fuel. The fuel movement may be for any purpose, including fuel consolidation.

Changes to Technical Specification 3.7.16:

Specification 3.7.16, Fuel Storage. Pool Water Level, requires that a water level of at least 23 feet be maintained "over the top of irradiated fuel assemblies seated in the storage racks." This requirement applies "During movement of irradiated fuel assemblies in the fuel storage pool." If the requirement is not met, Specification 3.7.16 requires that the movement of irradiated fuel assemblies be suspended immediately. The 23 foot minimum water level is subject to periodic surveillance.

It is proposed to make the following changes to Specification 3.7.16 to extend the existing requirements to consolidated fuel and fuel consolidation activities:

a. In LC0 3.7.16, change " fuel assemblies" to " fuel ."

l b. Under APPLICABILITY, change " movement of irradiated fuel assemblies" to " movement of irradiated fuel" and add the words "or cask pool" after the words " fuel storage pool."

c. Under REQUIRED ACTION, change " movement of irradiated fuel assemblies" to " movement of irradiated fuel." Add the words "and cask pool" after j the words " fuel storage pool."

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d. Under SURVEILLANCE, change " irradiated fuel assemblies" to " irradiated fuel."

J The term " movement," as used in b and c, has been defined in the applicable Bases (see B 3.7.16 in Attachment H) and includes fuel consolidation.

The Bases in B 3.7.16 identify two reasons for the 23 foot water depth:

a. It meets the minimum water depth stated in Regulatory Guide (RG) 1.25 l for applying the iodine decontamination factors assumed in the RG to i evaluate the consequences of a fuel handling accident.

b. It acts as a shield to minimize the general area dose when the storage l racks are filled to capacity. j i 1 j In addition, B 3.7.16 states that the water provides shielding during the  ;

movement of spent fuel. j Regarding item a, it is assumed in Regulatory Guide (RG) 1.25 that the fuel j handling accident occurs at the earliest time when fuel handling operations l

are allowed to begin. For SONGS 2 and 3 this is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown. The consolidated fuel accident analyses evaluated in Enclosure 2 4 assume a minimum spent fuel decay time of six months before consolidation.

i This minimum decay time will be implemented as an administrative control and i it will allow volatile fission products (iodine, krypton, and xenon isotopes) to decay to very low levels. Therefore, the 23 foot water depth is conservative for applying the iodine decontamination factors of Regulatory

Guide 1.25 to consolidated fuel, i l i Regarding item b, as pointed out in Section 6.2.3 of Enclosure 2, the j estimated exposure at the surface of the spent fuel pool from stored fuel

assemblies is on the order of 1 X 10-6 mrem /hr and no significant increase is i expected after fuel consolidation. Section 6.2.3 also notes that dose rates i in the spent fuel pool area are primarily driven by activated corrosion products such as Co-58 and Co-60 present in the water and deposited on the

!' pool walls, rather than spent fuel inventory, for which there is ample water shielding.

1 In addition, considering the low activity levels in the fuel after six months 4

of decay and that fuel consolidation and the movement of consolidated fuel canisters will take place under water, ample shielding will be available for the consolidated spent fuel.

Changes to Technical Specification 3.7.17:

Specification 3.7.17, Fuel Storage Pool Boron Concentration, requires a boron concentration of at least 1850 ppm in the SFP. This requirement applies when fuel assemblies are stored in the SFP and a verification to confirm that there is no misloaded fuel in the SFP has not been performed since the last movement 4

of fuel assemblies. The associated surveillance requirement is to verify the 4

boron concentration once every 7 days. If the concentration is not within its limit, Section 3.7.17 requires that the movement of fuel assemblies be

suspended and either actions to restore boron concentration be initiated or an administrative verification of fuel loading in the SFP be conducted. These corrective actions are required immediately.

It is explained in Section 3 of Enclosure 2 that consistent with the " double contingency principle" of ANSI /ANS-8.1-1983, the presence of soluble boron in the SFP water (and consequently in the cask pool water) has been assumed for those accidents that have the potential to add reactivity. In view of this assumption, the requirements of Section 3.7.17 must be extended to fuel consolidation operations in the cask pool and consolidated fuel storage in the SFP.

It is proposed to make the following changes to Specification 3.7.17:

a. In APPLICABILITY, change the words " fuel assemblies are stored" to

" fuel is stored" and " movement of fuel assemblies" to " movement of fuel." Add a new sentence "Also when fuel is present in the cask pool."

b. Under REQUIRED ACTIONS, add a Note 2 to clarify that the administrative verification of fuel loading stated in Action A.2.2 is not applicable when the sole reason for compliance with LC0 3.7.17 is the presence of fuel in the cask pool. Change A.1 from " Suspend movement of fuel assemblies in the fuel storage pool." to " Suspend movement of fuel in the fuel storage pool and cask pool." This requirement applies if the boron concentration is not within the limit I stated in the LCO. In addition, change " movement of fuel assemblies" l in A.2.2 to " movement of fuel."

l The UFSAR accident analyses assume a minimum boron concentration of 1800 ppm, whereas the Technical Specifications specify 1850 ppm (1800 ppm plus 50 ppm measurement uncertainty).

Changes to Technical Specification 3.7.18:

Specification 3.7.18, Spent Fuel Assembly Storage, restricts the initial ,

enrichment and burnup of SONGS 1, 2, and 3 fuel assemblies that may be stored l in Region II of the SFP. It requires that the combination of initial .

enrichment and achieved burnup for each assembly be within the acceptable  !

limits shown in Figure 3.7.18-1 for SONGS 1 fuel and Figure 3.7.18-2 for l SONGS 2 and 3 fuel. Fuel assemblies that do not meet these limits may still be stored in Region II if they comply with administrative controls that are implemented through Specification 4.3.1.1. If these requirements are not met, then action must immediately be initiated to move the noncomplying fuel assembly out of Region II or comply with the administrative controls for Region 11 storage. Specification 3.7.18 also contains a surveillance requirement to administratively verify that the fuel assembly complies with the initial fuel enrichment and burnup requirements prior to storage in Region II.

The purpose of storing in Region II only fuel which meets the burnup criteria of Figures 3.7.18-1 or 3.7.18-2 or complies with administrative controls implemented through Specification 4.3.1.1 is to protect the fuel against criticality. Section 9.1.2 of the UFSAR currently states:

"The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (k-eff) of the fuel  ;

assembly array will be less than or equal to 0.95 when the storage racks are )

fully loaded with spent fuel (the Region I racks are fully loaded with fresh 4.1 w/o fuel, and the Region II racks are fully loaded with spent fuel which meets the minimum burnup criterion) and flooded with unborated water as recommended in ANSI-57.2-1983 and in Standard Review Plan 3.8.4." l As demonstrated in Sections 3 and 6 of Enclosure 2, spent fuel consolidation and storage will not adversely affect the above quoted design basis, even with  !

assumed SONGS 2 and 3 fuel enrichments of up to 5.1%. The effective neutron  ;

multiplication factor will remain consistently less than 0.95 in all l postulated accident and storage conditions applicable to the spent fuel pool and the cask pool.

The following changes are proposed to Specification 3.7.18 to allow spent fuel consolidation and storage:

a. All page headers and the title will be changed from " Spent Fuel Assembly Storage" to " Spent Fuel Storage."

b. In the LCO, the words "each spent fuel assembly" will be changed to

" spent fuel." The words " Figure 3.7.18-1 and Figure 3.7.18-2" will be changed to " Figure 3.7.18-1 or Figure 3.7.18-2," consistent with ,

Figure 3.7.18-1 being applicable only to SONGS 1 fuel and Figure  !

3.7.18-2 being applicable only to SONGS 2 and 3 fuel. The words "or i in accordance with Specification 4.3.1.1" will be changed to "or the i fuel shall be stored in accordance with Licensee. Controlled )

Specification 4.0.100," which provides a more direct reference to the applicable administrative controls,

c. Under APPLICABILITY and the REQUIRED ACTION column, the words " fuel assembly" will be changed to " fuel".

d. Under " SURVEILLANCE," the words "in accordance with Figure 3.7.18-1 and Figure 3.7.18-2 or Specification 4.3.1.1" will be replaced with the words "in accordance with LC0 3.7.18," which covers the two I figures and the applicable administrative controls in Licensee l Controlled Specification 4.0.100. I

e. Under " SURVEILLANCE" and " FREQUENCY," the words " Prior to storing the fuel assembly" will be changed to " Prior to storing the fuel." This change will extend the existing surveillance to storage of consolidated fuel.

i i i Changes to Technical Specification 4.3:

j Specification 4.3, Fuel Storage, describes the design features of the SFPs.

It specifies fuel spacing and burnup requirements, maximum permissible values i of k-eff under different conditions, restrictions on pool drainage, and pool capacity (1542 assemblies).

l l It is proposed to make the following changes to Specification 4.3. ,

i j j a. In Specifications 4.3.1.1.a, 4.3.1.1.e, 4.3.1.1.f, and 4.3.1.1.g, the j term " Fuel assemblies" will be changed to " Fuel."

I

b. Specification 4.3.1.1.c will be changed from "A nominal 8.85 inch center to center distance between fuel assemblies placed in Region II;" to "A nominal 8.85 inch center to center distance between fuel ,

storage cells in Region II;." l Similarly, Specification 4.3.1.1.d will be changed from "A nominal 10.40 inch center to center distance between fuel assemblies placed in Region I;" to "A nominal 10.40 inch center to center distance between fuel storage cells in Region I;."

The above changes reflect the fact that the nominal center to center distances of 8.85 inches and 10.40 inches do not apply to fuel assemblies but rather apply to the as-designed fuel storage cells of I Region II and Region I respectively.

c. In Specification 4.3.1.1.e, which allows fuel storage in Region I with no restrictions, the phrase "with no restrictions" will be changed to "with no burnup restrictions." This change is needed to distinguish "burnup restrictions," which are intended in this specification to ,

apply to all fuel from " load restrictions," which will apply only to i consolidated fuel canisters. These load restrictions will be implemented as part of administrative controls which are described in Attachment G.

d. In Specification 4.3.1.1.f, the words " Figure 3.7.18-1 and Figure l 3.7.18-2" will be changed to " Figure 3.7.18-1 or Figure 3.7.18-2," for l the same reason given in proposed change "b" to Specification 3.7.18.

Also, the word "are" will be changed to "is" consistent with the change from " Fuel assemblies" to " Fuel."

e. In Specification, 4.3.1.1.g, the words " Figure 3.7.18-1 and Figure 3.7.18-2" will be changed to " Figure 3.7.18-1 or Figure 3.7.18-2," for the same reason given in proposed change "b" to Specification 3.7.18.

Also, the words "will be stored in compliance with the Licensee Controlled Specification" will be changed to "will be stored in compliance with Licensee Controlled Specification 4.0.100" to provide the specific reference.

1

i

f. In Specification 4.3.2, the water depth limit of "23 feet" above i stored fuel assemblies will be corrected to "a 23 feet." Also " fuel assemblies" will be changed to " fuel."

9 In Specification 4.3.3, the SFP storage capacity limit "no more than 1542 fuel assemblies" will be changed to "no more than 1542 fuel l assemblies (total) without storage of consolidated fuel assemblies, or 2867 fuel assemblies (total) with storage of consolidated fuel assemblies."

i The changes to the SONGS 2 and 3 Bases as the result of PCN-443 are shown in Attachment H. The changes to the Attachment H Bases pages as the result of PCN-443 and PCN-449 are shown in Attachment I. Attachments H and I are for .

information only. l

Discussion of Accidents l

Potential accidents that apply to fuel consolidation and storage are briefly )

discussed below. Some of these accidents are variations of accidents evaluated previously in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). For further discussion of the consequences of each accident, refer to Chapter 6, Section 6.3, " Accident Evaluation," of Enclosure 2.

l The fuel of reference in the discussion below is SONGS 2 and 3 fuel. Due to the longer decay time of SONGS 1 fuel, the consequences of damage to this fuel are negligible. l l

A. Design Basis Fuel Handling Accidents (Dropped Fuel Assembly or Consolidated Fuel Canister)

Dropped Fuel Assembly: i Of all the potential fuel assembly drop events, the most limiting event is a 254-inch drop onto the SFP floor of a vertically-oriented fuel assembly that has decayed for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, followed by rotation of the fuel assembly to the horizontal position. This event is described in the Updated Final Safety Analysis Report (UFSAR), Section 15.7. The postulated bounding event results in a total of 60 fuel rods failing in the dropped assembly, which will not change as the result of fuel consolidation and storage activities. The UFSAR shows that the resulting post-accident doses are well within (less than 25% of) the 10 CFR 100 limits and the doses to control room personnel meet 10 CFR 50, Appendix A, General Design

, Criterion (GDC) 19 limits.

The probability of a spent fuel assembly drop during movement of spent l

fuel is slightly increased by fuel consolidation because the candidate fuel assemblies for consolidation will be moved from their individual rack cell location to the cask storage pool. However, this increase in probability is not significant since the process and equipment used to move fuel assemblies will not be changed. Additionally, fuel movement activities are performed by personnel trained, qualified, and certified in

fuel handling operations. Therefore, the increase in probability of a spent fuel assembly drop due to fuel consolidation is not significant.

Since fuel assemblies selected for consolidation must have decayed at least six (6) months rather than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the consequences of the design basis accident discussed above have not increased.

This accident is further discussed in Section 6.3.1 of Enclosure 2.

Dropped Consolidated Fuel Canister:

The maximum drop of a consolidated fuel canister onto a fuel assembly is analyzed in Section 6.3.2 of Enclosure 2. The results indicate that the 1 fuel in the target fuel assembly, which is potentially more radioactive than fuel in the canister, would not be damaged. )

Of all the potential consolidated fuel canister drop events, the limiting l event is a 74-inch drop of a consolidated fuel canister from the Spent Fuel Handling Machine (SFHM) into a rack cell containing a consolidated fuel canister. Although the structural integrity of the racks would not be impacted and both consolidated fuel canisters would remain intact, it is conservatively assumed that all 944 fuel rods within the two canisters (472 rods / canister x 2 canisters) are damaged.

The probability of this accident is not expected to' vary significantly l from the probability of a fuel assembly drop accident because the methods and equipment used to move consolidated fuel canisters will not be significantly different from those used for fuel assemblies.

Additionally, effective training methods, administrative controls, and equipment design will be developed to minimize the likelihood of dropping a canister during the consolidation process.

Spent fuel will have decayed at least six months prior to being consolidated. The criticality calculations show that with 1800 ppm boron in the spent fuel pool water (Technical Specifications limit of 1850 ppm  !

includes 50 ppm measurement uncertainty) there are no criticality consequences of postulated consolidated fuel canister drops. In all cases the structural integrity of the racks will be maintained. The portions of the canisters where fuel is contained (above and inclusive of the bottom plate) will maintain their structural integrity in all drop cases. Hence, no fuel in the canisters or racks will be compromised by a dropped ,

canister.

The offsite doses which result from this dropped canister scenario are l bounded by the fuel assembly drop event discussed earlier (60 failed rods in an assembly which has decayed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) and are well within the limits (less than 25%) imposed by 10 CFR 100. The control room doses meet the l GDC 19 limits when crediting the control room emergency air cleanup I system. Therefore, the consequences of a consolidated fuel canister drop event remain enveloped by the limiting fuel assembly drop event discussed in the UFSAR and will not be increased by the proposed fuel consolidation j activity. )

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1 l

l J

This accident is further discussed in Sections 3.1.6.1 and 6.3.2 of Enclosure 2.

B. Cask Handling Crane Load Drops The types of loads currently lifted by the cask handling crane include spent fuel casks, transshipment casks, and the crane load block. To support consolidation activities, lifts of the fuel consolidation equipment will also be performed by the cask handling crane. As discussed in Section 9.1.4.3 of the UFSAR, the travel path of the cask handling crane does not extend over spent fuel in the spent fuel pool.

Administrative controls included in Attachment G will prohibit operation of the cask handling crane, including the crane load block, within ten feet of the edge of the cask pool, when fuel is present in the cask pool during consolidation. Other heavy load lifts will be prohibited over the cask pool when fuel is present in the cask pool during consolidation. The handling of heavy loads by the cask handling crane is governed by the SONGS heavy loads program which has received NRC approval. The movement of fuel consolidation equipment by the cask handling crane will be evaluated under the heavy loads program. Thus, an accident resulting from cask handling crane loads dropping into the spent fuel pool or the cask pool is not credible.

It is expected that the consolidation work station in the cask pool will be temporarily removed prior to lifting or moving a spent fuel cask or a transshipment cask over the cask pool. Other than when inserting or  !

removing the consolidation work station, the equipment and procedures used to lift and move cask handling crane loads will be unaffected by fuel consolidation. Therefore, the probability and consequences of a spent fuel cask or transshipment cask drop are not significantly increased by i the proposed fuel consolidation activity. l l

i This accident is further discussed in Section 6.3.10 of Enclosure 2. The cask drop analysis is presented in Section 6.3.11 of Enclosure 2.

C. Spent Fuel Pool Gate Drop Accident Current gate lift height restrictions (no more than 30 inches above the racks) will be maintained for fuel consolidation. With these restrictions, it has been determined that fuel in only one rack cell (either a spent fuel assembly with 236 rods or a consolidated fuel canister with 472 rods) would be impacted.

The probability of a spent fuel pool gate drop is not significantly increased by fuel consolidation because the process and equipment used to move the gate will not change and because the gate will be kept open and not moved or removed when fuel is located in the cask pool during consolidation (administrative control).

Despite the additional fuel rods in a consolidated fuel canister (472 versus 236 in a fuel assembly), the minimum six month decay time allows more than 99.9% of the radioactive gases to decay. Thus, the failure of

I l

l the 236 rods in one fuel assembly 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown is more limiting than the failure of 472 rods in a consolidated fuel canister. )

With the analysis demonstrating impact of fuel in only one cell, offsite doses remain well within (less than 25% of) the limits of 10 CFR 100 l

without crediting the FHB filters. The control room emergency air cleanup l system will maintain control room doses within 10 CFR 50, Appendix A, i GDC 19 limits. Therefore, the consequences of a gate drop will not be I significantly increased due to the proposed fuel consolidation activity.

This accident is further discussed in Sections 3.1.6.2 and 6.3.3 of Enclosure 2.

D. Test Equipment Skid Drop Current test equipment skid height restrictions (no more than 72 inches i above rack cells containing SONGS 2 and 3 fuel assemblies or 30 feet i 8 inches above those containing SONGS 1 fuel assemblies) will be maintained after fuel consolidation is implemented. Restricting the skid height above the racks will ensure that the potential depth of penetration into the racks is not sufficient to damage stored fuel.

The probability of a test equipment skid drop is not affected by fuel consolidation because the methods and equipment used to move the skid will not change. In addition, there are no adverse criticality consequences of a test equipment skid drop on a fuel assembly or a consolidated fuel canister, since the structural configuration of the fuel or of the impacted storage rack cells is not significantly changed because of the l drop impact.

Since no fuel is damaged, the consequences of a test equipment skid drop ,

will not be significantly increased due to the proposed fuel consolidation !

l activity, This accident is further discussed in Sections 3.1.6.2 and 6.3.3 of i l Enclosure 2.

E. SFP Leakage due to Load Drops I

)

l This accident is bounded by the postulated empty rack drop previously  ;

l evaluated in subsection 4.7.4.4 of the Reracking Licensing Report l l (Reference 1). The various loads considered for drops are a fuel assembly l i (dry weight 1540 pounds), a loaded consolidated fuel canister (dry weight l l 2904 pounds), and the SFP gate (4650 pounds, including accessories). Also  ;

l the maximum SFP water loss (49 gal / min) is well within the makeup water l supply capability (150 gal / min), thereby assuring that the required SFP l water level can be maintained. Analyses also show that FHB integrity is l maintained and no water leakage to soil occurs. l The probability of SFP leakage due to load drops is not significantly increased since the methods and equipment used to handle heavy loads l remain the same.

l l

l SFP leakage due to load drops is further discussed in Section 6.3.4 of Enclosure 2.

F. Mispositioning of a Consolidated Fuel Canister i The probability of mispositioning a consolidated fuel canister has not  !

been specifically determined. However, it is expected to be of the same order of magnitude as that for mispositioning of a spent fuel assembly.

The reason is that the methods and equipment used to move and position l consolidated fuel canisters in rack cells will not be significantly I different from those used for fuel assemblies. Additionally, fuel movement activities are and will continue to be performed by personnel trained, qualified, and certified in fuel handling operations.  ;

The potential consequences of a mispositioned consolidated fuel canister related to fuel criticality are acceptable. The burnups of the fuel stored in the SFP before, during, and after consolidation will conform to (

the criteria provided in the Technical Specifications. With a minimum ,

1800 ppm baron concentration in the pool and the Region II racks loaded l with fuel which meets the burnup criteria of Technical Specification 3.7.18, k-eff remains less than 0.90 for a consolidated fuel canister mispositioned in the Region II racks. Therefore, the consequences of I mispositioning a consolidated fuel canister are not significantly higher l than the consequences of mispositioning a fuel assembly.

1 This accident is further discussed in Sections 3.1.6.3 and 6.3.5 of l Enclosure 2. '

G. Maximum Flow Blockage to Cool Spent Fuel Flow blockage to a consolidated fuel canister may be caused by either damage to the canister or loose material in the spent fuel pool or cask pool. Canisters will be inspected prior to being placed in the cask pool (prior to loading with fuel), and if damaged during movement or placement in the spent fuel pool. Additionally, the existing foreign material exclusion control in the spent fuel pool area will be utilized for fuel consolidation. Therefore, the probability of blocking flow to a consolidated canister will not be significantly increased.

1 The temperature effects of a postulated flow blockage of a consolidated i fuel canister were evaluated relative to the anticipated maximum cladding temperature of 700 F during reactor full power. Each rack storage cell has either large or multiple flow holes to virtually eliminate the possibility that all flow in a cell would be blocked by debris or foreign material. In addition, the flow openings in the consolidated fuel canisters will be designed to maintain a clear flow area of at least 20%

under all postulated blockage conditions. For the postulated 80% flow blockage to fuel in the canister, the resulting maximum cladding temperature is 233.1*F, which is well below the anticipated maximum cladding temperature of 700*F. Therefore, the consequences of flow

I blockage will not be significantly increased by the proposed fuel l consolidation activity.

I This accident is further discussed in Section 6.3.6 of Enclosure 2.

Design criteria for the consolidated fuel canisters are discussed in ]

Section 4.3.2 of Enclosure 2.

H. Consolidation Work Station Accidents Fuel consolidation will require fuel handling operations at the work l station. Since fuel handling methods and equipment will not be significantly different from those currently used for fuel handling, including fuel reconstitution, no new failure mechanisms are expected. To minimize the probability of consolidation work station accidents, personnel training methods, equipment design, and administrative controls will be utilized.

Administrative controls will require a minimum decay time of six months for spent fuel prior to its movement into the cask pool for consolidation.

This restriction ensures that the limiting radiological offsite and control room dose consequences from a work station accident remain bounded by a fuel assembly drop. The results are well within (less than 25% of) 10 CFR 100 limits and meet 10 CFR 50, Appendix A, GDC 19 dose limits.

Fuel assemblies in the work station shall be separated by more than 12 inches of water from edge to edge to maintain neutronic isolation (administrative control). The total spent fuel which will be permitted in the cask pool at any given time is 553 fuel rods (administrative control).

This quantity of fuel is equivalent to two full SONGS 2 or 3 fuel assemblies plus a damaged fuel rod storage canister or basket containing up to 81 fuel rods. Criticality analysis for the consolidation work I station accidents, as described in section 3.1.6 of Enclosure 2, has shown l that in the worst case scenario, at 1800 ppm boron concentration, k-eff l will be below 0.95. Additional administrative controls will be imposed to ensure that a minimum of 400 fuel rods or non-fuel rods will be loaded ,

into a SONGS 2 or SONGS 3 consolidated fuel canister and a minimum of 324 l fuel rods or non-fuel rods will be loaded into a SONGS 1 consolidated fuel l canister. The canisters shall be designed for storage of fuel rods within a maximum allowed rod pitch. For consolidated fuel canisters not fully loaded, the rod pitch shall be maintained by restraints inserted within the canister to ensure against rod displacement during canister movement l (administrative control). These limitations ensure that the k-eff for a i loaded consolidated fuel canister will not exceed 0.95 with zero ppm boron '

concentration, considering the worst case pitch between consolidated rods.

With 1800 ppm boron concentration in the pool, k-eff will be below 0.88 assuming the worst case pitch between consolidated rods. )

l These accidents are further discussed in Sections 3.1.6.4, 3.1.6.5, 6.3.8, i 6.3.9 and 6.3.10 of Enclosure 2.

l

I I. Consolidated Fuel Canister Stuck in a Spent Fuel Rack The probability of a consolidated fuel canister being stuck in a spent l fuel rack is not known from experience since fuel consolidation on a large I commercial scale has not been implemented and fuel consolidation I demonstration projects to date have not reported this type of occurrence. l This subject will be reviewed as experience is gained. However, the  !

consolidated fuel canisters will have the same approximate cross sectional dimensions as fuel assemblies and similar handling equipment and methods  ;

will be used. Therefore, the probability of this accident is expected to be similar to that for a stuck fuel assembly.

The canisters shall be designed to be handled by the spent fuel handling machine (SFHM) and to accommodate all operational and handling loads. A  ;

design requirement will be imposed that the canisters be capable of l withstanding the maximum SFHM lift load of 6000 pounds and remain intact '

with no fuel spillage. This is consistent with the criteria utilized in i Section 4.6.3 of the Reracking Licensing Report (Reference 1) for the i spent fuel racks and a jammed fuel assembly. Using these design criteria and restrictions, while deformation of rack cell geometry is possible, compliance with the criticality acceptance criterion (k-eff :s 0.95) will still be maintained. Therefore, the consequences of a stuck consolidated fuel canister would be bounded by the consequences of a stuck fuel assembly, and there is no significant consequence increase due to the proposed fuel consolidation activity.

See Section 6.3.13 of Enclosure 2 for additional information on this accident.

J. Loss of Spent Fuel Pool Cooling Loss of spent fuel pool (SFP) cooling has been postulated in Section 9.1.3 of the UFSAR. The probability of this type of accident is not affected by fuel consolidation and storage because the existing SFP cooling system will perform its design function without modification. However, the overall design basis heat load will be increased due to an increased number of spent fuel elements stored. The cask pool may be used for ,

temporarily holding spent fuel assemblies during consolidation. Loss of  !

cooling flow to the cask pool has not been specifically analyzed. I However, because of administrative controls which limit the amount of fuel permitted in the cask pool during consolidation and require the gate between the cask pool und the SFP to be open when fuel is present in the cask pool, this accident scenario is bounded by the SFP boiling case discussed below.

An analysis of loss of SFP cooling has been performed using the highest expected fuel burnups, and design basis consolidated fuel heat load. This analysis shows that, without crediting the FHB filters, the offsite doses will remain well within (less than 25% of) the 10 CFR 100 limits. Since  ;

the reactivity will decrease with increasing temperature at 0 ppm boron i concentration, there will be no adverse criticality effects. l l

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1 Additionally, the normal makeup sources to the SFP (the refueling water storage tanks) will continue to maintain adequate inventory and flow l capacity (150 gallons per minute or gpm) to compensate for evaporative I losses due to boiling (< 112 gpm maximum). Therefore, the consequences of a loss of spent fuel pool cooling will not be significantly increased by ,

the proposed fuel consolidation activity. '

Loss of SFP cooling and makeup provisions are further discussed in Sections 3.2.2.7 and 3.2.3 of Enclosure 2. The temperature effects of SFo boiling on the SFP liner plate and concrete structure are discussed in  !

Section 4.1.3.1.D of Enclosure 2 and are acceptable.

K. Limiting Component Cooling Water (CCW) System Heat Load Effects on Spent Fuel Pool Cooling l 1

The maximum calculated heat load for the CCW system occurs during a loss of Coolant Accident (LOCA). The probability of a LOCA is not affected by fuel consolidation and storage, since these activities are not, of l themselves, LOCA initiators. For the purposes of assessing the heat load on the CCW system, the LOCA is divided into two phases, " safety injection" and " recirculation."

During the safety injection phase the SFP heat load is isolated from the l CCW system. During the recirculation phase CCW system cooling to the spent fuel pool may be reestablished manually. The recirculation phase represents the highest design heat load for the CCW system. Considering l the limiting consolidated fuel heat load contribution from the SFP, based l on a minimum of 60 days decay of the most recent half-core discharged into the SFP (where 60 days represents the refueling outage length), the CCW l system still has adequate capacity to remove its design basis heat load.

Therefore, the consequences of a limiting design basis event on the CCW system are not significantly increased by the proposed fuel consolidation activity.

Limiting CCW heat load is also discussed in Section 6.3.14 of Enclosure 2.

L. Design Basis Earthquake Among natural phenomena, the only one of concern inside the FHB is the design basis earthquake (DBE) event. ihe probability of occurrence of a DBE is unaffected by fuel consolidation and storage.

The spent fuel racks are designed, and the consc:;de J fuel canisters will be designed, to Seismic Category I requirements, as defined by Regulatory Guide 1.29 (Revision 3). Additionally, the individual storage cells, individual racks, and rack layout in the SFP comply in all respects with published NRC requirements. All portions of the SFP cooling system, and other systems which are not designed as Seismic Category I but are located close to essential portions of the SFP cooling system, satisfy Seismic II/I requirements and, therefore, would not affect the performance of essential functions during or following a seismic event. The Fuel Handling Building and the SFP and cask pool structures have been evaluated

for the increased loading from fully-loaded consolidated fuel canisters, consistent with the applicable UFSAR criteria, and the loads have been ,

found to be within the design allowables. The fuel consolidation )

equipment that will be installed within and adjacent to the cask pool will '

be analyzed and either restrained or anchored as appropriate to meet seismic interaction II/I requirements. Therefore, the consequences of a DBE event are not significantly increased by fuel consolidation and i storage.

The DBE event is further discussed in Section 6.3.12 of Enclosure 2.

Safety Analysis The proposed change shall be deemed to involve a significant hazards l consideration if there is a positive finding in any one of the following areas:

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No In the course of previous analyses and the analyses required to support the consolidation and storage of spent fuel assemblies generated by the San Onofre Nuclear Generating Station Units 1, 2 and 3 (SONGS 1, 2 and 3),

the enveloping scenarios described below have been considered. The limiting event or accident is considered that which produces the greatest  ;

radiological dose consequences.

1) Design Basis Fuel Handling Accidents l Postulated fuel handling accidents consider drops of either a spent fuel assembly or a consolidated fuel canister in the spent fuel pool  !

(SFP) or cask pool. In addition to damage to the dropped fuel }

assembly or consolidated fuel canister, a fuel assembly or ,

consolidated fuel canister seated in the SFP or the cask pool may be j impacted by the drop. Alternatively, the dropped assembly or canister  ;

may fall over an empty rack cell, or fall onto the pool floor / liner. l These various scenarios have been considered.

i The reference fuel in the analysis presented below is SONGS 2 and 3 i fuel. Due to the longer decay time, lower burnup, and lower operating power of SONGS 1 fuel, the consequences of damage to SONGS 1 fuel are bounded by the consequences of damage to SONGS 2 and 3 fuel. [

i a) Dropped Fuel Assembly l i

The limiting and design basis fuel assembly drop event is a 254-inch  !

drop of a vertically-oriented fuel assembly, which has decayed for  !

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, onto the SFP floor, followed by rotation of the fuel l assembly to the horizontal position. The postulated bounding event j

results in a total of 60 fuel rods failing, which will not change as a result of fuel consolidation.

The probability of a spent fuel assembly drop during movement of spent fuel is slightly increased by fuel consolidation because the candidate fuel assemblies are moved from their individual rack cell location to the cask pool for consolidation. However, this increase in probability is not significant since the process and equipment used to move fuel assemblies will not be changed. Additionally, fuel movement activities will be performed by personnel trained, qualified, and certified in fuel handling operations. Therefore, the increase in probability of a spent fuel assembly drop due to fuel consolidation is not significant.

The SFP water leakage consequences of a fuel assembly drop are bounded by the consequences of a postulated empty spent fuel rack drop. The resulting iaakage (approximately 49 gallons per minute) is well within the makeup water supply capability (150 gallons per minute). Additionally, the water loss would be contained within the  !

spent fuel pool leak chase system and would not be released to the soil or the environment.

Spent fuel assemblies will be decayed (subcritical) at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to being moved and at least 6 months prior to being cor,solidated. Administrative controls will require that fuel assemblies being moved to and from the consolidation work station, and when in the work station, be separated by more than 12 inches of water from edge to edge to maintain neutronic isolation.

Criticality calculations show that with 1800 parts per million (ppm) minimum boron concentration in the SFP water (Technical l Specifications limit of 1850 ppm includes 50 ppm measurement uncertainty), a dropped fuel assembly event will not result in fuel l criticality. i Without crediting filtration by the fuel handling building (FHB) post-accident cleanup units, the offsite doses which result from this scenario are sell within the required limits, i.e., less than l 25 percent (%) of the limits imposed by 10 CFR 100. The control )

room doses meet 10 CFR 50, Appendix A, General Design Criterion (GDC) 19 limits when crediting the control room emergency air cleanup system. Therefore, the consequences of a fuel handling accident remain enveloped by the fuel assembly drop event.

In conclusion, the probability and consequences of a fuel assembly drop event will not be significantly increased by the proposed fuel consolidation activity.

b) Dropped Consolidated fuel Canister A dropped consolidated fuel canister event does not involve significantly new failure mechanisms compared with a dropped fuel assembly event. The limiting event in this category is a 74-inch l

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l drop of a consolidated fuel canister from the spent fuel handling

machine (SFHM) into a rack cell containing a consolidated fuel l canister. The structural integrity of the racks would not be j impacted and both consolidated fuel canisters would remain intact. ,

i However, it is conservatively assumed that all 944 fuel rods within I the two canisters (472 rods / canister x 2 canisters) are damaged. ]

The probability of a consolidated fuel canister drop is not expected to vary significantly from that expected for a fuel assembly drop because the methods and equipment used to move consclidated fuel  !

canisters will not be significantly different.from those used for fuel assemblies. Additionally, effective training methods, administrative controls, and equipment design will be developed to minimize the likelihood of dropping a canister during the consolidation process.

The SFP water leakage consequences of a consolidated fuel canister '

drop are bounded by the consequences of a postulated empty spent fuel rack drop as discussed previously in Item 1.1)a). l l

The criticality calculations show that, with the required 1800 ppm  !

boron concentration in the SFP and cask pool water, there are no i criticality consequences of postulated consolidated fuel canister  ;

drops. In all cases, the structural integrity of the racks will be maintained. The portions of the canisters where fuel is contained (

(above and inclusive of the bottom plate) will maintain their '

structural integrity in all drop cases.

The offsite doses which result from this scenario are bounded by the fuel assembly drop event discussed previously in Item 1.1)a) (60 failed fuel rods in an assembly which has decayed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) and are well within (less than 25% of) the limits imposed by 10 CFR 100. l The control room doses meet the GDC 19 limits when crediting the control room emergency air cleanup system. Therefore, the consequences of a consolidated fuel canister drop event are enveloped by the limiting fuel assembly drop event.

In conclusion, the probability and consequences of the limiting fuel drop event will not be significantly increased by storing consolidated fuel in canisters.

2) Spent Fuel Pool (SFP) Gate Drop The limiting case is a SFP gate drop on a fuel assembly. Analysis has shown that only one assembly would be impacted and all 236 rods in the assembly potentially damaged subsequent to a drop of the SFP gate.

The radiological consequences are shown to be acceptable (less than 25% of 10 CFR 100 limits). l Current gate lift height restrictions (no more than 30 inches above l the racks) will be maintained for fuel consolidation. With these I restrictions, fuel in only one rack cell (either a spent fuel assembly l

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with 236 rods or a consolidated fuel canister with 472 rods) would be impacted with all rods in the fuel assembly or canister being  !

potentially damaged.

The probability of a SFP gate drop is not significantly increased by fuel consolidation because the process and equipment used to move the gate will not change and because the gate will be kept open and not moved or removed when fuel is located in the cask pool during consolidation (administrative control).

Despite the additional fuel rods in a consolidated fuel canister (472 rods versus 236 rods in a fuel assembly), the minimum six month decay time allows more than 99.9% of the radioactive gases to decay.

l Thus, a gate drop that'results in a damaged fuel assembly 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

! after shutdown is more limiting than a gate drop-that results in a l damaged consolidated fuel canister. With the analysis demonstrating l impact of feel in only one cell, offsite doses remain well within (less than 25% of) the limits of 10 CFR 100 without taking credit for the FHB filters. The control room emergency air cleanup system will maintain control room doses within GDC 19 limits.

l Therefore, the probability and consequences of a gate drop will not be

! significantly increased due to the proposed fuel consolidation activity.

3) Test Equipment Skid Drop Current test equipment skid height restrictions (no more than 72 inches above rack cells containing SONGS 2 and 3 fuel assemblies or 30 feet 8 inches above those containing SONGS 1 assemblies) will be maintained after fuel consolidation is implemented. These ,

restrictions will ensure that the potential depth of penetration of 1 l

test equipment skid into the racks is not sufficient to damage stored l fuel. '

The probability of a test equipment skid drop is not affected by fuel l consolidation because the methods and equipment used to move the skid will not change. In addition, there are no adverse criticality consequences of a test equipment skid drop on a fuel assembly or consolidated fuel canister, since the structural configuration of the fuel or of the impacted storage rack cells is not significantly changed because of the drop impact.

Since no fuel is damaged, the probability and consequences of a test equipment skid drop will not be significantly increased due to the proposed fuel consolidation activity.

l 4) Cask Handling Crane Load Drops

! The types of loads currently lifted by the cask handling crane include i

! spent fuel casks, transshipment casks, and the crane load block. To  !

support consolidation activities, lifts of the fuel consolidation i

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equipment will also be performed by the cask handling crane. The travel path of the cask handling crane does not extend over spent fuel in the SFP. Administrative controls will prohibit operation of the cask handling crane, including the crane load block, within ten feet of the edge of the cask pool when fuel is present in the cask pool during consolidation. The handling of heavy loads by the cask handling crane is governed by the SONGS heavy loads program which has received Nuclear Regulatory Commission (NRC) approval. The movement of fuel consolidation equipment by the cask handling crane will be evaluated under the heavy loads program. Thus, an accident resulting from cask handling crane load drops into the SFP or onto irradiated fuel in the cask pool is not credible.

l It is expected that the consolidation work station in the cask pool will be temporarily removed prior to any spent fuel cask, transshipment cask, or other load lifts / movements over the cask pool.

Other than insertion and removal of the consolidation work station, i the equipment and procedures used to lift and move cask handling crane l loads will be unaffected by fuel consolidation. l Therefore, the probability and consequences of a spent fuel cask or i transshipment cask drop are not significantly increased by the- ,

proposed fuel consolidation activity.

5) Mispositioning of a Consolidated Fuel Canister The probability of mispositioning a consolidated fuel canister is-expected to be comparable to that for mispositioning of a spent fuel assembly because the metheds and equipment used to move and position consolidated fuel canisters in rack cells will not be significantly different from those used for fuel assemblies. Additionally, fuel movement activities are and will continue to be performed by personnel trained, qualified, and certified in fuel handling operations.

The potential consequences of a mispositioned consolidated fuel canister relate to fuel criticality. The burnup of the fuel stored in the SFP before, during, and after consolidation will conform to the criteria provided in the Technical Specifications. With the minimum required 1800 ppm (1850 ppm plus 50 ppm measurement uncertainty) boron concentration in the SFP and the Region II racks loaded with fuel which meets the burnup criteria of Technical Specification 3.7.18, k-eff remains less than 0.90 for a consolidated fuel canister mispositioned in the Region II racks.

Therefore, the probability and consequences of mispositioning a consolidated fuel canister are not significantly higher than the i probability and consequences of mispositioning a fuel assembly. j l

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6) Maximum Flow Blockage to Cool Spent Fuel Flow blockage to a consolidated fuel canister may be caused by either i damage to the canister or loose material in the spent fuel pool or cask pool. Canisters will be inspected prior to being placed in the cask pool (prior to loading with fuel), and if damaged during movement or placement in the spent fuel pool. Additionally, the existing foreign material exclusion control in the spent fuel pool area will be utilized for fuel consolidation. Therefore, the probability of t blocking flow to a consolidated fuel canister will not be i

significantly increased.

The temperature effects of a postulated flow blockage of a l consolidated fuel canister were evaluated relative to the anticipated I j maximum cladding temperature of 700 degrees Fahrenheit (700*F) during I reactor full power. Each rack storage cell has large or multiple flow  !

holes to virtually eliminate the possibility that all flow in a cell would be blocked by debris or foreign material. The flow openings in the canisters will be designed to maintain a clear flow area of at ,

least 20% under all postulated blockage conditions. For the '

postulated 80% flow blockage, the resulting maximum cladding temperature is 233.1*F, which is well below the maximum anticipated  !

cladding temperature of 700*F during reactor full power. l Therefore, the probability and consequences of flow blockage will not be significantly increased by the proposed fuel consolidation  :

activity. l

7) Loss of Spent Fuel Pool (SFP) Cooling )

The probability of loss of SFP cooling is not affected by fuel consolidation because the existing SFP cooling system will perform its design function without modification.

The overall design basis (maximum abnormal) heat load will be increased due to an increased number of spent fuel elements stored.

The cask pool may be used for temporary storage of spent fuel assemblies during consolidation. Loss of cooling flow to the cask J pool has not been specifically analyzed. However, because of administrative controls which limit the amount of fuel permitted in the cask pool during consolidation and require the gate between the cask pool and the SFP to be open when fuel is present in the cask l pool, this accident scenario is bounded by the SFP boiling case  !

discussed below.

An analysis of loss of SFP cooling has been performed using the design basis consolidated fuel heat load. This analysis shows that, without crediting the FHB filters, the offsite doses will remain well within (less than 25% of) the 10 CFR 100 limits. Since the reactivity will decrease with increasing temperature at 0 ppm boron concentration, there will be no adverse criticality effects. Additionally, the 1

i normal makeup sources to the SFP will continue to maintain adequate inventory and flow capacity (150 gallons per minute or gpm) to compensate for evaporative losses due to boiling (< 112 gpm maximum).

The temperature effects of SFP boiling on the SFP liner plate and concrete structure have been determined to oe acceptable.

l Therefore, the probability and consequences of a loss of SFP cooling event will not be significantly increased by the proposed fuel consolidation activity.

8) Consolidation Work Station Accidents Fuel consolidation will require additional fuel handling operations.

However, since the fuel handling methods and equipment will not be significantly different from those currently used, consolidation work station accidents will be similar to fuel handling accidents already discussed in this Safety Analysis (dropped fuel assembly, dropped consolidated fuel canister, or other load drops). To avoid a significant increase in the probability of any of these accidents, personnel training methods, equipment design, and administrative controls will be util' N. Administrative controls will require a minimum decay time o. six months for spent fuel prior to its movement into the cask pool for consolidation. This restriction ensures that the limiting radiological offsite and control room dose consequences from a work station accident remain bounded by a fuel assembly drop.

The results are well within (less than 25% of) 10 CFR 100 and meet GDC 19 dose limits.

Fuel assemblies in the work station shall be separated by more than 12 inches of water from edge to edge to maintain neutronic isolation (administrative control). The total spent fuel which will be permitted in the cask pool at any given time is 553 fuel rods (administrative control). This quantity of fuel is equivalent to two full SONGS 2 or 3 fuel assemblies plus a damaged fuel rod storage canister or basket containing up to 81 fuel rods. A criticality analysis has shown that, in the worst case scenario, at 1800 ppm (Technical Specification limit of 1850 ppm includes 50 ppm measurement uncertainty) boron concentration, k-eff will be below 0.95.

Additional administrative controls will be imposed to ensure that a minimum of 400 fuel rods or non-fuel rods will be loaded into a SONGS 2 or SONGS 3 consolidated fuel canister and a minimum of 324 l

fuel rods or non-fuel rods will be loaded into a SONGS 1 consolidated fuel canister. The canisters shall be designed for storage of fuel rods within a maximum allowed rod pitch. For canisters not fully l loaded, the rod pitch shall be maintained by restraints inserted within the canister to ensure against rod displacement during canister movement (administrative control). These limitations ensure that the k-eff for a loaded consolidated fuel canister will not exceed 0.95 with zero ppm boron concentration, considering worst case pitch between consolidated rods. With 1800 ppm boron concentration in the pool, k-eff will be below 0.88 for the worst case canister pitch between rods. Thus, there are no adverse criticality consequences r

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since the minimum number of rods consolidated in a canister is administratively controlled and SFP and cask pool boron concentration will be maintained at or above 1800 ppm during consolidation.

Therefore, the consequences of a consolidation work station accident are not significantly increased as a result of the proposed fuel consolidation activity.

9) Seismic Events The probability of occurrence of a seismic event is unaffected by the proposed fuel consolidation activity. The consequences of a design basis earthquake (DBE) have been analyzed, and the fuel consolidation process and consolidated fuel canisters will not affect the ability of the racks to maintain their required design basis function during and after a DBE. The spent fuel racks are designed, and the consolidated fuel canisters will be designed, to Seismic Category I requirements, and the consolidation equipment will be designed to Seismic Category II/I requirements as defined by NRC Regulatory Guide 1.29, Revision 3.

The consolidation process provides the capability to store more spent fuel (up to approximately 2867 fuel assemblies) than previously approved by the NRC (up to 1542 fuel assemblies) in the SFP. The fuel handling building and the SFP and cask pool structures have been evaluated for the increased loading from fully-loaded consolidated fuel canisters and the loads found to be within the design allowables.

Thus, the probability or consequences of a seismic event are not significantly increased by the proposed fuel consolidation activity.

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10) Consolidated Fuel Canister Stuck in a Spent Fuel Rack 1

The probability of a consolidated fuel canister being stuck in a spent fuel rack is not known from experience since fuel consolidation l demonstration projects conducted to date have not reported this type I of occurrence. However, the canisters will be designed to be handled l by the spent fuel handling machine (SFliM), will have the same ,

approximate cross-sectional dimensions as spent fuel assemblies, and l similar handling equipment and methods will be used. Therefore, the failure mechanisms are expected to be comparable to those for a stuck fuel assembly. On this basis, the probability of a consolidated fuel canister being stuck in a spent fuel rack is estimated to be comparable to that for a stuck fuel assembly.

The canisters will be designed to accommodate all operational and handling loads. A design requirement will be imposed that the canisters be capable of withstanding the maximum SFHM lift load of 6000 pounds and remain intact with no fuel spillage. This is consistent with the criteria utilized previously during SFP reracking for the spent fuel racks and a jammed fuel assembly. With these design criteria and restrictions, deformation of rack cell geometry would not be sufficient to exceed the criticality acceptance criterion

j (k-eff 1 0.95). Therefore, the consequences of a stuck consolidated fuel canister would be bounded by the consequences of a stuck fuel assembly.

Therefore, there is no significant increase in the probability or consequences of an accident previously evaluated due to the proposed fuel consolidation activity.

11) Limiting Component Cooling Water (CCW) System Heat Load Effects on Spent Fuel Pool Cooling The maximum calculated heat load for the CCW system occurs during a Loss of Coolant Accident (LOCA). The probability of a LOCA, and therefore the probability of maximum heat load being imposed on the CCW system, is not affected by fuel consolidation. The reason is that l

l spent fuel handling operations in the SFP or the cask pool are not, of

! themselves, LOCA initiators. For the purposes of assessing the heat load on the CCW system, the LOCA is divided into two phases, " safety injection" and " recirculation."

During the safety injection phase, the SFP heat load is isolated from the CCW system. During the recirculation phase, CCW system cooling to l

the SFP may be reestablished manually. The recirculation phase l represents the highest design heat load for the CCW system.

l Considering the limiting consolidated fuel heat load contribution from the SFP (assuming a minimum of 60 days decay of the most recent half-core discharged into the SFP), the CCW system has adequate capacity to still remove its design basis heat load.

l Therefore, the probability or consequences of a limiting design basis heat load event on the CCW system are not significantly increased by the proposed fuel consolidation activity.

Therefore, operation of the facility in accordance with this proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any l

accident previously evaluated?

l Response: No The proposed change will allow the consolidation of San Onofre Units 1, 2 and 3 spent fuel in canisters and the storage of these canisters along with fuel assemblies in the Units 2 and 3 spent fuel pools. Fuel consolidation is similar in nature to fuel reconstitution within a fuel assembly since individual rods are manipulated in both processes.

Accidents involving consolidated fuel canisters are similar in nature to fuel assembly handling accidents since both use similar fuel handling processes and equipment. Administrative controls will be instituted to provide assurance that postulated events involving consolidated fuel will l

be enveloped by the spectrum of design basis fuel handling accidents.

Furthermore, heavy load drops during spent fuel handling operations are accidents that have been previously evaluated. Additional evaluations have been performed to demonstrate that when the minimum boron j concentration requirements of the Technical Specifications have been met, the criticality criterion is satisfied for all postulated accidents.

Therefore, operation of the facility in accordance with the proposed i change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No i The issue of " margin of safety," when applied to spent fuel consolidation i and storage, includes the following areas: i

1) Nuclear criticality, l

2) Thermal-hydraulics, j

3) Mechanical, material and structural aspects, and

4) Offsite doses.

l These four areas are addressed below.

1) Nuclear Criticality The margin of safety that has been established for nuclear criticality is that, including all uncertainties, there is a 95% probability at a 95% confidence level that the effective neutron multiplication factor ,

(k-eff) in spent fuel pools shall be less than or equal to 0.95, under l l all normal and postulated accident conditions. This margin of safety )

has been adhered to in the criticality analyses for fuel consolidation ,

l and the storage of consolidated fuel canisters.

Criticality of fuel assemblies and consolidated fuel canisters in fuel storage racks is prevented by the rack design which precludes interactions between two fuel assemblies or two consolidated fuel canisters or between a fuel assembly and a consolidated fuel canister.

l This is accomplished by fixing the minimum separation between storage cells containing fuel assemblies or consolidated fuel canisters, using Boraflex, a neutron absorbing material, and utilizing strict l

administrative controls.

i During the consolidation process, fuel rods which cannot be

! consolidated will be placed in a damaged fuel rod canister or basket.

Fuel assemblies, consolidated fuel canisters, and damaged fuel rod canisters or baskets moving to and from the consolidation work station j l

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3 or present in the work station shall be separated by more than 12 inches of water, measured edge to edge, to ensure that they are neutronically isolated (administrative control). The total spent fuel which will be permitted in the cask pool at any give time is 553 fuel rods (administrative control). This quantity of fuel is equivalent to two full SONGS 2 or 3 spent fuel assemblies plus 81 fuel rods in a damaged fuel rod canister or basket. Additionally, the rod pitch inside partially loaded canisters shall be maintained by restraints inserted within the canister to ensure against rod displacement during canister movement (administrative control). .

The analytical methods utilized in the criticality analyses conform ~

with American National Standards Institute (ANSI) Standard N18.2-1973,

" Nuclear Safety Criteria for the Design of Stationary Pressurizer Water Reactor Plants," Section 5.7, Fuel handling Systems; ANSI Standard 57.2-1983, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear-Power Stations," Section 6.4.2; ANSI Standard N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety;" NRC Standard Review Plan (NVREG-0800),

Section 9.1.2, " Spent Fuel Storage"; and the NRC guidance, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," (April 1978), as modified (January 1979).

The criticality analyses performed for normal conditions assume zero boron concentration in the SFP water and worst-case fuel enrichments and burnups. Most credible accident conditions will not result in an increase in k-eff of the spent fuel racks. However, accidents, such as a heavy load drop, misloading a consolidated fuel canister or dropping a fuel assembly, can be postulated to increase reactivity.

For these accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This principle states that it is not required to assume two unlikely, independent events to ensure protection against a criticality accident. Therefore, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since the absence of boron would be the second unlikely event.

Worst case accident analyses have been performed that show that -

1800 ppm of soluble baron will maintain the spent fuel pool and cask storage pool k-eff less than 0.95, including uncertainties, at the required 95%/95% probability / confidence level.

2) Thermal-Hydraulics The relevant thermal-hydraulics considerations for determining if there is significant reduction in a margin of safety are: (1) maximum fuel temperature, and (2) increase in temperature of the water in the pool, and (3) increase in heat load rejection to the environment.

Similar to the criticality analysis, the SFP decay heat load  :

calculation assumes worst-case fuel loading, enrichment, and burnup.  !

The calculation uses the same methodology as that used for the J l

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original decay heat analysis. Standard Review Plan (SRP)

Section 9.1.3 criteria for maximum normal and maximum abnormal heat load conditions were used in this evaluation.

The effect of the increased heat load has been evaluated and it has been shown that, under the SRP maximum normal heat load, the existing spent fuel pool cooling system will maintain the bulk pool water temperature below 145'F. This value considers a single active failure of one spent fuel pool cooling system pump, coincident with a loss of offsite power, and is consistent with Standard Review Plan, Section 9.1.3.III.1.d. The 145*F temperature represents a small increase in the currently approved SFP temperature of 140*F. However, this temperature limit was very conservatively calculated, considering only heat losses through the spent fuel pool heat exchangers, and conservatively neglecting losses through evaporation to the spent fuel pool area, as well as conduction to the fuel handling building structure mass. This increase in spent fuel pool temperature does not represent a significant reduction in the margin of safety, since the affected portions of the spent fuel pool cooling system and other important to safety equipment in the fuel handling building are qualified for this slightly higher temperature and will still perform the necessary safety functions when required.

A thermal-hydraulic analysis has been performed which shows that the maximum local water temperatures along the fuel channels will remain below the nucleate boiling condition values, even with the maximum postulated flow blockage (80%) of the consolidated fuel canisters.

The maximum calculated fuel cladding temperature for the design basis condition is 233.1*F, which is well below the anticipated maximum cladding temperature of 700*F during full power operation of the reactor.

SONGS 2 and 3 conduct refueling by offloading either half the core or the full core. The full core offload refueling provides the greater of the two heat loads. Therefore, in addition to the SRP criteria, the heat load during refueling operations was also evaluated. For this case the heat load was evaluated assuming a two year refueling cycle, the spent fuel pool completely filled with consolidated fuel (except for the last core offload), and the full core offloaded at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decay. Under these conditions, a single SFP cooling pump with two heat exchangers will maintain the SFP temperature below 160'F, assuming the component cooling water temperature is 88 F and the ocean water temperature is 76*F. Thus, the SFP cooling system meets the single active failure criterion for the maximum refueling heat load condition.

With the postulated SRP maximum abnormal heat load, the bulk pool temperature will reach a maximum of 160 F with two pumps and two heat exchangers in operation. This maximum temperature is well below the SRP maximum temperature limit of 212*F. Also, according to the SRP

1 l l guidance, a single active failure need not be considered for the I maximum abnormal heat load case.

l The shutdown cooling system (SDCS), if available, can be used as an alternate heat dissipation path for cooling the SFP. The SDCS has

! been evaluated for the maximum normal and maximum abnormal heat loads and it has been determined that the system and interconnecting ties l are adequate to maintain the SFP temperature below 145'F for the maximum normal heat load and below 160*F for the maximum abnormal heat load. Since the maximum abnormal heat load bounds the maximum refueling heat load, there is no need to evaluate the SDCS for the maximum refueling heat load. For the maximum refueling heat load, the SDCS does not meet the single failure criterion for SFP cooling; however, the use of the SDCS for SFP cooling during Modes 5 and 6 of plant operation has previously been evaluated and considered acceptable by the NRC.

The heat load rejection to the environment will only increase by approximately 0.03%.

Thus, there is no significant reduction in a margin of safety, as determined by thermal-hydraulics considerations.

3) Mechanical, material, and structural aspects The main safety function of the spent fuel pool and the storage racks is to maintain the spent fuel assemblies and consolidated fuel canisters in a safe configuration through normal and/or abnormal loadings. Abnormal loads include an earthquake, impact due to a cask drop, drop of a spent fuel assembly or consolidated fuel canister, or drop of a heavy load including a spent fuel pool gate. The mechanical, material, and structural design of the consolidation work station and consolidated fuel canisters will be in accordance with the applicable portions of the "NRC OT Position of Review and Acceptance of Spent Fuel Storage and Handling Applications" and other applicable NRC guidance and industry codes. The canisters will be designed to Seismic Category I requirements, and the consolidation equipment will be analyzed and either restrained or anchored as appropriate to meet  !

Seismic Category II/I requirements as defined by NRC Regulatory l Guide 1.29, Revision 3. The consolidation work station and I consolidated fuel canister materials will be compatible with the spent l fuel rods and spent fuel assemblies, and the spent fuel pool water I chemistry. Therefore, margins of safety relative to mechanical, material, and structural aspects of the proposed fuel consolidation activities will not be significantly reduced.

l 4) Offsite and Control Room Doses l

j The offsite and control room dose consequences of accidents involving l consolidated fuel canisters or fuel consolidation activities were evaluated. To determine the radiological consequences, all credible l

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! accidents related to fuel consolidation activities were considered.

The analyses assume that spent fuel has decayed a minimum of 6 months prior to commencing the consolidation process.

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The limiting accident for fuel consolidation is a 74-inch drop of a consolidated fuel canister from the Spent Fuel Handling Machine (SFHM) onto a rack cell containing a consolidated fuel canister. Although both consolidated fuel canisters would remain intact, it is conservatively assumed that all 944 fuel rods within the two canisters (472 rods / canister x 2 canisters) are damaged. The resultant release of radioactivity, after escaping from the spent fuel pool, is j exhausted from the fuel handling building (FHB) over a two-hour l

period; no credit for FHB isolation system or FHB filters was taken.

The results demonstrate that, with a minimum decay time of 6 months and no credit taken for isolation or filtration, the radiological consequences of the worst case consolidated fuel accident would not result in releases that would exceed 25% of the 10 CFR 100 limits.

The results also demonstrate that the control room doses would meet l the 10 CFR 50, Appendix A, GDC 19 limits when crediting the control i room emergency air cleanup system.

Therefore, operation of the facility according to this proposed change will not involve a significant reduction in a margin of safety.

Safety and Significant Hazards Consideration Determination Based on the above Safety Analysis, it is concluded that: (1) The proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92 and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change. Moreover, because this action does not involve a significant hazards consideration, it will also not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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l t References 1 " Spent Fuel Pool Reracking Licensing Report," Revisions 5 and 6, submitted by Edison's letters to the NRC dated January 18, 1990, and February 16, 1990, respectively.

• Attachments A. Amendment Nos. 127/128 Approved Technical Specifications, Unit 2 B. Amendment Nos. 116/117 Approved Technical Specifications, Unit 3 C. Amendment Nos. 127/128 Approved Technical Specifications as Revised by l PCN-443, Unit 2 l D. Amendment Nos. 116/117 Approved Technical Specifications as Revised by PCN-443, Unit 3 E. Amendment Nos. 127/128 Approved Technical Specifications as Revised by PCN-443 and PCN-449, Unit 2 l F. Amendment Nos. 116/117 Approved Technical Specifications as Revised by l

PCN-443 and PCN-449, Unit 3 G. Administrative Controls Pertaining to Fuel Consolidation H. Unit 2 Amendment Nos. 127/128 and Unit 3 Amendment Nos. 116/117 Approved Bases as Revised by PCN-443 I. Unit 2 Amendment Nos. 127/128 and Unit 3 Amendment Nos. 116/117 Approved Bases as Revised by PCN-443 and PCN-449 l

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