Text

Safety Evaluation for STP Units 1 & 2 Spent Fuel Storage Pool Rack Criticality Analysis W/Credit for Soluble Boron
	ML20236L617
Person / Time
Site:	South Texas
Issue date:	07/07/1998
From:	; HOUSTON LIGHTING & POWER CO.
To:	;
Shared Package
ML20236L615	List: ML20236L617; ML20236L619; ML20236L626; ML20236L631; NOC-AE-000178, Application for Amends to Licenses NPF-76 & NPF-80, Reflecting Rev of Criticality Analyses & Rack Utilization Schemes for Regions 1 & 2 of Spent Fuel Racks as Described in STP UFSAR & TSs;
References
	NUDOCS 9807130094
	Download: ML20236L617 (64)
	v • d • e

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l ATTACHMENT 1 SAFETY EVALUATION for SO'UTH TEXAS PROJECT UNITS 1 AND 2 SPENT FUEL STORAGE POOL RACK CRITICALITY ANALYSIS WITH CREDIT FOR SOLUBLE BORON l

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NOC-AE-00178 O O 00 98 P PDR I

Table of Contents l

1. S UMM ARY .... ........ ....................................................................................5

2. PURPOSE........................................................................ ...............6

3. DESCRIPTION OF CHANGE............ .... .. .. . .............................................................6

4. S AFETY EVALUATION ............ ...... ........ ....... ........ . ..... . .... . ....... ......... . ... . .. ... .... . .. . .... I 1 4.1 Criticality Analysis . . .. . ..... .......... . .. .. . ............ . .. . .... . . ... ....... .. .. . .. .. ... . .... . . ...... . ... .. .. . ... I 1 4.1.1 General Criticality Analysis Methodology...... ........... . . . . . . . . . . . . . . . . . . . . . . . . .. . . ... . . 12 4.1.1.1 No Soluble Boron 95/95 La Calculation Methods .. .. ................... ....... ........ .......13 4.1.1.2 Soluble Boron Credit Kg Calculation Methods ............ ... .... ..... . ...... ....... . . . ...16 4.1.1.3 Burnup Credit Reactivity Equivalencing Methods....... ........ ...... ..................17 4.1.1.4 IFB A Credit Reactivity Equivalencing Methods ... ..... ......... ....... .......... .... ... ..18 4.1.1.5 Decay Tirne Reactivity Equivalencing Methods.......... . . . ................. .. . .. ... .... 20 4.1.1.6 Reactivity Equivalencing Application...... . .... ...... ...... ............. .. ... .. ............ .. . . 21 4.1.2 Region 1 Rack Design. .... ....................................................................21 4.1.2.1 Criticality Analysis of Region 1 All Cell Storage...... .. . .... .... . ............. ...... .. . ... 21 4.1.2.1.1 No Soluble Boron 95/95 Na Calculation . .. ............. .. .............. .... ....... . .. .. 22 4.1.2.1.2 Soluble Boron Credit La Calculations . ..... ........ ........ ............... ... .. ............ . 22 4.1.2.1.3 Burnup Credit Reactivity Equivalencing.................... ... .... ..... .......... .. .. ...... 23 4.1.2.1.4 IFB A Credit Reactivity Equivalencing... ........... ....... . ........ ............. ..... ....... 23 4.1.2.1.5 Reactivity Equivalencing Application............................. . . ... ..... . .. .......... .... 23 4.1.2.2 Criticality Analysis of Region 1 Checkerboard #1 Storage....... ........................24 4.1.2.2.1 No Soluble Boron 95/95 Na Calculation .... . ......... ................ . ... . ............... 24 4.1.2.2.2 Soluble Boron Credit La Calculations ............. ................... ... ...... . . ...... .. . 25 4.1.2.2.3 Burnup Credit Reactivity Equivalencing........ . ... .. ...... . .. .. ..... ... ......... .. . 25 4.1.2.2.4 IFB A Cmdit Reactivity Equivalencing.......... ... ...... ... ....... . ... .......... . . ... ... 26 1

4.1.2.2.5 Reactivity Equivalencing Application... ........ . ............... .. ...... .. . . . . . . 26 4.1.2.3 Criticality Analysis of Region 1 Checkerboard #2 Storage...... .... . ........... .. ........ 27 1

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r 4.1.2.3.1 No Soluble Boron 95/95 Kg Calculation ... . ........... .... . .. ..... . ............... .. 27 4.1.2.3.2 Soluble Boron Credit La Calculations ........ . ...... ............. .. ....... .. .... . ......... . 27 4.1.2.3.3 Burnup Credit Reactivity Equivalencing......... ................ . ......... ..... ... . ....... 28 4.1.2.3.4 IFB A Credit Reactivity Equivalencing...................... . .. .. ............ ............ ...... 28 4.1.2.3.5 Reactivity Equivalencing Application...... .. ...............................................29 4.1.3 Region 2 Rack Design .. ....... .. ......... . .. . . .. . . .. . .. ...... . ..... . ................ ... . . . .. ...... . 29 4.1.3.1 Criticality Analysis of Region 2 All Cell Storage........ ... ................................ ...... 29 4.1.3.1.1 No Soluble Boron 95/95 La Calculation ......... . . ..... .. ......... ........ . .. .......... 30 4.1.3.1.2 Soluble Boron Credit Ng Calculations .. .... . ..................... . .. ..... .................... 30 4.1.3.1.3 Higher Enrichment in Peripheral Cells.. ...... . ............ ....................31 4.1.3.1.4 Burnup and Decay Time Reactivity Equivalencing....... ... ... . ......... .......... .. 31 4.1.3.2 Criticality Analysis of Region 2 3-out-of-4 Storage .......................... .......... ..... . .. 32 l 4.1.3.2.1 No Soluble Bomn 95/95 Kn Calculation ... ........ ... ......... .... .. . ...... ............. 32 l 4.1.3.2.2 Soluble Boron Credit Ng Calculations .... .. .......... . .............. ................ ........ 32 4.1.3.2.3 Burnup Credit Reactivity Equivalcueing...... . ............ ........... ........ . .......... .... 33

4.1.3.3 Criticality Analysis of Region 2 2-out-of-4 Storage ...................... ... .............. ........ 33 l

4.1.3.3.1 No Soluble Boron 95/95 La Calculation .... ............... ........................34 4.1.3.3.2 Soluble Boron Credit Ng Calculations .......... .. ...... .......... .......... ............ .. ... 34 4.1.3.3.3 Burnup Credit Reactivity Equivalencing......... .. ............... .............. ...... ... . .. 35 4.1.3.4 Criticality. Analysis of Region 2 RCCA #1 Checkerboard Storage.... ...................... 35 4.1.3.4.1 No Soluble Boron 95/95 Kg Calculation .. ........... ......... ..... .. .. . . ..... ......... . 35 4.1.3.4.2 Soluble Boron Credit Kg Calculations .. ............. ......... . ... . . . ..... ....... ... 36 4.1.3.4.3 Burnup Credit Reactivity Equivalencing.. .. ... ...... ... ....... . ... ......... .... .. . .. . 37 4.1.3.5 Criticality Analysis of Region 2 RCCA #2 Checkerboard Storage...... . ........ ....... . 37 4.1.3.5.1 No Soluble Boron 95/95 La Calculation .. ............................ .. .. . ....... ..... 38 4.1.3.5.2 Soluble Boron Credit La Calculations.... . .......... .............. . .... .... . .... ........ 38 l 4.1.3.5.3 Burnup Credit Reactivity Equivalencing... . .... .. . ..... ....... ...................39 2

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4.1.4 Fuel Rod Storage Canister Criticality . . .... .. . . . ... .. ...... .. . .. ...... 39 4.1.5 Storage Configuration Interface Requirements... .. . . . . . . . . . . . . . . . . . . . 40 4.1.5.1 Interface Requirements within Region 1. ... .. ...... . . . .. . . . . . . .. . . . .. . 40 4.1.5.2 Interface Requirement 3 within Region 2... .. . .. . . . .. . . . . . . . . . . 41 4.1.5.3 Interface Requirements within Region I without Water Box Insen . . . . ... 41 4.1.5.4 Interface Requirements within Region 1 and Region 2.. . . . . . . . . . . . . . ... ... 42 4.1.5.5 Interface Requirements between Region 1 and Region 2..... . .. . . . . . . . . .. . . 4 2 )

4.2 Spent Fuel Pool Dilution Analysis.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . 43 4.2.1 Spent Fuel Pool And Related System Features.. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . 43 4.2.1.1 Spent Fuel Pool. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . 43 4.2.1.2 Spent Fuel Pool Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . 44 4.2.1.3 Spent Fuel Pool Cooling Subsystem. .. . ... . ...... ..... . . . . . . . . . . . . . . . . . . . . . 44 4.2.1.4 Spent Fuel Pool Cleanup Subsystem.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . . . 44 4.2.1.5 Dilution Sources ... . ...... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .45 4.2.1.5.1 Boron Recycle System (BRS).... . . . . . . . . . . . . . .. . . . . . . . . . . . . ... 45 4.2.1.5.2 Reactor Makeup Water System. .... . . . ........ ... ..... ... . . . . . . . . . .. .. 46 4 4.2.1.5.3 Demineralized Water System.. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .46 4.2.1.5.4 Component Cooling Water System... .. .. .. . . . . . . . . . . . . . . . . . .... . 47 4.2.1.5.5 Drain Systems ... .... .... . ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... . 47 4.2.1.5.6 Fire Protection System ......... ..... . . .. .... ... . . . . ..... ............. . ... ..... .. ... 47 4.2.1.5.7 Spent Fuel Pool Demineralizers ... .. .. .. .... . . . . . . . . . . . . . ... . . . 48 4.2.1.5.8 Piping ... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . 4 8 4.2.1.5.9 Dilution Source and Flow Rate Summary... ... .. . . . . .... . . . .. ... .. . . 48 4.2.1.6 Boratior. Sources..... .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . 49 l

4.2.1.6.1 Refueling Water Storage Tank .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 4.2.1.6.2 Direct Addition of Boric Acid. . . ... .. . . . . . . . . . . . . . .. . . 49 l 4.2.1.7 Administrative Controls . ... . . . .. . . ... . .. . . . . . . . . . . . 50 3

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4.2.1.8 loss of Offsite Power Impact . .. . .. . ..... ........... ..... ..... .... .... ........... . .......... ..... . 50 4.2.2 Spent Fuel Pool Dilution Evaluation... ..... ..... . . . .... ... ............ ...... . ..... ..... ..... . 51 4.2.2.1 Calculation of Boron Dilution Times and Volumes ... .... ............... .. .... ... . ... .. 51 4.2.2.2 Evaluation of Boron Dilution Events . . .. .... ... ....... ......... ........ . ...... . . .... ... .52 4.2.2.2.1 Dilution From BRS Recycle Holdup Tanks... .......... .. . ........ .... . ......... ....... 52 4.2.2.2.2 Dilution From Reactor Makeup Water Storage Tank..... .... .. .. . . ......... ....... 52 4.2.2.2.3 Dilution From Demineralized Water System... ..... ..................... . . . 53 4.2.2.2.4 Dilution from Fire Protection System .. .... ... ......... .. ........ ...... ..... .........53 4.2.2.2.5 Dilution Resulting From Seismic Events or Random Pipe Breaks ........ . ..... ... 54 4.2.2.2.6 Dilution From Spent Fuel Pool Demineralized .... .. .............. . ... ........ .. .. .... 55 4.2.2.3 S ummary of Dilution Events ..... ............... . ...................... .... . . .................. .. ...... 55 4.2.3 Boron Dilution Conclusions ....... ........... ............. ... ......... ....... ........ .................. ... ..... 56 4.3 Accident Analysis ........ ... ........................... ... . . . . . . . . . . . . . . . . . . . . . ...................................57 4.3.1 Postulated Conditions with No increase in Rack Ln................. ....... ........... . ....... .... 57 4.3.1.1 Dropping a Fuel Assembly on Top of a Rack............... . . ... .. . ............... ..... ... ... . 57 4.3.1.2 Dropping a Fuel Assembly Between Rack Modules ........ ... .. ..... ...... ....... ... . ... . 57 4.3.1.3 Dmpping a Fuel Assembly Adjacent to a Rack Module .. ........ ....... ... ..... .. .. . . 58 4.3.2 Postulated Conditions with An Increase in Rack Ken................ .. ...... ...... .. ... ........... 58 i 4.3.2.1 Spent Fuel Pool Water Temperature Accident.. ........... .. . .... .......................... .... .. 58 4.3.2.2 Dropping of a Fuel Assembly into an Already Imaded Cell Accident... .. ... . . ...... 59 4.3.2.3 Misloaded Assembly Accident........... .... ......... ....... ............ .. ... . ...................59 4.3.2.4 Withdrawal of an RCCA from a Region 2 RCCA #1 or #2 Checkerboard..... . ..... 59 4.3.3 Boron Dilution Accidents........ . ....... .. ..... . .. ........... ... . . . . ...... . .. ......... .......... . . 60

5. CO N C L U S IO NS . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. .. . . . ... . . . ... . .. . . . . . .

6. RE FE RENCES . . . . . .. .. . . . . .. . .... . . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . .. .. . . . . . . . . . ... . . . . . . . . . . . . . . . . . ... . ... .. . 61 i l

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4 Attachnent I safety Evaluation NOC-AE-00178 i

1.

SUMMARY

The purpose of this license change is to reflect the revision of the criticality analyses and the rack utilization schemes for Regions 1 and 2 of the spent fuel racks described in the South Texas Project Updated Final Safety Analysis Repon (UFSAR) and the Technical Specifications. ,

i The South Texas Project's spent fuel pool contains two rack types. The Region I racks were '

designed to use Boraflex panels in a removable stainless steel box to absorb neutrons. The Region 2 racks are fabricated by trapping Boraflex panels between the cell walls. However, due to the observed degradation of the Boraflex panels, the pool criticality analysis was reperformed assuming the absence of Boraflex and taking panial credit for the presence of soluble boron in the pool water.

This submittal presents the results of the reperformed criticality analysis of the South Texas Regions I and 2 spent fuel storage racks with partial credit for spent fuel pool soluble boron (from Reference 1). The methodology employed to produce these results is contained in Reference 3.

The Regions 1 and 2 spent fuel racks have been reanalyzed to allow storage of Westinghouse 2

17x17XL fuel assemblies with nominal (design) enrichments up to 4.95 w/o U " in the storage cell locations using credit for checkerboard configurations, burnup credit, and Integral Fuel Burnable Absorber (IFBA) credit. The nominal fuel enrichment for the agion is the enrichment of the fuel ordered from the manufacturer. This analysis does not take any credit for the presence of the spent fuel rack Boraflex poison panels.

The Regions 1 and 2 spent fuel rack analysis is based on maintaining Kerr < l.0 including uncertainties and tolerances on a 95 percent probability at a 95 percent confidence level, including uncertainties (95/95) without the presence of any soluble baron in the storage pool (No Soluble Boron 95/95 Q condition). Soluble boron credit is used to provide safety margin by maintaining Ken 5 0.95 including uncenainties, tolerances, and accident conditions in the presence of spent fuel pool soluble boron.

Because the South Texas Project has concerns with the spent fuel pool silica content levels L resulting from the Boraflex degradation,it may become desirable to amove the Boraflex, and its l stainless steel water box, from the Region I racks. The analysis has been performed for the Region l 1 racks with and without the steel water box insen, and the limiting results are reponed.

! The effects of the proposed changes do not pose a significant increase in hazards.

5 Attachnent i safety Evaluation NOC-AE-00178

2. PURPOSE The purpose of this licensing change is to reflect the revision of the criticality analyses and the rack utilization schemes for Regions 1 and 2 of the spent fuel racks in the South Texas Project Updated Final Safety Analysis Report (UFSAR) and the Technical Specifications. The mechanical, seismic, and cooling propenies of the racks are not affected by this change.

The lower enrichment limit for the close-packed (or "all cell storage") use of the Region 1 racks, resulting from the revised criticality analysis, hinders efficient use of the racks and precludes the use of higher fuel enrichments. Therefore, the proposed license change also addresses alternative rack utilization schemes.

This submittal proposes a revised licensing basis for the spent fuel racks and includes changes to peninent sections of the Updated Final Safety Analysis Repon and Technical Specifications to reflect the revised analyses and a revised rack utilization plan.

3. DESCRIPTION OF CHANGE This submittal proposes to modify the South Texas Project Updated Final Safety Analysis Repon (UFSAR) and the Technical Specifications to reflect the revision of the spent fuel pool criticality analyses and the alternative rack utilization schemes for the spent fuel racks.

The proposed change modifies the UFSAR as delineated below and as shown on the marked-up UFSAR sections in Attachment 5. Proposed changes to the Technical Specifications are listed below. The proposed Technical Specification changes are presented in Attachment 6.

The following storage configurations and enrichment limits are considered in this proposed change:

Region 1 Enrichment Limits All Cell Storage Storage of 17x 17XL fuel assemblies in all cell locations-Fuel assemblies must have an initial nominal enrichment no greater than 23 2.50 w/o U s or satisfy a minimum bumup requirement for higher initial enrichments. Fuel assemblies can also contain a minimum number ofIFBAs. The soluble boron concentration that results in a Ken of 5 0.95 was calculated as 500 ppm. Including accidents, the l

soluble boron credit required for this storage configuration is 700 ppm.

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l Attachnent i safety Evaluation NOC-AE-00178

Checkerboard #1 Storage Storage of 17x17XL fuel assemblies si a 2X2 checkerboard arrangement-The checkerboard pattem contains fuel assemblies in two diagonally adjacent cells with an initial, nominal enrichment no g cater 235 than 1.70 w/o U and fuel assemblies in the two remaining cells with 235 a nominal enrichment no greater than 3.55 w/o U . Fuel assemblies with enrichments greater than these values must satisfy a minimum bumup requirement or contain a minimum number ofIFBAs. The soluble boron concentration that results in a Nrr of 5 0.95 was calculated as 300 ppm. Including accidents, the soluble boron credit requimd for this storage configuration is 550 ppm.

Checkerboard #2 Storage Storage of 17x17XL fuel assemblies in a 2X2 checkerboard arrangement-The checkerboard pattem contains fuel assemblies in two diagonally adjacent cells with an initial, nominal enriclunents no greater than 1.40 and 1.70 w/o U 235 respectively, a fuel assembly in one remaining 23 cell with an initial, nominal enrichment no gmater than 2.50 w/o U , and a fuel assembly in the other cell with an initial, nominal enrichment no greater than 4.95 w/o U235. Fuel assemblies with initial enrichments greater than these values must satisfy a minimum bumup mquirement or contain a minimum number of IFBAs. The soluble boron concentration that results in a Ng of s 0.95 was calculated as 400 ppm. Including accidents, the soluble boron credit required for this storage configuration is 700 ppm.

Region 2 Enrichment Limits All Cell Storg;e Storage of 17x17XL fuel assemblies in all cell locations-Fuel assemblies must have an initial nominal enrichment no greater than 235 1.20 w/o U or satisfy a minimum bumup requirement for higher initial enrichments. The soluble boron concentration that results in a Ng of 5 0.95 was calculated as 700 ppm. Including accidents, the soluble boron credit required for this storage configuration is 1700 ppm.

Periphery Locations For fuel assemblies on the periphery of the Region 2 rack modules, storage of 17x17XL fuel assemblies with an initial nominal enrichment no greater than 1.40 w/o U 235 or that satisfy a minimum bumup requirement for higher initial enrichments is permitted.

l Attachment I sarety Evaluation NOC-AE-00178 '

3-out-of-4 Checkerboard Storage of 17x17XL fuel assemblies in a 3-out-of-4 checkerboard Storage arrangement with empty cells-Fuel assemblies must have an initial nominal enrichment no greater than 1.70 w/o U 235 or satisfy a minimum burnup requirement for higher initial enrichments. A 3-out-of-4 checkerboard with empty cells means that no more than 3 fuel assemblies can occupy any 2x2 matrix of storage cells. The soluble boron concentration that results in a La of 5 0.95 was calculated as 550 ppm. Including acciden's, the soluble boron cmdit required for this storage configuration is 2050 ppm.

2-out-of-4 Checkerboard Storage of 17x17XL fuel assemblies in a 2-out-of-4 checkerboard Storage arrangement with empty cells. Fuel assemblies must have an initial nominal enrichment no greater than 4.85 w/o235 U . A 2-out-of-4 checkerboard with empty cells means that no 2 fuel assemblies may be stored face adjacent. Fuel assemblies may be stored comer adjacent.

The soluble boron concentration that results in a Laof s 0.95 was calculated as 300 ppm. Including accidents, the soluble boron credit requimd for this storage configuration is 2100 ppm.

RCCA #1 Checkerboard Storage of 17x17XL fuel assemblies in a 2X2 checkerboard where 1 of the 4 assemblies contains a silver-indium-cad.mium or hafnium (Ag-In-Cd or Hf) Rod Cluster Control Assembly (RCCA). Fuel assemblies must have an initial nominal enrichment no greater than 1.40 w/o U 235 or satisfy a minimum bumup requirement for higher initial enrichments. The soluble boron concentration that results in a La of 5 0.95 was calculated as 650 ppm. Including accidents, the soluble boron credit required for this storage configuration is 1950 ppm.

RCCA #2 Checkerboard Storage of 17x17XL fuel assemblies in a 2X2 checkerboard where 2 diagonally adjacent of the 4 assemblies contain a Ag-In-Cd or Hf RCCA. Fuel assemblies must have an initial nominal enrichment no greater than 1.65 w/o U235 or satisfy a minimum burnup requirement for higher initial enrichments. The soluble boron concentration that results in a La of s 0.95 was calculated as 700 ppm. Including accidents, the soluble boron credit required for this storage configuration is 2200 ppm.

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l To help decide among allowable storage locations for each fuel assembly, each fuel assembly will i be categorized by reactivity prior to insenion into the spent fuel storage racks. Fuel categories are defined by Technical Specification 5.6.1.2.

8 Attachment i safety Evaluanon NOC-AE-00178

When the two storage areas meet at an interface, the type ofinterface is explicitly defined to ensure the rack Na limit is met. The interface between a close packed fuel storage area and a checkerboard storage area in Region 1 is described, as is the same interface in Region 2 racks. The interface requirements for different storage area types at Region 1 - Region 2 rack module boundaries are also described. Technical Specification Section 5.6.1.5 describes the storage configuration interface requirements.

In all cases, administrative controls will be used to guide fuel assembly placement. To prevent a violation of the Technical Specification limit of the rack Nu in the event of the misloading of a fuel assembly. Technical Specification 3.9.13 is revised to provide for a minimum soluble boron concentration of 2500 ppm for the spent fuel pool. This minimum soluble boron concentration is 300 ppm above the 2200 ppm value mquired to satisfy requirements including tolerances, uncenainties, and accident conditions evaluated in Reference 1. This 300 ppm margin more than er:eeds the small additional uncenainties for the combination of soluble boron measurement and th P. use of linear interpolation with the affected Figures in the Technical Specifications.

The presence of this amount of boron is adequate to ensure that for all single misloadings of a fuel assembly, the Kn is maintained at less than or equal to 0.95.

To accommodate the above changes, the following changes to the Technical Specifications are proposed: ,

a. Sections 3.9.13 and 4.9.13 are revised to provide for a 2500 ppm minimum soluble boron j concentration for the spent fuel pool; )

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b. Section 5.6.1.1 is revised to reflect the requirement for Lnwith borated and unborated water; and the reference to Boraflex is deleted; )

c. Section 5.6.1.2 is revised to update the fuel reactivity categories;

d. Section 5.6.1.3 is revised to update the Region I rack loading;

e. Section 5.6.1.4 is revised to update the Region 2 rack loading;

f. Section 5.6.1.5 is revised to reflect new fuel storage interface requirements;

g. Section 5.6.1.6 is revised to discuss misloaded assemblies;

h. Section 5.6.1.8 is revised to reflect editorial changes, new figure numbers, and a new fuel i categorization; I i. Figure 5.6-1 is replaced by the minimum bumup curve for Category 2 fuel; 9

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j. Figure 5.6-2 is revised to provide the minimum burnup curve for Category 3 fuel;

k. Figure 5.6-3 is revised to provide the minimum IFBA credit curve for Category 3 fuel;

1. Figure 5.6-4 is revised to provide the minimum burnup curve for Category 4 fuel;

m. Figure 5.6-5 is revised to provide the minimum IFBA credit curve for Category 4 fuel;

n. Figure 5.6-6 is revised to provide the minimum bumup curve for Category 5 fuel;

o. Figure 5.6-7 is revised to provide the minimum bumup curve for Category 6 fuel;

p. Figures 5.6-8 through 5.6-12 are added to provide the minimum bumup curves for Category 7 through 11 fuel;

q. Figure 5.6-13 provides the requirements for Region I rack loadings;

r. Figure 5.6-14 provides the requiremerits for Region 2 rack loadings;

s. Figures 5.6-15 through 5.6-19 provide the Region Boundary Interface requirements;

t. Figure 5.6-20 provides the IFBA mquirements for In-Containment Fuel Storage Rack;

u. The BASES for Section 3/4.9.13 is revised;

v. The INDEX to the Technical Specifications is evised to reflect the above changes.

Accordingly, the following changes to the Updated Final Safety Analysis Report (UFSAR) are proposed:

a. Section 3.1.2.6.3.1 is revised to reflect the criteria for minimum Kctr;

b. The "Tah!c of Contents" for Chapter 4 is revised to reflect revision of Section 4.3.2.6.2;

c. Section 4.3.2.6.2.1 describing the criticality analysis methodology is added;

d. Section 4.3.2.6.2.2 describing the Region I rack design and the results ofits criticality analysis is revised;

e. Section 4.3.2.6.2.3 describing the Region 2 rack design and the results ofits criticality analysis is revised; j

i l f. Section 4.3.2.6.2.4 describing the storage configuration interface requirements is revised; 10 Attachnent I safety Evaluaton NOC-AE-00178 l

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g. Section 4.3.2.6.2.5 describing the rack utilization for both regions is revised;

h. Section 4.3.2.6.3 is revised with respect to references used;

i. The " References" section for Section 4.3 is updated;

j. Section 9.1.2.3 is revised to reGect the criteria for minimum La.

4. SAFETY EVALUATION The purpose of this section is to discuss the impact of the proposed change on the design and licensing basis of the plant.

First, the criticality analysis will be presented in Section 4.1, followed by a discussion of the boron dilution analysis (Section 4.2). A discussion of postulated accidents is presented in Section 4.3.

4.1 Criticality Analysis The reactivity of the spent fuel rack is analyzed in Reference 1. To provide safety margin in the criticality analysis of the spent fuel racks, credit is taken for the soluble boron present in the spent fuel pool. This parameter provides signiGcant negative reactivity in the criticality analysis of the spent fuel racks and will be used in conjunction with administrative controls to offset the reactivity increase resulting from removal of the spent fuel rack BoraGex poison panels. Soluble boron eredit provides sufficient mlaxation in the enrichment limits of the spent fuel racks.

The design basis for preventing criticality in the spent fuel pool is:

1. the effective neutron multiplication factor, Krr, of the fuel rack array will be less than 1.00 in pure, unborated water, with a 95 percent probability at a 95 percent confidence level, including c:nainties; and,

2. the effective neutron multiplication factor, Ng, of the fuel rack array will be 5 0.95 in the pool containing borated water, with a 95 peicent probability at a 95 percent confidence level, including uncertainties.

The purpose of this section is to pmsent the storage requirements, including maximum nominal enrichments, minimum burnup values, minimum decay times, minimum IFBA content, storage configurations, and the minimum pool soluble boron concentration.

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l With the simplifying assumptions employed in thi!, analysis (no grids, sleeves, axial blankets, etc.),

the various types of 17x17XL fuel do not contribute to any increase in the basic assembly reactivity. This includes small changes in guide tube and instrumentation tube dimensions.

Therefore, futun fuel assembly upgrades do not require a criticality analysis if the fuel rod outside diameter continues to be 0.374 inches and the rod pitch is 0.496 inches.

The fuel rod, guiac tube, and instrumentation tube claddings are modeled with zircaloy in this l l analysis. This is conservative with respect to the Westinghouse ZIRLO* product which is a l zirconium alloy containing additional elements including niobium. Niobium has a small absorption cross-section which causes more neutron capture in the cladding regions resulting in a lower reactivig. Therefore, this analysis is conservative with respect to fuel assemblies containing I ZIRLO cladding in fuel rods, guide tubes, and the instrumentation tube.

Empty water cells may be substituted for fresh or bumed fuel assemblies at any location. Any l positive reactivity effect due to the additional water in the 2x2 analytical cell is offset by the i

absence of uranium.

When storing fuel with an initial nominal enrichment greater than the maximum all-cell enrichment l for the respective rack region, a rack Ken less than or equal to 0.95 is ensured by the maintenance of a minimum amount of soluble boron and the use of administrative procedures to control the placement of burned and fmsh fuel and RCCAs. A rack Ken less than 1.00 in pure, unborated water is ensured by the use of administrative procedums to control the placement of bumed and fresh fuel l and RCCAs. Guidance on the close packed storage of fresh and bumed fuel, and fuel containing IFBAs, is provided in Section 5.6 of the Technical Specifications.

l l The licensing basis for the racks is met by the combination of the physical design and center-to-l center spacing of the storage cells, the required presence of soluble boron, and the use of l administrative procedures to guide the placement of fuel assemblies and RCCAs.

l I

4.1.1 General Criticality Analysis Methodology l

The design method which insures the criticality safety of fuel assemblies in the fuel storage rack is described in detail in the Westinghouse Spent Fuel Rack Criticality Analysis Methodology topical report (Reference 3). This report describes the computer codes, benchmarking, and methodology used to calculate the criticality safety limits presented in this mport for Regions 1 and 2.

The effects of various mechanical and thermal variances are conservatively treated by applying the most limiting parameter variances together at one time in a KENO model. These include pool water temperature, stainless steel thickness variation, fuel enrichment variation, and uranium dioxide pellet density. The Ken for all storage pattems is less than 1.00 including biases and uncemdnties in unborated water.

12 Attachnent I safety Evaluanan NOC-AE-(X)l78 t

f b____________________________

The criticality calculation method and cross-section values am verified by comparison with critical experiment data for fuel assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncenainty will apply to rack conditions which include strong neutron absorbers, large water gaps, low moderator densities and spent fuel pool soluble boron.

To acconunodate higher fuel enrichments in close packed storage in both rack regions, reactivity credit for assembly burnup and decay and for the presence of IFBAs or RCCAs in fuel assemblies is allowed. These methodologies are discussed in the following sections.

4.1.1.1 No Soluble Boron 95/95 Kn Calculation Methods To determine the enrichment required to maintain Na < l.0, KENO-Va is used to establish a l nominal mference reactivity and PHOENIX-P is used to assess the temperature bias of a nonnal pool temperature range and the effects of material and construction tolerance variations. A final 95/95 La is developed by statistically combining the individual tolerance impacts with the calculational and methodology uncertainties and summing this term with the temperature and method biases and the nominal KENO-Va reference reactivity. The equation for determining the final 95/95 Lais defined in Reference 1 and shown below.

i Ng = Knw + Bu.mp + Bmeina + Bseg + Bunen where:

Knonnoi = nonnal condition KENO-Va Ly Bu.mp = temperature bias for normal operating range Bn m = method bias determined from benchmark critical comparisons Bug = B' self-shielding bias Boneen = statistical summation of uncenainty components for n tolerances / uncertainties The following assumptions are used to develop the Nc Soluble Boron 95/95 Kg KENO-Va model for storage of fuel assemblies in all cells of the Region 1 spent fuel storage rack:

1. The fuel assembly parameters relevant to the criticality analysis are based on the l Westinghouse 17x17XL fuel design.

2. Fuel assemblies contain uranium dioxide at a suitable nominal enrichment over the entire l length of each rod.

i l

13 Atughment I safety Evaluatmn NOC-AE-00178

l

3. The fuel pellets are modeled assuming nominal values for theoretical density (95%) and dishing fraction.

4. No credit is taken for any natural or reduced enrichment axial blankets. This assumption I results in either equivalent or conservative calculations of reactivity for all fuel assemblies used at the South Texas Project, including those with annular pellets at the fuel rod ends.

234

5. No credit is taken for any U or U 236 in the fuel, nor is any credit taken for the buildup of fission product poison material. !

6. No credit is taken for any spacer grids or spacer sleeves. 4

7. No credit is taken for any bumable absorber in the fuel rods.

8. No credit is taken for the pmsence of spent fuel rack Boraflex poison panels. The i Boraflex volume is replaced with water.

9. The moderator is water with 0 ppm soluble boron at a temperature of 68 F. A water 3

density of 1.0 gm/cm is used.

10. The an ay is infinite in the lateral (x and y) extent and finite in the axial (vertical) extent.

I1. All available storage cells are loaded with symmetrically positione.d (centered within the storage cell) fuel assemblies. Figure 5 of Reference 1 shows the all cell configuration.

12. The water box inserts (upon which the Boraflex panels are mounted) are assumed to be removed from the Region I racks. The results from this case bound the results fmm a case with water boxes included; thus, either configuration is allowable.

With the above assumptions, the KENO-Va calculations of La under nominal conditions are then performed.

Temperature and methodology biases must be considered in the final Lg summation prior to comparing against the 1.0 La limit. The following biases were included:

Methodology: The benchmarking bias as determined for the Westinghouse KENO-Va methodology was considered.

1 Water Temperature: A reactivity bias determined in PHOENIX-P was applied to account for the effect of the normal range of spent fuel pool water temperatures (50 F to 160 F),

I l

14 Attachnent i Safety Evaluanon NOC-AE-00178 l

To evaluate the reactivity effects of possible variations in material characteristics and mechanical / construction dimensions, additional PHOENIX-P calculations were performed. For the all cell storage configuration, UO 2material tolerances were considered along with construction tolerances related to the cell inside diameter (celi ID), storage cell pitch, and stainless steel wall thickness. Uncenainties associated with calculation and methodology accuracy were also considered in the statistical summation of uncertainty components. To evaluate the reactivity effect of asymmetric assembly positioning within the storage cells, KENO-Va calculations were performed.

The following tolerance and uncenainty components were considered in the total uncenainty statistical summation:

U 235 Enrichment: The standard enrichment tolerance of10.05 w/o U235 about the nominal reference enrichment was considemd.

UO 2Density: A 12.0% variation about the nominal reference theoretical density was considered.

Fuel Pellet Dishing: A variation in fuel pellet dishing fraction from 0.0% to twice the nominal dishing was considered.

Storage Cell ID: The tolerance about the nominal reference cell ID was considered.

Storage Cell Pitch: The tolerance about the nominal refennce cell pitch was considered.

Stainless Steel Wall Thickness: The tolerance about the nominal reference stainless steel cell wall and cell separation thickness was considered.

Asymmetric Assembly Position: Conservative calculations show that an increase in reactivity can occur if the comers of four fuel assemblies were positioned together. This reactivity increase is considered in the statistical summation of the spent fuel rack tolerances.

Calculation Uncertainty: The 95 percent probability /95 percent confidence level uncenainty on the KENO-Va nominal reference Kerrwas considered.

Methodology Uncenainty: The 95 percent probability /95 percent confidence uncertainty in the benchmarking bias as determined for the Westinghouse KENO-Va methodology was considered. I 15 Attachnent I safety Evaluata:m NOC-AE-00178 L

i l

The 95/95 Kn for the spent fuel rack all cell storage configuration is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va mference reactivity.

Since N uwill be < l.0, the spent fuel racks will remain suberitical when all cells are loaded with l the proper enrichment 17x17XL fuel assemblies and no soluble boron is present in the spent fuel l pool water. Soluble boron credit is then used to provide safety margin by determining the amount of soluble boron required to maintain Nu s 0.95 including tolerances and uncenainties.

4.1.1.2 Soluble Boron Credit Nn Calculation Methods i To determine the amount of soluble boron required to maintain Nn 5 0.95, KENO-Va is used to l establish a nominal reference reactivity and PHOENIX-P is used to assess the temperature bias of a normal pool temperature range and the effects of material and construction tolerance variations. A final 95/95 &n is developed by statistically combining the individual tolerance impacts with the calculational and methodology uncertainties and summing this term with the temperature and method biases and the nominal KENO-Va refemnce mactivity.

The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for all cell storage spent fuel racks are similar to those in Section 4.1.1.1 except for assumption #9 l

mgarding the moderator soluble boron concentration. The moderator is replaced with water l containing an appropriate amount of soluble boron.

Temperature and methodology biases must be considered in the final Nusummation prior to

comparing against the 0.95 Nu limit. The following biases are included:

Methodology: The benchmarking bias as determined for the Westinghouse KENO-Va i methodology was considered.

Water Temperature: A reactivity bias determined in PHOENIX-P was applied to account for the effect of the normal range of spent fuel pool water temperatures (50 F to 160 F).

To evaluate the reactivity effects of possible variations in material characteristics and mechanical / construction dimensions, additional PHOENIX-P calculations were performed. For the spent fuel rack all cell storage configuration, UO 2material tolerances were considered along with I construction tolerances related to the cell ID, storage cell pitch, and stainless steel wall thickness.

I Uncertainties associated with calculation and methodology accuracy were also considered in the j statistical summation of uncenainty components. To evaluate the reactivity effect of asymmetric j assembly positioning within the storage cells, KENO-Va calculations were performed.

16 Attachnent I safety Evaluatmn NOC-AE-00178 i

f I

The same tolerance and uncertainty components as in the No Soluble Boron case were considered in the total uncenainty statistical summation.

The 95/95 Lais developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncenainties to the nominal KENO-Va reference reactivity.

! Since L awill be s 0.95 including soluble boron credit and uncertainties at a 95/95 probability / confidence level, the acceptance criterion for criticality will be met for all cell storage of l 17x17XL fuel assemblies in the spent fuel racks. Storage of fuel assemblies with nominal enrichments no greater than a specified value will be shown to be acceptable in all cells including the presence of a specified amount of soluble boron.

l 4.1.1.3 Burnup Credit Reactivity Equivalencing Methods

)

Storage of fuel assemblies with initial enrichments higher than the all cell maximum enrichment in i the spent fuel racks is achievable by means of burnup credit using reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with fuel I

depletion. For bumup credit, a set of reactivity calculations is performed to generate a set of enrichment-fuel assembly discharge bumup ordered pairs which all yield an equivalent La when stored in the spent fuel storage racks. !

l l The depletable, two-dimensional, transpon theory coda PIIOENIX is used to perform the reactivity equivalence calculations. PHOENIX has been val dated by comparisons with both benchmark l critical experiments and experiments where the isotopic fuel composition has been examined i

following discharge from a reactor.

Figures from Refemnce I show the const'mt contour generated for all cell storage in the each region of the spent fuel racks. The figures reprosent combinations of fuel enrichment and discharge l bumup which yield a conservative rack multiplication factor (Lg) as compared to the rack loaded

! with the maximum enrichment for all cell storage of 17x17XL fuel assemblies at zero burnup in all cell locations.

l Uncertainties associated with bumup credit include a reactivity uncenainty of 0.01 delta La at 30,000 MWD /MTU applied linearly wiJi burnup to account for calculation and depletion l uncertainties and 5% on the calculated bumup to account for bumup measurement uncertainty. This l uncenainty is considered to be very conservative and is based on consideration of the good i agreement between PHOENIX predictions an:1 measurements and on conservative estimates of fuel assembly isotopic buildup variances. The amount of additional soluble boron needed to account for these uncenainties in the bumup requiremcat is determined. This is additional boron above the amount determined to be required in Section 4.1.1.2. This results in a total soluble boron mquirement of the am of the amounts determined in Section 4.1.1.2 and this section.

17 Attachnerr. ?

safety Ew6uauon NOCAE-00178

__ ____ _ ________ _ _ - ____ __N

It is important to recognize that the curves developed by the methodology in this section are based on calculations of constant rack reactivity. In this way, the environment of the storage rack and its influence on assembly reactivity are implicitly considered. For convenience, the data from these figu es are also provided in tabular format. Use oflinear interpolation between the tabulated values is acceptable since the curves are approximately linear between the tabulated points.

Previous evaluations have been performed to quantify axial burnup reactivity effects and to confirm that the reactivity equivalencing methodology results in calculations of conservative bumup credit limits. The effect of axial bumup distribution on assembly reactivity has thus been addressed in the development of burnup credit limits.

4.1.1.4 IFBA Credit Reactivity Equivalencing Methods Storage of fuel asamblies with nominal enrichments greater than those determined in Section 4.1.2.1 is achievable by means of IFB A credit using the concept of reactivity equivalencing. The concept of reactivity eqtJvalencing is predicated upon the reactivity decrease associated with the addition of IFBAs. IFBAs consist of neutron absorbing material applied as a thin ZrB2 coating on the outside of the UO2fuel p-llet. As a result, the neutron absorbing material is a non-removable or integral part of the fuel assemb!y once it is manufactured.

A series of reactivity calculations ic performed to generate a set of IFBA rod number versus enrichment ordered pairs which all ybld the equivalent Kerr when the fuel is stomd in any of the configurations analyzed for the spent feel racks. The following assumptions were used for the IFBA rod assemblies in the PHOENIX-P models:

1. The fuel assembly parameters relevant to the criticality analysis are based on the Westinghouse 17x17XL fueldesign.

2. The fuel assembly is modeled at its tr.ost reactive point in life.

3. The fuel pellets are modeled assuraing nominal values for theoretical density and dishing fraction.

4. No credit is taken for any natural enrichment or reduced enrichment axial blankets.

5. No credit is taken for any U2 n or U236 in the fuel, nor is any credit taken for the buildup of fission product poison r.taterial.

6. No credit is taken for an;/ spacer grids or spacer sleeves.

18 Attachnent I safety Evaluauan NOC-AE-00178

7. The IFBA absorber material is a zirconium diboride (ZrB 2) coating on the fuel pellet.

NominalIFBA pin B loadings of 1.57 milligrams B' per inch (1.0X) and 2.35 milligrams B' per inch (1.5X) are used in determining the IFBA requirement.

8. For reduced length IFBA, the IFBA B' loading is reduced by 28.5% to conservatively model a minimum poison length of 120 inches.

9. The moderator is pure water (no boron) at a temperature of 68 F. A water density of 1.0 gm/cm3 is used.

10. The array is infinite in lateral (x and y) and axial (vertical) extent. This precludes any neutron leakage from the array.

I 1. Standard Westinghouse IFBA patterns (including previous standard pattems) for 17x17XL fuel assemblies were considered.

The results of the IFBA credit reactivity equivalencing for the spent fuel racks are provided in both figure and tabular formats. Figures in Reference I show the constant Ken contour generated for those configurations analyzed.

Results are provided for both full length (168 inch) IFBA, and reduced length (120 inch) IFB A. For intermediate length IFB A, linear interpolation between the full length and reduced length IFB A requirements is conservative.

It is imponant to recognize that the results are based on reactivity equivalence calculations (i.e.

holding rack Ken constant) for the specific enrichment and IFBA combinations in actual rack geometry (and not just on simple comparisons of individual fuel assembly infinite multiplication factors). In this way, the environment of the storage rack and its influence on assembly reactivity are implicitly considered.

Additional IFBA credit calculations were performed to examine the reactivity effects of higher IFBA linear B' loadings (2.36 and 3.14 mg/in). These calculations confirm that the assembly reactivity remains constant provided the net B'" material per assembly is preserved. Therefore with higher IFBA B' loadings, the required number ofIFB A pins per assembly can be reduced by the ratio of the higher loading to the nominal 1.57 mg/in loading.

The IFBA requirements were developed based on the standard IFBA pattems used by Westinghouse. However, since the worth ofindividual IFBA pins can change depending on position within the assembly (due to local variations in thermal flux), studies were performed to evaluate the effect and a conservative reactivity margin was included in the development of the 19 Attachment i Safety Evaluation NOC-AE-00178 j i

)

l l

l IFBA requirement to account for this effect. This assures that the IFBA requirement remains valid at intermediate enrichments where standard IFBA pattems may not be available.

Uncertainties associated with IFBA credit include a 5% manufacturing tolerance and a 10%

calculational uncertainty on the B' loading of the IFBA rods. A determination was made of the amount of additional soluble boron needed to account for these uncertainties in the IFBA credit mquirement. This is additional boron above the amount required from Section 4.1.1.2.

4.1.1.5 Decay Time Reactivity Equivalencing Methods Decay Time Credit is an extension of the Bumup Credit process which includes the time an assembly has been discharged as a variable. This methodology gains additional margin in reactivity and reduces the minimum burnup requirements. Spent fuel decay time credit results from the radioactive decay ofisotopes in the spent fuel to daughter isotopes, which results in reduced reactivity. One of the major contributors is the decay of Pu 24 to Am 241 . In this report, credit is taken only for the decay of actinides. Decay of the fission products has the effect of furtbr reducing the reactivity of the spent fuel.

In the decay time methodology reponed here, the fission products are frozen at tue isotop5 concentrations existing at the time of discLarge of the fuel (except Xe which is removed). These i calculations are performed at different discharge burnups. The actinide isotopes are allowed to decay based on their natural processes. The loss in reactivity due to the radioactive decay of the spent fuel results in reducing the minimum burnup needed to meet the reactivity requirements.

Thus for different decay times, a family of curves is generated which yield the desired equivalent Kerrwhen stored in the spent fuel storage racks. In the decay time methodology, the following assumptions are usedin the models:

1. The fael assemblies are modeled using the same criteria as in Section 41.3.1.

2. Fuel is depleted using a conservatively high soluble boron letdown curve to enhance the !

buildup of plutonium making the fuel more reactive in the spent fuel storage racks.

Sensitivity studies have shown that spectmm effects are also conservative for the decay tirne calculation.

3. No credit is taken for fission product isotopic decay.

4. Credit is taken for the decay of actinide isotopes only.

5. Nominal spent fuel rack configuration / dimensions are used.

With the above assumptions, the calculation of the decay time burnup credit curves are found to be conservative for use in the spent fuel pool criticality analysis.

20 Attachment I safety Evaluation NOC-AE-00178

(

1 I

l

i 1

4.1.1.6 Reactivity Equivalencing Application In Section 3.3 of Reference 1, the boron requirement to compensate for tolerances and uncertainties for burnup credit was determined to be 200 ppm. In Section 3.4 of Reference 1, the boron requirement to compensate for tolerances and uncertainties for IFBA credit was determined to be l 300 ppm. These boron values for burnup and IFBA credit tolerances and uncertainties were l calculated independently. That is, if all assemblies being stored in the racks utilized burnup credit, the additional boron required for tolerances and uncertainties in the burnup credit calculation would i be 200 ppm. Similarly,if all assemblies being stored in the racks utilized IFBA credit, the additional boron requi ed for tolerances and uncenainties in the IFBA credit calculation would be i 300 ppm. When some assemblies bei :g stored in the all cell configuration utilize bumup credit and some utilize IFBA aedit, the boron required for uncenainties and tolerances is not the sum of these two amounts. Instead, the more limiting of the two amounts bounds tolerances and uncertainties for both bumup and IFBA credit. Thus, the total required boron in the all cell configuration is 550 ppm (the most limiting of Sections 3.3 and 3.4 of Reference 1) for the Region I spent fuel racks.

4.1.2 Region 1 Rack Design l

l The Region I racks have a 10.95 inch-nominal center-to-center spacing with locked removable i

poison (Boraflex) assemblies between the cells. The Boraflex panels in the Region I racks are mounted on a stainless steel water box insert (Reference 1 Figure 1). Due to operational concems with the miease of silica as the Boraflex degrades, STP desires the flexibility to remove the Boraflex from the Region I racks. This may necessitate the removal of the stainless steel water box insert. Therefore, the criticality analysis covers the Region I racks with and without the water box insen in place. Figure 2 of Reference I shows the configuration with the water box removed.

l The impact of removing the water box insert will cause the reactivity of ecch configuration for Region 1 to increase since removing the insen will take out stainless steel material which acts to absorb neutrons from the system. As a result, to offset the reactivity increase, the enrichments for l each storage configurations in Region 1 are reduced to meet the criticality design limits for the !

spent fuel racks. Since the configuration in which the water box insert is removed is the limiting

]

scenario, this section describes the criticality analysis of the Region I spent fuel racks with the !

water box insert removed.

4.1.2.1 Criticality Analysis of Region 1 All Cell Storage This section describes the analytical techniques and models employed to perform the criticahty l analysis and reactivity equivalencing evaluations for the storage of fuel in all cells of the Region 1 l l spent fuel storage racks.

l Section 4.1.2.1.1 describes the no soluble boron 95/95 Ken KENO-Va calculations. Section 4.1.2.1.2 discusses the results of the spent fuel rack 95/95 Key soluble boron credit calculations.

l 21 Attachnent 1 safety Evaluauon NOC-AE-00178

Section 4.1.2.1.3 presents the results of calculations performed to show the minimum burnup requirements for assemblies with initial enrichments above those determined in Section 4.1.2.1.1.

Section 4.1.2.1.4 presents the results of calculations performed to show the minimum IFBA requirements for assemblies with initial enrichments above those determined in Section 4.1.2.1.1.

Finally, Section 4.1.2.1.5 presents how the application of the reactivity equivalencing methodology is used with multiple reactivity credit techniques.

4.1.2.1.1 No Soluble Boron 95/95 Kg Calculation The assumptions of Section 4.1.1.1.1 are used to develop the No Soluble Boron 95/95 La KENO-Va model for storage of fuel assemblies in all cells of the Region I spent fuel storage racks. For this case, the fuel essemblies contain uranium dioxide at a nominal enrichment of 2.50 w/o U235 over the entire length of each rod. For the Region I racks, the more limiting criticality cases occur when the water box insen is assumed to be removed. Throughout this report, the reponed Ng results will be from the cases that assume no water box insen.

With the above assumptions, the KENO-Va calculations of Ng under nominal conditions resulted in a La of 0.97070.

The 95/95 Kg for the Region I spent fuel rack all cell storage configuration is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncenainties to the nominal KENO-Va reference reactivity. The summation is shown in Reference 1-Table 7, and resmts in a 95/95 Kg of 0.99660.

Since Ng is < l.0, the Region 1 spent fuel racks will remain suberitical when all cells are loaded with 2.50 w/o U 233 17x17XL fuel assemblies and no soluble boron is present in the spent fuel pool water. l 4.1.2.1.2 Soluble Boron Credit La Calculations The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for all cell storage in the Region i spent fuel racks are similar to those in Section 4.1.1.1 (the No Soluble Boron case) except that the moderator soluble boron concentration is set at 200 ppm. With the above assumptions, the KENO-Va calculation resulted in a La of 0.92117.

The 95/95 Lais developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO Va reference reactivity. The summation is shown in Reference 1-Table 7, and results in a 95/95 Lao f 0.94579.

l Since Ng is s 0.95 including soluble boron credit and uncenainties at a 95/95 probability / confidence level, the acceptance criterion for criticality is met for all cell storage of 17x17XL fuel assemblies in the Region 1 spent fuel racks. Storage of fuel assemblies with nominal 22 Attachment I safety Evaluation NOC-AE4X)l78

(

L--__-_----_-_ _ - - - - - _ - - - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- ---- J

enrichments no greater than 2.50 w/o U235 is acceptable in all cells including the presence of 200 ppm soluble boron.

4.1.2.1.3 Bumup Credit Reactivity Equivalencing Storage of fuel assemblies with initial enrichments higher than 2.50 w/o U235 in all cells of the Region 1 spent fuel racks is achievable by means of bumup credit using reactivity equivalencing.

Figure 14 of Reference I shows the constant contour generated for all cell storage in the Region I spent fuel racks. The curve of Figure 14 represents combinations of fuel enrichment and discharge burnup which yield a conservative rack multiplication factor (Kn) as compared to the rack loaded 235 with 2.50 w/o U 17x17XL fuel assemblies at zero bumup in all cell locations.

Uncertainties associated with bumup credit include a reactivity uncertainty of 0.01 delta Kn at 30,000 MWD /MTU applied linearly with bumup to account for calculation and depletion uncertainties and 5% on the calculated bumup to account for burnup measurement uncertainty. The amount of additional soluble boron needed to account for these uncenainties in the bumup 4 requirement of Reference 1 Figure 14 was 300 ppm. This is additional boron above the 200 ppm required in Section 4.1.2.1.2. This results in a total soluble boron requirement of 500 ppm.

4.1.2.1.4 IFBA Credit Reactivity Equivalencing Storage of fuel assemblies with nominal enrichments greater than those determined in Section

, 4.1.2.1 is achievable by means ofIFBA credit using the concept of reactivity equivalencing. The results of the IFBA credit reactivity equivalencing for the Region I spent fuel racks are provided in Table 11 of Reference 1. The results applicable to the all cell configuration is also illustrated in Figure 18 of Reference 1, which shows the constant La contour generated for those configurations.

Uncertainties associated with IFBA credit include a 5% manufacturing tolemnce and a 10%

calculational uncertainty on the B' loading of the IFBA rods. The amount of additional soluble boron needed to account for these uncenainties in the IFBA credit requirement of Reference 1 l Table 11 is 300 ppm for the all cell configuration. This is additional boron above the 200 ppm required in Section 4.1.2.1.2. The soluble boron needed for IFBA credit is the same as the 200 ppm required for bumup credit in the Region 1 spent fuel racks as determined in Section 4.1.2.1.3. l Themfore, the total soluble boron credit required for the Region I spent fuel racks all cell configuration is 500 ppm.

4.1.2.1.5 Reactivity Equivalencing Application In Section 6.1.3 of Reference 1, the boron requirement to compensate for tolerances and l

uncenainties for burnup credit was determined to be 300 ppm. In Section 6.1.4 of Reference 1, the l boron requirement to compensate for tolerances and uncenainties for IFBA credit was determined to be 300 ppm. These boron values for burnup and IFBA credit tolerances and uncertainties were 23 Attachnent I safety Evaluauan '

NOC-AE-00178 l

L____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _

calculated independently. That is, if all assemblies being stored in the racks utilized burnup cmdit, the additior al boron required for tolerances and uncertainties in the bumup credit calculation would be 300 ppm. Similarly,if all assemblies being stored in the racks utilized IFBA cmdit, the additional boron required for tolerances and uncertainties in the IFBA credit calculation would be 300 ppm. When some assemblies being stored in the all cell configuration utilize burnup credit and some utilize IFBA credit, the boron required for uncenainties and tolerances is not the sum of these two amounts. Instead, the more limiting of the two amounts bounds tolerances and uncenainties for both burnup and IFBA credit. Thus, the total required boron in the all cell configuration without the water box insert is 500 ppm (the most limiting of Sections 6.1.3 and 6.1.4 of Reference 1) for the Region 1 spent fuel racks.

! 4.1.2.2 Criticality Analysis of Region 1 Checkerboard #1 Storage l

This section describes the analytical techniques and models employed to perform the criticality analysis and reactivity equivalencing evaluations for the storage of fuel in Checkerboard #1 cells of j

the Region I spent fuel storage racks.

{

Section 4.1.2.2.1 describes the no soluble boron 95/95 Kn KENO-Va calculations. Section 4.1.2.2.2 discusses the results of the spent fuel rack 95/95 Ensoluble boron credit calculations.

Section 4.1.2.2.3 presents the results of calculations performed to show the minimum burnup requirements for assemblies with initial enrichments above those determined in Section 4.1.2.2.2 Section 4.1.2.2.4 presents the results of calculations performed to show the minimum IFBA requirements for assemblies with initial enrichments above those determined in Section 4.1.2.2.2.

Finally, Section 4.1.2.2.5 presents now the application of the reactivity equivalencing methodology is used with multiple reactivity credit techniques.

4.1.2.2.1 No Soluble Boron 95/95 Kn Calculation

, The following assumptions, in addition to those in Section 4.1.1.1, are used to develop the No Soluble Boron 95/95 Kn KENO-Va model for storage of fuel assemblies in Checkerboard #1 cells of the Region 1 spent fuel storage rack:

1. Fuel assemblies contain uranium dioxide at a nominal enrichment of 1.70 and 3.55 w/o U235 over the entire length of each rod.

2. Fuel storage cells are loaded with symmetrically positioned (centered within the storage cell) fuel assemblies in a 2x2 matrix checkerboard arrangement. The checkerboard #1 ,

configuration contains two assemblies at 1.70 w/o U235 diagonally adjacent to each other l and the remaining two assemblies at 3.55 w/o U 235 Figure 13 of Reference 1 shows the checkerboard #1 configuration for the limiting case wherein the water box insert is l

removed. l 24 Attachnent I safety Evaluation NOC-AE-00178

With the above assumptions, the KENO-Va calculations of La under nominal conditions resulted in a La of 0.97266.

The 95/95 La for the Region 1 spent fuel rack checkerboard #1 configuration is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 8 of Reference 1 and results in a 95/95 La of 0.99852.

Since La is < l.0, the Region 1 spent fuel racks will remain suberitical when checkerboard #1 cells am loaded with 1.70 and 3.55 w/o U235 17x17XL fuel assemblies and no soluble boron is present in the spent fuel pool water.

4.1.2.2.2 Soluble Boron Credit Na Calculations The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for checkerboard #1 cell storage in the Region I s cent fuel racks are similar to those in Section 4.1.1.1 except for the assumption regarding the moderator soluble boron concentration. The moderatoris replaced with water containing 200 ppm soluble boron. With the above assumptions, the KENO-Va calculation for the nominal case with 200 ppm soluble boron in the moderator resulted in a La of 0.92500.

The 95/95 La is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 8 on page 76 and results in a 95/95 Laof 0.94889.

Since La is s 0.95 including soluble boron credit and uncertainties at a 95/95 probability / confidence level, the acceptance criterion for criticality is met for c checkerboard #1 storage of 17x17XL fuel assemblies in the Region 1 spent fuel racks. Storage of fuel assemblies with nominal enrichments no greater than 1.70 and 3.55 w/o U 235 is acceptable in checkerboard #1 cells including the presence of 200 ppm soluble boron.

I 4.1.2.2.3 Bumup Credit Reactivity Equivalenemg i Storage of fuel assemblies with initial enrichments high. r than 1.70 and 3.55 w/o U 235 in checkerboard #1 cells of the Region I spent fuel racks is achievable by means of burnup credit using reactivity equivalencing.

Figures 15 and 16 of Reference 1 show the constant La contours generated for checkerboard #1 cell storage in the Region I spent fuel racks. These curves represent combinations of fuel enrichment and discharge burnup which yield the same rack multiplication factor (La) as l compared to the rack loaded with 1.70 and 3.55 w/o U235 17x17XL fuel assemblies at zero burnup l

in checkerboard #1 celllocations.

Attachnumt I safety Evaluaum '

NOC-AE-00178 l

t

{

Uncertainties associated with burnup credit include a reactivity uncenainty of 0.01 delta Ng at 30,000 MWD /MTU applied linearly with bumup to account for calculation and depletion uncenainties and 5% on the calculated bumup to account for bumup measurement uncertainty. The amount of additional soluble boron needed to account for these uncertainties in the bumup requirement of Figures 15 and 16 (Reference 1) was 100 ppm. This is additional boron above the 200 ppm required in Section 4.1.2.2.2. This results in a total soluble boron requirement of 300 ppm.

4.1.2.2.4 IFBA Credit Reactivity Equivalencing Storage of fuel assemblies with nominal enrichments greater than those addressed in Section 4.1.2.2.2 is achievable by means ofIFBA credit ising the concept of reactivity equivalencing. The results of the IFBA credit reactivity equivalencing for the Region 1 spent fuel racks are provided in Table 11 of Reference 1. The results applicable to the checkerboard #1 configuration are also illustrated in Figure 19 of Reference 1, which shows the constant Lacontour generated for the 3.55 w/o U2n fuel.

Uncertainties associated with IFBA credit include a 5% manufacturing tolerance and a 10%

calculational uncertainty on the B" loading of the IFBA rods. The amount of additional soluble boron needed to account for these uncertainties in the IFBA credit requirement of Table 11 (Reference 1)is 50 ppm for the checkerboard #1 configuration. This is additional boron above the 200 ppm required in Section 4.1.2.2.2. The soluble boron needed for IFBA credit is bounded by the 100 ppm required for bumup credit in the Region 1 spent fuel racks as determined in Section 4.1.2.2.3. Therefore, the total soluble boron credit required for the Region 1 spent fuel rack.s checkerboard #1 configuration remains at 300 ppm.

4.1.2.2.5 Reactivity Equivalencing Application In Section 4.1.2.2.3 of the repon, the boron requirement to compensate for tolerances and uncertainties for burnup credit was determined to be 100 ppm. In Section 4.1.2.2.4, the boron requirement to compensate for tolerances and uncertainties for IFBA credit was determined to be 50 ppm. These boron values for bumup and IFBA credit tolerances and uncertainties were calculated independently. That is, if all assemblies being stored in the racks utilized burnup credit, the additional boron required for tolerances and uncenainties in the burnup credit calculation would be 100 ppm. Similarly, if all assemblies being stored in the racks utilized IFBA credit, the additional boron required for tolerances and uncenainties in the IFBA credit calculation would be 50 ppm.

When some assemblies being stored in the checkerboard #1 configuration utilize burnup cmdit and l some utilize IFBA credit, the boron required for uncertainties and tolerances is not the sum of these two amounts. Instead, the mc.e limiting of the two amounts bounds tolerances and uncenainties for both burnup and IFBA credit. Thus, the total required boron in the checkerboard #1 configuration is 300 ppm (the most limiting of Sections 4.1.2.2.3 and 4.1.2.2.4) for the spent fuel racks.

26 Attachnent I safety Evaluation NOC-AE-00178 1

l L )

l 4.1.2.3 Criticality Analysis of Region 1 Checkerboard #2 Storage This section describes the analytical techniques and models employed to perform the criticality analysis for the storage of fuel in Checkerboard #2 cells of the Region I spent fuel storage racks.

Section 4.1.2.3.1 describes the no soluble boron 95/95 La KENO-Va calculations and Section 4.1.2.3.2 discusses the results of the spent fuel rack 95/95 La soluble boron credit calculations.

Section 4.1.2.3.3 presents the msults of calculations performed to show the minimum bumup requirements for assemblies with initial enrichments above those determined in 4.1.2.3.1. Section 4.1.2.3.4 presents the results of calculations performed to show the minimum IFBA requirements for assemblies with initial enrichments above those determined in Section 4.1.2.3.1. Finally, Section 4.1.2.3.5 presents how the application of the reactivity equivalencing methodology is used with multiple reactivity credit techniques.

4.1.2.3.1 No Soluble Boron 95/95 La Calculation The following assumptions,in addition to those in Section 4.1.1.1, are used to develop the No Soluble Bomn 95/95 La KENO-Va model for storage of fuel assemblies in checkerboard #2 cells of the Region I spent fuel storage rack:

1. Fuel assemblies contain uranium dioxide at a nominal enr'chment of 1.4,1.7,2.5 or 4.95 w/o U* over the entire length of each rod, for the case wherein the water box insen is removed.

2. Fuel storage cells are loaded with symmetrically positioned (centered within the stomge cell) fuel assemblies. The configuration is shown in Figure 13 of Reference 1.

With the above assumptions, the KENO-Va calculations of Launder nominal conditions resulted in a La of 0.97636.

The 95/95 Kg for the Region I spent fuel rack checkerboard #2 configuration is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 9 of Reference 1 and results in a 95/95 Keg of 0.99868.

Since La is < l.0, the Region I spent fuel racks will remain suberitical when checkerboard #2 cells are loaded with 1.4,1.7,2.5, and 4.95 w/o U* 17x17XL fuel assemblies and no soluble boron is present in the spent fuel pool water.

4.1.2.3.2 Soluble Boron Credit La Calculations The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for checkerboard #2 cell storage in the Region 1 spent fuel racks are similar to those in Section 4.1.2.3.1 except for the assumption regarding the moderator soluble boron concentration. The j 27 Attachnent 1 I safety Evaluation NOC-AE-00178

moderator is replaced with water containing 250 ppm soluble baron. Given these assumptions, the KENO-Va calculation for the nominal case results in a Kg of 0.92358.

Temperature and methodology biases must be considered in the final Ng summation prior to comparing against the 0.95 Kglimit. ' he 95/95 Kgis developed by adding the temperature and methodology biases and the statisticM ; ofindependent tolerances and uncenainties to the nominal KENO-Va reference macti + 'he summation is shown in Table 9 of Reference 1 and results in a 95/95 Kg of 0.94405.

Since La is 5 0.95 including soluble boron credit and uncertainties at a 95/95 probability / confidence level, the acceptance criterion for criticality is met for checkerboard #2 cell storage of 17x17XL fuel assemblies in the Region I spent fuel racks. Storage of fuel assemblies with nominal enrichments no greater than 1.4,1.7,2.5 and 4.95 w/o U 235 s acceptable in checkerboard #2 cells including the presence of 250 ppm soluble boron.

4.1.2.3.3 Eurnup Credit Reactivity Equivalencing Storage of fuel assemblies with initial enrichments higher than 1.4,1.7 and 2.5 w/o U235 in checkerboard #2 cells of the Region I spent fuel racks is achievable by means of burnup credit using reactivity equivalencing.

Figures 14,15 and 17 of Reference I shows the constant Lacontours generated for checkerboard

2 cell storage in the Region I spent fuel racks. These curves represent combinations of fuel enrichment and discharge bumup which yield the same rack multiplication factor (La) as compared to the rack loaded with 1.4,1.7 and 2.5 ,v/o 23 U s17xl7XL fuel assemblies at zero burnup in checkerboard #2 cell locations.

Uncenainties associated with burnup credit include a reactivity uncenainty of 0.01 delta Naat 30,000 MWD /MTU applied linearly with burnup to account for calculation and depletion uncertainties and 5% on the calculated burnup to account for burnup measurement uncenainty. The amount of additional soluble boron needed to account for these uncenainties in the burnup requirement of Figures 14,15 and Figure 17 was 150 ppm. This is additional boron above the 250 ppm required in Section 4.1.2.3.2. This results in a total soluble boron requirement of 400 ppm.

4.1.2.3.4 IFBA Credit Reactivity Equivalencing I l

Storage of fuel assemblies with nominal enrichments greater than those determined in 4.1.2.3.2 is achievable by means ofIFBA credit using the concept of reactivity equivalencing. (

l A series of reactivity calculations is performed to generate a set ofIFBA rod number versus l enrichment orde ed pairs which all yield the equivalent Ng when the fuel is stored in any of the l

three configurations analyzed for the Region 1 spent fuel racks. The results of the IFBA credit 28 Attachnent I safety Evaluation NOC-AE-00178

reactivity equivalencing for the Region 1 gent fuel racks are provided in Table 11 of Reference 1.

The results applicable to the checkerboard #2 configuration are also illustrated in Figure 18 of Reference 1, which shows the constant Kar contour generated for the 2.5 w/o U* fuel.

Uncenainties associated with IFBA credit include a 5% manufacturing tolerance and a 10%

calculational uncenainty on the B' loading of the IFBA rods. The amount of additional soluble boron needed to account for these uncenainties in the IFBA credit requirement of Table 11 is 50 ppm for the checkerboard #2 configuration. This is additional boron above the 250 ppm required in Section 4.1.2.3.2. The soluble boron needed for IFBA credit is bounded by the 150 ppm required for burnup credit in the Region I spent fuel racks as determined in Section 4.1.2.3.3. Therefore, the total soluble boron credit required for the Region i spent fuel racks checkerboard #2 configuration remains at 400 ppm.

4.1.2.3.5 Reactivity Equivalencing Application In Section 4.1.2.3.3 of the repon, the boron requirement to compensate for tolerances and uncertainties for burnup credit was determined to be 150 ppm. In Section 4.1.2.3.4 the boron requirement to compensate for tolerances and uncenainties for IFBA credit was determined to be 50 ppm. These boron values for burnup and IFBA credit tolerances and uncenainties were calculated independently. That is, if all assemblies being stored in the racks utilized burnup credit, the additional boron requimd for tolerances and uncertainties in the burnup credit calculation would be 150 ppm. Similarly, if all assemblies being stored in the racks utilized IFBA credit, the additional boron required for tolerances and uncenainties in the IFBA credit calculation would be 50 ppm.

When some assemblies being stored in the checkerboard #2 configuration utilize burnup credit and some utilize IFBA credit, the boron required for uncer'.ainties and tolerances is not the sum of these two amounts. Instead, the more limiting of the two amounts bounds tolerances and uncenainties for both burnup and IFBA credit. Thus, the total required boron in the checkerboard #2 configuration is 400 ppm (the most limiting of Sections 4.1.2.3.3 and 4.1.2.3.4) for the Region I spent fuel racks.

4.1.3 Region 2 Rack Design The Region 2 racks have a 9.15 inch-nominal center-to-center spacing with fixed poison material surrounding each cell. The poison material is " sandwiched" between stainless r. eel cell walls as shown in Figure 3 of Reference 1.

4.1.3.1 Criticality Analysis of Region 2 All Cell Storage This section describes the analytical techniques and models employed to perform the criticality i analysis and reactivity equivalencing evaluations fer the storage of fuel in all cells of the Region 2 spent fuel storage racks.

29 Attachstent I safety Evaluauon NOC-AE-00178

Section 4.1.3.1.1 describes the no soluble boma 95/95 Ken KENO-Va calculations. Section 4.1.3.1.2 discusses the results of the spent fuel rack 95/95 Ken soluble boron credit calculations.

Section 4.1.3.1.3 presents the results of calculations performed to allow storage of assemblies with higher enrichments in peripheral cells of the Region 2 racks. Finally, Section 4.1.3.1.4 presents the results of calculations performed to show the minimum bumup requirements for assemblies with initial enrichments above those determined in Section 4.1.3.1.1.

4.1.3.1.1 No Soluble Boron 95/95 Ken Calculation The Section 4.1.1.1 assumptions are used to develop the No Soluble Boron 95/95 Ken KENO-Va model for storage of fuel assemblies in all cells of the Region 2 spent fuel storage rack. For thi s 2

case the fuel assemblies contain uranium dioxide at a nominal enrichment of 1.20 w/o U n over the entire length of each rod.

The KENO-Va calculations of Ken nder u nominal conditions resulted in a Ken of 0.97403, as shown in Table 12 of Reference 1.

The 95/95 Ken for the Region 2 spent fuel rack all cell storage configuration is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 12 and results in a 95/95 Kenof 0.99896.

Since Ken is < 1.0, the Region 2 spent fuel racks will remain subcritical when all cells are loaded 2

with 1.20 w/o U " 17x17XL fuel assemblies and no soluble boron is present in the spent fuel pool water.

4.1.3.1.2 Soluble Boron Credit Ken Calculations The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for all cell storage in the Region 2 spent fuel racks are similar to the No Soluble Boron Case assumptions (Sectior 4.1.1.1) except that the moderator soluble bomn concentration is set at 200 ppm soluble boron. With the above assumptions, the KENO-Va calculation resulted in a Ken of 0.91754.

The 95/95 Ken is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 12 of Reference 1 and results in a 95/95 Kenof 0.94259.

Since Ken is s 0.95 including soluble boron credit and uncenainties at a 95/95 probability / confidence level, the acceptance criterion for criticality is met for all cell storage of 17x17XL fuel assemblies in the Region 2 spent fuel racks. Storage of fuel assemblies with nominal 2

enrichments no greater than 1.20 w/o U " is acceptable in all cells including the presence of 200 ppm soluble boron.

30 Attachment i safety Evaluatmn NOC-AE-00178

4.1.3.1.3 Higher Enrichment in Peripheral Cells Calculations wem performed to assess the reactivity impact of loading a slightly higher fuel assembly enrichment in the peripheral cells (next to the spent fuel pool wall or separated from Region I fuel by an empty row). Two KENO-Va models were used to determine the reactivity effects. The first model is a finite layout of the Region 2 fuel racks loaded with 1.20 w/o U235 fuel assemblies. The second model is also a finite layout of the Region 2 fuel racks loaded with 1.20 w/o U 235 fuel asseniblies in the interior locations and surrounded by 1.40 w/o U 235 fuel assemblies on the outside edge of the racks. The reactivity increase between these two calculations is 0.00069 delta Kn. The difference in reactivity between these two calculations is less than the reactivity margin available to maintain the No Soluble Boron Kn < l.0 and less than the reactivity marge available to maintain the soluble boron credit Nn s 0.95. Therefore, it is concluded that the No Soluble Boron Lawill 235 remain < l.0 and the soluble boron credit Lnwill not exceed 0.95 when nominal 1.40 w/o U fuel assemblies are stored in the peripheral cells of Region 2.

4.1.3.1.4 Burnup and Decay Time Reactivity Equivalencing Storage of fuel assemblies with initial enrichments higher than 1.20 w/o U 235 or 1.40 w/o U 235 (periphery only) in all cells of the Region 2 spent fuel racks is achievable by means of burnup credit and Decay time credit using reactivity equivalencing.

Figures 21 and Figure 22 of Reference 1 show the constant Nncontours generated for all cell storage in the Region 2 spent fuel racks. The curves represent combinations of fuel enrichment and discharge burnup which 2 y5ield the same rack multiplication factor (Kn) as compared to the rack loaded with 1.20 w/o U or 1.40 w/o U235 (periphery only) 17x17XL fuel assemblies at zero burnup in all celllocations.

)

Decay Time Credit is an extension of the Bumup Credit process which includes the time an assembly has been discharged as a variable. This methodology gains additional margin in reactivity and reduces the minimum burnup requirements. The all cell decay time bumup cmdit curves are shown in Figure 23 of Reference 1.

L Uncertainties associated with burnup credit include a reactivity uncertainty of 0.01 delta Knat 30,000 MWD /MTU applied linearly with bumup to account for calculation and depletion uncertainties and 5% on the calculated burnup to account for burnup measurement uncertainty. The I amount of additional soluble boron needed to account for these uncertainties in the bumup requirement of Figure 21, Figure 22, and Figure 23 was 500 ppm. This is additional boron above the 200 ppm required in Section 4.1.3.1.2. This results in a total soluble boron requirement of 700 ppm.

l 4 l

31 Attachnent 1 1 safety Evaluanon NOC-AE-00178 f

4.1.3.2 Criticality Analysis of Region 2 3-out-of-4 Storage

' This section describes the analytical techniques and models employed to perform the criticality analysis and reactivity equivalencing evaluations for the storage of fuel in 3-out-of-4 cells of the Region 2 spent fuel storage racks. Section 4.1.3.2.1 describes the no soluble boron 95/95 Ku KENO-Va calculations. Section 4.1.3.2.2 discusses the results of the spent fuel rack 95/95 Kg soluble boron credit calculations. Finally, Section 4.1.3.2.3 presents the results of calculations performed to show the minimum bumup requirements for assemblies with initial enrichments above those determined in Section 4.1.3.2.1.

4.1.3.2.1 No Soluble Boron 95/95 La Calculation The following assumptions, in addition to those in Section 4.1.1.1, are used to develop the No l l

Soluble Boron 95/95 La KENO-Va model for storage of fuel assemblies in 3-out-of-4 cells of the Region 2 spent fuel storage rack:

1. Fuel assemblies contain uranium dioxide at a nominal enrichment of 1.70 w/o U235 over the entire length of each rod.

2. Fuel storage cells are loaded with symmetrically positioned (centered within the storage cell) fuel assemblies in a 3-out-of-4 checkerboard arrangement. A 3-out-of-4 checkerboard with empty cells means that no more than 3 fuel assemblies can occupy any 2x2 matrix of storage cells. Figure 20 of Reference 1 shows the 3-out-of-4 checkerboard configuration.

With the above assumptions, the KENO-Va calculations of Rn under nominal conditions resulted in a Kg of 0.97137, as shown in Table 16 of Reference 1.

The 95/95 Kg for the Region 2 spent fuel rack 3-out-of-4 checkerboard configuration is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in l Table 16 of Reference 1 and results in a 95/95 Ken of 0.98999.

Since Kg is < l.0, the Region 23 2 spent fuel racks will remain suberitical when 3-out-of-4 cells are loaded with 1.70 w/o U 17x 17XL fuel assemblies and no soluble boron is present in the spent !

fuel pool water.

4.1.3.2.2 Soluble Boron Credit La Calculations 1

The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for 3- .

out-of-4 storage in the Region 2 spent fuel racks are similar to those in Section 4.1.1.1 except that the moderator soluble boron concentration is set at 200 ppm. With the above assumptions, the 32 Attachnent I safety Evaluation NOC-AE-00178

l

(

KENO-Va calculation for the nominal case with 200 ppm soluble boron in the moderator resulted in a Kenof 0.92083.

The 95/95 Ken is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 16 of Reference 1 and results in a 95/95 Ken of 0.93974.

Since Ken is 5 0.95 including soluble boron credit and uncertainties at a 95/95 l

probability / confidence level, the acceptance criterion for criticality is met for 3-out-of-4 storage of 17x17XL fuel assemblies in the Region 2 spent fuel racks. Storage of fuel assemblies with nominal l enrichments no greater than 1.70 w/o U235 is acceptable in 3-out-of-4 cells including the presen;e of 200 ppm soluble boron.

4.1.3.2.3 Burnup Credit Reactivity Equivalencing Storage of fuel assemblies with initial enrichments higher than 1.70 w/o U 235 in 3-out-of-4 cells of the Region 2 spent fuel racks is achievable by means of burnup credit using reactivity equivalencing. Figure 24 of Reference I shows the constant Ken contour generated for 3-out-of-4 storage in the Region 2 spent fuel racks. The contour represents combinations of fuel enrichment and discharge bumup which yield the same rack multiplication factor (Keg) as compared to the rack i 235 loaded with 1.70 w/o U 17x17XL fuel assemblies at zero burnup in 3-out-of-4 cell locations.

Uncertainties associated with burnup credit include a reactivity uncertainty of 0.01 delta Ken at 30,000 MWD /MTU applied linearly with burnup to account for calculation and depletion uncertainties and 5% on the calculated bumup to account for burnup measurement uncertainty. The amount of additional soluble boron needed to account for these uncertainties in the burnup requirement of Figure 24 was 350 ppm. This is additional boron above the 200 ppm required in Section 4.1.3.2.2. This results in a total soluble boron requirement of 550 ppm.

4.1.3.3 Criticality Analysis of Region 2 2-out-of-4 Storage This section describes the analytical techniques and models employed to perform the criticality analysis for the storage of fuel in 2-out-of-4 cells of the Region 2 spent fuel storage racks.

l Section 4.1.3.3.1 describes the no soluble boron 95/95 Keg KENO-Va calculations and Section 4.1.3.3.2 discusses the results of the spent fuel rack 95/95 Ken soluble boron credit calculations.

Finally, Section 4.1.3.3.3 presents the results of calculations performed to show the minimum burnup requirements for assemblies with initial enrichments above those determined in Section 4.1.3.3.1.

I l

33 Attachnent I safety Evaluanon NOC-AE-00178

4.1.3.3.1 No Soluble Boron 95/95 La Calculation The following assumptions, in addition to those in Section 4.1.1.1, are used to develop the No Soluble Boron 95/95 Kg KENO-Va model for storage of fuel assemblies in 2-out-of-4 cells of the Region 2 spent fuel storage rack:

1. Fuel assemblies contain uranium dioxide at a nominal enrichment of 4.85 w/o U235 over the entire length of each rod.

i

2. Fuel storage cells are loaded with symmetrically positioned (centered within the storage cell) fuel assemblies in a 2-out-of-4 checkerboard arrangement as shown in Figure 20 of Reference a

1. A 2-out-of-4 checkerboard with empty cells means that no 2 fuel assemblies may be stored I face adjacent.

With the above assumptions, the KENO-Va calculations of La under nominal conditions resulted i in a La of 0.97875, as shown in Table 17 of Reference 1. '

The 95/95 La for the Region 2 spent fuel rack 2-out-of-4 checkerboard configuration is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 17 of Reference 1 and results in a 95/95 La of 0.99862. i Since Lais < l.0, the Region23 2 spent fuel racks will remain suberitical when 2-out-of-4 cells are loaded with 4.85 w/o U 17x17XL fuel assemblies and no soluble boron is present in the spent fuel pool water. In the next section, soluble boron credit will be used to provide safety margin by determining the amount of soluble boron required to maintain La 5 0.95 including tolerances and uncertainties.

4.1.3.3.2 Soluble Boron Credit La Calculations The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for 2-out-of-4 storage in the Region 2 spent fuel racks are similar to those in Section 4.1.1.1 except that the moderator soluble boron concentrations set at 250 ppm. With the above assumptions, the KENO-Va calculation for the nominal case results in a La of 0.93025.

l The 95/95 La is developed by adding the temperature and methodology biases and the statistical sum of independent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 17 of Reference 1 and results in a 95/95 Laof 0.94578.

Since En is s 0.95 including soluble boron credit and uncertainties at a 95/95 probability / confidence level, the acceptance criterion for criticality is met for 2-out-of-4 cell storage of 17x17XL fuel assemblies in the Region 2 spent fuel racks. Storage of fuel assemblies with 34 Attachment i Safety Evaluation l

NOC-AE-00178 '

I l

l

nominal enrichments no greater than 4.85 w/o U 235 is acceptable in 2-out-of-4 cells including the presence of 250 ppm soluble boron.

4.1.3.3.3 Burnup Credit Reactivity Equivalencing Storage of fuel assemblies with initial enrichments higher than 4.85 w/o U235 n 2-out-of-4 cells of the Region I spent fuel racks is achievable by means of burnup credit using reactivity equivalencing. Figure 25 of Refemnce 1 shows the constant La contour generated for 2-out-of-4 cell storage in the Region I spent fuel racks. The contour represents combinations of fuel enrichment and discharge burnup which yield the same rack multiplication factor (Kg) as compared to the rack loaded with 4.85 w/o U 235 17x17XL fuel assemblies at zero bumup in 2-out-of-4 celllocations.

Uncertainties associated with burnup credit include a reactivity uncertainty of 0.01 delta Na at 30,000 MWD /MTU applied m.carly with bumup to account for calculation and d-letion uncenainties and 5% on the calculated burnup to account for bumup measureme. uncertainty. The amount of additional soluble boron needed to account for these uncertainties in the bumup requirement of Figure 25 was 50 ppm. This is additional boron above the 250 ppm required in Section 4.1.3.3.2. This results in a total soluble boron requirement of 300 ppm.

4.1.3.4 Criticality Analysis of Region 2 RCCA #1 Checkerboard Storage This section describes the analytical techniques and models employed to perform the criticality analysis and reactivity equivalencing evaluations for the storage of fuel in a RCCA #1 checkerboard in the Region 2 spent fuel storage racks.

Section 4.1.3.4.1 describes the no soluble boron 95/95 Kg IGNO-Va calculations.

Section 4.1.3.4.2 discusses the results of the spent fuel rack 95/95 Kens oluble boron credit calculations. Finally, Section 4.1.3.4.3 presents the results of calculations performed to show the minimum burnup requirements for assemblies with initial enrichments above those determined in Section 4.1.3.4.1.

Units 1 and 2 have used both hafnium and silver-indium-cadmium RCCA absorber material. In the spent fuel storage rack environment, the hafnium RCCAs provide slightly less reactivity hold-down so the hafnium RCCAs were used in the criticality analysis. This bounds the effects of either type of RCCA in the Region 2 storage racks.

4.1.3.4.1 No Soluble Boron 95/95 Ng Calculation g

The following assumptions, in addition to those in Section 4.1.1.1, are used to develop the No !

Soluble Boron 95/95 Kg KENO-Va model for storage of fuel assemblies in a RCCA # 1 !

checkerboard in the Region 2 spent fuel storage rack.

35 l Attacharnt I !

safety Evaluanon NOC-AE-00178 !

i

! l l

1.

Fuel assemblies contain uranium dioxide at a fixed nominal enrichment of 1.40 w/o U2 U 235 over the entire length of each rod.

2. A conservative allowance for the worth of the RCCA absorber material is assumed. This is done by using conservative number densities for the absorber material equivalent to depleting the full length of the RCCA for 60,000 MWD /MTU exposure.

3. Fuel storage cells are loaded with symmetrically positioned (centered within the storage cell) fuel assemblies in a 2x2 matrix checkerboard arrangement. The RCCA #1 checkerboard contains the same fuel enrichment of 1.40 w/o U235 in all of the cells and an .

, RCCA in 1 of the 4 assemblies. Figure 20 of Reference i shows the RCCA #1

( checkerboard configuration.

With the above assumptions, the KENO-Va calculations of Ken under nominal conditions resulted in a La f o0.97006, as shown in Table 18 of Reference 1.

1 The 95/95 Kenis developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 18 of Reference 1 and results in a 95/95 Ken of 0.99917.

Since Keg is < l.0, the Region 2 spent fuel racks will remain subcritical when cells are loaded in a RCCA #1 checkerboard with a 1.40 w/o U235 17x17XL fuel assembly and no soluble boron is present in the spent fuel pool water.

l 4.1.3.4.2 Soluble Boron Credit Ken Calculations The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for RCCA #1 checkerboard storage in the Region 2 spent fuel racks are similar to those in Section 4.1.1.1 except that the moderator soluble boron concentration is set at 200 ppm. With the above

! assumptions, the KENO-Va calculation for the nominal case results in a Kenof 0.91905, as shown in Table 18 of Reference 1.

Temperature, methodology and RCCA depletion biases must be considered in the final Ken summation prior to comparing against the 95/95 Keg limit. The follow additional bias was included:

l RCCA Depletion: A reactivity bias determined in PHOENIX-P was applied to account for the effect of the depletion of the RCCA absorber material.

1 The 95/95 Ken is developed by adding the temperature and methodology biases and the statistical I

sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The i summation is shown in Table 18 of Reference 1 and results in a 95/95 Ken of 0.94795. j i

36 Attachment 1 i safety Evaluauon NOC-AE-00178 l

Since La is 5 0.95 including soluble boron credit and uncenainties at a 95/95 probability / confidence level, the acceptance criterion for criticality is met for the RCCA #1 checkerboard storage configuration of 17x17XL fuel assemblies in 23the Region 2 synt fuel racks.

Storage of fuel assemblies with nominal enrichment no gmater than 1.40 w/o U and one of the four assemblies containing a Ag-In-Cd or Hf RCCA is acceptable including the presence of 200 ppm soluble boron.

4.1.3.4.3 Burnup Credit Reactivity Equivalencing Storage of fuel assemblies with initial enrichments higher than 1.40 w/o U 235 in RCCA #1 checkerboard in the Region 2 spent fuel racks is achievable by means of bumup credit using reactivity equivalencing. Figure 26 of Reference 1 shows the constant La contour generated for RCCA #1 checkerboard in the Region 2 spent fuel racks. The contour represents combinations of fuel enrichment and discharge burnup which yield the same rack multiplication factor (Lg) as compared to the rack loaded with 1.40 w/o U235 fuel assemblies at zero burnup in RCCA #1 checkerboard locations.

Uncertainties associated with bumup credit include a reactivity uncertainty of 0.01 delta Ng at 30,000 MWD /MTU applied linearly with burnup to account for calculation and depletion uncenainties and 5% on the calculated burnup to account for burnup measurement uncenainty. The amount of additional soluble boron needed to account for these uncertainties in the burnup requirement of Figure 26 was 450 ppm. This is additional boron above the 200 ppm required in Section 4.1.3.4.2. This results in a total soluble boron requirement of 650 ppm.

4.1.3.5 Criticality Analysis of Region 2 RCCA #2 Checkerboard Storage This section describes the analytical techniques and models employed to perform the criticality analysis and mactivity equivalencing evaluations for the storage of fuel in an RCCA #2 checkerboard in the Region 2 spent fuel storage racks.

Section 4.1.3.5.1 describes the no soluble boron 95/95 Kg KENO-Va calculations.

Section 4.1.3.5.2 discusses the results of the spent fuel rack 95/95 La soluble boron credit calculations. Finally, Section 4.1.3.5.3 presents the results of calculations performed to show the minimum burnup requirements for assemblies with initial enrichments above those determined in Section 4.1.3.5.1.

Units 1 and 2 have used both hafnium and silver-indium-cadmium RCCA absorber material. In the spent fuel storage rack environment, the hafnium RCCAs provide slightly less reactivity hold-down; consequently, the hafnium RCCAs were used in the criticality analysis. This bounds the effects of either type of RCCA in the Region 2 storage racks.

37 Attachnent I safety Evaluation NOC-AE-00178 i

(

l l 4.1.3.5.1 No Soluble Boron 95/95 Ken Calculation The following assumptions, in addition to those in Section 4.1.1.1, are used to develop the. No Soluble Boron 95/95 Ken KENO-Va model for storage of fuel assemblies in a RCCA #2 l checkerboard in the Region 2 spent fuel storage rack:

1. Fuel assemblies contain uranium dioxide at a fixed nominal enrichment of 1.65 w/o U235 l over the entire length of each rod.

2. A conservative allowance for the wonh of the RCCA absorber material is assumed. This is l done by using conservative number densities for the absorber material equivalent to depleting the full length of the RCCA for 60,000 MWD /MTU exposure.

l 3. Fuel storage cells are loaded with symmetrically positioned (centered within the storage cell)

I fuel assemblies in a 2x2 matrix checkerboard arrangement. The RCCA #2 checkerboard I l contains the same fuel enrichment of 1.65 w/o U235 in all of the cells and RCCAs in two (diagonally adjacent) of the four assemblies. Figure 20 of Reference I shows the RCCA #2 checkerboard configuration.

With the above assumptions, the KENO-Va calculations of Ken under nominal conditions resulted in a Keg of 0.95734, as shown in Table 19 of Reference 1.

l l Temperature, methodology and RCCA depletion biases must be considered in the final Ken i summation prior to comparing against the 1.0 Ken limit. The following additional bias was l included:

! RCCA Depletion: A reactivity bias determined in PHOENIX-P was applied to account for l the effect of the depleton of the RCCA absorber material.

1 The 95/95 Ken is developed by adding the temperature and methodology biases and the statistical l l sum ofindependent tolerances and uncenainties to the nominal KENO-Va reference reactivity. The l summation is shown in Table 19 of Reference 1 and results in a 95/95 Ken of 0.99755.

i Since Keg is < l.0, the Region 2 spent fuel racks will remain suberitical when cel!s are loaded in a 235 RCCA #2 checkerboard with a 1.65 w/o U 17x17XL fuel assembly and no soluble boron is present in the spent fuel pool water.

4.1.3.5.2 Soluble Boron Credit Ken Calculations The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for RCCA #2 checkerboard storage in the Region 2 spent fuel racks are similar to those in Section 4.1.1.1 except that the moderator soluble boron concentration is set at 250 ppm soluble boron.

38 Attachment I safety Evaluation NOC-AE-00178

4.1.3.5.1 No Soluble Boron 95/95 Kn Calculation The following assumptions, in addition to those in Section 4.1.1.1, are used to develop the No Soluble Boron 95/95 Ken KENO-Va model for storage of fuel assemblies in a RCCA #2 checkerboard in the Region 2 spent fuel storage rack:

1. Fuel assemblies contain uranium dioxide at a fixed nominal enrichment of 1.65 w/o U235 over the entire length of each rod.

2. A conservative allowance for the worth of the RCCA absorber material is assumed. This is done by using conservative number densities for the absorber material equivalent to depleting the full length of the RCCA for 60,000 MWD /MTU exposure.

3. Fuel storage cells are loaded with symmetrically positioned (centered within the storage cell) fuel assemblies in a 2x2 matrix checkerboard arrangement. The RCCA #2 checkerboard 235 contains the same fuel enrichment of 1.65 w/o U in all of the cells and RCCAs in two (diagonally adjacent) of the four assemblies. Figure 20 of Reference 1 shows the RCCA #2 checkerboard configuration.

With the above assumptions, the KENO-Va calculations of Kn under nominal conditions resulted in a Nn of 0.95734, as shown in Table 19 of Reference 1.

Temperature, methodology and RCCA depletion biases must be considered in the final Ln summation prior to comparing against the 1.0 Kn limit. The following additional bias was included:

RCCA Depletion: A reactivity bias detemiined in PHOENIX-P was applied to account for the effect of the depletion of the RCCA absorber material. .

The 95/95 Knis developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 19 of Reference 1 and results in a 95/95 Kn of 0.99755. l

)

Since Ken is < l.0, the Region 2 spent fuel racks will remain suberitical when cells are loaded in a 235 RCCA #2 checkerboard with a 1.65 w/o U 17x17XL fuel assembly and no soluble boron is present in the spent fuel pool water.

4.1.3.5.2 Soluble Boron Credit Ken Calculations

}

The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for RCCA #2 checkerboard storage in the Region 2 spent fuel racks are similar to those in Section 4.1.1.1 except that the modemtor soluble boron concentration is set at 250 ppm soluble boron.

38 Attachnwnt i safety Evaluanon NOC-AE-00178

4.1.3.5.1 No Soluble Boron 95/95 &n Calculation The following assumptions, in addition to those in Section 4.1.1.1, are used to develop the No Soluble Boron 95/95 Ng KENO-Va model for storage of fuel assemblies in a RCCA #2 checkerboard in the Region 2 spent fuel storage rack:

1. 2 Fuel assemblies contain uranium dioxide at a fixed nominal enrichment of 1.65 w/o U "

over the entire length of each rod.

2. A conservative allowance for the worth of the RCCA absorber material is assumed. This is done by using conservative number densities for the absorber material equivalent to depleting the full length of the RCCA for 60,000 MWD /MTU exposure.

3. Fuel storage cells are loaded with symmetrically positioned (centered within the storage cell) fuel assemblies in a 2x2 matrix checkerboard arran contains the same fuel enrichment of 21.65 w/o U "gement. The RCCA #2 in all of the cells and RCCAs in two checkerboa (diagonally adjacent) of the four assemblies. Figum 20 of Reference I shows the RCCA #2 checkerboard configuration.

With the above assumptions, the KENO-Va calculations of Ng under nominal conditions resulted in a Nn fo0.95734, as shown in Table 19 of Reference 1.

Temperature, methodology and RCCA depletion biases must be considered in the final Nn summation prior to comparing against the 1.0 Nn limit. The following additional bias was included: l l

RCCA Depletion: A reactivity bias determined in PHOENIX-P was applied to account for the effect of the depletion of the RCCA absorber material.

The 95/95 Ntr is developed by adding the temperature and methodology biases and the statistical sum ofindependent tolerances and uncertainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 19 of Refemnce I and results in a 95/95 Nn of 0.99755.

Since La is < l.0, the Region 2 spent fuel racks wih remain suberitical when cells are loaded in a 2

RCCA #2 checkerboard with a 1.65 w/o U " 17x17XL fuel assembly and no soluble boron is present in the spent fuel pool water.

4.1.3.5.2 Soluble Boron Credit NnCalculations 5 The assumptions used to develop the nominal case KENO-Va model for soluble boron credit for RCCA #2 checkerboard storage in the Region 2 spent fuel racks are similar to those in Section 4.1.1.1 except Gat the moderator soluble boron concentration is set at 250 ppm soluble boron.

l 38 Attachment I safety Evaluanon NOC-AE-00178 l

l I

l

4 With the above assumptions, the KENO-Va calculation for the nominal case results in a La of 0.90205, as shown in Table 19 of Reference 1.

The 95/95 Kg is developed by adding the temperature and methodology biases and the statistical sum of independent tolerances and uncenainties to the nominal KENO-Va reference reactivity. The summation is shown in Table 19 of Reference 1 and results in a 95/95 Kg of 0.94148.

Since La is 5 0.95 including soluble boron credit and uncenainties at a 95/95 probability / confidence level, the acceptance criterion for criticality is met for the RCCA #2 checkerboard storage configuration of 17x17XL fuel assemblies in the 2 Region 2 synt fuel racks.

Storage of fuel assemblies with nominal enrichment no greater than 1.65 w/o U and two of the four assemblies containing a Ag-In-Cd or Hf RCCA is acceptable including the presence of 250 ppm soluble boron.

4.1.3.5.3 Bumup Credit Reactivity Equivalencing Storage of fuel assemblies with initial enrichments higher than 1.65 w/o U235 in the RCCA #2 checkerboard in the Region 2 spent fuel racks is achievable by means of burnup credit using reactivity equivalencing. Figure 27 of Reference I shows the constant La contour generated for RCCA #2 checkerboard in the Region 2 spent fuel racks. The contour represents combinations of fuel enrichment and discharge bumup which vield the same rack multiplication factor (Lg) as compared to the rack loaded with 1.65 w/o UI35 fuel assemblies at zero bumup in RCCA #2 checkerboard.

Uncertainties associated with bumup credit include a reactivity uncenainty of 0.01 delta La at 30,000 MWD /MTU applied linearly with burnup to account for calculation and depletion uncertainties and 5% on the calculated burnup to account for burnup measurement uncertainty. The amount of additional soluble boron needed to account for these uncenainties in the bumup requirement of Figure 27 was 450 ppm. This is additional boron above the 250 ppm required in Section 4.1.3.5.2. This results in a total soluble boron requirement of 700 ppm.

4.1.4 Fuel Rod Storage Canister Criticality In Section 12 of Reference 1 a criticality analysis was performed for the Fuel Rod Storage Canister (FRSC) which is available at South Texas. This report compared the FRSC, loaded with 5.0 w/o 2

U n fuel rods, to an intact assembly with 5.0 w/o U 235 fuel rods. The conclusion was that the FRSC is less reactive than an assembly with 5.0 w/o U 235 fuel rods. However, this analysis was done independent of any rack geometry. Therefore, for storage of the FRSC in the racks, the FRSC must be teated as ifit were an assembly with enrichment and bumup of the rod in the canister with the most limiting combination of enrichment and bumup.

39 Attachna .' I safety Evaluatum NOC-AE 00178 I

4.1.5 Storage Configuration Interface Requirements The spent fuel pool is composed of two different types of racks, designated as Region 1 and Region

2. Each of these spent fuel pool areas has been analyzed for all cell storage, where all cells share the same storage requirements and limits, and checkerboard storage, where neighboring cells have different requirements and limits. A schematic of the Region 1 checkerboard pattems are shown in Figure 5 (with box insert) and Figure 13 (without box insert) of Reference 1. A schematic of the Region 2 checkerboard patterns is shown in Figure 20 of Refemnce 1.

The boundary between checkerboard zones and the boundary between checkerboard zones and all cell stored fuel must be controlled to prevent an undesirable increase in reactivity. This is accomplished by examining each 2x2 matrix surrounding a fuel assembly and ensuring that each set of 2x2 neighboring cells conform to checkerboard restrictions for the given region.

For example, consider a fuel assembly location E in the following matrix of storage cells.

A B C D E F G H I Four 2x2 matrices of storage cells which include storage cell E are created in the above figure.

They include (A, B, D, E), (B, C, E, F), (E, F, H,1), and (D, E, G, H). Each of these 2x2 matrices of cells are required to meet the checkerboard requirements determined for the given region.

4.1.5.1 Interface Requirements within Region i Using the requimment that all 2x2 matrices within the storage racks must conform to every all cell and 2x2 checkerboard requirement, the following interface requirements am applicable to Region i storage cells:

Region 1 All Cell Storage Next The boundary between all cell storage and checkerboard #1 or #2 can be to Region 1 Checkerboard #1 either separated by a vacant row of cells or the interface must be or Checkerboard #2 configured such that the first row of carryover uses 2.7 and 1.7 w/o U 235 fuel assemblies. Figure 28 of Reference 1 illustrates the carryover configuration.

Region 1 Checkerboard #1 The boundary between checkerboard #1 and checkerboard #2 can be Next to Region 1 Checkerboard either separated by a vacant row of cells or the interface must be

2 configured such that the first row of carryover uses 2.8 and 1.7 w/o U 235 fuel assemblies. Figure 29 of Reference 1 illustrates the carryover configuration.

l 40 Attachment i safety Evaluabon NOC AE-00178

l l

l 4.1.5.2 Interface Requirements within Region 2 Using the requirement that all 2x2 matrices within the storage racks must conform to every all cell l and 2x2 checkerboard requirement, the following interface requirements are applicable to Region 2 storage cells:

Region 2 All Cell Storage Next The boundary between all cell storage and 2-out-of-4 or 3-out-of-4 to Region 2 2-out-of-4 Storage storage can be either separated by a vacant row of cells or the interface l or 3-out-of-4 Storage must be configured such that the first row of carryover uses 1.7 w/o U235 fuel assemblies and empty cells. Figure 30 of Reference 1 illustrates the carryover configuration.

Region 2 All Cell Storage Next The boundary between all cell storage and RCCA checkerboard #1 or to Region 2 RCCA RCCA checkerboard #2 can be either separated by a vacant row of cells Checkerboard #1 or RCCA or the interface must be configured such that the first row of carryover 235 Checkerboard #2 uses 1.40 w/o U fuel assemblies and empty cells for the RCCA checkerboard #1 and 1.65 w/o fuel assemblies and empty cells for the RCCA checkerboard #2. Figure 31 of Reference 1 illustrates the carryover configuration.

Region 2 2-out-of-4 Storage The boundary between 2-out-of-4 storage and 3-out-of-4 storage can be Next to Region 2 3-out-of-4 either separated by a vacant row of cells or the interface must be l Storage configured such that the first row of canyover uses 4.85 w/o U 235 fuel assemblies and empty cells. Figure 32 of Reference 1 illustrates the carryover configuration.

Region 2 RCCA Checkerboard The boundary between RCCA checkerboard storage patterns can be

1 Next to Region 2 RCCA either separated by a vacant row of cells or the interface must be Checkerboard #2 configured such that the first row of carryover uses 1.40 w/o U 235 fuel assemblies with an RCCA in every other storage cell location. Figure 32 of Reference 1 illustrates the carryover configuration.

Peripheral Cellin All Cell Assemblies with equivalent enrichment of 1.4% can be placed in all cell Configuration configuration only at the periphery of the Region 2 racks.

4.1.5.3 Interface Requirements within Region 1 without Water Box Insert Using the requirement that all 2x2 matrices within the storage racks must conform to every all cell and 2x2 checkerboard requirement, the following interface requirements are applicable to Region 1 storage cells without the water box insert:

l 41 Attachment I safety Evaluation NOC-AE-00178 1

Region 1 All Cell Storage Next The boundary between all cell storage and checkerboard #1 or #2 can be to Region 1 Checkerboard #1 either separated by a vacant row of cells w the interface must be or Checkerboard #2 configured such that the first row of carryover taes 1.4,2.5 and 1.7 w/o U 235 fuel assemblies for checkerboard #2 and 2.5 and 1.7 w/o U235 fuel assemblies for checkerboard #1 Figure 33 of Rr,ference 1 illustrates the carryover co0 figuration.

Region 1 Checkerboard #1 The boundary tetween checkerboard #1 and checkerboard #2 car. be Next to Region 1 Checkerboard either separated by a vacant row of cells or tre interface must be

2 configured such that the first few of carryover u. es 1.4,2.5 and 1.7 w/o U 235 fuel assemblies. Figure 34 of Reference 1 illustrates the cmyover configuration.

4.1.5.4 Interface Requirements within Region I and Region 2 The following interface requirements illustrate example conditions which are applicable to both Region I and Region 2 storage configurations:

Open Water Cells For all configurations, an open water cell is permitted in any location of the spent fuel pool to replace an assembly since the water cell will not cause any increase in reactivity in the spent fuel pool.

Non-fissile Items For all configurations, non-fissile items may be stored in open cells of the pool provided that they are not stored in cells face-adjacent to cells that store fuel. Non-fissile items can be stored in open cells that are face-adjacent to cells that store fuel provided an evaluation is performed.

Neutron Source in a Cell The placement of a neutron source will not cause any increase in reactivity in the spent fuel pool because the neutron source is an absorber which reduces reactivity. Therefore, neutron sources may be stored in an empty cell orin an assembly. i Non-Fuel Bea:ing Assembly Non-Fuel Bearing Assembly components (i.e. thimble plugs, RCCAF, Components discrete burnable absorbers, etc.) may be stored in assemblies without affecting the storage requirements of that assembly.

4.1.5.5 Interface Requimments between Region I and Region 2 The boundary between Region I and Region 2 must be configured such that one row of open water cells is maintained between the regions (the open water row can be positioned in either region).

This requirement is necessary because the removal of the Boraflex neutron absorber panels from the criticality analysis increases the amount of neutron interaction between Region 1 and Region 2.

l l Non-fissile items can be stored in the open water row provided an evaluation is performed.

42 Attachment I safety Evaluauon NOC-AE-00178 u

I I

i 1

4.2 Spent Fuel Pool Dilution Analysis A boron dilution analysis has been completed for crediting boron in the spent fuel rack criticality analysis. The boron dilution analysis includes an evaluation of the following plant specific features:

Spent Fuel Pool and Related System Features

Administrative Procedures

!

Dilution Sources

Boration Sources

Spent Fuel Pool Dilution Evaluation e Boron Dilution Initiating Events e Boron Dilution Times and Volumes

. Loss of Offsite Power Impact The boron dilution analysis was performed to ensure that sufficient time, administrative procedures, and instmmentation are available to detect and mitigate the dilution before the spent fuel rack criticality analysis 0.95 Kerr design basis is exceeded. The design basis assumes normal plant operations and fuel movement. No other accidents, such as the misloading of a fuel assembly, are assumed to occur during the dilution accident.

4.2.1 Spent Fuel Pool And Related System Features This section provides background information on the spent fuel pool and its related systems and features. A one-line diagram of the spent fuel pool related systems is provided as Figure 1 of Reference 3. A spent fuel pool is provided for each of the two Units at South Texas. For the purposes of this evaluation, the spent fuel pool and its related systems are sufficiently similar between the two Units that they will be treated as identical and this repon will be bounding for both Units.

4.2.1.1 Spent Fuel Pool The spent fuel pool is housed in the fuel handling building. This stmeture is a seismic Class I building. The spent fuel pool is approximately 45 feet deep. The top of the pit is located on the 68' .

elevation of the fuel handling building. The bottom of the pit is at the 21'-l1" elevation. As shown j in Figure 2 of Reference 3, a transfer canal lies adjacent to the pool and connects to the reactor refueling water cavity during refueling operations. The pool and the transfer canal are connected by fuel transfer slots that can be closed by pneumatically sealed gates. The transfer canal is normally filled. On the floor elevation there is a 2" to 3" curb surrounding the pool, except for an approximately 18" gap around the fuel elevator motor next to the transfer canal. The curb, in I addition to an open floor drain, minimizes any pool dilution source from the floor elevation level.

f l Anachnent i I safety Evaluauon NOC-AE-00178

l i

The volume of the pool is approximately 486,000 gallons to the low level alarm elevation of 66'-0".

No credit is taken for the transfer canal water volume. The spent fuel storage racks and the spent fuel assemblies displace a maximum of 63,900 gallons of water. Displacement of water caused by other objects and piping in the pool is considered negligible. The remaining water volume is !

conservatively rounded down to 420,000 gallons at the low level alarm setpoint elevation.

4.2.1.2 Spent Fuel Pool Instmmentation 1

Instrumentation is available to monitor spent fuel pool water level and temperature. Additional instrumentation is provided to monitor the pressure and flow of the spent fuel pool cleanup system, and pressure, flow, and temperature of the spent fuel pool cooling system.

The instrumentation provided to monitor the temperature of the water in the spent fuel pool is l

indicated locally and alarms are annunciated in the control roorrt. The water level instrumentation l indicates locally and the high and low alarms annunciate locally and in the control room. The l

instmmentation which monitors radiation levels in the spent fuel pool area provides high radiation '

alarms locally and in the control room.

A change of one foot in spent fuel pool level with the transfer canal isolated requires approximately l 10,300 gallons of water. If the pool level was raised from the low level alarm point to the high level alarm (12"), a dilution of approximately 10,300 gallons could occur before an alarm would be received in the control room. If the spent fuel pool boron concentration were at 2500 ppm initially, l a dilution using unborated water would only result in a reduction of the pool boron concentration of approximately 61 ppm.

4.2.1.3 Spent Fuel Pool Cooling Subsystem Each of the two trains of the cooling subsystem consists of a pump, a heat exchanger, valves, piping and instrumentation. The pump takes suction from the fuel pool at an inlet located four feet below the normal pool water level, transfers the pool water through a heat exchanger and returns it back into the pool through an discharge header with sparging nozzles located on the pool bottom and at the opposite wall from the cooling system inlet. The return line includes an anti-siphoning hole at elevation 65'- 6" to limit loss of pool inventory in the event that the return line breaks below the normal water level. The heat exchangers are cooled by component cooling water.

The ponion of the spent fuel pool cooling subsystem which, ifit failed, could result in a significant release of pool water, is seismically designed.

4.2.1.4 Spent Fuel Pool Cleanup Subsystem The spent fuel pool cleanup subsystem is connected to the spent fuel pool cooling system. About 250 gpm of the spent fuel pool cooling pump (s) discharge flow can be diverted to the cleanup loop, 44 Attachnv.nt I safety Evaluauon NOC-AE-00178

which includes the spent fuel pool demineralizers and filters. The filters remove particulate from the spent fuel pool water and the spuit fuel pool demineralized removes ionic impurities.

The refueling water purification loop also uses the spent fuel pool demineralized and filters to clean up the refueling water storage tank after refueling operaaons. The flow rate in the loop is administratively limited to 200 gpm to accommodutc the design flow of the spent fuel pool demineralized.

The spent fuel pool has a surface skimmer consisting of two surface skimmers, a single strainer, a f

single pump and one filter. The skimmer pump is a centrifugal pump with a 100 gpm capacity.

The pump discharge flow passes through the filter, retuming to the spent fuel pool.

4.2.1.5 Dilution Sources l1 4.2.1.5.1 Boron Recycle System (BRS) '

l Potential Dilution Path (sh

1. A line from the outlet of the spent fuel pool heat exchangers to the BRS recycle holdup tanks.

This connection is normally isolated and is used to transfer water from the spent fuel pool to the BRS recycle holdup tanks. The isolation is by one manual valve.

l There is no check valve between the BRS recycle holdup tanks and this connection to the spent fuel pool cooling system. However,it is not credible that water would flow from the tank to the spent fuel pool cooling system. In the situation where the BRS recycle holdup tank is misaligned to the spent fuel pool through this connection, water from the spent fuel j pool cooling system would flow by gravity to the tank due to the elevation difference. Thus, this path would only result in the loss of water from the pool if the normally closed valve were to fail or were to be left open. The amount of pool water which could be lost in this scenario is limited by the location of the suction line at an elevation 4' below the normal water level and the presence of the anti-siphon hole in the spent fuel pool cooling system pump discharge line. The holdup tanks also have a high level alarm, which annunciated in the Rad Waste Control Room.

2. The second connection between the spent fuel pool and the BRS is from the BRS recycle evaporator feed pump discharge header to the cleanup loop piping. This is a normally isolated 3" line that is an additional source of makeup water to the pool. The maximum rate of addition is approximately 255 gpm, assuming both feed pumps are operating in parallel.

Two normally closed and one lock-closed manual valves are used to isolate this connection.

The recycle evaporator feed pumps can take suction from either of the two BRS recycle holdup tanks. However, by procedure, only one holdup tank is aligned at a time. Manual 45 Attachment 1 safety Evaluation NOC-AE-00178 l

valve manipulations are required to switch the pump suction to another tank. Each BRS recycle holdup tank has a totai volume of approximately 84,000 gallons and can have a boron concentration from 0 to 3000 ppm.

4.2.1.5.2 Reactor Makeup Water System The Reactor Makeup Water (RMW) System includes one reactor makeup water storage tank (RMWST) and two RMW pumps per Unit. During normal operation, one RMW pump is mnning on recirculation to provide RMW on demand to multiple users. Each RMWST contains approximately 153,050 gallons of non-borated water. Makeup to the tank is provided automatically from the demineralized water system on a low tank level signal.

Potential Dilution Path (s):

The RMW System connects to the spent fuel pool cooling system directly in the return header to the pool. Using the direct connection, the contents of the RMWST can be transferred directly to the spent fuel pool via the RMW pumps. The direct connection is normally isolated from the RMW System by a closed manual valve. The flow rate through this path is estimated to be 240 gpm, assuming one RMW pump is operating. The direct connection is used as the normal water supply to the spent fuel pool and is a source of makeup water in case of a loss of spent fuel pool inventory.

The makeup source can be unborated water, since the evaporation pacess does not carry off the boron. Evaporation actually increases the boron concentration in the pool.

4.2.1.5.3 Demineralized Water System The demineralized water system includes a demineralized water storage tank and three transfer pumps. The storage tank capacity is 951,700 gallons.

Potential Dilution Path (s):

1

1. A flow path from the pump discharge header feeds directly into the spent fuel pool cooling l system retum header to the pool.

The maximum flow from the 2" line makeup line to the spent fuel pool is estimated to be 190 gpm, assuming one transfer pump is operating. I

2. An indirect path for demineralized water to the spent fuel pool is through a 1" line to the sluice path for the spent fuel pool and other system demineralizers.

The deminemlizer is isolated from the cleanup loop by one manual valve. If this valve wem l

left open, demineralized water cotdd be transferred into the spent fuel pool. The flow from this pathway is estimated to be 55 gpra, assaming one transfer pump is operating.

46 Attachnent i Safety Evaluauon NOC-AE-00178 l

l u - - - - - - - . - - - - .-- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

1

3. The demineralized water supply to the upender hydraulic unit in the spent fuel pool area is provided by 3/8" tubing. If this line were to break, its flow rate into the spent fuel pool would j be bounded by the flow available through the largerlines as described above. Therefom, this !

3/8" path is not considered funher in this analysis. 1 4.2.1.5.4 Component Cooling Water System Component cooling water is the cooling medium for the spent fuel pool cooling system heat

{

exchangers. There is no direct connection between the component cooling system and the spent i fuel pool cooling system. If, however, a le.ak were to develop in a heat exchanger that is in seivice, the connection would be made. The flow rate of any leakage of component cooling water into the ,

spent fuel pool cooling system would be very low due to the small difference in operating pressures l between the two systems. Even if there was significant leakage from the component cooling water '

system to the spent fuel pool, the impact on the spent fuel pool boron concentration would be 4 limited to the loss of component cooling water surge tank volume that would initiate alarms and control room indications to alert the control room operators. l A low surge tank level alarm would alert the control room operators of a component cooling water system leak. If this alarm were to fail and leakage from the component cooling water system to the spent fuel pool cooling system were to continue undetected, the component cooling water surge tank would be periodically refilled with water from the demineralized water system. The resulting dilution from the demineralized water system would be bounded by the dilution events discussed in Section 4.2.1.5.3.

Because a spent fuel pool heat exchanger leak is bounded by other analyzed events, it is not considered funherin this analysis.

J.2.1.5.5 Drain Systems The equipment or floor drain systems connect directly to the spent fuel pool cooling system and skimmer system at the drain connections for the spent fuel pool pumps, heat exchangers (tube side),

filters, demineralizers, demineralized filters, the skimmer pump, and skimmer filter. Each connection has a normally closed isolation valve. Backflow through these paths is not considered credible, because the situation would cause water to back up through floor drains in a number of locations before getting into the spent fuel pool cooling system.

4.2.1.5.6 Fire Protection System In the event of a loss of spent fuel pool inventory, two local fire hose stations are potential makeup sources. These stations are capable of providing a total flow of approximately 350 gpm of non-borated water. Any planned addition of fim system water to the spent fuel pool would be under the 47 Attachment I safety Evaluation NOC-AE40178 l

l contml of an approved procedure and the effect of the addition of the non-borated water from the fire system on the spent fuel pool boron concentration would be addressed.

There is a 6" fire protection hose supply piping header located under the hose stations outside the spent fuel pool area. If this line were to break, a significant amount of water would, if not isolatea by operator action, be released into the area outside and beneath the spent fuel pool area. The fire i protection system contains instrumentation which would alarm in the control room should this type of flow develop in the fire protection system. Thus, the break of any of the fire protection hose supply piping is not considered further in this analysis.

4.2.1.5.7 Spent Fuel Pool Demineralizers The two spent fuel pool demineralizers each have a maximum capacity of 75 ft3 of 1:1 equivalent mixed bed resin. This implies a volume ratio of 60%/40% anion to cation resin. Assuming the beds are loaded with 100% anion, it would bound the capacity to remove boron when it is first aligned to the system. Each demineralized would be operated at a 250 gpm maximum flow rate.

Dilution of the spent fuel pool resulting from operation of the demineralized will not result in a change in the spent fuel pool inventory.

4.2.1.5.8 Piping There are no systems (other than those listed in section 4.2.1.5.1 to 4.2.1.5.7) identified which have piping in the vicinity of the spent fuel pool which could result in a dilution of the spent fuel pool if they were to fail.

The fire protection, reactor makeup water, and demineralized water line stations, if damaged, could provide a source of spent fuel pool dilution. However, as discussed in Section 4.2.2.2, the physical arrangement of the area surrounding the spent fuel pool would limit the amount of water which ,

could flow into the spent fuel pool. l 4.2.1.5.9 Dilution Source and Flow Rate Summary l Based on the evaluation of potential spent fuel pool dilution soumes summarized above, the dilution sources of Table 4.2-1, below, were determined to be capable of providing a significant amount of non-borated water to the spent fuel pool. The potential for these sources to dilute the spent fuel pool boron concentration is evaluated in Section 4.2.2.

l 48 Attachment i safety Evaluauan NOC-AE-00178

l Table 4.2-1 Dilution Flow rate Summary l Approximate Flow Rate Source (gpm)

Boron Recycle System

-IIoldup Tank to cleanup subsystem 255 Reactor Makeup Water System

- 2" connection to return header 240 Demineralized Water System

- 2" connection to return header 190

- 1" makeup to spent resin sluice header 55 Fire Pmtection System

- Fire hose stations in spent fuel pool area 350 Spent Fuel Pool Demineralized 500 4.2.1.6 Boration Sources The normal source of borated water to the spent fuel pool is from the refueling water storage tank via the n fueling water purification pump. It is also possible to borate the spent fuel pool by the addition of dry boric acid directly to the spent fuel pool water. A discussion of each source follows.

4.2.1.6.1 Refueling Water Storage Tank The refueling water storage tank (RWST) connects to the spent fuel pool via the purification loop.

This connection is used to purify the RWST water when the purification loop is isolated from the spent fuel pool cooling system. Normally, this connection can supply borated water to the spent fuel pool via the refueling water purification pump to the inlet to the spent fuel pool cooling system purification loop. The refueling water purification pump is powered fram a non-safeguards bus power supply. It must be re-started manually following a loss of offsite power. The RWST is required by Technical Specifications to be kept at a minimum boron conce atration of 2800 ppm.

Alternatively, the low head safety injection pumps can be utilized to transf.'r RWST water to the 1 spent fuel pool. Temporary connections are made to the vent and dr$ Unes on one of the three ,

low head safety injection pumps. The hoses are routed to the spent tud (ool, the pump is operated on miniflow, the vent and drain valves are opened, and the water is tranch'ned to the pool.

4.2.1.6.2 Direct Addition of Boric Acid l If necessary, the boron concentration of the spent fuel pool can be increased by emptying drums of l dry L . c acid directly into the spent fuel pool. However, boric acid dissolves very slowly at room I Attachment i safety Evaluatmn NOC-AL-00178 I

w_____________________________________ -_ ______ __ -_ . _ _ _ _ _ _ _ _ .

l temperature and requires that the spent fuel pool cooling pumps be available for mixing throughout l the spent fuel pool water. (See section 4.2.2.1 for funher discussion on spent fuel pool mixing.) l 4.2.1.7 Administrative Controls The following administrative controls are in place to control the spent fuel pool boron concentration ,

and waterinventory:

1. Procedures are available to aid in the identification and termination of dilution events.

2. The procedures for loss of inventory (other than evaporation) specify that borated makeup sources be used as makeup sources. The procedures specify that non-borated sources only be used as a last reson.

3. In accordance with procedures, plant personnel perform rounds in the spent fuel pool enclosure once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The personnel making rounds to the spent fuel pool are trained to be aware of changes in the status of the spent fuel pool. They are instructed to check the temperature and level in the pool and conditions around the pool during plant i rounds. I 1

4. The spent fuel pool boron concentration is verified by sample analysis every seven days per l Technical Specification 3.9.13. I Administrative controls on the spent fuel pool boron concentration and water inventory ensure that l

the boron concentration is administratively controlled during both normal and accident situations. j The procedures ensure that the proper provisions, precautions and instructions are in place to ;

control the pool boron concentration and water inventory.

4.2.1.8 I.oss of Offsite Power Impact Of the dilution sources listed in Section 4.2.1.5.9, only the fire protection system is capable of providing non-borated water to the spent fuel pool during a loss of offsite power.

The loss of offsite power would affect the ability to respond to a dilution event. The spent fuel pool level instrumentation is not powered from emergency diesel generator-backed power supplies.

l The refueling water purification pump is not powered from a safeguards supply and would not be available to deliver borated water from the RWST. The RWST cannot be gravity-drained to the spent fuel pool through the refueling water purification pump, because the spent fuel pool minimum level is above the maximum level of the RWST. The low head safety injection pumps are powered from a safeguards bus, and can be used as an indirect source of RWST water to the spent fuel pool.

50 Attachnent I safety Evaluauon NOC-AE-00178

l 1

Finally, manual addition of dry boric acid to the pool could be used ifit became necessary to increase the spent fuel pool boron concentration during a loss of offsite power.

1 Currently, the spent fuel pool cooling pumps are not automatically restarted following a loss of {

offdte power and are supplied by power supplies backed by non-safeguards feeds from the diesel

{

generators. However, safeguards power supplies can be manually aligned to provide power to the j pumps, if necessary, to assure good mixing in the spent fuel pool.

j i

4.2.2 Spent Fuel Pool Dilution Evaluation l

4.2.2.1 Calculation of Boron Dilution Times and Volumes For the purposes of evaluating spent fuel pool dilution times and volumes, the total pool volume available for dilution, as described in section 4.2.1.1, is conservatively (low) based on 420,000 gallons.

Based on the criticality analysis (Section 4.1), the soluble boron concentration required to maintain the spent fuel pool boron concentration at La < 0.95, including uncertainties and bumup, with a 95% probability at a 95% confidence level (95/95) is 700 ppm. This concentration assumes no fuel mislaading accident.

For the purposes of calculating dilution times and volumes, the initial spent fuel pool boron concentration is assumed to be 2500 ppm. The evaluations are based on the spent fuel pool boron concentration being diluted from 2500 ppm to 700 ppm. To dilute the pool water volume of 420,000 gallons from 2500 ppm to 700 ppm would require 535,000 gallons of non-borated water, based on a feed-and-bleed operation (constant volume).

This analysis assumes thorough mixing of all the non-borated water added to the spent fuel pool with the contents of the spent fuel pool. Refer to Figure 3 of Reference 3. Based on the design flow of 2500 gpm per spent fuel pool cooling pump, the 420,000 gallon system volume is tumed over approximately every three hours with one pump running, which is the normal alignment. It is l

unlikely, with cooling flow and convection from the spent fuel decay heat, that thorough mixing I would not occur. However,if mixing was not adequate,it would be conceivable that a localized pocket of non-borated water could form somewhere in the spent fuel pool. This possibility is addressed by the calculation in Reference I which shows that the spent fuel rack Krr will be less than 1.0 on a 95/95 basis with the spent fuel pool filled with non-borated water. Thus, even if a pocket of non-borated water formed in the spent fuel pool, Krr would not exceed 1.0 anywhere m the pool.

51 Attachnrnt I safety Evaluat on NOC-AE-00178

l The time to dilute the spent fuel pool depends on the initial volume of the pool and the postulated rate of dilution. The dilution volumes and times for the dilution scenarios discussed in Sections 4.2.2.2 and 4.2.2.3 are calculated based on the equation developed in Section 3.1 of Reference 3.

4.2.2.2 Evaluation of Boron Dilution Events The potential spent fuel pool dilution events that could occur am evaluated below.

4.2.2.2.1 Dilution From BRS Recycle Holdup Tanks The contents of a BRS recycle holdup tank can be transfermd via the recycle evaporator feed pumps to the spent fuel pool via the purification loop piping. The flow path is isolated by one locked closed and two normally closed valves. This connection is a designated source of makeup water in a loss of spent fuel pool inventory event. Each of the two BRS recycle holdup tanks has a total volume of approximately 84,000 gallons. The water in the tanks can have a boron concentration

from 0 ppm to 3000 ppm, but is more typically in the range of 0 to 1500 ppm, consistent with the

reactor coolant system. Therefom, any amount of boron in the BRS recycle holdup tank water would reduce the dilution of the spent fuel pool msulting from the transfer of BRS recycle holdup

tank water to the spent fuel pool. To dilute the spent fuel pool from 2500 ppm to 700 ppm would require 535,000 gallons of unkorated water. The combined contents of the two BRS recycle holdup

, tanks (approximately 168,000 gallons) is less than the required dilution volume.

l The BRS recycle evaporator feed pumps can take suction from either of the two BRS recycle l holdup tanks. Manual valve manipulations are required to switch the pump suction to another tank.

l Thus, it is assumed for the purposes of this evaluation that only the contents of one BRS recycle holdup tank is available for a spent fuel pool dilution event. The 84,000 gallons of water contained in one BRS recycle holdup tank is less than the 535,000 gallons necessary to dilute the spent fuel

, pool fmm 2500 ppm to 700 ppm. The path from the recycle evaporator feed pumps to the spent fuel pool via the 3" connection to the spent fuel pool cooling return header can provide approximately 255 gpm. If the open path were left unattended, it would take 40 minutes to increase the spent fuel pool level from the low to high alarm setpoints, and 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months to provide the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron, if sufficient water were available.

4.2.2.2.2 Dilution From Reactor Makeup Water Storage Tank The reactor makeup water system consists of one reactor makeup water storage tank and two reactor makeup water pumps per Unit. The reactor makeup water storage tank contains approximately 150,000 gallons of non-borated, reactor grade water. However, makeup to the tank from the demineralized water system is automatically provided on a low level signal. Thus, with sufficient makeup, the tank contents could be sufficient to dilute the spent fuel pool from 2500 to 700 ppm.

52 Attachnent 1 Safety Evaluanon NOC-AE-00178 I

The contents of the reactor makeup water storage tank can be transferred via the reactor makeup water pemps to the spent fuel pool via the cooling loop return header. This connection is normally isolated from the reactor makeup water system by a closed manual valve. It can be used as the I normal makeup supply to the spent fuel pool and is a source of makeup water in case of a loss of spent fuel pool inventory event.

The path fmm the reactor makeup water pumps to the spent fuel pool via the 2" connection to the l spent fuel pool cooling retum header can provide approximately 240 gpm. If the open path were left unattended, it would take 43 minutes to increase the spent fuel poci level from the low to high i alarm setpoints, and 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months to provide the 535,000 gallons required 1 o dilute the pool from 2500 l to 700 ppm boron, if sufficient makeup water were available. l 4 2.2.2.3 Dilution From Demineralized Water System Trie demineralized water system includes a demineralized water storage tank and three transfer pumps. The non-borated contents of the demineralized water storage tank can be transferred directly to the spent fuel pool. The volume of the demineralized water storage tank (961,000 gallons) is greater than the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron.

The path from the demineralized water transfer pump to the spent fuel pool cooling return header via the 2" connection can provide approximately 190 gpm. If the flow path were left unattended, it would take 57 minutes to increase the spent fuel pool level from the low to high alarm setpoints, and 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months to provide the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron.

The path from the demineralized water pump to the spent fuel pool via spent resin sluice pump discharge header can provide approximately 55 gpm. If the flow path were left unattended, it would take three hours to increase the spent fuel pool level from the low to high alarm setpoints, and seven days to provide the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron.

4.2.2.2.4 Dilution from Fire Protection System The fire protection system draws from two 300,000 gallon tanks. The combined volume of the two tanks is greater than the 535,000 gallons required to dilute the spent fuel pool from 2500 to 700 ppm boron. The path from the fire water pump to the spent fuel pool via the two fire hose stations in the spent fuel pit area can provide approximately 350 gpm. If the hoses were left unattended,it would take 29 minutes to increase the spent fuel pool level from the low to high alarm setpoints,

! and 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months to provide the 535,000 gallons required to dilute the pool from 2500 to 700 ppm I boron.

53 Attachment I safety Evaluation NOC-AE-00178 l

1 l

L- _- - - - - - - - - - - - - - - i

{

I 4.2.2.2.5 Dilution Resulting From Seismic Events or Random Pipe Breaks A seismic event could cause piping ruptures in the vicinity of the spent fuel pool in piping that is not seismically qualified. For a seismic event with offsite power available, rupture of the reactor makeup and demineralized water supply lines to the spent fuel pit cooling loop will not result in a direct addition of unborated water to the spent fuel pool. If offsite power is not available, the reactor makeup and demineralized water systems would not operate and thus, there would be no dilution source.

In the event of a break in one of the fire protection hose station supply lines which are outside the l spent fuel pool enclosure but in the general area surrounding the spent fuel paol, water would l approach the spent fuel pool, but would be blocked by the 2" to 3" curb surrounding the pool. In l addition, there is an open stairwell and floor drains through which this water would drain to lower elevations of the fuel handling building. For the purposes of this analysis, it is conservatively assumed that a fire protection hose station line break floods the entire ama to a depth of three inches. This is conservative because of the openings to the cask pool, the cask connecting channel, the cask decontamination area, the new fuel storage, and the drop area openlug leading to bay doors in the building. Even before the water level mached three inches, the drop area would be capable of draining the full flow of any fire protection hose station supply line break.

l Once the water depth was equalized at three inches inside the curb (pool side) and outside curb (floor area), the driving head to force additional water into the enclosure would be significantly reduced. At that point, most of the flow from the pipe break would bypass the spent fuel pool enclosure, taking the path ofleast resistance around the enclosure to the drop area opening.

The total amount of water added to the spent fuel pool enclosure to raise the water level to three inches above the floor would be approximately 23,192 gallons assuming the spent fuel pool level was initially at the low level alann setpoint. This is much less than the 535,000 gallons required to dilute the spent fuel pool from 2500 ppm to 700 ppm. While a limited amount of flow through the enclosure would continue until the line break were isolated, a fire protection system line break of i this magnitude would be readily detected in the control room and break flow would be terminated long before enough water could enter the spent fuel pool enclosul. :o reduce the pool boron concentration to 700 ppm.

l Because of the limited flow into the spent fuel pool enclosum, and because a fire protection hose station supply line break would be terminated long before the spent fuel pool boron concentration would be reduced to 700 ppm, this event is not considered a credible event and is given no further consideration in this analysis.

l Attachment 1 54 l i

safety Evaluation !

NOC-AE-00178

l l 4.2.2.2.6 Dilution From Spent Fuel Pool Demineralized When the spent fuel pool demineralized is first placed in service after being recharged with fresh resin,it can initially remove boron from the water passing through it. In the worst case, assuming 3

75 ft of anion resin per demineralized,it is conservatively estimated that 58 ppm of boron could be

! removed from the spent fuel pool water before the resin becomes saturated. The deborating effect of the demineralizers is modeled by removing 250 gpm of borated water per train and mturning 250 gpm of deborated water per train until the ion exchange capacity is depleted. Since each demineralized normally utilizes a mixed bed of anion and cation resin, less boron would actually be removed before saturation. In addition, procedures are in place to flush a new resin bed with borated water prior to aligning it for service and verify boron saturation via sample analysis.

Because of the small amount of boron removed by the demineralizers, it is not considered a credible dilution source for the purposes of this evaluation.

4.2.2.3 Summary of Dilution Events Table 4.2-2 provides a summary of the dilution event results.

l Table 4.2-2 Summary of Dilution Events Approximate Flow Rate Dilution Time (hr)

Source (gpm)

Boron Recycle System 255 35 (limited source volume)

Reactor Makeup Water System

- 2" connection to return header 240 37 (limited source volume)

Demineralized Water System

- 2" connection to return header 190 49

- 1" makeup to spent resin sluice header 55 168 Fire Protection System

- fire hose stations in spent fuel pool area 350 25 Spent Fuel Pool Demineralizers 500 N/A (insufficient resin capacity) l

! 55 Attachnent I safety Evaluation NOC-AE-00178 l i

l The addition of unborated water from ta fire protection system hoses provides the shonest dilution l time. However, this requires that fire hoses be placed to deliver water from both fire protection supply tanks into the spent fuel pool, and that this off-normal operation continue for 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months .

l Operators would detect the unusually large drawdown from the fire protection system and would also notice the overflowing spent fuel pool during their rounds eveg 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The connection from the Boron Recycle System gives the next shortest dilution time, but the fluid is normally borated to some degree and the volume of one recycle holdup tank is less than tiac volume required for dilution from 2500 to 700 ppm. The next shortest dilution time and the limitirg scenario for normal operation is based on using the reactor makeup water connection to the sp:nt fuel pool cleanup subsystem for makeup when the process isolation valve is manually opened. This connection is the normal flowpath for unborated water authorized for use under norinal plant !

conditions by procedure. Although it is fed from a tank which has a capacity less thal the required volume to dilute the spent fuel pool from 2500 to 700 ppm, the tank receives automat,ic makeup from the demineralized water storage tank which has a volume gmater than the required dilution volume. 3 For the limiting credible scenario to result in the dilution of the spent fuel pool from 2500 ppm to 700 ppm, the addition of 535,000 gallons of water to the spent fuel pool over a period of 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months would have to go unnoticed. The first indication of such an event would be high level alarms in the control room from the. pool level instrumentation. If the high level alarms fail, it is reasonable to expect that the significant increase in pool level and eventual pool overflow that would result from a pool dilution event will ba madily detected by plant operators in time to take mitigative actions.

A pool overflow conditim would result in flooding of the fuel handling building sumps, and significant input flow rates would result in high sump level alarms. Although area radiation monitors are available, relatively clean spent fuel pool contents might not set off an alarm. In addition, it can be assumed that the operator rounds through the spent fuel pool area that occur once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months will detect the increase in the pool level even if the alarms fail and the flooding is not detected.

Funhermore, for any dilution scenario to successfully add 535,000 gallons of water to the spent fuel pool, plant operators would have to fail to question or investigate the continuous makeup of water to the reactor makeup water storage tank or fim protection system for the required time period, and fail to recognize that the need for 535,000 gallons of administrative makeup to the demineralized !

water storage tank was unusual.

4.2.3 Boron Dilution Conclusions As a result of the spent ftu gol boron dilution analysis, it is concluded that an unplanned or !

inadvertent event which would result in the dilution of the spent fuel pool boron concentration from 2500 ppm to 700 ppm is not a credible event. This conclusion is based on the following: ;

I I

56 Attachment 1 I safety Evaluatmn i NOC-AE-00178 l

l In order to dilute the spent fuel pool to the design Ng of 0.95, a substantial amount of water (535,000 gallons)is needed. To provide this volume, an operator would have to initiate the dilution flow, then abandon monitoring of pool level, ignore tagged valves, violate administrative pmcedures, and ignore spent fuel pool and building sump high level alarms.

l e

Since such a large water volume tumover is required, a spent fuel pool dilution event would be readily detected by plant personnel via alarms, flooding in the fuel handling building or by normal operator rounds through the spent fuel pool area.

It should be noted that this boron dilution evaluation was conducted by evaluating the time and water volumes requimd to dilute the spent fuel pool from 2500 ppm to 700 ppm. The 700 ppm end point was utilized to ensure that Kg for the spent fuel racks would remain less than or equal to 0.95. As part of the criticality analysis for the spent fuel racks (Reference 1), a calculation has been performed on a 95/95 basis to show that the spent fuel rack La remains less than 1.0 with non-borated water in the pool. Thus, even if the spent fuel pool were diluted to zero ppm, which would take significantly more water than evaluated above, the spent fuel would be expected to remain subtritical and the health and safety of the public would be assured.

4.3 Accident Analysis Accidents are postulated to occur at the most reactive rack condition, i.e., the pool is filled with unborated water. Section 4.3.1 covers accident conditions which do not result in an increase in La of the rack. Section 4 3.2 covers accident conditions which have the potential to result in an increase in La of the rack.

4.3.1 Postulated Conditions with No Increase in Rack En 4.3.1.1 Dropping a Fuel Assembly on Top of a Rack This accident analysis assumes an assembly is dropped onto the racks and falls to a horizontal position over the cells. The accident analysis further assumes that the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has sufficient w ater separating it from the active fuel height of nored assemblies to preclude neutronic interaction. This accident will not result in an increase in Lu of the rack.

4.3.1.2 Dropping a Fuel Assembly Between Rack Modules The design and installation of the spent fuel racks is such that it pmcludes the insertion of fuel j

assemblies in other than prescribed locations, including between the rack modules. j 57 Attachnent 1 i safety Evaluation I NOC-AE-00178 j l______._._. _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _

t 4.3.1.3 Dropping a Fuel Assembly Adjacent to a Rack Module The reactivity impact of dropping a fresh 4.95 w/o U235 assembly next to a rack module was evaluated. The Region 1 racks have two faces exposed to the pool walls. The gap between the rack and the pool wall at the Nonh face is inaccessible to the fuel handling machine. Likewise, the gap between the rack and the pool wall at the West face of the racks is also inaccessible. In both cases, the rack structure exists undemeath the fuel handling machine's funhermost extent of travel.

l Therefore, the dropping of an assembly adjacent to the outside of a Region I rack module is not

) considered credible.

Dropping an assembly on the outside of a Region 2 rack adjacent to a fresh, no IFBA,4.95 w/o U 235 fuel assembly is bounded by the misloaded fuel assembly accident discussed in Section 4.2.2 since placing a fuel assembly inside the racks next to other fuel assemblies will result in a higher La.

This is a change from the current analysis due to the fact that the revised criticality analysis does not include the Boraflex panels between the cells in the Region 2 racks.

4.3.2 Postulated Conditions with An Increase in Rack La Most accident conditions will not result in an increase in La of the rack. However, three accidents can be postulated for each storage configuration which can increase reactivity beyond the analyzed condition. The first postulated accident would be a change in the spent fuel pool water temperature outside the normal operating range. The second accident would be dropping an assembly into an already loaded cell. The third would be a misload of an assembly into a cell for which the restrictions on location, enrichment, or bumup are not satisfied.

I For an occurrence of the postulated accident conditions discussed below, the double contingency

{

principle of ANSI /ANS 8.1-1983 can be applied.This states that one is not required to assume two i unlikely, independent, concunent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the storage pool water (above the concentration required for normal conditions and reactivity equivalencing) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

4.3.2.1 Spent Fuel Pool Water Temperature Accident For the change in spent fuel pool water temperature accident, a temperature range of 32 F to 240 F is considered. Calculations were performed for the Regions 1 and 2 storage configurations to determine the reactivity change caused by a change in the spent fuel pool water temperature outside the normal range (50 F to 160 F). The results of these calculations are tabulated in Table 20 of Reference 1. In all cases sufficient reactivity margin is available to the 0.95 La limit to allow for

(

58 Attachnent 1 safety Evaluation l

j NOC-AE-00178 m____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ - _ . _ - - . _

l l

l temperatum accidents, without any additional change in the boron level as demonstrated in Table 20.

4.3.2.2 Drepping of a Fuel Assemblyinto an Alreadyleaded Cell Accident This accident analysis assumes an assembly is dropped onto the racks and lodges in a vertical position on top of an assembly in a cell. The axial leakage of the assembly in the cell would be reduced, however, the effect on the rack reactivity will be insignificant. The total axial leakage in both the upward and downward directions for the entire spent fuel arrayis worth about 0.003 delta Key, Thus, minimizing the upward-only leakage ofjust a single cell will not cause any significant inemase in rack reactivity. Also, the neutronic coupling between the dropped assembly and the assembly in the rack will be low due to several inches of assembly nozzle structure which would separate the active fuel regions. Therefore, this accident would be bounded by the mistoad accident.

4.3.2.3 Misloaded Assembly Accident For the misloaded assembly accident, calculations were performe:1 in each configuration to show the largest reactivity increase caused by a 4.95 w/o 17x17XL fuel assembly misplaced into a storage cell for which the restrictions on location, enrichment, or burnup are not satisfied. The results of these calculations are tabulated in Table 20 of Reference 1. The misloaded assembly accident can only occur during fuel handling operations in the spent fuel pool. In the postulated misload accident for the RCCA #1 and #2 checkerboard configurations, a 4.95 w/o assembly was placed in a location previously occupied by an RCCA. This results not only in a mistoad, but also the removal of an RCCA from the spent fuel pool.

The amount of soluble boron required to offset each of the postulated accidents was determined with P110ENIX-P calculations, where the impact of the reactivity equivalencing methodologies on the soluble boron is appropriately taken into account. The additional amount of soluble boron for accident conditions needed beyond the required boron for uncertainties and burnup is shown in Table 20 of Reference 1.

4.3.2.4 Withdrawal of an RCCA from a Region 2 RCCA #1 or #2 Checkerboard The withdrawal of an RCCA from a 1.40 w/o fuel assembly in an RCCA # 1 checkerboard or from a 1.65 w/o fuel assembly in a RCCA # 2 checkerboard has been evaluated and is bounded by the analysis discussed in Section 4.3.2.3, Misloaded Assembly Accident, above. In the postulated misload accident for the RCCA #1 and #2 checkerboard configurations, a 4.95 w/o assembly was placed in a location pmviously occupied by an RCCA. This results not only in a misload, but also bounds the removal of an RCCA from a 1.40 w/o fuel assembly in a RCCA # 1 checkerboard or from a 1.65 w/o fuel assembly in a RCCA # 2 checkerboard.

59 Attachneru !

safety Evaluation f

f NOC-AE-00178 1 !

- - - - - - - - - - - - - - - - - - - - - - --- ---- -- - - - - - --- -- - - - - - - - ----- --- a

The amount of soluble boron required to offset this postulated accident was determined with PHOENIX-P calculations, where the impact of the reactivity equivalencing methodologies on the soluble boron is appropriately taken into account. The additional amount of soluble boron for .

accident conditions needed beyond the required boron for uncertainties and burnup is shown in I Table 21 of Reference 1. J Note that the presence of the additional soluble boron, as required by Table 21 of Reference 1, negates the need for a positive restraint to prevent the withdrawal of an RCCA from either a RCCA

1 checkerboard pattern or from a RCCA # 2 checkerboard.

4.3.3 Boron Dilution Accidents Spent fuel pool soluble boron has been used in this criticality anan c to offset storage rack and fuel assembly tolerances, calculational uncenainties, uncenainty associated with reactivity equivalencing (burnup credit and IFBA credit) and the reactivity increase caused by postulated accident conditions. The total soluble boron concentration required to be maintained in the spent fuel pool is a summation of each of these components. Table 21 of Reference 1 summarizes the storage configurations and corresponding soluble boron credit requirements.

Based on the above discussion, should a spent fuel water temperature change accident or a fuel assembly misload accident occur in the Region 1 or Region 2 spent fuel racks, Kerr w ill be maintained 5 0.95 due to the presence of at least 2200 ppm of soluble boron in the spent fuel pool water. In addition, the boron dilution analysis described in Section 4.2 concluded that an unplanned or inadvertent event which would result in the dilution of the spent fuel pool boron concentration from 2500 ppm to 700 ppm is not a credible event.

l

5. CONCLUSIONS This report shows that the acceptance criteria for criticality is met for the Regions 1 and 2 spent fuel l racks for the storage of Westinghouse 17x17XL fuel assemblies under both normal and accident conditions with soluble boron credit and the appropriate storage configurations and enrichment i limits.

The proposed changes to the UFSAR and Technical Specifications, as described above, are acceptable because the proposed changes to the spent fuel storage pool criticality analyses and rack utilization do not pose a significant increase in hazard or involve a significant reduction in a margin of safety. The South Texas Project requests approval of the proposed changes.

60 Attachnu:nt I safety Evaluation NOC-AE-00178

6. REFERENCES

1. Wes*inghouse letter: ST-UB-NOC-1803, dated May 22,1998, " Final Report for Soluble Boron Criticality Report"(Attachment 2 herein)

2. Westinghouse letter: ST-UB-NOC-1790, dated March 10,1998, " Final Spent Fuel Pool Boron Dilution Analysis Report" (Attachment 3 herein)

3. Westinghouse Spent Fuel Rack Criticality Analysis Methodology, WCAP-14416-NP-A Revision 1, November 1996

)

l I

l Attachnent I safety Evaluation NOC-AE4X)l78 l

ATTACHMENT 2 South Texas Units 1 and 2 Spent Fuel Rack Criticality Analyses with Credit for Soluble Boron

)

Attachnent 2 NOC-AE-00178 ,

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Westinghouse Energy Systems J8",,, p% im Electric Company May 22,1998 98TG-G-0027 ST-UB-NOC-1803 Mr. D. F. Hoppes Supervisor, Nuclear Fuel STP Nuclear Operating Company South Texas Piroject Electric Generating Station P. O. Box 289 Wadsworth, TX 77483 STP NUCLEAR OPERATING COMPANY SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION UNIT 1 AND 2 FINAL REPORT FOR SOLUBLE BORON CRITICALITY REPORT

Dear Mr. Hoppes:

Attached is the final report entitled " South Texas Units 1 and 2 Spent Fuel Rack Criticality Analysis with Credit for Soluble Boron." This report documents the criticality efTorts performed in response to STPNOC Purchase Order ST-400700, Supplement 26.

Please note that comments received from STPNOC have been incorporated per our discussion. If you have questions or need additional information, please feel free to contact me.

Very truly yours, NV James P. Sechrist Project Engineer Fuel Marketing & Projects cc: D. Gore - STPNOC I R. Dunn - STPNOC L. L. Snell - STPNOC A. Lehing - W Houston Field Sales J. Wyble - W NSD Projects ECE 517A t ,

ML20236L617

Text

SUMMARY

Dear Mr. Hoppes:

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