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Non-proprietary Rev 2 to HI-971760, Licensing Rept for Expanding Storage Capacity in Harris Spent Fuel Pools 'C' & 'D'.
	ML18016A771
Person / Time
Site:	Harris
Issue date:	05/26/1998
From:	Pellet S; HOLTEC INTERNATIONAL
To:	;
Shared Package
ML18016A769	List: ML18016A768; ML18016A770; ML18016A771;
References
	HI-971760, HI-971760-R02, HI-971760-R2, NUDOCS 9812290061
	Download: ML18016A771 (307)
	v • d • e

Enclosure 7 to Serial: HNP-98-188 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT SPENT FUEL STORAGE LICENSING REPORT FOR EXPANDING STORAGE CAPACITY IN HARRIS SPENT FUEL POOLS 'C'ND VERSION) 'D'NON-PROPRIETARY 98i22'7006% 98i22S PDR ADOCK 05000400 P PDR

I IRSNI HOLTEC I N T E R N AT 0 I N AL Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (609) 797-0900 Fax (609) 797-0909 LICENSING REPORT for EXPANDING STORAGE CAPACITY HAjRRIS SPENT FUEL POOLS C AI'6) D HOLTEC IlVIXRNATIONAL 555 LINCOLNDRY WEST MARLTON, NJ 08053 HOLTEC PROJECT NO. 70324 HOLTEC REPORT HI-971760 REPORT CATEGORY: A, REPORT CLASS: SALTY RELATED CLIENT CONTRACT NO. XTA7000024

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REVIEW AND CERTIFICATION LOG FOR MULTIPLEAUTHORS Sheet I of2 REPORT NUMBER: PROJECT NUMBER:

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~ ~ of thc governing This document conforms to the requirements of the Design Spccilication and the applicable sections codes.

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Note: Signatures and printed names are rcquircd in thc review block, A revision of this document willbc ordcrcd by thc Project Manager and carried out ifany of its contents is materially affcctcd during evolution of this project.

Thc determination as to the nccd for revision will bc made by the Project Manager with input from others, as dccmcd ncccssary by him, Must bc Project Manager or his dcsignce.

Distribution: C: Client M: Dcsignatcd Manufacturer F: Florida Office THE AEVISION CONTROL OF THIS DOCUMENT IS BY A "

SUMMARY

OF REVISIONS LOG" PLACED BEFORE THE TEXT OF THE REPORT. Form: RCL.02

AND CERTIFICATIONLOG FOR MULTIPLEAUTHORS Sheet 2 of 2 NUMBER'EVIEW REPORT PROJECT NUMBER'EVISION 0 REVISION 1 REVISION 2 REVISION 3 Document Portion Author Reviewer Author Reviewer Author Reviewer Author Reviewer Chapter 9 -zz-'F -ZC- 4+ s p st -Ec-)f ~< <zo Chapter 190 .S. I 2. S.VI> 8 f 5 P.

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QA APPROVAL PROJECT MANAGER i

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of thc Design Specification and the applicablc sections of thc governing codes.

This document conforms to the rcquircmcnts Note: Signatures and printed names arc rcquircd in the rcvicw block.

A revision of this document will bc ordered by the Project Manager and carried out ifany of its contents is materially affcctcd during evolution of this project.

Thc determination as to the need for revision will be made by thc Project Manager with input from others, as dccmcd ncccssary by him.

Must bc Project Manager or his dcsigncc.

Distribution: C: Client M: Dcsignatcd Manufacturer F: Florida Office THE REVISION CONTROL OF THIS DOCUMENT IS BY A "

SUMMARY

"OF REVISIONS LOG" PLACED BEFORE THE TEXT OF THE REPORT. Form: RCL.O2

TABLE OF CONTENTS

1.0 INTRODUCTION

1.1 geee~ec ~ . 1-6 2.0 OVERVIEW OF THE PROPOSED CAPACITY EXPANSION .......... 2-1

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2.1 Jgg~uc~i 2-1 2.2 alDe i 2-2 2.3 2.4 ua'ra e e ce d ta r ra rd 2-11 2-11 2.5 ,

2.6 a cat' . 2-12 2.6.1 c i n ectve...........,,... . 2-12 2.6.2 fte 'P ac l 2-13 2.6.3 na f e ack dule . . 2-15 3.0 MATERIAL,HEAVYLOAD, AND CONSTRUCTION CONSIDERATIONS .. 3-1 3.1 @~due'QQQ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ .. 3-1 3.2 3-1 3.3 i ateri e rber 3-1 3.4 ack a ati wit .. 3-3 3.5 ea d 'de ~ ~ ~ ~ ~ 3 3 3.6 ~e at'~emcee 3-8 4.0 4.1 4,2 4.2.1 4.2.1.1 4.2.1.2 e'

u PWR Fuel Results BWR Fuel Results f

e CRITICALITYSAFETY EVALUATION o.e.

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4-1..'-1 4-3 4-3

.. 4-4 4.3 .. 4-5 4,3,1 efe e .. 4-5 4.3.2 e ce e e l an ae .. 4-6 4,4 'c l eth 1 .. 4-7 4.4.1 ee e Ic .. 4-7 4.4.2 I u a a ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ . 4-8 4.4.2.1 PWR Fuel Burnup Calculations . .. 4-8

'.4.2.2 BWR Fuel Burnup Calculations and Comparison to Vendor Calculations .. 4-8 4.4.3 e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 4-9 4.4.3.1 PWR Fuel Axial Burnup Distribution ..............,....,........ .. 4-9 4.4.3.2 BWR Fuel Axial Burnup Distribution . . 4-10 4,4 4 v . 4-10 4.5 . 4-11 4.5.1 i e e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 4-11 Holtec International Holtec Report HI-971760

TABLE OF CONTENTS 4.5.2 certain ie ue t Rac nufac le ce 4-11 4.5.2.1 Boron Loading Tolerances ., 4-12 4.5.2.2 Boral Width Tolerance . 4-12 4.5.2.3 Tolerance in Cell Lattice Spacing and Cell Box Inner Dimension ......... 4-12 4.5.2.4 4.5.2.5 4.6 4.6.1 Stainless Steel Thickness Tolerance Fuel Enrichment and Density Tolerances WR t

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........4-13

. 4-12 4-12 4-13 4.6.2 cert fc e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4-14 4.6.2.1 Boron Loading Variation . 4-14 4.6.2.2 Boral Width Tolerance Variation... ~ ~ ~ ~ 4-14 4.6.2.3 Tolerance in Cell Lattice Pitch and Inner Box Dimension ~ ~ ~ 4-14 4.6.2.4 Stainless Steel Thickness Tolerances .. 4-15 4.6.2.5 Fuel Enrichment and Density Variation ~..... .. ~..... ~ ~ ~ ~ ~ 4-15 4.6.2.6 Zirconium Flow Channel........ ~ ~ 4-15 4.7 n rma CC d ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ 4-15 4.7.1 e e n W n' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4-16

'l 4.7.2 e e 4-16 4.7.3 a c v ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ 4-17 4.7.4 norma ati n WR WR l em l 4-17 0 ~

4.7.5 c ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4-18 4.8 QCfeee~cg . 4-19 5.0 THERMAL-HYDRAULICCONSIDERATIONS 5-1 5.1 Lntmd JQQggf QQ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

5-1" 5.2 5-2 5.3 5-3 5.4 5-4 5.4.1 ~ ~

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~ 5-8 5.5 5-8 5.6 R [ggg o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-11 6.0 STRUCTURAL/SEISMIC CONSIDERATIONS.... '........... 6-1 6.1 QlfK~duc LQD ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 1 6.2 verview ack c ral i M I 6-1 6.2.1 e ~ ~ ~ ~ . 6-2 6.3 . 6-5 6.3.1 ~IW ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 6-5 6.4 tc 6-6 6.5 d ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-7 6.5.1 e e ue 6-7 6.5.1.1 .. 6-7 6.5.1.2 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 6-9 6.5.2 luid 6-10 Holtec International n Holtec Report HI-971760

TABLE OF CONTENTS 6.5.2.1 ul- lin P 6-11 6.5.3 ti e e e all ~ ~ ~ ~ 6-12 6.5.4 e cie ~ ~ ~ ~ 6-13 6.5.5 ve ~ ~ ~ ~ 6-14 6.6 C V ~ ~ ~ 6-15 6.6.1 ie 'c re Acce t n ~ ~ ~ ~ 6-15 6.6.2 va in 6-16 6.6.3 e act r ............. 6-18 6.6.4 ad a e Racks. 6-20 6.7 Para e 'c n ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-21 6.8 Time 6-23 6.8.1 ac e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ 6-23 6.8.2 ede e . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-25 6.8.3 ede c ce ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ d ~ ~ ~ ~ ~ ~ ~ 6-26 6.8.4 6-27 6.8.4.1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ d ~ d ~ ~ ~ ~ ~ 6-27 6.8.4.2 c W ll act 6-29 6,8.4.3 Fuel t ell W act Load 6-29 6.9 ack c v uti 6-30 6.9.1 Rac e

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~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ d 6-30 6-31 6.9.1.1 c e .

6.9.2 e t 6-32 6.9.3 ca I ac 6-33 6.9.4 e t Rc ati ue a 6-34 6,9.5 ~wld . 6 6.9.6 e 6-39 6.9.7 eve 6-40 6,10 dr a c Wal 6-41 6.11 Qggg~ujg~ 6-42 6.12 Ea ~ee nce ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-43 7.0 FUEL HANDLINGAND CONSTRUCTION ACCIDENTS .. ~ . ~......... 7-1 7.1 Q1~WCLQQ . . 7-1 7.2 c e ad'n c' 7-1 7.3 c . 7-3 7.4 at e c e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ . 7Q 7.5 ue . 7-5 7.5.1 V ~ ~ ~ ~ ~ ~ . 7-5 7.5.2 V . 7-5 7.6 ce . 7-6 7.7 . 7-7 7.8 k~efe e~ne . 7-8 Holtec International Holtec Report HI-971760

TABLE OF CONTENTS 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS .......... 8-1 8.1 Lnimduaiiau ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ I \~ ~ ~ ~ ~ ~ 8-1 8.2 8-1 8.3 8-2 8.3.1 c d 8-2 8.3.2 I c 8-2 8.3.3 l Ill 8-3 8.4 8.4.1 'e e 8-3 8-3 8.4.2 . 8-4 8.4.3 8-6 8.5 8-7 8.6 P~l ~ne ....,.... 8-8 8.7 Qgggcygjggg ....... 8-9 8.8 ~eence 8-10 9.0 RADIOLOGICALEVALUATION 9-1 9.1 SalldR d L.... 9-1 9.2 aeu e ~ ~ ~ ~ ~ ~ ~ ~ 9-1 9.3 e n e e 9-1 9.4 c ki 9-2 10.0 INSTALLATION 10-1 10.1 Inimdudiaa ... 10-1 10.2 10-4" 10.3 10-5 10.4 ~

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~ ~ ~ ~ ~ 10-6 10.4.1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ 10-6 10.4.2 PQQQQQQQ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 10-6 10.5 10-6 10.6 10-8 10.6,1 /sf nf Nr wr ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 10-8 10.6.2 10-8 10.6.3 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 10-8 10.7 10-9 Holtec International 1V Holtec Report HI-971760

TABLE OF CONTENTS 11.0 ENVIRONMENTALCOST/BENEFIT ASSESSMENT ......... . 11-1 11;1 . 11-1 11.2 . 11-1 11.3 . 11-1 11.3.1 . 11-6 11.4 . 11-6 11.5 . 11-7 11.6 . 11-8 11.7 Ra ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 11-9 Holtec International Holtec Report HI-971760

TABLE OF CONTENTS Key Haris Plant Information 1-7 2.1.1 Geometric and Physical Data for Pool C Rack Modules .. 2-17 and 2-18 2.1.2 Geometric and Physical Data for Pool D Rack Modules .. 2-19 2.5.1 Module Data for Harris Spent Fuel Racks 2-20 3.3.1 Boral Experience List - PWRs ~.....

~ .... 3-9 and 3-10 3.3.2 Boral Experience List - BWRs ~ ~ ~ ~ ~ ~ ~ 3 1 1 and 3-12 3.3.3 1100 Alloy Aluminum Physical Characteristics . 3-13 3.3.4 Chemical Composition - Aluminum (1100 Alloy) . ... 3-14 3.3.5 Chemical Composition and Physical Properties of Boron Carbide 3-15 3.5.1 Heavy Load Handling Compliance Matrix (NUREG-061) 3-16 4.2.1 Summary of Criticality Safety Calculations for PWR Fuel Racks ... 4-20 4.2.2 Summary of Criticality Safety Calculations for BWR Fuel Racks . 4-21 4.3.1 PWR Fuel Characteristics . 4-22 4.3.2 BWR Fuel Characteristics .... ~... ~ . ~ 4-23 4.4.1 Reactivity Allowance for Uncertainty in Burnup Calculations and the Effect of Axial Burnup Distributions for PWR Fuel ........ ~....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

4.5.1 Comparison of MCNP-4A and CASMO-3 Calculations . 4-25, 4.7.1 Reactivity Effects of Temperature and Void .. 4-26" 5.1.1 Partial Listing of Rerack Applications Using Similar Methods of Thermal-Hydraulic Analysis .. 5-12 and 5-13 5.2.1 Decay Periods for a Bounding Pools C and D Storage Configuration ... 5-14 5.2.2 Fuel Assemblies Input Data for Decay Heat Evaluation,, .. 5-15 5.2.3 Bounding Decay Heat Input from Stored Fuel in Pools C and D ... . . 5-16

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5.4.1 Bounding Fuel Assemblies Hydraulic Flow Resistance Parameters . 5-17 5.5.1 Pools C and D Dimensional Data . ... 5-18 5.5.2 Bulk and Local Temperature Summary................... ~... ... 5-19 6.2.1 Partial Listing of Fuel Rack Applications Using DYNARACK ......... 6-45 and 6-46 6.3.1 Rack Material Data (200'F) (ASME - Section II, Part D) 6-47 6.4.1 Time-History Statistical Correlation Results ~ ~ ~ ~ ~ ~ ~ o 648 6.5.1 Degrees-of-Freedom ....... ~ ~ ~ ~ ~ ~ ~ .. 6-49 6.5.2 (MR216) Numbering System for Gap Elements and Friction Elements in the Pool D Campaign I Model ... 6-50 and 6-51 6.9.1 Comparison of Bounding Calculated Loads/Stresses vs. Code Allowables at Impact and Weld Locations ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 52 Holtec International vi Holtec Report HI-971760

TABLE OF CONTENTS 7.1 Impact Event Data . . 7-9 7.2 Material Definition . . 7-10 8.5.1 Bending and Shear Strength Evaluation ........ ... 8-11 9.4.1 Preliminary Estimate of Person-REM Dose During Reracking 9-4 Holtec International Vll Holtec Report HI-971760

TABLE OF CONTENTS goyim.'~

1.1 Harris Fuel Handling Building Plan Layout . 1-8 1.2 Storage Configuration for Pool C . 1-9 1.3 Storage Configuration for Pool D 1-10 2.1.1 Pictorial View of Typical Harris Rack Structure 2-21 2.6.1 Seam Welded Precision Formed Channels . ~ ~ 2-22 2.6.2 Three PWR Cells in Elevation View . 2-23 2.6.3 Three BWR Cells in Elevation View......... 2-24 2.6.4 Composite Box Assembly 2-25 2.6.5 Typical Array of Storage Cells .. 2-26 2.6.6 Support Pedestal for Holtec PWR Rack ... .. ~ 2-27 2.6.7 Support Pedestal for Holtec BWR Rack 2-28 4.2.1 Burnup Versus Enrichment for PWR Fuel . 4-27 4.3.1 a Two Dimensional Representation of the Calculational Model Used for the PWR Storage Rack Analysis .........., . 4-28 4.3.2 a Two Dimensional Representation of the Calculational Model Used for the BWR Storage Rack Analysis . 4-29 5.3.1 C and D Pools Minimum Total Cooling System Requirements Curve at 137' Bulk Pool Temperature .... 5-20" 5.4.1 Hams C and D Pools Physical Configuration . 5-21 5.5.1 Plan View of the Harris Pools C and D and CFD Model .. 5-22 5.5.2 Perspective View of the Harris Pools C and D and CFD Model . 5-23 5.5.3 Peak Local Water Temperature in the Rack Cells 5-24 5.5.4 Pools Interconnecting Channel Flow Velocity Vectors Elevation View Plot 5-25 5.5.5 Pool Cooling Inlet/Outlet Piping Region Flow Velocity Vectors Plot ..... 5-26 6.3.1 Phased Storage Configuration for Pool C 6-53 6.3.2 Phased Storage Configuration for Pool D 6-54 6.4.1 SFP Time History Accelerogram (X-Direction, OBE) . 6-55 6.4.2 SFP Time History Accelerogram (Y-Direction, OBE) . 6-56 6.4.3 SFP Time History Accelerogram (Z-Direction, OBE) . 6-57 6.4.4 SFP Time History Accelerogram (X-Direction, SSE) 6-58 6.4.5 SFP Time History Accelerogram (Y-Direction, SSE) 6-59 6.4.6 SFP Time History Accelerogram (Z-Direction, SSE) 6-60 6.5.1 Schematic of the Dynamic Model for DYNARACK . 6-61 6.5.2 Fuel-to-Rack Impact Springs at Level of Rattling Mass 6-62 6.5.3 Two Dimensional View of the Spring-Mass Simulation 6-63 6.5,4 Rack Degrees-of-Freedom and Bending Springs 6-64 6.5.5 Rack-to-Rack Impact Springs ............ ~....... 6-65 Holtec International viii Holtec Report HI-971760

TABLE OF CONTENTS 6.5.6 Bottom Rack Impact Springs - Campaign I - Pool C 6-66 6.5.7 Top Rack Impact Springs - Campaign I - Pool C ".', 6-67 6.5.8 Bottom Rack Impact Springs - Campaigns II and III - Pool C 6-68 6.5.9 Top Rack Impact Springs - Campaigns II and III- Pool C 6-69 6.5.10 Bottom Rack Impact Springs - Campaign I - Pool D .. 6-70 6.5.11 Top Rack Impact Springs - Campaign I - Pool D . 6-71 6.5.12 Bottom Rack Impact Springs - Campaign II - Pool D . 6-72 6.5,13 Top Rack Impact Springs - Campaign II - Pool D.... .. 6-73 6.8,1 Vertical Pedestal Time History Loading Plot ........,... . 6-74 6,9.1 Rack Fatigue Finite Element Model ............... . 6-75 7.2.1 Shallow Drop on a Peripheral Cell 7-11 7.2.2 Plan View of Impactor and Impact Zone (Shallow Drop Event) 7-12 7.2.3 Deep Drop on a Support Leg Location . 7-13 7.2.4 Deep Drop on a Center Cell Location 7-14 7.3.1 Heaviest Rack Drop 7-15 7.5.1 Shallow Drop: Finite Element Model Detail - Impacted Region 7-16 7.5.2 Maximum Cell Deformation for Shallow Drop on Exterior Cell 7-17 7.5.3 Shallow Drop: Maximum Deformation - Impacted Region Plan 7-18 7.5.4 Plan View of Deep Drop Scenarios 7-19 7.5.5 Maximum Baseplate Deformation &om Deep Drop Scenario 7-20 7.6.1 Gate Drop Finite-Element Model . 7-21 7.6.2 Gate Drop Finite-Element Model, Detail of Impacted Region .. 7-22 7.6.3 Gate Drop Finite-Element Model, Detail of Impacted region (Plan) 7-23 7.6.4 Gate Drop Maximum Deformation 7 7.6.5 Gate Drop Maximum Deformation, Impacted Region Plan ........ 7-25 .

8.2.1 Pool Structure Dimensions .........,................ 8-12 8.4.1 Fuel Handling Building Finite Element Model 8-13 Holtec International 1X Holtec Report HI-971760

INTRODUCTION The Harris Nuclear Plant (HNP) is a single unit pressurized water reactor installation located in the extreme southwest corner of Wake County, North Carolina, and the southeast corner of Chatham County, North Carolina. The HNP installation is owned by the Carolina Power &

Light Company (CP&L) and the North Carolina Eastern Municipal Power Agency (NCEMPA), located in Raleigh, North Carolina. CP&L has the overall responsibility to ensure that plant operations are performed without undue risk to the health and safety of the public. Table 1.1 contains key overview data for HNP's PWR Unit.

HNP was originally named Shearon Harris Nuclear Power Plant (SHNPP) and was initially designed as a four unit nuclear reactor site, of which only Unit 1 was completed. The Fuel Handling Building (FHB), however, was constructed to service all four Units as originally envisioned. During initial licensing, the possibility of transhipment from other Units was recognized and consequently the Spent Fuel Pools were licensed to store both PWR and BWR fuel. Transhipped fuel from the Robinson and Brunswick plants is already in stored in pools A and B.

The FHB is a long narrow structure intended to be sandwiched between the nuclear plants, in order to service all four Units. Each end of the building contains two large pools, with the South end pools (A and B) originally intended to service Units 1 and 4 and the North end pools (C and D) designed to service Units 2 and 3. The layout of the FHB and pools in relationship with Unit 1 is shown in Figure 1~ 1. The two pools in each end of the building were originally designated as the "New Fuel Pool"for the smaller of the two pools and the "Spent Fuel Pool" for the larger pool. These four pools have since been re-designated as pools A, B, C, and D, where pools A and D represent the smaller pools. All four pools are interconnected through "gated" passages and are capable of storing spent fuel.

Holtec International Holtec Report HI-971760

Pools A and B, located at the South end of the building, have already been racked and are nearly full. Pool A contains six Region 1 type (6 x 10 cell) PWR racks and three (11 x 11 cell) BWR racks for a total storage capacity of 723 assemblies. Pool A has been, and will continue to be, used to store fresh (unburned) fuel, recently discharged Harris fuel and transshipped fuel. Pool B contains six (7 x 10 cell), five (6 x 10 cell), and one (6 x 8 cell)

PWR Region 1 style racks. Pool B also currently contains seventeen (11 x 11 cell) BWR racks, twelve of which have been supplied by Holtec International. Pool B is licensed to store one more (11 x 11 cell) Holtec BWR rack which would increase the total pool storage capacity to 2946 assemblies. The combined pool A and B licensed storage capacity is 3669 assemblies.

Projected operation of the Harris Unit and transhipments from the Robinson and Brunswick Units will continue to demand incremental increases in spent fuel storage capacity. The Carolina Power & Light Company, HNP's principal owner and operator, has entered into a contract with Holtec International of Marlton, N.J. to design maximum density spent fuel storage racks for pools C and D. Under the proposed capacity expansion, fuel storage racks

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This process is consistent with the

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will be installed in campaign phases on an as needed basis.

incremental capacity expansions already performed in pool B.

Pools C and D are unused and are located in the north end of the Harris Fuel Handling Building. Pool C will provide storage for both PWR and BWR fuel. This pool has nominal dimensions of 27 feet wide, by 50 feet long and at maximum storage density can accommodate 927 PWR and 2763 BWR assemblies. Pool D will contain only PWR fuel and with nominal dimensions of 20 feet wide by 32 feet long can accommodate 1025 maximum density storage cells. Proposed storage configurations for pools C and D are provided in Figures 1.2 and 1.3, respectively.

The configuration shown in Figure 1.2 represents the mixture of PWR and BWR storage which will accommodate future storage needs based on the best information currently available. To provide the greatest flexibilityin mixture of fuel types, the storage racks were Holtec International 1-2 Holtec Report HI-971760

The dimensions of the 9x9 PWR storage rack are nearly

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sized to allow interchangeability.

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identical to those of the 13x13 BWR rack. Therefore, configurations other than those shown in Figure 1.2 are possible by replacing one rack type by the other. The complete geometric fungibility between the 9x9 PWR and 13x13 BWR rack modules affords CPkL the latitude to alter the mix between PWR and BWR storage as the precise need for the two types of spent nuclear fuel storage become known. Interchanging of PWR and BWR modules would be performed after appropriate safety evaluations supported by reanalysis of the criticality, thermal-hydraulic, and structural analyses are successfully conducted to support such a substitution under Subpart 50.59.

The new Holtec racks are free-standing and self-supporting. The principal construction materials for the new racks are SA240-Type 304L stainless steel sheet and plate stock, and SA564-630 (precipitation hardened stainless steel) for the adjustable support spindles. The only non-stainless material utilized in the rack is the neutron absorber material which is a boron carbide and aluminum-composite sandwich available under the patented product name Boral .

The new Holtec racks are designed to the stress limits of, and analyzed in accordance with, Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV)

Code. The material procurement, analysis, and fabrication of the rack modules conform to 10CFR50 Appendix B requirements.

The rack design and analysis methodologies employed in the Harris storage capacity expansion are a direct evolution of previous rerack license applications. This Licensing Report documents the design and analyses performed to demonstrate that the new Holtec racks meet all governing requirements of the applicable codes and standards, in particular, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", USNRC, 1978 and the 1979 Addendum thereto [1.0. 1].

Holtec International 1-3 Holtec Report HI-971760

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Sections 2 and 3

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of this report provide an abstract of the design and material information on the new racks. ~

The criticality safety analysis requires that the neutron multiplication factor for the stored fuel array be bounded by the USNRC k,rf limit of 0.95 under assumptions of 95% probability and 95% confidence. The criticality safety analysis provided in Section 4 sets the requirements on the Boral panel length and the areal B-10 density for the new high density racks.

Thermal-hydraulic consideration requires that fuel cladding will not fail due to excessive thermal stress, and that the steady state pool bulk temperature will remain within the limits prescribed for the spent fuel pool to satisfy the pool structural strength, operational, and regulatory requirements. The thermal-hydraulic analyses carried out in support of this storage expansion effort are described in Section 5.

Demonstrations of seismic and structural adequacy are presented in Section 6.0. The analysis shows that the primary stresses in the rack module structure will remain below the allowable stresses of the ASME B&PV Code (Subsection NF) [1.0.2]. The structural qualification also .."

includes analytical demonstration that the subcriticality of the stored fuel will be maintained under all postulated accident scenarios in the Harris Final Safety Analysis Report (FSAR).

The structural consequences of these postulated accidents are evaluated and presented in Section 7 of this report.

Section 8 contains the structural analysis to demonstrate the adequacy of the spent fuel pool reinforced concrete structure. A synopsis of the geometry of the Harris reinforced concrete structure is also presented in Section 8.

The radiological considerations are documented in Section 9.0. Sections 10, and 11 discuss the salient considerations in the installation of the new racks, and a cost/benefit and Holtec International Holtec Report HI-971760

environmental assessment to establish the prudence of CPSs decision to exercise the wet

~ ~

storage expansion option, respectively. ~

All computer programs utilized to perform the analyses documented in this licensing report are benchmarked and verified. These programs have been utilized by Holtec International in numerous rerack applications over the past decade.

The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety in respect to all considerations of safety specified in the OT Position Paper, namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanical integrity.

Holtec International 1-5 Holtec Report HI-971760

[1.0.1] USNRC, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978, and Addendum dated January 18, 1979.

ASME Boiler & Pressure Vessel Code,Section III, Subsection NF, and Appendices (1995).

Holtec International Holtee Report HI-971760

Table 1.1 KEY HARRIS PLANT INFORMATION ITEM DATA Docket Number 50-400 Capacity, MWe 940 Applied to NRC 9-4-71 Construction Permit 1-27-78 Commercial Operation 1986 Present Capacity Pool A 723 Pool B 2946 TOTAL 3669 Holtec International 1-7 Holtec Report HI-971760

OF UNIT 4 CONTAINMENT FUEL TRANSFER (NOT CONSTRUCTED )

CANALS

~ NORTH .

CO C4 POOL B 21 I OF UNIT 1 CONTAINMENT OF UNIT 3 CONTAINMENT FUEL TRANSFER (NOT CONSTRUCTED ) EQUIPMENT CANALS HATCH N

CASK LOADING 0 L POOL POOL D l

COLUl(N LINE IDENTIFIERS

( NOT CONSTRUCTED )

FIGURE 1.1; HARRIS FUEL HANDLING BUILDING PLAN LAYOUT

~ ~

1-8

coo" 2.50" .mp SUPPORT LEG 2.50" OO 0 OO 0 4.06" 4.06" OO OO 0 OO 0 OO OO OO OO 324" 4.ae" 4.06" I

.625" MIN.(TYP) 5 Ol 2.50" TOTAL CELL COUNT:

927 CELLS - PER 2763 CELLS " BlR FIGORB 1.2; STOMB CONFIGlJRATION FOR POOL C HI-971760 1+

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2.0 OVERVIEW OF THE PROPOSED CAPACITY EXPANSION 2.1 Intl:mhZQaiI In its currently proposed fully implemented configuration, Pool C will contain eleven PWR racks and nineteen BWR racks. 'Pool D will contain twelve PWR racks. All storage racks arrays will consist of free-standing modules, made from Type 304L austenitic stainless steel containing prismatic storage cells interconnected through longitudinal welds. A panel of Boral cermet containing a high areal loading of the B-10 isotope provides appropriate neutron attenuation between adjacent storage cells. Figure 2.1.1 provides a schematic of the typical Region 2 storage module proposed for Harris. Data on the cross sectional dimensions, gross weight and cell count for each rack module in pools C and D are presented in Tables 2.1.1 and 2.1.2, respectively.

Each new rack module is supported by four legs which are remotely adjustable. Thus, the racks can be made vertical and the top of the racks can easily be made co-planar with each other. The rack module support legs are engineered to accommodate undulations in the fuel pool and cask pit floor flatness.

A bearing pad interposed between the rack pedestals and the pool liner serves to diffuse the dead load of the loaded racks into the reinforced concrete structure of the pool slab.

Holtec International 2-1 Holtec Report HI-971760

The overall design of the Harris racks is similar to those presently in service in the spent fuel pools at many other nuclear plants, among them Zion Nuclear Station of the Commonwealth Edison Company, Donald C. Cook of American Electric Power, and Connecticut Yankee of Northeast Utilities. Altogether, over 50 thousand storage cells of the Harris design have been provided by Holtec International to various nuclear plants around the world.

2.2 The key design criteria for the new Harris spent fuel racks are set forth in the classical USNRC memorandum entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", April 14, 1978 as modified by amendment dated January 18, 1979. The individual sections of this report expound on the specific design bases derived from the above-mentioned "OT Position Paper". Nevertheless, a brief summary of the design bases for the Harris racks are summarized in the following:

a. Disposition: All new rack modules are required to be free-standing.

b.

or

Ally -

if dig d I k (which I'Mly is of the Al (against tipping overturning) a seismic event 150%

postulated OBE or 110% of the postulated SSE) is imposed on any module.

C. :Ally' I k k dl 'ly k limits postulated in Section III subsection NF of the 1995 ASME Boiler and Pressure Vessel Code.

d. : The spatial average bulk pool temperature is required to remain under 137'F t in the wake of a normal refueling.

In addition to the limitations on the bulk pool temperature, the local water temperature in the Harris pools must remain subcooled (i.e., below the boiling temperature coincident with local hydraulic pressure conditions).

The 137'F limit is consistent with that currently in the Harris FSAR and procedures for pools A and B. CP&L is in the process of re-evaluating systems and components to allow for an increase the allowable bulk pool temperature.

Holtec Intermrtiorml 2-2 Holtec Report HI-971760

The reracking of Harris must not lead to violation of the off-site dose limits, or adversely affect the area dose environment as set forth in the Harris FSAR. The radiological implications of the installation of the g aLIIDUgm: Th billy f h I f H <<

new racks also need to be ascertained and deemed to be acceptable.

load combinations set forth in NUREG-0800, SRP 3.8.4 must be demonstrated.

'y h

h. : In addition to satisfying the primary stress criteria of Subsection NF, the alternating local stresses in the rack structure are evaluated to ensure that the "cumulative damage factor" due to at least ten SSE events does not exceed 1.0.

fgggglgTh I g y fh g d ylll- ~I I dlgd 'g seismic event must be demonstrated. A material fatigue evaluation is performed in accordance with ASME B&PV Code. The alternating local stresses in the liner are evaluated to ensure that the "cumulative damage factor" due to at least ten SSE events does not exceed 1.0.

3~

Thb'gpdbACI limits ffll lyhlkhdgp on the liner continues to satisfy the during and after a design basis seismic event.

k. :I h fp I XI p ( IIVI g of a fuel assembly, for instance), it is necessary to demonstrate that of the rack structure is not compromised. the'ubcriticality Th flld <<I I ~i&

within b

"state

'd of for executing the reracking must be demonstrated to be the proven art".

The foregoing design bases are further articulated in Sections 4 through 9 of this licensing report.

Holtec International 2-3 Holtec Report HI-971760

The following codes, standards and practices are used as applicable for the design, construction, and assembly of the Harris fuel storage racks. Additional specific references related to detailed analyses are given in each section.

ao (1) AISC Manual of Steel Construction, 1970 Edition and later.

(2) ANSI N210-1976, "Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations" (contains guidelines for fuel rack design).

(3) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code Section III, 1986 Edition; ASME Section V, 1986 edition; ASME Section VIII, 1986 Edition; ASME Section IX, 1986 Edition; and ASME Section XI, 1986 Edition.

(4) ASNT-TC-1A June, 1984 American Society for Nondestructive Testing (Recommended Practice for Personnel Qualifications).

(5) American Concrete Institute Building Code Requirements for Reinforced..

Concrete (ACI318-63) and (ACI318-71) ~

(6) Code Requirements for Nuclear Safety Related Concrete Structures, ACI349-85/ACI349R-85, aild ACI349.1R-80.

(7) ASME NQA-1, Quality Assurance Program Requirements for Nuclear Facilities (8) ASME NQA-2-1989, Quality Assurance Requirements for Nuclear Facility Applications.

(9) ANSI Y14.5M, Dimensioning and Tolerancing for Engineering Drawings and Related Documentation Practices.

(10) ACI Detailing Manual - 1980.

Holtec International Holtec Report HI-971760

b.

(1) E165 - Standard Methods for Liquid Penetrant Inspection.

(2) A240 - Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Fusion-Welded Unfired Pressure Vessels.

(3) A262 - Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel.

(4) A276 - Standard Specification for Stainless and Heat-Resisting Steel Bars and Shapes.

(5) A479 - Steel Bars for Boilers & Pressure Vessels.

(6) ASTM A564, Standard Specification for Hot-Rolled and Cold-Finished Age-Hardening Stainless and Heat-Resisting Steel Bars and Shapes.

(7) C750 - Standard Specification for Nuclear-Grade Boron Carbide Powder.

(8) A380 - Recommended Practice for Descaling, Cleaning and Marking Stainless Steel Parts and Equipment.

(9) C992 - Standard Specification for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.

(10) ASTM E3, Preparation of Metallographic Specimens.

(11) ASTM E190, Guided Bend Test for Ductility of Welds.

(12) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section II-Parts A and C, 1995 Edition.

(13) NCA3800 - Metallic Material Manufacturer's and Material Supplier's Quality System Program.

C. 'ABMBB ll dd and Brazing Qualifications, 1995 Edition.

V ld d,d l IX Will d Holtec International 2-5 Holtec Report HI-971760

d.

(1) ANSI 45.2.1 - Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants.

(2) ANSI N45.2.2 - Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (During the Construction Phase).

(3) ANSI - N45.2.6 - Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.58).

(4) ANSI-N45.2.8, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Plants.

(5) ANSI - N45.2.11, Quality Assurance Requirements for the Design of Nuclear Power Plants.

(6) ANSI-N45.2. 12, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants.

(7) ANSI N45.2.13 - Quality Assurance Requirements for Control of Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide I. 123).

(8) ANSI N45.2.15 Hoisting, Rigging, and Transporting of Items For Nuclear Power Plants.

(9) ANSI N45.2.23 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Regulatory Guide 1.146).

(10) ASME Boiler and Pressure Vessel,Section V, Nondestructive Examination, 1995 Edition.

(11) ANSI - N16.9-75 Validation of Calculation Methods for Nuclear Criticality Safety.

e.

(1) "OT Position for Review and Acceptance of Spent Fuel Storage and

. Handling Applications," dated April 14, 1978, and the modifications to this document of January 18, 1979.

Holtec International Holtec Report HI-971760

(2) NUREG 0612, "Control of Heavy Loads at Nuclear Power Plants",

USNRC, Washington, D.C., July, 1980.

(1) ANSI/ANS 8.1 (N16. 1) - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.

(2) ANSI/ANS 8.17, Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors.

(3) N45.2 - Quality Assurance Program Requirements for Nuclear Facilities

- 1971.

(4) N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants - 1974.

(5) N45.2.10 - Quality Assurance Terms and Definitions -1973.

(6) ANSI/ANS 57.2 (N210) - Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.

(7) N14.6 - American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or more for Nuclear Materials.

(8) ANSI/ASME N626-3, Qualification and Duties of Personnel Engaged in ASME Boiler and Pressure Vessel Code Section III, Div. 1, Certifying Activities.

(9) ANSI Y14.5M, Dimensioning and Tolerancing for Engineering Drawings and Related Documentation Practices.

g.

(1) 10CFR20 - Standards for Protection Against Radiation.

(2) 10CFR21 - Reporting of Defects and Non-compliance.

(3) 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants.

Holtec International 2-7 Holtee Report HI-971760

(4) 10CFR50 Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.

(5) 10CFR61 - Licensing Requirements for Land Disposal of Radioactive Material.

(6) 10CFR71 - Packaging and Transportation of Radioactive Material.

h.

(1) RG 1.13 - Spent Fuel Storage Facility Design Basis (Revision 2 Proposed).

(2) RG 1.25 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility of Boiling and Pressurized Water Reactors.

(3) RG 1.28 - (ANSI N45.2) - Quality Assurance Program Requirements .

(4) RG 1.29 - Seismic Design Classification (Rev. 3).

(5) RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Material.

(6) RG 1.38 - (ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants.

(7) RG 1.44 - Control of the Use of Sensitized Stainless Steel.

(8) RG 1.58 - (ANSI N45.2.6) Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel.

(9) RG 1.60 - Design Response Spectra for Seismic Design of Nuclear Power Plants.

(10) RG 1.61 - Damping Values for Seismic Design of Nuclear Power Plants, Rev. 0, 1973.

(11) RG 1.64 - (ANSI N45.2.11) Quality Assurance Requirements for the Design of Nuclear Power Plants.

(12) . RG 1.71 - Welder Qualifications for Areas of Limited Accessibility.

Holtec International 2-S Holtec Report HI-971760

(13) RG 1.74 - (ANSI N45.2.10) Quality Assurance Terms and Definitions.

(14) RG 1.85 - Materials Code Case Acceptability - ASME Section 3, Div.

l.

(15) RG 1.88 - (ANSI N45.2.9) Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records.

(16) RG 1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis.

(17) RG 1. 122 - Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components.

(18) RG 1.123 - (ANSI N45.2.13) Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants.

(19) RG 1. 124 - Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports, Revision 1, 1978.

(20) RG 3.4 - Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities.

(21) RG 3.41 - Validation of Calculational Methods for Nuclear Criticality Safety, Revision 1, 1977.

(22) RG 8.8 - Information Relative to Ensuring that Occupational Radiation Exposure at Nuclear Power Plants will be as Low as Reasonably Achievable (ALARA).

(23) DG-8006, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants".

(24) IE Information Notice 83 Fuel Binding Caused by Fuel Rack Deformation.

(25) RG 8.38 - Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, June, 1993.

(1) CPB 9.1 Criticality in Fuel Storage Facilities.

Holtee International 2-9 Holtec Report HI-971760

(2) ASB 9 Residual Decay Energy for Light-Water Reactors for Long-Term Cooling.

J~

(1) SRP 3.2.1 - Seismic Classification.

(2) SRP 3.2.2 - System Quality Group Classification.

(3) SRP 3.7.1 - Seismic Design Parameters.

(4) SRP 3.7.2 - Seismic System Analysis.

(5) SRP 3.7.3 - Seismic Subsystem Analysis.

(6) SRP 3.8.4 - Other Seismic Category I Structures (including Appendix D), Technical Position on Spent Fuel Rack.

(7) SRP 3.8.5 - Foundations for Seismic Category I Structures, Revision 1, 1981.

(8) SRP 9.1.2 - Spent Fuel Storage, Revision 3,, 1981.

(9) SRP 9.1.3 - Spent Fuel Pool Cooling and Cleanup System.

(10) SRP 9.1.4 - Light Load Handling System.

(11) SRP 9.1.5 - Heavy Load Handling System.

(12) SRP 15.7.4 - Radiological Consequences of Fuel Handling Accidents.

k.

(1) AWS D1.1 - Structural Welding Code, Steel.

(2) AWS D1.3 - Structure Welding Code - Sheet Steel.

(3) AWS D9.1 - Welding of Sheet Metal.

(4) AWS A2.4 - Standard Symbols for Welding, Brazing and Nondestructive Examination.

(5) AWS A3.0 - Standard Welding Terms and Definitions.

Holtec International 2-10 Holtec Report HI-971760

(6) AWS A5. 12 - Tungsten Arc-welding Electrodes.

(7) AWS QC1 - Standards and Guide for Qualification and Certification of Welding Inspectors.

2.4 The governing quality assurance requirements for design of the Harris spent fuel racks are enunciated in 10CFR50 Appendix B. The quality assurance program for design of the Harris racks is described in Holtec's Nuclear Quality Assurance Manual, which has been reviewed and approved by the Carolina Power & Light Company. This program is designed to provide a flexible but highly controlled system for the design, analysis and licensing of customized components in accordance with various codes, specifications, and regulatory requirements.

The manufacturing of the racks will be performed. in accordance with the requirements setforth in 10CFR50 Appendix B.

2S The Harris rack modules are designed as cellular structures such that each fuel assembly has a prismatic square opening with conformal lateral support and a flat horizontal bearing surface.

The basic characteristics of the Harris spent fuel racks are summarized in Table 2.5.1. The design of the PWR and BWR storage racks are very similar. The major differences are in the cell inside dimension and pitch, the baseplate flow holes, the support legs, and the poison width and length.

A central objective in the design of the new rack modules is to maximize their structural rigidity while minimizing their inertial mass, Accordingly, the Harris modules have been designed to simulate multi-fiange beam structures. The multiple flanges are formed from the numerous cell walls in the rack cross-sectional array. These cells are connected through intermittent welds. The weld lengths, location, and size were chosen during the original Holtec International 2-11 Holtec Report HI-971760

style/series to ensure adequate strength and to adjust the natural'frequency

~ ~

design of this rack

~

of the rack modules to avoid resonance. ~ In general, this effort has resulted in excellent detuning characteristics with respect to the applicable seismic events.

2.6 This subsection presents an item-by-item description of the anatomy of the Harris rack modules in the context of the fabrication methodology. The object of this section is to provide a self-contained description of rack module construction for the Harris fuel pool to enable an independent appraisal of the adequacy of design.

2.6.1 The requirements in manufacturing the high density storage racks for Harris may be stated in four interrelated points:

1. The rack modules are fabricated in such a manner that there is na weld splatter on the storage cell surfaces which would come in contact with the fuel assembly.

2. The storage locations are constructed so that redundant flow paths for the coolant are available.

3. The fabrication process involves operational sequences which permit immediate verification by the inspection staff.

4. The storage cells are connected to each other by austenitic stainless steel corner welds which leads to a honeycomb lattice construction. The extent of welding is selected to "detune" the racks from the stipulated seismic input motion.

Holtec International 2-12 Holtec Report HI-971760

2.6.2 In addition to the composite box assembly, the baseplate and the support legs constitute the principal components of the Harris fuel rack modules, The following description provides details of all of the major rack components.

Holtec International 2-13 Holtec Report Hl-971760

The assembly of the rack modules is carried out by welding the composite boxes in a vertical fixture with the precision fabricated baseplate serving as the bottom positioner.

Holtec International 2-14 Holtec Report HI-971760

2.6.3 Holtec International 2-15 Holtec Report HI-971760

The assembly of the rack modules is camed out by welding the composite boxes in a vertical fixture with the precision fabricated baseplate serving as the bottom positioner.

An elevation view of three PWR and BWR storage cells is shown in Figures 2.6.2 and 2.6.3, respectively.

Holtec International 2-16 Holtec Report HI-971760

Table 2.1.1 GEOMETRIC AND PHYSICAL DATA FOR POOL C RACK MODULESt Rack Number of Number of Dimension (inches) Shipping Submerged I.D. TYpe Cells Cells Per Weight ebs) Weight ebs) tt N-S E-W Module N-S Direction E-W Direction

)',9%8;~

+P,r/g<$ 4y% .

pPhWR'"..,

gjPWR N".-'.!:Pa%RQ3 e.':B4":gi 6PWR.'""'-'.':.

""".NN.';:'.'"..

':::::'PWRjgi yyyy~y y g y Ace yv 'y ~c yap AY yyN

<<$ iagfRy;.i;',

~B~R'jg All dimensions are rounded off to the nearest 0.5 inch, and all weights are rounded off to the nearest 10 lbs.

See Figure 1.2 for pool configuration.

Holtec International 2-17 Holtec Report HI-971760

Table 2.1.1 (Cont'd.)

GEOMETRIC AND PHYSICAL DATA FOR POOL C RACK MODULES Rack Number of Number of Dimension (inches) >>ppmg Submerged I.D. ape Cells Cells Per Weight ebs) Weight ebs) tt N-S E-W Module N-S Direction E-W Direction

" )NWp

@~~ew~i

" '{ %""i:."">>y'"r"',x,','~;,';~.~.

{<<4>. 'P~N::

<<.x<<{ x<<~<<:.<<..:.:,t xAz {..NKC': . '4'4: cc' ? "" 'w" ""

'8?{<<4':::.k/Vl":P<<'A?;:.v8.:((H?'.?':"""';8%R':.:;":.>

t(('.:$'<<g:{

~g'..'.."...",";.".',-".-",'-'ja zx'@WE?M~ "',+'.y~.i'~g K.

<<<<<<4%'.{<<'v.:...k.<<<<

's{g{'.<<{~A<<<< jg>'<<jj.{V<<(:?'. z ...'{ <<<<{(YN? <<<<P%:y"':{:.': g'NK 9<</+Rex.x '~+<<<<<:,.:,~vzz

.'WR@" '{":.i%+4

~: 8%R:~~ ',,':

(gg%<?{"',

i>,-::.',':ij::.",'NBV!fg+Zg

@'""'.A?<<"'g All dimensions are rounded off to the nearest 0.5 inch, and all weights are rounded off to the nearest 10 lbs.

See Figure 1.2 for pool configuration.

Holtec International 2-18 Holtec Report Hl-971760

Table 2.1.2 GEOMETRIC AND PHYSICAL DATA FOR POOL D RACK MODULESt Rack Number of Number of Dimension (inches) Shipping Submerged I.D. TYpe Cells Cells Per Weight ebs) Weight ebs) tt N-S E-W Module N-S Direction E-W Direction t~~PjjR~

W F>>W..& . C" X>>:< rA'C'vj

.."-:::PEVR '--":

?gA,K <

W>>? 0::?<?>>>>:

t.:>>p%R~:.i<

>>ONwvsV>>>>

(Wg> .~N( ...N....?...N%>>>>WY).'>>>>'?>>>>..>>>>?(>>+NP>> .'?O'V'.'>>>>)>>>>>>?

jjf%R'i"..;i It:,!PNR;P A6 -",jP%'.R'l',':-",.i

'>>."r'>ANN ?.".;~':(i>>:: >>>>?x'N:v 'i'. " '~>> ~%'A4""::.2:;'N+

% xQ:a

~PRE."".

'A 8l" ":>>>>'P<sMFy>>?>>+?Ã>> NN::%:>>.N>>?'.ca %?>>>>>>>>""5@/:: 94e.>>4>>:. j'>>'.:;:.'.".;

'NN~jc<jPj~gp g'-?'.j agxg

':,'PB3+ t-:';PWR~::',i

',.,"'5C'I,:j"."'.,';:-; j<q"::.PWR::.:",:,

" '>>;4~'~>>':" '~ ':~ """ t'~e",":""'4""%':"A~Yt' ll <

g ig gg'?;?

'w,, i,cv>>:

All dimensions are rounded off to the nearest 0.5 inch, and all weights are rounded off to the nearest 10 lbs.

See Figure 1.3 for pool configuration.

Holtec International 2-19 Holtec Report HI-971760

Table 2.5.1 MODULE DATA FOR HARRIS SPENT FUEL RACKS Parameter PWR BWR gB ceH.;:hetgllt(above:,tits;:1) aseplate) "t':>~';g~q,j+ ';4'.c<t t ".'~.'AM>gg@gp;,,';.,', : Rw'~g'j<

>Retllot8iltfting!8nd. flantlllng":>fovt~stofpVkFKNK+~%>'<tr~~ ~Cggi: "'"@~>'@r'<r> .>@

<"'%oltec International 2-20 Holtec Report HI-971760

FIGURE P..i.i; PICTORIAL VIEW OF TYPICAL HARRIS RACK STRUCTURE HI

- 971760

FIGURE 2.6.1; SEAM WELDED PRECISION .FORMED CHANNELS HOLTEC PROPRIETARY HI-971760 2-22

FIGURE 2.6.2; THREE PWR CELLS IN ELEVATION VIER Holtec Proprietary HI-971760

FIGURE P. 6 3; THREE BWR CELLS IN ELEVATION VIEW Holtec Proprietary 2-24

FIGURE 2 6 4; .COMPOSITE BOX ASSEMBLY Holtec Proprietary

FIGURE 2 6 5; TYPICAL ARRAY OF STORAGE CELLS

( NON FLUX TRAP CONSTRUCT ION )

OLTEC PROPRIETARY HI- 971760 2-26 ,

FIGURE 2.6.6; SUPPORT PEDESTAL FOR PWR RACK HOLTEC PROPRIETARY HI-971760

FIGURE 2.6.7; SUPPORT PEDESTAL FOR HOLTEC BWR RACKS HOLTEC PROPRIETARY HI-971760

3.0 MATERIAL,HEAVYLOAD, AND CONSTRUCTION CONSIDERATIONS 3.1 Intmdusihn Safe storage of nuclear fuel in the Harris pools requires that the materials utilized in the rack t

fabrication be of proven durability and be compatible with the pool water environment. Likewise, all activities during the rack installations must comply with the provisions of NUREG-0612 to eliminate the potential of construction accidents. This section provides a synopsis of the considerations with regard to long-term service life and short-term construction safety.

The following structural materials are utilized in the fabrication of the new spent fuel racks:

a. ASME SA240-304L for all sheet metal stock

b. Internally threaded support legs: ASME SA240-304L

c. Externally threaded support spindle: ASME SA564-630 precipitation hardened stainless steel (heat treated to 1100'F)

d. Weld material - per the following ASME specification: SFA 5.9 ER308L 3.3 The racks employ Boral~, a patented product of AAR Manufacturing, as the neutron absorber material. Boral is a thermal neutron poison material composed of boron carbide and 1100 alloy aluminum. Boron carbide is a compound having a high boron content in a physically stable and chemically inert form. The 1100 alloy aluminum is a lightweight metal with high tensile strength which is protected from corrosion by a highly resistant oxide film. The two materials, boron carbide and aluminum, are chemically compatible and ideally suited for long-term use in the radiation, thermal and chemical environment of a nuclear reactor or a spent fuel pool. Boral has been shown [3.3.1] to be superior to alternative materials previously used as neutron absorbers in storage racks.

Holtcc International 3-1 Holtec Rcport HI-971760

Boral has been the most widely used neutron absorbing material in fuel rack applications over the past 20 years. Its use in the spent fuel pools as the neutron absorbing material can be attributed to its proven performance (over 150 pool years of experience) and the following unique characteristics:

The content and placement of boron carbide provides a very high removal cross-section for thermal neutrons.

Boron carbide, in the form of fine particles, is homogeneously dispersed throughout the central layer of the Boral panels.

The boron carbide and aluminum materials in Boral do not degrade as a result of long-term exposure to radiation.

Iv. The neutron absorbing central layer of Boral is clad with permanently bonded surfaces of aluminum.

v. Boral is stable, strong, durable, and corrosion resistant.

Boral will be manufactured by AAR Manufacturing under the control and surveillance of a Quality Assurance/Quality Control Program that conforms to the requirements of 10CFR50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants". As indicated in Tables 3.3.1 and 3.3.2, Boral has been licensed by the USNRC for use in numerous PWR and BWR spent fuel storage racks and has been extensively used in international nuclear installations.

'l Aluminum: Aluminum is a silvery-white, ductile metallic element that is the most abundant in the earth's crust. The 1100 alloy aluminum is used extensively in heat exchangers, pressure vessels and storage tanks, chemical equipment, reflectors, and sheet metal work.

It has high resistance to corrosion in industrial and marine atmospheres. Aluminum has atomic number of 13, atomic weight of 26.98, specific gravity of 2.69 and valence of 3. The physical, Holtec International 3-2 HoltecReportHI-971760

of the

~

1100 alloy aluminum are listed in Tables 3.3.3 and

~ ~

mechanical and chemical properties 3.3.4.

The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting corrosion.

Boron Carbide: The boron carbide contained in Boral is a fine granulated powder that conforms to ASTM C-750-80 nuclear grade Type III. The material conforms to the chemical composition and properties listed in Table 3.3.5.

References [3.3.2], [3.3.3], and [3.3.4] provide further discussion as to the suitability of these materials for use in spent fuel storage module applications.

i 1 W All materials used in the construction of the Holtec racks have an established history of in-pool usage. Their physical, chemical and radiological compatibility with the pool environment is an established fact.

Austenitic stainless steel (304L) is perhaps the most widely used stainless alloy in nuclear power plants, since it provides both high strength and non-corrosive properties.

3.5 The Fuel Handling Building auxiliary crane will be used for installation of the new storage racks in pools C and D. The Spent Fuel Cask Handling Crane (CHC) cannot be used for rack installation, since travel limitations prohibit its movement over the spent fuel pools. Storage capacity will be Holtec International 3-3 Holtec Report HI-971760

increased starting in the south end of pool C and proceeding north. This installation pattern will enable the storage racks to be manipulated without lifts over spent fuel.

The auxiliary crane is a single failure proof crane and is currently rated for 10 tons. A 20 ton hoist will be attached to the auxiliary crane hook to prevent submergence of the auxiliary crane hook.

The auxiliary crane was used for installation of storage racks in pool B. Rigging and procedures for pools C and D rack installation will be similiar to those used previously.

The maximum lift weight during rack installation is determined by the following table.

Item Weight (Ibs)

Rack 15,700 (maximum)

LiftRig 1,200 Rigging 500 20 ton hoist 1,420 Total Lift 18,820 The rack sizes were limited to ensure that the crane and lifting components remain single failure proof and it may be seen that the maximum lift of 18,820 lbs is below the auxiliary crane rating of 20,000 lbs. As a result, the auxiliary crane, which can travel over both pools C and D, is qualified to accept the anticipated load during the rack installation project.

A remotely engagable lift rig, meeting NUREG-0612 [3.5.1] stress criteria, will be used to liftthe new modules. The rig is designed for handling both PWR and BWR racks. The new rack lift rig consists of independently loaded lift rods in a lift configuration which ensures that failure of one traction rod will not result in uncontrolled lowering of the load being carried by the rig (which complies with the duality feature called for in Section 5.1.6(3a) of NUI&G0612).

Holtec International Holtec Report HI-971760

The rigs have the following attributes:

a. The traction rod is designed to prevent loss of its engagement with the rig in the locked position. Moreover, the locked configuration can be directly verified from above the pool water without the aid of an underwater camera.

b. The stress analysis of the rigs will be carried out using a finite element code, and the primary stress limits in ANSI 14.6-1978 [3.5.2] will be shown to be met by detailed analysis.

The rigs will be load tested with 300% of the maximum weight to be lifted. The test weight will be maintained in the air for 10 minutes. All critical weld joints will be liquid penetrant examined to establish the soundness of all critical joints.

Pursuant to the defense-in-depth approach ofNUREG-0612, the following additional measures of safety will be undertaken for the racking operation.

The crane used in the project will be given a preventive maintenance checkup and inspection per the Harris maintenance procedures before the beginning of the racking operation.

Safe load paths will be developed for moving the new racks in the Fuel Handling Building. The racks will not be carried directly over any fuel located in the pool.

The rack upending and laying down will be carried out in an area which precludes any adverse interaction with safety related equipment.

lv. All crew members involved in the use of the liAing and upending equipment will be given training similar to that utilized in previous rack installation operations.

The rack installation activities will require Harris PNSC approval and will be conducted in accordance with written procedures which will be reviewed and approved by Carolina Power &

Light.

The proposed heavy loads compliance will be in accordance with the objectives of the CP&L, NRC-approved submittal to NUI&G-0612. The guidelines of NUREG-0612 call for measures to Holtcc Intcmational 3-5 Holtcc Rcport HI-971760

~

"provide an adequate defense-in-depth for handling a of heavy loads near spent fuel...". The

~ ~ ~ ~

NUEEG-0612 guidelines cite four major causes of load handling accidents, namely operator errors rigging failure lack of adequate inspection inadequate procedures The Harris racking program ensures maximum emphasis on mitigating the potential load drop accidents by implementing measures to eliminate shortcomings in all aspects of the operation including the four aforementioned areas. A summary of the measures specifically planned to deal with the major causes is provided below.

Operator errors: As mentioned above, CPAL plans to provide comprehensive training to the installation crew. All training shall be in compliance with ANSI B30.2 [3.5.3].

Rigging failure: The lifting device designed for handling and installation of the new racks at Harris has redundancies in the lift legs and lift eyes such that there are four independent load members. Failure of any one load bearing member would not lead to uncontrolled lowering of the load. The rig complies with all provisions of ANSI 14.6 [3.5.2], including compliance with the primary stress criteria, load testing at 300% of maximum lift load, and dye examination of critical welds.

The Harris rig design is similar to the rigs used in the initial racking or the rerack of numerous other plants, such as Hope Creek, Millstone Unit 1, Indian Point Unit Two, Ulchin II, Laguna Verde, J.A. FitzPatrick and Three Mile Island Unit 1.

Lack of adequate inspection: The designer of the racks will develop a set of QC hold points which will require inspections and approvals prior to proceeding. Additional hold points will be established for activities during the installatin process. These inspection points have been proven to significantly reduce any requirement for rework or instances of erroneous installation in numerous prior rerack projects.

Inadequate procedures: CPkL is developing various operating procedures to address operations pertaining to the rack installation e6ort, including, but not limited to, mobilization, rack handling, upending, liAing, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance. Many of the procedures will be the same or revisions to those developed and currently in use for rack installations in pool B.

Holtec International 3-6 Holtec Report HI-971760

~

The series of operating

~

procedures planned for Harris rack installations are the successors of the

~

procedures successfully implemented in previous projects.

Table 3.5.1 provides a synopsis of the requirements delineated in NUREG-0612, and their intended compliance.

Holtec International 3-7 Holtec Report HI-971760

[3.3.1] "Nuclear Engineering International," July 1997 issue, pp 20-23.

[3.3.2] "Spent Fuel Storage Module Corrosion Report," Brooks & Perkins Report 554, June 1, 1977.

[3.3.3] "Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Storage Pools," Brooks & Perkins Report 578, July 7, 1978.

[3.3.4] "Boral Neutron Absorbing/Shielding Material - Product Performance Report," Brooks &

Perkins Report 624, July 20, 1982.

[3.5.1] NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," July 1980.

[3.5.2] ANSI N14.6-1978, Standard for Special LiftingDevices for Shipping Containers Weighing 10000 Pounds or more for Nuclear Materials," American National Standard Institute, Inc., 1978.

[3.5.3] ANSVASME B30.2, "Overhead and Gantry Cranes, (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist)," American Society of Mechanical Engineers, 1983.

[3.5.4] ANSVASME B30.20, "Below-the-Hook LiAing Devices," American Society of

~ ~ ~

~

Mechanical Engineers, 1993.

[3.5.5] CMMASpecification 70, "Electrical Overhead Travelling Cranes," Crane Manufacturers Association of America, Inc., 1983.

[3.5.6] ANSVASME B30.20, "Below-the-Hook LiftingDevices," American Society of Mechanical Engineers, 1993.

Holtec International 3-8 Holtee Report HI-971760

Table 3.3.1 BORAL EXPERIENCE LIST - PWRs Plant Utility Docket No. Mfg. Year Maine Yankee Maine Yankee Atomic Power 50-309 1977 Donald C. Cook Indiana & Michigan Electric 50-315/316 1979 Sequoyah 1,2 Tennessee Valley Authority 50-327/328 1979 Salem 1,2 Public Service Electric & Gas 50-272/311 1980 Zion 1,2 Commonweal th Edison Co. 50-295/304 1980 Bellefonte 1, 2 Tennessee Valley Authority 50438/439 1981 Yankee Rowe Yankee Atomic Power 50-29 1964/1983 Indian Point 3 NY Power Authority 50-286 1987 Byron 1,2 Commonwealth Edison Co. 50454/455 1988 Braidwood 1,2 Commonwealth Edison Co. 50456/457 1988 Yankee Rowe Yankee Atomic Power 50-29 1988 Three Mile Island I GPU Nuclear 50-289 1990 Sequoyah (rerack) Tennessee Valley Authority 50-327 Donald C. Cook (rerack) American Electric Power 50-315/316 1992 Beaver Valley Unit 1 Duquesne Light Company 50-334 1993 Fort Calhoun Omaha Public Power District 50-285 1993 Zion 1 & 2 (rerack) Commonwealth Edison Co. 50-295/304 1993 Salem Units 1 & 2 (rerack) Public Gas and Electric Company 50-272/311 1995 Haddam Neck Connecticut Yankee Atomic Power 50-213 1996 Company Gosgen Kernkraftwerk Gosgen-Daniken AG 1984 (Switzerland)

Koeberg 1,2 ESCOM (South Africa) 1985 Beznau 1,2 Nordostschweizerische Kraftwerke 1985 AG (Switzerland)

Holtcc International 3-9 Holtec Report HI-971760

Table 3.3.1 (Cont'd.)

BORAL EXPERIENCE LIST - PWRs Plant Utility Docket No. Mfg. Year 12 various Plants Electricite de France (France) 1986 Ulchin Unit 1 Korea Electric Power Company 1995 (Korea)

Ulchin Unit 2 Korea Electric Power Company 1996 (Korea)

Korin Korea Electric Power Company 1996 (Korea)

Yonggwang 1,2 Korea Electric Power Company 1996 (Korea)

Sizewell B Nuclear Electric, pic (United 1997 Kingdom)

Angra 1 Furnas Centrais-Electricas SA 1997 (Brazil)

Holtec International 3-10 Holtec Report HI-971760

Table 3.3.2 BORAL EXPERIENCE LIST - BWRs Plant Utility Docket No. Mfg. Year Cooper Nebraska Public Power 50-298 1979 J.A. FitzPatrick NY Power Authority 50-333 1978 Duane Arnold Iowa Electric Light & Power 50-331 1979 Browns Ferry 1,2,3 Tennessee Valley Authority 50- 1980 259/260/296 Brunswick 1,2 Carolina Power & Light 50-324/325 1981 Clinton Illinois Power 50461/462 1981 Dresden 2,3 Commonwealth Edison Company 50-237/249 1981 E.I. Hatch 1,2 Georgia Power 50-321/366 1981 Hope Creek Public Service Electric & Gas 50-354/355 1985 Humboldt Bay Pacific Gas & Electric Company 50-133 1985 LaCrosse Dairyland Power 50409 1976 Limerick 1,2 Philadelphia Electric Company 50-352/353 1980 Monticello Northern States Power 50-263 1978 Peachbottom 2,3 Philadelphia Electric 50-277/278 1980 Perry 1,2 Cleveland Electric Illuminating 50M 0/441 1979 Pilgrim Boston Edison Company 50-293 1978 Susquehanna 1,2 Pennsylvania Power & Light 50-387,388 1979 Vermont Yankee Vermont Yankee Atomic Power 50-271 1978/1986 Hope Creek Public Service Electric & Gas 50-354/355 1989 Shearon Harris Pool B Carolina Power &, Light 50401 1991 Duane Arnold Iowa Electric Light & Power 50-331 1993 Pilgrim Boston Edison Company 50-293 1993 LaSalle 1 Commonwealth Ed ison Company 50-373 Millstone Unit 1 Northeast Utilities 50-245 1989 James A. FitzPatrick NY Power Authority 50-333 1990 Hope Creek Public Service Electric & Gas Company 50-354 1991 Holtcc International 3-11 Holtcc Rcport HI-971760

Table 3.3.2 (Cont'd.)

BORAL EXPEMENCE LIST - BWRs Plant Utility Docket No. Mfg. Year Duane Arnold Energy Iowa Electric Power Company 50-331 1994 Center Limerick Units 1,2 PECO Energy 50-352/50- 1994 353 Shearon Harris Pool 'B'arolina Power & Light Company 50-401 1996 Nine Mile Point Unit 1 Niagara Mohawk Power Corporation 50-220 1997 Chinshan 1,2 Taiwan Power Company (Taiwan) 1986 Kuosheng 1,2 Taiwan Power Company (Taiwan) 1991 Laguna Verde 1,2 Comision Federal de Electricidad 1991 (Mexico)

Holtec International 3-12 Holtec Report HI-971760

Table 3.3.3 1100 ALLOYALUMINUMPHYSICAL CHARACTERISTICS Density 0.098 lb/in3 Melting Range 1190'F - 1215'F Thermal Conductivity (77'F) 128 BTU/hr/ft /F/ft Coefficient of Thermal Expansion 13.1 x 10 in/in-'F (68'F - 212'F)

Specific Heat (221'F) 0.22 BTU/lb/'F Modulus of Elasticity 10 x 10'psi Tensile Strength (75'F) 13,000 psi (annealed) 18,000 psi (as rolled)

Yield Strength (75'F) 5,000 psi (annealed 17,000 psi (as rolled)

Elongation (75'F) 35-45% (annealed) 9-20% (as rolled)

Hardness (Brinell) 23 (annealed) 32 (as rolled)

Annealing Temperature 650'F Holtec International 3-13 Holtec Report HI-971760

Table 3.3.4 CHEMICALCOMPOSITION - ALUMINUM (1100 ALLOY) 99.00% min. Aluminum 1.00% max. Silicone and Iron 0.05-0.20% max. Copper 0.05% max. Manganese 0.10% max. Zinc 0.15% max. Other Holtec International 3-14 Holtec Report HI-971760

Table 3.3.5 CHEMICAL COMPOSITION AND PHYSICAL PROPERTIES OF BORON CARBIDE CHEMICAL COMPOSITION (WEIGHT PERCENT)

Total boron 70.0 min.

B" isotopic content in natural boron 18.0 Boric oxide 3.0 max.

Iron 2.0 max.

Total boron plus total carbon 94.0 min.

PHYSICAL PROPERTIES Chemical formula B4C Boron content (weight percent) 78.28%

Carbon content (weight percent) 21.72%

Crystal structure rhombohedral Density 0.0907 lb/in'elting Point 4442'F Boiling Point 6332'F Holtec International 3-15 Holtec Report HI-971760

Table 3.5.1 HEAVY LOAD HANDLINGCOMPLIANCE MATRIX(NUREG-0612)

Criterion Compliance Are safe load paths defined for the movement of heavy loads to minimize the potential of impact, if dropped, on irradiated fuel?

2. Will procedures be developed to cover:

identification of required equipment, inspection and acceptance criteria required before movement of load, steps and proper sequence for handling the load, defining the safe load paths, and special precautions?

3. Will crane operators be trained and qualified?

4. Will special lifting devices meet the Yes guidelines of ANSI 14.6-1978?

5. Will non-custom lifting devices be Yes installed and used in accordance with ANSI B30.20, latest edition?

6. Will the cranes be inspected and tested Yes prior to use in rack installation?

7. Does the crane meet the intent of ANSI B30.2-1976 and CMMA-70?

'oltec International 3-16 Holtec Report HI-971760

C 4.0 CRITICALITYSAFETY EVALUATION 4.1 D~iB The high density spent fuel PWR and BWR storage racks for Harris Pools C and D are designed in accordance with the applicable codes listed below. The rack design and fuel storage configuration acceptance criteria is to show that the effective neutron multiplication factor, k,rr, is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with un-borated water at a temperature corresponding to the highest

. reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including mechanical tolerances. Alluncertainties are statistically combined, with uncertainties applied conservatively to calculate the final k,rr which must be shown to be less than 0.95 with a 95% probability at a 95% confidence level [4.1.1], Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under credible abnormal and accident conditions, the reactivity willnot exceed the limiting design basis value.

Applicable codes, standards, and regulations or pertinent sections thereof, include the following:

~ General Design Criteria 62, Prevention of Criticality in Fuel Storage and Handling.

~ USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3

- July 1981.

~ USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.

~ USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981.

~ ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

~ ANSI/ANS-57.2-1983, Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.

Holtec International 4-1 ReportHI-971760

~ ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.

USNRC guidelines and the applicable ANSI standards specify that the maximum effective multiplication factor, k,ti, including uncertainties, shall be less than or equal to 0.95. The infinite multiplication factor, k;~ is calculated for an infinite array, neglecting neutron losses due to leakage &om the actual storage rack, and therefore is a higher and more conservative value. In the present criticality safety evaluation of the Harris storage racks, the design basis criterion was assumed to be a k, of less than 0.95, which is more conservative than the limit specified in the regulatory guidelines.

To ensure that the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made:

0* 4 f

I \

t C

The PWR spent fuel storage racks are designed to accommodate any and all of the fuel assemblies listed in Table 4.3.1 with a maximum enrichment of 5 wt% 'U. To assure the Holtec International 4-2 Report HI-971760

reactive fuel assembly type was identified and used as the design basis fuel assembly. The Westinghouse 15x15 assembly was determined to have the highest reactivity at zero burnup and as a function of burnup for an initial 5 wt% ~'U enrichment and therefore was used as the design basis PWR fuel assembly.

The BWR spent fuel storage racks are designed to accommodate any and all of the fuel assemblies listed in Table 4.3.2 with a maximum planar average enrichments of 4.6 wt.% ~'U.

Each fuel assembly type was analyzed independently to determine its acceptability in the rack. It is noted that individual fuel rods can have enrichments that are less than or greater than the maximum planar average enrichment.

4.2 Summ of Criticali Anal ses 4.2.1 Normal 0 eratin Conditions 4.2.1.1

~ ~ PWR Fuel Results The design basis PWR fuel assembly is a 15 x 15 Westinghouse fuel assembly containing UO, at a maximum initial enrichment of 5.0 wt% ~'U, Allfuel assembly types listed in Table 4.3.1 were also evaluated and the Westinghouse 15x15 assembly was shown to exhibit the highest reactivity for the high density PWR storage racks at Harris.

The NRC guidelines specify that the limiting ~ of 0.95 under normal storage conditions should be evaluated in the absence of soluble boron. Consequences of abnormal and accident conditions have also been evaluated assuming no soluble boron, where "abnormal" refers to conditions (such as higher water temperatures) which may reasonably be expected to occur during the lifetime of the plant and "accident" refers to conditions which are not expected to occur but nevertheless must be protected against.

~ ~

of the

~

The criticality analyses spent fuel storage pool are summarized in Table 4.2.1 for the

~ ~

design basis storage conditions. The maximum ~ is 0.9450 (95% probability at the 95% con-Holtec International 4-3 Report HI-971760

fidence level) for the enrichment-burnup combinations shown in Figure 4.2.1. The calculated maximum reactivity includes burnup-dependent allowances for uncertainty in depletion calculations and for the axial distribution in burnup. Reactivity allowances for manufacturing tolerances and calculational uncertainties are also included. As cooling time increases in long-term storage, decay of Pu-241 and growth of Am-241 results in a significant decrease in reac-tivity, which willprovide a contiriuously increasing subcriticality margin for the next'100 years.

The racks can safely accommodate fuel of various initial enrichments and discharge fuel burnups, provided the combination falls within the acceptable domain above the curve in Figure 4.2.1. For convenience, the minimum (limiting) burnup data for unrestricted storage can be described as a linear function of the initial enrichment (E, in weight percent 'U) which conservatively encompasses the limiting burnup data. The equation for this curve is shown in Figure 4.2.1 and provided below.

For Unrestricted Storage of the following PWR fuel assemblies Westinghouse 17x17 Std Westinghouse 17x17 V5 Westinghouse 15x15 Siemens 17x17 Siemens 15x15 the enrichment must be less than or equal to 5 wt% 'U and the burnup must satisfy the minimum burnup requirements Minimum Burnup in MWD/MTU= 12114*E-19123 The burnup criteria willbe implemented by appropriate administrative procedures to ensure verified burnup as specified in the proposed Regulatory Guide 1.13, Revision 2, prior to fuel transfer into Spent Fuel Pools C or D.

4.2.1.2 BWR Fuel Results P

AllBWR fuel assembly types being considered were explicitly-analyzed to determine the acceptability for storage in Spent Fuel Pool C. The maximum planar average enrichment was I

Holtec International Report Hl-971760

e assumed for all rods in the assembly and no credit was taken for gadolinia which might be present.

The criticality safety was evaluated at the burnup corresponding to a k, of 1.32 in the Standard Cold Core Geometry (SCCG), SCCG is defined as an infinite array of fuel assemblies on a 6-inch lattice spacing at 20'C, without any control absorber or voids.

The maximum k, in the BWR storage

~

rack was determined to be 0.9443 (95% probability at the 95% confidence level) including aii known caicuiationai and manufacturing uncertainties.

This allowance also encompasses any uncertainty in the burnup calculations.

The basic calculations supporting the criticality safety of the Harris fuel storage racks for the design basis fuel are summarized in Table 4.2.2. For the design basis fuel, the fuel storage rack The acceptance criteria for storage of spent BWR fuel in Harris Pool C can be summarized in the "

following manner.

For Unrestricted Storage of the following BWR fuel assemblies GE 3, GE 4, GE 5, GE 6, GE 7, GE 8, GE 9, GE 10, GE 13 the maximum planar average enrichment must be less than or equal to 4.6 wt.% 'U and the k; in standard cold core geometry must be less than or equal to 1.32 4.3 I~ P t

4.3.1 Reference PWR Fuel Assembl and Stora e Cell The design basis PWR fuel assembly is a 15x15 array of fuel rods with 21 rods replaced by 21 control rod guide tubes. Table 4.3.1 summarizes the PWR fuel assembly design specifications Holtec International 4-5 Report HI-971760

t for all fuel assemblies analyzed. Figure 4.3.1 shows the calculational model of the PWR spent fuel storage cell containing a 15x15 assembly.

The design basis for the Region 2 type storage cells is fuel of 5.0 wt.% ~'U maximum initial enrichment burned to 41,447 MWD/MTUI, t 4.3,2 Reference BWR Fuel Assembl and Stora e Cell The design basis BWR fuel assembly, used for uncertainty calculations, is a standard 8x8 array of BWR fuel rods containing UO, clad in Zircaloy (60 fuel rods with 4 water rods). Design parameters for all BWR fuel assemblies analyzed are summarized in Table 4.3.2. Figure 4.3.2 shows the calculational model of a BWR storage rack cell containing an 8x8 assembly.

Holtec International 4-6 Report HI-971760

4.4 Anal ical Methodolo 4.4.1 Reference Desi n Calculations In the fuel rack analyses, the primary criticality analyses of the high density spent fuel storage racks were performed with a two-dimensional multigroup transport theory technique, using the CASMO-3 computer code [4,4.1 - 4.4.4]. Since CASMO-3 can not be directly compared to critical experiments, a calculational bias is not available for CASMO-3. Therefore, independent verification calculations were made with a Monte Carlo technique utilizing the MCNP-4A computer code [4.4.5]. Benchmark calculations, presented in Appendix A, indicate a bias of 0.0009+ 0.0011 for MCNP-4A, evaluated at the 95% probability, 95% confidence level. The

~

MCNP-4A bias and uncertainty were included in the MCNPQA to CASMO-3 comparison as

~ ~

discussed in Section 4.5.~ ~

4 CASMO-3 was also used for burnup calculations and for evaluating small reactivity increments associated with manufacturing tolerances.

MCNP-4A was used to determine reactivity effects, to calculate the reactivity for fuel misloading outside the racks and to determine the effect of having PWR and BWR racks adjacent to each other. MCNP-4A Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking.,

k

,,1.>> ~,( p, Holtec International 4-7 Report HI-971760 II

4.4.2 Burnu Calculations and Uncertainties CASMO-3 was used for burnup calculations during core operations. CASMO-3 has been extensively verified [4.4.4, 4.4.6] against Monte Carlo calculations, reactor operations, and heavy-element concentrations in irradiated fuel, In addition, Johansson [4.4.7] has obtained very good agreement in calculations of close-packed, high-plutonium-content, experimental configurations.

4.4.2.1 PWR Fuel Burnup Calculations Since critical experiment data with spent fuel is not available for determining the uncertainty in burnup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations.

Table 4.4.1 summarizes results of the burnup analyses to determine the allowances for uncertainties in burnup calculations. The reactivity allowances for uncertainties in burnup are listed for three different burnup ranges: less than 30,000 MWD/MTU, between 30,000 and 40,0000 MWD/MTU,and between 40,000 and 45,000 MWD/MTU. The appropriate uncertainty was used for each burnup range in determining the acceptable burnup versus enrichment combinations depicted in Figure 4.2.1. The allowance for uncertainty in burnup calculations is believed to be a conservative estimate, particularly in view of the substan-tial reactivity decrease with aged fuel as discussed in Section 4.4.4.

4.4.2.2 BWR Fuel Burnup Calculations and Comparison to Vendor Calculations CASMO-3 was used to perform depletion calculations and to calculate the k;, in the SCCG. As discussed, there are no depleted fuel critical experiments with which to benchmark CASMO-3's depletion calculations. Therefore a reactivity allowance for uncertainty in depletion is needed.

Holtec International 4-8 Report HI-971760

The allowance is used to also encompass any potential differences between the SCCG calculations performed here and the vendor calculations.

4.4.3 Effect of Axial Burnu Distribution Initially, fuel loaded into the reactor willburn with a slightly skewed cosine power distribution.

As burnup progresses, the burnup distribution willtend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to neutron leakage. Consequently, it would be expected that over most of the burnup history, distributed burnup fuel assemblies would exhibit a slightly lower reactivity t than that calculated for the average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced Generic analytic results of the axial burnup effect have burnup.

been provided by Turner [4.4.8] based upon calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup, becoming positive at burnups greater than about 30,000 MWD/MTU. The trends observed [4.4.8] suggest the possi-bility of a small positive reactivity effect above 30,000 MWD/MTUincreasing to slightly over 1% dk at 40,000 MWD/MTU.

4.4.3.1 PWR Fuel Axial Burnup Distribution Calculations for the Harris storage racks with PWR fuel of three different average burnups were made using an axial burnup distribution representative of spent PWR fuel' At lower burnups, the The axial burnup distribution measured on spent fuel Rom the Surry plant was used as representative of PWR fuel.

Holtec International 4-9 Report HI-971760

~ ~ ~ ~

reactivity increment is smaller as indicated in Table 4.4.1, being negative at 30,000 MWD/MTU

~ ~

~ ~ ~

and at lower burnups. No credit is taken for this negative reactivity effect at the lower burnups other than the suggestion of additional conservatism. Furthermore, the reactivity signiGcantly decreases with time in storage (Section 4.4.4 below) providing a continuously increasing margin below the 0.95 limit.

The appropriate reactivity allowance for the effect of axial burnup distribution was used for each burnup range in determining the acceptable burnup versus enrichment values in Figure 4.2.1.

4.4.3.2 BWR Fuel Axial Burnup Distribution The burnup at which ~ in the SCCG reaches 1.32 is approximately 12,000 MWD/MTU.As discussed above and in [4.4.8] the effect of using the explicit axial burnup distribution as opposed to an average burnup distribution results in a negative effect on reactivity. Therefore, no

~ ~

reactivity allowance for axial burnup distribution is applied to the BWR fuel analysis.

4.4.4 Lon Term Reactivi Chan es At reactor shutdown, the reactivity of the fuel initially decreases due to the growth of Xe-135.

Subsequently, the Xenon decays and the reactivity increases to a maximum at several hundred hours when the Xenon is gone. Over the next 30 years, the reactivity continuously decreases due primarily to Pu-241 decay and Americium growth. At lower burnup, the reactivity decrease will be less pronounced since less Pu-241 would have been produced. No credit is taken for this long-term decrease in reactivity other than to indicate additional and increasing conservatism in the design criticality analysis.

Holtec International 4-10 ReportHI-971760

4.5 PWR Stora e Rack Criticali Anal ses and Tolerance Variations 4.5.1 Nominal Desi n Case The principal method of analysis for the racks was the CASMO-3 code, using the restart option in CASMO-3 to analytically transfer fuel of a specified burnup into the storage rack configuration at a reference temperature of 4'C (39'F). Calculations were made for fuel of several different initial enrichments and, at each enrichment, a limiting k, value was established which includes reactivity allowances for manufacturing tolerances, the uncertainty in the burnup analyses and for the effect of the axial burnup distribution on reactivity. The restart CASMO-3 calculations (cold, no-Xenon, rack geometry) were then interpolated to define the burnup value yielding the limiting k, value for each enrichment. A line was fitted to these converged burnup values and this line defines the boundary of the acceptable domain shown in Figure 4.2.1.

An independent MCNP-4A calculation was performed to verify the acceptability of the reference criticality analyses. Fuel of 5.0 wt% initial enrichment was analyzed by MCNP-4A and by CASMO-3. The results of this comparison are presented in Table 4.5.1. In comparing the MCNP values to the CASMO-3 values, the MCNP-4A calculational bias and calculational statistics were-included. In addition, the MCNP-4A model correctly included the effect of axial neutron leakage which the CASMO-3 calculations conservatively neglect. Since the MCNP-4A model is at 20 'C and the CASMO-3 model is at 4 'C, a temperature correction had to be applied to the MCNP-4A result. The MCNP-4A result confirms that the reference CASMO-3 calculations are conservative.

4.5.2 Uncertainties Due to PWR Rack Manufacturin Tolerances Allreactivity allowances for manufacturing tolerances are summarized below and listed in Table 4.2.1. Since the tolerances are statistically independent, the allowances are statistically combined into a single reactivity allowance which was used in the final calculations (see Table 4.2.1).

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Boron Loading Tolerances t

1 4.5.2.2 Boral Width Tolerance 4.5.2.3 Tolerance in Cell Lattice Spacing and Cell Box Inner Dimension e Since the Region 2 style racks do not utilize a water gap between storage cells, the manufacturing tolerance on inner box dimension is identical to the tolerance on the storage cell lattice spacing.

4.5.2.4 Stainless Steel Thickness Tolerance 4.5.2.5 Fuel Enrichment and Density Tolerances Holtec International 4-12 ReportHI-971760

4.6 BWR Stora e Rack Criticali Anal ses and Tolerance Variations 4.6.1 Nominal Desi Case The two-dimensional CASMO-3 code was used as the principal method of analysis for the Harris spent fuel pool BWR racks. CASMO-3 was used to perform depletion calculations on the fuel assembly and using the restart option in CASMO-3 the fuel of a specified burnup was analytically transferred into the storage rack at a reference temperature of 4'C (39'F). The same fuel of a specified burnup was also analytically transferred into the standard cold core geometry (SCCG) configuration which is an infinite lattice with 6 inch spacing at a temperature of 20'C without any burnable absorber or control blades and no voids. AllXenon which was present during the depletion calculations was removed during the restarts in the rack and SCCG. The reactivity effects of the natural uranium blanket normally located at the ends of the assemblies were conservatively neglected since an infinite fuel length was used.

All fuel assemblies specified were analyzed at the maximum enrichment specified. The maximum + in the SCCG was specified 1.32. Using the CASMO-3 results, the burnup as corresponding to a ~ in the SCCG of 1.32 was determined and the corresponding k. r in the rack was determined. The reactivity adjustments were added to the rack k; to determine the maximum value and this was compared against the 0.95 k,fr limit. Based on this analysis, all specified fuel assemblies are acceptable for storage as stated in Section 4.2.1.2. Table 4.2.2 provides the final results of the BWR fuel assembly calculations.

An independent MCNP-4A calculation was used to verify the acceptability of the reference criticality analyses. Fuel of 4.6 wt% initial enrichment was analyzed by MCNP-4A and by CASMO-3. The results of this comparison are presented in Table 4.5.1. In comparing the MCNP values to the CASMO values, the MCNP-4A calculational bias and calculational statistics were included. In addition, the MCNP-4A model correctly included the effect of axial neutron leakage which the CASMO-3 calculations conservatively neglect. Since the MCNP-4A model is at 20 'C and the CASMO-3 model is at 4

~

'C, a temperature correction had to be applied to the MCNPPA result. The MCNP-4A result confirm that the reference CASMO-3 calculations are conservative.

~

Holtec International 4-13 ReportHI-971760

4,6.2 Uncertainties Due to Manufacturin Tolerances The reactivity effects associated with manufacturing tolerances are discussed below and shown in Table 4.2.2. Since the tolerances are statistically independent, the allowances are statistically combined into a single reactivity allowance which was used in the final calculations (see Table 4.2.2).

4.6.2.1 Boron Loading Variation F

0 4.6.2.2 Boral Width Tolerance Variation 4.6.2.3 Tolerance in Cell Lattice Pitch and Inner Box Dimension Since the Region 2 style racks do not utilize a water gap between storage cells, the manufacturing tolerance on inner box dimension is identical to the tolerance on the storage cell lattice spacing. ', '. -*

=

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4.6.2.4

~ ~ ~ Stainless Steel Thickness Tolerances 4.6.2.5 Fuel Enrichment and Density Variation The maximum planar average fuel enrichment was specified for each fuel assembly analyzed.

Therefore, there is no reactivity allowance for variations in enrichment since the absolute maximum was used for all calculations.

The UO, density was specified for each fuel assembly analyzed.;,.'."..-.',:,:;: -"'".;~.-;.-:--, -';

4.6.2.6 Zirconium Flow Channel Elimination of the zirconium flow channel results in a small y'::.;<:->)., '~~'; '=<'-=';;~--,

) decrease in reactivity. More significant is a positive reactivity effect resulting Rom potential bulging of the zirconium channel, which moves the channel wall outward toward the Boral absorber. It is conservatively assumed that the maximum bulging that could occur would result in the channel touching the cell walls. Since this would not occur over the entire length of the channel, the model assumed that the entire channel was enlarged so that the mid-point of the channel wall was placed equidistant between the nominal channel outer dimension and the cell wall. This results in an incremental reactivity of ".,",',:y-, i >','-.j; as determined with MCNP-4A.

4.7 Abnormal and Accident Conditions Strict administrative controls on the fuel transfer to Pools C and D willpreclude fuel which is outside the range of the previously stated acceptance criteria &om being brought into the spent Holtec International 4-15 Report HI-971760

fuel pool. Therefore, the only potential abnormal and accident conditions that exist are the misplacement of a fuel assembly outside the rack or the dropping of a fuel assembly on top of the rack. It is also possible to inadvertently place a BWR spent fuel assembly in the PWR rack.

4.7.1 Tem erature and Water Densi Effects The spent fuel pool temperature coefficient of reactivity is negative. Using the minimum temperature of 4'C, therefore, assures that the true reactivity will always be lower than the calculated value regardless of the temperature. Temperature effects on reactivity have been calculated and the results are shown in Table 4.7.1. Introducing voids in the water internal to the storage cell (to simulate boiling) decreased reactivity, as shown in the table. Boiling at the submerged depth of the racks would occur at approximately 122'C.

4.7.2 Dro ed Fuel Assembl For a drop on top of the rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance &om the fuel in the rack of more than 12 inches (which is considered infinite), including an estimated allowance for deformation under seismic or accident conditions. At this separation distance, the effect on reactivity is insignificant.

It is also possible to vertically drop an assembly into a location occupied by another assembly.

Such a vertical impact would at most cause a small compression of the stored assembly, reducing the water-to-fuel ratio and thereby reducing reactivity. In addition the distance between the active fuel regions of both assemblies willbe more than sufficient to ensure no neutron interaction between the two assemblies.

Dropping an assembly into an unoccupied cell could result in a localized deformation of the baseplate of the rack. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in the active fuel height no longer being covered by the boral. The immediate eight surrounding fuel cells could also be affected.

Holtec International 4-16 ReportHI-971760

However, the amount of deformation for these cells would be considerably less. The amount of localized deformation would not exceed three inches for a PWR assembly and would therefore be considerably less for the lighter BWR assembly. The criticality eFect of this drop accident has been conservatively analyzed and it has been shown that this localized event (nine storage cells at most) has a negligible impact on reactivity.

4.7.3 Lateral Rack Movement Lateral motion of the rack modules under seismic conditions could potentially alter the spacing between rack modules. Region 2 storage cells do not use a flux-trap and the reactivity is therefore insensitive to the spacing between modules. The spacing between modules is suQi-ciently large to preclude adverse interaction even with the maximum seismically-induced reduction in spacing.

h 4.7.4

~ Abnormal Location of a PWR or BWR Fuel Assembl

~

Strict administrative controls willprevent an unacceptable assembly, as determined by the

~ ~ ~

acceptance criteria stated in Section 4,2, Rom being transferred to Harris Pools C and D.

Therefore, the only potential mislocation of a fuel assembly is the mislocation of a fuel assembly of equal or lower reactivity to the design basis outside a PWR or BWR rack.

Since the racks willhave a Boral panel on the outside face (when the outside face is not against a wall) the reactivity effect of a misloaded fuel assembly outside the rack is negligible because of the neutron leakage that occurs Rom the rack itself. Therefore, the conservative infinite lattice calculations that were performed have k, values that are higher than any potential mislocation accidents.

Another mislocation event could occur with a BWR assembly. This would be the inadvertent placement of a BWR assembly in the PWR racks. Since, the BWR assembly is significantly smaller than a PWR assembly, the reactivity effect of placing a BWR assembly in the PWR rack is negligible. The reverse scenario of misplacing a PWR Holtec International 4-17 Report HI-971760

4.7,5 Eccentric Fuel Positionin The fuel assembly is assumed to be normally located in the center of the storage rack cell and in the case of the BWR rack there are bottom fittings and spacers that mechanically restrict lateral movement of the fuel assemblies. Nevertheless, MCNP-4A calculations were made with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations indicated that eccentric fuel positioning results in a decrease in reactivity:;, , The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells.

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4.8 References

[4.1.1] M. G. Natrella, Ex erimental Statistics National Bureau of Standards Handbook 91, August 1963.

[4.4.1] A. Ahlin, M. Edenius, H. Haggblom, "CASMO - A Fuel Assembly Burnup Program," AE-RF-764158, Studsvik report (proprietary).

[4.4.2] A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604, 1977.

[4.4.3] M. Edenius, A. Ahlin, and B. H. Forssen, "CASMO-3 A Fuel Assembly Burnup Program, Users Manual", Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986.

[4.4.4] M. Edenius and A. Ahlin, "CASMO-3: New Features, Benchmarking, and Advanced Applications," Nuclear Science and Engineering, 100, 342-351, (1988).

[4.4.5] J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).

[4.4.6] E. E. Pilat, "Methods for the Analysis of Boiling Water Reactors (Lattice Physics),"

YAEC-1232, Yankee Atomic Electric Co., December 1980.

[4.4.7] E. Johansson, "Reactor Physics Calculations on Close-Packed Pressurized Water Reactor Lattices," Nuclear Technology, Vol. 68, pp. 263-268, February 1985.

[4.4.8] S. E. Turner, "Uncertainty Analysis - Burnup Distributions", presented at the DOE/SANDIATechnical Meeting on Fuel Burnup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988.

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Table 4.2.1 Summary of Criticality Safety Calculations for the PWR Fuel Racks Fuel Assembly Westinghouse 15x15 Enrichment 5%

Temperature 4'C Burnup &om Calculation (MWD/MTV) 41,352 Burnup &om Curve (MWD/MTU) 41,447 CASMO-3 Uncertainties

~ 0.9126 UO, density Inner box dimension Box wall thickness Boral width B-10 loading Burnup Total Uncertainty at 95%/95%

Effect of Axial Burnup Distribution Maximum Qr 0.9450 Regulatory Limit 0.9500 Notes:

1. Only the most reactive assembly is shown.

2. The total uncertainty is a statistical combination of the manufacturing uncertainties.

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Table 4.2.2 Summary of Criticality Safety Calculations for the BWR Fuel Racks Fuel Assembly GE3 GE4 GE7 GE8 GE9 GE10 GE13 Temperature 4'C 4'C 4'C 4'C 4'C 4'C 4'C SCCG +

Enrichment 1.32 4.6 1,32 4.6 1.32 4.6 1.32 4.6 1.32 4.6 1.32 4.6 1.32 4.6 CASMO-3 M 0.9163 0.9140 0.9192 0.9214 0.9207 0.9201 0.9227 Uncertainties UO, density Inner box dimension Box wall thickness Boral width B-10 loading Total uncertainty at 95%/95%

Channel bulging Uncertainty for burnup and vendor comparison Maximum k r 0.9379 0.9356 0.9408 0.9430 0.9423 0.9417 0.9443 Regulatory Limit 0.9500 0.9500 0.9500 0.9500 0.9500 0.9500 0.9500 Notes:

1. The total uncertainty is a statistical combination of the manufacturing uncertainties.

2. The GE 13 assembly has part length rods. Two CASMO-3 calculations were performed: one with all rods present and the other with only the full length rods present. The most reactive configuration was the second and the ~ &om this configuration is presented.

3. The GE 5 and GE 6 are identical to the GE 7 for the fuel parameters analyzed and therefore the GE 5 and GE 6 have a maximum ~ equivalent to the GE 7.

4. The enrichment is the planar average enrichment.

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Table 4.3.1 PWR Fuel Characteristics Fuel Assembly Westinghouse Westinghouse Westinghouse Siemens Siemens 17x17 Std 17x17 VS 15x15 17x17 15x15 NOTE: AllDimensions in inches Clad O.D.

Clad LD.

Clad Material Pellet Diameter Stack Density Maximum Enrichment Active Fuel Length Number Fuel Rods Fuel Rod Pitch Number of Thimbles Thimble O.D.

Thimble I.D.

The highlighted data in the table above is the property of Westinghouse or Siemens and is proprietary information provided in confidence. Access to this information shall be limited to those individuals having a need for such access and shall not be disclosed or transmitted to any Holtec International 4-22 ReportHI-971760

Table 4.3.2 BWR Fuel Characteristics Fuel Assembly GE3 GE4 GE7 GES GE9 GE 10 GE 13 NOTE: All dimensions in inches Clad O.D.

Clad I.D.

Clad Material Pellet Diameter Stack Density Maximum Enrichment SCCG k; Active Fuel Length Fuel Rod Array Number Fuel Rods Fuel Rod Pitch Number of Water Rods Water Rod O.D.

Water Rod I.D.

Channel I.D.

Channel Thickness Notes:

I

1. The GE 13 assembly has 8 part length rods.

2. The GE 5 and GE 6 are identical to the GE 7 for the fuel parameters listed.

3. The enrichment is the maximum planar average enrichment.

The highlighted data in the table above is the property of GE and is proprietary information provided in confidence. Access to this information shall be limited to those individuals having a need for such access and shall not be disclosed or transmitted to any organization without the written permission of GE.

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Table 4.4.1 Reactivity Allowance for Uncertainty in Burnup Calculations and the Effect of Axial Burnup Distributions for PWR Fuel Calculated Burnup Applicable Burnup Uncertainty in Burnup Effect of Axial (MWD/MTU) Range (MWD/MTU) Burnup Distribution 45,000 40,000-45,000 40,000 30,000-40,000 30,000 < 30,000 Notes:

1. The uncertainty in burnup was calculated by taking 5% of the reactivity decrement &om zero burnup to the calculated burnup using CASMO-3.

2. The effect of the axial burnup distribution was calculated using MCNP-4A by comparing results f'rom two cases: the first had a uniform axial burnup and the second had a distributed axial burnup distribution represented by 10 axial zones,

3. The effect of the axial burnup distribution is negative at and below 30,000 MWD/MTU, therefore, conservatively no reactivity adjustment was made.

Holtec International 4-24 Report HI-971760

Table 4,5.1 Comparison of MCNP-4A and CASMO-3 Calculations PWR Rack BWR Rack Fuel Assembly W 15x15 GE8 Enrichment 5.0 4.6 Temperature 4'C 4'C MCNP-4A k,rr 1.2004 0.9993 Uncertainties Calculational Statistics Bias Uncertainty Total Uncertainty at 95%/95%

Temperature correction

&om 20'C to 4'C Bias MCNP-4A Maximum k, 1.2056 1.0045 CASMO-3 k,r 1.2076 1.0126 Notes:

1. The MCNP-4A calculation correctly includes the effect of axial neutron leakage.

Holtec International 4-25 ReportHI-971760

Table 4,7.1 Reactivity Effects of Temperature and Void Incremental Reactivity Effect - LQc (relative to reference)

Temperature PWR Rack BWR Rack 4'C (39'F) reference reference 20'C (68'F) 60'C (140'F) 120'C (248'F) 120'C with 10% void Holtec International 4-26 ReportHI-971760

45000 40000 Acceptable Burnup Domain 35000 Burnup=12114~Enrichment-19123 30000 25000 20000 Unacceptab le Burnup Domain 15000 10000 5000 2.5 3 35 4 4.5 Enrichment (wt% U235)

Figure 4.2.1: Burnup Versus Enrichment for PWR Fuel Holtee International 4-27 Report HI-971760

Boral Box Wall G GG W QwQ W W W

wQQ wQ W W W W Q w W G W w QQ wQ W GQ W QwQ W W W = guide tubes Figure 4.3.1: This is a two dimensional representation of the calculational model used for the PWR storage rack analysis showing a Westinghouse 15x15 fuel design. This figure was drawn with the two dimensional plotter in MCNP-4A.

Report HI-971760

Boral Box Wall Channel OO OO OO OO OO OO OO OO OO OO O Qw Qw OO O Qw Qw OO O OO W = water rod Figure 4.3.2: This is a two dimensional representation of the calculational model used for the BWR storage rack analysis showing a GE 8 fuel design. This figure was drawn with the two dimensional plotter in MCNP-4A.

Holtec International 4-29 Report HI-971760

so I TI 5.1 ~n~iQflQl1 This section provides a summary of the methods, models, analyses and numerical results to demonstrate the compliance of Harris Spent Fuel Pools C and D with the provisions of Section IIIof the USNRC "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications", (April 14, 1978) for a bounding configuration. Similar methods of thermal-hydraulic analysis have been used in other rerack licensing projects (see Table 5.1.1).

The thermal-hydraulic qualification analyses for the rack array may be broken down into the following categories:

(i) Evaluation of the long-term decay heat load, which is the cumulative spent fuel decay heat generation from all fuel assemblies stored in the C and D pools.

Evaluation of the steady-state bulk pool temperatures when forced cooling is available. The bulk pool temperatures are required to be maintained z 137'F t under normal conditions with fuel pool cooling in operation.

(iii) Determination of the maximum pool local temperature at steady bulk pool temperatures.

(iv) Evaluation of the potential for flow.bypass &om pool inlet to outlet in the absence of a sparger line to the spent fuel pools racks.

(v) Evaluation of the "time-to-boil" ifall forced heat rejection paths &om the pool are lost.

The 137'F limit is consistent with that currently in the Harris FSAR and procedures for pools A and B. CP&L is in the process of reevaluating systems and components to allow for an increase in the allowable bulk pool temperature.

Holtec International 5-1 Holtec Report HI-971760

This section presents a synopsis of the analysis methods employed, and final results. The decay heat load calculation is conservatively performed in accordance with the provisions of USNRC Branch Technical Position ASB9-2, "Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July, 1981.

The Pool C and D fuel rack configurations for proposed expansion are depicted in Figures 1.2 and 1.3. A total of 1,952 PWR cells and 2,763 BWR cells will be available in a bounding configuration to maximize fuel storage capacity.

To determine the limiting decay heat in the Harris spent fuel pools, a projected bounding decay period for fuel scenario is considered as shown in Table 5.2.1. The in-core irradiation time and limiting assembly specific power inputs are provided in Table 5.2.2. The C and D spent fuel pools (SFPs) are designated to store old fuel which has been cooled for at least 5 years. The fuel is envisaged to be transhipped &om Brunswick and Robinson plants or shuffled from Harris'ools A and/or B.

Since the decay heat load &om the old assemblies varies very slowly as a function of time, the long-term decay heat in the bounding configuration is assumed to be constant. Based on the discharge scenario and fuel assemblies characteristics listed in Tables 5.2.1 and 5.2.2, the combined Pools C and D decay heat rates are determined and summarized in Table 5.2.3.

Holtec International 5-2 Holtec Report HI-971760

5.3 lk The decay heat load to the two pools (C and D) willbe removed by several passive and active

.heat rejection mechanisms, as listed below:

In the interest of conservatism, no credit is applied to removing heat by any of the mechanisms listed above &om (a) to (e). Consequently, all of the decay heat generated in the C and D pools is considered to be removed by the forced flow of SFP cooling water circulating through a heat exchanger, which transfers heat to the CCW system. In a forced SFP cooling scenario, hot water

&om the pool is circulated by a pump through an exchanger cooled by the CCW system. The cooled SFP water is then directed back to the C and D pools. The decay heat load in the C and D pools is Rom old fuel discharges, which is relatively constant (i.e., steady heat load). Therefore, at equilibrium conditions, the total decay heat load to the pool is equal to the heat removed by the cooling system and a constant bulk temperature is maintained in the C and D pools.

The heat removal capacity of the SFP cooling system is principally characterized by two .-

parameters, namely the water circulation flow rate and the fuel pool inlet water temperature. The pool temperature of pools C and D is required to be maintained at or below 137'F t. The

'ulk minimum SFP water flow rate required to comply with this bulk pool temperature criterion is thus a function of the fuel pool inlet water temperature. This requirement is graphically illustrated in Figure 5.3.1. A SFP cooling system design point, which is on the curve, satisfies the to~in'm cooling requirements. A design point above this curve ~eceed the SFP cooling The 137'F limit is consistent with that currently in the Harris FSAR and procedures for pools A and B. CP&L is in the process of re-evaluating systems and components to allow for an increase in the allowable bulk pool temperature.

Holtec International 5-3 Holtec Report HI-971760

~ ~

requirements. Therefore, Figure 5.3.1 establishes the thermal-hydraulic design basis for SFP

~ ~

~

cooling system capacity and the final cooling system design shall comply with these flow vs. ~

inlet temperature parameters.

5.4 c e e tue n l In this section, we present the methodology for calculating the local temperatures when forced JI cooling is available to the Pool C only. The results Rom evaluations performed with forced cooling in pool C only are conservative, since the pool cooling system will be connected to both pools and cooling water willbe discharged to both pools. Therefore, these evaluations predict conservative local temperatures, especially in pool D.

Truncation of sparger lines has become a standard pool modification procedure in rerack campaigns in recent years. Over a dozen SFPs reracked in the past several years have removed sparger lines to enable a high density storage layout and thus maximize pool capacity. Absence of a sparger in the Harris C and D pools removes the mechanistic feed of cold water into the bottom plenum of the fuel racks. It is not apparent Rom heuristic reasoning alone that the cooled water delivered to the pool would not bypass the hot fuel racks and the stored spent fuel in the two pools and exit through the outlet piping. To demonstrate adequate cooling of fuel in the two areas, it is therefore necessary to rigorously quantify the velocity field in the pool created by the interaction of buoyancy driven flows and water ingress/egress. A CFD analysis for this demonstration is required. The objective of this study is to demonstrate that the principal thermal-hydraulic criteria of ensuring local subcooled conditions in the pool is met for the bounding fuel storage configuration, An outline of the CFD approach is described in the following.

Figure 5.4.1 depicts the fuel Pools C and D physical configuration in plan view. The two pools are connected by a transfer canal. Pool piping connections for introducing cooling water and discharge of heated water are shown for both pools. Currently, SFP cooling system design work Holtec International Holtec Report HI-971760

is in progress to provide a forced cooling system which willprovide suction and discharge to both pools. Thermal-hydraulic adequacy of the two pools shall be conservatively demonstrated by assuming that forced cooling is available to only Pool C. Adequate cooling of Pool D is enabled by a buoyancy-driven flow of relatively cooler bulk Pool C water to Pool D through the interconnecting transfer canal. Decay heat inputs to both pools are based on a bounding fuel storage configuration and spent fuel cooling times. The buoyancy-induced cooling of Pool D is demonstrated by performing a rigorous Computational Fluid Dynamics (CFD) analysis of the temperature and flow fields in the two pools. The CFD methodology is discussed in the next subsection. An additional assumption about the location of cooling inlet and outlet piping is included in the analysis to result in an extremely conservative thermal-hydraulic portrayal of the two interconnected pools. The pool cooling inlet and outlet piping connections are assumed to be located on the southeast end of the pool. Thus, forced cooling of the pool is in a diagonally opposite (i.e., farthest) corner from the northwest location of the connection Rom Pool C to the transfer canal. The forced cooling ingress and egress locations are in close proximity to each other and at the same elevation. The potential for flow bypass from inlet to outlet is conservative, since the modeled locations are closer than the actual relative positions.

There are several significant geometric and thermal-hydraulic features of the Harris SFPs which need to be considered for a rigorous CFD analysis. From a fluid flow modeling standpoint, there I

are two regions to be considered. One region is the bulk pool region where the classical Navier-Stokes equations are solved with turbulence effects included. The other region is the heat generating fuel assemblies located in the spent fuel racks located near the bottom of the SFP. In this region, water flow is directed vertically upwards due to buoyancy forces through relatively small flow channels formed by stored fuel assembly rod arrays in each rack cell.

~

Holtec International 5-5 Holtec Report HI-971760

--.:::- '::: -:-.: - Bounding permeability and inertial resistance parameters for the rack cells loaded with PWR or BWR fuel is determined based on &iction factor correlations for laminar flow conditions typically encountered due to low buoyancy induced velocities and small size of the flow channels. A large number of fuel assembly types have been analyzed for hydraulic flow resistance [5.4.1] determination. Table 5.4.1 provides flow resistance parameters which bound all PWR and BWR fuel assembly types which were analyzed in this study.

The pool geometry requires an adequate portrayal of large scale and small scale features, spatially distributed heat sources in the spent fuel racks and water inlet/outlet configuration.

Relatively cooler bulk pool water normally flows down through the narrow fuel rack outline to pool wall liner clearance known as the downcomer. Near the bottom of the racks, the flow turns Rom a vertical to horizontal direction into the bottom plenum supplying cooling water to the rack cells. Heated water issuing out of the top of the racks mixes with the bulk pool water. An adequate modeling of these features on the CFD program involves meshing the large scale bulk pool region and small scale downcomer and bottom plenum regions with sufficient number of computational cells to capture the bulk and local features of the flow field.

Holtec International 5-6 Holtec Report HI-971760

Holtec International 5-7 Holtec Report HI-971760 5.4.1

~ ~ e-Ifall heat exchanger assisted forced pool cooling becomes unavailable, then the pool water will begin to rise in temperature and eventually willreach the normal bulk boiling temperature at 212'F. The time to reach the boiling point will be the shortest when the loss of forced cooling occurs at the point in time when the bulk pool temperature is at its maximum calculated value for a bounding fuel storage configuration. The calculation is conservatively performed for a bounding decay heat load to the pool, no credit for evaporation cooling and no credit for thermal inertia of racks. The amount of water holdup above the racks in the two pools is in excess of 48,000 fP (2.9 x 10'bs) of water. The maximum rate of temperature rise of bulk pools water at a bounding 15.63 million Btu/hr decay heat input (Table 5.2.3) is therefore less than 5.4'F/hr with no water makeup. Ifthe initial temperature is conservatively assumed to be at a uniform maximum bulk average limit of 140'F t, then the time to reach normal boiling point of the bulk pool is in excess of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. This is a relatively long time period for operator action to start makeup water and re-initiate forced cooling to the pool.

5.5 F is f A summary of pools dimensional data used to generate a Computational Fluid Dynamics (CFD) model of the two interconnected C and D pools is provided in Table 5.5.1. The CFD model provides a determination of the difference between the peak local and bulk pool temperatures.

The local temperature corresponding to the maximum bulk pool temperature can then be determined by adding this local temperature rise to the bulk temperature limit. In the CFD model, a minimum bounding downcomer gap between racks outline to pool liner is applied as noted in Table 5.5.1. In this manner, the downcomer water flow path hydraulic resistance is mmimized. Consequently, the local rack cell temperature predictions shall be conservatively f

maximized. The background constant decay heat input to the pool is modeled as a uniform The assumption of an initial temperature of 140'F is conservative, since the bulk pool temperature is currently limited to 137'F.

Holtec International Holtec Report HI-971760

~

volumetric heat source term in the active fuel region of the Pools C and D racks. The total heat generating volume is calculated to be 657 m'. Thus, &om the total decay heat input (Table 5.2.3),

~

the volumetric heat source term is determined to be 6,956 W/m'.

A plan view of the three-dimensional CFD model is presented in Figure 5.5.1. In this view, the two pools with an interconnected transfer canal is depicted. The water inlet/outlet connections are shown modeled in the top left end corner of the Pool C. The racks outline, modeled as a porous media, is depicted in blue color. A perspective view of the CFD model is presented in Figure 5.5.2. The bottom of the transfer canal, as shown in this figure, is at the same elevation as the top of the racks. The average background decay heat is applied to the model as a volumetric heat source term in the active fuel region of the fuel racks. The CFD model of the C and D pools is solved to obtain converged temperature and velocity profiles. The results obtained from the analysis are discussed next.

Peak local water temperature in the rack cells is shown as a contour plot in cross sectional plan view as shown in Figure 5.5.3. The plan view elevation is within the region of the racks above the active fuel region, but below the top of the racks.

An exchange of cold and hot water streams &om the Pool D to Pool C is determined by the CFD solution with only pool C cooled by a forced cooling system. This exchange of cold and hot water between the two bulk pools is illustrated as a flow velocity vectors plot (Figure 5.5.4) in the pools'nterconnecting channel. The peak local temperature is 6.8'F above the water temperature at the cooling system discharge &om pool. Consequently, the peak local temperature corresponding to the maximum bulk pool temperature limit is obtained by adding this local temperature rise. Table 5.5.2 provides the bulk and local temperature summaries. The peak 143.8'F local temperature is below the local water boiling temperature by a large margin.

Figure 5.5.5 provides a flow velocity vectors plot in the pool cooling inlet/outlet piping region.

The pool inlet piping is modeled to be 12 inches below the pool water level and the pool outlet Holtec International Holtec Report HI-971760

~ ~ ~ ~

piping suction is adjacent to the inlet piping discharge at the same elevation. From the velocity

~

vectors plot, it is apparent that no bypass of incoming water to outlet is indicated for an extremely conservative configuration. In the actual pool piping arrangement for Pool C, the water inlet and outlet connections are widely separated. Consequently, it is concluded that any water bypass &om inlet to outlet is not possible.

6 Holtec International 5-10 Holtec Report HI-971760

geeg~ecec;

[5.4.1] Holtec Report HI-951325, "HI-STAR 100 System Thermal Design Package".

[5.4.2] "QA Documentation and Validation of the FLUENT Version 4.3 CFD Analysis Program", Holtec Report HI-961444.

[5.4.3] Batchelor, G.K., "An Introduction to Fluid Dynamics", Cambridge University Press, 1967.

[5.4.4] Hinze, J.O., "Turbulence", McGraw HillPublishing Co., New York, NY, 1975.

[5.4.5] Launder, B.E., and Spalding, D.B., "Lectures in Mathematical Models of Turbulence", Academic Press, London, 1972.

Holtec International Holtec Report HI-971760

Table 5.1.1 PARTIAL LISTING OF RERACK APPLICATIONS USING SIMILARMETHODS OF THERMAL-HYDRAULICANALYSIS PLANT DOCKET NO.

Enrico Fermi Unit 2 USNRC 50-341 Quad Cities 1 and 2 USNRC 50-254, 50-265 Rancho Seco USNRC 50-312 Grand Gulf Unit 1 USNRC 50-416 Oyster Creek USNRC 50-219 Pilgrim USNRC 50-293 V.C. Summer USNRC 50-395 Diablo Canyon Units 1 and 2 USNRC 50-275, 50-455 Byron Units 1 and 2 USNRC 50-454, 50-455 Braidwood Units 1 and 2 USNRC 50-456, 50-457 Vogtle Unit 2 USNRC 50-425 St. Lucie Unit 1 USNRC 50-425 Millstone Point Unit 1 USNRC 50-245 D.C. Cook Units 1 and 2 USNRC 50-315, 50-316 Indian Point Unit 2 USNRC 50-247 Three Mile Island Unit 1 USNRC 50-289 J.A. FitzPatrick USNRC 50-333 Shearon Harris Unit 2 USNRC 50-401 Hope Creek USNRC 50-354 Kuosheng Units 1 and 2 Taiwan Power Company Chin Shan Units 1 and 2 Taiwan Power Company Holtec International 5-12 Holtec Report HI-971760

Table 5.1,1 (continued)

PARTIALLISTING OF RERACK APPLICATIONS USING SIMILARMETHODS OF THERMAL-HYDRAULICANALYSIS PLANT DOCKET NO.

Ulchin Unit 2 Korea Electric Power Corporation Laguna Verde Units 1 and 2 Comision Federal de Electricidad Zion Station Units 1 and 2 USNRC 50-295, 50-304 Sequoyah USNRC 50-327, 50-328 La Salle Unit One USNRC 50-373 Duane Arnold USNRC 50-331 Fort Calhoun USNRC 50-285 Nine Mile Point Unit One USNRC 50-220 Beaver Valley Unit One USNRC 50-334 Limerick Unit 2 USNRC 50-353 Ulchin Unit 1 Korea Electric Power Corporation Holtec International Holtec Report HI-971760

Table 5,2.1 DECAY PERIODS FOR A BOUNDlNG POOLS C AND D STORAGE CONFIGURATION PWR Fuel Assemblies BWR Fuel Assemblies Number of Assys Decay Period Number of Assys Decay Period 172 5 years 456 5 years 172 7 years 456 7 years 172 9 years 456 9 years 172 11 years 456 11 years 172 13 years 456 13 years 172 15 years 483 15 years 172 17 years 172 19 years 172 21 years 172 23 years 232 25 years Holtec International 5-14 Holtec Report HI-971760

Table 5.2.2 FUEL ASSEMBLIES INPUT DATAFOR DECAY HEAT EVALUATION Item Value PWR Assembly Irradiation Time 1,915 EFPDt PWR Assembly Specific Power 19.11 MWt BWR Assembly Irradiation Time 2,028 EFPD BWR Assembly Specific Power 4.66 MWt Effective Full Power Days Holtec International 5-15 Holtec Report HI-971760

Table 5.2.3 BOUNDING DECAY HEAT INPUT FROM STORED FUEL IN POOLS C AND D Decay Heat Load Fuel Assemblies (MillionBtu/hr)

BWR Fuel Assemblies 4.47 PWR Fuel Assemblies 11.16 Total 15.63 (4.57 MW)

Holtec International 5-16 Holtec Report HI-971760

Table 5.4.1 BOUNDING FUEL ASSEMBLIES HYDRAULICFLOW RESISTANCE PAIVQvKTERS Parameter Value Permeability 10~ m~

Inertial Resistance Factor 95 m'oltec International 5-17 Holtec Report HI-971760

Table 5.5.1 POOLS C AND D DIMENSIONALDATA Parameter Value Pool C: Length 597.88" Width 320.60" Pool D: Length 383.36" Width 237 79" Water Depth 38.5 ft Pools-to-Transfer Canal Channel Width 24" Bottom Plenum 6 II Pool C Downcomers North Wall 1.44"t South Wall 1.44" East Wall 2.36" West Wall 2.36" Pool D Downcomers North Wall 5.15" South Wall 5 OII East Wall 5 OII West Wall 5 Oll A minimum uniform downcomer gap equal to 1.44" applied to both pools for CFD analysis.

Holtec International 5-18 Holtec Report HI-971760

Table 5.5.2 BULKAND LOCAL TEMPERATURES

SUMMARY

Item Temperature ('F) t Local temperature rise above bulk 6.8 Bulk pool maximum temperature limit 137.0 Peak Local Temperature 143.8 Local temperature values are conservatively computed based on neglecting forced cooling to pool D, as discussed at the beginning of Section 5.4 Holtec International 5-19 Holtec Report HI-971760

3000 2500 E

0) o 2000 u ACCEPTABLE L REGION

~ 1500 E

E C

UN-ACCEPTAB E REGION 1000 500 100 105 110 115 120 125 Spent Fuel Pool Incoming Water Temperature [Deg FI FIGURE 5.3.1: C AND D POOLS MINIMUMTOTAL COOLING SYSTEM REQUIREMENTS CURVE AT 137 Deg. F BULK POOL TEMPERATURE 5-20

EL. 279'-6 MIT TO SCALE El. 279'-6 I

I 2M.O 174.0 I 224,0 l98.0 I 2I6.0 I

I I

I92.0

',EL 278-6 240.0 g EL 278'-6 EL 278'-3 I EL 278'-3 I.

I 279'-6

-',~-j 69.0 L EL

%.0 KN YIEY TRNSFER CNAL YATER LEYEL CKIMi SYSTEH PIPES OTIERS CAHAL 3SF!6-ISA-2&3 ( l6') 9 7SF4-218-2&3 m.x 3SFI6-2SB-2&3 ( l6') PB 7SF3-lm-Za3 CANAL 3SFI2-6SB-2&3 (12')

(9 3SFIHSA-2&3 ( l2')

3SF(2-I74SA-2Q ( l2')

3SFI2-I7IS8-2&3 ( l2')

3SFI2-I76SB-2&3 ( l2')

IS.BI25 RAES 3SFI2-179SA-2&3 ( 12')

KIT(I( EL 24b"-0 ELEYATlm YIB( FROH SRJTH TO MRTH FIGURE 5.4.1: HARRIS C AND D POOLS PHYSICAL CONFIGURATION 70324%197178h5 4 I 5-21

FlGURE 5.51; PLAN VIEW OF THE HARRIS POOLS C AND D CFD MODEL 5-22 HI-971760 Holtec Proprietary

FIGURE 5.5.2; PERSPECTIVE VIEW OF THE HARRIS POOLS C AND D CFD MODEL Holtec Proprietary 5-23 Hl-97 I760

FIGURE 5.53; PEAK LOCAL WATER TEMPERATURE IN THE RACK CELLS 5-24 Hl-971760 Holtec Proprietary

FIGURE 5.5.4; POOLS INTERCONNECTING CHANNEL FLOW VELOCITY VECTORS ELEVATION VIEW PLOT Holtec Proprietary 5-25 HI-971760

FIGURE 5.5.5; POOL COOLING INLET/OUTLET PIPING REGION FLOW VELOCITYVECTORS PLOT Holtec Proprietary 5-26 HI-971760

6.1 This section considers the structural adequacy of the new maximum density spent fuel racks under all loadings postulated for normal, seismic, and accident conditions at Harris. The existing spent fuel storage racks are also examined for stability during the installation process.

The analyzed storage rack configurations with the new racks in place are shown in Figures 1.2 and 1.3.

The analyses undertaken to confirm the structural integrity of the racks are performed in compliance with the USNRC Standard Review Plan [6.1.1] and the OT Position Paper [6.1.2].

For each of the analyses, an abstract of the methodology, modeling assumptions, key results, and summary of parametric evaluations are presented. Delineation of the relevant criteria are discussed in the text associated with each analysis.

The response of a free-standing rack module to seismic inputs is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning), resulting in impacts and friction effects. Some of the unique attributes of the rack dynamic behavior include a large fraction of the total structural mass in a confined rattling motion, friction support of rack pedestals against lateral motion, and large fluid coupling effects due to deep submergence and independent motion of closely spaced adjacent structures.

Linear methods, such as modal analysis and response spectrum techniques, cannot accurately simulate the structural response of such a highly nonlinear structure to seismic excitation. An accurate simulation is obtained only by direct integration of the nonlinear equations of motion with the three pool slab acceleration time-histories applied as the forcing functions acting simultaneously, Holtec International 6-1 Holtec Report HI-971760

Whole Pool Multi-Rack (WPMR) analysis is the vehicle utilized in this project to simulate the

~ ~

dynamic behavior of the complex storage rack structures. The following sections provide the basis for this selection and discussion on the development of the methodology.

6.2.1 Reliable assessment of the stress field and kinematic behavior of the rack modules calls for a conservative dynamic model incorporating all key attributes of the actual structure. This means that the model must feature the ability to execute the concurrent motion forms compatible with the free-standing installation of the modules.

The model must possess the capability to effect momentum transfers which occur due to rattling of fuel assemblies inside storage cells and the capability to simulate lift-offand subsequent impact of support pedestals with the pool liner (or bearing pad). The contribution of the water mass in the interstitial spaces around the rack modules and within the storage cells must be modeled in an accurate manner since erring in quantification of fiuid coupling on either side of the actual value is no guarantee of conservatism.

The Coulomb friction coefficient at the pedestal-to-pool liner (or bearing pad) interface may lie in a rather wide range and a conservative value of friction cannot be prescribed a priori. In fact, a perusal of results of rack dynamic analyses in numerous dockets (Table 6.2.1) indicates that an upper, bound value of the coefficient of friction often maximizes the computed rack displacements as well as the equivalent elastostatic stresses.

In short, there are a large number of parameters with potential influence on the rack kinematics. The comprehensive structural evaluation must deal with all of these without sacrificing conservatism.

Holtec International 6-2 Holtec Report HI-971760

~ ~ ~ ~

The three-dimensional single rack dynamic model introduced by Holtec International in the

~ ~

Enrico Fermi Unit 2 rack project (ca. 1980) and used in some 50 rerack projects since that

~

time (Table 6.2.1) addresses most of the above mentioned array of parameters. The details of this methodology are also published in the permanent literature [6.2. 1]. Despite the versatility of the 3-D seismic model, the accuracy of the single rack simulations has been suspect due to one key element; namely, hydrodynamic participation of water around the racks. During dynamic rack motion, hydraulic energy is either drawn from or added to the moving rack, modifying its submerged motion in a significant manner. Therefore, the dynamics of one rack affects the motion of all others in the pool.

A dynamic simulation which treats only one rack, or a small grouping of racks, is intrinsically inadequate to predict the motion of rack modules with any quantifiable level of accuracy.

Three-dimensional Whole Pool Multi-Rack analyses carried out on several previous plants demonstrate that single rack simulations under predict rack displacement during seismic responses [6.2.2].

Briefly, the 3-D rack model dynamic simulation, involving one or more spent fuel racks, handles the array of variables as follows:

Parametric runs are made with upper bound and lower bound values of the coefficient of friction, The limiting values are based on experimental data which have been found to be bounded by the values 0.2 and 0.8. Simulations are also performed with the array of pedestals having randomly chosen coefficients of friction in a Gaussian distribution with a mean of 0.5 and lower and upper limits of 6.2 and 0.8, respectively. In the fuel rack simulations, the Coulomb friction interface between rack support pedestal and liner is simulated by piecewise linear (friction) elements. These elements function only when the pedestal is physically in contact with the pool liner.

Holtec International 6-3 Holtec Report HI-971760

~

fl kl lyly,kl

~

fl

~

kk,l'dd'Pdlky

~

~ ~ ~

introducing linear springs to represent the elastic bending action, twisting, and extensions. ~

fl p l -Hygpl dp ldl p'gd closing of interfaces such as the pedestal-to-bearing pad interface, and the fuel assembly-to-cell wall interface. These interface gaps are modeled using nonlinear spring elements. The term "nonlinear spring" is a generic term used to denote the mathematical representation of the condition where a restoring force is not linearly proportional to displacement.

The fuel assemblies are conservatively assumed to rattle in unison which obviously exaggerates the contribution of impact against the cell wall.

kflflgypHH Hl I <<l l

d dl'l '

l \ -kdyfl ld pl' dl to multiple bodies and utilized it to perform the first two-dimensional multi-rack analysis (Diablo Canyon, ca. 1987). Subsequently, laboratory experiments were conducted to validate

~ ~

the multi-rack fluid coupling theory. This technology was incorporated in the computer code

~

DYNARACK(a.k.a. MR216) [6.2.4] which handles simultaneous simulation of all racks in the pool as a Whole Pool Multi-Rack 3-D analysis. This development was first utilized in Chinshan, Oyster Creek, in earlier projects at the Harris plant [6.2.1, 6.2.3] and, subsequently, in numerous other rerack projects. The WPMR analyses have corroborated the accuracy of the single rack 3-D solutions in predicting the maximum structural stresses, and also serve to improve predictions of rack kinematics.

For closely spaced racks, demonstration of kinematic compliance is verified by including all modules in one comprehensive simulation using a WPMR model. In WPMR analysis, all rack modules are modeled simultaneously and the coupling effect due to this multi-body motion is included in the analysis. Due to the superiority of this technique in predicting the dynamic behavior of closely spaced submerged storage racks, the Whole Pool Multi-Rack analysis methodology is used for this project.

Holtec International Holtec Report HI-971760

63 The implementation of the storage capacity increase in pools C and D will be performed on an as needed basis through incremental phases (campaigns). Figures 6.3.1 and 6.3.2 identify the fully implemented configuration and also designates which racks will be included in each of the campaigns. The new high density storage racks are analyzed for the anticipated configurations at the completion of each of the installation campaigns. Evaluated configurations of the two pools are also handled separately, since the pools are physically separated by the surrounding concrete walls. The analyzed configurations considered are described as follows:

Incremental Incremental Number of 14 1680 10 1260 6 750

~ 500 525 The materials utilized in fabrication of the rack components are identified in Table 6,3.1.

The cartesian coordinate system utilized within the rack dynamic model has the following nomenclature:

x = Horizontal axis along plant North y = Horizontal axis along plant West z = Vertical axis upward from the rack base 6.3.1 For the dynamic rack simulations, the dry PWR fuel weight is taken to be 1600 lbs and the dry BWR fuel weight is taken to be 680 lbs.

Holtec International 6-5 Holtec Report HI-971760

The synthetic time-histories in three orthogonal directions (N-S, E-W, and vertical) are generated in accordance with the provisions of SRP 3.7.1 [6.4.1]. In order to prepare an acceptable set of acceleration time-histories, Holtec International's proprietary code GENEQ f6.4.2] is utilized.

A preferred criterion for the synthetic time-histories in SRP 3.7.1 calls for both the response spectrum and the power spectral density corresponding to the generated acceleration time-history to envelope their target (design basis) counterparts with only finite enveloping infractions. The time-histories for the pools have been generated to satisfy this preferred (and more rigorous) criterion. The seismic files also satisfy the requirements of statistical independence mandated by SRP 3,7.1.

~

Figures 6.4.1 through 6.4.3 and 6.4.4 through 6.4.6 provide plots of the time-history

~ ~ ~ ~

accelerograms which were generated over a 20 second duration for OBE and SSE events, respectively.

Results of the correlation function of the three time-histories are given in Table 6.4.1.

Absolute values of the correlation coefficients are shown to be less than 0.15, indicating that the desired statistical independence of the three data sets has been met.

Holtec International 6-6 Holtec Report HI-971760

6S Recognizing that the analysis work effort must deal with both stress and displacement criteria, the sequence of model development and analysis steps that are undertaken are summarized in the following:

6.5.1 The dynamic modeling of the rack structure is prepared with special consideration of all nonlinearities and parametric variations. Particulars of modeling details and assumptions for the Whole Pool Multi-Rack analysis of racks are given in the following:

6.5.1.1 Holtec International 6-7 Holtec Report HI-971760

Holtec International 6-8 Holtec Report HI-971760 6.5.1.2 Figure 6.5 ~ 1 shows a schematic of the dynamic model of a single rack. The schematic depicts many of the characteristics of the model including all of the degrees-of-freedom and some of the spring restraint elements.

Table 6.5.1 provides a complete listing of each of the 22 degrees-of-freedom for a rack model.

Six translational and six rotational degrees-of-freedom (three of each type on each end) describe the motion of the rack structure. Rattling fuel mass motions (shown at nodes 1', 2',

3, 4, and 5'n Figure 6.5. 1) are described by ten horizontal translational degrees-of-freedom (two at each of the five fuel masses). The vertical fuel mass motion is assumed ( and modeled) to be the same as that of the rack baseplate.

Figure 6.5.2 depicts the fuel to rack impact springs (used to develop potential impact loads between the fuel assembly mass and rack cell inner walls) in a schematic isometric. Only one of the five fuel masses is shown in this figure. Pour compression only springs, acting in the horizontal direction, are provided at each fuel mass.

Figure 6.5.3 provides a 2-D schematic elevation of the storage rack model, discussed in more detail in Section 6.5.3. This view shows the vertical location of the five storage masses and some of the support pedestal spring members.

Figure 6.5.4 shows the modeling technique and degrees-of-freedom associated with rack elasticity. In each bending plane a shear and bending spring simulate elastic effects [6.5.4].

Holtec International 6-9 Holtec Report HI-971760

~ ~

Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also included

~

in the model. ~

Figure 6.5.5 depicts a single rack module with its surrounding impact springs (used to develop potential impact loads between racks or between rack and wall). Figures 6.5.6 through 6.5.13 show the rack numbering schemes used for the WPMR analyses of both pools. These figures also provide the numbering scheme for all of the rack periphery compression only gap elements.

6.5.2 In its simplest form, the so-called "fluid coupling effect" [6.5.2, 6.5.3] can be explained by considering the proximate motion of two bodies under water. Ifone body (mass m,) vibrates adjacent to a second body (mass re), and both bodies are submerged in frictionless fluid, then Newton's equations of motion for the two bodies are:

(mi + Mii) Xi + Mtg X2 applied forces on mass m, + 0 (X, )

M>> X, + (m, + M22) X, = applied forces on mass mt + 0 (Xt')

Xand X, denote absolute accelerations of masses m, and mrespectively, and the notation O(X ) denotes nonlinear terms.

Mti Mi2 +t aild M22 are fluid coupling coefficients which depend on body shape, relative disposition, etc. Fritz [6.5.3] gives data for M; for various body shapes and arrangements.

The fluid adds mass to the body (M<< to mass m,), and an inertial force proportional to acceleration of the adjacent body (mass aQ. Thus, acceleration of one body affects the force field on another. This force field is a function of inter-body gap, reaching large values for small gaps. Lateral motion of a fuel assembly inside a storage location encounters this effect.

Holtec International 6-10 Holtec Report HI-971760

For example, fluid coupling behavior will be experienced between nodes 2 and 2* in Figure 6.5.1. The rack analysis also contains inertial fluid coupling terms which model the effect of fluid in the gaps between adjacent racks.

Terms modeling the effects of fluid flowing between adjacent racks in a single rack analysis suffer from the inaccuracies described earlier. These terms are usually computed assuming that all racks adjacent to the rack being analyzed are vibrating in-phase or 180'ut of phase.

The WPMR analyses do not require any assumptions with regard to phase.

Rack-to-rack gap elements have initial gaps set to 100% of the physical gap between the racks or between outermost racks and the adjacent pool walls.

6.5.2.1 Holtec International 6-11 Holtec Report HI-971760

6.5.3 Table 6.5.2 lists all spring elements used in the 3-D, 22-DOF, rack model for Campaign I of pool'D. This set of elements is chosen since it represents the smallest of the models and provides a sufficient example to describe spring element numbering of Campaign II of pool D and the larger pool C models, which are similar. Three element types are used in the rack models. Type 1 are linear elastic elements used to represent the beam-like behavior of the integrated rack cell matrix. Type 2 elements are the piece-wise linear friction springs used to develop the appropriate forces between the rack pedestals and the supporting bearing pads.

Type 3 elements are non-linear gap elements which model gap closures and subsequent impact

~

loadings (i.e., between fuel assemblies and the storage cell inner walls, and rack outer

~

periphery spaces.

A detailed numbering scheme for the rack-to-rack and rack-to-wall gap elements for each of the pool models is provided in Figures 6.5.6 through 6.5.13.

Ifthe simulation model is restricted to two dimensions (one horizontal motion plus one vertical motion, for example), for the purposes of model clarification only, then Figure 6.5.3 describes the configuration. This simpler model is used to elaborate on the various stiffness modeling elements.

Type 3 gap elements modeling impacts between fuel assemblies and racks have local stiffness Kt in Figure 6.5.3. In Table 6.5.2, for example, type 3 gap elements 5 through 8 act on the rattling fuel mass at the rack top. Support pedestal spring rates K are modeled by type 3 gap Holtec International 6-12 Holtec Report HI-971760

~ ~ ~

elements 1 through 4, as listed in Table 6.5.2. Local compliance of the concrete floor is

~

~ ~

included in K,. The type 2 friction elements listed in Table 6.5.2 are shown in Figure 6.5.3 as

~ ~

The spring elements depicted in Figure 6.5.4 represent type 1 elements.

Friction at support/liner interface is modeled by the piecewise linear friction springs with suitably large stiffness K, up to the limiting lateral load pN, where N is the current compression load at the interface between support and liner. At every time-step during transient analysis, the current value of N (either zero ifthe pedestal has lifted off the liner, or a compressive finite value) is computed.

The gap element K, modeling the effective compression stiffness of the structure in the vicinity of the support, includes stiffness of the pedestal, local stiffness of the underlying pool slab, and local stiffness of the rack cellular structure above the pedestal.

The previous discussion is limited to a 2-D model solely for simplicity. Actual analyses incorporate 3-D'motions and include all stiffness elements listed in Table 6.5.2.

6.5.4 To eliminate the last significant element of uncertainty in rack dynamic analyses, multiple simulations are performed to adjust the friction coefficient ascribed to the support pedestal/pool bearing pad interface. These friction coefficients are chosen consistent with the two bounding extremes from Rabinowicz's data [6.5.1]. Simulations are also performed by imposing intermediate value friction coefficients developed by a random number generator with Gaussian normal distribution characteristics. The assigned values are then held constant during the Holtec International 6-13 Holtec Report HI-971760

entire simulation in order to obtain reproducible results. Thus, in this manner, the WPMR analysis results are brought closer to the realistic structural conditions.

6.5.5 Using the structural model discussed in the foregoing, equations of motion corresponding to each degree-of-freedom are obtained using Lagrange's Formulation [6.5.4]. The system kinetic energy includes contributions from solid structures and from trapped and surrounding fluid.

The final system of equations obtained have the matrix form:

1 (Et 2

= [Q] + [G]

where:

total mass matrix (including structural and fluid mass contributions). The size of this matrix will be 22n x22n for a WPMR analysis (n = number of racks in the model).

It is noted that MR216 has the capability to change the coefficient of friction at any pedestal at each instant'of contact based on a random reading of the computer clock cycle. However, exercising this option would yield results that could not be reproduced. Therefore, the random choice of coefficients is made only once per run.

Holtec International 6-14 Holtec Report HI-971760

the nodal displacement vector relative to the pool slab displacement (the term with q indicates the second derivative with respect to time, i.e., acceleration)

[G] a vector dependent on the given ground acceleration

[Ql a vector dependent on the spring forces (linear and nonlinear) and the coupling between degrees-of-freedom

'l The above column vectors have length 22n. The equations can be rewritten as follows:

Nl '0]

d q = + Wl dE This equation set is mass uncoupled, displacement coupled at each instant in time. The numerical solution uses a central difference scheme built into the proprietary computer program MR216 [6.2.4].

There are two sets of criteria to be satisfied by the rack modules:

a.

Per Reference [6.1. 1], in order to be qualified as a physically stable structure it is necessary to demonstrate that an isolated rack in water would not overturn when an event of magnitude:

~ 1.5 times the upset seismic loading condition is applied.

~ 1.1 times the faulted seismic loading condition is applied.

b.

Stress limits must not be exceeded under the postulated load combinations provided herein.

Holtec International 6-15 Holtec Report HI-971760

6.6.2 The stress limits presented below apply to the rack structure and are derived from the ASME Code, Section III, Subsection NF [6.6. 1]. Parameters and terminology are in accordance with the ASME Code. Material properties are obtained from the ASME Code, Section II, Part D

[6.6.2), and are listed in Table 6.3.1.

a. Allowable stress in tension on a net section is:

F, = 0.6 Sr Where, S= yield stress at temperature, and F, is equivalent to primary membrane stress.

Allowable stress in shear on a net section is:

F=

.4',

Allowable stress in compression on a net section F, =S .47- kt 444 r kl/r for the main'rack body is based on the full height and cross section of the honeycomb region and does not exceed 120 for all sections.

unsupported length of component length coefficient which gives influence of boundary conditions. The following values are appropriate for the described end conditions:

1 (simple support both ends) 2 (cantilever beam)

Vi (clamped at both ends) radius of gyration of component Holtec International 6-16 Holtec Report HI-971760

d. Maximum allowable bending stress at the outermost fiber of a net section, due

" to flexure about one plane of symmetry is; F, = 0.60 S(equivalent to primary bending)

e. Combined bending and compression on a net section satisfies:

~a mx fbx my fby where:

f, = Direct compressive stress in the section f== Maximum bending stress along x-axis Maximum bending stress along y-axis f>>

C = 0.85 Cy = 0.85 D= 1-(f,/F',)

F ~ (m E)/(2 1 5 (kl/r)z y)

E = Young's Modulus and subscripts x,y reflect the particular bending plane.

f. Combined flexure and compression (or tension) on a net section:

f.

0.6S

+

fbx + fby Fbx Fb

<10 The above requirements are to be met for both direct tension or compression.

g. Welds Allowable maximum shear stress on the net section of a weld is given by; F= 0.3 S where S is the weld material ultimate strength at temperature. For fillet weld legs in contact with base metal, the shear stress on the gross section is limited to 0.4S, 'where S is the base material yield strength at temperature.

Holtec International 6-17 Holtec Report HI-97,1760

Section F-1334 (ASME Section III, Appendix F) [6.6.2], states that the limits for the Level D condition are the minimum of 1.2 (QFP or (0.7S/Fg times the corresponding limits for the Level A condition, S is ultimate tensile stress at the specified rack design temperature. Examination of material properties for 304L stainless demonstrates that 1.2 times the yield strength is less than the 0.7 times the ultimate strength.

Exceptions to the above general multiplier are the following:

a) Stresses in shear shall not exceed the lesser of 0.72Sor 0.42S. In the case of the Austenitic Stainless material used here, 0.72Sgoverns.

b) Axial Compression Loads shall be limited to 2/3 of the calculated buckling load.

c) Combined Axial Compression and Bending - The equations for Level A conditions shall apply except that:

F, = 0.667 x Buckling Load/ Gross Section Area, and the terms F',and F'~ may be increased by the factor 1,65.

d) For welds, the Level D allowable maximum weld stress is not specified in Appendix F of the ASME Code. An appropriate limit for weld throat stress is conservatively set here as:

F= (0.3 Sg x factor where:

factor = (Level D shear stress limit)/(Level A shear stress limit) 6.6.3 For convenience, the stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value.

The limiting value of. each stress factor is 1.0, based on the allowable strengths for each level, for Levels A, B, and D (where 1.2$ ( .7'.

Holtec International 6-18 Holtec Report HI-971760

The stress factors reported are:

I R, = Ratio of direct tensile or compressive stress on a net section to its allowable value (note pedestals only resist compression)

R, = Ratio of gross shear on a net section in the x-direction to its allowable value R, = Ratio of maximum x-axis bending stress to its allowable value for the section R, = Ratio of maximum y-axis bending stress to its allowable value for the section R, = Combined fiexure and compressive factor (as defined in the foregoing)

R6 = Combined fiexure and tension (or compression) factor (as defined in the foregoing)

R, = Ratio of gross shear on a net section in the y-direction to its allowable value Holtec International 6-19 Holtec Report HI-971760

6.6.4 The applicable loads and their combinations which must be considered in the seismic analysis of rack modules are excerpted from Refs. [6.1.2] and [6.6.3].

The load combinations considered are identified below:

Loading Combination Service Level D+L A D+L+T, D+L+T +E D+L+T,+E'evel D+L+T,+E D+L+T +P Level B Level D D+L+T,+Fd The functional capability of the fuel racks must be demonstrated.

D Dead weight-induced loads (including fuel assembly weight)

L Live Load (not applicable for the fuel rack, since there are no moving objects in the rack load path)

P, Upward force on the racks caused by postulated stuck fuel assembly Fd = Impact force from accidental drop of the heaviest load from the maximum possible height.

Operating Basis Earthquake (OBE)

EI Safe Shutdown Earthquake (SSE)

T0 Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition)

Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions)

T, and T, produce local thermal stresses. The worst thermal stress field in a fuel rack is obtained when an isolated storage location has a fuel assembly generating heat at maximum postulated rate and surrounding storage locations contain no fuel. Heated water makes unobstructed contact 'with the inside of the storage walls, thereby producing maximum possible Holtec International 6-20 Holtec Report HI-971760

temperature difference between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience secondary (thermal) stresses.

6.7 Whole Pool Multi-Rack (WPMR) simulations have been performed to investigate the structural integrity of each rack array. Pools C and D had separate runs performed for the SSE seismic event considering pools filled and partially filled with racks. The partially filled pools represent interim configurations subsequent to the installation campaigns identified for each pool in Figures 6.3.1 and 6.3.2. The configurations were considered with friction coefficients of 0.8, 0.2, and a guassian distribution with a mean of 0.5 (i.e., random coefficient of friction (COF) with upper and lower limits of 0.8 and 0.2). The SSE simulations were performed and conservatively compared against the allowables for OBE events. This process eliminated the need for performing OBE simulations to significantly reduce the number of runs needed. Due

~

to the mild SSE earthquake postulated for Harris, this conservative evaluation technique

~ ~

yielded satisfactory design margins.

The overturning check simulations were performed to determine the behavior of the highest aspect (width/length) ratio racks under both the OBE and SSE events. The overturning check simulations considered a single rack (i.e., no dynamic fluid coupling to walls or other racks) half full with fuel all loaded along the long side of the rack.

The rack numbering schemes used to identify the racks in each simulation model are introduced in Figures 6.5.6 through 6.5.13. The circled rack numbers in the figures correspond to the rack numbers shown in the following tables.

Holtec International Holtec Report Hl-971760

The following table presents a complete listing of the simulations discussed herein.

Consideration of the parameters described above resulted in the following runs:

Pool C (Campaign I) 0.8 SSE Pool C (Campaign I) 0.2 SSE Pool C (Campaign I) Random SSE Pool C (Campaign II) 0.8 SSE Pool C (Campaign II) 0.2 SSE Pool C (Campaign II) Random SSE Pool C (Campaign III - Full) 0.8 SSE Pool C (Campaign Ill - Full) 0.2 SSE Pool C (Campaign III - Full) Random SSE

'.8 10 Pool D (Campaign I) SSE Pool D (Campaign I) 0.2 SSE 12 Pool D (Campaign I) Random SSE 13 Pool D (Campaign II - Full) 0.8 SSE 14 Pool D (Campaign II - Full) 0.2 SSE; 15 Pool D (Campaign II - Full) Random SSE 16 Single Holtec Rack 0.8 OBE x 1.5 Overturning Check 17 Single HoltecRack 0.8 SSE x 1.1 Overturning Check Holtec International 6-22 Holtec Report HI-971760

68 The results from the MR216 runs may be seen in the raw data output files. The MR216 output files archive all of the loads and displacements at key locations within each of the rack modules at every time step throughout the entire time history duration. However, due to the huge quantity of output data, a post-processor is used to scan for worst case conditions and develop the stress factors discussed in subsection 6.6.3.

Further reduction in this bulk of information is provided in this section by extracting the worst case values from the parameters of interest; namely displacements, support pedestal forces, impact loads, and stress factors. This section also summarizes other analyses performed to develop and evaluate structural member stresses which are not determined by the post processor.

6.8.1

~ ~

A tabulated summary of the maximum displacement for each simulation is provided below with the location/direction terms defined as follows:

uxt = displacement of top corner of rack, relative to the slab, in the East-West direction for pool C racks and in the North-South direction for pool D rack modules.

uyt = displacement of top corner of rack, relative to the slab, in the North-South direction for pool C racks and in the East-West direction for pool D rack modules.

Simulations 16 and 17 were performed to evaluate the potential for overturning of a single Holtec rack isolated in the pool without any fluid coupling to adjacent racks or walls. This simulation was performed to account for the unlikely possibility of a seismic event occurring during the installation process.

Holtec International 6-23 Holtec Report HI-971760

The following maximum rack displacements (in inches) are obtained for each of the runs:

Pool Event Run COF Maximum Location/ Rack Displacement Direction (inches)

Pool C Campaign I SSE 0.8 1.132 16 SSE 0.2 0.631 SSE Random 0.878 16 Pool C Campaign II SSE 0.8 1.494 28 SSE 0.2 0.917 SSE Random 0.878 16 Pool C Campaign III SSE 0.8 0.617 uyt 29 SSE 0.2 0.740 SSE Random 0.684 Uyt Pool D Campaign I SSE 10 0.8 0.520 SSE 0.2 0.390 uyt SSE 12 Random 0.521 Pool D Campaign II SSE 13 0.8 0.575 uyt SSE 14 0.2 0.595 SSE 15 Random 0.576 Tipover: Single Holtec Rack OBE 16 0.8 0.347 uyt PWR Tipover: Single Holtec Rack SSE 17 0.8 1.054 uyt The largest displacement of 1.494 occurs in run 4 for rack 28 in the X direction. Since this displacement maintains the centroid of the rack well within the boundaries represented by the support pedestals, there is no possibility of rack overturning (tipover).

Holtec International 6-24 Holtec Report HI-971760

6.8.2

~ ~

Pedestal number 1 for each rack is located in the +X, -Y corner of each rack. Numbering increases counterclockwise around the periphery of the rack. The following bounding vertical pedestal forces (in kips) are obtained for each run:

Pool Event Run COF Maximum Rack Ped.

Pedestal Load (kips)

Pool C Campaign I SSE 0.8 122 SSE 2 0.2 115 SSE 3 Random Pool C Campaign II SSE 4 0.8 153 SSE 5 0.2 121 SSE 6 Randottl 134 Pool C Campaign III SSE 7 0.8 113 SSE 8 0.2 110 SSE 9 Random 122 26 Pool D Campaign I SSE 10 0.8 118 SSE 11 0.2 112 SSE 12 Random 114 Pool D Campaign II SSE 13 0.8 135 SSE 14 0.2 116 SSE 15 Random 130 As may be seen, the highest pedestal load is 153,000 lbs and occurs in run 4 for pedestal 2 of rack 5. Figure 6.8,1 provides a plot of the vertical force of this pedestal transmitted to the bearing pad over the entire duration of the SFP, 0,8 COF, SSE, campaign II simulation.

Holtec International 6-25 Holtec Report HI-971760

6.8.3 The maximum (x or y direction) shear load (in kips) bounding all pedestals for each simulation are reported below and are obtained by inspection of the complete tabular data.

Pool Event Run COF Maximum Rack Friction Load (kips)

Pool C Campaign I SSE 0.8 46 SSE 2 0.2 22.3 SSE 3 Random 41.7 13 Pool C Campaign II SSE 4 0.8 44.2 13 SSE 5 0.2 22.2 SSE 6 Random 40.9 Pool C Campaign III ~

SSE 7 0.8 43.4 SSE 8 0.2 19.7 SSE 9 Random 45.8 26 Pool D Campaign I ~ SSE 10 0.8 45.6 SSE 11 0.2 19.7 SSE 12 Random 34.4 Pool D Campaign II SSE 13 0.8 42.3 SSE 14 22.3 SSE 15 Random 42.4 Holtec International 6-26 Holtec Report HI-971760

6.8.4

~ ~

A freestanding rack, by definition, is a structure subject to potential impacts during a seismic event. Impacts arise from rattling of the fuel assemblies in the storage rack locations and, in some instances, from localized impacts between the racks, or between a peripheral rack and the pool wall. The following sections discuss the bounding values of these impact loads.

6.8.4.1 As is often the case with close rack spacing, some rack to rack impacts occur. The following instantaneous maximum impact forces and locations are identified for each of the simulations performed. Listings are only given for those simulations within which an impact occurred.

The element numbering is identified in Figures 6.5.6 through 6.5.13.

Run Impact Load Element Location Run Impact Load Element Location (kips) (kips) 3.0 494 Top 11.3 814 Bottom 8.1 503 Top 8.1 817 Top 8.1 Top 8.1 '18 Top 8.1 583 Top 4.9 831 Bottom 8.1 Top 8.4 937 Bottom 8.1 493 Top 5.6 945 Bottom 8.1 494 Top 6.4 991 Bottom 6.7 493 Top 6.5 Bottom 8.1 494 Top 8.1 736 Top 8.1 503 Top 1.9 746 Top 8.1 504 Top 8.1 759 Top 3.0 539 Top 7.9 Top 2.1 540 Top 8.1 781 Top 8.1 583 Top 8.1 782 Top 8.1 Top 8.1 789 Top 8.1 599 To 8.1 To Holtec International 6-27 Holtec Report HI-971760

Run Impact Load Element Location Run Impact Load Element Location gdps) (keeps) 8.1 Top 4.9 799 Top 5.3 736 Top 8.1 817 Top 8.1 759 Top 8.1 818 Top 8.1 Top 1.2 Top 8.1 781 Top 8.1 828 Top 8.1 782 Top 8.1 835 Top 8.1 799 Top 8.1 836 Top 8.1 Top 1.9 914 Top 8.1 817 Top 1.8 Bottom 8.1 818 Top 8.1 949 Top 8.1 827 Top 8.1 950 Top 8.1 Top 8.1 979 Top 8.1 835 Top 8.1 980 Top 8.1 836 Top 2.6 982 Bottom 8.1 Top 8.1 986 Top 8.1 Top 12.9 Bottom 8.1 913 Top 8.1 913 Top 8.1 914 Top 8.1 914 Top 8.1 Top 5.3 949 Top 8.1 980 Top 0.8 950 Top 6.7 736 Top 8.0 991 Bottom 16.7 743 Bottom 11.3 Bottom 7.7 Bottom 8.1 913 Top 10.7 756 Bottom 8.1 914 Top 4.5 777 Bottom 8.1 949 Top 22.1 778 Bottom 7.8 950 Top 16.1 813 Bottom Holtec International 6-28 Holtec Report Hl-971760

6.8.4.2

~ ~ ~

Storage racks do not impact the pool walls under any simulation, 6.8.4.3 A review of the results from each simulation allows determination of the maximum instantaneous impact load between fuel assembly and fuel cell wall at any modeled impact site.

The maximum values obtained are reported in the following table.

Pool Event Run COF Maximum Fuel Rack Impact Load ebs)

Pool C Campaign I SSE 0.8 532 SSE 0.2 562 SSE Random Pool C Campaign II SSE 0.8 531 25 SSE 0.2 548 SSE Random 535 22 Pool C Campaign III SSE 0.8 17 SSE 0.2 527 17 SSE Random 515 Pool D Campaign I SSE 10 0.8 473 17'2 SSE 0.2 591 SSE 12 Random 473 Pool D Campaign II SSE 13 0.8 472 SSE 14 0.2 462 12 SSE 15 Random 472 12 The maximum fuel to cell wall impact load is 605 pounds. Based on fuel manufacturer's data, loads of this magnitude will not damage the fuel assembly.

Holtec International 6-29 Holtec Report HI-971760

69 6.9.1 The vertical and shear forces at the bottom casting-pedestal interface are available as a function of time. The maximum values for the stress factors defined in Section 6.6.3 can be determined for every pedestal in the array of racks by scanning this data to select the limiting loads and performing calculations to determine member stresses. These two tasks are performed by a post-processor. With this information available, the structural integrity of the pedestal can be assessed and reported. The net section maximum (in time) bending moments and shear forces can also be determined at the bottom casting-rack cellular structure interface for each spent fuel rack in the pool. This allows the evaluation of the maximum stress in the limiting rack cell (box).

The tables presented in this section provide limiting stress factor results for male and female pedestals, and for the entire spent fuel rack cellular cross section just above the bottom casting.

These locations are the most heavily loaded net sections in the structure so that satisfaction of the stress factor criteria at these locations ensures that the overall structural criteria set forth in Section 6.6.1 are met.

The tables below develop stress factors for all of the SSE (Level D) simulations based on the associated SSE allowables, However, as stated above the intent is to evaluate the stresses developed from the SSE loadings with the allowables associated with OBE (Level B). Since the OBE allowables are tA of the SSE allowables, this comparison may be conservatively performed by reducing the acceptable stress ratio to 0.5. This is very conservative, since the actual OBE loads which should be compared against the OBE allowable would be much lower than the SSE loads herein.

Holtec International 6-30 Holtec Report HI-971760

6.9.1.1

~ ~ ~

The rack cell dimensionless stress factors for each of the simulations are as follows:

Pool Event Run COF Maximum R6 Rack Stress Factor Pool C Campaign I SSE 0.8 0.494 w

SSE 2 0.2 0.289 SSE Random 0.384 Pool C Campaign II SSE 4 0.8 0.454 SSE 5 0.2 0.221 13 SSE 6 Random 0.452 Pool C Campaign III SSE 7 0.8 0.409 SSE 8 0.2 0.266 SSE 9 Random 0.432 24 Pool D Campaign I SSE 10 0.8 0.230 SSE 11 0.2 0.224 SSE 12 Random 0.230 Pool D Campaign II SSE 13 0.8 0.224 SSE 14 0.2 0.227 SSE 15 RandoIn 0.232 The values for all other defined stress factors are also archived. As may be seen, all of the stress factors are well below 1.0. Therefore, the stresses developed during SSE conditions remain below the allowable SSE range and the rack modules are satisfactory to withstand the loadings. Note that stress factors for these SSE simulations are calculated based on SSE allowable strengths. However, since none of the stress factors exceed 0.5, the rack structures also adequately withstand the OBE conditions.

Holtec International 6-31 Holtec Report Hl-971760

6.9.2

~ ~

The average shear stress in the thread engagement region is given below for the limiting pedestal in each simulation.

Pool Event Run COF Maximum Thread Rack Shear Stress (psi)

Pool C Campaign I SSE 0.8 4,682 SSE 0.2 4,382 SSE 3 Random 4,607 Pool C Campaign II SSE 4 0.8 5,731 5 SSE 5 0.2 4,532 SSE 6 Random 5,019 Pool C Campaign III SSE 7 0.8 4,232 SSE 8 0.2 4,120 SSE 9 Random 4,570 26 Pool D Campaign I SSE 10 0.8 3,003 SSE 11 0.2 2,850 SSE 12 Random 2,901 Pool D Campaign II SSE 13 0.8 3,435 SSE 14 0.2 2,952 SSE 15 Random 3,307 The ultimate strength of the female part of the pedestal is 66,200 psi. The yield stress for the female pedestal material is 21,300 psi, as shown in Table 6.3.1. The male pedestal material has much greater strength and is therefore not a controlling factor in the design. The allowable shear stress for Level B conditions is 0.4 times the yield stress which gives 8,520 psi. The allowable shear stress for Level D conditions is the lesser of: 0.72 5, = 15,336 psi or 0.42 S

= 27,804 psi . Therefore, the former criteria controls.

Holtec International Holtec Report HI-971760

The largest thread shear stress computed by the post-processor is 5,731 psi. Since this value is below the allowable stresses for both OBE and DBE conditions, the thread shear stresses are within the acceptable range.

6.9.3 D Impact loads at the pedestal base (discussed in subsection 6.8.2) produce stresses in the pedestal for which explicit stress limits are prescribed in the Code. The post-processor reports the stress factors in the pedestals which are developed, in part, from these impact stresses.

The reported pedestal stress factors are included in the discussion above in Section 6.9.1.1 along with the rack cell stress factors. However, the post-processor does not develop stress factors for the localized areas of the cellular and baseplate regions of the racks which experience fuel to cell wall, rack to rack, and rack to wall impact loads. These impact loads produce stresses which attenuate rapidly away from the loaded region. This behavior is characteristic of secondary stresses.

Even though limits on secondary stresses are not prescribed in the Code for Class 3 NF structures, evaluations were made to ensure that the localized impacts do not lead to plastic deformations in the storage cells which affect the subcriticality of the stored fuel array.

Local cell wall integrity is conservatively estimated from peak impact loads. Plastic analysis is used to obtain the limiting impact load which would lead to gross permanent deformation. Table 6.9.1 indicates that the limiting impact load (of 3,238 lbf, including a safety factor of 2.0) is much greater than the highest calculated impact load value (of 605 lbf, see subsection 6.8.4.3) obtained from any of the rack analyses. Therefore, fuel impacts do not represent a significant concern with respect to fuel rack cell deformation.

Holtec International Holtec Report HI-971760

As may be seen from subsection 6.8.4.1, the bottom (baseplate) of the storage racks will impact each other at a few locations during seismic events. Since the loading is presented edge-on to the 3/4" baseplate membrane, the distributed stresses after local deformation will be negligible. The impact loading will be distributed over a large area (a significant portion of the entire baseplate length of about 50.4 (minimum) inches by its 3/4 inch thickness). The resulting compressive stress from the highest impact load of 26,200 lbs distributed over 37 sq. inches is only 708 psi, which is negligible.

Therefore, any deformation will not effect the configuration of the stored fuel.

Additional impacts will be experienced at the tops of some storage racks. These impacts will result in local yielding of the rack cell walls whenever the load exceeds 8,100 lbs. However, localized damage from all of these impacts occurs above the fuel active region. The fuel configuration and poison areas remain unaffected. Therefore, these impacts are acceptable.

6.9.4 Deeply submerged high density spent fuel storage racks arrayed in close proximity to each other in a free-standing configuration behave primarily as a nonlinear cantilevered structure when subjected to 3-D seismic excitations. In addition to the pulsations in the vertical load at each pedestal, lateral friction forces at the pedestal/bearing pad-liner interface, which help prevent or mitigate lateral sliding of the rack, also exert a time-varying moment in the baseplate region of the rack, The friction-induced lateral forces act simultaneously in x and y directions with the requirement that their vectorial sum does not exceed pV, where p, is the limiting interface coefficient of friction and V is the concomitant vertical thrust on the liner (at the given time instant). As the vertical thrust at a pedestal location changes, so does the Holtec International 6-34 Holtec Report HI-971760

~ ~

maximum friction force, F, that the interface can exert. In other words, the lateral force at the

~ ~

pedestal/liner interface, F, is given by F s p N (w) where N (vertical thrust) is the time-varying function of ~. F does not always equal pN; rather, pN is the maximum value it can attain at any time; the actual value, of course, is determined by the dynamic equilibrium of the rack structure.

In summary, the horizontal friction force at the pedestal/liner interface is a function of time; its magnitude and direction of action varies during the earthquake event.

The time-varying lateral (horizontal) and vertical forces on the extremities of the support pedestals produce stresses at the root of the pedestals in the manner of an end-loaded cantilever. The stress field in the cellular region of the rack is quite complex, with its maximum values located in the region closest to the pedestal. The maximum magnitude of the stresses depends on the severity of the pedestal end loads and on the geometry of the pedestal/rack baseplate region.

Alternating stresses in metals produce metal fatigue ifthe amplitude of the stress cycles is sufficiently large. In high density racks designed for sites with moderate to high postulated seismic action, the stress intensity amplitudes frequently reach values above the material endurance limit, leading to expenditure of the fatigue "usage" reserve in the material.

Because the locations of maximum stress (viz., the pedestal/rack baseplate junction) and the close placement of racks, a post-earthquake inspection of the high stressed regions in the racks is not feasible. Therefore, the racks must be engineered to withstand multiple earthquakes without reliance of nondestructive inspections for post-earthquake integrity assessment. The fatigue life evaluation of racks is an integral aspect of a sound design.

Holtec International 6-35 Holtec Report HI-971760

The time-history method of analysis, deployed in this report, provides the means to obtain a complete cycle history of the stress intensities in the highly stressed regions of the rack.

Having determined the amplitude of the stress intensity cycles and their number, the cumulative damage factor, U, can be determined using the classical Miner's rule where n; is the number of stress intensity cycles of amplitude q, and N, is the permissible number of cycles corresponding to o, from the ASME fatigue curve for the material of construction. U must be less than or equal to 1.0, To evaluate the cumulative damage factor, a finite element model of a portion of the spent fuel rack in the vicinity of a support pedestal is constructed in sufficient detail to provide an accurate assessment of stress intensities. Figure 6.9.1 shows the essentials of the finite

~

element model. The finite element solutions for unit pedestal loads in three orthogonal

~

~ ~ ~

directions are combined to establish the maximum value of stress intensity as a function

~

of the three unit pedestal loads. Using the archived results of the spent fuel rack dynamic analyses (pedestal load histories versus time), enables a time-history of stress intensity to be established at the most limiting location. This permits establishing a set of alternating stress intensity ranges versus cycles for several seismic events. Following ASME Code guidelines for computing U, it is found that U =0.464 due to the combined effect of 21 SSE events. This cumulative damage factor is below the ASME Code limit of 1.0 and therefore, fatigue failure is not expected. Selection of 21 SSE events represents a conservative evaluation compared to other previous fatigue assessments which were based on the damage resulting from 10 SSE

/

events, as discussed in the Harris FSAR.

Holtec International 6-36 Holtec Report HI-971760

Weld locations subjected to significant seismic loading are at the bottom of the rack at the baseplate-to-cell connection, at the top of the pedestal support at the baseplate connection, and at cell-to-cell connections. Bounding values of resultant loads are used to qualify the connections.

a.

Reference [6.6.1] (ASME Code Section III, Subsection NF) permits, for Level A or B conditions, an allowable weld stress v = .3 S= 19860 psi. As stated in subsection 3.4.2 the allowable may be increased for Level D by the ratio (15336/8520) = 1.8, giving an allowable of 35,748 psi.

Weld dimensionless stress factors are produced through the use of a simple conversion (ratio) factor applied to the corresponding stress factor in the adjacent rack material. A 2.15 factor for PWR racks is based on the differences in material thickness and length versus weld throat dimension and length:

' 2.15165 Rat' 0.0625 + 0.7071 + 7 Similarly, a 1.49 factor for BWR racks is developed as follows:

' 1 48736 Ratio 0.0625 + 0.7071 + 7 The highest predicted weld stress for DBE is calculated from the highest R6 value (see subsection 6.9.1. 1) as follows:

Holtec International 6-37 Holtec Report HI-971760

R6 + [(0.6) F] + Ratio =

OA94 [(0.6) 21,300]+2.144 = 13,574 psi this value is less than the OBE allowable weld stress value, which is 19,860.

Therefore, all weld stresses between the baseplate and cell wall base are acceptable.

b.

The weld between the baseplate and support pedestal are evaluated by development of a finite element model of the bearing pad/base plate interface and appropriate application of the maximum pedestal loads. The maximum weld stress was determined to be 10,194 psi, which is much less than the OBE allowable weld stress value of 19,860 psi.

The results are also shown in Table 6.9.1.

C.

Cell-to-cell connections are made using a series of connecting welds along the cell height. Stresses in storage cell to cell welds develop due to fuel assembly impacts with the cell wall. These weld stresses are conservatively calculated by assuming that fuel assemblies in adjacent cells are moving out of phase with one another so that impact loads in two adjacent cells are in opposite directions; this tends to separate the two cells from each other at the weld.

Table 6.9.1 gives results for the maximum allowable load that can be transferred by these welds based on the available weld area. An upper bound on the load required to be transferred is also given in Table 6.9.1 and is much lower than the allowable load.

This upper bound value is very conservatively obtained by applying the bounding rack-to-fuel impact load from any simulation in two orthogonal directions simultaneously, Holtec International 6-38 Holtec Report HI-971760

and multiplying the result by 2 to account for the simultaneous impact of two assemblies. An equilibrium analysis at the connection then yields the upper bound load to be transferred. It is seen from the results in Table 6.9.1 that the calculated load is well below the allowable.

6.9.6 To protect the pool slab from high localized dynamic loadings, bearing pads are placed between the pedestal base and the slab. Fuel rack pedestals impact on these bearing pads during a seismic event and pedestal loading is transferred to the liner. Bearing pad dimensions are set to ensure that the average pressure on the slab surface due to a static load plus a dynamic impact load does not exceed the American Concrete Institute, ACI-349 [6.9.1] limit on bearing pressures. Section 10.17 of [6.9.2] gives the design bearing strength as fb = 4 (.85 f,')

where $ = .7 and g's the specified concrete strength for the spent fuel pool. e = 1, except when the supporting surface is wider on all sides than the loaded area. In that case, e =

(A,/At)', but not more than 2. A, is the actual loaded area, and At is an area greater than A, and is defined in [6.9.2]. Using a value of e ) 1 includes credit for the confining effect of the surrounding concrete. It is noted that this criteria is in conformance with the ultimate strength primary design methodology of the American Concrete Institute in use since 1971. For Harris, the concrete compressive strength is f,' 4,000 psi. The allowable bearing pressure is conservatively computed by taking @ =1 to account for lack of total concrete confinement in the leak chase region and a stress reduction factor of /=0.7. Thus, the maximum allowable concrete bearing pressure is 2,380 psi.

The maximum vertical pedestal load is 153,000 lbs (SSE event). The bearing pad selected is 1.5" thick, austenitic stainless steel plate stock. The average pressure at the pad to liner Holtec International 6-39 Holtec Report HI-971760

interface is computed and compared against the above-mentioned limit. Calculations show that the average pressure at the slab/liner interface is 2,168 psi which is below the allowable value of 2,380 psi, providing a factor of safety of 1.1.

Therefore, the bearing pad design devised for the Harris pools C and D is deemed appropriate for the prescribed loadings.

6.9.7 The Level A condition is not a governing condition for spent fuel racks since the general level of loading is far less than Level B loading. To illustrate this, the heaviest (fully loaded) spent fuel rack (which is an 11X9 PWR rack) is considered under the dead weight load. It is shown below that the maximum pedestal load is low and that further stress evaluations are unnecessary.

LEVEL A PEDESTAL LOAD Dry Weight of Largest PWR Holtec Rack 15,700 lbf t Dry Weight of 99 PWR Fuel Assemblies 158,400 lbf Total Dry Weight 174,100 lbf tt Total Buoyant Weight (0.87 x Total Dry Weight) = 151,467 lbf Load per Pedestal 37,867 lbf The stress allowables for the normal condition is the same as for the upset condition. An upset condition pedestal load may be conservatively (bounded on the low side) determined for the Conservative weight corresponding to the heaviest rack, which is a BWR storage rack. The heaviest PWR rack nominal weight is 15,620 lb.

This weight exceeds the weight of the heaviest fully loaded BWR rack, which is

[15,700 lb + (13x13) x 680 lb] = 130,620 lb.

Holtec International 6-40 Holtec Report HI-971760

~

purpose of comparing with the load above by dividing the DBE pedestal load by a factor of 2.0. This would result in an OBE pedestal load of 153,000 :2=76,500, which is still much

~

greater than the calculated Level A load. Since this load (and the corresponding stress throughout the rack members) is much greater than the 37,867 lb load calculated above, the Upset (OBE) condition controls over normal (Gravity) condition. Therefore, no further evaluation is necessary for Level A.

6.10 The maximum hydrodynamic pressures (in psi) that develop between the fuel racks and the spent fuel pool walls will occur at those conditions and locations of greatest relative displacements. The greatest displacement was shown in Section 6.8.1 to be 1.494 inches, which occurs in rack 28 under simulation number 4. The maximum hydrodynamic pressure during this simulation was determined to be 19 psi. This hydrodynamic pressure was

~ ~

considered in the evaluation of the Fuel Handling Building and Pool structure.

Holtec International 6-41 Holtec Report HI-971760

Time history simulations, including all non-linear'impact and interface fr'iction effects, have been applied to evaluate the structural margins in the Holtec spent fuel racks.

The totality of simulations provide an extensive set of results for loads, stresses, and displacements, which taken together, demonstrate that the spent fuel racks meet the input specification and the governing Code requirements.

Evaluation of structural margins have been performed for the array of racks in each pool with all racks loaded with fuel. The requirements of the specification and the governing Code documents are met for Level A, Level B, and Level D conditions.

~ Based on all results presented in tabular form above the spent fuel racks are demonstrated to be acceptable for the service intended.

Holtec International 6-42 Holtec Report HI-971760

6.12

[6.1.1] USNRC NUREG-0800, Standard Review Plan, June 1987.

[6.1.2] (USNRC Office of Technology) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, and January 18, 1979 amendment thereto.

[6.2.1] Soler, A.I. and Singh, K.P., "Seismic Responses of Free Standing Fuel Rack Constructions to 3-D Motions", Nuclear Engineering and Design, Vol. 80, pp. 315-329 (1984).

[6.2.2] Soler, A.I. and Singh, K.P., "Some Results from Simultaneous Seismic Simulations of All Racks in a Fuel Pool", INNM Spent Fuel Management Seminar X, January, 1993.

[6.2.3] Singh, K.P. and Soler, A.I., "Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the Chin Shan Experience, Nuclear Engineering International, UK (March 1991).

[6.2.4] Holtec Proprietary Report HI-961465 - WPMR Analysis User Manual for Pre&Post Processors & Solver, August, 1997.

USNRC Standard Review Plan, NUREG-0800 (Section 3,7.1, Rev. 2, 1989).

[6.4.2] Holtec Proprietary Report HI-89364 - Verification and User's Manual for Computer Code GENEQ, January, 1990.

Rabinowicz, E., "Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company, 1976.

[6.5.2] Singh, K.P. and Soler, A.I., "Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium: The Case of Fuel Racks," 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982.

[6.5.3] Fritz, R.J., "The Effects of Liquids on the Dynamic Motions of Immersed Solids," Journal of Engineering for Industry, Trans. of the

~

ASME, February 1972, pp 167-172.

Holtec International 6P3 Holtec Report HI-971760

[6.5.4] Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering,"

McGraw Hill, 1976.

[6.5.5] Paul, B., "Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment", (Proprietary), NUSCO/Holtec Report HI-88243.

[6.6.1] ASME Boiler 8c Pressure Vessel Code,Section III, Subsection NF, 1995 Edition.

[6.6,2] ASME Boiler Er, Pressure Vessel Code,Section II, Part D, 1995 Edition.

[6.6.3] USNRC Standard Review Plan, NUREG-0800 (Section 3.8.4, Rev. 2, 1989).

[6.9.1] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan, 1985.

[6.9.2] ACI 318-95, Building Code requirements for Structural Concrete,"

American Concrete Institute, Detroit, Michigan, 1995.

Holtec International Holtec Report HI-971760

PLANT DOCKET NUNIER(s)

Enrico Fermi Unit 2 USNRC 50-341 1980 Quad Cities 1 Ec 2 USNRC 50-254, 50-265 1981 Rancho Seco USNRC 50-312 1982 Grand Gulf Unit 1 USNRC 50-416 1984 0 ster Creek USNRC 50-219 1984 Pil rim USNRC 50-293 1985 V.C. Summer USNRC 50-395 1984 Diablo Can on Units 1 Et; 2 USNRC 50-275, 50-323 1986 B ron Units 1 k2 USNRC 50-454, 50-455 1987 Braidwood Units 1 &2 USNRC 50456, 50-457 1987 Vo tie Unit2 USNRC 50-425 1988 St. Lucie Unit 1 USNRC 50-335 1987 Millstone Point Unit 1 USNRC 50-245 1989 Chinshan Taiwan Power 1988 D.C. Cook Units 1 Er,2 USNRC 50-315, 50-316 1992 Indian Point Unit 2 USNRC 50-247 1990 Three Mile Island Unit 1 USNRC 50-289 1991 James A. FitzPatrick USNRC 50-333 1990 Shearon Harris Unit 2 USNRC 50-401 1991 Ho e Creek USNRC 50-354 1990 Holtec International 645 Holtec Report HI-971760

PLANT DOCKET NUMI)ER(s)

Kuoshen Units 1 &2 Taiwan Power Com an 1990 Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units 1 &2 Comision Federal de 1991 Electricidad Zion Station Units 1 &2 USNRC 50-295, 50-304 1992 Se uo ah USNRC 50-327, 50-328 1992 LaSalle Unit 1 USNRC 50-373 1992 Duane Arnold Ener Center USNRC 50-331 1992 Fort Calhoun USNRC 50-285 1992 Nine Mile Point Unit 1 USNRC 50-220 1993 Beaver Valle Unit 1 USNRC 50-334 1992 Salem Units 1 &2 USNRC 50-272, 50-311 1993 Limerick USNRC 50-352, 50-353 1994 Ulchin Unit 1 KINS 1995 Yon wan Units 1 &2 KINS 1996 Kori-4 KINS 1996 Connecticut Yankee USNRC 50-213 1996 An ra Unit1 Brazil 1996 Sizewell B United Kin dom 1996 Holtec International 6-46 Holtec Report HI-971760

Young's Modulus Strength Ultimate Strength Material E (psi) 10'ield S (psi)

Su (psi)

SA240; 304L S.S. 27.6 x 21,300 66,200 SA240, Type 304L (upper 27.6 x 10~ 66,200 part of support feet) 10'1,300 SA-564-630 (lower part of 28.5 x 106,300 140,000 support feet; age hardened at 1100 F)

'oltec International 6-47 Holtec Report HI-971760

OBE Datal to Data2 0.0295 Datal to Data3 0.0392 Data2 to Data3 0.0169 DBE Datal to Data2 0.0183 Datal to Data3 0.0588 Data2 to Data3 0.0299 Holtec International Holtec Report Hl-971760

Holtec International 649 Holtec Report HI-971760 Holtec International 6-50 Holtec Report HI-971760 Holtec International 6-51 Holtec Report Hl-971760

Item/Location DBE Calculated OBE Allowable Fuel assembly/cell wall impact, lbf. 605

3,238 **

Rack/baseplate weld, psi 13,574 19,860 Female pedestal/baseplate weld, psi 10,194 19,860 Cell/cell welds, lbf. ] 71] 3,195 See Section 6.8.4.3.

Based on the limit load for a cell wall. The allowable load on the fuel assembly itself may be less than this value but is greater than 605 lbs.

- - Based on the fuel assembly to cell wall impact load simultaneously applied in two orthogonal directions.

Holtec International 6-52 Holtec Rcport HI-971760

NORTH 597.88 HIN.

1.43" 2.28" PHASE III PHASE III PHASE III PWR PWR PWR 9 X 9 9 X 9 9 X 9 PHASE III PHASE III BWR BWR 13 X 13 13 X 13 320.60 MIN.

PHASE III BWR 13 X 13 44l ~

.625" KIN.( TYP ) 1.44" 1.43" PHASE I CELL COUNT : PHASE II CELL COUNT : PHASE III CELL COUNT :

PWR TOTAL CEI,L COUNT :

927 CELLS PWR 360 CELLS PWR 324 CELLS - PWR 243 CELLS 2783 CELLS BWR 507 CELLS BWR BVR 1320 CELLS BWR 936 CELLS FIGURE 6.3.1; PHASED STORAGE CONFIGURATION FOR POOL C HOLTEC INTERNATIONAL REPORT HI 971780 6-53

NORTH 5.865" 5.097" 237.79 HIN.

.625" ( TYP )

4.125" 5.1875" 625 "( TYP )

383.36 llIN.

PHASE I CELL COUNT:

PHASE II CELL COUNT: TOTAL CELL COUNT:

525 CELLS PNR 1025 CELLS PWR 500 CELLS PtIR FIGURE 6.3.2; PHASED STORAGE CONFIGURATION FOR POOL D HOLTEC INTERNATIONAL REPORT HI 971760 6-54

Hnrr t s Pl nn t Spen t Fue I Poo I T t.me H t.s tor y Acce er ogr am I

X d t.r ec t l.on Boun d t.n g Spec tr a ( 2% Damp t n g)

8. 28 N

8. 18 0

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o 8.18

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8. 88 588. 88 . 1888. 88 1588. 88 Time ( sec. X 'l88)

Figure 6.4.1 HI-971760

~ ~

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8. 28 0

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8. 88 688. 88 1888. 88 1588. 88 2888. 88 T irne ( sec . X 'I 88)

Figure 6.4.3 HI-971760 6-<7

Har r t.s PI an t Spen t Fue I Poo I 7 t.me H t.s tor y Acce er ogr am I

X d tr ect ton Bound l.ng Spec tr a ( AR Damp l.ng)

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8. 28 S

0 o 8. 28 588. 88 1 888. 88 1 588. 88 2888. 88 T mme ( sec . X 188)

Figure 6.4.4 HI-971760

Har r t.s Pl an Spen t Fue I Poo I T t me H t s tor y Acce er ogr am I

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Figure 6.4.5 HI-971760

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FIGURE 8.5.1; SCHEMATIC OF THE DYNAMIC eE~oer u ~r-g7imo MODEL FOR DYNARACK Holtec Proprietary 6%1

CD C)

Holtec Proprietary

FIGURE 6 5 3; TWO DIMENSIONAL VIEW OF THE SPRING-MASS SIMULATION REPORT 0 HI-97l760 r

Holtec Proprietary 6%3

FIGURE 6 5 4- RACK DEGREES-OF-FREEDOM FOR X-Z PLANE BENDING WITH SHEAR AND BENDING SPRING REPORT II HI-971760 Hoitec Proprietary

FIGURE 6 5 5; RACK-TO-RACK IMPACT SPRINGS PORT 8 HI-971760 Holtec Proprietary 6-6S

IIARRIS SPENT FUEL POOL C FIGURE 6.5.6; RACK INPACT SPRING NUSERING SCBENE ( BOTTON)

CAHPAIGN I I -9717$

Holtec Proprietary

HARRIS SPENT FUEL POOL C FIGURE 6.5.7; RACK IMPACT SPRING NUHBERING SCIIENE (TOP)

CARPAIGN I Holtec Proprietary 6%7

HARRIS SPENT FUEL POOL C FIGURE 6.5.8; RACK IMPACT SPRING NUMBERING SCHEME (BOTTOM )

CAMPAIGNS II ANB III I .-II717%

Holtec Proprietary

HARRIS SPENT FUEL POOL C FIGURE 6.5.9; RACK IMPACT SPRING NUMBERING SCHEME ( TOP )

CAMPAIGNS II AND III I -971750 Holtec Proprietary 6<9

HARRIS SPENT FUEL POOL D FIGURE 6.5.10; RACK IMPACT SPRING NUMBERING SCHEME ( BOTTOM )

CAMPAIGN I Holtec Proprietary HI 971760 6-70

HARRIS SPENT FUEL POOL 9 FIGURE 6.5.11; RACK IHPACT SPRING NUHBERING SCHEHE ( TOP )

CAbtPAIGN I Holtec Proprietary Hl 971760

HARRIS SPENT FUEL POOL D FIGURE 6.5.12; RACK IhIPACT SPRING NUIERING SCHEIIE ( BUTTON )

CALIPAIGN II Holtec Proprietary H1 971760 6-7'7

HARRIS SPENT FURL POOL D FIGURE 6.5.13; RACK IMPACT SPRING NUMBERING SCHEME ( TOP )

CAMPAIGN II Holtec Proprietary H1 971760

Har r Ls Poo I C Run Ver t (ca I Pedes ta S I T mme His tory Load (n g Rack 5, Pedes ta 2 I F

Figure 6.8.1 HI-971760 Holtec Proprietary

Figure 6.9.1; Rack Fatigue Analysis Model Holtec Proprietary HI-971760 6-75

7.0 FUEL HANDLINGAND CONSTRUCTION ACCIDENTS 7.1 Introduction The USNRC OT position paper {7.1] specifies that the design of the rack must ensure the functional integrity of the spent fuel racks under all credible drop events in the spent fuel pool.

This section contains synopses of the analyses carried out to demonstrate the regulatory compliance of the proposed racks under postulated fuel assembly drop scenarios germane to HNP pools C and D.

In addition to the postulated fuel assembly free-fall scenarios, a gate drop accident event was also considered. In this case, the ability of the pool structure to avert primary structural damage (leading to rapid loss of water) needs to be demonstrated.

7.2 Descri tion of Fuel Handlin Accidents Two categories of fuel assembly accidental drop events are considered. In the so-called "shallow drop" event, a fuel assembly, along with the portion of handling tool which is severable in the case of a single element failure, is assumed to drop vertically and hit the top of the rack. The "depth" of damage to the affected cell walls must be demonstrated to remain limited to the portion of the cell above the top of the "active fuel region", which is essentially the elevation of the top of the Boral neutron absorber. To meet this criterion, the plastic deformation of the rack cell wall should not extend more than 21.3 inches (downwards) from the top of a PWR rack. The distance separating the top of the rack from the Boral in the BWR racks is 13.75 inches.

Therefore, to be conservative the smaller BWR dimension of 13.75 inches is selected as the maximum depth of damage of an object falling onto the tops of storage racks.

Holtec International 7-1 Holtee Report HI-971760

By observation, the drop of a PWR assembly onto a PWR rack is more limiting than any other combination of the two fuel types (PWR vs. BWR) with the two rack (PWR vs.

BWR) types. This isobviousbecauseof tworeasons. The PWRassembly drop is amore severe case than the BWR assembly case, since the effect of the weight differences (approximately 1600 vs. 680 lbs, respectively) far exceeds the effect of the differences in the impact cross-section zone (about 8.4 vs. 5.5 inches, respectively). The PWR storage rack cell controls as an impact zone over the BWR cell because it is larger (8.4 vs. 6.06 inches, respectively) resulting in less capacity to withstand top of cell or baseplate impacts.

(The nominal cell wall thicknesses of the two rack types is identical).

In order to utilize an upper bound of kinetic energy at impact, the impactor is assumed to weigh 2,100 lbs and the free-fall height is assumed to be 36 inches. The impactor weight corresponds to the heaviest fuel (plus handling tool) which will be handled in pools C and It is readily apparent from the design of the rack modules described in Section 3, that the impact resistance of a rack at its periphery is less than its interior. Accordingly, the potential shallow drop scenario is postulated to occur at the periphery in the manner shown in Figure 7.2.1.

Finally, the fuel assembly assemblage is assumed to hit the rack in a manner to inflict maximum damage. The impact zone is chosen to minimize the cross sectional area which experiences the deformation. Placement of the impact at the corner would reduce the impact zone area, but actually increases the cross-sectional area experiencing deformation.

Impact at the corner would involve the crushing of two cell walls under the dynamic impact. Therefore, impact on only one cell wall is chosen to simulate the worst case accident. Figure 7.2.2 depicts the impacted rack in plan view.

Holtec International 7-2 Holtec Report HI-971760

The second class of "fuel drop event" postulates that the impactor falls through an empty storage cell impacting the rack baseplate. This so-called "deep drop" scenario threatens the structural integrity of the "baseplate". Ifthe baseplate is pierced, then the fuel assembly might damage the pool liner (which at 3/16" is rather thin) and create an abnormal condition of the enriched zone of fuel assembly outside the "poisoned" space of the fuel rack. To preclude damage to the pool liner, and to avoid the potential of an abnormal fuel storage configuration in the aftermath of a deep drop event, it is required that the baseplate remain unpierced and that the maximum lowering of the fuel assembly support surface is less than the distance from the bottom of the baseplate to the liner.

The deep drop event can be classified into two scenarios, namely, drop through cell located above a support leg (Figure 7.2.3), and drop in an interior cell away from the support pedestal (Figure 7.2.4).

In the former deep drop scenario (Figure 7,2.3), the baseplate is buttressed by the support pedestal and presents a hardened impact surface, resulting in a high impact load. The principal design objective is to ensure that the support pedestal does not pierce the lined, reinforced concrete pool slab.

The baseplate is not quite as stiff at cell locations away from the support pedestal (Figure 7.2.4) Baseplate severing and large deflection of the baseplate (such that the liner would

~

be impacted) would constitute an unacceptable result.

7.3 The drop of a rack above spent fuel stored within in-place rack modules is precluded, since racks will not be lifted above spent fuel. The drop of a rack module during installation is also extremely remote, due to the defense-in-depth approach discussed in Sections 3.5 and Holtec International 7-3 Holtec Report HI-971760

~ ~

of this possibility, a rack dropping to

~ ~

11.1.

~ ~ Despite the unlikelihood the pool floor has

~

been considered. ~ To evaluate the consequences of an accidental, uncontrolled lowering of the heaviest rack module, a 13x13 BWR module conservatively considered with a submerged weight of 16140 lb (actual maximum nominal dry weight is only 15700 lb),

from a height of 480 inches above the pool liner is considered (Figure 7.3.1). The objective of the analysis is to ensure that a rapid loss of pool water will not occur, leading to loss of shielding to the stored nuclear fuel.

7.4 In the first step of the solution process, the velocity of the dropped object (impactor) is computed for the condition of underwater free fall. Table 7.1 contains the results for the three drop events.

In the second step of the solution, an elasto-plastic finite element model of the impacted region on Holtec's computer Code PLASTIPACT (Los Alamos Laboratory's DYNA3D implemented on Holtec's QA system) is prepared. PLASTIPACT simulates the transient collision event with full consideration of plastic, large deformation, wave propagation, and elastic/plastic buckling modes. For conservatism, the impactor in all cases is assumed to be rigid. The physical properties of material types undergoing deformation in the postulated impact events are summarized in Table 7.2.

Holtec International Holtec Report HI-971760

I 1.1 Figure 7.5.1 shows the finite element model utilized in the shallow drop impact analysis.

Dynamic analyses show that the top of the impacted region undergoes severe localized deformation. Figure 7.5.2 shows an isometric view of the post-impact geometry of the rack for the shallow drop scenario. The maximum depth of plastic deformation is limited to 11 inches, which is below the design limit of 13,75 inches. Figure 7.5.3 shows the plan view of the post-collision geometry. Approximately 10% of the cell opening in the impacted cell is blocked.

7.5.2 The deep drop scenario depicted in Figure 7.5.4(b), wherein the impact region is located above the support pedestal, is found to produce a negligible deformation on the baseplate.

The vertical force in the support pedestal remains below the loads generated during seismic events (see Section 6). Therefore, it is concluded that the pool liner will not be damaged.

The deep drop condition through an interior cell depicted in Figure 7.5.4(a) does produce some deformation of the baseplate and localized severing of the baseplate/cell wall welds (Figure 7.5.5). However, the fuel assembly support surface is lowered by a maximum of 2.89 inches, which is less than the minimum distance of 6 inches from the bottom of the baseplate to the liner. Therefore, the deformed baseplate will not strike the liner during this drop event and the pool liner will not be damaged. As stated in Subsection 4.7.2, criticality evaluations performed for this baseplate deformation have shown that the storage configuration remains acceptable.

Holtec International 7-5 Holtec Report HI-971760

e 0

7.6 Since the primary structural integrity of the slab is unimpaired subsequent to a rack drop to the pool floor liner, catastophic loss of pool water would not occur. Therefore, catastrophic failure of the pool structure or rapid loss of pool water will not occur.

No other credible in-pool drops have been identified. An object potentially carried over the pools is one of the 4,000 pound gates which isolate the pools. These gates are long rectangular metallic structures with a base area of 8 inches by 41 inches. During handling the gate is lifted using a single failure proof crane and double rigging. The rigging complies with the safety margin requirements of NUREG-0612. An accidental drop of the gate is not a credible event, because of the above mentioned defense-in-depth approach to the lifting of this heavy load. Additionally the gates are not located within the pools, but are installed inside of slots within adjacent transfer canals. Nevertheless, analyses were carried out for this accident scenario. A gate drop during handling from 40 feet above the pool liner was evaluated and it has been determined that a primary failure of the water retaining concrete structure will not occur. A gate drop during handling from 15 inches II above the top of a PWR rack loaded with fuel was also evaluated. A schematic of the 3D finite element model is depicted in Figures 7.6.1 and 7.6.2. The gate is conservatively considered to strike only three rack storage cell walls, as shown in Figure 7.6.3. This impact zone is conservative, since the dimensions of the gate would span at least four cell walls. The gate is shown to penetrate the rack to a depth of less than 5 inches, as shown in Figures 7.6.4 and 7.6.5. Since this penetration remains above the tops of the stored fuel assemblies, no fuel damage occurs.

Holtec International 7-6 Holtec Report HI-971760

The fuel assembly and gate drop accident events evaluated for the HNP fuel pools were analyzed and found to produce localized damage well within the design limits for the racks. A construction accident event wherein the heaviest rack falls from a 40 foot height onto the pool floor was also considered. Analyses show that the pool structure will not suffer any primary structural damage. A similar conclusion is reached with regard to a gate drop event.

Holtec International 7-7 Holtec Report HI-971760

[7.1] "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978.

Holtec International 7-8 Holtec Report HI-971760

TABLE 7.1 IMPACT EVENT DATA Impactor Drop Impact Weight Imp actor Height Velocity Case (lbs) (inches) (inch/sec)

1. Shallow drop event 2,100 Fuel Assembly 36 152

2. Deep drop event 2,100 Fuel Assembly 205 353

3. Construction event 16,140 Rack Module 480 304 Holtec International Proprietary Information 7-9 Holtec Report HI-971760

TABLE 7.2 MATERIALDEFINITION Elastic Stress Strain Density Modulus (pcf) (psi) First Yield Failure Elastic Failure Material (psi) (psi)

Name Type Stainless SA240-304L steel Stainless SA240-304 steel Stainless SA564-630 steel Concrete 4000 psi Holtec International Proprietary Information 7-10 Holtec Report HI-971760

FLIEL ASSEHBLY IHPACT REGION I I I) I I

I I I I

( I I )I I I

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I )I I I

(I I 1 1 I )I )I (I Igl I (I I

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IQPACTOR

.Figure 722; Plan View of Impactor and Impact Zone (Shallow Drop Event)

HI-97l760

FL!EL ASSEHQL4 I

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RACK IMPACT REGION Figure 723; Deep Drop on a Support Leg Location 7-13

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RACK TGP OF S.F.P.

TGP GF MATERIEL I

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I I

I I IMPACT REGION BOTTOM OF S.F.P.

4 4

4 L

4

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4 LINER 4 ~

4 4

Figure 733; Heaviest Rack Drop 7-15 HI-971760

Figure 753; Shallow Drop: Finite Element Model Detail Impacted Region Holtec Proprietary 7-16 HI-971760

Figure 752; Maximum Cell Deformation for Shallow Drop on Exterior Cell Holtec Propnetary 7-17 HI-971760

Figure 753; Shallow Drop: Maximum Cell Deformation Impacted Region Plan Holtec Proprietary Hl-971760 7-18

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Figure 755; Maximum Baseplate Deformation from Deep Drop Scenario 7-20 Holtec Proprietary HI-971760

Figure 7.63; Gate Drop Finite-Element Model 7-21 HI-971760 Holtec Proprietary

Figurc 7.6.2; Gate Drop Finite-Element Model, Detail of Impacted Region 7-22 Holtec Proprietary HI-971760

Figure 7.63; Gate Drop Finite-Element Model Detail of Impacted Region (Plan)

HI-971760

Figure 7.6.4; Gate Drop Maximum Deformation HI-971760 Holtec Proprietary 7-24

Figure 7.65; Gate Drop Maximum Deformation Impacted Region Plan Holtec Proprietary HI-971760

0 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 J~tr ~cti The Harris Spent Fuel Pools (SFPs) C and D are safety related, seismic category I, reinforced concrete structures. Spent fuel is to be placed within storage racks located in both of these areas and they will be collectively referred to herein as the fuel pool structure. This section describes the analysis to demonstrate structural adequacy of the pool structure, as required by Section IV of the USNRC OT Position Paper [8.1.1].

The pool regions are analyzed using the finite element method. Results for individual load components are combined using factored load combinations mandated by SRP 3.8.4 [8.1.2]

based on the "ultimate strength" design method of the American Concrete Institute (ACI 318)

[8.1.3]. It is demonstrated that for the critical bounding factored load combinations, structural integrity is maintained when the pools are assumed to be fully loaded with spent fuel racks, as shown in Figures 1.2 and 1.3 with all storage locations occupied by fuel assemblies.

The regions examined in the SFPs are the floor slabs, and the highly loaded wall sections adjoining the slabs. Both moment and shear capabilities are checked for concrete structural integrity. Local punching and bearing integrity of the slab in the vicinity of a rack module support pedestal pad is evaluated. All structural capacity calculations are made using design formulas meeting the requirements of ACI 318.

8.2 c c e The SFPs are located inside the Fuel Handling Building and are supported by a two way, reinforced concrete slab, The minimum thickness of the slab is 12.0 feet, including grout. The SFPs are separated by reinforced concrete walls and transfer canals.

Holtec International 8-1 Report HI-971760

~

Figure 1.1 shows the layout

~ of the majority of the Fuel Handling Building. A plan of the building dimensions of the

~ ~

area of concern is shown in Figure 8.2.1, which shows the major structural

~

pools. The floor liner plate of the SFPs are located at elevation 246.0 The spent fuel area operating floor is at elevation 286.0.

8.3 Pool structural loading involves the following discrete components:

8.3.1 a ic 'a lid V

1) Dead weight of pool structure includes the weight of the Fuel Handling Building concrete upper structure.

2) Maximum dead weight of rack modules and fuel assemblies in the fully implemented storage configuration,'as shown in Figures 1.2 and 1.3.

3) Dead weight of a shipping cask including yoke of 250 kips.

4) The Cask Crane, Auxiliary Crane and Spent Fuel Handling Machine (Refueling Platform) are designed to move along the N-S direction. The dead weight and the rated liftweight of these cranes are considered as live load.

5) The hydrostatic water pressure.

8.3.2 'c

1) Vertical loads transmitted by the rack support pedestals to the slab during a SSE or OBE seismic event.

Holtec International 8-2 Report HI-971760

2) Hydrodynamic inertia loads due to the contained water mass and sloshing loads (considered in accordance with TID-7024 [8.3.1]) which arise during a seismic event.

3) Hydrodynamic pressures between racks and pool walls caused by rack motion in the pool during a seismic event.

4) Seismic inertia force of the walls and slab.

8.3.3 e Thermal loading is defined by the temperature existing at the faces of the pool concrete walls and slabs. Two thermal loading conditions are evaluated: The normal operating temperature and the accident temperature.

8.4 8.4.1 i ie e The finite element model encompasses the two SFPs, the Fuel Transfer Canal, the Cask Loading Pool, and adjacent transfer canals and building structure. The interaction with the rest of the Fuel Handling Building reinforced concrete, which is not included in the finite-element model, is simulated by imposing appropriate boundary conditions. The structural area of interest for the reracking project includes only two pools which are involved in the fuel storage capacity increase. However, by augmenting the area of interest, by considering in the constructed finite-element model and numerical investigation the additional areas described above, the perturbation induced by the boundary conditions on the stress field distribution for the area of interest is minimized. A finite element 3D view of the structural elements considered in the numerical investigation is shown in Figure 8.4.1.

Holtec International 8-3 Report HI-971760

The preprocessing capabilities of the STARDYNE computer code [8.4.1] are used to develop the 3-D finite-element model. The STARDYNE finite-element model contains 13,353 nodes, 3,564 solid type finite-elements, 7,991 plate type finite-elements and 24 hydro-dynamic masses. Figure 8.4.1 depicts an isometric view of the three-dimensional finite element model without the water and concentrated masses (racks, cask, etc.).

The dynamic behavior of the water mass contained in the SFPs and Transfer Canal during a seismic event is modeled according to the guidelines set in TID-7024.

8.4.2 i et The structural region of concern, from column lines 43 to 73 and from line L to N, is isolated from the Fuel Handling Building. This region is numerically investigated using the finite element method. The pool walls and their supporting reinforced concrete slab are represented by a 3-D finite-element model.

Holtec International 8-4 Report Hl-971760

The individual loads considered in the analysis are grouped in five categories: dead load (weight of the pool structure, dead weight of the rack modules and stored fuel, dead weight of the reinforced concrete Fuel Handling Building upper structure, the hydro-static pressure of the contained water), live loads (weights of the Cask Crane, Auxiliary Crane, and SFHM and their maximum suspended loads), thermal loads (the thermal gradient through the pool walls and slab for normal operating and accident conditions) and the seismic induced forces (structural seismic forces, interaction forces between the rack modules and the pool slab, seismic loads due to self-excitation of the pool structural elements and contained water, and seismic hydro-dynamic interaction forces between the rack modules and the pool walls for both OBE and SSE conditions) The dead and thermal loads are considered static acting loads, while the seismic

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induced loads are time-dependent.

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I I

Holtec International 8-5 Report Hl-971760

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Results for individual load cases are combined using the factored load combinations discussed

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below. The combined stress resultants are compared with the ultimate moments and shear

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capacities of all structural elements pertinent to the SFPs, which are calculated in accordance with the ACI 318 to develop the safety factors.

8.43 ~d The various individual load cases are combined in accordance with the NUREG-0800 Standard Review Plan [8.1.2] requirements with the intent to obtain the most critical stress fields for the investigated reinforced concrete structural elements.

For "Service Load Conditions" the following load combinations are:

- Load Combination No. 1 = 1.4* D + 1.7*L

- Load Combination No. 2 = 1.4* D + 1.7*L + 1.9*E

- Load Combination No. 3 = 1.4* D + 1.7*L - 1.9*E

- Load Combination No. 4 = 0.75* (1.4* D + 1.7*L + 1.9*E +1.7*To)

- Load Combination No. 5 = 0.75* (1.4* D + 1.7*L - 1.9*E +1.7*To)

- Load Combination No, 6 = 1.2*D + 1.9*E

- Load Combination No. 7 = 1.2*D - 1.9*E For "Factored Load Conditions" the following load combinations are:

Holtec International 8-6 Report HI-971760

- Load Combination No. 8 = D + L + To + E' Load Combination No, 9 = D + L + To - E' Load Combination No. 10 = D + L + Ta + 1.25*E

- Load Combination No. 11 =D + L+ Ta - 1.25*E

- Load Combination No. 12 =D + L+ Ta + E' Load Combination No. 13 = D + L + Ta -

E'here:

D= dead loads; live loads; To = thermal load during normal operation; Ta = thermal load under accident condition; OBE earthquake induced loads; E' SSE earthquake induced loads.

8.5 The STARDYNE computer code is used to obtain the stress and displacement fields for the 1 individual load cases.

The STARDYNE postprocessing capability is employed to form the appropriate load combinations and to establish the limiting bending moments and shear forces in various sections ofthepoolstructure. Atotalof13 loadcombinationsarecomputed. Sectionlimitstrength Holtec International 8-7 ReportHI-971760

formulas for bending loading are computed using appropriate concrete and reinforcement strengths. For Harris, the concrete and reinforcement allowable strengths are:

concrete 4,000 psi f,'einforcement f = 60,000 psi Table 8.5.1 shows results from potentially limiting load combinations for the bending and shear strength of the slab and walls. For each section, we define the limiting safety margins as the limited strength bending moment or shear force defined by ACI for that structural section divided by the calculated bending moment or shear force (from the finite element analyses). The major regions of the pool structure consist of the four concrete walls and floors delimiting each of the SFPs. Each area is searched independently for the maximum bending moments in different bending directions and for the maximum shear forces. Safety margins are determined &om the calculated maximum bending moments and shear forces based on the local strengths. The procedures are repeated for all the potential limiting load combinations, Therefore, limiting safety margins are determined. Table 8.5.1 demonstrates that the limiting safety margins for all sections are above 1.0, as required.

8.6 PuuLLhcz The pool liners are subject to in-plate strains due to movement of the rack support feet during the seismic event. Analyses are performed to establish that the liner will not tear or rupture under limiting loading conditions in the pool. These analyses are based on loadings imparted &om the most highly loaded pedestal in the pool assumed to be positioned in the most unfavorable position. Bearing strength requirements are shown to be satisfied by conservatively analyzing the most highly loaded pedestal located in the worst configuration with respect to underlying leak chases.

Holtec International 8-8 Report HI-971760

Regions affected by loading the fuel pool completely with high density racks are examined for structuml integrity under bending and shearing action. It is determined that adequate safety margins exist assuming that all racks are fully loaded with a bounding fuel weight and that the factored load combinations are checked against the appropriate structural design strengths. It is also shown that local loading on the liner does not compromise liner integrity under a postulated fatigue condition and that concrete bearing strength limits are not exceeded.

Holtec International 8-9 ReportHI-971760

0 0

[8.1.1] OT Position for Review and Acceptance of Spent Fuel Handling Applications, by B.K. Grimes, USNRC, Washington, D.C., April 14, 1978.

[8.1.2] NUI&G-0800, SRP-3.8.4, Rev. 1., July 1981.

[8.1.3] ACI 318-95 and ACI 318R-95, "Building Code Requirements for Structural Concrete and Commentary," American Concrete Institute, 1995.

[8.3.1] "Nuclear Reactors and Earthquakes, U.S. Department of Commerce, National Bureau of Standards, National Technical Information Service, Springfield, Virginia (TID 7024).

[8.4.1] STARDYNE User's Manual, Research Engineers, Inc., Rev. 4.4, July 1996.

[8.4.2] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit Michigan.

Holtec International 8-10 Report HI-971760

Table 8.5.1 BENDING AND SHEAR STRENGTH EVALUATION Pool Critical Flexure Critical Shear Location Limiting Load Combinations Limiting Load Combinations Safety Margin (see Section 8.4.3) Safety Margin (see Section 8.4.3)

North Wall 1.97 1.31 South Wall 3.51 2.20 East Wall 1.72 1.10 West Wall 1.05 10 1.06 Pool Floor Slab 2.1 2.71 D North Wall 2.32 3.43 South Wall 1.30 10 1.08 37 East Wall 1.48 2,6 1.07 37 West Wall 1.05 1.06 Pool Floor Slab 2.01 1.64 Holtec International S-11 Report HI-971760

~,

i

96~ l4l

~

l25 l 360

'Figure 8.2.1; Pool Structure Dimensions HI-971760

Figure 8.4.1; Fuel Handling Building Finite Element Model 5'h Holtec Proprietary Hl-971760 8-13

90 9.1 W No significant increase in the volume of solid radioactive wastes is expected from operating with the expanded storage capacity. The necessity for pool filtration resin replacement is determined primarily by the requirement for water clarity, and the resin is normally expected to be changed about once a year. During racking operations, a small amount of additional resins may be generated by the pool cleanup system on a one-time basis.

9.2 e e Gaseous releases &om the fuel storage area are combined with other plant exhausts. Normally, the contribution from the fuel storage area is negligible compared to the other releases and no significant increases are expected as a result of the expanded storage capacity.

9.3 nnel o e During normal operations, personnel working in the fuel storage area are exposed to radiation from the spent fuel pool. Operating experience has shown that area radiation dose rates originate primarily &om radionuclides in the pool water. As expected, subsequent to the removal of transhipped fuel from the shipping casks, Harris has experienced increases in the pool water radionuclide concentrations due to sloughing of crud and other contaminants associated with fuel handling. Additionally, radionuclide concentration increases are also experienced subsequent to the discharge of fuel &om the Harris Unit 1 reactor. These two conditions represent the the previously analyzed conditions for pool water radionuclide concentrations and will not be significantly changed by the capacity expansion of storing spent fuel in pools C and D.

Therefore, no additional evaluations for pool water radionuclides are required for the proposed

. change, Holtec International 9-1 ReportHI-971760

Radiation dose rates in accessible areas around the SFPs will be determined for comparison with existing zone designations. Any changes required to the zone designations will be identified and included in an update to the Harris FSAR, if necessary.

Operating experience has also shown that there have been negligible concentrations of airborne radioactivity in the Spent Fuel Pool area. No increase in airborne radioactivity is expected as a result of the expanded storage capacity.

9.4 All of the operations involved in racking will utilize detailed procedures prepared with full consideration of ALARAprinciples. Similar operations have been performed in a number of facilities in the past, and there is every reason to believe that racking can be safely and efficiently accomplished at Harris, with low radiation exposure to personnel. The Harris racking project represents lower radiological risks due to the fact that the pools currently contain no spent fuel.

The existing radiation protection program at Harris is adequate for the re-racking operations.

Where there is a potential for significant airborne activity, continuous air monitors will be in operation. Personnel will wear protective clothing as required and, if necessary, respiratory

~ protective equipment. Activities will be governed by a Radiation Work Permit, and personnel monitoring equipment will be issued to each individual. As a minimum, this will include Holtec International 9-2 Report HI-971760

~

thermoluminescent dosimeters (TLDs) and self-reading dosimeters. Additional personnel

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monitoring equipment (i.e., extremity TLDs or multiple TLDs) may be utilized

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~ ~ as required.

Work, personnel traffic, and the movement of equipment will be monitored and controlled to minunize contatnination and to assure that dose is maintained ALARA.

Holtec International 9-3 Report HI-971760

Table 9.4.1 PRELIMINARYESTIMATE OF PERSON-REM DOSE DURING RACKING Estimated Number of Person-Rem Step Personnel Hours Dose Clean and vacuum pool Remove underwater appurtenances Installation of new rack modules Total Dose, person-rem Holtec International ReportHI-971760

10.0 10.1 The construction phase of the Harris Spent fuel pool rack installation willbe executed by Carolina Power Ec Light. CP8cL willalso be responsible for specialized services, such as underwater diving and welding operations, ifrequired. All construction work at Harris will be performed in compliance with NUREG-0612 (refer to Section 3.0), and site-specific procedures.

Crane and fuel bridge operators are to be adequately trained in the operation of load handling machines per the requirements of ANSI/ASME B30,2, latest revision, and the plant's specific training program.

The lifting devices designed for handling and installation of the new racks and removal of the old racks at Harris are remotely engageable. The lifting devices comply with the provisions of ANSI

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N14.6-1978 and NUREG-0612, including compliance with the primary stress criteria, load

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testing at a multiplier of maximum working load, and nondestructive examination of critical

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welds.

An intensive surveillance and inspection program shall be maintained throughout the rack installation phase of the project. A set of inspection and QC hold points willbe implemented which have been proven to eliminate any incidence of rework or erroneous installation in numerous previous rack installation campaigns in Pools A and B.

Holtec International and CP&L have developed a complete set of operating procedures which cover the entire gamut of operations pertaining to the rack installation effort. Similar procedures have been utilized and successfully implemented by Holtec International on previous rack installation projects. These procedures assure that ALARApractices are followed and provide detailed requirements to assure equipment, personnel, and plant safety. The following is a list of Holtec International 10-1 Holtec Report HI-971760

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procedures which willbe available for use in implementing the rack installation phase

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of the

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project. ~

A. t r due:

This procedure provides direction for the handling/installation of the new high density modules.

The procedure delineates the steps necessary to receive a new high density rack on site, and the proper method for unloading and uprighting the rack, staging the rack prior to installation, and installation of the rack. The procedure also provides for the installation of new rack bearing pads, adjustment of the new rack pedestals and performance of the as-built field survey.

B. e' Pr ce e:

This procedure delineates the steps necessary to perform a thorough receipt inspection of a new

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rack module after its arrival on site. The receipt inspection includes dimensional measurements,

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cleanliness inspection, visual weld examination, and verticality measurements. ~

This procedure provides for the cleaning of a new rack module, ifit is required, in order to meet the requirements of ANSI 45.2,1, Level C. Permissible cleaning agents, methods and limitations on materials to be employed are provided.

D. e-I ti nD TetP c ue:

This procedure stipulates the requirements for performing a functional test on a new rack module prior to installation into Pools C or D. The procedure provides direction for inserting and withdrawing a "dummy" fuel assembly into designated cell locations, and establishes an acceptance criteria in terms of maximum kinetic drag force.

Holtec International 10-2 Holtec Report HI-971760

E. t-n 11

'a This procedure stipulates the requirements for performing a functional test on a new rack module following installation into Pools C or D. The procedure willprovide direction for inserting and withdrawing a "dummy" fuel assembly into designated cell locations, and establishes an acceptance criteria in terms of maximum kinetic drag force.

F. v Underwater diving operations may be required to assist in the positioning of new rack modules.

This procedure describes the method for introducing a diver into Pools C or D, provides for radiological monitoring during the operation, and defines the egress of the diver from the fuel pool following work completion. Furthermore, this procedure requires strict compliance with OSHA Standard 29CFR-1910, Subpart T, and establishes contingencies in the event of an emergency.

G.

Consistent with Holtec International's ALARAProgram, this procedure provides details to minimize the total man-rem received during the rack installation project, by accounting for time, distance, and shielding. Additionally, a pre-job checklist is established in order to mitigate the potential for an overexposure, In the event that a visual inspection of any submerged portion of the Spent Fuel Pool liner is deemed necessary, this procedure describes the method to perform such an inspection using an underwater camera and describes the requirements for documenting any observations.

Holtec International 10-3 Holtec Report HI-971760

c c e:

This procedure describes the method to test the Spent Fuel Pool liner for potential leakage using a vacuum box. This procedure may be applied to any suspect area of the pool liner.

J. n e ater We n cedure:

In the event of a positive leak test result, an underwater welding procedure will be implemented which willprovide for the placement of a stainless steel repair patch over the area in question.

The procedure contains appropriate qualification records documenting relevant variables, parameters, and limiting conditions. The weld procedure is qualified in accordance with AWS D3.6-93, Specification for Underwater Welding or may be qualified to an alternate code accepted by CPkL and Holtec International.

'te t e ce ure:

This procedure establishes the requirements for safely storing a new rack module on-site, in the event that long term job-site storage is necessary. This procedure provides / environmental restrictions, temperature limits, and packaging requirements.

10.2 ck e e Pools C and D at Hams have been previously unused. The new rack arangement has been prepared to maximize flexibilityin the number and type (PWR vs. BWR) fuel assemblies stored.

The new rack arrangement for Pool C consists of a mixture of &ee-standing PWR and BWR Holtec racks.

Holtec International Holtec Report HI-971760

A breakdown of the number of racks and storage cells in the first campaign and completely filled configuration of Pool C is as follows:

First Campaign Filled Pool ell +gill ~ac ~

PWR Cells 360 927 BWR Cells 1320 10 2763 19 Total 1680 14 3690 30 Pool D will store a maximum of 1025 PWR assemblies in 12 rack modules. Racks willbe added to the pools on an as needed basis, A schematic plan view depicting the Spent Fuel Pools in the new maximum density configuration can be seen in Figure 1,1.

10.3 e A pool inspection shall be performed to determine ifany items attached to 'the liner wall or floor will interfere with the placement of the new racks or prevent usage of any cell locations subsequent to installation.

In the event that protrusions are found which would pose any interference to the installation process, it is anticipated that underwater diving operations and mechanical cutting methods would be employed to remove the protrusions.

Holtec International 10-5 Holtec Report HI-971760

iO.4 P l 10.4.1 The pool cooling system shall be operated in order to maintain the pool water temperature at an acceptable level. It is anticipated that specific activities, such as bearing pad elevation measurements, may require the temporary shutdown of the Spent Fuel Pool cooling system. At no time, however, willpool cooling be terminated in a manner or for a duration which would create a violation of the Harris Technical Specification or procedures.

Prior to any shutdown of the Spent Fuel Pool cooling system, the duration to raise the pool bulk temperature to 137'F will be determined. A margin temperature of 112'F is chosen such that the cooling system may be restarted prior to reaching this temperature. This will ensure that the pool bulk temperature willalways remain below 137'F.

10.4.2

~ ~ Pgg~ct~i The existing Spent Fuel Pool filtration system shall be operational in order to maintain pool clarity. Additionally, an underwater vacuum system shall be used as necessary to supplement fuel pool purification. The vacuum system may be employed to remove extraneous debris, reduce general contamination levels prior to diving operations, and to assist in the restoration of pool clarity following any hydrolasing operations.

10.5 i no w ck The new high density racks shall be delivered in the horizontal position. A new rack module shall be removed from the shipping trailer using a suitably rated crane, while maintaining the horizontal configuration, and placed upon the upender and secured. Using two independent overhead hooks, or a single overhead hook and a spreader beam, the module shall be uprighted into vertical position.

Holtec International Holtec Report HI-971760

The new rack lifting device shall be installed into the rack and each liftrod successively engaged.

Thereafter, the rack shall be transported to a pre-levelled surface where the appropriate quality control receipt inspection shall be performed.

In preparing Pool C or D for the initial rack installation, the pool floor shall be inspected and any debris which may inhibit the installation of bearing pads willbe removed. New rack bearing pads shall be positioned in preparation for the rack modules which are to be installed. Elevation measurements willthen be performed in order to gage the amount of adjustment required, ifany, for the new rack pedestals.

The new rack module shall be lifted with the Auxiliary Crane and transported along the safe load path. The rack pedestals shall be adjusted in accordance with the bearing pad elevation measurements in order to achieve module levelness after installation.

It is anticipated that the rack modules shall be lowered into the Pools C and D using the Cask Handling Crane. A hoist with sufficient capacity willbe attached to the Auxiliary Crane for installation and removal activities in order to eliminate contamination of th'e main hook during lifting operations in the pools. The rack shall be carefully lowered onto its bearing pads.

Movements along the pool floor shall not exceed six inches above the liner, except to allow for clearance over floor projections.

Elevation readings shall be taken to confirm that the module is level and as-built rack-to-rack and rack-to-wall offsets shall be recorded. The lifting device shall be disengaged and removed &om the fuel pool under Radiation Protection direction.

Holtec International 10-7 Holtec Report HI-971760

d'0.6.1 During the rack installation phase of the project, personnel safety is of paramount importance, outweighing all other concerns. All work shall be carried out in strict compliance with applicable approved procedures.

10.6.2 diati ect Radiation Protection shall provide necessary coverage in order to provide radiological protection and monitor dose rates. The Radiation Protection department shall prepare Radiation Work permits (RWPs) that will instruct the project personnel in the areas of protective clothing, general dose rates, contamination levels, and dosimetry requirements.

In addition, no activity within the radiologically controlled area shall be carried out without the knowledge and approval of Radiation Protection. Radiation Protection shall also monitor items removed Qom the pool or provide for the use of alarming dosimetry and supply direction for the proper storage of radioactive material.

10.6.3 The key factors in maintaining project dose As Low As Reasonably Achievable (ALARA)are time, distance, and shielding. These factors are addressed by utilizing many mechanisms with respect to project planning and execution, Each member of the project team will be properly trained and willbe provided appropriate education and understanding of critical evolutions. Additionally, daily pre-job briefings will be employed to acquaint each team member with the scope of work to be performed and the proper Holtec International 10-8 Holtec Report HI-971760

means of executing such tasks. Such pre-planning devices reduce worker time within the

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radiologically controlled area and, therefore, project dose.

Qjgggce Remote tooling such as liftfixtures, pneumatic grippers, a support levelling device and a liftrod disengagement device have been developed to execute numerous activities from the pool surface, where dose rates are relatively low. For those evolutions requiring diving operations, diver movements shall be restricted by an umbilical, which will assist in maintaining a safe distance from irradiated sources, By maximizing the distance between a radioactive sources and project personnel, project dose is reduced.

m~ed~n During the course of the rack installation, primary shielding is provided by the water in the Spent Fuel Pool. The amount of water between an individual at the surface (or a diver in the pool) and an irradiated fuel assembly is an essential shield that reduces dose. Additionally, other shielding, may be employed to mitigate dose when work is performed around high dose I rate sources.

10.7 wae ea Radioactive waste generated from the rack installation effort shall include vacuum filter bags, miscellaneous tooling, and protective clothing.

Vacuum filter bags shall be removed Rom the pool and stored as appropriate in a suitable container in order to maintain low dose rates.

Contaminated tooling shall be properly stored per Radiation Protection direction throughout the project. At project completion, an effort will be made to decontaminate tooling to the most practical extent possible.

Holtec International 10-9 Holtec Report HI-971760

ENVIRONMENTALCOST/BENEFIT ASSESSMENT Introduction Article V of the USNRC OT Position Paper [11.1] requires the submittal of a cost/benefit analysis for the chosen fuel storage capacity enhancement method. This section provides justification for selecting rack installation in Pools C and D as the most viable alternative.

11.2 Im erative for Increased Stora e Ca acit The specific need to increase the limited existing storage capacity at the Harris facility is based on the continually increasing inventory in Pools A and B due to core offloads at Harris and transhipments from the Robinson and Brunswick plants, the prudent requirement to maintain full-core offload capability, and a lack of viable economic alternatives.

Based on the current number of stored assemblies and estimated discharge and transhipment rates, the Harris fuel pool is projected to lose the capacity to discharge one full core in 2001.

This projected loss of storage capacity in the Harris pool would affect CP&L's ability to operate the reactors. CP&L does not have an existing or planned contractual arrangement for third party fuel storage or fuel reprocessing.

11.3 A raisal of Alternative 0 tions CP&L has determined that rack installation at the Harris pools is by far the most viable option for increasing spent fuel storage capacity in comparison to other alternatives.

Holtec International Holtec Report HI-971760

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The key considerations in evaluating the alternative options are:

Safety: minimize the number of fuel handling steps Economy: minimize total installed and OAM cost Security: protection &om potential saboteurs, natural phenomena Non-intrusiveness: minimize required modification to existing systems Maturity: extent of industry experience with the technology ALARA:minimize cumulative dose due to handling of fuel Rack installation was found by CPAL to be the most attractive option in respect to each of the foregoing criteria. An overview of the alternatives is provided in the following.

~

Rod consolidation has been shown to be a potentially feasible technology. Rod consolidation involves disassembly of spent fuel, followed by the storage

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of the fuel rods Rom two assemblies into the volume of one and the disposal of the fuel assembly skeleton outside of the pool (this is considered a 2:1 compaction ratio), The rods are stored in a stainless steel can that has the outer dimensions of a fuel assembly. The can is stored in the spent fuel racks. The top of the can has an end fixture that matches up with the spent fuel handling tool. This permits moving the cans in an easy fashion.

Rod consolidation pilot project campaigns in the past have consisted of underwater tooling that is manipulated by an overhead crane and operated by a maintenance worker. This is a very slow and repetitive process.

Holtec International 11-2 Holtec Report HI-971760

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The industry experience with rod consolidation has been mixed thus far. The principal

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of this technology

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advantages are: the ability to modularize, compatibility with DOE waste management system, moderate cost, no need of additional land and no additional required surveillance. The disadvantages are: potential gap activity release due to rod breakage, potential'or increased fuel cladding corrosion due to some of the protective oxide layer being scraped off, potential interference of the (prolonged) consolidation activity which might interfere with ongoing plant operation, and lack of suQicient industry experience.

- 'e Dry cask storage is a method of storing spent nuclear fuel in a high capacity container. The cask provides radiation shielding and passive heat dissipation. Typical capacities for PWR fuel range Rom 21 to 37 assemblies that have been removed &om the reactor for at least five years. The casks, once loaded, are then stored outdoors on a seismically qualified concrete pad. The pad will have to be located away &om the secured boundary of the site because of site limitations. The

'torage location willbe required to have a high level of security which includes sequent tours, reliable lighting, intruder detection, (E-field), and continuous visual monitoring.

The casks, as presently licensed, are limited to 20-year storage service life. Once the 20 years has expired the cask manufacturer or the utility must recertify the cask or the utility must remove the spent fuel &om the container.

There are several plant modifications required to support cask use. Tap-ins must be made to the gaseous waste system and chilled water to support vacuum drying of the spent fuel and piping must be installed to return cask water back to the Spent Fuel Pools. A seismic concrete pad must be made to store the loaded casks. This pad must have a security fence, surveillance protection, a diesel generator for emergency power and video surveillance.

Holtec International Holtec Report HI-971760

Finally, the cask facility must have equipment required to vacuum dry the cask, backfill it with helium, make leak checks, remachine the gasket surfaces ifleaks persist, and assemble the cask on-site. For casks which have closure gaskets, the space between the inner and outer lid must be continuously monitored to check for inner seal failure.

Presently, no MPC cask has been licensed. Because of the continued uncertainty in the government's policy, the capital investment to develop a dry storage system is considered to be an inferior alternative for Harris at this time.

ar Vault Vault storage consists of storing spent fuel in shielded stainless steel cylinders in a horizontal configuration in a reinforced concrete vault. The concrete vault provides radiation shielding and missile protection. It must be designed to withstand the postulated seismic loadings for the site.

A transfer cask is needed to fetch the storage canisters &om the fuel pool. The plant must provide for a decontamination bay to decontaminate the transfer cask, and connection to its gaseous waste system and chilled water systems. A collection and delivery system must be installed to return the pool water entrained in the canisters back to the fuel pool. Provisions for canister drying, helium injection, handling, and automatic welding are also necessary.

The storage area must be designed to have a high level of security similar to that of the nuclear plant itself. Due to the required space, the vault secured area must be located outside the secured perimeter. Consideration of safety and security requires it to have its own video surveillance system, intrusion detection, and an autonomous backup diesel generator power source, Some other concerns relating to the vault storage system are: inherent eventual "repackaging" for shipment to the DOE gepository, the responsibility to eventually decommission the new facility, Holtec International 11-4 Holtec Report HI-971760

large "footprint" (land consumption), potential fuel handling accidents, potential fueVclad rupture due to high temperature and high cost.

At the present time, no MPC technology based vault system has yet been offered for licensing to the USNRC. Therefore, this option is considered to be unavailable at this time.

A variation of the horizontal vault storage technology is more aptly referred to as "horizontal silo" storage. This technology suffers Rom the same drawbacks which other dry cask technologies do, namely,

i. No fuel with cladding defects can be placed in the silo.

ii. Concern regarding long-term integrity of the fuel at elevated temperature.

iii. Potential for eventual repackaging at the site.

iv. Potential for fuel handling accidents.

v, Relatively high cumulative dose to personnel in eQecting fuel transfer (compared to rack installation).

vi. Compatibility of reactor/fuel building handling crane with fuel transfer hardware.

vii. Potential incompatibility with DOE shipment for eventual off-site shipment.

viii. Potential for sabotage.

Holtec International 11-5 Holtec Report HI-971760

11.3.1 Alternative 0 tion Summa An estimate of relative costs in 1997 dollars for the aforementioned options is provided in the following:

Rack Installation: $ 12 million Horizontal Silo: $ 35-45 million Rod consolidation: $ 25 million Metal cask (MPC): $ 6S-100 million Modular vault: $ 56 million The above estimates are consistent with estimates by EPRI and others [11.2, 11.3].

To summarize, there are no acceptable alternatives to increasing the on-site spent fuel storage capacity of Harris. First, there are no commercial independent spent fuel storage facilities operating in the U.S. Second, the adoption of the Nuclear Waste Policy Act (NWPA) created a de facto throw-away nuclear fuel cycle. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium, reprocessing represents an added cost for the nuclear fuel cycle which already includes the NWPA Nuclear Waste Fund fees. In any event, there are no domestic reprocessing facilities. Third, at over $ '/2 million per day replacement power cost, shutting down the Harris reactor is many times more expensive than simply installing racks in the existing Spent Fuel Pools.

11.4 Cost Estimate The proposed construction contemplates installation of storage modules in Harris Pools C and D using free-standing, high density, poisoned spent fuel racks. The engineering and design is completed for rack installation in the pools. This rack installation project willprovide sufficient pool storage capacity to maintain full-core offload capability until the end of the current plant license.

Holtec International 11-6 Holtec Report HI-971760

The total capital cost is estimated to be approximately $ 12 million as detailed below.

Engineering, design, project management: $2 million Rack fabrication: $7 million Rack installation: $3 million As described in the preceding section, many alternatives were considered prior to proceeding with rack installation, which is not the only technical option available to increase on-site storage capacity. Rack installation does, however, enjoy a definite cost advantage over other technologies.

11.5 The expansion of the Harris Spent Fuel Pool capacity is expected to require the following primary resources:

Stainless steel: 250 tons Boral neutron absorber: 20 tons, of which 15 tons is Boron Carbide powder and 5 tons are aluminum.

The requirements for stainless steel and aluminum represent a small &action of total world output of these metals (less than 0.001%). Although the &action of world production of Boron Carbide required for the fabrication is somewhat higher than that of stainless steel or aluminum, it is unlikely that the commitment of Boron Carbide to this project willacct other alternatives.

Experience has shown that the production of Boron Carbide is highly variable and depends upon need and can easily be expanded to accommodate worldwide needs.

Holtec International Holtec Report HI-971760

11.6 v al n 'deai ns Prior to the proposed modification, Pools C and D were maintained full of water with levels consistent with those of Pools A and B. Although water was allowed to be exchanged between all four pools at various times, there was no heat load associated with Pools C and D. Therefore, the bulk pool temperatues in Pools C and D have always been maintained at or below the temperatures in Pools A and B. Due to the heat load arising from the spent fuel inventory, the pool cooling system will be connected to Pools C and D to provide adequate heat removal capabilities. The maximum normal bulk pool temperature willbe realized when the capacity is maximized for Pools C and D, but will still be x 137'F t.

Maintaining four pools (instead of the previous two pools) in the Fuel Handling Building with bulk pool temperatures F137'F t will result in an increase in the pool water evaporation rate.

This pool water evaporation increase has been determined to increase the relative humidity of the Fuel Building atmosphere by less than 10%. This increase is within the capacity of both the normal and the ESF Ventilation Systems. The net result of the increased heat loss and water vapor emission to the environment is negligible.

The 137'F limit is consistent with that currently in the Harris FSAR and procedures for pools A and B. CP&L is in the process of re-evaluating systems and components to allow for an increase the allowable bulk pool temperature.

Holtec International 11-8 Holtec Report HI-971760

11,7 Pgfere~cg

[11.1] OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, USNRC (April 1978).

[11.2] Electric Power Research Institute, Report No. NF-3580, May 1984.

[11.3] "Spent Fuel Storage Options: A Critical Appraisal", Power Generation Technology, Sterling Publishers, pp, 137-140, U.K. (November 1990).

Holtec International 11-9 Holtec Report HI-971760

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Enclosure 8 to Serial: HNP-98-188 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT SPENT FUEL STORAGE 10CFR50.55a ALTERNATIVEPLAN

Enclosure 8 to Serial: HNP-98-188 Page 1 of 13 10CFR50.55a ALTERNATIVEPLAN I. Introduction Re ulato Back round 10CFR50.55a (Codes and Standards) requires that nuclear power facilities be subject to the licensing condition that (1) structures, systems and components are designed, fabricated, erected, constructed and inspected to quality standards commensurate with the importance of the safety function to be performed, and (2) that certain systems and components of nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code. 10CFR50.55a(a)(3) allows alternatives to these requirements if with the permission of the Office of Nuclear Reactor Regulation it can be demonstrated that the proposed alternative would provide an acceptable level of quality and safety, or if compliance with the requirements would result in hardship or unusual difficultywithout a compensating increase in the level of quality and safety.

The following is an outline of a "10CFR50.55a Alternative Plan" for licensing plant systems originally intended for use in cooling and storage of Harris Units 2 and 3 spent fuel. This portion of the plant was only partially completed under the Harris Plant construction program at the time that Unit 1 was completed and was never turned over as a part of the licensed and operating facility. The completion of this spent fuel storage capacity is now needed for long term storage of spent fuel from the Harris, Brunswick and Robinson Nuclear Plants in support of continued operation of these CP&L facilities.

However, continuing its construction on the basis of the original site construction program is not viable since (1) CP&L has discontinued its N certificate holder program, and (2) certain code required construction records associated with the field installation of this piping are no longer available. This 10CFR50.55a Alternative Plan is intended to provide the basis for construction requirements for the completion of this portion of the Harris Plant and to justify the acceptability of previously constructed equipment in light of missing documentation.

Construction Histo / Chronolo Carolina Power & Light filed an application with the Atomic Energy Commission in 1971 for licenses to construct and operate its proposed Shearon Harris Nuclear Power Plant Units 1, 2, 3 and 4, in Wake County, NC. After completion of preconstruction reviews and hearings, the AEC issued Construction Permit Nos. CPPR-158, CPPR-159, CPPR-160 and CPPR-161 on January, 1978. Construction proceeded on the four unit site until December 1981, when CP&L informed the NRC that Units 3 and 4 had been canceled, and requested that Units 1 and 2 be considered concurrently for operating licenses. NUREG-1038 was issued in November 1983 for Unit 1, and reflected ongoing construction and eventual completion of Unit 2. However, Unit 2 was canceled soon

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Enclosure 8 to Serial: HNP-98-188 Page 2 of 13 afterward in December 1983, leaving Unit 1 as the only Unit to be completed and licensed. The Unit 1 Full Power Operating License was issued in January 1987, with commercial operation beginning in May 1987.

The original design of the four unit Harris Nuclear Plant located Units 1 and 4 at the south end of the plant, and Units 2 and 3 on the north end. These four units were to share a common fuel handling building to serve the purposes of loading and offloading fuel, as well as storage of spent fuel. Two sets of fuel storage pools were located in the fuel handling building, each set containing a spent fuel pool and a new fuel pool. The spent fuel pools were intended to function primarily as spent fuel storage capacity, while the new fuel pools were provided for staging new fuel and offloading spent fuel from the reactor. In the initial design, Units 1 and 4 shared the south ('A'nd 'B') fuel pools, while the north ('C'nd 'D') fuel pools were intended to service Unit 2 and 3.

The Fuel Handling Building was a common feature to all units, and completion of the building itself was requisite for operation of the first unit placed into service. Logical progression of the Fuel Handling Building construction dictated that major pieces of equipment be installed early in the schedule. As a result, the full complement of Spent Fuel Pool Cooling pools, heat exchangers and pumps initially associated with four unit construction was installed. Many of the smaller pumps, filters, strainers and lesser pieces of equipment were installed as well. Fuel Handling Building construction also dictated that all of the piping to be embedded in concrete be installed at the logical interval as the building was erected. Since the pools were encased in concrete, the adjoining portions of piping providing cooling connections and auxiliaries were necessarily constructed, inspected and tested prior to the encasement concrete being poured.

Subsequent to the cancellation of Units 3 and 4, work on the 'C'nd 'D'pent Fuel Pools continued in support of the planned completion of Unit 2. By the time that Unit 2 was canceled, the majority of the mechanical piping and equipment associated with operation of the 'C'nd 'D'nd pools was already installed, including all of the embedded and most of the exposed portions of ASME Section IIIpiping associated with these fuel pools'ooling system. Work on the remaining equipment associated with the

'C'nd 'D'ools in the Fuel Handling Building was suspended when Unit 2 was canceled. Plant documents from that time describe plans to eventually complete the

'D'pent fuel pools and place them into service.

'C'nd Construction Records Issue The completed portion of the Unit 2 Fuel Pool Cooling and Cleanup System (FPCCS) and supporting facilities were constructed to the same codes and standards and using the same procedures and personnel as was Unit 1, which was fully completed and licensed.

Appropriate records documenting field activities were generated at the time of construction as required by the construction codes and plant procedures, and maintained in storage under the control of the construction Quality Assurance (QA) program pending system completion and turnover. When construction on Unit 2 was halted, these records

Enclosure 8 to Serial: HNP-98-188 Page3 of13 were transferred to temporary storage facilities maintained by the Harris Nuclear Plant Document Control. They were not microfilmed since they were associated with systems which were not fully completed and accepted under the site's N Certificate Program, and later were inadvertently discarded during a document control records cleanup effort.

Notably, these discarded records include the piping isometric packages for field installation of the completed portion of Unit 2 Fuel Pool Cooling and Cleanup System and Component Cooling Water System (CCWS) piping within Code boundaries. As a result, Code required records are no longer available for approximately 40 of the nearly 200 large bore welds in the completed ASME Section IIIportions of the Unit 2 FPCCS and CCWS.

II. Alternative Plan for Missing Construction Records (Piping Pedigree Plan)

The plan for addressing the missing construction documentation associated with the portion of the piping initially installed during plant construction and intended for the

'D'pent Fuel Pools'ooling systems consists of four elements. These are: (1)

'C'nd scoping, (2) records retrieval and review, (3) examination and testing, and (4) reconciliation. The intent of this plan is to develop the body of evidence which supports the quality of the previously completed constructed piping. Consistent with 10CFR50.55a, any deficiencies identified will be evaluated to determine whether a acceptable level of quality and safety can be provided through alternate methods, or if not, whether attaining full compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

(1) The scoping portion of the Piping Pedigree Plan defines the boundaries of piping within the plan, and basically consists of a review of the extent of existing construction vs. that required for completion of the system. The extent of previously completed construction is determined by conducting and documenting detailed field walkdowns.

Identification markings such as spoolpiece numbers, welder identification numbers, heat numbers, etc. are recorded at this time for use later in the records review and retrieval phase. Accessibility (both external and internal) are assessed for planning the examination / testing phase.

(2) The records review and retrieval phase of the project is an investigation of construction era documents to compile the archived body of evidence which substantiates the quality of the Unit 2 Spent Fuel Cooling piping. Specific sources of this information are discussed as follows:

A) Procurement documents for piping spool pieces. Requirements to which these spool pieces were fabricated were delineated on Purchase Order NY 435035, which invoked piping spec CAR-SH-M-30. Vendor Data Packages were supplied to the requirements of the pipe spool vendor's NPT program, and

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Enclosure 8 to Serial: HNP-98-188 Page 4 of 13 include records of material certification, welding activities and Nondestructive Examination (NDE) and hydrotesting. These records were retained by the Harris Nuclear Plant Document Control Program and are available on microfilm.

B) Construction era documents which defined requirements associated with the procurement, storage, handling and installation of the piping. Work procedures fall into this category, and include those for welding, weld material control, piping installation, concrete placement, hydrotesting, etc.

Development of the sequence of installation through controlling procedures establishes the activities related to quality (tests, inspections, reviews, etc.)

which by procedure would have to be satisfactorily completed in order to meet specific documented construction milestones, such as concrete placement and hydrotest.

C) Review of records which are available through the Harris Nuclear Plant Document Control System relating to construction of the Spent Fuel Pools and related equipment. Record types which fall into this category include, hydrotest records, concrete placement tickets, records relating to pipe spool modifications, etc. In many cases records may be found which do not directly establish quality, but rather serve to demonstrate that the construction of this piping was subject to the same level of scrutiny as was comparable Unit 1 piping, for which the appropriate quality records do exist.

D) Review of construction era records which are not quality assurance records, but which do serve to substantiate the quality of construction. This category would include documents such as engineering files, or quality control inspector log books which note specific inspections or records review.

(3) An examination and test phase will recreate, to the extent possible, any inspections or records which would have originally been required by plant procedures and the construction code and for which documentation is no longer available. The primary focus of this phase will consist of inspection and NDE of field welds for which weld data records are not available. Accessible ASME Section III welds will be subject to 100%

surface examination, and ANSI B31.1 welds will receive a visual examination. Where feasible, internal weld inspections will be performed to verify fitup and adequacy of shielding gas purge. Notably, this will include an internal remote camera inspection of a substantial portion of the embedded FPCCS piping. Alternate methods of attaining comparable assurance will be developed whenever code required inspections cannot be performed, or deficiency in code required records cannot be otherwise addressed. For example, since filler material traceability cannot be established by weld data records, examination and testing of weld filler material will be performed to verify the composition of filler material is consistent with weld requirements. Finally, system hydrotesting will be performed upon completion of the piping systems using ASME Section IIIhydrotest criteria.

Enclosure 8 to Serial: HNP-98-188 Page 5 of 13 (4) The reconciliation phase of the Piping Pedigree Plan is a review of the data collected in previous phases and assessment of the level to which original construction documentation requirements were met. This is accomplished by compiling the body of records retrieved from document control and those generated by the examination / testing effort, then reviewing this record set against code documentation requirements to determine the extent to which code requirements are met. For instances wherein deficiencies are identified, the body of evidence (alternate tests or inspections, construction procedures, etc) which substantiates the quality of the component would be evaluated to determine ifcomparable assurance of quality and safety exists.

Pi in Pedi reePlan-Im lementation ASME Section III Piping:

The elements of the Piping Pedigree Plan as described above are essentially complete for the ASME Section IIIpiping associated with the 'C'nd 'D'ools'PCCS.

The following is a summary of the results of this effort to date:

Scope Definition - The ASME Section III piping associated with the 'C'nd 'D'PF Cooling System has been walked down by CP&L engineering and Harris Nuclear Plant Quality Control personnel to compare the plant configuration with construction isometric drawings and ensure that all welds, both vendor and field constructed, have been identified. Pipe spool identification numbers and welder symbols were inspected and recorded for review and comparison against vendor data packages. The scope of the ASME Section III piping within the plan has been defined based on field walkdowns, a review of modification design and results of the records retrieval effort.

Basically, the plan will cover the large bore ASME Section III piping in the FPCCS and CCWS, leaving the small bore pipe welds (vents, drains, etc.) to be cut out and redone as part of the modification effort. A total of 40 large bore piping field welds and 12 pipe hanger attachment welds are being addressed within this portion of the Alternative Plan scope. Of this total, 37 are FPCCS piping welds (15 of which are embedded in concrete) and 3 are CCWS piping welds. All 12 hanger attachment welds are in the FPCCS piping.

Vendor Data Package review - All of the 44 vendor data packages associated with the ASME Section IIIportions of the 'C'nd 'D'PCCS have been retrieved and reviewed to ensure that the requisite paperwork is in hand. These packages account for approximately 80% of the large bore piping welds in the previously constructed portions of this system . Of the nearly 200 existing large bore (12" and 16") ASME Section III FPCCS piping welds, approximately 160 are vendor welds for which all required records exist. As noted above, these vendor data packages also account for all but 12 of the hanger attachments welds existing in the FPCCS piping. Only 2 vendor data packages are associated with the portion of the previously installed Unit 2

Enclosure 8 to Serial: HNP-98-188 Page 6 of 13 CCW System which will be used in the design to tie in Unit 1 CCW to the 'C'nd

'D'pent Fuel Pool Cooling Heat Exchangers. These packages account for all but 3 of the existing large bore piping welds in this piping.

Review of other documentation - A review of other Construction Quality Control (QC) documentation in the document control system has identified that some construction information does exist for the piping in question. Notably, hydrotest records were located which show that all of the embedded piping was in fact subject to hydrotest. Completion of weldments within the hydrotest boundary and review of Weld Data Reports (WDRs) was a procedural prerequisite for conducting these hydrotests. Of these 15 embedded field welds, hydrotest records contain specific signoffs attesting to satisfactory review of completed WDRs for 9. An additional 4 embedded welds are specifically identified as being within the hydrotest boundary with a general signoff attesting to satisfactory review of weld records, while the remaining 2 can be shown to be within a hydrotest boundary with a signoff for review of welding documentation, although not specifically identified by name.

Additional information pertaining to the quality of the 15 embedded field welds can be found in QC reports (ie., nonconformance reports or deficiency disposition reports*) associated with construction of this piping. Notably, several of these records contain WDR and repair WDRs for embedded welds, providing information pertaining to welder id, filler material and / or NDE for those welds. Pipe Spool Modification packages were located on microfilm; these have been reviewed to determine ifany field changes had been made to the pipe spools as supplied from the vendor. Construction era procedures and specifications have been reviewed to identify programmatic requirements pertinent to construction quality.

(* Note - These QC records address routine construction issues which were satisfactorily resolved, and do not have any adverse implications on overall construction quality. On the contrary, the existence of such records serves to strengthen the position that construction was subject to the appropriate level of QC scrutiny.)

Field inspections - Reinspection and NDE of the 37 piping field welds and 12 hanger attachment field welds within the ASME Section III SFP Cooling System portion of the plan scope has been completed. WDRs were generated to document the inspection results; these will be reviewed by both Harris Nuclear Plant Quality Control personnel and the site Authorized Nuclear Inspector (ANI). These inspections also located and recorded weld symbols from each field weld to verify which welds were performed by the pipe spool vendor and to identify the specific welder responsible for field welds. This information was reviewed against pipe spool modification records and vendor data packages to determine that the original vendor welds were intact (had not been replaced or altered by field work), and to ensure that all welds had been identified and their origin accounted for. A total of 4 externally

Enclosure 8 to Serial: HNP-98-188 Page 7 of 13 accessible field welds were also subject to internal examination by engineering and welding craft supervisory personnel, with no anomalies being identified which might indicate substandard weld quality.

The internal examination of externally inaccessible field welds is an integral component of the Piping Pedigree Plan These inspections will be completed prior to post-modification acceptance testing. CP&L has contracted with a specialty vendor to provide remote camera inspections of a substantial portion of the embedded piping and field welds. An inspection procedure willbe developed specifically for this activity and will include detailed inspection and acceptance criteria. Based on a feasibility walkdown with the vendor, it is anticipated that greater than one third of the embedded field welds will be subject to an internal inspection in this manner.

These inspections will take place at the appropriate interval in the modification process, when pool levels are lowered and the welded piping blanks are removed.

Any discrepancies will be appropriately dispositioned at that time, including any necessary supplemental submittals to this 10CFR50.55a Alternative Plan.

Filler Material Analysis - All of the accessible large bore FPCCS piping field welds were subject to examination and/or testing to ascertain the composition of filler material. Generally, this was done using a nondestructive x-ray diffraction "alloy analyzer". In addition, chip samples were taken from three welds at random to support the validity of the alloy analyzer results. The results of this effort support that filler material alloy used in these field welds is consistent with that required by site specifications and welding procedures. The carbon steel CCWS piping welds do not lend themselves to conclusive identification using an x-ray diffraction analyzer, so the three field welds in this piping will either be subject to chemical analysis of chip samples, or as an alternative, cut out and replaced.

B31.1 Piping:

The non-safety related piping and equipment providing skimmer, purification and other support functions for the 'C'nd 'D'pent fuel pools was very nearly completed at the time of original construction. All of this piping which will be retained in the final design is considered in the scope of the piping pedigree plan. As with the ASME Section III piping, vendor records can be located for this piping, but not the construction records associated with field installation. Under B31.1 and plant welding procedures, this piping would have been subject to external visual inspection at the time of construction.

Reinspections have been performed on a large number of these field welds, with none being rejected. A complete reinspection of this piping will be accomplished as part of the modification effort, and a full system hydrotest to original construction requirements will be completed as part of post-modification acceptance testing.

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Enclosure 8 to Serial: HNP-98-188 Page 8 of 13 Pi in Pedi ree Plan Conclusion - an acce table level of ualit and safet 10CFR50.55a(a)(3) allows for the development of an alternative plan with the permission of the Office of Nuclear Reactor Regulation ifit can be demonstrated that the proposed alternative would provide an acceptable level of quality and safety, or ifcompliance with the requirements would result in hardship or unusual difficultywithout a compensating increase in the level of quality and safety. In the case of unavailable Unit 2 construction records, a great deal of evidence can be compiled to demonstrate that this piping was indeed constructed to the quality requirements consistent with the construction codes.

These are summarized as follows:

Design - CP8cL held the N certificate over the ASME Section III portion of Harris Nuclear Plant Construction. A single N Certificate program was developed and implemented uniformly to ensure code compliance for the entire site. All materials were specified to a common program using the same procurement specifications. The same welder qualification program and weld procedures, weld engineering, NDE program, and QC program were common to the site.

Work and Document Control - The Harris Nuclear Plant was designed and constructed (to the extent that it was completed) under a single construction program.

Common work control procedures, document control, warehousing and storage facilities were used throughout the site. Generally, the same pool of craft and supervisory personnel, QC personnel and engineering staff was available for construction of all four units.

Welder Qualification - Welder identification symbols have been identified at each of the externally accessible field welds, and can be traced to welders qualified to perform that weld. The chronology of precisely when a welder was qualified vs. when the weld was made is difficultto establish since the precise time the weld was performed cannot be determined, but the work control procedures ensure that the appropriate qualifications were established prior to performing weld, particularly with regard to welds within ASME Section III boundaries.

Obviously, welder identification symbols cannot be inspected and recorded for the 15 embedded welds, but again, the same program and procedures would have applied.

Work procedures specifically directed the creation of WDR packages for all welds within code boundaries and required that the supervisor ensure that welders were appropriately qualified. Besides the craft supervisor, welder qualification would have been subject to scrutiny by QC and the ANI upon review of the weld records. Of the 15 embedded field welds, QC construction reports provide the identification of welders associated with at least 3 of these welds. No direct records of welder identification have yet been located for the remaining 12 embedded field welds, but hydrostatic test records have been located which attest to the existence of completed WDR packages for these welds at the time of construction. These records contain

Enclosure 8 to Serial: HNP-98-188 Page 9 of 13 signatures individually attesting to satisfactory review of completed WDRs for 9 of the 15 embedded field welds, with an additional 4 welds being specifically identified as being within the test boundary with a general signoff attesting to satisfactory review of weld records. The remaining 2 embedded field welds were also shown to be within a hydrotest boundary, although not specifically identified by name.

Generally, the same pool of welders was available for work on Unit 2 as was for the completed Unit 1 at any point during construction. A programmatic lack of appropriate welder qualification would have represented a quality assurance breakdown in the welder qualification program for the site, not just for a given unit.

Thus, the satisfactory completion and subsequent operation of Unit 1 using a common craft pool qualified under a single welder qualification pro'gram provides strong assurance that the Unit 2 welders were also appropriately qualified.

Filler Material Identification - The WDR package generated for each field weld contained the heat number of weld filler metal which provided the traceability for this material. Since the WDRs are typically the only historical source of this information, material certification cannot be directly established for field welds without these records. However, assurance that the filler material was procured to ASME Section III requirements and supplied with traceability records is provided in Site Specification SS-021 (Purchasing Welding Materials for Permanent Plant Construction). Per this procedure, austenetic stainless steel weld filler material procured for permanent plant welding (such as would have been used in the embedded FPCCS piping) was purchased to ASME Section III requirements, including those requirements associated with traceability and certification.

Issuance and control of weld filler material was strictly controlled through the site materials control program. This program and its implementing procedures were common to all Harris units under construction. The site materials control program was regularly subject to QC audit to ensure compliance with the site ASME Section IIIProgram Manual.

An examination and testing program has been completed for the accessible large bore piping welds in the ASME Section IIIportion of the 'C'nd 'D'ools'PCCS, as well as 12 hanger welds on this piping. Each of these welds was tested either by use of a non-destructive alloy analyzer or by removing chip samples for chemical assay.

In each case, the results supported that the filler material alloy was consistent with that required by site specifications and welding procedures. Such inspections cannot be performed for the inaccessible welds, but the quality of filler metal in these welds is supported by the existence of hydrotest records as discussed above, the existence of QC records for several of these welds which do provide certification and traceability information, the procurement requirements of Site Specification SS-021, as well as satisfactory test results from the 22 accessible welds. The 3 carbon steel CCW field

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Enclosure 8 to Serial: HNP-98-188 Page 10 of 13 welds in the Piping Pedigree Plan will also be subject to chemical analysis of chip samples to verify composition.

NDE - The WDR package generated for each field weld contained the record of code required inspections and non-destructive examination. The specification of required NDE was a line item on the WDR, and completion of these examinations was affirmed by signature on the WDRs and supported by NDE records included in the respective piping isometric package. Site work control procedures required that these examinations be performed and appropriately documented, and it is clear from interviewing plant personnel that these piping isometric packages were generated and did exist until recently discarded. Since the WDRs are again the only source of this information, the completion of original construction NDE cannot be directly established for the field welds in question.

To address the issue of NDE records, each of the accessible field welds identified as being in the Piping Pedigree Plan scope has been subjected to reinspection and NDE consistent with that which would have been originally performed and found to be acceptable. Obviously, this level of NDE cannot be reperformed on the field welds embedded in concrete, but the existence of hydrotest records attesting to review of completed WDR, QC records for several of these welds which do contain the appropriate NDE records, and the satisfactory NDE of accessible field welds with no rejections provides assurance that the NDE was satisfactorily completed for the embedded welds as well.

The internal camera inspection of a large percentage of embedded field welds will also be performed against inspection criteria developed to provide both subjective examination of weld quality and, to the extent feasible, objective compliance with code and procedural requirements. While an inspection of this nature is not a Code requirement, it is significant in that it willprovide direct physical evidence of quality for the embedded field welds. These inspections will take place at the appropriate interval in the modification process, when pool levels are lowered and the welded piping caps are removed. Any discrepancies will be appropriately dispositioned at that time, including any necessary supplemental submittals to this 10CFR50.55a Alternative Plan.

In summary, the portion of the 'C'nd 'D'PCCS which were installed at the time of original plant construction were constructed under CPEcL's N Certificate program, using sitewide programs and controls for quality assurance and a common pool of craft, quality control and engineering resources. There is no evidence to support that the level of quality in this portion of Harris plant construction is any less than that of Unit 1, and indeed, it would be difficultto conceive of an unacceptable deficiency which might exist in the partially completed Spent Fuel Cooling facilities without implicating the possibility of its existence in Unit 1 as well. That Unit 1 was completed, licensed and has been in commercial operation for approximately 12 years without cause to suspect construction

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Enclosure 8 to Serial: HNP-98-188 Page 11 of 13 quality provides strong assurance of that the quality assurance programs for the site were suitably comprehensive and fully implemented. It follows that a comparable level of quality exists in the partially completed Unit 2 facilities, including those for spent fuel storage.

Beyond programmatic assurances, a large body of evidence has been compiled which directly attest to quality of construction. Vendor data packages, hydrostatic test records, QC records and other construction era documentation has been retrieved which constitute substantial proof of compliance with site programs and procedures. An examination effort has been completed in which code required external NDE of accessible welds has been reperformed with no rejectable indications, and material examinations provide proof that the filler metal used in field welds was appropriate for the weldment. These results provide direct evidence of the quality of accessible field welds, and by extension, the smaller group of welds which are embedded. Internal examination of a significant percentage of these embedded field welds provides an additional measure of quality assurance beyond that required by the Code.

There is no evidence that supports that the missing records were never generated, and to the contrary, document control records indexes indicate that these piping isometric packages were transferred to QA storage and maintained there until they were inadvertently discarded in a document control "cleanup effort". Adverse Condition Report 93-354 was generated at that time which specifically identifies that installation documentation for the 'C'nd 'D'PCCS, including installation verification data and field weld records, was inadvertently discarded during Sept. 1993.

It is concluded that the Piping Pedigree Plan outlined above provides ample evidence exists to support that the portion of the Harris plant associated with the 'C'nd 'D'pent Fuel Pools which was completed during the original site construction effort was indeed constructed to the appropriate level of quality and safety and in compliance with construction code requirements. It follows that the issue of missing code documentation is simply that, a documentation issue, and does not infer a physical lack of quality in the field.

III.Alternative Plan for Continuance of Design and Construction The original construction of the Harris Nuclear Plant was subject to the full requirements of ASME Section III of the ASME Boiler and Pressure Vessel Code under the authorization of a single N Certificate program maintained by CP&L. This site ASME Section III QA program was discontinued shortly after completion and turnover of Unit 1, and a corporate QA program meeting 10CFR50 Appendix B requirements was implemented as required to address plant operation, including Section XI requirements regarding inspection, repair and replacement activities. Thus, the original construction program no longer exists and it is not possible to complete construction of the 'C'nd

Enclosure 8 to Serial: HNP-98-188 Page 12 of 13

'D'PCCS as a continuance of this program. Further, since a Code data report was not prepared by CP&L for this partially completed piping and equipment under its N certificate holder program at the time it was constructed, responsibility for its construction cannot be now assumed by another N certificate holder under a current program. It follows that it is not possible to N stamp the previously completed portion plant associated with the 'C'nd 'D'pent Fuel Pools. Given this, and considering that the majority of construction has been completed, it is the opinion of CP&L and code authorities within the Hartford Steam Boiler Inspection and Insurance Co. and Bechtel Power Corporation that there is no benefit with invoking an N certificate program to govern the completion of the relatively small outstanding portion of construction vs.

using another suitable quality assurance program of comparable rigor.

Since this portion of the plant was never turned over at the time of construction, it is not considered part of the operating facility from the perspective of the ASME code and its completion could not be interpreted as a replacement activity as defined in Section XI.

However, the site Section XI Repair and Replacement Program as implemented under the Corporate 10CFR50, Appendix B QA Program does contain many elements of quality control (ie., welder qualification, weld procedures, inspections, documentation, etc.)

consistent with the original construction program. Therefore, CP&L proposes to complete the design of this portion of the plant to appropriate ASME Section III requirements, but utilize the Corporate 10CFR50, Appendix B QA Program and site procedures for those elements of quality assurance for which it is appropriate to provide.

Generally, any conflicts between the ASME Section III requirements and that of the Corporate 10CFR50, Appendix B QA Program (and the corporate and site procedures which invoke it) would be conservatively dispositioned, such as the use of ASME Section III hydrotest requirements vs. those requirements found in Section XI.

A set of supplemental quality assurance requirements has also been developed to augment the Corporate 10CFR50, Appendix B QA Program in completion of the Code portions of the plant associated with the 'C'nd 'D'pent Fuel Pools. These requirements were obtained by a close review of the requirements in the approved ASME Section III Construction QA Program Manual as it existed at the time of completion of construction vs. those of the currently existing Corporate 10CFR50, Appendix B QA Program, and are specifically intended to identify and conservatively reconcile deficiencies in the corporate program with ASME Section III requirements. For instance, the supplemental requirements specify a level of ANI involvement commensurate with ASME Section III requirements, including review of work packages prior to field issuance, integration of ANI involvement into the work control process, and final review and approval of documentation subsequent to work completion. Other highlights of the supplemental quality assurance requirements include integration of comparable requirements for design specifications and a process for system documentation review and turnover similar to that of N Stamping. These supplemental quality assurance requirements will be implemented by integration into the modification package, or when necessary, by procedure revision.

Enclosure 8 to Serial: HNP-98-188 Page 13 of 13 Since the current Corporate 10CFR50, Appendix B QA Program is sufficient to govern ongoing operation of the Harris Plant (including Section XI repair and replacement activities), it follows that it is of sufficient rigor for the construction effort to complete and activate the portion of the plant associated with the 'C'nd 'D'pent fuel pools.

There are instances wherein the Corporate 10CFR50, Appendix B QA Program does not address specific ASME Section III quality assurance requirements, and a set of supplemental quality assurance requirements has been developed specifically for the purpose of addressing these items. This approach for continuance of construction is both technically acceptable and commercially viable, and will ensure the requisite level of quality and safety in the completed systems as discussed in 10CFR50.55a(a)(3)(i).

Enclosure 9 to Serial: HNP-98-188 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT SPENT FUEL STORAGE UNREVIEWED SAFETY QUESTION ANALYSIS

Enclosure 9 to Serial: HNP-98-188 Page1of 4 CCW VNRKVIKWKDSAFETY QUESTION DISCUSSION As part of the preparation of the design change package for the tie-in of the existing Component Cooling Water (CCW) system, a 10CFR50.59 Safety Evaluation was prepared. The scope of the evaluation addressed the tie-in of the Unit 1 CCW system to the heat exchangers of the 'C'nd 'D'uel Pool Cooling and Cleanup System (FPCCS).

This evaluation considered a heat load of no more than 1.0 MBtu/hr'n the 'C'nd Fuel Pools (SFP). In support of this design change package, a thermal-hydraulic

'D'pent model was created to analyze the overall impact of this additional heat load, including its impact on the Emergency Service Water (ESW) system and the Ultimate Heat Sink (UHS). This analysis demonstrated that adequate thermal margin exists in the CCW system to accommodate the proposed additional heat load in Spent Fuel Pools 'C'nd 'D'.

However, it was determined that while the post-modification configuration was safe it was potentially an Unreviewed Safety Question (USQ). The following discussion delineates the methodology used in this analysis and the reasoning behind its classification as a USQ.

CURRENT SYSTEM CONFIGURATION The CCW system serves as an intermediate closed cooling water system between the radioactive or potentially radioactive systems and the non-radioactive service water system. The FPCCS rejects its heat via the CCW system which in turn rejects its heat via the station service water system to the Ultimate Heat Sink. The Ultimate Heat Sink is comprised of three separate possible cooling sources that are used independently: the main cooling towers for normal service and the auxiliary or main reservoir for emergency service.

The CCW system provides cooling to various safety related (RHR Heat Exchangers, RHR Pump, and Spent Fuel Pool Heat Exchangers) and non-safety related heat loads.

The CCW system contains two separate trains, each containing a component cooling water system heat exchanger. There are three component cooling water pumps for the two trains. Two pumps are normally operated during cooldown, with each pump supplying half of the total component cooling water flow. Normal power operation only requires one pump for operation with another on standby. In the event of a LOCA, only one pump is required although two CCW pumps start to ensure cooling flow to the safeguards loads in the event of a single failure.

When the Emergency Core Cooling System is aligned to recirculate from the containment sump to the Reactor Coolant System, the CCW trains are separated from each other and from the non-essential header to maintain protection against a single passive failure and to, provide sufficient flow to their respective RHR trains. In this alignment, each CCW train

'ontrolled by revised Technical Specification 5.6

Enclosure 9 to Serial: HNP-98-188 Page2of 4 is balanced to provide greater than 5 gpm to the RHR pump for cooling the pump and 6050 gpm is available to the RHR heat exchanger.

The minimum CCW flow that must be maintained through the RHR Heat Exchanger and the RHR Pump subsequent to alignment to recirculation is 5600 gpm and 5 gpm respectively. Subsequent to alignment to recirculation the operators are directed by Operating Procedures to restore sufficient CCW flow from one CCW train to the SFP heat exchangers to maintain the temperature of the spent fuel pools to less than 150'F.

Based on the CCW flows established to the RHR heat exchanger and the RHR pump when the non-essential header is isolated, each train is capable of individually providing the specified 5600 gpm and 5 gpm in addition to the minimum flow of 1789 gpm through the SFP heat exchangers 'A'nd 'B'.

10CFR50.59 SAFETY EVALUATIONOVERVIEW Performance of the 10CFR50.59 Safety Evaluation requires that certain questions must be answered to determine ifthe proposed activity will require the completion of an Unreviewed Safety Question Determination (USQD). Since this design change involved a change to the Technical Specifications (to facilitate the control of the heat loads in Spent Fuel Pools 'C'nd 'D') it could not be implemented without prior NRC approval.

Nonetheless it was determined that a USQD be performed since this modification involves a change to the facility, a change to procedures described in the SAR, a change to the licensed operator training program, etc. and no previously approved USQ determination fully bounds this activity.

UNREVIEWED SAFETY UESTION DETERMINATION The USQD analysis performed yielded an affirmative answer to the question concerning whether the proposed activity may reduce the margin of safety as defined in the basis for any Technical Specification. The portion of the design change which triggered this affirmative response centered on the analysis methodology used in the thermal-hydraulic analysis to verify that adequate excess thermal capacity existed in the CCW system to accommodate the additional heat loads from Spent Fuel Pools 'C'nd 'D'. The following is a discussion of the subject thermal-hydraulic analysis and the logic that prompted the decision to categorize this activity as a USQ.

The new thermal-hydraulic analysis was performed to evaluate the 1.0 MBtu/hr heat load that would be added to Spent Fuel Pools 'C'nd 'D's a result of this activity. This thermal-hydraulic analysis includes an assessment of Core Shuffle and Abnormal Full Core Offload scenario heat loads to satisfy the analysis requirements of NUREG-0800 (Standard Review Plan). The analysis demonstrates that adequate margin exists during all normal and accident modes of system operation and that the CCW system has

Enclosure 9 to Serial: I-INP-98-188 Page3of 4 adequate thermal-hydraulic capacity to provide the minimum flow required by the fuel pool heat exchangers after the activation of Pools 'C'nd 'D'. As a result of the analysis, the minimum CCW flow to the RHR heat exchangers and the minimum ESW flow to the CCW heat exchanger change from the current requirements.

The analysis considered the additional spent fuel pool cooling heat load well as a 6%

modeling uncertainty and degraded IST pump performance. The new analysis also accounts for the change in RHR heat exchanger performance as it relates to the variation in fluid properties. This is a departure from the current licensing basis with regard to RHR heat exchanger performance. Current analyses assume that the performance of the RHR heat exchanger is fixed based on the design values associated with the heat exchanger data sheet. The data sheet fixes the tubeside inlet temperature to the RHR heat exchanger to 139'F, however, during the development of the new thermal-hydraulic analysis it was noted that RHR tube side inlet temperature is postulated to rise to 244.1'F during the initial phase of containment sump recirculation. This increase in the tube side fluid temperature is predicted to increase the overall heat transfer coefficient approximately 10% due to the change in tube side fluid viscosity. These conditions tend to increase heat transfer through the RHR heat exchanger and might otherwise increase CCW system supply temperatures above the maximum of 120'F under limiting conditions of minimum CCW heat exchanger ESW flow and maximum ESW supply temperature. The two previously mentioned changes in minimum CCW flow to the RHR heat exchangers and the minimum ESW flow to the CCW heat exchanger are specified to address this issue.

The minimum specified CCW system flow to the RHR heat exchanger is reduced to a level consistent with a heat rejection of 111.1 MBtu/hr under the new analysis. It is important to note that this heat rejection rate is consistent with the existing post-LOCA containment pressure/temperature calculations, such that no change in containment heat removal is prescribed. The thermal-hydraulic calculation includes an analysis of RHR heat exchanger performance to determine the minimum shell side flow rate to maintain 120'F shell side inlet temperature, 244.1'F tube side inlet temperature and 1.846E6 lbm/hr tube side flow rate to maintain the aforementioned consistency. It was shown that a minimum CCW system flow rate of 4874 gpm at 120'F is required at the beginning of the sump recirculation phase. The specified CCW system flow to the RHR heat exchanger under these conditions; assuming 6% model uncertainty consistent with previously developed hydraulic models is 5166 gpm, or approximately 5200 gpm. As the containment sump temperature decreases, the minimum required CCW system flow rate decreases based on maintaining a maximum RHR heat exchanger tube side outlet temperature of 180'F. The CCW system was initially rebalanced in the model in the LOCA recirculation (RHR only) alignment, with a 10% degraded CCW pump curve.

When the nominal CCW pump curve is applied to this alignment CCW system flow to the RHR heat exchanger increases to approximately 5440 gpm, resulting in an increased RHR heat exchanger heat duty of 118 MBtu/hr. Under the most limiting postulated conditions, the increased RHR heat exchanger duty could increase CCW system supply

Enclosure 9 to Serial: HNP-98-188 Page 4 of 4 temperature marginally above its 120'F design limit. This concern is addressed by increasing the current minimum required ESW flow to the CCW system heat exchanger from 8250 gpm to a slightly higher value of 8500 gpm.

Summarizing the preceding discussion, a reduction in the minimum specified RHR heat exchanger CCW system flow from 5600 gpm to 5200 gpm and an increase in the minimum specified CCW heat exchanger ESW system flow from 8250 gpm to 8500 gpm are prescribed by the new thermal-hydraulic analysis in order to maintain all thermal/hydraulic assumptions which are used in the HNP containment analysis. A minimum specified ESW system flow of 8500 gpm to the CCW heat exchangers was verified to be within the capacity of the current system even considering the most limiting ESW system single failure.

Per CP&L's Draft SER OI 365 - ASB Question 9.2.2(1) Revised Response, 5600 gpm was the number specified to the NRC as that which was "...sufficient capacity..." from one train of CCW "...to carry the heat loads from the ... RHR heat exchanger".

Section 92 2 of the SER(NUREG-103S) states that "5600gpm would be ~re wiredfor the RHR heat exchanger" and that ".. flow remaining from one operating CCS'train would be sufficient to keep the Unit 1 SFP at a temperature of150'F or less". In this context, it follows that the NRC's acceptance of the CCW system is based, in part, on ensuring that 5600 gpm CCW system flow is provided to the RHR heat exchangers under these conditions. Therefore, the decrease in minimum required CCW system flow to the RHR heat exchangers is deemed to be a reduction in the acceptance limit. The change in the minimum specified RHR heat exchanger CCW system flow from 5600 gpm to 5200 gpm as a result of the new thermal-hydraulic analysis does not prevent the CCW system from meeting the previously defined criteria in any way. The addition of Spent Fuel Pools 'C'nd 'D'o the CCW system does not directly result in changing the minimum specified RHR heat exchanger CCW system flow. As previously discussed, an increase in the minimum specified CCW heat exchanger ESW system flow from 8250 gpm to 8500 gpm also results from the new thermal-hydraulic analysis but unlike the minimum specified RHR heat exchanger CCW system flow, this value is not mentioned in the SER.

SUMMARY

In determining whether or not the proposed activity reduces the margin of safety, as defined in the basis of any Technical Speciflcation, the only item which could not be ruled out was that associated with the reduction in the minimum CCW flow to the RHR heat exchanger. Since this is deemed to be a change in the acceptance limit, this activity is considered to be a USQ.

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