ML20083H638

Draft Licensing Rept on High Density Spent Fuel Racks for VC Summer Nuclear Station

ML20083H638
Person / Time
Site:	Summer
Issue date:	12/31/1983
From:	SOUTH CAROLINA ELECTRIC & GAS CO.
To:
Shared Package
ML20083H625	List: ML20083H622 ML20083H638
References
NUDOCS 8401040597
Download: ML20083H638 (182)

Text

{{#Wiki_filter:. LICENSING REPORT Q-ON HIGH-DENSITY SPENT FUEL RACKS MAFT VIRGIL C. SUMMER' NUCLEAR STATION NRC DOCKET NO. 50-395 O SOUTH CAROLINA ELECTRIC & GAS COLUMBIA, SOUTH CAROLINA 29218 DECEMBER,1983 1 D ADO K O O O 5 p PDR _ _ _ _. - _. _ _ _ _ _ _.. _ _ _. _ - - _ _ _ _ - _. _ _ _.. _ _ _ _, _. - - _.. ~ _. _, _.. _ _ _ _ _ _,. _, _, _ _.

l TABLE OF CONTENTS (,). Page SECTION 1 - INTRODUCTION 1.0 Introduction 1-1 SECTION 2 - GENERAL ARRANGEMENT 2.0 General Arrangement 2-1 SECTION 3 - RACK CONSTRUCTION 3.1 Fabrication Details 3-1 3.1.1 Regions 1 and 2 3.1.2 Region 3 3.2 Codes, Standards and Practices 3-7 for the Spent Fuel Pool Modification SECTION 4 - NUCLEAR CRITICALITY ANALYSIS 4.1 Design Bases 4-1 4.2 Summary of Criticality Analyses 4-3 ,( ) 4.2.1 Normal Operating Conditions 4-3 4.2.2 Abnormal and Accident Conditions 4-3 4.3 Reference Fuel Storage Cell 4-8 4.'3.1 Reference Fuel Assembly 4-8 4.3.2 Region 1 Storage Rack 4-9 4.3.3 Region 2 Storage Cells 4-9 4.3.4 Region 3 Storage Cells 4-9 4.4 Analytical Methodology 4-14 4.4.1 Reference Analytical Method and Bias 4-14 ,4.4.2 Manufacturing Tolerances and Uncertainties 4-15 4.4.3 Fuel Burnup Calculations 4-15 ,O \\,_./ i i e.-.

TABLE OF CONTENTS (Continued) U 4.4.4 Long-Term Decay 4-17 4.4.5 Effect of Axial Burnup Distribution 4-18 4.5 Reference Subcriticality and Mechanical Tolerance Variations 4-23 4.5.1 Nominal Design Cases 4-23 4.5.2 Boron Loading Variation 4-24 4.5.3 Storage Cell Lattice Pitch Variations 4-25 4.5.3.1 Inner Water Thickness variations 4-25 4.5.3.2 Outer (Flux-Trap) Water Thickness Variation 4-25 4.5.4 Stainless Steel Thickness Variations 4-26.' 4.5.5 Fuel Enrichment and Density Variation 4-26 4.5.6 Boraflex Width Tolerance Variation 4-27 O 4.6 Abnormal and Accident Conditions 4-28 4.6.1 Eccentric Positioning of Fuel Assembly in 4-28 Storage Rack 4.6.2 Temperature and Water Density Effects 4-28 4.6.3 Dropped Fuel Assembly Accident 4-29 4.6.4 Abnormal Positioning of Fuel Assembly Outside Storage Rack 4-29 4.6.5 Lateral Rack Movement 4-30 SECTION 5 - THERMAL-HYDRAULIC CONSIDERATIONS 5-1 l 5.1 Decay Heat Calculations for the Spent Fuel 5-1 Cooling 5.1.1 Basis 5-1 5.1.2 Model Description 5-3 5.1.3 Decay Heat Calculation Results 5-6 1 1 iii

TABLE OF CONTENTS (Continued) ) O 5.2 Thermal Hydraulic Analysis for Spent Fuel 5-10 Cooling 5.2.1 Basis 5-10 i 5.2.2 Model Description 5-11 5.2.3 Results 5-13 REFERENCES TO SECTION 5 5-16 SECTION 6 - STRUCTURAL ANALYSIS 6-1 6.1 Analysis Outline 6-1 6.2 Fuel Rack - Fuel Assembly Model 6-3 6.2.1 Assumptions 6-3 6.2.2 Model Description 6-5 6.2.3 Fluid Coupling 6-6 6.2.4 Damping 6-7 6.2.5 Impact 6-8 () Assembly of the Dynamic Model 6-8 6.2.6 6.3 Stress Analysis 6-12 6.3.1 Stiffness Characteristics 6-12 6.3.2 Combined Stresses and Corner Displacements 6-13 6.4 Time Integration of the Equations of Motion 6-14 6.5 Structural Acceptance Criteria 6-17 6.6 Results 6-22 6.7 Summary of Mechanical Analyses 6-24 REF.ERENCES TO SECTION 6 6-27 SECTION 7 - SPENT FUEL POOL FLOOR STRUCTL \\L ANALYSIS 7.1 Introduction 7-1 7.2 Analysis Methods 7-1 O iv

TABLE OF CONTENTS (Continued) 7.3 Assumptions 7-2 i 7.4 - Load Combinations 7-2 7.5 Results 7-3 7.6 Conclusions 7-4 SECTION 8 - ENVIRONMENTAL EVALUATION 8.1 Summary 8-1 8.2 Characteristics of Stored Fuel 8-2 8.3 Related Industry Experience 8-3 8.4 V.C. Summer Operating-Expetience 8-4 8.5 Spent Fuel Pool Cooling and Clean-Up System (FPCC) 8-4 8.6 Fuel Pool Radiation Shie,lding 8-6 () 8.7 Radiological Consequences 8-6 8.8 Re-racking Operation 8-7 8.9 Conclusions 8-7 References to Section 8 8-9 SECTION 9 INSERVICE SURVEILLANCE PROGRAM FOR 10-1 l BORAFLEX NEUTRON ABSORBING MATERIAL 9.1 Program Intent 10-1 9.2 Description of Specimens 10-1 .9.3 Test 10-2 9.4 Specimen Evaluation 10-2 SECTION 10 COST / BENEFIT ASSESSMENT 10-1 10.1 Specific Needs for Spent Fuel Stor, age 10-1 O. l l i

e-TABLE OF CONTENTS (Continued) 10.2 Cost of Spent Fuel Storage 10-2 10.3' Alternatives to Spent Fuel Storage 10-2 10.4 Resource Commitments 10-4 REFERENCES TO SECTION 10 10-5 SECTION ll Quality Assurance Program 11-1 11.1 Introduction 11-1 11.2 General 11-1 11.3 System Highlights 11-1 11.4 Summary 11-? O i e i 4 vi ..,,,..-..-,y e, ...---.-.-.__--_,....--,,---.w- ..,r---..,es-- - - - + - ' - -

LIST OF FIGURES w (_) Page SECTION 2 Fig. 2.1 Module Layout 2-5 SECTION 3 Fig. 3.lA 3x3 Typical Array (Poisoned Cells) Region 1 3-10 3.lB 3x3 Typical Array (Poisoned Cells) Region 2 3-11 3.2(a) Small Angular Subelement 3-11 3.2(b) Large Angular Subelement 3-11 3.3 Basic Cell Element 3-12 3.4A Welding Sequence of Composite Boxes 3-14 3.4B Typical Cell Elevation 3-15 3.5 Adjustable Support 3-16 3.6 Fixed Support 3-17 3.7 3x3 Typical Array (Unpoisoned Cells) Region 3 3-18 SECTION 4 3 ~ s,) Fig. 4.1 Limiting discharge fuel burnup for Region 2 storage rack for fuel of various initial enrich-ments 4-6 4.2 I.imiting discharge fuel burnup for Region 3 4-7 storage rack for fuel of various initial enrich-ments 4.3 Configuration of Region 1 spent fuel storage cell 4-11 4.4 Configuration of Region 2 spent fuel storage cell 4-12 4.5 Configuration of Region 3 spent fuel storage cell 4-13 l 4.6 Decrease in K. with fuel burnup under hot 4-19 operating conditions 4.7 Long-term change in infinite multiplication factor (k.) of 4.3% enriched fuel burned to 20 Mwd /kgU 4-20 O vii

LIST OF FIGURES (Continued) O 4.8 Long-term change in infinite multiplication factor (k.) for 4'.3% enriched fuel burned to 42 Mwd /kgU 4-21 4.9 Relative axial burnup distribution from NUREG/ CR-0722 4-22 4.10 Infinite multiplication factor (k.) of 4.3% enriched fuel assemblies separated only by water. 4-31 SECTION 5 Fig. 5.1.1 (a) Pool Bulk Temperature, Normal Discharge 5-17 (b) Pool Bulk Temperature, Normal Discharge 5-18 (c) Power Discharge, Normal-Discharge 5-19 (d) Power Discharge, Normal Discharge 5-20 5.1.2 (a) Pool Bulk Temperature, Full Core Discharge 5-21 (b) Pool Bulk Temperature, Full Core Discharge 5-22 (c) Power Discharge, Full Core Discharge 5-23 (d) Power Discharge, Full Core Discharge 5-24 5.1.3 (a) Pool Bulk Temperature, Normal Discharge 5-25 With Loss of 1 SFPHX l (b) Pool Bulk Temperature, Normal Discharge 5-26 with loss of 1 SPPHX (c) Power Discharge, Normal Discharge with loss of 1 SFPHX 5-27 (d) Power Discharge, Normal Discharge with loss of 1 SFPHX 5-28 5.1.4 (a) Pool Bulk Temperature, 1/3 core Discharge j with loss of 1 SFPHX 5-29 (b) Pool Bulk Temperature, 1/3 Core Discharge 5-30 with Loss of 1 SFPHX O \\# viii

LIST OF FIGURES (continued) (G_j (c) Power Discharge, 1/3 Core Discharge 5-31 with Loss of 1 SFPHX (d) Power Discharge, 1/3 Core Discharge 5-32 with Loss of 1 SFPHX 5.2.1 Idealization of Rack Assembly 5-33 5.2.2 Thermal Chimney Flow Model 5-34 SECTION 6 Fig. 6.1 Dynamic Model 6-29 6.2 Impact Springs and Fluid Dampers 6-30 6.3 Spring Mass Simulation for Two Dimensional Motion 6-31 6.4(a) Horizontal Cross Section of Rack 6-32 6.4(b) Vertical Cross Section of Rack 6-32 r 6.5 Dynamic Model 6-33 () 6.6 Stress Resultants Orientation 3-34 6.7 Subdivision of a Typical Rack 6-35 6.8 Finite Elements Model Cross Section 6-36 LIST OF FIGURES j (continued) i 6.9 Horizontal OBE, 4 percent Damping 6-36 6-10 Horizontal Y, OBE, 4 percent Damping 6-38 6-11 Vertical OBE, 4 percent damping 6-39 9 ix l l

LIST OF FIGURES (continued) SECTION 7 Spent Fuel Pool Structural Analysis Fig. 7.1 Plan on Spent Fuel Pool 7-7 7.2 Overall Foundation Plan 7-8 7.3 Section Looking West through Spent Fuel Pool 7-9 7.4 Section Looking North through Spent Fuel Pool 7-10 SECTION 9 Fig. 9-1 Test Coupon 9-5 l [ i O x I

3 LIST OF TABLES f. Page SECTION 1 Table 1.1 V.C. Summer Nuclear Station Fuel Assembly Discharges (Tentative Schedule 3 1-3 SECTION 2 Table 2.1 Design Module Data 2-3 Table 2.2' Module Data 2-4 SECTION 3 Table 3.1 Boraflex Experience for High Density Racks 3-2 SECTION 4 I Table 4.1 Summary of Criticality Safety Analyses 4-4 4.2 Reactivity Effects of Abnormal and Accident Conditions 4-5 4.3 Fuel Assembly Design Specifications 4-8 () 4.4 Comparison of Cold, Clean Reactivities Calculated at 20 Mwd /kgU and 42 Mwd /kgU 4-16 4.5 Long-Term Changes in Reactivity 4-17 4.6 Effect of Temperature and Void on Calculated Reactivity of Storage Rack 4-29 SECTION 5 Table 5.1.1 List of Cases Analyzed 5-7 5.1.2 Maximum Pool Bulk Temperature, t, Coincident 5-8 Total Power Qi and Coincident Specific Power for the Hottest Assembly -5.1.3 Time (Hrs) to Boiling and Boiling Vaporiza-5-9 tion Rate From the Instant All Cooling is Lost 5.2.1 Maximum Local Pool Water Temperature and 5-14 Local Fuel Cladding Temperature At Instance of Maximum Pool Bulk Temperature xi r- ~,,, _ _, - - _,p,, ,_.-,,_____m__,_,.s,,.-.,,,,,,,,,,,,,_,,,,_,..,,,y.. .,,-_,.,-,y,,m._,,-, ,..-,,..-,_,m

LIST OF TABLE (Continued) 5.2.2 Pool and Maximum Cladding Testperature At the 5-15 Instance Fuel Assembly Transfer Begins SECTION 6 Table 6.1 Degrees of Freedom 6-5 6.2 Numbering System for Springs, Gap Elements, 6-10 Friction Elements 6.3 Phys'ical Property 1httu - e 6-19 6.4 Support Material Data 6-19 6.5 File DSCLOl Module A (llxll), Coef = .8, 6-28.1 Full Rack 6-28.2 SECTION 7 Table 7.1 Caisson Evaluation-Required vs. Minimum Available Capacity 7-5 7.2 Required vs. Available Capacity In Spent Fuel Pool Walls and Slab 7-6 Section 9 Table 9.1 Time Schedule for Removing Coupons 9-4 v xii

1. INTRODUCTION The purpose of thi.s rreport is to describe the design, fabrication, and safety *. analysis of High Density Spent Fuel Storage Racks produced by.Itneph Gat Corporation for Virgil C. Summer Nuclear Station. 'VJC Stummer Nuclear Station is co-owned. by South Carolina Elec trjit racd Gas Company and South Carolina Public Service Authoritrf, with the former serving as the principal owner and oper ating age at. The plant is located in Jenkinsville, South Carol'mra, ipproximately 26 miles northwest of Columbia, South Carolina.. V.C. Summer is a single innit Pressurized Water Reactor (Westinghouse design) witib 'a design capacity of 900 Mwt(e). The reactor core contains 157 fuel assemblies rated to produce 2775 thermal megawatts. At present there are no stored fuel assemblies in the pool. The power station is slated to go into commercial operation in January, 1984. The plant is currently licensed for the storage of 682 spent fuel assemblies at a maximum of 3.5% enrichment. As shown in Table 1.1, the storage pool will lose full core discharge capability in the year 1994. The proposed pool storage densification will equip the pool with 1276 storage locations. Table 1.1 indicates that the proposed reracking of the pool will provide adequate storage with full core discharge capability well into the next century (c. 2008). Table 1.1 is based on a conservatively estimated 18 month fuel cycle. Current trends towards extended burn-up and higher enrichment would further extend the time span of on-site storage. The racks proposed herein are of free

standing, self supporting variety.

The principal materials

d. construction are

+ ASTM SA240, Type 304 for the storage locations and "BOR4 FLEX", a patented product of BISCO (a division of Brand, Inc.). OV 1-1

B The specifications for design, construction and quality assurance for the high-density spent fuel storage racks were O ><9ra =v se ea c re11=- =1eceric

1. a c-ce 9 r-The

' mechanical design, seismic / structural analysis, thermal-hydraulic

analysis, and other related calculations as well as the fabrication of the hardware are performed by Joseph Oat Corporation.

Southern Science, a division of Black & Veatch, is serving as a consultant to Joseph Oat Corporation in the area of criticality analysis. Gilbert Commonwealth of

Reading, Pennsylvania is providing the analytical support in the areas of pool slab / wall qualification, and radiological considerations.

The analyses performed by Joseph Oat Corporation in conjunction with Gilbert Commonwealth and Black and Veatch demonstrate that acceptable margins of safety exist with respect to appropriate NRC and ASME acceptance criteria. A, cost-benefit comparison of several potential spent fuel disposition. alternatives indicate that reracking of the V.C. Summer Pool is the lowest risk and most cost-effective alternative and that neither the reracking operation nor the increased on-site storage of irradiated material pose an increased hazard to the Plant Staff or the public. 7 l l The following sections provide a synopsis of the de s ig'n, fabrication, neutronics

analysis, thermal / hydraulic

analysis, l

structural analysis, accident analysis, environmental analysis and cost-benefit appraisal of the High Density Spent Fuel Racks. In particolar, the integrity of the rack structure under the specified combinations of inertial, seismic, and mechanical loads and thermal gradient per NUREG-0800 is demonstrated. Also l included are concise descriptions of the rack In-service Surveillance Program and the Joseph Oat Corporation Quality Assurance Program. The Joseph Opt Corporation Quality Assurance Program has been reviewed and found acceptable for engineered j fabrication of ASME Section III Class 1, 2, 3 and MC components by both ASME and NRC. O l-2 l-

f-'s Table 1.1 (,) V.C. Summer Nuclear Station Fuel Assembly Discharge (Tentative Schedule) Remaining Remaining Storage Storage Total Discharged Capability with-Capability Refueling Discharge Assemblies in Spent Fuel out Proposed with Proposed Date Assembly Pool Following Refueling Expansion Expansion Fall 1984 44 44 638 1232 Fall 1985 72 116 566 1160 Spring 1987 68 184 498 1092 Fall 1988 72 256 426 1020 Spring 1990 68 324 358 952 Fall 1991 72 396 286 880 Spring 1993 68 464 218 812 Fall 1994 72 536 146* 740 Spring 1996 68 604 178 672 Fall 1997 72 676 6** 600 Spring 1999 68 744 532 Fall 2000 72 816 460 Spring 2002 68 884 392 Fall 2003 72 956 320 Spring 2005 68 1024 252 Fall 2006 72 1096 180 Spring 2008 68 1164 112* 40** Fall 2009 72 1236 Full core discharge capability lost (157 assemblies)

• • Normal discharge capability lost (= 72 assemblies) n m

1-3

2. GENERAL ARRANGEMENT A The high density spent fuel racks consist of individual cells with 8.85" x 8.85" (nominal) square cross section, each of which accommodates a single PWR (Westinghouse) fuel assembly. The cells are arranged in modules of varying size. A total of 1276 cells are arranged in 11 distinct modules in 3 regions. Region 1 is designated for storage of freshly discharged fuel assemblies with enrichments up to 4.3 weight percent U-235. The cells in Region 2 are reserved for accommodating fuel assemblies with initial enrichments of 4.3 weight percent U-235 and a minimum burnup of 20,000 MWD /MTU. The remaining cells, i.e. Region 3 cells, are capable of accommodating fuel assemblies with initial enrichments of 4.3 weight percent U-235 and a minimum burnup of 42,000 MWD /MTU. Fig. 2.1 shows the arrangement of the rack modules in,the V.C. Summer pool in the 3 regions described above. The high density racks are engineered to achieve the dual objective of maximum protection against structural loadings (arising from ground motion, taermal

stresses, etc.)

and the maximization of available storage locations. In general, a greater width to height aspect ratio provides greater margin against rigid body tipping. Hence the modules are made as 'large as possible within the constraints of transportation and site handling capabilities. As shown in Fig. 2.1, there are 11 discrete modules arranged in the fuel pool at 1-7/8" minimum inter-module gap. Table 2.1 gives the relevant design data on each region. The modules in the three regions are of 4 different types. Table 2.2 summarizes the physical data for each module type. O 2-1

The modules a,re not anchored to the pool floor, to each other, or to the pool walls. A minimum gap of 1-7/8" is provided between the modules to ensure that kinematic movements of the modules during the Plant Design Basis Earthquake will not cause inter-module impact. Adequate clearance with other pool hardware, eg. light fixtures, etc. is also provided. c O e e O O b 4 O 2-2

l 1

O Table 2.1 Design Data 10 Cell' Pitch Mic.-B Flux Trap Region (nominal)

Loading Gap (nominal) 4 2 1 10.4025" .022 gm/cm 1.1605 2 2 10.4025" x 10.1675" .0015 gm/cm 1.2605 x 1.0455 i 3 10.116" unpoisoned 1.086 e O I i w 9 2-3

4 Table 2.2 Module Data Approximate Region Module Celisiper Array Weight No. Type Quantity module Size (lb/ module) 1 A 2 121 llxil 36300 2 B 1 99 llx9 28300 3 C 5 121 lixll 25500 3 D 3 110 .llx10 23100 J l i ~' O

O O O u-if ic 120 ~ --{co }W w =W ^ 'm g m ee in 4 R h c t a d % w =< REGION 3 REGION 1 e REGION 3 fl RESERVED h O ll X 11 11 X 11 Z ll X 11 AREA ^ 5 g / o a a a ~4 r<> lw j 'd-REGION 3 REGIO N I e REGION 3 REGIO N 3 o --g 9 ll X ll ll X ll Z ll X ll II X 11 = ,1 l ro co i 5 -he _W i =g, y g n a Inte _8 R E G I O N '3 R EGI O N 2 :g REGION 3 RE6 ION 3 O 61XlO II X 9 11 X 10 11 X 10 i O O =c } p n n l h 1

• n$

i im t llOk 11 4 [

• llOk llOk 2 7 "R E F.

a 27 77 = =,, -7 = = = 3 32 7,, 7 g /7 ' 1/8 1'8 8 39-0 = = FI G. 2 1 MODULE LAYOUT

3. RACK CONSTRUCTION pV 3.1 Fabrication Details: 3.1.1 Regions 1 and 2: The rack module is fabricated from ASTM 240-304 austenitic stainless steel sheet and plate material, and ASTM A182-Type F304 forging material. The weld filler material utilized in body welds is ASME SFA-5.9 Type 308. Boraflex, a patented brand name product of BISCO

• serves as the neutron absorber material. The detailed radiological properties of Boraflex may be found in Section 4.

The experience list of Boraflex is given in Table 3.1. A typical module contains storage cells which have an 8.,85" nominal cross sectional opening. This dimension ensures that fuel assemblies with maximum expected axial bow can be inserted and removed from the storage cells without any damage to the fuel ) assemblies or the Tack nodules. Figs. 3.1A and 3.lB show a horizontal cross section of a 3x3 array. The cells provide a smooth and continuous surface for lateral contact with the fuel assembly. The anatomy of the rack modules is best exposed by describing the basic building blocks of the design, namely (a) Internal square tube (b) Neutron Absorber Plate (Boraflex) envelope angular elements (c) Angular structural element (d) Base plate (e) Support assembly l (f) Top lead-in i i

BISCO, a Division of Brand, Inc.,

1420 Renaissance Drive, Park Ridge, Illinois J-3-1 1

Table 3.1 -s .) BORAFLEX EXPERIENCE FOR HIGH DENSITY RACKS Plant NRC Licensing ^ Site Type Docket # Status Point Beach -l & 2 PWR 50-226 & 301 Issued Nine Mile Point - 1 BWR 50-220 Issued Oconee 1 & 2 PWR 50-269 & 270 Issued Prairie Island 1 & 2 PWR 50-282 & 306 Issued Calvert Cliffs - 2 PWR 50-318 Issued

• Quad Cities - 1 & 2 BWR 50-254 & 265 Issued Midland - 1 & 2 PWR 50-329 & 330 Pending Watts Bar 1 & 2 PWR 50-390 & 391 Pending Waterford - 3 PWR 50-382 Pending

( )

• Fermi - 2 BWR 50-341 Issued H.B.

Robinson - 2 PWR 50-261 Pending River Bend - 1 BWR 50-458 Pending

• Rancho Seco - 1 PWR 50-312 Pending Nine Mile Point - 2 BWR 50-410 To be applied for Shearon Harris - 1 PWR 50-400 To be applied for Millstone - 3 PWR 50-423 To be applied for
• Grand Gulf

-1 BWR 50-416 Pending 4

• Oyster Creek' BWR 50-219 Pending
• Joseph Oat Corporation fabricated racks.

O 3-2 e v~ --r w w m--

• mm e

-ew-- m--rw--+----

---

+---------w

-.-ir w---- e- -we------ + - - - - - - - - - - - - - - -.

(a) Internal square tube: _b V This element provides: the lateral bearing surf ace to the fuel asssmbly. 'It is fabricated by joining two formed channels usiing a controlled seam welding operation. The weld : penetration in the seam welded zone is required to be 90% minimum. This element is 8.85" square (nominal.) cross section x 169" long. (b) Neutron Absorber Plaese. (Boraflex) envelope angular elements: Boraflex surrounds thre square tube on all four sides over a length of 138" which completely envelopes the active fuel length except the top 3 and bottom 3 inches. (c) Angular structural elcments: Two angular subelements, illustrated in Fig. 3.2 (a) and (b) comprise the structural support gridwork for the fuel racks. One set of large and small angular subelements is placed around the square tube with the Neutron Absorber interposed in-between, as shown in the cross section in Fig. 3.3. The fillet welds indicated in Fig. 3.3 are made while the angular subelements exert a contact pressure on the neutron absorber sheets in the welding

fixture, thereby ensuring a continuous surface contact in a macroscopic sense, between the constituent elements of the sandwick.

As shown in Fig. 3.46, bottom spacer sheets (also made from ASTM 240-304 material) 3-3 p) \\.

position the Boraflex sheets in the vertical r% direction. The top of the angular sub-elements la \\v) \\ welded to the square tube using a suitable spacer, j In this ma'nner a composite box assembly is j fabricated. An array of composite box assemblies welded as indicated in Figs. 3.lA and 3.lB form the " honey-comb" gridwork of cells which harnesses the structural strength of all sheet and plate type members in an efficient manner. Figure 3.4.A illustrates a typical welding sequence to obtain a honey comb construction. The array of composite boxes has overall bending, torsional and axial rigidities which are an order of magnitude greater than configurations utilizing grid bar type of construction. (d) Base Plate: The base plate is a 5/8" thick plate type member [ which has 5" diameter holes concentrically located with respect to the internal square tube. These holes provide the primary path for coolant flow. Secondary flow paths are available between adjacent cells via the lateral flow holes (1.0" diameter) near the root of the " honey-comb" (Figure 3.4B). The honey comb is welded to the base plate with 1/8" fillet welds. (e) Support Assembly: Each module has 4 support legs. One support leg is of fixed height,(Fig. 3.6) the other three are adjustable in length to enable leveling of the rack. The variable height support assembly l consists of a flat-footed spindle which rides into an internally threaded cylindrical member. The cylindrical member is attached to the underside of the base l l 3-4 i

pl'te thr:ough double fillet and partial penetration a weld. Thee base of the flat-footed spindle sits on the pool floor. Leveling of the rack modules is accomplished by turning the hex sprocket in the spindle using a long arm (approximately 16' long) hex head wrench. Fig 3.5 shows a vertical cross section of the adjustable support assembly. The supports elevate the module base plate approxiahtely 4 '.if4" above pool

floor, thus creating the water plenum for coolant flow.

The lateral holes in the cylindrical member provide the coolant entry path leading into the bottom of the storage locations. (f) Top Lead-In: Contiguous walls of adjacent cells are connected by a suitably designed lead-in for fuel assembly insertion. These lead-in joints also aid in reducing the lateral deflection of the inner square tube due to the impact of fuel assemblies during the ground motion (postulated seismic motion specified in the PSAR). This construction procedure leads to natural venting locations for the inter-cell space where the neutron absorber material is located. The fabrication of the rack modules is performed under a strict quality assurance system suitable for ASME Section III, Class 1, 2 and 3 manufacturing which has been in place at Joseph Oat Corporation for over 10 years. 3-5 y

1 3.1.2 Region 3: (a The rack modules in, Region 3 are fabricated from the same material as that used for Regions 1 and 2

modules, i.e.

-ASTM-240-304 austenitic stainless steel and ASTM A182 Type F-304 forging material. No neutron absorber material is used. A typical module also contains storage cells which have an 8.85" nominal cross-sectional opening. Fig. 3.7 shows a horizontal cross-section of a 3x3 array. The rack construction ' varies from that for Regions 1 and 2 in as much as the internal square tube and the neutron absorber plate are eliminated. Hence the basic components of this design ~is as follows: (a) Angular structural element (b) Baseplate (c) Support Assembly (d) Top Lead-in V In this construction, two angular structural elements form the cell of an 8.85" nominal cross-sectional opening in addition to functioning as part of the structural support gridwork as illustrated in Fig. 3.7. The fillet welds for these unpoisoned cells are also shown in Fig. 3.7. The baseplate and support assemblies are exactly the same as those described for Regions 1 and 2. Contiguous walls of adjacent cells are also connected by a suitably designed lead-in for fuel assembly insertion. i O 3-6 1

3.2 CODES, STANDARDS, AND PRACTICES FOR THE SPENT FUEL POOL /~N MODIFICATION \\~)3 The following are the public domain codes, standards, and practices to which the fuel storage racks are

designed, constructed and assembled, and/or pool structure analyzed.

Additional problem-specific references related to detailed analyses are given at the end of each section and at the beginning of the section for Section 4. I. Design Codes (a) AISC Manual of Steel Construction, 8th edition (1980) (b) ANSI N210-1976 Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations. (O V (c) American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code, Section III, 1980 Edition up to and including Winter 1982 addenda. (Subsection NF) (d) ASNT-TC-1A June,

1980, American Society for Nondestructive Testing (Recommended Practice for Personnel Qualifications)

II. Material Codes (a) American Society for Testing and Materials (ASTM). Standards - A240. (b) American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code, Section ) II, Parts A and C, 1980 Edition up to and including Summer 1983 addenda. 3-7

i III. Welding Codes V (a) ASME Doiler and Pressure Vessel Code, Section IX-Welding and Brazing Qualific'ations, 1980 Edition up to and including Summer 1983 addenda. IV. Quality Assurance, Cleanliness Packaging, Shipping, \\ Receiving, Storege, and Handling Requireiients 'N ' (a) ANSI 45.2.2, Packaging,

Shipping, Receiving =,

Storage and Handling of Items for Nuclear Power Plants. i (b) ANSI 45.2.1, Cleaning-of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants. l (c) ASME Boiler and Pressure Vessel, Section V, Non-destructive Examination, 1980

Edition, including Summer 1983.

-(d) ANSI-N16.1-75 Nuclear Criticality Safety Operations with fissionable materials outside s*~ reactors. D,' ; (e) ANSI-N16.9-75 Validat-lon of Calculation Methods .K for Nuclear Criticality Safety. [ N (f) ANSI-N45.2.ll-1974 t Quality Assurance Requirements 4for the ' Design of Nu$ lear Power L Plants. ~ V. Other References 4 (a) NRC Regulatory Guides, division 1, regulatory guides 1.13, 1.29, 1.31, 1.61, 1.71, 1.85, 1.92, and 1.124 (revisions as applicable). 3-8

(b) General Design Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 50, Appendix A (GDC Nos. 1, 2, 61, 62, and 63). (c) NUREG-0800, Standard Review Plan (1981). (d) "NRC 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. S O I O 3-9 i

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== m i' d ,I,, k l l l / / / / Y S / / / / s i / l / / / / p n / / / / d I d d J f f f J R k 3, nr z s f n I ~f f I f I w L Q 1' g ]P 1 k k {- L 3 K I / f I K I J f E ,f g .r / / / y3 _z 7 i i / / x ',s i i / / / / + t + s / / / / l / j / l e m ,4 / / t / l f / / //////I n L L' / / / s // I k V /. / / / / /F1 i v i / / s / / l / / / / ) i + + / / 1 l / s \\ / l V( l

[

p / / (. I; /,. / f.. l l i _1. POISON INNER BOX OUTER BOX MATERIAL I t l FIG. 3.1 A 3X3 TYP ARRAY (POlSONED CELLS) REGION l O 3 -10

- 10.4025" ~ T Y P. ( NO Ma .-,) T, .Ti Fm / / l / I O s h 4 + + / / / s s / p i e L r 1 .,.1,^e e t k .\\ r 1 1 L L ' P 4.... .,g ,;p ,, g / / / l / / ,/ / l / / / + 4-r L / / i o -e s m z P' v' s' ' ' ' - o 4 n m m D ' '} }' _[' o b n f / t + + e / / I Q /.-ss 7 7 y' 1 = l 7 i POISON INNER BOX OUTER BOX MATERIAL l l FIG 3.18 3X3 TYP ARR AY ( POISONED CELLS ) REGION 2 i O 3 -11 i I

n STAINLESS STEEL ASTM 240-304 (.0 65 TH K ) O U s s N A* N s A s h s 90* .150"R

7

) y s(s s s s s s s s s s s FIGURE 3 2(a) SMALL ANGULAR SUBELEMENT O m B* 2 c i l s s s s. B s sh 90* s s .150"R s s s s x s s s s s s

• DIMENSIONS "A" AND "B"

DEPEND ON REGI,0N FIGURE 3 2 (b) LARGE ANGULAR SUBELEMENT 3-12

O LARGE ANGULAR SUBELEMENT I 3 l

d {' { ( x's' l'!'$'

7 [7 ' /l' NEUTRON l] h A BSORBER.- PLATE j; ^ SMALL ANGULAR O J / s "NA\\ \\ x x x x x x x x x x d3 L V / / / / // // / / / /1 L O / / \\ FIGURE 3. 3 BASIC CELL ELEMENT 0 3 - 13

('M ,- m v) 42 44 46 48 ysxxxsy neu pxxxxxgegne//,u,uxxwq s s s 4 s x u v W X s Y / x sf40 ' /./41 x 43 / / 45 s 39 (c e - - -- cx us,f f e--a xxxxspfuun en -g _sx x x u s% N // ra/777,7 x m x sy 32/ c u - u /e 47 \\ / 34 /2 Q 36/ '3 38 x ~j D P ,3 Q t / 3 R 3 x S T 3 s S / s / s s ,p,g ~ Q's O 31 / / 33 N 3 s u / f ,' m x v eo/ uni)qiw m;fu, m /qfa/35 29 pu x wj xm _gyx u xy(l 2/ / u_g/ s u x s.33;9 37 / 22'\\ x 24'- 26 \\ 28 \\ x s M 1 ;N f f g K q L / N N 5 ; O- ; u Zy 20 2. /m N / ff21 x N 2 f3 f 5,25 s A \\g mu/7 x l / n wja//n px xs, x/un/ 8 x N' q"'Q [',"'}yff"(s ;-),#h[/'z'"'"'7, N zt ,g s' F $ j 'G b $ H j j l J j glh~v{)bo7r $}Nsx{h'i,nu$,M'*ay$ mmvn' _ y n n u spis,' s m s wo n u,o u m p 5' s 6'> / 7 N' s s 8'I 3 3 x x x N / s' / 3 A 5 t B 3 j C 1 R D j 3 E x f s x s .x 4 h////////8N NNNNCd s x / x

1. /777/77/MNNx Nxxx(1[/ //fu//I I

2-3 4 FI G. 3.4 A WELDING SEQUENCE OF i CO M PO SIIE BOXES I l l

f Q y - 3",_,,,------19 / / / l i / p / / / / / u / .'f /. / h ,1 1, 1 1 1 n _w uzz (v,,,,,/1,: FUEL ASSEMBLY l I p /[ R E F. OUTER BOX Ib OP SPACER U l l 1. c INNER BOX 4 i l d l ? s 16 9 4 ,l 4 E i ll BOTTOM SPACER j i l l BASE PLATE d I FLOW HOLES s L-_ -:p fi -d:. :j: ?, b i V 3 \\ O- \\ DIA. HOLE 5 FIG 3.4 8 TYPICA.L CELL ELEVATION i 3-15 1 O

l ' O tt CELL 7' l" Ii 5/ d 4 l g/$ % HO LE S BASE i = S* O l! (4 PLATE 7 I I y I I I d o7/8 Td f l34 .-g f x ,,/ 3 4/ h f 5" 6/4 I i p / i 1/4

NOM, O

2- / lr i ^ / l1 9/S v v a a g6 4/% 2 a 4,,d-4 TH R,D S PER INCH I* 72% / (UNC) z CLASS l A FIG. 3 5 ADJUSTABLE SUPPORT e O O l 3 - 16

O tT T. '.) BASE I o l l, ,5d PLATE i / ~ ) ? i ,y k 1 l_ ' ' I!D l0" U i 6 /k I h-- SCH160 I PIPE - O l l 2 l k' , r I" l ' u u ~ 12 S G, = FIG.3.6 FlX ED SUPPORT 6 O 3-17

i e 4 e e t O i o 6" c' TYP. I t 1f 7I g 1 7 1 S 1 ~ g a s. I t -L g k k i i i 8 e 8 e e 8 I I m _m g k Q i i h L -L I e O FIG.3.7 3X3 TYP ARR AY (UNPOISONED CELLS) REGION 3 O 3 -18 ~

4. NUCLEAR CRITICALITY ANALYSIS 4.1 Design Bases The high density spent fuel storage racks for the V. C. k,gg equal to or Summer Station are designed to assure that a less than 0.95 is maintained with the racks fully loaded with fuel of the highest anticipated reactivity in each of three regions and flooded with unborated water at a temperature corre-sponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity cal-culations and in mechanical tolerances, statistically combined, such that the true k gg will be equal to or less than 0.95 with a e 95% probability at a 95% confidence level. l Applicable codes, standards and regulations, or pertinent sections thereof, include the following. O Prevention of Criticality l e General Design Criterion 62 ( in Fuel Storage and Handling. e USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage. e USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis (proposed), December 1981. I e USNRC letter of April 14,

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

e USNRC Regulatory Guide 3.41, validation of Calculational l Method for Nuclear Criticality Safety (and related ANSI N16.9-1975). e ANSI /ANS-57.2-1983, Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants. O I 4-1 l

e ANSI N210-1976, Design Objectives for Light Water Reactor hw Spent Fuel Storage Facilities at Nuclear Power Plants, i e ANSI N18.2-1973, Nucl' ear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants. To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made. e Moderator is pure, unborated water at a temperature cor-responding to the highest reactivity. e Lattice of' storage racks is assumed infinite in all di-rections; i.e., no credit is taken for axial or radial neutron leakage (except in the assessment of certain abnormal / accident conditions where leakage is inherent). e Neutron absorption in minor structural members is neglected; i.e., spacer grids are replaced by water. The design basis fuel assembly is a 17 x 17 array of fuel rods (Westinghouse design) containing 002 at a maximum initial enrichment of 4.3% U-235 by weight, corresponding to 54.30 grams U-235 per axial centimeter of fuel assembly. Three independent regions are provided in the spent fuel storage

pool, with separate criteria defining the highest anticipated reactivity in each of the three regions as follows.

e Region 1 is designed to accommodate new unirradiated fuel with a maximum enrichment of 4.3 wt.% U-235, or spent fuel regardless of the discharge fuel burnup. e Region 2 is designed to accommodate spent fuel of 4.3 wt.% U-235 initial enrichment, which has accumulated a - minimum burnup of 20,000 Mwd /mtU. Region 2 will also safely accept fuel of lower discharge fuel burnup pro-vided the initial enrichment is correspondingly lower. e Region 3 is designed to accommodate fuel of 4.3 wt.% U-235 initial enrichment which has accumulated a minimum burnup of 42,000 Mwd /mtU. Region 3 will also safely accept fuel of lower discharge fuel burnup provided the initial enrichment is correspondingly lower. 4-2 I

l Summary of Criticality Analyses O 4.2 4.2.1 Normal Operating Conditions The criticality analyses of each of the three regions of the spent fuel storage pool are summarized in Table 4.1 for the an-ticipated ncrmal storage conditions. The calculated maximum reactivity in Regions 2 and 3 is ~0.94 including all uncer-tainties, which provides an additional margin of 1% Ak below the limiting value of 0.95. Both long-term storage and axial burnup distribution were investigated and found to result in negative reactivity effects. Regions 2 and 3 can accommodate fuel of lower discharge fuel burnup provided the initial enrichment is correspondingly I lower. Figures 4.1 and 4.2 illustrate the minimum acceptable burnup as a function of the initial fuel enrichment, which yields the maximum reactivity given in Table 4.1 for Regions 2 and 3, respectively. These curves are intended to be incorporated in the Technical Specifications supplemented with appropriate administrative procedures to assure verified burnup as specified in draf t Regulatory Guide 1.13, Revision 2. 4.2.2 Abnormal and Accident Conditions Although credit for soluble poison normally present in the pool water is permitted under abnormal / accident conditions, most abnormal or accident conditions will not result in exceeding the limiting reactivity of 0.95 even in the absence of soluble poi-son. The single accident condition that could potentially exceed the limiting reactivity is the inadvertent loading of a new fuel assembly (4.3% enrichment) into Region 2 or Region

3. storage cells, with the simultaneous occurrence of a loss of soluble poi-son.

Administrative procedures will be necessary to preclude the 4-3

Table 4.1 Summary of Criticality Safety Analyses Region 1 Region 2 Region 3 Minimum'burnup with 4.3% 0 20 Mwd /kgU 42 Mwd /kgU initial fuel enrichment Temperature assumed 40*F 40*F 150*F for analysis Rederence k, 0.9357 0.9108 0.9203 (AMPX-KENO) ~ Calculational bias 0 0 0 Uncertainties Bias

• 0.0030
• 0.0030
• 0.0030 Calculational
• 0.0065

+0.0058

• 0.0035 B-10 concentration 70.0017 70.0019 NA Boraflex thickness 70.0032 70.0121 NA Boraflex width T0.0006 70.0006 NA

() Inner box dimension

• 0.0008 70.0013 70.0033 Flux-trap water gap 70.0059 T0.0071 70.0031 SS tolerance
• 0.0006
• 0.0006 70.0040 Fuel enrichment
• 0.0020
• 0.0020
• 0.0020 Fuel density
• 0.0023
• 0.0023

+0.0023 Statistical

• 0.0105
• 0.0160
• 0.0072 combination Total 0.9357 0.9108 0.9203
• 0.0105
• 0.0160
• 0.0072 Extra allowance for NA 0.01 0.01 burnup uncertainty Maximum reactivity 0.946 0.938 0.938 O

4-4

I possibility of simultaneous occurrence of these two independent accident conditions.- i Effects on reactivity of other abnormal and accident condi-tions evaluated are summarized in Table 4.2. ' Table 4.2 Reactivity Ef fects of Abnormal and Accident Conditions Accident / Abnormal Conditions Reactivity Effect Temperature increase Negative in Regions 1 and 2; positive in Region 3 void (boiling @ 248'F) Negative in all regions Eccentric positioning of assembly Negative f Assembly outside rack Negligible Assembly lying on top of rack Negligible Lateral rack module movement Negligible The single positive reactivity effect results from an increase in temperature above the nominal maximum pool l temperature of 150*F assumed for criticality evaluation of the Region 3 storage rack. Temperatures above 150*F are considered accident conditions, in which case the soluble poison would f maintain a low reactivity. Nevertheless, in the absence of l soluble poison, the increase in reactivity is calculated to be only 0.01 Ak at 248'F (approximate boiling temperature of the bulk coolant at the submerged depth of the fuel racks).

Thus, even. in the absence of soluble poison, the maximum. reactivity does not exceed the limiting value of 0.95.

At 248'F, voids resulting from boiling have a negative reactivity effect. !O l 4-5 i

g i'

i,k. j l 4 li ii!l:2' jll i III .lii, 77 ,f j {J1' l l l l u } i I i /, Q L i u l j .r No l, l l i I l U u _ a ,f i g f K ) [ t _s, J gf D i W i eg>O l 4 M i I P U N I) R i l 4 Q U 1 1 B E _ o i G R A H C 1 S i I ) D YL B M E S S A m l'1[' l l f l, l l l .l. l )% [; !_ { l 1lI[l! l I O il- 'I,lf hS u N .b Chau Z_4sr,z=OIEm2j

• ~

W0 h mY'OD a " .eP'sP'tVU'4 ([ g ] mdmH yctUCDrP@P mDNOhOU r( s L 1 r tn D 'PD* c Z P D

t l ~, l O ',,. ~ 40 _.. _. 35- _f 3 A C C F A.T Afsi-F lm: O 30-R 2 vmN-acc E== = stE ~ 1. .i 3 I Z a: 25-3 83 W / o 20- -o 2 O >= .a O 15-2 W m m< ^ l 10-5- iz= a g-O 0-i i kJ d 3 4 l INITI AL ENRI C H MENT, wt. % U-235 . Fig. 4.2 Limiting ~ discharge fuel burnup for Region 3 storage rack for fuel of various initial enrichments.

Reference Fuel Storage Cell .O 4.3 4.3.1 Reference Fuel Assembly The reference design basis fuel assembly, illustrated in Fig. 4.3, is a 17 x 17 array of fuel rods with 21 rods replaced by 20 control rod guide tubes and one instrument thimble. Table 4.3 summarizes the fuel. assembly design specifications and the expected range of significant parameters. Table 4.3 Fuel Assembly Design Specifications Fuel Rod Data Outside dimension, in. 0.374' Cladding thickness, in. 0.0225 ~ Cladding material Zr-4 b01 D sh ng f or 002 density, % T.D. 95

• 2 3

002 stack density, g/cm 10.296

• 0.2172 Enrichment, wt. % U-235 4.3
• 0.05 Fuel Assembly Data Number of fuel rods 264 (17 x 17 array)

Fuel rod pitch, in. 0.496 Control rod guide tube Number 20 0.D., in. 0.482 Thickness, in. 0.016 Material Zr-4 Instrument thimble Number 1 0.D., in. 0.482 ~ Thickness, in. 0.016 Material Zr-4 U-235 loading O, g/ axial em of assembly 54.30

• 1.78 4-8 I

4.3.2 Region 1 Storage Rack The nominal spent fuel storage cell used for the criticality analyses of Region 1 storage cells is shown in Fig. 4.3. The rack is composed of Boraflex absorber material sandwiched between a 0.049-in. inner stainless-steel box and a 0.065-in. outer stainless-steel box. The fuel assemblies are centrally located j in each storage cell on a nominal lattice spacing of 10.4025

• I 0.0625 in.

Stainless-steel tabs connect one storage cell box to another in a rigid structure and define an outer water space between boxes. This outer water space constitutes a flux-tra~p between the two Boraflex absorber plates that are essentially opaque (black) to thermal neutrons. The Boraflex absorber has a thickness of 0.082

• 0.007 in, and a nominal B-10 areal density 2

7 of 0.0265 grams per cm, 4.3.3 Region 2 Storage Cells O ^ Figure 4.4 illustrates the storage cell design used for the . Region 2 storage cells. In Region 2, the rectangular storage cells are located on a lattice spacing of 10.4025

• 0.0625 in. in one direction and 10.1875
• 0.0625 ' in. in the other direction.

Boraflex absorber material of 0.032-in, thickness, sandwiched between 0.049-in. and 0.065-in. stainless-steel plates, consti-tutes the walls of each storage cell. Th; outer water-gap flux-trap is 1.0455

• 0.0625 in. in one dimension and 1.2605
• 0.0625 in. in the other direction, as indicated in Fig.

4.4. For fuel of 4.3 wt.t U-235 initial enrichment burned to 20 Mwd /kgU, the nominal design B-10 areal density is 0.002 g/cm2, 4.3.4 Region 3 Storage Cells Region 3 storage cells, designed for fuel of 4.3 wt.% U-235 initial enrichment burned to 42 Mwd /kgU, is unpoisoned, other l than the 0.090-in. thick stainless-steel plates forming the walls 4-9

of the storage cell. These cells, shown in Fig. 4.5, are located on a lattice spacing of 10.116

• 0.032 in, defining a 1.086
• 0.032-in water gap between the steel walls.

4 e S O O 4-10

,f Iv) ?

8. 4 50 t.0 6 2 5 IBORAFLEX O.082

.007 THlCK O.02'56 g/ cm 2 f k ^ wmsmmmm3m++mmwxwe++mpyy; s m s mm mmmmmms s m mm mms m f,g,,,, s 00000000000000000 h 00000000000000000 ! s OUTER SS

R 00000000900800000 q

g O0000000000000000 t t .os51.co4 Rn $h- 00000000000000000 S' h e 00000000000000000 t s INNER SS s t[& 00000000000000000 4 t 'o492.005 10 4o25 stR 00000000000000000 m.os25 (PITCH) N 00000000000000000 t . ? tl% 00000000000000000 t $ WATER s O0000000000000000 !! q' GAP 5; Rf% 00000000000000000 t . 'i 1.1so5 i a t;~1 00000000000000000 t 8 ( 4 00000000000000000 t - t t.osas O s!t o0000*O000000000o E -s t t O0000000000000000 t t 00000000000000000 3 g }& t ti t aggewusmu umu smwumu s s u ggp h m xm x m m -. e . ss m s, l d 8 850 t. 032 1 (IN SIDE B OX) l c 9.242 7 1 10.4 025 t.0625 7 (PITCH ) (NOT TO SCALE) (V Fig. 4.3 Configuration of Region i spent fuel storage cell. 4-11

(V~) 8.4501.0625 ? BOR AFLEX-0.0322.007 THICK 0.00 20 g/cm 2 l l _g;;3 m. m u x u x s ususwwuu uvuuswwwww<wvt ( p$ hg,0.146 s t~: 00000000000000000 ' i t 00000000000000000 3 3 00000000900000000 f UTER SS 3 t O000000000000000 .o651.004 ,-t O000000000000000 !t v R - t 00900900000000000 FD E 00000000000000000 t!$ $ERS l 10.4025 9

• OO3
• .o625 t

. t O0000000000000000 bj t (PITC H) d ' t O0000900900000000 (; ; E . ? 00000000000000000 h RWATER i t ;t O0000000000000000 D R GAP S 't O0000000000000000 J t .l O0000000000000000 'I.o 4 5 5 Et.06 25 - t OOO 0000000000000 x 'l - t OOO 0000000000000 t s ' S 00000000000000000 l 00000000000000000 s { s s s Lux ms uu su sumuuu u xuuum u x u 5 2 s RM imemesm<ewswamwwwm mw+J W e f I' 8.8 50 t. 032 g v i d I (INSIDE B OX ) WATER GAP l 1.2605 y 9.142 M2.0625 ---u----- c 10.1875 2.0625 ?' (PITCH) ( N OT TO SC A LE). ,O Fig. 4.4 Configuration of Region 2 spent fuel storage cell. 4-12

O, ~ ~ g n s mummmmmummmmmmmmmmw 00000000000000000 00000000000000000 s O0000000000000000 00000000000000000 t S 00000000000000000 t S 00000000900900000 E 00000000000000000 $ TAP" t O0000000000000000 t < its lo'o32 00000000000000000 t i.oss 00000000000000000 t ^' s$ 00000000000000000 e 'o32 (PlTCH) s 00000000000000000 00000000000000000 S 000o000000000eO00 t Q' 00000000000000000 E OOOgOOgOOOOOOOOOO $ ss Box t I j OOO OO 0000000000 ,gggg,gg3 lmmmmmmmmmmmmmmmuummmh l S 8.850 2.032 (INSIDE BOX) v 9.'030 l0.ll6 2.032 7 ( PIT C H ) (NOT TO SCALE) Fig. 4.5 Configuration of Region 3 spent fuel storage cell. _ _ _ _ _ _ _ _ _ _ - _4 13

1 4.4 Analytical Methodology 4.4.1 Reference Analytical Method and Bias The reference nuclear criticality analyses of the high den-1 2 sity spent fuel storage rack were performed with the-AMPX -KENO computer package, using the 123-group GAM-THERMOS cross-section set and the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treatment). AMPX-KENO has been ex-tensively benchmarked against a number of critical experiments (e.g., Refs. 3, 4, 5, and 6), including those,6 most representa-4 tive of spent fuel storage racks. In the geometric model used in KENO, each fuel rod and cladding were described explicitly. For two-dimensional X-Y' analysis, a zero current (white albedo) boundary condition was applied in the axial direction and at the centerline through the outer water space (flux-trap) on all four sides of the cell, effectively creating an infinite array of storage cells. 1 6 on a series of criti-Results of the benchmark calculations cal experiments indicate a calculational bias of 0, with an uncertainty of

• 0.003 (954 probability at a

95% confidence level). In addition, a small correction in the calculational bias might be necessary to account for the internal water-gap l thickness (0.418 in.) between rack walls and the fuel assemblies in the V. C. Summer spent fuel rack compared to the corresponding thickness (0.644 in.) in the benchmark critical experiments. Based upon the correlation developed in Ref. 6, the correction for water-gap thickness in the V. C. Summer spent fuel storage rack indicates a small overp,rediction of ~0.002 Ak. For conser-vatism, the overprediction is neglected and the net calculational bias is taken as 0.000

• 0.003, including the effect of the water-gap thickness.

4-14

4.4.2 Manufacturing Tolerances and Uncertainties O For investigation of small reactivity effects due to manu-7 was used to calcu-facturing tolerances, the CASMO computer code late small incremental reactivity changes that would otherwise be i lost in the normal statistical variation associated with Monte Carlo techniques (i.e., KENO). CASMO is a two-dimensional trans-port-theory code, based on capture probabilities, that allows an explicit description of each fuel pin in an assembly, as well as an approximate description of the storage cell geometry. Geomet-rical approximations necessary in CASMO include the following.- e The outer stainless-steel plates were homogenized with the water in the flux-trap. region. e The Boraflex absorber plate was necessarily described as.' 8.948 in wide, rather than the actual 8.45-in. width, o In Region 2, the rectangular cell geometry was represented as a square of equal area. Despite these approximations, reactivities calculated by CASMO were within 0.005 ak of the reference AMPX-KENO calculation for Region 1. In Region 2, the square approximation resulted in an underprediction of 0.01 Ak by CASMO. For Region 3, where the i approximations above were not needed and an exact geometrical representation was possible, the CASMO-calculated reactivity was within the statistical uncertainty of the corresponding AMPX-KENO calculation. \\ 4.4.3 Fuel Burnup Calculations Fuel burnup calculations were performed primarily by the CASMO code. However, to enhance the credibility of the burnup calculations in lieu of critical experiments, the CASMO results 8 were confirmed by independent calculations with the EPRI-CELL 9 codes. Figure 4.6 compares results of the three and NULIF 4-15 - -, _ _ _. _ _ _ _ _ _.. _. _ _ _. _ - ~ _ _ _ _..

independent methods of burnup analysis under reactor operating O conditions. Agreement between CASMO and EPRI-CELL is very good (0.002 Ak at 20 Mwd /kgU and ' 42 Mwd /kgU), although reactivities calculated by NULIF are somewhat lower, probably due to dif fer-ences in treatment of temperature and xenon ef fects. Also shown on Fig. 4.6, for additional information, are the burnup-dependent 10 calculations for a ' com-reactivities extracted from CHEETAH-P parable reactor system with essentially the same core nuclear properties. Of more importance for storage rack cr.iticality analyses is the comparison under conditions more representative of fuel to be stored in the racks (cold, xenon-free). Table 4.4 compares the cold, xenon-free reactivities calculated, at 68'F, by the three codes at the two reference fuel burnups. Table 4.4 Comparison of Cold, Clean Reactivities Calculated at 20 Mwd /kgU and 42 Mwd /kgu Burnup k, Xe-Free @ 68'F Calculational Method @ 20 Mwd /kgU @ 42 Mwd /kgU CASMO 1.2597 1.0743 EPRI-CELL

• 1.2610 1.0747*

NULIF (Later] [Later]

• At maximum reactivity during long-term decay (see Section 4.4.4).

[ Conclusions - later] To be provided at time of licensing amendment request. 4-16

4.4.4 Long-Term Decay Since the fuel racks (Regions 2 and 3) are intended to con-tain spent fuel for long periods of time, calculations were made using EPRI-CELL (which incorporates the CINDER code as a subroutine) to follow the long-term changes in reactivity of spent fuel. Figures 4.7, and 4.8 show these changes for fuel burned to 20 Mwd /kg and 42 Mwd /kg, respectively. Early in the decay period, xenon grows in and subsequently decays, with the reactivity (k,,) reaching a maximum at about 10 days after reactor shutdown. (This maximum value is listed in Table - 4.4 above. ) For' longer. storage periods, tihe decay of Pu-241 (13-year half-life) again reduces reactivity, as shown in Figs. 4.7 and 4.8. The design criticality safety calculations do not take credit for this long-term reduction in reactivity, although this effect - would afford an increasing subcriticality margin in Regions 2 and 3 of the spent fuel storage pool. For illustrative purposes, Table 4.5 lists the long-term reduction in reactivity below the maximum used for the criticality safety evaluation. Table 4.5 Long-Term Changes in Reactivity Ak from Maximum Reactivity Storage Time, years @ 20 Mwd /kg @ 42 Mwd /kg 0.5 -0.0016 -0.005 1.0 -0.0032 -0.011 -0.034 4.0 10.0 -0.0192 -0.060 -0.0283 -0.086 20.0. 30.0 -0.0335 -0.101 k O 4-17 i

7 c, L 4.4.5 Effect of Axial Burnup Distribution O., ,g-_ 5 Initially, fuel loaded into the reactor will burn with a ^' nearly ~ cosine power

• distribution.

As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. To investigate the reactivity effect of the axial burnup ' g ( ..fe distribution of spent fuel to be stored in the racks, the experi-mentally detarmined burnupil in fuel removed from the H. B. Robinson reactor (similar to the V. C. Summer Station).was assumed to be representative of the burnup distribution in high burnup fuel. ~ 1 1 + 'g(

• *P-,'

- 1, a w I i' ? r ' "1 nd ^ &r *,: L a.w a ') l 't s j {Leter] y l To be provided at time of licensing amendment request. 4-18 l

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• F RO M REACTOR Fig. 4.7 Long-term change in infinite inultiplication factor (kog ) of 4.3% enriched fuel burned to 20 Mwd /kgU.

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O O O OR NL OWG.7 7-60 A DIS TA N C E FROM BOTTOM OF ROD (f t ) O I 2 3 4 5 6 7 8 9 10 ll 12 13 Ju-cl l l l 1 1 I i i i l l i .i i i i i i i. i I l l .I I i i i i l 4 w l l 1 I I I l l I l l l- -+

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\\ 4.5 Reference Subcriticality and Mechanical Tolerance Variations pO 4.5.1 Nominal Design Cases Under normal conditions, with nominal dimensions, the calcu-lated k, for 4.3% enriched fuel in Region 1 is 0.9357

• 0.0034 (la with 100 generations of 500 neutrons each) for the nominal l2 of 1.924, corresponding case.

For a one-sided tolerance factor to 95% probability at a 95% confidence limit with 100 genera-tions, the maximum deviation of k, is *0.0065. The corresponding k,, from the CASMO calculation is 0.9337,

which, despite the necessary geometrical approximations in CASMO, provides addi-tional confidence in the validity of the reference AMPX-KENO calculation.

In Region 2, CASMO calculations, with fuel burned to 20 Mwd /kgU in the reference design storage cell at 40*F, yielded a k, of 0.9007. Iterative CASMO calculations with fresh fuel of varying enrichments resulted in an enrichment of 2.30 wt.%-U-235 yielding the same k, value. AMPX-KENO calculations were then made on fresh fuel of 2.30% enrichment (to compensate for the geometric approximations in CASMO), yielding a k, of 0.9108

• 0.0058 (95% probability at a 95%

confidence level). Subse-quently, iterative burnup and storage cell calculations were made l with CASMO for fuel of varying initial enrichments (3.6%, 3.0%, and 2.5%), in each case searching for the burnup which gave the same k, as the reference fuel at 20 Mwd /kgU. These converged burnup values are those shown in Fig. 4.1. At the design basis burnup (20 Mwd /kgU), the sensitivity to burnup is calculated to be 0.0066 Ak per Mwd /kgU. A similar procedure was used for the Region 3 storage cells at a reference temperature of 150*F, giving a k, from the CASMO f I calculation of 0.9254. Iterative CASMO calculations determined U an equivalent enrichment of 1.42% U-235 for this same value of 4-23 l

t reactivity. This enrichment, in an AMPX-KENO calculation, gave a O

k. of 0.e203

• 0.0035 <e5% groeab111er at a e5% confidence 1 eve 1) as the reference design reactivity for the Region 3 storage cells..

Iterative CASMO calculations, as described above (at initial enrichments of 3.6%, 3.0%, 2.5%, and 1.8% E) determined the burnup values presented in Fig. 4.2 for various initial en-richments. At the design basis burnup of 42 Mwd /kgU, the calcu-lated sensitivity to burnup is 0.0072 ak per Mwd /kgU. 4.5.2 Boron Loading Variation The Boraflex absorber plates used in Region 1 storage cells are nominally 0.082 in. thick, with a B-10 areal density of 0.0265 g/cm2 Independent manufacturing tolerance limits are 2 in B-10 content. This*

• 0.007 in in thickness and *0.00245 g/cm assures

that, at any point where the minimum boron loading 2

(0.02405 grams B-10/cm ) and minimum Boraflex thickness (0.075 in.) may coincide, the boron areal density will not be less than 2 0.022 grams B-10/cm. CASMO calculations indicate that these tolerance limits result in an incremental reactivity change (un-certainty) of *0.0017 Ak for boron conter.t and

• 0.0032 for Bora-flex thickness variations.

In Region 2, che Boraflex absorber plates are nominally 0.032 in. thick with a B-10 concentration corresponding to an areal density of 0.0020

• 4%.

For an independent thickness tol-erance of *0.007 in., the minimum B-10 areal density at any point where the minimum thickness (0.025 in.) and minimum concentration 2 (0.00192 g/cm ) might coincide, the areal density will not be less 'than 0.0015 grams B-10/cm2 CASMO calculations show an incremental reactivity effect (unce'rtainty) of *0.0019 ak for B-10 concentration and *0.0191 Ak for Boraflex thickness. Boraflex poison sheets are not used in the Region 3 storage O ce11 - 4-24 l

4 Storage Cell Lattice Pitch Variations O .5.3 The design storage cell lattice spacing between fuel assem-blies ranges from 10.1875 in Region 3 to 10.4025 in Region 1. An increase in storage cell -lattice spacing may or may not reduce that may be reactivity depending upon other dimensional changes associated with the increase in lattice - spacing. Decreasing lattice spacing by decreasing the outer (flux-trap) water thick-ness always in creases reactivity, although decreasing the inner water thickness (between the fuel and the inner stainless-steel box) may result in a small increase or decrease in reactivity. The reactivity ef fect of the outer (flux-trap) water thickness, however, is usually more significant. Both of these effects have been evaluated for the independent design tolerances in each of the three regions. 4.5.3.1 Inner Water Thickness Variations O The inner stainless-steel box dimension, 8.850

• 0.032 in.

for all three regions, defines the inner water thickness between the fuel and the inside box. For the stated tolerance limit, the calculated uncertainty in reactivity is *0.0008 ak in Region 1,

• 0.0013 ak in Region 2, and *0.0033 ak in Region 3.

In Region 1, as the inner stainless-steel box dimension (and de-k, increases rivative lattice spacing) increases, while in Regions 2 and 3, increasing the box dimension reduces reactivity. 4.5.3.2 Outer (Flux-Trap) Water Thickness Variation water thickness is In Region 1, the design outer (flux-trap) 1.1605 0.0625 in., which results in an uncertainty of

• 0.0059 ak due to the tolerance in flux-trap water thickness, assuming the water thickness is simultaneously reduced on all four sides.

Region 2 is rectangular with a flux-trap water thickness of 1.2605

• 0.0625 in.

in one direction and 1.0455

• 4-25

l 0.0625 in, in the other direction. Assuming the thickenss tolerance is simultaneously applied on all four

sides, the uncertainty in reactivity is **0.0071.

For Region 3, which uses only a-single steel box and no Boraflex, the tolerance in gap water thickness results in a reactivity uncertainty of

• 0.0031 ak.

Since the fabrication tolerance on each of the four sides is statistically independent, the actual reactivity uncertainties would be approximately one-fourth of the values shown, although the more conservative values have been used in the criticality evaluation. '4. 5. 4 Stainless-Steel Thickness variations The nominal stainless-steel thickness is 0.049 in, for the inner box and 0.065 in, for the outer box. The maximum positive reactivity effect of the expected stainless-steel thickness tolerance variation (i0.004 in.) was calculated by CASMO to be

• 0.0006 ak.

In Region 3, the *0.005-in. tolerance on the single stainless-steel box wall has a larger effect and is calculated to result in a *0.0040 ak uncertainty. 4.5.5 Fuel Enrichment and Density Variation The design maximum enrichment is 4.30

• 0.05 wt.% U-235.

Calculations of the sensitivity to small enrichment variations by CASMO yielded a coefficient of 0.0041 Ak per 0.1 wt.% U-235 in Region 1 at the design enrichment. For the tolerance on U-235 enrichment of

• 0.05 in wt.%,

the uncertainty on k, is

• 0.0020 Ak.

The same value has been assumed for Regions 2 and 3. Calcuiations were made with the UO2 fuel density ranging from the minimum of 93% theoretical density to a maximum value of 97% theoretical density. For the mid-range value (95% T.D.) used for the reference design calculations, the uncertainty in re-densities activity is *0.0023 a k over the maximum range of UO2 expected. This uncertainty is assumed to apply in all three -regions. 4-26

4.5.6 Boraflex Width Tolerance Variation The reference storage ceil design for Regions 1 and 2 (Figs. 4.1 and 4.2) uses a Boraflex blade width of 8.45

• 0.0625 in.

A positive increment in reactivity occurs for a decrease in Bora-flex absorber width. For the width tolerance of -0.0625 in., the re'ctivity increment is +0.0006 Ak. In-maximum calculated a creasing the Boraflex width decreases reactivity. b O e O 4-27

l 4.6 Abnormal and Accident Conditions 4.6.1 Eccentric Positioning'of Fuel Assembly in Storage Rack The fuel assembly is normally located in the center of the storage rack cell with bottom fittings and spacers that mechan-ically limit lateral movement of the fuel assemblies. Neverthe-less, calculations were made with the fuel assemblies moved into the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations resulted in a negative 13 with diffusion coefficients reactivity effect using PDQ$7 generated by CASMO. Fuel assembly bowing will produc'e a small negative reactivity effect locally. Thus, the nominal case, with ~ the fuel assemblies positioned in the center of the storage rack cell, yields the maximum reactivity. 4.6.2 Temperature and Water Density Effects The temperature coefficients of reactivity in Regions 1 and 2 are negative, and a temperature of 40*F, corresponding to a water density of 1.0, was assumed for the reference design. In Region 3, however, the temperature coefficient of reactivity is positive in the temperature range to which the racks are normally exposed. For this reason, a design basis temperature of 150*F was assumed for the criticality evaluation. Temperatures above 150*F are considered accident conditions, and credit for the soluble poison actually present would maintain a low reactivity. Temperature effects or reactivity have been calculated and the results are shown in Table 4.6. Introducing voids in the water internal to th'e storage cell (to simulate boiling) de-creased reactivity, as shown in the table. voids due to boiling will not occur in the outer (flux-trap) water region. O 4-28

L l i l -. Table 4.6 Effect of Temperature and Void on Calculated Reactivity of Storage Rack Incremental Reactivity Change, Ak Case Region 1 Region 2 Region 3 40*F Reference. Reference 68'F ( 20

• C)

-0.0021 -0.0082 104*F (40

• C)

-0.0055 -0.0004 150*F (65.56*C) Reference 248*F (120

• C)

-0.0205 -0.0051 +0.0108 248'F with 20% void -0.1111 -0.0566 -0.0015 4.6.3 Dropped Fuel Assembly Accident i l To investigate the possible reactivity ef fect of a postu-lated fuel assembly drop accident, calculations were made for unpoisoned assemblies separated only by water. Figure 4.10 shows the results of these calculations. From these data, the reactiv-ity (k.) will be less than 0.95 for any water-gap spacing greater than ~6 in. in the absence of any absorber material other than water between assemblies. For a drop on top of the rack, the i fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance of >12 in. Maximum expected l deformation under seismic or accident conditions (see Sections 6 and 7) will not reduce the minimum spacing between fuel assem-blies to less tl.an 12 in. Consequently, fuel assembly drop accidents will not result in an increase in reactivity above that l calculated for the infinite nominal design storage rack. 4.6.4 Abnormal Positioning of Fuel ' Assembly Outside Stotage Rack If a fresh fuel assembly of the highest initial reactivity were to be positioned outside and adjacent to a fuel rack, the 4-29

l reactivity could potentially be increased by a percent or more. If the fuel element had accumulated some burnup prior to dis-charge, the reactivity increase would be less. The fuel racks, however, are designed so that the space between the fuel rack and the pool wall (<5 in.) is not sufficient to permit a fuel assem-bly being abnormally positioned outside a fuel rack. Similarly, the spacing between rack modules is too small to permit inserting an extraneous assembly in these positions. The Region 3 cells facing the cask area represent the only module faces where an extraneous assembly might conceivably be positioned. To preclude

this, those modules facing the cask area will have mechanical stops to assure that a minimum space of 8 in, will be maintained between the rack module and a' potential extraneous assembly.

From Fig. 4.10, this spacing is more than adequate to maintain k, less than 0.95. Furthermore, soluble boron is normally present in the spent fuel pool (f r which credit is permitted in this condition) and would reduce the maximum k to substantially less O ehan 0.95. Therefore, it is conc 1uded that the aenorma1 gesi-tioning of a fuel assembly outside and immediately adjacent to the storage rack is not a credible occurrence. 4.6.5 Lateral Rack Movement Lateral motion of the rack modules under seismic conditions could alter the spacing between rack modules. However, the lat-eral motion is not of suf ficient magnitude to reduce the spacing to less than the nominal spacing of the flux-trap water gaps in the reference storage cell. In addition, soluble boron would substantially reduce the k,, under the postulated conditions. O 4-30

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=====r=---

.....__.__2___ - - -n E _~_ _ ,92 _____-~~-~-2 = =: _ _ ---+ - -TT:iEEE = EEM = b ~1-L~i --=u--- g 3 ---. L .90 i i i -f g O 2 4 6 8 10 !2 E W ATER G AP BETWEEN FUEL ASSEMBLIES, inches a Fig. 4.10 Infinite multiplication factor (kgo ) or. i 4.3% enriched fuel assemblies separated only bi water. 4-31 ~

l 1 REFERENCES n V) t 1. Green, Lucious, Petrie, Ford, White, Wright, PSR-63/AMPX-1 (code package), AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B, ORNL-TM-3706, Oak Ridge National Laboratory, March 1976. 2. L. M. Petrie and N. F. Cross, KENO-IV, An Improved Monte Carlo Criticality Program, ORNL-4938, Oak Ridge National Laboratory, November 1975. 3. S. R. Bierman et al., Criticb5 Subcritical Clusters of 4.29 wt% U Enriched UO2 Rods in Water with Fixed Neutron Poisons, NUREG/CR-0073, Battelle Pacific Northwest Laboratories, May 1978, with errata sheet issued by the USNRC August 14, 1979. 4. M. N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, The Babcock & Wilcox Company, July 1979. ~ 5. R. M. Westfall and J. R. Knight, Scale System Cross-Section Validation with Shipping-Cask Critical Experiments, ANS Transactions, Vol. 33, p. 368, November 1979. s 6. S. E. Turner and M. K.

Gurley, Evaluation of AMPX-KENO Benchmark Calculations for High Density Spent Fuel Storage w

Racks, Nuclear Science and Engineering, 80(2):

230-237, February 1982. A Fast Transport Theory 7. A. Ahlin and M.

Edenius, CASMO Depletion Code for LWR Analysis, ANS Transactions, Vol. 26,

p. 604, 1977 CASMO-2E Nuclear Fuel Assembly Analysis, Application Users Manual, Rev. A, Control Data Corporation, 1982.

8. W. J. Eich, Advanced Recycle Methodology Program, CEM-3, Electric Power Research Institute, 1976. Neutron Spectrum Generator, Few-9. W. A. Wittkopf, NULIF Group Constant Generator and Fuel Depletion Code, BAW-426, The Babcock & Wilcox Company, August 1976. 10' NAI Modified

LEOPARD, Rev.

2, NAI Report 71-13 (Proprietary), Nuclear Associates International Corporation, December 10, 1973. " CHEETAH-P" report module within the LEAHS Nuclear Fuel Management and Analysis Package, Publication No. 84004100 (Proprietary), Nuclear Associates International Corporation, July 1974.

REFERENCES ( Continued) 11. R. A. Lorenz et al., F'ission Product Release From Highly Irradiated LWR Fuel, NUREG/CR-0722, February 1980. 12. M. G.'Natrella, Experimental Statistics National Bureau of Standards, Handbook 91, August 1963. 13. W. R. Cadwell, PDQ-7 Reference Manual, WAPD-TM-678, Bettis Atomic Power Laboratory, January 1967. G 0 O

5. THERMAL-HYDRAULIC CONSIDERATIONS O A central objective in the design of the high-density fuel rack is to ensure adequate cooling of the fuel assembly cladding. In the following, a brief synopsis of the design basis, the method of analysis, and computed results are given. Similar analysis has been used in previous licensing reports on high density spent fuel racks for Fermi II (Docket 50-341), Quad Cities I and II (Dockets 50-254 and 50-265), Rancho Seco (Docket 50-312), Grand - Gulf Unit l' (Docket 50-416), and Oyster Creek (Docket 50-219). 5.1 Decay Heat Calculations for the Spent Fuel This report section covers requirement III.l.5(2) of the,NRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" issued on April 14, 1978. This requirement states that calculations for the amount of thermal energy removed by the spent fuel cooling system shall be made in accordance with Branch Technical Position APCSB 9-2 " Residual Decay Energy for Light Water Reactors for Long Term Cooling"I. The calculations contained herein have been made in accordance with this requirement. I 5.1.1 Basis The V.C. Summer reactor is rated at 2775 Megawatt-Thermal (MWT). The core contains 157 fuel assemblies. Thus, the average operating power per fuel assembly, Po, is 17.675 MW. The fuel assemblies are assumed to be removed from the reactor after a '000 EFPD (Effective full power days) in accordance with the 1 plant system diagram. The fuel discharge can be made in one of the following two modes: (i) Normal discharge - Mode (i) (ii) Pull Core discharge - Mode (ii) O 5-1

As shown in Table 1.1, the anticipated fuel batch size for normal discharge may vary from 6.$ to 72.

However, for analysis O

ourgesee, we es eme the bat *,sise te ae 8. fue1 essemb11ee. The fuel transfer begins atten 144 hours of cool-off time in the reactor (time after shut (deurJ. It is assumed that the time psriod of discharge of thiss thatch is 27 hours. (Three assemblies transferred to the pool per: ihmr). The cooling system consists of' two seismic category I spent: fuel cooling circuits. The bulk temperature analysis assumes the typical

• operating condition of two spent fuel-pool coolers working in parallel.

For comparison purposes, the bulk temperatures corresponding to batch sizes of 1/3 rd core (53 assemblies) and 76 assemblies are also calculated.

However, for these cases the normal

• condition of only one cooler in operating condition is analyzed and reported herein.

Mode (ii) corresponds to a full core discharge (157 assemblies). Full core off-load condition implies that the -s reactor core has no remaining fuel. It is assumed that the total time period for the discharge of one full core is 52 hours (after 144 hours of shut down time in the reactor). The discharge rate to the pool is assumed to be continuous and uniform. The bulk temperature analysis assumes two spent fuel pool coolers working in parallel. The water inventory in the reactor cavity cooled by the RHR heat exchanger exchanges heat with the fuel pool water mass through the refueling canal. This important source of heat removal is also neglected in the analysis. Thus, the results obtained for both modes (i) and (ii) are highly conservative. Normal condition refers to normal discharge assuming only one cooler in operating condition;

however, typically two coolers are available.

p G 5-2

In the,following, aII relevant performance data for the () spent fuel pool heat exchangnrs is given. a. Spent Fuel Poma 3 eat Exchanger: Type Tube and shell Quantity 2 Performance data s. o Heat transferred 14.02 x 106 Btu /hr Tube Side o Fluid flow 1,800 gpm o Pool water inlet temperature 135'F o Outlet temperature 119.2*F Shell side o Fluid flow 1800 gpm o Coolant inlet temperature 105*F o Outlet temperature 120.8'F o Touling factor 0.0005 The above data enables complete characterization of the thermal performance of the fuel pool heat exchanger. 5.1.2 Model Description Reference (l) is utilized to compute the heat dissipation requirements in the pool. The total decay power consists of " fission products decay" and " heavy element decay". Total decay power P for a fuel assembly is given as a linear function of Po and an exponential function of to and ts* i.e.: P = Po f(to,ts) where P= linear function of P O o 5-3 c..........

P,o= average operating power per fuel assembly to=cumulatiw expcaure time of the fuel assembly in the reactor t = Time elapsed since reactor shutdown s The uncertainty factor K, which occurs in the functional relationship f (to,ts) is set equal to 0.1 for t > 107 sec *in the Interest of conservatism. Furthermore, s the operating power Po is taken equal to the rated power, even though the reactor may be operating at less than its rated power during most of the period of exposure of the batch of fuel assemblies. Finally, the computations and results reported here are based on the discharge taking p' lace when the inventory of fuel in the pool will be at its maximum resulting in an upper bound on the computed decay heat rate. O Having determined the heat dissipation rate, the next v task is to evaluate the time-temperature history of the pool water. Table 5.1.1 identifies the loading cases examined. The pool bulk temperature time history is determined using the first law of thermodynamics (conservation of energy). A number of simplifying assumptions are made which render the analysis conservative. The principal ones are: 1. The cooling water temperature in the fuel pool cooler is based on the maximum postulated values given in the FSAR.2 2. The heat-exchangers are assumed to have maximun fouling. Thus, the temperature effectiveness, s, for the heat exchangers utilized in the analysis are the lowest postulated values: S= 0.526 for 3(V s-4

fuel pool

coolers, S

is calculated from heat exchanger te chnical data sheets.No heat loss is assumed to talke place through the concrete floor. 4. No credit is uken for the improvement in the film coefficients of the heat exchangers as the operating temperature rises.

Thus, the film coefficient used in the computations are lower bounds.

5. No credit is taken for evaporation of the pool water. The basic energy conservation relationship for the pool heat exchanger system yields:

== 0 03 (5.1.2) Ct 1 dT OV) where Ct: Thermal capacitance of stcred water in the pool. t: Temperature of pool water at time, T 0: Heat generation rate due to stored fuel 1 assemblies in the pool. 01 is a known function of

time, T

from the preceding section. 02; Heat removed in the fuel pool cooler. 03: ' Heat removed in the RHR heat exchanger (03=0 if RHR is not used). 5-5 hau r i

The po,ol has total unter inventory of 38162 cubic feet when all racks are in pl.nce in the pool and every storage , location is occupied. 5.1.3 Decay Heat Calculation Results: The calculations were performed for the pool disregarding the additional thermal capacity and cooling syst.em available in the transfer channel, and the reactor cavity. For a specified coolant inlet temperature and flow

rate, the quantities 02 and Q3 are shown to be linear function of t

in a recent paper by Singh (3). As stated 0, is an exponential function of T. Thus Equation

earlier, 1

(5.1.2) can be integrated to determiite t directly as a function ~ of T. The results are plotted in Figures (5.1.1) to (5.1.4). The results show that the pool water never approaches the boiling point under the most adverse conditions. These figures also give 01 as a function of T. Four plots are generated for each case. The first and third plots for each case shows temperature and power generation respectively for a period extending from t= 2t where T is the total time of fuel transfer. The 0 + r = n n second and fourth plots show the same quantities (i.e. temperature and power generation respectively) over a long period. The long-term plots are produced to indicate the required operating time for the heat exchangers. Summarized results are given in Table 5.1.2. Finally, computations are made to determine the time interval to boiling after all heat dissipation paths are lost. Computations are made for each case under the following two assumptions: (i) All cooling sources lost at the instant pool bulk temperature reaches the maximum value. O 5-6

o o o TABLE 5.1.1 a LIST OF CASES ANALYZED i Case No. Condition No. of No. of No. of Total Time Cool off time i fuel spent fuel RHR's to transfer before transfer I assemblies pool HXS in-service fuel into begins, hrs. discharged the pool l N thi hrs. 1 Normal discharge 80 2 0 27 '144 4 i 4 1 2 Full core 157 2 0 52 144 l discharge 1 t 3 Normal discharge 76 1 0 25 144 i 4 Normal discharge 53 1 0 18 144 1 4 t 6 -s e E

1 O O O l TABLE 5.1.2 MAXIMUM POOL BULK TEMPERATURE t, COINCIDENT TOTAL POWER Q1 and COINCIDENT SPECIFIC POWER FOR THE HOTTEST ASSEMBLY Case No. of Time Maximum Coincident Coincident Q1x10-6 Notes No. Assemblies to transfer pool bulk time (since specific BTU / hour fuel into temp.*F initiation power q, pool, hrs. of fuel BTU /sec. transfer, hts. 1 80 27 123.3 34 49.77 17.018 Maximum normal batch size 2 157 52 136.7 58 47.37 29.454 Full core offload l 3 76 25 139.7 40 49.13 16.123 Normal condition l l 4 53 18 131.2 34 49.77 12.179 Normal condition. l 1/3 rd core off-load l O e S

O O O TABLE 5.1.3 TIME (Hrs) TO BOILING AND BOILING VAPORIZATION RATE FROM THE INSTANT ALL COOLING %S LOST Case No. CONDITION 1 CONDITION 2 Loss of Cooling at maximum Loss of Cooling at maximum pool bulk temperature power discharge rate Time (Hrs) Vap. Rate Time (Hrs) Vap. Rate Ib./hr. Ib./hr. 1 13 17135 12 17380 Y WP 2 6 30050 6 30360 3 10 16332 11 i 167'38 ~ ~ i 1 4 16 '12230 16 12553 l l l

(ii) A,ll cooling paths lost at the instant the heat G dissipation power reaches its maximum value in the U pool. Results are sunusarized -in Table 5.1.3. Table 5.1.3 gives the bulk boiling vaporization rate for both cases at the instant the boiling commences. This rate will decrease with time due to reduced heat generation in the fuel. 5.2 Thermal-Bydraulics Analyses for Spent Fuel Cooling This report section covers requirement III.1.5(3) of the NRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" issued on April 14, 1978. Conservative methods have been used t6 calculate the maximum fuel cladding temperature as required therein.

Also, it has been determined that nucleate boiling or voiding' of coolant on the surface of the fuel rods does not occur.

O

5.2.1 Basis

I r. order to determine an upper bound on the maximum fuel cladding temperature, a series of conservative assumptions are made. The most important assumptions are listed below: l a. As stated above, the fuel pool will contain spent l fuel with varying " time-after-shutdown" (ts)- l Since the heat emission falls off rapidly with increasing ts, it is obviously conservative to assume that all fuel assemblies are fresh 144 hours) and they all have had 24000 (ts = hours.of operating time in the reactor. The heat emission rate of each fuel assembly is assumed to be equal.2 O l 5-10

b. As shown in Figures 2.1 in Section 2, the modules occupy an irregular floor space in the pool. For Oy purposes of the hydrothermal analysis, a circle circumscribing the actual rack floor space is drawn. It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal pitch of 10.405 inches (see Figure 5.2.1). c. The downcomer space around the rack module group varies, as shown in Figure 5.2.1. The nominal downcomer gap available in the pool is assumed to be the total gap available around the idealized cylindrical rack; thus, the maximum resistance to downward flow is incorporated into the analysis. d. No downcomer flow is assumed to exist-between the rack modules. 5.2.2 Model Description In this manner, a conservative idealized model for the rack assemblage is devised. The water flow is axisymmetric abo'ut the vertical axis of the circular rack assemblage, and thus, the flow is two-dimensional (axisymmetric three-dimensional). Fig. 5.2.2 shows a typical " flow chimney" rendering of the thermal hydraulics model. The governing equation to characterize the flow field in the pool can now be written. The resulting integ ral-equation can be salved for the lower plenum velocity field (in the radial direction) and axial velocity (in-cell velocity field), by using the method of collocation. It should be added that the hydrodynamic loss coefficients which enter into the formulation of the integral equation are also taken from well-recognized sources 4 and wherever discrepancies in reported values exist, the conservative values are consistently used. Reference [5] gives the details of mathematical analysis used in O' this solution process. 5-11

l Af ter,the axial velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding . temperature. The knowledge of the overall flow field enables pinpointing the storage ' location with the minimum axial flow (i.e. maximum water outlet temperature). This is called the most " choked", location. It is recognized that some storage locations, where rack module supports are located, have some additional hydraulic resistance not encountered in other cells. In order to. find an upper bound on the temperature in such a cell, it is assumed that it' is located at the most " choked" location. Knowing the global plenum velocity field, the revised axial flow through this choked call can be calculated by solving the Bernoulli's equation for the ~ circuit through this cell. flow Thus, an absolute upper bound on the water exit temperature and maximum fuel cladding temperature is obtained. It is believed that in view of the aforementioned assumptions, the temperatures calculated in this manner overestimate the temperature rise that will actually occur in the pool. The maximum pool bulk temperature t is computed in Section 5.1.3 and reported in Table 5.1.2. The corresponding average power output from the hottest fuel assembly, q is also reported in that table. The maximum radial peaking factor is 1.55 for the V.C. , Summer installation. Thus, it is conservative to assume that the maximum specific power of a fuel assembly is given by 9A = q ar r = 1.55 where a The maximum temperature rise of pool water in the most disad.vantageously placed fuel assembly is given in Table 5.2.1 for all loading cases. Having determined the maximum " local" water temperature in the pool, it is now possible to determine the maximum fuel cladding temperature. It is conservatively assumed that the total peaking factor aT is 2.32. Thus, a fuel rod can produce 2.32 times the average heat emission rate over a 5-12

s ^ small length. The axial heat dissipation in a rod is known to reach a maximum in the central region, and taper off at its two extremities. For the sake a:E added conservatism it is assumed s that the peak heat emission occurs at the top where the local water temperature also reaches its maximum. Furthermore, no credit is taken for axial conduction of heat along the rod. The highly conservative model thus constructed leads to simple algebraic equations which directly give the maximum local cladding temperature, tc* 5.2 3 Results: Table 5.2.1 gives the maximum local cladding temperature,, tc, at the instant the pool bulk temperature has attained its maximum value. It is qtiite possible, however, that the peak cladding temperature occurs at the instant of maximum value of qA, i.e., at the instant when the fuel assembly is - first placed i n' '. a storage location. Table 5.2.2 gives the maximum local cladding temperature at t = 0. It is to be noted that there are wide margins to local boiling in all cases. The ~ local boiling temperature near the top of the fuel cladding is 240*F. Furthermore, the cladding temperature must be somewhat higher than the boiling temperature to initiate and sustain nucleate boiling. The above considerations indicate that a comfortable margin against the initiation of localized boiling exists in all cases. \\s 4 \\ t v 5-13

O O O TABLE 5,2.1 MAXIMUM LOCAL POOL WATER TEMPERATURE AND LOCAL FUEL CLADDING TEMPERATURE AT INSTANCE OF MAXIMUM POOL BULK TEMPERATURE Case No. Max. Local Pool Maximum Coincident Local Case Water Temperature 'F Cladding Temperature 'F Identified 1 155.5 184.0 80 Assemblies Cooling Mode A I l 2 139.8 167.1 157 Assemblies Cooling Mode A Y 5 3 171.6 199.8 76 Assemblies Cooling Mode B 4 163.4 191.9 53 Assemblies Cooling Mode B

• Coolfng Mode A means 2 spent fuel pool coolers (SFPHX).

Cooling Mode B means only 1 SFPHX in operation. s 6e

(:) TABLE 5.2.2 POOL AND MAKIMUR CLADDING TEMPERATURE AT THE INSTANCE FUEL ASSEMBLY TRANSFER BEGINS Case No. Cladding Coincident Pool Temp.

• F

Temp,

'F Bulk Local' 1 or 2 172.8 108 142.0 () .3 or 4 175.7 110.9 145.0 9 e 5-15

REFERENCES TO SECTION 5 O 1. NUREG 0800 U.S. Nuclear Regulatory Commission, Standard Review Plan, Branch' Technical Position, ASB 9-2, Rev. 2, July 1981. 2. FSAR, V.C. Summer Nuclear Station. 3. Journal of Heat Transfer, Transactions of the ASME August,1981, Vol. 103, "Some Fundmental Relationships for Tubular Heat Exchanger Thermal Performance," K.P. Singh. 4. General Electric Corporat' ion, R&D Data Books, " Heat Transfer and Fluid Flow, " 1974 and updates. 5. 4th National Congress of the ASME, "A Method for Computing the Maximum Water Temperature in a Fuel Pool Containing Spent Nuclear Fuel", paper 83-NE-7, Portland, Oregon (June 1PS3). O t 5-16

O o* e N' r E. 0 PEAK VALUE = 123.3 F AT 34 HRS. wI" o o m$- (.D j 3 CASE 1 o W, Number of Assemblies 80 O = I $E Time of Discharge 27 hrs. = l [ Spent Fuel Pool Heat Exchangers= 2 E R.H.R. Heat Exchanger 0 = Wo (1.o l EQ m._ F--

• o o,

e h" oo T k.co ib.co ab.co 3b.co 4b.co sb.co s'o.co TIME (HOURS) FIG. 5.1.1 (a) POOL BllLK TEMDERATtIRE; NORMAL DISCHARGE 1 5-17 .._._________.---_--,---_-.-.--.__.__-..----,,.,=a

' O A 8 i 3-o 0 PEAK VALUE = 123.3 F AT 34 HRS. 5- / o o C.D N-L1J" f Q ~ CASE 1 O m:d E Number of Assemblies 80 = U" p Time of Discharge 27 hrs. = Spent Fuel Pool Heat Exchangers = 2 Wg R.H.R. Heat Exchanger 0 = zi w._ HM 8 i o. - 8 3.00 2'.00 /.00 6'.00 8'.00 1'O.00 i2.00 TIME (DAYS) O ero. s 1 '<8) eoo' Bu'x Teneea^'uae: "oa"^' orsc"^aat 5-n

---

w

-m -.m.m.- e

O ) C9 E. N o9oO. h PEAK VALUE = 17.25 x 10 BTU /HR. 0 ,1 AT 27 HRS.

1. o no crg I to Na.

3 l-(D CASE 1 a go Number of Assemblies 80 = ~ Time of Discharge 27 hrs. = C S.F.P. Heat Exchangers 2 = 4Io R.H.R. Heat Exchangers 0 = co tn Hg-O C tux8 O CL.0-o 9 %00 1'O.00 2'0.00 3'0.00 4'0.00 5'0.00 6'0.00 TIME (HOURS) O l FIG. 5.1.1'(c) POWER DISCl!ARGE; NORMAL DISC!!ARGE t 5-19 4 -N "e--- ,,a,--~,---,,--------a t -9 -,----a-e ,,m,n-e-w,-~---,--,--w .,-,a- ---,,--we-- r,,--s,,---,,e,-----,--,v,----e

8 N 8 5. 0 v4 PEAK VALUE = 17.25 x 10 BTU /HR. AT 27 HRS. M a o n Im;. x Na 3 l-o "R C Q "- W CASE 1 to E Number of Assemblies 80 = Zo Time of Discharge 27 = Oo S.F.P. Heat Exchangers (/)o. 2 = g l m o R.H.R. Heat Exchangers 0 = W W f za. o 8" l 8 %.00 2'.00 /.00 6'.00 8'.00 i0.00 l'2.00 TIME (DAYS) FIG. 5.1.1 (d) POWER DISCHARGE; NORMAL DISCHARGE 5-20

• • w-*

O 0 PEAK VALUE = 136.65 F AT 58 HRS. 8 i g. 8 4 0" CASE 2 o Number of Assemblies 157 = CD N-Time of Discharge 52 = W Q Spent Fuel Pool Heat Exchangers= 2 o R.H.R. Heat Exchanger 0 = O W *. To D ru. F" EWo O.o Ef W.. l F-

• l l

8 d .8 i d 1.00 2'0.00 /0.00 6'0.00 8'0.00 1'00.00 l'20.00 TIME (HOURS) O rio 5 i 2 (e) Poo' au'x Teata^T"ae: ru u coae oisc"^ ace 5-21 l

O 0 PEAK VALUE = 136.65 F AT 58 HRS. i2 .~ o 8. w t -8

u..

Q CASE 2 Number of Assemblies = 157 ~ Time of Discharge 52 = W SFP Heat Exchanger 2 = [o O m. R.H.R. Heat Exchanger 0 = F* 4: EWo Q.o Ed w [ H* 8 O w' l 8 l 5 1.00 4'.00 8'.00 1'2.00 i6.,00 2'O.00 2'4.00 TIME (DAYS) FIG. 5.1.2 (b) P0OL BULK TEMPERATURE; FULL CORE DISCHARGE s-22

s O-o PEAK VALUE = 29.77 x 10 BTU /HR. O AT 52 HRS. o 8" o9o8-CASE 2 wt Number of Assemblies 157 = Time of Discharge 52 = E*o S.F.P. Heat Exchangers 2 = ro. NN R.H.R. Heat Exchangers 0 = 3 P-G3 0 08 we CD E<28 c. (J)O HO-o I EW38 O-Q.g-8 i.00 2'0.00 /0.00 6'0.00 8'0.00 1'00.00 i20.00 TIME (HOURS) FIG. 5.1.2 (c) POWER DISCHARGE; FULL CORE DISCHARGE. 5-23

O ~ o3 6 PEAK VALUE = 29.77 x 10 BRl/HR. AT 52 HRS. oo. M o S. .N 'O ,e oo mg Zo. NN CASE 2 3l-- CDw@ Number of Assemblies 157 = O@" Time of Discharge = 52 Wg" S.F.P. Heat Exchangers 2 = E R.H.R. Heat Exchangers 0 = 4IS u. (Do HO-O EWES O. Q.@- S ~ .00 4'.00 8',00 i2.00 i6.00 2'O.00 2'4.00 TIME (DAYS) FIG. 5.1.2 (d) POWER DISCIIARGE: FULL CORE DISCl!ARGE 5-2.4

O~ 0 PEAK VALUE = 139.7 F at 40 hrs. o k. m oo ed CASE 3 m. m Number of Assemblies = 76 Time of Discharge = 25 hrs. LL. o Spent Fuel Pool Heat Exchangers = 1 mm. W* R.H. R. Heat Exchanger = 0 _O o O lil*. Em Dan. 1* 4 E ELEo a.o Z*N. F-* o9 mm. m o '3 3.00 1'0.00 2'0.00 3'0.00 [0.00 5'0.00 6'0.00 TIME (HOURS) FIG. 5.1.3(a) Pool Bulk Temperature; Normal Discharge with Loss of 1 SFPHX. 5-25

O PEAK VALUE = 139.6 at 40 hrs. o .ov. w it. o mg. o CASE 3 m LL. o t.O tn. w Number of Assemblies e = 76 = 25 hrs. O Time of Discharge Spent Fuel Pool Heat Exchangers = 1 = 0 R.H.R. Heat Exchanger En Ocu. W* 4EWo (1.o Z'o f Wtv. Fw o9 mw. w o ? S 4.00 a'.oo 4'.00 s'. co s'. co io.co s'a.co TIME (DAYS) FIG. 5.1.3 (b) Pool Bulk Temperature; Normal Discharge O With Loss of 1 SFPHX 5-26 l I

O N 8 4 o-6 m N PEAK VALUE = 16.5875 x 10 BTU /HR. at 25 hrs. O v4*o o Ef Im_ Na D t-8, CASE 3 Y. Number of Assemblies = 76 Time of Discharge = 25 hrs. To Spent Fuel Pool Heat Exchanger = 1 Uo CD

• R.H.R. Heat Exchanger

= D t-4$- O E 10x8 O G.O-o9 %.00 i0.00 2'O.00 3'0.00 4'0.00 5'0.00 6'O.00 TIME (HOURS) FIG. 5.1.3 (c) Power Discharge; Normal Discharge IUith Loss of 1 SFPHX. 5-27

~ O 8 S. N 8 4 m Oy PEAK VALUE = 16.5875 x 10 BTU /hr. oo E* re 3 CASE 3 CD E <C Number of Assemblies = 76 $)*~ U Time of Discharge = 25 hrs. Q Spent fuel Pool Heat Exchangers = 1 = 0 g R.H.R. Heat Exchanger to28 O. O O. ~ 8 %.00 2'.00 4'.00 6'.00 8'.00 l'0.00 i2.00 TIME (DAYS) FIG. 5.1.3 (d) Power Discharge; Normal Discharge tuith Loss of 1 SFPHX. 5-28

O PEAK VALUE = 131.2 F at 34 hrs. C-8 i E" , CEE 4 5. Number of Assemblies 53 = (,9 W 18 hrs. o Time of Discharge = ~ 1 Spent Fuel Pool Heat Exchangers = W. R.H.R. Heat Exchanger 0 = l Eo l 3ni.

• C E

E13o 0.0 Zf W w. 6--

• 8

l-8 5

3.00 i0.00 2'0.00 3'0.00 4'O.00 5'0.00 6'O.00 TIME (HOURS) FIG. 5.1.4 (a) Pool Bulk Temperature,1/3 Core Discharge with loss of 1 SFPHX 5-29

O 8 0 PEAK VALUE = 131.2 F at 34 hrs. 8 i. 2- .8 u v (DN. w" CASE 4 8 O kg Number of Assemblies 53 = U$- Time of Discharge 18 hrs. = g 4 Spent Fuel Pool Heat Exchangers 1 = EWo R.H.R. Heat Exchanger 0 = (1.0 i Ef w._ t--

• 8 l

8 5 1.00 2'.00 4'.00 6'.00 8'.00 i0.00 i2.00 TIME (DAYS) FIG. 5.1.4 (b) Pool Bulk Temperature; 1/3 Core Discharge With loss of 1 SFPHX 5, 30

O 8 l l a w. 6 W4DE e 12.55 x 10 BTU /HR. at 18 hrs. 8 E. o ~ l ,1 % O Q9 IE. CASE 4 Nw 3 Number of Assemblies 53 = ~g Time of Discharge 18 hrs. = Of, Spent Fuel Pool Heat Exchanger 1 = We CD R.H.R. Heat Exchanger 0 = E<Zo Oo m@. M O EWx8 O. Q.O-8 a "b.co sb.co a'o.oo ab.co 4'o. co sb.co s'o.oo TIME (HOURS) O FIG. 5.1.4 (c) Power Discharget 1/3 Core Discharge flith loss of 1 SFPHX 5 31 I

O J 8 E. i w 6 PERK TREUE = 12.55 x 10 BTU /hr. 8 / S. Ow*o Q*. IE Nw. 3 CASE 4 E1 i I 8 Nurbber of Assemblies 53 = O '. Time of Discharge l 18 hrs. g = CD Spent fuel Pool Heat Exchangers 1 = K< l R.H.R. Heat Exchanger 0 = Zo Uo U3

• HE-i t

O E Id 3:@ ~ l O. Q.0-8 l 6 %.00 2'.00 4'.00 6'.00 8'.00 i0.00 i2.00 TIME (DAYS) O-FIG. 5.1.4 (d) Power Discharge; l/3 Core Discharge Hith Loss of 1 SFPHX 5-32 l

I a-O ^ i d e a li z e d ' O mt5 fin e Of Rac k.A ssembly Ra k s embly r ~ / \\ //EA o RACK ASSEMBLY l s i 8 Actual Outline l of Pool idealized l Outline of Assumed Added Fuel Assemblies Pool Boundary O FIG. 5.2.1 IDEALIZATION OF RACK ASSEMBLY 5-E

l O Water Assumed At The I Pool Bulk Temperature ,r - + ~. U J L ^ / / l~} / _/ Tout"[] ^ l l c VO I IN7 w O e a t W 2 jp 3' g J a O w d + O Heat Addition W V 3' S Z 4i __/ E v / Bi

L y

I IN d / / = q0t I /) / ' ~ s j FIG. 5.2.2 THERMAL CHIMNEY FLOW MODEL 5-34 1

6. STRUCTURAL ANALYSIS The purpose of this sectinc is to demonstrate the structural adequacy of the spent fuel radk design under normal and accident loading ~ conditions. The method of analysis presented herein is similar to that previously used in the Licensing Reports on High Density Fuel, Racks for Fermi II (Docket 50-341), Quad Cities I and II (Dockets 50-254 and 50-265), Rancho Seco (Docket No. 50-312), Grand Gulf Unit 1 (Docket No. 50-416) and Oyster Creek (Docket No. 50-219). The results show that the high density spent fuel racks are structurally adequate to resist the postulated stress combinations associated with normal and accident conditions. 6.1 Analysis outline: The spent fuel storage racks are seismic

Category, I

equipment. Thus, they are required to remain functional during and after an SSE (Safe Shutdown Earthquake).1 As noted previously, 9 these racks are neither anchored to the pool floor, nor are they attached to the side walls. The individual rack modules are not interconnected. Furthermore, a particular rack may be completely loaded with fuel assemblies (which corresponds to greatest rack inertia), or it may be partially loaded so as to produce maximum geometric eccentricity in the structure. The coefficient of

friction, y,

between the supports and pool floor is another indeterminate factor. According to Rabinowicz,2 the results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of p to be 0.503 with a standard deviation of 0.125. The upper and lower bounds (T20) are thus 0.753 and 0.253, respectively. Two separate analyses are performed for this rack assembly with values of p equal to 0.2 (lower limit) and 0.8 (upper limit) respectively. Initially, the following four separate analyses are performed on the largest rack module (Module A in Table 2.21 1. Fully loaded rack (all storage locations occupied), y = 0.8 ( p = coefficient of friction). 2. Fully loaded rack, p = 0.2. 6-1

l O 2-aer r cx - a 4. Empty rack, p = 0.2. 1 I ~ Based on the results of these runs, additional analyses are performed. The actual studies performed for the different rack modules are summarized in Section 6.6. I' The method of analysis employed is the time history met { hod. The pool slab acceleration data are developed from the original pool floor response spectra. The object of the seismic analysis is to determine the structural response (stresses, deformation, rigid body motion, etc.) due to simultaneous application of the three orthogonal excitations. Thus, recourse to approximate statistical summation techniques such as " Squ a re-Roo t-o f-t he-S um-o f-t he-Squ a re s " method 3 is avoided and the dependability of computed results is ensured. The seismic analysis is performed in four steps; namely 1. Development of nonlinear dynamic model consisting of

beam, gap,

spring, damper and inertial coupling

. elements. 2. . Derivation and computation of element stiffnesses using a sophisticated elastostatic model. O l 6-2

3. Layout, of the equations of

motion, and inertial decoupling and solution of the equations using the

" component element time integration" procedure 4,5 to determine nodal' and element forces and displacements of nodes. 4. Computation of the detailed stress field in the rack structure, using the detailed elastostatic model, from the nodal forces calculated in Step III above. Determine 1E the stress and displacement limits,given in Section 6.5, are satisfied. A brief description of the dynamic model follows. l 1 6.2 Fuel Rack - Fuel Assembly Model 6.2.1 Assumptions )

a. The fuel rack metal structure is represented by five lumped masses connected by appropriate elastic springs as shown in Figure 6.1.

The spring rates sir.ulate the elastic behavior of the fuel rack as a beamlike structure.

b. The fuel assemblies are represented by five lumped masses located, relative to the rack, in a manner which simulates either fully or partially loaded conditions.

c. The local flexibility of the rack-support interface is modeled ccnservatively in the analysis.

d. The rack base support may slide or lift off the pool floor.

O l 6-3

s

e. The pool floor is assumed to have a known time history of ground accelerations along the three orthogonal directions.

f. Fluid coupling between rack and assemblies, and between rack and adjacent racks is simulated by introducing appropriate inertial coupling into the system kinetic energy.

g. Potential impacts between rack and assemblies are accounted for by appropriate spring gap connectors between masses involved.

h. Fluid damping between rack and assemblies, and between rack and adjacent rack is conservatively neglected.

i. The supports are modeled as extensional elements for dynamic analysis.

The bottom of a support leg is attached to a frictional spring as described in Section 6.2.2. The cross section properties of the support beams are derived and used in the final l computations to determine support leg stresses.

j. The effect of sloshing can be shown to be negligible at the bottom of a pool and is hence neglected.

l i / 6-4

6.2.2 Model Description O The absolute degrees of freedom associated with each of the mass locations i, i* are a~s follows (see Figure 6.1): Table 6.1 Degrees of Freedom Location Displacement Rotation 0 O O (Node) ux uy uz x y z 1 p1 p2 P3 94 95 96 B'Y,2=0 1* Point is assumed fixed to base at X B 2 p7 pg 411 912 0 2* P8 P10 3 P13 P15 917 918 3* P14 P16 4 P19 P21 923 924 4* P20 P22 5 p25 P27 P32 429 930 931 5* p26 P28 O 6-5

• /

/ Thus, there are 3a degrees of freedom in the system. Note that elastic motion of the rack in extension is represented by generalized coordinates P3 and P32 This is due to the relatively high axial rigidity of the rack. Torsional motion of the rack relative to its base is governed by q31 The members joining nodes 1 to 2, 2 to 3, etc., are the beam elements with deflection capability due. to bending and shear (see Reference 4,pp. 156-161.). The elements of the stiffness matrix of these beam elements are readily computed if the ef fective flexure modulus, torsion modulus, etc., for the rack structure are known. These coefficients follow from the elastostatic model as described later. The nodal points i (i = 1,2..

5) denote the fuel rack mass at the 5

elevations. ,The node points i* (i* = 1,2..

5) denote the cumulative mass for all the fuel assembl,ies distributed at 5 elevations.

The element stiffnesses of the fuel assembly are obtained from the structural properties of the OCNS fuel assemblies. The nodes i* are located at x = XBe Y YB in the global coordinate system shown in Figure 6.1. 6.2.3 Fluid Coupling An effect of some significance requiring careful modeling is the so-called " fluid coupling effect." If one body of mass mi vibrates adjacent to another body (mass m2), and both bodies are submerged in a frictionless fluid

medium, then the Newton's J

O 6-6 l

~ g equation of motion for the two bodies have the form (mi + M11) XI-M12 X2 = applied forces on mass mi M21 X1 + (m2 + M22) X2 = applied forces on mass m2

M11, M12e

M21, and M22 are fluid coupling coefficients which depend on the shape of the -two bodies, their relative disposition, etc.

-Eritzbg4Nec data for Mj for various body i shapes and arrangements. It is1to be noted that the above equation indicates that effect of the fluid is to add a certain amount of mass to the body (M11 to body 1), and an external force which is proportional to the acceleration of the adjacent body (mass m2)-

Thus, the acceleration of one body a.ffects the force field on another.

This force is a strong function of the interbody gep, reaching large values forms very small gaps. This inertial coupling is called fluid coupling. It has an important effect in rack dynamics. The lateral motion of a fuel assembly inside the storage location will encounter this effect. So will the motion of a rack adjacent to another rack. These effects are included in the equations of motion. The fluid coupling is between nodes i and i* (i 2,3.. 5) in Figure 6.1. Furthermore, nodal masses i = contain coupling terms which model the effect of fluid in the gaps between adjacent racks. Finally, fluid virtual mass is included in vertical direction vibration equations of the rack; virtual inertia is added to the 2~ governing equations corresponding to rotational degrees of freedom, e such as qq, q5e 96r 911, etc. (7 6.2.4 Damping In reality, damping of the rack motion arises from material t hysteresis '(material damping), relative intercomponent motion in structures (structural damping), and fluid drag effects (fluid damping). The fluid damping acts on the i and i* nodal masses. In d sg

a m,aximum of 44 structural damping is imposed on the analysis, a O elements of the rack structure during SSE seismic simulations. -d Actual structural damping values used in the analysis are provided in -Table 6.4. This is in accordance with the FSAR and NRC guidelines 7 Material and fluid damping are conservatively neglected. 6.2.5 Impact The fuel assembly nodes i* will impact the corresponding structural mass node i. To simulate this impact, 4 impact springs around each fuel assembly node are provided (see Figure 6.2 ). The fluid dampers are also provided in~ parallel with the springs. The spring constant of the impac. springs is assumed equal to the local stiffness of the vertical panel comput'ed by evaluating the p,eak deflection of a six inch diameter circular plate subject to a specified uniform pressure, and built in around the edge. The spring constant calculated in this manner should provide an upper bound on the local stiffnesses of the structure. 6.2.6 Assembly of the Dynamic Model The dynamic model of the rcck, rack base plus supports, and internal fuel assemblies, is modeled for the general three dimensional (3-D) motion simulation, by five lumped masses and inertial nodes for the rack, base, and supports, and by five lumped masses for the assemblage of fuel assemblies. To simulate the connectivity and the elasticity of the configuration, a total of 18 linear spring dampers, 20 nonlinear gap elements, and 16 nonlinear friction elements are used. A summary of spring-damper, gap, and friction elements with their connectivity and purpose is presented in Table 6.2. If we restrict the simulation model to two dimensions (one horizontal motion plus vertical

motion, for example) for the 0g) purposes of model clarification only, then a descriptive model of the simulated structure which includes all necessary spring, gap, y

6-8

and friction elements is shown in Figure 6.3. The beam springs, K' KB at each level, which represeni. a rack segment treated as A a structural beam,4 are located in Table 6.2 as linear springs 2, 3,. 6 ', 7, 10, 14, and 15. The extensional spring, K which e, simulates the lowest elastic motion of the rack in extension relative to the rack base, is given by linear spring 18 in Table 6.2. The remaining spring-dampers either have zero coefficients (fluid damping is neglected), or do not enter into the two-dimensional (2-D) motion shown in Figure 6.3. The rack mass and inertia, active in rack bending, is apportioned to the five levels of rack mass; the rack mass active for vertical motions is apportioned to location: 1 and 5 in the ratic 2 to 1. The mass and inertia of the rack base and the support legs is concentrated at node 1. The impacts between fuel assemblies and rack show up in the gap elements, having local stiffness K, in Figure 6.3. In Table I 6.2, these elements are gap elements 3, 4, 7, 8, 15, 16, 19 and

20. The support leg spring rates K6 are modelled by elements 9 and 10 in Table 6.2 for the 2-D case.

Note that the local K. To simulate elasticity of the concrete floor is included in 6 sliding potential, friction elements 2 plus 8 and 4 plus 6 (Table 6.2) are shown in Figure 6.3. The local spring rates Kg reflect the lateral elasticity of the support legs. Finally, the support rotational friction springs KR, reflect the rotational elasticity of the foundation. The nonlinearity of these springs (friction elements 9 plus 15 and 11 plus-13 in Table 6.2) reflects the edging limitation imposed on the base of the rack support legs. For the 3-D simulation, carried out in detail for this analysis, additional springs-and support elements (listed in Table 6.2), are included in the model.. Coupling between the two horizontal seismic motions is provided by th,e offset of the fuel assembly group centroid which causes the rotation of the entire rack. The potential exists for the assemblage to be supported on 1 6-9

l Table 6.2 Numbering System for Springs, Gap Elements, Friction Elements I. Spring Dampers (18 total) Number Node Location Description 1 1-2 X-Z rack shear spring 2 1-2 Y-2 rack shear l 3 1-2 Y-Z rack bending spring 4 1-2 X-2 rack bending 5 2-3 X-2 l rack shear 6 2-3 Y-Z 7 2-3 Y-2 O rack bending 8 2-3 X-Z 9 3-4 X-Z rack shear 10 3-4 Y-2 11 3-4 Y-Z rack bending 12 3-4 X-Z 13 4-5 X-2 rack shear 14 4-5 Y-Z 15 4-5 Y-Z rack bending 16 4-5 X-Z 17 1-5 Rack torsion spring 18 l-5 z rack extensional spring. 3 O 6-10

Table 6.2 (continued) O II. Nonlinear Springs (Gap ' Elements) (20 total) ~ Number Node Location Description 1 2,2* X rack / fuel assembly impact spring 2 2,2* X rack / fuel assembly impact 3 2,2* Y rack / fuel assembly impact 4 2,2* Y rack / fuel assembly impact 5 3,3* X rack / fuel assembly impact 6 3,3* - X" rack / fuel assembly impact 7 3,3* Y rack / fuel assembly impact 8 3,3* Y rack / fuel assemlby impact 9 Support S1 2 compression spring 10 Support S2 2 compression spring 11 Support S3 2 compression spring 12 Support S4 Z compression spring 13 4,4* X rack / fuel assembly impact spring 14 4,4* X rack / fuel assembly impact spring 15 4,4* Y rack / fuel assembly impact spring 16 4,4* Y rack / fuel assembly impact spring 17 5,5* X rack / fuel assembly impact spring 18 5,5* X rack / fuel assembly impact spring (g 19 5,5* Y rack / fuel assembly impact spring (_) 20 5,5* Y rack / fuel assembly impact spring III. Friction Elements {l6 total) Number Node Location Description 1 Support S1 X direction support friction 2 Support S1 Y direction friction 3 Support S2 X direction friction 4 Support S2 Y direction friction 5 Support S3 X direction friction 6 Support S3 Y direction friction 7 Support S4 X direction friction 8 Support S4 Y direction friction 9 S1 X Floor Moment 10 S1 Y Floor Moment 11 S2 X Floor Moment 12 S2 Y Floor Moment 13 S3 X Floor Moment 14 S3 Y Floor Moment 15 S4 X Floor Moment 16 S4 Y Floor Moment O 6-11

to 4 rack suppor,ts during any instant of a complex 3-D seismic event. All of these potentia.'J events may be simulated during a 3-D motion and have been observedl in the results. A brief description of the elastostatic model now follows. This detailed model is used to obtain overall beam stiffness formulae for the rack dynamic model, and to determine detailed stress distributions in the rack from a knowledge of the results of the time history analysis. 6.3 Stress Analysis 6.3.1 Stiffness Characteristics: The fuel rack is a multic' ell, folded-plate struc,ture which has what is colloquially called a " honey-comb" configuration. This type of construction is very similar to the so-called " stressed-skin" construction of ribs, spars, and cover f') plates which are widely used in aircraft construction. Techniques ud developed in the field of aircraft structural analysis are utilized herein to find the stresses and deformations in such structures. These methods have been thoroughly tested and their reliability has been documented in a number of publications.8-12 Figure 6.4 shows two cross-sections of the fuel rack which is modeled as a' rectangular network of plates interconnected along nodal lines shown as points in Figure 6.1. An arbitrary load with components

Fxi, F i, Fzi acts at an arbitrary elevation on y

one of the nodal lines. We find the displacements and stresses due to such a typicci load according to the stressed-skin model as follows. The torsional deformations are solved for by using the classical theory of torsion for multicelled, thin-walled, cross sections.13 O 6-12

) 1 (] The bending deformation is found by using the theory of shear flo>:12 wherein all axial, stresses are carried by the effective flanges (or stringers) formed by the intersections of the plates and' all transverse shears are carried by the plates modeled as shear panels. From a knowledge of-the shear flows, the bending and torsional deformations, it is possible to provide a set of influence functions or the' Yo11owing section properties for the fuel rack as a whole: (EI)eq Bending rigidity (in two places) = Torsional rigidity (GJ)eq = (AE)egs Extensional rigidity = Shear deformation coefficient k = s Such properties are used for the dynamic analysis of seismic loads and serve to establish values for the spring rates of the elastic beam elements representing each rack section. l 6.3.2 Combined Stresses and Corner Displacements The cross-sectional properties and the Timoshenko shear correction factor calculated in the previous section are fed into a dynamic analysis of the system shown in Figure 6.5, with a specified ground motion simulating earthquake loading. From the dynamic

analysis, the stress resultants (Fx, Fy, Fz, Mx, M

Mz) act as shown in Figure 6.6 are computed for a large y, At, 24t, etc., at a selected number of cross l number of times t = sections. The displacements (Ux, U Uz) at selected nodal y, ~ points on the z axis are also provided by the dynamic analysis as well as the rotations (O O x, y, e) of the cross sections at z I the nodes. Oo l [ 6-13

m I (y. Figure 6.7 shows a typical subdivision of the structure into elements, nodes, and sections. The stresses are calculated at all sections and the displacements at all four corners of the racks are calculated at P.hese elevations. Since the axial stress varies linearly over the cross section and achieves it's extreme values at one of the four corners of the

rack, the shear, stresses Anec.~to torsional loads (M )

achieve z their extreme values near the middle of each side. The shear stresses due to lateral forces (F Fy) will achieve their x, extreme values at the center of the cross section or at the middle of each side. Thus, candidates for the most critical point on any section will be the points labelled 1 through 9 in Figure 6.8. The expression for the combined stress and kinematic displacement f'or each of these points is written out. Similarly, the stresses in the support legs are evaluated. A validated Joseph Oat Corporation proprietary computer program "EGELAST"i computes the stresses at the candidate points at each level. It sorts out the most stressed location in space as well as time. The highest stress and maximum kinematic displacements are thus readily found. 6.4 Time Integration of the Equations of Motion Having assembled the structural model, the dynamic equations of motion corresponding to each degree of freedom can be written by using Newton's second law of motion; or using Lagrange's equation. For. example, the motion of node 2 in y-direction (governed by the generalized coordinate pg) is written as follows: t This code has been previously utilized in ' licensing of similar I, racks for Fermi II (Docket No. 50-341), Quad Cities I and II (Docket Nos. 50-254 and 265), and Rancho Seco (Docket No. 50-312). 6-14

The inertial mass,is: O-m22 + A211 + B211 where m22 is the' mass of node 2 for y-directional motion. A211 is the fluid coupling mass due to interaction with node 2*, and B211 is the fluid coupling mass due to interaction of node 2 with the reference frame (interaction between adjacent racks). Hence, Newton's law gives (m22 + A211 + B211)-99 + A212 P10 + B212 u = 09 where 09. represents all the beam spring and damper forces on node 2, and A212 is the cross term fluid coupling effect of node 2*; B212 is the cross term fluid coupling effect of the adjacent racks, and u represents ground motion. Let 99 " P9 - u 410 " P10 - u That is, q9 is the relative displacement of node 2 in x-direction with respect to the ground. Substituting in the above equation, and rearranging, we have: (m22 + A211 + B211) 99 + A212 910 = 09 - (m22 + 211 + A 12 + B 12) 'u' A211 + B 2 2 6-15

A similar equation for each one of the 32 degrees of freedom can be written.- The system of egarations can be represented in matrix notation as: [M] {q'} = [0] + {G} whe're the vector (0) is a function of nodal displacements and velocities, and {G} depends on the coupling inertia and the ground acceleration. Premultiplying above equation by (M ]-1 renders the resulting equations uncoupled in mass. We haves {ki} = (M]-1 (0) + [M]-1 {G} The generalized forca 0, which contains the effects 9 of all spring elements acting on node 2 in the " direction" of coordinate q9 (the relative displ,acement of node 2 in the y direction), can easily be obtained from a free body analysis of node 2. For example, in the 2-D model shown in Figure 6.3, contributions to 09 are obtained from the two shear springs of the rack structure, and the two impact springs which couple node 2* and node 2. Since each of these four spring elements contain couplings with other component deformations through the spring force-deformation relations, considerable static coupling of the complete set of equations results. The level of static coupling of the equations further increases when 3-D motions are considered due to the inclusion of rack torsion and general fuel assembly group centroid ef fect. For example, referring to Figure 6.3, and Table 6.1, a 2-D simulation introduces static coupling between coordinates 2,9 and 15 in the expression for 0 ; this coupling comes from the shear 9 springs simulating the rack elasticity which have constitutive relations of the form F =K (99 - 92) K (915 - 99) Further, the s s impact springs introduce two additional forces having constitutive equations of the form F = Ky (q9 - ql0) Of course, at any instant, these forces may be zero if the local gap is open. p The local gap depends on the current value of q9 - q10 b 6-16

It should be, noted that in the numerical simulations run to. verify structural integrity during a seismic event all elements of ~ the fuel assemblies are assumed to move in phase. This will t. provide maximum impact force level, and hence induce additional conservatism in the time history analysis. l This equation set is mass uncoupled, displacement coupled, and is ideally suited for numerical solution using the central difference scheme. The computer program named "DYNAHIS"t, developed by General Electric Company and further enhanced by Joseph Oat Corporation, performs this task in an efficient manner. Having determined the internal forces as a function of time, the computer program "EGELAST"i computes the detailed stress and displacement fields for the rack structure as described in,the preceding section. 6.5 Structural Acceptance Criteria There are two sets of criteria to be satisfied by the rack modules: (a) Kinematic Criterion: This criterion seeks to ensure that adjacent racks will not impact during SSE (condition E'14 ) assuming the lower bound value of the pool floor surface friction coefficient. It is further required that the factors of safety against tiltingl5 are met (1.5 for OBE, 1.1 for SSE). (b) Stress Limits (1) The stress limits of the ASME Code, Section III, Subsection NF, 1983 Edition were chosen to be met, since licensing of similar t These codes has been previously utilized ir] Quad racks for Fermi II (Docket No. 50-341), Cities I and II (Docket Nos. 50-254 and 265), and Rancho Seco (Docket No. 50-312). O 6-17

thic Codo providas the most consistent cot of limits for various stress types, and various loading conditions. The following loading combinations are applicable (ref. ~ 14, Sec. 3.8.4). . Load Combination Acceptance Limit D+L Level A service limits D + L + To D + L + To + E D + L + Ta + E Level B service limits D + L + To + Pg D + L + Ta + E ' Level D service limits D + L + Fd The functional capability of the fuel racks should be demonstrated where D= Dead weight induced stresses L= Live load induced stresses; in this case stresses these are developed during lifting. Fd: Force caused by the accidental drop of the heaviest l'oad from the maximum possible height Pf: Upward force on the racks caused by postulated stuck fuel assembly E: Operating Basis Earthquake E': Safe Shutdown Earthquake To: Differential temperature induced loads (normal or upset condition) Ta: Differential temperature induced loads (abnormal design conditions) The conditions Ta and To cause local thermal stresses to be produced. The worst situation will be obtained when an isolated storage location has a fuel assembly which is generating heat at the maximum postulated rate. The surrounding storage locations are assumed to contain no fuel.- The, heated water makes unobstructed contact with the inside of.the storage walls, thereby producing maximum possible temperature difference between the adjacent cells. The secondary stresses thus produced are limited to the body of the rack; that is, the support legs do not experience the secondary (thermal) stresses. f 6-18 l

(2) Basic Data: The following data on the physical properties of the rack material are obtained from the ASME Codes, Section III, appendices. Table 6.3 Physical Property Data

• Property Young's Yield Ultimate Modulus Strength Strength 9 2000F 92000F 92000F E

S Su y Value 28.3 x 106 25 KSI 71 KSI psi Section III Table Table Table Reference I-6.0 I-2.2 I-3.2

• Evaluated at 2000F.

This temperature is higher than the pool water bulk temperature under any of the loading conditions under consideration. Table 6.4 Support Material Data (Table 6.4 gives the equivalent data for the support materials.) Young's Modulus Yield Strength Material at 200*F at 200*F 6 1 ASTM 479-521800 27.5 x 10 50,000 psi 2 SA564-630 (hardened at 27.6 x 106 125,000 1075'F) (3.1) Normal and upset conditions (level A or les0. 1 B): (i) Allowable stress in tension on a net section =Ft =0.6 Sy or Ft =(0.6) (25000) =15000 psi (rack material) Ft is equivalent to primary membrane stresses Ft (.6) (50,000) = 30,000 psi for support feet) (ii) On the gross section, allowable stress in shear is Fy = 0.4 S (0.y) (25000) = 10000 psi (main 4 = rack body) i Ft= (.4)(50,000) = 20,000 psi for support feet) O 6-19

,(iii) Allowable stress in compression, Fa ( [1 - ( b) / 2C )8 c y F = ((5) + - (3 (S) 8C ) 8C ) c] - ((r c 3 r where 2 ,J (2x E) J s Y ouus u cu ung n u m v e s. 3, we

obtain, for both support leg and " honey-comb" region:

Fa = 15000 psi (main rack body) Pa = 30,000 psi (support legs) (iv) Maximum bending stress at the oute rmos t fiber due to flexure about one plane of symmetry: Pb = 0.60 Sy = 15000 psi (rack body) j Pa = 30,000 psi (support feet) l (v) Combined flexure and compression: f f f, C,x bx C,y by 4 y F, DF DF x bx y by where fa: Direct compressive stress in the section. f x: Maximum flexural stress along x-axis b by: Maximum flexural stress along y-axis f Cmx = Cmy = 0.85 6-20

a o 1, F',, fa D =1-Y ye ey where 2 12w E p. ex 2 kl 23 ( )

1. b (vi) Combined fldxure and compression (or tension) f E

f a bx by < l.0 0.6 S F P y bx by The above requirement should be met for both direct tension or compression case. 3 (3.2) Faulted Condition: F-1370 (Section III, Appendix F), states that the limits for the faulted condition are 1.2 (S /F ) times the corresponding limits for y t normal condition.

Thus, the multiplication factor is 000 (1.2)

= 2.0 Factor = 15000 O 6-21

6.6 Results O Figures 6.9, 6.10, and 6.11 show the pool slab motion in horizontal x, horizontal y, and vertical directions, respectively. These plots correspond to the Operating Basis Earthquake. The corresponding SSE time history motions are conservatively assumed to be 1.62 times the OBE values. Since there are several rack module configurations (Fig. 2.1) it was decided to make an exhaustive analysis of one rack type. We note that module A is an above-average size module, and hence will produce above-average floor reaction and support stress levels. Therefore, module A is chosca for performing extensive analyses. The support locations for module A are taken to be three cells in-bo,a rd of the edge of the rack along one side. This assumption will maximize rack displacement and rotation. Appropriate simulations are also carried out for other limiting rack /~(T geometrics (e.g. tipping study for rack with low cross / section to height aspect

ratio, stress evaluations for the heaviest module, etc.).

To determine the magnitude of structural dampers, free lateral vibration plots of the top of rack A (in X and Y directions) for fully loaded and empty conditions were developed. The dominant natural frequency of vibration thus evaluated enables computations of the linear structural dampers. The percentage structural damping for SSE condition is assumed to be 4% and modifications to the stiffness matrix to incorporate damping is based on the d dominant frequency of 10 cps. Having determined the damper characteristic data, the dynamic analysis of the rack module is performed using the computer program DYNAHIS. A complete synopsis of the analysi,s of mocule A subject to the safe shutdown earthquake motions is presented in the-photocopied computer print-outs labelled as Table 6.5. Table 6-22

m 6.5 gives ,the maximum values of stress factors (Ri O-1.2.3.4.5.e). 's 1ues 2 en in the tadies are the 1 (i maximum values in time and space (all sections of the rack). The various stress factors are listed below for convenience of reference. R: Ratio of tensile stress on a net section to its t allowable OBE value R: Ratio of gross shear on a net section to its allowable 2 OBE value R3: Ratio of net compressive stress to its allowable OBE value for the section R: Ratio of maximum bending stress in one plane to its 4 allowable value in OBE R: Combined flexure and compressive factor 5 R6: Combined flexure and tension (or compression) factor 1,2,3,4,5,6) is 1 for CBE The allowable value of Ri (i = O condition, and is 2 for SSE condition (see Section 6.5). The displacement and stress tables given herein are for the SSE condition. It is noted that the maximum displacements are less than the limiting value for inter-rack impact. The maximum stress factors (Ri) are well below limiting value for SSE condition for all sections. Five plots (Figs. 6.12 through 6.15) show the variation of support reactions with time. Figure 6.12 shows the variation of the total rack support reaction (sum of four feet) with time. Figures 6-13 through 6.15 show the reaction variation in individual feet. The plots vividly portray the module motion. We note that support feet never lose contact with the ground. However, the module 3 displacements are not infinitesimal. This is due to the in-board location of support feet assumed in the analysis. Subsequent analyses 6-23

S with support feet located at the outermost corner cell () locations produce sashstantially lower stresses and displacements. Seismic simulatioes for the tipping conditions are carried out by increasiatg the horizontal SSE accelerations by-50%.15 The calculatime:s indicate that the rack remains stable, and the gross movement remains within the limit of small motion theory. Thus the rack module is seen to satisfy both kinematic and stress criteria with large margins of safety. Analysis of welded joints in the rack also show large margins of safety. 6.7 Summary of Mechanical Analyses: The mathematical model constructed to determine the impact velocity of the above falling objects is based on several conservative assumptions, such as

1. The virtual mass of the body is conservatively assumed to be equal to its displaced fluid mass.

17 Evidence in the literature indicates that the virtual mass can be many times higher.

2. The minimum frontal area is used for evaluating drag coefficient.

3. The drag coefficients utilized in the analysis are l8 lower bound values reported in the literature In particular, at the beginning of the fall when the velocity of the body is small, the corresponding Reynolds number is low resulting in a large drag coefficient.

6-24

4. The,f alling bodies are assumed to be rigid for the A

purposes of impact stress calculation on the rack. V The solution of the immersed body motion problem is found analytically. The impact velocity thus computed is used to determine the maximum stress generated due to stress wave propagation. With this model, the following analyses are performed. (i) Dropped Fuel Accident I A fuel asncmbly (weight - 1616 pounds with control rod assembly) is dropped from 36 inches above the module which impacts the base. Local piercing of the base plate is not' found to occur. Direct impact with the pool liner does not occur. The subcriticality of the adjacent fuel assemblies is not violated. O (ii) Dropped Fuel Accident II One fuel assembly dropping from 36 inches above the rack and hitting the top of the rack. Permanent deformation of the rack is found to be limited to the top region such that the rack cross-sectional geometry at the level of the top of the active fuel (and below) is not altered. (iii) Jammed Fuel-Handling Equipment and Horizontal Force A 4400-pound uplift force and a 1100-pound horizontal force are applied at the top of the rack at the " weakest" storage location; the force is assumed to be applied on one wall of the storage cell boundary as an upward shear force. The damage 6-25

is found to be limited to the region above the top of the active fuel. These analyses prove that the rack modules are engineered to provide maximum safety against all postulated abnormal and accident conditions. O O O e O 6-26

-() REFERENCES TO SECT' ION 6 1. .USNRC Regulatory Guide 1.29, " Seismic Design Classification," Rev.-3, 1978. 2. " Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," by Prof. Ernest Rabinowicz, MIT, a report for Boston Edison Company, 1976. 3. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis," Rev. 1, February 1976. 4. "The Component Element Method in Dynamics ~with Application to Earthquake and Vehicle Engineering" by S. Levy and J.P.D. Wilkinson, McGraw Hill, 1976. 5. " Dynamics of Structures" R.W. Clough & J. Penzien, McGraw Hill (1975). 6. R.J. Fritz, "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. 7. USNRC Regulatory Guide 1.61, Damping Values for Seismic (} Design of Nuclear Power Plants, 1973. 8. J.T. Oden, " Mechanics of Elastic Structures," McGraw Hill, N.Y., 1967. ~ 9. R.M. Rivello, " Theory and Analysis of Flight Structures," McGraw-Hill, N.Y., 1969. 10. M.F. Rubinstein, " Matrix Computer Analysis of Structures," Prentice-Hall, Englewood Cliffs, N.J., 1966.

11..

J.S. Przemienicki, " Theory of Matrix Structural Analysis," McGraw-Hill, N.Y., 1966. 12. P. Kuhn, " Stresses in Aircraft and Shell Structures," McGraw-Hill, N.Y., 1956. 13. S.P. Timoshenko and J.N. Goodier, " Theory of Elasticity," McGraw-Hill, N.Y., 1970, Chapter 10. 14. U.S. Nuclear Regulatory Commission, Standard Review Plan, NUREG-0800 (1981). U.S. Nuclear Regulatory Commission, Sta'ndard Review Plan, 15.. Section 3.8.5, Rev. 1, 1981. O 6-27

16. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.124, O " Design Limits and Loading Combinations for Class 1 Linear-Type Component Supports, November 1976.

17..

" Flow Induced Vibration" by R.D. Blevins, VonNostrant (1977). 18. " Fluid Mechanics" by M.C. Potter and J.F. Foss, Ronald press,

p. 459 (1975).

e O e V 6-28

A / TABLE 6.5 FILE DSCLO1 MODULE A (11 X 11) COEF.= .8, FULL RACK i, 9 j '1 WED, DEC 14',~1983, 11:45 AM' ~~~ PROGRAM EGELAST' ~ ~] , ' ~ ~- ~ ~ ~ ^~ ~ ~ ~ ~~ ~ 1 Y SCL A2,C0Fn.8,30s4) e10HZ,DATAzDSCL01,11X11 FULL RACK SSE QUAKE l ',. INPUT.. PAPA 4ETERS If NO. OF NODES ( Nil 9 N O D) = 5 fi No. OF Et,EMENTS(NUMEL)s 12 3 PRINT _ OPTION.(10P_T)___m._0 X-MALFw!DTH (A2) a 5.713E+01 Y-HALFWIOTH (h7) a 5.713E+01 FLEN. E?Lfi44FTH2fPZ) = 2 L10J+01 2d O STOESS CO,,FFICIF%TS FOR RACK 1 3 CfX._=_9.900E-03___CFYs P,900F-01 (FZ m 2.970E_-0) u CMX = 1.406e-04 CMY = 1.406E-04 CTX s 2.702C-04 CTY a 2.702E-04 m. O STRFSS__C9(FFICIfNI4 FOR SUPPORTS [ CF12 9.990F-02 CFY2 = 9.990F-02 CF7,2 = 4.440E-02 C=X2 = 1.693F-02 CMY2 a 1.693E-02 CTX2 = 0.000E+00 CTY2a 0.000E+00 c'! ~ ~-*'" m {j f 0 STRESS COEFFICIENTS FOR SUPP0HT BOTTOM a O STRESS COEFFICIE'NTS FOR SifPPORTS i CFX2 = 1.791E-01 CFY2 a 1.791E-01 CFZ2 m 7 96_0E-02 t CwX2 a 1.590E-01 CMY2 = 1.590E-01 CTX2 = 0.000E+00 CTY2= 0.000E+00 n i 2p___0 . STATIC STRFSS COEFFICIENTS. 0.C00E+00 0.000F+00 0.000E+00 0.000E+00 0.000E+00 CCtP,CCX".CCYP,CCYM.CXSH CYSH= 0.000E+00 FTT, FAT,FVT,Fet= 19000.0 18u00.0 12000.0 1R000.0 2-FT0aff0,FV0,FBot 30000,0 300no.0 20000.0 10000.0 rj FTP.,FAP,FVA.FPPs 75000.0 75000.0 55000.0 75000.0 SECTION h0, OF dO7T (JROOT) = 5 TOTAL NO,_UL SECTIONS (NUMSE_C)=_13_. a 1 TABLE OF MAXIMAX EQUIVAI.ENT STRESS ?. _E_fCTa_.__T_IMg_ POINT w A X,

SEO, DJR.STRFSS X-PENO SFR Y-RENO STRESYLATg MEAR YLAT.SHFAR NET SHEAR) 0 NO.

NO. NO. (SEMYMX) (50) (SPX) (39Y) (TC) (TY) (T6) i p* _ _ 1_.____404__ 6.__ 2.19AE+03 - 6. 6 6 5 E + 0 0____1. 6 2 7 E + 01 -7.209E+01 -2.093E+02 -4,91Rr+01 -1.0995'+03 2 404 8 4.4R1F+03 -1.333F+01 7.916E+01 -3.005E+02 -4.602E+02 -1.352E+02 -2.241E+03

• d 3

404 8 6.552F+03 -1.949F+01 2.117E+02 -6.603E+07 -6.026R+02 -2.603E+02 -3.274F+03 4 401 8 9.506V+03 -2.666E+01 4,196E+02 -1.045E+03 - 6,e% 7'.t+02 -3.544R+02 -4.247M+03

• (_

-6.as7'+02 -3.549E+n2 -4.247E+03 5 404 8 8.513E+03 -2.666E+01 5.394F+02 -1.325E+03 'l 6 174 8 3.15dE+04 -1.447E+04 1.010E+os 1.05tE+04 9.926t+03 -9.2dRE+03 9.926E+03 7 396 2 3,524F+04 -1.31$f+04 -1.297Efo.4 1,12pr+04 1.0687+04 1.213E+0* -1.610E+04 8 746 1 3.556E+04 -1.256E+04 -1.357E+04 -1.047E+04 -1.029E+04 1.271E+04 1.626E+04 "{ 9 453 4 4.494E+04 -1.392V+04 1.832E+04 -1.281E+04 -1.202r+04 -1.716E+04 -2.063E+04 M to _7, _7 4 __,J (602E+04_ -2,593E+04 -2,426F103 4.R42 Q01_ ~ ~ 1,7 R 0l*+ 0 4 1,66%5;+04 2.415E+0,4 11 396 4 6.2h3E+04 -2.35NE+04 1.11bE+03 -5.179E+01 -1,007R+04 -2.174E+04 -2.896E+04 12 746 3 6.262E+04 -2.251E+04 1.101E+03 6.012E+02 1.94St+04 -2.27AE+04 2.915E+04

p TABLE s ( inued) 007 Clot.DATARACK. ENGR THU, DEC 15, 1083, 3:24 P4 PACE 10 _.1 1 _. 4 5 3 _ _._'L_ I. P J 9 E + 0 4 -2.49.6n+04 -1.12_0D 0 3 7.929r+02 2.154c+04 3.077E+04 -3.699E+04 II 1 00 N r#IT IO N S wuEM x-DtSRLACEMENT OT A CORNF.R ISMAXI4AX j NDDE TJ K_F_ _9 A X. C D P N ER_IE.N_T*D i nil __CJN TR o tp A [:,,, TQR$_IQNAL I NO. X-D1SP. X-0ISP. Y-DISP. ANGLE

1. _7 4 L __E. 611 E-01_-1.J 4 68:-01

5. 91_R g-01

-2.600.E-03 2 745 7.142F-01 -5.657E-01 4.47pE-01 -2.600E-03 L 3 745 5.618E-01 -4.161E-01 3.930E-01 -2.550E-01 I ', 4 ___7 4 4___.4.13 2 E-01 - 2. _6 5 9 E - 0.1 3.757E-01 -2,5785 03 &jC 5 697 2.724E-01 -1.147E-01 2.740E-01 -2.760E-03 l h* J ].'j cDNDITlGNE_aSEN Y-DIEEk& CEMENT AT CORNER 15 MAXIM &X ~ l'l 40cc TIME MAX. CORNER CENTRotDAL CENTROID 4L TORSIONAL ft-ND-Y -D I S P.. X-RISP-Y _0 !EP_, AM#ki l' 1 784 9.312r-01 -4.19aE-01 -8.107E-01 -2.110E-03 f, 2 a19 7.463E-01 -1.197r-01 6.972E-01 -8,5u'E-94 1_.__B20.._6.272E-01 _-1.715f.91 5.7P7h01 -8.4955-94 f[]1 4 755 5.254E-01 -1.230E-01 3.962E-01 -2.314E-03 5 754 4.39tE-01 -8.450E-02 3.0448-01 -2.366E-01 [i STRUCTUPA5 'C'CEPTANCF=ASVC KF SFCTION r.U=BER 1 cn I RI= _.001AT_ TIME 10C5__R2=.. 01247 Tl"M___117 R3= 00!LAT TIME 757 ~ ~ ~ ~ ' ~ ~ ~ ~ ~ (. R4= 006AT TIME 763 R$= 0104T TIME 756 R6= .011Af TIME 756 SECTIO *! Niih B FR 2 co { i ??

1. 1 =__ _. 0 0 2 A T. T i w F_,j o R $ _,,9 2 =,_ 0J O A T_71 M E 756 R3=

037A1_TI2_E.757 t g R4= 026AT T1*E 763 R$= 041AT TIVE 756 R6= .0444T TIME 756 SFCTION 7:U p p F P 3 i< R1=_.003AT_TIH5.10RS..P2=__.042AT_ TIM 6_ 156 __R3= a006AT Time 757 ~ l] R4= 059AT TIME 764 R5= 090AT TIMF 756 R6= .106AT TIME 756 ~~- ~ SECTION ht!waFR 4 M a.1 = __4 0 0 d A I.T I " E.19 5 5__9 2 = 0 4 7 4 I..T TF E 73.6 P3= .1.174t_ITME 757-94= .100AT T!*E 764 as: .1504T TIME 39" R6= 176AT TIME 397 %fCT10u humdER 5 91=_ 0044T TIwC.1095__R2=__.997AT_ TIME __756.__.83=___ 1794I_TIut 757 N R4= .122AT TTWE 706 R5= .182AT TIME 397 R6= .214AT TIWE 397 ~~~

• l SFCTION NUM9ER 6

R1= .4Y7AT TIME _775._32=_.24R AT Tite 17 4.__8L3 =. 3.4 2 4 T_T I M E 779 R4= 350AT TI"E 774 95= .895AT TIME 774 R6= 9684T Tide 774 SECTIon NUd3EP 7

• _.01

_.449AT II*iE. 397...R2=__.364AT_T.IMU__183__B3=___.47847_.T140 _156 i] R4= .51447 Tide 383 R5= 925AT TIME 196 R6= 1.0114f TIME 396

1. 1 SFCTION NU40ER 8

R1=__. 414AT T I w S _,,7 4 6 J ? =_,. 3 2 R A T_T I w f,_7 5 7 D)= . 4 514 T _T I M F._. 7 57 E1= .3953T TIME 707 PS 913AT 7IME 746 k6= 1.00047 TIwE 746 SECTION NU"9FP 9 f-it1=__.464AT TIME __453. B2=__.42947_TIMF. _453__B3= . 611 B T. T.T.MI _4 b 3 l"; a4= 429AT TIwF 152 R5= 1.00047 TIPE 453 R6= 1.209AT TIME 453 h"j' _ _._R 1 =... '. 3 4 9 A T SECTION NU4PER 10 TIvr. 775__F2=_.16247__.TT.MF_,,?74_){J= _.03 RAT TJ4r 77R r l"; R4= 033AT TIwE 666 R5: .378AT TIMF 775 R6= .3834T TTME 775 s SFCTION NU"8ER 11 D L

O "z "a 3 s g / / C O U PLIN G ELEMENTS 44 TYPICAL FU EL ASS EM B LY 3 GROUP M AS S H TYPIC AL FUEL RACK MAS S FU EL R AC K B A S E O 2'N2 )~ AY a p / \\ / Ax 1 l Y ly e 1, Y y-h ;3 XB /-+-% i t i i j a I I h ,), S FUEL R ACK SUPPORT ~ \\ x XB, YB - LOCATION OF CE N TROID O F FU EL ROD GROUP M ASSES - REL ATIVE TO Q CENTER OF FU E L R A C K ni = UNIT VECTORS FIG. 6.1 D Y N A M I C MODEL 6-29

O t IMPACT J L SPRIN G S I? l-T [.H MASS 5* + d] } t F LUID DAM PER S RIGID FRAME X = O F I G. 6.2 I M PA C T S P R I N G S AND F LUI D D A M P E R.S. 30

5 W M $. K Uyp.l N g 4 ~ Seismic l b WM K Motions E k 4 Z N d Fuel Assembly W Group Lumped Mass y Urp.) 3 l H WWW + ~At Rack Lumped. 3 Mass & Inertia Fe,r Horizontal Motions Ryp.) 9 g + 2 , K Uyp.) g k n K g h 3 K 11f f Mk- ,a & A x, ~ i 4, h Figure 6.3 Spring Mass Simulation For Two Dimensional Motion 6-31 j:

~ O ~ i j'FY Yd B-B L =Fh ~<. "X ( a) TOP VIEW Zo a I F a yz 9 I = Fh ( 3) AXIAL CROSS SECTlON ( B-B ) mu my ,9 ^ FIG. 6.4 (a) HORIZONTAL CROSS 7. SECTION OF RACK (b) VERTICAL CROSS SECTION +" OF RACK /, / 6-32

CELL z(W) Q~ - WALLS l [C '^^EA t / ^ ^x o o /, o's'i /^ ^^^^'C -T ' ?,-(%,(%'/(W i b= NyC .a ~ a:NxC7 C A' y(V) A RIGID PLATE B /

x X (u) 9 l

P77 f ^~ A _=6 A =-d' O l i SUPPORTS FI G. 6.5 oynamic Modei hMz M /y d p s. C' A' / Fx o B C A 8 a O F G. 6.6 s tress nesuit ant s o r i e n t a t i o n 6-33 i

O. hZ NODEI / g E L.I S E C. l -- / NOD E 2- - E L,2 S E C. 2 -- - NODE 3 - E L,3 S E C,3 - NO DE 4 - Pi = - E L,4 [- y EL5 S E C.4 --.- / ,/ ,. rz / 'X e-d S E C,6 S E C. 5 - + - 4'g, N O D E 5 k- ' d--ef g -EL,8 EL,7-ROOT OF RACK ~ S E C, 8 S E C,9 N O, O F E L E M E N T S = 8 N O. 0F S EC TI ON S =9 N'O. 0 F N O'D E S =5 i O = G. 6.7 SUBDIVISION OF A TYPIC AL R ACK 1 6-34

Y t G e G ()~ Y b @X o g 0 = = F G. 6.8 FINITE ELEMENTS MODEL CROSS-SECTION O 6-35

11\\l1ll1l1ll ll 1 0 O. 0 '6 i 00 '4 i l 0 0 2 '1 l 0 o l 0 o 0 i 'l P I ) I C E le S ( 0 9 u e 0 O F i 'B.E 6 1 f M I G r 1 I T F e X ml I 0 0 m a ) '6 t u n 1 ( S o z 0 .C r l 0 i '4 o .V H l 0 0 '2 0 0 ~ ~ 0 O o ;- .oo qo 0 nB zO agc mE 1l1 l

o O. c.s 's l g 0 0 ' s. 4 1 oc a 'l t l o o c o '1o P l 0 J 1 e 0 u Y i 0 6 O F a G l F I l r a e t i o m n c ' s. m o i u z \\ f S i r \\ 0 )\\ o 0 C H 1 '4 V l oc l 'a l M Sp ~ oc o 8 o 8.i O "g z ni US

• a~

ll 0 O 0 '6 1 0 0 '4 i 0 0 2 'l l l 0 o 0 o I '1 O P ) f C l i E 1 S 1 e 0 ( 6 0 O u '8.E F M G I I F T r e 0 m l 0 '6 m a u c S i j t 0 r 0 C e l '4 V V 0 0 '2 0 0 ~ ~ 8 o

i e.

0 O Bz j0E a 9 l 1

I l ~ O O 0 0 4 l s 05 i 3 I: K I L 0 O 6 / 3 O P L I 1 0 E h E S U S F S h 2 R1 i .EE6 O O bM MG S 2 I N C TMF I E O S U I S T 5 C 1 0 h A b = i 3 E C R D O T I R R E 0 V O P '0 i P P L U AT S O T F h O S M U 0 S 0 0 9I 1i f O j dr i ,'O l [ l lli

1 1 O 00 [_ 4 l y A 0k3 j 1 1 l {\\ 0 h3 n 1 0 S 5 L D 2 O f N O O P C s E .E L 0 S 6M E O 2 I U 5 T F t 1 / E R S = 'E 0 M S i 5 D p M 1 O U I S R A i E G P 0 C 6 E r 1 L L A V T T p O R w O T b 3 P S 1 P 6 U G S IF 0 0 0

• .?

.( _ i O O 3s 5; w 2

I O O O SUPPORT LEG 2 - SSE - TOTAL PERIOD = 15 SEC. i ~ E-m j I h >"V\\P V L 5 f L 3 J \\ ) d k ~. i gi j r W E j i k I_ i I 'O.000 Sb. 100. 150. 200. 250. 300. 350. 400. TIME FIG. 6.14 .V. C. SUMMER FUEL POOL 1

-y 0 O O 4 i SUPPORT LEG 3 - SSE TOTAL PERIOD = 15 S$C. o l ? 1 1 1 k i l n i a i 1 b 3d I I I l g?- f i 5 j A Yd 4 I i j 07 j d -d i g l y I i i E ^ l '0.000 5b. 160. 150. 2'00. 250. 300. 350. 400. l TIME FIG. 6.15 V. C. SUMMER FUEL POOL m

t O i s i f l y C g { w 4 i N g a, 6 B O ~ O l H o g A _J N o W O Q. CL l L1J d O. 6 -o Z W l-- i N HD o I H 11. I TW .o Z W 2 E cn D c/) l ~ CD i d O co m .= _J k i-a a e e O 1 cf G. iE D o gn o o. 9'E-9'd-9'F-9'd-9'N-9'l-9'8 o i O a ,0TM (81) NOI13V38 6-43

s I 7. SPENT FUEL POOL STRUCTURAL ANALYSIS O v.1 1 erea ettee The high density rack mo&ules for long term fuel storage ~ described in Sections 2 and 3 are located in the Spent Fuel Pool of the Fuel Handling Building shown on Figure 7.1. The Spent Fuel Pool structure is a, reinforced concrete structure supported on caissons down to competent rock and is integrated with the remainder of the building. Figure 7.2 shows a plan of caisson layout, and Figures 7.3-7.4 show cross sections through the structure. The pool walls and slab are 6'-0" thick and the caissons are J'-0" and 4'-0" in diameter. It has been demonstrated that the existing Spent Fuel Pool structure and foundations maintain their structural integrity for p all postulated loading conditions for the new high density racks. In particular, the requirements of NUREG-0800, Standard Review Plan Section 3.8.4 and the FSAR have been met. 7.2 Analysis Methods The structural analysis methods used for the existing Spent Fuel Pool are described in the PSAR Section 3.8 and the dynamic analysis is in Section 3.7. Investigation of the existing design and analysis indicates that reasonable margins existed in the walls, slab, and caissons. This condition allows a simplified conservative approach to be taken for analyzing the structure for increased loads from the new racks and spent fuel. The models used for the analysis were idealized to produce upper bound results for tension and compression conditions in the structure at any given location. The walls and slab were analyzed as a continuous two-dimensional frame supported by the caissons and surrounding _ J 7-1

integral concretq floors. The design of the caissons was I re-evaluated to include the effects of the additional loads from the new racks and spent fuel. 7.3' Assumptions 1. The loading used to qualify the pool structure and caissons assumes that all racks are fully loaded and that the loads are evenly distributed. 2. All caisson supports to the pool slab are considered pinned for analyzing the slab and walls;

however, moments in the caissons from horizontal forces are considered in evaluating ti e caisson design.

h 3. The seismic forces from the racks are represented by equivalent static loads on the pool slab. An equivalent static load equal to two times the weight of the racks is used for. seismic vertical loads for the Safe Shutdown Earthquake (SSE) condition. 4. The Maximum Operating Temperature (To) of spent fuel pool is assumed to be 150'. In Section 5, the maximum water temperature is shown to be limited to 137'F (Fig. 5.1.2(a), p. 5-21) for full cc re off-load condition. Even in the normal condition of one (out of two) spent fuel pool cooling loops in operating condition, the maximum temperature is limited to 140'F (Fig. 5.1.3(a),

p. 5-25).

7.4 Load Combinations In compliance with USNRC Standard Review Plan section 3.8.4 the following load combinations were evaluated for the reinforced concrete Spent Fuel Pool structures: A 7-2

U = 1. (D + 1.7L + 1.3E () U= (0.75) (1.4D + 1.7L + 1.9E + 1.7To) l U = D + L + To f E U = D + L + Ta + 1.2SE U= (0.75) (1.4D + 1.7L + 1.7To) U = D + L + Ta Where D = Dead Load (Including Hydrostatic) L = Live load E = Operating Basis Earthquake l E = Safe Shutdown Earthquake To = Operating Temperature Ta = Accident Temperature Note: The spent fuel pool cooling sy' stem is seismic category I which allows accident temperature (Ta) to be replaced by Operating temperature To in accordance with s Standard Review Plan 9.1.3. 7.5 Results 1. Caisson The results of the caisson evaluation are shown on Table 7.1. The capacity of the caisson is dependent upon the interaction curve of Bending Moments and Axial Forces. T,he results show that all the affected caissons have suf ficient capacity to sustain the loading for the new rack conditions with further margin available. 2. Walls and Slab The results of the wall and slab evaluation are shown on Table 7.2. The capacity of the walls and slab is 7-3 L

I j dependent upon the Interaction curve of Bending Moments ( and Membrane Forces. The results show that the walls and slab have sufficient capacity to sustain the loading from the new rack conditions. Considerable reduction in bending due to thermal gradient is possible if a cracked-section analysis is performed. However, except in the case of slab north / south with Bending causing tension in the bottom (total bending 352 Kip ft), it was not necessary to utilize reduction due to a cracked j section in order to show sufficient capacity exists. l 4 7.6 Conclusions Review of the rack seismic analysis results in Section 6 shows that assumption 3 in Section 7.3 of this report.was conservative. Assumption 4, that the maximum operating temperature To will not exceed 150*F, is confirmed by the cooling analysis result of Section 5. The results of the circulation of the Spent Fuel Pool walls, slab and caissons show that adequate margin exists in the structure to meet the requirements of NUREG-0800, Standard Review Plan Section 3.8.4 and FSAR requirements for the high density rack modules. I 7-4 ____._______________________________J

O O O Table 7.1 Celsson Evolvation - Required vs. Minimum Avellebte Capecity Atfected Celsson REQUIRED CAPACITY AVAILABLE CAPACITY Allowable bending en Caisson # Dia.(FT) Compression (KlPS) Moment (KIP-FT) Compression (KIPS) Moment (KIP-FT) Rock Sockets (KIPS) 7 4.0 2525 939 2699 1600 3817 8 4.0 2508 939 2699 1600 3817 i 9 4.0 2400 939 2699 1600 3230 l y 10 4.0 2591 939 2699 1600 32b l l US 16 4.0 1160 939 2699 1600 3f30 l l 17 3.0 976 657 1615 675 207 I 18 4.0 1804 939 2547 1755 380 19 3.0 1044 657 1615 675 2257 20 3.0 951 657 1615 675 2257 21 4.0 1764 939 2547 1755 3817 's In all cases the governing loed combination Is 1.40 + 1.7L + 1.9E

q O O O Table 7.2i Required vs. Avelleblo Cepecity in Spent Fugl Pool Wells and Sielp REQUIRED CAPACITY AVAILAGl.E CWACITY l l Bending Membrane Shear Bending Memlrone Sheer l WORSE CASE LOCATION GO'ERNING LOAD COMBINATION (KIP-FT) (KIPS) (KIPS) (KIP-FT) (KJfS) (KIPS) l WALLS Bending Tension on Outside Face (0.75)(1.4D + 1.7L + 1.7T ) 913 53.6(Comp) 930 1669(Comp) O Bending Tension on inside Face 1.4D + 1.7L + 1.9E 304 23.8(Comp) 449 166$tComp) (0.75) (1.4D + l.7L+1.9E+1.7T ) 45.5 77.4 I l y C m SLAB-NORTH / SOUTH l Bending Tension on Bottom (0.75)(1.4D + 1.7L+ 1.9E + 1.7T ) 352 42.1(Tens) 67.8 899 168(Tens) 74.8 O l Bending Tension on Top 1.4D + 1.7L + 1.9E 200 26.6(Tens) 444 168(Tens) SLAB-EAST / WEST l 1176 1665(Comp) l Bending Tension on Bottom (0.75)(1.4D + 1.7L + 1.9E+ 1.7T ) 988 617.2(Comp) C Bending Tension on Top 1.40 + 1.7L + 1.9E 171 3.9(Tens) 61.0 463 168(Tens) 74.6 I i Only limiting values of shears and moments are reported. l l l

) PLANT NORTH I i l= 123'-3" c m Il I -G l l l l 3 _r.______.3.__________.r.__________.i-______--.o n I i 1 i W I_________.I_________I__________ I,_, -c... 7_ f,_ _ _ _ _ _ _ _ _ _ _I x. l l.________l,_______.!,.__________.l,_I,,. .1 l l __________.,___________4 ______________.________...._____..,__________.y, 2 9'. o .________7, W fN /,)'rt_. g. prr I _ A /. in g I. 'l r- .Y["~ $PEMr ' a. Poet. Veh-I.' hi .I \\ lt >, 'mi e i l/ \\ / \\ 4e \\ l ' m g a -- t,-----_- _________.l___________.,.____ _____... _ _ _ _ _. _ _ L _.;

n e s

- ;- -- -- - --- 4-- -- - - - - -l-- -- - - - - - - -/:_ _ _ _ _ _ _ _ _ _ _3.. _ _ _ _ _ _ _ _ _ _.: o 1,. _ _ _ _ _ _ _ _ _.y _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l i n _____..e a,--------- L w.,__._,_ _ _i j '~^t.__ 7 ' NJ STEEL y-FRANDG f v REACTOR a BUILDING PLAbi El 463' FUEL HAhlDLING BUILDING ) PLAN ON SPENT FUEL POOL FIGURE 7_1 7_7

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Lj :Ii ' ' COMPETENT il:: u tp g 3 SECTION 2-2 D I i SECTION LODKING bj0RTH THROUGH SPENT FUEL POOL. FIGURE 7_4 7-10

8.0 ENVIRONMENTAL EVALUATION '8.1 Summary Installation of High Density Spent Fuel Storage Racks at V.C. Summer Nuclear Station (VCSNS) will increase the licensed storage capacity of the spent fuel from 682 to a maximum of 1276 assemblies. Radiological consequences of expanding the capacity have been evaluated

• with the ebjective of. determining if there is significant additional on site or off site radiological impact relative to that previously reviewed and evaluatedl. In addition, radiological impact to operating personnel has been evaluated to ensure that exposures remain As Low As Is Reasonably Achievable (ALARA).

e, Tne decay heat loading and the radiological burden to the spent fuel pool water are determined almost entirely by refueling /" operations. The frequency of refueling operations and the conduct of refueling are independent of the increased capacity of the storage pool, except that the increased capacity will reduce fuel movement and allow continued normal operation. Since the fuel assemblies which will utilize the bulk of the storage capacity (and will ultimately fill all incremental capacity above that of the existing design) are aged, their contribution to either the peak decay-heat load or the increased radiological impact, in terms of increased dose,.is negligibly small. A study 2 performed by the NRC supports this conclusion. Consequently, the increase in the storage capacity of the spent fuel pool will neither significantly alter the operating characteristics of the current pool nor result in a measurable change in impact on the environment. 8-1

8.2 Characteristics of Stored Fuel O Because of radioactive decay, the heat generation rate and the intensity of gamma radiation from the spent fuel assemblies decreases substantially with cooling time. After a cooling time of about 4 years,3 the decay heat generation rate is less than 2% of the rate at 7 days--the nominal time at which depleted fuel assemblies are transferred to the spent fuel pool. The intensity of gamma radiation is very nearly proportional to the decay heat and decreases with cooling time in a similar' manner. The bulk of heat loading is due to freshly discharged fuel; moreover, aged fuel contributes very little to the total heat load. Therefore, it is not expected that this expansion will significantly increase the therdial dissipation to the environment. Since the intensity of gamma radiation follows the decline in decay heat generation rate, it is similarly concluded l that there would be no significant increase in gamma radiation due to the expanded storage. t It is important to note that the aged fuel in the expanded storage capacity will not contain cignificant amounts of radioactive iodine or short-lived gaseous fission products, since these would have decayed during the refueling period. The Krypton-85 which might escape from defective fuel assemblies has 2 been shown to do so quickly (i.e. within a short time after discharge from the core). Further, the residual Krypton-85 will l be contained within the fuel pellet matrix and hence any leakage would occur at very low rates 2 Cesium 134/137 2, is strongly bound within the fuel pellet matrix and its dissolution rate in water is extremely small. Any cesium dissolved in the pool water is easily controllable in the clean up system (demineralizer-ion exchanger resin bed)2 Thus the planned storage expansion will not significantly increase the release of gaseous radionuclides. O 8-2

} 8.3 Related Industry Experietce c l q* Experience with storing spent fuel underwater has been substantial 2,3,s. These references show that the pool water activity, normally low, during refueling periods experiences a small increase which decays rapidly with time. Typical concentrations" of radionuclides in spent fuel pool water range from 10-4pCi/ml, with the higher value associated with refueling operations. References 2 and 4 also state that the increase in pool water activity during refueling can be attributed to:

a. dislodging (sloughing off) of corrosion products on the fuel assembly during transfer and handling operations

b. the possible short-term exposure of fuel pellets to pool water via'a cladding defect, and tC

c. mixing of the spent fuel pool water with the higher activity reactor coolant.

Upon cessation of the refueling operations the fuel pool water and the reactor coolant system would be isolated from each

other, thereby terminating transport of corrosion products from the Reactor Coolant System.

Thus, deposition of crud is a

function of refueling operations and is not impacted by the expanded storage. Furthermore, it has been shown6 that release of fission products from failed fuel decreases rapidly after shutdown to essentially negligible levels. The fuel pellets are made of inert UO2 that have very low solubility in water and the propensity for corrosion of the cladding (Zircaloy 2) at speqt fuel ' pool water temperatures is virtually nil 2,4 Thus the only mechanism available for the release of the gaseous ,( 8-3

fission products is dif fusion through the UO2 Pellet. It has been shown that at low water temperatures (<l50*F) the diffusion coefficient is extremely 7 small. Therefore, the small increase in activity of the spent fuel pool water is due to either crud transport, fission products release, or cross flow from the reactor coolant system and is only a function of refueling operations. It is reasonable to assume that the increased capacity of the spent fuel pool will reduce fuel handling operations, thereby reducing the probability of increased pool water activity due to crud dislodging.

Thus, the expansion of fuel pool storage capacity will not cause a significant increase in dose either on site or off site.

The corrosion properties of irradiated Zircalloy-5 cladding have been reviewed in References 2 and 5, and the conclusion is drawn that the corrosion of the cladding in spent fuel pool water is negligibly small. The minor incremental heating of pool water, due to the expansion of storage capacity, is far too small to materially affect the corrosion properties of Zircalloy-2 cladding. 8.4 V.C. Summer Operating Experience At present there are no stored fuel assemblies in the V.C. Summer fuel pool. 8.5 Spent Fuel Pool Cooling and Cleanup System (FPCC) It has been shown previously (Section 5 and V.C. Summer FSAR that the cooling system at the V.C. Summer Nuclear Station is adequate to handle the expected heat loads and maintain the temperature peaks within acceptable limits. It has been shown in 8-4 \\ ..._._J

Section 5 that the,small increa.no in heat load due to the storage O cagacitr o n ie-wi11. it io ificantiv 1 cre se the tliermal dissipation to the environment nor increase the propensity for corrosion of the cladding. It has also been shown that the crud deposition in the spent fuel pool water occurs during refueling outages and that the planned expansion will not increase crud deposition. The fuel pool clean-up system (filter and demineralizer) is designed to maintain fuel pool water clarity and is operated and maintained in accordance with V.C. Summer Nuclear Station operation procedures. The clean-up system takes a surface skim from the fuel pool and cleans it through a process of filtration and demineralization to prevent crud build-up on the fuel pool walls at the water-to-air interface. The spent fuel pool water is samp1'ed and analyzed periodically to confirm proper operation of the pool clean-up s system. The frequency of filter and resin replacement is determined primarily by requirements for water clarity rather than the loading of fission products radionuclides. The fuel peol demineralizer contains 54 cubic feet of mixed bed bead type resin. The cation resin is a strongly acidic, highly cross

linked, sulfonated styrene-divinyl benzene copolymer.

The anion resin is a strongly basic, quaternary ammonium-poly (styrene-divinyl benzene) resin. The demineralizer discharge piping is instrumented with a conductivity monitor which sounds an alarm in the control room on high conductivity. The fuel pool filter cartridge consists of epoxy impregnated cellulose fiber media and stainless steel hardware, with integral seals. The filter is monitored for pressure drop and alarms to control room on a high differential pressure. 8-5

The present, expected ac.nual quantity of solid radwaste generated by the Spent Fuel Pool Purification System is about 54 f t;.3 (FSAR, Table 11.5.1). The SFP modification is not expected to result in a significantly higher quantity of solid radwaste. 8.6 Fuel Pool Radiation Shielding Radiation shielding for the spent fuel pool is provided by six foot thick concrete pool walls and by approximately 24 feet of water above the spent'tuel storage racks. A three-dimensional shielding analysis was performed on the spent fuel pool assuming the pool is filled to capacity with the proposed storage densification arrangement. This ' analysis shows that radiation dose levels will be less,than 1 mr/hr on the outside of.the pool walls and at the pool s'arface fr,om the' stored spent fuel. This radiation level meets the - V.C. Summer Nuclear Station design radiation zoning for the fuel handling building. ("3 V Thus it is concluded that the proposed fuel storage modification will not appreciably increase radiation levels or personnel doses in the fuel handling building and surrounding areas above current estimates. 8.7 Radiological Consequences As stated earlier and confirmed by obher studies 2,4,5,6,8,9, it can be shown that there will be no significant increase in activity due to Krypton-85, Cesium 134/137 or crud buildup on pool walls. It is concluded that the incremental impact from the release of either volatile fission products or crud with the expanded capacity of the spent fuel pool will be negligibly.small. /m 8-6

r V, a e rs 8.8 Reracking Operation '.Or The existing spent fuel racks, in the spent fuel pool, are to be removed prior to the discharge of any spent fuel from the core'. These racks have not been exposed to spent fuel and are only minitaallh c'ontaminated. Therefore, it is concluded that ~ significant radiation dose to individuals involved in the reracking is not anticipated. i t/ '8.9 Conclusioriq Based upon the industry experience and evaluations discussed in previous sections, the following conclusions are made. e i e /[o }Ainor increases in radiological burden to the pool ,4 ,e ster;, if any, can be adequately handled by the fuel s ~ pool clean up system (filter and demineralizer), thereby maintaining the radionuclide concentration in the water at an acceptably low level, o No appreciable increase in solid radioactive wastes (i.e. filter media and demineralizer resin) is anticipated. o No increase in release of radioactive gases is expected, since any long-lived inert radioactive gas potentially available for release (i.e., Kr-85) will have leaked from the fuel either in the reactor core during operation or during the first few months of reside'ce in the pool. Further, Vol. 1, Ref. 3, (pp. n 4-16) ha,s shown airborne activity to be considerably lower than that allowable'by Table 1 of 10CFR Part 20, Appendix B. Therefore, the planned expansion will not significantly increase the release of. radioactive O. gases. 1 8-7

r~ + k ,Q i o The existing spent fuel pool cooling system will keep ^ the pool water temperature at'an acceptable level (see Section 5 - Thermal liydraulic Considerations). o The existing radiation protection monitoring systems [ and program are adequate to detect and warn of any unexpected abnormal increases in radiation level. This provides sufficient assurance that personnel exposures e can be, maintained As Low As is Reasonably Achievable. s E since the re-racking operations will be performed prior o to any spent fuel being placed in the fuel pool, the existing racks are expected to be only minimally cont!bninated. .\\ I ~ f Hencel removal and ~ disposal of the existing racks will h'a've'only minor radiological impact. <<d o Expanding the storage capacity of the spent fuel pool will not significantly increase the onsite or offsite radiological impact above that of the currently authorized storage capacity, nor is any significant increase in environmental radiological or non-radiological impact anticipated. n l .j ' ' + 3 1 -,s, f ,\\ ~ L 'J i i c: f I \\ i 8-8

REFERENCES TO SECTION 8 O 1. "FSAR", V.C. Summer Nuclear Station, Docket No. 50-395. 2. NUREG 0575, " Handling and Storage of Spent Light Water Power Reactor Fuel, Vol.. 1, Executive Summary and Text, USNRC August 1979. 3. NUREG 0800, USNRC Standard Review Plan - Branch Technical Position ASB9-2, Rev. 2, July 1981. 4. A.B.

Johnson, Jr.,

" Behavior of Spent Nuclear Fuel in Water Pool Storagen BNWL-2256, September 1977. 5. J.R. Weeks, " Corrosion of Materials in Spent Fuel Storage Pools", BNL-NUREG-2021, July 1977. 6. J.M.

Wright,

" Expected Air and Water Activities in the Fuel Storage Canal", WAPD-PWR-CP 1723, (with addendum) undated. 7. ANS 5.4 Proposed Standard, " Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel", American Nuclear Society, issued for review 1981. 8. " Licensing Report on High Density Spent Fuel Racks for G Quad Cities Units 1 and 2,", Docket Nos. 50-254 and 50-265, Commonwealth Edison Company, June 1981. 9. " Licensing Report for High Density Spent Fuel Storage Racks", Rancho Seco Nuclear Generating Station, Sacramento Municipal Utilities

District, Docket No.

50-312, June 1982. 8-9 O

p.., s 9. . INSERVICE SURVEILLANCE PROGRAM FOR BORAFLEX NEUTRCN ABSORBING MATERIAL 9.1 Program Intent: A sampling program to verify the integrity of the neutron absorber material employed in the high-density fuel racks in the long-term environment is described in this section. The program is intended to be conducted in a manner which allows access to the representative absorber material samples without disrupting the integrity of the entire fuel storage system. The program is tailored to evaluate the material in normal use mode, and to forecast future changes using the data base developed. 9.2 Description of Specimens: The absorber material, henceforth referred to as " poison", used in the surveillance program must be representative of the L material used within the storage system. It must be of the same composition, produced by the same method, and certified to the same criteria as the production lot poison. The sample coupon must be of similar thickness as the poison used within the storage system and not less than 5 3/4 x3 inches on a side. Figure 9.1 shows a typical coupon. Each poison specimen must be encased in a stainless steel jacket of an identical alloy to that used in the storage system, formed so as to encase the poison material and fix it in a position and with tolerances similar to that design used for the storage system. The jacket has to be closed by tack welding in such a manner as to retain its form throughout the test period and still allow rapid and easy opening without causing mechanical, damage to the poison specimen contained within. The jacket should permit wetting and venting of the specimen similar to the actual. rack environment. O 9-1

9.3 Test: The test conditions represent the vented conditions of the box elements. The samples are to be located in one of the cells of a rack. Eighteen test samples are to be fabricated in accordance with Figdre 9.1 and installed in the pool when the racks are installed. The procedure for fabrication and testing of samples is as given below:

a. The samples should be cut to size and weighed carefully in milligrams,

b. The length, width, and the average thickness of each specimen is to be measured and recorded,

c. The samples should be fabricated in accordance with Figure 9.1 and installed in a cell (Region I).

d. Two samples should be removed at each time interval according to the schedule shown in Table 9.1.

9.4 Specimen Evaluation: After the removal of the jacketed poison specimen from the cell at a designated time, a careful evaluation of that specimen should be made to determine its actual condition as well as its apparent durability for continued function. Separation of the poison from the stainless steel specimen jacket must be performed carefully to avoid mechanical damage to the poison specimen. Immediately after the removal, the specimen and jacket section should visually be examined for any effects of environmental exposure. Specific attention should be directed to the examination of the stainless steel jacket for any evidence of physical degradation. Functional evaluation of the poison material can be accomplished by the following measurements: i n_ 9-2

a. A ne,utron radiograph of the poison specimen aids in the determination of the maintenance of uniformity of the boron distribution.

b. Neutron attenuation measurements will allow evaluation of the continued nuclear effectiveness of the poison.

Consideration must be given, in the analysis of the attenuation measurements, for the level of accuracy of such measurements as indicated by the degree of repeatability normally observed by the testing agency.

c. A measuren.ent of the hardness of the poison material will establish the continuance of physical and structural durabiity.

The hardness acceptability criterion requires that the specimen hardness will not exceed the hardness listed in the qualifying test document for laboratory test specimen irradiated to 1011 rads. The actual hardness measurement should be made after the specimen has been withdrawn from the pool and allowed to air dry for not less than 48 hours to allow for a meaningful correlation with the preirradiated sample.

d. Measurement of the

length, the

width, and the average thickness and comparison with the pre-exposure data will indicate dimensional stability within the variation range reported in the Boraflex laboratory test reports.

A detailed procedure paraphrasing the intent of this program will be prepared for step-by-step execution of the test procedure and interpretation of the test data. O 9-3

TABLE 9.1 Time Schedule for Removing Coupons g Date Installed INITIAL FINAL WEIGHT PIT WEIGHT WEIGHT CHANGE PENETRATION 2 2 2 SCHEDULE (mg/Cm -Yr) (mg/Cm -Yr) (mg/Cm -Yr) mil /Yr 1 2 90 day 3 4 180 day 5 E 6 1 Year O' 8 5 Year T 9 10 10 Yearl' 11 I 12 15 Year 13 P 14 20 Year 15 16 30 Year " 17 18 40 YearF g TIME SCHEDULE FOR REMOVING COUPONS 9-4

t 5 O 1 <t Si F 0: 2 1 h TACK WELD / / I _F N / l ~N

b. 7~. g.05$THK.X fjy 7,gw,SST.

u 7 2 / 304 STRIP / ^ ~ /, (4 SlDES-) ( g e N i i I ) \\ g NEUTRON l 3 / / / ABSORBER / / / kh ,/ S%S ';h l' / / + N's / .06&THK S.S. -304 9 FIG. 9.1 - TEST COUPON 9-5

10.0 COST / BENEFIT ASSESSMENT 0- + A cost / benefit assessment has been prepared in accordance with the requirements. of ' reference 1 Section V, Part 1. The -purpose of the assessment is to demonstrate that the installation of high-density spent fuel storage racks is the most advantageous means of handling spent

fuel, considering the needs of our customers for a dependable source of electric power.

The material is presente60to' satisfy the NRC's need for information; it is the position of SCE&G that no environmental impact statement need be prepared in support of the request, because there will be no significant impact on the human environment. NRC precedent establishes that alternatives and economic costs need not be discussed when there is no significant environmental impact.

However, for the sake of completeness, alternatives to re-racking, for additional spent fuel storage capacity, are discussed in Section 10.3.

O 10.1 Specific Needs for Spent Fuel Storage Disposal of V.C. Summer spent nuclear fuel is scheduled to be carried out by the Department of Energy in or after 1998 in accordance with Public Law 97-425; Nuclear Waste Policy Act of 1982. As V.C. Summer spent fuel may not be accorded a high priority under the DOE program, SCE&G is seeking to provide a spent fuel storage capacity to support approximately twenty-five years of nominal operation. No other contractual arrangements exist for the interim storage or reprocessing of spent fuel from V.C. Summer Nuclear Station; therefore, increased storage capacity in the V.C. Summer fuel pool is the only viable option under ' consideration. Table 1.1, the fuel discharge schedule, indicates that with the high density spent fuel racks, loss of full core discharge capability (FCDC) will occur in 2008. O 10-1

d 10.2 Cost of Spen't Fuel Storage V The design and manufacture of the spent fuel storage racks will.be undertaken by the organizations described in Section 1. It is expected that the total project cost will be between 1.2 and 1.4 million dollars. 10.3 Alternatives to Spent Fuel Storage South Carolina Electric & Gas has considered the various alternatives to the proposed onsite spent fuel storage. These alternatives are-as follows: o Shipment of fuel to a reproc'essing or independent spent fuel storage / disposal facility No commercial spent fuel reprocessing facilities are presently operating in the United States. SCPSA and SCE&G have made contractual arrangements whereby spent nuclear fuel and/or high level nuclear waste will be accepted and disposed of by the U.S. Department of Energy; but such services are not expected to be available before 1998. The V.C. Summer Nuclear Station existing spent fuel storage capacity will not provide full core discharge capability beyond 1993. Spent fuel acceptance and disposal by the Department of Energy is

not, therefore, an alternative to increased on-site pool storage capacity.

O Shipment of fuel to another reactor site Shipment of V.C. Summer Nuclear Station fuel to another reactor site O ' ~' J

could pro, vide short term relief to the storage capacity problem. However, transshipment of spent fuel merely serves to transfer the problem to another site and does not result in any additional net long-term storage capacity. Accordingly, SCEEG does not consider the transshipment of spent fuel to be an appropriate . alternative to high-density spent fuel storage at the site. o Not completing the reactor plant /not operating the plant after the current spent fuel storage capacity is exhausted As indicated in NUREG-0575,," Final Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel," the replacement of nuclear power by coal-generating capacity would cause excess mortality to rise from 0.59-1.70 to 15-120 per year for 0.8 GWY(e). Based on these facts, not operating the plant or shutting down the plant after exhaustion of spent fuel discharge capacity are not viable alternatives to high density storage in the spent fuel pool. The prospective 1983 expenditure of approximately $1.4 million for the high density racks is small compared to the estimated value of replacement power equivalent to the plant's energy output: approximately $9 million per month in 1983 and between $18.1 and $22.7 million per month in 1990-1991. Ilie subject of the comparative economics associated with various spent fuel options is the subject of Chapter 6 of NUREG-0575. Although the material presented is generic, it is of value in comparing the costs of the various options. Of the options presented in Chapter 6 of NUREG-05752, high-density spent fuel storage at the site is the most economic option at $18 per O 10-3

KgU. The price o,f "Away From Keactor (AFR)" fuel storage, if available, would be $115 per KgU. This corresponds to 0.5 mill /Kwh from a 1000 MWe power reactor for AFR storage. The marginal cost per KgU of high density spent fuel racks for V.C. Summer Nuclear Station is S5. 10.4 Resource Commitments The expansion of the V.C. Summer Nuclear Station spent fuel storage capacity will require the following primary resources: o Stainless steel - 290,600 pounds o Boraflex neutron absorber 6900 pounds of which 3200 pounds is Boron Carbide (BgC) powder. The requirement for stainless steel represents a small f raction of the total domestic production for 1983.3 Although the fraction of domestic production of BgC, required for the fabrication, is somewhat higher than that for stainless steel, it is unlikely that the commitment of BgC to this project will affect other alternatives. Experience has shown that the production of BgC is highly variable and depends on need, but could easily be expanded to accommodate additional domestic needs. l 10-4

~ REFERENCES TO SECTION 10 1. B.K. Grimes, "OT Position for Review and Acceptance of Spent Fuel Storages and Handifng Applications," April 14, 1978. 2. NUREG-0575, " Handling and Storage of Spent Light Water Power Reactor Fuel", Vol. 1-3, USNRC, August, 1979. 3. " Mineral Facts and Problems," Bureau of' Mines Bulletin 671, 1980. l O l O O 10-5

11. QUALITY ASSURANCE PROGRAM O 11.1 Introduction This chapter provides a general description of the Quality Assurance Program that is implemented to assure that the quality objectives of the contract specification are met. 11.2 General Tne Quality Assurance Program used on this project is based upon the system described in Joseph Oat's Nuclear Quality Assurance Manual. This system is designed to provide a flexible, but a highly controlled system for the design, manufacture and testing of customized components in accordance with various

Codes, specifications, and regulatory requir,ements. The Joseph Oat Nuclear Quality Assurance Program has been accepte.d by ASME and found tg be adequate by NRC audit team.

The philosophy behind Oat's Quality Assurance System is that it shall provide for all controls necessary to fulfill the contract requirements with sufficient simplicity to make it functional on a day to day basis. As this system is applied to most of the contracts which Joseph Oat obtains, implementation of it is almost second nature to Oat's personnel. The system readily adapts to different designs and component configurations, making possible the construction of many varied forms of equipment. The highlights of this system, as addressed in the following paragraphs, provide an overview of the system and how it has been applied to the customer specifications and regulations. 11.3 System Highlights: The design control is organized to provide for careful review of all contract requirements to extract each indivi' dual design and quality criteria. These criteria are translated into design and quality control documents customized to the contract requirements and completely reviewed and approved by responsible personnel. 11-1

The system for control of purchased material entails g generating detailed descriptions of each individual item of material along with specifications for any special requirements \\ such as impact

testing, corrosion

testing, monitoring, or witnessing of chemical analysis, provision of overcheck specimens, special treatments or conditioning of material, source inspection, and provision of documentation of performance of any of the above.

Material receipt inspection includes a complete check of all material and its ~ ~ 6ecumentation. Upon acceptance, each item of material is individually listed on a control sheet issued once a week to assure that only accepted material goes into fabrication. The fabrication control system provides that a shop traveler is prepared for each subassembly and ' assembly in each contract. The traveler is generated specifically to provide step by step instructions for fabricetion, inspection,

testing, cleaning, packaging, etc. which address all standard and special requirements h

of the contract specifications. Special attention is given to deployment of fabrication sequence and inspection steps to preclude the possibility of missing poison sheets or incorrect sheets (incorrect B10 loading). Due to the tendency of contract specifications to require special examination techniques or test procedures, all nondestntetive examination procedures and test procedures are custom written to apply to each given component within a contract. The system provides f o r, qualification and written certification of personnel performing quality related activities including nondestructive examination and fabrication inspection, welding, engineering, production supervision and auditing. Other requirements of a solid quality control system are fully covered as specified in the Quality Assurance Manual including 11-2

document control,, control of measuring and test equipment, control of nonconforming material and parts, corrective action auditing and other areas as specified. 11.4 Summary: Joseph Oat Corporation's Quality Assurance System provides the full measure of quality assurance required by the contract. All special requirements of the specifications are covered including source inspection of material and witnessing of material testing by the

Engineer, furnishing of material certifications and test reports within five days of shipment, and obtaining verification of qualification testing of poison materials.

Oat has a long history of providing excellent quality control over a wide range of equipment types such as the high density fuel racks. O 9 5 0 11-3 . _ _ _ _ _ _ _ _ _}}