Latest revision as of 01:34, 16 March 2020

Text

Spent Fuel Storage Pool B Rack Mod, SAR
	ML19325E888
Person / Time
Site:	Crystal River
Issue date:	11/01/1989
From:	; FLORIDA POWER CORP.
To:	;
Shared Package
ML19325E886	List: ML19325E885; ML19325E887; ML19325E888;
References
	NUDOCS 8911090319
	Download: ML19325E888 (105)
	v • d • e

. ._ . - . . _ . _ _ _ _ _ _ _ - _

, 4

?

r ATTACHMENT 2 FLORIDA POWER CORPORATION .

CRYSTAL RIVER ENIT 3 r

l l

L SPENT FUEL STORAGE POOL B !

RACK MODIFICATION u

o SAFETY ANALYSIS REPORT I

DOCKET NO. 50 - 302 l

8911090319 DR 991o33 p ADOCK 05000305 j;. PDC ,

l

. _ _ . . -____m _ _ _ _ _ _ _ _ _ _ _ _ _ . _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ -- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . . - - --. _ _ _ _ , ~ . . ~ . , _- . - - -'

1 l

.\/,:- , ,

v.

, FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SPENT FUEL POOL B STORAGE RACKS LICENSING REPORT November 1, 1989 1

l I

l ;

)

I a l

'I i

i 1.

l:

l l

1.

l 1..

l.

1 l

1 a

l i

I l

l l'

t l ;

i. ;

u. )

r L

TABLE OF CONTENTF Page

1.0 INTRODUCTION

1-1 1.1 License Amendment. Requested 1-1 1.2 Current Status 1-1 1.3 Summary of Report 1-1 j 1.4 Conclusions 1-2 1.5 References 1-2 2.0 DESIGN BASES 2-1 b g.. 2.1 Description of Spent Fuel Racks 2-2 2.1.1 Region 1 Rack Design 2-2 !

2.1.2 Region 2 Rack Design 2-3 2.2 Neutron Multiplication Factor 2-4 L

I

'2.2.1 Nonnal Storage 2-4 2.2.2 Postulated Accidents 2-5 2.2.3 Calculational Methods 2-6 Region l' 2-6 Region 2 2-7 2.2.3.1 Method Validation 2' 7 j

, Region 1 Calculation Method 2-7 l l Region 2 Calculation Method 2-8 j 2.2.3.2 Criticality Analysis for 2-8 Region 1 ;

l. 2.2.3.3 Criticality Analysis for 2-11 l Region 2 !

2.2.3.4 Storage of Consolidated Fuel 2-12 j' '

2.2.4 Acceptance Criteria for Criticality 2-13 2.2.4.1 Neutron Poison Material 2-13 Surveillance Program t s 2.2.4;2 Decay Heat Calculations for 2-14 Spent Fuel 2.2.4.3 Thermal Hydraulic Analysis 2-15 for Spent Fuel Cooling 2.2.4.4 Potential Fuel and Rack 2-19 Handling Accidents 3.0 MECHANICAL, MATERIAL AND STRUCTURAL CONSIDERATIONS 3-1 3.1 Description of Spent Fuel Pool 3-1 i

~~

pf l

i.- ,:

i .i 7

T TABLE OF CONTENTS (Continued) t f.ASLR 3.2 Structural Design of the Racks 3-1

3.3 Integrity'of Fuel Racks Under Fuel Handling 3 -1 ;

i Accident Conditions L 3.3.1 Spent Fuel Handling Machine Uplift 3-1 l Analysis 3.3.2 Fuel Assembly Drop Accident Analysis 3-1 3.3.2.1 Drop Orientations 3-2 3.3.2.2 Acceptance Criteria 3-2 !

3.3.2.3 Assumptions for Energy 3-2 c' Dissipation i

3.3.2.4 Drop Analysis Results 3-2 3.4. Applicable Codes, Standards and Specifications 3-3 3.5 Seismic Analysis Procedures for Spent Fuel 3-5 Storage Racks. ,

3.5.1 Analysis Overview 3-5 l 3.5.2 Seismic Model 3-6 l 3.5.2.1 Model Description 3-7 s 3.5.2.2 Structural Model 3-7 3.5.2.3 Nonlinear Seismic Model 3-10 3.5.2.4 Damping 3-11 3.5.2.5 Fluid Coupling 3-12 3.5.2.6 Friction Coefficient 3-13 3.5.3 Time History Evaluation 3-13 3.6 Structural Acceptance Criteria and Analysis 3-17 l Results for Spent. Fuel Storage Racks 3.6.1 Criteria 3-17 l 3.6.2 Stress Limits for Specifie6 Conditions 3-18 i 3.6.3 Results for Rack Analysis 3-19 1 3.6.4 Rack Displacements 3-19 l'

L 3.7 Materials, Quality Control, and Special 3-20 l Construction Techniques L 3.7.1 Construction Materials 3-20 ;

l 3.7.2 Neutron Absorbing Material 3-20 l 3.7.3 Quality Assurance 3-20 -

l~ \

Testing and Inservice Surveillance 3.8 3-21 I 11 1

I l

TABLE OF CONTENTS (Continued)

Q' ;;

R&EA 4.0 SAFETY EVALUATION 4-1 4.1 Degree of Subcriticality 4-1 4.2 Governing Codes for Design 4-1 '

4.3 Ability to Withstand External Loads and Forces 4-1

! 4.4 Ability to Ensure Continuous Cooling 4-2 l 4.5 Provisions to Avoid Accidental Dropping of Heavy 4-3 objects on Spent Fuel l 4.6 Material Compatibility 4-3 l 4.7 Radiological Considerations 4-3 4.8 Ability of Racks to Withstand Accidenta.'. 4-3 I

Lift Forces i

5.0 COST / BENEFIT ASSESSMENT 5-1 5.1 Cost / Benefit Assessment 5-1 i

5.1.1 Need for Increased Storage Capacity 5-1 5.1.2 Estimated Costs 5-2 5.1.3 Consideration of Alternatives 5-2 -

5.1.4 Resources Committed 5-2 5.1.5 Thermal Impact on ths Environment 5-3 ,

5.2 Radiological Evaluation 5-3 '

( 5.2.1 Solid Radwaste 5-3 l' l 5.2.2 Gascous Radwaste 5-3 l

5.2.3 Personnel Exposure 5-3 5.2.4 Radiation Protection During Re-Rack 5-5 i Activir.ies 5.2.5 Rach Decontamination and Disposal 5-5 5.3 Accident Evaluation 5-6 5.3.1 Cask Drop /Tip Analysis 5-6

5.3.2 NUREG-0612 Results 5-7 5.3.3 Safety Evaluation Report and Final 5-7 Environmenta) Statement 5.3.4 Conclusion 5-7 l'

6.0 REFERENCES

6-1 L

111

~ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _-- _ --

LIST OF 'rABLIS ERER 2-1 Rack Module Data 2-20 2-2 Benchmark Critical Experiments 2-21 2-3 Comparison of Phoenix Isotopic Prediction 2-24 to Yankee Core 5 Measurements 2-4 Benchmark Critical Experiments Phoenix Comparison 2-25 2-5 Data for U Metal & UO2 Critical Experiments 2-26 3-1 Loads and Icad Combinations for Spent Fuel Racks 3-22 3-2 Minimum Margin to Allowable Region 1 3-23 3-3 Minimum Margin to Allowable Region 2 3-24 5-1 Spent Fuel Pool Capacity Without Reracking B Pool 5-8 5-2 Components Stored in Spent Fuel Pool 5-9 5-3 Spent Fuel Pool Capacity After Reracking B Pool 5-10 5-4 System Production Costs 5-11 5-5 Gamma Isotcpic Analysis of Spent Fuel Pool Water 5-12 iv

i i

[ LIST OF FIGURES l

l e Enun !

L 2-1 Crystal River 3 Pool "B" Spent Fuel 2-28 {

Storage Rack Arrangement j l

2-2' Region 1 Fuel Storage Rack Module 2-29 2-3 Region 1 Module Cross-Section 2-30 l 2-4 Region 2 Fuel Storage Rack Module 2-31 )

l i 2-5 Region 2 Module Cross-Section 2-32 I 2-6 Region 2 Module Top View 2-33 l 1

2-7 Region 1 Cell Layout 2-34 2-8 Region 2 Cell Layout 2-35 l 2-9 Minimum Burnup vs Initial Enrichment for Region 2 Racks 2-36 1

2-10 Spent Fuel Pool Natural Circulation Model (Elevation View) 2-37 i 2-11 Spent. Fuel Pool Natural circulation Model (Plan View) 2-38 2

2-12 Spent Fuel Rack Inlet Flow Area (Plan View) 2-39 2-13 Intercell Flow Area 2-40 i l 3-1 Design vs Tine History Response Spectra N-S SSE 4% Damping 3-25 3-2 Design vs Time History Response Spectra E-W SSE 4% Damping 3-26 4

3-3 Design vs Time History Response Spectra Vertical 3-27 l SSE 4% Damping :

l l 3-4 Acceleration Time History N-S SSE 3-28 j l

3-5 Acceleration Time History E-W SSE 3-29

'3-6 Acceleration Time History Vertical SSE 3-30 :

3-7 Structural.Models Regions 1 & 2 3-31 u

3-8 . Structural Model (9 x 10 Mod) 3-32 l 3-9 Effective Structural Models Regions 1 & 2 3-33 '

l 3-10 3-D Nonlinear Seismic Model 3-34 j l

v 1

i

L

{.

I LIST OF FIGURES (Continued)

( Page l' 3 13 Nonlinear Seismic Model Region 1 (2-D View of 3-D Model) 3-35 3-12 Nonlinear Seismic Model Region 2 (2-D View of 3-D Model) 3-36 l 3-13 Auxiliary Building-Fuel Handling Area 3-37 (Elevations 143' & 119')

3-14' Auxiliary Building-Fuel Handling Area 3-38 (lkongitudinal Section A-A) l i

A l

t f

{

F r

e h

f L vi .

L ,

i

'l t

s

>w w w p. _ - . , _ , , -_ .-_.g , .- -y

l l

1.0 INTRODUCTION

i l

1.1 LICENSE AMENDNENT REQUESTED i l

Florida Power Corporation (FPC) is currently pursuing the design <

and manufacture of new spent fuel storage racks to be placed into the spent fuel pool at crystal River Unit $3. The purpose of these new racks is to increase the amount of spent fuel that can j be stored in the existing spent fuel pools. The racks are ;

designed so that they can store spent fuel assemblies in a high l density array. Therefore, FPC hereby requests a License Amendment l be issued to the Crystal River Unit #3 facility Operating License DPR-72 to include installation and use of new storage racks that meet the criteria contained herein. This report has been prepared to support tnis request for license amendment.

1.2 CURRENT STATUS There are two spent fuel pools at Crystal River Unit 3. Spent fuel pool A features high density storage racks while spent fuel ;

j pool B utilizes standard racks. Pool B presently contains racks ;

for 120 spent fuel assemblies. With this present spent fuel j storage capacity, FPC will lose full core reserve storage capacity after Refuel VII in 1990. Therefore, to ensure that sufficient capacity continues to exist at Crystal River to store discharged j fuel assemblies, FPC plans to replace the existing storage racks :

with new spent fuel storage racks who design will allow for more .I dense storage of cpent fuel, thus enabling the existing pool to j store more fuel in the same space as occupied by the current i I

racks. Spent fuel pool B maintains an eversafe geometric !

centerline to centerline spacing of 21 1/8 inches between !

assemblie.s and is designed to seismic category I criteria. i

, i i

1.3

SUMMARY

OF REPORT j This licensing report follows the guidance of the NRC position paper entitled, "OT Position for Review and Acceptance of Spent l Fuel Storage and Handling Applications," dated April 14, 1978, as i amended by the NRC letter dated January 18, 1979, with the _

exception that crerH t is taken for burnup for all storage i l locations other than those reserved for full core discharge. ;

L 1

The Design Bases (section 2.0) includes a descriptien of the new spent fuel racks, the effective neutron multiplication factor, and the acceptance criteria for criticality. This section considers l normal storage and handling of spent fuel as well as postulated accidents with respect to criticality and the ebility of tha spent fuel pool cooling system to maintain sufficient cooling.

Mechanical, material and structural aspects (section 3.0) of the ,

report concern the capability of the fuel assemblies, storage !

racks, and spent fuel pool system to withstand the effects of 1-1 l

t

L . 1 1

4 l

natural phenomena such as earthquakes, tornadoes, etc. and other service loading conditions. The design, procurement and i fabrication of the storage racks comply with the pertinent Quality l Assurance Requirements of Appendix B to 10 CFR 50.

i Section 4 contains a brief summary of the safety evaluations ,

related to the deuign criteria, material compatibility, degree of l subcriticality, radiological considerations, accidental drop !

provisions and the ability to ensure continuous cooling. This )

section ensures the spent fuel racks are safe and capable of performing their intended function.

The environmental aspects of the report (section 5.0) concern the j thermal and radiological release from the f acility under normal !

and accident conditions. This section also addresses the '

occupational rodiation exposures. generation of radioactive waste, need for expansion, commitment of material and nonmaterial i resources, and a cost-benefit asnessment. l l i

1.4 CONCLUSION

S l i

On the basis of the design requirements presented in this report, i operating experience with high density fuel storage, and material j referenced in this report, it is concluded that the proposed ;

modification of the Crystal River Unit 3 spent fuel storage i facility will continue to provide safe spent fuel storage, and j that the modification is consistent with the facility design and 1 operating criteria as provided in the Crystal River Updated FSAR L l

and Operating Licenses. l l '

i

1.5 REFERENCES

l ;

1. Crystal River Unit 3 Facility Operating Licenses DPR-72, Docket !

No. 50-302

{

2. "O'? position for Review and Acceptance of Spent Fuel Storage ;

l e'.d Handling Applications", dated April 14, 1978 as amended i < anuary 18, 1979. .

L i

3. Crystal River Unit 3 Final Safety Analysis Report.

}

l l

r l

l 1-2 l

I

i 3.0 DESIGN BASES The function of the spent fuel storage re.cks is to provide for storage of new, spent, and consolidated fuel assemblies in the appropriate region of a flooded pool, while maintaining a coolable geometry, preventing criticality, and protecting the fuel assemblies from excess mechanical or thermal loadings.

} A likt of design criteria is given below:

1. The racks ars designed in accordance with the NRC "OT Position for Review and Acceptance of Spent Puel Storage and Handling Applications," dated April 14, 1978 and revised January 18, 1979, with the exception that credit is taken for burnup for all storage locations other than those racerved for full core discharge (Region 1).

2. The racks are designed to meet the nuclect requirements of ANEI-N210-1976. The effective multiplicatior, factor, 4,n, in the spent fuel pool is less than or equal to 0.9S, including all uncertainties and under all credible conditions.

3. The racks are designed to allow coolant flow such that boiling in the water channels between the fuel assemblies in the rack does not occur. Maximum fuel cladding temperatures are calculated for various pool cooling conditions.

4. The racks are designed to Seismic Category I requirements, and are classified as ANS Safety Class 3 and ASME Code Class 3 Component Support structures. The structural evaluation and seismic analyses are performed using the specified loads and load .

combinations in Table 3-1.

5. The racks are designed to withstand loads which may result from

l fuel handling accidents and from the maximum uplift force of the :

fuel handling crane without violating the criticality acceptance ;

criterion.

6. Each storage position in the racks is designed to support and I guide the fuel assembly or consolidation fuel in a manner that will minimize the possibility of application of excessive lateral, axial and bending loads to fuel assemblies or consolidated fuel during fuel assembly handling and storage.

l '

l 7. The materials used in construction of the racks are compatible with the storage pool environment and do not contaminate the fuel assemblies. -

8. The spent fuel racks consist of two designs varying with storage capability. Region 1 racks are poison racks designed to store fresh and spent fuel and consolidated arrays of fuel at a max Sum 2-1

! I l

ratio of 2:1. Region 2 racks are designed, to take credit for l burnup and is based on criterias established in sub section ;

2.2.3.3 (Initial enrichment vs. Burnup requirements), and !

consolidated arrays of fuel at a maximum ratio of 2:1. l l

2.1 DF.SCRIPTION OF SPENT FUEL RACKS The spent fuel storage pool rack arrangement is shown in Figure !

2-1. l l

Fuel storage is divided into two regions within the pool.

Region 1 (174 locations) consists of high density fuel assembly l l spacing obtained by utilizing a noutron absorbing material and

is reserved for core off loading (177 fuel assemblies) . Region !

2 (641 locations) also consists of high density fuel assembly !

spacing and provides normal r.torage for spent fuel assemblies.

Region 1 is designed to accommodate non-irradiated, 4.2 weight percent U235 enriched fuel or fuel which has not achieved a '

pre-determined burnup. Region 2 is designed to accommodate irradiated fuel. Placement of fuel in Region 2 is determined i

by burnup calculations and is controlled administratively. Fuel ;

consolidated at a 2:1 consolidation ratio can also be stored in !

the racks. The racks meet the requirements of the NRC "OT l Position for Review and Acceptance of Spent Fuel Storage and -

Handling Applications," dated April 14, 1978, and modified ,

January 18, 1979, with the exc9ption that, for Region 2 storage, l credit is taken for fuel burnup based on the proposed Revidon ;

2 of USNRC Regulatory Guide 1.13. Fuel is assumed to bc l consolidated two years after it has been removed from the core.

l 2.1.1 Region 1 Rack Design The Region 1 storage rack modules are composed of individual storage cells made of stainless steel. These racks utilize a '

neutron absorbing raterial, Boraflex, which is attached to each cell. The cells within a module are interconnected by grid assemblies to form an integral structure as shown in Figure 2-2.

Each rack module is provided with leveling pads which contact the spent fuel pool floor or pool floor plates and are remotely adjustable from above, through the cells, at installation. The ,

modules are neither anchored to the floor nor braced to the pool walls.

The fuel rack ass 6mbly consists of three major sections which '

are the leveling pad assembly, the top and bottom grid assemblies, and the cell essembly. Figure 243 illustrates these sections.

The major components of the leveling pad assembly are the ;

support block, the leveling pad, and the leveling pad screw. ;

The top of the support block is we'ted to the base plate. The leveling pad assemblies transmit the loado to the pool floor and ,

2-3 i

4

.-,+n - . , , - - . - - - - n , - - , ,

~ . _ . _ . . _ _ _ _ _ _ _ . _ _ . _ _

i I

I provide a sliding contact. The leveling pad screw permits the ;

leveling adjustment of the rack.

l The lower grid assembly consists of box-beam members, side !

plates and the base plate. The bottom of the cell assembly is [

welded to the lower grid. The upper grid assembly consists of ,

box + beam members and side plates. The upper part of the cell :

assembly is welded to the upper grid. The upper and lover grid j assemblies maintain the conter-line to center-line spacing i between the cells and provide the ctructural conneccions between

. the cells to form a fuel rack assembly, i

The major components of the cell assembly are the fuel assembly :

cell, the Boraflex (neutron absorbing) material, and the t wrapper. The wrapper is attached to the outside of the cell by i spot welding the entiro length of the wrapper. The wrapper '

covers the Boraflex material and also provides for venting of the Boraflex to the pool environment. Depeading on the ,

criticality prevention requirements, some calle have a Boraflex :

wrapper on all four sides, some on three sides, and some on two sides.

2.1.2 Region 2 Rack Design !

The Region 2 storage rack modules consist of ctainless steel cells assembled in a checkerboard pattern, producing a honeycomb ;

type atructure as shown in Figure 2-4. Each cell is of the same basic /Stign as described for Region 1; i.e., the major t componeuss are the cell, the Bornflex (noutron absorbing) '

t ma%erial, and the wrapper. The cells are welded to a base

! support assembly and to one another to form an integral structure without use of grids as used in Region 1 racks. This design is also provided with leveling pads which contact the spent fuel pool floor or pool floor plates and are remotely -

adjustable from above, through the cells, at installation. The ;

modules are neither e7 chored to the floor nor braced to the pool ;

walla. The fuel rack assembly consists of two major sections [

which are the base support assembly and the cell assembly.

Figures 2-5 and 2-6 illustrate these sections. ;

The major components of the base support assembly are the leveling pad, the leveling pad screw, and the support plate.

1 The top of the support plate ik weldad to the fuel rac.k base plate. The leveling pads transmit the loads to the pool floor and provide a sliding contact. The leveling pad screw permits l

the leveling adjustment of the rack.

Rack module data is described in Table 2-1.

2-3

I l i, . I I

l 1

2.2 NEUTRON MULTIPLICATION FACTOR l f

2.2.1 Normal Storage j l

Criticality of fuel assemblies in the spent fuel storage rack l is prevented by the design of the rack limits fuel assembly >

interaction. This is done by fixing the minimum separation l between assemblies and inserting neutron absorbing material (poison) betwuen assemblies. {

The design bas is for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent ;

probability at a 95 percent confidence level that the effective i multiplication factor (K,n) of the fuel assembly array will be :

less than or equal to 0.95 when the storage racks which are l fully loaded with spent fuel, consolidated fuel, or a ;

combination of spent fuel and consolidated fuel, and flooded j with unborated water as recommended in ANSI 57.2-1983 and in i Standard Review Plan 3.8.4. !

The following are the conditions that are assumed in meeting !

this design basis.

a. The fuel assembly contains the highest enrichaent authorized 1 without any control rods or any noncontained burnable poison .

and is at its most reactive point in life. The B&W 15x15 !

fuel assembly is more reactive than its consolidated array.

The assembly is conservatively modeled with water replacing i i the assembly grid volume and no U-234 or U-236 in the fuel

pellet. No U-235 burnup is assumed in Region 1. ;

I

b. Tha storage cell nominal geometry !s shown on Figure 2-7 for Region 1 and Figure 2-8 for Regior. 2.

c. The moderator is pure water at the temperature within the design limits of the pool which yields the 2argest t reactivity. A conservative value of 1.0 gn/cm8 is used for !

the danalty of water. No dissolved boron is included in the i water. ;

e

d. The nominal case calculation is infinite in lateral and axial extent. However, poison plates are not necessary on the periphery of the modular array and between widely spaced >

nodules because calculations show that this finite array is less reactive than the nominal case infinite array.

Therefore, the nominal case of an infinite array of poison cells is a conservative assumption, y

e. Machanical uncertainties and biases due to mechanical tolerances during construction are treated by either using 2-4

p ,

')

i

" worst case" conditions or by performing sensitivity studies 1 and obtaining appropriate values, The items included in the j analysis are: )

- Poison pocket thickness ,

stainless steel thickness 1 Cell ID l Center-to-center spacing i Cell bowing The calculated method uncertainty and bias are discussed in Section 2.2.3.

f. Credit is taken for the neutron absorption in full length i structural materials and in solid materials added specifically for neutron absorption. A minimum po.%cn i leading is assumed in the poison plates and B 4 C particle self ,

shielding is included as a bias in the reactivity t calculation. ;

i 2.2.2 Postulatud Accidents i Most credible accident conditions will not result in an increase in 4 of the rack. Examples are the loss of cooling systems !

(reactivity decreases with decreaming water density) and dropping a fuel assembly on top of the rack.

Mcwever, accidents can be postulated which would increase !

reactivhy (i.e. , dropping a fuel assembly between the pool wall i and the fuel racks or misloading an assembly with a burnup and enrichment combination outside of the acceptable area in Figure 2-9), if the presence of boron was not assumed. Therefore, for .

accident conditions, the double contingency principle er ANSI l N16.1-1975 is applied. This states that one is not required to ,

assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident

conditions, the p.o.ence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. l The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by about 30 percent delta k. In ,

! perspective, this is more negative reactivity than is present l

in the poison plates (15 percent delta k), so 4 for the rack i would be 1 css than 0.95 even if the poison plates were not i present. Thus, for all postulated accidents involving mis.-placement and damage to fuel assemblies, should there be a reactivity increase, Q would be less than or equal to 0.95 due to the combined offects of the dissolved boron and the poison plates. )-

2-5 L !

I l

l The " optimum moderation" accident is not a problem in spent fuel I storage racks because the presence of poison plates removes the i conditions necessary for " optimum noderation". The Q !

! , continually decreasps as moderator density decreases to values p less than 1.0 gn/cm ,

2.2.3 . Calculational Methods

, Realen 1 The criticality calculation method and cross-section values are i l verified by comparison with critical experiment data for ;

I assemblies 21.ailar to those for which the racks are designed.

This benchmarking data is sufficiently diverse to establish that !

the method bias and uncertainty will apply to rack conditions I which include strong neutron absorbers, large water gaps and low I l moderator densities.

The design method which ensures the criticality safety of fuel i assemblieg of codes (

infor thecross-section spent fuel storage rack uses the At:PX pstem generation and KENO IV for j

reactivity determination. I l

The 227 energy group cross-section library that is the common i starting point forallcross-sectionsusedfortpUe benchmarks and the storage rack is generated from ENDF/B-V data. The i

i NITAWL" program includes, in this library, the self-shielding !

resonance cross-sections that are appropriate for each particular geometry. The Nordheim Integral Treatment is used. i Energy and spatial g ighting the XSDRNPM program of cross-sketions is performed by which d.s a one-dimensional SN transport I

I theory code. These multigroup cross-section sets are then used )

as input to KENO IV A which is a three dimensional Monte Carlo criticality theory program designed for reactivity calculations. .

t A set of 33 critical experiments has been enalyzed using the '

above method to demonstrate its applicability to criticality analysis and to establish the method bias and variability. The )

experiments range from water moderated, oxide fuel arrays separated by various materials (B4c, steel, water, Metc.)

simulate LWR fuel shipp!ng and storage conditions that to dry, harder spectrum uranium metal cylinder arrays with various -

interspersed anatorials W (Plexiglas, stoel and air) that ,

demonstrate the wide range of applicability of the method. ,

Table 2-2 summarizes these experiments.

The final result of the analysis is that the criticality design ,

criterion is met when the calculated effective multiplication factor, plus the total uncertainty and any biases, is less than 0.95.

l l

a-s l

i

+

l Reaion 2 Spent fuel storage, in the Region 2 spent fuel stoDage racks, is achievable by means of the concept of reectivity i e equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with fuel j i depletion. A series of reactivity calculations are performed l to generate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield the equivalent Kw when the fuel 1 is stored in the Region 2 racks. l l

Figure 2-9 chows the constant Kg contour generated for the l Crystal River Unit 3 Region 2 racks. Note in Figure 2-9 the 1 endpoint at 0 MWD /MTU where the enrichment is 1.63 w/o and at i 33,000 n'WD/MTU where the enrichment is 4.20 w/o. The j interpretation of the endpoint data is as follows: the

, reactivity of the Region 2 racks containing fuel at 33,000 L MWD /MTU burnup which had an initial enrichnet.c of 4.20 w/o is

! equivalent.to the reactivity of the Region 2 racks containing I fresh fuel having an initial enrichment of 1.63 w/o. It is important to recognize that the curve in Figure 2-9 is based on a constant rack reactivity for that region and not on a constant !

fuel assembly reactivity. )

l The data points on the reactivity equivalence curve wage l l generated with a transport theory computer code, PHOENIX .

PHOENIX is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code. A 25 energy group nuclea l

data library based on a modified version of the Britnsh WIMS(g 4 I library is used with PHCENIX. l A study was done to examine fuel reactivity as a function of i time following discharge from the reactor. Fission product 1 decay was accounted for using CINDER *. CINDER is a one point-depletion computer code used to determine fission product 1 activities. The fission products were permitted to decay for j 30 years after discharge. The fuel reactivity was found to reach a maximum at approximately 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after discharg . At this point in time, the major fiscion product poison, Xe , has nearly completely decayed away. Furthermore, the fuel ;

reactivity was found to decrease continuously from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 30 yects following dischurge. Therefore, the most reactivo i reint in time for a fuel assembly af ter discharge from go reactor can be conservatively approximated by removing the Xe .

1' 2.2.3.1 Method Validation l Realon 1 Calculation Method !

Methof.s confarm with I.NSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stetionary Pressurit.ed Water Reactor Plants,"

i I 2-7 1

l

L I

c? i b

Section 5.7, Fuel Handling Systear ANSI 57.2-1983, " Design 1 Objectives for IMR Spent Fuel Storage Facilities At Nuclear Power Stations," Section 6.4.2; ANSI N16.9-1975, " Validation of Calculational Kathods for Nuclear Criticality Safety," NRC ;

Standard Review Plun, Section 9.1.2, " Spent Fuel Storage"; and j the NRC guidance, )RC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications."

Reaion 2 Calculation Method g In the Re(aon e calculational method, the PHOENIX code has been validated by comparisons with experiments where isotopic fuel i composition has been examined following discharge from a l reactor. In addition, an extensive set of benchmark critical l experiments has been analyzed with PHOENIX. Comparisons between j measured and predicted uranium and plutonium isotopic fuel ;

compositions are shown in Table 2-3. gpe measuraments were. made on fuel discharged from Yankee Core 5 . The data in Table 2-3 j

\ shows that the agreement between PHOENIX predictions and t measured isotopic compositions is good.

]

The agreement between react?.ities computed with PHOENIX and the results of 81 critical benchmark experiments is summarized in l Table 2-4. Key parameters describing each of the 81 experiserts ]

are given in Table 2-5. These reactivity comparisons again show :

good agreement between experiment and PHOENIX calculations. )

i An uncertainty associated with the burnup-dependent reactivities computed with PHOENIX is accounted for in the development of the 1 Region 2 burnup requirements. A bias of 0.01 delta k at 30,000 ;

MWD /NTU is considered to be very conservative since comparison l between PHOENIX results and the Yankee Core experiments and 81 :

benchmark experiments indicates closer agreement. '

2.2.3.2 Criticality Analysis For Region 1 i The spongB fuel storage racks are described plates is in0.023 Section"2.1 2,The minimum loading in the poison gm B/cm ,

l The following assumptions were used to develop the nominal case ;

t KENO model for the Region 1 spent fuel rack storage of fresh ;

fuelt l

1. The B&W 15x15 fuel ansembly contains the highest enrichment authorized, is at its most roactive point in life, and no l credit is taken for any burnablo poison in the fuel rods. l l 2. All f 1 rodr, contain uranium dioxide at an enrichment of 4.2 '

l w/o over the entire length of each rod.

1 1

l I

2-8

,\.

l 1

0 l

l l

l

3. No credit is taken for any U* or U28 ' in the fuel, nor is any '

credit taken for the buildup of fission product poison material.

4. The moderator is pure water at a temperature of 68'F. A conservative value of 1.0 gn/cm3 is used for the density of water.

5. No credit is taken for any spacer grids or spacer sleeves.

t

6. The array is infinite in lateral and finite axial extent which allows neutron leakage from only the axial direction.

l- ,

7. The minimum poison material loading of 0.023 grams B-10 per '

square centimeter is used throughout the array.

The KENO calculation for the nominni case resulted in a Kg of O.9291 with a 95 percent probability /95 percent confidence level j uncertainty of 10.0054. !

The maximum Kg under normal conditions arises from consideration of mechanical and material thickness tolerances resulting from ,

the manufacturing process in addition to asymmetric positioning of fuel assemblies within - the storage cells. Studies of asymmetric positioning of fuel assemblies within t he storage '

cells has shown that symmetrically placed fuel assemblies yield i conservative results in rack Kg. The sheet metal tolerances are !

considered along with conctruction tolerances related to the l

cell I.D., and cell center-to-center spacing. For the Region j 1 racks this resulted in a reduction of the nominal 1.20 inch ,

water gaps to their minimum values. Thus, the " worst case" KENO mode.1 of the Region 1 storage racks contains minimum water gaps I j of 1.112 inch with symmetrically placed fuel assemblies. ,

t For normal operation and using the method in the above sections, the Kg for the Region 1 rack is determined in the following manner.

I M " borst + Baethod + Bpart + [@sworst) + @smethod) ,

l' where: :

rut ase KENO borst = wassemblies Kg includes symmetrically placed fuel material tolerances, and mechanical l

tolerances which result in spacings between assemblies L less than nominal.

B method bias determined from benchmark critical mothod = coleparisons.

l l

i

~

l

\

B = bias to account fcr poison particle scif-shielding.

part I

95/95 uncertainty in the worst case KENO Km.

ksworst =

v5/95 uncertainty in the method bias.

kanethod =

1 Substituting calculated values in the order 11sted above, the result ist j i

Kg = 0. 9 3 6 5 + 0. 008 3 + 0. 0 012 + (( ( 0. 003 5 ) # + (0.0018)2 ) = 0.9499 l 1

Since Xw is not greater than 0. 95 including uncertainties at a ;

95/9b probability / confidence level, the acceptance criteria for :

g criticality is met with fuel enriched to 4.2 w/o. )

i The average Xg of the benchmarks is 0.992. The standard deviation l of the bias value la 0.0008 delta k. The 95/95 one sided !

tolerance limit factor for 33 values is 2.19. Thus, there is a !

95 percent probability with a f 5 percent confidence level that the ,

uncertainty in reactivity, due to the method, is not greater than 0.0018 delta k.

Uncertaintics and biases due to mechanical tolerances during construction will have an effect or, the reactivity. The uost i important effect on re. activity of the mechanical tole,ratices is the l possible reduction in the water gap between 'se poison plates. 1 The worst combination of mechanical toleran.es is that which results in the maximum reduction in the water gap. For a single cell it ;s found that reactivity does not increase significantly j

! because the increase in reactivity due to the water gap reduction '

l on one side of the cell is offset by the decrease in reactivity due to the increased water gap on the opposite side of this cell. I j

The analysis, for the effect of nachanical tolerances, however, i assumed a worst case of a rack compor.d of an array of groups of ,

four cells with the minimum water gap between the four cells. The )

reactivity increasu of this configuration is included in the bese case Kg of the rack. It is included in the baso case since cells ,

, can be welded to a common (Jrid during manufacturing which is the l l ljkely cause of the water gap reduction.

Other nochanical tolerances which are not included in the analysis because worst case e,csumptions are used in the base case analysis include eccentric assembly positioning. calculations were performed which show that the most reactive condition is the assembly centored in the cell which is assumed in the nominal r case. Another example is the reduced widte of the poison plates.

l l

2-10 I

l l

1 i

No bias iA. included here since the nominal KENO case models the i reduced width explicitly. ]

The value of K s for Region 1 from this analysis is less than 0.95, {

including all uncertainties at a 95/95 probability / confidence ,

l level. Therefore, the acceptance criterion for criticality is ;

l tet. !

l.

l 2.2.3.3 Criticality Analysis For Region 2 l The nominal and maximum K.n for storage of spent fuel in Region 2 !

is determined usin; the methods described for Region 1 in addition l l to the methods described for Region 2. The actual conditions for .

l this determination are defined by the zero burnup intercept point i i in Figure 2-9. The KENO-IV computer code is used to calculate the ,

! storage rack multiplication ! actor with an equivalent fresh fuel I i enrichment of 1.63 w/o. Combinations of fuel enrichment and

! discharge burnup yielding the same rack multiplication factor as j at the zero burnup intercept <.ra determined with PMOENIX. t The following assumptions were used to develop che nominal cave ;

KENO model for the Region 2 storage of spent fuel:

1. The B&W 15x15 Staniard (STD) fuel assembly was analyzed for e

Region 2.

l l

2. The B&W 15x15 STD spent fuel assembly contains uranium j

dioxig fuel at an equivalent " fresh fuel" enrichment of 1.53 l w/o.U r

3. The moderator is pure water at a temperature of 68'F. A conservative value of 1.0 gm/cm3 is used for the density of i water.

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

I

j. 5. The array is infinita in lateral and axial exterit which ,

L procludes any neutron leakage from the array. >

w

6. The minimum poison material loading of 0.015 grams B-10 per l l square centimeter is used throughout the array.

t The KENO calculation for the nominal case resulted in a Kg of 0.9211 with a 95 percent probability /95 percent confidence 5.cVel :

uncerteinty of 0.0042.

The maximum K.,, under normal conditions was determined with a

" worst case" KENO model, in tb4 same manner as for the Region 1 storage racke. Thus, the " worst case" KENO model of the Region 2 storage racks contains minimum center to center spacing with symmetrically placed fuel assemblies. The uncertsinty ascociated ,

2-11 y

E 1 i

l with the reactivity equivalence methodology was included in the 1 development of the burnup requirements. I l

Based on the analysis described above, the following equation is ,

used to develop the maximum Q for the storage of spent fuel in j the crystal River Unit "I Region 2 spent fuel storage racks 4=Kworst + Baethod + ((ks)2 worst + IN"I method j l

where Ew rst = wpositions, rst case KENO Q that includes centered fuel assembly material to3 erances, and mechanical tolerance which can result in spacing between assemblien less than l nominal i Beethod = method bias determined from benchmark critical ;

comparisons .

B part = bias to account for poison partical self-uhielding

, Ksworst = 95/95 uncertainty it; the worst case KENO Kg Ksmethod= 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result ist j l Kw = 0. 9 2 59 + 0. 00 8 3 + 0. 0019 + ( ( 0. 0 04 3 ) 2 + (0. 0018)2)m = 0. 94 08 f l

i l The maximum 4 for Region 2 for this configuration is less than 1 0.95, including all uncertainties at a 95/95

probability / confidence level. Therefore, the acceptance criteria :

for criticality are met for storage of ent fue4 at an equivalent l

" fresh fuel" enrichment of 1.63 w/o U >

, 2.2.3.4 Storage of Censolldated Fuel The storage racks in Crysta) River Unit 3 are capable of storing an array of consolidated fuel rods (consolidation ratio of 2:1) l in each storage location. Analysis invo3ving consolidated fuel ,

storage show that if the consolidated fuel storage canister is i designed such that a consolidation ratio of 2:1 is maintained, the .

consolidated array is always less reactive than the unconsolidated array if the same type of fuel is stored in the same rack cell.

Therefore it can be concluded that the storage of arrays of single

, B&W 15x15 standard fuel assemblies with enrichments no greater ,

l than 4.2 w/o in Region 1, or meeting the initial burnup vs l

1 l 2-12 l

l

f' ] ,

f )

I I

enrichment requirements of Figure 2-9 for Region 2, will meet criticality criteria. Arrays of consolidated fuel rods meeting :

these same requirements will also meet criticality criteria. .

( '

2.2.4 Acceptance criteria For Cri:icality

)

2.2.4.1 Woutron Poison Material Surveillance program !

t The neutron absorber rack design includes a poison verification !

view-hole in the cell wall so that the presence of poison waterial i may be visually confirmed at any time over the life of the racks. !

Upon completion of rack fabrication, such an inspection is !

performed. This visual inspection, coupled with the Westinghouse >

quality assurance program controls and the use of qualified ;

Boraflex neutron absorbing material, satisfies an initial i verification test to assure that the proper quantity and placement l of material was achieved during fabrication of the racks. This ;

precludes the necessity for on-site poison verification.

The poison coupons used in the surveillance program will be '

r representative of the material used within Region 1 and Region 2 locations. They will be chosen from the production lot poison and certified to the same criteria as the production lot poison. The sample coupons will be of the same thickness and width as the poison used within the storage system. Coupon length will be at i least twice its width. Each poison specimen will be encased in l' 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 similar to that designed into the storage system. :

The jacket will be mechanically closed without welding in such a

manner as to retain its form throughout the use period yet allow rapid and easy opening without contributing mechanica) detaage to the poison specimen contained within.

A series of the jacketed poison specimens shall be suspended from rigid straps so. designed as to be hung on the outside periphery of e. Region 1 or Region 2 rack module. There is at least one coupon specimen from each production lot of poison. The specimens will be located in the spent fuel pool auch that they will receive l a representative exposure of gamma radiation. The spociman .

location will be adjacent to a designated storage cell with design ability to allow for removal of the strap, providing access to a particular specimen. In addition, one surveillance assembly is stored closc to the pool water surface in order to study the effect of pool water exposure.

As discussed in Section 3.7, irradiation tests have been previously performed to test the stability cnd structural

integrity of Boraflex in boric acid solution under irradiation.

These tests have concluded that there is no evidence of deterioration of the suitability of the Boraflex poison material L 2-13 l

i

I I

through a cumulative irradiation in excess of 1 x 10" rads gamma radiation. As more data on the service life performance of .

t Boraflex becomes available in the nuclear industry in the coming years through both experimentation and operating experience, FPC l

will evaluate this information and will. modify the surveillance :

program as determined warranted and justified.

FPC plans to perform an initial surveillance of the specimons I after approximately two years of exposure in the pool environment. '

During this surveillance, several specimens will be removed from the pool and examined. This examination is expected to include i visual inspection as well as other tests determined necessary to

verify that the performance of the Boraflex is consistent with the .

reported test recults. Based on the results of this initial surveillance, FPC will determine the scheduling and extent of additional surveillances so as to assure acceptable material performance throughout the life of the plant.

2.2.4.2 Decay Heat calculations for Spent Fuel The addition of high density spent fuel racks to Spent Fuel Pool "B" increaser the total storage capability of Spent Fuel Pt."ils "A" and "B" to 19/3 cores normal storage plus the reserved stocage ,

spaces for the full core discharge condition. The maximum heat i

, load generated in the pools is based on off-loading the full ,

This reactor core at the and of a two i maxipum decay heat load of 33.5 x 10, year refueling cycle. Btu /hr. l

. x 10 Btu /hr. for the off-loaded full core plus 0.5 x 10 Btu /hr. '

for all the previously-removed spent fuel assemblies stored in the l spent fuel pools. For the off-loaded full core, infinite ,

irradiation is assumed prior to refueling and a cooling period of seventy two hours is conservatively assumed after reactor shutdown. I l

Decay heat load calculations have been generated to determine the maxirum steady state temperature of the spent fuel pools for the established decay heat load of 33.5 x 10' Btu /hr. This calculated maximum pool temperature is 157'F for the case of ooth spent fuel cooling loops in operation for removal of the aforementioned decay >

heat load. Thirt calculated pool temperature is also based on the normal design cooling water (from the Nuclear Services Closed .

Cycle Cooling System) temperature of 9 5'F . The 157'F pool '

temperature is acceptable since, as indicated in GAI Report 1949, the structural capacity of the spent fuel pools is considered adequate for a steady state water temperature of 160'F with water in both pools.

For the single failure case or loss of one spent fuel cooling ,

loop, the pool temperature could exceed 210'F if supplemental cooling were not available. Starting with a pool temperature of 16 0*F , the minimum time for the pools cc reach 190*F is 2-14 c- - w e- - TMr-* --mg---- Y

t approximately eight hours. This is sufficient time to perform the necessary repairs or provide an alternate source of cooling water, such as the Plant Fire Protection Syston (using temporary connections and hoses) or by utilizing the Decay Heat Removal System which is permanently piped and valved to the Spent Fuel Cooling System for maintaining acceptable spent fuel pool temperatures.

2.2.4.3 Thermal Hydraulic Analysis for Spent Fuel Cooling Local Puel Bundle Thermal-Hydraulics The purpose of thermal-hydraulle analysis is to determine the maximum fuel clad temperatures which may occur as a result of using the spent fuel racks in the Crystal River 3 spent fuel pool "B".

Criteria The criteria used to determine the acceptability of the design from a thermal-hydraulic viewpoint is summarized as follows:

1. The design utst allow adequate cooling by natural circulation ,

and by flow provided by the spent fuel pool cooling system. ;

The coolant should remain subcooled at all points within the !

pool when the cooling system is operational. When the i cooling system is postulated to be inoperable, adequate cooling implies that the temperature of the fuel cladding !

should be sufficient.ly low that no structural failures would occur and that no safety concerns would exist. j

2. For normal operations, the maximum pool temperature shall not exceed 14 u ' i. For conservatism, the temperatures of the storage racks and the stored fuel are evaluated assuming that the temperature of the water at the inlet to the storage cells is 150'F during normal operation.

r

3. Thw rack design must not allow trapped air or steam. Direct !

gamma heating of the storage cell walls and the intercell water must be cot.nid: red.

4.

I The heat output of a consolidated fuel assnmbly, assuming the consolidation of 2 year old fuel, is enveloped by the normal storage assumptions.

t Kev Assumntions ,

1. The nominal water level is 24 feet above the top of the fuel l- storage racks.

l l

2. The maximum fuel assembly decay heat output is 50.42 BTU /sec per assembly following 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months decay after shutdown. For ,

2-15

l

/

conservatism, this value will be used for all Region 1 and Region 2 storage locations.

3. The navinum temperature of the water at the inlet to the storage cells is 150'T when the cooling system is operational. I

4. Under postulated accident conditions, when no pool cooling systems are operational, the maximum temperature at the inlet to the cells is assumed to be equal to the saturation '

temperature of 212'F at the top of the pool. This number is !

conservative, since it is assumed that sufficient cold water I is constantly being added to the pool to maintain water j level. !

Analvtical Method and calculations A natural circulation calculation is employed to determine the therag-hydraulic cells .

conditions within the spent fuel storago The model used assumes that all downflow occurs in the :

peripheral gap between the pool walls and the outermost storage l cells and all lateral flow occurs in the space between the bottom of the racks and the bottom of the pool. The effect of flow area !

blockage in the region is conservatively accounted for and a I' multi-channel formulation is used to determine the variation in axial flow velocities through the various storage cells. The )

hydraulic resistance of the storage cells and the fuel assemblies i' is conservatively modeled by applying large uncertainty factors to loss coefficients obtained from various sources. Where necessary, thu effect of Reynolds Number on the hydraulic resistance is considered, and the variation in momentum and elevation head pressure drops with fluid density is also determined.

The solution is obtained by iteratively solving the conservation equations (mass, momentum and energy) for the natural circulation loops. The flow velocities and fluid temperatures that are obtained are then used to determine the fuel cladding temperatures. An elevation view of a typical model is sketched in Figure 2-10 where the flow paths are indicated by arrows. Note that each cell shown in that sketch actually corresponds to a row of cells that is located at the same distance from the pool walls.

This 30 more clearly shown in a plan view, Figure 2-11.

As shown in thst sketch, the lateral flow area underneath the p storage cells decreases as the distance from the wall increases.

This counteracts the decrease in the total lateral flow that occurs because of flow that branches up and flows into the cells.

This is significant because the lateral flow velocity affects both the lateral pressure drop underneath the cells and the turning losaes that are experienced as the flow branches up into the 2-16 i

Il d

cells. These effects are considered in the natural circulation !

analysis.

The most recently discharged or " hottest" fuel assemblies are i assmaed to be located in various rows during diffarent l calculations in order to ensure that they may be placed anywhere within the pool without violating safety limits. In order to ,

simplify the calculations, each row of the model must be composed '

of storage cells having a uniform decay heat level. This decay heat level may or may not correspond to a specific batch of fuel, '

but the model is constructed so that the total heat input is correct. The " hottest" fuel assemblies are all assumed to be l placed in a given row of the model in order to ensure that ;

conservatively accurate results are obtained for those assemblies.

In fact th2 nost conservative analysis that can be performed is to assume that all assemblies in the pool (or rows in the model) t have the same maximum decay heat rate. This maximizes the total i natural circulation flowrate which leads to conservatively large ;

pressure drops in the downconer and lateral flow regions which l

reduces the driving pressure drop across the limiting storage !

i locations.

I Since the natural circulation velocity strongly affects the ;

temperature rise of the water and the heat transfer coefficient within a storage cell, the hydraulic resistance experienced by the !

flow is a significant parameter in the evaluation. In order to i minimize the resistance, the design of the inlet region of the racks has been chosen to maximize this flow area. Each storage '

cell has one or more flow openings as shown in Figure 2-12. The use of these large or multiple flow holes virtually eliminates the '

possibility that all flow into the inlet of a given cell can be !

4 blocked by debris or .,ther foreign material that may get into the '

pool. In order to determine the impact of a partial blockage on the thermal-hydraulic conditions in the cells, an analysis is also performed for various assumed blockages.

The analyses that have been described only address the flow !

through the storage cells. As noted in the discussion of 5 criteria, it is also required that the flow and temperatures in the axial gap between adjacent storage cells for Region 1 racks be evaluated. In order to preclude the possibility of stagnant ;

conditions in these gaps, flow relief areas are provided at the '

location of the grid support structures as shown in Figure 2-13.

This flow area also ensures that air or steam cannot be trapped ,'

in the rack structure. The thermal-hydraulic conditions in the gap region are evaluated by using a parallel path thermal-hydraulic model of the gap and cell under consideration.

3-17

Results l

Normal Operation Basist a) Cooling System Operational b) 150 hours after shutdown-Decay Heat = 50.42 BTU /second/ assembly c) Uniform decay heat loading in pool - No credit for lower actual heat input d) Peak rod has 60 percent more heat output than average rod e) All storage cells filled.

Results of the analysis show that no boiling occurs at any point within the storage racks when the normal cooling system ..s in operation or whenever pool temperature is maintained within its i allowable limits. Water temperatures in the gap between cells are lower than inside the cells, and boiling does not occur in the inter-cell gaps.

Flow Blockage Analysis Basist a) 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months after shutdown '-

b) Temperature of water at inlet to storage racks = 150*F L Results of the analysis show that should up to 80% flow blockage occer, there would be no boiling in the water channels between the cells or inside the cells. Because of the large or multiple flow openings that are used in the Westinghouse storage racks, it j is very improbable that a complete blockage could occur.

Abnormal Condition Under postulated accident conditions where all non-Category 1 spent fuel pool cooling systems become inoperative, there is an alternative method for cooling the spent fuel pool water.

Although it is highly unlikely that a conplete loss of cooling capability could occur, the racks are analyzed to this condition.

i l- 2-is 1

l i

m. __ __. _ ._ .,e- - - . -

Basist l i

a) No pool cooling implies that temperature of water at inlet to opent fuel racks is 212'F which corresponds to ,

the saturation temperature at the top of the pool.

b) The nominal water level of 24 feet above the top of the i racks is maintained. l c) Uniform decay heat loading in pool - No credit for lower ,

actual heat input.

d) The assemblies that are evaluated are initially put into the pool at 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months after shutdown - Decay Heat = l 50.42 B1U/Second/Assen6bly. ;

e) The peak rods are assumed to have 60 percent graatar !

heat output than average rods.

l' f) All storage cells are filled and all downflow occurs in the peripheral gap.

Results of this analysis show that due to the effects of natural ;

circulation, the fuel cladding temperatures are sufficiently low j to preclude structural failures.

2.2.4.4 Potential Fuel and Rack Handli ng Accidents y The Fuel Handling Accident outs.de the Reactor Building is !

postulated as the dropping of a fuul assembly into the spent fuel storage pool which results in darage to a fuel assembly and the :

release of the volatile gaseous fission products. The original ;

analysis of this accident as drcumented in the Crystal River 3 i FSAR, section 14.2.2.3, was bast.d on the assumption that the outer row of fuel rods (56) in the dropped assembly were damaged. The USAEC Division of Reactor Licensing (DRL) acknowledge the use of j the 56 fuel pins in December, 1965. Subsequently, the DRL issued j

.the CR3 Safety Evaluation in 1974 in which the staff assumed that :

all 208 rods in the dropped assembly experienced damage. The Fuel '

Handling Accident analysis was updated in 1978 to reflect damage

f. '

to 208 fuel pins and the results incorporated into an FSAR update and associated 10CFR50.59 safety evaluation in 1988. As documented therein the results of the analysis met the applicable i dose acceptance criteria as provided in 10CFR100.ll.

Since the radiological consequences of this accident are only l based on the failure of the dropped assembly and the latest analysis assumes that all 208 fuel pins in the dropped assembly are damaged, the installation of the high density racks in the .

spent fuel storage pool has no effect on the dose analysis and the resultant conclusions. I L

2-19 l l

t _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _. .- - - . - - . . _ - -

TABLE 2-1 RACK MODULE DATA Realon 1 Reaion 2 Number of Storage 174 641 Locations Number of Rack (1) 9 x 10 (1) 9x11 Arrays (1) 9 x 10 mod (4) 9x12 (modified) (1) 10x11 Center-to-Center 10.60 9.17 Spacing (Inches) cell Inside Width 9.00 9.00 (Inches)

Type of Fuel D&W 15 x 15 and/or B&W 15 x 15 and/or Consolidated fuel Consolidated fuel (2:1 Ratio) (2:1 Ratio)

Rack Assembly (9x10) (9x11)

Dimensions (Inches) 96x107x167 84x102x167 (9x10 mod) (9x12) 96x107x167 84x111x157 (10x11) 93x102x167 Dry Weights (1bs) 17,800 (9x10) 9,700(9x11)

Per Rack Assembly 16,900 (9x10 mod) 10,500(9x12) 10,800(10x11)

L 1

3~20 l

TABLE 2-2 BENCHMARK CRITICAL EXPERIMENTS UN General Enrichment Separatimy Soluble K Descriotion w/o U235 Reflector Material B-10 DDa eff

1. UO2 rod lattice 2.46 water water 0 0.9857 i .0028

2. UO2 rod lattice 2.46- water water 1037 0.9906 i .0018

3. UO2 rod lattice 2.46 water water 764 0.9996 i .0015

4. UOz rod lattice 2.46 water B4C pins 0 0.9914 i .0025

5. UO2 rod lattice 2.46 water B4C pins O O.9891 i .0026

6. UO2 rod lattice 2.46 water B4C pins 0 0.9955 + .0020

7. UO2 rud lattice 2.46 water B4C pins 0 0.9889 + .0026

8. UO2 rod lattice 2.46 water BAC pins O O.9983 i .0025

9. UO2 rod lattice 2.46 water water O O.9931 i .0028

10. UO2 rod lattice 2.46 water water 143 0.9928 i .0025

11. UO2 rod lattice 2.46 water stainless steel 514 0.9967 i .0020

12. UO2 rod lattice 2.46 water stainless steel 217 0.9943 i .0019

13. UO2 rod lattice 2.46 water borated aluminum 15 0.98S2 i .0023

14. UOg rod lattice 2.46 water borated aluminism 92 0.9884 i .0023

15. UOg rod lattice 2.46 water borated aluminum 395 0.9832 i 0021 2-21

__ . .,- - . ___ _ - . _ - . - - _ _ . - - _ _ . - . _ .-- ._, _ -.- _-._. _ ~ _ _ . - _ _ _ _ .

TABLE 2-2 (CONTINUED)

BENCHMARK CRITICAL EXPERIMENTS U*N General Enrichment Separating Soluble K

. Descrintion w/o U235 Reflector Material B-10 Dom eff 4

16. UO2 rod lattice 2.46 water borated aluminum 121 0.9848 i .0024

17. UO2 rod lattice 2.46 water borated aluminam 487 0.9895 i .0020

18. UO2 rod lattice 2.46 water borated aluminum 197 0.9885 i .0022

19. UO2 rod lattice 2.46 water borated aluminum 634 0.9921 i .0019

20. UO2 rod lattice 2.46 water borated aluminum 320 0.9920 i .0020 21., UO2 rod lattice 2.46 water borated aluminum 72 0.9939 i .0020

22. U metal cylinders 93.2 bare air 0 0.9905 i .0020

23. U metal cylinders 93.2 bare air 0 0.9978 i .0020

24. U metal cylinders 93.2 bare air 0 0.9947 i .0025

25. U metal cylinders 93.2 bare air O O.9928 i .0019

26. U metal cylinders 93.2 bare air O O.9922 i .0026

27. U metal cylinders 93.2 bare air 0 0.9950 i .0027

, 28. U metal cylinders 93.2 bare plexiglass 0 0.9941 i .0030

29. U metal cylinders 93.2 paraffin plexiglass 0 0.9928 i .0041

30. U metal cylinders 93.2 bare plexiglass 0 0.9968 i .0018 2-22

_- . ...;-. .s

_- ._y~ . _

~ ' ~

~ - . .

,t

~ -

TABLE 2 (MNTINUED)

BENCHMARK CRITICAL EXPERIMENTS U'6I s-General Enrichment- Separating Soluble K Description w/o U235 Reflector -Material B-10 Don eff ,.

31. U metal cylinders 93.2 paraffin plexiglass 0 1.0042 1,.0019

'32. U metal cylinders 93.2 paraffin plexiglass 0- 0.9963 i .0030 33.- U metal cylinders 93.2 paraffin plexiglass 0 0.9919 1 0032 3

2-23

_ . . , , . _ _ . _ _ _ - _ . _ _ - . . _ _ . . , . ~

l.-

!~

t.

k: f hI TABLE 2-3

<, COMPARISON OF PHOENIX ISOTOPIC PREDICTION TO YANKEE CORE 5 MEASUREMENTS Quantity % Difference l (Atom Ratio)

U235/U -0.67 U236/U -0.28 i

U238/U -0.03 PU239/U +3.27 PU240/U +3.63 PU241/U -7.01 3 PU242/U -0.20 PU239/U238 +3.24 MASS (PU/U) +1.41

, .FISS-PU/ TOT-PU -0.02 ;

- POrcent difference is average difference of ten compP.risons for each isotope.

P 2-24

~ , - . . - ,

g= --

i I N

TABLE 2-4 Benchmark Critical Experiments, ,

PHOENIX Comparison '

De2cription of Number of PHOENIX krr Using Exoeriments Exoeriments Egoerimen,t Bucklinas UO2 ' ,

Al clad 14 .9947 SS clad 19 .9944 Borated H 2 O 7 .9940 Subtotal 40 .9944 U-M-tal.

Al clad 41 1.0012 l

TOTAL 81 .9978 l ,

l' L

E l r 1 ;

I i'

2-25 l

_--~ -

.= - -

9 TABLE 2-5 DATA FOR U METAL AND U02 CRmCAL EXPERIMENTS Case Ce:1 AO ICS/U Fuel Penet Material CInd Gad Imrtice ' B-10 Number Type - U-235 Ratio Dermry Diameter Oad OD Hrkness Pitch PPM (G/d') (CM) (CM) (CM) (CW) 1 licza 1328 102 7.53 13265 Nmmuni L6916 .07110 12050 0.0 2 IIcza 1328 1 95 7.53 13265 . unmune 1.6916 .07110 23590 00 3 Hexa 1328 4.95 7.53 L5265 Ahnnment L6916 .07110 23120 0.0 4 IIcza 1328 3.92 7.52 .9855 Nummum 1.1506 .07110 1.5580 0.0 5 ficxa 1328 4.89 7.52 .9855 Aluminum L1506 .07110 1 6520 0.0 6 Ikza 1328 2.88 10.53 .7728 Ahnmous 1.1506 .07110 L5580 0.0 7 licxa 1328 3.58 10.53 .9728 Alummum 1.1506 .07110 1.6520 0.0 8 Ilexa 1328 4E3 10.53 .9728 Alummum 1.1506 .07110 13060 0.0 9 Square 2.734 11C 10.18 .7620 SS.304 E594 .04085 LO287 0.0 10 Square 2.734 2.92 10.18 .7620 SS-304 E594 .04085 1.1049 0.0 11 Square 2.734 3M 10.18 .7620 SS-294 E594 .04085 L1938 - 0.0 12 Square 2.734 7D2 10.18 .7620 SS-304 .8594 .04085 L4554 0.0 13 Square 2.734 8.47 10.18 .7620 SS.304 .8594 .04085 L5621 0.0 14 Square 1734 1038 10.18 .7620 SS-304 E594 .04085 1.6891 0.0 15 Square 2.7M 230 10.18- .7620 SS-304 3594 .065 1.0617 0.0 13 Square 2.734 4.51 10.18 7520 55-304 E594 .04085 L2522 0.0 17 Square 1 745 2.50 10.27 .7544 55-304 .86no .04060 1.0617 0.0 18 Square 3.745 431 1037 .7544 55-304 E600 .04060 L2522 0.0 19 Square 3.745 4.51 1037 .7544 SS.304 E6ne .04060 L2522 0.0 20 Square 1 745 4.51 1037 .7544 SS-304 E600 .04060 L2522 456.0 21 Square 3."*U 4.51 1037 '.7548 55-304 E600 .04060 L2522 709.0 12 Square 3.745 4.51 1037 .7544 55-304 8600 .04060 1.2522 1260.0 23 Square 1 745 431 1037 .7544 55-304 .8600 .04060 1.2522 1334.0 24 Square 3.745 4.51 1037 .7544 SS-334 E600 .04060 1.2522 167.0 25 Square 4.069 235 9.46 L1278 SS-3% 1.2090 .04060 L5!13 0.0 26 Square 4.069 2.55 9.46 1.1278 55-394 1.2090 .04060 1.5113 33910 27 Square 4.069 2.14 9.46 1.1278 SS-304 12090 .04060 1.4500 0.0 28 Square 1490 234 10.24 1.0297 Ahunmum L2060 .08130 1.5113 0.0 29 Square 1037 164 9.28 L1268 $5-304 L1701 .07163 13550 0.0 30 Square 3.037 8 16 9.28 L1268 55 304 1.2701 .07163 11990 0.0 31 Square 4.069 159 9.45 1.1268 55-304 L2701 .07163 1.5550 00 32 Square 4.069 3.53 9.45 1.1268 SS 304 L2701 .07163 L6840 0.0 33 Square 4.069 8.02 9.45 1.1268 SS-304 L2701 .07163 11980 0.0 34 Square 4.069 9.90 9.45 1.1268 55 304 1.2701 .07163 23810 0.0 35 Square 2.490 2.84 10.24 1.0277 Alununus- 1.2060 .0R130 L5113 1677.e 36 IIcza 1996 2.06 1938 '1.5240 Ahnmaam 16916 .07112 11737 0.0 37 ficxa 2.096 3.09 1938 1.5240 Mummau 1.6916 .07112 14052 0.0 38 Ilexa 1096 4.12 1038 L5240 Alummum 1.6916 .07112 2.6162 0.0 39 11cza 2.096 6.14 1938 L5240 Alummum 1.6916 .07112 19691 0.0 40 IIcza 1096 8.20 1038 15240 Aluennum L6916 .07112 33255 0.0 2-26 m - _ ;

s -

TAB'.E 2-5 DATA FOR U METAL AND UO2 CRTTICAL EXPERIMENTS (Continued)

Case Ceft NO H2%U Fuci Pellet Material CInd Gad Lattre B-10 Nunser Type U.235 Ratio Densery Diameter Clad OD Thickness Pitch PPM (G!CC) - (CM) (CM) (CM) (CM) 41 licza 1307 1.01 18.90 15240 Alummum 1.6916 .07112 11742 0.0 42 licza 1307 1.51 18.90 1.5240 Alunnnum 1.6916 #7112 2.4054 0.0 43 Hexa 1307 2.02 18.90 13240 Alum mum 1.6916 .07112 16162 OS 44 Ilexa 1307 3.01 18.90 1.5240 Alummum 1.6916 .07112 2.9896 0.0 45 11cxa 1307 4.02 18.90 1.5240 Almmme. 1.6916 .07112 33249 0.0 46 Ilexa 1.160 1.01 18.90 1.5240 Mummum 1.6916 .67112 11742 0.0 47 licza 1.160 1.51 18.90 13240 A!ammum 1.6916 .07112 2.4054 0.0 48 licza 1.1M 2.02 1&90 13240 Alummum 1.6916 E7112 2.6162 0.0 49 licxa 1.160 3.01 1190 1.5240 Alummum 1.6916 .07112 2.9906 0.0 50 licza 1.160 4.02 1190 1.5240 Alummum 1.6916 .07112 33249 0.0 S1 IIca 1.040 1.01 18.90 15240 Aluminum 1.6916 .07112 11742 0.0 52 IIcza 1.040 1.51 18.90 1.5240 Alummum 1.6916 .07112 14054 0.0 53 IIcza 1.040 102 18.90 1.5240 Alummum 1.6916 .07112 2.6162 0.0 54 Ilexa 1.040 3.01 1130 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 55 licra 1.040 4.02 18.90 1.5240 Mummum 1.6916 .07112 33249 0.0 56 Ilexa 1307 1.00 18.90 .?R30 Nummum 1.1506 A7112 1.4412 0.0 57 Ilexa 1307 1.52 18.90 .9630 Alummum 1.1506 .07112 1.5926 0.0 58 IIcza 1307 102 1&90 .9830 Naa inum 1.1506 .07112 1.7247 0.0 S9 Ilexa 1307 3.02 18.90 .9830 Alt minum 1.1506 .07112 1.9609 0.0 60 licxa 1307 4.02 18.90 .9830 Alumine 1.1506 .07112 11742 0.0 61 licxa 1.160 132 18.90 .9830 Alemamm 1.1506 .07112 1.5926 0.0 62 Ilexa 1.160 102 18.90 .9830 . .ninum .".1506 .07112 1.7247 0.0 63 IIcza 1.160 3.02 13.90 .9833 Aluminum 1.1506 .07112 1.9509 0.0 64 IIcza 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 11742 0.0 65 IIcza 1.160 1.00 18.90 .9830 Ahmmmm I? ~06 .07112 1.4412 0.0 66 IIcza 1.160 I5i 1190 .9830 Alummum 1.1506 .07112 13926 0.0 07 IIcza 1.160 102 18.90 .9830 Alemmum 1.1506 .07112 1.7747 0.0 68 Ilexa 1.160 3.02 1&90 .9830 Alummum 1.1506 97112 1.9609 0.0 69 Ilexa 1.160 4.02 18.90 .9830 Alummum 1.1506 .07112 11'42 0.0 70 IIcza 1.Os0 133 18.90 19.050 Aluminum 10574 .07620 23687 0.0 71 licra 1.040 1.58 18.90 19.050 Alummum 10574 .07620 3.0086 0.0 72 IIcza 1.040 1.83 18.90 19.050 Alummum 2.0574 .07620 3.1425 0.0 73 IIcza 1.040 231 18.90 19.050 Alum mum 2.0574 .07620 33942 0.0 74 licza 1.040 183 18.90 19.050 Alummum 10574 .07620 3.6284 0.0 75 IIcza 1.040 3.83 18.90 19.050 Alummum 10574 .07620 4.0566 0.0 76 IIcza 1310 102 18.88 1.5240 Almmone 1.6916 .07112 2.6760 0.0 77 IIcza 1310 3.01 18.88 1.5240 Alummum 1.6916 .07112 2.9900 0.0 78 licxa 1.159 2.02 18.88 13240 Alummum 1.6916 .07112 2.6160 0.0 4 79 licza 1.159 3.01 18E8 15240 Aluminum 1.6916 .07112 2.9900 0.0 80 licxa 1312 1 03 18.88 S 830 Aluminum 1.1506 .07112 1.7250 0.0 81 IIcza 1312 3.02 48.88 .9830 Alummum 1.1506 .07112 1.9610 0.0 2-27

.+

.e

-m ,e , . _

~ - - .

.s -=.__.

-n 3y -

~ - -

=

390.00 REF.

4.00 f

- '-1.30 TYP. -- - 4.01 4

- 92.53 - 83 38 " -

109.40 =*r 95.00 -

1, 4 i ,

I i i

! R2 R2 R1 95.80 R1 to X 11 e X 11 to X 9 9 X 10 - 33 8 08 NEF.

m 101.70 i N 906.40 i 2 1.29 -

8 TYP. {7 q, .

4 i 3 [ ^ '

g g term REr. j , 1 i.. iso -

t* .< l m $ 83.36 R2 -R2 .

y yy 1YP. 9 X 12 9 X 12 = 166.00 91EF. -

N l

b. $M A M -

7 h R2 e O" 9 X 11 R2 S X 12 i

y @

O O t*

o 6 l

_I to =

J g 8.00 8 - -- 110.8T TYP. - -

1.50' - + 6.76

^'

234.00 REF.

CRYSTAL RIVER 3 POOL "B" RACK LAYOUT REGION 1: 174 LOCATIONS @ 10.60" CTC PIGION 2: 641 LOCATIONS @ 9.17" CTC BT5' TOTAL STORAGE LOCATIONS i

. _ - ,4.- ~ - f m-~ g3,- % . - .wy., a +p..s.g ,,99 y .g -

---m._ v. ,%-i. w as e *.m e.

-ww.e_ e * - - - - - - - - - - - - - - = - 'm- - - - -- ==-----i=

b 3 i 1

s k

g

g

% # \

\ # \ * %

% # \

A

\ # %

% % # \

\ # % # %

% # % # \

- # s # % ** % # %

$#g # % # \ #% # %

s %.

g

% # \

e%

de g # % # % e g a %

T P #% # %

% /

- % # % h

% # g I

) N # % # % N

$ # % # % g

% 9 - g #% -

{ # %

4 .

1 - A

\ #.% >

k. .g 4"

?,s'$,y)' -

% \<*p '- ks q n 0 (18' I

$%g9 9 i

[

l - -

I l

g s ,I i

i ,

I v ,

a

'y

- 7 gtj ,***

%[ ,

h

/g i ll FUEL STORAGE RACK MODULE (REGION 1)

, 'RE 2-2 2-29 l , ,

l.

$ d'c

.-. 1

,,,, _ 10060 =

= 1 il P

?

t

% ADJACCHT CELL

- 9.00 50V AR E - , AS S E M B LY

% TOP GA10 <

ASSEMBLY 1 .. -

l-

.] .i I

,, -- Po l%C N M AT ER I At.

CELL.

1Q.00 REF. - ASSEMBLY l' ...

l . . __ &

l d .

t o g BOTTC u GR10

$l jQ ASSEMBLY t : :

4 l l .

D ! l l

e 5 _

,, v, .y a .

< as. wi . ./ . m i 4 g

i. B ASE P L AT E

f. 5.50 RET.' ,

[ ~

. SUPPORT BLOCX 7

/ LEVELING P AQ A / 3- ASSEMBLY NN _ wrs

~ ~C +- POOL LINER g'

, g - [,4 CONCRE7E PLATE teeTTSSCALCl

~

t REGION 1 MODULE CROSS-SECTION TIGURE 2-3 !

2-30 V

8 _

- . . - . . . . . _ . , . - _ ~ -

-, g y4 6m e 6 mue - as A +Jn4 A ^ 6-MvA- =^ e m o*a-- 6 A '

'fg L N"OW' -,M'

' :13 ., l' , . i L. ; t

, -. J l' ,f's

-w

t. i

l. ,y -

f 1- t N

. 1 1 i 1 1

' . l J J l I I

( J 1 S

1 ,

i J

j J

s i

J J A

s t J

Y' l i'

jg A . i i -

3 i i s d J J#

i

. i l t i I a a a i "O ! i j 3 i J i i: i J A A i i

g J t a j i 1:

i g<t A J

j J i

i

l. i J i i

f

]e

%s&%

c ,8 ,

f.

i e b J .

L b 4

e e '

1 \

s REGION 2 FUEL STORAGE RACK MODULE FIGURE 2-4

.t.

g *h-2-31 ,

- r, -+, . - - - . , - - - -,,,,m_,r,.,-

, .. a

}{'k, ' 3.

2,'l,. 9.17 RE F. )

4 '

o CELL --

I vN ~

l sA _

> ASSE MBL m

I I

1 l

! l Y -

i PolSON h I l .

W AT E R I AL i i- ~

! m# 5

. N. --

l e 162.00 REF. { !

\

l I .

l 1- P j . i WRAPPER j )

'I A!!ille I

i i l'

I v .

l

),

g

[

r,n c n wwm v

. ww//in

\ BASE PL ATE

5. 5 0, R E F . SUPPORT BLOCK E. ,.

k

\

. ----- L E YE L IN G P AD ASSE MBL Y

-_2 POOL L mtR

[4 l CCNCRETE J PL ATE REGION 2 MODULE CROSB-SECTION FIGURE 2-5 g.

2-32

.--+++s -,-..e, ,e . - - - . , - -. . , , . - . , . . - , <--.-..----r - = ~ - - - , ,-

.. - . ~ . . - . .. . - . . . . . - . . . . . . . . . - .

8 l

d

..t'. .

i <

j 9.00 l cett 1.o. tresc u.

T YP.

- . .- e.

[ l _ I i

t.814 4 '50. '

NON.CRLL l.0.

J TYP.

i I

+ 1 1 +

l M.11 vve.'

. . I f %

t

< l + i

, 1 L

l-

{ j L i 8l.17 7 Tf.

k REGION 2 F.C,DULE TOP VIEW i j.

FIGURE 2-6 l ;.

1, '

2-33

,.s

) ',..

0 Q= a .. i

c- ,

W. ,

i

[yi "l

4 10.60 CENTth-to-CDfTER $?ACINC --

l fr Q -

q ;- l l

l Q 4-- 7.50 BORATLEX + < g* ss 5 : !

s h !

- l : -

i h !

___ j ._ _ _

+ __

! .+

- l -

- l- -

3 - 9.00 CELI. I .D. - L g j l : j l

I' : :

3 w r,

~ '

s. l g I A 4 d e .

~

< >; 1.20 CELL-To-CELL CAF l

l l' -

l L L I

, .060 INNER CELL WALL

.120 CAP y .CB5 BORATLEX l- i

)y

_ \, _.

4 4(\ \ \ \ \ \ \ \ \ \ \ \ V i, d 6 t

l^ .020 VRAPPER l

l

~

l REGJ.ON 1 CELL LAYOUT FIGURE 2-7 i

2-34

' k w <.e.~ . . - - - - - - r,-.,, - , - , . . - , - - - - - - . - - , - - -

.si i

N[ e--- 9e17 CENTER-TO-CENTER SPAC1pC -+

.I L

.V~

il I . - ..

I a . ;

I 1

I d 0.994 NON-CELL 1.D. ;

9.17 -

p o s i r

Y D%

asumammiA- ,ff .

.032 I e 7.50 BORAlLEX

~ 15 8 p

5 3

[ .

~

l i

~

t. .

.y._

4 -

- l '

E

. g c 9.00 CE!.L 1.D. ;-

r w

e pas 0 -

m D 5- % y;;- e

-Q e c .

Y dl H f'

.060 INNER CELL WALL

, .056 BORA" LEX

,, .093 CAP i e 4 5 y ((T\\\\\\ \\\\\ V t 'A i 1

.020 VUJPER REGION 2 CELL LAYOUT FIGURE 2-8

.x-2-35 s

\'

,_ 3. ,

. -a, -

. r _

ts y< y

.y, w v

.\f '\

t1 -c

~.

p

, c il 35 i

i. i i i i e

l '. >

'1

\l .l

l l l D l i l .I I I i 1 1

1

. 1.

1 ;

s i 1 1 'l 1 1 4 i.

)g

i. t- t I - t / -e I I I I I i I i i i i I

\ i

_ .l l . .l. . . . 4 _ .. .. ..4. .

\ l 'I l l I - l u n.

i ACCEPTABLCl i I I I '

.; y:s25

> , , , t 7 . ,

' s >

1 I l l 1 I

y. 1 y,o E I l-

, _ .L. .T*--

i I i l

,q - '* -

**

I - - '

-T---

I I I i I I 20 I s I 1

~~

/e 1 1 I

l i I I I l l l l l l l I i

,w , .J .. L . J. .. .J . L., . .'. _

I O I i 1 l l

< ac '

l I I I l- 1 i ;

\, y -

; 3.. ;

'.I l i i Y

O 3 .__,.

i I

-r-1 I [Ir l l

---r----T---

I l

a '

'g i I'

l 1 1 NOTACCEPTABLEl1

. I w l I I vi n,

to .I i i i 1

C l I i i l I

. ..J . l_ .. . . .L . . .J. .. L .. .. .L .

I I I I I I i 1 i l i I I

f s. I I l' l I l ll y '

'l l l I l I a .. p .. .t-- --l--- --*i------i---

1 I l l I l l l l l ,

l i 1 l

.5 2.0 2.5 3.c 3.5 4.0 4.5 y U-235ENRICdENT(t/0) . . . . . _ . . . - - .-

l MINIMUM BURNUP V.S. INITIAL ENRICHMENT l

1' FOR REGION 2 RACKS

,, FIGURE 2-9 '

s

% 2-36 l p, r

= v -. .- _ . . _ _ . 3 . __ . _ . . _ _ _ _ _ _ _ - _ . _ _ . _ _ - , _ _ _ _ _ , _ _ _ _ _______.a__ . _ _ _ , _ - . .

.c - . . . - .. .- .__.- .-

p ,:.

CJ6TCe guaracg ;

l

, g;

. l Q. l

~

j 0tkACr5 N A I s, p cVT

<, o ,

4, o o o o o f

,s l

r a- A x- ., _g- , o c e --

I ll o ,

t -

, I c

ACW ;

ib '

/

r I: 2 3 -4 5 E 8 S a, o A i, o o 1 ,

o o

.[

l :

t t . t l

1 --

a y

p

!' t I O

1 '

v I ..

i z y .

8 .

j ' ,\ . y N\ili . . . . . . ,

P 7 i HCICHT Asuyg rLCCR l[

N/ _

~

7_ ;

L Af ra At rLow a tG10N l l t

A SPENT FUEL POOL NATURAL OIRCULF. TION MOD 1:L Q. I

'N (ELEVATION VIEW) u\_ + '

FIGURE 2-10 '

r 4

I 2-37 a

, . .s i I 1 .' s. _ _ _ , _ _ . ., , _ _ _ . _ _ . . . . , , , ,

a'.' .. .. ... .. ..

. l' ,

p >

j , ? ., .

e t 4 3 0 $ 8 7 0 't l a00 i

l

, g

.. s a ; -

c , ., j ;

..s..

. e l

n .) ,, .

s 6, . . ,. j *

. . * . . i,

..,... l

' .'. .b.

s j

. . . w e

.. . ..'.. a.

2 j, i

. I e

, e g ;

' 'b.. ... . .. ..

W ' ;

. . e 9 ,

... .

w ,

. I .

.. i

..a g ,

g ! 1

" 7

..,' ..e'..

.t w 4

, 1

\\ .

D. '.. .* . e w

Q s .

3

. 3 $

., . e '

.. w n ,. -s

. ,'. . . gl i

, .. 4 e i * ..

2 .'

. . e

. :. . t

....b 2e i ,

. . ' ,

7 l

. . . s

.... row 1 i s

, . ..

a l- * ** COwsiCCWCR ! CM CRC *C !

i l:

ec:.L w.g *

6. ... .. :. ... .. ... i'

. c

=

++,

q../3....,. ,,.,..

, . g . . . ,

. .. . . . 4 l'

(

'i s

'L i

SPENT FUEL POOL NATURAL CIRCULATION MODEL K ,

1 10' 1

(PLAN VIEW)

'4 t.-

FIGURE 2-11 l ' i

[m,lt 1 I . >

(.' 37 e

[ 2-38 lF lm '

( .t l ,

n +

3 %,z' n1' l I. a, ,.), i t '. .jF m,h. ';I _ t I gi--nak_v,2 i ', { ] QL,g . w[*d

~. 'ig,~. i s

7,.

i

s 1

I

! ! I i I

c w .

V ,

o f

l l fM

%)

rh M _

\ /

%) %J

p. m.

VJ 1

) i SPENT FUEL RACK INLET FLOW AREA (PLAN VIEW) 1:,

,< FIGURE 2-12

,' ,f ,

2-39 O

yll(

u!. , ' '

.s

.-,s. ,,w, ,, , , .,_,e- w - . . , , , . . - - . - , ., , . - - - -

9.MlSt.

A r

.060ctLLest q.oo .ra.

.se a d ivP. .

rLew agg

~

1 L. - .. . . . s d

INTERCELL FLOW AREA

FIGURE 2-13 2-40 4M

3.0- MECEAMICAL, MATERIAL. AND STRUCTURAL CONSIDERATIONS 3.1 DESCRIPTICN OF SPENT FUEL POOL Two individual spent fuel storage pools are located within the fuel handling area of the Auxiliary Building. Both pools are rectangular in plant Pool A is 32'- 2" by 24'- 0" and Pool B is 3 2 8 -7" by 2 4 '- 0" . Both pools have a depth of 43 '- 8". A 10 feet by 10 feet c.sk storage pool is located in the south-eastern corncr of pool B. The walls and bottom slab of the spent fuel pools have nominal thickness of 5'- 0", except for the commom wall between the two pools which is 4'- 0" thick. Walls forming the cask storage pool within pool B are 3 '- 0" thick. Wetted surfaces

-of the pools are lined with stainless steel to ensure. water tight integrity.

The' spent fuel storage pools are suppcrted on reinforced concrete walls extending downward to elevation 93 8- 0", the top of the-structural foundation mat. 'The bottom slab of the spent fuel storage pools is at elevation 118'- 4". The operating floor at the spent fuel pools is at elevation 162'- 0". Above elevation l l'

1628- 0", the spent fuel pool is enclosed within the limits of the j steel framed and metal sided Auxiliary Building structure.

L Missils shield structures are normally in place over top of the spent-fuel _ pool. 1 1

General layout and arrangement of the spent fuel pools and fuel l handling area is presented ir, figures 3-13 and 3-14. I I

3.2 STRUCTURAL DESIGN OF RACKS The Region 1 and Region 2 spent ?luel storage rack configurations j g- are described in Section 2.1. l 1

l 3.3 INTEGRITY OF FUEL RACKS UNDER FUEL HANDLING ACCIDENT CONDITIONS l L 3.3.1 S}snt Fuel Handling Machine Upliit Analysis An analysis was performed to demonstrate that a rack can withstand l

( an uplift load of 350 pounds produced by a jammed fuel assembly. l L 'Using worst geometry assumptions, the stresses resulting from this i l load were calculated and compared to the acceptance limits. This L loading condition was determined not to be a governing condition i and is covered by the results reported in Tables 3 2 and 3-3 for l l' the limiting loading combinations. In addition, since the gross '

L stresses remained within the elastic regime, there is no change l of rack cell geometry of a magnitude sufficient to cause the criticality acceptance criterion te be violated.

3.3.2 Fuel Aseembly Drop Accident Analysis Two storagr situations were considered for the accidental drop of a fuel assembly ento or into the racks. These were: ;

a 3-1

\ ;.

f A. B&W 15x15 standard fuel assembly with handling tool, total dry weight of 2500 pounds, dropped from a height of 24. inches above the top of the rack.

B. B&W 15x15 consolidated fuel essembly with handling tool, total dry weight of 4100 pounds, dropped from a height of 24 inches above the top of'the rack.

- 3.3.2.1 Drop Orientations Three orientations of drop were considered. These were:

A. Drop of an assembly onto the top of. the racks with the assembly in a vertical position, B. Drop of an assembly onto the top of the racks with the assembly in an inclined position, and i

C. Drop of a fuel assembly through an empty cell to the bottom of !

the pool.

3.3.2.2 Acceptance Criteria l The acceptance criteria used were:

A. Fuel criticality does not occur, and B.. Perforation of the pool liner ioes not occur.

p L 3.3.2.3 Assumptions for Energy Tissipation 1

L For evaluation of the drop orientation cases defined above, the following general assumptions for energy dissipation were made:

A. The fuel assembly falls freely in an infinite pool of static j

-water with hydrodynamic drag being considered, l B. No energy is dissipated in the ;2ck structure during the drop, v

L C. The pool liner and floor flexibilities are neglected. The only j i flexibilities considered are those component parts at the ;

7 bottom (impact) end of the fuel assembly, '

1 D. No' energy is dissipated in the fuel rods, and

. E. The kinetic energy of the fuel assembly is totally converted l, into strain energy.of the assembly structure. l 1

3.3.2.4 Drop Analysis Results l Satisfaction of Criticality Criterion 4

Criticality calculations show that with 2000 ppm Boron in the fuel pool water, fuel criticality does not occur. Thus, for the fuel 3-2 l M

lLj

i "s.

j f~

drop accicant the presence of the Boron insures that the criticality criterion is satisfied for all cases.

Satisfaction of Pool Liner Intearity Criterion Drop orientations A and B are considered together since the same philosophy covers both cases. For these cases, either the fuel i assembly will remain on top of the racks after impact or. will impact the pool floor with a lower velocity than the drop through case since part of the potential energy will be absorbed by the initia3 impact with the top of the rack. Therefore, perforation of the pool liner is enveloped by the case of drop of a fuel assembly through a cell. l l

Each of the three cases was evaluated to determine the velocity of impact with the pool liner. In each case the structure at the lower and of the assembly, i.e. , bottom nozzle, guide tubes, etc. ,

had enough strain energy capacity to absorb the drop kine ,

energy. When. consideration was given to the " footprint" of 17 <

L dropped assembly, the stresses japosed on the pool liner we. 1 determined to be less than the ASML Code Allowable Limit foi Faulted Conditions. The pool liner will therefore not be l l perforated for any of the drop accidents. j i

33.4 APPLICABLE CODES, STANDARDS AND SPECIFICATIONS The Westinghouse poison spent fuel storage racks are designed in I

i. accordance with the applicable provisions of the following codes, i standards, and regulations:

l NRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and revised January 18, 1979.

l- l l l l' NRC Reaulatory Guides l l . i R.G. 1.13 Spent Fuel Storage Facility Design Basis L R.G. 1.29 Seismic Design Classifications l

l R.G. 1.31 Control of Ferrite Content in Stainless Steel Welding l Metal, Rev. 3, April 1978 I

R.G. 1.32 Quality Assurance Requirements for Packaging, Shipping, I Receiving, Storage, and Handling of Items for l Water-Cooled Nuclear Power Planta, Rev. April 1973 l

/

l R.G. 1.60 Design Response Spectra for Seismic Design of Nuclear

, Power Planta Rev. 1 j R.G. 1.61 Damping Values for Seismic Design of Nuclear Power j l Plants, October 1973 4 l l l

l 3-3 l l

l ,

- _ _ _ _ - _ . _ - _ _ . _ _ _ _ ~,

L ,

R.G. 1.70 Standard tormat and Content of Safety Analysis' Report

.for Nuclear. Power Plants, Rev. 3 Nov. 1978, Section 9.1.2, " Spent Fuel Storage" R.G. 1.92 Combining. Modal Responses and Spatial Components in

~ Seismic Response Analysis, Rev. 1 R.G. 1.124 Service Limits and Loading Combinations for Class I Lineer-Type Component Supports NUREG-0612 Control of Heavy Loads at Nuclear Power Plants, 7 July 1980 r,

NRC Standard Review Plans

'SRP 3.7 Seismic Design SRP 3.8.4 Other Category I Structures SRP'3.8.5 Foundations F.RP.9.1.2 Spent Fuel Storage SRP 9.2.5 Ultimate Heat Sink NRC Branch Technical Positions

a. CPB 9.1-1, " Criticality in Fuel Storage Facilities."

b. APCSB 9. 2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling."

c. RDT Standard F6-6T, " Welding of Structural Componentt.", latest revision.

Industry Codes and Standards American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section. III, Division 1, Subsection NF. 1980, Summer 1982 Addenda.

American National Standards Institute /American Nuclear Society (ANSI /ANS):

n. ANSI /ANS-8.1-1983, " Nuclear Criticality Safety in Operations with Fissionable Materials outside Reactors."

b. ANSI /ANS-51.1-1983, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."

c. ANDI/ANS-57.2-1983, " Design Requirements for LWR Spent Fuel Storage Facilities at Nuclear Power Plants."

, , 3-4

(

._ 1 1

> Q" ,

j n

l l

[ American National Standards Institute (ANSI):

a. N45.2.1-1980, " Cleaning of Fluid Systems and Associated

. Components for Nuclear Power Plunts."

b. N45.2.2-1972, " Packaging, Shipping, Receiving, Storage, and ,

Handling of Items for Nuclear Power Plants." l I

American National Standards Institute /American Society cf l

. Mechanical Engineers (ANSI /ASME): N45.2-1971, " Quality Assurance l Program Requirements for Nuclear Facilities." l 3.5 SEISMIC ANALYSIS PROCEDURES FOR SPENT FUEL STORAGE RACKS The purpose of this subsection is to demonstrate the adequacy of ,

the spent fuel rack design to store B&W 15x15 fuel assemblies under normal and' accident loading conditions. The seismic analysis, which produces the governing loading conditions, i produces results which are divided into two types (stresses and displecements).

The stresses were checked against the design limits to ensure the structural adequacy of the design. The horizontal displacement -

results .wera checked against the rack clearances to determine whether the racks collide with another rack or the pool wall. The vertical displacements of the support pads, due to rack rocking and lift-off, were used to show that the racks do not overturn.

The results showed that the high density spent fuel racks are structurally adequate to resist the postulated stress combinations shown in Table 3-1, and the racks have margin against overturning and against rack collision. ,

3.5.1 -Analysis Overview The spent fuel storage racks are seismic category I equipment. !

Therefore, they are required to remain functional during and after an SSE (Reference 19) . Since the racks are free standing (neither anchored to the pool floor, to the pool wall, nor structurally interconnected) and the fuel is free to move inside the cell within the limits of the clearance between the fuel and cel), a nonlinear seismic analysis was performed. To ensure the limiting values have been obtained with a nonlinear analysis, numerous conditionti with different conibinations of parameters were analyzed in order to identify the combination of parameters which produces the limiting conditions. The parameters within the pool configuration which affect the response of the fuel racks are:

Damping

Impact between the fuel assembly and cell

Fluid coupling >etween the fuel racks and pool wall

+ Coefficient of friction between the fuel rack support points ar.d pool ~ floor

Structural characteristics of the fuel racks (Region 1, Region 2) 3-5 ;

[

I 1

To obtain the rack responses (rigid body motion, structural deformation, loads, stresses, etc.) the analysis was performed on '

a 3-D nonlinear dynamic finite element model of a fuel rack which j was subjected to the simultaneous application -of three i statistically independant, orthogonal acceleration time histories '

at the pool floor. The response spectrum from each corresponding time history record is shown in Figures 3-1 through 3-3. The design spectra is also plotted on each figure to show that the ,

,. time history spectra enveloped the design spectra. The pool floor l

. acceleration time history data developed for the SSE are shown in

. Figures 3-4 through 3-6.

The steps in the seisraic analysis were:

A. Develop 3-D nonlinear dynamic finite element models of the fuel

' rack modules consisting of beam, mass, dynamic gap, and

. friction elements.

B. Perform time history analyses on the nonlinear dynamic models for the bounding cases using the dynamic capabilities of the nonlinear. modal superposition sclution the Hestinghouse Electric Computer Analysis (WECAN) techniqueCode(g) .

C. Compute the stresses in the fuel rack at the critical structural locations using the loads from the previous step.

These atresses were checked against the design limits given in ;

Table 3-1 to ensure the adequacy of-the design.

3.5.2 Seismic Medal Since the fuel assembly is not structurally connected to the cell wall and the fuel racks are free standing, the seismic model must l have the capability to address a wide variety ' of rigid body motions: .

The fuel moving in the clearance between the fuel and cell with subsequent impact on the cell wall (" fuel rattling condition");

}

+

The rack sliding on the pool floor (" sliding conditiond); i l

One or more support pads momentarily losing contact with the j pool floor (" tipping condition") ;

l Rack rocking onto one support pad and pivoting about that pad <

(" torsional rotating condition") ;

The rack may also experience a combination of sliding and tipping conditions.

l l

These mechanical nonlinearities in the fuel rack dynamic responses required a strong emphasis on the modeling of Doth the linear and nonlinear elements.

l L 3-6 l s

To determine ~ the dynamic response of a system with multiple i nonlinearities, as described above, the analysis must be conducted '

on a model which realistically represents the dynamic characteristics. That is: the model must have realistic linear stiffness and mass properties as well as the correct number and geometric location of impact points, a realistic value of impact stiffness, and sufficient Dynamic Degrees of Freedom (DDOF) in order to develop the higher modes of response which are associated with impact forces. The linear .and nonlinear elements used in the model' must have the capabilities described in the following paragraphs.

The linear elements represent both the fuel and rack stiffnesses as well as the geometric properties (support pad spacing, fuel grids locations, etc.) with sufficient DDOF to capture the dynamic 1 response of the combined fuel and rack system. More specifically, the number and location of fuel to cell contact points were accurately modeled in order to provide the correct impact force )

magnitude and location'so that the correct modes of response of l the fuel and. cell were produced. The fuel rack height, mass distribution, and support pad spacing were modeled to provide a q representation of'the rack rigid body stability characteristics. l

. The nonlinear elements which model the impacting between the fuel J l' assembly grids and cell included the impact stiffness, impact l l damping, and gap size. The nonlinear elements that model the support pads which may lift off and impact the floor, or may slide relative to the pool liner, included the impact stiffness, impact L

damping, and Coulomb friction to appropriately simulate, this interface between the fuel rack and pool floor.

3.5.2.1 Model Description The time history analysis was performed on a 3-D nonlinear finite L element model which represents one rack in the pool. The p following sections describe in detail the structural model and the nonlinear single rack model.

l To compile the finite element model, the properties of the linear elements and nonlinear elements must be calculated. The L properties of the linear elements are obtained from a 3-D finite I element structural model of the fuel rack. The linear properties, t

referred to as the effectivo structural properties, are calculated L from the r.tructural model and used as the basis of the nonlinear l: ,

model. The nonlinear elements, with properties such as impact stiffness, gap, impact damping, and friction coefficient, are

l. added to the effective structural properties to produce the l nonlinear model.- The details of the structural model, effective structural properties, and the nonlinear model are presented in the following paragraphs.

1.

3.5.2.2 Structural Model The structural model is a 3-D finite element model of the fuel rack. Since the spent fuel racks are a two region design, and the j- 3-7 l

. . -__ _ _-._..._._ . _ _ _ _ _ __ _. . _ b

(F, - ,

13 i 1

racks'- for each region are structurally different, two different structural codels are used in the analycis as shown in Figure 3-7 for Region 1 and Region 2. ]

The fu'el rack is composed of the following structural componentas support ' pad assembly, base plate, cells, and call to . cell J connections. These components are attached in a manner which produces an overall. rack structure. The structural input proporties of each component are discussed in the following paragraphs.

The support pad assembly is composed of a leveling pad, leveling screw, and support block. The structural components in the auinsmbly, leveling screw and support block, are modeled as beams with area and inertia values obtained from cross sectional properties.

The base plate assembly in the structural model has the effective

'> stiffness values of the base plate, support blocks, and cells or bottom grid welded to the base plate. The base plate is modeled as effective beams which connect the support pads and the cells.

L Since the cells (Region 2) or bottom grid (Region 1) are welded p to the ' base plate, the inertia and area of the base plate p effective beams are based upon the cross section which includes l a portion of the base plate and the bottom grid (Region 1) or a -

portion of the cell wall (Region 2).

The cells are square (0.060 inch wall thickness) tubular sections which have 0.020-inch thick wrapper plates on the four sides to support the Boraflex sheets. The cells are modeled as beams with .

I area and inertia based upon the cross section of the 0.060-inch thick cell wall and only one wrapper plate on the tensile side of the cell. The use of one wrapper on the tensile side of the cell is based upon the post buckled condition which produces structurally ineffective members of the wrapper plates which are loaded in compression. It is noted that the dynamic analysis is based upon section properties which include the wrapper plate on the tensile side, but the stress analysis is based upon section properties which do not include the wrapper plcte, thereby L producing a conservative analysis. The cell to cell connections L are the members which form the structural connection between the L cells and produce the overall shear connection of the rack l assembly. In Region 1 the connection is produced by the top and bottom grid assemblies and in Region 2 the connection is produced by cell to cell welds at the cell corners. This connection is L modeled with effective beams which connect between cells. The properties of these beams for Region 2 are obtained from a finite element analysis of the cell wall and the weld. '

l The properties of the effective structural model are obtained from the results of the structural model. The effective model shown in Figure 3-9 for Regions 1 and 2, is composed of elements which represent the cell assembly, cell to cell connection, and support i

pad / base plate. The properties of these elements are calculated 3-8

l l

from the load and displacement results of the structural model, '

l l' as. discussed in the following paragraphs.

l, A' finite element model of the effective structural' model is I compiled and run, and the mode shapes, including the higher cell. .

modes, of the effective model are. compared with those of the

' structural model to ensure that the effective structural model is '

an adequate dynamic representation of the structural model and has a sufficient number of modes to produce the higher mode response -

due to fuel impact loads.

i The properties of the cell assembly in the effective structural L model are the same as those in the structural model.

The properties of the cell to cell connection in the effective model are rotational stiffnesses (Ke) in units of in-lb/ rad. These values are the ratio of bending moments reacted by the cell to cell connection members divided by the angular rotation of the connection at the cell from the structural model. At each ;~

elevation of cell to cell connection, this calculation is performed at each cell location and. averaged for the total cross section to obtain the effective rotational stiffness at that elevation. The cell to cell connections, which produce the overall shear connection of the rack, are placed at locations ,

along the length of the cells to produce the required frequency of the rack. As shown in Figure 3-9, there are 2 cell to cell connections, one top grid and one bottom grid connection for the Region 1 effective structural model, and there are 3 cell to cell connections (welds) for the Region 2 effective structural model.

The value of the rotational stiffness between the bottom of the l cell and base plate is calculated by the same method as the cell

! to cell. rotational stiffness values. It is included in the I stiffness matrix element, [K), between the bottom of the cell and l base plate.

L The effective support pad / base plate properties are calculated in two stages. The overall base plate rotational stiffness is I calculated from the structural model, and then thr) v Pination of l rigid beams with vertical spring elements on the . is used to L represent the rotational stiffness while providir.g a effective li vertical stiffness and geometric spacing of the corner support l pads. The overall base plate rotational stiffness is calculated by dividing the rotation of the rack base by the total moment c applied to the base. The rotation of the rack base is calculated I by dividing the vertical displacement at each cell location on the L- base by the distance from that point to the rack centerline, and l calculating the average for all the cell locations in the rack base. This accounts for both the support pad deformation and base plate deformation at the locations between support pads.

1 The vertical stiffness of the effective corner pads is calculated by equating the rotational stiffness of the vertical spring / rigid beam assembly to that of the base plate rotational stiffness (Ke) .

Using four support pads, one at each corner, and a spacing of (L) 1 3-9

l l

i A

inches between support pads, the' effective vertical stiffness per

! corner pad is Ky = Ke/L ,

3.5.2.3' Nonlinear Seismic Model The nonlinear model, shown in Figure 3-10, is a'3-D model composed of effective proparties from the structural model, to account for the rack structure, with additional elements to account for the fuel assembly, fuel to cell gap, fuel hydrodynamic mass, support pad boundary ' conditions of a free standing rack, and the hydrodynamic mass between the fuel rack and the pool wall or ,

between fuel racks. A better illustration of the elements of a 2-D view of the 3-D model is shown in Figures 3-11 and 3-12 for Regions 1 and 2 respectively.

The effective components / properties from the structural model are:

[

The cell assembly represented by 3-D beam elements with '

effective stiffness properties and mass density; ;

a Cell to cell connections represented by rotary stiffness '

elements in both horizontal axes;

Cell to base connection represented ' by a s.tiffness matrix ,

element with rotary stiffness in both horizontal axes, i lateral stiffness in both horizontal axes, and rigid connection in the vertical axis;

Support pad corner spacing represented by rigid 3-D beam elements connecting the center and the four corners;

The effective support pad / base plate vertical stiffness represented by a 3-D friction element which has the nonlinear capability to slide on the horizontal plane or lose contact in the vertical direction (support pad lif t-off) and impact

[ upon contact.

The . fuel assembly is represented by a combination of 3-D beam elements and rotational stiffness elcments. The beam elements have properties of inertia, area, and mass density. The area and '

<. inertia values are based upon the fuel rod cladding and fuel skeleton cross section. The rotational stiffness elements have rotational stiffness about both horizontal axes to account for the l

connection of the fuel rods at the fuel grid locations. The connection between the fuel base and fuel rack base plate is I modeled with a 3-D friction element which has the nonlinear l capability to allow the fuel assembly to lift-off the base plate I and produce impact forces.

I The fuel to cell gaps are modeled by 3-D dynamic elements with l gaps, impact stiffness, and impact damping values to represent the 4

gap between the fuel and cell. The fuel assembly impact stiffness d

and damping values for the B&W 15x15 fuel assemblies were used in the analysis. The maximum value of the gap between the fuel and l cell is conservatively used in the model. This is done by using l

3-10 l l

1

the tolerances to produce the minimum fuel grid dimension and maximum cell opening. The gap is input as a concentric gap where

'the fuel is initially in the center of the cell and can impact any of the four sides of the cell. The impact stiffness between the 2 fuel and cell is based upon the series combination of the fuel ,

grid' impact stiffness and the local stiffness of the cell wall. '

The fuel grid impact stiffnoss is supplied by the fuel vendor, and !

the local cell wall stiffness is calculated from the displacements !

of_the cell wall, modeled as a flat plate, due to loads applied at the corners of the fuel grid. The impact damping of the fuel

i. grid is based upon values supplied by the fuel vendor. There are 9 gap elements located along the length of the fuel assembly to ;

represent 8 intermediate grids and one top fitting.

The hydrodynamic mass between the fuel assembly and cell is

modeled by a 3-D mass matrix element in both horizontal i' directions. Since the fuel assembly is an open array of rods, !

the proximity effect of the cell wall on the rods only affects the !

edge row of rods and does not produce a significant hydrodynamic l mass on the fuel assembly. Thus, the flow around the rods is the j condition which produces the governing hydrodynamic mass. The !

calculationofthehydrodynamicmassduetoflowaroundghefuel rods is based upon the kinetic energy of the fluid . The I hydrodynamic mass elements are modeled at 8 locations along the

length of the fuel with values proportional to the effective fuel t

length at each location.

j The hydrodynamic mass between the fuel rack and the pool wall is l

, modeled by 3-D mass matrix elements in both horizontal directions.

'. The hydrodynamic mass elements are modeled at 8 locations along ;

the length of the cell with values proportional to the effective j

, cell length at each location. The values of hydrodynamic mass for j the fuel rack are based upon potential flow theory and are I calculated by evaluating the effects of the gap between the rack and the pool wall or betweengacks by using the method outlined in the paper by R. J. Fritz

( . A more detailed discussion on hydrodynamic mass is presented in Paragraph 3.5.2.5 Fluid coupling.

3.5.2.4 Damping There are two types of damping present in the dynamic response of a nonlinear structure such as the fuel rack (structural damping and impact damping).

The structural damping values used for the seismic analysis of fuel rack structures are 2% for OBE and 4% for SSE. These values are in accordance with Regulatory Guide 1.61A for welded steel structures.

The damping which occurs during impact is produced by the small amount of local plastic deformation which takes place. The two types of locations in the rack which have impact are the fuel grid to cell wall and the support pad to the pool floor. The damping value used for the fuel grid impact, which was cuppliect by the l l

3-11 l

l:

i fuel vendor, is 15%. The damping value used for the support pad ,

i >

impact is based upon steel on steel impact. The ' range of l

l coefficient of restitution for steel on steel in Reference 26 is 0.5 to 0.8. Thus, conservatively using a value 'of 0.85 coefficient of restitution produces an effective impact damping of 4% for the support pads.

3.5.2.5 Fluid coupling l'

The effect of water on the dynamic response of a submerged l

structura is significant, and must be included in the modeling. ,

If one body of mass (mi) vibrates adjacent to another body of mass (m2), and both bodies are submerged in a frictionless fluid medium, i then Newton's equations of motion for the two bodies have the ,

form: '

(mi + Mn) Ni - M12 'N2 = applied forces on mass mi

-M21 Xi + (m2 + M22) 2 = applied forces on mass m2 -

5, 3 *X2 denote absoluto accelerations of mass mi and 3, 2

l respectively.

l Fluid coupling coefficients, Mu, M12, M21 and M22, depend on'the shape of the two bodies, their relative disposition, etc. Fritz" gives data for My for various body shapes and arrangements. It is noted that the above equations indicate that the effect of the fluid is to add a certain anount of mass to the body (Mn to body ,

i 1), and an external force which is proportional to the ,

I acceleration of the adjacent body (mass m2) . Thus, the i acceleration of the one body affects the force field on another.

This force is a strong function of the interbody gap, reaching l large values for very small gaps. This inertial coupling is l

., called fluid coupling. It has an important effect in rack j L dynamics. The motion of the rack as well as the lateral motion ,

I- of a fuel assembly inside the storage location will encounter this !

effect. 1 The hydrodynamic coupling between the fuel and cell and between i ' the rack and pool wall is represented in the acdels by mass matrix l.. elements. The hydrodynamic mass between the fuel asser.bly and the I cell walls is based upon the fuel rod array size and cell inside dimensions using the technique of potential flow and kinetic energy. The hydrodynamic mass is calculated by equating the i kinetic energy of the hydrodynamic mass with the kinetic energy of the fluid flowing around the fuel rods. The concept of kinetic i F.

energyof.thehydrodynamicmassisdiscussedinapaperbyD.

DeSanto l

The hydrodynamic mass between the racks and pool wall was l calculated by evaluating the effects of the gaps between the racks l and the pool wall using the method based upon potential flow theory outlined in the paper by R. J. Fritz W . To account for the flow in three dimensions, the hydrodynamic mass for flow in the horizontal direction around the racks and for flow in the vertical 3-12

= direction up over the top of the rackt and down below the bottom of the racks are calculated independently, and combined to produce an overall hydrodynamic mass value.

3.5.2.6 Friction Coefficient Since the fuel racks are free standing, the frictional resistance in the interface between the support pads and pool floor is the only horizontal constraint. Thus, the value of friction coefficient must be accurately represented in a manner to ;

conservatively calculate the displacements of the fuel rack. The values useg inAccording Rabinowicz ..

the analysis are based to Rabinowicz, upon tests performed by the results of 199 tests performed on austenitic stainless steel Aates submerged in water show a mean value of friction coeffic:,ent to be 0.503 with a standard deviation of 0.125. The upper and lower bounds (based on two. standard deviations) are 0.753 and 0.253 respectively. In order to address the upper and lower bounds of friction, two separate analyses are performed on thn rack model with friction 1 coefficient values of 0.8 and 0.2. The results of these analyses showed that different dynamic behavior was produced for each coefficient, thus producing the bounding response.

3.5.3 Time History Evaluation 1 The seismic analysis of a free standing fuel rack is a time I history analysis performed on a 3-D nonlinear finite element model l subjected to the simultaneous input of three statistically l independent acceleration time histories at the pool floor l elevation. The analysis was performe g' n the Westinghouse i Electric Computer Analysis (WECAN) Code using the dynamic l capabilities of the nonlinear modal superposition method. I WECAN is a general purpose finite element code which has been reviewed and approved by the NRC. The general reviews of the l overall code by the NRC are: )

i Documents submitted to the NRC for re. view are: i 1

Westinghouse Report WCAP-8252(@ '

Westinghouse Report WCAP-8929(W NRC Review at Pittsburgh, PA on October 1-5, 1984 by Messrs.

P. Sears and P. Milano, Reference Document No. 99900404/84-03 In addition to the general reviews, the application of the nonlinear modal superposition method in the WECAN Code for spent fuel rack seismic analysis has been specifically reviewed by the NRC during the licensing review for the following fuel rack dockets.

Duke Power Co., Oconee Unit No. 1 & 2, Docket No. 50-269, 50-270, 50-287, during 1980.

3-13

i .

i i

Duke Power Co., McGuire Unit No. 1&2, Docket No. 50-369, 50-370 during 1094.

Consumers Power Co. , Palisades Plant, Docket No. 50-255, E during 1986.

The following is a list of rarack projects in which the seismic analysis was performed on the H WECAN Ccde.

Utility Site Name l- Arkansas Power and Light Arkansas 1 & 2- l

) Carolina Power and Light Shearon Harris 1, 2, 3, &4 H. B. Robinson Carolina Power and Light l Consumers Power Palisades Duke Power Oconee 1, 2&3 McGuire 1 & 2 '

Florida Power and Light Turkey Pt. 3 Georgia Power A. W. Vogtle 1 Gulf States Utilities River Bend 1 l Northeast Utilities Millstone 1 & 3 l Philadelphia'Elec. Co. Peach Bottom 2 & 3 !

Public Service of New Hampshire Seabrook Tennessee Valley Authority Bellefonte 1 & 2 Texas Utilities Comanche Peak 1 & 2 {

The nonlinear modal superposition method was developed to analyze nonlinear structural dynamics problems involving impact between !

components and Coulomb friction. The finite element method is used to erpress the equations of motions with the nonlinearities represented by a pseudo force vector. The details of the nonlinear modal superposition method used in WECAN are presented below. References 24 and 25 provide additional information and application of the nonlinear modal superposition method.

l The natural frequencies and mode shapes for the nonlinear structure are obcained by reduced modal analysis or full modal l analysis. These frequencies and mode shapes represent the :

reference state of the nonlinear structure. During the time history analysis, as the nonlinear behavior comes into action, the true frequencies and mode shapes change. The effect of the variation of the true frequencies and mode shapes from the original ones is represented by pseudo forces on the right-hand side of the equation of motion.

The generalized equation of motion of a structure is:

(M)(X) + (Cnt)(X) + [Kn1) ( X) = (F) (1)

Where, (h ) is a mass matrix.

3-14

-_____________-___u

L l

(Cm) is a nonlinear damping matrix, dependent upon velocity and displacement.

(Km) is a nonlinear stiffness matrix, dependent upon displacement (X), ($), (lK) and {F) are displacement, velocity, acceleration and applied force vectors.

Let (cm) = (C) + ['C) and (Km) = (K) + (K) (2)

Where (C) and (K) are the damping and stiffness matrices

( rJe resenting the reference state of the structure. The [6] and l

[K) are the damping ar.d stiffness matrices dependent on velocity and displacement. Equations (2) are substituted in equation (1).

This gives:

(M](X) + (C)(k) + [K](X) = (F) - (Fm) (3)

Where, the pseudo force vector is defined by (Fm) = (6)(X) + (R)(X) (4) l As the nonlinear properties are normally restricted to a small portion of the structure, only a small number of finite elements with nonlinear properties are used to model the structure. So the matrices (E) and ['K') are sparse, and the cost of calculation ;

of the pseudo force is small if computations are performed at '

element level. The homogeneous, undamped equation of motion representing the reference state of the structure is:

(M](k)+[K](X)= (0) (5)

Let (w) and (9) be the natural frequency and normalized mode shape l matrix associated with equation (5). The following transformation,

)

(X) = [9)(q) (6) is substituted in equation (3) pre-multiplied by (t]T, employing I the orthogonality relations expressed by 1 (6)f(H][t]= [1]

[9] (C)[4] = (2 fj wj]

and j

(, (4]T [K)[9] = [wj2) 3-15

___ _ t

i l

the resulting modal equations become (k) + -[2. fj uj)(q) + (w[] {q) = (Q) - (Qnl) (7) where, I fj = percentage of the critical damping for the j* mode.

0 '

(Q) = [6)T(F) = generalized applied force vector (Qng) = [4)T(Fni) = generalized pseudo force vector Arrays (q), (q)and(k) are the modal displacement, velocity and acceleration vector respectively. The generalized pseudo force vector is a function of displacement and velocity. For a given time step, it can be approximated by Taylor series as follows:

oo (Qng ) t= E L.

k=o kl (k

dt (Qn3 })y (t - T)k (8) '

T < t < T + AT k=0,1,2.......

Equation (7) represents a set of uncoupled equations. These equations are integrated analytically to eliminate numerical damping or frequency distortion during integration. Equation (8) represents the extrapolation of generalized pseudo force vector i by Taylor series. The number of terms that can be included in the Taylor series is determined by the continuity of (Qni)and its time

! derivatives. The more the number of terms, the larger the allowable integration time step. The extrapolation is done in the l

modal space, which is of relatively small size. Each additional term of the series will require small additional storage in computer core. For the most practical applications, it suffices '

L to include only the first two terms of the Taylor series.

1.

For a given time step, modal equations of motion are integrated analytically. Then the displacement and velocities of the nodes associated with the nonlinear elements are calculated. This l information is used to calculate the generalized pseudo force l- vector and its time derivatives. Then the modal equations are '

l integrated for the next time step. -

1 The nonlinear model was run for the bounding cases which account for the variation of parameters such as friction coefficient (0.2 and 0.8) and Region 1 and Region 2 rack structure. The results i from these runs include the fuel to cell impact loads, support pad loads, fuel rack structure internal loads and moments, support pad lift-off, fuel rack sliding and structural displacements. Since l' the seismic analysis was conducted on a single rack model, the

[ absolute (where absolute is the displacement of a rack with respect to the pool floor) displacements at the bottom and top of the racks were obtained, but the relative (where relative is the I displacement of one rack with respect to an adjacent rack) l' l 3-16 I

l.

l -

[,

i N

i l- ' displacement between racks is not calcolated by the model. To l l' obtain the relative displacement between racks it was conservatively assumed that the adjacent racks respond 180' !

out-of-phase, thus producing a relative displacement equal to {

twice the absolute displacement of one rack.

l The values of these results were searched through the 20 seconds I duration of the time history to obtain the maximum values. The j maximum values of the loads and moments were used in the stress -

analysis, and _the displacement results were used to show that significant separation margin against collision with an adjacent rack or the pool wall remains and that there is ample margin against overturn.

A number of conservatisms have been incorporated in the analysis and are listed below.

A. All fuel assemblies are treated as if they respond in phase l which results in the maximum rack response. .

B. Friction coefficients of 0.8 maximum and 0.2 minimum are used in the analysis.

l C. Adjacent fuel racks respond 180' out-of-phase.

D. Hydrodynamic mass is based upon constant gaps. As the gap decreases the hydrodynamic mass restoring force increases; L but since the analysis is based upon constant gaps, the I

displacements which close the gaps are conservative because the restoring forces increase.

E. No friction is used in support pad swivel joints to resist rotation when the rack rocks onto one pad.

F. Gaps between fuel and cell were maximized and produce the maximum impact forces.

1~

G. The magnitude of the separation between the racks and between r a rack and the pool wall, which is maintained during the seismic event, shows the conservatism in the overall design.

L 3.6 STRUCTURAL ACCEPTANCE CRITERIA AND ANALYSIS RESULTS FOR SPENT FUEL STORAGE RACKS l

3.6.1 Criteria There are two sets of criteria to be satisfied by the racks:

A. Kinematic Criterion This criterion seeks to ensure that the rack is a physically .

stable structure. The racks are evaluated for margin against overturning and also for rack displacement to ensure that rack to rack and rack to pool wall impact does not occur.

3-17

i l I

B. Stress Limits ,

a The stress limits of the ASME Code,Section III, Subsection NF

! are used since this code provides the most appropriate and consistent set of limits for various stress types and various loading conditions, i

3.6.2 Stress Limits for Specified Conditions

]

The structural analysis of the Crystal River racks is an elastic 1 type analysis for linear type supports per ASME Code,Section III, Subsection NF. The Load Combination and Acceptance Limit table j is shown in Table 3-1 and repeated below.

]

Elastic Analysis Load Combination Acceptance Limit

1. D+L Level A service limits of NF

2. D+L+E Level A service limits of NF J l- 3. D+L+To Lesser of 2Sy or Su stress range

! 4. D+L+E+To Lesser of 2Sy or Su stress range

5. D+L+E+T. Lesser of 2Sy or Su stress range l C. D + L + To + P, Lesser of 2Sy or Su stress range i 7. D + L + T. + E ' Level D service limits of NF l

Margins to Allowable shown in Tables 3-2 and 3-3 are for the limiting load combinations 2, 5 and 7. The Margin to Allowable (MA) is calculated, as shown in equation form below, by comparing the acceptance limit with the applied stress. It can be seen in the table above that there are three different acceptance limits

- one for load combinations 1 and 2, another for load combinations 3 through 6, and a third for load combination 7. Thus, it is possible to meet the requirements of the load combinations by I

addressing the three limiting combinations. Load combination 7 '

is limiting because it is the only combination involving the SSE condition. Load combinations 2 and 5 are the limiting combinations of combinations 1 through 6 as shown in the following paragraphs.

i Load combination 2[D + L + E) envelops load combination 1[D + L).

Load combination 5(D + L+ E + T.] envelops load combinations 3 ( D + L + To] and 4 [ D + L + E + To) . Since stresses caused by the stuck fuel assembly load condition ( P,) are much lower than stresses caused by the OBE (E), load combination 5 also envelops load combination 6(D + L + To + P,) .

The two columns on Tables 3-2 and 3-3 are labeled OBE and SSE.

The column labeled OBE is the margin to allowable for either load combination 2[D + L + E) or load combination 5[D + L + E + T.],

whichever is the more limiting condition. Except where indicated on the tables, load combination 2[D + L + E) is more limiting.

The column labeled SSE is the margin to allowable for load combination 7 (D + L + T. + E ' ) .

3-18

q

' j 1

The MA shown in Tables 3-2 and 3-3 is defined as '

MA = Allowable Stress - 1 -

Appl.ied Stress 3.6.3 Results for Rack Analysis i Tables 3-2 and 3-3 show the minimum margin to allowable for the various components and welds on the racks. The adequate margin .

in each case shows that the racks meet the structural requirements '

of the ASME Code.

In addition, the impact loads on the fuel assemblies due to the ,

interaction with the rack during a seismic event have been '

determined. The maximum calculated seismic impact load at a spacer grid location is 305 pounds, which is less than the allowable spacer grid strength for the B&W 15x15 fuel assemblies.

3.6.4 Rack Displacements From the nonlinear time history analysis, the maximum Region 1 rack displacement (absolute displacement) was determined to be 0.31 inches in the east-west direction and 0.32 inches in the north-south direction. For the Region 2 racks, the maximum displacement was determined to be 0.32 inches in the east-west direction and 0.29 inches in the north-south direction. The remaining rack to pool wall gap is calculated by taking the

minimum initial clearance between the rack and the pool wall and i then subtracting the installation tolerance (0.25 inches), total l thermal growth of one rack (0.08 inches), and the seismic l displacement of the rack. The minimum remaining rack to pool wall 1 gap is determined to be 3.35 inches and is based on the nominal

initial rack to wall gap of 4.00 inches in the east-west direction

! for a Region 2 rack and the seismic displacement of 0.32 inches.

l-The most limiting relative displacement between racks is determined by conservatively doubling the absolute displacement of one rack. Using the appropriate minimum initial clearance between racks of 1.29 inches and then subtracting the thermal I growth of two racks (0.16 inches total due to 0.08 inches per i rack), and the seismic relative displacenant between racks (2 x 0.32 inches), remaining rack to rack gap is determined to be 0.49 inches.

From these results it is concluded that the racks are spaced with l sufficient clearance so that rack to rack and rack to pool wall I impact does not occur.

Also extracted from the time history results is the maximum support pad vertical displacement (lif t-of f) . The maximum support l pad lift-off is found to be 0.003 inches. For this magnitude of I pad lift-off the factor of safety against rack overturning is determined to be greater than 50 which satisfies the requirements of Section 3.8.5.II.5 of the SRP.

3-19

I i

3.7 MATERIALS, QUALITY CONTROL, AND SPECIAL CONSTRUCTION TECHIQUES 3.7.1 Construction Materials Construction materials conform to the requirements of ASME Boiler ,

and Pressure Vessel Code,Section III, Subsection NF. All the materials used in the construction are compatible with the storage pool environment and do not contaminate the fuel assemblies or the pool water. The racks are constructed from Type 304LN etainless steel except the leveling screws which are Type 17-4 PH stainless steel.

3.7.2 Neutron Absorbing Material The neutron absorbing material, Boraflex, used in the Crystal s River spent fuel rack construction is manufactured by Brand Industrial Services, Inc., and fabricated to safety related quality assurance criteria of 10CFR50, Appendix B. Boraflex is a silicone based polymer containing fine particles of boron carbide i Boraflex contains a l

minimum B'gareal a homogeneous, density of stable 0.023 matrig.

gm/cm for Region 1 racks and 0.015 gm/cm' for Region 2 racks.

Boraflex has undergone extensive testing to study the effects of gamma irradiation in various environments, and to verify its structural integrity and suitability as a neutron absorbing material."

Tests were performed at the University of Michigan exposing Boraflex to 1.03 x 10" rads gamma radiation with a substantial concurrent neutron flux in borated water. These tests <

indicate that Boraflex maintains its neutron attenuation capabilities before and after being subjected to an of borated water and 1.03 x 10" rads gamma radiation.gvironment Long term conducted.gIratedIt water soakthat was shown tests Boraflex at high temperatures withstands were also a borated water immersion of 240*F for 260 days without visible distortion or softening. Boraflex maintains its functional performance <

characteristics and shows no evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide.

The actual tests verify that Boraflex maintains a long-term material stability and mechanical integrity, and can be safely utilized as a poison material for neutron absorption in spent fuel storage racks.

3.7.3 Qttality Assurance The design, procurement, and fabrication of the new high density

( spent fuel storage racks comply with the pertinent Quality L Assurance requirements of Appendix B to 10 CFR 50. In addition, the Quality Assurance Program of the Westinghouse Nuclear Components Division (WNCD) , at which the racks are manufactured, also conforms to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, l Section III and VIII (Subsection NCA-4000), MIL-I-45208, 3-20 )

t,

..% m -,4 , . . - .

>;. ,s i

e-MIL-Q-9858A, and RDT F2-2. The Quality Assurance Program at WNCD I is implemented through the Westinghouse Water Rgetors Division

~

i. Quality Assurance Plan as described'in WCAP 8370 1 1

3.8 TESTING AND INSERVICE SURVEILIANCE Testing and inservice surveillance is ' discussed in section 2.2.4.1, " Neutron Poison Surveillance Program".

l 1

I l

l l'

1 i-l' I

1 i

N, 1

l l

3-21

m I

u s ..

l TABLE 3-1 LOADS'AND LOAD COMBINATIONS FOR SPENT FUEL RACKS l

Load Combination- ,_

Acceptance Limit 1

.D + L- Level A service limits of NF '

D +'L + E Level A service-limits of NF L.

1 1

D+L+To Lesser of 2Sy or Su stress range 1:

D+L+E+T, Lesser of 2Sy or Su stress range r

D+L+E+T. Lesser of 2Sy or Su stress range

D + L + T. . + . P, . Lesser of 2Sy or Su stress range l

D + L + T. + E ' Level D service limits of NF

1. The abbreviations in the table above are those used in Section 3.8.4 of the Standard Review Plan (SRP) where each term is defined l except for T. and P t. The term T. in defined here as the highest L temperature associated with the postulat.ed abnormal design .'

l conditions. The ' term P is the upward force on the racks caused l by a postulated stuck fuel assembly. ,

1; 2.- The provisions of NF-3231.1 of ASME Section III, Division I, shall be amended by the requirements of Paragraph c.2.3 and 4 of 1 Regulatory Guide' 1.124, entitled " Design Limits and Load l: Combinations for Class A Linear-Type Component Supports."

l l

3. For the faulted load combination, thermal loads will be neglected -

when they are secondary and self limiting in nature and the material is ductile.

L 1.

3-22

n'

)..

i

>~ ,

t l

1 1

TABLE 3-2 l MINIMUM MARGIN TO ALLOWABLE Region 1 l

)

OBE" SSE D Support Pads 0.64 0.46 Cells 0.30 0.25 L

Grids 0.70 >1.00 L Welds-Cell to Grid 0.21' O.20 Grid to Gr.ld 0.40 0.30 Grid to Base Plate >1.00 >1.00 .

Cell Seam >1.00 .70 g' Cell to Wrapper >1.00 >1.00 t

a - Load combination D + L + E, unless otherwise indicated b - Load combination D + L + T. + E' c - Load combination D + L + T. + E

+

( l ,-

l l:

l 3-23 p.

f-EN

^

p

, ' tt NA ,

TABLE 3-3 I MINIMUM MARGIN TO ALIDWABLE :

' Region 2- !

l i

i OBE* SSE D Support Pads >1.00 0.45 j

^ ~

Cells 0.64 0.37- ,

Welds 1 Cell to Base Plate 0.70 0.27 I J

Cell to Cell 0.55 0.21 i l

Cell' Seam 0.50' O.19 Cell to Wrapper 0.78 0.32 a - Load combination D + L + E, unless otharwise indicated

. i b - Load combination D + L + T. + E' c' -' Load combination' D + L + T. + E ,

I 4

e e

3-24

a i

?

' ?

8.5719 0.5500 0.5000 I A

0. 45= . .

0.4000 t 0.5500 0.5000 0.2560

- 0.2000 S-l- .J W 0.1500 V

U

~

0. ien m_

0..

i.

L I l I I I I I I I I I a

a =i i i i i i i i i

. FREQUENCY (HZ) i DESIGN VS. TIME HISTORY RESPONSE SPECTRA

> N-S SSE 4% DAMPING FIGURE 3-1

?-

! .i.

I 3-25 1

l ..

l 1-l._

_. - _ _ . - . . _ . - ~ . _ _ _ . _ . _ . . - . _ . _ _ _ _ . . . _ . . _ _ _ . _ . . _ . . . . _ . , . _ _ . . . _ _ _ . - . , _ . _ . , - , _ _ _ _ . . . - . , . . . . _ _ , _ - . _ _ _ _ _ . . _ - - _ . . , , _ . - _ . .

gp , ;

[:.

l+ '

p' ;

h r I

}'.

l i

. O 57N j.,

0.5500 -

A h ' O .SSN - I 1 'l

' )

i S'

e O . 45N '

{ 4 l 1

. . .N l 8.5800.

l.

l C.25N '

f

'* ~

a S.2000 3

a L

$ 0.15N 3 l- U e.isas a,u,-

i

e. sees 3 3 3 8 3 3 8 E E t l

L 3 3 e e e e = * * "! * -

3*g i J e w a * *

. i J m e m w w *

. a - - m l'

FREQUENCY (HZ)

L. :

l .

DESIGN VS. TIME HISTORY RESPONSE SPECTRA r E-W SSE 4% DAMPING ,

FIGURE 3-2

.,., I l

3-26 j;

1.

l i

.i i

o ,

I 8.NN '

f s ssa

, [J t 1

[ ;

e.ma ;

l

\

..am ;

8.N88 a

s. ism L 6

d5.1000 '

V ,

U i 4

g,gggi

.wam I- dI *I. 2 2

4 3 2

2 i 3 FREQUENCY (HZ)

L l '.

DESIGN VS. TIME HISTORY RESPONSE SPECTRA i

VERTICAL SSE 4% DAMPING FIGURE 3-3 L 3-27 1

- . , - - , - - . . . , - , , . . . . . ~ . . , ..- .. . ..--..... , --, -

l l

. ia, , . . . i. ia i. is ie a ACCELERATION T E HISTORY

~

__,1;,z;.1_

FIGURE 3-4 3-28

- _ . - . . - . . - - _ . . . - _ - ~ . . . - . - . - . - - _ . _ - . . . - - . _ . - - - . - - . . ._. .

. In- ;

i

.1 .

i i tl . : >

4.. .

.m -

I.

, . x t- :. ._

3. i -

.sE =z

} ; ,f-8 4 : l - ,- - '; _ l'[ f-

-' . ,2

.- ., J,. .v

. z

.= . .

!)

'. [h ) :

...i !, [ - - -

I .;I :

- ' . 1 ; g.

g ;;g ;7.

3 .

3 825 .

L <-

l' If l ![:- . ;- . ,: -

t:c - . : : ;; - .

l g: . .<-i - I--

1-

{, ,

l' Id: i

- - r f ; ' :"

,.. l.

- t , . ':; - ; -,

-l . 1 ::. . ;: _ g r ; - .

. I ; l

9. f [ -l ) f ff! e-(l 4

r ; 4 ::_ : : j. ; -; , -: _ ;i- y ;;

( ' ';

e..

'. O h

. ' .[:

q .

i -

. . j.. 4 }/ ..'. i ' i .

'i '! : . . . .. 3 0 .' ..l .:p .-

D( .Ig"

-- :,- E s

- 4: . 1. 3 .l : ;

!s 4-

.'.i y'.826 I 'i '

U 1.

-1 l

l [

. ..;.,."=- .-

! =;; ". , .-

. . . ] g{' . q . e i . }:

' " - [~

,}

5 (I .- , ) ~. 5-

,g

.'.* )-

~:

( p, .

. [ :' h _

'i 5 : ' ' ' j. .

- 1 I. . Et , .,

j. -g -

p . : .- .

.; .t ; ; ... ,

f' .'

3 ,. ! ' ; - - ., ' . - .. , :

.C75 l '! 'l : .

i

l i

In s 16 19 29 j 2 4 6 6 le 12 14 IIMC ISCCI 1 ACCELERATION TIME HISTORY l

->..... 4

. l o ACCELERATION TIME HISTORY l

E-w ssE FIGURE 3-5 3-29 l

l l

L

t

'"e 2 4 s e ie i: i4 is is n ACCELERATION TIME HISTOR1' ACCELERATION TIME HISTORY R 3 i

3-30 i

)

i l

\

REGION It REGIONI i

l

^$

No ,

m

*

p ,

L '4 m y * ' -

4l l l

'i

/ -' l b -a 1

Top Grid ( ,

A o

Jm Cell to Cell Connection Cell

\ l l

-- s -s: j l . .

At '

Base Plate Nu N- '

e e

s= =- y m y ""1

%/ ""

) $ up ds**

f ,

' e s ,.

$Upport Pad ,

h-

((kjh! -

N/

T STRUCTURAL MODELS REGION 1 AND 2 FIGURE 3-7 3-31

,, ..- .. - . . . . - - . - _ - . . - . - . - . . . - . . - . ~ . . - - - _ . . . - , - . . . . - .

i. 4 3

L 1

4 3

)

c 5

4l f '

l P s

.i P

4 k l 1

y ETRUCTURAL MODEL (9x10 MOD)

FIGURE 3-8 s l 3-32

'~'

c: , .

L 4 ;. s

.* 1 REGION I . REGION II 1

1 CELL Asstmetvm ~kNke$Ike ;

88" '55'*k'- I r-(M,ittt" I

i

)

f- '

4 l

) I A

a ,

j f'

ll /

l l

l ErT[ss" E fsg'ss " I av ,, -

f k s

EFFECTIVE STRUCTURAL MODELS REGION 1 AND 2 1

FIGURE 3-9 i

3-33

. . r

,v-. d. c. . . - - ., -. , , . ,-- . . , _ . - ..-.w . - . - - - - - . - - - , , - _ , ,. --,.,.a.- , - - - - , ~ - , - - - - , ,

. - - . -_. . _ ._. - ..~ _..___

k l

[ ;I$

N vb

~

I >>- /

@= <-

<1-

+ % v \,

h E3 ,

, s Nf >

i

.u .

s'

[ h N #

g ,

/

7 ;y. .

l

\ 'g \ f N/ +

I l

3-D NONLINEAR SEISMIC MODEL FIGURE 3-10 l

1 3-34 v s e -rs, ,.,,,-ven-- aww--+,- e,-., ,,en-o,w--

t l

j avm$ga,*gg,wass. i

!b!J8#A']

[

cett .r..t, i t t run. .. :

le

..c, l

7&u"ct" g '

MD -

i M,h,.5.P.a

.ti .

r4L.To.Ctgt

/ :

cu - -

. , l

/ !

ND. ..

7 ;

o h

MD

/

l e -

I iu.D

+,

i

+.

9

.%. / /

+

r 2D

+-

.- ,-, l t 7 ;

l

'#V.tM' Ltutut 0/ g,

/ i t?l'n,'r * ~ h & '

l i-f///////////// -

r

! NONLINEAR SEISMIC MODEL REGION 1 (2-D VIEW OF 3-D MODEL) .

FIGURE 3-11 .

3-35

1 i>

i WT W W& WAIIe

!hkJitik '

ctLL asscusLv

{

i ruct asscwso 1

w 'c N g-i w i@,'TriA G:

2D ' :-:

/

/ / /

,o<...d<-

s ia.c c+L,b. - w.,- /

eb :.: 7-

+.

mb

/

+.

(-)

/, !,

+. .

, mb.

+ :.: y/

mb

-: /

s6

+' :_: /

~ /

v.tw-kLtutut

/ [r p" S

/ .

l

+ N f

,pont pao-

[] D lll f '

NONLINEAR SEISMIC MODEL REGION 2 i

(2-D VIEW OF 3-D MODEL) )

FIGURE 3-12 3-36 l

. , , _ . . . . , - - - - - - , - , - ~ . - - - - - - - . .

~_ -

.. o

-t r\

i i t

{

c. '

t i ..

E g

t gL

E _

t l 1

s. ,. ,

[. g t i t i 4 AUXILIARY BU_lLDING_ - FUEL HANDLIRO AREA VALLS AT ELEVATION 143'-0' t t t L 1:. -

I

-3 _

)

d i ,

(

Q -s

, ..(

A L 9

ib

'2 J

,1.

c_ ,

a.

J i f

i ' 1, ' 0 ,

AUXILI ARY B1)_ll.J)1NG - T UE L HANDLING ARE A VALLS BELDV ELEVAT10N 119'-0' r!Guet 3 AUXILIARY BUILDING / FUEL HANDLING AREA (ELEVATIONS 143' & 1198)

FIGURE 3-13 3-37 ,

F, l

)

L

, . l l ... ,

t 1 f w >I ,,

h I

i f'

k 1

C -

(

ri ett .

sh.. !

( ,

f

T m e,,

Kh1 3

.i..i I:

t

., : t l ,

l f I._ i t

L"" 3 )

l [

l LONGITUDINAL SECTION A - A- ,

r

(. 1 i

f t

AUXILIARY BUILDING / FUEL HANDLING AREA i (LONGITUDINAL SECTION A-A)

FIGURE 3-14 ;

3-38 l

l 1

l

I c

SPJRTY EVALUATION l

4.0 I

4.1 DEGREE OF SUBCRITICALITY l The design of the racks is such that kw remains less than or equal to 0.95 under all conditions, including fuel handling accidents.

The close spacing of the racks precludes insertion of fuel assemblies in other than design storage locations, except in a area south of the Region 1 racks where a fuel assembly ray be i inserted between the pool wall and the racks. Such inadvertent !

insertion of a fuel into this location, or the placement of a fuel ;

assembly across the top of a fuel rack, is considered a postulated I accident, and as such, realistic initial conditions such as boron )

in the water can be taken into account. This condition has nn ;

acceptable Kg of less than 0.95.

4.2 GOVERNING CODES FOR DESIGN ;

Design of the facility in accordance with applicable USNRC Regulatory Guides, Standard Review Plans and the OT Position for i Review and Acceptance of Spent Fuel Handling Applications, ensures ]

adequate safety under normal and postulated accident conditions.

4.3 ABILITY TO WITHSTAND EXTERNAL LOADS AND FORCES i

The racks, being Nuclear Safety Class 3 and Seismic Category 1 structures, are designed to withstand all design basis loads. The pool structure has been evaluated for loads and loads combinations specified in USNRC SRP 3.8.4 and is adequate for all postulated loadings. Additionally, the racks are design with adequate energy absorption capability to withstand the impact of a dropped fuel ,

assembly from the maximum lift height of the fuel handling !

machine. Loads considered in the evaluation were !

+ Dead Load e f

Live Load

+ Wind Load ;

Equipment Loads

Operating Basis Earthquake

Safe Shutdown Earthquake

Tornado Loads

Thermal Loads Development of the above loads considered the submerged weight of the high density storage racks and their increased seismic loading '

on the fuel pool structure. Thermal load development was governed by the heat generated by a full core off load. Accident thermal load conditions were defined as a full core off load combined with .

a failure of the spent fuel pool cooling system. !

i Interaction of the free standing fuel storage rack structure and ,

the spent fuel pool structure were accounted for by the application of adynamic load factor to the submerge weight of the i racks and fuel elements.

4-1

[

I e ;

Detailed information on the postulation of loads and the !

evaluation of the structure's capacity to resist the imposed loads !

is documented in GAI Report 1949 and design calculation 5500- i 089.1.

4.4 ABILITY TO ENSURE CONTINUOUS COOLING l

One of the major features for ensuring continuous cooling to the !

spent fuel pools is the provision of two independent and redundant i spent fuel cooling subsystems which can be supplied with on-site ;

emergency povar. The partial or complete loss of spent fuel j cooling is highly unlir:ely because of the following reasons: ,

1. All piping and components in the spent fuel cooling loops are l Seismic class I. ,

i

2. The spent fuel cooling loops contain redundant pumps and heat exchangers. l i

3. Cooling pumps are supplied from separate electrical sources, each with the capability of being powered by separate emergency '

I diesel generators.

4. The systems that provide the utlimate heat sink for the spent fuel cooling heat exchangers are also Seismic class I systems ;

and are provided with redundant pumps and heat exchangers. ;

In addition to the Spent Puel cooling System, there are four f supplemental means of providing cooling water to the spent fuel pools:

1

1. The Decay Heat Removal System can be used to cool the spent l fuel pools. ;

2. The forced ventilation system above the pools can be uced to enhance the cooling effects of pool surface evaporation. ,

3. The Borated Water Storage Tank (BWST) water can be used for pool water makeup as well as for its cooling effect. ,

4. Time delays can be imposed on the transfer of spent fuel t assemblies into the fuel pools. The decay heat analysis very i I conservatively assumes that a full core is discharged into the ,

l spent fuel pools within seventy-two hoors of reactor shutdown.

No credit is taken for additional decay.

Inclusion of the BWST into the system as a second supplemental cooling scheme was analyzed. The BWST can be valved in parallel ,

to a spent fuel cooling heat exchanger. This arrangement utilizes !

in the extended BWST times water forand theprovides fuel pools a much to attain largera heat sink resulting'F.

temperature of 190 l' It is noted that inclusion of the BWST water decreases the pool '

temperature for the first several hours of the transient.

l 4~t 1

k 1

\

n 1 i

d.

4.5 PROVISIONS TO AVOID ACCIDENTAL DROPPING OF HEAVY OIL 7ECTS ON GPENT l FUEL As discussed in Section 5.3, the NRC has appioved Phase 1 of NUREG-0612 " Control of Heavy Loader at Nuclear Power Plants" for CR-3. Additionally, FPC is required to prohibit loads greater than 2750 pounds (the nominal weighc of c fuel assembly and handling tool) to be transported over spent fuel in the SFP, except during the removal of old racks and placement of new racks ;

in the pools. During this activity the fuel will be in the '

alternate pool which will be covered with missile shields. With j the upgrade of the Auxiliary Building overhead crane to be dorated

]

by 60 and 75 percent of its present capacity, the likelihood of 1 any load handling accident is suf ficiently small and no additional i restrictions on load handling operations in the SFP area are necessary. )

i 4.6 MATERIAL COMPATIBILITY l

1 All materials used in the construction of the racks are compatible '

with the storage pool environment and all surfaces that come into !

contact with the fuel assemblies are made of annealed aus;tenitic !

stainless steel. All the materials are corrosion resistant and !'

will not contaminate the fuel assemblies or the pool environment.

Venting of the Boraflex is accomplished through holes in the l stainless steel wrapper, intended for this purpose.

4.7 RADIOLOGICAL CONSIDERATIONS I I

Although the new high density racks will accommodate a larger j inventory of spent fuel, FPC has determined as discussed in '

Section 5.0 that the installation and use of the racks will not change the radiological consequences of a postulated fuel handling accident in the SFP area.

Additionally, expansion of the storage capacity of the SFP will !

not create any significant additional radiological effects. The i offsite integrated thyroid and whole body doses were calculated I using the calculation models and assumptions presented in R.G. I 1.25. The whole body and thyroid doses are well within the limits of 10 CFR 100. l 4.8 ABILITY OF RACKS TO WITHSTAND ACCIDENTAL LIFT FORCES t Analysis of the fuel racks demonstrates that the racks can withstand a maximum uplift load from the fuel handling machine 4 without violating the criticality acceptance criterion. The resulting stresses from this load are with acceptable limits and there is no criticality acceptance criterion to be violated. ;

4-3

1 5.0 C3ST/BRMEFIT ASSESSMENT i

l 5.1 COST / BENEFIT ASSESSMENT The cost / benefit of the chosen raracking alteration is ,

demonstrated in the following sections, j 5.1.1 l Need for Increased Storage capacity

a. Florida Power Corporation (FPC) currently has no contractual I arrangements with any fuel reprocessing facility. l

b. Crystal River 3 (CR-3) has two adjacent spent fuel pools on site, A and B. Fuel Pool A has been previously reracked. The l total number of cells available in both pools is 676. Of this i number, 120 cells are located in spent fuel Pool B. Because j of physical interferences in some areas of the A pool, only 642 cells of the 676 are usable for spent fuel storage.

Table 5-1 provides the proposed refueling schedules for future J fuel cycles at CR-3. The table shows the dates of refuelings and projected number of assemblies that will be transferred into the .

spent fuel pools until the total existing capacity la reached. !

All calculations in the table for loss of full core reserve )

(FCR) are based on the number of usable cells, not the total ;

number of cells in the pools, i

c. As of April 30, 1989, the Crystal River Unit 3 spent fuel pools !

contained 400 spent fuel assemblies. I i

d. Currently, the storage of components other than fuel has not affected the total number of available storage locations in i each pool because, for the most part, these components are ;

inserted in the fuel assemblies. An itemized list of components ;

stored in each pool is given in Table 5-2. !

i

e. Adoption of this proposed spent fuel storage expansion would ,

not necessarily extend the time period that spent fuel -

assemblies would be stored on site. Spent fuel could be sent off-site for final disposition under existing legislation. The government facility is expected to become available no sooner t than 2003, according to the June 1987 office of civilian Radioactive Waste Manacement (OCRWM) Mission Plan Amendment. f As matters now stand and until alternate storae facilities are available, spent fuel assemblies on site will remain there.

f. Table 5-3 references the spent fuel storage capacity for CR-1 after rerecking. Based on the current FPC fuel management policy and raracking the D pool, CR-3 will lose full core reserve after the refueling for cycle 18, in 2010, based on useable cell locations.

P 5-1

r. ;

1 I

G 5.1.2 Estimated Costs The costs associated with the proposed spent fuel pool !

modification are estimated to be approximately 2.5 million '

dollars. This figure includes the cost of; 1) engineering studies of spent fuel disposal alternatives, 2) design, engineering, manufacture, and installation of new spent fuel storage racks, and .

3) removal and offsite disposal (as low level radioactive vaste) !

of the existing spent fuel storage racks. The estimated values !

for uncertainty factors, cost escalation, and allowance for funds !

used during construction (AFUDC) are not included in this sum. !

l 5.1.3 Consideration of Alternatives

a. There are no operational commercial reprocessing facilitics !

available to meet FPC's needs, nor are there expected to be any in the forseeable future.

b. At the present time, there are no existing available independent spent fuel storage facilities. There are no firm commitments by either commercial firms or government agencies ,

to construct or operate an independent spent fuel storage 1 facility. In addition, cost and/or schedule considerations make ,

an independent spent fuel storage facility on site unacceptable l to meet the spent fuel storage needs at Crystal River 3.

I

c. Since Crystal River 3 is the only nuclear unit of Florida Power ,

Corporation, FPC does not have the option of tranushipping fuel J within the FPC system. l

)

d. The Cetober 1988 version (ver 28.8) of PROMOD was used to l simulate the annual replacement energy cost which would be incurred if CR-3 were prematurely shutdown. For the comparison, Cycle 10 was considered the last cycle before refueling capability is lost. Table 5-4 indicates the average yearly system cost increases for replacement power costs for i 20 years af ter reactor shutdown. This is the time through which I the plant is currently licensed to operate. Plant shutdown I would place a heavy financial burden on Florida residents within FPC's service area and cannot be justified.

5.1.4 Resources Committed Reracking of the spent fuel pools will not result in any irreversible and irretrievable commitments of water, land or air i resources. The land area now used for the spent fuel pools will i be used more efficiently by safely increasing the density of fuel storage.

l The materials used for new rack fabrication are discussed in l Section 3.7.1. These materials are not expected to significantly foreclose alternatives available with respect to any other licensing actions designed to improve the possible shortage of spent fuel storage capacity.

l >

s-a i l

_ - . , ----1

r i

5.1.5 Thermal Impact on the Etvironment Section 2.2.4.2 contains a description of the following considerations: the additional heat load and the anticipated maximum temperature of water in the spent fuel pool that would result from the proposed expansion, the resulting increase in evaporation rates, the additional heat load on component and/or plant cooling vatar systems, and whether there will be any significant increase in the ar.ou nt of heat released to the environment. As discussed in Section 2.2.4.2, the proposed increase in storage capacity will result in an insignificant impact on the environment.

5.2 RADIOlhGICAL EVALUATION 5.2.1 Solid Radwaste Currently, approximately 42 cubic

leet (2 cubic feet of filter and 40 cubic feet of resins) of solid radioactive wastes has been generated by the spent fuel pool purification system. No significant increase in volume of solid radioactive wastes is expected as a result of the expansion of the capacity of the spent fuel pool. It is estimated that an additional 20 cubic feet of resins will be generated by the spent fuel pool cleanup system but only during raracking.

5.2.2. Gaseous Radwaste There has been no Krypton-85 measured from the fuel building ventilation system for the last two years. The Semi-Annual l Radioactive Effluent Release Report includes a summary of the l l

quantities of radioactive liquid and gaseous effluents and solid waste released from Crystal River Unit 3 ar outlined in Regulatory Guide 1.21 (Revision 1, 1974).

5.2.3. Personnel Exposure The folloving discussion addresses expected increases in the doses :

to personnel from radionuclide concentrations in the spent fuel pool due to the proposed expansion

a. The range of values for recent gamma isotopic analyses of spent -

fuel pool waste is shown on Table 5-5.

b. Operating experience shows dose rates of 0.5 to 2.0 mrem / hour either at the edge or above the center of the spent fuel pools I regardless of the quantity of fuel stored. This is not ,

expected to change with the proposed reracking because radiation levels above the pool are due primarily to radioactivity in the water, which experience shows to return to a level of equilibrium. Stored spent fuel is so well shielded by the water above the fuel that dose rates at the top of the pool from this source are negligible.

5-3

i i

l

c. There is no routine concentrations of airborne radioactivity i from the spent fuel pools. The spent fuel pool ventilation i system dos:,gn provides a continuous sweep of air across the '

top 0* the spent fuel pools and cask loading pit. The system includes a continuous row of supply diffusers along the j southside of the pools and a continuous row of exhaust outlets ;

along the northside of the pools. Additionally, a continuous l exhaust flow is maintained from the enclosed top portion of the l pools when the top enclosing shield slabs are in place. All :

exhaust flow is directed to the main Auxiliary Building filter i system where it passes through roughing, HEPA, and charcoal filters before being discharged to the plant vent. The proposed raracking is not expected to increase this activity.

I

d. Operating plant experience with dense fuel storage has shown I no noticeable increases in airborne radioactivity above the l spent fuel pool or at the site boundary. No significant '

increases are expected from more dense storage. See section 5.2.3.c.

e. As stated in section 5.2.1 and based on operating experience with dense fuel storage racks, there is no significant increase in the radwaste generated by the spent fuel pool cleanup system. This is because operating experience has shown that with high density storage racks, there is no significant increase in the radioactivity levels in the spent fuel pool )

water. Operating experience with high density storage racks i has also shown no significant increase in the annual man-rem due to the increased fuel storage, including the changing of spent fuel pool cooling system resins and filters. Changing the racks to an even higher density will not change these conclusions. The spent fuel pool filters and domineralizers are located in a shielded cubicle not in a readily accessible area, and thus, will not present any radiation hazards should they become contaminated.

f. There has been no crud (e.g., "Co, "Co) built up alcng the sides of the pool. The highest possible water level is maintained in the spent fuel pool to keep exposure as low as reasonably achievable. Should crud buildup ever be detected on the spent fuel pool walls around the pool edge, it could easily be washed down or hydrolased.

g. During normal operation, the radiation zone designation of areac around the sides of the pools will not change due to raracking. The expected total man-rem to be received by personnel occupying the fuel pool area on all operations in that area is less that 200 man-rems. Two radiation criteria were considered in the analysis of the spent fuel pool to ensure shielding adequacy i.e., first, the concrete shielding surrounding the pool and second, the minimum water depth above the fuel elements. The shielding is adequate to maintain below tolerance dose levels for normal contamination of the pool l water by particulates. Similarly gaseous activity coming out s-4 l

t 1 i

l l

of solution from the pool water is picked up by a push-pull l type ventilation system over the pool. j l

As discussion in sections 5.2.3.c and 5.2.3.d, operating l experience has shown no noticeable increase in airborne i radioactivity for high density storage racks. j

\

operating experience has shown no increase in man-rem due to the j increased fuel storage with high density racks. Thereforo, no i increase in the annual man-rem is expected at Crystal River Unit 3 as a result of the increased storage capacity of the spent fuel pools with the higher density racks.

The existing Crystal River health physics program did not have to i be modified as a result of the previous increase in storage of !

spent fuel. It is not anticipated that the health physica program will need to be modified for this increase in storage capability.

]

5.2.4 Radiation Protection During Re-Rack Activities f

General Description of Protectivw Measures i The radiation protection aspects of the spent fuel pool modification are the responsibility of the Plant Health Physicist, who is assisted by his staff. Gamma radiation levels in the pool !

area are constantly monitored by the station Area Radiation Monitoring System, which has a high level alarm feature.

Personnel working in radiologically controlled areas shall wear protective clothing and respiratory protective equipment, I depending on work conditions, as required by the applicable Radiation Work Permit (RWP). Personnel monitoring equipment is assigned to and worn by all personnel in the work area. At a :

minimum, this equipment consists of a thermoluminescent dosimeter (TLD) and self-reading pocket dosimeter. Additional personnel monitoring equipment, such as externity badges, are utilized as required. ,

contamination control measures are used to protect persons from ;

internal exposures to radioactive material and to prevent the [

spread of contamination. Work, personnel traffic, and the i movement of material and equipment in and out of the area are ;

controlled so as to minimize contamination problems. The station

~

radiation protective staff closely monitors and controls all !

aspects of the work so that personnel exposures, both internal and external, are maintained as low as reasonably achievable (ALARA).

5.2.5. Rack Decontamination and Disposal There are six spent fuel storage rack modules that will be removed from the spent fuel pool. The wight of the racks is as follows: l

1 rack weighing approximately 3000 pounds e 3 racks weighing approximately 3300 pounds

2 racks weighing approximately 3700 pounds 5-5 1

m l

l The racks vill be decontaminated and disposed of at the J

Westinghouse Waltz Mill Service Center Decontamination, Disposal ;

and Recycle (DDR) Facility near Madison, Pennsylvania. The DDR i Facility is NRC licensed to transport, receive, store and process i l radioactively contaminated equipment and material for the purpose !

of decontamination and/or volume reduction and burial as radioactive waste. l I

Westinghouse will provide strong tight containers to be used for i exclus:.ve use, LSA shipment of the spent fuel racks. Westinghouse i will also provide transportation of the spent fuel racks from the !

Crystal River Nuclear Station to the DDR Facility. It is understood by Westinghouse that the material to be disposed of is :

as follows: i ii

+ Six (6) Spent Fuel Racks, to be removed from the Crystal River Unit 3 Nuclear Station, resulting in about 5,400 cubic feet of i' material requiring disposal.

. The Spent Fuel Racks will be hydrolazed at site and that radiation levels will be reduced sufficiently to allow for ;

exclusive use, LSA, unshielded shipnents. !

= Radiation levels will be less than 500-mrem /hr average on !

contact on any one side of the spent fuel rack. ',

5.3 ACCIDENT EVALUATION The accident aspects of the environmental review includes the !

o radiological consequences of the cask drop /tip accident and an !

l evaluation of the Safety Evaluation Report and Final Environmental l Statement to ensure that none of the conclusions reached in these documents with regard to spent fuel storage have changed i significantly.

5.3.1 Cask Drop /Tip Analysis The overhead crane used to handle the spent fuel shipping cask is ;

equipped with interlocks and/or physical stops which prevent i transferral of the crane and shipping cask over the spent fuel ,

pool. The operability of these interlocks and/or physical stops 1 are demonstrated once per seven (7) days during cask handling as -

required by CR-3 Technical Specification 4.9.7.1. In addition, I Technical Specifications 4.9.7.2 requires that prior to operating i the crane in the cask handling mode, we must verify that no fuel assemblies are in the storage pool adjacent to the cask loading area, and that the watertight gate between the two storage pools is in place and sealed. Maximum loads which can be transported l

over the spent fuel pools are limited by CR-3 Technical Specifications 3.9.6 and 3.9.7. The design of the spent fuel storage facility for CR-3 has been reviewed with regard to the cask drop /tip event by the NRC and has been found acceptable as ,

described in Section 9.1.2 of the Safety Evaluation Report for CR-

3. ] l l

5-6 1

FPC has upgraded the Auxiliary Building overhead Crane (FHCR-5) to enhance safe operation. The main hoist (120 ton) and e.uxiliary hoist (15 ton) will be derated to 60 percent of their present capacity. This will result in a safety factor being increased to a factor of 1.67. The third hoist (2 ton) will be dorated to 75 percent of its present capacity. This reduction of allowable loads will greatly reduce any probability for load drop accident.

Additionally, a redundant upper limit switch will be installed to prevent "two blocking" events.

5.3.2 NUREG-0612 Results The NRC cpproved Phase I of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", for Florida Power Corporation by letter dated July 13, 1984. The NRC letter of June 28, 1985 (generic Letter 85-11) indicated that satisfaction of Phase I guidelines assures that the potential for a load drop is extremely small.

The objective identified in section 5.1 of NUREG-0612 for providing " maximum practical defense in depth" is satisfied by Phase I compliance, and that the Phase II analyses did not indicate the need to require further generic actions. The NRC conclusion has been confirmed by the results obtained from the Phase II pilot program and additional Phase II reviews, which identified no residual heavy load handling concerns of sufficient significence to demand further generic actions.

5.3.3 Safety Evaluation Report and Final Environmental Statement FPC has reviewed the Safety Evaluation Report and the Final Environmental Statment and have concluded that the determinations reached by the Commission in these documents are not significantly l

altered as a result of our proposed modifications to the spent fuel storage facility at CR-3. This conclusion is supported by

the information provided within the Environmental and Cost / Benefit Assessment in section 5.0 of this safety analysis report.

5.3.4 Conclusion

The alternatives described herein do not offer the operating flexibility of the proposed action nor could they be completed as rapidly as the proposed action. These alternatives, i.e.,

reprocessing, independent storage facilities, or shipment to another reactor site, are more costly than the proposed action and might preempt storage space needed by another utility. The alternative of coesing operation of CR-3 is also more expensive

-than the proposed expansion spent fuel storage capacity because

l. of the need to provide replacement power. In addition to the economic advantages of our propossd action, we have determined that the expansion of the spent fuel pool would have negligible environmental impact and not proceeding with the expansion of the spent fuel pool at CR-3 would result in substantial harm to the public interest.

5-7

n-. -

s l

TABLE 5-1 i SPENT FUEL POOL CAPACITY WITHOUT RERACKING B POOL ~

i f

No. Assy in Pool Reload Capacity ,

Refuelinn Date After Refuel Cycle No. Site After Refuel -

September 1987 400 7 ** 242 March 1990 472 8 72 170' ;

March 1992 544- 9 72 98 '

March 1994 616 10 72 26 l l-March 1996 688 11 72 March 1998 760 12 72 March 2000 832 13 72 -

t March 2002 904 14 72 March 2004 976 15 72 -

March 2006 1048 16 72 March 2008 1120 17 72 1192 18 72 March'2010 March 2012 1264 19 72 March 2014 1336 20 72

{

March 2016 1408 21 72 l

4

'1 i

i l

' Full Core Reserve Lost ;

5-s 1

's I

TABLE 5-2 COMPONENTS STORED ~IN SPENT FUEL POOL Component Number Orifice Assemblies 2 Burnable Poison Rod Assemblies 163 Gamma Sample Holder 1 Control Rod Assemblies 61

" Dummy" Test Fuel Assembly 1 Control Component Storage Assemblies 3 Primary Sources 2 Axial Power Shaping Rods 8 i

5-9

p i

r TABLE 5-3 SPENT TUEL POOL CAPACITY AFTER RERACKING B POOL No. Assy in Pool Reload Capacity.

Refuelina Date After Refuel Cycle Jfat Size After Refuel-September 1987 400 7 80 242 March 1990 472 8 72 871 March 1992 544 9 72 799 March 1994 616 10 72 727 March 1996 688 11 72 655 March 1998 760 12 72 583 March 2000 832 13 72 511 March 2002 904 14 72 439 March 2004 976 15 72 367 i

March 2006 1048 16 72 295 March 2008 1120 17 72 223 March 2010 1192 18 72 151 2 March 2012 1264 19 72 79 March 2014 1336 20 72 7 March 2016 1408 21 72 -

l l

l

' Full Core Reserve Lost 5-10 ;

i L

! TABLE 5-4 SYSTEM PRODUCTION COSTS

($000)

Annual Replacement Energy Cost lgar For CR-3 1996 $143,815

, 1997 $216,628 1998 $184,061 1999 $238,209 2000 $205,126 2001 $272,047 2002 $233,383 2003 $303,880 2004 $261,991 2005 $345,929 2006 $299,086 2007 $389,151 2008 $322,804 2009 $416,068 2010 $351,496 2011 $450,857 2012 $380,187 2013 $485,646 1

2014 $408,879 2015 $520,435 2016 $437,570 l-1 l

5-11 l

u.

^

Q '

, )

! l TABLE 5-5 GAMMA ISOTOPIC ANALYSIS OF SPENT FUEL POOL WATER i' l ISOTOPES CONCENTRATION uCi/cc l

' l Xenon - 133 1.76 x 10 4 l i

Xenon - 133m 4 1.79 x 10 l 1

4 Manganese - 54 1.06 x 10 I

)

Cobalt - 60 4.88 x 10 4 I j

4 Cobalt - 58 1.16 x 10 !

1 4

kntimony - 122 2.42 x 10 Technetium - 99m 2.66 x 10 4 I Iodine - 131 3.49 x 10 4 !,

4 Cesium - 134 1.94 x 10 - f Cesium - 137 5.53 x 10 4 !

Antimony - 125 1.87 x 10 4 i

i

[

1 b

i I

1 P

5-12

, - a. , , ,

r l.

s.0 REFERENCES

{

1

1. W. E. Ford, III, CSRL-V, " Processed ENDF/B-V 227-Neutron-Group and '

Pointwise Cross-Section Libraries for Criticality Safety, Reactor i and Shielding Studies," ORNL/CSD/TM-160 (June 1982) .

2. N. M. George, et al, "AMPX A Modular Code System for Generating Coupled Multi-group Neutron-Gamma Libraries from ENDF/ B , " '

ORNI/TM-3706 (March 1976). ;

3. L. M. Petrie and N. F. Cross, " KENO IV--An Improved Monte Carlo criticality Program," ORNL-4938 (November 1975).

4. M. N. Baldwin, " Critical Experiments Supp0rting Close Proximity Water Storage of Power Reactor Fuel," BAW-1484-7, July 1979.

5. Askew, J. R., Fayers, F. J., and Kemshell, P. B., "A General e Description of the Lattice Code WINS," Journal of British Nuclear Energy Society, 5, pages 564-584, 1966.

6. J. T. Thomas, " Critical Three-Dimensional Arrays of U (93.2) --

Metal cylinders," Nuclear Science and Engineering, Volume 52, !

pages 350-359 (1973).

7. Harris, A. J., et al, "A Description of the Nuclear Design and Analysis Programs for Boiling Water Renctors," WACAP 10106, June 1982.

8. England, T. R., " CINDER -

A One-Point Depletion and Fission Product Program," WAPD-TM-334, August 1962. l

, 9 .- Melehan, J. B., " Yankee Core Evaluation Program Final Report,"

WCAP-3017-6094, January 1971.

10. WECAN - " Documentation of Selected Westinghouse Structural -

Analysis computer Codes," WCAP-8252. I

11. WECAN " Benchmark Problem Solution Employed for Verification of the WECAN Computer Program," WCAP-8929. ;

12. J. S. Anderson, "Boraflex Neutron Shielding Material -- Product Forformance Data," Brand Industries, Inc., Report 748-30-2 (August 1981). ;

13. J. S. Anderson, " Irradiation Study of Boraflex Neutron Shielding Materials," Brand Industries, Inc., Report 748-10-1 (August ,

1991).

5-1 )

f.

14. J. S. Anderson, "A Final Report on thu Effects of High Temperature Borated Water Exposure on BISCO Boraflex Neutron Absorbing Materials," Brand Industries, Inc., Report 748-21-1 (August 1978).

15. WCAP 8370, The Westinghouse Electric Corporation Quality Assurance Plan, Revision 9, Amendment 1, February 1978.

16. Barry, R. F., " LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094," WCAP-3269-26, September 1963.

17. Altomare, S. and Barry, R. F., "The TURTLE 24.0 Diffusion Depletion Code," WCAP-7758-A, January, 1975.

18. J. S. Anderson, " Irradiation Study of Boraflex Neutron Absorber-Interin Test Data," Brand Industries, Inc., Report No.

NS-1-050 (INTERIM), Rev. 1, 11/25/87.

19. USNRC Regulatory Guide 1.29, " Seismic Design Classification,"

Rev. 3, 1978.

20. D. F. DeSanto, "Added Mass and Hydrodynamic Damping of Perforated Plates Vibrating in Water", ASME Journal of Pressure Vessel T*chnoloay, May 1981.

21. 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. f

22. USNRC Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants," Rev. 1, 1973.

23. " Friction Coefficients of Water Lubricated Stainless Steels for a Spent Puel Rack Facility", Prof. Ernest Rabinowicz, MIT, a report for Boston Edison Company, 1976.

24. V. N. Shah, G. J. Bohn, and A. H. Nahavandi, " Modal Superposition Method for computationally Econonical Nonlinear Structural Analysis", ASME Journal of Pressure Vessel Technology, Vol.101, May 1979, pp. 134-141.

T l 25. V. N. Shah and C. B. Gilmore, " Dynamic Analysis of a Structure 7 with Coulomb Friction", Submitted for Presentation at the 1982 i I

ASME Pressure Vessel Piping Conference, Orlando, Florida, June '

27-July 2, 1982.

, 26. A. Higdon and W. B. Stiles, " Engineering Mechanics, Vector l' l Edition", Prentice-Hall, Inc. , Enge1 wood Clif fs, N. J. , 1962, p.

639.

t

27. " TRAM-Ihermal-Hydraulic Back Analysis Model; Description and Verification" by J.C. Buker, WNEP-8530, May 1985. .

l s-2 L

l

ML19325E888: Difference between revisions

Latest revision as of 01:34, 16 March 2020

Contents

Text

1.0 INTRODUCTION

6.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

1.4 CONCLUSION

1.5 REFERENCES

Navigation menu

ML19325E888: Difference between revisions

Latest revision as of 01:34, 16 March 2020

Text

1.0 INTRODUCTION

6.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

1.4 CONCLUSION

1.5 REFERENCES

Navigation menu

Search