ML20236N346

Fuel Consolidation Demonstration at Prairie Island Nuclear Generating Station

ML20236N346
Person / Time
Site:	Prairie Island
Issue date:	08/31/1987
From:	NORTHERN STATES POWER CO.
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ML20236N341	List: ML20236N338 ML20236N346 ML20236N359
References
NUDOCS 8708110463
Download: ML20236N346 (87)

Text

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l TOPICAL REPORT FUEL CONSOLIDATION DEMONSTRATION AT

-PRAIRIE ISLAND NUCLEAR GENERATING STATION Northern States Power Company August, 1987 8708110463 G70B06  !

PDR ADOCK 05000282 P PDR

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TABLE OF CONTENTS y

l. 1

-1.0 - Introduction' 1;

1.1 Consolidation of Spent Fuel at Prairie Island 1.2'. Purpose of; Report

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~2.0 Consolidation Process -

2.1 Equipment Description l 2.2 Process Description

3.0 -Criticality Analysis - 2:1 Consolidation

'3.1 Introduction l 3.2 Acceptance Criterion,for Criticality 3.3 Criticality Analytical Method

, 3.4 Consolidation Operations l

3.4.1 Design Description

.3.4.2- Design Criter_ia 3.4.3 Criticality Analysis 3.4.3.1 Fue1' Transition Canister 3.4.3.2 Cons. Rod Storage Canister and FA 3.4.3.2.1 Reactivity Equivalency & Methods 3.4.3.2.2 Reactivity. Calculations 3.5L Consolidated Fuel Storage in Racks 3.5.1 Design Description 3.5.2 Design Criteria

, 3. 5. 3 . Criticality Analysis - Spent Fuel Racks 3.6 Postulated Accidents

-3.7 Reference 4.0 Thermal / Hydraulic Analysis of Consolidated Fuel Storage Canister 4.1 Summary 4.2 Thermal-Hydraulic Analysis 4.3 Letdown Plate / Bottom Plate / Canister Hydraulics 4.4 Heating Rates .

4.5 Results and Discussion 4.6 Conclusions '

L 4.7.. References i

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5.0 Spent Fuel Pool Cooling System Evaluation 5.1 Introduction 5.2 Methods and Assumptions 5.3 Forced Cooling Conditions 5.4 Total Loss of Forced Cooling 5.5 References 6.0 Radiological Analysis of Handling Accidents 6.1 Canister Drop on Floor 6.2 Canister Drop onto Storage Racks 6.3 Summary of Results 6.4 Other Potential Accidents 6.5 Conclusions 7.0 Mechanical, Material, and Structural Considerations 8.0 Consolidation Demonstration Experience of Prairie Island 9.0 Conclusions 9.1 Limiting Conditions for Consolidation 9.1.1 Fuel Assembly Characteristics 9.1.2 Pool Conditiens 9.2 Conclusion of No Unreviewed Safety Question 4

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i LIST OF TABLES I l

TABLE NO. TITLE I

3-1 Benchmark Critical Experiments l 3-2 Fuel Parameters Employed in Criticality Analysis 3-3 Comparison of PHOENIX Isotopic Prediction to Yankee Core 5 Measurements i

3-4 Benchmark Critical Experiments PHOENIX Comparison 3-5 Data for U Metal and U02 Critical Experiments i

5-1 Summary of Pool Cooling System Evaluation - Forced Cooling 5-2 Summary of Pool Cooling System Evaluation - Loss of Forced Cooling i

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, . LIST OF FIGURES FIGURE NO 2, LTITLE-F 2-1. Fuel Assembly Dismantling Station 9

'2-2 .Storag'e Can Loading.. Station (A) 2-3. Storage'Can' Loading Station (B) 2-4 Location of Work Stations.

'3-1 Mir.imum Burnup cys. Initial U-235 Enrichment for Fuel Assembly

'Consolidad on Operations 3-2 Consolidated' Rod Storage Canister Nominal Dimensions

'3-3 Prairie Island Spent Fuel Storage Cell l 4-1 Basic Canister Geometry Showing Natural Circulation Cooling L,

4-2 Rod-to-Rod Flow Passageway 4-3 Letdown Plate Flow Hole. Pattern 4-4 Decay Heat / Rod vs. Time After Shutdown 4-5 Cecay Heat / Rod.vs. Time After Shutdown 4-6 Time After Shutdown vs. Operating Time 5-1 Approximate PI Spent Fuel Pool Cooling System Capability -Steady State Evaluation ,

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FUEL CONSOLIDATION DEMONSTRATION AT PRAIRIE ISLAND NUCLEAR GENERATING PLANT

1.0 INTRODUCTION

1.1 CONSOLIDATION.0F SPENT. FUEL AT PRAIRIE ISLAND

' Northern States Power's (NSP) Prairie Island Nuclear Generating Plant consists: of two pressurized water reactors,. Units 1 and 2. The current NRC Operating . Licenses expire in. 2013, for Unit 1, and 2014, for Unit 2. The Prairie Island (PI) Spent Fuel-Storage Facility includes two storage pools shared-by both reactors. When both pools'are filled with racks, 1582 i storaget locations are provided. However, PI's Technical Specifications

-limit total storage to 1386 spent fuel assemblies, not including those

. assemblies which can be returned to the reactor. Under current cycle l 1

manage' ment plans, PI's spent fuel storage capacity will be exhausted by 4 about 1994.

i NSP has entered into contract with the Department of Energy (DOE) for i disposal of spent fuel, as required by the Nuclear Waste Policy Act (NWPA) of 1982. The DOE is required to take title to spent fuel in 1998, when the first repository was scheduled to begin operating. However, the DOE is currently. seeking a five year delay for this milestone. It is possible the-DOE may be able to accept spent fuel from utilities by 2000, if Congress -

..will authorize the Monitored Retrievable Storage facility proposed by the 1

00E. 'In.the interim, each fuel owner has the primary responsibility for 1-1 L - _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ -

, , storage for i_ts own spent fuel by maximizing the effective use of existing l storage facilities at its sites, and by adding new onsite storage capacity in.a timely manner. )

'.Thus Prairie' Island will exhaust' its existing spent fuel storage capacity before the DOE will be ready to accept spent fuel for either monitored retrievable storage or disposal. The spent fuel consolidation demonstration

- program is part of NSP's efforts to meet Prairie. Island's spent fuel storage needs.

NSP's consolidation demonstration will take place in the PI fuel transfer canar adjacent to the spent fuel storage pools. Westinghouse technicians, using equipment and processes developed by Westinghouse and reviewed and approved by NSP, will consolidate up to a maximum of 50 spent fuel assemblies. Previous efforts in the industry have demonstrated the technical feasibility of fuel consolidation, achieving a 2 to 1 consolidation ratio.

NSP's demonstration will show whether large scale consolidation (1000 or more assemblies) can be implemented at PI both efficiently and economically.

Once the demonstration is complete, NSP will evaluate the results and decide whether to implement consolidation at PI on a large scale. If the decision is made to proceed, NSP will then apply to the NRC for a License Amendment to increase PI's spent fuel storage capacity. NSP would al'so apply to the State of Minnesota for a certificate of Need, as required by state law.

1-2

1.2 PURPOSE OF REPORT The purpose of this' report is to fully document all aspects of the Prairie Island spent fuel consolidation program and to provide verification that the program does not involve an unreviewed safety question. .

The report discusses the following specific aspects of the spent fuel consolidation demonstration program:

o Description of the consolidation equipment, including mechanical  !

and structural design features and material considerations.

o Description of consolidation process.

o Criticality considerations during consolidation process, and of storing consolidated fuel canisters in PI's racks.

o Thermal-hydraulic effects of consolidating fuel rods from two assemblies into a single canister.

o Evaluation of' radiological effects of postulated fuel handling accidents during consolidation.

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This report also contains a pool cooling system evaluation for the increaied heat load assuming maximum utilization of consolidation. Although this 1-3

.________-___________a

I analysis does not pertain to the demonstration, it is included here # -

convenience.

1 This report does not contain any structural analyses. NSP has had structural analyses performed on the pool and the racks for the increased fuel load, assuming maximum utilization of consolidation. Again, these analyses do not pertain to the demonstration. The results are available in two separate reports.

1-4 u_______________ ________

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, . 2.0 CONSOLIDATION PROCESS The consolidation process involves the removal of the fuel assembly nozzles and the fuel rods from a fuel assembly and preparing them for storage in the spent fuel pool. The nozzles are stacked on top of each other and are placed in the fuel racks. The fuel rods are packaged closely together in storage canisters and the canisters are placed in the spent fuel racks. The nozzles from 40 fuel assemblies fill slightly more than two cells in a rack

'and the storage canisters occupy twenty cells. The empty fuel assembly skeletons will be stored for volume reduction at a later date.

-2.1 Equipment Description The equipment used to accomplish fuel consolidation is listed below, followed by a brief description of the equipment items and their components.

o Fuel assembly dismantling station o Storage can loading station o Nozzle stacking station o Debris removal, water filtration, and underwater TV systems ]

i o Other tools and equipment 1 l

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A. Fuel Assembly Dismantling Station (Figure 2-1) l l

At this station, the top and bottom nozzles are removed from the fuel assembly, the fuel assembly is rotated to the upside down position, 'nd a

the fuel rods are removed. The following sub-assemblies / components make up the full compliment of equipment used at this station:

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, , 1. Elevator / Rotator This supports the fuel assembly during dismantling. It allows the fuel assembly to be raised and lowered for access to the nozzles and rods. Rotation allows access to the bottom nozzle. It hangs from the deck and is bolted to the deck to provide horizontal position stability. The design of this elevator / rotator is basically the same as the Multi-function repair station (MFRS) which Westinghouse has used at several utilities for fuel assembly repair and reconstitution.

2. Fuel Transition Canister (FTC)

The fuel rods are pushed from the fuel assembly into the FTC in preparation for loading into the consolidated rod storage canister. The FTC changes the fuel rod array from an open square array to a closed triangular array.

3. Transition Canister Support Stand The stand positions the transition canister under the fuel l

assembly and also positions it where it can be lifted by an overhead crane.

4. Rod Push Tool s l

The rod push tool is used to push the rods out of the fuel ]

assembly and into the FTC. Rods will be pushed two at a time. 'A guide plate locates the tool over the fuel rods.

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.. . 5. ' Nozzle Removal Tools u

L These.are special cutting tools that are designed to remove the 4 top and bottom nozzles. They cut the thimbles and the thimble 1

screws.

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6. -Handling. Tools

.There'are.several long handled tools which extend from the deck to the' fuel assembly and equipment below water level. They are used to move items and to actuate equipment that is under water.

B. Storage Can Loading Station (Figures 2-2 and' 2-3)

The fuel rods are transferred from the transition canister to the storage can at this station. Following is a' brief description of the sub-assemblies of this station and other equipment.used at this point:

1. Winch The winch is mounted on the deck,'across the feel transfer canal, t

and it is used to raise and lower the consolidated rod-storage canister (CRSC) loading frame and to move the~ frame laterally.

I 1 2. CRSC Loading Frame The CRSC loading . frame supports the fuel transition canister (FTC) and the CRSC. It is an angle iron structure.

3. Rod Position Indicator I

2-3

i The rod position indicator is used to assure that each fuel rod is l exiting the FTC and entering the storage canister during fuel rod transfer. It is made of a bundle of bars, one for each fuel rod.

The bundle hangs from the winch.

4. Consolidated Rod Storage Canister (CRSC) l The objective of the consolidation process is to place the fuel rods in'the CRSC with a 2 to I ratio. To do this, the CRSC is designed to contain the fuel rods from two fuel assemblies (358 ,

rods) and yet fit into a fuel rack cell. This dictates that the cross sectional area of the CRSC be as small as structural design will allow. To this end, the CRSC is made of .050-inch thick stainless steel sheet metal.

A partition divides the canister cross section in half. This allows the fuel rods from one fuel assembly (179) to be placed in the canister at a time. The partition extends above the top end of the canister to serve as a lifting lug. Plates are riveted to both sides of the extension for reinforcement.

Snap-in covers are used on each side of the lifting lug to close  !

off the top end of the storage canister. Fuel assembly and storage canister identification numbers are located on the covers and canister walls for fuel rod accountability in accordance with ANS-57.10, " Design Criteria for Consolidation of LWR Spent Fuel".

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, The bottom of the storage canister includes a letdown pan which is a plate on which the fuel rods rest on. The pan is also used to l l

support the rods during the earlier stages of the consolidation process. It is initially installed in the bottom of the i j

transition canister to support the fuel rods as they are pushed j l

from the fuel assembly. It is transferred from the transition canister to the storage canister with the rods, where it remains.

It rests on an open plate at the bottom of the storage canister.

The open plate allows the letdown support columns to extend into the canister during rod loading.

5. Letdown Pan Support Columns When the transition canister is in the storage canister loading frame, the letdown pan rests on the columns. The columns are two lengths of 2.1/2-inch piping which extend through two holes in the bottom of the canister. There are two holes on each side of the canister partition.

I C. Nozzle Stacking Station The nozzles are stacked on top of each other as they are removed from the fuel assemblies. The top nozzles are stacked separately from the bottom nozzles in groups of ten per stack. Tie rods align the nozzles and hold the completed stacks together. The completed stacks will fit into a fuel rack cel' one stack upon another. A lifting eye is attached to the top for stack handling.

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The~ work ' station wi11 hang from the deck beside.the fuel assembly dismantling station. The top of the ' stack will be about five' feet below water level. Various long handled tools will be used to stack the nozzles and to operate and lock fasteners.

D. -Debris Removal, Water Flitration, and Underwater TV Systems These systems will support the process. For example, during some of l the operations, such as cutting the nozzles from the fuel assembly, ij' ' metal chips will develop. An underwater vacuum system will be used to

-[ ' pick up this debris. . Also, there.is a possibility of. crud clouding the water during rod removal. In this' instance, filtration will be used since underwater visibility is needed to monitor the process with TV.

These systems will be used mainly around the fuel assembly dismantling

. station. -The TV'will'also be used to monitor loading of the storage cenisters and' installation of their covers.

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E. Other Tools and Equipment All of the equipment described previously, with the exception of the storage canisters, is equipment owned by Westinghouse. It will be removed fr6m'the Prairie Island plant'upon completion of each phase of

-the fuel-consolidation program including'this demonstration phase.

Ot'er h equipment, however,.is being supplied to Northern States Power l Company and will remain at the Prairie Island site. These are:

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Fuel rod storage canister handling tool'

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2. -Stray and broken rod storage can 3.s Fuel assembly skeleton handling tool 4., Nozzle stack handling tool

'The storage cani. ster handling tool is a custom design that is suitable for underwater . engagement with the lifting lug on the canister. The load being lifted will be 2500 pounds and, therefore, the tool will be

-designed in accordance with the required. standards.

i The stray and broken rod storage can is a design that has been used'on previous projects. It has 57 tubes each capable of holding a single fuel: rod. The tubes are supported vertically by a plate and angle iron structure. 'The whole assembly fits in a fuel rack cell. A top, similar to.a fuel. assembly top nozzle, is attached so the canister can be. lifted with the fuel handling tool.

A tool that is presently at Prairie Island may be used for fuel assembly skeleton handling. It uses a hook for engaging the grids on the skeleton. The tool will be used to move the skeleton from the fuel assembly dismantling station to the fuel racks.

l The' nozzle stack handling tool will be used to move the stacks from the stacking station to the fuel racks. A nozzle stack will weigh about 500 pounds. The tool is designed to engage a lug at the top of the -

stack. The tool for lifting the storage cans may also be used to lift the nozzle stacks.

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. 2.2': Process Description-

- The equipment .will be located in the fuel transfer canal (Figure 2-4). The location isolates the consolidation equipment from the stored fuel assemblies. At all times during consolidation processes, at least five feet of water shielding will be

!' maintained above,the fuel. pins. This amount of shielding is based on' previous Prairie Island experience. The minimum water shield will be., assured by a combination of equipment design, administra'tive controls, and local and control room alarms.

Following is a step by step description of'the consolidation process.  !

1. :The basket for holding the fuel assembly is lowered on the elevator.to allow fuel assembly insertion.

2. The fuel assembly is moved from the racks to the basket.

-3. The fuel assembly is raised closer to the surface to facilitate dismantling.

4. The bottom and top nozzles are cut loose from the thimbles. l

5. The top nozzle is moved to the nozzle stacking station.  !

6. A catch grid is placed over the fuel assembly.

7, The fuel assembly is rotated to the upside down position and the bottom nozzle removed.

8. The bottom nozzle is moved to the nozzle stacking station.'

9. A guide pla+.e'is placed over the fuel assembly.

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10. The fuel transition canister (FTC) is rolled under the fuel <

assembly.

11. The fuel rods are pushed from the fuel assembly and into the ,

FTC.

12. The FTC is rolled from under the fuel assembly.

13. The FTC is lifted and placed in the consolidated fuel storage canister (CRSC) loading frame.

14. The rod position indicator is placed over the FTC.

15. The CRSC loading frame is raised (Figure 2-3). This lifts the FTC off the rods and pulls the CRSC over the rods. The rod position indicator above and the support columns below

. pravent the fuel rods from moving up or down.

16. The frame is moved horizontally about four inches to place the empty half of the CRSC over the letdown pan support columns.

17. The frame is lowered. At this stage, fuel rods are in one half of the CRSC and the letdown pan support columns extend into the other half of the can.

18. The cover is installed on the loaded half of the CRSC.

19. The rod position indicator is removed.

20. The FTC is removed, a new letdown pan is installed, and the FTC is placed on the transition canister support frame.

21. The guide plate is removed from the fuel assembly dismantling station.

22. The basket is lowered and the fuel assembly skeleton is -

picked up and delivered to storage. I 2-9

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. . 23. The empty basket is rotated to the upright position and the catch grid is reaioved. _

This completes the consolidation of'one fuel assembly. Since the consolidation ratio is 2 to 1, the consolidation process involves pairs of fuel assemblies. The process for the second fuel assembly is the same as the first, except that it changes slightly.

. after step'15; the frame is moved horizontally to a third position to place the letdown pan support columns outside of the storage can.

24. The frame is lowered a second time. At this stage, both sides of the storage can are loaded with fuel rods.

25. The cover is. installed on the other half of the CRSC.

26. The rod position indicator is removed.

27. The FTC is removed, a new letdown pan is installed, and the

, canister is placed on the transition canister support frame.

28. The CRSC is moved to the fuel rack.

29. ;The guide plate is removed from the fuel assembly dismantling  ;

station.  !

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30. The' basket is lowered and the fuel assembly skeleton is

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picked up and. delivered to storage. i

31. ..The empty basket is rotated to the upright position and the catch grid 'is removed.

. l This completes the consolidation of a pair of fuel assemblies and ]

completes the loading of one CRSC. The objective is to perform )

these operations in a 10-hour work shift. 1 i

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- Rod Push Tool

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Deck 7,

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MFRS-- , _

, Guide Plate -

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4, x-hi --Fuel Assembly

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, .i Transition Canister Support Frarne L 3n u u 4

Mransition Canister

.e et Down Pan ,

. l, E k 1 Figure 2-1 luel Assembly Dismantling Station

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M,n. i k, Winch (2) d l . Deck 4

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gf et Down Pan (c= 7 p Storage can Loading Frame r

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Let Down Pan Support Columns 1

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'I t d' Figure 2-3 Storage Can I.cading Station (B)

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.. i Pool #2 Pool fl Location of " - -

q Work Stations' 7 -

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i I New Fuel L--, ,_l -  !

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I' l-Figure 2-4 Incation of Work Stations ,

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- 3.0 CRITICALITY ANALYSIS - 2:1 CONSOLIDATION 1

3.1 Introduction I l

i The consolidation equipment, i..e., the fuel transition canister (FTC) and l

the consolidated rod storage canister (CRSC), criticality analysis is based I l

on maintaining Keff 5 0.95 during the consolidation operations involving the use of this equipment with fuel at 4.0 w/o U-235 and an initial enrichment /

3 burnup combination in the acceptable area of Figure 3-1.

The Prairie Island spent fuel rack design described herein was analyzed for criticality to show that fully or partially loaded consolidated rod storage canisters (CRSC) can be stored in the fuel racks.

The spent fuel rack analysis is based on maintaining K,ff 5 0.95 for storage of Westinghouse 14x14 0FA and STO fuel rods and EXXON 14x14 HI-PAR, LO-PAR and TOPROD fuel rods at 4.0'w/o with utilization of every cell permitted for storage of the CRSC, i l

l 3.2 Acceptance Criterion For Criticality The neutron multiplication factor of the fuel handled in the fuel pool during the-fuel consolidation operations and of the fuel stored in the pool l

for any possible configuration shall be less than or equal to 0.95, t

including all uncertainties, under all conditions.

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' The analytical methods employed herein conform with ANSI N18.2-1973,

" Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Handling System; ANSI 57.2-1983, " Design

~ 0bjectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations,"

Section 6.4.2; ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety,"; NRC Standard Review Plan, Section 9.1.2,

" Spent Fuel Storage"; and the NRC guidance, "NRC Position for Review and 8.17-1984, Acceptance of Spent Fuel Storage and Handling Application," ANSI

" Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors."

3.3 Criticality Analytical Method The criticality calculation method and cross-section values are verified by I comparison with critical experiment data for fuel rods similar to those to be consolidated. This benchmarking data is sufficiently diverse to establish I that the method bias and uncertainty will apply to conditions which include strong neutron absorbers, large water gaps and low moderator densities.

The design method which insures the criticality safety of fuel in the spent fuel storage pool uses the AMPX system of codes (2,3) for cross-section generation and KENO IV(4) for reactivity determination.

The 227 energy group cross-section library (2) that is the common starting point for all cross-sections used for the benchmarks and the storage rack is The NITAWL program (3) includes, in this generated from ENDF/B-V data. 2 library, the self-shielded resonance cross-sections that are appropriate for 3-2

g

-< ~ each particular geometry. The Nordheim Integral Treatment is used. Energy and spatial- weighting of cross-sections is performed by the XSDRNPM

- program (3) which.is a one-dimensional nS transport theory code. These multigroup cross-section sets are then used as input.to KENO IV(4) which is

. a three dimensional Monte Carlo theory program designed for reactivity calculations.

A set of 33 critical experiments has been analyzed 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 (boroflex, steel, water, etc) that simulate LWR fuel shipping and storage conditions (5) to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials (6) (Plexiglas and air) that demonstrate the wide range of applicability of the method. Table 3-1 summarizes these experiments.

The average K,ff of the benchmarks is 0.992. The standard deviation of the bias value is 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 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0018 delta k.

3.4 CONSOLIDATION OPERATIONS ,

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.3.4.1 Design Description 3-3

The FTC and CRSC design drawings used in the analysis are shown in

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l References 11 and 12. The CRSC design is depicted schematically in Figure 3-2 with nominal dimensions given on the figure. The sequence of operations involving the use of this equipment in the fuel consolidation process is given in Section 2.2.

3.4.2 Design Criteria Criticality of the fuel in the consolidation process is prevented by the design and operation'of the consolidation equipment which limits fuel interaction. This is done by performing operations with a number of fuel ,

1 rods that is less than that required to achieve criticality in the normal operating environment.

l The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (Keff) f the fuel array will be less than 0.95 as recommended in ANSI 57.2-1983, 3.4.3 Criticality Analysis The consolidation process will be performed using the sequence of operations given in Section 2.2. In this sequence of operations, the fuel configuration will change significantly under normal operating conditions.

To bound these fuel configurations from a criticality bases, the following fully or partially loaded configurations will be considered under normal operating conditions:

3-4

, . a) Fully and partially loaded FTC b) Fully and partially loaded CRSC c) Fully and partially loaded fuel assembly The following assumptions were used to develop the nominal case KENO models for analysis of the configurations described above:

j a) The fuel contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any l 1

burnable poison in the fuel rods. Calculations have shown that

, t the W 14x14 STD fuel pins yields a equal or larger Keff than does the W 14x14 0FA or the EXXON HI-PAR, LOW-PAR, and TOPROD fuel when all pins have the same enrichment. Thus, only the W 4 14x14 STD fuel pins were analyzed (see Table 3-2 for fuel parameters).

The moderator is pure water at a temperature of 68 F. A b) 3 conservative value of 1.0 gm/cm is used for the density of water.

c) No credit is taken for any spacer grids or spacer sleeves.

d) The arrays are infinite in the axial extent which precludes any neutron leakage from the ends of the arrays.

As a result of these basic assumptions and the similar sizes of the fuel-assembly and CRSC fuel envelops, the fully and partially loaded fuel assembly and CRSC cases can be bounded by the same case.

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, 3.4.3.1 Fuel Transition Canister 'l l

The fully or partia11y' loaded fuel transition canister (FTC) is bounded by l

ensuring that the maximum K,ff that can result from any amount of fuel-loaded'into the FTC envelop or area does not exceed the design. limit of 0.95. :The FTC. envelop or area is a function of elevation, therefore for conservatism the fuel area of the largest grid opening'in the FTC is used to

~ determine the maximum K,ff. 'he maximum FTC fuel area arises from consideration of mechanical tolerances resulting from the manufacturing

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process on the large'st grid opening. The tolerances are stacked in such a way as to maximize the grid ' opening.

2 As a result,.the' largest FTC fuel area is determined to-be 97.2 in . The 2

model consists of a' number of evenly spaced fuel rods in the 97.2 in area, each surrounded by the stainless' steel tubes in the FTC. -No credit is taken

'for any other structural material in the FTC. -The number of fuel rods in

' the fuel . envelop is varied until the maximum Keff is reached with a fuel rod enrichment of. 4.0 w/o U-235.

Results from the sensitivity study show that the maximum reactivity occurs when there are 185 fuel rods in the FTC maximum fuel area.

l Based on the analysis described above, the following equation is used to develop the maximum K,ff for the partially or fully loaded FTC.

.K,ff = Kworst- +O method worst +(ks[ method

+ [(ks)2 3-6 L_ ___

- Where:

l w rst case KEN 0 K eff that includes material tolerances, K

=

worst and mechanical tolerances which result in the largest FTC

- - fuel area. -

B = method bias determined from benchmark critical method comparisons ks = 95/95 uncertainty in the worst case KENO Keff orst ks = 95/95 uncertainty in the method bias method Substituting calculated values in the order listed above, the result is:

K,77 = 0.8800 + 0.0083 + [(0.0054)2 + (0.0016)2 3 1/?. = 0.8940 The K,ff for the configuration is less than 0.95 including uncertainties at a 95/95 probability / confidence level. Therefore, the acceptance criteria for criticality are met for the transfer of fuel rods to the FTC with an enrichment of 4.0 w/o U-235.

4 3.4.3.2 Consolidated Rod Storage Canister and Fuel Assembly The fully and partially loaded fuel assembly and consolidated rod storage canister (CRSn are bounded by ensuring that the maximum Keff that can i result from any amount of fuel loaded into the fuel assembly or CRSC fuel 3-7

_ _ _ _ ~ .. . .

- envelop or area does not exceed the design limit of 0.95. To meet this V design limit, credit is taken for the reactivity decrease caused by fuel

' depletion'in fuel a semblies'that have initial enrichments greater than 3.0 w/o U-235 This methodology, known as reactivity equivalencing, is f-described below.

i 3.4.3.2.1 Reactivity Equivalencing and Methods l

l Spent. fuel. consolidation.of fuel assemblies with initial enrichments greater than'3.0 w/o are achievable by means of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decreass associated with fuel depletion. A series of reactivity calculations are performed to generate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield the equivalent K,ff. .

Figure 3-1 shows the constant K,ff contour generated for the Prairie Island fuel. Note tae endpoint at 0' MWD /MTV where the enrichment is 3.0 w/o and at 4,000 MWD /MTU where the enrichment is 4.0 w/o. The interpretation of the endpoint data is as follows: the maximum reactivity of the fuel assembly or CRSC, containing fuel at 4,000 MWD /MTV burnup which had an initial l

enrichment of 4.0 w/o is equivalent to the reactivity of the fuel assembly It or CRSC containing fresh fuel having an initial enrichment of 3.0 w/o.

is important to recognize that the curve in Figure 3-1 is based on a constant maximum reactivity resulting from a fully or partially loaded fuel

. ' assembly or CRSC.

i 3-8

! -j 4 The data points'on the reactivity equivalence curve were generated with a transport theory computer code, PHOENIX U) .

PHOENIX is a depletable, i

two-dimensional, multigroup, discrete ordinates, transport theory code. A l.

1 i -25 energy group nuclear data library based on a modified version of the British WIMS(8) library is used with PHOENIX.

~

A study was 'done to examine fuel reactivity as a function of' time following discharge from the reactor. Fission product decay was accounted for using CINDER (9) CINDER is.'a point-depletion computer code used to determine fission product activities. The fission products were permitted to decay j

for 30 years after discharge. The fuel reactivity was found to reach a J l maximum at approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after discharge. At this point in time, 135 , has nearly completely decayed away.

the inajor fission product poison, Xe l

Furthermore, the fuel reactivity was found to' decrease continuously from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 30 years fo' lowing discharge. Therefore, the most reactive point h

in time for a fuel assembly after discharge from the reactor can be 135 conservatively approximated by removing the Xe ,

'The PHOENIX code has been validated by comparisons with experiments where 1 isotopic fuel composition has been examined following discharge from a f

, j reactor. In addition, an extensive set of benchmark critical experiments has been analyzed with PH0ENIX. Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 3-3.

The n, measurements were made on fuel discharged from Yankee . TheCore 5(10) data in Table 3-3 shows that the agreement between PH0ENIX predictions and . .I measured isotopic compositions is good. i 3-9 1 i

_L x _ __ _ __ _ _ _ _ _ _ _ - _ .

, _ _ _ _ _ _ ._. ____._- _ _ ___ ___-_ _ _ A

Thel agreement between reactivities computed with PH0ENIX and the results of

' 81' critical benchmark experiments is summarized in. Table 3-4. Key parameters describing each ofLthe 81' experiments are given in Table 3-5.

These reactivity-comparisons again show good agreement between experiment and PH0ENIX calculations. .

~

An uncertainty. associated with the burnup-dependent reactivities computed with PHOENIX is accounted for in the development of the maximum Region 2 multiplication factor. An. uncertainty of 0.01 delta k is considered to be very conservative since comparison between PH0ENIX results and the Yankee Core experiments and 81 benchmark experiments indicates closer agreement.

'3.4.3.2.2 Reactiv'ity Calculations Since the fuel assembly and the CRSC have nearly the same nominal fuel envelope .or area, the largest area is used to bound both the fuel assembly l

and the CRSC. The envelop of the largest fuel area is determined to be the L CRSC. The maximum CRSC fuel area ari.ses from consideration of mechanical and material thickness tolerances resulting from the manufacturing process. The

~

i tolerances are stacked in such a manner to maximize the fuel area. The CRSC inside diameter (ID) is increased from its nominal value to a maximum of 7.963 inches. Therefore the area to be loaded with fuel for the sensitivity -

1 2

study is 63.4 in ,

2

-The model consists of a number of evenly . spaced fuel rods in the 63.4 in .

area. Cold " clean" water is placed around the fuel rods and around the fuel envelop. There are no structural materials used for the boundary of the 3-10

l fuel area. In this way no credit is taken for the structural material that 1

I would be in a fuel assembly or the CRSC. The number of fuel rods in the 2

-63.4 in fuel area is varied until the maximum K,ff is reached.

1 Results from the sensitivity study show that the maximurn reactivity occurs when 153 fuel rods are evenly spaced in the fuel envelop.

Based on the analysis described above, the following equation is used to develop the maximum K eff for the partially or fully loaded CRSC and fuel assembly.

K,ff = Kworst +Bmethod + @ sworst) +(ksmethod) + (ksre) where:

K = w rst case KEND K eff that includes material tolerances, worst and mechanical tolerances which result in the largest CRSC fuel area.

B = method bias determined from benchmark critical method comparisons ks orst

= 95/95 uncertainty in the worst case KEN 0 Keff ks = 95/95 uncertainty in the method bias -

method ks = uncertainty in the reactivity equivalence methodology l re 3-11

Substituting calculated values in the order listed above, the result is:

K,ff = 0.8912 + 0.0083 + [(0.0058)2 + (0.0018)2 + (0.01)231/2 = 0.9112 The maximum K,ff for this configuration is less than 0.95, including all uncertainties at a 95/95 probability / confidence level. Therefore, the acceptance criteria for criticality are met for consolidation of spent fuel  !

235 in the CRSC at an equivalent " fresh fuel" enrichment of 3.0 w/o U ,

3.5 CONSOLIDATED FUEL STORAGE IN RACKS The Prairie Island spent fuel rack design described herein was analyzed for criticality to show that full:/ or partially loaded consolidated rod storage canisters (CRSC) can be stored in the fuel racks.

The spent fuel rack analysis is based on maintaining Keff 50.95 for storage of Westinghouse 14x14 0FA and STO fuel rods and EXXON 14x14 HI-PAR, LO-PAR and TOPROD fuel rods at 4.0 w/o with utilization of every cell permitted for  ;

storage of the CRSC.

3.5.1 Design Description The sp.ent fuel storage cell design is depicted schematically in Figure 3-3 and shown in detail in the design drawings in Reference 14. Nomini dimensions for the poison storage cell are shown on the figure. The CRSC design is depicted schematically in Figure 3-2 with nominal dimensions given on the figure. ,

3-12

)

3.5.2 Design Criteria Criticality of the CRSC 'in a fuel storage rack is prevented by the design of the rack which limits fuel interaction. This is done by fixing the minimum

-separation between storage locations.

.The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (K,ff) of the fuel array will be less than 0.95 as recommended in ANSI 57.2-1983, and in Reference 13.

3.5.3 Criticality Analysis - Spent Fuel Rack The following assumptions were used to. develop the nominal case KEN 0 model for the spent fuel rack storage of consolidated fuel rod canisters:

a) The. fuel rods'contain the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burnable poison in the fuel rods. Calculations have shown that the W 14x14 STD fuel-rods yield a larger K,ff than does the W 14x14 0FA or the EXXON HI-PAR, LOW-PAR, and TOPROD fuel pins when all fuel pins have the same U 235 enrichment. Thus, only the W 14x14 STD fuel pins were analyzed for storage. (See Table 3-2 for fuel parameters). _

3-13

7 .

~b) ' All fuel rods contain uranium dioxide at an enrichment of 4.0 w/o 235 U -over the' infinite length of each rod.

c)! No credit is taken for any U 234 or U 236 in the fuel,'nor is any-credit taken for the buildup df fission product poison material.

l:

d) .The moderator is pure water.at a temperature of 68 F. ' A 3

conservative value of 1.0 gm/cm is used for the density of water, e)' The array is infinite in the axial and radial extent which precludes any neutron leakage from the array.

f) The minimum poison material loading (i.e., 0.04 grams B-10 per 1 . ,

f <

square centimeter) is used throughout the array. '

A sensitivity analysis was performed to determine the minimum number of fuel

. rods that can be placed in the CRSC at a uniform pitch and meet'the spent

~

fuel rack K,ff limit of 0.95. Results of the study show that down to 113 fuel rods can be placed in each half of the CRSC for a total minimum number 4 of 226 fuel rods in a canister. Calculations have shown that the most reactive configuration is with fuel rods in both halves of the canister.

Therefore calculations were performed with fuel rods in both halves' of the CRSC.

-The KEN 0 calculation for the nominal case resulted in a K,ff of 0.8419 with a 95 percent probability /95 percent confidence level uncertainty of 1 0.0037.

3-14 1 I

_ _ _ _ _ _ _ _ _ - _. l

i

, , The maximum _k,ff.under normal conditions arises from. consideration of ]

mechanical and material thickness tolerances resulting from the manufacturing process in addition to asymmetric positioning of CRSC within the storage cells. The manufacturing tolerances are stacked in such a ,

. manner to minimize the water sap between adjacent cells, thereby causing an increase in reactiv-ity. The sheet metal tolerances are considered along with construction tolerances related to the cell I.D., and cell I l

E center-to-center spacing. For the spent fuel storage racks, the water gap  ;

1' is reduced from a nominal'value of 0.682" to a minimum of 0.492". Thus, the most conservative, or " worst case", KEN 0 model of the storage racks contains a minimum water gap of 0.492" with symmetrically.placed CRSC's. j Based on-the analysis described above, the .following equation is used to develop the maximum k,ff for the Prairie Island spent fuel storage racks with consolidated rod storage canisters: '

1

+

K,ff = Kwrst + Bmethod + Bpart s)2 worst + Gs)2 method Where:

K = w rst case KEN 0 K eff that includes material tolerances, worst and . a

. i mechanical tolerances which can result in spacings between l canisters less than nominal.

B = method bias determined from benchmark critical comparisons method 3-15 l

1

)

S

= bias to account for poison partical self-shielding part ks = 95/95 uncertainty in the worst case KENO Keff worst ks = 95/95 uncertainty in the method bias method Substituting calculated values in the order listed above, the result is:

K,ff = 0.9038 + 0.0083 + 0.0010 + [(0.0042)2 3+ (0.0018)2 1/2 = 0.917 Since K,ff is less than 0.95 including uncertainties at a 95/95 probability /

confidence. level, the acceptance criteria for criticality is met.

3.6 POSTULATED ACCIDENTS Accidente can be postulated which would increase reactivity through the uncontrolled spreading of the fuel rods as outlined in Rcftreate 1. This uncontrolled spread could occur as a result of spillage from a damaged or ,

I

. mishandled canister, or misloading of the FTC or the CRSC. These accident conditions are bounded by the uncontrolled release of two assemblies worth (358) of fuel rods. At no time are more rods than this being handled in the i

fuel consolidation area. The maximum K eff that can result from the uncontrolled release of 358 fuel rods at 4.0 w/o enrichment with no burnup in cold unborated water is 1.1508 with an uncertainty of + 0.0025. l For these accident conditions however, the double contingency principle of s

ANSI 8.17-1984 is applied. This states that one is not required to assume .

l 3-16

1 two unlikely,~ independent, concurrent events to ensure protection against a l criticality accident. Thus, for accident conditions,.the presence of soluble boron in the storage pool water can be assumed as a realist'.c initial condition since not assuming its presence would be a second unlikely event. .

l l

The presence of 1000 ppm boron in the pool water will decrease reactivity by approximately 30 percent delta K to 0.850. Thus, for postulated accidents, l l

should there be a reactivity increase K eff w uld be less than or equal to l 0.95 if the baron concentration in the pool water is greater than or equal to 1000 ppm.

Accidents can also be postulated which may damage fuel assemblies and result  !

in the release of fuel rods from the CRSC. Calculations have shown that any fuel rack geometry change that results in a decrease in the average fuel rod pitch from that of a normal fuel assembly or axillary misaligns a fuel assembly in the fuel racks will result in a decrease in the fuel rack k,ff.

Any increase in an assembly fuel rod pitch caused by a dropped CRSC will be bounded by the loss of containment of fuel rods in CRSC discussed above.

Thus, for these po;,tulated accidents, should ther, be a reactivity increase, k,ff would be less than 0.95 if the boron concentration in t'Te pool water is greater than 1000 ppm.

3.7 REFERENCES

'1. Proposed ANSI /ANS-57.10, " Design Criteria for Consolidation of LWR Spent Fuel."

3-17

.t F

j

2. - . W. E. Ford III,.et al';, "CSRL-V: Processed ENDF/B-V 227-Neutron-Group J and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies," ORNL/CSD/TM-160.(June 1982).

~ 3. N.-M. Greene, cat.al., "AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," 0RNL/TM-3706

.(March 1976).

4. L. M. Petrie and N. F. Cross, " KEN 0 IV--An Improved Monte Carlo Criticality Program," ORNL-4938 (November 1975).

5. M. N. Baldwin, et al., " Critical Experiments Supporting Cletr Proximity Water Storage of-Power Reactor Fuel," BAW-1484-7, (July 1979).

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.. A. J. Harris, et al'., "A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors," WCAP-10106, June, 1982.

8. Askew, J. R. , Fayers, F. J. , and Kemshell, P. B. , "A General 4'

Description of the Lattice Code WIMS," Journal of British Nuclear Energy Society, 5, pp. 564-584 (1966).

9. England, T. R., " CINDER - A One-Point Depletion and Fission Product Program," WAPD-TM-334, August 1962.

3-18

b' I

i J

- 10. Melehan, J. B., " Yankee Core Evaluation Program Final Report,"

WCAP-3017-6094, January,.1971.  ;

l J

j

11. Fuel Transition Canister Dwg. No.1873E90. i

!ji 1

12. Consolidated Rod Storage Canister Dwg. No. 1875E53, 1875E54, 1875E08.

13. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes, April 14, 1978, "0T Position for Review and ,

Acceptance of Spent Fuel Storage and Handling Applications."

14. Northern States Power Fuel Storage Dwg. No. NF-39213, NF-90046, NF-90051.

9 e

i 3-19

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. Table 3-2 Pbel Para:eters E: ployed in C:-iticality l

. . . ' . Analysis -.

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l Table 3-3 Wrison of HEDTIX Isc?qic Predicticn to Yankee Core 5 MeastLw.ts Quantity (Atom Ratio)  % Difference L U235/U -0.67 U236/U -0.28 U238/U -0.03 PU239/U +3.27 PU240/U +3.63

. PU241/U .-7.01 PU242/U -0.20 PU239/U238 +3.24 MASS (PU/U) +1.41 FISS-PU/ TOT-PU -0.02 Percent difference is average difference of ten comparisons for each isotope.

G

= I I

1

l ..

Table 3-4 Benct= ark Critical Dge_h PHOE'IX

& , arisen ,__

Description of: Number of PHOENIX k,ff Using Experiments -Experiments Exceriment Bucklines 00 2

Al clad -14 .9947 SS clad 19 .9944 Berated H 2O 7 .9940 Subtotal -

40 .9944 U-Metal

'Al clad 41 1.0012 .

TOTAL. 81 .9978 .

.l 1

.i l

l l

q

{,

i

1 l

i l

l j

s r

Table 3-5 Data for Metal arx1 U02 Critical WMs __

. 7uel ' Pellet Clad Clad Lattice .

Thickness Pitch 8-10

Case Cell. . A/O H20/U Density. Diameter Waterial 00 .

Number Type _U-235. Ratio (G/CC) ' (CM) Clad .(CM)_ . (CM) (CM) PPM

{

l Hexa ~ 1.328 3.02- 7.53 1.5265 Aluminum 1.6916 .07110 2,2050 0.0 f l'

2 Hexa 1.328' 3.95 7.53 1.5265 Aluminum 1.6916 .07110 2.3590 0.0  !

3 Hexa 1.328 4.95 7.53 1.5265 Aluminum 1.6916 .07110 . 2.5120 0.0 4 Hexa 1.328 3.92 7.52 .9855 Aluminum 1.1506 .07110 1.5580 0.0

, -5 ' Hexa 1.328- 4.89 7.52 .9855 Aluminum 1.1506 .07110 1.6520 0.0 6

Hexa 1.328 2.88 10.53 ,9728 Aluminum 1.1506 .07110 1.5580 0.0 7, Hexa 1.328 3.58 10.53 .9728 Aluminum 1.1506 .07110 1.6520 0.0 8- Hexa -1.328 4.83 10.53 .9728- Aluminum' 1.1506 .07110 1.8060 0.0  !

'l 9 Square.2.734. 2,18 10.18 .7620 SS-304 .8594 .04085 1.0287 0.0 10- Square 2.734 2.92 10.18 .7620 SS-304 .8594 .04085 1.1049 0.0  !

1 11 - Square 2.734 ' 3.86 10.18 .7620 SS-304 .8594 .04085 1.1938 0.0 12 . Square 2.734 7.02. 10.18 .7620 55-304 .8594 .04085 1.4554 0.0 '

13 Square 2.734 8.49 .10.18 .7620 55-304 .8594 .04085 1.5621 0.0 i 84 Square 2.734 10.38. 10.18 .7620 55-304 .8594 .04085 1.6891 0.0 15< Square 2.734 2.50 10.18 .7620 55-304 .8594 .04085 1.0617 0.0 16 Square 2.734 4.51- 10.18 .7620 55-304 .8594 .04085 1.2522 0.0  !

17 - Square 3.745- 2.50 10.27 .7544. SS-304 .8600 .04060 1.0617 0.0

- 18 Square 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 0.0 19 Square.3.745 4.51 .10.37 .7544 SS-304 .8600 .04060 1.2522 0.0

.I 20' Square 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 456.0-21 ~ Square 3.745 4.51 10.37 .7544 S5-304 .8600 .04060 1.2522 709.0 22 Square 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 1260.0 i 23- Square 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 1334.0 24 - Square 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 1477.0 25 ,

Square 4.069 .2.55 ~ 9.46 1.1278 55-304 1.2090' .04060 1.5113 0.0 26 Square 4.069 2.55 9.46 1.1278 SS-304 1.2090 .04060 1.5113 3392.0

. 27- Souare 4.069 2.14 9.46 1.1278 55-304 1.2090 .04060 1.4500 0.0 28 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 0.0 29 Square 3.037 2.64 9.28 1.1268 $5-304 1.1701 .07153 1.5550 0.0 30 Square 3.037 8.16 9.28 1.1268 55-304 1.2701 .07163 2.1980 0.0 31 Square 4.069 2.59 9.45 1.1268 -SS-304 1.2701- .07163 1.5550 0.0

'32 Square 4.069 3.53 9.45 1.1268 55-304 1.2701 .07153 -1.6840 0.0 33'- . Square 4.069 8.02 9.45 1.1268 S5-304 1.2701 .07163 2.1980 0.0 34: - Square 4.069 9.90 9.45 1.1268 SS-304 -1.2701 .07163 2.3810 0.0

35. Square 2.490- 2.84 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 1677.0 36- Hexa 2.096 2.06 10.38 1.5240 Aluminum 1.6916 .07112 2.1737 0.0 37- Hexa L2.096- 3.09 10.38 1.5240 Aluminum 1.6916 .07112 2.4052 0.0

'38 Hexa 2.096 4.12 10.38 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 39 - Hexa 2.096 6.14 10.38 1.5240 Aluminum 1.6916 .07112 2.9891 0.0

'40 - Hexa LO96 8.20 10.38 1.5240 Aluminum 1.6916 .07112 3.3255 0.0 48' ' Hexa 1.307- 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 42 Hexa 1.307 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0 43 Hexa ^1.307 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 9

t Table 3-5 Data for Metal and ID2 Critical Experiments (Omt) i Fuel Pellet Clad Clad ' Lattice Casa Cell A/O H20/U Density Diameter Waterial 00 Thickness Pitch 8-10 Number Type U-235 Ratio (G/CC) (CM) Clad (CM) (CM) (CM) PPM e

144 Hexa 1.307 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 45 Hexa 1.307 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 46 Hexa 1.'160 1.01 18.90 1.5240. Aluminum 1.6916 .07112 2.1742 0.0 47 Hexa 1.160 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0 48 Hexa 1.160 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 49 Hexa 1.160 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9E96 0.0 i 50 Hexa 1.160 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 l 1.01 18.90 51 Hexa 1.040 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 52 Hexa 1.040 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0 53 Hexa 1.040 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 54 -Hexa. 1.040 . 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 55 Hexa 1.040 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0

. 56 Hexa 1.307 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 57 Hexa 1.307 1.52 18.90 .9830 Al uminum 1.1506 .07112 1.5926 0.0 58 Hexa 1.307 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 59 Hexa 1.307 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 60 Hexa 1.307 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 61 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 62 Hexa 1.160 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 ' O.0 63 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 ,

64 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 65 Hexa 1.160 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 66 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 67 Hexa 1.160 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 '

68 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 69 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 70 Hexa 1.040 1.33 18.90 15.050 Aluminum 2.0574 .07620 2.8687 0.0 71 Hexa 1.040 1.58 18.90 19.050 Aluminum 2.0574 .07620 3.0086 0.0 72 Hexa 1.040 1.83 18.90 19.050 Aluminum 2.0574 .07620 3.1425 0.0 73 Hexa 1.040 2.33' 18.90 19.050 Aluminum 2.0574 .07620 3.3942 0.0 74 Hexa 1.040 2.83 18.90 19.050 Aluminum 2.0574 .07620 3.6284 0.0 75- Hexa 1,040 3.83 18.90 19.050 Aluminum - 2.0574 .07620 4.0566 0.0 76 Hexa 1.310 2.02 18.88 1.5240 Aluminum 1.6916 .07112 2.6160 0.0 77 Hexa 1.310 3.01 18.88 1.5240 Aluminum 1.6916 .07112 2.9900 0.0 78 Hexa 1.159 2.02 18.88 1.5240 Aluminum 1.6916 .07112 2.6160 0.0 '

79 Hexa 1.159 3.01 18.88 1.5240 Aluminum 1.6916 07112 2.9900 0.0 80 Hexa 1.312 2.03 18.88 .9830 Aluminum 1.1506 .07112 1.7250 0.0

! 81. Hexa 1.312 3.02 18.B8 .9830 Aluminum 1.1506 .07112 1.9610 0.0 i' - _ -

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Figure 3-1 Minir.un Barnup vs. Initial U-235 Enric.ht For Fuel Assembly Consolidation Operations

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Figure 3-2 Consolidated Red Storage Canister Ncrninal Dimensions I

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.4.0:L THERMAL / HYDRAULIC ANALYSIS OF CONSOLIDATED FUEL STORAGE CANISTER i

4.1

SUMMARY

The fuel consolidation canister (CRSC) provides a means for long-term-storage'of compacted spent fur.i rods in a spent fuel pool. Because the i canister / fuel, arrangement has a much higher fuel-to-water ratio than that of a typical fuel assembly, specific areas for evaluation are: a)the' time

. after shutdown which is: required before natural circulation.can provide adequate cooling without boiling (Ref. 4-1),'and b) ensuring that the canister. thermal / hydraulic design efficiently minimizes this time, t

The time after shutdown which satisfies the no-boiling requirement is a  !

function not only:of the canister-fuel hydraulics, but also of the power operating' history of the fuel. ; Consequently, analyses were performed for a q

range'of effective full power operating histories, and also for peaking

~

factors which conservatively place an upper bound on the amount of power a given fuel rod can generate, compared to the average.

Ensuring an efficient thermal / hydraulic design of the canister reduces to a) i

, verifying that the hole pattern in the canister letdown plate maintains a low hydraulic loss and provides adequate cooling to all fuel rods, and b) verifying that the' offset distance between the letdown plate and the bottom plate allows uniformity of the flow into the canister. The hole pattern, which. keeps the. letdown plate hydraulic losses low relative to the compacted fuel rods while still maintaining structural integrity, was found to be l

1 4_1

7. _ _ _ _ - _ - - _ _ _ _ - - - - - - - - - - ----- a

j b

adequate.- Similarly, the offset distance was found to be satisfactory to .

. permit flow uni.formity without necessitating an increase in CRSC design I

length.

1 l

4.2 THERMAL'-HYDRAULIC:ANALY$IS I

1 I

LThe mechanism for cooling the fuel rods in the canister is natural  !

circulation. The decay heat.of the fuel rods heats the fluid, creating buoyancy forces which. produce flow through the canister. The higher the decay heat level, the greater the flow. However, the canister fluid exit tempera-ture increases with increasing decay heat level faster than does the~

i flow rate. Consequently, the decay' heat level must be at or below a certain limit before boiling can be prevented. This limit is inversely proportional to the total' hydraulic losses in the CRSC.

Figure 4-1'shows the basic canister geometry. The flow enters the bottom of the rack cell, passes across the cell's support cross bars, into the CRSC through the bottom plate, through the letdown plate and into the passages between fuel rods. After being heated, the flow exits these passages, passes through the top plates and discharges into the ' spent fuel pool. The limiting region, from a thermal-hydraulic point of view, is a single passageway in the center of the assembly formed by three adjacent rods (see Figure 4-2). The reason for this is that in the center of the assembly, little energy is lost by radial heat transfer to the pool; the fluid absorbs most of it. Assuming incompressible flow, a momentum balance of the

. buoyancy driving forces and the hydraulic losses gives:

. . . 4~2

~

2 Kw

= pb[Z (T - T ) + 0.5 Z (T -T )]

2 p p a f p a 2pgA

= pb[(Z +0.5Z)(T-T))

.p f p a (4-1) where w = flow rate in the passageway A = flow area of.the passageway T

p

= water temperature exiting the canister T, = temperature in pool Z

p

= height of top of canister above fuel rods Z

f

= fuel rod length K = overall hydraulic loss factor (referenced to passageway area) p = fluid density b = -(1/p) dp/dT The quantity K can be expressed as:

Z f

K = (f-) + IR D fuel (4-2) passageway where l

l 1

4-3

IR = irreversible losses (expansions and contractions of the bottom, top, and letdown plates) f = passageway friction factor D = passageway hydraulic diameter The reason for expressing equation (4-2) in the manner indicated is because the (f Z f/D) portion is much larger than the IR portion. To simplify the calculations, the IR portion is initially neglected and a correction made afterwards. This will be discussed later.

The friction factor f is calculated from the following formula:

f = N/(Re)" (4-3) where Re = Reynolds number and N = 96 for laminar flow (per Rcf. 4-3) n = 1.0 N = 0.316 for turbulent flow n = 0.25 1

In Figure 4-2 it can be seen that the passageway receives the energy input of three, sixty degree, pie-segments of each adjacent rod, or the equivalent of one-half a rod. If Q is the power output of one rod, an energy balance yields Q/2 = w Cp (T p - T,) (4-4) 4-4

I f

- where l C := fluid specific heat' p

Equations (4-1),(4-2)'(neglectingIR),(4-3)and'(4-4)'canbecombinedto y-b' yield thecallowable power per rod as a function of the exit temperature Tp :

Q=2 PAC p [h8(o 2) (Tp - T,) h 3 k (4-5) 1 where:

v = . flu 13 kinematic viscosity g =. acceleration of gravity

-' Equation (4-5) was used to calculate Q-values assuming laminar and turbulent flow. Whichever of these was lowest was taken to be the limiting value.

Corrections for the irreversible losses IR were made by using equation (4-4) to calculate w and thus the Reynolds number and velocity after the first iteration. From this, equation (4-2) was used to calculate a new equivalent friction factor which accounted for tl4 JR contribution. This equivalent friction-factor, manifested as an equivalent "N" value, was used in equation (4-5) to determine a corrected "Q" value. However, friction losses clearly dominate the total loss through the CRSC, and this correction was -

insignificant. Consequently, only one. iteration on aquation (4-5) was necessary. For the calculations, Ta was assumed to be 150 deg F, and Tp was 4-5

. taken to be the boiling point of water at an elevation corresponding to the top of the CRSC at rest in its storage position in a rack cell. This elevation is approximately 26 ft below the spent fuel pool surface, and the corresponding Tp is approximately 241.2 deg F. Note that the results of this analysis, illustrated in Figures 4-4 through 4-6, also apply to the CRSC transfer process where the CRSC actually would have as little as 10 ft of water cover, but where the spent fuel pool temperature is limited to 120 deg F during CRSC transfer from the consolidation area to the rack cells.

i 4.3 LETDOWN PLATE / BOTTOM PLATE / CANISTER HYORAULICS  ;

The hole pattern shown in Figure 4-3 was determined to be suitable for providing flow uniformly to all coolant passageways between fuel rods without compromising structural integrity. Each hole in Figure 4-3 feeds into two coolant passageways. This allows the number of holes to be minimized and maximizes the ligament dimension between holes, thus enhancing structural integrity, and manufacturability. Also, a smaller number of holes than would be required if each passageway were fed individually means that the hole diameter is larger for a given letdown plate flow area. This, in turn, diminishes the possibility of the holes being completely plugged by crud or other obstructions.

It was determined that 356 unchambered holes, 0.125 inch in diameter are adequate to permit cooling flow to pass through the letdown plate without incurring significant hydraulic losses. This permits letdown plate _

thick-nesses on the order of 0.5 inch to 0.75 inch. Figure 4-3 depicts the hole pattern for one of the two le.down plates that comprise a single 4-6

1 I

canister assembly. A canister assembly is divided in half by a divider plate (Figure 4-3). The letdown plate in each half therefore contains 178 holes.

Unlike the letdown plate, the flow area in each half of the bottom plate consists of two 3.1 inch diameter holes. If the bottom plate and letdown plate are too close together, the flow passing through the large bottom plate holes will not be distributed adequately to those letdown plate holes on the outer periphery of the canist er, and near the divider plate. It was found that the bottom plate to letdown plate offset distance of 0.4 inches was large enough and that the hydraulic loss incurred as the flow turns from the bottom plate to the peripheral letdown plate holes, was at least an order of magnitude less than the hydraulic loss of the holes themselves.

4.4 HEATING RATES The method used to calculate decay heat levels is described in (Ref. 4-2).

The analysis conservatively assumed the maximum fuel pool temperature of 150 F as the water temperature at the canister inlet, a reactor design power of 1650 MWt as applied to the fuel burnup histories, and a peaking factor of ,

1.8 for all rods. That is, the power generated at any given time, for any <

burnup history, was assumed to be 1.8 times the value calculated using the i approved methodology in (Ref. 4-2). This accounted for radial and axial core power distributions pertinent to actual fuel burnup histories.

1 Figures 4-4 and 4-5 show the decay heat rate per rod as a function of time after shutdown for various core operating times prior to shutdown. Core 4-7

_. _ _ _ _ _ _ _____-________-._.____________-_ m

1 operating times between 5000 hrs. and 60,000 hrs. are considered. The 1.8 peaking factor .is included in all of the curves shown in Figures 4-4 and 4-5.

4.5 RESULTS AND DISCUSSION -

The limiting power per rod was calculated for laminar and turbulent flows.

The laminar flow power _ limit of.0.0225 (Btu /sec)/ rod was lower and was therefore used as the peak rod power allowed to preclude boiling within a CRSC. Note that this limiting power per rod applies to the following two scenarios:

1. CRSC in place in a rack cell where the top of the CRSC is approximately 26 ft below the spent fuel pool surface, and where the maximum pool temperature is 150 deg F.

2. CRSC is in transit where the top of the CRSC comes up to 10 ft below the spent fuel pool surface, and where the maximum pool temperature during transit is 120 deg F.

In effect, maintaining these pool temperature limits during those respective scenarios enables determination bf how much time must elapse after shutdown before consolidation can be performed. From another perspective, this approach addresses both storage (long term CRSC position) and transit (short term CRSC position).

4-8

, To determine an allowable time after shutdown beyond which the power level per rod would be below .0225 Btu /sec., Figures 4-4 and 4-5 were used. From the. curve for each effective full power operating time, the time at which

, the power level.per rod was below 0225 Btu /sec. was determined. These values are plotted in Figure'4-6 as allowable time after shutdown as a function of the effective full power hours of core operation. This figuie '

L will be used as a guide for the minimum times after shutdown required before

< fuel consolidation is initiated.

4.6 CONCLUSION

S A thermal-hydraulic evaluation of the NSP spent fuel consolidation canister was performed resulting in the following conclusions:

1. A letdown plate hole pattern design consisting of 178 holes (per plate), .0.125 inch in diameter, is acceptable. This design ensures flow uniformity and maintains low hydraulic losses in the letdown plate.

2. The offset distance between the bottom plate and the letdown plate of 0.4 ' inch avoids significant turning losses between the bottom plate and letdown plate.

3. Based on maximum spent fuel pool temperatures of 150 deg F (CRSC in storage in rack cell) and 120 deg F (CRSC in transit through pool), and the corresponding acceptable (no boiling) rod power level (0.0225 (Btu /sec)/ rod), it was determined that between 400 and 1000 days must 4-9

l elapse after shutdown prior to fuel consolidation. This range of times 1

after shutdown correspond to a range of effective core full power operating hours of 5000 hours0.0579 days <br />1.389 hours <br />0.00827 weeks <br />0.0019 months <br /> to 60000 hours.

4.7 REFERENCES

4-1. ANS-57.10, " Design Criteria for Consolidation of LWR Spent Fuel,"

Section 6.11.1, October, 1986.

4-2. NUREG-0800, Branch Technical Position ASB-9-2, Rev. 2, July 1981.

4-3. Heat and Mass Transfer, E. R. Eckert, R. M. Drake, 2nd Edition, copyright 1959, Maple Press Company, York, Pa. (p. 159).

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Pools 1 and 2 contain fuel accumulated'according to the normal discharge schedule, and utilizing consolidation to the maximum extent. .1377 storage cells are filled: 240 intact assemblies from the last 4 discharges and.1 assembly discharged more than 15 years earlier; and 1136 corisolidated fuel storage canisters holding fuel from 2272 assemblies. Both the.following cooling' system modes were analyzed.

a. Main heat' exchanger and one pump

b. Backup heat exchanger and two pumps i

- 2. Abnormal-Same fuel storage condition as above, plus a freshly off-loaded core . consisting of 121 assemblies. The pools are cooled by both the main and backup heat exchanger and both_ pumps.

3. Faulted Same as abnormal, but either 4

a. Main heat exchanger and one pump, or

b. Backup heat exchanger and two pumps in service i

' Table 5-1 presents a summary of the results. For normal operation either condition a. or b. can be used to keep the pool exit temperature below 5-4 150 F. For the abnormal condition, with both heat exchangers and pumps in ) operation, the maximum pool temperature can also be maintained well below 150 F. For the faulted condition, with the backup heat exchanger and two pumps in operation only, a maximum temperature of 183.79 F is reached. Without corrective action, the temperature will remain above 150 F for this 1 condition for 616 more hours, or 25.7 days. i 5.4 TOTAL LOSS OF FORCED COOLING l The worst time a complete loss of cooling could occur is at the time of maximum heat load for the cases analyzed with forced cooling. Unig identical criteria for the normal, abnormal and faulted conditions as used with forced cooling, the results of Table 5-2 were generated using this assumption. It is noticed that the faulted condition represents the limiting condition, in that Tsat will be reached in /3 hours following the loss of all forced cooling. Even in this case, however, the top of the racks will not be uncovered (without corrective action) for /66 hours after the cooling loss. In the unlikely event that all spent fuel pool cooling is lost and boiling occurs, six sources of makeup water are available. Ten minutes or less is required to line up the valves or to carry out the steps necessary in order to make the water available. The six sources of makeup water and their makeup rates are as follows: (a) Chemical and Vol'me u Control System - 300 gpm, (b) Chemical and Volume Control System Blender -100 gpm, (c) Refueling Water Storage Tank - 80 gpm, (d) Reactor Makeup 5-5 l Storage Tanks - 80 pm, (e) four demineralized water hose stations, each station rated at 20 gpm, and (f) the fire protection system - there are two i fire hose stations near the spent fuel pool each rated at 95 gpe.

5.5 REFERENCES

'1. U. S. Nuclear Regulatory Commission (USNRC) Branch Technical Position APCSB 9.2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling.

2. " TRAM - Thermal-Hydraulic Rack Analysis Model; Description and Verification, by J. C. Buker, WNEP-8530, May 1985.

3. " TRAM-T - Thermal-Hydraulic Rack Analysis Model - Transient Version; Description and Verification, by J. C. Buker, WNEP-8563, 12/4/85.

1

~l 5-6 8

l' s.

I Table 5-1 Stu: nary of Pool Cbolirq System Evaluation .

Tczad Cooling

1) Normal 6

Maximum heat load at 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> after shutdown = 15.79 10 BTU /Hr.

Condition Maximum F_221 Elil Temo. Ilma After Shutdown

a. 142.350 F 130 Hrs

b. 144.370 F 132 Hrs

~

2) Abnormal 6

Maximum heat load at 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown = 28.25 10 BTU /Hr Condition Ma ximum f.221 E_x.il x Temo. Ilmt After Shutdown Main + backup heat 139.400 F 128 Hrs exchang' ers + both pumps in service

3) Fauited 6

Maximum heat load at 120 hou'rs after shutdown = 28.25 10 BTU /Hr.

Condition Ma ximum P_ cal Exil Temo. Ilmi After Shutdown l

a. 180.160 F 137 Hrs j

, b. 183.79 F 138 Hrs' l 1

._. _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . ___ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . . . _ _ . _ _______.9

4 Table 5-2 Sumary of Pool Cbolirg Syste Evaluaticn Icss of FcIced Coolirg

1) Norms 1 .

Forced cooling lost at !!O hours after shutdown with condition b.

Average Surface Ig Resehed in Evaporation E111 R acks 122 Resehed in i

11 Hours 14200 lbm/Hr 131 Hours I

2) Abnormal Forced cooling lost at 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown.

Average Surface Ig Reached in Evaporation Ella Racks 122 Resched in 6 Hours 27000 lbm/Hr 70 Hours

3) Faulted l

Forced cooling lost at 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown with condition b. I

. Average Surface Ig. Reached in Evaporation Eals Racks ing Resched in

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. 3 Hours 27000 lbm/Hr 66 Hours i- -

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kOLS HhA LOAD - 106'g[U/ d Figure 5-1 Approximate FI spent Fuel Pools cooling system capability steady _ state Evaluation )

, , 6.0 RADIOLOGICAL ANALYSIS OF HANDLING ACCIDENTS Radiological analyses were performed to determine the thyroid and whole body doses at the exclusion area boundary resulting from postulated drop accidents involving a consolidated rod storage canister (CRSC) and comparing th,e results with previously analyzed accident conditions. The following two accidents were analyzed:

.1. Canister drop on floor

2. Canister drop onto storage racks The postulated accidents were analyzed only for the fuel handling building (not containment) since that is where fuel consolidation activities take place.

6.1 CANISTER DROP ON FLOOR The assumptions used in this analysis were as follows:

1. The canister contains 358 fuel rods (from two assemblies of 179 rods each) and all fuel rods suffer clad damage in the drop.

2. For the postulated accident there would be a sudden release of the gaseous fission products held in the void space between the pellets and

~

the cladding.

6-1

l

]

]t '

. . 3. The activity inL the void. space of the. fuel rods in-the canister is

' based onf the core activities given in Table D.1-1 of the Prairie Island .

) 4 Updated Safety Analysis. Report.(USAR). In addition, the assumptions'in US.NRC Regulatory Guide 1.25 (i.e., 30% Kr-85, 10% other noble gases and ' halogens in the gap; and a radial peaking factor of 1.65) are used, i

4. ~~~4 3

.The. accident x/Q value of 6.5 x l0 sec/m is used. This is the same-x/Q_value which'is used in the fuel handling accident described in the I

USAR. j i

4

5. .A decay period of 2 years is used.  !

6. . Dose conversion factors from Regulatory Guide.l.109 are used (Table B-1 I for noble gases).

7. Retention of noble gases in the fuel pool is negligible (i.e., DF-1).

3 Under these assumptions, an activity release of 2.70 x 10 curies of

-Krypton-85 occurs (all other noble gases and halogens decay to negligible levels in 2 years). This activity releases results in the following doses at the' exclusion area boundary:

Thyroid Dose = 0 ,

Beta Skin Dose = 74.6 Millirem

~

. Gamma' Body Dose = 0.9 Millirem

-I s .

6 _ - - - - - _ - -

< ~

6.2 CANISTER DROP ONTO STORAGE RACKS In addition to the assumptions above (1 through 7) for the " canister drop on floor" accident, the following are included:  !

8. In addition to the damage to all of the fuel rods in the canister, the dropped load causes damage to all of the rods in a freshly stored fuel i i

assembly (i.e., 179 rods).

i

9. The fuel assembly damage is also assumed to be the highest rated fuel assembly-(see Assumption 3) and damage occurs at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after i

reactor shutdown.

10. Additional assumptions are taken from Regulatory Guide 1.25 for thyroid does determination. These are:

6 I-131 Dose Conversion Factor = 1.48 x 10 REM / Curie 5

I-133 Dose Conversion Factor = 4.0 x 10 REM / Curie ,

Standard Breathing Rate = 3.47 x 10 -4 3 m /sec Overall Effective Pool Decontamination Factor (DF) for Iodine = l 100 -

1

(

l

11. A charcoal filter efficiency for halogen removal of 95 percent is used

' 1 i

(USAR Section 14.5.1.3).

l 1

i i

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i

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. . With these additional assumptions included, the significant activities released from the fuel building are as follows:

3 3 Kv-85 = 4.23 x 10 Curies (2.70 x 103 + 1.53 x 10 )

Xe-133m = 8'.7 x 102 Curies Xe-133 = 7.32 x 104 Curies I-131 = 20.2 Curies I-133 = 2.36 Curies ,

These released activities result in the following exclusion area boundary doses:

Thyroid Dose = 6.96 REM Beta Skin Oose = 0.60 REM Gamma Body Dose = 0.45 REM 6.3

SUMMARY

OF RESULTS The results of these two radiological analyses are summarized in the following table and compared with the USAR fuel handling accident analysis.

\*

6-4

.I . -

q. , RADIOLOGICAL ANALYSIS COMPARISON w,

Whole Body Thyroid Line Analysis Dose (REM) Dose (REM) 1 1 1 assembly'(USAR) 1.0 0.58

. 7 '2- 1 assembly (USAR S.G.25)- 1.8 14 3 '1. assembly (USAR updated 2.6 13.8 evaluation) l i

4 358 rods (2 yr. decay) 0.076 0 l 5 358 + 179 rods 1.05 6.96 (358 with 2 yr decay)

(179 with 100-hr decay) l 6 10 CFR 100 dose guidelines 25 300  :

~

7 NUREG 0800 dose limits 6.25 -75 (25% of 10 CFR 100) a

'A.s can be seen in the comparison table, the whole body' dose and thyroid ddse

.for both analyses involving consolidated fuel are enveloped by all the Prairie Island updated safety analysis report fuel handling analyses except

'for the initial analysis reported in the USAR, the results of which are given on. Line 1. This USAR analysis, involving a single assembly, although

somewhat' conservative did not include all of the conservatism that were

, included in the other analyses (Lines'2, 3, 4 & 5). However, as can be 1 seen, all of the analyses are well within the 10 CFR 100 guidelines and the NUREG 0800 dose limits.

6.4 0THER POTENTIAL ACCIDENTS l Other potential accidents which may occur during the consolidation process "are as-follows: '

1. Rupturing of the' cladding of one or more fuel rods.

2. Jamming of one or more fuel rods in the. consolidation equipment.

3. Dropping of fuel rods.

1

4. Dropping of equipment.

f Although equipment design and personnel training are aimed at prevention of L accidents, one or more of these potential accidents may happen over the course of consolidating 1,000 fuel assemblies. A discussion of each of these accidents follows:

1. Rupturing of Fuel Rod Cladding Rupture can occur due to rod handling because the rods have to b'e deflected in order to consolidate them. Deflection would occur on straight rods as they are moved from the fuel assembly to the 6-6

, transition canister and as they'are moved from the transition car.ister to the storage can. Another cause of rod deflection will be the removal of bowed rods from the fuel assembly. The likelihood of rod rupture due to deflection is low based on data from the Westinghouse Nuclear Fuel Division which shows that irradiated fuel rods can be severely deflected without causing rupture. Radiation effects from such an accident would be enveloped by those of the accidents discussed in 6.1 and 6.2.

2. Jamming of Rods in Equipment Debris can cause jamming of the rods in the equipment. Examples are jamming of a fuel rod as it is being pushed into the transition canister, or jamming of a rod as it is being removed from this canister. Special techniques are used to disengage the rods.

The techniques depend upon the elevation of the rods at the point of jamming, and upon the water depth over the jammed rod and the other rods in the equipment. Jamming would not create a radiological concern.

3. Dropping of Fuel Rods The proposed consolidation process minimizes the possibility of' .

dropping individual rods because the rods are pushed down instead I i

6-7

1

. of lifted. However, a pushed rod could free-fall in the equipment if drag devices malfunction.

l l'

The most severe case of rod drop is dropping of a full storage can with 358 rods (the rods from two fuel assemblies). To preclude such a drop, the lifting equipment is being designed in accordance with ANSI N14.6-1978, "American National Standard ~for Special Lifting Devices for Shipping Containers 10,000 Pounds (4500 Kg) or More for Nuclear Materials", even though a full storage can will weigh less (2500 pounds). Should such a drop occur, the radiological effects are enveloped by the results of the accident analyses discussed in 6.1 and 6.2. The criticality effects are discussed in 3.6.

4. Dropping of Equipment The consolidation equipment could be accidentally dropped during setup, use, and teardown. However, the equipment will be moved and located where, if it is dropped, it will not fall on stored spent fuel. Therefore, dropping of the equipment should not damage fuel other than the fuel being consolidated, nor should it compromise any safety related system-6.5 Conclusions The analyses presented in this section show that the spent fuel consolidation which will take place at Prairie Island does not create a 6-8 I

-.-__.__._w

i possibility for an accident or malfunction of a different type than l evaluated previously. Nor does consolidation increase the probability of l 1

occurrence of a previously analyzed accident. This analysis assumes a probability of 1, as does the previous analysis.

Finally, consolidation does not increa>e the consequences of a previously analyzed accident. In fact, the radiological consequences cf dropping a canister holding two assemblies with of spent fuel with a minimum of two years of cooling / decay time, are significantly less than the consequences of dropping a single assembly with only 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of cooling / decay time.  !

9 l

l 1

1 6-9 l l

_ _ _ _ _ _ _ _ _ _ _ _ _ __ }

i 7.0 ME.CHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS i

r i The use of previously tried designs was a mechanical design consideration for the fuel consolidation equipment to be used at Prairie Island. Some of the consolidation equipment is based, in part, on equipment that has been used before by Westinghouse. The elevator / rotator in the fuel assembly dismantling station is a major part of the Multi-Function Repair System that is used to repair fuel assemblies. Also utilized was Westinghouse's experience and lessons learned from consolidating four fuel assemblies, in 1982, at Duk.e Power's Oconee plant. The storage canisters have metal thicknesses which were demonstrated to be suitable when used for the storage canisters in the consolidation of Oconee fuel. A major component in the consolidation equipment is the trarsition canister. Its design is based on the transition. canister developed for consolidating Oconee fuel. The long handled tools for moving equipment underwater have been used extensively.

Underwater television systems are also used frequentiy to repair fuel assemblies.

Another mechanical design consideration was the application of ANS-57.10, I l

" Design Criteria for Consolidation of LWR Spent Fuel", which states that the 1

equipment is to be designed in accordance with commercial codes and )

i standards. The codes and standards provide mechanical design guidance. J I

. The majority of the materials used below water are made of stainless stee'l and aluminum. Small amounts of bronze and brass are used in threads and l

7-1 l

.__________________a

l l-

. . bearings to prevent galling. Above the surface, on the storage can loading frame crane, for example, structural steels are used.

l The materials were also selected in accordance with ANS-57.10 which, in turn, specifies materials per the A;ME Bailer and Pressure Vessel Code, l Section 111, Division 1, Subsection NF, or AISC-5326 1978 Specification for Design, Fabrication and Erection of Structural Steel for Buildings.

Structural design is based upon the codes and standards referenced in >

ANS-57.10. Allowable stresses and strengths of section members or i connections are as given in ANSI /AISC N690, " Specifications for Design, l

Fabrication, and Erection of Safety related Structural Steel in Nuclear Service", 1984.

The heaviest load to be lifted, in the vicinity of the fuel racks, will be the loaded consolidated rod storage canister (2500 pounds). Since a load greater than the weight of one fuel assembly (1200 lb) is considered to be a

" heavy load", NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants",

applies to the storage canister handling tool. NUREG-0612 states that special lifting devices should satisfy the guidelines of ANSI N14.6-1978,

" Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More for Nuclear Materials". Therefore, the guidelines in ANSI N14.6-1978 will be used to design and test the storage can lifting tool. j

.The consolidation equipment will not be moved over stored spent fuel. It will be set up in the fuel transfer canal, which is separated by a wall from l

7-2

the pool area. Therefore, accidental dropping of the equipment should not compromise the operation of safety and cooling systems, or damage stored fuel.

l l

l 4

l l .

i 7-3

)

8.0 CONSOLIDATION DEMONSTRATION EXPERIENCE AT PRAIRIE ISLAND (to be supplied later)

[

e e

8-1

ble 9:- CONCLUSIONS preparation for conducting a demonstration program of spent fuel thin bsolidation, Northern States Power Company has thoroughly examined all tects of the process to provide assurance of its feasibility, reliability Qsafety. Analyses were conducted to verify that the criticality and fall mmal-hydraulic design bases of the Prairie Island' Nuclear Generating int's spent fuel storage pool were not exceeded and that the radiological 7

ects, resulting from postulated accidents occurring during the

)solidation demonstrate.on, are bounded by those of accidents already  !

3tulated in the Prairie Island Updated Safety Analysis Report.

31tionally, the compatibility' of the materials of construction of the

)solidation equipment with spent fuel pool coolant was verified. The hability of accidents occurring as a result of the fuel consolidation is

)imized by testing and demonstration of the equipment under simulated an

) solid: tion procedures and through the training of personnel in these xsedures prior to actual application.

LIMITING CONDITIONS FOR CONSOLIDATION tanalyses presented in this report have identified certain constraints on t fuel which can be consolidated, and on the pool conditions during colidation activities. These are referred to here as Limiting Conditions

~

Consolidation. i n

.1 Fuel Assembly Characteristics .o 9-1

t

, the consolidation demonstration program defined in .Section 1.1 of this report:

1. It does notfereate a possibility for an accident or malfunction of a different' type than evaluated previously in the USAR or subsequent commitments 41 The . types of accidents which may result.during Jaent fuel consolidation

.are ' discussed in Section 3 and 5 of this report,: and as can be seen

_ from these discussions, their types.are similar.to those.previously evaluated. Equipment malfunctions, such as consolidation equipment falling while in the process of erection, are found to be of no consequence due to the safe location in which the consolidation process isLto be performed.

' 2. It does not increase the probability of occurrence of an accident or malfunction of equipment important to safety previously analyzed in the USAR or subsequent commitments.

The accident analyses 1.9 Section 3 and 6 assume a probability of 100%.

.For example, in Section 6, the assumptions made are that a loaded canister does drop onto a freshly discharged fuel assembly, and that l all fuel pins' involved do break. The USAR fuel handling accident evaluation also assumes a probability of 1.

9-3

_..__m______. _______. -_.--__-----------

g______---- , _-

3. It does not increase the consequences of any accident or malfunction of equipment important to safety previously analyzed in the USAR or subsequent commitments.

The consequences of accidents occurring as a result of consolidation are shown in Section 6 to be bounded by those of accidents previously evaluated in the USAR.

4. It does not reduce the margin of safety defined in the bases for any t Technical Specification.

The margin of safety as defined in the basis for any Technical Specification is not reduced as a result of implementing spent fuel consolidation.

Therefore, based on the foregoing, it can be concluded that conducting a demonstration of spent fuel consolidation by transferring the fuel pins from fifty spent fuel assemblies to twenty-five consolidated rod storage canisters does not involve an unreviewed safety question.

4

. 1 l

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