ML20235G601

From kanterella
Revision as of 17:29, 20 March 2021 by StriderTol (talk | contribs) (StriderTol Bot change)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Fuel Consolidation Demonstration at Prairie Island Nuclear Generating Station. Westinghouse 870903 Memo Re Fuel Rod Clad Surface Temp in Util Fuel Canisters & Nutech to L Mccarten at Util Encl
ML20235G601
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 09/30/1987
From:
NORTHERN STATES POWER CO.
To:
Shared Package
ML20235G494 List:
References
NUDOCS 8709300119
Download: ML20235G601 (89)


Text

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

1

' TOPICAL REPORT l

I FUEL CONSOLIDATION DEMONSTRATION AT PRAIRIE ISLAND NUCLEAR GENERATING STATION

\

Northern States Power Company ser e ler, 1987 1

d e

l 9709300119 070922 2 ppR ADOCK 0500 -

P

I TABLE OF CONTENTS l

. 1 1.0 Introduction

]

1 1.1 Consolidation of Spent Fuel at Prairie Island l 1.2 Purpose of Report '

2.0 Consolidation Process 2.1 Equipment Description 2.2 Process Description 3.0 Criticality Analysis - 2:1 Consolidation 3.1 Introduction 3.2 Acceptance Criterion for Criticality 3.3 Criticality Analytical Method 3.4 Consolidation Operations 3.4.1 Design Description 3.4.2 Design Criteria 3.4.3 Criticality Analysis ,,

3.4.3.1 Fuel Transition Canister 3.4.3.2 Consolidated Rod Storage Canister and FA 3.4.3.2.1 Reactivity Equivalency & Methods 3.4.3.2.2 Reactivity Calculations 3.5 Consolidated Fuel Storage in Racks l

3.5.1 Design Description 3.5.2 Design Criteria 3.5.3 Criticality Analysis - Spent Fuel Racks 3.6 Postulated Accidents I

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 4.7 References i

t - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

5.0 Radiological Analysis of Handling Accidents 5.1 Canister Drop on Floor -

5.2 Other Potential Accidents 6.0 Mechanical, Material, and Structural Considerations '

7.0 Consolidation Demonstration Experience of Prairie Island 8.0 Conclusions 8.1 Limiting Conditions for Consolidation 8.1.1 Fuel Assembly Characteristics 8.1.2 Pool Conditions 8.1.3 Consolidated Fuel Storage Canister 4

4 I

E 11 )

_-- - - _ - _ - _ - - - - - - - - - - - - - J

i LIST OF TABLES TABLE NO. TITLE -

3-1 Benchmark Critical Experiments 3-2 Fuel Parameters Employed in Criticality Analysis i 3-3 Comparison of PHOENIX Isotopic Prediction to Yankee Core 5 Measurements 3-4 Benchmark Critical Experiments PHOENIX Comparison i

3-5 Data for U Metal and UO2 Critical Experiments 5-1 Fuel Handling Accidents Radiological Effects ,

l i

4 1

iii l

I i

LIST OF FIGURES TITLE l FIGURE NO.

i 2-1 Fuel Assembly Dismantling Station l

2-2 Storage can Loading Station (A) l 2-3 Storage Can loading Statior. (B) i 2-4 Location of Work Stations ,

3-1 Minimum Burnup vs. Initial U-235 Enrichment for Fuel Assembly Consolidation Operations 3-2 Consolidated Rod Storage Canister Nominal Dimensions 3-3 Prairie Island Spent Fuel Storage Cell i

4-1 Basic Canister Geometry Showing Natural Circulation Cooling 4-2 Rod-to-Rod Flow Passageway ,

I 4-3 Letdown Plate Flow Hole Pattern 4-4 Decay Heat / Rod vs. Time After Shutdown 4-5 Decay Heat / Rod vs. Time After Shutdown 4-6 Time After Shutdown vs. Operating Time l

e

- - - - - - _ _ . - _ _ _ - - _ _ . - - _ _ _ . . x.- 3

i l

FUEL CONSOLIDATION DEMONSTRATION l AT PRAIRIE ISLAND NUCLEAR GENERATING PLANT

1.0 INTRODUCTION

1.1 CONSOLIDATION OF SPENT FUEL AT PRAIRIE ISLAND l I

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 t.wo storage pools shared by both reactors. When both pools are filled with racks, 1582  ;

storage locations are provided. However, PI's Technical Specifications j limit total storage to 1386 spent fuel assemblies, not including those assemblies which can be returned to the reactor. Under current cycle .

management plans, PI's spent fuel storage capacity will be exhausted by l about 1994.

NSP has entered into contract with the Department of Energy (DOE) for '

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 1 DOE may be able to accept spent fuel from utilities by 2000, if Congress will authorize the Monitored Retrievable Storage facility proposed by the DOE. In the interim, each fuel owner has the primary responsibility for 1-1 i

\

h i

storage for its own spent fuel by maximizing the effective use of existing storage facilities at its sites, and by adding new onsite storage capacity f

in a timely manner.

P 1

l 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 canal adjacent to the spent fuel storage pools. Westingho.use 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 I consolidation ratio.

NSP's demonstration will show wh. ether 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 also apply to the State of Minnesota for a Certificate of Need, as required by state law.

6 l m _ - --

1.2 PURPOSE OF REPORT The purpose of this report is to fully document all aspects of the ' Prairie Island spent fuel consolidation demonstration 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.

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

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

il

'i 1

- _ _ _ _ _ _ _ _ _ 1-3. i

1 4

i

]

l j

2.0 CONSOLIDATION PROCESS . 1 i

I i

The consolidation orocess 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 i

and the storage canisters occupy twenty cells. The empty fuel assembly skeletons will be stored for volume reduction at a later date, i 2.1 Equipment Description The equipment us'ed 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 o Other tools and equipment

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

At this station, the top and bottom nozzles are removea from the fuel assembly, the fuel assembly is rotated to the upside down position, and l,

l l .

the fuel rods are removed. The following sub-Issemblies/compor,ents make up the full compliment of equipment used at thi's station:

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 squars array to a closed triangular array.

3. Transition Canister Support Stand The stand positions the transition canister under the fuel assembly and also positions it where it can be lifted by an overhead crane.
4. Rod Push Tool l'

_____ _ _ _ _ _ _ _ _ - - - _ _ _ - 2

. l

. i 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 t'me. A f

guide plate locates the tool over the fuel rods.

5. Nozzle Removal Tools These are special catting tools that cre designed to remove the i i

top and bottom nozzles. They cut the thi.mbles and the thimble screws.

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 transitica canister to the storage can at this station. Following is a brief description of the sub-assemblies of thir station and other equipment used at this point:

i l

l

1. Winch The winch is mounted on the deck, across the fuel transfer canal, and it is used to raise and lower the consolidated rod storage canister (CRSC) loading frame and to move the frame laterally.
2. CRSC Leading Frame 1

2-3

]

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

3. Rod Position Indicator The rod position indicator is used to assure that each fuel rod is 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) -

The objective of the consolidation process is to place the fuel rods in the CRSC with a 2 to 1 ratio. To do this, the CRSC is designed to contain the fuel rods from two fuel assemblies (358 i

rods) and yet fit into a fuel rack cell. 1his 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 i stainless steel sheet metal. '

A partition divides the canister cross section in half. This 1 allows the fuel rods frem one fuel assembly (J79) to be placed in the caniste- at a time. The partition extends above the top end of the canister to serve as a lifting lug. Plates arn riveted to I 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 cf the storage canister. Fuel assembly and storago canister identificatier, numbers are lor.ated on the covers 1

2-4

b and canister walls for fuel rod & accountability in accordance with l ANS-57.10, "Cesign Criteria for Consolidation of LWR Spent Fuel".

The bottom of the storage canister irH udes & letdown pan which is a plate on which the fuel reds rest on. The pan is also used to support the rods during th3 aarlier stages of the consolidation process. It is initially installed in the bottom of the transition canitter to Scpport the fuel rods as they are pushed from the fuel assembly. . It is transferred from the transition canister to the storage canister with the rods, where it remains.

It rests nn 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. Letoown Pan Support Colcans i When the transition canister is in the storage canister loading I frame, the letdown pan rests on the columns. The columns are two 1engths 6f 21/2-inch piping which extend through two holes in the i bottom of the canister. There are two holes on each side of the canister partition. j I

i C. Nozzle Stacking Si.ation 1

4' The nozzles are stacked on top of each other as they are removed from the fuel assemolies. The top nozzles are stacked separately from the bottom nozzles in groups of ten per stack. Tie rods align the nozzles 1 l

(

1

_ - - - _ _ - - - - - - - . u

=

1 l

l and hold the completed stacks together. The completed stacks will fit '

~into a fuel rack cell one stack upon another. A lifting eye is attached to the top for stack handling.

i The work station will 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 Filtration, and Underwater TV Systems These systems will support the process. For example, during some of the operations, such as cutting the nozzles from the fuel assembly, 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 4

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

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 from the Prairie Island plant upon completion of each phase of the fuel consolidation program including this demonstration phase.

2-6

F l i

Other equipment, however, is being supplied to Northern States Power f Company and will remain at the Prairie Island site. These are:

1. Fuel rod storage canister handling tool
2. Stray and broken rod storage can
3. Fuel assembly skeleton handling tool
4. Nozzle stack handling tool i

The storage canister 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.

The stray and broken rod storage can is a design that has been used on previous projects.

It hcs 57 tubes each capable of holding a single fuel rod.

The tubes are supported vertically by a plate and angle fron j 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. {

I I

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.

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 2-7 I 1

l 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.

jl 2.2 Process Description The equipment will be located in the fuel transfer canal (Figure i 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 d sign, I administrative 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 tne 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.
5. i The top nozzle is moved to the nozzle stacking station. '
6. A catch grid is placed over the fuel assembly.
  • 2-8 . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - -
g. .

s 7.

The fuel assembly is rotated to the upside down position and s

s the bottom nozzle removed. ,

j 5

8.

The bottom nozzle is moved to the nozzle stacking station. I

9. A guide plate is placed over the fuel assembly.

! 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 prevent 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. '

i

18. The rod position indicator is removed.

t 19.

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

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

}

j i

I 2-9

I k

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

U picked up and delivered to storage. .

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

This completes the consolidation of one fuel assembly. Since the 1 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.

23. The frame is lowered a second time. At this stage, both

. sides of the storage can are loaded with fuel rods.

24. The rod position indicator is removed.
25. .The FTC is removed, a new letdown pan is installed, and the canister is placed on the transition canister support frame.
26. The covers are installed on the CRSC.
27. The CRSC is moved to the fuel rack.
28. The guide plate is removed from the fuel assembly dism&ntling station.
29. The basket is lowered and the fuel assembly skeleton is

, picked up and delivered to storage. -

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

~ ~ ~

2-10 A

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.

l l

E________________ _ 2-11

e Rod Push Tool 5

6 i Deck i

l

~

s. A  :

i MFRS i

1. Guide Plate  :===. "

.I .

gl.i t ., -

..t t w Fuel Assembly O J

'w .

1. :-

40' 2" .

. I Transition Canister '- ~ -

Support Frame p 3 F, u a i

M ransition Canister  !

~

- ~

/ et Dcwn Pan .-

I

, l f

e

_- Y Fig m 2-1 Fuel Asse.bly Dicantidng Station

, i 1 l

- 1

[fg a -

Winch (2) a .

Deck 11' p Red Position Indicator IP il j E I o

p

' i l

j p 7ransition Canister

. Fuel Rods } ,

11 1 N

-. q ics a: I. '

y Q Let Down Pan s x= =:r 4

l t

I

!pStorage Can Leading Frame l 1

--Storage Can f

b 8

i et Down~ Pan Support Columns

- t

~ ' t; ,

~

- ;j I

i

-n i P i

..- - - l i

I -

Fig.:In 2-2 Ste:. ge can Icading Statien (A)

=- _ _ _ _ _ _ - _ _ - _ _ .

4 l .

l i

Pool f2 l

, 4 Pool fl '

1 l

Location of

~~

Work Stations' I' l .

I i 1

1li I l 3 New Fuel L, __ .J .

M M 9

e 8 e

?

,. e e e e

.p

- . _ _ - - - _ . _ . . - - - . . - _ . _ _n . . _ _ _ - - _ . . - . _ _ . _ . _ _--

-Winch (2) ys t Deck

. g  ; .

b>

.. . . , r a

t [ l Transition Canister a

11'

  • i f

?  ?  ?

r

w

a ,. e, n- ,

s tu.

s b hi

L' -Storage Can

> b

.i l

r /_ -Storage Can Leading Frame Fuel Rods ' '

d

! A *1 l

i

,I .-

1 L y n / et Down Pan

Q ' M' -

d l

. .' . .a.

Let Down Pan Support Columns

~

l .

t Figt:re 2-3 St.cmge cri I.cading Staticn (B)

l

. l' l

t I

j 3.0 CRITICALITY ANALYSIS - 2:1 CONSOLIDATION t l 3.1 Introduction 1

The consolidation equipment, i.e., the fuel transition canister (FTC) and the consolidated rod storage canister (CRSC), criticality analysis is based on maintaining K,ff 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 /

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 consolidat-d 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.

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 for any possible configuration shall be less than or equal to 0.9E, including all uncertainties, under all conditions.

1 3-1

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

" Nuclear Safety Criteria for the Design of Stationary Pressurized Water i

Reactor Plants," Section 5.7, Fuel Handling System; ANSI 57.2-1983, " Design Objectives 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 Acceptance of Spent Fuel Storage and Handling Application," ANSI 8.17-1984,

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

3.3 Criticality Analytical Method The criticality calculation method and cross-section values are verified by comparison with critical experiment data for fuel rods similar to those to be consolidated. This benchmarking data is sufficiently diverse to establish 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 i

fuel storage pool uses the AMPX system of codes (2' ) for cross-section generation and KENO IV(4) for reactivity determination.

The 227 energy group cross-section library (2) that is the common starting j

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. I library, the self-shielded resonance cross-sections that are appropriate for

)

j

- - - - _ _ . J

f .

each particular geometry. The Nordheim Integral Treatment is used. Energy and spatial weighting of cross-sections is perfo'medr by the XSDRNPM program ( ) which is a one-dimensional Sn transport theory code. These multigroup cross-section sets are then used as input to KENO IV O) which is a three dimensional Monte Carlo theory program designed for teactivity 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 c/linder arrays with various'i' 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 CONSOLI0ATION OPERATIONS 3.4.1 Design Description I

f

  • 3-3

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

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

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) of the fuel 57.2-1983.

array will be less than 0.95 as recommended in ANSI 3.4.3 Criticality Analysis The consolidation process will be performed using the sequence of operations 1 In this sequence of operations, the fuel

given in Section 2.2.

configuration will change significantly under normal operating conditions.

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

3-4

o f e I

l a) Fully and partially loaded FTC b) Fully and partially loaded CRSC.

c) Fully and partially loaded fuel assembly I 4

q The following assumptions were used to develop the nominal case KENO models for analysis of the configurations described above:

a)

The fuel contains the highest enrichment acchorized, is at its most reactive point in life, and no crejit is taken for any burnable poison in-the fuel rods. Calculations have shown that the W 14x14 STD fuel pins yields a equal or larger K,ff 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 14x14 STO fuel pins were analyzed (see Table 3-2 for fuel t

parameters).

A b)

The moderator is pure' water at a temperature of 68 F.

3 conservative value of 1.0 gm/cm is used for the density of water.

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

c) d)

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

~ li 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 i assembly and CRSC cases can be bounded by the same case.

1 l

l 3 - _ - - -- _ ----_ __ __ J

3.4.3.1 Fuel Transition Canister The fully or partially loaded fuel transition canister (FTC) is bounded by that can result from any amount of fuel ensuring that the maximum Keff 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 Keff. The maximum FTC fuel area arises from consideration of mechanical tolerances resulting from the manufacturing process on the largest grid opening. The tolerances are stacked in such a way as to maximize the grid opening.

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

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

l 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.

Based on the analysis described above, the following equation is used to develop the maximum Keff f r the partially or fully loaded FTC.

i .

+O method + ( s)2 method Keff

  • Kwo m + Uworst ks)2 l

I o--------- - -

3-6

E Where:

K

= w rst case KENO K,ff that includes material tolerances, wrst 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 KEN 0 K,ff worst ks = 95/95 uncertainty in the method bias method l'

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

K,ff = 0.8800 + 0.0083 + [(0.0054)2 + (0.0018)2 1/2 = 0.8940 3

for the configuration is less than 0.95 including uncertainties at The K,ff a 95/95 probability / confidence level. Therefore, the acceptance criteria for criticality are met for the transfer of fuel rods to the FTC assuming a fuel enrichment of 4.0 w/o U-235 at zero burnup.

i 3.4.3.2 Consolidated Rod Storage Canister and Fuel Assembly The . fully and partially loaded fuel assembly and consolidated rod storage that can canister (CRSC) are bounded by ensuring that the maximum Keff result from any amount of fuel loaded into the fuel assembly or CRSC fuel 1

3-7

I J

To meet this envelop or area does not exceed the design limit of 0.95.

design limit, credit is taken for the reactivity decrease caused by fuel

depletion in fuel assemblies that have initial enrichments greater than 3.0 is This methodology, known as reactivity equivalencing, w/o U-235.

described below.

Reactivity Equivalencing and Methods 3.4.3.2.1 Spent fuel consolidation of fuel assemblies with initial enrichments greater than 3.0 w/o are achievable by means of the concept of reactivity The concept of reactivity equivalencing is predicated upon equivalencing.

A series of the reactivity decrease associated with fuel depletion.

reactivity calculations are performed to generate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield the equivalent K,ff.

contour generated for the Prairie Island Figure 3-1 shows the constant Keff fuel. Note the endpoint at 0 MWD /MTV where the enrichment is 3.0 w/o and a The interpretation of the j 4,000 MWD /MTU where the enrichment is 4.0 w/o.

the maximum reactivity of the fuel assembly or endpoint data is as follows:

CRSC, containing fuel at 4,000 MWD /MTV burnup which had an initial bly  ;

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

is important to recognize that the curve in Figure 3-1 is based on a l

constant maximum reactivity resulting from a fully or partially loaded fue assembly or CRSC.

' 3-8

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

PHOENIX is a depletable, A

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

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 for 30 years after discharge. The fuel reactivity was found to reach a 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 major fission product poison, Xe 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 following discharge. Therefore, the most reactive point 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 isotopic fuel composition has been examined following discharge from a reactor.

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

The The measurements were made on fuel discharged from Yankee .

Core 5(10) data in Table 3-3 shows that the agreement between PHOENIX predictions and measured isotopic compositions is good.

l l

l 3-9

The agreement between reactivities computed with PHOENIX and the results of Key 81 critical benchmark experiments is summarized in Table 3-4.

parameters describing each of the 81 experiments are given in Table 3-5.

These reactivity comparisons again show good agreement between experim.ent and PHOENIX 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 PHOENIX results and the Yankee Core experiments and 81 benchmark experiments indicates closer agreement.

3.4.3.2.2 Reactivity 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 and the CRSC.

The envelop of the largest fuel area is determined to be the CRSC. The maximum CRSC fuel area arises from consideration of me The material thickness tolerances resulting from the manufacturing process.

The CRSC tolerances are stacked in such a manner to maximize the fuel area.

inside diameter (ID) is increased from its nominal value to a maximum o 7.963 inches.

Therefore the area to be loaded with fuel for 'the sensitivity study is 63.4 in .

i . 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 I

l 3-10

I fuel area. In this way no credit is taken for the structural material that l would be in a fuel assembly or the CRSC. The number of fuel rods in the I

2 63.4 in fuel area is varied until the maximum Keff is reached. l Results from the sensitivity study show that the maximum 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.

I'e f f

  • Kworst +Bmethod + Uksworst) + @smethod) + J ksre) .

where:

K =

worst w rst case KENO K,77 that includes material tolerances, and mechanical tolerances which result in the largest CRSC fuel area.

B = method bias determined from benchmark critical method comparisanc ks worst

= -95/95 uncertainty in the worst case KENO K eff ks method

= 95/95 uncertainty in the method bias ks re

= uncertainty in the reactivity equivalence methodology 3-11

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

i K,7f = 0.8912 + 0.0083 + [(0.0058)2 + (0.0018)2 + (0.01)2 31/2 = 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 in the CRSC at an equivalent " fresh fuel" enrichment of 3.0 w/o U 235 ,

3.5 CONSOLI0ATED FUEL STORAGE IN RACKS 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 Keff <0.95 for storage of Westinghouse 14x14 0FA and STD fuel rods and EXXON 14x14 HI-PAR, LO-PAR and TOPR00 fuel rods at 4.0 w/o with utilization of every cell permitted for storage of the CRSC.

3.5.1 Design Description .

I The spent fuel storage cell design is depicted schematically in Figure 3-3 l - and shown in detail in the design drawings in Reference 14. Nominal 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.

l 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 4 i

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 Keff than does the W 14x14 0FA or the EXXON HI-PAR, LOW-PAR, and TOPROD fuel pins when 235 Thus, only the W all fuel pins have the same U enrichment j

14x14 STD fuel pins were analyzed for storage. (See Table 3-2 for I .  !

fuel parameters).

1 i

1 I

3-13 j

a^

b)

All fuel rods contain uranium dioxide at an enrichment of 4.0 w/o 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 of fission product poison material.

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) Theminimumpoisonmaterialloading(i.e.,0.04gIams8-10per square centimeter) is used throughout the array.

A sensitivity analys!s 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 eff 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 of 226 fuel roos 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 wit'h fuel rods in both halves of the i CRSC.

1 1

,The KENO calculation for the nominal case resulted in a K f 0.8419 with eff a 95 percent probability /95 percent confidence level uncertainty of +

0.0037.

l 3-14

- _ _ - _ - _ - - - - - - - a

l l

'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 gap 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 center-to-center spacing. For the spent fuel storage racks, the water gap is reduced'from a nominal value of 0.682" to a minimum of 0.492". Thus, the most conservative, or " worst case", KENO model of the storage racks contains a minimum water gap of 0.492" with symmetrically placed CRSC's.

Baseriontheanalysisdescribedabove,thefollowingequat1Inisusedto develop the maximum k f r the Prairie Island spent fuel storage racks eff with consolidated rod storage canisters:

K,ff = Kworst +Omethod + Bpart + Uks)2 worst + Gs)2 method -

Vnere:

K worst = w rst case KENO K,ff that includes material tolerances, and mechanical tolerances which can result in spacings between canisters less than nominal.

B method = method bias determined from becchmark critical comparisons 3-15

___-__-_____________-_____------m s

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

K eff = 0.9038 + 0.0083 + 0.0010 + [(0.0042)2 + (0.0018)2 3 1/2 = 0.9177 The maximum K eff f r 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 storage o'f' spent fuel in the

[

CRSC a the PI spent fuel storage racks, assuming a fuel enrichment of 4.0 w/o U-235 at zero burnup, and assuming each half of the CRSC contains no fewer than 113 rods.

3.6 POSTULATED ACCIDENTS Accidents can be postulated which would increase reactivity through the uncontrolled spreading of the fuel rods as outlined in Reference 1. This uncontrolled spread could occur as'a result of spillage from a damaged or mishandled canister, or misloading of the FTC or the CRSC. These accident conditions are bounded by the uncontrolled release of two assemblies worth (358) o,f fuel rods. At no time are more rods than this being handled in the fuel consolidation area. The maximum K eff that can result from the 3 -16 L_-_-________---__-___--_. - - -

L ,

l 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.

For these accident conditions however, the double contingency principle of ANSI 8.17-1984 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 presence of soluble baron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

i The presence of 1000 ppm baron in the pool water will decrease reactivity for this postulated accident condition by approximately 30'bercent delta K to 0.850. Therefore, the acceptance criteria for criticality is met for a postulated accident during consolidation involving spillage of a CRSC, assuming a fuel enrichment of 4.0 w/o U-235 at zero burnup, and 1000 ppm boron in the pool water.

Accidents can also be postulated which may damage fuel rods and result in the release of fuel pellets during the consolidation operation. Generic studies at Westinghouse have shown that more than 100 lbs of UO 2 pellets at i 3.0 w/o are needed to form a criticci mass in unborated water. As a result, small releases of fuel debris from broken fuel rods will not cause K eff to exceed 0.95.

Large accumulations of fuel debris (>100 lbs) will most likely not cause K,ff exceed 0.95 with the consideration of the soluble boron in the pool water but must be evaluated on a case by case bases.

i

l

~ Accidents can also be postulated'which may damage fuel assemblies and result l

in the release of fuel rods from,the CRSC. Calculations have shown that any i'uel 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 postulated accidents, should there be a reactivity increase, k

eff w uld be less than 0.95 if the baron concentration in the pool water is greater than 1000 ppm.

3.7 REFERENCES

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

1

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

1

3. N. M. Greene, et al., "AMPX: A Modular Code System for Generating l

Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," ORNL/TM-3706 j (March 1976). i 1

J

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

3-18 i i

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

L6. 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 1 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.

{

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

WCAP-3017-6094, January, 1971.

11. Fuel Transition Canister Dwg. No. 1873E90.
12. Consolidated Rod Storage Canister Dwg. No. 1875E53, 1875E54, 1875E08.

1 I

13. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes, April 14, 1978, "0T Position for Review and 1

Acceptance of Spent Fuel Storage and Handling Applications."  !

3-19

\

(

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

l e

3-20

r7 uh [1b h QI[Jp I. ]

1}3t 3  ;[

t '

885580658509331402900005g67088902 21 1222222221 2222221 22222122341 133 000000000000000000000000000000000 00000000000000000000000000000000,0 K +TTTTTTTTTTTTTTTTTTTTTTTTTTTTTITT 766e8 509l 951 5931 t 8732428551 832649034 9022307 095678201 4225426461 88239 8009tl 9i999l 91 11 0080999999999999099 t f 1

999999999999990900999990999999099 O00OOOOOO 000000O00000OOOOOOO0Ol OO m

ep l p -

b lb0 1 O74OOOOOO 36 347525I 48 1 1 99289327 77402OOOOOOOOOOOO o- 02 1 52 31 4163 SB 1

. 1 6 tl mmmmmmmmm eeuuuuuuuuu 5

[ g __ eennnnnnnnni *

's's ssssss ssss ssmmal immii i i t ti i t

s nl I a sssss nnnnn u u mmmm aaaaaa n

e ti ar rrriiiiirrssl l l l l eeepppppeessaaaaaaaaaiiiiiit 1t i 1 l uut ut luuuul rrrrrrg0og99 l l l l l l t t m re t t t t t ee aaaaaau=nM = m i

r at pa a a a C C C C C a a l. l ddddddddd www44 4 4 4 wwrneeeeeeeee eeeeee l l lll l e eM ' 8DD80 l i t t tl t t tt t i ' ;- pppppp p S aaaaaaaaaaa m t t rrrrrrrrr E ssooooooooo l

bbbbbbbbb a

c in innn i

t r i i i

o f fff r t rrrrrrrrrrrrrrrrrrrrr f fff C c e eeeeeeeeeeeeeeeeeeeeeeeeeeeeaeaaa t t ttt t t t t t t t t t t t t t t t t rrrrrrrrrrrr k

r l

f aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa w w w w w w w w w w w w w w w w w w w w w b b b b b b b pb p p p e a ie m l h

c n

e H t _

n5 e3 m2 666666666666666666666 '

hU 444444444444444444444222222222222 Ic r> 222222222222222222222333333333333 r/ 999999999999 nw E

. ssssssssssss rrrrrrrrrrrr

, eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee cccccccccccccccccccccdddddddddddd ii i I i i I II i iiiiiI i I i Ii nnnonnnnnnnn ttttt tt t t tttt tt tttt t ti I il i i i iiiti o tt tt tt tt t t t t t t t t t t t t tl il l l l ll ll tl

- o aaaaaaaaaaaaaaaaaaaaayyyyyyyyyyyy lt at l l l llll l l ll l ll l l lll l l cccccccccccc b rn dddddddddddddddddddddl llll l llllll eI nr oooooooooooooooooooooaaaaaaaaaaaa rrrrrrrrrrrrrrrrrrrrrtI t l t tt t t t tt ec e, nei e e e e e en e e e e Gs e 0nG00000O0000000O00On s m mms ms ss mm n i D UUUU01 0UU0UUU00UUUUUUUuOOUUOOUOUU 1

1 234567890I 2345628908 234567890423 1 I 1 1 1 1 1 1 1 1 22222222223333

!l

Table 3-2 Ital Pr'~ tars E=plcyed in C:-/*ticality

  • = - .
  • Efillp$,5 -.

t .

w

- e a e e = tn  !

x e m w e C. us -- n - = m 1 w

=

m

= w n v vw m ~

m. m. C. C w

n >

  • C C C C + C-

~ C m= C C 31 C- C w

i

. e.

v1 w .

, =u m To e e e e e S > - m m n w us . a C. g) m = "0

  • v = w n e - an . = an. C. .

n n w

= .

O* C.

C c-J. O D C

~ C C C C 3i ns e

==

m

>e

.= .,

w a -

C oc z e <<

E

== c.: .z.

= -N%

== ww

=

c m==

w er e n  :-)

m uu u e C m W O e a

  • w

.u. ww m n m m m u) = m. C. - C- - . -- - -.

., = = = = w W1 = C C C C m.

== . C. m*  % N N N 6 2: O C C = 5 w m v cc C w - N e xx W1 C w- C

  • C xx a

-- ww C C- C C 2

.c=

t w

w m =

b K w wc

., == c c tn > an m e C. e, # == > =~ w m n e zz n -w =n =m m o a <=

u - m w e =

an. C.

6 cx esc. .

Q an. Ce n x C gw}.

O O C C. C A u,a C- O m.

.I b C  % .C * .C U C 3 u e v m e an e= C .C C ,

. .C= .== y = m+,.==manm c

=se .c e e spam an, d'* U h "U Z .A e,

.

  • C C == v m me . U M Q* La * > D 3 C m C Q* 4

=

V C w es am. Q == W o O == == .C. == C Q >

a == . m9 4 a w .s

  • T Cu e 4 e . U h (# #

.- m C. gs - = v U C == == a C W .a

- m C C a= Q C 4.

  • Z e 3 3 3h U W e == = C > C > e-C ** es b em auss N **

6 == **

  • .E GA G g 9 9 ma*

L' e. 's W a- == g a. c a3 e- C C a B Q&m W =C = W =3 e e-

== W Q O Q

== b e == b O w **

&m = >= A & & == b I g< u - a a em ou e 6

-u 6:

. ..= o

u tu c. - .:,

C mC C-

. - 2 =,

Ei I n. E W h w= m E .i G f " "-

--_____e_____________ . _ _ _ _ _ . _ _ _ _ _

4 e Table 3-3 Wh7 cf HEDCX Isc: pic Predictica 4

to Yardme C=s 5 Measents Quantity.

(Atem Ratio)-

% Difference U235/U

-0.57 U23s/U

-0.28 U238/U -

-0.03

~

PU239/U '

+3.27 PU240/U

' +3.53 PU241/U

-f.01 PU242/U

-0.20

- - - - ~~' ~

PU239/U238

~ +3.24 MASS (PU/U) +1.41 FISS-PU/ TOT-FU -0.02 Percant difference is average differenca of ten. con =arisens for each isetcpe.

e e

e .

o 4 e a

g +

l-4 Table ~1-4 Ed:rf.. ark Critic:al Experi::ents EErr(

.. C ._ arisen - .

l Description of Nt=ber of Ex eriments PHOENIX k,f, Using Ex:eriment:

Exterimenthucklines UO 2

Al clad 14 ,

,g947 SS clad 19

,g944 Sc' rated H O 7 2

.9940 Subtetal -

40 ,9944 U-Metal - - - -- - - - -- - ~ ~ - ~ ~ ~ ~ ' " ~ ~ ~

Al clad 41 1.0012 ,

TOTAL 31 ,gg73

  • g e

O 4

e W

4 9

e

. W O

en b

s Table 3-5 Cata for 1%t 2. and UC2 6.h

. W ~e2.3

. . Fuel Pellet Clad Cla~s m LaI I"

~

Case Call .

/0 H20/U Density. Diameter Material 00 Thickness Nu=cer Type U-235 Ratio (G/CC) - (CM) Clad (CM) (CM) pi (C$[- p 1 Hexa 1.328 3.02 7.53 1.5255 Aluminum 1.5915 .07110 2.2n~0  !

2 Hexa 1.328 3.95 7.53 0'0 1.5255 Aluminum 1.5915 .07110 2'250 0'O 3 Hexa 1.328 4.95 7.53 1.5255 Aluminum 1.5915 .07110 2'EED O"O 4 Hexa 1.328 3.92 7.52 .9855 Aluminum 1.1505 .07110 EE2 0 1.328 4.89 0'O 5 Hexa 7.52 .9255 Aluminum 1.1505 .07110 ll5E 0 1 0'O 5 Hexa 1.328 2.88 10.53 .9728 Aluminum 1.1505 .07110 1' *Nn O'O 7 Hexa 1.328 3.58 10.53 .9728 Aluminum 1.1505 .07110 EE50 0'O O Hexa 1.328 4.83 10.53 .9728 Aluminum 1.1505 .071.10 l5Eco 1 O~O 9 Scuare 2.734 2.18 10.18 .7520 55-304 .8594 .04085 1.0257 0.0 10 Scuare 2.734 2.92 10.18 .7520 55-304 .8594 .04085 1.104c 0.0 11 Scuare 2.734 3.85 10.18 .7520 55-304 .8594 .04085 1. led O

12 Square 2.734 7.02 10.18 .7520 55-304 .8594 .04085 la Scuare 2.734 8.49 10.18 1.4552 0O0 '

.7520 55-304 .8594 .0408: 1 c:2' O^O 14 Scuare 2.734 10.38 10.18 .7620 55-304 .8594 .04085 1559[ 0.0 lo Scuare 2.734 2.50 10.18 .7520 55-304 .8594 .04085 1.0517 0.0 15 Square 2./34 4.51 10.18 .7520 55-304 .8594 .04085 1.2522 0.0 17 Scuare 3.745 2.50 10.27 .7544 55-304 .8500 .04050 ~ 1.0517 ~ 0. 0 ~

8 --Scuare 3.745 '4.51 "10.37 .7544 '55-304 .8500 '.04050 1.2522 0.0 3

}9 Scuare 3.745 4.51 10.37 .7544 55-304 .8500 .04050 1.2522 0.0 g Scuare 3.745 Scuare 3.745 4.51 10.37 4.51 10.37

.7544 S5-304 .8500 .04050 1.2522 455.0 c;

f .7544 S5-304 .8500 .04050 1.2522 709.0 Square 3./45 4.:1 10.37 .7544 55-304 .8500 .04050 1.2522 1250.0 23 Scuare 3.745 4.51 10.37 .7544 S5-304 .8500 .04050 1.2522 1334.0 g Scuare 3./4c Scuare 4.059 4.51 10.37 .7544 55-304 .8500 .04050 1.2522 1477.0

g. 2.:: 9.45 1.1278 55-304 1.2090 .04050 1.5113 0.0 g5 Scuare 4.05e 2.5: 9.45 1.1278 55-304 1.2090 .04050 1.5113 3392.0 47 Square 4.059 2.14 9.46 1.1278 55-304 1.2090 .04050 1.4500 0.0 1 28 Scuare 2.490 2.84 10.24 1.0297 Aluminum 1.2050 .08130 1.5113 0.0 1 49 Scuare 3.037 2.54 9.28 1.1258 55-304 1.1701 .07153 1.5550 0.0 I 30 Square 3.037 S.15 9.28 1.1258 55-304 1.2701 .07153 2.1950 0.0 1 31 Scuare 4.059 2.59 9.45 1.12cm ec 204 1.2701 .07153 1.5550 0.0 l

.:c Square 4.059 3.53 9.45 1.1258 55-304 1.2701' .07153 1.5540 0.0 33 , Square 4.059 8.02 9.45 1.1258 55-304 1.2701 .07153 2.1950 0.0 34 Scuare 4.059 9.90 9.45 1.1258 S5-304 1.2701 .07153 2.3810 0.0 35 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2050 .08130 1.5113 1577.0 I 35 Hexa 2.095 2.05 10.38 1.5240 Aluminum 1.5915 .07112 2.1737 0.0 l 37 . Hexa 2.095 3.09 10.38 1.5240 Aluminum 1.5915 .07112 2.4052 0.0 l 38 Hexa 2.095 4.12 10.38 1.5240 Aluminum 1.5915 .07112 2.5152 0.0 39 Hexa 2.095 5.14 10.38 1.5240 Aluminum 1.5915 .07112 2.9591 0.0 40 Hexa 2.095 8.20 10.38 1.5240 Aluminum 1.5915 .07112 3.3255 0.0 41 Hexa 1.307 1.01 18.90 1.5240 Aluminum 1.5915 .07112 2.1742 0.0 42 Hexa 1.307 1.51 18.90 1.5240 Aluminum 1.6915 .07112 2.4054 0.0 43 Hexa 1.307 2.02 18.90 1.5240 Aluminum 1.5915 .07117. 2,5152 0.0 l

-- --- Table 3-5 nata fer Metal ard It2 c:itic:11 -

Expe'iner:ts (C::nt)

Fuel .

Pellet Clad Clad 'Lattica Case Cell A/O H20/U Density Diameter Waterial 00 Number Type U-235 Ratio (G/CC) Thickness Pitch B-10 (CM) Clad (CM) (CM) (CH) PPM 44 Hexa 1.307 3.01 18.50 1.5240 Aluminum 1.5915 .07112 45 Hexa 1.307 4.02 18.90 1.5240 2.9595 0.0 45 1.150 Aluminum 1.5915 .07112 3.32a9 0.0 Hexa 1.01 18.90 1.5240. Aluminum 1.5915 .07112 2.1742 47 Hexa 1.150 1.51 18.90 1.5240 0.0 48 Hexa 1.150 Aluminen 1.5915 .07112 2.4054 0.0 2.02 18.90 1.5240 Aluminum 1.5915 .07112 49 Hexa 1.150 3.01 18.90 2.5152 0.0 1.5240 Aluminum 1.5915 .07112 50 Hexa 1.150 4.02 18.90 1.5240 2.9895 0.0 51 Hexa 1.040 Aluminem 1.5915 .07112 3.3249 0.0 1.01 18.90 1.5240 Aluminum 1.5915 .07112 52 Hexa 1.040 1.51 18.90 1.5240 2.1742 0.0 53 1.040 Aluminum 1.5915 .07112 2.4054 0.0 Hexa 2.02 18.90 1.5240 Aluminum 1.5915 .07112 54 Hexa 1.040 . 3.01 18.50 2.5152 0.0 1.5240 Altminum 1.5916 _.07112 55 Hexa 1.040 4.02 18.50 2.9595 0.0 1.5240 Aluminum 1.5915' .07112 3.3249

. 55 Hexa 1.307 1.00 18.90 .9830 0.0 57 Hexa 1.307 Aluminum 1.1505 .07112 1.4412 0.0 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5925 53 Hexa 1.307 2.02 18.90 .5830 0.0 59 Hexa 1.307 Aluminum 1.1505 .07112 1.7247 3.02 18.50 .5830 Aluminum 1.1505 .07112 - 1.9509

~ ~ ~0.0 50 Hexa 1,307 4.02 18.90 .5830 0.0 51 Heta 1.150 Aluminum 1.1505 .07112 2.1742 0.0 1.52 18.90 .9830 Alumines 1.1505 .07112 1.5925

'52 Hexa 1.150 2.02 18.90 .9530 0.0 53 Hexa 1.150 Aluminum 1.1505 .07112 1.7247 0.0 3.02 18.90 .S830 Aluminum 1.1505 .07112 1.9509 54 Hexa 1.150 4.02 18.90 .9830 0.0 55 Hexa 1.150 Aluminum 1.'1505 .07112 2.1742 0.0 1.00 18.90 .9830 Aluminum 1.1505 .07112 1.4412 55 Hexa 1.150 1.52 18.90 .9830 0.0 57 Hexa 1.150 Aluminum 1.1505 .07112 1.5925 0.0 2.02 18.90 .9830 Aluminum 1.1505 .07112 1.7247 58 Hexa 1.150 3.02 18.90 .9830 0.0 59 Hexa 1.150 Aluminum 1.1506 .07112 1.9509 0.0 4.02 18.50 .9830 Aluminum 1.1505 .07112 2.1742 70 Hexa 1.040 1.33 18.50 19.050 0.0 71 Hexa 1.040 Aluminum 2.0574 .07520 2.8587 0.0 1.58 18.90 19.050 Aluminum 2.0574 .07520 3.00E5 72 Hexa 1.040 1.83 18.90 19.050 0.0 73 Aluminum 2.0574 .07520 3.1425 0.0 Hexa 1.040 2.33 18.90 19.050 Aluminum 2.0574 74 .07520 3.3942 0.0 Hexa 1.040 2.83 18.90 19.050 Al uminum 2.0574 75 .07520 3.5254 0.0 Hexa 1.040 3.83 18.90 19.050 Aluminum 2.0574 j 75 .07520 4.0555 0.0 Hexa 1.310 2.02 18.88 1.5240 77 Hexa 1.310 Aluminum 1.5915 .07112 2.5150 0.0 3.01 18.88 1.5240 Aluminum 1.5915 .07112 2.9900 0.0

{

70 Hexa 1.159 2.02 18.88 j 1.5240 Al u=inum 1.5915 .07112 2.5150 0.0 79 - Hexa 1.159 3.01 18.88 1.5240 j

Aluminum 1.5915 .07112 2.9900 0.0 80 Hexa 1.312 2.03 18.88 .9830 Aluminum 1.1505 .07112 1.7250 0.0 81 Hexa 1.312 3.02 18.BS .9830 Aluminum 1.1505 .07112 1.9510 0.0 1

i l

]

. .i

- t

-1 1

  • l l

-4 gg \

.- i 1

(

1.3 j l

i A D u / -

m i I N

C ACCI?IAE.C /

it:

a 1.s

{

v ,

C D.

= s.s  !

C::

C l C'

>- . /

l J

C l l-2 NOT ACCI? TABLE I

w m ...

D

/ ,

t

< ,/ {t

. )

. i ,!

\

. I

s. . l u I u u u u u u u u o 1

{

IN171 A1 ENRICHMENT ('#/0 U-235) .

l l

' l

. 1

. l

. l e

- 3 Tigre 2-1 Miri'm Burz:tp vs. Initial U-235 E. rich:mnt Fcr Fuel Asse:.bly Ccnsclidatica Cperatiens

4 l

l i

\

j

\

O N

-0.05O

" l l I 4 4 -

)

A l' 1 1 1 1 1 i  ;  ; i ,  :.

! 1

  • I *I 1 i xx x x i 1 1 1 1 ,c

.~

/ 1 1 x xx ili x i i 1 i ,

s I i i x x x x 1 I I I xI y y N

~7 9 8 3" t w & I I I I I W,n ,,,J- -

py'y , ,

IlI I I 1 I I I )

c s'I >I 1 I I 1

~ %,1 I

1 1

x 1

y y

x, y y y y

s I I f'

-I A I Y Y Y Y YIy y y y y y y y hr 8 i O

I O

e Fi m 3-2 Comclidstad R::d stcrage wyg y;c;,,i al Di.~ensions


._CL___._.________________

o , ,, e .

-~ r-

i 4 i l e'- s s 1 i t .

s I).s

.n L

+ -

cn

? -

A

~

c:

W e +

W E e

W a

y

.._. C

, s g C:

A s

f W r; s =

Tisl d , , w e .

U i

.A ..

l a

i i

- I y N u ai '

" l l -

1 8

)i N i

N N I .

. t v

i '

. l

.  % f o .

< m l.

i.

jN cd (lq  ;

li

. -  : . "; I t

r% -

-s 1 '

9 -

N. j I cc .

1 . a tt F p n 1,

e. e a

. ,e . '

! = c

.S

- N y e e s m 6 f:

- uou uzu y

,- - oe v

=mu

  • c .: =

~<c

=

1

. c =.n =a i Ch N N

.. . e. . c.

4 ece Figure 3-3 prairie Island spe-T Ibel Storage Cell Ncr.u.al Dw.ciens

- _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ - -- _ _ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ - _ _

4.0 THERMAL / HYDRAULIC ANALYSIS OF CONSOLIDATED FUEL STORAGE CANISTER 4.1

SUMMARY

The fuel consolidation canister (CRSC) provides a means for long-term storage of compacted spent fuel rods in a spent fuel pool. Because the 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.

The time after shutdown which satisfies the no-boiling requirement is a function not only of the canister-fuel hy2raulics, but also of the power operating history of the fuel. Consequently, analyses were performed for a 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) verifying that the hole pattern in the canister letdown plate maintains a low hydraulic loss and provides adequate cooling to all fuel reds, and b) -

. verifying that the offset distance between the letdown plate and the bottom plate allows Uniformity of the flow into the caniste . 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 4-1

adequate. Similarly, the offset distance was. found t'o be satisfactory to permit flow uniformity without necessitating an increase;in CRSC design' length.

4.2 THERMAL-HYDRAULIC ANALYSIS The mechanism for cooling the fuel rods in the canister is natural 4 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 j ' temperature increases with increasing decay heat level faster than does the 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 poo.l. 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

1 I

1 .

2 Kw

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

2 p p a f p a 2pgA l

l l = pb[(Z + 0.5 Z )(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 X can be expressed as:

  • Z f  !

K = (f--) + IR 0 fuel (4-2) i passageway where -

1 I

I 4-3 {

i

1 IR = irreversible losses (expansions and contractions of the 1 bottom, top,andletdownpla.tes) f = passageway friction factor D = passageway hydraulic diameter The reason for expressing equation (4-2) in the manner indicated is because i the (f Z f/0) 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 U

i Re = Reynolds number and I j

N = 96 for laminar flow (per Ref 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 pC p(T -T) a (4-4) 1

! I f 44 4 l-___-_________________. l

i l

l J

{

)

where '

i

,C p

= . fluid specific heat 1

Equations (4-1), (4-2) (neglecting IR), (4-3) and (4-4) can be combined to l yield the allowable power per rod as a function of the exit temperature T :

p 1

Z Z 1+n 3-n 2 p+ f/2 D (T - T ) 2-n Q = 2 PAC ([_ab ) p a ] (4-5) p N Z n f v where l

v = fluid 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 the IR 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 equation (4-5) was necessary. For the calculations, Ta was assumed to be 150 deg F, and Tp was 4-5

S 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 tempuature is limited to 120 deg F during CRSC transfer from the consolidation area to the rack cells.

4.3 LETDOWN PLATE / BOTTOM PLATE / CANISTER HYDRAULICS The hole pattern shown in Figure 4-3 was determined to be suitable for providing flow uniformly to all coolant passageways between fuel rods 1

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 1

l , adequate to permit cooling flow to pass through the letdown plate without incurring significant hydraulic losses. This permits letdown plate 1

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

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 i 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 i plate holes will not be distributed adequately to those letdown plate holes on the outer periphery of the canister, 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 l-burnup history, was assumed to be 1.8 times the value calculated using the approved methodology in (Ref. 4-2). This accounted for radial and axial

, core power distributions pertinent to actual fuel burnup histories. ,

Figures 4-4 and 4-5 show the decay heat rate per rod as a function of time l j

after shutdown for various core operating times prior to shutdown. Core 1

4-7

i 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 of 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).

I 4-8

A 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 figure will be used as a guide for the minimum times after shutdown required before fuel consolidation is initiated.

If external spent fuel cooling is lost, the pool water temperature would rise until boiling at the pool surface began. For this condition, boiling would take place in the CRSC, but a conservative check on canister thermal hydraulic conditions determined that the maximum fuel cla[ ding temperature would not exceed 263 F. This is well below the nonnal fuel operating temperatures .

4.6 CONCLUSION

S t

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 l plate), 0.125 inch in diameter, is acceptable. This design ensures flow uniformity and maintains low hydraulic losses in the letdown

! plate.

1

{

4-9 )

w__- _ - - _ 1

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

4 i

f 4-10

. - _ _ _ _ _ _ t

2. ;z ;- _ .  ;. . - - - -

7,~ 7, M rack cell wall

./

/ -- ,

e.. ..

. . . . . . / --

- .. -i . ..- --

.. c. y s,, / . , ,

., /

- " ' ^ '

. ,J '

/ ,

,e ./

\ opt plate

/ '

/

/ /

' /

/

/ /

/ l

/ /

/ /

/ / l

/

< / /

, /

/ fuel fuel y canister wall retis

h rods FUEL /

POOL .

/ /

(150*F) / / -

j ..-

y divider plate

/ /

/ /

/ /

/ s

/ /

' /

l /

/

/

,/ /

/ /

/ /

/ .

' letdown plate

./

/

/

/>/

~

/

/ /

'/ it ti ti i t; ?! it bottom plate rs y$W s .

t

?

$ [

~

~ ~

) [ ) 6 6 cross bars

- ~

- t FLOW n .

Figten 4-1 Basic We Gecmetry Showing Natural circulation cccling 1 u__.___.6 _

. . t

. (

l l

i Flow Hole Rod Rod

/ Rod

, a ~

,,N <

's / s f' 's /

N g

f

\., ,

  1. s, , s :fg '-

,' \ , , , ' ) Flow Passageway M q

/ ~3f r

'y.'L~hh$.,kk('T/ 3, t ,1 p e

N e \ ,/' ,# g j\

Rod Rod i Flow Hole Fuel Fuel Rod Rod

~ ~

/\ /\ /\

'fA

  1. 4 Vis YM/

\ etdown L Plate

, i p

I I

Figre 4-2 Red to Red Flcw Passageway ]

+- -'

s.-  :' # -4. -

__m r x -

ri o om ,

om o G e me o,e ce 11 j G QQ O' C 0 Q Qi @ C )

, O G G G G GlO O O t l O O 9 G O 9 @l;- N 9_@ @ 9 0 9 GIO o G e e l e o o e l ij

, 1lc o io D o OCCO jG, QO}G G Q G G 'lG Q {.G SQ@ @ Gj@_@ e'G 2@ @ @!@ @ @ %1l mee n.

gG e e ~e le e: s

- 6, (1, th t%L

^

-Q Q - -Q , ,

n,wn a -

i 3 6

  • 6 y[7
  • @l [ t d_ (M .

{C A sj ay_t s _ q/ t 4

' 4 pgig g pj l

_l7 ,_open volume T

divider plate FUEL RCD 0.125" FLOW HOLE 4 l

(178PERPLATE)

~ ~

. J l

i Fig:re 4-3 Ic+& n Plate Flew Hele pa _czn

l

. I c

. .s ~

4 + tg

. i' .

i t l 3

E. o db. .$

  • ' .. .O T;:--.:. { 5 . _..

..'"..,ou ' ' ~ .. . .. -

. . 'O .

~

l.~ : . . . .

D ,, , , >~

~$ - l 4

55 5h {

(7) 2222 8888

o. o. o. o.

_ m.

m ]2:

  • C -o o C 4 j

g c2 + o 4

{f m

J 0 ,

I'.

<EF .

l 2a lug 1a I'g 4

i 0 -

h5 bd c-2 6

CQ U ic civ %a C v oa>0 $

4 m

oM e il

(

LU

~

C "I

<C o .

LU

-Q -

N, 4 4 4 . i , i i iO O O *O O O o b to O v M N o

e o

. ~

(Des /nig CO-3) COW /1r.r4 AYO3C Figure 4-4 Decay Heat / Red v . Tire Mr whn s

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

I

=

a x *29t '

2 U5

c. ~

!J o .a . r Q

.g e )

O .

Y

_. l W ~ )

e\

i r- EEEE EE8E a .= .c .=

Cn 8888 -a c

m n: =. =. c. c. _

LU 5 R888 m n D+ o4 R

'3 ,

- a

c 4 oaE -

w E

OE E

r-ei Ld i

, o .

LU N.

i i e i a e 4 A e e e e e e eus e

c w r1 N e b .

l'

\

(:os/nE EC-3) CCM/1YF" AY03C l

Figure 4-5 Decay Heat /Texi vs. Ti:ne er Shutdem

(_ _. .. I

,l~ii i

m _

E M

I

! . , i. i 0 6

T .

G N

T -

i S

D _

A O

t R R u l E

Eb i

U F

PN A E.

I F O

OCN - '

B A

T P i 0 E

)

s r .,

O E 4)MaI u o m SIT C d T l

C I ,

vtM I L

A T

nG a

sN o f

O I O uT s NS N I

I hoMd n TD a WOC s

Ii (P u _

O oh O D R

DO R E (T W

TL O P

UE U -

L lF 0 t

U SS NP E.

2 F F

I R B A U _

T E P E

C

- C A

F i A _

E .

M l

- - - - - ~ - - r 0

- 1 9 5 7 6 5 4 3 0 O 0 0 0 0 0 .

Og

. ^

~ ' o E 8.t.

m.

g aId , C c,Wg8e Nfa uJ' a!T IP f mi i s* ,

1.,I' d. g d 5

, \l

5.0 RADIOLOGICAL ANALYSIS OF HANDLING ACCIDENTS A Radiological analysis was performed to determine the thyroid and whole body doses at the exclusion area boundary resulting from a postulated dr accident involving a consolidated rod storage canister (CRSC).

The postulated accident was analyzed only for the fuel handling building (not containment) since that is where fuel consolidation activities place.

5.1 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 r each) and all fuel rods suffer clad damage in the drop.

2.

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

3.

The activity in the void space of the fuel rods in the canister is based on the core activities given in Table A.4 Section 14A of the Prairie Island 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 are used). A radial peaking factor 5-1

t 4

of 1.70 is used, which is conservative compared'to the value of 1.65

.. allowed by Reg Guide 1.25.

4. 3 The accident x/Q value of 6.5 x 10-4 sec/m is used. This is the same x/Q value which is used in the fuel handling accident described in the USAR.
5. A decay period of 2 years is used.

6.

Dose convcesion factors from Regulatory Guide 1.109 are used (Table B-1 for noble gases).

7. ~

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

Under these assumptions, an activity release of 3.308 x 103 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 '

1 at the exclusion area boundary:

Thyroid Dose = 0 Beta Skin Dose = 91.4 Millirem Gamma Body Dose = 1.1 Millirem .

w-Table 5.1 compares the radiological effects of a canister drop with a freshly discharged fuel assembly drop, and with 10CFR100 guidelines and NUREG 0800 dose limits.

5-2

There is no thryoid dose from the canister drop because there is no Iodine present after two years of decay. Thus, the consequences.of dropping a CRSC full of 2 year decayed fuel rods do not exceed 10CFR100 nor NUREG 0800 dose limits, and are insignificant in comparison with the consequences of the fuel handling accident evaluated in the PI USAR.

For a postulated accident involving a canister dropping onto a freshly discharged assembly, the total radiological effect would be bounded by the sum of the releases from the CRSC and the fuel assembly, and would not exceed 10CFR100 nor NUREG 0800 dose limits, iM l

l 5-3 l

9 TABLE 5-1 ,

l l

FUEL FANDLING ACCIDENTS RADIOLOGICAL EFFECTS Whole Body Thyroid Line Analysis Dose (REM) Dose (REM) l 1 358 rods (2 yr decay) 0.093 0

-2 1 assembly (USAR updated 2.6 13.8 evaluation, high-burnup fuel 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay) 3 10 CFR 100 dose guidelines 25 300 4- NUREG 0800 dose limits 6.25 75 (25% of 10 CFR 100)

)

w h

e w

1 E----____-----___--. - .5-4

5.2 OTHER POTENTIAL ACCIDENTS Other potential accidents which may occur during the consolidation process l 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.
4. Dropping of equipment.

Although equipment design and personnel training are aimed at prevention of 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 i

~

Rupture can occur due to rod handling because the rods have to be deflected. in order to consolidate them. Deflection would occur en

. straight rods as they are moved from the fuel assembly to the

, transition canister and as they are moved from the transition canister to the storage can. Another cause Lf red deflection will be the removal of bowed rods from the fuel assembly. The 5-5

likelihood of rod rupture due to deflection is low based on data i 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 l -

The proposed consolidation process minimizes the possibility of dropping individual rods because the rods are pushed down instead-of lifted. However, a pushed rod could free-fall in the equipment if drag devices malfunction.

5-6

I

(

i' 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.

9 l

l L

l 6.0 MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS l 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 ecuipment 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 Wrmtinghouse's experience and lessons learned from consolidating four fuel assemblien, in 1982, at Duke Power's Oconee plant. Thestoragecanistebshavemet'al 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 transition canister. Its design is based on the transitior canister developed for consolidating Oconee fuel. The long handled tools for moving equipment underwater have been used extensively.

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

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

" Design Criteria for Consolidation of LWR Spent Fuel", whic!7 states that the equipment is to be designed in accordance with commercial codes and i

l standards. The codes and standards provide mechanical design guidance.

The majority of the materials used below water are made of stainless steel and aluminum. Small amounts of bronze and brass are used in threads and 1

l l 6-1

-1 I

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

) i The materials were also selected in accordance with ANS-57.10 which, in

1. turn, specifies materials per the ASME Boiler and Pressure Vessel Code, Section 111, Division 1, Subsection NF, or AISC-53261978 Specification for Design, Fabrication and Erection of Structural Steel for Buildings. t Structural design is based upon the codes and standards referenced in ANS-57.10, Allowable stresses and strengths'of section members or connections are as given in ANSI /AISC N690, " Specifications for Design, Fabrication, and Erection of Safety related Structural Steel in Nuclear I Service", 1984.

l 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 fer 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.

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 w> . _ _ _ _ . ___. _ _ _ _ _ _ _ _

t the pool area. Therefore, accidental' dropping of the equipment should not compromise the operation of safety and cooling systems,' o'r damage stored

- fuel.

.M

)

~

e e

6-3

9 i

8.0 CONCLUSION

S I

In preparation for conducting a demonstration program of spent fuel consolidation, Northern States Power Company has thoroughly examined all aspects of the process to provide assurance of its feasibility, reliability and safety. Analyses were conducted to verify that the criticality and thermal-hydraulic design bases of the Prairie Island Nuclear Generating Plant's spent fuel storage pool were not exceeded and that the radiological effects, resulting from postulated accidetns occurring during the [

consolidation demonstration, are bounded by those of accidents already postulated in the Prairie Island Updated Safety Analysis Report.

Additionally the compatibility of the material of construr: tion of the consolidation equipment with spent fuel pool coolant was verified The probability of accidents occurring as a result of the fuel consolidation is minimized by testing and demonstration of the equipment under simulated consolidation procedures and through the training of personnel in these procedures prior to actual application.

8.1 LIMITING CONDITIONS FOR CONSOLIDATION The analyses presented in this report have identifed certain constraints on the fuel which can be consolidated, on the pool conditior - during

_ consolidation activities, and on the final pin loading of a consolidated

, fuel storage canister. These are referred to here as Limiting Conditions for Consolidation.

1 l

1 I

8-1 l

k Spent fuel assemblies which meet all of the following criteria are eligible for consolidation.

a) Both the initial enrichment and burnup of the assembly fall within the acceptable region of Figure 3.1 of this report.

b) Both the core operating time and cooling time of the assembly fall within the acceptable region of Figure 4.6 of this report.

c) The assembly out of core decay period is two years or greater, so as to comply with assumptions of Section 5, Radiological Analysis of Fuel Handling Accidents.

4 8.1.2 Pool Conditions .

The following conditions shall be met whenever consolidation is taking place, a) The spent fuel pool water will be at a boron concentration of either 1000 ppm, or at a lesser value as long as the calculation is performed to show that keff will be less than 0.95 for the postulated accident conditions defined in Section 3.6.

b) The spent fuel pool water temperature is less than or equal to 120*F, as discussed in Section 4.2.

8.1.3 Consolidated Fuel Storage Canister Each half of a consolidated fuel storage canister shall contain no fewer than 113 rods, as discussed in Section 3.5.3 of this report.

The limiting conditions pertaining to fuel assembly characteristics will be verified by reviewing the appropriate fuel assembly records for each assembly to be consolidated, prior to it being consolidated.

The limiting conditions pertaining to the pool and consolidated fuel storage canister will be administratively controlled.

i i

e 1

8-3

l , MED-RPV-1524 From . Mechanical Equipment Design WlN 236-6168 Date September 3,1987 -

Subject:

Fuel Rod Clad Surface

' Temperature in NSP Fuel Canisters Ref: a) MED-RPV-1450, June 29,1987

! to L. R. Benson oc: R. Laubham C, H. Boyd E. J. Rusnica E. Bassler Central File RPVSA-87-T131/1 A In Reference (a), acceptable times after shutdown were determined for consolidating NSP gpent fuel into canisters. The Reference (a) analysis was based on a 150 F pool temperature, assmed that the canisters were at the bottom of the pool, and required a no-boiling criterion. The question has been raised whether acceptable cooling performance of the spent fuel in the canisters can be shown if the pool itself is in a saturated liquid condition. The criterion for acceptability in this case would clearly not be a no-boiling condition, but would be a limit on clad surface temperature. For scoping gpurposes, it is assmed that a clad surface temperature less than 500 F is acceptable.

- The effect of allowing boiling in the analysis of Reference (a) would be to significantly increase the flow rates through the limiting channels in the canister. This would, in turn, promote cooling. However, for the present evaluation, it will be assumed that the flow rate remains at the value in the limiting channel presented in Reference (a). Using this flow rate and the limiting heat generation rate found in the Reference (a) analysis, two quantities were estimated: a) the quality and temperature of the fluid existing the channel, and b) the elevation of the clad temperature above the fluid exit temperature. From these, the peak clad surface temperature is determined. l Because the varying saturatiog temperature through the canisters in the racks is much higher than 212 F, there is a region near the channel inlet in which the fluid is strictly liquid. A conservative length of this _

re61on is approximately 45 inches from the botte of the fuel rods. After flowing 45 inches, steam begins to form. Because the fluid above 45 inches is two-phase, however, the heat transfer from the adjacent rods would only be effective in steam production, and would certainly not raise the fluid temperature abgve the saturation temperature at the botte of the fuel canister (M55 F). At the top of the canister, steam quality was estimated to be approximately 7 5%.

Discard Date:

..m.,

g_.

,f c.

a Though, with a-quality of only 7.55, it'can be reasonably argued that a liquid beat transfer. coefficient between the fluid and the-clad is appropriate, the heat transfer coefficient was based on steam properties to-insure conservatism. The clad-to-fluid temperature difference calculated 255 + 7 3 = 262 3;F, well below 500 F.from this was 7 3 F, which means thgt a maxim Questions on this evaluation should be directed to R. E. Schwirian, WIN 236-6168.

Please forward this information to NSP's Laura McCarten.

R. E. Schwirian RPV System Analysis l

l

r I i t '

3955 ANNAPOLIS LANE e PLYMOUTH, MN 55441

  • PHONE (612) 559-4400 September 18, 1987 l

JKS-87-057 Ms. Laura McCarten Northern States Power Company 414 Nicollet Mall IN04 Minneapolis, MN 55401 Subj ect: Evaluation of Spent Fuel Racks for Prairie Island Rod Consolidation Demonstration Proj ect

Dear Laura:

The purpose of this letter is to summarize the results of a recently completed canister / storage tube impact evaluation and the reasons why the impending rod consolidation project continues to satisfy the requirements of the USNRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978. More detail on these subj ects is provided in the enclosure to this letter.

The current licensing basis for the 7x8 spent fuel storage racks is the capability to store 56 spent fuel assemblies. This means that the racks are designed to transfer all dead weight and seis-mic loads associated with the 56 fuel assemblies to the floor of the spent fuel pool while maintaining stresses in the rack within the limits defined by the OT position referred to in the previous paragraph. The consolidation project will store in one 7x8 rack a maximum total mass no greater than that represented by 56 spent fuel assemblies. Therefore, the overall mass and dead weight for the demonstration project is bounded by the current licensing basis evaluation of the rack. However, each rack storage tube will contain the fuel rods from two assemblies inside a single canister. The licensing basis evaluation considers only one fuel assembly per tube. Because of this, it was necessary to calcu-late the local impact stresses in the wall of the consolidated canister that would be caused by a postulated Safe Shutdown Earthquake (SSE), since the licensing basis evaluation does not consider this particular situation. The maximum stress in the canister wall due to this impact was calculated to be 55% of yield, and satisfies the criteria of the referenced OT Position.

Ms. Laura McCarten _

September 18, 1987 Northern States Power Company JKS-87-057 If you have any questions.or comments on the above, please con--

. tact me.

Very truly yours, kha~ \

J. K., Smith, P.E. 1 Engineering Manager JKS/ma enclosure-I nutech

(,

1

,?

! Enclosure September 18, 1987 Page 1 of 3 JKS-87-057 Evaluation of Prairie Island Spent Fuel Racks The purpose of this evaluation is to demonstrate that the impen-ding rod consolidation demonstration project at the Prairie Island Nuclear Generating Plant satisfies the design criteria stated in the USNRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications",. dated April 14, 1978. The demonstration project will consist of consolidating 50 fuel assemblies into 25 consolidated canisters. The 25 canisters will then be stored in an existing 7x8 spent fuel storage rack.

The canisters will be installed in such a manner as to maintain the center of mass of the spent fuel as close to the center of rigidity of the rack as possible. In addition, the total mass added to the rack used to store the canisters will not exceed the mass of 56 spent fuel-assemblies.

The current licensing basis structural evaluation of the spent i fuel storage racks is documented in report no. QUAD-1-79-509,

" Licensing Report for Prairie Island Nuclear Generating Plant Units 1 and 2 Spent Fucl Storage Modification", dated December, 1979. This report assumes a maximum of 56 fuel assemblies stored in a 7x8 rack. The sliding, dead load, seismic, overturning and rack-to-rack impact analyses documented in this report remain the bounding, licensing basis analyses, since the demonstration pro-ject will be controlled so that the maximum mass added to the spent fuel storage rack will not exceed the maximum added mass used in these analyces. The purpose for minimizing the eccen-tricity between the center of mass and the center of rigidity of the rack is to ensure that the licensing basis analysis remains the bounding analysis for the demonstration project.

The current licensing basis analysis does not address the local impact forces and associated stresses that would occur between the consolidated canister and spent fuel storage tube. Due to a maximum possible gap of .393 inches between the canister and sto-rage tube, additional loads caused by impact of the canister against the storage tube may be generated during a postulated SSE. The referenced OT Position states that "the additional loads.... may be determined by estimating the kinetic energy of the fuel assembly. The maximum velocity of the fuel assembly may be estimated to be the spectral velocity associated with the nat-ural frequency of the submerged fuel assembly. Loads thus gener-ated should be considered for local as well as overall effects on the walls of the rack and the supporting framework. It should be demonstrated that the consequent loads on the fuel assembly do not lead to a damage of the fuel." Based on this, the natural nutech

e l

1 .

l Encl'osure September 18,-1987 Page 2 of 3 JKS-87-057 frequency of the submerged fuel assembly must first~ be calcu-lated. The canister / storage - tube beam model' shown in Figure 1 was used to calculate this natural frequency. The mass of the two fuel assemblies, canister, water, and the storage tube was uniformly distributed along the length of the beam. The stiff-ness (moment of inertia) of the beam is based on the .09" thick inner casing of the storage tube. The stiffness of the canister is neglected. Node points 5 and 1 correspond to the upper and lower grids, respectively. The springs at node 5 model the stiffness of the rack in the north / south and east / west direc-tions. The fundamental mode of vibration.for this model is 7.4 cps. Note that consideration.of the added stiffness provided by the canister would increase this number. .If the global rack stiffness and mass is considered, an overhil natural frequency associated with the rack structure may also be calculated. From Figure 5-2 of NUTECH Report No. NSP-49-101, the first mode of vibration which is important to overall rack behavior occurs at 5.1 cps. This assumes a 7x8 rack with all spaces occupied by consolidated canisters. The maximum total mass carried by a rack in the demonstration project will be 50% of this. Since the frequency is proportional to 1/ mass, the frequency of 5.1 cps would be increased to 5.1/ .5 = 7.2 cps. Since this is slightly lower than what was obtained in the previous paragraph, and will yield slightly higher spectral accelerations and velocities, a natural frequency of 7.2 cps will be used in this evaluation.

The OBE spectral acceleration (SA) and spectral velocity (Sy) are related by the following equation:

S y = S3 /6.28f For a given value of f, S A may be obtained from the appropriate OBE horizontal response spectrum. S, is then calculated from the above equation. For f = 7.2 cps, N y = 1.11 in./sec. A higher  !

value for f would have yielded a lower value for S y. Since SSE l is twice OBE, the SSE S y = 2.22 in./sec. The kinetic energy (KE) i is then:

KE= 1/2 M S y where M is the mass of the canister, two fuel assemblies, and the water inside the canister. For a postulated SSE, KE = 16.6 in-pounds.

The impact loading between the canister and storage tube is then calculated by equating the kinetic energy of the canister to the strain energy of the storage tube, assuming the storage tube to be a simply supported beam with the impact loading being a uni-formly applied load, w, along the length of the storage tube.

nutech

Enclosure September 18, 1987 Page 3 of 3 JKS-87-0 When -the strain energy. of the storage tube is equated ' to_ the kinetic energy of the canister, the impact load w may be solved for.. In this. case, w was found to be'5.67 pound / inch. Based on this impact loading, a maximum thru-wall' bending stress of 13.5 ksi was calculated for the consolidated' canister. Since the yield stress for A240 TP304 sgainless steel is 24.7 ksi at a max-imum pool temperature of ' 212 F , the maximum canister stress due to impact is 55% of yield.

Because'the stresses due to impact in the canister are substan-tially below the yield stress, the canister will continue to perform its function of confining the spent fuel pins as required by the 0T Position.

Based on the above, the maximum storage tube bending stress due to a postulated SSE,. including the effects of canister impact, is ,

2507 psi, or 10% of the yield stress of 24,700 psi.

nutech

_m._ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ . _ _ , _ - _ _ . . _ - _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. N 4

I Y

o

... j

. s wn--; '

1 e, o Gi l x

r- 2 CANISTER / STORAGE TUBE BEAM MODEL

\

Figure 1 i

1 e

- - - - - - - - - - - _ - ~