ML20212P287

Responds to Requesting Interviews W/Degrassi & Philippacopoulos of BNL & Availability for Upcoming Hearing Re Spent Fuel Pool Rerack Application.Degrassi Will Be Presented at Hearing as Witness.Related Correspondence

ML20212P287
Person / Time
Site:	Diablo Canyon
Issue date:	03/10/1987
From:	Chandler L NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:	Lowry E GRUENEICH, D.M. (FORMERLY GRUENEICH & LOWRY)
References
CON-#187-2785 OLA, NUDOCS 8703160055
Download: ML20212P287 (245)

Text

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" UNITED STATES 7 pN NUCLEAR REGULATORY COMMISSION 00LMETED g ,., WASHINGTON, D. C. 20665 USNRC

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v March 10,1987 87 MR 12 P1 :48 Eowin F. Lowry, Esq.

Grueneich & Lowry gFFI g BRANCH 345 Franklin Street San Francisco, CA 94102 In the Matter of PACIFIC GAS AND ELECTRIC COMPANY (Diablo Canyon Nuclear Power Plant, Units 1 & 2)

Docket Nos. 50-275 OLA and 50-323 OLA (Spent Fuel Pool)

Dear Mr. Lowry:

This is in response to your 1ctter of February 17, 1987, which requests that the Sitff make Messrs. DeGrassi and Philippacopoulos of Brookhaven National Laboratory available to you for interviewing and produce them at the upcoming hearing. You also request " copies of all documents that relate to the decision not to recommend approval at this time of the Byron reracking plan, as well as the licensee's application and proffered justification of its .,

proposal." These matters were briefly discussed with you on ' February 20 and 21,1987.

As you recognize, Messrs. DeGrassi and Philippacoupolos are serving as consultants to the Staff in connection with the review of the Byron Station spent fuel pool rerack amendment application, as such, they are viewed as NRC personnel in accordance with 10 C.F.R. I 2.4(p). It is, therefore, appreciated that you have not contacted them directly but rather have sought to make arrangements with Jay and I as counsel for the Staff.

The Staff has considered your request for the participation of Messrs. DeGrassi and Philippacoupolos in the context of its efforts to resolve the multi-rack impact issue as it bears on the Diablo Canyon rerack application. As I informed you on March 4,1987, the Staff has decided that, in order to avoid any question as to the adequacy of the record in this proceeding, it will present Mr. DeGrassi as a witness at the hearing in j

addition to the other Staff witnesses previously identified. Mr. DeGrassi is i

the principal structural reviewer assigned by Brookhaven to the review of the Eyron rcrack application. As such, it is expected that he would be able to he responsive to any questions you might raise at the upcoming Diablo Canyon hearing, to the extent they are relevant to this proceeding. In i addition, in regard to your request for an interview, the staff will make

Mr. DeGrassi available for a brief conference call at a mutually agreeable 1 time. As we discussed , your expectation is that the call would be i approximately one half hour in length. Counsel for PG&E will also be invited to participate.

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, Edwin J. Lowry, Esq. March 10,1987 With respect to the Byron rerack application, your letter reflecta a misunderstanding of"the status of the Staff's review. The Staff has not

" returned as inadequate the modeling and other justification advanced by Coramenwealth Edison and its consultant, Sargent and Lundy", nor has the Staff made a " decision not to recommend approval at this time of the Byron rcracking plan ," as your letter states. Rather, the Staff has simply requested that the licensee provide additional information in support of its license amendment application, as routinely occurs in conjunction with Staff review efforts when questions arise. This should not be construed as a decision not to approve the Byron rerack proposal, the review of which is continuing.

Insofar as you request documents pertaining to the Byron application, the Staff's Board Notification, DN 87-01, dated January 29, 1987, appended a copy of the Staff's cuestion to Commonwealth Edison Company regarding the multi-rack impact issue. In addition, in response to your telephone request of February 10, 1987, Jay has already sent to you several documents bearing on the multi-rack impact issue, by letter dated February 12, 1987. While not conceding relevance, as a courtesy, I am enclosing a copy of the Byron rerack amendment application dated September 3, 1986 and Attachment D thereto, entitled " Licensing Report on High Density Spent Fuel Racks for Byron Unita 1 and 2 ... " dated August 1986, as well as Attachments A and C; e letter from K. A. Ainger, Commonwealth Edison Ocmpany (CECO) to ^

!!arold Denton, NRC, dated November 7,1986; a letter from K. A. Ainger, CECO, tc Parold Denton, NRC, dated November 24, 1986; a letter from

Leonard Olshan, NRC, to D.L. Farrar CECO, dated November 25, 1980; a letter from K. A. Ainger, CECO, to Harold Denton, NRC, dated December 11,

' 1986; a ! citer from Leonard Olshan, NRC, to D.L. Farrar, CECO, dated February 17, 1987; and, a letter from Leonard Olshan, NRC, to D.L.

Farrar. CECO, deted February 25, 1987.

Sincerely, cac Lawrence J. Chandler wA Special Litigation Counsel Enc!csures: As stated cc w/ encl.: B. Faul Cotter, Jr.

Olenn O. Bright, Esq.

Dr. Jerry !! arbour Bruce Norton, Esq.

Richard Locke, Esq.

Dr. Richard Ferguson Dlan !\1. Grueneich, Esq.

cc w/out encl.: Remainder of Service List

! o e i . Commonwealth Etligen b* - }'7 Address Reply to: Post Omco Boa 757 "5 one Fire Nabonel Mesa. CheeGo. Enos Chicago,IEnos 80000 0767 September 3, 1986 Mr. Harold R. Denton, Director office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Byron Station Units 1 and 2 Application for Amendment to Pacility operating License NPF-37, Appendix A, Technical Specifications NRC Docket Nos. 50-454 and 50-455

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Dear Mr. Denton:

Pursuant to 10 CFR 50.90, commonwealth Edison proposes to amend Appendix A, 2echnical specifications, of Facility operating License NFF-37. .

The proposed change, which involves an increase in the storage capacity of the spent fuel racks, is contained in Attachment A. A licensing report regarding the proposed change is contained in Attachment B.

Chapter 8 of the licensing report describes an inservice surveillance program to verify the integrity of the neutron absorber material utilized in the high density fuel racks. In addition to this .

program, neutron attenuation measurements will be made on the neutron absorber material during installation of the new racks in the spent fuel

. pool.

The proposed amendment has been reviewed and approved by both on-site and off-site review in accordance with Commonwealth Edison Company procedures. We have reviewed this proposed amendment in accordance with 10 CFR 50.92(c) and determined that no significant hazards considerations exist.

Our evaluation is documented in Attachment C.

Prompt NRC review and approval of this proposed license amendment is requested. Sixty-four new fuel assemblies for Byron Unit 2 initial core load are being stored in the current spent fuel racks. In the event Byron Unit 2 fuel load is delayed, these Unit 2 fuel assemblies will remain in the spent fuel racks. Storage of this Unit 2 fuel in the spent fuel racks has the potential of impacting the installation schedule of the new spent fuel racks.

• According to the present delivery schedule for the new spent fuel racks, installation of the new racks could proceed until about December 1.

1986. At that point, if Unit 2 fuel remained storrsd in the present spent fuel racks, further installation of the new racks would be suspended.

Approval of this proposed license amendment by December 1, 1986 would permit movement of the Unit 2 new fuel into the new spent fuel racks. Installation limitations on the remainder of the new spent fuel racks would thereby be eliminated.

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Mr. H. R. Denton September 3, 1986

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! Maintaining the present installation schedule of the new spent fuel racks is important to Commonwealth Edison because it will minimize the possi-bility of contaminating the present spent fuel racks. Unit 1 spent fuel will be placed in the spent fuel pool during the Unit I refueling outage which is presently scheduled for early 1987. Disposal of contaminated racks, as well as installing the new racks under wet conditions, would result in additional costs to commonwealth Edison. For these reasons, NRC approval of this proposed amendment is requested by December 1, 1986.

Commonwealth Edison is notifying the State of Illinois of our application for this amendment by transmitting a copy of this letter and its attachments to the designated State official.

In accordance with 10 CFR 170, a fee remittance in the amount of

$150.00 is enclosed.

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_ l Please direct any questions you may have concerning this matter to this office.

Three (3) signed originals and thirty-seven (37) copies of this letter and attachkents are provided for your use.

Very truly yours, gf#'

K. A. Ainger Nuclear Licensing Administrator 1m Enclosure Fee Remittance Attachments (A): Proposed Technical Specification Change (B): Licensing Report

, (C): Evaluation of Significant Hazards Considerations cc: Resident Inspector - Byron L. N. 01shan - NRR M. C. parker - State of Ill.

SUBSCRIDED AMD SWORN to before!mejis ~'d day of .ar f, +- , 1986 "Y$ s. .

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Notary Public 2058K i

4.-

ATTACHMENT A PROPOSED CHANGE TO APPENDIX A TECHNICAL SPECIFICATIONS OF

[ fAqlkITY OPERATING LICENSE NPF-37 Revised Pace: 5-5 Y

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r-DESfGN FEATURES i

5. 5 FUE!. STORAGE CRITICALITY

5. 6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. k,ff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance e meer for uncertainties  ;

and i

l b. >lA ncmio.; l' i r - a*=r-to-center distance betw..n %e :::;.;;,; ;,,

IOacid " tt.; n w .we racxs.

5. 6.1. 2 The k,ff for new fuel for the first core loading stored dry in the i

- I

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spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is

• assumed. i DRA!NAGE ,, ,- ,

5.6.2 The scent fuel storage pool is designec and shall be maintained to -

prevent inaovertent draining of the pool below elevation 423 feet 2 inches.

! CAPACITY I S.6.3 The scent fuel storage pool is designed and shall be maintained with a l storage caeacity limited to no more than 38(7 fuel assemblies.

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5.7' COMPCNE'4T CYCLIC OR "'RANSIENT LIMIT

5. 7'.1 The commenents ioentifiec in Taole 5.7-1 are designed anc shall be maintainec sitnin the cycite or transient limits of Table 5.7-1.

l A nominal 10.32 inch north-south and 10.42 inch east-west, center-to center distance between fuel assemblies placed in Region 1 spent fuel storage racks and a nominal 9.011 inch center-to-center distance between fuel assemblies placed in

! Region 2 spent fuel storage racks. -

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! SYRON - UNITS 1 1 : 5-!

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ATTACOSNT 5 D 6 e ,

a LICENSING REPORT 6 4 6

0 2050K

3

) ATTACISIENT C IVALyATION OF SIGNIFICANT HAZARDS CONSIDERATIONS commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations.

According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The proposed amendment involves the replacement of the present spent fuel racks with high density racks. This will increase the spent fuel storage capacity at afron Station from 1060 to 2940 fuel assemblies including six locations for failed fuel storage.

Accidents previously evaluated for radiological consequences are a spent fuel assembly drop onto the spent fuel pit floor and a spent fuel cask drop. Abnormal conditions which have been previously evaluated with respect

. to. potential for criticality are (1) a dropped fuel assembly laying across the, top of a fuel rack and (2) a fuel assembly in transport accidently dropped into a position parallel with stored fuel in the most reactive co'rner of the racks. The effect of a seismic event was also evaluated with respect to potential for criticality.

All of these events could occur independent of the design and installed configuration of the spent fuel racks. As a result, the probabi-lity of these events occurring is not affected by replacement of the spent fuel racks.

The consequences of the previously evaluated events have been evaluated considering the new spent fuel racks. Our review has concluded that the criticality acceptance criterion of maintaining X,gg less than or equal to 0.95 will not be excoeded for these events.

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The radiological consequences were also evaluated to determine the I impact on offsite and onsite doses previously determined. The increase in the storage capacity of the spent fuel pool will neither significantly alter the operating characteristics of the current spent fuel pool nor result in a measurable change in impact on the environment. The design basis fuel handling accidents, described in FSAR Section 15.7.4, were reviewed for possible effects on radiological dose consequences. The review determined that the conclusions in the FSAR will remain valid and that offsite radiological dose consequences will remain within 10 CFR 100 limits. As a result, the consequences of previously evaluated events will not significantly increase as a result of replacing the spent fuel racks.

With respect to the second standard of 10 CFR 50.92(c), the new spent fuel racks only allow closer spacing of the fuel assemblies. No new or different kind of accidents result from this.

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Although the new spent fuel racks contain two storage regions, the placement of a fuel assembly in the wrong region is not a different kind of accident from the placement of a fuel assembly parallel to stored fuel in the most reagtive. corner among the current spent fuel racks. Both of these accidents fall within the category of abnormal placement of a fuel assembly. For these reasons, the replacement of the spent fuel racks does not create the possibility of a new or different kind of accident from any accidents previously evaluated.

Installation of the new spent fuel racks will not result in a significant reduction in a margin of safety as contemplated by the third standard of 10 CFR 50.92(c). A small increase in the spent fuel pool heat' load is expected due to the storage capacity expansion. However, the spent

, fuel pool cooling system design can handle the increased heat load and maintain the temperature peaks of the pool below design values. Installation of the new spent fuel racks will also result in a small increase in the pool re' activity as measured by the neutron multiplication factor (K,gg).

However, the maximum neutron multiplication factor will be maintained less than or equal to 0.95. For these reasons, increasing the spent fuel pool storage capacity will not significantly reduce a margin of safety.

The NRC has provided guidance concerning the application of the standards of 10 CFR 50.92(c) for determining whether significant hazards considerations exist by providing examples (48 FR 14870 and 51 FR 7751) of amendments that are considered not likely to involve significant hazards considerations. Example (x) relates to an expansion of the storage capacity of a apent fuel pool when all of the following are satisfied:

(1) The storage expansion method consists of either replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies or placing additional racks of the original design on the pool floor if space permits,

g (2) The storage expansion method does not involve rod consolidation or double tiering.

(3) T,he K,gg of the pool is maintained less than or equal to 0.95, and (4) No new technology or unproven technology is utilized in either the construction or the analytical techniques necessary to justify the expansion.

The proposed reracking of Byrcn Station will replace the existing racks with a design which allows closer spacing between stored spent fuel assemblics and does not involve rod consolidation or double tiering. The K,gg will be maintained less than or equal to 0.95. Similar analyses and construction techniques have been used previously for plants that have

- l' licensed high density spent fuel racks. These plants include Fermi 2, Quad

-- Cities 1 and 2, Rancho Seco, Oyster Creek, Virgil C. Summer, and Diablo Canyon 1 and 2. Therefore, no new technology or unproven technology has been util,12ed..in this case. .

For all the reasons stated above, Commonwealth Edison believes the proposed installation of high density spent fuel racks does not involve any significant hazards considerations.

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e 2058K

i LICENSING REPORT a

.ON HIGH DENSITY SPENT FUEL RACKS FOR BYRON UNITS I AND 2 -

NRC DOCKET NO. 50 - 454 -

50 - 455 I =

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j. COMMONWEALTH EDISON COMPANY l 1.

CHICAGO, ILLINOIS 60603

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i AUGUST,1986 1

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i -- .- . . _ - - - - - . - - - . - _ . - - . - . . - - - - - - . - . . - _ - . - . - - . - - . . -

. TABLE OF CONTENTS

. r Section ', , g -

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1.0 INTRODUCTION

1-1 , i l 2.0 CENERAL ARRANGENENT 2-1

... 3.0 RACK CONSTRUCTION 3,-t 3.1 Fabrication Details 3-1

. 3.1.1 Region 1 3-1 l l'*

3.1.2 Region 2 3-3

! 3.2 Codes, Standards, and Practices for the

, a- Spent Fuel Pool Nodification 3-4 l

)\ 4.0 4-1 NUCLEAR CRITICALITY ANALYSIS

! i.

i 4.1 Design Bases 4-1 11 4.2 Summary of Crittoa11ty Analyses 4-3

> [, 4.2.1 Normal Operating Conditions 4-3 j u 4.2.2 Abnormal and Aeoident Conditions 4-7 4.2.3 New Fuel Storage 4-8

{ 4.3 Reference Fuel Storage Cell 4-10 i.* 4.3.1 Reference Fuel Assembly 4-10 i ,. 4.3.2 Region 1 Storage Cells 4-10 lp 4.3.3 Region 2 Storage Colls 4-10 l

j 4.4 Analytloal Nethodology 4 15

..f 4.4.1 Reference Analyttoal Nethods and Bias 4.4.2 Fuel Burnup Calculations 4-15 4-17

. 4.4.3 Effect of Antal Burn-up Distribution 4 21

- 4.4.4 Long-term Decay 4 22 l- 4.5 Region 1 Critloality Analysis and Tolerance Variations 4 24

l. , 4.5.1 Nominal Design Case 4-24 8

4.5.2 Boron Leading Variation 4-24 4.5.3 Storage Cell Lattice Pitch Variation 4-25 J' 4.5.4 Stainless Steel Thickness Tolerances- 4-24

,, 4.5.5 Fuel Enrichment and Density Variation 4-24 Boraflex Width Tolerance Variation 4.5.4 4-24

,. 4.5.7 Axial Cutback of Sorafien 4-27 4.4 Region 2 Critloality Analysis and Tolerance Variations 4-24 1

4.4.1 Nominal Design Case 4 28 i 4.4.2 Boron Loading Variation 4 29 4.4.3 Storage Cell Lattice Pitch Variations 4-30 4.4.4 Stainless Steel Thickness Tolerance 4-30 1

• TABLE OF CONTENTS (Continued) ,

1 Section

• f.agg 1

4.4.5 Fuel Enrichment and Density Yariation 4-30 4.4.4 Boraflex Width Tolerance 4-30 -

4.7 Abnormal and Accident Conditions 4 31 }

4.7.1 Eccentric Positioning of Fuel Assembly 4-31 9 3 in Storage Rack 4.7.2 Temperature and Water Density Effects 4-31 7 l 4.7.3 Dropped Fuel Assembly Accident 4-32 1 4.7.4 Abnormal Location of a Fuel Assembly 4-34 4.7.5 Lateral Rack Movement 4-35 i, 4.8 New Fuel Storage 4-34 i

4.8.1 Storage in Region 1, Dry 4-34 4.8.2 Storage in Region 2, Flooded 4-34 f

4.8.3 Storage in Region 2, Dry 4-37 4[

References to Section 4 -

4-38 *u

[ 5.0 THERMAL-NYORAULIC CONSIDERATIONS 5-1 5.1 Decay Heat Calculations for the Spent Fuel 5-1 '

5.1.1 Sasis 5-1 -

j 5.1.2 Model Description 5-3 ,,

j S.1.3 Decay Heat Calculation Results 5-4 i

i 5.2 Thermal-Hydraulio Analyses for Spent Fuel 5-7

, Cooling t;

! 5.2.1 Sasis 5-7 l . 5.2.2 Model Description -

5-4 i

5.2.3 Results 5-7 .

References to Section 5 5-11 -

! 4.0 STRUCTURAL ANALYSIS d-1 I'

! 4.1 Analysis Outline 4-1 l l .

. I d.2 Fuel Rack Fuel Assembly Model 4-3 l 4.2.1 Outline of Model **

i 4-4 4.2.2 Model Description 6-4 t,  ;

i 4.2.3 Fluid Coupling 4-7 ,,

i 4.2.4 Damping 4-8

• j d.2.5 Impact 4-8 ,

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{ d.) Assembly of the Dynamic Model ,

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--,,.,,,-,..,_,n_,,_-, ,n.,_, , _ _ _ - , , , , , ,,,,w-n._,_ _

TABLE OF CONTENTS (Continusu)

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Section Page i

4.4 Time Integration of the Equations of Motion 4-11 4.4.1 Time-History Analysis Using 14 00F

i. Rack Model .

4-11 4.4.2 Evaluation of Potential for Inter-Rack 4-13 Impact

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4.5 Structural Acceptance Criteria 4-13 4

4.4 Material Properties 4-15

] 4.7 Stress Limits for Various conditions 4.7.1 Normal and Upset Conditions 4-15

,. (Level A or Level 8) 6-15 4.7.2 Level 0 Service Limits 4-18 4.8 Results 4-18

,, 4.9 Impact Analyses 4-21 4.9.1 Impact Loading Between Fuel Assembly

and Cell Wall 4-21 4.9.2 Impacts Between Adjacent Racks 4-21 i . 4.10 Weld Stresses 4-22 c

4.11 Sum; nary of Mechanical Analyses . d-22 d.12 Definition of Terms Used in Section d 6-24

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References to Section d 6-24

(. 7.0 ENVIRONMENTAL ANALYSIS 7-1

l l.- 7.1 Summary 7-1 7.2 Characteristics of Stored Fuel 7-1

( Related Indu'stry Experience 7-2 7.3 J.

, 7.4 Syron Nuclear Power Station Experience 7-4 -

7.5 Spent Fuel Pool 'ooling C and Cleanup System 7-4 l

7.4 Fuel Pool Radiation Shielding 7-5 I 7.4.1 Source Terms 7-5 l 7.4.2 Radiation Shielding 7-6 ,

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, , TABLE OF CONTENTS (Continued)

Section Pace 7 .'7 Radiological Consequences 77 7.8 Riracking Operation 7-8 T:

7.9 Conclusions 7-9 .h References to Section 7 7-11 i

'8.0 IN-5ERVICE SU'RVEILLANCE PROGRAM FOR BORAFLEX i NEUTRON ABSORBIHC MATERIAL 8-1 .,

8.1 Program Intent 8-1 8.2, Description of Specimens 8-1

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8.3 Specimen Evaluation 8-2 9.0 COST / BENEFIT ASSESSMENT 9-1 9.1 Specific Needs for Spent Fuel Storage 9-1 9.2 Cost of Spent Fuel Storage -

9-2

. . p j 9.3 Alternatives to Spent Fuel Storage 9-2 9.4 Resource Commitments 9-3 /,

i References to Section 9 9-5

• 10.0 QUALITY ASSURANCE PROGRAM 10-1 .'

10.1 Introduction 10-1 9 J

10.2 Ceneral 10-1 10.3 System Highlights 10-1 10.4 Summary 10-3 ,

1 Appendix A - Benchmark Calculations A-1 ,1 n

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. LIST OF TABLES Table Pace 1.1 Byron Unit 1 & Unit 2 Fuel Assembly Discharge 1-3

., -- (Tentative Schedule) 2.1 Design Data 2-2 l ,

2.2 Module Data 2-3 i ).

3.1 Boraflex Experience for High Density Racks 3-7 4.1 Summary of Criticality Safety Analyses 4-5

<- 4.2 Reactivity Effects of Abnormal and Accident 1

Conditions 4-7 L, 4.3 Fuel Assembly Design Specifications 4-12 .

i l c, 4.4 Comparison of Cold, Clean Reactivities 4-19

' Calculated at 25,000 MWD /HTU Burnup and 3.25 I<* Enrichment 4.5 Estimated Uncertainties in Reactivity due to 4,-21

<. Fuel Depletion Effects 4.4 Long-term Changes in Reactivity in Storage Rack 4-23 (Xenon-Free)

.. 4.7 Fuel Burnup Values for Required Reactivities 4-28 (k.) with Fuel of Various Initial Enrichments 4.8 Effect of Temperature and Void on Calculated 4-32

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Reactivity of Storage Rack 5.1 List of Cases Analyzed 5-12 5.2 Haximum Pool Bulk Temperature, t, Coincident 5-13

• Total Power, Q1, and Coincident Specific Power, l.. q, for the Hottest Assembly 5.3 Time (Hrs) to Boiling and Boiling Vaporization l- Rate From the Instant All Cooling is Lost 5-14

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5.4 Maximum Local Pool Water Temperature and Local Fuel Cladding Temperature at Instant l of Maximum Pool Bulk Temperature 5-15 I

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LIST OF TABLES (Continued) , .

8 Table Page J

5.5 Pool and Maximum Cladding Temperature at the

, Instant Fuel Assembly Transfer Begins- 5-16 .,

6.1 Degrees of Freedom ,

6-27 ,

, 6.2 Humbering System for Cap Elements and Friction 6-28 Elements 1

4.3 Rack Material Data 6-29 I?

4.4 Support Material Data 6-29

f 4.5 Byron Racks - Bounding Values for Stress Factors . 6-30 7.1 Photon Energy Produ'ction Rates of an Average 7-12

• i Spent Fuel Assembly 7.2 Photon Energy Production Rates of Peak 7-13 [

l Spent Fuel Assembly v 7.3 Calculated Dose Rates in Areas Adjacent to 7-14  ! ,

the Spent Fuel Pool , t i

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.. LIST OF FIGURES .

Tigure Page 2.1 Pool Layout 2-4

... 2.2a Typical Rack Elevation - Region 1 ,

2-5 2.2b Typical Rack Elevation - Region 2 2-6

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3.1 3x3 Typ. Array Region 1 3-8 3.2 Channel Element - Region 1 & 2 (2 for Square Cell) 3-9 3.3 Cap Elements Region 1 3-9

., 3.4 Composite Box Assembly . Region 1 '3-10 b 3-11 3.5a Typical Cell Elevation - Region 1

, j' 3.5b Typical Cell Elevation . Region 2 3-12 s

3.6 Adjustable Support 3-13 3.7 3x3 Typical Array Region;2 3-14 4.1 Acceptable Burnup Domain in Region 2 of the 4-6

-4 3

Byron Station Spent Fuel Storage Racks 4.2 Region 1 Storage Cell Geometry 4-13 -

.. 4.3 Region 2 Storage Cell Geometry 4-14 4.4 Comparison of Depletion Calculations for 4-16 Fuel of 3.2% and 4.2% Initial Enrichments

. 4.5 Reactivity Effect of Water Spacing Between 4-33 Fuel Assemblies .

5.1A Spent Fuel Pool Bulk Temperature (0-700 hours) 5-17

(' (Normal' Refueling Discharge) 1.,

5.1B Spent Fuel Pool Bulk Temperature (0-44 days) 5-18

- (Normal Refueling Discharge) l Power Discharged in Spent Fuel Pool (0-700 hours) 5-19 5.1C (Normal Refueling Discharge) l

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LIST OF FIGURES (Continued)

. 3 Figure Pace 5.10 Power Discharged in Spent Fuel Pool (0-44 days) 5-20 (Normal Refueling Discharge) ,

5.2A 5-21 Spent Fuel Pool Bulk Temperature (0-1200 hours) 4, (Full Core Discharge) .

5.28 Spent Fuel Pool Bulk Temperature (0-115 days) 5-22 i

(Full Core Discharge) 7 5.2C Power Discharged in Spent Fuel Pool (0-1200 hours) 5-23

,(Full Core Discharge) 't 5.2D Power Discharged in Spent Fuel Pool (0-115 days) ,

5-24 (Full Core Discharge) .

! 5.3 Idealization of Rack Assembly 5-25 -

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5.4 Thermal Chimney Flow Model 5-26 6.1 Byron Fuel Rack - $$E East / West 6-34 j s'

6.2 Byron Fuel Rack - SSE North / South 6-35 {

6.3 Byron Fuel Rack - SSE Vertical 6-36 .

t 6.4 Dynamic Model 6-37 -

4.5 Cap Elements to Simulate Inter-Rack Impacts 6-38

6.6 Impact Springs and Fluid Dampers 6-39 6.7 Spring Mass simulation for Two-Dimensional Motion 6-40 ,,

8.1 Test Coupon 8-3

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1.0 INTRODUCTION

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• This report describes the design, fabrication, and safety analysis of high density spent fuel storage racks manufactured by Joseph Oat Corporation (Oat) for the Byron Station Unit 1 and Unit 2. The plant, which is located two milss east of the Ro.ck River and approximately three miles southwest of Byron in Ogle County, is owned and operated by Commonwealt'h Edison Company (CECO).

,- Byron is a two-unit pressurized water reactor (PWR) with a net design capacity of 1120 megawatts electric for each unit. Each of

,, the'two reactor cores contains 193 fuel assemblies and is rated to produce 3411 thermal megawatts (HWt). At present, there are no spent fuel assemblies stored in the spent fuel pool. Unit 1 went

[ into commercial operation in September of 1985. Unit 2 is scheduled to go into commercial operation in June, 1987.

The two units share one common spent fuel storage pool which is currently licensed for the storage of 1060 spent fuel

. assemblies. As shown in Table 1.1, the storage pool would lose

. full core discharge capability in 1994. The proposed reracking will increase the number of pool storage locations to 2940 (includes six failed fuel locations). Table 1.1 indicates that the new racks will provide adequate storage ,

with full core discharge capability well into the next century (circa 2011).

Table 1.1 is based on an. estimated 18-month fuel cycle. Current trends toward longer cycles, extended *burnup, and higher q enrichment would further extend the time span of onsite storage.

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. The proposed racks are free-standing and self-supporting. The principal construction materials are ASTM A-240, Type 304L stainless steel for the structural members and shapes, and i "Boraflex," a patented product of BISCO (a division of Brand, Inc.), for neutron attenuation.

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The specifications for design, construction, and quality

- assurance for the high density spent fuel storage racks were prepared by Sargent & Lundy Engineers (S&L) of Chicago, Illinois.

The mechanical design, seismic / structural, analysis, thermal-hydraulic analysis, and other related calculations, and the fabrication of the hardware, were performed by Cat. S&L provided the seismic response spectra and performed the spent fuel pool 7' structure evaluation. S&L performed the radiation shielding -

analysis. Southern Science, a division of Black & Veatch, served ,

as a consultant to Oat in the area of. criticality analysis. The (,

analyses performed by Oat in conjunction with Black and Veatch ,

and S&L demonstrate that acceptable margins of safety exist with ,

respect to appropriat's NRC and ASME acceptance criteria. A cost-benefit comparison of several potential spent fuel disposition alternatives indicates that (1) reracking of the Byron pool is the lowest risk and most cost-effective

alternative, and (ii) that neither the reracking , operation nor ,

4:

the increased onsite storage of irradiated material pose an undue i hazard to the plant staff or the public. .

The following sections provide a synopsis of the design, fabrication, nuclear criticality analysis, thermal / hydraulic 0

analysis, structural analysis, accident analysis, environmental analysis, and cost-benefit appraisal of the high density spent fuel racks. In particular, the integrity of the rack structure under ,the specified combinations of inertial, seismic, and ~

mechanical loads and . thermal gradient per NUREC-0800 is demonstrated. .

Also included are descriptions of the rack In-Service Surveillance Program and the Oat Quality Assurance Program. This .

l Quality Assurance Program has been reviewed and found acceptable '

for engineered fabrication of ASME Section III, Class 1, 2, 3 and MC Components by both ASME and the NRC.

~

1-2

_ _ . - - _ - = .

[- ,

I

~ Table 1.1

(' BYRON UNIT 1 & UNIT 2 FUEL ASSEMBLY DISCHARGE c (TENTATIVE SCHEDULE)

,l Remaining Remaining Total Discharged Storage Storage

!< Assemblies in Capability Capability

~

Number of Spent Fuel Pool Without With Assemblies Following Proposed Proposed Refueling Date Discharged Refueling Expansion Expansion l

i February 1987 88 88 972 2852 Unit #1 '

August 1988 88 176 884 2764 l~ Unit #1

~~

2A 796 2676 Unit #2 December 1988 88 88 352 708 2588

<~

Unit #1 , February. 1990 88 440 620 2500 Unit #2 Dune 1990 528 532 2412

.- Unit #1 August 1991 88 Unit #2 December 1991 88 616 4% 2324 February 1993 84 700 360 2240 Unit #1 788 272 2152 Unit #2 June 1993 88 Unit #1 August 1994 84 872 188** 2068 1984 i Unit #2 December 1994 84 956 104 Unit #1 February 1996 84 1040 20

• 1900 84 1124 1816 Unit #2 June 1996 84 1208 1732 l

Unit #1 August 1997 84 1292 - - 1648 -

g-Unit #2 December 1997 s.

l l .-

Il l

• Partial core discharge capability lost - 84 assemblies

• • Full core discharge capability lost - 193 assemblies

)

I l; '

1-3 l1 l

l. - - - - --

s 3

l Table 1.1 (Continued)

BYRON UNIT 1 & UNIT 2 FUEL ASSEMBLY DISCHARGE "

(TENTATIVE SCHEDULE)

'{,

Remaining Remaining Total Discharged Storage Storage '

, Assemblies in Capability Capability ,,

Number of Spent Fuel Pool Without With

, Assemblies Following Proposed Proposed Refueling Date Discharged Refueling Expansion Expansion Unit #1 February 1999 84 - 1376 15A -

Unit #2 June 1999 84 1MO 1480 .

Unit #1 August 2000 84 1544 1396 .

Unit #2 December 2000 84 - 1628 1312 ..

Unit #1 February 2002 84 1712 1228

• Unit #2 ~ June 2002 M 1796 11 %

Unit #1 -August 2003' 84 1880 1060 Unit #2 December 2003 84 1964 976 Unit #1 February 2005 84 2048 892 '

Unit #2 Uune 2005 84 2132 808 -

Unit #1 August 2006 84 2216 724 ..

Unit #2 December 2006 84 2300 640 .

Unit #1 February 2006 84 2384 556 ,,

Unit #2 June 2008 84 2M8 472 Unit #1 August 2009 84 2552 388 Unit #2

~

December 2009 84 2636 .

. 304

! Unit #1 February 2011 84 2720 220 'l l Unit #2 June 2011 84 2804 136*

• Unit #1 August 2012 84 2888 52' O

e l

1-4

)

2.0 GENERAL ARRANGEMENT

- The high density spent fuel racks consist of individual cells -

4 .w ith 8.85-inch by 8.85-inch (nominal) square cross-section, each of which accommodates a single Westinghouse PWR fuel assembly or equivalent. A total of 2940 cells are arranged in 23 distinct modules of varying sizes in two regions. Region 1 is designed for storage of new fuel assemblies with enrichments up to 4.2 weight percent U-235. Region 1 is also designed to rtore fuel l

assemblies with enrichments up to 4.2 weight percent U-235 that

'have not achieved adequate burnup for Region 2. The Region 2

- cells are capable of accommodating fuel assemblies with various

- initial enrichments which have accumulated minimum burnups within l , an acceptable bound as depicted in Figure 4.1. Figure 2.1 shows l

the arrangement of the rack mo'dules in the spent fuel pool.

The high density racks are engineered to achieve the dual objective of maximum protection against stru.ctural loadings

~

(arising from ground motion, thermal stresses, etc.) and the maximization of available storage locations. In general, a greater width-to-height aspect ratio provides greater margin

.. against rigid body tipping. Hence, the modules are made as large as p%ssible within the constraints of transportation and site handling capabilities.

As shown in Figure 2.1, there are 23 discrete modules arranged in the fuel pool. Each rack module is equipped (see Figures 2.2a f and 2.2b) with girdle bars, 5/8-l'nch-thick for Region 1 I

(1/4-inch-thick for Region 2) by 3-1/2 inches high. The nominal gap between adjacent modules is 1-1/4 inches for Region 1 and

E 3/4-inch for Region 2. The modules make surface contact between their contiguous walls at the girdle bar locations and thus maintain a specified gap between them. Table 2.1 gives the relevant design data on each region. The modules in the two

( Table 2.2 summarizes the j regions are of 11 different types.

physical data for each module type.

ll

~< . - - . - . .

)

E. e

,I Table 2.1 DESIGN DATA _

.f (Cell Pitch) Flux Trap )

Nominal' Min. B-10 Cap (Nominal

?

Region in. Loading in.

1 10.32 NAS .020 gm/cm 2 1.16 ,

& 10.42 E&W 1.26

, 2 9.011 .010'gm/cm* 0.0 .

e h

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,3 Table 2.2 I. MODULE DATA i 1 *'

Approximate Module Number of Cells per Module Weight Region Type Modules Module Size (1b/ module)

J.

, 1 A1 1 104 ,

13x8 20,800.0 1 B1-3 3 96 12x8 19,200.0 2 C1-8 8 168 14x12 26,900.0 4 .

2 D1-3 3 126 14x9 20,150.0

.. 2 04 1 116 14x9 18,550.0

l. -(2x2+3x2) f 2 D5 1 114 14x9-(4x3) 18,250.0 2 E1-2 2 112 14x8 17,900.0 2 F1 1 132 11x12 21,100.0

. 2 C1 -

1 . 143 11x13 22,900.0 l b.

2 M1 1 e56 7x8 8,950.0 t,

}. ~I 2 31 1 35+6 7x5 10,150.0

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FIGURE 2.2a TYPICAL RACK ELEVATION-REGION 1

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FIGURE 2.2b TYPfCAL RACK ELEVATION-REGION 2 -

2-6 I

l .

i 3.0 RACK CONSTRUCTION i .

3.1 FABRICATION DETAILS 3.1.1 Region 1

<. The rack module is' fabricated from ASTM A-240-304L austenitic

. stainless steel sheet and plate material, SA351CF3 and SA217CA15

.. casting material. The weld filler material utilized in body welds is ASME SFA-5.9, Type 308L and 308LSI. Boraflex serves as the neutron absorber material. The detailed neutronic properties of The,Boraflex' experience list Boraflex may be found in Section 6, l

is given in Table 3.1. .

A typical module contains storage cells which have an 8.85-inch l.. nominal square cross-sectional opening. This dimension ensures 7

. that fuel assemblies with maximum expected axial bow can be inserted and removed from th's storage cells without any damage to the fuel assemblies or the rack modules.

f Figure 3.1 shows a horizontal cross-section of a 3 x 3 array.

!' The cells provide a smooth and continuous surface for lateral contact with the fuel assembly. The anatomy of the rack modules is best explained by describing the components of the design, namely l' 0 Internal square tube l[

O Neutron absorber material (Boraflex)

l. O Poison sheathing 0 Cap element i )-

t l 0 Baseplate

,1 e

0 Support assembly

. O Top lead-in 3-1 .

I _ - _ . _

. )

)

a. Internal Square Tube -

i This element provides the lateral bearing surface to the fuel ass'embly. It is fabricated by joining two formed ,'

channels (Figure 3.2) using a controlled seam welding operation. This element is an 8.85-inch square 1 (nominal) cross-section by 168-7/8 inches long.

I Neutron Absorber Material (Boraflex)

~

b. ,

Boraflex is placed on all four sides of a square tube 3,t over a length of 139-1/2 inches, which covers the active '

fuel length except the top and bottom 3 inches, for Region 1. For Region 2 the Boraflex length is 144 s 4

inches and it covers the entire length of the . active i fuel length. D ci. Poison Sheathing J' The poison sheathing (cover plats)', shown in Figure 3.5, i serves to position and retain the poison material in its -

i designated space. This is accomplished by spot welding  !'

4 the ' cover sheet to the square tube along the former's edges at numerous (at least 20) locations. This manner .

of attachment ensures that the poison material will not

! sag or laterally displace during fabrication processes I and under any subsequent loading condition.

It

d. Cap..' Element i

Cap . elements, illustrated in Figure 3.3, position two n inner boxes at a predetermined distance to maintain the l ,

minimum flux trap gap required between two boxes. The '

4 gap element is welded to the inner box by fillet welds. ",

An array of composite box assemblies welded as indicated '

I in Figure 3.1 form the honeycomb.gridwork of cells which &

harnesses the structural strength of all sheet and plate -

type members in an efficient manner. The array of composite boxes has overall bending, torsional, and -

axial rigidities which are an order of magnitude greater >

than configurations utilizing . grid bar type of .

construction. .

i

e. Baseplate , , ,.

The baseplate is a 5/8-inch thick plate type member .

which has 6-inch diameter holes concentrically located I with respect to the internal square tube, except at l support leg locations, where the hole size is 5 inches ,' ,

in diameter. These holes provide the primary path for coolant flow. Secondary flow paths are available .

between adjacent cells via the lateral flow holes (1 inch in diameter) near the root of the honeycomb ~'

(Figures 3.5a and 3.5b) which preclude flow blockages.

The honeycomb is welded to the baseplate with 3/32-inch .fI fillet welds. l

. 3-2 [

'k

l. . '

f. Support Assembly

,i The supports are l

Each module has four support legs.

adjustable in length to enable levelling of the rack.

support assembly consists of a The variable height rides- into an internally-t-

flat-footed spindle which

,, threaded cylindrical member. ofThe cylindrical member is the baseplate through attached to the underside The base of the fillet and partial penetration welds.

flat-footed. spindle sits on the pool floor. Levelling of is accomplished by turning the square the rack modulesin the spindle using a long arm (approximately sprocket 46 feet long) square head wrench. Figure 3.6 shows a l .

the adjustable support d.

vertical cross-section of assembly.

1*

The supports elevate the module baseplate approximately 7-1/2 inches above the pool floor, The thus lateral creating holes int.hthe e water plenum for coolant flow. the coolant entry path

~-

member provide cylindricalleading into the bottom of the storage locations.

I J. g. Top Lead-in 4

' Lead-ins are provided on each cell to facilitate fuel q assembly insertion.* The lead-ins of structurally contiguous at

. connected wallsthe of.

adjacent cells are in reducing the These lead-in joints aid lead-in.

lateral deflection of the inner square tube due to the during the ground motion l

impact of fuel assemblies This l .

(postulated seismic motion specified in the FSAR).

type of construction leads to natural venting locations I

for the inter-cell space where the neutron absorber material is located.

l' 3.1.2 Recion 2 .

(- same modules in Region 2 are fabricated from the The rack i.e., ASTM A-240-304L q material as that used for Region 1 modules,

- austenitic stainless steel.

A typical Region 2 module storage cell also has an 8.85-inch 3.7 shows a l Figure cross-sectional opening.

nominal square horizontal cross-section of a 3 x 3 array. The rack construction varies from that for Region 1 in as much as the stainless steel 4

3-3 L

I

I;

- cover plates, gap elements, and top lead-ins are eliminated.

Hence, the basic components of.this design are as follows:

~ '

O Inner tube .

O Neutron absorber material 0 Side strips .

0 Baseplate ,

O Support assembly .,

In this construction, two channel elements form the cell of an 8.85-inch nominal square cross-sectional opening. The poison material is placed between two boxes as shown in Figure 3.7. 't 6

Stainless s t e e l "s i d e strips are inserted on both sides of the poison material to firmly locate it in the lateral direction.

The bottom strip positions the poison material in the vertical '

direction to envelope the entire active fuel length of a fuel ,

assembly (Figure 3.5b). Two adjacent boxes and the side strip ,

between boxes are welded together as shown in Figure 3.7, to form the honey-comb rack module.

9 1 1 .

t .

The baseplate and support assemblies are incorporated in l

exactly the same manner as described for Region 1 in the' -

preceding section. q 3.2 CODES, STANDARDS, AND PRACTICES FOR THE SPENT FUEL POOL i MODIFICATION 1 -

The fabrication of the rack modules is performed under a strict .

. quality assurance system suitable for ASME Section III, Class 1, _

2, and 3 manufacturing which has been in place at Oat for over 10 ,,

years.

The following codes, standards, and practices were used as

~'

applicable for the design, construction, and assembly of the spent fuel storage racks and analysis of the pool structure. '.

Additional specific references related to detailed analyses are -

given in each section. .

3-4 i j

i

c 1

1

h. .

I a. Design Codes 1

l (1) AI5bManualofSteelConstruction,8thEdition, 1980.

i .

a, (2) ANSI N210-1976, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear

.- Power Stations."

American Society of Mechanical Engineers (ASME), '

(3)

(

Boiler and Pressure Vessel Code,Section III, 1983 Edition up to and including Summer 1983 Addenda

< (Subsection NF).

(4) ASNT-TC-1A 3une, 1980 American Society for l i' Nondestructive Testing (Recommended Practice for Personnel Qualifications).

l

b. Material Codes '

(1) American Society for Testing and Materials (ASTM)

Standards - A-240.

l (2) American Society of Mechanical Engineers (A5ME),

r Boiler and Pressure Vess'el Code,Section II - Parts A and C, 1983 Edition, up.to and including Summer l l*

1983 Addenda.

I c.

c. Welding Codes I

A5ME Soiler and Pressure Vessel Code, Section IX ' -

i' Welding and Brazing Qualifications, 1983 Edition up to and including Summer, 1983 Addenda.

Assurance, Cleanliness, Packaging, Shipping,

d. Quality l- Receiving. Storage, and Handling Requirements N45.2.2 Packaging, Shipping, Receiving, (1) ANSI - .

'. Storage and Handling of Items for Nuclear Power I. . Plants. . .

of Fluid Systems and I

(2) ANSI 45.2.1 - Cleaning t'

Associated Components during Construction Phase of Nuclear Power Plants.

and Pressure Vessel, Section V, (3) A5ME Boiler

} Hondestructive Examination, 1983 Edition, including Summer and Winter 1983.

t N16.1-75 Nuclear Criticality Safety

! (4) ANSI -

Fissionable Materials Outside Operations with

. Reactors.

}

I 3-5 r

1 1

- (5) ANSI - N16.9-75 Validation of Calculation Methods for~ Nuclear Criticality Safety. 7 4

(6) ANSI . N45.2.11, 1974 Quality Assurance Require-ments for the Design of Nuclear Power Plants.

3

e. Other References I

(1) HRC Ragulatory Guides, Division 1, Regulatory I' Guides 1.'43, Rev. 2 (proposed): 1.29, Rev. 3; 1.31, ..

Rev. 3 '; 1.61, Rev. Os 1.71, Rev. Os 1.85, Rev. 22; 1.92, Rev. 13 1.124, Rev. Il and 3.41, Rev.1. -

(2) Ceneral Design Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 50, .

Appendix A (CDC Nos. 1, 2, 61, 62, and 63)

(3) NUREC40800, Standard Review Plan (1981).

1 (4) "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, and the modifications to this '-

document of January 18, 1979. , .

4

)

. l i

4' l

1 1

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h I e

e

.(

3-6 i

i

,_--..,?__._"**'$ '

Table 3.1

}, BORAFLEX EXPERIENCE FOR HICH DENSITY RACKS ,

F Plant NRC Licensing

i Site Type Docket No. Status Point Beach 1 and 2 PWR 50-226 & 301 Licensed j <.

. Nine Mile Point 1 BWR 50-220 Licensed 4 1 Oconee 1 and 2 PWR 50-269 & 270 Licensed 1* Prairie Island 1 and 2 PWR 50-282 & 306 Licensed Calvert Cliffs 2 . PWR 50-318 Licensed c.

Quad Cities

• 1 and 2 BWR 50-254 & 265 Licensed Watts Bar 1 and 2 PWR 50-390 & 391 Pending

.. Waterford 3 PWR 50-382 Pending Fermi

• 2 BWR 50-341 Licensed

{ ,

H. B. Robinson - 2 PWR 50-241 Licensed l?

River Band 1 BWR 50-458 Licensed I.

i Rancho Seco

• 1 PWR 50-312 Licensed

l. Nine Mile Point 2 BWR 50-410 To be ap-plied for 1'

Shearon Harris 1 PWR 50-400 To be ap-plied for Millstone 3 PWR 50-423 To be ap-f!. - plJed for (t Grand Gulf

• 1 BWR 50-416 Pending I.

Oyster Creek

• BWR 50-219 Licensed V. C. Summer

• PWR 50-395 Licensed Diablo Canyon

• 1 and 2 PWR 50-275 & 323 Licensed I

• Joseph Oat Corporation fabricated racks

(

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SHEET (BORAFLEX) ,/

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! FIGURE 3.7 3 X 3 TYPtCAL ARRAY REGION 2 ~)

i 3-14 ,

I l l 1

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l

. I' i

I 4.O NUCLEAR CRITICALITY ANALYSIS I ..

1 .

4.1 DESIGN BASES l, The high density spent fuel storage racks for the Byron j Nuclear Power Station are designed to assure that the neutron

'I multiplication factor (k,gg) is equal to or less than 0.95 with the racks fully loaded with fuel of c'he highest anticipated C reactivity in each of two regions, and flooded with unborated 1-water at a temperature corresponding to the highest reactivity.

,- The maximum calculated reactivity includes a margin for uncertainty in reactivi,ty calculations' and in mechanical

.. tolerances, statistically combined, such that the true k gg will be equal to or less than 0.95 with a 954 probability at a 954 confidence level.

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

r e General Design Criterion 62, Prevention of Criticality in J. Fuel, Storage and Handling.

e USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, l

1 New Fuel Storage, and Section 9.1.2, Spent Fuel Storage.

i.

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

a USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility

' ff

• Design Basis, Rev. 2 (proposed), December 1981.

O e USNRC Regulatory Guide 3.41, validation of Calculational

-. Methods for Nuclear Criticality Safety (and related ANSI N16.9-1975).

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

4-1

i l.

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

e ANSI N18. 2-1973, Nuclear Safety Criteria for the Design

, of Stationary Pressurized Water Reactor Plants. i To assure tl$e true reactivity will always be less than the .)

l calculated reactivity, the. following conservative assumptions were mades

,h e Moderator is pure, unborated water at a temperature cor- I responding to the highest reactivity. '

L e Lattice of storage racks is assumed infinite in all di-

i rections, i.e., no credit is taken for axial or radial (except neutron leakage in the assessment of certaitt abnormal / accident conditions).

. e Neutron absorption in minor structural members

is neglected, i.e., spacer grids are replaced by water. .< .

i i The design basis fuel assembly is a 17 . x 17 Westinghouse v,

optimized fuel assembly containing U02 at a maximum initial enrichment of 4.2% U-235 by weight, corresponding to 48.6 grams U-235 per axial centimeter of fuel assembly. Two separate J' storage regions are provided in the spent fuel storage pool, with

separate criteria defining the highest anticipated reactivity in '!

each of the two regions as follows: -

e Region 1 is designed to accommodate new fuel with a maximum enrichment of 4.2 wtt U-235, or spent fuel regardless of the discharge fuel burnup. ,

i e Region 2 is designed to accommodate fuel of various '

initial enrichments which have accumulated minimum burnups within an acceptable bound as depicted in Fig.

4.1. }

n l

M l

. J

?

I 4.2

SUMMARY

OF CRITICALITY ANALYSES 1 .,

1

, 4.2.1 Normal' Ope'ratino Conditions

{ The criticality analyses of each of the two separate regions of the spent fuel storage pool previously described are sum-F marized in Table 4.1 for the anticipated normal storage condi-tions. The ca'lculated maximum reactivity in Region 2 includes a burnup-dependent allowance for uncertainty in depletion calcu-

d. lations and, furthermore, provides an additional margin of more

, than,2% Ak below the limiting effective multiplication factor '

, (k,gg) of 0.95. As cooling time increases in long-term storage, decay of Pu-241 results in a significant decrease in reactivity, I which will provide an increasing suberiticality margin and tends to further compensate for any uncertainty in depletion calcula-tions. Spacing between the two different rack modules is suf-ficient to preclude adverse nuclear interaction between modules.

. Region 2 can accommodate fuel of various initial enrichments

, and discharge fuel burnups, provided the combination falls within l; the acceptable domain illustrated in Fig. 4.1. For convenient

, reference, the minimum burnup values in Fig. 4.1 have been fitted by linear tangents at various values and the results are tabulated below.

f I

Initial Minimum Initial Minimum

. Enrichment, 4 Burnup, MWD /MTU Enrichment, 4 Burnuts , MWD,1MTU -

f. s 1.52 0 . 3.00 . 22,490 l

t' l.80 5,230 3.20 25,000 -

! L 2.00 8,570 3.40 '27,510 '

i 2.20 11,340 3.60 , 30,020 '

r 2.40 14,390 3.80 32,540 , ;

2.60 17,050 4.00 34,960 2.80 19,700 4.20 37,370 i

k I

4-3

Linear interpolation between the tabulated values will always yield values on or conservatively above the curve of limiting burnups.

These data will be , implemented in appropriate administrative

, procedures to assure verified burnup as specified in draf t Regu-latory Guide 1.13, Revision 2. Administrative procedures will \

also be employed to confirm and assure the presence of soluble '

poison in the pool water during fuel handling operations, as a 9 further margin of safety and as a precaution in the event of fuel U misplacement during fuel handling operations as discussed in 't Section 4.2.2.

I e'

e

'I I

n I

e 1

.i

.i F

I 4-4 l i f

I

~

L-

~

Table'4.1 1

SUMMARY

OF CRITICALITY SAFETY ANALYSES i 't ' '

l Region 1 Region 2

'C' i I. -

', Minimum acceptable burnup 0 37,370 MWD /MTU 4 4.2% initial enrichment id

Temperature assumed O'C 0*C l1' for analysis Reference k, (nominal) 0.9374 0.8999 I

(.

Calculational bias 0.0013 0.0013 I

Uncertainties

f. Bias 40.0018 *0.0018

. B-10 concentration *0.0021 *0.0028 i ,; Boraflex thickness *0.0047 *0.0078 Boraflex width 10.0007 *0.0009 Inner box dimension to.0018 to.0011 Water gap thickness AO.0038 NA

[ SS_ thickness *0.0025 *0.0001 Fuel enrichment

~

• 0.0024 *0.0024 Fuel density *0.0026 to.0026

i. Eccentric assembly negative negative position l'

*0.0082 to.0003 Statistical combination (l)

Allowance for +0.0187 f- burnup uncertainty NA

( Total 0.9387

• 0.0082 0.9199

• 0.0093

)

Maximum reactivity 0.9469 0.9292 f

III Square root of sum of squares.

4-5 1

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l '1 INITIAL ENRICHMENT,WTX U-235 Fig. 1 ACCEPTABLE BURNUP DOMAIN IN REGION 2 OF THE BYRON '

STATION SPENT FUEL STORAGE RACKS. L 4-6. f 1__.__- _ _ __ _ _ _ _ _ _ _ __- _- _ .__ _ _ _ _ . _ _ . _ _ _ . . - _ . _ _ . _ _ _ _ _ _ - _ _ _ . . . . _

j ,.

I i 4 4.2.2 Abnormal and Accident Conditions '

1 .

I Although credit for the soluble poison normally present in

-~

< the spent fuel pool water is permitted under abnormal or accident ,

i conditions,* most abnormal or accident conditions will not result

, , . in exceeding the limiting reactivity (k,gg of 0.95) even in the .

f, absence of soluble poison. The effects on reactivity of credible l, abnormal and accident conditions are summarized in Table 4.2 below. Of these abnormal / accident conditions, only one has the potential for a more than negligible positive reactivity effect.

l l' " Table 4.2 p.

i- REACTIVITY EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS c.

Accident / Abnormal Conditions Reactivity Effect p.

Temperature increase Negative in both regions l l' ll Void (boiling) Negative in both regions

,. Assembly dropped on top of rack Negligible Lateral rack module movement Negligible Misplacement of a fuel assembly Positive i

l_

• Double contingency pririciple of ANSI N16.1-1975, as specified in f the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision (draft) to Reg. Guide 1.13 (Section 1.4, f Appendix A).

1 -

4-7 f

l. . - - - - - - ..- - - - - - . - - - - . . . - . - . -- -

4

( The inadvertent misplacemen't of a new fuel assembly (either ~

into a Region 2 ; storage

• cell or outside and adjacent to a rack module) has the potential for exceeding the limiting reactivity 1 should there be a concurrent and independent accident condition

• I resulting in the loss of all soluble poison. Administrative .f procedures. to assure the presence of soluble poison during fuel handling operations will* preclude the possibility of the simul- b taneous occurrence of these two independent accident condi-tions. The largest reactivity increase occurs for accidentally

placing a new fuel assembly into a Region 2 storage cell with all U other cells fully loaded. Under this condition, the presence of ll only 300 ppm soluble boron assures that the infinite multipli- '

cation factor would not exceed the design basis reactivity for 3 j Region 2. With the nominal concentration of soluble poison ,,

! present (2000 ppm boron), the maximum reactivity, k,, , is less

. than 0.95 even if Region 2 were to be fully loaded with fresh

fuel of 4.2% enrichment. ..

t\

4.2.3 New Fuel Storace *

[

Region 1 of the storage racks is designed to safely ~accommo- '

date new unirradiated fuel of 4.2% enrichment, when fully flooded ri with clean unborated water. Under certain circumstances, it may .s be desirable to store new fuel in the dry condition in Region 1 ,

or to utilize Region 2 for the temporary storage of new fuel, ,

either dry or fully flooded. These conditions were analyzed to i assure the acceptability of Region 1 in the dry condition and to '

determine an arrangement in Region 2 that would assure criti-cality sa}'ety in conformance with the requirements of SRP 9.1.1, "New Fuel Storage." -}

ti criticality analyses confirmed that Region 1 does not -.

i exhibit a peak in reactivity at low moderator densities (e.g., .

fog or foam moderation) and that the optimum moderation (highest k gg) occurs for the fully flooded condition. 'This condition is *

. i

• • h I

r,-,--,-..,,_.-n,- -

i i .

1 .

the design basis for Region 1 where the maximum k,, including all *

; uncertainties, is less than 0.947.

In Mgion 2, it was determined that a checkerboard pattern '

! (fuel assemblies aligned diagonally) provided an acceptable k, in

, . either the fully flooded or the dry (low density moderation) l [, condition for news fuel assemblies of 4.2% enrichment. These .

calculations indicated a nominal k, of 0.813

• 0.014 (la ) when j fully flooded with clean unborated water--a value substantially 1ess than the limiting k gg of 0.95, even with an additional l I' allowance for uncertainties (maximum k, of ~0.86 at 954/954 tol-erarice limits) . .

{ .

t. Calculations, using Monte Carlo techniques, did not reveal a

. peak in reactivity at low moderator densities, and the fully

, flooded condition corresponds to the highest reactivity (optimum l moderation). Thus, the checkerboard pattern of new 4.2% enriched

{ fuel in Region 2 represents a safe configuration in conformance with SRP 9.1.1 and 9.1.2.

(I u

l' O

n 4

f I

c d

l 4-9 l

4.3 REFERENCE FUEL STORAGE CELL t  ;., . .

4.3.1 Reference ' Fuel Assembly " -

3 The design basis fuel assembly, illustrated in Fig. 4.2, is a 17 x 17 array of fuel rods with 25 rods replaced by 24 control ,

rod guide tubes and 1 instrument thimble. Table 4.3. summarizes 4 [

the Westi.nghouse optimized fuel assembly (OFA) design specifi-cations and the expected range of significant variations.

f l 4.3.2 Recion 1 Storace Cells I The nominal spent fuel storage cell used for the criticality -)

l analyses of Region 1 stiorage cells is shown in Fig. 4.2. The .l

' rack is composed of Boraflex absorber material sandwiched between ,

a 0.060-inch inner stainless steel box and a 0.020-inch outer ,,

stainless steel (SS) coverplate (0.125-inch coverplate for module periphery cell walls). The fuel assemblies are centrally located '

l in each storage cell on a nominal lattice spacing of 10.320

• ll 0.050

+

other

~ inches in one direction and 10.420

• 0.050 inches in the direction. Stainless steel gap channels connect one

[

storage cell box to another in a rigid structure and define an '!

j- outer water space between boxes. This outer water space con- -

stitutes a flux-trap between the two Boraflex absorber sheets .

that are essentially opaque (black) to thermal neutrons. The ..

>Boraflex absorber has a thickness of 0.075

• 0.007 inch and a i

nominal B-10 areal density of 0.0238 gram per em2 , -

4.3.3 Recion 2 Storace Cells -

U j Region 2 storage cells were initially designed for fuel of 11 3.2 wtt U-235 initial enrichment burned to 25,000 MWD /MTU and -'

extended to encompass fuel of 4.2% initial enrichment burned to 37,370 MWD /MTU. )

i l

In this region, the storage cells are composed .

of a single Boraflex absorber sandwiched between the 0.060-inch .,

t  !

4-10 i \

l

I -

1- .

stainless steel walls of adjacent storage cells. These cells,

]

shown,in Fig. 4.3, are located on a lattice spacing of 9.011 i .

! 0.040 inches.

l.

d.

e i., .

e s 1.

i. .

e i

1, e

i 4

e 1 .

• O 1

e h

,a -

G I '

I 4-11

k e I Table 4.3 FUEL ASSEMBLY DESIGN SPECIFICATIONS

}

I Fuel Rod Data -

Outside diameter, in. 0.360 tl Cladding thickness, in. 0.0225

[

Cladding material 2ir'caloy-4 .,

Pellet diameter, in. 0.3088 t

UO2 pellet density, t TD 95

• 2 ,

UO2 stack density, g/cm 3 10.288

• 0.217 .i

.l Enrichment, wtt U-235 4.2

• 0.05 li ,

Fuel Assembly Data .

Number of fuel rods 264 (17 x 17 array) [f ;

Fuel rod pitch, in. 0.496

'I Control rod guide tube Number . 24

Outside diameter, in. 0.474 Thickness, in. 0.016 .'

Material zircaloy-4 -

Instrument thimble t Number 1 .

Outside diamatar, in. 0.474 Thickness, in.

• 0.016 q Material 2ircaloy-4 g U-235 loading g/ axial em of assembly 48.6

• 1.0

1. 1 4-12 l

i

i I

h*

i d

8.85' i o.o32" l box I.D. D l~  !!_ sonvLEx i l----

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vate cap

=

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00000000000000000 l l

00000000000000000 l l e l l l 1 Water Cap y Cap Chame!

y I

(

t Fig. 4.2 REGION 1 STORAGE CELL GEOMETRY.

I------- __ , -

______ly A

t.

}

}

LATTICE SPACINC _

5.011' 20.040' l .,

Ip s.as0 2 0.032 sax I.o. - =; .

g Ii i I

00000000000000000 " O 0'0"

• 0'00'"

0000000000000000 u nx -

lllllllllllllllll0 00000000000000000 00000000000000000 l

i 00000000000000500 88888888888888888 i I

00000000000000000 -

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_ . _ .t_. . - . __ . _ . ._ . _ _ t -

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0.041" 1 0.007' Thick O.01302 0.0009 3 810/ cat '

4 l

Fig. 4.3 REGION 2 STORAGE CELL GEOMETRY. .

4-14 i

j

,_ y. __

i 4.4 ANALYTICAL METHODOLOGY '

!h i

4.4.1 Reference Analytical Methods and Bias The CASMO-2E computer code (Refs. 1, 2, and 3), a two-l dimensional multigroup transport theory code for fuel assemblies, l ,

has been. benchmarked (see Appendix A) and is used both as a l' primary method of analysis and as a means of evaluating small l7 reactivity increments associated with manufacturing tolerances.

l'* CASMO-2E benchmarking resulted in a calculational bias of 0.0013

!,

• 0.0018 (954/954). ' '

,, In fuel rack analyses, for independent verification, criti-l ,

cality analyses of the high density spent fuel storage racks were i also performed with the AMPX-KENO computer package (Refs. 4 and l' 5), using the 27-group SCALE

• cross-section library (Ref. 6) with the NITAWL subroutine for U-238 resonance shielding effects l (Nordheim integral treatment). Details of the benchmark calcu-l- lations with the 27-group SCALE cross-section library are also ,

4 presented in Appendix A. These benchmark calculations resulted i

in a bias of 0.0106

• 0.0048 (954/954).

In the geometric model used in KENO, each fuel rod and its cladding were described explicitly. For two-dimensional X-Y analysis, a zero current (white albedo) boundary condition was applied in the axial direction and, for Region 1, at the center-line through the outer water space (flux-trap) on all four sides of the cell, effectively creating an infinite array of storage q

t .

cells. In Region 2, the zero current boundary condition was

,. applied at the center of the Boraflex absorber sheets between f  !

f

• SCALE is an acronym for Standardized Computer Analysis for licensing Evaluation.

4-15 l

I

_-____-7-___.

l 4

storage 'c ells. The AMPX-KENO Monte Carlo calculations inherently

- include a statistical uncertainty due to the random nature of

{ ,

i neutron tracking. To minimize the statistical uncertainty of the KENO-calculated reactivity, a total of 50,000 neutron histories q 3

is normally accumulated for each calculation, in 100 generations of 500 neutrons each.

.i CASMO-2E is also used for burnup calculations, with inde-pendent verification by EPRI-CELL and NULIF calculations. In

. to tracking long-term (30-year) reactivity effects of spent fuel

) stored in Region 2 of the fuel storage rack, EPRI-CELL calcula-

tions indicate a continuous reduction in reactivity with time (after Xe decay) due primarily to Pu-241 decay and Am-241 growth. t A third independent method of criticality analysis, util- ,,

ising diffusion / blackness theory, was also used for additional .i co.nfidence in results of the primary calculational methods, ,

although no reliance for criticality safety is placed on the l,I reactivity value from the diffus, ion / blackness theory technique. l l

i This technique, however, is used for auxiliary calculations of small incremental reactivity effects (e.g., axial cutback or f

mechanical tolerances) that would otherwise be lost in normal h i

KENO statistical variations, or would be, inconsistent with CASMO-

• l 2E geometry limitations. -

I j Cross sections for the diffusion / blackness theory calcula-l tions were derived from CASMO-2E or calculated by the NULIF com- ,

puter, code (Ref. 7), , supplemented by a blackness theory routine that effectively imposes a transport theory boundary condition at

~

the surface of the Boraflex neutron absorber. Two different

! spatial diffusion theory codes, PDQ07 (Ref. 8) in two dimensions I' l ..

l h 4-16 t

i I

i .

and SNEID* in one diniension, we're used to calculate reactivi-i '

ties. The two-dimensional PDQ07 code was used to describe *the

)< actual storage cell geometry, with NULIF cell-homogenized con-i stants representing each fuel rod and its associated water

!(g- ,

moderator. SNEID is a one-dimensional model, in s'ylindrical or

!, slab geometry, used for the calculation of axial cutback re-activity effects and in the assessment of abnormal occurrences.

l

!'# Fuel Burnup Calculations 4.4.2 j2

! r

! Fuel burnup calculations in the hot operating' condition were i

performed primarily with the CASMO-2E code. However, to enhance l c- the credibility of the ',burnup calculations (in lieu of critical ,

l ?, experiments), the CASMO-2E results were independently checked by i

(, calculations with the NULIF code (Ref. 7) a n'd w i t h EPRI-CELL f

[ (Ref. 9). Figure 4.4 compares results of these independent l

!p methods of burnup analysis under hot reactor operating condi-tions. The results agree within 0.008 Ak in the hot operating

!", ' condition. .

C '

'I '

In addition to' depletion calculations under hot operating

i-conditions, reactivity comparisons under conditions more repre- f 4

. sentative of fuel to be stored in the racks (cold, xenon-free)

. are also significant in storage rack criticality analyses. Table

, 4.4 compares the~ cold, xenon-free reactivities calculated by l

CASMO-2E, NULIF/PDQ07, and EPRI-CELL. In the cold condition, the l,.

CASMO-2E calculations gave a slightly higher reactivity valu,e for

. the Region 2 fuel storage cellt and the good agreement generally ,

,1 observed lends credibility to the calculations, particularly in view of the known bias and uncertainty in CASMO-2E calculations (Appendix A).

,t

(*

!p lL l

! *SNEID is a one-dimensional diffusion theory routine developed by

,4 Black & Veatch and verified by comparison with PDQ07 one-dimen- l l sional calculations.

!' 4-17

~ , - . _ , - _ - - .

I t.

4I&IIiiII Ii4444445 6444IIii4 44e' j 4ii44 h a

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=

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- . j -

0 c

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= e

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.i e

ff fI f f f I f f f f t t t tt tt t t t tt tt 9999 9 9 , , ,

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,_. =

[ ,

l ,

. s 1

3 Table 4.4 l '.

1

1 COMPARISON OF COLD, CLEAN REACTIVITIES CALCULATED i AT 25,000 MWD /MTU BURNUP AND 3.2% ENRICHMENT '

j .

j '

l 4

d

k. Xe-free O'C  :

Calculational Method Fuel Assembiv In Recion 2 Cell i 1 <.

! CASMO-2E 1.1206 0.9061 j <

NULIF/PDQ07 1.1294 0.9017 l' EPRI-CELL

• 1.1201(1) -

1 i L'  !

l I

III EPRI-CELL k,at maximum value during long-term (30-year) storage.

l 1

l l'. No definitive method exists for determining the uncertainty  ;

i .' in burnup-dependent reactivity calculations. All of the codes i

L discussed above have been used to accurately follow reactivity i

{, loss rates in operating reactors. CASMO-2E has been extensively l, benchmarked (Appendix At Refs. 1, 2, 3, and 10) against cold, l clean, critical experiments (including plutonium-bearing fuel), .

' 7 Monte Carlo calculations, reactor operations, and heavy-element l In particular, the analyses  !

concentrations in irradiated fuel.

f' (Ref. 10) of 11 critical experiments with plutonium-bearing fuel gave an average k,gg of 1.002 4 0.011 (954/954), showing adequate

!,. treatment of the , plutonium nuclides. In addition, Johansson

! (Ref. 11) has obtained very good agreement in calculations of ; I close-packed, high-plutonium-content, experimental configura-

{ f, tions.

I I s ji since critical-experiment data with spent fuel is not avail-j able, it is necessary to assign an uncertainty in reactivity ,

}  ;

I 4-19  : r

.- - . - - . - .a.. - . - -

based on other' considerations, supported by the close agreement I 5 between different calculational methods and the general industry j experience in predicting reactivity loss rates in operating plants. Over a considerable portion of the burnup, the reac- -

1 i

tivity loss rate in PWRs is approximately 0.01 Ak for each 1,000 j o MWD /MTU, becoming somewhat smaller at the higher burnups. By conservatively assuming an uncertainty in reactivity

• of 0.5 x *I 10-6 times the burnup in MWD /MTU, a burnup-dependent uncertainty '

l is defined that increases with increasing fuel burnup, as would  ?'

be reasonably expected.

, This assumption provides an estimate of - I the burnup uncertainty that is more conservative and bounds ,<f estimates frequently employed in other fuel rack licensing applications (i.e., 5% of the total reactivity decrement). Table ,

1 4.5 summarises results of the burnup analyses and estimated I uncertainties. These uncertainties are appreciably larger, in l general, than would be suggested by the industry experience in predicting reactivity loss rates and boron let-down curves over '

many cycles in operating plants. The increasing level of l conservatism at the higher fuel burnups provides an adequate ll l margin in the uncertainty estimate to accommodate the possible l

existence of a small positive reactivity increment from the axial ll I .

distribution in burnup (see section 4.4.3). In addition, , , ,

although the burnup uncertainty may be either positive or , ,

negative, it is treated as an additive ' term rather than being combined statistically with other uncertainties.

Thus, the allowance for uncertainty in burnup calculations is believed to .

be a conservative estimate, particularly in view of the substantial reactivity decrease with aged fuel as discussed in

• Section 4.4.4. -

i ,

• 0nly that portion of the uncertainty due to burnup. Other un-certainties are accounted for elsewhere.

G 4-20

. l

. _ _ ~ __ __ __ _

t Table 4.5 ,

I .

I ESTIMATED UNCERTAINTIES IN REACTIVITY DUE TO FUEL DEPLETION EFFECTS i

1 Design 0.5 x 10-6 Initial Burnup Times i

Enrichment MWD /MTU Burnuo, Ak Reac.tivit{1)

Loss, Ak 1.8 5,230 0.0026 0.0475

,I' 2.5 15,720 0.0079 0.1575 3.2 25,000 0.0125 0.2337

3.7 31,280 0.0156 . 0.2757 g- 4.2 37,370 0.0187 0.3107 l <.

(II Total reactivity decrease, calculated for the cold, Xe-free condition in the fuel storage rack, from the beginning-of-life

'. to the design burnup.

a

'e 4.4.3 Effect of Axial Burnup Distribution

• i Initially, fuel loaded into the reactor will burn with a I slightly skewed cosine power distribution. As burnup progresses,

' the burnup distribution will tend to flatten, becoming more ,

highly burned in the central regions than in the upper and lower ends. This effect may be clearly seen in the curves compiled in

.. Ref. 12. At high burnup, the more reactive fuel near the ends of '

] the fuel assembly (less than average burned) occurs in regions of lower reactivity worth due to neutron leakage. Consequently, it ,

is expected that distributed-burnup' fuel assemblies would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power

( distribution, precluding the existence of large regions of signi-I ficantly reduced burnup.

4-21

, A number of one-dimensional diffusion theory analyses have been made based upon calculated and measured axial burnup distri-butions. These analyses confirm the minor and generally negative reactivity effect of axially distributed burnup. The trends observed, however, suggest the possibility of a small positive reactivity effect at the high burnup values, and the uncertainty -

in k, due to burnup, assigned at the hig.her burnups (Section  ;

4.4.2), is adequately conservative to encompass the potential for a small positive . reactivity effect of postulated axial burnup y distributions. Furthermore, reactivity decreases with time in storage (Section 4.4.4), and, in addition, there is a large margin in reactivity (>0.02 A k) below the limiting k,gg value (0.95) which can accommodate any reasonab1'e reactivity effects '

. l that might be larger than expected. '

r 4.4.4 Leno-term Decay 4-fi Since the fuel racks in Region 2 are intended to contain  !

spent fuel for long periods of time, calculations were made using EPRI-CELL (which incorporates the CINDER code) to follow the long-term changes in reactivity of spent fuel over a 30-year period. CINDER tracks the decay and burnup dependence of some

179 fission products. Early in the decay period, xenon grows in (reducing reactivity) and subsequently decays, with the reactiv-ity reaching a maximum at 100-200 hours. The decay of Pu-241 -

(13-year half-life) and growth of Am-241 substantially reduce -

reactivity during long term storage, as indicated in Table 4.6. _

Th.e reference design criticality calculations do not take credit ,

for this long-term reduction in reactivity, 'other than to indi-cate an increasing subcriticality margin in Region 2 of the scent fuel storage pool.

4-22 e

4

~

Table 4.6 LONG-TERM CHANGES IN REACTIVITY IN STORAGE RACK (XENON-FREE)

, <~

Ak from Shutdown (Xenon-free)

Storage 1

Time, 3.2%E 4.2%E

,4, years 025,000 MWD /MTU G37,000 MWD /MTU 0.5 -0.0046 ,

-0.0057

1. 0 -0.0080 -0.0103 10.0 -0.0406 -0.0529 20.0 -0.0588 -0.0756

.* 30.0 -0.0692 -0.0886

'S

\

e l I e

e I

1,- - -

}- . .

e e.

d

.I

(

,. I .

4-23 I

i

., 4.5 REGI'ON 1 CRITICALITY ANALYSIS AND 'lVLERANCE VARIATIONS 4.5.1 Nominal Desion Case l Under normal conditions, with nominal dimensions, the 3;

' ~

k, values calculated by the three methods of analysis are as fol- '

lows. '

.l l

t Maximum k,

Analytical Method Blas-corrected k_ (954/954) ;l 4 CASMO-2E 0.9387

• 0.0018 0.9405 i; AMPX-KENO -

0.9301

• 0.0061 0.9362 I l Diffusion blackness 0.9393 0.9393  !

l theory i

i .

The AMPX-KENO calculations include a one-sided tolerance factor *

(Ref. 13) of 1.799 corresponding to 95%, probability at a 954 con- l

fidence limit. For the nominal design case, the CASMo-2E calcu- .! '

! lation yields the highest (most conservative) reactivity, and, ,

therefore, the independent verification calculations substantiate f'

i CASMO-2E as the primary calculational method.

i 4.5.2 Boron Leading variation The Boraflex absorber sheets used in Region 1 storage cells

• are nominally 0.075-inch thick, with a B-10 areal density of O.0238 g/cm 2,. Independent manufacturing tolerance limits are .

• 0.007 inch in thickness and *0.0017 g/cm2 in B-10 content. This .

assures that at any point where the minimum boron concentration .

(0.0221 gram B-10/cm 2) and minimum Boraflex thickness (0.068 inch) may coincide, the boron-10 areal density will not be less "

than 0.020 gram /cm 2. Differential CASMO-2E calculations indicate that these tolerance limits result in an incremental reactivity J uncertainty of *0.0021 Ak for boron content and *0.0047 for Bora-flex thickness variations. r 1

, 4-24 -

r

' ,- 4.5.3 Storace cell Lattice Pitch variation ,

4 s .

l The design storage cell lattice spacing between fuel assen-

. blies in Region 1 is 10.32 inches in one direction and 10.42 1 ,

inches in the other direction. A decrease in storage cell lattice spacing may or may not 16 crease reactivity depending upon

other dimensional changes that ,may be associated with the

~

decrease in lattice spacing. Increasing the water thickness

'1' between the fuel and the inner stainless steel box results in a small increase in reactivity. The reactivity effect of the flux-

, ,' trap water thickness, however, is more significant, and de-

! creasing the flux-trap water thickness increases reactivity. ,

4

p. Both of these effects have been evaluated for independent design

,, tolerances.

i i

• The inner stainless steel box dimension, 8.850

• 0.032 inches, defines the inner water thickness between the fuel and the inside box. For the tolerance limit, the uncertainty in

{* reactivity is *0.0018 Ak as determined by differential CASMO-2E calculations, with k, increasing as the inner stainless steel box l

dimension (and derivative lattice spacing) increases.

l The design flux-trap water thicknesses are 1.160

• 0.040 i inches and 1.260

• 0.040 inches, which result in an uncertainty of *0.0038 Ak due to the tolerance in flux-trap water thickness, assuming the water thickness is simultaneously reduced on all

'l four sides. since the manufacturing tolerances on each of the four sides are statistically independent, the actual reactivity uncertainties would be less than *0.0038, although the more con-servative value has been used in the criticality evaluation.

L l.

l t i

l 4-25 L

l i

. 4.5.4 _ stainless steel Thickness Tolerances

• 1 1[

The nominal stainless steel thickness in Region 1 is 0.060 inch for the inner stainless steel box and 0.020 inch for the  !

Boraflex coverplate (Q.125 inch on module boundary). The maximum i positive reactivity effect of the expected stainless steel thick-  ;,

i ness ' tolerance variations, statistically combined, was calculated  ;  !

(CAsMo-2E) to be 40.0025 Ak.

y,

! 4.5.5 Fuel Enrichment and Density Variation {'

. '1 The design maximum enrichment is 4.20 4 0.05 wtt U-235. i Calculations of the sensitivity to small enrichment variations by J, j

CASMo-2E yielded a coefficient of 0.0047 Ak per 0.1 wtt U-235 at k' '

l the design enrichment.

For a tolerance on U-235 enrichment of r

} 40.05 in wtt, the uncertainty on k, is 40.0024 Ak. .

j i

calculations were made 'with the 00 2 fuel density increased lI ,

to a maximum value of 974 theoretical density (TD). For the mid- '

t range value (954 TD) used for the reference design calculations, I 1

'the uncertainty in reactivity is 60.0026 Ak over the range of Uo2 l densities expected.

'l i

4.5.6 5craflex Width Tolerance variation

• i The reference storage cell design for Region 1 (Fig. 4.2)  ;< j uses a Boraflex blade width of 7.75 4 0.0625 inches. A positive ,1 ;

increment in reactivity occurs for a decrease in Boraflex absorber width. For a reduction in width of the maximum toler-ance, 0.0625 inch, ' the calculated positive reactivity increment .

I is +0.0007 Ak. '

I l

4-26 i

9 i

, i .

1 .

4.5.7 Axial Cutback of Boraflex i .

The axial length of the Boraflex poison material is less than the active fuel length by three inches at the top and at the bottom of the Region 1 storage rack modules. To account for the

. reactivity effect of this axial cutback, one-dimensional (slab)

, , diffusion theory calculat. ions were made using flux-weighted homogenized diffusion theory constan,ts edited from CASMO-2E cal-l, culations of the array of storage cells, with and without Boraflex present. In the one-dimensional calculations, an

)

infinite (30-cm) water reflector was used above and below the assembly, fuel with the lengths of the unpoisoned " cutback" l[* regions, top and botton, varied in a series of parametric cal-culations. Results of these calculations showed that the k,gg

. remains less than the k, of the reference central storage cell 1, region, until the axial cutback exceeds four inches top and  !

, bottom. Thus, the actual axial neutron leakage more than compen-sates for the three-inch design cutback, and the reference

, infinite multiplication factor, k,, remains a conservative over-estimate of the true reactivity.

~

i 4 t

I i

! i 4-27 I

l '

4.6 REGION 2 CRITICALITY ANALYSIS AND TOLERANCE VARIATIONS 4.4.1 NeminalDNsionCase The principal method of analysis in Region 2 was the CASMO-2E code, using the restart option in CASMO to transfer fuel of a specified burnup into the storage rack configuration at a refer- 'l once temperature of 0'C. Calculations were made for fuel of

several different initial enrichments and, for each enrichment, a *-

limiting k, value established which included an additional factor .

for uncertainty in the burnup analysis and for the axial burnup .

distribution. The restart CASMO-2E calculations (cold, clean, l rack geometry) were then interpolated to define the burnup value yielding the limiting k, value for each enrichment, as indicated ,

in Table 4.7. These converged burnup values define the boundary

of the acceptable domain shown in Fig. 4.1.

Table 4.7 '

FUEL BURNUP VALUES FOR REQUIRED REACTIVITIES (k,)

. WITH FUEL OF VARICUS INITIAL ENRICHMENTS l

Fuel ,'

Initial Reference UncertaintyIII Design turnup, Enrichment k, in Burnup, Ak Limit k. MWD /MTU -

1.58 0.9106 0 0.9186 0 - .

1.8 0.9186 0.0026 0.9160 5,230 2.5 0.9106 0.0079 0'.9107 15,720 , .,

3.2 0.9106 0.0125 0.9061 25,000 .

3.7 0.9186 0.0156 0.9030 31,200 4.2 0.9106 0.0187 0.8999 37,370 (1)see section 4.4.2.

4-20

l-i At a burnup of 37,000 MWD /MTU, the sensitivity to burnup is 4 calculated to be -0.0079 Ak per 1000 PWD/MTU. During long-term storage, the k, values of the Region 2 fuel rack will decrease

. continuously from decay of Pu-241 as indicated in Section 4.4.4. -

Two independent calculational methods were used to provide t'

' additional confidence in' the reference Region 2 criticality analyses. Fuel of 1.5% initial enrichment (approximately equiva-

~

lent to the reference rack design for burned fuel) was analyzed by AMPX-KENO (27-group SCALE cross-secti6n library) and by the CASMO-2E model used for the Region 2 rack analysis. For this

(

case, the CA5M0-2E k, (0.9014) was within the statistical uncer-

. tainty of the bias-c,orrected value (0.9043 6 0.0030 (le))

obtained in the AMPX-KENO calculations. This agreement confirms i

the validity of the primary CASMO-2E calculations.

! The second independent method of analysis used the NULIF code for burnup analysis, and for generating diffusion theory

, constants (cold, clean) for the NULIF-calculated composition at These l*

25,000 MWD /MTU with fuel of 3.2% initial enrichment.

constants, together with blackness theory constants for the

<- Boraflex absorber, were then used in a two-dimensional PD007

i.  ! calculation for the storage rack configuration. Results of this

{ calculation (k. of 0.9017) compared favorably with the CASMO-2E ,

I calculation for the same conditions (k of 0.9061) ,and thus tend j, to , confirm the val,idity of the primary calculational method.

\' '

l' 4.6.2 Boron Leadine variation '

l

- The Boraflex absorber sheets used in the Region 2 storage l

j - cells are nominally 0.041 inch thick with a 5-10 areal density of l 0.0130 g/cm2. Independent manuf acturing limits are *0.007 inch I in thickness and *0.0009 g/cm2 in 5-10 content. This assures ,

! that at any point where the minimum boron concentration (0.01206 B-10/cm 2 ) and the minimum Boraflex thickness (0.034 inch) may l4 i

) 4-29 l

coincide, the boron-10 areal density will ..ot b'e 1ess than 0.010 i

g/cm 2. Differential CASMO-2E calculations indicate that these tolerance limits result in an incremental reactivity uncertainty of *0.0028 Ak for boron content and *0.0078 Ak for Boraflex l

thickness. I 4.6.3 storace cell Lattice Pitch Variations s i

t' l

The design storage cell lattice spacing between fuel assem- [i blies in Region 2 is 9.011

• 0.040 ine,hes, corresponding to an uncertainty in reactivity of 0.0011 Ak.

l 4.6.4 stainless steel Thickness Tolerance i .

The nominal thickness of the stainless steel box wall is e 0.060 inch with a tolerance limit of 40.005 inch, resulting in an .

] uncertainty in reactivity of *0.0001 Ak.

l 4.6.5 Fuel Enrichment and Density Variation 9

.i .I Uncertainties in reactivity due to tolerances on fuel en-richment and U0 2 density in Region 2 are assumed to be the same

• l! as those. determined for Region 1.

( 4.6.6 Boraflex Width Tolerance

• i The reference storage cell design for Region 2 (Fig. 4.3) .

uses a Boraflex absorber width of 7.25

• 0.0625 inches. For a ,

reduction in width of the maximum tolerance, the calculated posi- ,;

I tive reactivity increment is 0.0009 Ak.

l l

l i

3' i  :

l 4-30

Ir i

i

.. , e w , . .

4.7 ABMORMAL- AND ACCIDENT C'ONDITIONS *

a. .

4.7.1 Eccentric Positionino of Fuel Assembly in Storace Rack f

4 The fuel assembly is normally located in the center of the

, stora'ge. rack cell with bottom fittings and spacers that mechan-2 ically limit lateral movement of the fuel assemblies. Neverthe-l less, calculations were made with the fuel assemblies moved into the corner of the storage rack cell (four-assembly cluster at j

closest approach). These calculations indicated that the reac-tivity decreases very slightly in both regions, as determined by

} PD007 calculations with diffusion coefficients

• generated by NULIF and a blackness , theory routine. The highest reactivity

, therefore corresponds to the reference design with the fuel i , assemblies positioned in the center of the storage cells.

4.7.2 Temperature and water Density Effects i

(

l The moderator temperature coefficient of reactivity in both -

regions is negativer a moderator temperature of O'C, with a water -

density of 1.0 g/cm ,3 was assumed for the refere6ce designs, ,

l- which assures that the true reactivity will always be lower, regardless of temperature.

ll.

Temperature effects on reactivity have been calculated and

, the results are shown in Table 4.8. Introducing voids in the water internal to the storage cell (to simulate boiling) de-creased reactivity, as shown in the table. Voids due to boiling j' will not occur in the outer (flux-trap) water region of Region 1.

l. -

i .

l

i

! *This calculational approach was necessary since the reactivity' l effacts are too small to be calculated by KENO, and CASMO-22 geometry is not readily amenable to eccentric positioning of a fuel assembly.

I '

I 4-31 -

l

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

~

b

, Table 4.8 - - l

-s .. .

EFFECT'OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF STORAGE RACK -

Case Incremental Reactivity Chance, Ak ,

, . Region 1 Region 2

{

O'C Reference Reference t

g 4

20*C -0.0022 -0.0047 50*C -0.0084 -0.0081 80*C -0.0165 -0.0121 l 120*C -0.0298 -0.0178 120*C + 208 void -0.0953 -0.0520 -

i 1

With soluble poison present, the temperature coefficients of l reactivity would be expected to differ from those inferred from ~

the data in Table 4.8. However, the reactivities would also be

{

l substantially lower at all temperatures with soluble boron -

present, and the data in Table 4.8 is pertine tn to the higher- -

reactivity unborated case. .

4.7.3 Dropped Fuel Assembly Accident To investigate the possible reactivity effect of a postu-lated fuel assembly drop accident, calculations were made for ~

unpoisoned assemblies separated only by clean unborated water.

Figure 4.5 shows the results of these calculations. From these ~

data, the reactivity (k ) will be less than 0.95 for any water e

4-32

--- -~->v, - , , , , - - - , .,_,,---.nw ,r-,---m- mww,w-,,,,- __~y,-,mewww,e v

i . .

I

• 1.50 i ia eiia 6 6 i e iei4 i e i i eeii l . . _

.l -

1.40 4?

a. .

1.30

. \ -

u .

>=

- ~

.; u z - -

? u 1.20 z -

. H h

1,go

\

lij i

- 1.00 0.90 0 2 4 & 3 to 12 i

r l WATER GAP SETWEEN ASSEMBLIES, In.

' d Fig. 4.5 REACTIVITY IFFECT OF WATER SPACING BETWEEN fuel assemblies.

f l

gap spacing greater than 6 to 7 inches in the absen'ce of any i .

absorber material, other than water, ,between assemblies. For a drop on top of 'the rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance of more than 12 inches. Maximum expected deformation ,l under seismic or accident conditions will not reduce the minimum spacing between fuel assemblies to less than 12 inches. Con-

sequently, fuel assembly drop accidents will not result in an '

increase in reactivity above that calculated for the infinite  ?

nominal design storage rack. Furthermore, soluble boron in the i pool water would substantially reduce the reactivity and assure  ;

that the true reactivity is "always less than the limiting value .

l for any conceivable fuel handling accident. .

l 4.7.4 Abnormal Location of a Fuel Assembly 9

The abnormal location of a fresh unirradiated fuel assembly '

of 4.2% enrichment could, in the absence o'f soluble poison, I result in exceeding the design reactivity limitation (k,,'of  !

, 0.95). This could occur if the assembly were to be either posi- r i

tiened outside and adjacent to a storage rack module or loaded -

}

into a Region 2 storage cell,~with the latter condition producing 7 the larger positive reactivity increment. Soluble poison, how- ,.

ever, is normally present in the spent fuel pool water (for which .

credit is permitted under these conditions) and would maintain the reactivity substantially less than the design limitation. '

i t

The largest reactivity increase occurs for accidentally #

placing a new fuel assembly into'a Region 2 storage cell with all

• other cells fully loaded. Under this condition, the presence of -

300 ppm soluble boron assures that the infinite multiplication  :

f actor would not exceed the design basis reactivity. With the _

' nominal concentration of soluble poison present (2000 ppm boron),

the maximum reactivity, k,, is less than 0.95 even if Region 2 were to be fully loaded with fresh fuel of 4.2% enrichment.

4-34

=

• A----re- , w --wwy- w ww---- ----w~ ~ , - -w- e ,mn

Administrative procedures will be used to corifirm and assure the

, continued presence of soluble poison in the spent fuel pool water during fuel handling operations.

i 4.7.5 Lateral Rack Movement l Lateral motion of the rack modules under seismic conditions .

could potentially alter the spacing between rack modules. How-f ever, girdle bars on the modules prevent closing the spacing to less than 1.25 inches, whicli is approximately the normal flux-trap water gap in the Region 1 reference design. Region 2 storage cells do not use flux-trap and the reactivity is

~

insensitive to the spacing between module's. Furthermore, soluble poison would assure that a reactivity less than the design limi-~

, . tation is maintained under all conditions.

9 I

t e

• e e

l l.

h.

• ~

l I

I 4-35

i

. 4.8 NEW FUEL STORAGE . -

4.8.1 Storace in Region 1, Dry Region 1 is normally designed to accommodate new unirra- e I

diated fuel assemblies in the fully flooded condition. For

,I storage in the dry co ndition, the racks must also conform to the

[ ,

requirements of SRP 9.1.1 which specify a limiting k,gg value of 0.98 under optimum low density moderation. Calculations were ,l l

made, using AMPX-KENO, for several hypothetical low-moderator al densities down to 0.05 g/cc simulating fog or foam moderation.

These calculations showed a . continuously decreasing k, as the moderator density decreased,, yielding a k, of 0.546 i 0.008 (la) I at lot moderator density. Axial leakage was neglected in these ..

calculations, but would substantially reduce the already ,,

low k, values. These results are consistent with the general ,l observation that a low-density optimum-moderation peak in I

reactivity does not exist in poisoned racks (Ref. 14).

4.8.2 Storage in Region 2, Flooded l

In a succession of trial-and-error calculations, it was ~

found that a checkerboard storage pattern in Region 2 would allow -

new fuel assemblies of 4.2% enrichment to be safely accommodated without exceeding the limiting 0.95 k,gg value. In this checker- .

board loading pattern, the fuel assemblies are located on a diagonal array, as illustrated' on the next page, with alternate , ,

storage cells empty of any fuel.

e m

4-36

, , . - y- w..c , _.-..-w - , _ , _ . . - . , . - - - , _ . . . , . - , -

t I .

l . .

N //// '///. //// //// . '///d f///' r///, -

///) //// '///< l r///, ///s //// '///. ///s r/// '///d //// r/// '///s

//// '///d //// //// '///<

~ '///,

f/// '///i ///) f/// _

l f///, '///s //// '///, ///d gggggg r/// '///; //// r/// .///s _

//// '///4 //// //// '///d f/// '///. //// f/// '///d r///. '///s //// f///. '///J 7777 NON-r///, ///s //// r/// ///s (./J.L US A BLE

//// '///a //// //// '///a I //// '///, //// //// '///' (empty) f///. ///4 1/// f///, .///s r/// ///s //// r/// ///s

. //// '///a //// f/// '///s

//// '///, ///) //// '///<

r///4 ///s //// f///< '///s

'///s r///, '///s 1//// f///.

Monte Carlo calculations (AMPX-KENO) resulted in a k, of

. C.8133

• 0.0138. With a one sided K-factor (Ref. 13) for 95%

a probability at a . 954 confidence level and a Ak of 0.009 for uncertainties (T'ble a 4.1 for Region 2), the maximum k, is 0.863, which is substantially less than the 0.95 limiting value. Thus, ,

l l Region 2 may be safely used for the temporary storage of new fuel assemblies provided the storage configuration is restricted to the checkerboard pattern indicated above.

l 4.8.3 Storace in Recion 2, Dry

. As indicated in Section 4.8.1 above, a peak in reactivity f (k,gg) at low moderator densities is not expected fo,r poisoned rack designs. AMPX-KENO calculations confirmed the absence of a -

i low-moderator-density peak in Region 2 with 4.2% enriched fuel arranged in the checkerboard pattern. At lot moderator density, the calculated k, was 0.552, which would be substantially reduced if axial leakage were to be included. Thus, Region 2 conforms to the requirements of SRP 9.1.1 (k, <0.98 at optimum moderation) for the safe storage of 4.2% enriched fuel, dry, in the checker-board loading pattern.

4-37 i

l e

t-REFERENCES

1. A. Ahlin, M. Edenius, H. Haggblom, "CASMO - A Fuel Assembly

Burnup Program," AE-RF-76-4158, Studsvik report (proprietary). l I;

2. A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory  :

Depletion Code for LWR Analysis," ANS Transactions, Vol. 26,

p. 604, 1977. ,

3.

?

! M. Edenius et al., "CASMO Benchmark Report," Studsvik/RF '

6293, Aktiebolaget Atomenergi, March 1978. I

4. Green, Lucious, Petrie, Ford, White, Wright, "PSR-63/AMPX-l ~ *

(code package), AMPX Modular Code System for Generating l Coupled Multigroup N6utron-Gamma Libraries from ENDF/B,"

ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.

5. L. M. Petrie and N. F. Cross, "KENC-lV, An Improved Monte Carlo Criticality Program," ORNL-4938, Oak Ridge National Laboratory, November 1975. .

6. R. M. Westfall et al., " SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing  !

Evaluation," NUREG/CR-0200, 1979. 6 j 7. W. A. Wittkopf, "NULIF - Neutron Spectrum Generator, Few- '

Group Constant Generator and Fuel Depletion Code," BAW-426,  !

The Babcock & Wilcox Company, August 1976. '

8. W. R. Cadwell, PDQ07 Reference Manual, WAPD-TM-678, Bettis Atomic Power Laboratory, January 1967. -

9. W. J. Eich, " Advanced Recycle Methodology Program, CEM-3," "

Electric Power Research Institute, 1976. -

10. E. E. Pilat, " Methods, for the Analysis of Boiling Water Reactors (Lattice Physics)," YAEC-1232, Yankee Atomic Electric Co., December 1980. *

~

11. E. Johansson, " Reactor Physics Calculations on Close-Packed Pressurized Water Reactor Lattices," _ Nuclear Technolocy, -

Vol. 68, pp. 263-268, February 1985.

12. H. Richings, Some Notes on PWR (W) Power Distribution Probabilfstic Probabilities for LOCA Analyses, NRC Memorandum to P. S. Check, dated July 5, 1977.

4-38

. REFERENCES (Continued) t.

I

13. M. G. Natrella, Experimental Statistics National Bureau of j, Standards, Handbook 91, August 1963.

i

14. J. M. Cano et al., "Supercriticality Through optimum Modsration in Nuclear Fuel Storage," Nuclear Technolocy, j, Vol. 48, pp. 251-260, May 1980.

4.

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6 4

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l 1 -

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i 4-39

' f.

. 5.0 THERMAL-HYDRAULIC CONSIDERATIONS A primary objective in the design of the high density spent fuel ,

storage racks is to ensure adequate cooling of the fuel assembly cladding. In the following, a brief synopsis of the , design basis, the method of analysis, and computed results are given. -

l Simila'r analysis has been used in previous licensing reports on l -

high density spent fuel racks for Fermi 2 (D,ocket 50-341), Quad

. Cities 1 and 2 (Dockets 50-254 and 50-265), Rancho Seco (Docket

. 50-312), Grand Gulf Unit 1 (Docket 50-416), Oyster Creek (Docket l 50-219), Virgil C. Summer (Docket 50-395), and Diablo Canyon 1 and 2 (Docket Hos. 50-275 and 50-323).

~

5.1 DECAY HEAT CALCULATIONS FOR THE SPENT FUEL This report section covers requirement III.1.5(2) of the NRC's

<9 "0T Position for Review and Acceptance of Spent Fuel Storage and

- Handling Applications" issued on April 14, 1978. This requirement states that calculations for the amount of thermal Ia energy rencoved by the spent fuel cooling system shall be made in a .'

, accordance with Branch Technical Position APCSB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling" (Ref. 1). The calculations contained herein have been made in accordance with this requirement.

5.1.1 Basis 14

~

7 The Byron Nuc1 ear Power Station Units 1 and 2 reactors are both rated at 3411 megawatts thermal (MWt). Each core contains 193

' .. fuel assemblies. Thus, the average operating power per fuel assembly, Po, is 17.6736 MW. The fuel discharge can be made in one of the following two modes i

! O Normal refueling discharge 0 Full core discharge l

l l

r-, -- . N

1 l

I An equilibrium 1 reload consists of 84 assemblies (with 18-month l cycles). The four-transitional reloads for each unit consist of  !

88 assemblies. The fuel transfer begins after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of decay

, time in the reactor (time after shutdown). It is assumed that the time period of discharge of this batch is 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> (three assemblies transferred to the pool per hour). The cooling system .,.

consists of a Seismic. Category I spent fuel cooling circuit. The bulk temperature analysis assumes a 105'F coolant inlet temperature to the spent fuel pool heat exchanger for these refueling cases. y l *

.t i

For the full core discharge, it is assumed that the total time period for the discharge of the full core is 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br /> (after 100  ;

hours of shutdown time.in the reactor). The discharge rate to ,

the pool is assumed to be continuous and uniform.  ;

I The fuel assemblies' are removed from the reactor af ter a maximum postulated time'of 4.5 years of cumulative operating time. Since the decay heat load is a monotonically increasing function of the j cumulative reactor operating I

, time, To, it is conservatively assumed that every fuel assembly discharged has had the maximum -

postulated to of 4.5 years for the batch size of 84. .

The water inventory in the reactor cavity cooled by the residual ,

heat removal (RHR) heat exchanger exchanges heat with the fuel pool water mass through the refueling canal. This source of heat removal is neglected in the analysis. Thus, the results obtained for both normal refueling discharge and full core discharge are conservative. -'

e1 5-2 ,

t

,,. The fuel pool cooling system (FC) consists of two independent trains, each consisting of one p0 rep and heat exchanger. Either -

train is capable of providing sufficient cooling for the pool.

~

The following list identifies all relevant design data for the

'* spent fuel pool heat exchangers:

l.

i-O Type ,

Tube and shell

. 0 Quantity 2 O Performance data

- Heat transferred 15.833 x 106 Stu/hr

, l. Tube Side

- Fluid flow 2.23 x 10' lb/hr 1 - Pool water inlet temperature 120*F

- Outlet temperature 112.9'F

- Fouling Factor .0005

! Shell Side

- Fluid flow 2.72 x 10' lb/hr

- Coolant inlet temperature 105*

- Outlet temperature 110.82*F

- Fouling factor .

0.0005 The above data enables complete characterization of the thermal performance of a fuel pool heat exchanger.

~

5.1.2 Model Description ~

l? -

Reference 1 is utilized to compute the heat dissipation r;.1uirements in the pool. The total decay heat consists of fission product and heavy element decay heat. Total decay heat, i ,

P, for a fuel assembly is given as a linear function of Po and 5-3 .

I i as an exponential function ,of , and .s8 P = Po f( o, s) (5.1-1) .

where:

P

. . 'l

= total decay heat per fuel assembly, linear .1 function of P o P, = average operating power per fuel assembly ,

[

= cumulative exposure time of the fuel assembly in the reactor *l

}

, = time elapsed since reactor shutdown The appropriate uncertainty factor, K, was applied in accordance with NUREG-0800 (Ref. 1). Furthermore, the operating power, P, o .

. is taken equal to the rated power, even though the reactor may be ,,

operating at less than its rated power during much of the [

} exposure period for the batch of fuel assemblies. Finally, the ,

computations and results. reported here are based on the discharge I taking place when the inventory of fuel in the pool will' be at its maximum resulting in an upper bound on the computed decay heat rate.

• l -

Having determined the heat dissipation rate, the next task is to -'

evaluate the time-dependent temperature of the pool water. Table 5.1 identifies the loading cases examined. The pool bulk .

temperature is determined using the first law of thermodynamics ,

(conservation of energy).

l l

A number of simplifying assumptions are made which render the

~

! analysis conservative, principally:

l

.i l

5-4 l

l

. O The heat exchangers are assumed to have maximum fouling.

'7 Thus, ~.the temperature effectiveness, 5, for the heat exchanger utilized in the analysis is the lowest postulated value: $= .3875 for fuel pool cooler. 5 is i i-' calculated from heat exchanger technical data sheets. No heat loss is assumed to take place through the concrete floor.

f' i

g, 0- No credit is taken for the improvement in the film I. coefficients of the heat exchanger as the operating temperature rises. Thus, the film coefficien't used in P- the computations are lower bounds.

O No credit is taken for heat loss by evaporation of the

.- pool water.

0 No credit is taken for heat loss to pool walls and pool

j. floor slab.

l':

The basic energy conservation relationship for the pool heat exchanger system yields:

l C d,,t, ,gg 2 (5.1-2) t t

dt

, I where:

l

• I

!- Cg = Thermal capacity of stored water in the pool t = Temperature of pool water at time, t

l. Q3 = Heat generation rate due to stored fuel assemblies in the pool; Q1 is a known function of time, t from

- the preceding section.

Q = Heat removed in the fuel pool cooler 2 -

I The pool has a total water inventory of 63444.0 cubic feet when

~

i all racks are in place in the pool and every storage location is occupied.

I l1 5-5 l

l

5.1.3 Decay Heat Calculation Results

• i

\.

The calculations were performed for the pool, disregarding 'the 4

additional thermal capacity and cooling system available in the transfer canal, and the reactor cavity.

.I I'

For a specified coolant inlet temperature and flow rate, the ),

quantity Q2 is shown to be a linear function of T in a recent I1 '

paper by Singh (Ref. 3). As stated earlier, Q1 is an :i exponential function of T. Thus, Equation 5.1-2 can be 1 integrated to determine t directly as ' a function of T. The .

results are plotted in Figures 5.1 through 5.2. The results show that the pool water never approaches the boiling point even with j the most adverse heat load, under normal operating conditions.

J These figures also give Q1 as a function of T. Four plots are generated for each case. The first and third plots for each case shows temperature and power generation, respectively, for a period extending from T = 0 + T =2T n where Tn is the total p time of fuel transfer. The second and fourth plots show the same I quantities (i.e., temperature and power generation, respectively) over a longer period. The long-term plots are produced to show .

I the temperature drop with time. Summarized results are given in ,,

Table 5.2.

i Finally, computations are made to determine the time interval to boiling after all heat dissipation paths are lost. Computations are made for each case under the following two assumptions:

(* 0* All coofing systems lost at the instant pool bulk temperature reaches the maximum value. ,1 I O All cooling systems lost at the instant the heat <

dissipation power reaches its maximum value in the pool.

Results are summarized in Table 5.3. Table 5.3 gives the bulk boiling vaporization rate for all cases at the instant the boiling commences. This rate will decrease with time due to reduced heat generation in the fuel. In all cases, adequate time exists to take corrective action.

1 5-6 l '

)' . , ,

5.2 THERHAL-HY'DRAULIC ANALYSES FOR SPENT FUEL COOLING l

This report section covers requirement III.1.5(3) of the NRC's "0T Position for Review and Acceptance of Spent Fuel Storage and l

Handling Applications," issued on April 14, 1978. Conservative methods have been used to calculate the maximum fuel cladding temperature as required therein. Also, it has been determined that nucleate boiling or voiding of coolant on the surface of the I'

fuel rods occurs only at the locations where freshly discharged fuel assemblies are stored.

l i e

,. 5.2.1 Basis -

In order to determine an upper bou on the maximum fuel cladding temperature, a series of conservative assumptions are made. The most important assumptions are listed below:

O As stated above, the fuellpool will contain spent fuel with varying time-after-shutdown (Ts).

Since the heat

- emission falls off rapidly with increasing T it is a1,1 fuel obviously ' conservative to assume that

~

assemblies are fresh ( Ts = 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) and they all have had 4.5 years of operating time in the reactor for cases 1 and 2. The heat emission rate of each fuel

- assembly is assumed to be equal (Ref. 2).

I

'I O As shown in Figure 2.1 in Section 2, the modules occupy an irregular floor space in the pool. For the I

hydrothermal analysis, .a circle circumscribing the actual rack floor space is drawn. It is further assumed l- that the cylinder with this circle as its base is packed.

l with fuel assemblies at the nominal pitch of 9.011 inches (see Figure 5.3). . ,

- 0 The downcomer space around the rack module group varies, as shown in Figure 2.1. The nominal downcomer gap j' available in the pool is assumed to be the total gap available around the idealized cylindrical rack; thus, I

the maximum resistance to downward flow is incorporated into the analysis.

f l

l 5-7 l

I O No downcomer flow is assumed to exist between the rack '

modules.

5.2.2 Model Description In this manner, a conservative idealized model for the rack assemblage is ob,tained. The water flow is axisymmetric about the 't j vertical axis of the circular rack assemblage, and thus, the flow I is two-dimensional (axisymmetric three-dimensional). Figure 5.4 -

shows a typical " flow chimney" rendering of the ' thermal U hydraulics model. The governing equation to characterize the flow .; ,

field in the pool can now be written. The resulting integral i equation can be solved for the lower plenum velocity field (in the radial direction) and axial velocity (in-cell velocity field), by using the method of collocation. It should be added

! that the hydrodynamic loss coefficients which enter into the formulation of the integral equation are also taken from '

well-recognized sources (Ref. 4) and wherever discrepancies in t i 1

~

reported values exist, the con,servative values are" consistently l '

used. Reference 5 gives the details of mathematical analysis ,

used in this solution process. .

After the axial' velocity field is evaluated, it is a straight- 1 forward matter to compute the ' fuel assembly cla d dir.g temperature. The knowledge of the overall flow field enables pinpointing the storage location with the minimum axial flow

] *

(i.e., maximum water outlet temperature). This is called the most " choked" location. In order to find an upper bound on the -

tiemp erature in a typical cell, it is assumed that it is located ,-

at the most choked location. Knowing the global plenum velocit'y 5 field, the revised axial flow through this choked cell can be calculated by solving the Bernoulli's equation for the flow circuit through this cell. Thus, an absolute upper bound on the water exit temperature and maximum fuel cladding temperature is i obtained. It is believed that, in view of the aforementioned

'l assumptions, the temperatures calculated in this manner overestimate the temperature rise that will actually occur in the f pool.

l l

. 5-8

1 I

The maximum pool bulk temperature, t, is computed in Section

,2 5.1.3 and reported in Table 5.2. The corresponding average power output from thi hottest fuel assembly, q, is also reported in g ,,

that table. The maximum radial peaking factor, F xy, is 1.55 i f or the Byron Nuclear Power Station. Thus, it is conservative to assume that the maximum specific power of a fuel assembly,. qA,

, is given by:

9A = q Fxy (5.2-1) 1- where:

Fxy = 1.55 8-The maximum temperature rise of pool water in the most disadvantageously placed fuel assembly is given in Table 5.4 for all loading cases. Having determined the maximum local water temperature in the pool, it is now possible to determine the l' maximum fuel cladding temperature. It is conservatively assumed

that the total peaking factor F is 2.32. Thus, a fuel rod can produce 2.32 times the average heat emission rate over a small i length. The axial heat dissipation in a rod is known to' reach a

- maximum in the central region, and taper off at its two extremities. For the sake of added conservatism it is assumed

~

that the peak heat emission occurs at the top where the local water temperature also reaches its maximum. Furthermore, no .

'[

j credit is taken for axial conduction of heat along the rod. The highly conservative model thus constructed leads l to simple

algebraic equations which directly give the maximum local n cladding temperature, te".

5.2.3 Results i

Table 5.4 gives the maximum local cladding temperature, te, at

,I

l the instant the pool bulk temperature has attained its maximum

! value. It is quite possible, however, that the peak cladding temperature occurs at the instant of maximum value of gA, 5-9 lI

the instant when the fuel assembly is first placed in a storage locat' ion. Table 5.5 gives the maximum local cladding temperature at t= 0. The local boiling temperature near the top of the. fuel cladding is 240*F. However, the cladding temperature must be somewhat higher than the boiling temperature to initiate and .

sustain nucleate boiling. The above considerations indicate that a comfortable margin.against the initiation of localized boiling s

.]

exists in case 1. For full core discharge (case 2) under the described assumptions, the maximum cladding temperature will give rise to localized nucleate boiling, but not to bulk pool boiling (5.4). 'l

. I m

D 9

I e

a O

Ui t

t k

5-10 l r - - -- . - - , , - - - ,-- , _ , , , _ _ . , , , . , , , - , - - , _ . , - . , , . , _ . - - , - , , - , - . , ,- - - - , , . , - _

l l

t

/

' ~

. REFERENCES TO SECTION 5 1

1. NUREC-0800 U.S. Nuclear Regulatory Commission, Standard Review Plan, Branch Technical Position ASB 9-2, Rev. 2, July 1981.

2. FSAR, Byron Nuclear Power Station.

<- 3. Journal of H. eat Transfer, Transactions of the ASHE, August 1981, Vol. 103, "Some Fundmental Relationships for Tubular

, Heat Exchanger Thermal Performance," K.P. Singh.

4. General Electric Corporation, R&D Data Books, " Heat Transfer and Fluid Flow," 1974 and updates.

5. 4th National Congress of the ASME, "A Method for Computing

• the Maximum Water Temperature in a Fuel Pool Containing Spent Nuclear Fuel," by K.P. Singh and A.I. Soler, paper 83-NE-7, Portland, Oregon (3une 1983).

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5-11

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Table 5.1 LIST OF CASES ANALYZED I

l No. of Total Time Fuel to Transfer -

, Assemblies Fuel Into' Decay Time Discharged,. the Pool Before Transfer Case No. Condition N th, hrs Begins, hrs 1 Normal refueling discharge

• 84 28 100 y 2 Full core" 193 64 100 a discharge to Discharge is assumed to be l'nto a pool containing fuel from 17 previous discharges -

of 168 assemblies.

L I - ' * ' ' * ' ~ * *

~ L _' ' -'

-A ...- _ . . _. ..

- ~

. . . . . - , ,m ..

.--, ,, 3 -

Table 5.2 MAXIMUM POOL BULK TEMPERATURE, t, COINCIDENT TOTAL POWER, 01, AND COINCIDENT SPECIFIC POWER, q, FOR THE HOTTEST ASSEMBLY i

1 Coincident Coincident Time to Maximum Time (After Coincident Total .

Transfer Pool Bulk Initiation Specific Power j Case No. of Fuel Into Temp., t, of Fuel- Power, q, On (10")

No. Assemblies Pool, hrs *F Transfer), hrs Stu/sec Btu / hour Notes t

1 84 28 138 37.0 55.15 .1985 Normal

,, discharge

• 2 '

193 64 155 71.0 50.30 .1811 Full Core Discharge *

• Discharge is assumed to be into a pool containing fuel from 17 previous discharges of 168 ensemblies.

G 4

9

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

. . . l t

• \

Table 5.3

~ '

TIME (HRS) TO BOILING AND BOILING VAPORIZATION RATE'FROM THE S INSTANT ALL COOLING IS LOST .

'l

?!

!? ,

CONDITION 1 CONDITION 2

\

, . Loss of Cooling at Loss of Cooling at Maximum Maximum Power Discharge Pool Bulk Temperature Rate .

Case Time (Hrs) Yap. Rate Time (Hrs) Vap. Rate ' '

No. Ib/hr lb/hr i.- ,

1 9 35329.0 9 35788.0 t

2 4 54233.0

• 4 54827.0 -

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Table 5.4 1'

HAXIMUM LOCAL POOL WATER TEMPERATURE AND LOCAL FUEL CLADDING TEMPERATURE AT INSTANT OF MAXIMUM POOL BULK TEMPERATURE f

Maximum Local Maximum Local Fuel Claddin Case Water No. Temperature, 'F Temperature,g F Case Identified 84 assemblies 4

1 194 239

f. ,

l' 2 208 250 193 assemblies 9

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5-15

i i

i Table 5.5

'l POOL AND MAXIMUM CLADDING TEMPERATURE AT THE .

INSTANT FUEL ASSEMBLY TRANSFER BEGINS ;l I l'

. Coincident Pool Cladding Temperature, 'F Case No. Temperature, 'F Bulk Local _

1 236.6 122.6 186.0 ,

2 236.6 122.8 186.2 '

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v.0 y.0Poweg,re,ru/s53,,10..

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. g b es iso M Z X m

> M 0" f* 3 g'mP4>$o

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• Tl llp Ic m

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d ae em h o" . *

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'o 2 0 4 .

I, a .e

~-

• o e

I ACTUAL OUTLINE CF POOL L , ,

ACTUAL OUTLINE F RACK ASSEMBLY ASSUMED.ADDED r},FUEL ASSEMBLIES l.

'* RACK ASSEMBLY ,

l

+t+1 l, +++

+++__

l 1 A w

!i DEALIZED OUTLINE.

IDEAllZED OUTLINE OF RACK ASSEMBLY OF POOL BOUNDARY

g

'l -

FIGURE 5.5 IDEALIZATION OF RACK ASSEMBLY

4

,r-< ,

)

f n V h il"

/l /

g - n out

/

/

/

7[T o

i,

&_?

in ~ y' . ,

e N1 3= . ,,

e m e -

H . E e t o

o d . ..

> .

• Q Heat Addition g ,

o .

c' c

V ,,

, L, 3 y o . ..

c . -

t l

./ .

f_

V '

[u yin-

/ u >

1 6 .

, 7 ,

< 0 r~7 i

s'

/ *1 .'

~'

/ \

J l l

' f

. 1 FIGURE 5.4 Thermal Chimney Flow Model .

5-26

)

1 , . , , ,

i

1

\ ,

6.0 STRUCTURAL ANALYSIS ,

i , .' , ,

4 d

The purpose of 'this section is to demonstrate the structural, a . adequacy of the spent fuel rack design under normal and accident loading conditions. The method of analysis presented herein uses a time-history integration method similar to that previously used.in t,

,' I. the Licensing Reports on High Density Fttel Racks for Fermi 2

. r- .

l (Docket No. 50-341), Quay Cities 1 and 2 (Docket Nos. 50-254 and L' 50-265), Rancho Seco (Docket No. 50-312), Grand Gulf Unit 1 (Docket l

1' No. 50-416), Oyster Creek (Docket No. 50-219), V.C. Summer (Docket No. 50-395), and Diablo Canyon 1 and 2 (Docket Hos. 50-275 and 50-323). The results show that the high density spent fuel racks

g. are structurally adequate to resist the postulated stress combinations associated with level A, B, C, and D conditions as

, defined in References 1 and 2.

l1

', 6.1 ANALYSIS OUTLINE The spent fuel storage racks are . Seismic Category I equJpment.

~

,' Thus, they are required to remain functional during and after a i

'I Safe Shutdown Earthquake (Ref. 3). As noted previously, these racks

' are neither anchored to the pool floor nor attached to the side

. walls. The individual rack modules are not interconnected.

Furthermore, a particular rack may be completely loaded with fuel assemblies (which corresponds to greatest rack inertia), or it may be completely empty. The coefficient of friction, y , between the supports and pool floor is another indeterminate factor. According to Rabinowicz (Ref. 4) the results of 199 tests performed on 1 austenitic stainless steel plates submerged in water show a mean value of u to be 0.503 with a standard deviation of 0.125. The p- upper and lower bounds (based on twice the standard deviation) are i thus 0.753 and 0.253, respectively. Two separate analyses are performed for the rack assemblies with values of the coefficient of I

l friction equal to 0.2 (lower limit) and 0.8 (upper limit),

respectively. Analyses performed for the geometrically limiting ,

i e

4-1

l rack modules focus on limiting values oi' the coefficient of friction, and the number of fuel assemblies stored. Typical cases studied are f

0 Fully loaded rack (all storage locations occupied),

. = 0.83 0.2 ( = coefficient of friction) h l 0 Hearly empty rack = 0.8, 0.2 .,

i

, d 0 Rack half full, = 0,8 l .

i The method of analysis employed' is the time-history method. The j pool slab, acceleration ~ data were developed from the response l spectra provided by the Sargent and Lundy Company, Chicago,

'I i Illinois.

The objective of the seismic analysis is to determine the t l t

! structural response (stresses, deformation, rigid body motion, -

etc.) due to simultaneous application of the three independent, ,

I orthogonal excitations. Thus, recourse to approximate statistical ,

summation techniques such as the " Square-Root-of-the-Sum-of-the- ,.

Squares" method (Ref. 5) is avoided. For nonlinear analysis, the .

only practical method is simultaneous application.

'l Pool slab acceleration data are provided for two, earthquakes:

Operating Basis Earthquake (0BE) and Safe Shutdown Earthquake (SSE). Figures 6.1 - 6.3 show the time-histories corresponding to -

the SSE condition. ,

The seismic analysis is performed in three steps, namely: .

1. Development of a nonlinear dynamic model consisting of inertial mass elements and gap and friction elements.

I i

6-2

l l .

, appropriate inertial coupling into the system kinetic i energy. Inclusion of these effects uses the methods of References 4 and 6 for rack / assembly coupling and for j rack / rack coupling (see Section 6.2.3 of this report).

I g. Potential impacts between rack and assemblies are

.r -

accounted for by appropriate " compression only" gap l l.

~

elements between masses involved. -

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

1. The supports are modeled as " compression only" elements l[ for the vertical direction and as " rigid links" for f*

jc' dynamic analysis. The bottom of a support leg is. attached to a frictional spring as described in Section 6.2.2. The ld- cross-section inertial properties of the support legs are

,. computed and used in the final computations to determine i

i; supportleg} stresses.

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

\-

~

k. Inter-rack impact, if it occurs, is simulated by a series of gap elements at the top and bottom of the rack in the two horizontal directions. The most conservative case of i adjacent rack movement is assumed; each adjacent rack is

i. assumed to move completely out of phase with the rack

, being analyzed.

i,

, 1. The form drag opposing the motion of the fuel assemblies l, in the storage locations is conservatively neglected in the results reported herein.

m. The form drag opposing the motion of the fuel rack in the water is also consarvatively neglected in the results reported herein.

• l

i

- n. The rattling of the fuel assemblies inside the storage locatio$s causes the " gap" between the fuel assemblies and the cell wall to change from a maximum of twice the l nominal gap to a theoretical zero gap. However, the fluid t l coupling coefficients (Ref. 8) utilized are based on linear vibration theory (Ref. 9). Studies in the 'q literature show that inclusion of the nonlinear effect d l

I (viz, vibration amplitude of the same order of magnitude ,,

as the gap) drastically lowers the equipment response  ;.

l (Ref. 10). ,

l'

. Figure 6.4 shows a schematic of the model. Six degrees-of- freedom are used to 1,l ack the motion

• of the rack structure. Figures 6.5 and 6.6, respectively, show the inter-rack impact springs a n d f u'e l-assembly / storage cell impact springs. t The production run model for simulating fuel assembly motion ,,

l

, incorporates five lumped masses. The lower mass is assumed to be .

attached to the baseplate and to move with the baseplate. The

four rattling masses are located at quarter height, half height,  !'

f three quarter height, and top of the rack. Two degrees-of-freedom i are used to track the motion of each rattling mass.

4.2.2 Model Description The ab' solute degrees-of-freedom associated with each of the mass locations are identified in Figure 6.4 and Table 6.1. Note that .'

~

for clarity, only the top rattling mass (node ' 2*), is shown in the ,

figure. The remaining rattling masses (nodes 3*,4*, 5*) are ,

located at x = 3/4H, 1/2H, 1/4H, respectively, and are described by translational degrees-of-freedom qg-qp,.

Ug(t) is the pool floor slab displacement seismic time-history.

Thus, there are fourteen degrees-of-freedom in the system. Not shown in-Fig. 6.4 are the gap elements used to model the support legs and the impacts with adjacent racks.

l>

. 4-6

l l

d

. 2. Generation of the equations of motion and inertial coupling and solution of the equations using the " component element time integrat;1on scheme" (References 6 and 7) to determine nodal forces and displacements l' 3. Computation of the detailed stress field in the rack (at the I- critical locatilon) and ,in the support legs using the nodal c forces calcula'ted in the previous step. These stresses are 1

checked against the design limits given in Section 6.5.

f, A brief description of the dynamic model follows.

1 -

l ' , ,. .

6.2 FUEL RACK - FUEL ASSEMBLY MODEL

,[ Since the racks are not anchored to the pool slab or attached to E- the pool walls or to each other, they can execute a wide variety of 4

c- rigid body motions. For example, the rack may slide on the pool I floor (so-called " sliding condition"); one or more legs may f ,, momentarily lose contact with the liner (" tipping condition"); or

,,l the rack may experience a combination of sliding and tipping conditions. The structural model should permit simulation of these kinematic events with inherent built-in conservatisms. Since the Byron racks are equipped with girdle bars to dissipate energy due to inter-rack impact (if it occurs), it is also necessary to model

~

the inter-rack impact phenomena in a conservative manner.

Similarly, lift off of the support legs and subsequent impacts must ~

<I be modelled using appropriate impact element s, and Coulomb friction between the- rack and the pool liner .must be simulated

• by l, appropriate piecewise linear springs. These special attributes of l ,.

the rack dynamics require a strong emphasis on the modeling of the

! linear and nonlinear springs, dampers, and stop elements. The model outline in the remainder of this section, and the model description I in the following section describe tha detailed modeling technique to simulate these effects, with emphasis placed on the nonlinearity of the rack seismic response.

4-3

I I

g, 6.2.1 Outline of Model

a. The fuel , rack structure is a folded metal plate assemblage .i j welded to a baseplate and supported on four legs. The ,

l . rack structure itself is a very rigid structure. Dynamic l

. analysis of typical multicell racks has shown that the ,

l motion of the structure is captured almost completely by o

,f[

the behavior of a six degrees-of-freedom structure; there-fore, the movement of the rack cross-section at any height II

~

is described 1n terms of the six degrees-of-freedom. of the rack base.

,l

b. The seismic motion of a fuel rack is characterized by ,

random rattling of fuel assemblies in their individual .

storage locations. Assuming that all assemblies vibrate y in phase obviously exaggerates the computed dynamic  ;}

loading on.the rack structure. This assumption, however, greatly reduces the required degrees-of-freedom needed to model the fuel assemblies which are represented by two lumped masses located at different levels of. the rack.

l The centroid of each fuel assembly mass can be located, relative to the rack structure centroid at that level, so -

1 as to simulate a partially loaded rack. .

I

c. The local flexibility of the rack-support interface is }

l modeled conservatively in the analysis.

'1

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

floor.

.i

e. The pool floor has a specified time-history of seismic :n accelerations along the three orthogonal directions. -!

i l

i f. Fluid coupling between rack and assemblies, and between f 1

i rack and adjacent racks, is simulated by introducing ,

I i

6-4 ,'

l "e-e-er --

'*'a"sv ee- w wc- --ww._ _.._owwa- w www+ = ww.e eg-*eee-o-e.e--.-=-c- a y- + - - - - - e--n-,pg-yf--*exm wNr

I . .

a . .

,. 6.2.3 Fluid Coupling

, t.

An effect of some significance requiring careful modeling is the I so-called " fluid coupling offact". If one body of mass (m t)

. vibrates adjacent to another body (mass m 2) , and both bodies' are l (' submerged in a frictionless fluid medium, then Newton's e qu.a t'io n s I. of motion for the two bodies have the f'orm:

, c.

i 4

(m i + M 11) X 1 - M12 X2= applied forces on mass mi l -H 21 X 1+ (m 2 + H 22) X 2= applied forces on mass m2 l .

e ' accelerations of mass mi and m 2, X,1 X2 de' note absolute respectively. .

I i' M 11, M 12, M 21, and M22 are fluid coupling coef ficients which depend

<- on the shape of the two bodies, their relative, disposition, etc.

.. Fritz (Ref. 6) gives data for Mij for various body shapes ' and arrangements. It is to be noted that the above equation indicates

.' that the effect of the fluid is to add a certain amount of mass to j

\

the bo'dy (M11 to body 1), and an external force which 'si l proportional to the acceleration of the adjacent body (mass m2).

Thus, the acceleration of one body affects the force field on another. This force is a strong function of the interbody gap, i

reaching large values for very small gaps. This inertial coupling c is called fluid coupling. It has an important effect in rack ,

14 dynamics. The lategal motion of a fuel assembly inside the storage

. location will encounter this effect. So will the motion of a rack adjacent to another rack. These effects 'are included in the equations of motion. The fluid coupling is between nodes 2 and 2*

in Figure 6.4. Furthermore, the rack equations contain coupling terms which model the effect of fluid in the gaps between adjacent I racks. The coupling terms modeling the effects of fluid flowing

-l

' between adjacent racks are computed assuming that all adjacent

!( racks are vibrating 180* out of phase from the rack being 6-7 i

,--,,-,,-,,,-,-.-,_,,_v_m,__,

s, analyzed. Therefore, only one rack is considered surrounded by a hydrodynamic massi computed _ as if there were a plane of symmetry located in the middle of the gap region.

1  !

l Finally, fluid virtual mass is i~ncluded in the vertical direction  !

vibration equations of the rack; virtual. inertia is.also added to ,

I the governing equation corresponding to the rotational ,;

degree-of-freedom, qs(t).

I 1 1 4 6.2.4 Damping In reality, damping of 'the rack motion arises from material hysteresis (material damping), relative intercomponent motion in i structures (structural damping), and fluid drag effects (fluid .

damping). In the analysis, a maximum of 45 structural damping is

,h imposed on elements of the rack structure during SSE seismic ,;

simulations. This is in accordan,ce with the FSAR and NRC fi

, guidelines (Ref. 11). Material and fluid damping'are conservatively l

neglected. The dynamic model has the provision to incorporate fluid damping effects however, no fluid damping has been used for this ('

l analysis.

t 6.2.5 Impact -

Any fuel assembly node (e.g. 2*) may impact the corresponding ,

structural mass node 2. To simulate this impact, four .

c omp r e s sio n-o nly gap elements around each rattling fuel assembly node are provided (see Figure 6.6). As noted previously, fluid dampers may~ also be provided in parallel with the springs. The 'I

'o compressive loads developed in these springs provide the necessary data to evaluate the integrity of the cell wall structure and 'l stored array during the seismic event. Figure 6.5 shows the location of the impact springs used to simulate any potential for .

inter-rack impacts. Section 6.4.2 gives more details on these additional impact springs. .

6-8

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

I I, 6.3 ASSEMBLY OF THE DYNAMIC H0 DEL )

i The cartesian coordinate system associated with the rack has the

' i' following nomenclature

! g-j'

, 0 x = Horizontal coordinate along the short. direction of rack rectangular platform ,

i 0 y = Horizontal coordinate along the long direction of the

' I- rack rectangular platform ,

0 z = Vertically upward As described in the preceding section, the rack, along with the base, supports, and stored fuel assemblies, is modeled for the

L' general three-dimensional (3-D) motion simulation by a fourteen f- degree-of-freedom model. To simulate the impact and sliding phenomena expected, 60 nonlinear gap elements and 16 nonlinear 1 friction elements are used. Cap and friction elements, with their

, connectivity and purpose, are presented'in Table 6.2.

If the simulation model is restricted to two dimensions (one horizontal motion plus vertical motion, for example) for the

~

purposes of model clarification only, then a descriptive model of the simulated structure which includes gap and friction elements is shown in Figure 6.7. (Note that only the top rattling mass is shown for clarity.)

1 The impacts between . fuel assemblies and rack sho,w up in the gap elements, having local stiffness X, I in Figure 6.7. In Table 6.2, gap elements 5 through 8 are for the vibrating mass at the top

,[.

l of the rack. The support leg spring rates X 6 are m deled by elements 1 through 4 in Table 6.2. Note that the local compliance

! of the concrete floor is included in X. g To simulate sliding potential, friction elements 2 plus 8 and 4 plus 6 (Table 6.2) are i

shown in Figure 6.7. The friction of the support / liner interface is i modeled by a piecewise linear spring with a suitably large l

6-9

I stiffness Kf up to the limiting lateral load, tN , where N is the l

current compression load at the interface between support and -

liner. At every time step during the transient analysis, the ,

current value of N (either zero for liftoff condition, or a i compressive finite value) is computed. Finally, the support

,I rotational friction springs Kg reflect any rotational restraint ,g that may be offered by the foundation. This spring rate is calculated using a modified Bousinesq equation (Ref. 4) and is 1 included to simulate the resistive moment of the support to counteract rotation of the rack leg in .a vertical plane. This i rotation spring is also nonlinear, with a zero spring constant ,

value assigned after a certain limiting condition of slab moment -

t loading is reached. . .,

The nonlinearity of these springs (friction elements 9, 11, 13, and ,','i ,

15 in Table 6.2) reflects the edging limitation imposed on the base f.

of the rack support legs. In this analysis, this effect is neglected; any support leg bending, induced by liner / baseplate friction forces, is resisted by the leg acting as a beam i' '

cantilevered from the rack baseplate.

l The spring rate K6 modeling the effective compression stiffness -

of the structure in the vicinity of the support, is computed from the equations i i

1

= 1 .1 . ..

K y K1 K2 K3 ,

where: ,

i K1= spring rate of the support leg treated as a j tension-compression member = ESUPPORT

• ASUPPORT/h (h = length of support leg) '

l l

6-10

< , , , . - , - , , . . . , - - , . . . - - - - - . - - - - - - - - - - - - - , - - - - - - - -~~--~--..n--- - - - - - - - - - - , , - ~ ~ - - ~ - - - - -~- - --'-~ ~r w-------'

1 I

e s .

' 2 pool slab K2= 1.05Ec8/(1- v ) = local spring rate of

[ .

(Ee = Young's modulus of concrete, and 8 = length of i

J bearing surface)

K 3= spring rate of folded plate cell structure above support l ,

leg (same form as Kz with E chosen to reflect the local stiffness of the honeycomb structure above the leg)

For the 3-D simulation, all support elements (listed in Table j, 6.2) are included in the model. Coupling between the two horizontal seismic motions is provided both by the of fset of the fuel assembly group centroid which causes the rotation of the entire rack and by the possibility of liftoff of one or more

! support legs. The potential exists for the rack to be supported -

l' on one or more support legs or to liftoff complete'ly during any 1

'- instant of a complex 3-D seismic event. All of these potential events may be simulated during a 3-D motion and have been observed f7L. in the results.

+

c-6.4 TIME INTEGRATION OF THE EQUATIONS OF MOTION 6.4.1 Time-History Analysis Using 14 00F Rack Model 4 i.

L' Having assembled the structural model, the dynamic equations of motion corresponding to each degree-of-freedom can be written by

, using Newton's second law of motion; or by using Lagrange's

< equation. The system of equations can be represented in matrix

. notation as

, [H ] {q } = {Q } + {C }

li f -

I

{Q } is a function of nodal displacements and where the vector j

velocities, and {C } depends on the coupling inertia and the ground acceleration. Premultiplying the above equations by (M ]-1 renders the resulting equation uncoupled in mass.

1 i

l 6-11

I i We haves' {q } = [H ]-1 {Q } + [H ]-1 {G }

As noted earlier, in the numerical simulations run to verify structural integrity during a seismic event, all elements of the l fuel assemblies are assumed to move in phase. This wil1' provide maximum impact force level, and induce additional conservatism in h

the time-history analysis. -

!i 1  ?-

This equation set is mass uncoupled, displacement coupled, and is y ideally suited for numerical solution using a central difference scheme. The computer program "DYNAHIS"* is utilized for this I purpose. .

Stresses in various portions of the structure are computed from -

known element forces at each instant of time. rl 3

j Dynamic analysis of typical multicell racks has shown that the ,

motion of the structure is captured almost completely by the behavior of a six-degree-of-freedom structure; therefore, in this

'1 analysis model, the movement of the rack cross-section at any height is described in terms of the rack base degrees-of-freedom i

(q1(t),...qs(t)). The remaining degrees-of-freedom are associated with horizontal movements of the fuel assembly masses. In this dynamic model, four rattling masses are used to represent fuel 'i assembly movement. Therefore, the final dynamic model consists of I six degrees-of-fre'edom for the rack plus eight additional mass degrees-of-freedom for the four rattling masses. The remaining [g portion of the fuel assembly mass is assumed to move with the rack ,

base. Thus, the totality of fuel mass is included in the ,' ,

l simulation.

l ,

• This code has been previously utilized in licensing of similar racks for Fermi 2 (Docket No. 50-341), Quad Cities 1 and 2 '

(Docket Nos. 50-254 and 265), Rancho Seco (Docket No. 50-312),

Oyster Creek (Docket No. 50-219), V.C. Summer (Docket No. .

50-395), and Diablo Canyon 1 and 2 (Docket Nos. 50-275 and

• 50-323).

6-12

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

=

, 6.4.2 Evaluation of Potential for Inter-Rack Impact

. Since the racks are closely spaced, the simulation includes impact springs to model'the potential for inter-rack impact, especially l'. for low values of the friction coefficient between the support'and I

the pool liner. To account for this potential, yet still retain the simplicity of simulating only a single rack, gap elements were ,

i '-

located at the corners of the rack at the top and at the I

- baseplate. Figure 6.5 shows the location of these gap elements.

Loads in these elements,* computed during the dynamic analysis, are used to assess rack integrity if inter-rack impact occurs.

i4 .

6.5 STRUCTURAL ACCEPTANCE CRITERIA .

i l1 There are two sets of criteria to be satisfied by the rack modules

a. Kinematic Criterion

. This criterion seeks to ensure that the rack is a '

. physically stable structure. Byron racks are designed

. to sustain certain inter-rack impact at designated l, locations in the rack modules. Therefore, physical l stability of the rack is considered along with the localized inter-rack impacts. Localized permanent i' deformation of the module is permissible, so long as the subcriticality of the stored fuel array is not violated.

. b. Stress Limits The stress limits of the ASME Code, Section III, Subsection NF, 1983 Edition are used since this code provides the most appropriate and consistent set of l limits for various stress types and various loading conditions. The following loading combinations are l

applicable (Ref. 1).

l -

6-13

-,r-.,, - , . _ . - - , , - - . . .,___.n.--.,,, _ , . _ _ - - - _ , , , , , _ ---..,,n.,,___ _ _ - , , . . - , . . _ , - - - - - - - , .

i

. i Loading Combination Stress Limit I D+L Level A service limits D+L+To l D+L+To+E '

D + L + Ta + E Level B service limits I' D + L + To + Pf .-

l D + L + Ta + E' Level D service limits '-

D+L+Fd The functional capability .

of the fuel racks should be demonstrated

,f a

where D = Dead weight-induced stresses (including fuel 'l assembly weight) ,

l L = Live Load (0 for the structure, since there are no ,,

moving objects in the rack load path). -

F Force caused by the accidental drop of the d = heaviest load from the maximum possible height .i l Pg = Upward force on the racks caused by postulated ,

! stuck fuel assembly

! E = Operating Basis Earthquake ,.

E' = Safe Shutdown Earthq;ake To = Differential temperature induced loads (normal '

or upset condition)

T* = Differential temperature induced loads 'l' i

(abnormal design conditidns)

The conditions Ta and To cause local thermal stresses to be ,

produced. The worst situation will be obtained when an isolated , ,!

storage location has a fuel assembly which is generating heat at l

6-14

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

.1 l\ ', .

the maximum postulated rate. The surrounding storage locations are

,_ assumed'to contain no fuel. The heated water makes unobstructed contact with the inside of the storage walls, thereby producing the

, maximum possible temperature difference between the adjacent I" cells. The secondary stresses thus produced are limited to the l body of the racks that is, the support legs do not experience the l -

secondary (thermal) stresses.

l 6.6 MATERIAL PROPERTIES 1

l ,. The data on the physical properties of the rack and support l materials, obtained from the ASME Boiler & Pressure Vessel Code, j Section III, appendices, and supplier's catalog, are listed in Tables 6.3 and 6.4. Since the maximum pool bulk temperature (except i

for the full core discharge case) is 150*, this is used as the

' reference design temperature for evaluation of material properties.

l<.

6.7 STRESS LIMITS FOR VARIOUS CONDITIONS '

,1 .

,, The following stress limits are derived from the guidelines of the ASME Code,Section III, Subsection NF, in conjunction with the material properties data of the preceding section.

6.7.1 Normal and Upset Conditions (Level A or Level B) i

! <} a. Allowable stress in tension on a net section

=Ft = 0.6 Sy or .

a

{ Ft= (0.6) (23,150) = 13,890 psi (rack material) l lI ti

! 6-15

- ~

Ft = is equivalent to primary membrane stresses l

s -

Fg = (.6) (27,500) = 16,500 psi (upper part of support feet) i

= (.6) (62,400) = 37,440 psi (lower part of l support feet) I

b. On the gross section, allowable stress in shear is:

}1l F,= .4 S y

(.4) (23,150) = 9,260 psi (main rack body) -

{l F g= (.4) (27,500) = 11,000' psi (upper part of g.

support feet)

= (.4) (62,400) = 24,960 psi (lower part of support feet)

c. Allowable stress in compression, Fa8

. .I

[1 - (U) 2 2C, * ]S y ,,

r = r '

(1)+3

[3 (U)r 8C, ) - ((U )

r 8C,3]

g t

I where

• 2 1/ 2

• Ce = [ (2v E) ] .,

5 y ..

k A/r for the main rack body is based on the full height

, and cross section of the honeycomb. region. Substituting numbers, we obtain, for both support leg and honeycomb regions

• Fa = 13,890 psi (main rack body) -

Fa = 16,500 psi (upper part of support feet)

= 37,440 psi (lower part of support feet)

O 6-16 .

s

d. Maximum allowable bending stress at the outermost, fiber

f. due to flexure about one plane of symmetry:

Fb 13,890 psi (rack body)

Fb = 0.60 Syps=i (upper part of support feet)

= 16,500

= 37,440 psi (lower part of support feet)

-- e. Combined flexure and compression: -

1. a + Umx#b x Umy#b y

+ <1 ~

F, D,Fbx DFy by where:

f, = Direct compressive stress in the section fbx = Maximum flexural stress along y-axis f by = Maximum flexural stress along y-axis

,, C,, = C,y = 0.85

c. f

,, D , = 1 - - *--

F'ex f

a D =1- ,

y 7,ey where:

12: 2E F,,x,,y =

, 23 ( bx,y ,)

bx,y and the subscripts x,y reflect the particular bending plane of interest.

6-17 ,

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

f." Combined flexure and compression (or tension):  :

- l

' f a ,

I bx .

1. y b

< 1.0 .

0.6 S y F F bx by The above requirement should be met for both the direct i tension or compression case. ,

.i 6.7.2 Level 0 Service Limits 1-F-1370 (Section III, Appendix F), states that the limits for the 1.evel D condition are the minim,um of 1.2 (Sy /Ft ) or P (0.75u/Fg) times .the corresponding limits for Level A condition. Since 1.2 Sy is less than 0.7 Su for the rack ,

material, and for the upper part of the support feet, the .i multiplying factor for the limits is 2.0 for the SSE condition for .,

the upper section. The factor is 1.48 for the lower section under ,

SSE conditions. .

li

.I Instead of tabulating the results of these six different stresses as dimensioned values, they are presented in a dimensionless form. l These so-called stress factors are defined as the ratio of the '

actual developed stress to its specified limiting value. With this definition, the limiting value of each stress factor is 1.0 for 08E ,

and 2.0 or 1.48 for the $$E condition. ,

4.8 RESULTS l

, Figures 6.1, 4.2, and 6.3 show the pool slab motion in horizontal x, horizontal y, and vertical directions. This motion is for the SSE earthquake.

Results are abstracted here for a 12x14 module (the largest module) and for an 8 x 14 module (largest aspect ratio).

6-18 9

I

.. A complete synopsis of the analysis of the 12x14 and the 4x14 module subject to the

• SSE earthquake motions is presen'ted in a summary table 4.5 which gives the bounding values of stress factors

~

R1 (i = 1,2,3,4,5,4). The stress factors are defined ass R 3 = Ratio of direct tensile or compressive stress on a not section to its allowable value (note support feet only

- support compression)

R 2 = Rat 1o of gross shear on a not *section to its allowable

,, value ll R3 = Ratio of maximum bending stress due to bending about the x-axis to its . allowable value.for the section Rg = Ratio of maximum bending stress due to bending about the

y-axis to its allowable value l Rs = Combined flexure and compressive factor (as defined in 4.7.1e above) l Rs a Combined flexure and tension (or compression) factor

. (as defined in 6.7.1f above)

!I As stated before, the allowable value of R1 (i =1,2,3,4,5,4) is j 1 for the OBE condition. (except for the lower section of the

. support where the factor is 1.65), and 2 for the 55E.

f* The . dynamic analysis gives the maximax (maximum in time and in space) values of the stress factors at critical locations in the

! rack module. Since these maximax values are subject to minor

(under 5%) varia' tion if the input data (viz rack baseplate height, l 1 cell inside dimension) is perturbed within the range of manufacturing tolerances, the bounding values,,instead .of ,the

, actual values, are presented in Table 6.5. The terms in Table 6.5 l, have the following meaning

I a implies R < 1.0 b implies R < 1.5 R < 1.75 i

I c implies l d implies Ri < 2.0 6-17

It is found that,the results corresponding to SSE are most critical i vis-a-vis the corresponding allowable limits. The results given ,

, herein are for the SSE. The maximum stress factors (Ri ) are below l the limiting value for the SSE condition for all sections. It is I

noted that the critical load factors reported for the support feet j are all for the upper segment of the foot and are to be compared with the limiting value of 2.0.  !!

il Analyses (not included here) have been carried out to show that  ?-

il significant margins of safety exist against local deformation of a the fuel storage cell due to rattling impact of fuel assemblies and -

against local overstress of imp,act bars due to inter-rack impact.

Analyses (not presented here) have also been carried out for the .l OBE condition to demonstrate that the stress factors are below 1.0. Results obtained for all rack sizes and shapes are enveloped '

by the data presented herein. Overturning has also been considered for the cases where racks are adjacent to open areas. '

l e

1 .

. ,t i

, e' e

l .8 l

4-20

I

. ,. 6.9 IMPACT ANALYSES 6.9.1 Impact Loading Setween Fuel Assembly and Cell Wall .

The local stress in a cell wall is estimated from peak impact loads

~~

obtained from the dynamic simulations. Plastic analysis is used to obtain the limiting impact load that can be tolerated. Inciuding a

- safety margin of 2'.0, we find that the total limit load for the

. number of cells (NC) ist a ,

i QL = 9030 NC 4.9.2 Impacts Between Adjacent Racks All of the dynamic analyses assume, conservatively, that adjacent ,

. racks move completely out of phase. Thus, the highest potential

, for inter-rack impact is achieved. Based on the dynamic loads obtained in the gap elements simulating adjacent racks, we can I study rack integrity in the vicinity of'the impact point. The use of framing material around the top of the rack allows us to

~

i withstand impact loads of 57000 lbs. at a corner of the rack prior to reaching the fully yielded state.above the active fuel region.

It is shown that rack-to-rack corner impact loads' can be accommodated. Thus, impacts bet; ween racks can be accommodated

. without violating rack integrity. .

I d-21

l  !

l 4

4

, 4.10 WELD STRESSES 4

The critical weld locations under seismic loading are at the connection of the rack to the baseplate and in the support leg welds. For the rack welds, the allowable weld stress is the ASME code value of 24000 psi (Table NF-3324.5(a)-1, Subsection NF). For -

the support legs, the allowable weld stress is goveriod by the .

levels outlined in Section 6.7 (see NF-3324.5 for- partial  ;

penetration welds). ,

! Weld stresses due to heating of an isolated hot cell are also computed. The assumptio'n used is that a single cell is heated, -

over its entire length, to a temperature above the value associated with all surrounding cells. No thermal gradient in the vertical direction is assumed so that the results are conservative. Using e l

j the temperatures associated with this unit, we show that the skip .  :

welds along the entire cell length do not exceed the allowable ,

' value for a thermal loading condition. ,

i f 4.11

SUMMARY

OF MECHANICAL ANALYSES -

The mathematical model constructed to determine the impact velocity 1 l

l of falling objects is based on several conservative assumptions, '

such as:

1. The virtual mass (see ref. 8-10 for further mat'e rial on this subject) of the body is conservatively assumed to be equal to its displaced fluid mass. Ev1.dence in the ,

literature (Ref. 12), indicates that the virtual mass can be many times higher.

'~

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

G 4

i . l l

6-22 l1 i

i i

l e

i

, l .

. 3. The drag coefficients utilized in the analysis a'r e the lower bound values reported in the literature (Ref. 13).

f' In particular, at the beginning of the fall when the velocity of the body is small, the corresponding Reynolds number is low, resulting in a large drag coefficient.

l- falling bodies are assumed to be ricid for the

4. The purposes of impact stress calculation on the rack. The lr

' solution of the immersed body motion problem is found analytically. The impact velocity thus computed is used i

to determine the maximum stress generated due to stress l ,,

wave propagation.

~

With this model, the fo11'owing analyses are performed l

a. Dropped Fuel Accident I i .

- A fuel assembly (weight = 1545 pounds with control rod assembly) is dropped from 34 inches above the module and' impacts the base. The final velocity of the dropped fuel I;-

assembly (just prior to impact) is calculated and, thus, the total energy at impact is known. To study baseplate integrity, we assume that this energy is all directed baseplate in shear and thus toward punching of the work done by the supporting shear transformed into

stresses. It is determined that shearing deformation of l'-

the baseplate is less than the thickness of the base-i 1 - plate so that we conclude, that local piercing o f - t h.e baseplate will not occur. Direct impact with the pool The subcriticality of the adjacent liner does not occur.

fuel assemblies is not violated.

l

,4-23

e

,y

, b. Oropped Fue_1 Accident II e .A '

One fuel assembly drops from 36 inches above the rack and .

hit s: the. top of the rack.' Permanent deformation of the rack is found to be limited to the top ragion such that 4

, the rack cross-sectional geometry at the level of the top .;

of tho active fuel (and below) is not altered. The region , l of local ' permanent deformation does not extend below 6

. inches from the rack top. An energy balance approach !,s

, ,, '$.l-used here to obtain the results.

,)

c. 3ammed Fuel-handlino Equipment A 4000-pound uplitt force is applied at the top of the -lJ rack at the " weakest" storage location; the force is .

assumed to be applied on one wall of the storage cell .ll boundary. as an upward shear force. The plastic deformation is found to be limited to the region well

]l above the top of the active fuel.

1 i, These analyses prove that the rack modules are engineered to

~ '

provide maximum safety against all postulated abnormal and accident cond1tions.,

]'

6.12 DEFINITION OF TERHS USED IN SECTION 6 .!

51, 52, $3, 54 Support designations .

j pi Absolute degree-of-freedom number i r

41 Relative degree-of-freedom number 1 u Coefficient of friction U1 Pool floor slab displacement time history in the 1-th direction x,y coordinates horizontal direction z coordinate vertical direction 6-24 i

f ,

.---,__..m.---- - , . . - - - -

r - . - - - - - - --

KI Impact spring between fuel

,t ,

assemblies and cell Kr ,

Linear component of friction spring Kg Axial spring of support leg 3

locations .

N Compression load in a support foot

~

Kg Rotational spring provided by the

,. pool slab Subscript i When used with Il or X indicates direction (i = 1 x-direction, i = 2

- y-direction; 1 = 3 z-direction)

I l.

'e b

4 ie J. -

l l ..

l ..

T 1.

ie I

4-25 I ~ --- - - - _ - . _ _ _ , _ . _ _ _ _ _ , . _ _. __ _ . .

REFERENCES TO SECTION 6 7 .

1. USNRC Standard Review Plan, NUREC-0800 (1981).

2. ASME Boiler & Pressure Vessel Code,Section III, Subsection NF (1983). ..

. 3. USNRC Regulatory Guide 1.29, " Seismic Design Classification," 1; -

Rev. 3, 1978. ]

4. " Friction Coefficients of Water Lubricated Stainless Steels ,

P for a Spent Fuel Rack Facility," Prof. Ernest Rabinowicz, HIT, a report for Boston Edison Company, 1976.  !!

5. USNRC Regulatory Guide 1.92, " Combining Modal Responses and ',

j Spatial Components in Seismic Response Analysis," Rev. 1, j February 1976.

6. "The component Element Method in Dynamics with Application to ~

Earthquake and Vehicle Engineering," 5. Levy and 3.P.D.

Wilkinson, McGraw Hill, 1976. ,.

7. " Dynamics of Structures," R.W. Clough and 3. Penzien, McGraw '

J -

Hill (1975). .

8. " Mechanical Design of Heat Exchangers and Pressure Vessel .;

Components," Chapter 16, K.P. Singh and A.I. Soler, Arcturus Publishers, Inc., 1984. 'i l

~l

9. R.3. Fritz, "The Effects of Liquids on the Dynamic Motions of Immersed Solids," 3ournal of Engineering for Industry, .,

Trans, of the ASME, February 1972, pp 167-172.

10. " Dynamic Coupling in a closely Spaced Two-Body System

! Vibrating in Liquid Medium: The Case of Fuel Racks," K.P.

Singh and A.I. Soler, 3rd International Conference'on Nuclear .!

Powe,r Safety, Keswick, England, May 1982.

f

11. 11SNRC Regulatory Guide 1.61, " Damping Values for Seismic '

Design of Nuclear Power Plants," 1973.

12. " Flow Induced Vibration," R.D. Blevins,'VonNostrant (1977).

13. " Fluid Mechanics," M.C. Potter and 3.F. Foss, Ronald P r'es s ,

p 459 (1975). .

1 l

I  ;

! 6-26 4

,. - - . - . - - - - - , , , - - ,e,,-, m,-

e l

,. Table 6.1 1

.' DECREES OF FREEDOM Displacement Rotation

<. Location u, u y

ug 6, t y

O g

(Hode)

, I

1. .

1 p1 P2 P3 44 45 qs 1* Point 1* is assumed fixed to base at X BsV B , Z=0 2 point.2 is assumed attached to rigid rack at the top most point..

~

2* P7 ps

! i pi = qi(t) + U i(t) o Otter p g, plo Rattling { p it, p12 Hode points 3*, 5*, 5' '

Hasses p t s, p ig b

I l

l. -

I o

k i

I 6-27

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

Table 6.2 i NUMBERING SYSTEM FOR CAP ELEMENTS AND FRICTION ELEMENTS l I. Nonlinear Springs (Cap Elements) (60 total)

Humber Node Location Description i

1 Support 51 Z compression only element .i 2 Support S2 Z compression only element 3 Support $3 Z compression only element  ;

4 Support $4 Z compression only element s*

5 2,2+ X rack / fuel assembly impact element

{ 6 2,2+ X rack / fuel assembly impact element ,I 7 2,2+ Y rack / fuel assembly impact element '

8 2,2+ Y rack / fuel assembly impact element U 9-20 Other rattling masses ,' l 21 Bottom cross-section Inter-rack impact elements of rack (around edge)

Inter-rack impact elements l y

. Inter-rack impact elements

• Inter-rack impact elements 1

• Inter-rack impact elements

- Inter-rack impact elements '

-

• Inter-rack impact elements 40 Inter-rack impact elements y

.o 41 Top cross-section Inter-rack impact elements i

= of rack (around edge) Inter-rack impact elements l

- Inter-rack impact elements '

'

• Inter-rack impact elements

- Inter-rack impact elements ,,

• Inter-rack impact elements

. Inter-rack impact elements -

60 Inter-rack impact elements 1

t II. Friction Elements (16 total) .I

~

Number Node Location Description 1 Support 51 X direction support friction "'

2 Support $1 Y direction friction l

3 Support S2 X direction friction '.

4 Support $2 Y direction friction -

5 Support S3 X direction friction .

6 Support S3 Y direction friction 7 Support S4 X direction friction ,

t 8 Support $4 Y direction friction ,

l 9 S1 X Slab moment l

10 51 Y Slab moment 11 52 X Slab moment J

12 52 Y Slab moment 13 S3 X Slab moment ,

14 S3 Y Slab moment 15 S4 X Slab moment 16 54 Y Slab moment 6-28 t.

.i *

- Table 6.3 RACK MATERIAL DATA

. Young's Yield Ultimate Modulus Strength Strength Material E (psi) Sy (psi)

Su (Psi) 304L S.S.,

27.9 x 10' 23150 68100 Section III Table Table Table

. Reference I-6.0 I-2.2 I-3.2

~

Table 6.4 SUPPORT MATERIAL DATA t

I Young's Yield Ultimate

'  ! Material Modulus- Strength Strength 1 SA-351-CF3 k7.9 x 10' psi 27,500 psi 68,100 psi u (upper part of support feet) *

? IT 8 2 SA-217-CA15 27.9 x 10 62,400 psi 90,000 ps1

[-

(lower part of support feet) i i

6-29

l .

e

. Table 6.5 BYRON RACKS - BOUNDINC VALUES FOR STRESS FACTORS Stress Factorst R R2 R3 Rg Rs Rs I

Run No. (Upper values for rack base - lower values fo'r support feet)

C012, SSE a a a a . a a y= .8, full 12 x 14 a a b b b b

. C013, SSE a a a a a a e p= .2, full i$ 12 x 14 * * * * * *

~ ~

t C014, SSE a a a a a a

, W= .8, 16 a a a a b b cells filled '

12 x 14 i The terms a, b, c and d imply the stress factors R1 (i = 1,2.. 6) are bounded by the following limiting values:

at 1.0 , .

b: 1.5 c: 1.75 di 2 i

... L'

' ' " ~

.? *

.. ... :O  :. -~ .. ._ .

i

-~ -'

, .. . .r. .- . ., ..4 . .. . ..-. .

9 . ~ . - ~, ,.

I Tablo 6.5 (centinesd)

BYRON RACKS - BOUNDINC VALUES FOR STRESS FACTORS .

Stress Factors Ri R2 R3 Rg R5 Rs

Run No. (Upper values for rack base - lower i values for support feet) l .

C015, SSE a a a a a a y= .2, 16 a a a a a a

! cells filled .

12 x 14 I

l C016, SSE a a a a a a es y= .8, 1/2 Full O Pos. X Quadrant a a a a b b a

12 x 14 i

a i

C017, SSE a a a a a a y= .2, 1/2 Full l Pos. X Guadrant * * *

• 12 x 14 C018, SSE a a a a a a p= .8, a a b b c c C1 Edge Rack

Table 6.5 (continued)

BYROM RACKS - BOUNDINC VALUES FOR STRESS FACTORS i Stress Factors

?

R R2 Rg R3 R5 RE i

2 Run No. (Upper values for rack base - lower values for support feet)

C019, y= .2 a a a a' a a C1 Edge Rack a -

a a a a a 12 x 14 m CO20, SSE a a a a a a O p= .8, Full m a a b b c 8 x 14 c CO21, SSE a a a a a a y= .2, Full a a a a a a 8x 14 , ,

l CO22, SSE a a a a a a

"- *2' a a a a b b 11 Cells Filled 8 x 14

• M

' ' * * ~~

_ . _ _ .___ .__._ .t .  :  :..~

~~

. . , m .r. .~, . " .- .. . . ~, ~. ,, i - " ~~

e Table 6.5 (continued)

BYRON RACKS - BOUNDING VALUES FOR STRESS FACTORS ,

Stress Factors R R 2- Rg R3 R5 Rs Run No. (Upper values for rack base - lower ~

values for support feet)

CO23, SSE a a a a- a a

• y= .8 a a a a b ,

b 11 Cells Filled 8 x 14 os e

U CO24, SSE a a a a a a '-

p= .8, 1/2 Full a a a a b b in Heg. X Half -

8 x 14 CO25, SSE a a a a a a p= .2, 1/2 Full , ..

in Neg.X Half 8 x 14 b

I I - I e I s

s - . I I

.)

'l

_ g *1

) e

~

e ,,

m

• u .

]

- 4W .I e

w l'

= - .

ar .

D m

E I.

l W

,l -

.. )

~

t/3 *.

"I l -

i 2m: .e- * -

E s 4 l M w a -

n -

u. - -

2 ~

O CC Q -

2 m q To

.I

.I-i E.

00t 8 00E 8 004*s 00l*0 000*6 Oct*d- 004*d- 00t's- 00F'0 *

(S) NOI1YW31333Y J 6 - 3 f+

l l

. - -~ ~~ ~, ',: . c. -

, , .m .. . .. . . ,

5

=

2 BYRON FUELRACK - SSE NORTH / SOUTH E

~; v 1- . j e l

l

?  !

i~

O ,

g .,. .

I 4' s

e.ees abe.e abe.e do.e abe.e one.e ebe.e As.e ebe.e shoo.e s'see.e shoo.e she.e s'ess.e s m.e FIG 6.2 ,

c.y e

e.

so

. n e

e.

s 7 e

' 4 s

e.

s e

t s

e.

s e

ts e.

s e

' s s

- e.

o ,

o hs

- e. .

e b .

o .

1 3 .

e. 6 L

n A ,

M ,

C '

G I -

I T e.

F .

R ' e .

b E a .

V ,

E l e.

e S b S e L

e.

K e ba C -

._ A '

_ R - -

L e.

E b e r U ,

a F '

N e.

O e o

R 'a Y '-

B e.

e hs e

e

_ c. .

e N

i gi g i E- g- 3 O E

• R Y

D

? .

I

~

CouplinD Elements I

u 2

e f, e d

• - - =

48

/g 2*

47 l l

. 43 1 H r . I

<. =

I AY

=

• { l

. l 1 -

AX  :

S4

[ p .

~

j r}w - .1l'X B r}r N- y -_9

. / f 94 YB 1 f 45 l\ upport S

S1 S2 + h

. 7)n nr x

,[ XB, YB - Location of fuel rod q, group mass centroid - relative to centerline of fuel rack r .

/

I l t.

t FIGURE s.4 Dynamic Model 6-37 .

- TYP.TOPIMPACT -

EtWWENT_

.N . D '

, .Y^fr^

.- f Q( ,

Q

. e e

4

_BACK STRUCTURE ,i TYP. BOTTOM IMPACT ELEMENT i,

H n ~s .

W

.e@_

M W

^~f ,

M

/,.. _

. i y

i #

l .?

! FIG. 6.5 '

G AP. ELEMENTS TO SIMUL ATE INTER-R A CK IMPACTS 6-38 I

l f

1 i

1 l

.. y Impact Springs 3

i .

. 1

' < L

<r d -

, U en

[.. ..

rl

' Mass u

-=A a

g*

Q

. .  ! ! \

<  ! / \ Fluid Dampers

'Lj 4 , (not used in

. this analysis)

Rigid l[

l Frame I.

'I -

i 1

1 .

X g ..

l .

FIGURE.'.6.6 IMPACT SPRINGS AND I

FLUID D AMPERS l1 l

6-39 -

=

'Kz z .

n Il K- w ,- / Kw _,'

{Wl 2 < ~-

W !;^~4 lM ** '

'l 2

i

{L Gap' Elements .

Il l

?,

To Simulate  !

ji Inter-Rack Impacts (4 for 2-D Motion Rack Centroid -

20 for 3-D Motion) (Assumed At H/2)

E .i 4

H a I,i

[ Rigid Rack _

\-

Baseplate ' [i

, Kw K

, j ,. ,

h 3 0 ' H, ,

~

. g Kg Kg h K

j *g^

r!,w K K R [ wl Kg }

FIGURE e.7 Spring Mass Simulation For .

Two-DimensionalMotion '

'6-40 l .

^ ~

7.0 ENVIRONMENTAL EVALUATION

'.' 7.1

SUMMARY

- Installation of high density spent fuel storage racks at the a Byron Nuclear Power Station will increase the licensed storage

. capacity of the spent fuel pool from 1060 to a maximum of 2940 assemblies. Radiological consequences of expanding the capacity have been evaluated with the objective of determining if there is

~

a significant additional onsite or offsite radiological impact relative to that previously reviewed and evaluated (Ref. 1). In addition, radiological impact to operating personnel has been evaluated to ensure that exposures remain as low as is reasonably achievable (ALARA). ,.

4

<. The decay heat loading and the radiological burden to the spent

{, fuel pool water are determined almost entirely by refueling operations. The frequency of refueling operations N and the conduct of refueling are independent of the increased capacity of the storage pool, e xcep t- that the increased capacity should reduce fuel movement and allow continued normal operation. Since

'~ -

the fuel assemblies which will utilize the bulk of the storage

! - capacity (and will ultimately fill all incremental capacity above I - that of the existing design) are aged, their contribution to

.- either the peak decay-heat load or the increased radiological 1 impact, in terms of increased doses, is negligible. A study performed by the NRC (Ref. 2) supports this conclusion. ,

l.. Consequently, the increase in the storage capacity of the spent fuel pool will neither significantly. alter the operating

~

characteristics of the current pool nor result in a measurable change in impact on the environment.

r 7.2 CHARACTERISTICS OF STORED FUEL f

I Because of radioactive decay, the heat generation rate and the in' tensity of gamma radiation from the spent fuel assemblies l

l l

7-1 l

~

~

decreases substantially with decay time. After a cooling time of about-4 years (R f. 3), the decay heat generation rate is less than 2% of the rate at 7 days, the nominal time at which depleted fuel assemblies are transferred to the spent fuel pool. The  !

intensity of gamma radiation is very nearly proportional to the

. decay heat and-decreases with cooling time in a similar manner. 9I The bulk of the heat load is due to freshly discharged fuel; .. '

aged fuel contributes relatively little to the total heat load.  !

Therefore, this expansion will not significantly increase the

, thermal dissipation to the environment. Since the. Intensity of gamma radiation follows the decline in decay . heat generation rate, it is similarly concluded that there will be no significant ,

increase in gamma radiation beyond the pool environment due to the expanded storage. ,

i8 It is important to note .that the aged fuel in the expanded yt storage . capacity will not contain significant amounts of .'

radioactive iodine or short-lived gaseous fission products, since f,

these would have decayed during the storage period. The .,

Krypton-85 which might escape from defective fuel assemblies has ,

been shown to do so quickly (Ref. 2) (i.e., within a short time .

i after discharge from the. core). Further, the residual Krypton-85 l will be contained within the fuel pellet matrix and hence any leakage would occur at very low rates (Ref. 2). Cesium 134/137 I

(Ref. 2) is strongly bound within the fuel pellet matrix and its  !

dissolution rate in water is extremely small. Any Cesium -

dissolved in the pool water is easily controllable in ,the cleanup .

system (demineralizer-lon exchanger resin bed) (Ref. 2). Thus, .

! the planned storage expansion will not significantly increase the l release of gaseous radionuclides. ],

7.3 RELATED INDUSTRY EXPERIENCE [

.i '

Experience with storing spent fuel underwater has been substantial (Refs. 2, 3, and 4). These references show that the 7-2

,i .

1

. pool water activity, normally low, experiences ~ small increase

' -# during refueling , periods, which theN decays rapidly with time.

Typical concentrations (Ref. 5) of radionuclides in spent fuel t- pool water range from 10 "pC1/ml to 10 3pci/ml, with the higher d- value associated with refueling operations. References 2 and 5

<- also state that the ' increase in pool water activity during -

l, refueling can be attributed to 0 Dislodging (sloughing off) of corrosion products

.. on the fuel assembly during transfer and handling operations.

O The possible short-term exposure of fuel pellets to pool water via a cladding defect.

0 Hixing of the ' spent fuel pool water with the higher activity reactor coolant. Upon cessation of the refueling operations, the fuel pool water and the reactor coolant system would be isolated from each

. other, thereby terminating transports of corrosion products from the reactor coolant system. Thus, I- deposition of crud is a function of refueling operations and is not impacted by the expanded storage.

,, , O Furthermore, it has been shown (Ref. 6) that release of

u. fission products from. failed fuel decreases rapidly '

'- after shutdown to essentially negligible levels. The dissolution of exposed fuel pellets (made of UO 2 ) is l' very slow in water at fuel pool temperatures and the

. corrosion of the cladding (Zircaloy 4) at ' spent fuel pool water temperatures is virtually nil (Refs. 2 and 5). Another mechanism available for the release of the gaseous fission products is diffusion through the UO 2

~

pellet. It has been shown that at low water

,, temperatures (<150*F), the diffusion coefficient is i

j extremely small (Ref.7). Therefore, the small increase 2 in activity of the spent fuel pool water is due to either cr0d transport, fission products release, or 11 cross-flow from the reactor coolant system, and is only r a function of refueling operations. The expansion of fuel pool storage capacity will not cause a significant p- increase in doses either onsite or offsite.

l The corrosion properties of irradiated Zircaloy cladding have been reviewed in References 2 and 4 and the h conclusion is drawn that the corrosion of the cladding d

in spent fuel pool water is negligible. The minor incremental heating of pool water, due to the expansion of storage capacity, is far too small to materially

[l affect the corrosion properties of Zircaloy cladding.

7-3

5 I!

d 7.4 BYROM HUCLEAR POWER STATION EXPERIENCE At present there are no spent fuel assemblies in the spent fuel I

pool.

.i

. t 7.5 SPENT FUEL P0OL COOLING AND CLEANUP SYSTEM  :

. r.

It has been shown previously in Section 5 of this licensing 2 report that the cooling system at Byron is adequate to handle the b expected heat loads and maintain the pool temperature peaks j

within acceptable limits. It has been shown in Section 5 that the small increase in heat load due to the storage capacity expansion will neither significantly increase the thermal ,

dissipation to the environment nor increase the propensity for corrosion of the cladding. *l t-It has also been shown that the crud deposition in the spent fuel q

pool water occurs during refueling outages and that the planned 5 expansion will not increase long-term crud deposition." The fuel ,

pool cleanup system (filter and domineralizer) is designed to ,,

maintain fuel pool water clarity and is operatep and maintained in accordance with the Byron operating procedures.

system takes a surface skim from the fuel pool and cleans it The cleanup

}

through a process of filtration and demineralization to prevent ')

i crud buildup on the fuel pool walls at the water-to-air

l interface. -

The spent fuel pool water is sampled and analyzed periodically to .

confirm proper operation of the pool cleanup system. The spent .,

fuel pool modification will not result in a significantly higher i

quantity of solid radwaste. ,;

l 7-4

. I

p d .

.1 7.6 FUEL POOL RADIATION SHIELDING 7.6.1 Source Terms The spent fuel gamma source terms used for the fuel pool

'- shielding evaluation were generated using the point reactor 1

fission product inventory code RIBD. The following' assumptions

- were used in the an.alysis:

i .

• 0 Initial fuel enrichment = 4.2%

0 Net Reactor core power level = 3411 MWt Net ,

,. O Average assembly discharge burnup = 38,000 MWD /NTU t- 0 Power level for average assembly = 17.67 MWt O Power level for peak power assembly = 29.16 MWt L (peaking factor of 1.65) i 0 Burnup for peak power assembly = 33,000 MWD /MTU '

(one 18-month cycle at maximum power)

O Start of Refueling. Process = 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown 4

The peaking factor of 1.65 and burnup for peak power assembly of

!1' 33,000 MWD /MTU for one cycle were chosen to produce the highest possible ,

gamma source term attainable under operational

- conditions. The average assembly burnup and power level were chosen to represent a conservative gamma source term for the spent fuel. The 100-hour decay time is the minimum permitted

[ period before refueling can begin per the Technical Specifications. ,

i The photon energy production rates of an average assembly and of a peak power spent fuel assembly are given in Tables 7.1 and 7.2, respectively.

i f l

7-5

,-~c --- ,,,,,,.--,--.,-,,,,,,rn.,--,,,- -

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

o

- 7.6.2 Radiation Shielding For an equilibrium refueling cycle 84 fuel assemblies will be discharged into the pool starting no earlier than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after ,.

reactor shutdown at a rate of 1 fuel assembly per hour. For a full core discharge, 193 assemblies are discharged. g D

!- To evaluate the adequacy of the shielding capability of the spent y fuel pool Walls, the radiation dose from 104 freshly discharged [{

! fuel assemblies arranged in the A1 Region 1 fuel storage rack Pe (see Figure 2.1) is calculated. It is not necessary to consider the effects of the remaining 89 fuel assemblies in the B1 Region 1 rack on the north pool wall since the A1 rack effectively i shields the adjacent area to the north from the B1 storage rack -

assemblies. The radiation effects on the east / west walls are q l based on 386 assemblies. The photon production rates used in the .I calculations are tabulated in Table 7.2.

,I.

l The north and south perimeter walls of the spent fuel pool are 5 ,

feet thick. The east and west perimeter walls of the pool are 6 l, l

feet thick. The wall separating the spent fuel pool from the transfer canal is 5 feet 6 inches thick. These walls, together 'l

~I with the water and fuel storage racks are incorporated in the dose rate calculations for the adjacent areas. The results of  ;

I the calculations are provided in Table 7.3.

l Except for the area immediately adjacent to the A1 rack, the .!

l maximum calculated dose rates through t h e 'p o o l walls are less than the currently designated radiation level limit of 20-mrem /hr y for these areas. The area adjacent to the A1 rack is normally occupied only briefly during the local and integrated leak rate

~

tests which are conducted every 5 years. Access to the area is controlled.

f 7-6

.e --,------ - . . . - - . . , , , _ _ - , , , _ , . , . , , _ , , , - - - - - - - - , - . . - . . _ _ -..,,.--n -

I 1

i

., The dose rates in Table 7.3 are considered an upper limit since they are calculat.ed for freshly discharged fuel. The dose rates I

will reduce by 'a factor of six, 60 days after the fuel is I discharged into the pool. .

The above discussion ind1 cates that the shielding available

. around the spent fuel pool is adequate for installation of the

, high density storage racks.

l The radiation dose level at the side of the pool and on the spent l' fuel pit crane bridge due to the transfer of a peak power fuel assembly are 2.5 mrem /hr and 2.0 ares /hr, respectively. These i calculations assume a minimum of 10 feet of water cover over the

active fuel. If the transfer of fuel assemblies into,the spent fuel cask pit becomes necessary, the water cover will be less

! l- than 10 feet over the active fuel, however, the dose rate during 1 ,this fuel movement will be much lower than for the transfer of

,s, the peak power assembly noted above since the radiation level' of

, the fuel assembly will have had time to decay to a level which i ,! would more than compensate for the loss of water cover shielding.

l l.

I Since the fuel transfer operation normally lasts less than 4 days (88. assemblies at 1 assembly per hour), the above radiation field r does not create exc'essive operator exposure.

l.

4 . 7.7 RADIOLOGICAL CONSEQUENCES i

The design, basis fuel handling accident (dropped assembly) in the

,1 Fuel Handling Building in Section 15.7 of the FSAR was reviewed for possible effects on radiological dose consequences. The review determined that the conclusions in the FSAR were still valid and that offsite radiological dose consequences were well f within 10CFR100 limits.

J f

I 7-7

3,

. 54

., 7.8 RERACKING OPERATION i

Installation of the fuel racks will include removal of the existing racks, making minor pool modifications, and cleaning and '

installing the new racks. The existing racks are bolted to the pool floor. T; s:

Th,e new racks will be cleaned prior to installation. The fuel y handling building overhead crane will be used to place the racks ,)

in the pool. This effort is scheduled to'be performed prior to ,

the first refueling for Unit 1, which will allow a " dry" f

installation with no water or spent fuel in the pool. In this case, the existing fuel racks will not have been exposed to spent 'l I fuel and will only be nominally contaminated, if at all. I Therefore, doses to individuals involved in the raracking will be 'l I

negligible. d I

If there is a de' lay' in installing the high density racks until ,,

after the first refueling, then a " wet" installation will be required. All pool modifications that can be completed prior to filling the pool with water will be done to minimize underwater work. Although divers may be needed for some tasks, all of the

. work associated with the installation will be sequenced to minimize potential radiation exposure of personnel due to the 'i spent fuel located in the pool. ALARA considerations will be -I fully incorporated in the installation ~ procedures for this ,

condition. If the fuel. handling building overhead crane is used .

over the pool, electri~ cal interlocks will be adjuste.d on the l

.\

i crane to preclude carrying racks over any stored fuel assemblies. j Exact disposition of the existing racks has not been determined. I They will be decontaminated and/or packaged and disposed of in accordance with the applicable Federal and State regulations.  ;

1 e

i 78 -

i i

7.9 CONCLUSION

S .

Based upon the industry experience and evalua'tions discussed in

~

l previous sections, the following conclusions are made:

l <

l Minor increases in radiological burden to the pool

. water, if any, can be ' adequately handled by the fuel pool cleanup system (filter and demineralizer), thereby

l. maintaining the radionuclide concentration in the water

;, at an acceptably low level.

1 No appreciable increase in solid radioactive wastes t (i.e., filter. media and domineralizer resin) is ,

anticipated.

lI- No increase in release of radioactive gases is expected l.

l since any long-lived inert radioactive gas potentially .

t

, available for reluase (i.e. Kr-85) will have leaked from

, the fuel either in the reactor co~re during operation or l4 during the first few months o,f residence in'the pool.

j Further, Vol. 1, ' Reference 3 (pp. 4-16) has shown

~

airborne in activity to be considerably lower than that l allowable by Table 1 of 10CFR Part 20, Appendix B.

? Therefore, the planned expansion will not significantly increase the release of radioactive ga'ses.

. The existing spent fuel pool cooling system will keep the pool, water temperature at an acceptable level (See l i, Section 5, Thermal-Hydraulic Considerations).

t The existing radiation protection monitoring systems and program are adequate to detect and to warn of any q unexpected abnormal increases in radiation level. This provides sufficient assurance that personnel exposures g

can be maintained as low as is reasonably achievable.

l 1-9

,-.,,,,-,.--,.,_,,n,..,,----vr- - ,

I

. . . t

, For a dr,y reracking operation, radiation exposures will 'r ,

be extremely low. .If the reracking occurs after the ,

. first' refueling, procedural controls- and necessary precautions will be taken to reduce radiation exposure to as low as is reasonably achievable, and hence,1 i radiological impact will be minimized. ,

, it - '

Expanding the storage capacity of the spent fuel , pool s, will not significantly increase the onsite or offsite II-radiological impact above that of the currentl'y .

I authorized storage capacity, nor is any significant )

increase in environmental radiological 'or nonradiological'. impact anticipated. .l ,

'l .

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9 7-10 1

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REFERENCES TO SECTION 7 I

1. . FSAR, Byron Nuclear Power Station.

2. - NUREC-0575, " Handling and Storage' of Spent Light Water

' Power Reactor Fuel," Vol. 1, Executive Summary and Text, USNRC, August 1979.

I, -

3. HUREC-0800, USNRC Standard Review Plan, Branch Techni-

<* cal Position.ASB9-2, Rev. 2, July 1981.

I

4. 3. R. Weeks, " Corrosion of Haterials in Spent Fuel i

Storage Pools," BNL-NUREC-2021, July 1977.

5.' A. B. Johnson, 3r., " Behavior of Spent Nuclear Fuel in

(,._ Water Pool Storage," BNWL-2256, September 1977.

3. H. Wright, "E'xpected Air and Water Activities in the l

'. 6.

Fuel Storage Canal," WAPD-PWR-CP 1723 (with Addendum),

undated.

l 7. ANS 5.4 Proposed Standard, "Hethod for Calculating the l i, Fractional Release of Volatile Fission Products from

. Oxide Fuel," American Nuclear Society, issued for

d- review, 1981. .

8. " Licensing Report on High Density Spent Fuel Racks for l 1 Quad Cities, Units 1 and 2," Docket Nos. 50-254 and l 50-265, Commonwealth Edison Company, June 1981.

. p.

9. " Licensing Report for High Density Spent Fuel Storage l

Racks," Rancho Seco Nuclear Generating Station, Sacra-

. me'nto Municipal Utilities District, Docket No. 50-312, June 1982.

10. Final Safety Analysis Report, Limerick Cenerating Station l

,l Units 1 and 2, Section 9.1

11. Safety Evaluation Report Related to the Operation of t* Limerick Cenerating . Station Units 1 'and 2, NUREC-0991, j August 1983.

12. Source Term Data for Westinghcuse Pressurized Water l,, Reactors, WCAP-8253, July 1975.

7-11 l

-- -_-___ =.

t

. i Table 7.1 i g'

PHOTON ENERGY PRODUCTION RATES OF Il 9 AN AVERACE SPENT FUEL ASSEMBLY "i l

Photon Energy Production Rate '!

(MeV) Photon Energy (MeV /sec) , , .

f.

1.50E-002' 2.38E+016 . .

2.50E-002 9.64E+015 3.50E-002 1.52E+016 7, 5.50E-002 9.94E+015 ,gi ~

8.50E-002 1.08E+016 1.50E-001 1.75E+016- .

. 2.50E-001 4.82E+015 I' 3.50E-001

• 9.70E+015 5.60E-001 4.72E+016 9.10E-001 , 4.10E+016 l 1.35E+000 1.50E+016 .

1.80E+000 2.07E+014 2.20E+000 4.09E+014

• 2.60E+000 4.96E+014

• 3.00E+000 2.05E+013

. TOTAL Z.03E+017

.  ; f'

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7-12

. __-.-..._._._,...t__

i 1

i

_ Table 7.2 P'H0 TON ENERGY PRODUCTION RATES OF PEAK SPENT FUEL ASSEMBLY Photon Energy Photon Energy Production Rate (MeV) (MeV/sec) .

a. 1.50E-002 3.92E+016 -

2.50E-002 1.59E+016 1- 3.50E-002 2.50E+016 5.50E-002 1.64E+016 8.50E-002 1.79E+016

c. 1.50E-001 2.88E+016 2.50E-001 7.96E+015

- 3.50E-001 1.60E+016 5.60E-001 -

7.78E+016 9.10E-001 6.77E+016

. i.. 1.35E+000 2.47E+016 1.60E+000 3.41E+014 2.20E+000 6.75E+014 j~ 2.60E+000 8.18E+014 3.00E+000 2.60E+013

, TOTAL 3.3?E+017 s.

l

.9 l .

p P-i j

(

l l l' 7-13 ,

l,

t Table 7.3 CALCULATED DOSE RATES IN AREAS A03ACENT TO THE SPENT FUEL POOL -

I High Density Rack Location Dose Rate (mrem /hr) . ,

n 54.0 Floor el. 401 ft, 0 in., areas adjacent to the north walls' ,

Floor el. 426 ft, 0 in., areas 2.3 i adjacent to the edge of the pool. .

r Floor el. 401 ft, 0 in., spent fuel 4 .'0 pool heat exchanger area **

Fuel transfer canal ** -

20.0  !

.I

'I I a

• A design water gap -of 4-7/8 inches between the high density rack and the wall is used. {'

• • A design water gap of 4 inches between the high. density rack

, and the wall is used. -

q s.

4 O

O

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7-14 8 e

---

1. _ _ _ _ _ _ _ _ _ . _ _

i 4

I

.,. 8.0 IN-SERVICE SURVEILLANCE PROGRAM FOR'BORAFLEX

'S NEUTRON' ABSORBING MATERIAL I 8.1 PROGRAM INTENT a

A sampling program to verify the integrity of the neutron

i. absorber material employed in the high density fuel racks in the

,, long-term environment is described in this section.

The program is conducted in a manner which allows access to the representative absorber material samples without disrupting the integrity of the entire fuel storage system. The program is tailored to evaluate the material in normal use mode and to

<~

forecast future changes using the data base developed.

~

8.2 DESCRIPTION

OF SPECIMENS

, L- .

J. The absorber material used in the surveillance program,

[ henceforth referred to as poison, is representative of the

),f material used within the storage system. It is of the same composition, produced by the' same method, and certified to the same criteria as the production lot poison. The sample coupon is

~

of similar thickness as the poison used within the storage system, and not less than 4 by 2 inches on a side. Figure 8.1 shows a

~ '

typical coupon. Each poison specimen is encas'ed in a stainless ,

1; steel jacket of an ide,ntical alloy to that used 1n the storage 1 system, formed so as to encase the poison, material and fix it in

, a position and wit,h tolerances similan to the design used for the

[ storage system. The jacket has to be closed by tack welding in such a manner as to retain its form throughout the test period and still allow rapid and easy opening without causing mechanical damage to the poison specimen contained within. The jacket f should permit wetting and venting of the specimen similar to the actual rack environment. -

(

O

- 8-1 ,

Tl i

3 8.3 SPECIMEN EVALUATION After the removal of the jacketed poison specimen from the cell at a designated time, a careful evaluation of that specimen l should be made to determine its actual condition as well as its ..

apparen't durability for continued function. Separation of the .,

poison from the stainless steel specimen jacket must be performed ,i carefully to avoid mechanical damage to the poison specimen.

I 7,,

Immediately after the removal, the specimen and jacket section ,

should be visually examined for any effects of environmental exposure. Specific attention should be directed to the examina- 'I tion of the sthinless steel jacket for any evidence of physical '

~ degradation. Functional evaluation of the poison material can be 'I accomplished by the following measurements: '

0 A neutron radiograph of the poison specimen aids in, $

the determination of the maintenance of uniformity of j

the boron distribution.

O Heutron att'enuatio'n measurements will allow evaluation v crf the continued nuclear' effectiveness of the poison.

Consideration must be given in the analysis of the l attenuation measurements for the level of. accuracy of (:

l such measurements, as indicated by the degree of l repeatability normally observed by~the testing agency. q l 'O A measurement of the hardness of the poison material will establish the continuance of physical and .

structural durability. The hardness acceptability '

criterion requires that the s'pecimen hardness will not ..

reduce the hardness listed in .the qualifying tegtg document for laboratory test specimen irradiated to 10  :!

rads. The actual hardness measurement should be made after the specimen has been withdrawn from the pool and

• allowed to air dry for not less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to allow .

for a meaningful correlation with the pre-irradiated L sample. #

0 Heasurement of the length, the width, and the average )

thickness and comparison with the preexposure data will -.

Indicate dimensional stability within the variation range reported in the Boraflex laboratory test reports. )

A procedure will be prepared for execution of the test procedure ,

and interpretation of the test data.  !

i I

8-2 ,

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!~ o 040"THK FILLER A-

{ FOR-REGION 2 s 1- A ( 304 L S.ST. ) .

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FIG.' 8'.1 - TEST COUPON 8-3 .

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i l

l l 9.0 COST / BENEFIT ASSESSMENT i

,' A cost / benefit ssessment has been prepared in accordance with the requirements of Reference 1,Section V - Part 1. The l

- assessment demonstrates that the installation of high density spent fuel storage racks is the most advantageous means 'o f

( i. handling spent fuel.

g The material is presented merely for informational purposes. It is CEC 0's position that an environmental impact statement need i' not be prepared because , the installation of high density fuel racks provides no significant impact on the environment. NRC

'~

precedent establishes that alternatives and economic costs need

{

not be discussed when there is no significant environmental impact. However, for completeness, alternatives to reracking for

. additional spent fuel storage capacity are discussed in Section 9.3.

i. i.

i

<. ~

9.1 SPECIFIC NEEDS FOR $ PENT FUEL STORACE C

Disposal of Byron nuclear fuel is scheduled to be carried out by i

1" the Department of Energy in or after 1998 in accordance with Public Law 97-425, Nuclear Waste Policy Act of 1982. As Byron l' spent fuel may not be accorded a high priority under the DOE ,

I. Program, CECO is seeking to provide a spent fuel storage capacity

i. to support approximately 25 years of nominal operation. No other contractual arrangements exist for the interim storage, or r'eprocessing of speat fuel from Byron Nuclear Power Station. ,

Therefore, increased storage capacity in the spent fuel pool is the only viable option under consideration. Table 1.1, the Fuel

( Discharge Schedule, indicates that with the high density spent fuel racks, loss of full core discharge capability (FCDC) will f occur in 2011, 17 years beyond the current capability and 13 years beyond the scheduled repository fuel receipt date, per the g

DOE Mission Plan.

! 9-1 f

d 3 .9.2 COST OF SPENT FUEL STORAGE

, . .t The design and manufacture of the spent fuel storage racks will be undertaken by the organizations described in Section 1. It is expected that the t.otal project cost will be between $2.6 and

$2.9 million. p i

9.3 ALTERNATIVES TO SPENT FUEL STORAGE q U

CECO has considered the various alternatives to the proposed .

I onsite spent fuel storage. These alternatives are discussed  ;

below: -

l'

a. Shipment of Fuel to a Reprocessi~ng or Independent Spent Fuel Storage / Disposal Facility No commercial spent fuel reprocessing facilities are o presently operating in the United States. CECO has made contractual arrangements whereby spent nuclear fuel v, and/or high level nuclear waste will be accepted and 'l disposed of by the .U.S. Department of Energy. However, such services are not expected to be available before ,

~

1998.* The existing Byron spent fuel storage capacity will not provide full core discharge capability beyond

[ '

1994. Spent fuel acceptance and disposal by the-Department of Energy is not, therefore, an alternative *1 ,

to increased onsite pool storage capacity. .i

b. Shipment of Fuel to Another Reactor Site ,

r Shipment of Byron fuel to another reactor site could provide short-term relief to the storage capacity problem. However, . transshipment of spent fuel merely

(

( serves to transfer the problem to another site and does -

not result in any additional net long-term storage '

capacity. Accordingly, CECO does not consider the transshipment of spent fuel to be an appropriate . I alternative to high density spent fuel storage at the site.

c. Not Operating the Plant after the Current Spent Fuel Storage Capacity is Exhausted As indicated in NUREG-0575, " Final Environmental Impact Statement on Handling and Storage of Spent Light '

Water Power Reactor Fuel," (Ref.2) the replacement of 9-2 6

g .

J. ,. .

,.,, nuclear power by coal generating capacity would cause

. . excess n,ortality to rise from 0.59 - 1.70 to ,15 - 120 a . per year- for 0.8 GWY(e). Based on these facts, not operating the plant or shutting down tht plant after

[- exhaustion of spent fuel discharge capa. city is not a viable alternative to high density storage in the spent 4~

fuel pool. The prospective 1986 ex diture of

- approximately $ 2.6 million for the high' pen density racks is small compared to the estimated value of replacement

• power equivalent to the plant's energy output approximately $ 21 million per month in 1994 and $ 32 1

million per month in 2011.

1- The subject of the comparative economics associated with various d

spent fuel options is the subject, of Chapter 6 of HUREG-0575 lg (Ref. 2). Although the material presented is' generic, it is of l a, value in comparing the costs of the various options. Of the options presented in that chapter, high density spent fuel storage at the site is the most economic option at $ 3.50 per Kgu. The price of Independent Fuel Storage Facilities (IFSF), if I' available, would be $54.35 per KgU. .

u 9.4 RESOURCE COMMITHENTS f

li -

[,. The expansion of the Byron spent fuel storage capacity will L. require the following primary rescuces: .

l 0 Stainless steel - 284,815 lb/ unit i.

l 0 Boraflex neutron absorber - 22,000 lb/ unit of which

'd 10,925 lb is boron carbide (BgC) powder I li The requirement for stainless steel represents a. small fraction ,

j of the total domestic production for 1986 (Ref. 3). Although the fraction of domestic production of B gC required for the.

f, fabrication is somewhat higher than that for stainless steel, it I

. l 9-3 l 1

- - . - . . , - - - , , - - , , - . - - ,,-,--,---.,_-n.-,--,,.m..._-..n- - + , - , , , , , , , . , , _ . . - - . - . -

i l

.* is unlikely that the' commitment of B gC to this project will affect other alternatives. Experience has shown that the pro-  !

duction of BgC is highly variable and depends on need but could easily be expanded to accommodate additional domestic needs. l I!

1:

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..I 9-4

-i i

REFERENCES TO SECTION 9

1. B. K. Crimes, "0T Position for Review and Acceptance of Spent Fuel Storages and Handling Applications," April 14, -

( 1978.

,4 -

2. HUREC-0575, " Final . Environmental Impact Statement on

}. Handling and Storage of Spent Light Water Power Reactor

1. Fuel," Vols. 1-3, USNRC, August 1979. ,

3. " Mineral Fac'ts and Problems," Bureau of Mines Bulletin 671, 1980.

v.

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10.0 QUALITY ASSURANCE PROGRAM

.I

10.1 INTRODUCTION

I~

l* This chapter provides a general des'cription of the quality assurance program that is, implemented to assure that the quality

. objectives of the contract specification are met.

(3 il, .10.2 CENERAL The quality assurance program used on this project is based upon l

the system described in Oat's Nuclear Quality Assurance Manual. -

This system is designed to provide a flexible but highly si

• controlled system for the design, manufacture, and testing of ji- customized components in accordance with various codes, ii. specifications, and regulatory requirements. The Oat Nuclear.

,. Quality Assurance Program has been accepted by ASME and has been

.. approved by the CEC 0 Quality Assurance Department and placed on CECO's Qualified Suppliers List.

The philosophy behind Cat's Quality Assurance System is that it shall provide for all controls necessary to fulfill the contract requirements with sufficient simplicity to make it functional on The system . readily adapts to different l'

I a

designs day-to-day basis.

and component configurations, making possible the

. construction of many varied forms of equipment. The following

,} paragraphs provide an overview of the system and how it has been applied to Commonwealth Edison's specifications.

10.3 SYSTEM HICHLICHTS ii l The design control is organized to provide for careful review of 4 all contract requirements to extract each individual design and

L U 10-1

_ _. . _ ~

- q'uality criterion. These criteria are translated into design and quality control documents customized to the contract requirement and completely reviewed and approved by responsible and qualified personnel.

The system for control of purchased material includes generating ,

detailed descriptions of each individual item of material along -[

with specifications for any special requirements such as impact testing, corrosion testing, monitoring or witnessing of chemical h analysis, provision of over-check specimens, special treatments or conditioning of material, source inspection, and provision of ,

performance documentation on any of the above.

Material receipt inspection includes a complete check of all .

material and its documentation. Upon acceptance, each item of ,

, material is individually listed on a control sheet issued once a ,

! week to assure that ,only accepted material goes into fabrication.

The fabrication control system provides that a shop traveller is prepared for each subassembly and assembly in each contract. The h traveller is generated specifically to provide step-by-step instructions for fabrication, inspection, testing, cleaning, -

packaging, etc., which address all standard and special .

requirements of the contract specifications. Special attention is given to deployment of fabrication sequence and inspection _

steps to preclude the possibility of missing poison sheets or incorrect sheets (incorrect 8 10 loading).

All nondestructive examination procedures and test procedures are custo.m written to apply to CECO's requirements.

The system provides for qualification and written certification of personnel performing quality-related activities including -

nondestructive examination and fabrication inspection, welding, engineering, production supervision, and auditing.

l 10-2

I i Other CEC 0 requirements are fully covered in the Quality Program, including document control, control of Assurance measuring and test equipment, control of nonconforming material corrective action auditing, and other areas as and, parts, f~ specified by CECO.

er 10.4

SUMMARY

,. Dat's quality assurance system provides the full measure of quality assurance required by the contract. All special 1,

requirements of the specifications are covered, including source

inspection of material and witnessing of material testing by the

~

engineer, furnishing of material certifications and test reports within 5 days of shipment ~, and obtaining verification of qualification testing of poison materials.

cf i e e

e I

1 p

!O I

.,10-3

i i

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k.

1

's 1

APPENDIX A I.

BENCHMARK CALCULATIONS i

'O

~

W O

\

l I

t I

i l

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A-1 l

.j

5. INTRODUCTION AND

SUMMARY

The objective of this benchmarking study is to verify both '

the AMPX (NITAWL)-KENO (Refs. 1 and 2) methodology with the 27-group SCALE cross-section library (Refs. 3 and 4) and the CASMO-2E code (Refs. 5, 6, 7, and 8) for use in criticality calcula-tions of high density spent fuel storage racks. Both calcu- U lational methods are based on transport theory and have been '

benchmarked against critical experiments that simulate typical 2 spent fuel storage rack designs as realistically as possible. -

~

Results of these.-benchmark calculations with both methodologies are consistent with corresponding calculations reported in the  !

literature and with the requirements of Regulatory Guide 3.41,*

Rev. 1, May 1977. -

I Results of these benchmark calculations show that the U

?

27-group (SCALE) AMPX-KENO calculations consistently underpredict the critical eigenvalue by 0.0106

• 0.0048 Ak (with a 954 proba-  !

bility at a 9 5 t'. confidence level) for critical experiments -

selected to be representative of realistic spent fuel storage j rack configurations and poison worths. Similar calculations by Westinghouse suggest a bias of 0.012

• 0.0023, and the results of .

ORNL analyses of 54 relatively " clean" critical experiments show ,

a bias of 0.0100

• 0.0013.

Similar calculations with CASMO-2E for clean critical experiments resulted in a bias of 0.0013

• 0.0018 (95t/954).

CASMO-2E and AMPX-KENO intercomparison calculations of infinite

• arrays of poisoned cell configurations show very good agreement

• and suggest that a bias ,of 0.0013

• 0.0018 is the reasonably J expected bias and uncertainty for CASMO-2E calculations.

" Validation of Calcylational Methods for Nuclear Criticality Safety. (See also ANSI N16.9-1975.)

A-2

-,r-, - -n--e, ,nn-----_---- _. _ . , - , , _ , _ _ . _ - . - - - _ _ . - - - . , - - . - . . - - - , - , - - - . - - . .-

The benchmark calculations reported here indicate- that

, , .. either the 27-group (SCALE) AMPX-KENO or CASMO-2E calculations are acceptable for criticality analysis of high density spent fuel storage racks. The preferred methodology, however, is to .

I perform independent calculations with both code packages and to i('

utilize the higher, more conservative value for the reference design infinite multiplication factor.

t 2. AMPX (NITAWLl-KENO BENCHMARK CALCULATIONS

, ,. Analysis of a series of Babcock & Wilcon (B&W) critical experiments (Ref. 9), which include some with absorber sheets typical of a poisoned s' pent fuel rack, is summarized in Table 1,

, as calculated with AMPX-KENO using the 27-group SCALE cross-section library and the Nordheim resonance integral treatment in 11 NITAWL. The mean for these calculations is 0.9894

• 0.0019, 1-conservatively assuming the larger standard deviation calculated e from the k,gg values. With .a one-sided tolerance factor

! .1 (K = 2.502), corresponding to 954 probability at a 954 confidence

., level (Ref. 10), the calculational bias is +0.0106 with an uncer-l tainty of *0.0048.

j similar calculational deviations reported by Westinghouse

, (Ref. 11) are also shown in Table 1 and suggest a bias of 0.012 *

!' O.0023 (954/954). In addition, ORNL (Ref. 13) has analyzed some l- 54 critical experiments using the same methodology, obtaining a l_ mean bias of 0.0100

• 0.0013 (954/954). These published resulta j'l are in good agreement with the results obtained in the present

', , analysis and lend further credence t'o the v'alidity of the 27-

,, group AMPX-KENO calculational model for use in criticality analy-l ,,

sis of high density spent fuel storage racks. Variance analysis l, of the data in Table 1 suggests the possibility that an unknown j f actor may be causing a slightly larger variance than might be lI expected from the Monte Carlo statistics alone. However, such a l<

l A-3

i l

Table 1 RESULTS OF 27-GROUP (SCALE) AMPX-KENO CALCULATIONS OF B&W CRITICAL EXPERIMENTS Westinghouse

Experiment Calculated Calculated-meas.

Number k,gg a k,gg

• I.

I 0.9889 -

10.0049 -0.008 1 II 1.0040 10.0037 -0.012 .

III 0.9985 10.0046 -0.008  :

IXIII 0.9924. *0.0046 -0.016

X XI 0.9907' 0.9989

• 0.0039

• 0.0044

-0.008

+0.002

]

~

l XII 0.9932 10.0046 -0.013 XIII 0.9890 10.0054 -0.007 XIV 0.9830 ,

• 0.0038 -0.013  !

XV ,

0.9852 *0.0044 -0.016 -

XVI 0.9875 10.0042 -0.015 3 XVII 0.9811 *0,0041 -O'.015 . d XVIII 0.9784 *0.0050 -0.015 .,

XIX 0.9888 10.0033 -0.016 'l '

t-XX 0.9922 *0.0048 -0.011 XXI 0.9783 *0.0039 - 0.017 Mean 0.9894 *0.0011(2) -

0.0120

• 0.0010 l Bias 0.0106 *0.0019(3) 0.0120

• 0.0010 :1 Bias (954/954) 0.0106 *0.0048 0.0120

• 0.0023 '

Maximum Bias 0.0154 0.0143 '

(1) Experiments IV through VIII used B 4C pin absorbers and were ,

(2)not considered representative of poisoned storage racks.

(3) Calculated Calculatedfrom fromk,gg individual valuesstandard and useddeviations. as reference. ,

i A-4 e

---

.m. ..w--,---*- - , _ - , . m-_-___. _m--__-,.- _w,-- - -,w- ,7%_.y-.-vn,. y-m , , . e w y 3---- - -mit - w ,wy- - ---g

I l i factor, if one truly exists, is too small to be resolved on the

.; basis of critical-experiment data presently available. No trends in k,gg with intra-assembly water gap, with absorber sheet reactivity worth, or with soluble poison concentration were

{ identified.*

~

3. CASMO-2E BENCHMARK CALCULATIONS .

3.1 GENERAL 4

. The CASMO-2E code is a multigroup transport theory code l i, utilizing transmission probabilities to accomplish two-dimen-

! sional calculations of reactivity and depletion for BWR and PWR fuel assemblies. As such, CASMO-2E is well-suited to the criti-i cality analysis of spent fuel storage racks, since general i

practice is to treat the racks as an infinite medium of storage l* cells, neglecting leakage effects.

lg. .

'. CASMO-2E is. closely analogous to the EPRI-CPM code (Ref. 13)

, and has been extensively benchmarked against hot and cold crit-

.- ical experiments by 'Studsvik Energiteknik (Refs. 5, 6, 7, and 8). Reported analyses of 26 critical experiments indicate a mean k,gg of 1.000 t 0.0037 (le). Yankee Atomic (Ref. 14) has also reported results of extensive benchmark calculations with CASMO-2E. Their analysis of 54 Strawbridge and Barry critical experi-ments (Ref. 15) using the reported buckling indicates a mean of

.; 0.9987

• 0.0009 (le), or a bias of 0.0013~t 0.0018 (with 95%

I probability' at a 95% confidence level). Calculations were

,j repeated for seven of the Strawbridge and Barry experiments 1

Significantly large trends in k with water gap and with ab-c sorbersheetreactivityworthhave,gbeenreported (Ref. 16) for i

AMPX-KENO calculations with the 123-group GAM-THERMOS library.-

A-5 l

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

selected at random, yielding a mean k,gg of 0.9987

• 0.0021 (la),

- thereby confirming that the cross-section library and analytical methodology being used for the present calculations are the same as those used in the Yankee analyses. Thus, the expected bias for CASMO-2E in the analysis of " clean" critical experiments is 0.0013

• 0.0018 (954/954). i

1.

3.2 BENCHMARK CALCULATIONS ,

I CASMO-2E benchmark calculations have also been made for the g B&W series of critical experiments with absorber sheets, simu-lating high density spent fuel storage racks. However, CASMO-2E, as an assembly code, cannot directly represent an entire core configuration

• without introducing uncertainty due to reflector  ::

constants and the appropriateness of their spectral weighting. -

For this reason, the poisoned cell configurations of the central .

assembly, as calculated by CASMO-2E, were benchmarked against ..

corresponding calculations with the 27-group (SCALE) AMPX-KENO ,

code package. Results of this comparison are shown in Table 2. j, since the differences are well within the normal KENO statistical variation, these calculations confirm the validity of CASMO-2E lj calculations for the typical high density poisoned spent fuel rack configurations. The differences shown in Table 2 are also consistent with a bias of 0.0013

• 0.0018, determined in Section 3.1 as the expected bias and uncertainty of CASMO-2E calcula-tions. .

i S

• Yankee has attempted such calculations (Ref. 14) using CASMO-2E-generated constants in a two-dimensional, four-group PDQ model, obtaining a mean k.gg of 1.005 for 11 poisoned cases and 1.009 for 5 unpoisoned cases. Thus, Yankee benchmark calculations suggest that CASMO-2E tends to slightly overpredict reactivity.

A-6

k i

d Table 2 '

~ '

i RESULTS OF CASNo-2E BENCHMARK (INTERCOMPARISON) CALCULATIONS k,III B&W Experiment No.III AMPX-KENO (2) CASMO-2E Ak XIX ,

1.1203

• 0.0032 1.1193 0.0010 1

XVII 1.1149

• 0.0039 1.1129 0.0020 XV 1.1059

• 0.0038. 1.1052 0.0007

- InterpolatedI3) 1.1024

• 0.0042* 1.1011 0.0013

+

2 XIV '

,, 1.0983

• 0.0041 1.0979 0.0004 3

XIII '. 1.0992

• 0.0034 1.0979 0.0013

• ;* Mean ,

• 0.0038 0.0011 Uncertainty *0.0006

'1 BWR fuel rack

• 0.9212

• 0.0027 0.9218 -0.006 E ',i-r .

lt (1) Infinite array of central assemblies of 9-assembly B&W criti- '

(2) cal configuration (Ref. 9).

(3)kInterpolated from AMPX-KENO corrected for bias of 0.0106 Ak.

from Fig. 28 of Ref. 9 for soluble boron concen-tration at critical condition. *

[.

9 1 .

y .

de T

l' I, e A-7 <

9 9

0

..w., ,,,n--- - - - -. .,,m --w~

t l

REFERENCES TO APPENDIX A

}

1.

~

Green, Lucious, Petrie, Ford, White, Wright, "PSR-63/AMPX-1  :

, (code package), AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," ,

ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.  ;

i

~

2. L. M. Petrie and N. F. Cross, " KENO-IV, An Improved Monte Carlo . Criticality Program," ORNL-4938, Oak Ridge National I:

Laboratory, November 1975. I;

3. R. M. Westf all et al., " SCALE: A Modular Code System for Jj Performing Standardized Computer Analyses for Licensing gi Evaluation," NUREG/CR-0200, 1979.

4. W. E. Ford, III et al., "A 218-Neutron Group Master Cross-i section Library for Criticality Safety Studies," ORNL/TM-4, 1976.

ll

, 5. A. Ahlin, M. Edenius, R. Haggblom, "CASMO - A Fuel Assembly e Burnup Program," AE-RF-76-4158, Studsvik report (proprietary). t l ,

6. A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory i

Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, g

p. 604, 1977. s e i

7. M. Edenius et al., "CASMO Benchmark Report," Studsvik/RF-  ;

( 78/6293, Aktiebolaget Atomenergi, March 1978.

1

8. "CASMO-2E Nuclear Fuel Assembly Analysis, Application Users Manual," Rev. A, Control Data Corporation, 1982. *

^

9. M. N. Baldwin et al., " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel," BAW-1484-7, .

. The Babcock & Wilcox Company, July 1979.

10. M. G. Natrella, Experimental Statistics,. National Bureau of l Standards, Handbook 91, August 1963.

11. B. F. Cooney et al., " Comparisons of Experiments and Calculations for LWR Storage Geometries," Westinghouse NES, l

ANS Transactions, Vol. 39, p. 531, November 1981. ],j

12. R. M. Westfall and J. R. Knight, " Scale System Cross-section i validation with shipping-cask Critical Experiments," ANS Transactions, Vol. 33, p. 368, November 1979. -

i

13. "The EPRI-CPM Data Library," ARMP Computer Code Manuals, Part II, Chapter 4, CCM3, Electric Power Research Institute, November 1975.

i a

l A-8 4

REFERENCES TO APPENDIX A (Continu d) l.. . .

" Methods for the Analysis of Yankee Boiling Water

14. E. E. P il~at , YAEC-1232, Atomic ,

Reactors '(Lattice Physics),"

{.'- Electric Co., December 1980.

15. L. E. Strawbridge and R. F. Barry, " Criticality,Calcu -

Enaineerina, Vol. 23, p.,58, September 1965. .

i s.

S. E. Turner and M. K. Gurley, " Evaluation of AMPX-KENO

16. Benchmark Calculations for High Density Spent Fuel 230-237, Storage Racks," Nuclear Science and Encineerinc, 80(2):

February 1982.

6

< 9

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'I 9

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. A s.

/

l i 4) Commonwealth One First National Plaza. Checa0o, EdisonHlinois k Address Reply to: Post Omca son 757

. \ Chicago, Hlinois 60690 0767 i

% l November 7, 1986 Mr. Harold R. Denton, Director office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Byron Station Units 1 and 2 Application for Amendment to Facility operating License NPF-37 Appendix A, Technical Specifications NRC Docket Nos. 50-454 and 50-455 Reference (a): September 3, 1986 letter from K. A. Ainger to H. R. Denton

Dear Mr. Denton:

Reference (a) requested a license amendment to increase the storage capacity of the spent fuel racks at Byron Station. Attachment C of reference (a) contained an evaluation of significant hazards considerations for the reracking of the spent fuel pool. The evaluation concluded that the proposed amendment presented no significant hazards considerations.

Attached to this letter is a more detailed determination of no significant hazards considerations. This determination addresses the change from floor nounted to free standing racks, as well as other significant

~ # changes resulting from installation of high density spent fuel racks.

please direct any questions regarding this matter to this office.

One signed original and fifteen copies of this letter and attachment are provided for NRC review.

Very truly yours, K. A. Ainger Nuclear Licensing Administrator im Attachment cc: Byron Resident Inspector 2320K

y DETERMINATION OF NO SIGNIFICANT HAZARDS

. CONSIDERATIONS The proposed amendment would authorize the replacement of the currently installed spent fuel racks with free standing, high density racks. This rack replacement will increase the capacity for storing spent fuel at Byron Station from 1060 to 2940 fuel assemblies. Included in the new rack configuration are six locations for storing failed fuel. The details of the proposed reracking are described in Attachment B of reference (a).

Edison has evaluated the proposed reracking in accordance with 10 CFR 50.92(c) and determined that it presents no significant hazards considerations. This determination was based on an evaluation of the significant changes that would resuIt from the proposed reracking. As discussed in detail below, the conclusions of this evaluation were that the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated (Criterion 1);

2. Create the possibility of a new or different kind of accident from any accident previously evaluated (Criterion 2); or

3. Involve a significant reduction in a margin of safety (Criterion 3).

Because the criteria of 10 CFR 50.92(c) are satisfied, this proposed amend-ment should be noticed as involving no significant hazards considerations.

The significant changes that would result from the proposed

- ' reracking are:

A - free standing instead of bolted racks; B - increased density of spent fuel storage due to an increased quantity of fuel stored and a decreased spacing between the racks; i

l C - differences in rack design resulting from differences in materials

! including the incorporation of neutron absorbing materials j containing boron; and D - separation into two burnup dependent regions i

i l

i

• i o j y

For each of these significant changes, the criteria in 10 CFR 50.92(c) are evaluated below. This evaluation included consideration of the following accidents which were evaluated previously:

(a)' Radiologic consequences of the drop of a spent fuel assembly onto the floor of the spent fuel pit (accident a);

(b) Radiologic consequences of the drop of a spent fuel cask (accident b);

(c) The potential for criticality from a dropped fuel assembly laying I across the top of a fuel rack (accident c);

i

(d) The potential for criticality from a fuel assembly accidentally dropped into a position parallel to fuel stored in the most reactive corner of the racks (accident d); and i

(e) The effect of a seismic event on rack integrity and the potential for criticality (accident e).

BVALUATION j A. Free Standing Instead of Bolted Racks f* -

Criterion 1: This change has no effect on accidents (a) through (d).

As for accident (e), Edison's analysis shows that a seismic event will not significantly increase the probability or consequences of either the

potential for criticality or a challenge to rack integrity.

The new racks, like the old ones, are seismic Category I structures and, thus, have been designed to maintain their integrity during a seismic event. Even though the racks are free standing, that integrity will not be challenged significantly because the safe shutdown earthquake for Byron is not capable of accelerating the racks sufficiently to cause them to experience significant forces, either directly or on collision with each other or with the pool's walls. The safe shutdown earthquake (ssE) for Byron is capable of resulting in a horizontal acceleration of no more than 0.39 The corresponding seismically induced maximum rack movement (including tipping or sliding) is less than 3/16". But the i minimum distance between a rack and a wall of the spent fuel pool is 2 5/8". Therefore, an earthquake would not result in a collision '

i between a rack and a wall of the spent fuel pool. Thus, the free

standing character of the racks will not increase the probability or consequences of a seismically induced challenge to rack integrity due to a collision between a rack and a wall of the spent fuel pool.

I

.w --,e---, +,v- c- ,,,,,..,-,w ,,-e,,cy , ,,e_.

,,,--,e,,- -,__ -,-,,enn_,-----

l ,

, a Analysis also shows that the free standing character of the racks will not increase the probability or consequences of a seismically induced ,

challenge to rack integrity due to a collision between adjacent racks.

The low acceleration associated with the SSE, the extension of the base plate beyond the last row of cells and the use of framing material (girdle bars) at the top of the racks ensure that rack-to-rack impact loads due to an sst will not significantly increase a seismically induced challenge to rack integrity.

In addition, the use of base plates and girdle bars also ensures that an -

SSE will not reduce the spacing between cells in adjacent, either in Region 1 or in Regions 1 and 2, racks to a distance less than the minimum spacing assumed in the criticality analysis. Intercellular spacing is not a concern in Region 2 because all cells in that region are separated by the neutron absorber sheet. Thus, the free standing character of the racks will not cause a significant increase in the probability or consequences of a criticality event.

For the foregoing reasons, criterion 1 is satisfied in light of the change to free standing racks.

Criterion 2: The circumstance that the new racks are free standing does not create the possibility of a new or different kind of accident. The only new consequence of the free standing nature of the racks is their ability to move in the event of a seismically induced acceleration.

However, as discussed above, the maximum accelerations associated with the SSE for Byron are not capable of inducing significant rack motions.

Moreover, rack design precludes any impact between a rack and a wall of the spent fuel pool, and limits the effects of rack-to-rack collisions to forces well below the force necessary to challenge rack integrity.

Therefore, the new free standing racks satisfy criterion 2.

Criterion 3: For the reasons discussed above, the fact that the racks are free standing does not result in a significant reduction in a margin of safety. The SSE for Byron is simply not powerful enough to cause the unbolted racks to experience stresses which would challenge their structural integrity to any significantly greater extent than stresses would have challenged the integrity of the old ranks.

B. Increased Density of stored spent Puel criterion 1: The increased density of stored spent fuel does not increase the probability of any of the accidents previously considered.

The probabilities of those accidents aru controlled by the probabilities of the initiating events and these are independent of the density of spent fuel stored. Increased storage density also does not increase significantly the consequences of the previously analyzed accidents.

The new racks have been designed to ensure that none of the accidents will result in an exceelance of the criticality acceptance criterion "that X,gg will not exceed 0.95. Thus, criterion 1 is satisfied.

. .E" ,

i Criterion 2: No new or different kinds of accidents are presented by the circumstances that the new racks are closer together and hold more spent fuel. Weither the decrease in rack spacing nor the increase in quantity of spent fuel has led to the identification of any type of accident not already considered in the initial safety analysis for the spent fuel pool. Thus, Criterion 2 is satisfied.

Criterion 3: Installation of'the new spent fuel racks will not result in a significant reduction in a margin of safety. A small increase in i the spent fuel pool heat load is expected due to the storage capacity expansion. However, the spent fuel pool cooling system design can handle the increased heat load and maintain the temperature peaks of the pool below design values. Installation of the new spent fuel racks will also result in a small increase'in the pool reactivity as measured by the neutron multiplication factor (K,gg). However, the maximum f

neutron multiplication factor will be maintained less than or equal to

! 0.95.

The radiological consequences also were evaluated to determine the impact on off-site and on-site doses previously deterr'ned. The increase in the storage capacity of the spent fuel pov4 will neither significantly alter the operating characteristics of the current spent fuel pool nor result in a measurable change in impact on the

, environment. The design basis fuel handling accidents, described in FSAR Section 15.7.4, were reviewed for possible effects on radiological l

dose consequences. The review determined that the conclusions in the PSAR will remain valid and that off-site radiological dose consequences will remain within 10 CFR 100 limits.

4 For these reasons, increasing the spent fuel pool storage capacity will not significantly reduce a margin of safety. Therefore, criterion 3 is

satisfied.

i C. Differences in Rack Design criterion 1: The new racks are made of the same materials as the old

(

ones except that they also include a neutron-absorbing borated l material. The inclusion of this material has no effects on accidents (a) through (d) and is relevant to accident (e) only to the extent that the potential for criticality is affected. By its nature, the borated

- neutron absorbing material cannot significantly increase the probability +

or consequences of a criticality accident. Therefore, the inclusion of i this material satisfies criterion 1.

i I

4 I

i

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

.,b t

Criterion 2: Nothing about the materials in the new racks can create >

the possibility of a new or different kind of accident. Only the borated neutron absorber is a new or different kind of material in the racks. That material is passive and will perform its function as long as it remains in place. And the construction and testing of racks render insignificant the possibility that enough borated material could

• shift to result in a significant increase in criticality. Accordingly, this criterion is satisfied.

criterion 3: Because shifting of the borated material is the only kind

! of safety-related occurrence associated with the design of the new racks, and the probability of that occurrence is very low, these racks do not present a significant reduction in the margin of safety.

Accordingly, this critorion is satisfied.

l D. separation into Two Burnup Dependent Recions Criterion 1: of the accidents previously analyzed only accident (d),

the accidental drop of a fuel assembly parallel to fuel stored in the most reactive corner of.the racks, requires a re-evaluation of its i consequences. The probability of occurrence of this accident is not affected by the presence of two burnup dependent regions for storing spent fuel. As for the consequences of this accident, calculations show that it will not result in an exceedance of Keff above the acceptance i criterion of 0.95. Thus, Criterion 1 is satisfied by the separate locations.

Criterion 2: Thi use of separate locations for storing different burnup 4 fuels will not create the possibility of a new or different kind of accident. The two region design does create the possibility that a fuel assembly may be placed in the wrong region. However, the probability of such a misplacement is very low due to the obvious differences in design of the two regions, design differences which are readily discernable 4

from'the position above the pool which would be taken by an operator loading fuel into the pool. In any event, the misplacement of a fuel assembly would not lead to a new or different kind of accident because it is in essence just a version of accident (d). Therefore, the steps which have been taken to maintain Keff below the acceptance criteria l'

' for the present racks will also maintain Keff below the acceptance t criteria for the new racks. Thus, criterion 2 is satisfied by the separate locations.

Criterion 3: Nothing about the use of separate locations for storing

fuel with differing burnup can affect the margins of safety.

Accordingly, criterion 3 is satisfied.

1 1

2320K-I t

i

~l 1

} C4NnnMNMUGelth Esfloon One Frat National Plaza. Chcago, IEinois Adareas Reply to: Post omos som 757 o CNCago lEinois 80000-0767 November 24, 1986

• Mr.-Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Consaission Washington, DC. 20555 Eabject: Byrcr. Station Units 1 and 2 4 plication for Amendment to Facility Operating License NpF-37, Appendix A. Technical Specifications t References (a): September 3, 1986 letter from K.A. Ainger to H.R. Denton (b): November 7, 1986 letter from K.A. Ainger to H.R. Denton

Dear Mr. Denton:

Reference (a) requested a license amendment to increase the storage

' capacity of the spent fuel racks at Byron Station and reference (b) transmitted a detailed deterinination of no significant hazards considerations for the proposed amendment.

This letter contains revisions to the technical specification changes requested in reference (a). The revisions are primarily due to a change in the manufacturing process of the high density racks. In addition, the revised amendment includes a new Figure 5.6-1 which provides criteria for fuel storage in Region 2 of the high density spent fuel racks. Attachment A of this letter contains the revisions to the proposed amendment.

n This revised amendment has been reviewed and approved by both on-site and Off-site review in accordance with Commonwealth Edison Company procedures. The determination of no significant hazards considerations transmitted in reference (b) is not affected by this revised amendment.

A surveillance will be conducted to verify the proper location of each fuel assembly in the high density spent fuel racks after each unit's refueling and at intervals of approximately six months thereafter. However, when a unit is scheduled for a refueling which is expected to be completed within nine months of the previous pool inventory, the six month interval may i

l be extended to coincide with the inventory required after the refueling. This requirement is presently included in Byron Station procedure BAp 2000-4.

Chapter 9 of the Byron /Braidwood FSAR will be updated to reflect the j

design of the high density spent fuel racks after approval of the proposed

' license amendment. This, update will include the surveillance commitment outlined above, h

O 00 4

'll

. - _. .. _ _ __,___ J _.___.___._., , . _ _ . . . _ . _ . . _ . _ _ _____. _ _ _ _ . _ . _ . . - - _ _ _ _ . . _ _ _ . . _ _

9

(* .

~f 2 Please direct'any questions regarding this matter to this office.

One signed original and fifteen (15) copies of this letter and attachment are provided for NRC review.

Very truly yours,

/ .

K. A. Ainger Nuclear Licensing Administrator

/klj att.

cc: Byron Resident Inspector 2420K i

6 i

e. k DESIGN FEATURES Thi.s is based on spenF fuel stcrp e 5.6 FUEL STORAGE in Reg;dH a wHh enn'chenenf-s an burnup in acccedance with R vre S G-1 CRITICALITY or~ in a checkerboard pa He n 3 and )

5. 6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. A k,ff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance s**sSEE.

ahes.for end- uncertainties as described in Section 9.1 of the FSARK h

b. A nointnab.14_ inch center-to-centse di:t: ace bet ;ee Tu.i ...siiibTies lI Eplec.u in ine n.orace racks. -

t 5.6.1.2 The k,ff for new fuel for the'first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

ORAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 423 feet 2. inches.

CAPACITY -

5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3066. fuel ass'emblies.

28d t

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT " -

5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. .

L A nominql 10 32. inch nceth- soeA and 1o.'4 2. inc h eaJf -We.s h, m cenwr - to - cenfee diHance be+ ween hel effemblie.s placed in i

Reg fo spenF hel rForage. racks ad a norwinal *!.03 inch

+ ion cen er-I cen*r dis fance. be hacen N41 a mmbUca placed in Region 2 spani- Gel s) ora 34. rac)cs.

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BYRON - UNITS 1 & 2 5-5 l

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. ACCEPTABLE .

. REGION -

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1.5 2.0 2.5 3.0 3.5 4.0 4.5 FUEL ASSEMBLY INITIAL ENRICIDIENT (W/0 U-235) l FIGURE 5.6-1 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 STORAGE BYRON - UNITS 1 & 2 5-Sa

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Novemb:r 25, 1986

~'

. Docket Nos. STN 50-454 and STN 50-455 Mr. D. L. Farrar Director of Nuclear Licensing Comonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

Dear Mr. Farrar:

SUBJECT:

BYRON SPENT FUEL POOL EXPANSION - RE0 VEST FOR ADDITIONAL INFORMATION.'

Enclosed is a request for additional information that we need to complete

~

o'ur review of your September 3J,1986 amendment to rerack the Byron spent fuel pool. In order to accomodate your request that we approve the amendment by January 1, 1987, we need the responses to these questions by December 5, 1986.

If any further clarification is needed please contact me at (301) 492-4937.

Sincerely, UMainal signe2 tv Leonard N. Olshan, Project Manager Project Directorate #3 Division of PWR Licensing-A

Enclosure:

As stated cc: See next page i

O

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Mr.. Dennis L. Farrar . . Byron Station Commonwealth Edison Company , Units 1 and 2

' CC*

\.

' Mr. William Kortier Ms. Diane Chavez

. Atomic Power Distribution 528 Gregory Street Westinghouse Electric Corporation Rockford, Illinois 61108 Post Office Box 355 Pittsburgh, Pennsylvania 15230 Regional Administrator, Region III U. S. Nuclear Regulatory Commission Michael Miller 799 Roosevelt Road Isham, Lincoln & Beale Glen Ellyn, Illinois 60I37 One First National Plaza 42nd Floor Joseph Gallo, Esq. -

Chicago, Illinois 60603 Isham, Lincoln & Beale -

Suite 1100 Mrs. Phillip B. Johnson 1150 Connecticut Avenue, N.W.

1907 Stratford Lane . Washington, D. C. 20036 Rockford, Illinois 61107 Douglass Cassel, Esq.

Dr. Bruce von Zellen 109 N. Dearborn Street Department of Biological Sciences Suite 1300 Northern Illinois University Chicago, Illinois 60602

_. DeKalb, Illinois 61107 Ms. Pat Morrison Mr. Edward R. Crass 5568 Thunderidge Drive Nuclear Safeguards & Licensing Rockford, Illinois 61107 Sargent & Lundy Engineers 55 East Monroe Street Ms. Lorraine Creek Chicago, Illinois 60603 Rt. 1, Box 182 Mr. Julian Hinds Manteno. Illinois 60950 U. S. Nuclear Regulatory Commission

. Byron / Resident Inspectors Offices 4448 German Church Road Byron, Illinois 61010 .

Mr. Michael C. Parker, Chief Division of Engineering l Illinois Department of l Nuclear Safety 1035 Outer Park Drive Springfield, Illinois 62704

• i o.

ENCLOSURE s

REQUEST FOR ADDITIONAL INFORMATION BYRON RERACK AMENDMENT l '

1. For the spent fuel pool heat exchangers provide the following :

a. heat transfer conductivity in BTU /Hr.

FT3/"F ,

b. tube surface area in square feet 4

2. Section 5.1 of the Licensing Report (Attachment B to the September 3, 1986 letter) indicated that Branch Technical Position (BTP) APCSB 9-2 was used, but Reference 1 to Section 5 indicated BTP ASB 9-2, Revision 1. July 1981 was used. Correct this discrepancy. Also confirm that the guidance in Standard Review Plan Section 9.1.3,III, regarding uncertainty factors, has been used.

3. With respect to the spent fuel pool structure, the licensee provided virtually no documentation attesting to the adequacy of the analytical procedures, the load combination criteria, or the selection of allowable loads and stresses, other than to reference the original spent fuel pool analyses included in the FSAR. Accordingly, the licensee is requested to l provide the following for review:

a. Sketches and/or drawings of any changes to the spent fuel pool structure not considered in the FSAR analysis.

b. A description of the mathematical model of the pool structure, including the finite-element model if used, and the method of analysis.

Describe the assumptions employed and the limitations of the model.

L c. Ample description of the loadings used, and justification for the load combinations used.

L d. The source of the acceptance criteria and method of determinina the allowable loads and stresses in various parts of the structure.

e. A description of how the dynamic interaction between the pool structure and the rack modules was considered, including the key assumptions used in assessing the interaction effects and the value of any dynamic amplification factors. Also include all assumptions made regarding the summation and phasing consideration of all rack module

~

dynamic loads.

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2

f. An analysis of the adeauacy of the pool floor and liner under rack sliding and impact loads,

g. The critical regions of the pool structure, l.ist the loads or stresses as appropriate. Compare the loads and/or stresses to allowable values and indicate the source of the allowables in accordance with Question 5.d. above.

h. A description of any changes to be made to the leak monitoring system for the fuel pool.

4. It is not quite clear from the report how the support legs of the racks are proposed to be constructed. Provide the following information with .

respect to the design, construction, and installation of the support legs:

a. Sketch (or sketches) showing the upper and lower parts of the support feet including where the material transition from SA-217-CA15 to SA-351-CF3 occurs.

b.' How is the transition accounted for in the analysis of fuel racks?

c. Adjusting the support legs during installation would require a long am wrench. Provide information regarding the installation of racks when an adequate clearance to adjust the supports may not be available. How is the unevenness (if any) in the installation of the racks accounted for in the analysis?

5. The following information is needed for us to perform independent review of the analysis:

a. A set of fuel rack and fuel assembly drawings.

b. All analytical modeling parameters including:

- dimensions

- masses

- spring constants

- gap element properties

- fluid coupling coefficients

- coefficient for restitution for impacts

- multi-racks to rack and multi-racks to pool wall impacts

c. All supporting calculations for model parameters.

d. Digitized time histories (OBE A SSE) in three directions, as applicable.

e. Calculations supporting the assumption that the fuel racks are rigid.

. l l

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k

f. Detailed seismic analysis and results.

g. Detailed seismic stress analysis and results '

h. Thermal stress evaluations.

1. Various limiting rack and multi-rack displacements due to seismic loads.

6. Specific questions related to the analysis:

I a. Provide justification for ignoring the flexural rigidity of _. fuel assemblies and modeling them as five separated " rattling" masses. -

Provide information on the flexural rigidity of fuel assemblies.

Also discuss the manner in which the flexibilities of the multi-element fuel assemblies and cells are accounted for in the rack module analysis. -

b. How is the vertical mass of the fuel assemblies accounted for in the model?

c. How is the mass of the water within the fuel accounted for in the

! model?

d. Have the effects of hot gaps and cold gaps been considered?

e. The model in Figure 6.7 shows the gap elements between the rack module edges and fixed boundaries are used to simulate inter-rack impacts. This does not simulate the potential increase of gaps due to sliding of a row of rack modules and development of higher impact velocities with larger gaps. Provide further justification that s such worst case effects have been accounted for,

f. Demonstrate that inter-rack impacts at mid-height elevations are not possible.

4

g. Explain the significant frequency difference between the EW and the NS SSE time histories shown in Figures 6.1 and 6.2.

7. Provide the parameters and constants used in the analysis for impact loadino due to the drop of a fuel assembly. Also, provide a sumary of ductility ratios utilized to absorb kinetic energy in the tensile, flexural, compressive and shearing modes. Provide typical calculations indicating the input constants, equations used, and the results of the impact analysis.

8. Provide the considerations given regarding the potential impact on the functionality of fuel rack modules due to bowing and localized deformations of fuel assemblies and fuel rack cells.

l l

1

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9. Provide the estimated occupational dose expected from perfoming the

" dry" reracking of the Byron spent fuel pool (SFP). These doses should be quantified (e.g.,10 mrem, less than 0.1 mrem) and include the following:

a. occupational dose for each phase of the SFP modification;

b. the basis for the estimate (methodology), including dose rates and manpower;

c. maximum individual dose expected.

L

10. Provide the infomation in 9 above for the " wet" reracking, additionally -

addressing doses to divers. .(NOTE: If wet reracking is not addressed, the safety evaluation will consider dry reracking only.)

11. Identify, for " dry" reracking (and " wet" reracking if planned):

1. radwaste sources (e.g., racks, filters) volumes, and types (e.g.,

solid liquid, gaseous) expected from:

a. the reracking operation, and

b. from subsequent spent fuel pool operations; and

2. the expected occupational dose increases resulting from increased spent fuel storage (e.g., compare pre-mod radwaste doses and post-mod radwaste doses).

12. Quantify the gaseous effluent releases (i.e., Kr 85) and compare the releases expected due to increased spent fuel storage quantitatively with pre-mod and post-mod annual spent fuel pool and overall plant releases i (actual or calculated).

13. Provide a quantitative assessment of the impact of releases identified in question 12. above on individual and population doses offsite.

14. Compare the doses from 9 and IC above with the overall doses projected (or actually experienced) for spent fuel operations in the Byron (or a similar plant) SFP and with doses experienced or project for overall plant operations.

15. . Verify that no changes to the SFP ventilation and SFP water cleanup systems will be required for radiological purposes (e.g., need for increased capacit l

design or layout)y, higher flow rates, additional components, revised '

i l

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.(-

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-, h Com'monwealth Edison A7 One Fast National Plaza. Chicago, Ilhnois )

• -

• 7 Address Reply to: Post Offce Box 767 I x

\d Chicago, Illinois 60690 0767

.f l

December 11, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC. 20555

Subject:

Byron Station Units 1 and 2 Spent Fuel Pool Expansion NRC Docket Nos. 50-454 and 50-455

Reference:

November 25, 1986 letter from L.N. Olshan to D.L. Farrar

Dear Mr. Denton:

The referenced letter requested additional information regarding our proposed license amendment of September 3, 1986 to rerack the spent fuel pool at Byron Station. The responses to the request for additional information are enclosed with this letter.

Please direct any questions regarding this matter to this office.

One signed original and fifteen copies of this letter and enclosure are provided for NRC review.

Very truly yours, K. A. Ainger Nuclear Licensing Administrator 1m Enclosure cc: Byron Resident Inspector 2504K

1 l

p ,/ s

/ - Coennunnwashh Edloon l I m_

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t One FW: National Plaza. Chicago, Enois  !

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/ Address Reply to: Post Ofhce Box 767 4 \d ' Chcago, Illinos 60690 - 0767 f December 11, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC. 20555

Subject:

Byron Station Units 1 and 2 Spent Fuel Pool Expansion NRC Docket Nos. 50-454 and 50-455

Reference:

November 25,.1986 letter from L.N. Olshan to D.L. Farrar

Dear Mr. Denton:

The referenced letter requested additional information regarding our proposed license amendment of September 3, 1986 to rerack the spent fuel pool at Byron Station. The responses to the request for additional information are enclosed with this letter.

Please direct any questions regarding this matter to this office.

One signed original and fifteen copies of this letter and enclosure are provided for NRC review.

Very truly yours, K. A. Ainger Nuclear Licensing Administrator 1m Enclosure cc: Byron Resident Inspector 2504K

o Response to NRC Qunstions Item 1: For the spent fuel pool heat exchangers provide the following:

a. heat transfer conductivity in BTU /Hr.

FT4/0F

b. tube surface area in square feet

Response

a. The heat transfer conductivity in BTU /HR FTd/0F Clean Condition: 501 Fouled Condition: 325

b. The tube surface area is 6463 sq. ft.

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' Item 2:- Section 5.1 of the Licensing Report (Attachment B to the September 3, 1986 letter) indicated that Branch Technical Position (BTP) APCSB 9-2 was used, but Reference 1 to Section 5 indicated BTP ASB 9-2, Revision 2, July 1981 was used. Correct this discrepancy. Also confirm that the guidance in Standard Review Plan Section 9.1.3.III, regarding uncertainty factors, has been used.

BTP ASB 9-2, Rev. 2, July 1981 was used as cited in the reference. The text in section 5.1 as in error on this point.

The use of uncertainty factors was in accordance with Standard Review Plan Section 9.1.3.III.

- .- -m . . - . _ _ , _ _ _ , _ _ . . - . _ _ _ _ _ _ , . , y_-_-.._,. - _ , , . , - , . - - - . - _ - - , , - - . - - -

-s ' CECO RESPONSE TO

.. . NRC REQUEST FOR ADDITIONAL INFORMATION SPENT FUEL POOL RERACKING Item 3:-

With respect to the spent fuel' pool structure, the Licensee provided virtually no documentation attesting to the adequacy of the analytical procedures, the load combination criteria, or the selection of allowable loads and stresses, other than to reference the original spent fuel pool analyses included in the FSAR. Accordingly, the Licensee is requested to provide the following for review:

A. Provide sketches and/or drawings of any changes to the spent fuel pool structure not considered in the FSAR analysis.

Response 3A: ,

The installation of the high density fuel storage racks does not necessitate any change to the spent fuel pool structure.

The Byron high density fuel racks are free standing and apply loads directly to the bottom pool slab. The. bottom pool slab is a part of the~ building-foundation basemat which is not highly stressed from the pool loads.

B. Provide a description of the mathematical model of the pool structure, including the finite-element model if used, and the method of analysis. Describe the assumptions employed and the limitations of the model.

Response 3B:

The pool was analyzed by the finite element method of analysis using the SLSAP-IV computer program. The pool slab and walls are modelled utilizing a mesh of quadrilateral plate /shell elements. Spring boundary elements are used to I

model the supporting foundation material (Figure 1).

The assumptions made in the analysis are:

1..

a) Concentrated loads are distributed as nodal point loads.

b) The stiffness of the foundation boundary elements is conservatively based on the Braidwood foundation properties. The foundation at Braidwood is softer than Byron which results in higher stresses in the structure.

1

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

WEST WALL

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, WALL EAST Ie -

WALL l y) yfu '2 l @bg_ o .y O DEMOTES CRITICAL

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c) Since the pool structure is symmetrical about column

-line 18, half of the pool is modelled. The nodal restraints along the line of. symmetry are adjusted depending on the loading condition to reflect symmetry or antisymmetry.

d. Horizontal displacements are restrained for the east and west walls in the areas where the fuel handling building slab frames into the walls.

C. Provide ample description of the loadings used, and justification for the load combinations used.

Response 3C:

The following loading conditions are considered and used in the analysis: -

a. Thermal expansion of the liner

b. Horizontal hydrostatic pressure

c. Thermal gradient

. d. Axial expansion

e. Dead loads including fuel, fuel racks, self weight of structure, and vertical hydrostatic pressure

f. Dead loads from the cask pit side of the pool

g. Accidential cask drop

h. Hydrodynamic forces and seismic excitation of dead load for OBE and SSE The loading conditions above are combined into 15 load combinations per the B/B FSAR (Table 3. 8-10) . The following controlling. combinations are evaluated.

l

a. Normal

^

b. Severe environmental

c. Abnormal (cask drop)

d. Extreme environmental D. Document the source of the acceptance criteria and metho'd of c determining the allowable loads and stresses in various i parts of the structure.

. Response 3D:

The allowable stress acceptance criteria is found in the B/B l FSAR. The computer program TEMCO is used to check the l reinforced concrete sections for the factored design moments and forces. The resulting stresses are evaluated against

(

t the following acceptance criteria:

j Maximum allowable steel stress = 54.0 ksi Maximum allowable concrete stress = 3,000 psi 1

.0-E. Describe-how the dynamic interaction between the pool i-structure and the rack modules was considered, including the key assumptions used in assessing the-interaction effects and the value of any dynamic amplification factors. Also include all assumptions made regarding the summation and-phasing consideration of all rack module dynamic loads.

Response 3E:

Dynamic interaction between the pool structure and rack

modules is not significant, and the rack analysis is performed assuming the pool structure is rigid. The Byron high density fuel racks are free standing racks which may slide and uplift under earthquake. A nonlinear seismic analysis is performed on,the racks to determine rack response. The input to this analysis are the responses

calculated from the overall building seismic model.

The Byron /Braidwood building seismic models are described in FSAR 3.7.2.3 and the fuel pool is included in the auxiliary-fuel handling-turbine building model. The magnitude of the pool mass is small enough in comparison to participating building mass so that a change to this mass is not significant to the overall auxiliary-fuel-turbine building

+

seismic response. Therefore, overall dynamic interaction between the pool constituents and-the building is not significant.

Evaluation of local interaction requires comparison of the local stiffness of the supporting structure and the rack stiffness. Because the pool slab is a foundation slab, it is rigid in comparison to the rack fuel system and need not be included in the rack analysis. Furthermore, since the racks are free standing, there is no significant interaction I with the pool walls. The results of the rack analysis are used to design the pool structures. Dynamic amplification of the rack and fuel mass is accounted for in the dynamic analysis of the rack system. The maximum rack vertical loads and horizontal leads are conservatively assumed to be in phase for the evaluation of the pool structure.

(

I i F. Provide analysis of the adequacy of the pool floor and liner

under rack sliding and impact loads.

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Response 3F:

The rack sliding and impact loads are included in the maximum support loads supplied by the rack vendor. A local check of the pool floor and stainless steel liner is made for maximum rack foot. impact loads by confirming the adequacy of these-elements for the resulting bearing stresses. Vendor supplied rack loads are also accounted for in overall' behavior of the pool structure through use of the finite element analysis.

G. Identify the critical regions of the pool structure. List the loads or stresses as appropriate. Compare the loads and/or stresses to allowable values and indicate the source of the allowables in accordance with Question 3.d. above.

Response 3G:

Critical regions of the pool structure are identifed on Figure 1. The design stresses occuring at these locations are given in Table 1.

i Table 1 Summary of Stresses Rebar Stress (ksi) Max. Concrete Critical Horz. Bar Vert. Bar Compressive Stress (ksi)

Location Inside Outside Inside Outside Horz. Vert.

l.. West Wall -0.57 -22.51 46.54 -2.47 1.97 0.55

2. North Wall -14.51 1.61 -5.03 30.61 1.42 0.90

3. East Wall -19.77 -3.87 -5.01 31.25- 1.78 0.89

4. East Wall -30.11 -7.11 -11.08 -1.66 2.38 1.12 j 5. Basemat -23.66 0.78 -18.35 14.60 2.09 1.74 l

Notes:

1. For the basemat horz. bar is the east-west reinforcing and vert. bar is the north-south reinforcing.

2. Concrete shear stess is not critical.

i l

l The above stresses are well within the allowable values per the acceptance critiera given in Response 3.D.

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

H. Provide descr-iption of any changes to be made to the leak I monitoring system for the fuel pool.

Response 3H:

No changes are required for the leak monitoring system.

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. Item 4:

It is not quite clear from the report how the support legs of the racks are, proposed to be constructed. provide the following information with respect to the design, construction, and installation of the support legs:

a. Sketch (or sketches) showing the upper and lower parts of the support feet including where the material transition from SA-217-CA15 to SA-351-CF3 occurs,

b. How is the transition accounted for in the analysis of fuel racks?

c. Adjusting the support legs during installation would require a long arm wrench. provide information regarding the installation of racks when an adequate clearance to adjust the supports may not be available. How is the unevenness (if any) in the installation of the racks accounted for in the analysis?

Rtsponse:

a,b. Fig. 3.6 (p. 3-13 of licensing report) shows details of the support leg. The upper part of the support leg is constructed of SA-351-CF3 material and has eight additional gussets. The upper piece is the female part of the support leg. The male piece of the support leg, a threaded connection, is made of 3A-217-CA15 material. In the dynamic analysis, different material allowable stress values are used when checking for structural integrity.

c. The adjustment tool is shown in the attached JO-drawing. The use of the tool is independent of clearances. Therefore, all levelling of the rack can be carried out at installation. The analysis, therefore, presumes a level rack. The installation procedure Jp-2481-22 Rev. O was sent under separate cover.

i

Item 5:

Tha following information is nosdsd for us to parform indspandant review of_the analysis:

a. A set of fuel rack and fuel assembly drawings.

b. All analytical modeling parameters including:

- dimensions

- masses

- spring constants

- gap element properties

- fluid coupling coefficients

- coefficient for restitution for impacts

- multi-racks to rack and multi-racks to pool wall impacts

c. All supporting calculations for model parameters

d. Digitized time histories (OBE & SSE) in three directions, as applicable,

e. Calculations supporting the assumption that the fuel racks are rigid.

f. Detailed seismic analysis and results.

g. Detailed seismic stress analysis and results.

h. Thermal stress evaluations.

i. Various limiting rack and multi-rack displacements due to seismic loads.

Response

a. The following Joseph Oat drawings were sent to Brookhaven National Laboratory under separate cover:

C-8155 Rev. 1 D-8152 Rev. 1 C-8120 Rev. 4 D-8153 Rev. 1 D-8121 Rev. 5 D-8154 Rev. 2 D-8122 Rev. 3 D-8243 Rev. 1 D-8123 Rev. 2 E-8119 Rev. 5 D-8150 Rev. 0 E-8149 Rev. 1 D-8151 Rev. 1 D-8261 Rev. O

b. Referencing Figures 6.2.1, 6.2.2, 6.2.3, and 6.3.1 of the licensing document, along with Table 6.3.1 (p. 6-28), we provide an attachment which summarizes the spring constant values used for the model. This attachment is labelled as Table 3.la. There are four levels of rattling fuel assembly.

The uppermost mass is 12.5% of the total assembly mass; the remaining three rattling masses are each 25% of the total fuel assembly mass, and 12.5% of the total fuel assembly mass is assumed to be attached to the fuel rack base plate. Fluid coupling coefficients are calculated internally in the dynamic analysis code in accordance with the Fritz model. [R.J. Fritz, "The Effects of Liquids on the Dynamic Motions of Immersed Solids, " Journal of Engineering for Industry, ASME, 2/72, pp.

167-172.]

__ - A

I.

Item,5: (Cont'd)

Response

b. The analysis considers only a single rack subjected to a 3-D seismic event. The analysis assumes that all adjacent racks will have equal and opposite local velocities to the rack being studied. This leads to maximum impact force predictions, c,d Model parameters are calculated internal to the computer code based on inputs of cell configuration and geometry. A portion of a computer output is provided showing the mass matrix for a full rack with 168 modules computer output for.th SSE EW, NS, VERT are also provided. The data is in "g" units with .01 seconds between each data point in a row.

e. The fuel racks.are considered as rigid in the time history analysis by virtue of the fact that the lowest frequencies of vibration of the rack,' treated as a cantilever beam, are well above the dominant earthquake frequency. .The seismic forcing frequency is about 5 HZ maximum. Using an 8 x 14 module bending in the weak direction, we find that the lowest cantilever frequency is 80 redians per seconds. The inertia of the rack is about the weak axis, and the mass assumed to i

vibrate (for the purposes of the preceding calculation) is the rack metal mass, all of the mass of water inside the rack plus

all of the fuel assembly mass, f,g Tables 6.1, 6.2 attached summarize maximum loads, displacements, and stress factors (see licensing document p.

6-19) for all of the runs. These are abstracted from the computer outputs. All calculations leading to these values are done internal to the code and have been previously verified by hand computations.

h. The only thermal load of consequence to the rack analysis is to examine the effect of an isolated hot cell on the welds. This is done by considering the effects of heating up a long strip (the cell wall) and retraining the expansion by the edge weld.

We find stresses from this condition to be below the allowable.

i. See Tables 6.1-6.2 for items f,g.

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\ e Table 3.1a (Refer to Figure 3.1)

RACK 8x 14 12 x 14 5

. 84 x 10 .576 x 10 5 Ky(#/In) 7 K,(#/in) .1 x 10 .1 x 10 7 K 7 (#/in) .113 x 10 l'

.113 x 10 l' ,

7 7 K 6(#/In) .246 x 10 .246 x 10 Kg (# ") .314 x 10 8

.314 x 10 8

rad h (in.) 7.25 7.25 t

H (in.) 169 169 l

1 Hominal gap values for impact springs . between fuel assembly and cell wall are .151" on each side.

2 Cap values used are the approximate distances between the wall and the girdle bar, or 50% of the spacing between girdle bars and adjacent racks. For a rack completely surrounded by other racks, the gap for K, springs is .125".

Results are given horo for a 12x14 codulo (the largost codule), and for a 8x14 configuration (which is a module with the largest aspect ratio) next to the cask pit.

A complete synopsis of the analysis of the 12x14 module subject to the SSE earthquake motions is presented in a summary table labelled ,

as Table 6.1. Table 6.1 gives the maximum values of stress factors R1 (i = 1,2,3,4,5,6).

The values given in the tables are the maximum values in time and space (all sections of the rack). Table 6.2 gives typical results for the 8x14 rack. The stress factors are defined as:

Rg = Ratio of direct tenslie or compressive stress on a net sectioncompression) support to its allowable value (note support feet only i

R 2 =v&lue Ratio of gross shear on a not section to its allowable R

3 = Ratio of maximum bending stress due , to bending about the x-axis to its allowable value for the section Rg = Ratio of maximum bending stress due to bending about the y-axis to its allowable value R 5 = Combined flexure and compressive f actor (as defined in 5.3.1e above)

R s = Combined flexure and tension (or compression) factor (as defined in 5.3.1f above) l 1

As stated before, the allowable value of Rp (i =1.2.3.4.5 e) is 1 for the OBE condition and 2 for the SSE earthquake.

The results given in the tables are for the SSE earthquake and have been given in the licensing document. The maximum stress factors '

(R i) are below the limiting value for the SSE condition for all sections. It is noted that the critical load factors reported for  ;

the support feet are all for the upper segment of the foot and are to be compared with the limiting value of 2.0.

e l

Tablos 6.1 -

6.2 also present results chich are used to show that significant margins of safety exist against local deformation of the fuel storage cell due to rattling impact of fuel assemblies and against local overstress of impact bars due to inter-rack impact.

Thebased are tabularonresults a totalshown metal assume that the rack metal thicknesses structure area. thickness of 0.06 inch in the cell l

l

4 Table 6.1 BYROM RACKS -

SUMMARY

OF RESULTS-12 x 14 (Surrounded by Other Racks) 1 q MAX. DISP. D x MAX. DISP. D y MAX. FLOOR LOAD MAX. FLOOR RUN f REMARKS (In.) (In.) (4 feet) (f) 1 Foot (f) V/H" I

1 C012 y= .8 .1683 (IMP) .0877 5

.' SSE, full 4.538 x 10 1.996x10*/1.304x10*

l C016 p= .8 .1591 (IMP) .0528 5

! SSE, half 2.348 x 10 1.159x10*/69374 j full, pos. x .

C013 p= .2 .1534 (IMP) 5 SSE full

.0868 4.915 x 10 1.499 x 10 5/29525 i

j C017 p= .2 .1172 .0724 5 i

SSE half 2.386 x 10 1.018 x 10 5/18065 full, pos. x i C014 p= .8 .1425 (IMP) .0452 6.631 x 10 "

i 10% full 4.129 x 10"/30467 i C018 p= .8 C1 .1722 (IMP) .0843 5 5 edge rack 4.633 x 10 1.985 x 10 /152457 full load i l C015 y= .2 .1291 (IMP) .1191 87830 4.058 x 10"/6835 j 10% full

C019 y= .2 .1317 (IMP) .1470 5 1.862 x 10 5/31884 4.289 x 10 C1 edge rack (full) 1 V = Vertical Direction '

H = Horizontal Direction I

i l

e  :.

Table 6.1 (continued)

BYRON RACKS LOAD FACIORS s (UPPER VALUES FOR RACK BASE -

! FUEL ASSEMBLY i

TO CELL LOWER VALUES FOR SUPPORT FEET)

INTER RACK i RUN f IMPACT LOAD IMPACT LOAD Rg R 2 Rg R3 R3 R 6 l

C012 55190 43670 .083 .108 .176 .323 .356 .405

.545 .594 .662 .853 1.032 .1.163

)

j C016 36270 34970 .044 .052 .068 .187 .213 .243 l .316 .322 .269 i

.410 .570 .619 C013 53970 27980 .092 .040 .099 .224 .267 .306

.416 .136 .202 .179 .536 .558

C017 22800 .04? .015 .062 .157 .176 .199 j .281 .084 .106 .120 .302 .306 l

k C014 17570 17530 .011 .019 .060 .049 .071 .082

, .115 .136 .199 .142 .239 .270 C018 55160 46670 .088 .123 .142 .285 .351 .400

.529 .702 .429 1.011 1.189 1.342

j. C015 19930 4310 .015 .006 .058 .049 .064 .074

.095 .031 .051 .055 .128 .134 i

l C019 53480 5825 .081 .033 .088 .216 .258 .295

.503 .137 .168 .232 .570 .595

~

f

Table 6.2 BYROM RACKS -

SUMMARY

OF RESULTS E2 RACK 8 x 14 MAX. DISP. D, MAX. DISP. D MAX. FLOOR LOAD MAX. FLOOR

• RUN # REMARKS (In.) (In.) (4 feet) (#) 1 Foot (#)

Vertical / Shear C020 p= .8 SSE .1731 (IMP) .0691 3.149 x 10

• full load 1.495 x 10*/98932 l

i CO23 p= .8 SSE .6263 (IMP) .1324 7.055 x 10

• 10% filled 5.415 x 10"/38994 j no cent. offset l

, CO21 p= .2 .4567 (IMP) .0877 2.765'x 10 5 SSE full load 1.321 x 10 S/25827 CO24 p= .8 SSE .4767 (IMP) .1127 3 50% neg. x-axis 2.345 x 10 1.286 x 10 S/66013 CO25 p= .2, SSE .1407 (IMP) .1043 1.532 x 10

• 50% neg. x-axis 8.485 x 10"/16513

CO22 p= .2 .8012 .1312 5 SSC 10% f u l.1 1.129 x 10 7.863 x 10*/13382

• VERT = Vertical reaction i SHEAR = Horizontal Reaction l

I 1

a

Table 6.2 (continued)

BYRON RACKS l E2 RACK 8 x 14 i

\

LOAD FACTORS FUEL ASSEMBLY (UPPER VALUES FOR RACK BASE -

TO CELL INTER RACK LOWER VALUES FOR SUPPORT FEET)

RUN # IMPACT LOAD IMPACT LOAD Rg R2 .R 3 Rg R3 R,t i

j CO20 2.974 x 10 " 4.812 x 10 4 .088 .115 .233 .278 .366 .419

, .412 .392 .674 .555 .878 .996 i

CO23 4 .018 .024 l 1.265 x 10 " 1.398 x 10 .063 .096 .141 .16'3 l .135 .160 .178 .286 .358 .407

, CO21 3.767 x 10 " 6039. .074 .032 .133 .283 .343 .391

.359 .119 .179 .464 .637 .704 j CO24 2.521 x 10 " 4.386 x 10 4 .065 .079 .147 .221 .316 .362 j .350 .309 .422 .482 .602 .678 j CO25 2.720 x 10 " 1.485 x 10 " .042 .019 .097 .182 .217 4

.249

.229 .073 .111 .275 .382 .422 1

CO22 1.609 x 10 " 2.806 x 10 4 .021 .011 .107 .098 .171 .197 1

.186 .046 .093 .258 .285 l .323 1

1 1

i

Itcm 6:

Specific questions related to the analysis:

a. Provide justification for ignoring the flexural rigidity of fuel. assemblies and modeling them as five separated " rattling" masses. Provide-information on the flexural rigidity of fuel assemblies. Also discuss.the manner in which the flexibilities of the multi-element fuel assemblies and cells are accounted for1in the rack module analysis.

-b. How is the vertical mass of the fuel assemblies accounted for in the model?

c. How-is the mass of the water within the fuel accounted for in the model?

d. Have the effects of hot gaps and cold gaps been considered?

e. The model in Figure 6.7 shows the gap elements between the rack module edges and fixed' boundaries are used to simulate.

inter-rack impacts. This does not' simulate the potential increase of gaps due to sliding of a row of rack modules and development of higher impact velocities with larger gaps.

Provide further justification that.such worst case effects have been accounted for.

f. Demonstrate that inter-rack impacts at mid-height-elevations are not possible.

g. Explain the significant frequency difference between the EW and the NS SSE time histories shown in Figures 6.1 and 6.2.

Response

a. The rack inertia property for bending is calculated to be at least 65000 in4 Since all cells are connecFed together to.

provide a complete grid structure, comparisons of individual fuel cell inertia should be made with the above value.

Assuming a fuel assembly as a solid bar 9" x 9" in cross section gives an inertia, per assembly, of 547 in4 This leads us to neglect the assembly stiffness in the calculation and consider only the mass distribution. The rack inertia is calculated by considering an equivalent rectangular gridwork with thicknesses equal to the cell wall, and having the same metal area as the rack structure cross section. This easily

leads to a determination of the necessary inertia characteristics for mass calculations.

6

b. All of the fuel assembly mass is assumed to move with the rack base in the calculation of vertical inertia contributions,

c. Mass of water within the racks is automatically included when

the Fritz coupling model is used. The fluid mass effects are felt both in an "added mass" type term; and in a " hydrodynamic mass" type term which is inversely proportional to the annular gap.

4

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

l l

Item 6: (Cont'd) i

Response

d. All calculations have been based on nominal gaps. On some previous jobs, we have had occasion to vary the assembly - cell wall gap to a small degree; results showed that final values were not sensitive to small variations in gap dimension.

e. Certain cases are run with larger gaps on one or more wall (simulating a rack at the edge of the group). This generally turns out to be the limiting case since the rack-rack gap also affects the hydrodynamic mass that is present due to having the rack itself considered moving in an enclosed area (surrounded by other racks). Since the use of a hypothetical fixed boundary (based on symmetry considerations) is based on the conservative assumption of out of phase motion of all adjacent structures, we feel results are certainly conservative and are bounded by the extreme case of a vibrating edge rack.

f. Mid height rack to rack impact is not considered since the rack is essentially rigid. The rack will never deform enough to impact at mid height. The girdle bars at the top prevent rack cell walls from ever coming in contact except at the top and at the base plate. The exciting frequencies are not high enough to cause other than rigid body motions of the rack,

g. Frequency difference of about 2 HZ is due to our strenuous efforts to insure statistically independent horizontal movements from time history data developed from the specified response spectra. Different envelopes were used as input data to the time history generation code, l

l

l l

Item 7:

Provide the parameters and constants used in the analysis for impact loading due to the drop of a fuel assembly. Also, provide a summary y of ductility ratios utilized to absorb kinetic energy in the tensile, flexural, compressive and shearing modes. Provide typical calculations indicating the input constants, equations used, and the results of the impact analysis.

Rosponse:

Drop analysis used sample fluid model to estimate drag acting on cell assembly, Numerical analysis is then used to calculate final velocity of assembly at the bottom of the rack. This can be used to get the input kinetic energy which must be absorbed by the base plate. We show that shear stresses that develop are not sufficient to cause a " push out" of the base plate. A draft of typical calculation is attached. No ductility factors are used.

l Let:

4 W =

weight of fuel assembly (1616 lb) h =

height above base plate (201 in) a =

side of (square) effective solid cross-section of fuel assembly = 8.548" T =

thickness of base plate (0.625 in)

Y =

Yield strees of base plate material (23,150 psi) 6 =

distance that fuel assembly penetrates into the base plate Figure 7-2 shows the final deformed configuration of the base plate with fuel assembly sitting on it.

The work dissipated in plastic deformation is F4 where F is the average resisting force exerted on the fuel assembly by the base plate.

i Using the model which has proved accurate for the l

punching of slugs out of plates- (see Paul and Zaid i (1958)) the force F may be conservatively estimated by I

i F= ( Y) 4aT

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

where j Y is the yield stress in shear according to the distortion energy theory of yielding.

Then the work dissipated in plastic deformation is F6= YaT6 73 Upon equating this work to the kinetic energy gained during the fall, we find i YaT6 = W._ V g 2

/3 2g so that '

6= 3 N I 4 YaT 2g To calculate the final velocity of a body dropping through a channel, we account for virtual mass, gravity and fluid drag. We assume that the virtual mass is equal to the buoyant mass, and that the drag i

coefficient is based on exposed frontal area of the

, fuel rods. The governing equation for a mass element

) in free fall subject only to gravity and drag ef fects is C*

• D 2, gg _ g 39 (M + My) v+ ,g y 2

where Cg = effective drag coefficient due to all con-tributing effects, and My = virtual mass of object.

If c = Wy/W, then 1- c) gw P D^ " 9 l

g = ((1+c) 2 ( 1+ c) l

{

l

I subject to v = 0 at x = 0

.. V =v g at x = h l

In finite difference form '

2 C* A*v i avt ,g = At [ 1-

• g" D

]

1+ c 2w (1+eg )

1+1 " #1 + A'i+1 I *i+1 * *i + IVi+Vi+1)At/2 For a rod like body with characteristics similar to a fuel assembly, Cg = 1.0 + A where A represents the incremental increase in ef fective drag coefficient due to the fluid being confined in a narrow channel.

Consider Figure 8-3.

A*v = A 3v3; A cell V f*A'3 3 Therefore v3= gv;vg=ggy Sihe re u=A --  ; ul =

A A3 Ac ,ty Assuming that the expansion at A 3 is to a very large l area, the'n the pressure p3 is essentially equal to the l fluid pressure outside of the

• cell. Neglecting any depth effects, an energy balance yields the pressure difference across the fuel assembly as 2

o ov w

ap = 1 (v3 2_y g 2

) ,

( ,2 _ y3 2) 2 2 l

l

,,,e . m-. , - - . - -y. , . , - ,,y.-,,-.,.v. _ - - < - , . _ , _ , - , . _ , , _ , , _ , - . , , _ , - . . , . - - - - _ - _ _ _ _ ,-_,_,.m--._.--.._. -_-._.-.-,e.. - .

as long as u 2 ut. The effective incremental resist-ing force due to cell geometry is AF = ApA* so that 2

2 2, A A= 2 v -

u1 (1-A 3 /A c,11)

A3 The BASIC program on the next page has been written for an IBM micro computer. The accuracy of the results can easily be verified by considering the drag free case with vg = /2gh for any typical set of input data.

For a fuel cell dtop to the base of the rack through a distance of 201",

the final velocity at impact with the rack base is V1 = 202 in/sec Therefore, the maximum depth of penetration is N VI .443 x 1616 x 202 2 6= .443 =

YaT 2g 23150 x 8.548 x .625 x 386.4 x 2 6= .306" < .625" Therefore, the base plate will not be penetrated by the drop. -

3 l

4 l

)

l

~

y The center-to-center distance between adjacent storage cells is not dependent on the presence or absence of support from the base plate. The purpose of the above calculation is only to show that there is no danger to the liner. In the event of a dropped fuel assembly, it is correct to say that the base plate will separate from the tube in the immediate vicinity of the affected tube. While this would result in base plate plastic bending, it would not affect center-to-center spacing since there would be no effect'on the welds between adjacent tubes nor on the base plate-to-tube wel,ds away from the immediately vicinity of the dropped assembly. The base plate, even with plastic bending occurring, will not touch the liner floor in the event of a dropped fuel assembly hitting the base plate.

't l

1 _

Drop to Top of Rack acell~= 78.72291 a3 =

50.26544 astar=70.9638

ell length =165 weight = .1616 time step = .001 drop distance = 36 eps= .268757 ed= 1

1 = 36.0581 v1=125.3354 Drop With No Virtual Mass Effect and Cd= 0 acell= 78.1456 astar=70.82908 a3 = 28.27431 cell length =165 weight = 1616 time step = .001 drop distance = 36 eps= 0 cd= 0

': 1 = 76.05587 v1=166.9243

Drop to Base of Rack acell= 78.32251 astar=73.06831

-a3 =

28.27431 cell length =165 weight = 1616 time step = .001 drop distance = 201 eps= .2763172 cd= 6.8081 x1= 201.1655 v1=202.3124

a 10 rem fue. re:4 ceco esievic te s

.  : N.:L'T celt cim..gao.n le ractut.' :CC.'30,hR Ac=CD+CD: AS=(CD-GD)+(CD-G )to3=2.s4159+"7+wa

-13 .:4;NT* ace 11=",AC,"astar=".c5 14 acIN- a3= ",A3 15 Lie = t AS/ A3)"2* ( 1-( A3/ AC) ^2) 16 PRINT "Im = "._M 17 pqINT " g e, o u t a oove . va l ..te . for im er t *' o u t 0 if oraq c..wr m '. . r "

10 INAUT LM CD=1'+r_w 20 IN UT'"1,w,ct xf" L,W,DT,X:

2 LDRINT " fuel length ="tL 22 LPRINT " weight =",W," time stec=",DT 23'LARINT "droo cistance=".XC 54 INDUT" READ IN FUEL SOLIDITY"gFR 25 Ep=FR*AS*L*64/1728*18/W 23 LDRINT "mos=",ED:LARINT "cd=",CD 40 C2=(1-EP)/(1+EA) 50 C1=64'/1728'*CD/2!*AS/(W*(1+EP))

CO-X1=0 70 V1=0:X=O-CO VmV1 X=X1 90 DV=DT*(386.4*C2-Cl*V*V) 100 V1=V+DV X1=X+.5*(V+V1)*DT

• 105 PRINT " x 1 = " . X 1, v1=",V1 109 DRINT "x1=",X1," v1=",V1 110 IF X1)=XF THEN GOTC 140 E'.SE GOTO 80 t ORINT "x1=".X1," v1=",V1

• _:9 INT "x1=",X1," v1=",Vi

. 0 LORINr" " L3RINT" "

200 END 4

1

e- 3. ~

. f, '/

j P

W . v. ,

F U E L A S S E M B L Y 6-1 e .

Y 0 ' {

{.'./YI (

i i

c BASE. ' . PL A S TIC PLATE DEFORMATION l

i FIGURE 7.2 ,

4 l

E_ _ . _ . - __ ___ _ _ _ _ . . . _ _ _ _ _ _ _ _ ___ _ _ ..._ . . _ _ _ _ _ . _ _ _ _ _ _

9

.4 t, ,I n

l I L_

=

G U .

i

(

O e g P

\

> 1 o

m e

b

~

NN =

N CD A

\ .

h 4,6 4

0 9

i i -

. . Item 8:

Provide the considerations given regarding the potential impact on the functionality of fuel rack modules due to bowing and localized deformations of fuel assemblies and fuel rack cells.

Response

Sufficient fuel assembly - cell wall clearance is provided so that bowing and localized deformation is not a problem. We show that the cell wall has sufficient strength to withstand any seismic impact load without causing any safety related affect. The dynamic analyses gives predictions of maximum assembly to cell wall impact loads. Thereafter, our concern is simply that the cell wall does not undergo significant plastic deformation.

SB/dg/0290B 1

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

I l .-

! Pequest:

9. Provide the estimated occupational dose expected from performing the " dry" reracking of the Byron spent fuel pool. These doses should be quantified l (e.g., 10 mrem, 0.1 mrem, less than 0.1 mrem) and include the following:

a. occupational dose for each phase of the SFP modification;

b. the basis for the estimate (methodology), including dose rates

and manpower;

c. maximum individual dose expected.

10. Provide the information in 9, above for the " wet" reracking, additionally addressing doses to divers. (NOTE: If wet reracking is not addressed, the safety evaluation will consider dry reracking only.)

Response

Byron's spent fuel pool is presently dry. All but eight or nine of the old NUS supplied spent fuel storage racks (i.e., those which provide low density L

fuel assembly storage) have been removed from the pool. There are several scenarios for removal of the remaining racks and installation of the new racks which allow high density fuel storage. These scenarios are dependent upon the swif tness with w ich new racks are procured and licensed.

Conceivably, the remainder of the old racks can be removed and the new racks installed prior to water being added to the spent fuel pool (i.e., " dry" reracking). In this scenario there are no radioactive sources within the fuel pool and thus the occupational dose due to reracking will be minimal.

Another likely scenario is that the remainder of the old racks will be removed and the new racks installed after water has been added to the fuel pool and spent fuel has been stored therein (i.e., " wet" reracking).

For the purposes of estimating occupational doses during reracking, two scenarios will be considered. The first is the dry reracking scenario and the remaining scenario is for wet reracking. The wet reracking scenario assumes all work (i.e., removal and installation) takes place after spent

! fuel is introduced into the spent fuel pool.

I The following assumptions are used in determining the occupational doses involved with the reracking operation.

• The dose rate environment to the work area outside the spent fuel pool is assumed to be less than or equal to two mrem /hr (Figure 12.3-32 of the Byron /Braidwood Station FSAR).

l

• The dose rate to a diver submerged within the spent fuel pool water when spent fuel is located within the pool is assumed to be five mrem /hr.

This value is simply two times the maximum expected dose rate (i.e., 2.5 mrem /hr at the spent fuel pool water surface (pg. 9.1-26 of the Byron /

Braidwood Station FSAR, Amendment 43, September 1983).

Response (Cont'd):

• It is assumed that the diver is at a sufficient underwater distance from the stored spent fuel assemblies such that the direct dose rate to the diver from the spent fuel assemblies is insigni#icant compared to the immersion dose rate of five mrem /hr due to radioactive sources suspended in the spent fuel pool water.

• When there is no water within the spent fuel pool (i.e., " dry" reracking) or wnen there is water but no spent fuel located within the pool, the dose rate within the pool will be assumed to be equal to the dose rate environ-ment outside the pool under normal operating conditions i.e., two mrem /hr.

This is a conservative assumption since dissolved activation products due to stored spent fuel is a major contributor to the dose rate environment outside the pool.

• Twenty-three new spent fuel racks are to be installed (pg. 2-1, " Licensing Report on High Density Spent Fuel Racks for Byron Units 1 and 2," NRC Docket No. 50-454, 50-455, July 25, 1986).

• The total construction time for installation of the new racks is about 24 man-hours-(i.e., 3 man-days) per rack (in the refueling building - top floor). Of this, about eight man-hours (one man-day) per rack involves use of a diver within the pool installing the new racks. The three man-day per rack figure also includes the crane operator who will be maneuvering the racks. (R. Salsbury calling A. G. Klazura, " Construction Time Involved with the Spent Fuel Pool Modification." Memorandum of Telephone Conversation dated 11-19-86).

Occupational Dose Analysis

1. Dry Reracking The occupational dose associated with dry reracking is about 1.5 man rems and is determined as follows:

Dose due to removal of nine existing (i.e., old) storage racks:

(9 racks)x(3 man-days / rack)x(2 mrem /hr)x(8 hr/ man-day) = 432 mrem Dose due to installation of 23 new storage racks:

(23 racks)x(3 man-days / rack)x(2 mrem /hr)x(8 hr/ man-day) = 1104 mrem Total dose for dry reracking: 432

+ 1104 1536 mrem or ~ 1. 5 ma n Rems During dry reracking the phases of work were not treated separately because a constant dose rate value of two mrem /hr was assumed for all phases of activity. The assumed dose rate value of two mrem /hr is conservative because the dose rate to the area immediately around the spent fuel pool would be expected to have a dose rate considerably less than two mrem /hr when there are no spent fuel assemblies within the spent fuel pool. The actual dose rate will be only slightly above that due to natural background. The

Occupational Dose Analysis (Cont'd) calculated dose 'value of 1.5 man rems is thus a conservatively determined upper limit to the' exposure associated with dry reracking.

If it is assumed that a particular individual is always present during the reracking operation and that the reracking operation requires the presence of three people, then the maximum individual dose expected would be less than 0.5 rems.

2. Wet Reracking Scenario:

This scenario assumes that nine old racks are removed a 23 new racks are installed after spent fuel has been placed in the si fuel pool.

The occupational dose associated with the wet racking scenario is about 2.3 man-rems. This dose is determined as follows:

Dose to diver during removal of nine old racks:

ma d (9 racks)x(1 k )x(5*['r)x(8 d

)= 360 mrem Dose to non-d; vers during handling of nine old racks:

ma d (9 racks)x(2 k $)*(2 'r )x(8 d

)= 288 mrem Dose to diver during installation of 23 new racks:

ma (23 racks)x(1 r k )x(5 "'r h )x(8 o

)= 920 mrem Dose to non-divers during handling of 23 new racks:

(23 racks)x(2 man;dYS)x(2"r)x(8 ,, k d

) = 736 mrem Total dose for second wet reracking scenario: 360 288 920

+ 736 2304 mrem or ~ 2.3 man Rems if it is assumed that one diver handles both the removal of the old racks and installation of the new racks then this would result in the maximum individual dose which would be 360 + 920 = 1280 mrems.

Request:

11 '. Identify, for " dry" reracking (and " wet" reracking if planned):

1. radwaste sources (e.g., racks, filters) volumes, and types (e.g.,

solid, liquid gaseous) expected from:

t Recuest'(Cont'd):

a. the reracking operation, and

b. from subsequent spent fuel pool operations; and

2. the expected occupational dose increases resulting from increased spent fuel storage (e.g., compare pre-mod radwaste doses and post-mod radwaste doses).

Response

11.la Radwaste Sources Resulting from the Reracking Operation.

Dry Reracking There will be no expected radwaste sources associated with dry reracking. During dry reracking the spent fuel pool contains no spent fuel assemblies and is empty of water. Thus, there are no radioactive sources available to produce contamination.

Wet Reracking -

The radwaste sources associated with wet reracking (i.e., those above and beyond the specific activity normally expected in the spent fuel pool water and air above the water) would primarily be of the solid and liquid types. The solid radwaste sources would include contaminated set suits, diving gear, and tools which couldn't be sufficiently decontaminated. It is assumed tools, diving gear, etc., will be reused as much as possible to reduce the quantity of solid radwaste. The primary source of liquid radwaste would be water used to decontaminate.

This liquid radwaste is expected to have extremely low specific activity and would most likely be passed through the floor drain system.

The volume of liquid radwaste produced depends on the quantity of items contaminated and the amount of water required to sufficiently decontami-nate. It is impossible to specify volumes of liquid radwaste expected.

For the situation in which spent fuel is stored in the spent fuel pool prior to the removal of all the old spent fuel racks, the old spent fuel racks upon removal from the pool would have to be sprayed down to remove contamination. This fluid (most likely water) which comprises the spray would be treated as liquid radwaste. It would have a very low specific activity and would most likely be directed to the floor drain system. Any racks which could not be sufficiently decontaminated would have to be treated as solid radwaste.

A potential source of gaseous radwaste during reracking is noble gases which could be released from solution due to agitation of the water by the divers. Most of the noble gases which are able to be released from the fuel will be released upon depressurization of the reactor vessel or during transit to the spent fuel storage pool. The gases in solution alluded to here are daughter products of halogens. Any gases released from solution due to agitation of the water would be handled by the fuel pool gaseous exhaust system. Agitation of the water doesn't affect the quantity of radioactive gases from the spent fuel pool which have to be handled by the radwaste system, it simply quickens their release.

Response (Conc d):

ll.lb The radwaste sources resulting from subsequent spent f1el pool operation for the situation involving the new spent fuel racks

~

(i.e., in the high-density storage configuration) are expected to be virtually' identical to the radwaste sources which were-expected from the fuel pool operation for the situation involving the old spent fuel racks.

Radioactivity in spent fuel pool water is due to mixing of some reactor coolant water with the spent fuel pool water during refueling operations, to the release of surface crud on the spent fuel assemblies, and to fission product leakage from defective spent fuel elements (pg. 4-11 of NUREG-0575 Vol. 1, Executive Summary Text, Project No. M-4, August,1979).

Modification to the storage capacity of the spent fuel pool has no impact on the fuel transfer operation. The amount of reactor coolant water mixed in with the spent fuel pool water during fuel transfer operations is unaffected by the modification as is the reactor coolant water contribution to the radioactivity in the spent fuel pool water.

Modification to the storage capacity also has no affect on the amount of crud on the surfaces of spent fuel assemblies nor on the crud contribu-tion to the radioactivity in the spent fuel pool water. The spent fuel pocl storage rack modification for Byron increases the number of spent fuel assemblies which can be stored within the pool from 1060 assemblies to 2940 assemblies. The greater number of fuel assemblies means there is greater potential for leakage from defective fuel elements.

This could lead to increased radioactivity in the spent fuel pool water.

Studies have shown however, that the leakage of radioactivty into spent fuel pool water from defective fuel elements which have been stored for several months is nearly nonexistant (pg. 4-12 of NURGE-0575, Vol.1, Executive Summary Text, Project No. M-4, August,1979). For the Byron spent fuel pool modification, the spent fuel storage capacity is increased from nine years to about 24 years. The additional fuel elements stored in the additional space provided by the modified spent fuel pool storage rack configuration are thus effectively nine years old. The effect on sources due to the increased capacity of the spent fuel pool is insignificant.

11.2 Occupational doses due to transfer and storage of spent fuel are the doses received during transfer of spent fuel to the storage pool and arrangement of the fuel within the storage pool, doses above the pool water surface due to radioactive sources within the spent fuel pool water, doses due to radwaste sources generated during the transfer and storage of spent fuel, doses above the water surface due to radiation shine from the stored fuel assemblies, and doses due to radiation shine through the spent fuel pool walls.

As stated in response toll.lb, modification of the spent fuel pool storage capacity has no impact on the fuel transfer operation and no significant impact on radioactivity in spent fuel pool water nor on the quantities of radwaste generated. The spent fuel pool storage modification scheme thus has no affect on the occupational doses due to transfer of spent fuel, no affect on the doses above the pool water surface due to radioactive sources within the spent fuel pool water, or no affect on the doses due to radwaste sources generated during the transfer and storage of spent fuel.

Response (Cont'd):

? The radiation doses above the pool water surface due to radiation shine from the stored fuel assemblies are an immeasurably small fraction of the doses due to natural background radiation. The higher density storage of spent fuel assemblies allowed by the modification will result in doses above the pool water surface (due to radiation shine from the stored fuel assemblies) which are still only an immeasurably small fraction of the doses due to natural background radiation. The spent fuel pool storage modification scheme has no affect on the occupa-tional doses above the pool water surface due to radiation shine from the stored fuel assemblies.

~ Table 7.3 (of the " Licensing Report on High Density Spent Fuel Racks for Byron Units 1 and 2," NRC Docket No. 50-454, 50-455, CECO, July 25, 1986) lists the dose rates expected to the areas around the spent fuel pool due to radiation shine through the fuel pool walls and the high density storage configuration. These are the maximum expected dose rates due to freshly discharged fuel which is located as close as the storage configuration allows to the pool walls. The dose rate values in Table 7.3 of the stated report will be reduced by a factor of six, 60 days after the fuel is,placed within the pool.

The east wall of the spent fuel pool is 5.5 f t. thick concrete and separates the spent fuel pool and the fuel transfer canal. The north and south walls of the spent fuel pool are 5 ft. thick concrete. They lie between the spent fuel pool and pipe penetration areas. The west spent fuel pool wall is 6 ft. thick. It lies between the spent fuel pool and the fuel pool heat exchanger area. The pipe penetration areas and the fuel pool heat exchanger area have normal operation radiation zones of 20 to 100 mrem /hr. These areas are thus radiation areas.

Access to them is controlled. As shown in Table 7.3 of the stated report, the dose rate (for the modified storage configuration) adjacent to the north spent fuel pool wall is 54 mrem /hr and the dose rate adjacent to the west wall is 4 mrem /hr. The dose rate adjacent to the north spent fuel pool wall due to the old storage configuration was calculated to be about 23 mrem /hr and the dose rate adjacent to the west wall was calculated to be about 0.24 mrem /hr. The differences in dose rates when comparing the old to new high density storage configura-tion are mainly attributed to the reduced distances between the fuel assemblies and the spent fuel pool walls which are necessary to accommodate the increased storage capacity.

As previously stated the areas outside the north, south, and west walls of the spent fuel pool are not normally occupied and access to the areas is controlled. Additionally, the dose rates outside these walls reduce drastically with time. In light of these considerations, the high density configuration for storing spent fuel would have no significant impact on the occupational dose due to radiation shine through the spent fuel pool walls.

Request:

• 12 . Quantify the gaseous effluent releases (i.e., Kr-85) and compare the releases expected due to increased spent fuel storage quantitatively with pre-mod and post-mod annual spent fuel pool and overall plant releases- (actual ^or calculated).

Response

The storage capacity of Byron's spent fuel pool was originally intended to accommodate storage for nine years worth of spent fuel assemblies resulting from operation of units 1 and 2. The modified storage capacity is intended to increase storage capacity another 14 years. The added storage space is effectively used to store fuel assemblies which have been removed from the core for at least nine years.

As stated in Section 7.2 of the " Licensing Report on High Density Spent Fuel Racks for Byron Units 1 and 2" (NRC Docket No. 50-454, 50-455, July 25, 1986),

the aged fuel in the expanded storage capacity will not contain significant amounts of radioactive iodine or short lived gaseous fission products, since these would have decayed during the storage period. Gases contained within the voids of defective spent fuel elements tend to leak out of the fuel elements quickly during depressurization of the reactor vessel and subsequent handling of the fuel. The gases which are within the fuel pellet matrix and not in the void have a very low diffusion rate (Section 4.2.2.2 of NUREG-0575 Vol.1 August,1979). Thus, the release rate for gases from defective spent fuel elements in storage, particularly the noble gas Kr-85 which is the most prominent, is expected to be extremely low in comparison to the most recently stored fuel elements.

The increased spent fuel pool storage capacity will increase the yearly quantity of Kr-85 released from the spent fuel pool into the atmosphere by no more than 3.6 Curies / year per unit. In obtaining this value it was conservatively assumed that:

• The entire auxiliary building gaseous release rate per year for Kr-85 is due to fuel handling building sources. (The fuel handlin system is part of the auxiliary building HVAC system)g . Thebuilding expected exhaust annual release rate of Kr-85 ventilated from the auxiliary building is 2.0 Curies /

year per unit (Table 11.3-6, Byron /Braidwood FSAR, Amendment 20, April, 1979).

• The value of 2 Curies / year released via the auxiliary building ventilation system is assumed to be representative of the release for the spent fuel pool with the low density storage capacity.

• The spent fuel pool activity released per year is directly proportional to the ratio of the number of fuel assemblies stored in the spent fuel pool.

This assumes that the activity released per assembly is not reduced with time of storage and is thus a very conservative assumption. (Maximum storage capacities were used in applying the ratio.)

As seen from Table 11.3-6 of the stated F3AR, the total airborne Kr-85 released per year per unit for the entire plant is 700 Curies. The calculated estimate of Kr-85 increase per year (i.e., 3.6 Curies) due to the modified spent fuel pool storage capacity is less than one percent of the total Kr-85 released per year per unit for the entire plant.

I

Request:

13. Provide a quantitative assessment of the impact of releases identified in Question 12, above on. individual and population doses offsite.

Response

The response to request numberl2, states that the additional Kr-85 released to the atmosphere as a result of the increased capacity for the spent fuel pool is less than one percent of the Kr-85 released for the entire plant (one unit).-

This is arrived at using very conservative assumptions. The predominant gas released from the stored spent fuel is expected to be Kr-85. It follows that the increased storage capacity modification will result in less than one percent increases to the offsite doses.

i

e Request:

-s 14. Compare the doses from 9 and 10above with the overall doses projected (or actually experienced) for spent fuel operations in the Byron (or a similar plant) SFP and with doses experienced or projected for overall plant operations.

Response

Typical doses experienced at the Zion nuclear power station (a PWR) due to refueling operations (i.t., removing and replacing about 1/3 of the core) are around 4.5 man rems. The dose of 1.5 man rems acquired during dry reracking as determined in response number one is conservatively estimated as 1/3 the dose expected during a single refueling operation. The actual dose is expected .to be not much greater than that acquired from natural background radiation. The dose acquired due to wet reracking was determined in response number.two to be about 2.3 man rems. This is slightly more than half the dose expected during a single refueling operation.

The yearly radiation exposure associated with routine operation of the Zion station lies around 700 to 750 man rems. The calculated doses associated with dry and wet reracking are less than one percent of the yearly plant exposures.

The reracking operation (even if performed under " wet" conditions when following ALARA practices) results in doses which are on the order of those received during a single refueling operation. These doses add insignificantly to the expected yearly plant exposures.

9

w Recuest:

15. Verify that no changes to the SFP ventilation and SFP water cleanup systems will be required for radiological purposes (e.g., need for increased capacity, higher flow rates, additional components, revised design or layout).

Response

As stated in response number two, radioactivity in the spent fuel pool water is due to mixing of some reactor coolant water with the spent fuel pool water during refueling operations, to the release of surface crud on the spent fuel assemblies, and to fission product leakage from defective spent fuel rods. Studies have shown that the leakage of fission products into spent fuel pool water from defective fuel rods which have been stored for several months is nearly nonexistant. The increased storage capacity enabled by the modified spent fuel pool storage configuration effectively handles fuel which has been out of the core at least nine years.

As such, the radioactivity in spent fuel pool water due to additional storage of fuel assemblies made possible by the modification is insignificant in comparison to the radioactivi,ty contribution due to the most recently stored fuel assemblies. The existing water cleanup system (i.e., that designed to accommodate the premodification spent fuel pool capacity) is sufficient for the modified spent fuel pool with increased storage capacity.

The gases released to the atmosphere from the modified spent fuel pool are expected to be insignificant 1y different in types and quantities from those expected to be released from the premodified spent fuel pool (see response number four). Even upon application of conservative analysis, the amount of Kr-85 expected to be released per year is less than one percent of the total Kr-85 released for the entire plant (Kr-85 is the principal gas expected to be released from stored spent fuel). The effect of the spent fuel pool modification on offsite doses due to atmospheric releases is insignificant when compared to the expected releases from the premodified spent fuel pool. The existing spent fuel pool ventilation system is sufficient.

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February 17, 1987 Docket Nos. STN 50-454 and STN 50-455 Pr. D. L. Farrar Director of Nuclear Licensing Comonwealth Edison Company Post Office Box 767 Chicaco, Illinois 60690

Dear Mr. Farrar:

I SURJECT: BYRON SPENT FUEL POOL EXPANSION On December 22, 1986, a meeting was held in Bethesda, Maryland, to discu.is the structural analysis of the spent fuel expansion for Byron Station, IJnits 1 and 2. The meeting sumary was issued January 12, 1987. As noted in Item 4 of Enclosure 2 to the meeting summary, a concern was raised regarding the conservatism of the single rack model and the potential pile up of racks against the pool wall (multi-rack impact). At the meeting, we were undecided as to whether you needed to address this concern. Subsequently, you were infomed by telephone that this concern had to be addressed on Byron. This letter confims the previous telephone reouest that the concern regarding multi-rack impact must be addressed for the Byron spent fuel pool expansinn.

Leonard N. Olshan, Project Manager Project Directorate #3 Division of PWR Licensing-A cc: See next page l

- R

F ,

1 g3 ,/ jo,, UNITED STATES 3

o NUCLEAR REGULATORY COMMISSION

( $ WASHINGTON, D. C. 20555 i_ \ / February 25, 1987 Docket Nos. STN 50-454 STN 50-455 Mr. Dennis L. Farrar

- Director or Nuclear Licensing Comonwealth Edison Company ,

Post Office Box 767 Chicago, Illinois 60690

Dear Mr. Farrar:

SUB.1ECT: BYRON SPENT FUEL POOL EXPANSION On January 13, 1987, you telecopied a proposed response to our cuestion on

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multi-rack behavior during a seismic event (Enclosure 1). Enc'osure 2 is a request for addditional information that we need to satisfy our concerrs -

regarding multi-rack behavior.

The reporting and/or recordkeeping reouirements of this letter affect fewer than ten respondents; therefore, OMB clearance is not required under PL 06-511.

Sincerely, Q .

gfeo& il.Eb.

Leonard N. Olshan, Project Mananer Project Directorate #3 Division of PWR Licensing-A Encl osure: -

As stated cc: See next page l

- - - e e _ __

M --- ~.~ ENCLOS*JRE 1 ,

S. Futman - Januery M.198f

[

Npesed Respor.se to NRC Question on N1tf rack Sthevf er Dur':; 2 U1::::tc hcnt Question d4 ll How was the conservatism of the single rack model demonstrated? The model appears to Ifmit the amount of sliding and tilting of the rack between small gaps.

wall. Has This would not account for potential pfleup of racks against the pool this possibility been investigated? 1

Response

Rack pile up will not occur because seismic acceleration levels are below the threshold that pile up. necessary for the large s1fding motton of the racks necessary for

• It may be seen from the analysts of an indivfdual isolated rack that the input acceleration levels are low enough so that sufficient displacement produce such adoes not occur to obtain the free sHdin:1 behavior necessary to pileup.

tilting of an isolated rackThe maxf aum displacement, f ac' uding sliding and

• may tend to separate the rac,ks. is 0.122 inches. However, multiple rack behavior

- This condition may be simulated through analysts considering larger gaps than 1 prevfously assumed so that the sin 11e rack analysis impact force bound the '

impacts which may occur in the poo' .

Additional analysis designed to obtain enveloping rack responses in terms of s'

impact question. force between racks and the Ifner is proposed to respond to this Multiple single rack analysts may be used to maxfafze the impact force between racks.

the condition where the racks are out of phase.The boundary The rigid stop springs atconditions the in s end of the gap stop the rack and allow 11ttle energy to be absorbed by the boundary and maximizes impact force. In reality the adjacent rack will not be force. exactly out of phase and will absorb some ener,gy tending to reduce impact F

If theliner.

wall racks spread apart in the pool, they may come in contact with the pool i* h~- The pool liner is backed by concrete and is stronger than the rack structure and thus is capablis of resisting blows larger than the rack can deliver.

l t .

.,e /

Accordingly, the rack which produces the highest im be reanalyzed considering an increase in gap Thesize. pactwillforce gap sfra be (12 x 14) will varied in 1/4" increments startin until racks do non impact. In order to maximize rack response, g at 1/4the fit 11y loaded rack case and the e half loaded rack case will be analyzed. The above analyses will be carried out for the limiting coefficients of friction;/t= 0.2 and 0.8. The above set of parameters will yield an upper bound on impact force. The maximized impact force will be used to evaluate the racks and pool wall liner.

SP:atk

Copfes

! L fsher - NRC telecony K. Singh - Holtee telecopy S. Gubtn - CECO 35 FNW T. Ryan - 28

K. Anfger - Ceco 34 FNE R. Salsbury - 22 l

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_c a w sus. 5- " - * * * '

3

's r

y ENCLOSU:.E ?

( REOUEST FOR ADDITIONAL INFORf>ATION M.'LTI DACK "EHAVIOD

. OF SPEPT FilEL POUL RACKS

1. The proposed approach could be acceptable if it can be shown that seismic acceleration levels are below the threshold necessary for large sliding motion and the results of additional analysis demonstrate significant safety margins.

, 2. The statement that the maximum displacement of an isolated rack is only

, 0.127. inches appears inconsistent with previous results which had shown

" maximum displacenents of 0.1722 inch for the 12X14 rack, and 0.8012 inch for 8X14 rack. Explain the disgrepancy. How will fluid coupling be treated in the isolated rack model?

3. If the proposed additional analysis is performed:

a) The 8X14 rack should also be analyzed since previous analyses showed that this rack experienced b'oth - maximum displacements and maximum impact loads.

b) The zero initial gap conditions should also be analyzed since it is the nominal condition. -

• [

Provide a list of additional cases proposed to be analyzed. Will both 4.

interior and edge racks be analyzed? For edge racks, will gaps to the pool -

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wall be reduced as paos between racks increase?

5. Provide the minimum safety margins including the margins on girdle bar impact.

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