ML20207E801
| ML20207E801 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/19/1999 |
| From: | Gupta V, Pellet S HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML20137S422 | List: |
| References | |
| HI-981933, HI-981933-R02, HI-981933-R2, NUDOCS 9906070155 | |
| Download: ML20207E801 (280) | |
Text
-______ - _ - _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ - - _ _ - - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ .
Docket Number 50-346 License Number NPF-3 j Serial Number 2550
. Attachment 5 DESIGN AND LICENSING REPORT DAVIS-BESSE UNIT 1 CASK PIT RACK INSTALLATION PROJECT HOLTEC INTERNATIONAL NON-PROPRIETARY VERSION (277 pages follow) fok D K 346 P PM .
, Holtec Center,555 l.incoln Drive West, Marlton, NJ 08053
{ Telephone (609) 797-0900 Fax (609) 797-0909 DESIGN AND LICENSING REPORT DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 CASK PIT RACK INSTALLATION PROJECT for the FIRST ENERGY NUCLEAR OPERATING COMPANY by HOLTEC INTERNATIONAL 555 LINCOLN DRIVE WEST MARLTON, NJ 08053 HOLTEC PROJECT NO. 80284 HOLTEC REPORT Hl-981933 REPORT CATEGORY: A REPORT CLASS: SAFETY RELATED This document version has all proprietary information removed and has replaced those sections, figures, and tables with highlighting and/or notes to designate the removal of such information. This document is to be used only in connection with the performance of work by Holtec Interr mional or its designated subcontractors. Reproduction, publication or presentation, in whole or in part, for any other purpose by any party other than the Client is expressly forbidden.
f][ l Jg[ {y I Hohec Center,555 Lincoln Drive West, Marlton, NJ 08053 HOLTEC 1NTERNATIONAL Telephone (609) 797- 090 REVIEW AND CERTIFICATION LOG DOCUMENT NAME : DESIGN AND LICENSING REPORT DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 CASK PIT RACK INSTALLATION PROJECT HOLTEC DOCUMENT l.D. NUMBER 981933 HOLTEC PROJECT NUMBER 80284 CUSTOMER / CLIENT: FIRST ENERGY REVISION BLOC 8 REVISION AUTHOR & REVIEWER &
NUMBER, QA & DATE APPROVED & DIST.
DATE DATE , DATE ORIGINAL NN LY? a M'N SH z-e5 99 v. 2-t.sr/97
% 5-*ks \ESW s S- 9q C REVISION 1
~
SW 4-o t-99 t' 6- 4 !& 94
- Tdr. +-n-o *N S*4P A-lz.99 C REVISION 2 W'
- sWV S-tT-79 M-W #- ~T* c.by c om -99
- N F+61 ' 5- 19 9 7 C
S'/'Y.ff REVISION 3 REVISION 4 REVISION S REVISION 6 This document conforms to the requirements of the design specification and the applicable sections of the goveming codes.
Note : Signatures and printed names required in the review block.
A revision of this document will be ordered by the Project Manager and carried out if any of its contents is materially affected during evolution of this project. The determination as to the need for revision will be made by the Project Manager with input from others, as deemed necessary by him.
I Must be Project Manager or his designee.
x Distribution : C: Client M: Designated Manufacturer F: Florida Office
"* Report category on the cover page indicates the contractual status of this document as "*
A = to H nobmitted to client for approval I = for client's information N = not submitted extemally THE REVISION CONTROL OF THIS DOCUMENT IS BY A
SUMMARY
OF REVISIONS LOG" PLACED BEFORE THE TEXT OF THE REPORT.
i Holtec Center,555 Lincoln Drive West, Marlton, NJ 08053 HOLTEC i N T E R N A T l O N A 1.
Telephone (609) 797- 0900 QA AND ADMINISTRATIVE INFORMATION LOG (To Be Filled in By the Principal Author of the Document and Placed After ths Title Page)
CATEGORY: Generic Document No: Hl-981933
% ProjectSpecific Holtec Project No: 80284 In accordance with the Holtec Quality Assurance Manual, and associated Holtec Quality Procedures (HQPs), this document is categorized as a :
Calculation Package
- Technical Report (Per HQP 3.2)
(Per HQP 3.2) (Such as a Licensing report)
Design Criterion Document Design Specification (Per HQP 3.4) (Per HQP 3.4)
Other(Specify):
The formatting of the contents of this document is in accordance with the instructions of HOP 3.2 or 3.4 except as noted below:
l This documentis labelled :
Nonproprietary Holtec Proprietary Privileged Intellectual Property (PIP) 4 Documents labelled Privileged Intellectual Property contains extremely valuable intellectual / commercial property of Holtec International. They can not be released to external organizations or entites without explicit '
approval of a company corporate officer. The recipient of Holtec's proprietary or Privileged Intellectual Property (PIP) document bears full and undivided responsibility to safeguard it against loss or duplication.
Revisions to the calculaton Packages may be made by adding supplements to the document and replacing the
" Table of Contents" the " Review and Certfacation" page and the " Revision Log".
r-
SUMMARY
OF REVISIONS Revision 2 contains the following pages:
COVER PAGE 1 page-REVIEW AND CERTIFICATION LOG 1 page OA AND ADMINISTRATIVE INFORMATION LOG 1 page
SUMMARY
OF REVISIONS 1 page l TABLE OF CONTENTS 9 pages
1.0 INTRODUCTION
8 pages 2.0 OVERVIEW OF THE PROPOSED CAPACITY EXPANSION 21 pages 3.0 MATERIAL, HEAVY LOAD, AND CONSTRUCTION 19 pages CONSIDERATIONS 4.0 CRITICALITY SAFETY EVALUATION 24 pages APPENDIX 4A BENCHMARK CALCULATIONS 25 pages 5.0 THERMAI HYDRAUL.IC CONSIDERATIONS 35 pages 6.0 STRUCTURAL / SEISMIC CONSIDERATIONS 63 pages 7.0 FUEL HANDLING AND MECHANICAL ACCIDENTS 29 pages 8.0 CASK PIT STRUCTURE INTEGRITY CONSIDERATIONS 17 pages 9.0 RADIOLOGICAL EVALUATION 7 pages _
10.0 INSTALLATION 9 pages 11.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT 7 pages TOTAL 277 pages Revision 1 primarily incorporates client editorial comments. Revision 1 also incorporates the results of supporting calculation changes based on client comments.
Revision 2 primarily incorporates client editorial comments, transmitted via memo dated May 14,1999.
1 l
Holtec Report HI-981933 R1 80284 l 1
)
TABLE OF CONTENTS j
4 1.0 INTR ODU CTION . . . . .. . . .. . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .
1.1 R e fe re n ce s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
........................1-5 2.0 2.1 OVERVIEW OF THE PROPOSED CAPACITY EXPANSION... ... . .. ....... .. 2-1 '
2.2 I n t rod u c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Summarv of Principal Desicn Criteria ..... ...... . .. .. . ..... . .... . ...... ...... ... ... 2-2 2.3 A onlicable Code s an d St andard s ......... .... ........ .... ......... ....... . . ........... ....... . 2-4 2.4 Ou all ty A ssu rance Pro cram ..... . .. . .... ....... . .. ....... . ... ....... .. .. .. .. . .. .....
2.5 (
2.6 M ecb an i cal De s i en . . . . .. . . . .. .. . .. .. . . . .. . ... . . . . ... 2-
. 10. . . . . . .. . I Rack Fabricat ion Met bod s ....... . .... . . .. . .. .. . .... . ......... .. . .. ... ... ... . . ... . . .. ... .. . 2 2.7 R ack Mod ule Descrint ion ..... ...... .. ....... . .. . . . . ... .... ... ... . . ... . . . . . . . .. . ...... 2- 1 I(
3.0 MATERIAL, HEAVY LOAD, AND CONSTRUCTION !
j CONSIDERATIONS .... . ......... .. . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3.1 3.2 I n t rod u c t i on . . .. . . . . . . . . . . . . . . . . . . . . . ........3-1 Structural Materials... . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..................3-1 3.3 Poison Material (Neutron Absorbert .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.3.1 Boral Material Characteiistics... . ... ...... .
.........................3-3 3.4 Comnatibilitv with Coolant...... . . . . . . . . . . . . . . . . . l
.........................3-4 3.5 Heavy Load Considerations for the Proposed Rack Installations .. . ... ..... . . .. 3-5 3.6 References ..... . .. . . . . . . . . . . . . . . . . . .
...................................3-10
.4.0 -
CRITICALITY SAFETY EVALUATION ... ... ... .... ....... ... ... .... .......... .. 4- 1 4.1 Desinn B ases ....... .... . . ... . ... . . . . . . . . . . . . . . . . . . . . . . ..... .. ... . .. 4- 1 4.2 Surhmary of Criticality Analvses .. . .. ...... ..... . . . . . . . . . . . . . . . . . . . . . . . . . 4 -4 4.2.1 Normal Operatine Conditions .. ........ . ....... ..... ...... ... . .. .....................4-4 4.2.2 Abnormal and Accident Oneratinc Conditions .... ........... . .. .... . .. . .... ......... .. 4-4 4.3 Referenee Fuel Storace Cells ... .... .. ....... ... ............ .. ....... .......................4-6 4.3.1 Re ferenee Fu el A s sem blv... . ........... . .. . .. .. .... .. ..... .. .. . .... .. .. ... . ... . . ... .... .. .. 4-6 4.3.2 Fuel S tora ce Cells . ...... .. .. ... . .. ... . ..... . . . ... ... .. .. .. . ... . . ........................4-6 l 4.4 An al vtical Met h odolo nv .. .... . . .. . ... . . . .. .. . . .. . . ... . .. ... .. . . .. . ..... .. . . . . . . . . . . . 4 4.4.1 Reference Desinn Calculations .. .............................................. ..4-7 4.4.2 Fuel Burnun Calculations and Uncertainties ..... . ..............................4-8 4.4.3 Effect of Axial Burnun Distribution.. . .. .. .. . . . . . . . . . . . . . . 4-9 4.4.4 Lone Term Chances in Reactivity...... .. . . . . . . . . . . . . . . . . . . . . . . . . . ... .... 4-10 4.5 Criticality Analyses and Tolerances. . .. .. . .. . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . 4- 1 1 4.5.1 Nominal Desien Case.. .. . . . . . . . . . .
. . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . 4-11 4.5.2 Determination of Acceptable Burnun and Enrichment Combinations.. . . ... . 4-11 4.5.3 Uncertainties Due to Tolerances .. .. . .. . ... . .. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-12 4.5.4 Eccentric Fuel Positionin e . .............. . . . ... .. .. .. . . ....... . .. ....... . . 4- 12 4.5.5 Water-Gap Spacine Between Racks.... ........ .. .... . . . . . . . . . ..... ... ... 4-12 Holtec Report HI-981933 i 80284
TABLE OF CONTENTS 4.6 '
4.6.1 Abnormal and Accident Conditions ... ......................................................... ....,4- 13
. Temocrature and Water Density Effect s ........... ...............................................; 4-13 4.6.2 4.6.3 Lateral R ack M ove me nt .... ..... ..... .. ... .. ... ... .. . . . ... .... .. . . . . . .. ... . . . . ...
Abnormal location of a Fuel Assembly ........................................ .................. 4- 14 4.6.4 4.7 Drooned Fuel A ssembi v .. ... .... .... . . .. ... .. .. .... .... . .. ..... ........ ... . . .. . . . .. . ... .
Re fe re n ce s . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. . . . . . .. . . . . . .
Appendix 4A Benchmark Calculations .......................... ..... Total of 25 Pages including 6 figures 4A.1 4A.2 Introduction and Su mmarv .............................. ........................... . ...... ........... 4 A- 1 4A.3 Effect of Sn richme n t .. . .. . ... ... .. .. ... . . . . .. . . ... .. . . . .. . .. . ... . ... . . . . .. . . . .
Effect of 4A4 "B Load i n n . . . . ...... .. . . .. . .. . . . . ... . . . . ... ... .. . . . . . . . . .. . . . .
4 A.4.1 M iscellaneous and M inor Parameters ................................. . .... ..................... 4 A-5 Reflector M aterial and S pacin e s .............. ............. ...... ....................... ........... 4 A-5 4 A.4.2 4A.4.3 Fuel Pellet Diameter and Lattice Piteh.. ............. ............................. .............. 4 A-5 Soluble Boron Concentration Effects.............. . ........ ................. .......
4A.5 4A.6 MOXFuel.......................................................................................
R e fe re n ce s . . . . . . . . . . . . . . .. . . . . . . . . ... ... ... ... . .4 A-7 .........
5.0 THERMAL-HYDRAULIC CONSIDERATIONS .........................
5.1 5.2 1 n t rod u c t i o n . . . . . . . . . . . . . . . . . . . . . ... ... . ... .... ... . .. 5 .- 1. . . . . . . .
Cooli n e S vstems De sc ript ion .... .... ..... ................ ... . .. . . ......... ........... .......... ....... .. 5-3 5.3 Discharee/Cooline Alignment Scenarios .... .. ...... .................... . ...................... 5-5 l 5.4 Maximum Bulk Pool Temperature Methodolonv ....... ......................... .......... . 5-7 5.5 Minimum Time-to-Boil and Maximum Boiloff Rate Methodolonv. ............... 5-10 5.6_ tocal Water Temocrature Methodolony .......... . .......... ..... ............................. 5-12 5.6.I Local Temoerature Evaluation Methodoloev.................................... ............... 5- 13
- 5.7 Fuel Rod Claddinn Temperature Methodoloev.......................... ...................... 5-17 5.8 Results............................................................................................................5-19 i 5.8.1 M ax i m u m B ulk Pool Te moeratures .. ...................................... ................. ........ 5- 19 i
5.8.2 t
Minimum Time-to-Boil and Maximum Boiloff Rate ................................ ...... 5-20 l 5.8.3 Local Water and Fuel Claddine Temperatures.................................................. 5-20 i 5.9 Fuel Handlinc Area Ventilation (FH AV)................. ......... ......................
5.10 i
. R e fe re n c e s . . . . . . . . . . . . . . . . . . . . . . . . ..... . . .5-23 ..........
6.0 STRUCTURA1) SEISMIC CONSIDERATIONS ...... .... ..... . . .... .. .. ....... ... 6-1 6.1 In t rod u c t i o n . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .
6.2 Overview of Rack Structural Analysis Methodolonv...... ........ ............... .... ... . 6-1 6.2.1 Backcround of Analysis Methodology...... .... ..... ......... .. .. ...... ........... .. . ... . 6-2 6.3 De scri pt ion of R ac k s . . .. . ..... ... . . .. .. .. .. . . . . . . . . .. . . ... . .... .. . ...... .. . . . .. ... .. .... . . . . . . . . . . . . . . . . . 6 -5 6.3.I Fu e l We i eh t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
6.4 S vn thet ic Ti me-H i stori es . . .. .. . . . ... . .... ................. . . . .. .. ... . . .. . . . . . .. . .. . . . . . . . . .. . . . .. . .
Holtec Repon HI-981933. ii 80284
TA.BLE OF CONTENTS 6.5 6.5.1 W PM R Me t h odolo ev .. .. ... .. .. .. . .. . . .. . . . . . .. . . . . . .. . . . . . .. . . . . . . .. . . . .. .
Model Details for Snent Fuel Racks ............... . .... ....... ............. . . . . . . . . . . ... . . :. 6-l 6.5.1.1 Assumotions........................................................
6.5.1.2 Element Details . ... ........ ..... ........ .. . . ... ............................................6-10 6.5.2 F1 ui d Cou pli n e Effect ............ ............... ...... .... . .. ..... . . .... .. . . . . .. . . . . . . .. 6- 1 1 6.5.2.1 Multi-Bodv F1uid Counlina Phenomena .......... ..... .............. . .. ..... ............ 6-12 6.5.3 S ti ffness Element Detall s .. ............... .. ......... ....... ....... . ... ........... .. .. . . . ... . . . 6- 13 6.5.4 Coefficients of Friction ...................... .. . . . ... .. ........... .. ... ... ... ..... . . . . . . 6- 14 6.5.5 Gove rnin 2 Eau ation s of M otion .......... ........ ............... ...... ............. ...... . ...... 6- 15 6.6 Structural Evaluation of Snent Fuel Rack Desinn......... .............. ..... .......... ... 6-16 6.6.1 Kinematic and Stress Accentance Criteria ... ... .... ........ .. ....... ..... ...... . ..... 6-16 6.6.2 S t res s Li m it Evaluation s .. .............. ....... ... ... .. .... ... .... ........ . .......... ... ... ... . . 6- 17 6.6.3 Dimensionless Stress Factors ... .. ..... .... ....... ...... .. . . .. .. . ..... . . . . . . .. .... 6-20 6.6.4 Loads and toadine Combinations for Snent Fuel Racks.............. ...... . . ....... 6-21 6.7 Parametric Simulations ...... ................. ........................................6-22 6.8 Time History Simulation Results.... ............................................6-25 6.8.I Rack Disn1acements ..... .... .... ... ................................................6-25 6.8.2 Pedestal Vertical Forces . .. . .. ........ ..... ... .. ..... . . . . . . . . . . . . . . . . . . . . . . . . . . 6 -2 6 6.8.3 Pedestal Friction Forces ..... ...... . . . ... .. ....... . ... ................................6-26 6.8.4 R ack Im pact lead s . ...... . . . ... . . . ....... ... . . .. . .. . . . .. . . . . . . . . . . . . . .. 6-27 6.8.4.I Rack to Rack Impacts. .. ....
.........................................................6-27 6.8.4.2 Rack to Wall Imnact s .... . ......... ...... .......... . .... . . . . . . . . . ................6-27 6.8.4.3 Fuel to Cell Wall Imnact leads ....... . ... ........ . . ... .... . ......... . .. .. . ... .. ... .. 6-28 6.9 Rack Structural Evaluation.. ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-29 6.9.1 R ack Stress Factors ... . . ... . .. ..... . .. .... . ... . .... .........................6-29 6.9.2 Pedestal Thread S hear Stress ........ .. . ... .... . . ......... .. ..... .. ......... . . . . . . . .6-31 6.9.3 Local Stresses Due to Impacts... ... ..... ..... .... ...... ..... ...........................6-31 6.9.4 Wel d S t re s ses .. ........... .. ... . . ......... ... ..............................................6-32 6.9.5 Bearin e Pad An alvsis .............. ....... .... ..... . .. .......... ..........................6-35 6.10 Le vel A E val u a tio n . . . . . . . . ... .. . . .. .. . . . . . . . .. . . . . ... .. . . . . . . . . . . . . . . . . . . . .. . . . .. . . . ..
6.11 Hydrodvnamic Loads on Cask Pit Walls....... ... .... .... .... ......... .. ........ . .. . 6~37 6.12 Local Stress Considerations .. . .......... ............. . .. .............................6-37 6.12.1 Cel l Wal l B u c k l i n c .. . . .. . . .. . . . . . .. . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . . 6 6.12.2 Analvsis of Welded Joints in Racks.. ..... .... ........ . . . . . . . . . . . . . ... 6-38 6.13 Re fere nce s .. ........ ......... . . . . . . . . . . . . . . . . . . . . . . . . .. .. .. .. . 6-39 7.0 FUEL IIANDLING AND MECHANICAL ACCIDENTS.. .. . ... . .... ..... 7-1 7.1 In t rod u c ti on .. . . . . . . . . . . . . . . . . .. . . .. . . .. .. . . . . . . . . .. . .. . ... . . .. . . . . . . ..... .7-
.. . 1 7.2 Description of Accidents..... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... 7- 1 7.2.1 Shallow Dron Events....... .. .. . . . . . . . . . . . . . . . .... . . . . . . . . . . . . . .. . . 7-2 7.2.2 Dee o D ron Eve n t s . . . . . . . . . . . .. . . . .. . . . . . .. .. . . . . . . . . . . . . . . . . . .. .. .. .. .. . . . . ... . 7-3
) 7.2.3 Rack Dron Ev.e_nl .......... ...... .. . . . ..... . ...............................7-4 7.2.4 Unlift Force Evaluation. . .... . .................................... ... . .... 7-5 7.3 l
M at h e m at ic al M od e l . .. . ... . . . .. . . . . . .. .. . . . . . . . .... ... . .. . . . .. . . . . . . . . . . . . . . . . . . . . 7-5 I
\
lloltec Repon HI-981933 iii 80284
4 TABLE OF CONTENTS 7.4 Besults............................................................................ . . . . . .. .. 7-5 7.4.1 Shallow Dron Event Results ..... .... ........... ...... .... ......... . . . . . . . . . . . . . . . . . . . . . . . ,c . 7-5 7.4.2 Deen Dron Even t R es ult s ........ ........ .. . ......... . ......... .............................7-6 7.4.3 R ack Dron Event R esu lt s . ....... .. .. .. . .. . ..... .. . . . .. ... .. . ...... . ... .. ... . . . . . . . . . . .
7.4.4 Unli ft Force Evaluation Results ....... ... ............ ... . . .. .. .. ... . .... .. ................ . 7-8 j 7.5 Clo su re .. . ... . . .. .... . . . . ... .. ... . ... ............................................................7-8 7.6 R e fe re n c e s . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .;
I 8.0 CASK PIT STRUCTURE INTEGRITY CONSIDERATIONS . ... .. .. . ... ... .. 8-1 {
8.1 In t rod u c t i o n . . .. . . . . .. . . . . . . . . . . .. . . . . . . .... .. .. .. ... ... .. .. .. .. . . . . . . . . . . . . . . . . . . . . . . . 8- 1 j 8.2 Description of Cask Pit Structures ....... ... .. .......... .... .... .............8-1 i
8.3 De fi n i t i on o f Loa d s . . . . . .. . . . .. . .. . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . .
8.3.1 Static leadina (Dead Loads and Live leads).. .. . . .. . . . . .... ............ 8-2 '
8.3.2 Seismic Induced leads.. .. . . .... ..... . .................................8-3 8.3.3 Thermal Loadin c ....... . .. . . ..... . . . . .................................8-3 8.4 Arnlvs is Procedu res ... . . . . . . ... .. . ... . ... ... . .. . ... . . .. .. . . . . . . . .............. 8-5 8.4.1 Boundarv Conditions....... ..... . .. . .. . .. . .. .. . . . . . . . . . . . . . . . . . . . . . . . ..8-6 '
8.4.2 Material Properties .... ... ..
...................................................8-6 8.4.3 Load Combinations .......... . .. ........ . ... .... . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 8-6 8.5 Results of Reinforced Concrete Analysis.... . . . . . . . ...................8-8 8.6 Pool Liner.... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................8-8 8.7 Conclusions.... ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..............8-8 8.8 References . .. ... . .... ....... .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 -9 ;
i i
9.0 RADIOLOGICAL EVALUATION .. . . . . . .. . .
l
. . . . . . . . . . . . . . . 9- 1 j 9.I Solid Radwaste.. .. .. ... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . ... ...... 9-1 9.2 {
Liauid Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......... .. ....... . . ....... .. 9-1 '
9.3 Oaseous Releases ...... .. . ,. ... . . ........ . . . . . . . . . . . . . . . . . . . .. ....... .... .. 9-1 9.4 Personnel Doses . ... . ...... . ..... .... .. .. ..... . . . . . .
...................... ... 9-3 9.5 Anticinated Dose Durine Rack Installation.... . .. .............................9-4 10.0 INSTALLATION .. ..
...................................................10-1 10.1 Introduction .. . . . . . . . . . . . .
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10- 1 10.2 Rack Arran cement......... ..... . . . . . . . . . . . . . . . . . . . . . . . . . . .10-4 10.3 Cask Pit Survey and Inspection . . . . . . . . ... . . . . . .... 10-5 10.4 Cask Pit Cooline and Purification. .. . . .. . . . . . . . . . . ... ............ 10-5 10.4.1 Cask Pit Coolin n .......... . . . .... ..... . . .. . . . . . . . . . . . .. . . . . . . . .10-5 10.4.2 Purification. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 10-6 10.5 Fuel Movement .. ...... ...... . . . . . . . . . . . . . . . . . . . . . ..... .. .. ... . 10-6 10.6 Installation of New Racks . .. .. . . . . . . . . . . . . . . . . .. . . . . ... .. 10-6 10.7 Safety. Health Physics. and ALARA Methods...... . . . . . . . . . . . . ... 10-7 10.7.1 S afet v . . . . . . . . . . ... . .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-7 10.7.2 Health Phvsics . . ......... .. . . ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-8 10.7.3 ALARA... . . . . . . . . . . . . . . . . . . ... .. . .. ...10-8 10.8 Radwaste Material Control...... .. ... . .. . . .. .. . . . . . . .. . .. . . . ... 10-9 Ilottec Report Hl.981933 iv 80284
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TABLE OF CONTENTS i- 11.0' ENVIRONMENTAL COST / BENEFIT ASSESSMENT ........... ..............,:11-1 11.1.
11.2 I n trod u c ti on . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Imperative for Rack Installation ..................................... ............................. I 1 - 1 '
11.3 Appraisal of Alternative Octions ............................................ ..... .......
11.3.1 Alternati ve Option Su mmarv .............. .... .. ....... . ..............
11.4
....... I..l-4 11.5 Co s t E stimat e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
11.6 Re source Commit me n t . .. . .. ........ ... . . .. . . . ....... .. . . .. . .. . ... .. . . . . . . . . . . ,
Environ men tal Consideration s ........... ....... ...... .......................................... I 1 -6 l 11.7 R e fe re n ce s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . .
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TABLE OF CONTENTS g .
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2.1.1 Geometric and Physical Data for High Density Racks.................................................. 2-14 2.5.1 ' ? Module Data for Spent Fuel Racks ............................................................................... 2- 15
- 3.3.1 ' Boral Experience List - PWRs ....................................................................... 3-11 and 3-12 l 3.3.2 - Boral Experience List - BWRs ................................... .................................. 3-13 and 3- 141 3.3.3 11100 Alloy Aluminum Physical Characteristics .... ........ ............................................ 3-15 L 3.3.4 L Chemical Composition - Aluminum (l 100 Alloy) ................................................ ...... 3 -
3.3.5 Chemical Composition and Physical Properties of Boron Carbide ........... .............. ... 3-17 3.5.1 - Heavy Load Handling Compliance Matrix (NUREG-0612)........... ............................ 3-18 4.1.1 Fuel A'ssembly Specifications ........... ................. ......... . .. .... ....... .. ... . ............. ............ 4- 18 4.2.1 Summary of Criticality Safety Analyses ......................................... ........ .................... 4-19 4.2.2 - Reactivity Effects of Abnormal and Accident Conditions.......... .................. .............. 4-20
'4.5.1 Reactivity Effects of Manufacturing Tolerances..... .............. ......... ..... ..................... 4-21 4.6.1. Reactivity Effects of Temperature and Void................ ................... ...... ..................... 4-22
- 4A.1 ' Summary of Criticality Benchmark Calculations....... .. .................... 4A-9 through 4A-13 1
.'4A.2; Comparison of MCNP4a and Keno 5a Calculated Reactivities j
- for Various Enrich ment s . ... . . .... .. . . . ... ... . ..... . . . .. ... . ... . ... . . . ....... . ... ......... . .!
4A.3 MCNP4a Calculated Reactivities for Critical Experiments wi th Ne u tr,on Absorbe rs . ........... ....... ....... ... .. . ........... . . .... . . .... ... .. .... ....... .. . . .... .... . . . . . . . . .. 4 A-
/- 4A.4. Comparison of MCNP4a and KEN 05a Calculated Reactivities for Various
- BLoadings..............................................................................................................4A-16.
4A.5 . . Calculations for Critical Experiments with Thick Lead anu Steel Reflectors ........... 4A-17 l
'4A.6 j Calculations for Critical Experiments with Various Soluble Boron '
Conce nt rat io n s . . . . . . . . . . . . .. . . . . . . . .~. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4A,7 Calculations for Critical Experiments with MOX Fuel ............................................. 4A-19
- 5.3.1 - Davis-Besse Historic and Projected Fuel Discharge Schedule ........ ........ ......... ........ 5-24 L 5.4.1 Data for SFP Bulk Temperature Evaluation.... ................ ......... ...................... .. ...... 5-25
'5.5.1 Data for Time-to-Boil Evaluation ............ . ........................... ..... .... .... . ... ............... . ... . 5-26 5.6.1 : Data for SFP/ Cask Pit Local Temperature Evaluation................... ...................... ....... 5-27 5.8.1 Results of Bulk Temperature Transient.......................................
................ ................ 5-28 5.8.2 L Results of Minimum Time-to-Boil and Maximum Boiloff Rate Evaluation................ 5-29 F
6.2.1 . Partial Listing of Fuel Rack Applications Using DYNARACK.......... .. 6-41 through 6-43
- 6.3.1 ; Rack Material Data (200EF) (ASME - Section II, Part D) . ....................... . ........ .... 6-44
~ 6.4.1 Time-History Statistical Correlation Results ............................ . . . . . . . . . . . . . . . . . . . . . . . .
' 6.5 .1 ' Degrees-of-Freedom .' ....... .... . .. .... .. .. . ..... . .. . ........ .. .. ........ ..... .... 6-4 . ...6 . .. . .. . .... .
6.9.1 Comparison of Bounding Calculated Loads / Stresses vs. Code Allowables at Impact Locations and at Welds .................. ................ ........ .......... ........ .. . ...... ....... 6-47 L
Holtec Report Hi-981933 vi_ 80284 y-
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Tables (continued) 7.2.1 M aterial De finition. . ..... ... .... .... .. ... .. . . ... .. .. . . .. .. ... . . . ... ... . .. . . . . . ... .. .. .. . . ... ... . ... .. . . .... . . ..... . . . . . .. 7- 10 l 7.3.1 Impactor Weight and Impactor Velocity Calculations ................................................. 7 y -7.3.2 Structural and Material Definition ofImpactor and Target............. ........................... 7-12 !
l 8.4.1 ~ Concrete and Rebar Properties . ........ . . ... ......... ...... . . .... . . . . ...... . .. ...... . ...... ..... ... ... .. ...... .. . ... 8- 10 l 8.4.2 Material Properties . . ...... .. ..... ....... . .... . .. ... .. ............... ... . . . .. . . ... ......... ... .... ... .. . . . . .. .... . . . ...... 8- 1 1 l 8.4.3 - Individual Load Case Description ................................................................. .............. 8-12 8.5.1 North Wall Safety Factors . ... . ... . ... ... ... . . . .. . ... ..... ... .. . . .. . ... ....... ..... .. . .... . ... . . ... ... .. .... .. .. .... 8- 13 8.5.2 South Wall S afety Factors . . .. . .. . ... ... .. . .... ... ...... .. .. .. . .. ... .. .. . . . ........ .... . .. .... ...... .. . .... .. . ... .. 8- 14 8.5.3 West Wall S afety Factors .. .. .. . ... ... ... ...... . . .. .. . . ... .. .... . ..... .... . . . .. . . . .. .. . . . . . . . .. .. .. . . . .. .. .. .. . .... ... 8 8.5.4 ' East Wall S afety Factors .. ...... .... .. ... . . .. .. . .... .. .. ..... .. .. ... . .. ..... . . .. ... . ... .. .. ... ..... .... ... . ............ 8- 16 l
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9.3.1, Average Activity of Weekly S FP Samples. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 9-6 9.5.1 Preliminary Estimate of Person-REM Dose During Cask Pit Rack Installation......... 9-7 l
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l Holtec Report HI-981933 vii . 80284 1
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(;gf Figures -
i1 II Cask Pit Layout - Phase li
' l .2 , Cask Pit Layout- Phase 2 11.3 " { Cask Pit Layout - Phase 3
-2.1L Schematic of Typical Davis-Besse Rack Structure
- 2.2 Seam Welded Precision Fonned Channels L 2.3 Composite Box' Assembly'
. 2.4 Typical Array of Storage Cells 02.5- ~ Support Pedestal for Holtec PWR Rack -
- 2.6 Three PWR Cells in Elevation View
.3.5.1 Auxiliary Building Spent Fuel Pool Area Plan View 4.2.1. Minimum Required Fuel Assembly Burnup as a Function of Nominal Initial Enrichment
' to Permit Storage in the Cask Pit 4.3.1 j A Two Dimen",ional Representation of the Calculational Model Used for the Cask Pit Rack Analyses 4A.1 MCNP Calculated k-eff Values for Various Values of the Spectral Index 4A.2 ' KENO 5a Calculated k-efrValues for Various Values of the SpectralIndex 4A.3 MCNP Calculated k-efrValues at Various U-235 Enrichments- 4 4A.4 KENO Calculated k-eff Values at Various U-235 Enrichments -
4A.5. Comparison of MCNP and KENO 5a Calculations for Various Fuel' '
I
^ Enrichments . _
4A.6 Comparison of MCNP and KENO 5a Calculations for Various Boron-10 1 Areal Densities-5.6.1 Two-Dimensional Spent Fuel Pool Geometry Grid 5.6.2 - Two-Dimensional Cask Pit Geometry Grid -
5.6.3 -Two-Dimensional Spent Fuel Pool Model'- Temperature Contours 5.6.4 Two-Dimensional Spent Fuel Pool Model - Velocity Vectors 5.6.5 Two-Dimensional Cask Pit Model - Temperature Contours 5.6.6 Two-Dimensional Cask Pit Model - Velocity Vectors 6.4.1 Davis-Besse SFP and Cask Pit Time History Accelerogram (X-Direction, SSE) 6.4.2 Davis-Besse SFP and Cask Pit Time History Accelerogram (Y-Direction, SSE) 6.4.3 Davis-Besse SFP and Cask Pit Time History Accelerogram (Z-Direction, SSE) 6.5.1 - Schematic of the Dynamic Model of a Single Rack Module Used in DYNARACK q 6.5.2 Fuel-to-Rack Gap / Impact Elements at Level of Rattling Mass 6.5.3 Two Dimensional View of the Spring-Mass Simulation 6,.5.4 Rack Degrees-of-Freedom with Shear and Bending Springs ;
Holtec Report HI-981933 viii 80284 l
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. TABLE OF CONTENTS H, , Fisures (continued) {
6.5.5 Rack Periphery Gap / Impact Elements
~
. 6.5.6. Phase I Rack Layout 6.5.7 ~ Phase II Rack Layout .
6.5.8 -' _ Phase III Rack Layout 6.9.1, . Bearing Pad Finite Element Model 6.9.2 yBearing Pad Analysis Concrete Compressive Stress
, 6.9.3 , Bearing Pad Analysis Bearing Pad Stress
' 6.11.1 Rack Hydrodynamic Pressures j
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6.12.1- Welded Joint In Rack
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7.2.1.. c Plan View of Shallow Drops 7.2.2 . Plan View of Deep Drope Scenario _1 7.2.3 ' Plan View of Deep Drope Scenario 2 7.4.ll - Isometric of Shallow Drope Finite Element Model ~
' 7.4.2 lsometric of Scenario 1,AShallow Drope Vor. Mises Stress 7.4.3 Isometric of Scenario 2,AShallow Drope Von Mises Stress ,
27.4.4 Isometric of Over-Pedestal " Deep Drop" Finite Element Model
' 7.4.5 Over-Pedestal " Deep Drop" Pedestal Von Mises Stress 7.4.6 Over-Pedestal " Deep Drop" Bearing Pad Von Mises Stress 7.4.7 'Over-Pedestal " Deep Drop" Liner Von Mises Stress :
7.4.8. Over-Pedestal " Deep Drop" Concrete Von Mises Stress
'7.4.9 Plan of On-Center Deep Drope Finite Element Model 7 4.10 On-Center " Deep Drop" Baseplate Von Mises Stress 7.4.1 I . On-Center " Deep Drop" Baseplate Plastic Strain 7.4.12 Isometric of On-Center Deep Drope Baseplate Deformation
' 7.4.13 " Heaviest Rack" Drop: Maximum Von Mises Stress - Liner .
7.4.14 ~ " Heaviest Rack" Drop: Maximum Vertical Stress - Concrete j
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, i 8.f.1 Plan View of Cask Pit Area '!
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. Holtec Report 111-981933 ix 80284 b
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1.0 INTRODUCTION
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- The Davis-Besse Nuclear Power Station (DBNPS) Spent Fuel Pool (SFP) has not had enough storage capacity to allow full core offload capabilities since the discharge of Cycle 11 fuel in L
. April,1998. This report was prepared to support a license amendment to add temporary storage capacity to the DBNPS Unit 1 Cask Pit in order to ngain full core offload capabilities for the current Fuel Cycle 12 and Fuel Cycle 13. The discussions and results of the design and analyses 1
of the maximum density racks to be supplied by Holtec International are provided herem. 4 L
The Davis-Besse Nuclear Power Station is a single unit pressurized water reactor (PWR) facility located 21 miles east of Toledo near Oak Harbor, Ohio. The Babcock & Wilcox (B&W)
Company designed the nuclear steam supply system. The facility, capable of an electrical output of_873 net Megawatts-clectric, received its operating license from the NRC in April 1977, and commenced commercial operations in January 1978.
The new maximum storage rack array proposed for the DBNPS Unit 1 Cask Pit is shown in the plan views provided by Figures 1.1,1.2, and 1.3. Figure 1.1 shows the completion of the first phase of the rack installation effort by placing one rack in the Cask Pit to regain full core offload
. storage capacity for the current Fuel Cycle 12 operation, which is scheduled to be completed in April,2000. A complete offload of the reactor core in April 2000 is necessary to complete the required 10-year In-Service-Inspection of the reactor vessel. Figure 1.2 shows the completion of the second phase of the rack installation effort by placing an additional rack in the Cask Pit to maintain full core offload storage capacity for Fuel Cycle 13 operation, scheduled to occur
! i between May,2000 and April,2002. Installation phases one and two were completed in April 1 1999 as a plant modification, after evaluation in accordance with 10CFR50.59 demonstrated that
\
\
g mstallation of two empty racks did not involve an unreviewed safety question. These two racks will remain unused until a license amendment is approved by the United States Nuclear Regulatory Commission (USNRC). The final phase is the installation of two additional racks as
- shown in Figure 1.3 in order to support the necessary fuel movements that would be required in a full SFP rack replacement effort. It is expected that these two racks will be installed during a !
l Holtec Report HI-981933 1-1 80284 i
SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION. I
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futum re-rack effort, which is presently planned for completion during Fuel Cycle 13. These four j
racks will be emptied and relocated to the SFP during the latter stages of the SFP re-rack. .This - l I
licensing submittal addresses only the installation and the use of these four racks in the DBNPS
' Unit 1 Cask Pit.
The new Holtec racks are freestanding and self-supponing. The principal construction materials L for the new racks are ASME SA-240-Type 304 stainless steel sheet and plate stock, and ASME SA-564-630 (precipitation hardened stainless steel) for the adjustable support spindles. The only non-stainless material utilized in the rack is the neutron absorber material, which is a boron .
carbide and aluminum-composite sandwich available under the patented product name Boral *.
The new Holtec racks are designed to the stress limits of, and analyzed in accordance with,-
..Section HI, Division 1, Subsection NF of the American Society of Mechanical Engineers
' (ASME) Boiler and Pressure Vessel (B&PV) Code [1). The material procurement,' analysis, and fabrication of the rack modules conform to 10CFR50 Appendix B requirements.
The rack design and analysis methodologies employed in the storage capacity expansion are a
~ direct evolution of previous re-rack license applications. This Design and Licensing Report documents the design and analyses performed to demonstrate that the new Holtec supplied racks .
L meet all goveming requirements of the applicable codes and standards. This repon also ,
- documents that the racks meet the USNRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", and the addendum thereto [2].
- Sections 2 and 3 of this report provide an abstract of the design and material information on the new racks.
i The criticality safety analysis requires that the neutron multiplication factor for the stored fuel
. array be bounded by the USNRC kertlimit of 0.95 under assumptions of 95% probability and l 95% confidence. The criticality safety analysis provided in Section 4 sets the requirements on
{
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? Holtec Report HI-981933 1-2 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
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i the Boral panel length and the amount of B per unit area (i.e., loading density) of the Boral panel for the new high density racks.
Thermal-hydraulic considerations require that the fuel cladding will not fail due to excessive temperature, and that the steady state pool bulk temperature will remain within the limits prescribed for the cask pit and spent fuel pool to satisfy the pool structural strength, operational, and regulatory requirements. The thermal-hydraulic analyses carried out in support of this storage expansion effort are described in Section 5.
Rack module stmetural analysis requires that the primary stresses in the rack module structure will remain below the ASME B&PV Code (Subsection NF) [1] allowables. Demonstrations of seismic and structural adequacy are presented in Section 6.0. The structural qualification also requires that the suberiticality of the stored fuel will be maintained under all postulated accident scenarios. The structural consequences of these postulated accidents are evaluated and presented in Section 7 of this report.
Section 8 contains the structural analysis to demonstrate the adequacy of the Cask Pit reinforced concrete structure. A synopsis of the geometry of the reinforced concrete structure is also presented in Section 8.
The radiological considerations are documented in Section 9.0. Sections 10, and 11, respectively, discuss the salient considerations in the installation of the new racks, and a cost / benefit and environmental assessment to establish the superiority of the wet storage expansion option.
All computer programs utilized to perform the analyses documented in this Design and Licensing Report are benchmarked and verified. These programs have been utilized by Hollec Intemational l
in numerous re-rack license tpplications over the past decade.
lloltec Report 111-981933 1-3 l 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
l
i The analyses presented herein clearly demonstrate that the rack module arrays possess wide,
, 1 margins of safety in respect to all considerations of safety specified in the OT Position Paper [2],
namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanical integrity.
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I Holtec Report HI-981933 14 80284
- SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
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i L 1.I' References
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1
- [1] American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code,Section III,1986 Edition, including up to 1988 addenda, Subsection NF, and Appendices.
[2] USNRC, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978, and Addendum dated January 18,1979.
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. Holtec Report HI-981933 1-5 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION. j i
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HDLTEE REPORI.H1- 981933 L_ _.
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N WELD SE M W'LE R N 5E FIGURE 1.3; CASK PIT LAYOUT - PHASE 3 1
1 110LTEC REPORT 111-981933 l
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, c 2.0 OVERVIEW OF THE PROPOSED CAPACITY EXPANSION
- l 1 2.1 Introduction j
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i In its fully implemented configuration, the DBNPS Cask Pit will contain four racks with a total cell
, -1 count of 289 cells. All storage rack arrays will consist of free-standing modules, made primarily from -!
1 Type 304 austenitic stainless steel containing honeycomb storage cells interconnected through
, longitudinal welds. A panel of Boral cermet containing a high areal loading of the Boron-10 (B-10) isotope provides appropriate neutron attenuation between adjacent storage cells. Figure 2.1 provides a schematic of the typical storage rack module. Data on the cross sectional dimensions, weight and cell -
count for each rack module in the cask pit are presented in Table 2.1.1.
Since the new rack modules will not utilize flux traps between storage cells, in wet storage technology terminology, they are referred to as Region 2 style racks. The baseplates on all rack modules extend out beyond the rack module periphery wall such that the plate protrusions act to set a required minimum separation between the facing cells in adjacent rack modules. This separation between rack modules serves to establish a " flux trap" space between the peripheral cells of adjacent modules. In other words, although there is a single panel of neutron absorber between any two fuel assemblies stored in the same rack, there are two poison panels with a specified water flux trap between them, separating fuel assemblies located in the cells of two facing rack modules.
Each new rack module is supported by a minimum of four pedestals, which are remotely adjustable.
Thus, the racks can be made vertical and the top of the racks can easily be made co-planar with each I other. The rack module support pedestals are engineered to accommodate minor level variations in the pool floor flatness.
' Between the rack module pedestals and the Cask Pit liner is a bearing pad, which serves to diffuse the dead load of the loaded racks into the reinforced concrete structure of the pool slab.
l 4
Holtec Report HI-981933 2-1 80284 5- l SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
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The overall design of the rack modules is similar to those presently in service in the spent fuel ools at many other nuclear plants, among them Donald C. Cook of American Electric Power, and Connecticut Yankee of Northeast Utilities. Altogether, over 50 thousand storage cells of this design have been pmvided by Holtec International to various nuclear plants around the world.
2.2 Summary of Principal Desien Criteria The key design criteria for the new spent fuel racks are set forth in the USNRC memorandum entitled
{
"OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14,1978 as modified by amendment dated January 18,1979. The individual sections of this report expound on the specific design bases derived from the above-mentioned "OT Position Paper". A brief summary of the design bases for the Cask Pit racks are summarized in the following:
a.
Disposition: All new rack modules are required to be freestanding.
b.
Kinematic Stability: All freestanding modules must be kinematically stable (against tipping or overturning) if a seismic event is imposed on any module,
- c. Structural Compliance: All primary stresses in the rack modules must satisfy the limits postulated in Section III subsection NF of the 1986 ASME B & PV Code.
d.
Thermal-Hydraulic Compliance: The spatial average bulk pool temperature is required to remain under 140"F in the wake of a partial offload, with two SFP Cooling System trains
{
in operation.
- e. Criticality Compliance: Region 2 cells must be able to store the Zircaloy clad fuel of 5.05
! weight percent (w/o) nominal enrichment and 53.51 GWD/MTU burnup while !
t maintam' ing the reactivity less than 0.95.
l l Holtec Report HI-981933 2-2 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
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- f. Radiolonical Comoliance: The re-racking must not lead to a violation of the off-site dose limits, or adversely affect the area dose environment as set forth in the DBNPS Updated
. Safety Analysis Report (USAR). The radiological implications of the installation of the j
new racks also need to be ascenained and deemed to be acceptable.
g.
Cask Pit Stnicture: The ability of the reinforced concrete structure to satisfy the load combinations set forth in the DBNPS USAR must be demonstrated.
h.
Liner Intenrity: The integrity of the liner under cyclic in-plane loading during a seismic event must be demonstrated.
- i. Bearine Pads: The bearing pad size and thickness must ensure that the pressure on the liner continues to satisfy the American Concrete Institute (ACI) limits during and after a design basis seismic event.
- j. Accident Events: In the event of postulated drop events (uncontrolled lowering of a fuel assembly, for instance), it is necessary to demonstrate that the suberitical geometry of the rack structure is not compromised.
- k. Construction Events: The field construction services required to be carried out for executing the rack installation must be demonstrated to be within the " state of the proven an". !
l The foregoing design bases are further articulated in Sections 4 through 9 of this licensing report.
Holtec Report HI-981933 2-3 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION, l
L --- --
i 1
}'
. 2.3 Anolicable Codes and Standards '
The following codes, standards and practices are used as applicable for the design, construction, and
- assembly of the fuel storage racks. Additional specific references n: lated to detailed analyses are given in each section.
a.- Design Codes
-{
u (1) ..American Institute of Steel Construction (AISC) Manual of Steel Construction, 8* I Edition,1980.
(2) American National Standards Institute (ANSI) N210-1976, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations"
' (contains guidelines for fuel rack design).
.(3) . ASME B & PV Code Section III,1986 Edition, up to and including 1988 Addenda; ASME Section VIII,1986 Edition; ASME Section IX, latest version.
(4) American Society for Nondestructive Testing SNT-TC-1 A June,1980 Recommended Practice for Personnel Qualifications and Certification in Non-destructive Testing.
(5). American Concrete Institute Building Code Requirements for Reinforced Concrete (ACI 318-63).
(6) Code Requirements for Nuclear Safety Related Concrete Structures, ACI 349-85/ACI 349R-85, and ACI 349.1R-80.
'(7) ASME Y14.5M, Dimensioning and Tolerancing .
(8). ACI Detailing Manual - 1980.
(9) ASME B & PV Code, Section II-Parts A and C,1986 Edition up to and including 1988 Addenda.
(10) ASME B & PV Code NCA3800 - Metallic Material Organization's Quality System Program.
- b. Standards of American Society for Testing and Materials (ASTM)
(1). ~ ASTM E165 - Standard Test Method for Liquid Penetrant Examination.
- Holtec Report HI-981933 - 2-4 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
i i
)
I im .
(2)' ASTM'A240 - Standard Specification for Heat-Resisting Chromium ani Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Pressure Vessels.
(3) ~ ASTM A262 - Standard Practices for Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel.'
-(4)- 1 ASTM A276 - Standard Specification for Stainless Steel Bars and Shapes.
1 (5) ASTM A479 - Standard Specification for Stainless Steel Bars and Shapes for use t !
in Boilers and other Pressure Vessels. !
u (6) ASTM A564.- Standard Specification for Hot-Rolled and Cold-Finished Age-Hardening Stainless Steel Bars and Shapes.
(7) . ASTM C750 - Standard Specification for Nuclear-Grade Boron Carbide Powder.
(8) ASTM A380 - Standard Practice for Cleaning, Descaling, and Passivation of Stainless Steel Parts, Equipment and Systems.
i (9) . ASTM C992 - Standard Specification for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.
~
(10) ASTM E3 - Standard Practice for Preparation of Metallographic Specimens.
(11)- -ASTM E190 - Standard Test Method for Guided Bend Test for Ductility of Welds.
{
- c. Weldine Code:
i ASME B & PV Code,Section IX - Welding and Brazing Qualifications, latest l version. !
- d. Ouality Assurance. Cleanliness. Packaging. Shipping, Receivine. Storage. and Handline (1) ANSI N45.2.1 - Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants - 1973 (R.G.1.37).
(2) ANSI N45.2.2 - Packaging, Shipping, Receiving, Storage and Handling ofitems for Nuclear Power Plants - 1972 (R.G.1.38).
, i (3) ANSI N45.2.6 - Qualifications of Inspection, Examination, and Testing Personnel for the Construction Phase of Nuclear Power Plants - 1978 (Regulatory Guide 1.58).
Holtec Report HI-981933 - 2-5 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
r L
i (4) ANSI N45.2.8 - Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the i
- Construction Phase of Nuclear Plants - 1975 (R.G.1.116). i b
i (5). ANSI N45.2.11 - Quality Assurance Requirements for the Design of Nuclear Power Plants - 1974 (R.G.1.64).
l (6) - ANSI N45.2.12 - Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants - 1977 (R.G.1.144).
L (7) ' ANSI N45.2.13 - Quality Assurance R.equirements for Control of Procurement of Items and Services for Nuclear Power Plants - 1976 (R. G.1.123).
- (8) ANSI N45.2.23 - Qualification of Quality Assurance Program Audit Personnel for l Nuclear Power Plants - 1978 (R.G.1.146).
L (9) l ASME B & PV Code,Section V, Nondestructive Examination, latest version. l (10) ANSI N16.9 Validation of Calculation Methods for Nuclear Criticality Safety.
- e. : USNRC Documents l (1) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling l
Applications," dated April 14,1978, and the modifications to this document of j- -
January 18,1979.
(2) NUREG 0612, " Control of Heavy Loads at Nuclear Power Plants", USNRC' l Washington, D.C., July,1980.
- f. Other ANSI Standards (not listed in the orecedine)
(1) ANSI /ANS 8.1 - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.
-(2) ANSI /ANS 8.17 - Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors. I i
(3). ANSI N45.2 - Quality Assurance Program Requirements for Nuclear Power Plants - 1977.
4 i
(4) ANSI N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants - 1974, {
i l Holtec Report HI-981933 2-6 80284 i
. SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION. I E i^
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(5)- ANSI N45.2.10 - Quality Assurance Terms and Defm' itions - 1973. '
(6) ANSI N14.6 - American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or more for Nuclear Materials - 1978.
(7) ANSI /ASME N626 Qualification and Duties of Specialized Professional
- Engineers.
- g. Code-of-Federal Regulations (CFR)
(1) 10CFR20 - Standards for Protection Against Radiation.
(2) 10CFR21 - Reporting of Defects and Non-compliance.
(3) 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants.
(4) 10CFR50 Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
(5) 10CFR61 - Licensing Requirements for Land Disposal of Radioactive Waste.
(6) 10CFR71 - Packaging and Transportation of Radioactive Material.
- h. Regulatory Guides (RG)
(1) RG 1.13 - Spent Fuel Storage Facility Design Basis (Revision 2 Proposed).
(2) RG 1.25 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors, Rev. 0 - March,1972.
(3) RG 1.28 - Quality Assurance Program Requirements - Design and Constmetion, Rev. 2 - February,1979 (endorses ANSI N45.2).
(4) RG 1.29 - Seismic Design Classification, Rev. 2 - February,1976.
(5) RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Metal.
(6) RG 1,38 - Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling ofItems for Water-Cooled Nuclear Power Plants, Rev. 2 -
May,1977 (endorses ANSI N45.2.2).
Hollec Report HI-981933 2-7 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
)
(7) RG 1.44 - Control of the Use of Sensitized Stainless Steel. ,
(8) RG 1.58 --Qualification of Nuclear Power Plant Inspection, Examination, and -
' Testing Personnel, Rev. I _- September 1980 (endorses ANSI N45.2.6).
(9) RG 1,61 - Damping Values for Seismic Design of Nuclear Power Plants, Rev. O, 1973.
(10). RG 1.64 - Quality Assurance Requirements for the Design of Nuclear Power Plants, Rev. 2 - June,1976 (endosses ANSI N45.2.11).
(11) RG 1.71 - Welder Qualifications for Areas of Limited Accessibility.
(12) RG 1.74 - Quality Assurance Terms and Definitions, Rev. 2 - February,1974 (endorses ANSIN45.2.10).
(13) RG 1.85 - Materials Code Case Acceptability - ASME Section III, Division 1.
(14) RG 1.88 - Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records, Rev. 2 - October,1976 (endorses ANSI N45.2.9).
(15) RG 1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis, Rev.1 - February,1976.
(16) RG 1.116 - Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems, Rev. 0-R - May,1977 (endorses ANSI N45.2.8-1975)
(17) RG 1.123 - Quality Assurance Requirements for Control of Procurement ofItems and Services for Nuclear Power Plants, Rev.1 - July,1977 (endorses ANSI N45.2.13).
(18) RG 1.124 - Service Limits and loading Combinations for Class 1 Linear-Type Component Supports, Revision 1, January,1978.
(19) RG 1.144 - Auditing of Quality Assurance Programs for Nuclear Power Plants, Rev.1 - September,1980 (endorses ANSI N45.2.12-1977)
(20) RG 3.4 - Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities.
(21) RG 8.8 -Information Relative to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as Reasonably Achievable (ALARA).
Holtec Report HI-981933 2-8 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMA'I10N.
r (22). IE Information Notice 83 Fuel Binding Caused by Fuel Rack Deformation.
. (23) - ' RG 8.38 - Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, June,1993.
- i. Branch Terhaie=1 Position (1) - . CPB 9.1-1 -. Criticality in Fuel Storage Facilities.
L(2) APCSB 9 Residual Decay Energy for. Light-Water Reactors for Long-Term Cooling - November,1975.
- j. American Weldinn Society (AWS) Standards (1) 'AWS Dl.1 - Structural Welding Code - Steel.
(2) AWS D1.3 - Stmeture Welding Code - Sheet Steel.
(3) AWS D9.1 - Sheet Metal Welding Code.
(4) AWS A2.4 - Standard Symbols for Welding, Brazing and Nondestructive Examination.
(5) AWS A3.0 - Standard Welding Terms and Definitions.
(6) _. AWS A5.12 - Specification for Tungsten and T.mgsten Alloy Electrodes for Arc-welding and Cutting
.(7) AWS QCl - Standard for AWS Certification of Welding Inspectors.
- 2.4 - Ouality Assurance Program
. The governing quality assurance requirements for fabrication of the spent fuel racks are stated in j 10CFR50 Appendix B. Holtec's Nuclear Quality Assurance program has been reviewed ano approved l by the DBNPS Nuclear Assurance Department. This program is designed to provide a flexible but highly controlled system for the design, analysis and licensing of customized components in accordance with various codes, specifications, and regulatory requirements.
1 The manufacturing of the racks.will be carried out by Holtec's designated manufacturer, U.S. Tool &
Die, Inc. (UST&D). The Quality Assurance system enforced on the manufacturer's shop floor shall Holtec Report HI-981933 2-9 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
+-
provide for all controls necessary to fulfill all quality assurance requirements. UST&D has .
manufactured high-density racks for over 60 nuclear plants around the world. UST&D has been audited by the nuclear industry group Nuclear Procurement Issues Committee (NUPIC), and the Quality Assurance branch of the USNRC Office of Nuclear Material Safety and Safeguards (NMSS) with satisfactory results.
The Quality Assurance System that will be used by Holtec to install the racks is also controlled by the Holtec Nuclear Quality Assurance Manual and by the DBNPS site-specific requirements.
2.5 - Mechanical Desian !
The rack modules are designed as cellular structures such that each fuel assembly has a square opening with conforming lateral support and a Hat horizontal-bearing surface. All of the storage locations are constructed with multiple cooling flow holes to ensure that redundant flow paths for the coolant are available. The basic characteristics of the spent fuel racks are summarized in Table 2.5.1.
A central objective in the design of the new rack modules is to maximize structural strength while minimizing inertial mass and dynamic response. Accordingly, the rack modules have been designed to
- simulate multi-flange beam structures resulting in excellent de-tuning characteristics with respect to the applicable seismic events. The next subsection presents an item-by-item description of the rack modules in the context of the fabrication methodology.
i 2.6 Rack Fabrication Methods I The object of this section is to provide a brief description of the rack module construction activities, which enable an independent appraisal of the adequacy of design. The pertinent methods used in manufacturing the high-density storage racks may be stated as follows: I I
~ Holtec Report HI-981933 2-10 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
- 1. . The rack modules are fabricated in such a manner that the storage cell surfaces, vyhich would come in contact with the fuel assembly, will be free of harmful chemicals and
- projections (e.g., weld splatter).
- 2. The component connection sequence and welding processes are selected to reduce fabrication distortions.
- 3. The fabrication process involves operational sequences that permit immediate accessibility for verification by the inspection staff.
- 4. The racks are fabricated per the UST&D Appendix B Quality Assurance program, which ensures, and documents, that the fabricated rack modules meet all of the requirements of the design and fabrication documents.
- 2.7 Rack Module Description The composite box assembly, the baseplate, and the support pedestals constitute the principal components of the fuel rack modules. The following description provides details of all of the major rack components..
- i. Composite box subassembly: The rack module manufacturing begins with fabrication of the " box" from ASME SA-240-304 stainless steel. The boxes are fabricated from two precision formed channels by seam welding in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. The minimum weld penetration is 80% of the box metal gage. This process results in a square cross section box, as shown in Figure 2.2. The clear inside nominal dimension of the PWR box cell is 9.0".
Sheathing of ASME SA-240-304 stainless steel is attached to each side of the box with the poison material installed in the sheathing cavity. The sheathing design objective calls Holtec Report HI-981933 2-11 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
i for securing Boral to the bo.x surface. This is accomplished by die forming the intern and external boral sheathings to provide end flares with smooth edges, as shown in Figure 2.3. The flanges of the sheathing are welded to the box using skip welds and spot welds.
The sheathings serve to locate and position the poison sheet accurately, and to preclude its movement under seismic conditions. The sheathing also isolates the Boral from the fuel assembly.
The square cross section box with Boral panels affixed to its external surfaces is referred to as the " composite box assembly". Each composite box has at least two one inch diameter lateral holes punched near its bottom edge to provide auxiliary flow holes. For those cells located over support legs, four flow holes are required to compensate for the
. loss of the baseplate flow holes described below.
The composite boxes are arranged in a checkerboard array and welded edge-to-edge to form an assemblage of storage cell locations, as shown in Figure 2.4. Austenitic stainless steel corner welds connect the storage cells to each other. The extent of welding is selected to "detune" the racks from the stipulated seismic input motion. Filler panels and corner angles are welded to the edges of boxes at the outside boundary of the rack to complete the formation of the peripheral cells. The inter-box welding and pitch adjustment is accomplished by small longitudinal connectors. The connectors are sized
- and placed to ensure that the 9.0" inside cell clear dimension on developed boxes is maintained after inclusion of any reductions from the sheathing. This assemblage of box assemblies results in a honeycomb structure with axial, flexural and torsional rigidity depending on the extent of intercell welding provided. It can be seen from Figure 2.4 that all four comers of each interior box are connected to the contiguous boxes resulting in a well-defined path for " shear flow".
i
. ii. Baseolate: A 3/4 inch thick baseplate of ASME SA-240-304 provides a contmuous i horizontal surface for supporting the fuel assemblies. The baseplate has a 5 inch diameter I hole in each cell location, except at lift locations. For the four lift locations, the flow Holtec Report HI-981933 2-12 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
l 1
i I
l.
7 holes are a 3.12 inch diameter hole with a coincidental 2.625 inch by 5.125 inci t slot to allow insertion and engagement of the lifting rig. The location of all baseplate holes coincide with' the cell centerlines. The baseplate is attached to the base of the cell assemblage by fillet welds and extends horizontally approximately 1" beyond the -
L periphery of the rack cell assemblage at locations where racks interface. The baseplate extensions beyond rack edges, located around the periphery of the Cask Pit, vary between
%" and I" iii. The neutron absorber material: As mentioned in the preceding section, Boral is used as the neutron absorber material. Each storage cell side is equipped with one integral Boral sheet (poison material).
iv. Sheathine: As described earlier, the sheathing serves as the locator and retainer of the poison material and isolates the Boral from the fuel assembly, v.
Support Pedestals: All support pedestals are the adjustable type as shown in Figure 2.5.
The 10 inch square top (female threaded) portion is made of austenitic steel material. The bottom (male threaded) part is made of ASME SA-564-630 (17:4 Ph series) stainless steel to avoid galling problems. Each support pedestal is equipped with a readily I i
accessible socket to enable remote leveling of the rack after its placement in the pool. I The support pedestals are located at the centerlines of cells to ensure accessibility of the leveling tool through the 5 inch diameter flow hole in the baseplate.
The assembly of the rack modules is carried out by welding the composite boxes in a vertical fixture with the baseplate serving as the bottom positioner. l An elevation view of the PWR storage cell is shown in Figure 2.6.
-l Holtec Report HI-981933 2-13 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
i
4 8 .
2 0
8 R
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R J K .
C H 2 4 1
2 C A 8 7 6 7 S
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5 0 0 0 0 0 O F
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s
Table 2.5.1 l l
MODULE DATA FOR SPENT FUEL RACKS
- Storage cell inside nominal dimension 9.0 in.
Cell pitch 9.22 in.
Storage cell height (above the plate) 161.625 in.
l Baseplate hole size (away from pedestal) 5.0 in. *
- i Baseplate thickness 0.75 in.
Support pedestal height 4.25 in.
Support pedestal type Remotely adjustable pedestals Number of support pedestals 4 Number of cell walls containing %" diameter j
supplemental flow holes at base for cells located away from pedestals 2 Number of cell walls containing %" diameter flow holes at base for cells located above pedestals 4 Remote lifting and handling provisions Yes I Poison material Boral Poison length 148 in.
Poison width 7.5 in.
All dimensions provide nominal values
- Except at lifting locations Holtec Report HI-981933 2-15 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION.
- !!ll 11!!: ~
s N /
/
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FIGURE 2.1;SCllEWATIC0FTYPICAl, DAVIS-DESSERACKSTRUCTURE i 111-981933
AUXILIARY FLOW HDLE
( TYPIEAL )
WELD SEAM FIGURE 2.2; SEAM WELDED PRECISION FORMED CHANNELS HI-981933 l
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FIGURE 2.4;-TYPICAL ARRAY OF STORAGE CELLS 0 NON-FLUX TRAP CONSTRUCTION )
Hi-98l933 !
- CELL '
BASEPLATE
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RACK CORNER PLAN FIGURE 2.5; SUPPORT PEDESTAL FOR HOLTEC PWR RACK
( HI-981933
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- EVEL[ PED [ ELL
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~ BAS R ATE HDLE FIGURE 2.6: THREE PWR CELLS IN ELEVATION VIEW L
1 HI981933 NPRDIUS\980Bl1KEENREKhflaRESt6
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L E 3.0
! MATERIAL. HEAVY LOAD. AND CONSTRUCTION CONSIDERATIONS .
l 3.1- Introduslip.11 h Safe storage of nuclear fuel in'the Cask pit requims that the materials utilized in the rack
' fabrication be of proven durability and compatible with the pool water environment. Likewise, all activities in the rack installations must comply with the provisions of NUREG-0612 [3.1.1] to
. eliminate the potential for damage to fuel pmsently stomd in the SFP or any safety related equipment. This section provides a synopsis of the considerations with regard to long-term .
service life and short-term construction safety.
3.2 ' Structural Materials The following structural materials are utilized in the fabrication of the new spent fuel racks:
-a.
ASME SA-240-304 for all composite box subassembly sheet metal, baseplate and cell connecting bar stock
- b. Internally threaded support pedestals: ASME SA-240-304
- c. Externally threaded spindle for the support pedestal: ASME SA-564-630 pmcipitation hardened stynless steel (heat treated to 1100"F) d.- Weld material: . ASME Type 308 and Type 308L l' ' 3.3 . Poison Material (Neutron Absorber) l In addition to the structural and non-structural stainless steel material, the racks employ Boral*,
a patented product of AAR Manufacturing, as the neutron absorber material. Boral is a hot- I
, ' rolled cermet of aluminum and boron carbide, clad in aluminum. A brief description of Boral l and its pool experience list follows.
I
-Holtec Report HI-981933 3-1 80284 j SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION I
l-
t i
1 Boral is a thermal neutron poison material composed of boron carbide and 1100 allo y aluminum.
Boron carbide is a compound having a high boron content in a physically stable and chemica inent form.~ The 1100 alloy aluminum is a lightweight metal with high tensile stmngth, which is protected from corrosion by a highly resistant o ' xide film. The oxide film is formed by a strongly adhering film ~ of impervious hydrated aluminum oxide, which passivates the surface of the aluminum in the SFP environment. The corrosion layer penetrates the aluminum surface of the ' '
boral only a few microns. There is no net loss of aluminum cladding through the passivation -
process. The central matrix of the cermet is not affected by corrosion.~ The two materials, boron carbide and aluminum, are chemically compatible and ideally suited for long-term use in the -
radiation, thermal and chemical environment of the SFP. Boral has been shown [3.3.1] to be -
superior to alternative materials previously used as neutron absorbers in storage racks.
Boral has been extensively used in fuel rack applications in recent years. Its use in the spent fuel pools as the neutron absorbing material can be attributed to its proven performance (over 150 pool years of experience) and the following unique characteristics:
- i. The content and placement of boron carbide provides a very high removal cross-section for thermal neutrons.
. ii.-
Boron carbide, in the form of fine particles, is homogeneously dispersed throughout the central layer of the Boral panels.
1 iii. . The boron carbide and aluminum materials in Boral do not degrade as a result of I long-term exposure to radiation.
I iv. The neutron absorbing central layer of Boral is clad with perrnanently bonded surfaces of aluminum.
i
- v. Boral is stable, strong, durable, and corrosion resistant.
Hohec Report HI-981933, 2 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
Boral is manufactured by AAR Manufacturing under the control and surveillance of Holtec International's Quality Assurance / Quality Control Program that conforms to the requirements of
- 10CFR50 Appendix B, " Quality Assurance Criteria for Nuclear Power Plants". Holtec International has been evaluated by the DBNPS Nuclear Assurance Department and is an j approved supplier for the design, fabrication and installation of the Cask Pit racks.
As indicated in Tables 3.3.1 and 3.3.2, Boral has been licensed by the USNRC for use in numerous BWR and PWR spent fuel storage racks and has been extensively used in international nuclearinstallations.
-3.3.1 Boral Material Characteristics 4 Aluminum: Aluminum is a silvery-white, ductile metallic element. The 1100 alloy aluminum is used extensively in heat exchangers, pressure and storage tanks, chemical equipment, reflectors and sheet metal work.
It has high resistance to corrosion in industrial and marine atmospheres. Aluminum has an atomic number of 13, atomic weight of 26.98, specific gravity of 2.69 and valence of 3. The physical, mechanical and chemical properties of the 1100 alloy aluminum are listed in Tables 3.3.3 and 3.3.4.
The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that'quickly develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting corrosion.
Boron Carbide: The boron carbide contained in Boral is a fine granulated powder that conforms to ASTM C-750-80 nuclear grade Type III. The material conforms to the chemical composition and properties listed in Table 3.3.5.
References [3.3.2], [3.3.3], and [3.3.4] provide further discussion as to the suitability of these l-materials for use in spent fuel storage module applications.
l <
' Holtec Report H1-981933 3-3 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION f
y 3.4 - Comnatibility with Coolant
' All materials used in the construction of the Holtec racks have bece determined ko be compat with the DBNPS Spent Fuel Pool / Cask Pit, and have an established history ofin-per' usage.' A evidenced in Tables 3.3.1 and 3.3.2, Boral has been successfully used in fuel pools. Austenitic stainless steel (304) is perhaps the most widely used sta'mless alloy in nuclear power plants.
i I
l i
! l l \
L Holtec Report 111-981933 3-4 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION i-
p L 3.5 - Heavy load Considerations for the PmonceA Rek Installations
_ The Spent Fuel Cask Crane (SFCC) will be used for the installation of the new storage ' racks in the Cask Pit and is subject to the requirements of NUREG-0612, " Control ~of Heavy Ioads at Nuclear Power Plants". Safe handling of heavy loads by the SFCC will be ensured by following the defense in depth approach guidelines of NUREG 0612:
.-' Defined safe load paths in accordance with approved procedures --
e: Supe; vision of heavy load lifts by designated individuals Crane operator training and~ qualification that satisfies the requirements of ANSUASME B30.2-1976 [3.5.1]
e
' Use of lifting devices (slings) that are selected, inspected and maintained in accordance with ANSIB30.9-1971 [3.5.2]
e
,. Inspection, testing and maintenance of cranes in accordance with ANSI /ASME B30.2-1976 Ensuring the design of the SFCC is equivalent to the requirements of CMAA-70
[3.5.3] and ANSUASME B30.2-1976 e
Reliability of special lifting devices by application of design safety margins, and periodic inspection and examinations using approved procedures The salient features of the lifting devices and associated procedures are described as follows:
- a. - Safe Load Paths and Procedures Safe load paths will be defined for moving the new racks in the Fuel Building. As shown in Figure 3.5.1, the Cask Pit is located west of the SFP, between the
- Auxiliary Building Train Bay / Loading Area and the SFP. This location precludes any heavy load from being lifted over the SFP or any safety-related equipment. The SFCC is interlocked to prohibit travel over the Spent Fuel Pool.
Therefore, during installation of the new racks in the Cask Pit, the new racks will not be carried directly over any portion of the SFP.
L r - Holtec Report HI-981933 3-5 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
(-. . . _ .
The SFCC interlocks will be modified to further prohibit lifting of a heavy I ad over the Cask Pit when there is fuel stored in the Cask Pit. The rack upending or laying down will be carried out in an area which is not overlapping to any safety-related system or component.
All phases of rack installation activities will be conducted in accordance with written procedures, which will be reviewed and approved by the owner.
- b. Supervision of Lifts Procedures used during the installation of the Cask Pit Racks require supervision of heavy load lifts by a designated individual who is responsible for ensuring procedure compliance and safe lifting practices.
- c. Crane Operator Training All crew members involved in the use of the lifting and upending equipment will be given training by Holtec International using a videotape-aided instruction course which has been utilized in previous rerack operations.
- d. Lifting Devices Design and Reliability The SFCC is comprised of a main hook rated for 140 tons as well as an auxiliary hook rated for 20 tons. A temporary hoist with an appropriate capacity will be attached to the SFCC hook to prevent submergence of the hook.
l Iloltec Repon 111-981933 3-6 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION !
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i The following table detennines the maximum lift weight during rack installatjon.
Item Weight (lbs)
Rack 12,150 (maximum)
Lift Rig 1,000 Rigging . 500 Temporary hoist 2,000 Total Lift
- 15,650 It is clear, based on the heaviest rack weight to be lifted, that the heaviest load being lifted is well below the rating of the SFCC hooks. The temporary hoist to be used in conjunction with the SFCC hook will be selected to provide an adequate load capacity and comply with NUREG-0612.
L Remotely engaging lift rigs, meeting all requirements of NUREG4612, will be used to lift the new rack modules. The new rack lift rig consists of four independently loaded traction rods in a lift configuration, which ensures that failure of one traction rod will not result in uncontrolled lowering of the load.
Therefore, the lift rig complies with the duality feature called for in Section 5.1.6 (3a) of NUREG 0612.
The rig has the following attributes:
e-The traction rod is designed to prevent loss of its engagement with the rig in l- ..
the locked position. Moreover, the locked configuration can be directly verified from above the pool water without the aid of an underwater camera.
Holtec Report HI-981933 37 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORM ATION i
m e
The stress analysis of the rig is carried out and the primary stress limits ,
postulated in ANSI N14.6 {3.5.4] are met.'
e ' The rig is load tested with 300% of the maximum weight to be lifted. The test weight is maintained in the air for 10 minutes. All critical weldjoints are
. liquid penetrant examined to establish the soundness of all critical joints.
- e. Crane Maintenance
' The SFCC is maintained functional per the DBNPS preventative maintenance procedures.
The proposed heavy loads compliance will be in accordance with the guidelines of NUREG-0612, which calls for measures to " provide an adequate defense-in-depth for handling of heavy loads near spent fuel...". The NUREG-0612 guidelines cite four major causes ofload handling accidents, namely-i.' operator errors ii. rigging failure iii. lack of adequate inspection iv. inadequate procedures The racking program ensures maximum emphasis on mitigating the potential load drop accidents by implementing measures to eliminate shortcomings in all aspects of the operation including the four aforementioned areas. A summary of the measures specifically planned to deal with the major causes is provided below.
1 Operator errorsr As mentioned above, comprehensive training will be provided to the installation
- crew. All training shall be in compliance with ANSI B30.2.
)
L
)
I Rigging failure: The lifting device designed for handling and installation of the new racks has I redundancies in the lift legs and lift eyes such that there are four independent load members in the new rack lift rig,'and three independent load members in the existing rack lifting rig. Failure of any one load bearing member would not lead to uncontrolled lowering of the load. The rig j Holtec Report HI-981933 3-8 80284
- SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION J
i Leomplies with all provisions of ANSI 14.6-1978, including compliance with the primary stress criteria, load testing at 300% of maximum lift load, and dye examination of critical welds.
The rig designs are similar to the rigs used in the initial racking or the rerack of numerous other plants, such as Hope Creek, Millstone Unit 1, Indian Point Unit Two, Ulchin II, Laguna' Verde, J.A. FitzPatrick, and Three Mile Island Unit 1.- 1 Lack of adequate inspection: The designer of the racks has developed a set ofinspection points that have been proven to eliminate any incidence of rework or erroneous installation in numerous prior rerack projects. Surveys and measurements are performed on the storage racks prior to and
. subsequent to placement into the pools to ensure that the as-built dimensions and installed locations are acceptable. Measurements of the pool and floor elevations are also performed to determine actual pool configuration and to allow height adjustments of the pedestals prior to rack installation. These inspections minimize rack manipulation during placement into the pool.
Inadequate procedures: Procedures will be developed to address operations pertaining to the rack installation effort, including, but not limited to, mobilization, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance.
The procedures will be the successors of the procedures successfully implemented in previous projects.
Table 3.5.1 provides a synopsis of the requirements delineated in NUREG-0612, and its intended compliance.
l-Holtec Report HI-981933 - 3-9 80284 l~ SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
g -
p-3.61 .-
References
[3.1.1] NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," July 1980.
l . [3.3.1] " Nuclear Engineering International," July 1997 issue, pp 20-23.
1 e
" [3.3.2] " Spent Fuel Storage Module Corrosion Report," Brooks & Perkins Report 554, June 1, 1977.
[3.3.3] " Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Storage '
Pools," Brooks & Perkins Report 578, July 7,1978.
[3.3.4] "Boral Neutron Absorbing / Shielding Material - Product Performance Report," Brooks A .
Perkins Report 624, July 20,1982.
1
- [3.5.1] ANSI /ASME B30.2; " Overhead and Gantry Cranes, (Top Running Bridge, Single or
~ " Multiple Girder, Top Running Trolley Holst)," American Society of Mechanical Engineers,1976.
- [3.5.2] ANSI B30.9, " Safety Standards for Slings," 1971.
[3.5.3] CMMA Specification 70," Electrical Overhead Traveling Cranes," Crane Manufacturers l , - Association of America,Inc.,1983.
- [3.5.4] ANSI N14.6-19.78, Standard for Special Lifting Devices for Shipping Containers Weighing 10000 Pounds or more for Nuclear Materials," American National Standard
-Institute, Inc.,1978.
1
' ' [3.5.5] ANSI /ASME B30.20, "Below-the-Hook Lifting Devices," American Society of
- Mechanical Engineers,1993.
i i
j l 1 Holtec Report HI-981933 3-10 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION ;
L
Table 3.3.1 BORAL EXPERIENCE LIST- PWRs Plant Utility Docket No. Mfg. Year Maine Yankee Maine Yankee Atomic Power 50-309 1977 Donald C. Cook Indiana & Michigan Electric 50-315/316 1979 Sequoyah 1,2 Tennessee Valley Authority 50-327/328 1979 Salem 1,2 Public Service Electric & Gas 50-272/311 1980 Zion 1,2 Commonwealth Edison 50-295/304 1980 Bellefonte 1,2 Tennessee Valley Authority 50-438/439 1981 Yankee Rowe Yankee Atomic Power 50-29 1964/1983 Gosgen Kernkraftwerk Gosgen-Daniken 1984 AG (Switzerland)
Koeberg 1,2 ESCOM (South Africa) 1985 Beznau 1,2 Nordostschweizerische Kraftwerke 1985 AG (Switzerland) 12 various Plants Electricite de France (France) --
1986 Indian Point 3 NY Power Authority 50-286 1987 Byron 1,2 Commonwealth Edison 50-454/455 1988 Braidwood 1,2 Commonwealth Edison 50-456/457 1988 Yankee Rowe Yankee Atomic Power 50-29 1988 Three Mile Island I GPU Nuclear 50-289 1990 Sequoyah (rerack) Tennessee Valley Authority 50-327 1992 Donald C. Cook American Electric Power 50-315/316 1992 l (rerack)
Beaver Valley Unit 1 Duquesne Light Company 50-334 1993 Fort Calhoun Omaha Public Power District 50-285 1993
( Holtec Report HI-981933 3-11 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION i _.
Table 3.3.1 .- :
BORAL EXPERIENCE LIST - PWRs Plant Utility Docket No.
1 Mfg. Year Zion 1 & 2 (rerack) - Commonwealth Edison 50-295/304 1993 Salem Units 1 & 2 Public Gas and Electric Company 50-272/311 1995 ;
(rerack) '
Ulchin Unit 1 Korea Electric Power Company --
1995 (Korea)
Haddam Neck Connecticut Yankee Atomic Power 50-213 1996 Company Ulchin Unit 2 Korea Electric Power Company --
1996 (Koma)
Kori-4 Korea Electric Power Company --
1996 I (Korea)
Yonggwang 1,2 Korea Electric Power Company --
1996 (Korea) j Sizewell B Nuclear Electric, plc (United -
1997 Kingdom)
Angra 1 Furnas Centrais-Electricas SA --
1997 (Brazil)
Waterford 3 Entergy Operations 50-382 1997 Callaway Union Electric 50-483 1998 4
IIoltec Report 111-981933 3-12 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
1 Table 3.3.2 '
ROR AI EXPRRIRNC'R I.IST RWRn Plant Utility Docket No. Mfg. Year Cooper Nebraska Public Power 50-298 1979 J.A. FitzPatrick NY Power Authority 50-333 1978 Duane Arnold Iowa Electric Light & Power 50-331 1979 Browns Ferry 1,2,3 Tennessee Valley Authority 50-259/260/296 1980 .
Bmnswick 1,2 Carolina Power & Light 50-324/325 1981 Clinton Illinois Power 50-461/462 1981 Dresden 2,3 Commonwealth Edison 50-237/249 1981 E.I. Hatch 1,2 Georgia Power 50-321/366 1981 Hope Creek Public Service Electric & Gas 50-354/355 1985 Humboldt Bay Pacific Gas & Electric Company 50-133 1985 Lacrosse Dairyland Power 50-409 1976 Limerick 1,2 Philadelphia Electric Company 50-352/353 1980 f l
Monticello Northern States Power 50-263 1978 Peachbottom 2,3 Philadelphia Electric 50-277/278 1980 Perry 1,2 Cleveland Electric Illuminating 50-440/441 1979 I
Pilgrim Boston Edison Company 50-293 1978 Susquehanna 1,2 Pennsylvania Power & Light 50-387,388 1979 I Vermont Yankee Vermont Yankee Atomic Power 50-271 1978/1986 Hope Creek Public Service Electric & Gas 50-354/355 1989 Harris Pool'B' t Carolina Power & Light 50-401 1991 Duane Arnold Iowa Electric Light & Power 50-331 1993 Pilgrim Boston Edison Company 50-293 1993 l
Holtec Report HI-981933 3-13 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
Table 3.3.2 .-
BORAL EXPERIENCE LIST- BWRs Plant Utility Docket No. Mfg. Year LaSalle 1 Commonwealth Edison 50-373 1992 Millstone Unit 1 Northeast Utilities 50-245 1989 James A.FitzPatrick NY Power Authority 50-333 1990.
Hope Creek Public Service Electric & Gas 50-354 1991 Company Duane Amold Energy Iowa Electric Power Company 50-331 1994 Center Limerick Units 1,2 PECO Energy 50-352/50-353 1994 Harris Pool 'B' t Carolina Power & Light Company 50-401 1996 Chinshan 1,2 Taiwan Power Company (Taiwan) --
1986 Kuosheng 1,2 Taiwan Power Company (Taiwan) --
1991 Lagtma Verde 1,2 Comision Federal de Electricidad --
1991 (Mexico)
Harris Pool 'B' t Carolina Power & Light Company 50-401 1996
, James A.FitzPatrick NY Power Authority 50-333 1998 h
t Fabricated racks for storage of spent fuel transhipped from Brunswick.
Holtec Report HI-981933 3-14 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION
l l
l Table 3.3.3 -
l 1100 AIIDY ATIIMINITM PHYS'CAI CHAR ACTFRISTICS Density 0.098 lbfm 3
2.713 g/cm 3 Melting Range 1190*F- 1215*F 643" - 657"C Thermal Conductivity (77*F) 2 128 BTU /hr/ft /F/ft 2
0.53 cal /sec/cm /"C/cm Coefficient of Thermal Expansion 4 13.1 x 10 in/in *F (68*F - 212*F) 4 23.6 x 10 cm/cm *C Specific Heat (221*F) 0.22 BTU /lbf'F 0.23 cal /g/*C Modulus of Elasticity 6 10 x 10 ps; Tensile Strength (75"F) 13,000 psi (annealed) 18,000 psi (as rolled)
Yield Strength (75"F) 5,000 psi (annealed) 17,000 psi (as rolled) blongation (75"F) 35-45% (annealed) 9-20% (as rolled)
Hardness (Brinell) 23 (annealed) 32 (as rolled)
Annealing Temperature 650"F 343*C i
Holtec Report Hi-981933 3-15 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
Table 3.3.4 '
CHEMICAL COMPOSITION - ALUMINUM l (1100 AlIDY) 99.00% min. Aluminum 1.00% max. Silicone and Iron 0.05-0.20% max. Copper 0.05% max. Manganese 0.10% max. Zinc 0.15% max. Other -
l l
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Holtec Report HI-981933 3-16 80284 Sl{ADED REGIONS DESIGNATE PROPRIETARY INFORMATION :
L-F i I
i Table 3.3.5 #
CHEMICAL COMPOSITION AND PHYSICAL PROPERTIES OFRORON CARRinF CHEMICAL COMPOSITION (WEIGHT PERCENT)
Total boron 70.0 min.
B' isotopic content in natural boron 18.0 Boric oxide- 3.0 max. ,
Iron 2.0 max.
Total boron plus total carbon 94.0 min.
I PHYSICAL PROPERTIES j Chemical formula - BC4 Boron content (weight percent) 78.28 %
Carbon content (weight percent) 21.72 % j Crystal structure rhombohedral l Density 0.0907 lblin' 2.51 g/cm' Melting Point 4442*F 2450"C l I
Boiling Point 6332*F l 3500*C 2
Boral Loading (minimum grams B'Oper cm ) 0.030 l
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Holtec Report HI-981933 3-17 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION i
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1-l-
Table 3.5.1 1 HRAVY I D AD M ANDI JNG COMPI JANCR M ATRTY (NIIRRG 061M l
Criterion Compliance
- 1. Are safe load paths defined for the Yes
, movement of heavy loads to minimize the l potential ofimpact, if dropped, on irradiated fuel?
- 2. Will procedures be developed to cover: Yes identification of required equipment, l inspection and acceptance criteria required before movement of load, steps and proper sequence for handling the load, defining the safe load paths, and special precautions?
- 3. Will crane operators be trained and Yes qualified?
- 4. Will special lifting devices meet the Yes guidelines of ANSI 14.6-1978?
l 5. Will non-custom lifting devices be Yes installed and used in accordance with ANSI B30.20 [3.5.5], latest edition?
l
- 6. Will the cranes be inspected and tested Yes prior to use in rack installation?
- 7. Does the crane meet the intent of ANSI Yes B30.2-1976 and CMMA-707 l
Holtec Repon HI-981933 3 18 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION L
3 %
SP[NT FEL POOL I I I L FUEL TRANSr[R CASK TUBE Pli Pli
~
M Q-.
NIW FKL CASK STORAGE WASH AR[A AR[Aj Muv.a q l
~
bNb
".L*i Auxiliary Building Train Bay / l Loading Area l
_ tsa 1 ---
_ P !
m :::
(DOOR 3001 ,
i i i FIGURE 3.5.1; AUXILIARY BUILDING SPENT FUEL POOL AREA PLAN VIEW llOLTEC REPORT HI-981933 m
9
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v 4.0 CRITICALITY SAFETY EVALUATION - -
p L4.1- Design Bases The high density spent fuel storage racks in the Cask Pit at the Davis-Besse Nuclear Power Station are designed to assure that the effective neutron multiplication factor, k.tr, is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and
. flooded with un-borated water at a temperature within the Cask Pit temperature operating range corresponding to the highest reactivity. Including all applicable uncertainties, the maximum kerr is shown' to be less than 0.95 with a 95% probability at a P5% confidence level [4.1.11 Reactivity effects of abnormal and accident conditions have also been evaluated to assure that-under credible abnormal and accident conditions, the reactivity will not exceed 0.95.
Applicable codes, standards, and regulations or pertinent sections thereof, include the following:
. ' Code ofFederal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.
USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3 - July 1981.
USNRC letter of April 14,1978, to all Power Reactor Licensees - OT Position for U
Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18,1979.
L.L Kopp, "Ggidance on the Regulatory Requirements for Criticality Analysis of Fuel !
Storage at Light-Water Reactor Power Plants," June 1998. !
1 e
- USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981.
e i ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and
_ Transportation of LWR Fuel Outside Reactors. )
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USNRC guidelines [4.1.2) and the applicable ANSI standards specify that the maximum effective multiplication factor, k.er, including bias, uncertainties, and calculational statistics, shall be less than or equal to 0.95, with 95% probability at the 95% confidence level. In the present criticality safety
- evaluation of the storage racks, the design basis target maximum k.,r was selected to be 0.945, which is more conservative than the limit specified in the regulatory guidelines. o To ensure that the true reactivity will always be less than the calculated reactivity, the following -
conservative assumptions were made:
Moderator is un-borated water at a temperature within the Cask Pit temperature operating
- range that results in the highest reactivity (4*C, corresponding to the maximum possible moderator density).
. The racks are assumed to be fully loaded with the most reactive fuel authorized to be stored in the facility without any control rods or burnable poison.
' No soluble poison (boron)is assumed to be present in the Cask Pit water under normal operating conditions.
Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.
The effective multiplication factor of an intmite tsdial array of storage cells containing fuel ,
i assemblies is used, except for the assessment of peripheral effects and certain abnormal / accident conditions where neutron leakage is inherent.
No credit is assumed for the water gap between the racks (2.0 inches, as limited by the base plate extensions) or the additional Boral panel between adjacent racks.
Holtec Report HI-981933 4-2. 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION 4
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e
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e-In-core depletion calculations assumed conservative operating conditions: highest fuel and moderator temperature (1300 'F and 610 F, respectively), a conservative allowance for the soluble boron concentrations (1000 ppm), and bumable poison (4.0 wt% B4C) rods present in each guide tube (removed at 35 GWD/MTU). These conditions produce Plutonium in excess of normal operating conditions.
The spent fuel storage racks are designed to accommodate B&W 15x15 Mark B fuel assemblies characterized by the dimensions listed in Table 4.1.1. The fuel specifications in Table 4.1.1 allow for variations in the cladding thickness,'and thus_, calculations were performed to demonstrate that the most reactive assembly design corresponds to the minimum cladding thickness (minimum clad O.D.). The design basis fuel assembly is the most reactive (minimum cladding thickness) B&W 15x15 Mark B assembly containing UO2 at a maximum initial enrichment of 5.05 i 0.05 wt% 2 "U.
The water in the Cask Pit normally contains soluble boron, which would result in a large sub-criticality margin under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting kerrof 0.95 for normal storage be evaluated for the accident condition that assumes the loss of soluble boron. The double contingency principle of ANSI N16.1-1975 and of the April 1978 NRC letter allows credit for soluble boron under other abnormal or accident conditions, since only a single independent accident need be considered at one time. Consequences of abnormal and accident conditions have also been evaluated, where " abnormal" refers to conditions which may reasonably be expected to occur during the lifetime of the plant, and " accident" refers to conditions which are not expected to occur but nevertheless must be protected against.
Holtec Report HI-981933 4-3 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
l 4
l
' 4.2 ' Summary of Criticality Analyses
?4.2.1 Normal Operatine Conditions Calculations have been performed to qualify the racks for storage of fuel assemblies with a .i 2
maximum nominal initial enrichment of 5.05 wt% "U which have accumulated a minimum burnup of 53.51 GWD/MTU, or fuel of initial nominal enrichment and burnup combinations within the acceptable domain depicted in Figure 4.2.1. For burnup-enrichment combinations within the acceptable domain depicted in Figure 4.2.1, the maximum kerrvalue is shown to be less than 0.95 (95% probability at the 95% confidence level). The criticality analyses for the Cask Pit are summarized in Table 4.2.1. The calculated maximum reactivity includes the -
reactivity effect of the axial distribution in bumup and provides an additional margin of
. uncertainty for the depletion calculations.
L The burnup criteria identified in Figure 4.2.1 for acceptable storage will be implemented by
. appropriate administrative procedures.
l l
l 4.2.2 Abnormal and Accident Conditions 1
- Although credit for the soluble poison normally present in the Cask Pit water is permitted under i abnormal or accident conditions, most abnormal or accident conditions will not result in exceeding L
the limiting reactivity even in the absence of soluble poison. The effects on reactivity of credible -
abnormal and accident conditions are discussed in Section 4.6 and summarized in Table 4.2.2. !
Administrative procedures, to assure the presence of soluble poison during fuel handling operations, preclude the possibility of the simultaneous occurrence of two independent accident ' I conditions.
1
! Assuring the presence of soluble poison'during fuel handling operations will preclude the possibility of the simultaneous occurrence of the two independent accident conditions. The largest i
' Holtec Relmrt HI-981933 4-4 80284
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reactivity increase would occur if a fresh fuel assembly of 5.05 wt%2 nU enrichment were to be inadvertently loaded into a cell with the remainder of the rack fully loaded with fuel of the highest 7
L permissible reactivity (i.e., minimum bumup of the initial enrichment). Under these accident
- l. conditions, credit for the presence of soluble poison is permitted by tim NRC guidelinest Calculations were performed to demonstrate that 650 ppm soluble boron is adequate to assure that
' the maximum korr remains below 0.945,-
. [-
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' Double contingency principle of ANSI N16.1 1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Reg. Guide 1.13 (Section 1.4, Appendix A).
Holtec Repon HI-981933 4-5 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION l h
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, rk 4.3 ' Reference Fuel Storage Cells
'f
.# d '
Reference Fuel Ad=mbly
'4.3.1
! The design basis fuel assembly, illubrated in Figure 4.3.1, is the B&W 15x15 Mark B assemb Table 4.1.1 summarizes the fuel assembly design specifications.
4.3.2 Fuel Storaan Cells -
4 Figure 4.3.1 shows the calculational model of the nominal spent fuel storage cell containing a l l .B&W 15x15 Mark B assembly. The storage cells are composed of stainless steel walls with a L
4 single fixed neutron absorber panel, Boral, (held in place by a 0.035 inch stainless steel sheathing) _
{
. centered on each side in a 0.11 inch channel. Stainless steel boxes am arranged in an altemating l
pattem such that the connection of the box corners form storage cells between those of the stainless steel boxes. These cells are located on a lattice spacing of 9.22 ' M inches. Tiie 0.075 I TM] inch thick steel walls define a storage cell which has a 9.0 inches nominal inside <
i dimension. The Boral absorber has a thickness of 0.101
- M inches and a nominal B-10 areal -
g density of 0.0324 g/cm 2 (minimum of E g/im'), The Boral absorber panels are 7.5
- M i inches in width and 148 +59/-E inches in length. Boral panels are installed on all exterior .
! walls facing other racks, as well as, non-fueled regions, i.e., the Cask Pit walls. The minimum gap between neighboring racks is 2.0 inches, as assured by the base plate extensions.
Holtec Report HI-981933 4-6 80284 ;
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- 4.4 ' .- Analytical Methodolorv 4.4.1 ' Reference Design Calculations The principal method for criticality analysis of the high density storage racks is the three-dimensional Monte Carlo code MCNP4a [4.4.1]. MCNP4a is a continuous energy three-
-- dimensional Monte Carlo code developed at the los Alamos National Laboratory. MCNP4a calculations used continuous energy cross-section data based on ENDF/B-V, as distributed with
~
the code. Independent verification calculations were performed with KEN 05a [4.4.2], which is a three-dimensional multigroup Monte Carlo code developed at the Oak Ridge National Laboratory.
The KENO 5a calculations used the 238-group cross-section library, which is based on ENDF/B-V data and is distributed as part of the SCALE-4.3 package [4.4.3], coupled with the NITAWI II program [4.4.4], which adjusts the uranium-238 cross sections to compensate for resonance self-
' shielding effects. Benchmark calculations, presented in Appendix A, indicate a bias of 0.0009 with an uncertainty ofi 0.0011 for MCNP4a and 0.0030i 0.0012 for KEN 05a, both evaluated at the 95% probability,95% confidence level [4.1.1].
Fuel depletion analyses during core operation were performed with CASMO-4, a two-dimensional l multigroup transport theory code based on capture probabilities [4.4.5 - 4.4.7]. Restarting the CASMO-4 calculations in the storage rack geometry at 4 C yields the two-dimensional infm' ite
- multiplication factor (k ) for the storage rack. Parallel calculations with CASMO-4 for the storage rack at various enrichments enable a reactivity equivalent enrichment (fresh fuel) to be determined that provides the same reactivity in the rack as the depleted fuel. CASMO-4 was also used to determine the small reactivity uncertainties (differential calculations) of manufacturing tolerances.
In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions were used in the radial direction, which has the effect of creating an infinite radial array of storage cells. Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical
. Holtec Repon HI-981933 4-7 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION I
uncertainty of the MCNP4a and KEN 05a calculated reactivities and to assure convergence, a.
e , minimum of 1 million neutmn histories were accumulated in each calculation.
4.4.2 Fuel Burnuo Calenia' ions and Uncertainties CASMO-4 was used for bumup calculations in the hot operating condition. CASMO-4 has been extensively benchmarked [4.4.6,4,4.7] against cold, clean, critical experiments (including plutonium-bearing fuel), Monte Carlo calculations, mactor operations, and heavy element concentrations in irradiated fuel.
In the CASMO-4 geometric models, each fuel rod and its cladding were described explicitly and reflective boundary conditions wem used between storage cells. These boundary conditions have the effect of creating an infmite array of storage cells.
Conservative assumptions of moderator and fuel temperatums and the average operating soluble boron concentration, along with the presence of burnable poison rods, were used to assum the highest plutonium production and hence conservatively high values of reactivity during burnup. ;
Since critical experiment data with spent fuel is not available for determining the uncertainty in t
depletion calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations [4.1.2]. Assuming the uncertainty in depletion calculations is less than 5% of the total reactivity decrement, a burnup dependent uncertainty in reactivity for burnup calculations
. was assigned. Thus, the burnup uncertainty varies (increases) with burnup. This allowance for l
burnup uncertainty was included in determination of the acceptable burnup versus enrichment !
combinations.
I i
' The majority of the uncertainty in depletion calculations derives from uncertainties in fuel and moderator temperatures and the effect of reactivity control methods (e.g., soluble boron). For depletion calculations, bounding values of these operating parameters were assumed to assure conservative results in the analyses.
Hohec Report HI-981933 4-8 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION I
4.4.3 Effect of Axial Burnuo Distribution Initially, fuel loaded into a reactor will bum with a slightly skewed cosine power distribution. As bumup progresses, the bumup distribution tends to flatten, becoming more highly burned in the central region than in the upper and lower regions. At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in n gions of high neutron leakage.
Consequently, it is expected that over most of the bumup history, fuel assemblies with distributed j burnups will exhibit a slightly lower reactivity than that calculated for the uniform average bumup.
As bumup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, preclading the existence of large regions of significantly reduced I burnup. l Among others, Turner [4.4.8] has provided generic analytic results of the axial bumup effect based upon calculated and measured axial bumup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnups at values less than about 27 GWD/MTU with small positive reactivity effects at higher burnup values. However, for the present criticality analyses, a very conservative bounding axial burnup distribution, as supplied by Toledo Edison, was used, which resulted in a larger than typical positive reactivity effect. This distribution was developed by incorporating the most reactive top and bottom regions from all assemblies (including essemblies with only one cycle burnup), and thus is not based on any single assembly. Moreover, this distribution includes the effect of partially inserted control rods, and therefore, is not typical and is very conservative. Bumup-equivalent enrichments were I i
determined with CASMO-4 for each of I8 equally spaced axial zones (a very consen'ative I representation) and used in three-dimensional Monte Carlo calculations. Results of these l
0 calculations, therefore, inherently include the effect of the axial distribution in bumup. Companson of these results to results of calculations with uniform axial bumup allows the reactivity effect of the axial bumup distribution to be quantified. This reactivity effect is included in the calculation of ;
1 the maximum kenvalues.
Holtec Report HI-981933 4-9 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION
F '. 1 l
)
- i. .
l= 4 4.4 - ~ Lonn-Term Chanaes in Rertivity At reactor shutdown, the reactivity of the fuel initially decreases due to the growth of Xe-135.
Subsequently, the Xenon decays and the reactivity increases to a maximum at several hundred hours when the Xenon is gone. Therefore, for conservatism, the Xe is set to zero in the calculations to assure maximum reactivity. During the next 50 years, the reactivity continuously.
i decreases due primarily to Pu-241 decay and Am-241 growth. No credit is taken for this long-l term decrease in reactivity other than to indicate additional and increasing conservatism in the i design criticality analysis, lI I
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)
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e : Holtec Report HI-981933 4-10 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION - i h
I L ,
L_.
, 1 4
4.5 Criticality Analyses and Tolerances -
.4.5.1 Nominal Desian Cae i
For the nominal storage cell design in the Cask Pit, the criticality safety analyses are summarized in
~
Table 4.2.1. These data confirm that the maximum reactivity remains conservatively less than the
)
regulatory limit (kwr of 0.95). An independent calculation with the KENO 5a code provides jI confirmation of the validity of the reference MCNP4a calculations.
I
(
4.5.2 Determination of Acceptable Burnuo and Enrichment Combinations CASMO-4 was used for the depletion analysis, and the restart option was used to analytically
. transfer the spent fuel into the storage rack configuration at a reference temperature of 4 *C. l i
Calculations were also made for fuel of several different initial enrichments and interpolated to define the bumup-dependent equivalent enrichmentst , at each burnup. An MCNP4a calculation
- was then made for the equivalent enrichment to establish the limiting kwr value, which includes all l applicable uncertainties and the effect of the axial burnup distribution. This calculation was used to define the boundary of the acceptable domain shown in Figure 4.2.1. Assuming the uncertainty in depletion calculations is 5% of the total reactivity decrement, a burnup dependent uncertainty in
. reactivity for burnup calculations was assigned. Thus, the burnup uncertainty varies (increases)' !
with burnup. This allowance for bumup uncertainty was included in determination of the acceptable burnup versus enrichment combinations.
l
' 'The (reactivity) equivalent enrichment is the fresh un-burned fuel enrichment that yields the same reactivity as the
' depleted fuel, both evaluated in the storage rack configuration. The equivalent enrichment may then be used in
. three-dimensional MCNP4a or KENO 5a calculations. i Holtec Report H1-981933 4-11 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION l
y 4.5.3 ' Uncertainties Due to Tolerances ~
The reactivity effects of manufacturing tolerances are tabulated, along with the actual tolerances, in Table 4.5.1. To determine the Ak associated with a specific manufacturing tolerance, the-reference kw was compared to the kwr from a calculation with the tolerance included. All of the positive Ak values from the vmbus tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity uncertainty allowance for manufacturing tolerances. All of the individual reactivity allowances were calculated for the reference fresh -
. unburned fuel assembly and for burnups enveloping the required burnup. The largest final statistically combined reactivity uncertainty allowance was conservatively used in the determination of the maximum kerr. The individual reactivity allowances are shown in Table 4.5.1.
4.5.4 Eccentric Fuel Positioninn =
The fuel assembly is assumed to be normally located in the center of the storage rack cell.
' However, calculations were also made with the fuel assemblies assumed' to be in the corner of the I storage rack cell (four-assembly cluster at closest approach). Thew calculations iridicated that I the reactivity effect is small and negative. Therefore, the reference case in which the fuel assemblies are centered is controlling and no uncertainty for eccentricity is necessary. 1
'4.5.5 Water-Gao Spacinn Between Racks The minimum water-gap between racks, which is 2.0 inches between neighboring racks,
- . constitutes a neutron flux-trap for the storage cells of facing racks. The racks are constructed with the base plates extending beyond the edge of the cells which assures that the minimum i
spacing between storage racks is maintained under all credible conditions. However, no credit is i taken for the water-gaps between racks.
i Holtec Report HI-981933 3 4-12 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION j
i l
1
[l
, ; .- vy
- e 4.6 L Abnormal and Accident Conditions
,, - 4.6.1; Temperature and Water Density Effects .
The temperature and void coefficients of reactivity in the Cask Pit are negative. Therefore, a water temperature of 4 C (39 F) was assumed for the reference calculations, which assures that the
' true reactivity will always be lower over the expected range of the Cask Pit water temperatuits.
. Temperature effects on reactivity have been calculated (CASMO-4) and the results are shown b Table 4.6.1. In addition, the introduction of voids in the waterinternal to the storage cell (to simulate boiling) decreased reactivity, as shown in Table 4.6.1.
With soluble boron present, the temperatuie coefficients of reactivity would differ from those listed in Table 4.6.1. However, the reactivities would also be substantially lower at all temperatures with soluble boron present. The data in Table 4.6.1 is pertinent to the higher-reactivity unborated case.
Since the Monte Carlo codes, MCNP4a and KEN 05a, cannot handle temperature dependence, all
-MCNP4a and KENO 5a calculations were performed at 20 C and a positive temperature correction factor (the value of Ak between CASMO-4 calculations at 20 C and 4'C) was applied to the results.
-4.6.2 Lateral Rack Movement i
l Lateral motion of the storage racks under seismic conditions could potentially alter the spacing between racks. However, no credit for the flux-trap is assumed in the analysis, and thus, the ,
calculated maximum reactivity does not rely on the spacing between racks. The minimum water gap between the racks (2.0 inches, as limited by the base plate extensions) and the Boral panels, which are installed on all exterior walls of the racks, assure that the reactivity is always less than i the design limitation. Therefore, there is no positive reactivity effect of lateral rack movement.
l Holtec Report HI-981933' 4-13 80284 i SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION j l
\.
4.6.3 AbnormalIecation of a Fuel Assembly The misplacement of a fresh un-irradiated fuel assembly could,in the absence of soluble poison, result in exceeding the regulatory limito(ken f 0.95). This analysis is based on a fresh fuel assembly of the highest permissible enrichment (5.05 wt%) being inadvertently misloaded into one of the storage cells, which are intended for bumed fuel. Soluble boron in the Cask Pit water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. Calculations were performed to demonstrate that a soluble boron concentration of 650 ppm is more than adequate to assure that the maximum ken remains below 0.945.
In addition, the mislocation of a fresh unirradiated fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (kenof 0.95). This analysis is based on a fresh fuel assembly of the highest permissible enrichment (5.05 wt%) being accidentally mislocated outside of a storage rack adjacent to other fuel assemblies. The worst case would be an assembly mislocated in a corner formed by three storage racks. Calculations were performed for this condition to demonstrate that a soluble boron concentration of 550 ppm is more than adequate to assure that the maximum ken remains below 0.945.
4.6.4 Dropped Fuel Assembly For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel in the rack of more than 12 inches. At this separation distance, the effect on reactivity is insignificant. Furthermore, the soluble boron in the Cask Pit water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.
Hollec Report HI-981933 4 14 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION l
{W o ,
- it is also possible to vertically drop an assembly into a location occupied by another assembly.
Such a vertical impact would at most cause a small compression of the stored assembly, reducing thcl water-to-fuel ratio and thereby reducing stactivity. In addition, the distance between the active fuel regions of both assemblies will be more than sufficient to ensure no neutron interaction between the two assemblies.
L S nctural analysis has shown that dropping an assembly into an unoccupied cell could result in a j localized deformation of the baseplate of the rack. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in the h
active fuel height of that assembly no longer being completely covered by the Boral. The l . immediate eight surrounding fuel cells could also be affected. However, the amount of deformation for these cells would be considerably less. Structural analysis has shown that the arc Junt of localized deformation may be as great as 3.36 inches. The reactivity consequence of 1
l this situation was calculated and found to be statistically insignificant. For simplicity in modeling, .
I the calculation conservatively assumed an infinite array of assemblies in this damaged condition, and denionstrated the reactivity effect to be negligible. Since this is a localized event (nine storage cells at most) the actual reactivity effect will be even less than the calculated value. Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the
!' . limiting value for this dropped fuel accident. Consequently, a dropped fuel bundle will have a
{
negligible impact on reactivity.
I l
l Holtec Report HI-981933 4-15 80284
, .. SliADED REGIONS DPSIGNATE PROPRIETARY INFORMATION o
j'
7 4.7 ~ References
[4.1.1] M. G. Natrella, Exnerimental Statistics. National Bureau of Standards Handbook 91, August 1%3.
'[4.1.2] LL Kopp, " Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," June 1998.
r
[4.4.1]' J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, los Alamos National Laboratory (1993).
{4.4.2] L.M. Petrie and N.F. Landers, " KENO Va - An Improved Monte Carlo Criticality Program with Supergrouping," Volume 2, Section F11 from " SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR-0200, Rev. 4,~ January 1990.
1
[4.4.3] " SCALE 4.3: A Modular System for Performing Standardized Computer Analysis for i Licensing Evaluation For Workstations and Personal Computers, Volume 0," CCC-545 ORNL-RSICC, Oak Ridge National Laboratory (1995).
1-
- [4.4.4] N.M. Greene, L.M. Petric and R.M. Westfall, "NITAWL-II: Scale System Module for Performing Shielding and Working Library Production," Volume 1, Section F1 from
" SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR-0200, Rev. 4, January 1990.
[4.4.5] M. Edenius, K. Ekberg, B.H. Forssen, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1, Studsvik of America (1995).
1 i
Holtec Report HI-981933 4-16 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION 6.
(4.4.6] D. Knott, "CASMO-4 Benchmark Against Critical Experiments," Studsvik/SOA-94/13 (proprietary), Studsvik of America (1995).
[4.4.7] D. Knott, "CASMO-4 Benchmark Against MCNP," Studsvik/SOA-94/12 (proprietary),
Studsvikof America (1995).
[4.4.8) S.E. Turner, " Uncertainty Analysis - Burnup Distributions", presented at the DOE /SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ ENS Conference, Washington, D.C., November 2,1988.
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Ho'tec Repon Hi-981933 4-17 80284 ,
SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION -
Table 4.1.1 Fuel Assembly Specifications Fuel Rod Data Fuel pellet outside diameter, in. 0.370 Cladding thickness, in.
0.0195 - 0.0265 Cladding outside diameter, in. 0.416 - 0.430 Cladding inside diameter, in. 0.377 Cladding material Zr-4 Pellet density, g/cc 10.522 2
Maximum enrichment, wt% "U 5.05 i 0.05 Fuel Assembly Data Fuel rod array 15x15
{
Number of fuel rods 208 Fuel rod pitch, in. 0.568 N
' umber of guide tubes 16 Guide tube outside diameter, in. 0.530 Guide tube inside diameter, in. 0.498 ;
Instrument tube outside diameter, in. 0.493 Instntment tube inside diameter, in. 0.441 Active fuel length, in. 145 Holtec Report HI-981933 4-18 80284 SHADED IEGIONS DESIGNATE PROPRIETARY INFORMATION
r- -
I Table 4.2.1 -
Summary of the Criticality Safety Analyses Design Basis Burnup at 5.05 wt%2 "U 53.51 GWD/MTU Uncertainties Bias Uncertainty (95%/95%) i0.0011 Calculational Statistics' (95%/95%) t 0.0010 Depletion Uncertainty ' i 0.0176 Fuel Eccentricity negative Manufacturing Tolerances 1 0.0055 Statistical Combination of Uncertainties" t 0.0185 Reference kerr (MCNP4a) 0.8521 i
Total Uncertainty (above) 0.0185 Axial Burnup Distribution 0.0714 Calculational Bias (see Appendix A) 0.0009 Temperature Cotrection to 4 C (39 F) 0.0023 Maximum kerr 0.9452t n l Regulatory Limiting kerr 0.9500 l
l i
The value used for the MCNP4a (or KENO 5a) statistical uncertainty is 1.84 times the estimated standard deviation.
Each final k value calculated by MCNP4a (or KENO 5a) is the result of averaging a minimum of 200 cycle k values, and thus, is based on a minimum sample size of 200. The K multiplier, for a one-sided statistical tolerance with 95%
probability at the 95% confidence level, corresponding to a sample size of 200, is 1.84 [6].
" Square root of the sum of the squares.
Ui KENO 5a verification calculation resulted in a maximum ek no f 0.9456.
IIoltec Report 111-981933 4-19 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION r
Table 4.2.2 .
Reactivity Eficcts of Abnormal and Accident Conditions
= Abnormal / Accident Conditions Reactivity Effect Temperature Increase (above 4 C) Negative (Table 4.6.1)
Void (boiling) Negative (Table 4.6.1)
Assembly Drops Negligible or Negative Lateral Rack Movement Negative Misplacement or Mislocation of a Fresh Positive - controlled by less than 650 ppm Fuel Assembly soluble boron i
Holtec Report HI-981933 4-20 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION i
- l. i l
r Table 4.5.1 .
Reactivity Effects of Manufacturing Tolerances 1
Tolerance Reactivity Effect, Ak 2
Minimum Boral loading (E g/cm ,0.0324 g/cm 2nominal) i0.0026 '
Minimum Boral width (M", 7.5" nominal) i0.0008 Minimum Cell Pitch (E",9.22" nominal) 10.0011 Box wall thickness (Effj" max., E" min.; 0.075" nominal) Negative' l 1
2 2 Enriclunent (5.10 wt% "U,5.05 wt% "U nominal) i0.0030 i
2 Density tolerance (10.722 g/cm',10.522 g/cm nominal) 10.0036 f 1
Total (statistical sum)" i0.0055
\
I
Ihe nominal box wall dimension results in the highest reactivity.
" Square root of the stun of the squares.
IIoltec Report 111-981933 4-21 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION
4 1
i
~
Table 4.6.1 .
Reactivity Effects of Temperature and Void l
Temperature Reactivity Effect, Ak 4 C (39 F) reference 1
20 C (68 F) -0.0023 i
60 C(140 F) -0.0092 120 C(248 F) -0.0218 j 1
120 C w/10% void -0.0448 1
1 l
l l
l i
l l
l l-1 Holtec Report HI-981933 4-22 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
e-3 5 g g g ,
. e. . . ..
. ' !, j i ; (5.05,53.51)" '
50 - - - - - -
b ~ - - - - - - - A - - ~ - - - b, - - - - - - - l -. ~ - - - - - - b .- - - - - .
ICCENABLE DOMAIN..
s-----
f l- l (4'50' 43.29) 40 -- - -
9.- ---- -
p -- ---,-> - -- . s. . - - -- -;-
[ j ,(4.00,34.51) i ;
2 . 30 -
N- ---- -- " - ~ -- b -
~~- i- - b- - - (-
c l l . :
. . (3.50,25.17) :
4 4 g 20 - -~
-e -
l (3.00,18.80) '
P ,
l- 3
, (2.50,12.43) : UNACCEPTABLE DOMAIN 30 . - .
4 .:.
- l : :
i (2.00,4.92) ; : -: ,
i ' i i i i 0
2 2.5 3- 3.5 4 4.5 5 Initial Enrichmert(wt% U-235)
Figure 4.2.1 Minimum Required Fuel Assembly Burnup as a Function of Nominal
- Initial Enrichment to Permit Storage in the Cask Pit.
Note: Fuel assemblies with initial enrichments less than 2.0 wt%2 "U will conservatively be required to meet the burnup requirements of 2.0 wt%2"U assemblies.
H *' c Report HI-981933 4-23 80284
. SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
Boral Panel Box Wall Reflective Boundary Condition 000000000000000 0 0 0 0 0 0 0 0 0 0 0 0 0 00 l 0 0 0 0 0 @O 0 0 @ O 0 0 0 0 !
o o O @ o o o o o o o @ o oo
.: 0 0 0 0 0 0 0 0 0 0 0 0 0 00 o O@O O@o o O@o O@O O 1 1 Y 000000000000000 t l 1 0 0 0 0 0 0 000 0 0 0 0 0 0 1 j o00 0 00000 0 00 0 0 0 i !
l 3 O O @ o o @O o O @ o O @ O o 1
e o00000000000000 e o 0 0 @O 0 0 0 0 0 0 @O 0 0 O O O O O @ o O O @O O O O o o00 0 00000000 000 000000000000000 _
~
Reflective Boundary Condition ;
G = Guide Tube I = Instrument Tube Figure 4.3.1 A Two-Dimensional Representation of the Calculational Model Used for the !
Cask Pit Rack Analyses. !
lioltec Report HI-981933 4-24 80284 l SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
APPENDIX 4A: BENCHMARK CALCULATIONS ,
4 A.1 INTRODUCTION AND
SUMMARY
Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP4a [4A.1] is a continuous energy Monte Carlo code and KENO 5a (4A.2]
uses group-dependent cross sections. For the KENO 5a analyses reported here, the 238-group library was chosen, processed through the NITAWL-II [4A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errors' (trends) that have been reported (e.g., [4A.3 through 4A.5]) for calculations with collapsed cross section sets.
In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the ' B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.
Table 4A.1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.
One possible way of representing the data is through a q,ectrum index that incorporates all of the variations in parameters. KEN 05a computes and prints the " energy of the average lethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KEN 05a, the number of fissions in each group may be collected and the EALF determined (post-processing).
t Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices.
Holtec Repert HI-981933 Appendix 4A, Page 1
I I
Figures 4A.1 and 4A.2 show the calculated k,y for the benchmark critical experiments ,as a function of the EALF for MCNP4a and KENO 5a, respectively (UO2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental errort n performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals.
Linear regression analysis of the data in Figures 4A.1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KEN 05a). The total bias (systematic error, or mean of the deviation from a k,y of exactly 1.000) for the two methods of analysis are shown in the table below.
Calculational Bias of MCNP4a and KEN 05a MCNP4a 0.0009 0.0011 KENO 5a 0.0030 i 0.0012 The bias and standard error of the bias were derived directly from the calculated k,y values in Table 4A.1 using the following equations", with the standard error multiplied by the one-sided K-factor for 95% probability at the 95% confidence level from NBS Handbook 91 [4A.18] (for the number of cases analyzed, the K-factor is ~2.05 or slightly more than 2).
k=1 k, (4A.1) n s A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.
These equations may be found in any standard text on statistics, for example, reference
[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENO 5a.
Holtec Report HI-981933 Appendix 4A, Page 2
V k,* ' - ' (f ' k,)2f, s.:- s.: (4A.2)
I n (n-1)
/
~
Blas = (1- k ) e K og (4A.3)
- where ki are the calculated reactivities of n critical experiments;,oris the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias
. (mean)); K is the one-sided multiplier for 95% probability at the 95% confidence level .
(NBS Handbook 91 [4A.18]).
~
Formula 4.A.3 is based on the methodology of the National Bureau of Standards (now
- NIST) and is used to calculate the values presented on page 4.A-2. The first portion of the equation, ( l- E ), is the actual bias which is added to the MCNP4a and KENO 5a results.
The second term, Kog, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for -
one-sided statistical tolerance limits for 95 % probability at the 95% confidence level. The
. actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KEN 05a are 2.04 and 2.05, respectively.
~
The bias values are used to evaluate the maximum k,, values for the rack designs.
- KENO 5a has a slightlyi larger systematic error than MCNP4a, but both result in greater.
precision than published data [4A.3 through 4A.5] would indicate for collapsed cross section sets in KENO 5a (SCALE) calculations.
4A.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46 w/o
. to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 and
~4A.4 show the calculated k,y values (Table 4A.1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO 5a). Thus, there are no corrections to the bias for the various Holtec Report HI-981933 Appendix 4A, Page 3 l i g -f~i m j
')
enrichments. ,
As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENO 5a for various enrichments.
The cross-comparison of calculations with codes of comparable sophistication is suggested '
in Reg. Guide 331. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of k, for the two independent codes as evidenced by the 45 slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is E
considered confirmation of the absence of an enrichment effect (trend) in the bias.
. 4A 3
. ' Effect of % imding :
Several laboratories have performed critical experiments with a variety of thin absorber panels.similar to the Boral panels in the rack designs. ' Of these critical experiments, those '
performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment),
the reactivity worth of the absorbers in the PNL tests is very low and any signincant errors
' that might exist in the treatment of strong thin absorbers could not be revealed. t
- Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A.1) and shows the reactivity worth (Ak) of the absorber.t No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table 4A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.
I To further confirm the absence of a significant trend with ' B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO 5a (as suggested in Reg.
Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry.
These data substantiate the absence of any error (trend) in either of the two codes for the-conditions analyzed (data points fall on a 45 line, within an expected 95% probability
- limit).
' The reactivity' worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.
1 Holtec Report HI-981933 Appendix 4A, Page 4 3
p5 ,.
4A.4 ; Miscellaneous and Minnr Parameters 4A.4.1 Reflector Meterial and Snneings
.PNL has performed a nurmber of critical experiments with thick steel and lead reflectors.t
. Analysis of these critical e.<periments are listed in Table 4A.5 (subset of data in Table 4A.1). There appears to be a small tendency toward overprediction of k, at the lower
- spacing,'although there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward overprediction at close
- spacing means'that the rack ~ calculations may be slightly more conservative than otherwise.
4A.4.2 Fuel Pellet Diameter and I nuice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from
- 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch
. lattice' spacing) for BWR fuel. Thus, the critical experiments analyzed provide a reasonable representation of power reactor fuel. Based on the data in Table 4A.1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs.
l l
' 4A.4.3 ~ Sobible Boron Concentration Effects l 1
Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KEN 05a) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments (> 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative.
T A
Parallel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not pertinent to the Holtec rack design.
Holtec Report HI-981933 Appendix 4A, Page 5 4,.
' ' '-- --' ii _ _ . . . _,_.________.._._m__
4A.5 MOX Fuel ,
The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for UO2fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 4A.7. Results of these analyses are generally above a k,y of 1.00, indicating that when Pu is present, both MCNP4a and KENO 5a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KENO 5a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO 5a. It is also possible that the overprediction in k,y for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated k,y over a wide range of the spectral index (energy of the average lethargy causing fission).
l 4
Holtec Report HI-981933 Appendix 4A, Page 6
f p
.4A.6 References
_ [4A.1] .J.F. Briesmeister, Ed., "MCNP4a - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).
' [4A.2] SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200
-(ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995.
[4A'.3]' M.D. DeHart and S.M. Bowman, " Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use.in
, Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460)
Oak Ridge National Laboratory, September 1994.
[4A.4] W.C. Jordan et al., " Validation of KENOV.a", CSD/TM-238, Martin Marietta Energy Syste'ms, Inc., Oak Ridge National Laboratory, December 1986.
[4A.5] O.W. Hermann et al., " Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated.
[4A.6] R.J. Larsen and M.L. Marx, An Introduction to Mathematical Statistics and its Applications, Prentice-Hall,1986.
[4A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.
[4A.8] G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW-1645-4, Babcock & Wilcox Company, November 1991.
[4A.9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984.
~ Holtec Report HI-981933 Appendix 4A, Page 7 I
e 1
4
[4A.10] J.C. Manaranche et al., L" Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods," Trans.
Am. Nucl. Soc. 33: 362-364 (1979).
1
[4A.11] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 235 U Enriched UO2 Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981. l
[4A.12] S.R. Bierman et al., Criticality Experiments with'Suberitical
. Clusters of 2.35 w/o and 4.31 w/o 233 U Enriched UO2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December,1981.
[4A.13] S.R. Bierman et al., Critical Separation Between Suberitical Clusters of 4.31 w/o 233U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.
3
[4A.14] S.R.' Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.
2
[4A.15] B.M. Durst et al., Critical Experiments with 4.31 wt % 35U Enriched UO2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982. ,
t
[4A.16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803,' Battelle Pacific Northwest Laboratory, December 1981.
[4A.17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium Core, WCAP-3385-54,
' Westinghouse Electric Corp., Atomic Power Division, December 1%5.
[4A.18] M.G. Natrella, Exnerimental Statistics, National Bureau of Standards, Handbook 91, August 1%3.
l Holtec Report HI-981933 Appendix 4A, Page 8 i e
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0 3
3-3 3-3 3 3
3 3
3 3-3 3- aie xA R 8 8 8 3- t sF e . . t r
5 5 5 P P P P- P- P P o I
- i r i e A C
A C
A C
A C
A C
A C
A C C L Ah eoi s r s r rs a p
e N N N N ETeb R P P P W W W W W W W :
s c et " t e
3 4 5 6' 7 8 9 0 1 2 t o
l o
5 5 5 5 5 5 5 6 6 6 N H
o Table 4A.2 COMPARISON OF MCNP4a AND KEN 05a CALCULATED REACTIVITIESt FOR VARIOUS ENRICHMENTS.
Calculated k,y i la Enrichment MCNP4a KENO 5a 3.0 0.8465 0.0011 0.8478 0.0004 3.5 0.8820 i 0.0011 0.8841 0.0004 3.75 0.9019 i 0.0011 0.8987 0.0004 4.0- 0.9132 i 0.0010 0.9140 0.0004 4.2 0.9276 i 0.0011 0.9237 0.0004 4.5 0.9400 i 0.0011 0.9388 i 0.0004 1
i l
t Based on the GE 8x8R fuel assembly.
1Holtec Report HI-981933 . Appendix 4A, Page 14 L -<
n e
d Table 4A.3 MCNP4a CALCULATED REACTIVITIES FOR -
CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS '
l Ak MCNP4a Worth of Calculated EALF' Ref. Experknent : Absorber k, - (eV) - l 4A.13 - PNL-2615 Boral Sheet 0.0139 0.9994 i 0.0012 0.1165 4A.7 B&W-1484 Core XX . 0.0165 1.0008 i 0.0011- 0.1724 4A.13 J PNL-2615 :- 1.62% Boron-steel 0.0165 0.9996 i 0.0012 0.I161 4A.7 - B&W-1484 ' Core XIX 0.0202 0.9961 i 0.0012 0.2103 .
4A.7 B&W-1484 Cve XXI 0.0243 0.9994 0.0010- 0.1544 4 A.7 - B&W-1484 Core XVII 0.0519 0.9962 i 0.0012 0.2083 4A.I1 PNL-3602 Boral Sheet - 0.0708 0.9941 i 0.0011 0.3135 4 A~.7 ' B&W-1484 Core XV- 0.0786 0.9910 i 0.0011 0.2092 4A.7 ? B&W-1484- Core XVI 0.0845 0.9935 i 0.0010 0.1757 4A.7 B&W-1484 Core XIV 0.1575 0.9953 i 0.0011 0.2022 4 A.7 L B&W-1484 Core XIII - 0.1738 1.0020 i 0.0011 0.1988 4A.14, PNL-7167 Expt 214R flux trap 0.1931 0.9991 i 0.0011 0.3722
'EALF is the energy of the average lethargy causing fission.
- Holtec Report HI-981933. Appendix 4A, Page 15 E
p 1 4 L.
L -
k.-
. Table 4A.4 COMPARISON OF MCNP4a AND KENO 5a CALCULATED REACTIVITIES' FOR VARIOUS B LOADINGS 1
- Calculated k, 10
'"B, g/cm 2.. MCNP4a KENO 5a
! O.005 1.0381 0.0012 1.0340 i 0.0004 0.010 . 0.9960 0.0010 0.9941 i 0.0004 O.015 0.9727 0.0009 0.9713 0.0004-0.020 0.9541 0.0012 0.9560 0.0004 0.025 .0.9433 i 0.0011 0.9428 0.0004 l
' O.03 0.9325 0.0011 0.9338 0.0004 0.035 0.9234 i 0.0011 0.9251 0.0004 0.04 ~0.9173 0.0011 0.9179 i 0.0004 1
l-L !
I L
l-i i
l
. Based on a 4.5% enriched GE 8x8R fuel assembly.
I%
L Holtec Report HI-981933 Appendix 4A, Page 16 l L'
p i.
L l
\
Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORSt i Separation, Ref. Case E, wt% em MCNP4a k,, KENO 5a k,y 4A.I1 Steel 2.35 1.321 0.9980 i 0.0009 0.9992 i 0.0006 I Reflector
- 2.35 - 2.616 0.9968 i 0.0009 0.9964 1 0.0006 !
i 2.35 3.912 0.9974 i 0.0010 0.9980 i 0.0006 2.35 = 0.9962 i 0.0008 0.9939 i 0.0006 1 I
J I
4A.I1 Steel 4.306 1.321 0.9997 i 0.0010 1.0012 i 0.0007 Reflector 4.306 2.616 0.9994 i 0.0012 - 0.9974 t0.0007 <
l 4.306 3.405 0.9969 i 0.0011 0.9951 i 0.0007
{ 4.306 = 0.9910 1 0.0020 0.9947 i 0.0007 l
l 4A.12 Lead 4.306 0.55 1.0025 0.0011 0.9997 i 0.0007 Reflector 4.306 1.956- 1.0000 i 0.0012 0.9985 i 0.0007 4.306 5.405 0.9971 i 0.0012 0.9946 i 0.0007 i
i i
l
- t Arranged in order ofincreasing reflector-fuel spacing.
Holtec Report HI-981933 Appendix 4A, Page 17
Table 4A.6 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated k, Boron Concentration,
- Reference Experiment ppm MCNP4a KENO 5a.
4A.15 PNL-4267 0 0.9974 0.0012 -
4A.8 B&W-1645 -886 0.9970 0.0010 0.9924 0.0006 4A.9 B&W-1810 1337 1.0023 0.0010 -
4A.9 B&W-1810 1899 1.0060 0.0009 -
4A.15 PNL-4267 2550 1.0057 0.0010 -
i
)
i l
l Holtec Report HI-981933 Appendix 4A, Page 18 C
y Y.
l l
l Table 4A.7
' CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENO 5a-Reference ' Case' ' k,, 'EALF" k,, EALF"-
PNI 5803 MOX Fuel - Exp. No. 21 1.004I i 0.0011 0.9171 1.0046 1 0.0006 0.8868
[4A.16]
MOX Fuel - Exp. No. 43 1.0058 1 0.0012 0.2%8 1.00?6 1 0.0006 0.2944 MOX Fuel - Exp. No.13 1.0083 1 0.0011 0.1665 0.9989 1 0.0006 0.1706 MOX Fuel - Exp. No. 32 . 1.0079 i 0.0011 0.1139 0.9966 1 0.0006 -- 0.1165 WCAP_ Saxton @ 0.52" pitch 0.9996 1 0.0011 0.8665 1.0005 i 0.0006 0.8417 3385 54
[4A.17] Saxton @ 0.56" pitch 1.0036 1 0.0011 0.5289 1.0047 i 0.0006 0.5197 Saxton @ 0.56" pitch borated 1.0008 1 0.0010 0.6389 NC NC Saston @ 0.79" pitch 1.0063 i 0.0011 0.1520 1.0133 i 0.0006 0.1555 l
Note: NC stands for not calculated L t 1 Arranged in order ofincreasing lattice spacing.
EALF is the energy of the average lethargy causing fission.
Holtec Report HI-981933 Appendix 4A, Page 19 l
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0.84 ii , iii, iii, iii, iii, iiii iiii iiii iiii iiii iiii 0.84 0.86 0.88 0.90 0.92 0.94 i MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KEN 05A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS Report Hi-981933
c 1.04 1.03 _ inns g/,.m.q g 1.02 ~~
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l 0.900 0.920 0.940 0.960 0.980 1.000 1.020 1.0 Reactivity Calculated with KEN 05a l
FIGURE .4A.6 COMPARISON OF MCNP AND KEN 05a CALCULATIONS FOR VARIOUS BORON-10 AREAL DENSITIES l
l Report HI-981933 l
' 5.0 - THERMAL-HYDRhULIC CONSIDERATIONS 5.1 Introduction This chapter provides a summary of the methods, models, analyses, and numerical results of the
. thermal hydraulic evaluations performed for installation of fuel storage racks in the Cask Pit.
These evaluations demonstrate compliance to the provisions of Section IIIof the USNRC "OT Position Paper for Review and Acceptance o[ Spent Fuel Storage and Handling Applications,"
' dated April 14,1978. Evaluations were performed for the Spent Fuel Pool Cooling System .
(SFPCS), Decay Heat Removal System (DHRS), Cask Pit and SFP.
The DBNPS is requesting approval to place four high density fuel racks in the DBNPS Cask Pit.
. These four racks will add a total of 289 fuel storage spaces. The racks can be installed in three phases as described in Chapter 1.of this report. The installation of additional racks will re-establish full core offload capability, allow defueling of the reactor in April of the year 2000 for
- the required 10 year in-service inspection of the reactor vessel, and facilitate the planned
~
complete re-racking of the SFP. The re-racking will increase the total fuel storage capacity of the SFP to approximately 1650 fuel assemblies.
There is no direct, forced cooling of the Cask Pit. The heat produced by the fuel stored in the Cask Pit will be transferred to the SFP by an exchange of water through the open gate, which
{
connects the two bodies. Therefore, the Cask Pit gate must be open at all times that fuel !
assemblies are stored in the Cask Pit. The SFP heat removal systems are shown to have adequate f
capacity to remove the additional heat load of the fuel placed in the Cask Pit.
The thermal hydraulic qualification analyses for the Cask Pit racks were performed to show that .
fuel stored in the Cask Pit will be adequately coole 1 and the pit structure temperature will be
- appropriately limited. The analyses can be further described as follows:
Holtec Report HI-9.81933 . 5-1 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION 1
l q
L: ;
' discharge schedule was performed. This analysis was performed to establish that maximum bulk coolant temperature limits and Cask Pit structural temperature limits would not be exceeded. To account for the future re-racking of the SFP, the bulk temperature analysis was performed using a conservative storage capacity of approximately 1,650 fuel assemblies in the 5FP. Since the current capacity of the SFP is 735 spaces, the selection of 1,650 spaces will bound the addition of 289 spaces to the Cask Pit. The selection of 1,650 spaces will also bound the final SFP storage capacity
' because the 4 Cask Pit racks will be relocated to the SFP near the end of the future SFP re-racking.
ii. An evaluation of loss-of-forced cooling scenarios in the Spent Fuel Pool was completed to establish the minimum time to peform corrective actions to prevent boiling and maximum makeup water requirements. This analysis was also performed based on the assumption that approximately 1650 fuel assemblies were stored in the SFP. The time-to-boil and water makeup requirement analyses are conservative for the addition of 289 storage spaces to the current SFP capacity. 1 iii. An evaluation of the temperature gradient between the Cask Pit and the SFP was completed for an assumed Cask Pit heat load. This analysis verified that the Cask Pit will I be sufficiently cooled by the natural circulation driven exchange of water between the two bodies, such that the resulting water temperature will not exceed the maximum temperature limit for the Cask Pit structure. As a limiting case, the maximum bulk temperature of the SFP was used as the starting temperature for this evaluation.
Holtec Report HI-981933 5-2 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION u
C; e
iv. The maximum fuel rod cladding temperature for fuel stored in the Cask Pit was ,
determined to establish that departure from nucleate boiling at any location around the fuel is not possible. This evaluation was based conservatively on the maximum water temperature in the Cask Pit and the water-to-clad temperatum difference for the hottest fuel assembly in the SFP.
The following sections present plant system descriptions, analysis assumptions, a synopsis of the analysis methods employed, and the final results.
5.2 Cooline Systems Description A complete description of the SFPCS is found in the DBNPS USAR, Section 9.1.3. The SFPCS is designed to remove decay heat from the fuel stored in the SFP. The SFPCS at the DBNPS consists of two half-capacity recirculating pumps, two half-capacity heat exchangers, the i associated valves and piping. The SFPCS pumps are horizontal, centrifugal units with a rated capacity of 1,100 gpm. The SFPCS heat exchangers are shell and tube units. The cold cooling water shell side flow is supplied from the plant Component Cooling Water (CCW) system. SFP water is pumped through the heat exchange tube side. The SFPCS heat exchanger design performance is listed below.
6 Heat Transferred: 5.25x10 Btu /hr Shell Side Flow Rate: 650 gpm Shell Side Inlet Temperature: 95 F Shell Side Outlet Temperature: 111.2 F Tube Side Flow Rate: 1000 gpm Tube Side Inlet Temperature: 120 F Tube Side Outlet Temperature: 109.5 F Holtec Report HI-981933 5-3 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION
m ]
l f
The DHRS, described in the DBNPS USAR Section 9.3.5, serves as the Seismic Class I backup cooling system to the SFPCS. The DHRS consists of two recirculating pumps and two heat exchangers. The DHRS is permanently connected to the SFPCS via a 10 inch line. _Two
{
i normally closed gate valves provide isolation between the DHRS and the SFPCS. The DHRS pumps are single-stage, centrifugal units with a rated capacity of 3,000 gpm. The DHRS heat exchangers, which are also cooled by the CCW system, are shell and tube units with the following design performance:
6 Heat Transferred: 26.9x10 Btu /hr SFP Water Flow Rate: 3000 gpm SFP WaterInlet Temperature: 140 F CCW Flow Rate: 6000 gpm CCW Inlet Temperature: 95 F Loss of water from the SFP is unlikely since the SFP and piping within the SFP are Seismic Class I. Makeup water is readily available. The DHRS is permanently connected to the Class I boundary of the SFP. This system can provide borated make up water to the SFP from the Borated Water Storage Tank. SFP makeup water is also available from the Seismic Class II Demineralized Water Storage Tank or Clean Waste Receiver Tank.
Holtec Report HI-981933 5-4 80284 SIIADED REGIONS DESIGNATE PROPRIETARY INFORM ATION 1
l r
r i
1 5.3 Discharce/Cooline Alignment Scenarios l
A total of six reactor core discharge / cooling scenarios were postulated. These scenarios are:
i Scenario Discharge Type Cooling System Alignment i I Partial Core 2 SFPCS Pumps and Heat Exchangers 2 Partial Core 1 SFPCS Pump and Heat Exchanger 3A Type A Full Core 2 SFPCS Pumps and Heat Exchangers (65 days at power) j 3B. Type B Full Core l 2 SFPCS Pumps and Heat Exchangers I (2 years at power) 4A Type A Full Core 1 DHRS Train (65 days at power) 4B Type B Full Core 1 DHRS Train (2 years at power) i The DBNPS does not routinely perform a full core discharge at each refueling outage. Scenarios '
2,3A an'b 3B correspond to discharge type and cooling alignment combinations which are not typically used during fuel discharge operations. These scenarios are included to demonstrate that l
the bulk temperature will remain below boiling even under extreme circumstances. Time-to-boil, boiloff rate, and local temperature analyses are performed for the most limiting (i.e., highest bulk temperature and decay heat flux) of the full core discharge Scenarios 4A and 4B. !
Thermal hydraulic analyses were performed to conservatively account for the future re-racking of the SFP. The re-racked SFP will have a total capacity of approximately 1,650 spaces. Since the current SFP capacity is 735 spaces, the analyses performed will bound the addition of 289 spaces to the Cask Pit.
l Hohec Report HI-981933 5-5 80284 S11ADED REGIONS DESIGNATE PROPRIETARY INFORMATION '
l l
A partial core discharge is comprised of 72 assemblies discharged into the SFP, which aire y contains 1609 previously discharged assemblies. This analyzed stored fuel inventory (1681) conservatively exceeds the maximum possible inventory. The minimum decay time of the previously discharged fuel assemblies for these scenarios is 2 ' years.
A " Type A" full core discharge is comprised of 177 assemblies discharged into the SFP, which already contains 1537 previously discharged assemblies. This analyzed fuel inventory (1714)-
conservatively exceeds the maximum possible inventory. This full core discharge takes place after 65 days of full power operation since the last partial core discharge, The minimum decay time of the previously discharged fuel assemblies for these scenarios is 65 days.
A " Type B" full core discharge is comprised of 177 assemblies discharged into an SFP that aheady contains 1537 previously discharged assemblies. This analyzed fuel inventory (1714)
-conservatively exceeds the maximum possible inventory. This full core discharge takes place after 2 years of full power operation since the last partial core discharge. The minimum decay -
time of the previously discharged fuel assemblics for these scenarios is 2 years.
~
Table 5.3.1 presents the historic and projected discharge schedule used for these analyses.
In all scenarios, the cooling water which removes heat from the SFPCS and DHRS heat exchangers is assumed to be at its design maximum temperature and design basis flow rate.
I 1
Holtec Report HI-981933, 5-6 80284
( SIIADED REGIONS DESIGNATE PROPRIETARY INFORMATION
p 5.4 Maximum Bulk Pool Temperature Methodoloey This section presents the methodology for calculating the maximum SFP bulk temperatures for the scenarios presented in the preceding section. The maximum SFP bulk temperature will be used as the inlet temperature to the Cask Pit. The following conservatisms are applied in the maximum pool bulk temperature calculations:
The decay heat load is based on a discharge schedule with bounding projected fuel parameters.
The minimum initial enrichment for projected discharged batches is used for previously discharged fuel decay heat calculations.
The thermal capacity of the SFP is based on the net SFP water volume only. The considerable energy storage capability of the fuel racks, fuel assemblies, and pool structure is neglected.
The cooling effects of evaporation heat losses and all other passive heat removal mechanisms (i.e., conduction through walls and slab) are neglected.
The SFP and Cask Pit are treated as a " lumped" system with a single bulk temperature, however no credit is taken for the thermal capacity of the Cask Pit. This maximizes the applied decay heat load and minimizes the thermal energy storage.
Holtec Report HI-981933 5-7 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
v .
The transient thermal response of the SFP and the attendant cooling systems is governed by a -
first-order, ordinary differential equation.: The governing differential equation can be written by utilizing conservation of energy as:
Cdr.dT. = Q(r) - Qa(T) - Qu(T) where:
C = Pool thermal capacity, Btu / F T = Pool bulk temperature, F -
T = Time after reactor shutdown, hr Q(t) = Time varying decay heat generation rate, Btu /hr Qax(T) = Temperature dependent SFPCS or DHRS heat rejection rate, Btu /hr Qsv (T) = Temperature dependent passive heat losses to the surrounding environment, Btu /hr Qax(T) is a function of the SFP temperature and the cooling water flow rate and temperature can be written in terms of the temperature effectiveness (p) as follows:
. Qu(T) = W, C, p (T - to) where:
W i = CCW water flow rate,Ib/hr Ci = CCW water specific heat capacity, Btu /(lbx F) p = SFPCS or DHRS heat exchanger temperature effectiveness T = Bulk pool water temperature,"F ti = CCW water inlet temperature, F Holtec Report HI-981933 5-8 80284 SIIADED REGIONS DESIGNATE PROPRIETARY INFORMATION i
k
. I b.
The temperature effectiveness, a measure of the heat transfer efficiency of the SFPCS or DHRS heat exchangers,is defined as:
t , - to .
P=
T - to 4
where to is the CCW outlet temperatur: (*F) and all other terms are as defm' ed above.
. Qsv(T)is a nonlinear function of the pool temperature and ambient temperature. This term is conservatively neglected in the maximum pool bulk temperature calculations. However, a
- discussion of this term is provided for understanding of the conservatism applied tc, this calculation. Qsv contains the passive heat sosses from the pool surface which includes
- evaporation, natural convection and thermal radiation from the pool surface, and heat conduction
- through the pool walls and slab. Experiments show that heat conduction through the pool walls I and slab takes only about 4% of the total heat loss and is conservatively neglected [5.4.1]. The temperature dependent passive heat loss can be expressed as [5.4.2]:
Og(T) = h A (T-1,) + e GA (T'-tl) + aA (P -P,)
where:
h = Natural convection heat transfer coefficient, Btu /(hrxft2 x F)
A = Pool surface area, ft 2 t.= Ambient SFP building temperature, F e = Emissivity of pool water o = Stephan-Boltzmann constant 2
(x = Evaporation rate constant, Btu /(hrxft xp3;)
= P. = Vapor pressure of water at pool temperature, psi
- P, = Vapor pressure of water at ambient temperature, psi Holtec Report HI-981933 5-9 80284
. SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION 4
J
~
g I:
' The differential equation that defines the transient thermal msponse of the pool is solved f numerically. The decay heat load from previously discharged fuel assemblies is calculated using
~ Holtec's QA validated LONGOR program [5.4.3]. This program incorporates the ORIGEN2 isotope generation and depletion code [5.4.4) to perform the decay heat calculations. The
. transient decay heat loads and pool bulk temperatures are calculated using Holtec's QA validated BULKTEM program [5.4.5), which also incorporates the ORIGEN2 code. The maximum SFP bulk temperature is extracted from the results of the transient evaluations. The major input -
values for this analysis are summarized in Table 5.4.1.
5.5 Minimum Time-to-Boil and Maximum Boiloff Rate Methodolorv This section presents the methodology for calculating the minimum time-to-boil and '
corresponding maximum boiloff rate for the scenarios presented in Section 5.3. This analysis is conducted for the number of fuel assemblies in a re-racked SFP and therefore will bound the addition of 289 spaces in the Cask Pit.
The following conservatisms are applied in the SFP time-to-boil and boiloff rate calculations:
- The SFP bulk temperature and decay heat generation rates are assumed to be the calculated maximum bulk temperature and the coincident decay heat generation rates.
Maximizing the initial temperature and utilizing the coincident decay heat generation rates will conservatively minimize the time-to-boil.
I The thermal capacity of the SFP is based on the net water volume only. The considerable energy storage capability of the fuel racks, fuel assemblies, and pool stmeture is
' neglected.
Holtec Report HI-981933 5-10 80284 SHADED h 5sGIONS DESIGNATE PROPRIETARY INFORMATION
n p'
Heat losses through the pool walls and slab are neglected. ,
In calculating the spent fuel pool evaporation heat losses, the building housing the spent fuel pool is assumed to have a conservative ambient air temperature of 110 F and 100%
relative humidity. These conditions yield a conservative time-reducing pool thermal capacity while minimizing the credit for evaporative and other passive heat losses.
The SFP and Cask Pit are treated as a " lumped" system with a single bulk temperature, however no credit is taken for the thermal capacity of the Cask Pit. This maximizes the applied decay heat load and minimizes the thermal energy storage.
i y x
The governing enthalpy balance equation for this condition, subject to these conservative t
l assumptions, can be written as:
1 l
l ,
C(r)dT = Q(r + ro )- Gy(T)
- di L : where C(t) is the time-reducing thermal capacity, T is the time after cooling is lost (hr) and to is the loss of cooling time after shutdown (hr). The other terms of this equation are defm' ed in Section 5.4, including a discussion of Qsv(T). Temperature dependent passive heat losses from the pool surface are accounted for in this analysis.
This differential equation is solved using a numerical solution technique to obtain the bulk SFP temperature as a function of time. This analysis is performed using Holtec's QA validated TBOIL program [5.5.1]. This program utilizes the highly conservative correlations of ASB 9r2
[5.5.2] to perform the decay heat calculations, thereby imparting even more conservatism to the results. The major input values for this analysis are summarized in Table 5.5.1.
. Holtec Report HI-981933 5-11 80284 SIIADED REGIONS DESIGNATE PROPRIETARY INFORMATION l'
i
- l. v Ll
w ,
)t l9 s/
p ;5.6 local Water Teingr&ure Methodolorv y
- This: section summarizes the methodolohy for evaluating the' maximum local water for the SFP and Cask Pit. A conservative evaluation for a bounding amalgam of conditions is performed.
The result of this evaluation is a bounding temperature difference between the maximum lodal water temperature and the bulk pool temperature. The maximum temperature -
difference is added to'the maximum' bulk SFP temperature to determine the maximum loc'a l temperature in the SFP. The maximum SFP local temperatum is determined to ensure the SFPCS and DHRS heat removal capacity is acceptable to remove the additional heat of the fuel Stored in'the Cask Pit. The maximum Cask Pit local temperature is compared to the maximum SFP bulk temperature to ensure the Cask Pit temperature is not excessive and will demonstrate
- the exchange of water with the SFP.
In order to determine thy - m mum local water temperature, a series of conservative assumptions
- are made. The most important of these assumptions are:
. With a full core discharged into the SFP racks, approximately equidistant from the coolant water inlet and outlet, the remaining cells in the spent fuel pool are postulated to be occupied with previously discharged fuel.
The hottest assemblies, located together in the pool, are assumed to be located in pedestal cells of the racks. These cells have a reduced water entrance area, caused by the pedestal blocking the baseplate hole, and a correspondingly increased hydraulic resistance.
, No downcomer flow is assumed to exist between the rack modules.
All rack cells are conservatively assumed to be 50% blocked at the cell outlet to account for drop accidents resulting in damage to the upper end of the cells.
Holtec Report HI-981933 5-12 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
E The hydraulic resistance parameters for the rack cells, permeability and inertial resistance, are worsened by 15% and 25%, respectively.
.* L For evaluating maximum local water and fuel cladding temperatures, the SFP is modeled as separate from the Cask Pit. The fuel decay heat that would normally be in the Cask Pit is assumed to be in the SFP.
For evaluating the exchange of water between the Cask Pit and the SFP, a separate model of the Cask Pit and a large, pseudo-constant temperature fluid reservoiris used. While the depth of the reservoir is taken as the SFP depth, the other details of the reservoir are selected to minimize reservoir temperature gradients.
For evaluating the exchange of water between the Cask Pit and the SFP, the decay heat generation rate in the Cask Pit racks is calculated based on the maximum heat generation rate from the fuel discharge schedule of Table 5.3.1 (1,609 fuel assemblies). The resultant heat generation rate is 1,404,009 watts. This yields a total Cask Pit heat generation rate of 252,200 watts or 860,759 Blu/hr for 289 fuel assemblies.
5.6.1 Local Temperature Evaluation Methodology The inlet piping that returns cooled water from the SFPCS terminates above the level of the fuel racks. It is not apparent from heuristic reasoning alone that the cooled water delivered to the SFP would not bypass the hot fuel racks and exit through the outlet piping. To demonstrate adequate cooling of hot fuel in the SFP, it is therefore necessary to rigorously quantify the velocity field in
- the pool created by the interaction of buoyancy driven flows and water injection / egress. A Computational Fluid Dynamics (CFD) analysis for this demonstration is required. The objective of this study is to demonstrate that the principal thermal-hydraulic criterion of ensuring local subcooled conditions in the SFP is met for all postulated fuel discharge / cooling alignment Holtec Report HI-981933 .
5-13 80284 'l SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION 1
1
L t scenarios. The local thermal-hydraulic analysis is performed such that partial cell blockage and slight fuel assembly variations are bounded. An outline of the CFD approach is described in the following.
J There are several significant geometric and thermal-hydraulic features of the DBNPS SFP which need to be considered for a rigorous CFD analysis. From a fluid flow modeling standpoint, there l- are two regions to be considered. One region is the bulk SFP/ Cask Pit region whem the classical Navier-Stokes equations are solved with turbulence effects included. The other region is the heat generating fuel assemblies located in the spent fuel racks located near the botto'm of the SFP. In this region, water flow is directed vertically upwards due to buoyancy forces through relatively small flow channels formed by the Babcock and Wilcox (B&W) 15x15 fuel assembly rod arrays in each rack cell. This situation shall be modeled as a porous solid region in which the classical -
Darcy's Law, given below, governs fluid flow:
=-
8 Xi K(i) Vi - C p IVl 2 3 where BP/BXi is the pressure gradient, K(i), Vi and C are the corresponding permeability, velocity and inertial resistance parameters and is the fluid viscosity. The permeability and inertial resistance parameters for the rack cells loaded with B&W 15x15 fuel were determined based on friction factor correlations for the laminar flow conditions typically encountered due to the low buoyancy induced velocities and the small size of the flow channels, i
{
The DBNPS SFP geometry required an adequate portrayal oflarge scale and small scale features, {
spatially distributed heat sources in the spent fuel racks, and water inlet / outlet configuration.
Relatively cooler bulk pool water normally flows down between the fuel rack outline and pool wall liner clearance known as the downcomer. Near the bottom of the racks, the flow turns from a vertical to horizontal direction into the bottom plenum supplying cooling water to the rack l
cells. Heated waterissuing out of the top of the racks mixes with the bulk pool water. An
- Holtec Report HI-981933 5-14 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION L ;
~:
adequate modeling of these features on the CFD program involves meshing the large scale bulk
_ pool region and small scale downcomer and bottom plenum regions with sufficient number of computational cells to' capture the bulk and local features of the flow field.
. ..Two distinct CFD models havd been developed for the DBNPS SFP. The first model addresses
- the local thermal hydraulic acceptability of storing hot, recently discharged fuel assemblies in the -
SFP, which is supplied with forced cooling. The second model addmsses the adequacy of the cooling of the low heat generation rate previously discharged fuel assemblies to be stored in the Cask Pit. A synopsis of both models is provided in the following.
4 .. .
The distributed heat sources in the SFP racks are modeled by identifying distinct heat generation zones considering full-core discharge, bounding peak effects, and presence of background decay heat from previous discharges. Three heat generating zones were modeled. The first zone contains the heat generated by fuel from previous discharges and the second and third zones contain the decay heat generated by fuel from a bounding full-core-discharge scenario. The two full core discharge zones are differentiated by one zone with higher than average decay heat generation and the other with less than average decay heat generationc This is a conservative model, since all of the fuel with higher than average decay heat is placed in a contiguous area. - A uniformly distributed heat generation rate was applied throughout each distinct zone.
In the Cask Pit water exchange model, the entire fuel storage rack region in the Cask Pit is modeled as containing decay heat from previous discharges. A uniform volumetric decay heat generation rate is applied to the fuel racks region. A pseudo-constant temperature reservoir representing the SFP bulk temperature is included in the model.
. The CFD analysis was performed on the FLUENT [5.6.4] fluid flow and heat transfer modeling program. The FLUENT code enables buoyancy flow and turbulence effects to be included in the CFD analysis. Turbulence effects are modeled by relating time-varying Reynolds' Stresses to the mean bulk flow quantities with the k-e turbulence model. The k-e model is appropriate for the Holtec Report HI-981933 5-15 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION 1
L
~
E -
o 1 i
DBNPS CFD analysis. The k-c turbulence model is a time-tested, general-purpose turbulence model. This model has been demonstrated to give good results for the majority of turbulent fluid '
flow phenomena. I Rigorous modeling of fluid flow problems requires a solution to the classical Navier-Stokes equations of fluid motion [5.6.1]. The governing equations (in modified form for turbulent flows with buoyancy effects included) are written as:
~
B p. us + d p. {u's u'i) = Bp '&u,+&ui ~
dt d x, O x, d x, B x, > .
_ r dp d p.{u'au'ih
-( p - p,, ) g, +
- dx dxj l
where ui are the three time-averaged velocity components. p (u'i u'j) are time-averaged Reynolds stresses derived from the turbulence induced fluctuating velocity components u'i, p, is the fluid density at temperature To, is the fluid viscosity, gi are the components of gravitational acceleration and x; are the Cartesian coordinate directions. The Reynolds stress j tensor is expressed in terms of the mean flow quantities by defining a turbulent viscosity F, and a turbulent velocity scale ku2 as shown below [5.6.2]:
duj_
p (u'iu'j) = 2 /3p k 63 - #, .dui >
i
_ d xj + d xi.
The procedure to obtain the turbulent viscosity and velocity length scales involves a solution of two additional transport equations for kinetic energy (k) and rate of energy dissipation (c). This methodology, known as the k-c model for turbulent flows, is described by Launder and Spalding
[5.6.3]. !
l Holtec Report HI-981933 5-16 80284 SHADED REGIONS DESIGbiATE PROPRIETARY INFORMATION I
l l
)
g Some of the major input values for this analysis are summarized in Table 5.6.1. Views of th,e assembled CFD models for the SFP and the Cask Pit are presented in Figures 5.6.1 and 5.6.2.
Figures 5.6.3 and 5.6.4 present temperature contours and velocity vectors, respectively, in the SFP model. Figures 5.6.5 and 5.6.6 present temperature contours and velocity vectors, respectively,in the Cask Pit model.
5.7 Fuel Rod Cladding Temocrature Methodology This section summarizes the method to calculate the temperature of the fuel rod cladding.
Similar to the local water temperature calculation methodology presented in the preceding section, this evaluation is performed for a single, bounding scenario. The maximum temperature difference between the fuel cladding and the local water temperature is calculated for the hottest fuel assembly in the SFP. This temperature difference is used to conservatively show that the cooling systems can acceptably remove from the SFP the heat generated by 289 additional fuel assemblies in the Cask Pit.
Y The maximum specific power of a fuel assembly (qA) Can be given by:
ga = q F,,
where:
Fxy = Radial peaking factor q = Average fuel assembly specific power, Btu /hr s e The peaking factors are given in Table 5.6.1. The maximum temperature rise of pool water is computed for the most disadvantageously located fuel assembly, described in the assumptions to Section 5.6 as the one which is subject to the highest local pool water temperature. Having determined the maximum local water temperature in the pool, it is possible to determine the maximum fuel cladding tempcrature. A fuel rod can produce F, times the average heat emission rate over a small length, where F, is the axial rod peaking factor. The axial heat distribution in a
- Holtec Report HI-981933 5-17 80284 :
SIIADED REGIONS DESIGNATE PROPRIETARY INFORMATION !
l u_ __ _ _ __ --
U
rod is generally a maximum in the central region, and tapers off at its two extremities. Thus,
_ peak cladding heat flux over an infinitesimal area is given by the equation:
q, = q F,y F.
A, where Ac is the total cladding extemal heat transfer area in the active fuel length region.
Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially upwards from bottom to top under laminar flow conditions. Rohsenow and Hartnett
[5.7.1] report a Nusselt-number based heat transfer correlation for laminar flow in a heated channel.LThe film temperature driving force (ATr) at the peak cladding flux location is calculated
- as follows:
D, hr = Nu Kw .
ATr= i he where, hr is the water side film heat transfer coefficient, Dn is sub-channel hydraulic diameter, Kw is water thermal conductivity and Nu is the Nusselt number for laminar flow heat transfer.
In order to introduce some additional conservatism in the analysis, we assume that the fuel '
cladding has a crud deposit resistance Re (equal to 0.0005 ft2 -hr- F/ Btu) that covers the entire surface. Thus, including the temperature drop across the crud resistance, the cladding to water local temperature difference (ATc) is given by:
AT, = ATr + Rc q,
- Holtec Report HI-981933 5-18 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
r 5.8 - Results
- This section contains results from the analyses performed for the postulated discharge scenarios.
5.8.1 Maximum Bulk Pool Temocratures For the discharge / cooling scenarios postulated in Section 5.3, the maximum calculated pool bulk
. temperatures are summarized in Table 5.8.1.' The worst case decay heat load in the SFP for the 6
full' core discharge scenario 4A was determined to be 30.15x10 btu /hr. For Scenarios 1,4A, and 4B, SFP bulk .emperatures must remain within the limits of the American Concrete Institute (ACI) Code Reqairements for Nuclear Safety Related Concrete Structures ACI-349, to protect -
the integrity of the SFP structure. The ACI Code permits long-term temperatures of up to 150 F and short-term temperature excursions in localized areas (e.g., skin effects) up to 350 F. As discussed in Section 5.3, Scenarios 2,3A and 3B are considered accident conditions and are only compared to the bulk boiling temperature of 212 F.
The results presented in Table 5.8.1 demonstrate that calculated bulk temperatures for the first
- four scenarios listed remain below their respective allowable limits. The calculated peak bulk temperatures for Scenarios 4A and 4B exceed the 150 F concrete temperature limit for long term normal operating conditions by less than 1.5'F. In both scenarios, the bulk pool temperature will remain above 150 F for less than 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. The effect of this bulk SFP temperature condition is evaluated and determined to be acceptable in the structural evaluations in Section 8. Given the i conservatisms incorporated into the calculations, actual SFP bulk temperatures will be lower than '
the calculated values reported in Table 5.8.1.
i Holtec Report HI-981933 . ,
5-19 -80284 SIIADED REGIONS DESIGNATE PROPRIETARY INFORMATION
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C ]
5.8.2 Minimum Time-to-Boil and Maximum Boiloff Rate For discharge / cooling Scenarios 1 and 4A, the calculated time-to-boil and maximum boiloff rates are summarized in Table 5.8.2. These results show that, in the extremely unlikely event of a complete failure of both the SFPCS and DHRS, there would be at least 3.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> available for corrective actions. The maximum water boiloff rate is less than 70 gpm.
5.8.3 Imcal Water and Fuel Claddina Temperatures The CFD study has analyzed a bounding local thermal-hydraulic scenario. In this scenario, a bounding full-core discharge is considered in which the 177 assemblies are located in the pool, approximately equidistant from the water inlet and outlet, while the balance of the rack cells are postulated to be occupied by fuel from previous discharges. In this analysis, the difference between the peak local temperature and the coincident bulk pool temperature was conservatively calculated to be 42.75 F.
I The peak fuel cladding superheat is determined for the hottest cell location in the pool as obtained from the CFD model for the DBNPS pool. The maximum temperature difference l
between the fuel cladding and the local water (ATc)is calculated to be 36.1 F. This calculated cladding ATc is applied, along with the maximum temperature difference between the local water l 1
temperature and the bulk SFP temperature, to the calculated maximum SFP bulk temperature (Scenarios 4A and 4B) of approximately 151.5 F. This yields a conservatively bounding I 194.25 F maximum local water temperature and a conservatively bounding 230.35 F peak cladding temperature. These conservative bounding maximum local temperatures are less than the 239 F local boiling temperature on top of the racks. Thus, boiling does not occur anywhere within the DBNPS SFP. Based on these results, the SFPCS and DHRS will acceptably remove the heat generated from fuel placed in the Cask Pit.
Holtec Report HI-981933 5-20 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION L.
y The evaluation of the buoyancy driven, natural convection water exchange between the Cask Pit
- and the SFP, which was modeled as a pseudo-constant temperature reservoir, yields a maximum temperature difference of 4*F. Note that this is the difference between the Cask Pit maximum
. local temperature and the SFP bulk temperature. The maximum temperature of the water in the V
Cask Pit, based on the calculated maximum SFP bulk temperature of approximately 151.5 F, would themfore be 155.5 F. This is well below the local saturation temperatum at the top of the Cask Pit racks which, due to the greater depth of the Cask Pit, is even greater than the 239*F .
saturation temperature at the top of the SFP racks.
The maximum Cask Pit water temperature of 155.5 F stated above is based on the maximum local temperature at the top of the Cask Pit racks. The Cask Pit bulk water temperatum would be -
l approximately 154.5 F. As stated above, the ACI code permits long-term temperatures of up to 150*F and short-term temperature excursions in localized areas up to 350 F. Based on the SFP bulk temperature analyses for scenarios 4A and 4B, the long-term limit would be exceeded by less than 4.5*F for approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The effect of this bulk Cask Pit temperature condition is evaluated as acceptable to the Cask Pit structure in Chapter 8. Given the conservatisms assumed in the thermal hydraulic calculations, the actual Cask Pit bulk temperatums will be less than the calculated value of 154.5 F.
Due to the low heat generation rate of the background fuel stored in the Cask Pit, fuel cladding temperatures will be only slightly greater than the local water temperature. As the bounding fuel cladding temperature in the SFP is based on maximum decay heat fluxes from freshly discharged
. fuel, the fuel cladding temperatures in the Cask Pit are bounded by the previously calculated value. This demonstrates the adequacy of cooling the Cask Pit via the buoyancy driven exchange of water between the pit and the SFP.
l I
l Holtec Report HI 981933 21 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
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V. .
)
t
- 5.9 Fuel Handline ' Area Ventilation (FHAV) ~
' An evaluation'of the FHAV system was performed. This evaluation was performed for the full com discharge' scenario 4A,' which provides the greatest heat load burden to the FHAV system.
Using the design inlet air parameters from the DBNPS USAR, the maximum calculated building temperature is 103*F. The relative humidity was calculated to increase by less than 25 percent mlative humidity. Therefore, it is concluded that the additional burden on the FHAV system, as .
a result of the peak heat loads from the SFP, is within the design capability of the FHAV system.
i l
l Holtec Report HI-981933 5-22 80284 i
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-5.10 References -
[5.4.1]." Heat Loss to the Ambient from Spent Fuel Pools: Correlation of Theory with Experiment" Holtec Report HI-90477, Rev. O, April 3,1990.
[5.4.2] "An Improved Correlation for Evaporation from Spent Fuel Pools", Holtec Report HI-971664,Rev.O.
~ [5.4.3] "QA Documentation for LONGOR v10 ., " Holtec Report HI 951390 Revision 0 -
[5.4.4] Croff, A.G., "ORIGEN2 - A Revised and Updated Version of the Oak Ridge Isotope Generation and Depletion Code, ORNL-5621, Oak Ridge National Laboratory,1980.
[5.4.5] "QA Documentation for BULKTEM v3.0," Holtec Report HI-951391, Revision 1. -i
[5.5.1] "QA Validation for TBOIL v1.6," Holtec Report HI-92832, Revision 2.
[5.5.2] USNRC Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water
~ Reactors for Long Term Cooling," Revision 2, July 1981.
[5.6.1] Batchelor, G.K., "An Introduction to Fluid Dynamics", Cambridge University Press, -
1967.
[5.6.2] Hinze, J.O.,'" Turbulence", McGraw Hill Publishing Co., New York, NY,1975.
[5.6.3] Launder, B.E., and Spalding, D.B., " Lectures in Mathematical Models of Turbulence"..
- Academic Press, London,1972.
[' 5.6.4] ."QA Documentation and Validation of the FLUENT Version 4.32 CFD Analysis Program" Holtec Report HI-961444, Revision 0.
[5.7.1] Rohsenow, N.M., and Hartnett, J.P., " Handbook of Heat Transfer", McGraw Hill Book Company, New York,1973.
Holtec Report HI-981933 5-23 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
> , j
c Table 5.3.1 Davis Besse Historic and Projected Fuel Discharge Schedule Number of Discharge Date Average Burnup *U Enrichment Uranium Weight Assemblies (Month & Year) (mwd /MTU) (wt%) (kgU) 53 March 1982 23888 2.48 472.16 85 July 1983 26996 2.67 472.21 65 September 1984 28153 2.64 471.06 65 March 1988 34190 3.00 468.75 60 January 1990 31142 3.02 468.21 59 August 1991 36254 3.18 468.25 61 March 1993 38046 3.15 467.85 65 October 1994 41039 3.45 468.37 74 April 1996 42948 3.71 467.88 77 April 1998 46492 3.90 467.89-77 March 2000 49491 4.32 467.93 73 March 2002 51134 4.43 467.83 73 March 2004 52972 4.20 479.86 73- March 2006 55782 3.99 489.51 73 March 2008 55783 3.99 489.51 72 March 2010 55881 4.00 489.80 72 March 2012 - 55881 4.00 489.80 72 March 2014 55881 4.00 489.80 72 March 2016 55881 4.00 489.80 72 March 2018 55881 4.00 489.80 72 March 2020 55881 4.00 489.80 72 March 2022 55881 4.00 489.80 72 March 2024 55881 4.00 489.80 Note: In performing calculations, the listed burnup values are increased by 2% to include uncertainties in the reactor thermal power.
Holtec Report H1-981933 5-24 80284 SilADED REGIONS DESIGNATE PROPRIETARY INFORMATION !
l l
r I
l.
TABLE 5.4.1 DATA FOR SFP BULK TEMPERATURE EVALUATION Reactor Thermal Power 2827.5 MWt Reactor Core Size 177 assemblies SFPCS HX Coolant Flow Rate 650 gpm SFPCS HX Coolant Temperature 95 F DHRL .X Coolant Flow Rate 6000 gpm DHRS HX Coolant Temperature 95 F Minimum In-Core Hold Time 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> i
Fuel Assembly Discharge Rate 4 g Spent Fuel Pool length (N-S) 635.5 inches Spent Fuel Poollength (E-W) 239.5 inches l Spent Fuel Pool Depth 36.86 feet SFPCS HX Design Conditions l Coolant Inlet Temperature 95 F Coolant Outlet Temperature l' 1 1.2 F SFP Water Inlet Temperature 120 F DHRS HX Design Conditions 4 Coolant Inlet Temperature 95 F l SFP Water Inlet Temperature 140 F {
Coelant Flow Rate 6000 gptn Heat Removal Rate 6 26.9x10 Btu /hr
_Beanding Fuel Assembly Weight 1682 pounds I
i Holtec Report HI-981933 5-25 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION J
I TABLE S.5.1 DATA FOR TIME-TO-BOIL EVALUATION Spent Fuel Pool Length (N-S) 635.5 inches Spent Fuel Pool Length (E-W) 239.5 inches Spent Fuel Pool Depth 36.9 feet Total Rack Weight 268,000 lb Bounding Fuel Aaembly Weight 1682 pounds Pool Building Ambient i10'F Temperature Emissivity of Water 0.96 Pool Net Water Volume 31,580 ft' l
l Holtec Report HI-981933 5-26 80284 SIIADED REGIONS DESIGNATE PDOPRIETARY INFORMATION l
L_..
w-
]
TABLE 5.6.1 DATA FOR SFP/ CASK PIT LOCAL TEMPERATURE EVALUATION l Bounding Assembly Weight'- 1682 pounds ,
Radial Peaking Factor - 1.64 1 Axial Peaking Factor 1.52' Maximum Number of Fuel 1714/289 <
Assemblies Assumed for Analysis (SFP/ Cask Pit). 1 Cooled Water Flow Rate 3000 gpm Type of fuel asserdbly. Babcock and Wilcox 15x15
. -l Fuel Rod Outer Diameter. _0.430 inches max.
0.416 inches min.
Rack Cell Inner Dimension ' 9.0 inches Active Fuellength - __145 inches Number of Rods per Assembly
- 225 rods Rack'Celllength 1615/8 inches Bottom Plenum Height 6 inches
- Note Fuel assembly is modeled as a square array with all locations containing fuel rods for permeability determinations. 208 fuel rods are used for heat transfer calculations.
Holtec Report HI-981933
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5-27 80284
}. SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
~
)
I
.~
TABLE 5.8.1 RESULTS OF BULK TEMPERATURE TRANSIENT i Maximum Bulk Coincident Decay Heat Time After Reactor Scenario Temperature ( F) Load (Btu /hr) Shutdown (hrs)
I 132.98 6 g;. 15.89x10 183 2 169.32 6 15.55x10 397 3A 165.87 29.66x106 205*
s 3B 164.90 29.28x10 6
205 4A 151.42 29.75x106 203*
4B 150.67 29.38x106 203
- Note: Time for these scenarios is measured from the second reactor shutdown.
Holtec Report HI-981933 5-28 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION h
F TABLE 5.8.2 RESULTS OFMINIMUM TIME-TO BOIL AND MAXIMUM BOILOFF RATE EVALUATION Scenario * ' Minimum Time-to-Boil (hrs) Maximum Boiloff Rate (gpm) al -
10.42 34.45 4A 3.78 69.57
' Note: As discussed in Section 5.3, boiling evaluations are not performed for Scenarios 2,3A and 3B, and only performed for the most limiting of Scenarios 4A and 4B.
l l
Holtec Report HI-981933 5-29 80284 SIIADED REGIONS DESIGNATE PROPRIETARY INFORMATION ,
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88l
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qiig %:dp x . ;gp; . pg 5a t_. ptt ,c
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1 includes credit for the confining effect of the surrounding
. concrete. It is noted that this criterion is in conformance with the ultimate strength primary design methodology of the American Concrete Institute in use since 1971. For the DBNPS, fe' =
4,000 psi and the allowable static bearing pressure is fb = 4,760 psi, assuming full concrete 1 confinement. The allowable bearing pressure computed above is conservatively computed by .
taking e=1 to account for lack of total concrete confinement in the leak chase region. Thus, the maximum allowable concrete bearing pressure is 2,380 psi. The acceptance criterion for the !
- bearing pad is to show that this primarily compressive component remains in the clastic range. I i
The analysis is performed with ANSYS using a finite element model, which places a bearing pad
- and rack pedestal directly above a leak chase location. This configuration is selected with the.
l intent of bounding all other possible bearing pad / leak chase interfaces by removing a substantial f portion of the concrete directly beneath the pedestal. The liner plate is conservatively neglected L
in order to maximize bearing pad and concrete stresses. The analysis applies the maximum vertical pedestal load from results for all pedestals scanned from the time-history solutions from l
. all simulations. The maximum vertical pedestal load is taken to be 150,000 lbs (which is conservative, since the maximum SSE event pedestal impact load is actually 121,000 lbs).
. Holtec Report HI-981933 6-35 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION
m n
- The bearing pad selected is 1.5" thick,'austenitic stainless steel plate stock. Figure 6.9.1 provides an isometric of the' ANSYS fm' ite element model. The model permits the bearing pad to deform and lose contact with the liner,if the conditions of clastostatics so dictate. Figure 6.9.1 shows the bearing pad and underlying leak chase located within the supporting concrete. The slab is modeled as an elastic foundation. Figure 6.9.2 shows the pressure profile in the underlying concrete computed by the ANSYS analysis.
.The average pressure at the pad to liner interface is computed and compared against the above-mentioned limit. Calculations 'show that the average pressure at the slab / liner interface is 1,006 psi, which is well below the allowable value of 2,380 psi, providing a factor of safety of 2.36.
The stress distribution in the bearing pad is also evaluated, with the results shown in Figure 6.9.3.
The peak bending stress in the bearing pad under the maximum venical load is 16,345 psi. The material yield strength of 25,000 psi at 200 F provides factor of safety against yield of about 1.53.
Section 7.0 also discusses an alternate pedestal / leak chase configuntion considering a pedestal adjacent to multiple leak chases under a more extreme load condition resulting from a dropped fuel assembly. The instantaneous peak force from this conservatively analyzed shon duration accident is approximately 9 million pounds under which the bearing pad is still shown to be acceptable. Therefore, the bearing pad design devised for the DBNPS Cask Pit is deemed appropriate for the prescribed loadings.
6.10. -level A Evaluation The level A condition is not a governing condition for spent fuel racks since the general level of loading is far less than Izvel D loading. Additionally, the material stresses computed for the Level D loadings were compared against Level A allowables. This practice ensures that both level B and level A conditions are bounded.
Holtec Report HI-981933 - 6-36 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION t 1
6.11 Hydrodynamic L=ic on Camk Pit Walls
. The hydrodynamic pressures that develop between adjacent racks and the Cask Pit walls can be
. developed from the archived results produced by the WPMR analysis. The time dependent pressures am determined for the rack that resulted in the maximum displacement. The maximum insta'nt'aneous hydrodynamic pressure plots for the SSE event are shown in Figure 6.11.1.
6.12 Imal Stress Considerations This section presents the results of evaluations for the possibility'of cell wall buckling and the -
secondary stresses produced by temperature effects.
6.12.1 Cell Wall Buckline
. The allowable local buckling stresses in the fuel cell walls are obtained by using classical plate buckling analysis. The evaluation for cell wall buckling is based on the applied stress being
- uniform along the entire length of the cell wall. In the actual fuel rack, the compressive stress comes from consideration df overall bending of the rack structures during a seismic event, and as such is negligible at the rack top, and maximum at the rack bottom.
The critical buckling stress, with a safety factor of 1.5, is determined to be 6,799 psi. The computed compressive stress in the cell wall, based on the R6 stress factor, is 6,480 psi.
Therefore, there is a sufficient margin of safety against local cell wall buckling.
Holtec Report HI-981933 - 6-37 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION
6.12.2 - Analysis of Welded Joints in Rack Cell-to-cell welded joints are examined under the loading conditions arising from thermal effects due to an isolated hot cell in this subsection. This secondary stress condition is evaluated alone and not combined with primary stresses from other load conditions.
A thermal gradient between cells will develop when an isolated storage location contains a fuel assembly emitting maximum postulated heat, while the surrounding locations are empty. We can obtain a conservative estimate of weld stresses along the length of an isolated hot cell by considering a beam strip uniformly heated by the thermal gradient, and restrained from growth along one long edge. This thermal gradient is based on the results of the thermal-hydraulic evaluations, which show that the difference between the local cell maximum temperatures and the bulk temperature in the pool is 4.5 F. The analyzed configuration is shown in Figure 6.12.1. I Using shear beam theory, an estimate of the maximum value of the average shear stress in the strip is given as % = 1,240 psi. Since this is a secondary thermal stress, we use the allowable shear stress criteria for faulted conditions (0.42*So =29,820 psi) as a guide to indicate that the
. maximum shear is acceptable. The margin of safety against cell wall shear failure due to cell wall growth is greater than 24 for the worst case hot cell conditions.
l i
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Holtec Report HI-981933 6-38 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION j
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' 6.13 References
[6.1.1] USNRC NUREG-0800, Standard Review Plan, June 1987.
-[6.1.2]- (USNRC Office of Technology) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14,1978, and January 18,1979 amendment thereto.
[6.2.1]. Soler, A.I. and Singh, K.P., " Seismic Responses of Free Standing Fuel Rack Constmetions to 3-D Motions", Nuclear Engineering and Design, Vol. 80, pp. 315-329 (1984).
[6.2.2] Soler, A.I. and Singh, K.P.,"Some Results from Simultaneous Seismic Simulations of All Racks in a Fuel Pool", INNM Spent Fuel Management Seminar X, January,1993.
[6.2.3] Singh, K.P. and Soler, A.I., " Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the Chin Shan Experience, Nuclear Engineering International, UK (March 1991).
[6.2.4] Holtec Proprietary Report HI-961465 - WPMR' Analysis User Manual for Pre & Post Processors & Solver, August,1997.
[6.4.1] USNRC Standard Review Plan, NUREG-0800 (Section 3.7.1, Rev. 2, .
1989).
1 [6.4.2]' Holtec Proprietary Report HI-89364 - Verification and User's Manual for - l
. Computer Code GENEQ, January,1990.
[6.5.1] Rabinowicz, E., " Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company,1976.
[6.5.2] Singh, K.P. and Soler, A.I., " Dynamic Coupling in a Closely Spaced Two- :
Body System Vibrating in Liquid Medium: The Case of Fuel Racks," 3rd j International Conference on Nuclear Power Safety, Keswick, England, May 1982.'
[6.5.3] - Fritz, R.J., "The Effects of Liquids on the Dynamic Motions of Immersed I Solids," Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172. '
. Holtec Report HI-98t933 . 6-39 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION l
l i
[6.5.4] Levy, S. and Wilkinson, J.P.D., "The Component Element Method in.
Dynamics with Application to Earthquake and Vehicle Engineering,"'
McGraw Hill,1976.
1[6.5.5] . Paul, B., " Fluid Coupling in Fuel Racks: Corn:lation of Theory and Experiment", (Proprietary), NUSCO/Holtec Report HI-88243.
[6.6.1] - ASME Boiler & Pressure Vessel Code,Section III, Subsection NF,1989 Edition.
[6.6.2] ASME Boiler & Pressure Vessel Code,Section III, Appendices,1989 Edition.
[6.6.3] USNRC Standard Review Plan, NUREG-0800 (Section 3.8.4, Rev. 2, 1989).
[6.9.1] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan,1985.
[6.9.2] ACI 318-95, Building Code requirements for Structural Concrete,"
American Concrete Institute, Detroit, Michigan,1995.
I L
l Hohec Report HI-981933 6-40 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION
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R fe PLANT DOCKET NUMBER (s) YEAR Enrico Fermi Unit 2 USNRC 50-341 1980 Quad Cities 1 & 2 USNRC 50-254,50-265 1981 Rancho Seco USNRC 50-312 1982 Grand Gulf Unit 1 USNRC 50-416 1984 Oyster Creek USNRC 50-219 1984 Pilgrim USNRC 50-293 1985 V.C. Summer USNRC 50-395 1984 Diablo Canyon Units 1 & 2 USNRC 50-275,50-323 1986 Byron Units 1 & 2 USNRC 50-454,50-455 1987 Braidwood Units 1 & 2 USNRC 50-456,50-457 1987 Vogtle Unit 2 USNRC 50-425 1988 St. Lucie Unit 1 USNRC 50-335 1987 Millstone Point Unit 1 USNRC 50-245 1989 Chinshan Taiwan Power 1988 D.C. Cook Units 1 & 2 USNRC 50-315,50-316 1992 Indian Point Unit 2 USNRC 50-247 1990 Three Mile Island Unit 1 USNRC 50-289 1991 James A. FitzPatrick USNRC 50-333 1990 Shearon Harris Unit 2 USNRC 50-401 1991 Hope Creek USNRC 50-354 1990 Kuosheng Units 1 & 2 ,
Taiwan Power Company 1990 Holtec Report 111-981933 6-41 80284 SilADED AREAS DESIGNATE PROPRIETARY INFORMATION
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hy ypp PLANT DOCKET NUMBER (s) YEAR Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units 1 & 2 Comision Federal de 1991 Electricidad Zion Station Units 1 & 2 USNRC 50-295,50-304 1992 Sequoyah USNRC 50-327,50-328 1992 LaSalle Unit 1 USNRC 50-373 1992 Duane Arnold Energy Center USNRC 50-331 1992 Fort Calhoun USNRC 50-285 1992 Nine Mile Point Unit 1 USNRC 50-220 1993 Beaver Valley Unit 1 USNRC 50-334 1992 Salem Units 1 & 2 USNRC 50-272,50-311 1993 Limerick USNRC 50-352,50-353 1994 Ulchin Unit i KINS 1995 Yonggwang Units 1 & 2 KINS 1996 Kori-4 KINS 1996 Connecticut Yankee USNRC 50-213 1996 Angra Unit i Brazil 1996 Sizewell B United Kingdom 1996 Waterford 3 USNRC 50-382 1997 J.A. Fitzpatrick USNRC 50-333 1998 Callaway USNRC 50-483 1998 Nine Mile Unit 1 USNRC 50-220 1998 Holtec Repon H1-981933 6-42 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION
r-l aman m ma* PLANT DOCKET NUMBER (s) YEAR Chin Shan Taiwan Power Company 1998 Vermont Yankee USNRC 50-271 1998 Millstone 3 USNRC 50-423 1998 Byron /Braidwood USNRC 50-454,50-455, 1999 50-567,50-457 Wolf Creek USNRC 50-482 1999 Plant liatch Units 1 & 2 USNRC 50-321,50-366 1999 liarris Pools C and D USNRC 50-401 1999 l
l i
Holtec Report HI-981933 6-43 80284 SilADED AREAS DESIGNATE PROPRIETARY INFORMATION
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Data 3 corresponds to the time-history acceleration values along the Z axis (Vertical) l Ilottee Report 111-981933 6-45 80284 SilADED AREAS DESIGNATE PROPRIETARY INFORMNTION
l k_
. U, Uy U, 0, By 0, t Pi p2 P3 94 95 96 2 P7 Ps P9 910 9 11 92 Node 1 is assumed to be attached to the rack at the bottom most point.
Node 2 is assumed to be attached to the rack at the top most point.
Refer to Figure 6.5.1 for node identification.
2' P3 pi4 3' pas pts.
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pi = q,(t) + Ux(t) i = 1,7,13,15,17,19,21
= qi(t) + Uy(t) i = 2,8,14,16,18,20,22
=. q,(t) + U,(t) i = 3,9
= qi(t) i = 4,5,6,10,11,12 pi d enotes absolute displacement (or rotation) with respect to inertial space d
qi enotes relative displacement (or rotation) with respect to the floor slab
- denotes fuel mass nodes U(t) are the three known earthquake displacements Holtec Report HI-981933 6-46 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION
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%h #n4lk bsiydgd fk&h,k sW SSE Item / Location Calculated Allowable i Fuel assembly / cell wall impact, Ibf. 594 3,031" Rack / baseplate weld, psi 19,800 21,300 Female pedestal / baseplate weld, psi 6,421 21,300 Cell / cell welds, psi, based on impact loads 2,640 10,000 Cell / cell welds, psi, based on shear flow 8,7001 " 10,000 l
1 i
1 l'
Note that Level A condition allowables were conservatively applied against SSE loads.
Based on the limit load for a cell wall. The allowable load on the fuel assembly itself may be less than this value, but will be greater than 840 lbs.
t" l Based on the base metal stresses adjacent to weld placements resulting from the l
maximum shear flow developed between two adjacent cells.
i l
lloltec Report HI-981933 6-47 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION 1
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REPORT HI-981933 l-
+ -
7.0 - FUEL HANDLING AND MECHANICAL ACCIDENTS 7.1 ; Introduction The USNRC OT position' paper [7.1] specifies that the design of the rack must ensure the functional integrity of the spent fuel racks under the postulated load drop events in the Cask Pit. - -'
This section contains synopses of the analyses carried out to demonstrate the regulatory
. compliance of the proposed racks under postulated mechanical accidents germac to the DBNPS.
. 7.2 Description of Accidents' In the evaluation of fuel handling accidents discussed herein, the concern is with the damage to the storage racks, and the Cask Pit structure. The configuration of the rack cell size, spacing, and neutron absorber material must remain consistent with the configurations used in the criticality and thermal-hydraulic evaluations. Maintaining these designed configurations will ensure that the results of the criticality and thermal-hydraulic evaluations remain valid.
Two categories of fuel assembly drop accidents are evaluated: a shallow drop and a deep drop, both of which are discussed in detail below. Each of the fuel handling accidents considers the
' drop of a fuel assembly, along with the portion of handling tool, which may be severed due to a single element failure. The total dropped weight is 2,482 pounds. The origin of the dropping trajectory is chosen as the highest elevation that the load can be lifted, by the Fuel Storage
. Handling Bridge, which is 98.13 inches above the upper elevation of the fuel storage rack.
j l
' Additional evaluations were also performed to consider the ability of the rack to withstand a 500 g i
pound uplift force and the Cask Pit to withstand a rack dropd ' uring installation. Material
-l definitions are provided in Table 7.2.1.
'1 Holtec Report HI-981933 7-1 80284 1
SHADED AREAS DESIGNATE PROPRIETARY INFORMATION l
m,
~
The radiological consequences resulting from fuel damage are not an issue for th'e proposed changes. The previously evaluated fuel handling design basis accident for the DBNPS continues to bound the radiological consequences of dropping a fuel assembly.
7.2.1 Shallow Droo Events -
The first category of fuel handling accidents considers a fuel assembly striking either the top of .
stored fuel or the top of the storage rack and is referred to herein as a " shallow drop" event. The first shallow drop scenario considers a falling fuel assembly travelling vertically through the stratum of water before striking the top of a stored fuel assembly and subsequently impacting the -
. top,of the weakest module, which was determined to be an 8x8 cell rack. A portion of the kinetic energy of the falling assembly is absorbed by damage to the rack.
This first impact scenario determines the depth and extent of plastic deformation of the 0.075 inch thick cell wall. Since the new racks are of honeycomb construction, the deformation produced by the impact will be confined to the region of collision. However, the depth of gross 1 deformation to the cell walls must be demonstrated to remain limited to the portion of the cell above the top of the active fuel region, which is essentially the elevation of the top of the Boral neutron absorber. To meet this criterion, the plastic deformation of the rack cell wall is conservatively specified to not extend more than 4.75 inches downward from the top of the rack.
This is the minimum distance down to the top of the Boral, including tolerances. The active fuel area begins at approximately 5.25 inches below the top of the rack cell. Maintaining the Boral and surrounding storage cell will ensure that the configurations considered in the criticality evaluations are not compromised.
The impact zone is chosen to maximize penetration of the falling assembly. From the description of the rack modules in Section 3, the impact resistance of a single vertical cell wall at the rack corner is less than any other potential impact zone represented by multiple cell walls or interior walls. Accordingly, the potential shallow drop scenario is postulated to occur at a rack corner i i
cell in the manner shown in Figure 7.2.1. This impact zone is chosen to minimize the cross lioltec Report HI-981933 7-2 80284 SilADED AREAS DESIGNATE PROPRIETARY INFORMATION 1
J
sectional area. In order to maximize the penetration into the top of the rack by the falling assembly, the rack is considered empty, with the exception of the impacted corner cell, where an irradiated fuel assembly is stored.
The second shallow drop accident scenario considers a fuel assembly striking the top of an empty
- rack cell to maximize cell wall deformation. This drop scenario is performed to maximize cell - "'
blockage. As discussed in Section 5.6, the thermal hydraulic evaluations, performed to support the additional Cask Pit storage racks, considered 50 percent cell blockage. Therefore, the rack will be considered acceptable under this drop scenario if 50 percent or more storage cell area remains open for cooling flow subsequent to the event. In this scenario, all other elements of the impacting fuel assembly and the impacted rack assembly are identical to those used in the first shallow drop scenario. Since the rack is considered empty in this scenario, criticality consequences need not be considered.
7.2.2 Deep Drop Events The second category of fuel assembly drop events postulate that the 2482 lb. impactor falls through an empty storage cell and impacts the rack base-plate. The origin of the dropping trajectory is again chosen as the highest elevation that the load can be lifted by the Fuel Storage Handling Bridge, which is 98.13 inches above the upper elevation of the fuel storage rack. This so-called deep drop scenario evaluates the structuralintegrity of the rack baseplate. If the baseplate is pierced or deforms sufficiently, then the fuel assembly or base-plate might damage the pool liner and/or create an abnormal condition of the enriched zone of fuel assembly outside the poisoned space of the fuel rack. To preclude damage to the pool liner, and to avoid the potential of an abnormal fuel storage configuration in the aftermath of a deep drop event, it is required that the base-plate remain unpierced. It is also required that the maximum lowering of the fuel assembly support surface is less than the distance from the bottom of the rack base-plate )
to the liner. l Holtec Report HI-981933 7-3 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION l
I l 1
(; ,
The deep drop event is classified into two scenarios. The first scenario considers dropping an
~ assembly through a cell located above a support pedestal, which is located directly above a leak
" chase, as shown in Figure 7.2.2. The relative location of the pedestal and leak chase are chosen
. to account for all possible occurrences ofleak chases located beneath pedestals. The second scenario considers dropping the impactor at an interior cell near the center of the rack as shown in Figure 7.2.3. - '
In the first scenario, the base-plate is buttressed by the support pedestal and presents a hardened impact surface, resulting in a high impact load. The principal design objective is to ensure that the support pedestal does not cause catastrophic damage to the liner and underlying reinforced conen:te pool slab such that rapid loss of pool water occurs.
For the second deep drop scenario, the base-plate is not as stiff at cell locations away from the support pedestal. This scenario is evaluated to determine the damage and deformation to the rack baseplate. Baseplate severing or large deflection of the base-plate, such that the liner would be impacted,' would constitute an unacceptable result. The deformation must be shown to be less than the distance from the bottom of the baseplate to the pool floor liner, which is 6 inches, including tolerances.
7.2.3 Rack Droo Event 1 i
The rack drop event is analyzed to show that dropping a rack into the Cask Pit during installation 1 will not result in catastrophic leakage of the Cask Pit. Damage must not lead to development of cracks through the entire Cask Pit floor section. Although this scenario is evaluated,
' implementation of the control of heavy loads will preclude its consequences.
l
)
. Holtec Report HI-981933 7-4 80284 i SIIADED AREAS DESIGNATE PROPRIETARY INFORMATION I
7.2.4 Uplift Force Evaluation The 500 pound uplift force is evaluated to ensure the rack cell wall is able to withstand this load without deforming the rack cell such that it no longer satisfies dimensional requirements. The acceptance criterion for this evaluation is that local cell wall stress shall remain below the yield - -'
point.
7.3 Mathematical Model
. In the first step of the solution process, the velocity of the dropped object (impactor) is computed for the condition of underwater free fall. Table 7.3.1 summarizes the results for the fuel assembly drop events. In the second step of the solution, an clasto-plastic finite element model of the impacted region on Holtec's computer Code PLA.STIPACT (Lawrence Livermore National Laboratory's DYNA 3D implemented on Holtec's QA system) is prepared. PLASTIPACT simulates the transient collision event with full consideration of plastic, large deformation, wave propagation, and elastic / plastic buckling modes. The physical properties of material types l undergoing deformation in the postulated impact events are summarized in Table 7.3.2.
7.4 Results i7.4.1 Shallow Droo Event Results -
' Fi lture 7.4.1 provides an isometric view of the finite elemera model utilized in the shallow drop impact analysis.-
1 I
The first shallow drop scenario dynamic analysis shows that the top of the impacted region undergoes localized deformation. The impacting fuel assembly has an initial velocity of 250 i
- in/sec. Figure _7.4.2 shows an isometric view of the post-impact geometry of the rack for this shallow drop scenario, as well as a plot of the Von-Mises stresses. The maximum Von Mises Holtec Report HI-981933 7-5 80284 l SHADED AREAS DESIGNATE PROPRIETARY INFORMATION i
n i
stress in the cell wall, recorded at maximum displacement time,is 38.39 km end the maximum
' plastic strain is O'.106. . Approximately 10% of the cell opening in the impacted cell is blocked.
u The maximum gross deformation is limited to 3 inches, which is below the acceptance criteria of 4.75 inches. Therefore, the penetration is determined to be acceptable from a criticality perspective an'd the racks will remain subcritical,
, ..i The study of residual plastic strain for the second shallow drop analysis shows that damage remains local to the impacted cell, but is significantly more extensive than the first scenario.
iFigure 7.4.3 shows an isometric view of the post-impact geometry of the rack for this scenario as
- well as a plot of the Von-Mises stresses. Deformation of the impacted cell extends 18 inches
- downward from the top of the undeformed cells. The maximum Von-Mises stress in the cell wall is 40.96 ksi and the maximum plastic strain is 0.264. The effective damaged area measures 12 inches and can obstruct approximately 50 percent of the cross section of the cell. Thus, the acceptance criterion for blockage is met. Since the percentage of obstruction recorded is for an empty cell, it is concluded that this analysis would bound the damage sustained by a loaded cell.
Therefore, the partial blockage assumption of 50 percent is shown to be acceptable.
7.4.2 Deep Droo Event Result's
, The first deep drop scenario considers the impacted area to be over a pedestal that is resting on
!- the % inch thick liner and located near the convergence of two leak chases. Figure 7.4.4 shows an isometric view of the finite element model for the impactor, pedestal, bearing pad, liner and underlying concrete. As shown in Figure 7.4.5, a Von-Mises stress of 106 ksi is observed in the pedestal cylinder at the contact surface with the bearing pad, which is below the failure stress of L 140 ksi for the pedestal material.- The bearing pad registers a Von-Mises stress of approximately l 30 ksi, as shown in Figure 7.4.6.
i The numerical analysis of this event shows that the liner is not pierced during the collision, since the maximum Von-Mises liner stress, as shown in Figure 7.4.7, is 27 ksi, which is less than the failure stress of 71 ksi. Therefore, the acceptance criteria ~is satisfied. The concrete stratum Holtec Report HI-981933 7-6 80284 l t
SHADED AREAS DESIGNATE PROPRIETARY INFORMATION l
. 2
s directly below the pedestal sustains a very localized compressive stress of 21 ksi, as shown in Figure 7.4.8, which results in only localized damage to the concrete.
A plan view of the finite element model for the second deep drop scenario is shown in Figure 7.4.9. This scenario considers the dropped assembly to fall through an interior cell striking the
. base-plate at a point near the middle of the rack This drop scenario produces some deformation -
of the base-plate and localized severing of the base-plate to cell wall welds. The collision between the 2482 lb. impactor and the 0.75 inch thick rack base-plate occurs at 406 in/sec initial
. velocity and results in an accentuated local deformation of the base-plate extending over a 18
- square inch area around the impact zone. Due to the proximity of the fuel assembly lower end '
' fitting, the shock of the initial impact is carried into the walls of the centrally located cell, and
. fails the connecting welds to the adjoining cells. The base-plate does not break during the impact, but the welds connecting the cells located in the vicinity of the collision area to the plate are severed.
The structural damage resulting from this scenario has no adverse effect on the coolant flow
- through the storage cells. The maximum calculated Von-Mises stress in the base-plate as shown in Figure 7.4.10 is 46.04 ksi and the maximum calculated plastic strain in the base-plate is 0.109,
'as shown in Figure 7.4.11. Figure 7.4.12 shows the deformed shape of the base-plate. The maximum displacement of the base-plate is 3.36 inches, which develops 0.0135 seconds after the initial collision. The lower assembly storage position due to the deformed basplate is shown to be acceptable by the crititicality evaluations as discussed in section 4.6.4. This displacement does not result in the baseplate striking the liner, Therefore, the structural consequences are also acceptable.
7.4.3 : Rack Dron Event Results Numerical analysis of the drop of a 12,150 pound rack into the Cask Pit shows that the rack does ,
not pierce the % inch liner. The maximum calculated Von-Mises stress for the liner of about 45 ksi,' as shown in Figure 7.4.13, is less than the failure stress of 71 ksi for the liner material. The Holtec Report HI-981933 7-7 80284 SHADED AREAS DESIGNATE PROPRIETAR INFORMATION
r concrete stratum directly beneath the pedestal sustains a very localized compressive stress, as shown in Figure 7.4.14, with a maximum value of 23 ksi. This results in only localized damage to the concrete below the liner.
7.4.4 Unlift Force Evaluation Results
. .s
. This evaluation shows that the rack is able to withstand the uplift force of 500 pounds. For this scenario, the critical location for the load to be applied is at the top of a cell. For a load applied venically anywhere along a cell wall, the resultant stress is only 1,100 psi, which is'well below the yield stress of the material. For a load applied at a 45 degree angle to the top of a cell wall,.
tear out of the cell wall is evaluated. The damaged region extends no greater than 0.24 inches down the cell wall, which is well above the top edge of the neutron absorber material.
7.5 Closure The fuel assembly drop accident events postulated for the pools were analyzed and found to produce localized damage well within the design limits for the racks. The configuration of the fuel and poison (Boral) is not compromised from the configurations analyzed in the criticality
. evaluations discussed in Section 4.0. The base-plate deformation and corresponding fuel displacement is considered in the criticality evaluations. These evaluations concluded there are no criticality concerns for these accidents. The damage to the top of the racks reduces the cross sectional area available for coolant flow. However, the reduction of area is less than that considered in the thermal-hydraulic evaluations. Therefore, the accidents do not represent any thermal-hydraulic concerns.' Analyses show that the pool liner will not be pierced by the
- pedestals, but the underlying concrete will experience local cmshing. However, the pool stmeture will not suffer catastrophic damage. Therefore, there are no significant structural consequences.
IIoltec Repon HI 981933 7-8 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION
7.6 ' References
[7.1] "OT Position for Review and A' ceptance e of Spent Fuel Storage and Handling
' Applications," dated April 14,1978.
+
l Holtec Report HI-981933 7-9 80284 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION i
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" Shallow Drop" Von Mises Stress Report No. 981933 :
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" Shallow Drop" Von Mises Stress Report No. 981933
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Report No. 981933
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X Figure 7.4.7; Over-Pedestal" Deep Drop" Liner Von Mises Stress Report No. 981933
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Report No.981933
STEP 27 TIME = 1.3499969E-002 2 4991 -0 Z COORDINATE DISPt.ACEMENT
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" Heaviest Rack" Drop: Maximum Von Mises Stress - Liner Report No. 981933
RACK IMPACT ON POOL FLOOR 1.5871E+00 STEP 16 TIME = 3.1999566E-003 1.5871E+0 SIGZZ(MID) -2.1717E+0
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- Report No. 981933
n L 8.0 " CASK PIT STRUCTURE INTEGRITY CONSIDERATIONS -
i
' 8.1 - ' Introduction The DBNPS Cask Pit represents a small portion of the Auxiliary Building, which is a safety related, seismic category I, reinforced concrete structure.' Spent fuel is to be placed within the new storage racks located in the Cask Pit. This section discusses the analysis to demonstrate i
structural adequacy ~of the Cask Pit, herein also referred to as the pool structure. The analysis is
' performed in accordance with Section IV of the USNRC OT Position Paper [8.1.1].
l l
The numerical investigation is conducted considering the walls as plan plates and using " closed form" solutions available in the technical literature. Results for individual load components are ,
I combined using the factored load combinations mandated by SRP 3.8.4 [8.1.2], which are based on the " ultimate strength" design method. It is demonstrated that structural integrity is maintained for the critical factored load combinations. These evaluations were performed for the bounding case when the pools are fully loaded with spent fuel racks, as shown in Figure 1.3 with all storage locations occupied by fuel assemblies.
~
Both moment and shear capabilities are checked for concrete structural integrity. Bearing integrity of the slab in the vicinity of a rack module support pedestal pad was also evaluated. All structural capacity calculations are made using design formulas meeting the requirements of the American Concrete Institute (ACI).
8.2 Descriotion of Cask Pit Stmetures The analyzed reinforced concrete structure, which is comprised of the four perimeter walls of the Cask Pit,is isolated from the remainder of the Auxiliary Building and the Spent Fuel Pool and conservatively considered as an independent structure. The structural evaluation focused on the four reinforced concrete walls surrounding the Cask Pit. These four 46'-2" high reinforced concrete walls are supported at elevation 557'-0".by a massive (15'-0" thick) reinforced concrete
= Holtec Report HI-981933 8-1 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
mat, which extends down to bedrock. Figure 8.1.1 shows the area ofinterest and the major i
structural dimensions of the pool. The floor liner plate of the Cask Pit is located at elevation 557'. The operating floor is at elevation 603'-0".
The Cask Pit investigation concentrated on a portion of the monolithically constructed reinforced i concrete Auxiliary Building structure. The pertinent portion is the area located in the vicinity of the Cask Pit where the storage capacity is proposed to be increased. The thickness of the walls surrounding the Cask Pit are 3'-0 "at North and East, and 5'-6" at South and West. The continuity of the Cask Pit East wall is interrupted by the existence of the fuel gate opening, i
8.3 Definition of Loads Pool structural loading involves the following discrete components:
1
- 8.3.l' Static loading (Dead Loads and Live Loads)
- 1) Dead weight of pool structure includes the weight of the Auxiliary Building j concrete upper structure.
- 2) Maximum dead weight of rack modules and fuel assemblies stored in the modules -
based on 289 storage locations, as shown in Figure 1.3.
- 3) The Spent Fuel Cask Crane and Fuel Storage Handling Bridge (Refueling Platform) - The dead weight and the rated lift weight of these cranes are considered as dead load and live load, respectively.
- 4) The hydrostatic water pressure.
Holtec Report HI-981933 - 8-2 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
I- l
! 8.3.2 Seismic Induced I.oads 1); - Vertical loads transmitted by the rack support pedestals to the slab during a SSE or OBE seismic event.
j 2)' Hydrodynemic inertia loads due to the contained water mass and sloshing loads (considered in accordance with [8.3.1]) which arise during a seismic event.
- 3) Hydrodynamic pressures between racks and pool walls caused by rack motion in - .I the pool during a seismic event.
- 4) Seismic inertia force of the walls 8.3.3 Thermal Loading The temperatutes at the faces of the pool concrete walls and slabs define the thermal loading.
Two thermal loading conditions are evaluated and are defined by the bulk pool temperatures determined in the thermal-hydraulic evaluations, as described in Table 5.8.1. The normal operating condition considers the bulk pool temperature To to be 150 F. The accident condition conservatively considers the bulk pool temperature T,to be 180 F. The ambient temperature outside of the structure is considered to be -10 F. The temperature in the rooms of the Auxiliary Building and the Transfer Canal is considered to be 50 F. The concrete surface temperature on i I
the side exposed to air is elevated from the air temperatures to account for surface film behavior. i The concrete surface temperature on the water side is not adjusted.
1 i
The actual bulk pool water temperature under the limiting conditions considered for nomial !
operating conditions exceeds the 150 F concrete temperature limit imposed by the ACI code. i However, the use of 150 F for the normal condition in this calculation is acceptable because: !
- a. The maximum bulk temperature of the Spent Fuel Pool (SFP) water, which enters the l Cask Pit through the connecting gate,is determined to be 151.5 F. The maximum peak
. Holtec Report HI-981933 8-3 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORM ATION
local water temperature in the Cask Pit is determined to be 155.5 F, which occurs at the top of the storage rack cells, away from the concrete walls. The bulk water temperature in the Cask Pit will be between these two values, and probably closer to the 155.5 F.
~ However, the concrete wall surface temperature will be slightly lower, due to the )
insulating barrier provided by the liner and any trapped air beneath the liner, and the film effect of the water at the liner water interface. Therefore, the concrete surface temperature under the limiting normal conditions is expected to be only slightly higher
(~4.5 F) than the 150 F ACIlimit.
- b. The SFP bulk temperature will be above 150 F only under the worst case transient conditions and the duration of this temperature in excess of 150 F will occur for less than 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. The Cask Pit bulk temperature duration in excess of 150 F will be longer
(~100 hours) under the worst case conditions. These durations are not significant when considering the thermal inertia of the concrete walls and slab. In other words, the concrete temperature will lag such that the bulk of the concrete cross-section will remain well below the 150 F range. In fact, a very small depth,if any, of the concrete will experience temperatures in excess of 150' F.
l
- c. The normal condition evaluation, which includes the 150 F temperature, has large design margins. A comparison with the accident condition evaluations, with a 30 F higher temperature, indicates that a concrete temperature increase of only 4.5 F will not produce a significant effect on the computed results. Therefore, the evaluation of the concrete for a surface temperature of 150 F instead of 154.5 F produces results of sufficient accuracy. ,
- d. The corresponding limiting conditions are conservative for storage of spent fuel in the Cask Pit, since the evaluation considers stored fuel to completely fill a completely reracked maximum density SFP (to be sought in a future amendment) and a filled Cask Pit. ' However, the storage of fuel in the Cask Pit is temporary and will not occur along 1
with a filled reracked SFP, as discussed briefly in Section 1. i Holtec Report HI-981933 84 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION l
y l-L The concrete temperature in excess of 150 F under the worst case conditions is acceptable, since it will be experienced for a short duration. Thus, there will be no significant deterioration of the
- 1 l- concrete material properties. Based on ACI publication SP25 [8.3.2] and numerous other l technical papers, concrete compressive strength decreases about 10% at temperature 200 F as L
compared with the design specified strength of 4,000 psi. However, based on an ASCE paper
[8.3.3), the concrete compressive strength even at boiling conditions of 212 F is higher than the
\
28 day fc', if the strength margin from age is considered. At 212 F the concrete residual modulus of elasticity (Ec) is about 96.5% of the Ec ambient temperature value and the rebar ;
l modulus of elasticity (E.) is about 95% of the Es ambient temperature value according to ACI 216 [8.3.4]. 1 l
In general, both thermal expansion in the cross-section and the water pressure tend to create moments which cause the tension side of the concrete in the Cask Pit structure to be on the side away from the elevated temperatures. The rebars on the outside face are not affected by !
l temperature increases within the pool. Therefore, the short term elevated temperatures above the 150'F range do not significantly affect the material properties and evaluation of the cross-l sections at 150 F for normal' conditions is justified.
8.4 ' ' A' nalysis Procedures The Cask Pit reinforced concrete walls are subjected to various individual load cases covering
- l. the service conditions (the stmetural weight of the concrete structure, the weights of the upper portion of the Auxiliary Building concrete structure, the Spent Fuel Cask Crane, the hydro-static water pressure and the temperature gradients for normal operating and accident conditions) and L seismic induced loads (structural seismic loads, hydro-dynamic water loads, and rack-structure interaction dynamic loads) for OBE and SSE conditions. The service condition loads were considered as static acting loads, while the seismic induced loads for both OBE and SSE seismic l
l events are obtained from the simultaneous application of the three-directional acceleration spectra appropriate to elevation 603'-0".
Holtec Report HI-981933 8-5 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION l
b
e t-8.4.1 . Boundary Conditions To simplify the analysis and make it possible to use " closed form" solution results existing in the technical literature, the four walls surrounding the Cask Pit are considered as planar plates having three edges fixed at the contact borders with the adjoined walls and supporting mat. The remaining edge at the upper elevation is considered as a free edge.
8.4.2 Material Properties The behavior of the reinforced concrete existing in the structural walls is considered elastic and isotropic. The elastic characteristics of the concrete are independent of the reinforcement contained in each stmetural element for the case when the un-cracked cross-section is assumed.
This assumption is valid for all load cases with the exception of the thermal loads, where for a more realistic description of the reinforced concrete cross-section behavior the assumption of cracked concrete is used.' The elastic characteristics for the concrete and reinforcement used in this calcul'ation are summarized in Table 8.4.1. To simulate the variation and the degree of cracking patterns, the original elastic modulus of the concrete is modified in accordance with the methodology provided by ACI 349 [8.1.3]. Table 8.4.2 contains the elastic isotropic material properties and the . reduced elastic modulus (Ecra) pertinent to each wall.
8.4.3 lead Combinations The various individual load cases are provided in Table 8.4.3. These load cases are combined in accordance with the NUREG-0800 Standard Review Plan [8.1.2] requirements with the intent to obtain the most critical stress fields for the investigated reinforced concrete structural elements.
4 i
I Holtec Report HI-981933 8-6 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION i
-]
n i
l l' )
y
- For " Service Imad Conditions" the following load combinations are:
I l
l - Load Combination No.1 = 1.4* D + 1.7*L' l
-Ioad Combination No. 2 = 1.4* D + 1.7*L + 1.9*E l
- Load Combination No. 3 = 1.4* D + 1.7*L - 1.9*E
- Load Combination No. 4 = 0.75* (1.4* D + 1.7*L + 1.9*E +1.7*To)
- Load Combination No. 5 = 0.75* (1.4* D + 1.7*L - 1.9*E +1.7*To)
- Load Combination No. 6 = 1.2*D + 1.9*E
- Imad Combination No. 7 = 1.2*D - 1.9*E !
For " Factored Load Conditions" the following load combinations are:
- Load Combination No. 8 = D + L + To + E' l
- Load Combination No. 9 = D + L + To - E'
- Load Combination No.10 = D + L + Ta + 1.25*E !
- Load Combination No. I 1 = D + L + Ta - 1.25*E
- Load Combination No.12 = D + L + Ta + E'
- Load Combination No. I 3 = D + L + Ta - E' where:
D= dead loads; L= live loads; To = thermal load during normal operation; Ta = thermal load under accident condition; E= . OBE carthquake induced loads; E' = . SSE earthquake induced loads.
Holtec Report HI.981933 8-7 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
r i
- 1
?
o 8.5 ~ ' Results of Reinforced Concrete AnaI'vses The structural evaluation focused on the four reinforced concrete walls pertaining to the Cask Pit.
The axial forces, bending moments and shear forces were computed for each significant cross-
, section of the structural elements. The reinforced concrete cross-sectional capacities were determined and used to obtain the safety margins of the structural elements. Safety margins are defined as the allowable load divided by the computed load and acceptability is ensured if the safety margin is in excess of 1.0. The calculated safety margins for all four walls are shown in :
Tables 8.5.1 through 8.5.4. The limiting safety margin is 1.41.
8.6 ' Pool Liner i
t The pool liner is subject to in-plate strains due to movement of the rack support feet during the seismic event. Analyses are performed to establish that the liner will not tear or rupture under limiting loading conditions in the pool, and that there is no fatigue problem under the condition -
of I SSE event plus 20 OBE events. These analyses are based on loadings imparted from rack )
pedestals in the pool assumed to be positioned in the most unfavorable position. Bearing I strength requirements are shown to be satisfied by conservatively analyzing the most highly ic.aded pedestal located in the worst configuration with respect to underlying leak chases. j 8.7 - Conclusions
~ Regions affected by loading the Cask Pit completely with high density racks are examined for structural integrity under bending and shearing action. It is determined that adequate safety j margins exist when the factored load combinations are checked against the appropriate structural
' design strengths. It is also shown that local loading on the liner does not compromise liner !
a' integrity under a postulated fatigue condition and that concrete bearing ~ strength limits are not h exceeded.
t-l Holtec Report HI-981933 8-8 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION j
p l
l 8.8 References -
i . .
. [8.1.1] USNRC, "OT Position for Review and Acceptance of Spent Fuel Storage and l Handling Applications," April 14,1978, and Addendum dated January 18,1979 l [8.1.2] NUREG-0800, SRP-3.8.4, Rev.1., July 1981, i
[8.1.3] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures,
[ 'American Concrete Institute, Detroit, Michigan.
, [8.3.1] " Nuclear Reactors and Earthquakes, U.S. Department of Commerce, National Bureau of Standards, National Technical Information Service, Springfield, Virginia (TID 7024).
l l [8.3.2] ACI Publication SP25," Temperature and Concrete".
[8.3.3] ASCE Convention Paper," Strength Properties of Concrete at Elevated 3
L Temperatures," Boston, Mass., April 1979.
)
l
_ [8.3.4] ACI 216, R-81, " Guidelines for Determining the Fire Endurance of, Concrete l
Elements". ,
L l
I l
k t
Holtec Report HI-981933 8-9 80284 ;
SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
I
, Table No. 8.4.1 Concrete and Rebar Propenies Parameter Notation Value Concrete Compressive Strength (psi) fe' 4.000E+03 Un-Cracked Concrete Elastic Modulus (psi) Ee 3.605E+06 Concrete Poisson's Ratio y 0.167 ,
I 1
i Concrete Weight Density (Ib/ft3) D, 150.0 l l
Concrete Thermal Expansion Coefficient a 5.500E-%
Reinforcement Yield Strength (psi) Fy 6.000E+04 Reinforcement Elastic Modulus (psi) En.i,, 2.900E+07 l l l-Holtec Report HI-981933 8-10 80:284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION -
i
Table No. 8.4.2 Material Properties Structural Thickness E v 7, cx Ecr.ck Element (in) (psi) 3 (lb/ft ) (psi)
North Wall 36.00 3.605E+06 0.167 150. 5.500E-06 1.250E+06 South Wall 66.00 3.605E+06 4.059E+05 0.167 150. 5.500E-06 West Wall 3.605EM)6 4.059E+05 66.00 0.167 150. 5.500E-06 East Wall 3.605E+06 1.250E+06 36.00 0.167 150. 5.500E-06 Holtec Report Hi-981933 8-11 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION i
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i-Table No. 8.4.3 Individual Load Case Description Imad No. Type Description 1 -D Structural Concrete Weight 2 D Water Hydro-Static Pressure 3 L Auxiliary Building Live Loads
'4 E OBE Rack to Wall Coupling Pressure -
l 5 E' SSE Rack to Wall Coupling Pressure l- 6 E OBE Convective (Sloshing) Pressure
- 7. E' SSE Convective (Sloshing) Pressure j l
8 E OBE Impulsive Pressure 9 E' SSE Impulsive Pressure 1
10 E OBE Hydro-Dynamic Vertical Pressure j 11' E' SSE Hydro-Dynamic Vertical Pressure I l ,
12 E~ OBE StructuralInertia Loads l 13 E' SSE StructuralInertia Loads L
l 14 To Temperature for Operating Condition 15 Ta Temperature for Accident Condition i
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Table No. 8.5.1 North Wall Safety Factors
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Combination Reinforcement Orientation Load X Direction -- Y Direction Axial + Bending Shear Axial + Bending Shear _
1 412.59 96.34 93.09 85.82
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2 18.69 _4.28 21.50 4.63 3 20.08 4.69 39.71 5.19 I
4 27.43 5.70 28.50 6.18 5 31.74 6.26 52.69 6.92 6- 18.81 4.30 22.17 4.67 7 19.94 4.66 37.31 5.15 l
8 27.43 5.90 30.03 6.61 '
.9 .31.25 6.47 55.64 7.43 10 30.98 6.48 31.83 7.01 11 36.61 7.17 62.31 7.93 12' 27.94 5.90 30.03 6.61 13 32.35 6.47 55.64 7.43 Min 18.69 4.28 21.50 4.63 i
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Table No. 8.5.2 South Wall Safety Factors l
Combination Reinforcement Orientation bad X Direction Y Direction Case Axial + Bending Shear Axial + Bending Shear -
1 830.97 186.97 2 % .16 163.13 2 -33.17 7.25 46.51 7.74 3 32.35 7.86 77.37 8.55 4 3.51 9.66 58.89 10.32 l- 5 3.77 10.48 98.51 11.40 1
6 33.35 7.29 43.81 7.79 7 32.16 7.81 67.15 8.49 8 4.31 9.77 59.97 10.75 9 4.74 10.56 98.46 11.87 l
10 3.81 10.98 64.81 11.72 3 l
11 4.08 11.99 114.04 13.06 12 3.79 9.77 59.97 10.75 .
I 13 4.08 10.56' 98.46 11.87 Min 3.51 7.25 43.81 7.74 l
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Table No. 8.5.3 West Wall Safety Factors Combination Reinforcement Orientation Imad- ' X Direction Y Direction Case Axial + Bending Shear - Axial + Bending Shear 1 830.97 186.97 194.08 163.13 2 32.73 7.25 40.90 7.74 3 -33.50 i 7.86 68.72 8.55
'4 5.78 9.66 53.28 10.32 5 6.78 10.48 89.95 11.40 6 32.92 7.29 41.32 7.79 7 33.30 7.81 64.10 8.49
, 8 6.93 9.77 54.85 10.75 9 8.52 10.56 91.05 11.87 10 5.68 10.98 59.38 11.72 11 6.51 11.99 105.26 13.06 12 5.62 9.77 54.85 10.75 13 6.54- 10.56 91.05 11.87 Min 3502 7.25 40.90 7.74 Holtec Report HI-981933 8-15 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
Table No. 8.5.4 East Wall Safety Factors Combination Reinforcement Orientation Load X Direction Y Direction Case -
Axial + Bending Shear Axial + Bending Shear 1 1000 1000 1000 1000 2 7.96 1.41 31.43 4.90 3 7.78 1.41 31.28 4.90
'4 12.67 1.87 41.65 6.53 5 12.59 1.87 41.51 6.53 6 7.96 1.41 31.33 4.90 7 7.78 1.41 31.18 4.90 8 11.36 1.73 40.63 7.02 9 11.27 1.73 40.54 7.02 10 14.48 2.14- 47.44 7.44 11 14.40 2.14 47.29 7.44 12 11.81 1.73 40.63 7.02 13 11.72 1.73 40.54 7.02 Min 7.78 1.41 31.18 4.90 i
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I En IATE El SV - r fl3 IliVAID %T- F 14'T Elev. 603'-0" T OperatingDeck i
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S'f Figure 8.1.1; Plan View of Cask Pit Area lloitec Report HI-981933
9.0 RADIOLOGICAL EVALUATION 9.1 Solid Radwaste
= The SFP Purification System currently generates approximately 50 cubic feet of solid radioactive waste annually at the DBNPS. No significant increase in the volume of solid radioactive ~ wastes is expected from operating with the expanded storage capacity. The necessity for pool filtration resin replacement is determined primarily by the requirement for water clarity, and the resin is ]
normally changed about once every 18 months. The additional number of fuel assemblies in storage will not significantly affect the frequency of resin replacement.
9.2 Liauid Releases 1
The number of spent fuel assemblies in storage does not affect the release of radioactive liquids from the plant. The contribution of radioactive materials in the SFP water from the stored assemblies is insignificant relative to other sources of activity, such as the reactor coolant system.
The volume of SFP water processed for discharge is independent of the number of fuel assemblies stored. 1 9.3 Gaseous Releases Gaseous releases from the fuel storage area are combined with other plant exhausts. Currently there is no detectable contribution from the fuel storage area, and no significant increases are expected as a result of the expanded storage capacity.
Release of radioactive gases by the DBNPS will remain a small fraction of the limits of 10 CFR 20.1301 and the design objectives of Appendix I to 10 CFR 50 following the implementation of the proposed modification to increase spent fuel storage capacity. This conclusion is based on
- the following supporting statements:
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' a) sThe half-lives of short-lived nuclides such as I-131 are short in comparison to fuel cycle length; therefore, short-lived nuclides are present ohly in freshly offloaded fuel. The quantity of freshly offloaded fuel placed into the SFP each refueling outage is independent of the number of spent fuel assemblies being stored. Therefore, the inventory of I-131 in the SFP
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L : and Cask Pit will not be affected by the increased fuel storage capacity.
t b) 'Invento' ries oflong-lived fission products (e.g. Kr-85 and ternary tritium) in spent fuel assemblies will decrease slowly within individual fuel assemblies over years in storage.
' Therefore, an increase in the number of stored spent fuel assemblies would increase the total inventory of these radionuclides. However, these radionuclides are not released in significant amounts from the stored fuel to the SFP water, even for failed fuel, since the fuel pellet
. . temperature of stored fuel is not high enough to create sufficient gas pressure in the gap to overcome the static pressure of the SFP water.
c) . The radioactivity in the SFP water is independent of the number of stored assemblies. The SFP water activity is primarily dependent on the amount of fuel assembly movement within the SFP. The number of fuel assembly movements required for a refueling outage is generally limited to the movements required to complete the outage. The number of plant refueling outages should not change. Typical SFP activities are listed in Table 9.1.1.
d) The increased number of spent fuel assemblies in storage will raise the heat load on the SFP
. and could result in an increase in the evaporation rate. Other than a small amount of tritiated water released by evaporation, the radionuclides are non-volatile and consequently are not released from the pool water. The increased evaporation rate of tritiated water would result in an increase in gaseous tritium released in the plant's effluents. However, the discharge of gaseous radioactive effluents will continue to be a small fraction of the limits of 10 CFR 20.1301 and the design objectives of Appendix I to 10 CFR 50.
Holtec Report HI-981933 9- 2 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
9.4 Personnel Doses During normal operations, personnel working in the fuel storage area are exposed to radiation from the SFP. Operating experience has shown that area radiation dose rates originate primarily from radionuclides in the pool water.-
' l During refueling and other fuel-movement operations, pool water concentrations might be expected to increase somewhat due to crud deposits spalling from spent fuel assemblies and due to activities carried into the pool from the primary system. With respect to the rack installation, fuel movements in the SFP may be required in support of this project to reduce the possible dose to personnel during the rack installation. For this reason, although dose rates above and around the Cask Pit perimeter may increase marginally, the dose fields will still approximate conditions !
seen during normal operating conditions. Routine radiation surveys would identify any change to .
dose rates, and the appropriate radiological controls would be revised as required.
Radiation dose rates in accessible areas around the spent fuel storage and transfer zones were evaluated based on conservative fuel parameters and were found acceptable. For five year-l cooled fuel with design source gammas, the dose rate was determined to be approximately 28 millirems per hour. The DBNPS USAR Chapter 12 describes the current evaluations of personnel dose. USAR Figure 12.1-1 provides the radiation zones for normal operation at elevation 545 feet in the Auxiliary Building. The floor elevation of the Cask Pit is approximately j 557 feet elevation. Ground elevation at the DBNPS is approximately 585 feet elevation. The south and west walls of the Cask Pit are underground and present no possibility of exposure. The Auxiliary Building room to the north (Room 106) is designated as Radiation Zone D, with a designated dose rate of s 100 millirems per hour. The Auxiliary Building room to the east i (Room 109) is designated as Radiation Zone El, with a designated dose rate of s 1000 millirems per hour. The dose rate contribution from 3 year cooled fuel, at elevation 558 feet 6 inches, near the ceiling of rooms 1% and 109, was determined by the DBNPS to be approximately 70 millirems per hour.
Holtec Report HI-981933 9- 3 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
The Cask Wash Pit floor elevation, to the west of the Cask Pit, is 585 feet. Dose rates at this elevation are not expected to increase. This area is designated as a Radiation Zone D, which should not be affected by storage of spent fuel in the Cask Pit.
As a result, no changes are expected to the Radiation Zone designations evaluated in the DBNPS l USAR. Routine radiation surveys will be conducted to confirm the actual dose rates in rooms l 106 and 109 when fuel is transferred to the racks in the Cask Pit.
t Operating experience at the DBNPS has also shown that there are no detectable concentrations of airborne radioactivity in the SFP area except tritium, at approximately 3E-3 Derived Air Concentration (DAC). No increase in airborne radioactivity is expected as a result of the expanded storage capacity.
9.5 Anticipated Dose During Rack Installation All of the operations involved in the rack installation will utilize detailed procedures prepared with full consideration of ALARA principles. Similar operations have been performed in a l number of facilities in the past, and there is every reason to believe that re-racking can be safely and efficiently accomplished at the DBNPS, with low radiation exposure to personnel. !
l Total dose for the re-racking operation is estimated to be between 1.85 and 4.0 person-rem, as l
indicated in Table 9.5.1. While individual task efforts and doses may differ from those in Table 9.5.1, the total is believed to be a reasonable estimate for planning purposes. Though divers will be used only as necessary, the estimated person-rem burden for rack installation takes into consideration their possible dose. Radiation surveys will be conducted in the Cask Pit to confirm dose rates prior to diving activities. Cleanup of source material, which would contribute to an excessive dose for the divers will be performed, as necessary, in accordance with good practices i
I to limit dose ALARA.
4 Holtec Report HI-981933 9- 4 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
l The existing radiation protection program at the DBNPS is adequate for the rack installation c
operations. Where there is a potential for significant airborne activity, continuous air monitors will be in operation; Personnel will wear protective clothing as required and, if necessary, respiratory protective equipment. Activities will be governed by a Radiation Work Permit, and personnel monitonng equipment will be issued to each individual. Divers will be equipped with the appropriate personal dosimetry. As a minimum, this will include thermoluminescent , l dosimeters (TLDs) and self-reading dosimete'rs. Additional personnel monitoring equipment (i.e., extremity TLDs 'or multiple TLDs) may be utilized as required.
Work, personnel traffic, and the movement of equipment will be monitored and controlled to minimize contamination and to assure that dose is maintained ALARA.
1 After the rack installations, the lifting device will be washed with demineralized water and wrapped for contamination cont'r ols. The lift rig will be stored at the DBNPS site for future planned reracking of the SFP.
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, Table 9.3.1
- AVERAGE ACTIVITY OF WEEKLY SFP SAMPLES (From February,1999)
Nuclide Average Microcuries / cc Co-57 4.40 E-07 Co-58 1.57E-05 Co-60 8.65E-06 Ag-110M 3.66E-06 Sb-125 2.66E-05 Cs-134 9.88E-06 Cs-137 4.71E-05 Total - 1.12E-04 l
Hollec Report HI-981933 9- 6 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION 1
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Table 9.5.1 PRELIMINARY ESTIMATE OF PERSON-REM DOSE DURING CASK PIT RACK INSTALLATION Estimated Number of Person-Rem Step Personnel Hours Dose Installation of new rack module - 5 -20 0.25 to 0.5 Phase 1 Installation of new rack module -
5 20 0.8 to 1.5 Phase 2 Install remaining new rack 5 35 0.8 to 2.0 modules - Phase 3 l
Total Dose, person-rem - 1.85 to 4.0 l
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Holtec Report HI-981933 9- 7 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION l
10.0 INSTALLATION 10.1 Introduction The installation phase of the DBNPS Unit 1 Cask Pit rack project is executed by Holtec International's Field Services Division. Holtec, serving as the installer, is responsible for performance of specialized services, such as underwater diving and welding operations, as necessary. All installation work at the DBNPS is performed in compliance with NUREG-0612 (refer to Section 3.0), Holtec Quality Assurance Procedure 19.2, DBNPS project specific procedures, and applicable DBNPS procedures.
Crane operators are trained in the operation of overhead cranes per the requirements of ANSI /ASME B30.2, and the plant's specific training program. Consistent with the installer's past practices, a videotape aided training session is presented to the installation team, all of whom are required to successfully complete a written examination prior to the commencement of work.' Fuel handling bridge operations are performed by the DBNPS personnel, who are trained in accordance with DBNPS procedures.
The lifting device designed for handling and installation of the new racks at the DBNPS is engaged and disengaged on lift points at the bottom of the rack. The lifting device complies with the provisions of ANSI N14.6-1978 and NUREG-0612, including compliance with the design stress criteria, load testing at a multiplier of maximum working load, and nondestructive examination of critical welds.
A surveillance and inspection program shall be maintained as part of the installation of the racks.
A set of inspection points, which have been proven to eliminate any incidence of rework or erroneous installation in previous rack projects, is implemented by the installer.
Underwater diving operations are required to remove underwater obstructions, to aid in the rack installation by assisting in the positioning of new rack modules, and to verify installation per Holtec Repon HI-981933 10-1 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION U
p design. The DBNPS procedures for control of diving and radiological controls for diving operations are utilized. The DBNPS procedures are supplemented by the safe-practices guidance provided by the diving company. These documents describe the precautions and controls for dive operations and were developed utilizing OSHA Standard 29CFR-1910, Subpart T.
Holtec International developed procedures, to be used in conjunction with the DBNPS procedures, which cover the scope of activities for the rack installation effort. Similar procedures have been utilized and successfully implemented by Holtec on previous rack installation projects. These procedures are w:itten to include ALARA practices and provide requirements to assure eqmpment, personnel, and plant safety. These procedures are reviewed -
and approved in accordance with DBNPS administrative procedures prior to use on site. The following is a list of the Holtec procedures, used in addition to the DBNPS procedures to implement the installation phase of the project.
A. ' Installation / Handling Procedure: !
This procedure provides overall direction for the handling and installation of the new maximum density fuel storage rack modules in the Cask Pit. This procedure delineates the steps necessary to receive the new maximum density racks on site, the proper method for unloading and I uprighting the racks, staging the racks prior to installation, and installation of the racks. The procedure also provides for the installation of rack bearing pads, adjustment of the rack pedestals and verification of the as-built field configuration to ensure compliance with design documents.
B. Receipt Insocction Procedure:
This procedure delineates the steps necessary to perform a thorough receipt inspection of a new rack' module after its arrival on site. The receipt inspection includes dimensional measurements,
. cleanliness inspection, visual weld examination, and verticality measurements.
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Holtec Report HI-981933 10-2 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION i
. C. Cleaning Procedure:
This procedure provides for the cleaning of a new rack module, if required. The modules are to l
meet the requirements of ANSI N45.2.1, Level B, prior to placement in the Cask Pit. Methods j and limitations on materials to be utilized are provided. ~
D. Pre- and Post-Installation Dran Test Procedure:
These two procedures stipulate the requirements for performing a functional test on a new rack module prior to and following installation into the Cask Pit. The procedures provide direction for inserting and withdrawing an insenion gage into designated cell locations, and establishes an acceptance criteria in terms of maximum drag force.
E.- ALARA Procedure:
Consistent with Holtec International's AIARA Program, this procedure provides guidance to minimize the total man-rem received during the rack installation project, by accounting for time, distance, and shielding F.< Liner Inspection Procedure:
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In the event that a visual inspection of any submerged portion of the Cask Pit liner is deemed necessary, this procedure describes the method to perform such an inspection using an underwater camera and describes the requirements for documenting any observations.
'G. Ieak Detection Procedure:
i This procedure describes the method to test the Cask Pit liner for potential leakage using a !
vacuum box. This procedure may be applied to any suspect area of the Cask Pit liner. f i
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H.- Liner Repair and Underwater Welding Procedure:
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. In the event of a positive leak test result, underwater welding procedures may be implemented which provide for a weld repair, or placement of a stainless steel repair patch, over the area in question. The procedures contain appropriate qualification records documenting relevant '
variables, parameters, and limiting ~ conditions. The_ weld procedure is qualified in accordance with AWS D3.6-93, Specification for Underwater Welding or may be qualified to an alternate j
l code accepted by the DBNPS and Holtec International. j i
l 10.2 Rack Arrangement l The final rack arrangement allows for a total of four freestanding Holtec racks in the Cask Pit, l which provides a total of 289 storage locations. Two new fuel storage racks were installed in the )
Cask Pit to add 153 fuel storage cells, in April 1999. These two rack modules were installed as a i plant modification after evaluation in accordance with 10CFR50.59 demonstrated that -
installation of the empty racks did not involve an unreviewed safety question. The installed racks will remain unused until a license amendment application is approved by the NRC. The two racks were placed in the Cask Pit for the remaining duration of the DBNPS Fuel Cycle 12, which is scheduled to be completed in April 2000. One of these racks provides sufficient storage capacity for full core offload capability for the remainder of Fuel Cycle 12 and the ten year in j service inspection (ISI) of the reactor vessel. The ten year ISI is required to be completed during )
l the Cycle 12 Refueling Outage. The two installed racks also provide full core offload capability . i during Fuel Cycle 13, which is scheduled to occur between May,2000 and April,2002. The -
remaining two racks, consisting of 136 cells, will be installed into the Cask Pit in a future l campaign, during Fuel Cycle 13, to support fuel movements required for a full re-racking of the l SFP. The SFP re-racking is expected to take place during Fuel Cycle 13. A schematic plan view depicting the Cask Pit with the two newly installed maximum density racks can be seen in Figures 1.2. Figure 1.3 depicts the Cask Pit layout with all four maximum density fuel storage r :
< racks installed.
l Holtec Report HI-981933 10-4 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION L
l 10.3 Cask Pit Survey and Inspection A Cask Pit survey was performed to determine if any items attached to the liner wall or floor of the Cask Pit would interfere with the placement of the new racks or prevent usage of any cell locations subsequent to installation. This survey determined that an unused light pole support bracket on the south wall of the Cask Pit must be removed. This bracket was originally mounted on a one-halfinch mounting plate attached to the Cask Pit liner. The bracket was removed from the mounting plate for installation of the first two racks, leaving the mounting plate intact, without affecting the liner. On the north wall of the Cask Pit is another light pole mounting bracket that will be similarly removed from it's mounting plate before installation of racks N3 i and N4. Also on the north Cask Pit wall is a fuel handling bridge load test fixture that will be removed and relocated. Finally, the northwest corner of the Cask Pit floor has a sump for l draining the Cask Pit. The drain piping enters the pit through the west wall of the pit, takes a 90 degree bend downward, and extends into this sump. To install the N4 rack, this drain line l protrusion into the pit and associated supports will have to be removed. The pipe will be cut off parallel to the west wall such that a flange can be welded to the pipe as it enters the pit. The piping will be flanged so that the drain pipe extension can be temporarily removed to accommodate rack N4 and reinstalled in the future after removal of the N4 rack. All four rack I
modules will be eventually relocated into the SFP during the re-racking on the entire SFP. A stack of solid, stainless steel plates will be placed in the sump on which one leg of the N4 rack I will rest. The removal of these interferences will involve underwater diving and mechanical
- cutting operations.
l 10.4 Cask Pit Cooling and Purification 10.4.1 Cask Pit Cooling l There is no forced cooling in the Cask Pit. When fuel is transferred into the Cask Pit, the water l in the pit will be cooled by natural circulation mixing with the SFP water through the open gate.
The SFP water temperature is maintained by forced circulation cooling. During any installation Holtec Report HI-981933 10-5 80284 i
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. of racks in the Cask Pit, there will be no fuel in the Cask Pit, and the gate between the SFP and l
l Cask Pit will be installed to prevent the diver from entering the SFP. Since there is no forced cooling in the Cask Pit and the pit will contain no fuel, it is not anticipated that any rack installation activities will require the temporary shutdown of the SFP cooling system.
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1 10.4.2' Purification l A portable vacuum system may be employed to remove extraneous debris, reduce general
- contamination levels prior to diving operations, and to assist in the restoration of Cask Pit clarity i i
following any installation processes.
l 10.5' Fuel Movement i
Necessary fuel movements are performed prior to Cask Pit rack installation activities. Fuel movement operations are conducted in accordance with DBNPS procedures. Any fuel stored in the Cask Pit racks for the Cycle 12 refueling outage will be returned to the SFP prior to installation of rack modules N3 and N4.
10.6 Installation of New Racks Installation of the new high' density racks, supplied by Holtec International,' involves the following activities. The racks are delivered in the horizontal position. A new rack module is
- . removed from the shipping trailer using a suitably rated crane, while maintaining the horizontal configuration. The rack is placed on the up-ender and secured. Using two independent overhead hooks, or a single overhead hook and a spreader beam, the module is up-righted into a vertical
!- position.
The new rack lifting device is engaged in the lift points at the bottom of the rack. The rack is then transported to a pre-leveled surface where, after leveling the rack, the appropriate quality
_ control receipt inspection is performed.
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To address ALARA considerations, fuel in the adjacent SFP may be moved away from the gate l area in preparation for rack installation. Additionally, the Cask Pit floor is inspected and any l debris, which may inhibit the installation of bearing pads, is removed.
After Cask Pit floor preparation, new rack bearing pads are positioned on the pit floor before the module is lowered into the pit. The new rack module is lifted with the Spent Fuel Cask Crane (SFCC) and transported along the pre-established safe load path. The rack module is cautiously lowered into the Cask Pit unto the bearing pads using the SFCC. . A temporary hoist, with an appropriate capacity, is attached to the SFCC for installation in order to eliminate contamination of the main hook during lifting operations in the Cask Pit.
l Elevation readings are taken to confirm that the module is level. In addition, rack-to-rack and rack-to-wall off-set distances are also measured. Adjustments are made as necessary to ensure compliance with design documents. The lifting device is then disengaged and removed from the Cask Pit under Health Physics direction. Post-installation free path verification is performed using an inspection gage in order to ensure that no cell location poses excessive resistance to the insertion or withdrawal of a fuel assembly. This test confirms final acceptability of the installed rack module.
10.7 Safety. Health Physics, and ALARA Methods 10.7.1 Safety l
i During the installation phase of the Cask Pit rack project, personnel safety is of paramount importance, outweighing all other concerns. All work shall be carried out in compliance with applicable approved procedures.
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10.7.2 Health Physics Health Physics is carried out per the requirements of the DBNPS Radiation Protection Program.
10.7.3 ALARA The key factors in maintaining project dose As low As Reasonably Achievable (ALARA) are time, distance, and shielding. These factors are addressed by utilizing many mechanisms with respect to project planning and execution.
Time Each member of the project team is trained and provided appropriate education and understanding of critical evolutions. Additionally, daily pre-job briefings are employed to acquaint each team member with the scope of work to be performed and the proper means of executing such tasks. Such pre-planning devices reduce worker time within the radiologically controlled area and, therefore, project dose.
Distance t
. Remote tooling such as lift fixtures, pneumatic grippers, a support leveling device and a lift rod j
disengagement device have been developed to execute numerous activities from the Cask Pit surface, where dose rates are relatively low. For those evolutions requiring diving operations, diver movements shall be restricted to the empty Cask Pit by installation of the gate between the SFP and the Cask Pit. If necessary, additional fuel in the adjacent SFP may be moved to satisfy ALARA principles.
Holtec Report HI-981933 - 10-8 80284 SHA'JdD REGIONS DESIGNATE PROPRIETARY INFORMATION
j Shielding During the course of the rack installation, the concrete wall between the SFP and the Cask Pit, and the water in the Cask Pit provides shielding, If necessary, additional shielding may be utilized to meet ALARA principles. j i
10.8 Radwaste Material Control '
Radioactive waste generated from the rack installation will be controlled in accordance with established DBNPS procedures.
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11.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT 11.1 ' Introduction Article V of the USNRC OT Position Paper [l 1.1] requires the submittal of a cost / benefit analysis for a fuel storar,e capacity enhancement. This section provides justification for selecting installation of racks in the Cask Pit as the most viable alternative.
I 11.2' Imperative for Rack Installation The DBNPS lost full core offload capability (FCOC) in April 1998, during the refueling outage conducted after Fuel Cycle 11. Although FCOC is neither a license condition nor commitment for the DBNPS, it is considered a prudent operating practice. In January of 1996, the DBNPS completed storage of 72 spent fuel assemblies in the certified NUHOMS@ dry spent fuel storage system in accordance with the requirements of 10CFR72 Subpart K. After the vendor for the NUHOMS@ system temporarily stopped production, a decision had to be made to implement another spent fuel storage plan.
At the present time, the SFP has 114 open storage cells. The DBNPS reactor core contains 177 fuel assemblies, and is currently operating without FCOC. The present fuel storage rack l arrangement contains 735 storage cells and there is no available area for installation of additional racks in the SFP. The SFP currently contains 601 irradiated fuel assemblies, one dummy fuel assembly, and 2 inaccessible storage cells. An additional 17 storage cell locations are used or reserved for failed fuel assemblies, surveillance specimen storage, control component handling containers, abandoned fuel assembly cages from fuel assembly reconstitution campaigns, and radioactive trash containers. Control components are generally stored within irradiated fuel assemblies.
, Placing two fuel rack modules in the Cask Pit adds 153 fuel assembly storage locations. This addition will serve to regain FCOC during Fuel Cycle 12. Cycle 12 started in May of 1998, and Holtec Report HI-981933 11-1 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION i
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L is scheduled to be complete in April of the year 2000. FCOC at the end of Fuel Cycle 12 is required to allow completion of the 10-year reactor vessel in-service inspection (ISI). As the 10-year ISI is a regulatory requirement for operation, without additional storage capacity, operation of the DBNPS can not continue. Approximately 72 fuel storage locations will be used for the next refueling. As a result,195 storage cells should be available to maintain FCOC during Fuel Cycle 13 which is scheduled to begin in May,2000 and end for the thirteenth refueling outage in April of the year 2002.
The remaining two rack modules (total of four) addressed by this license amendment request are required to provide fuel storage during a future re-racking of the entire SFP, which is scheduled
. to take place during Fuel Cycle 13. As part of the SFP re-racking, the four Cask Pit rack modules will be placed in the SFP near the end of the re-racking sequence.
11.3 Appraisal of Alternative Options
' Adding fuel storage space to the DBNPS Cask Pit is the most viable option for temporarily increasing spent fuel storage capacity.
The key considerations in evaluating the alternative options included:
Safety: Minimize the risk to the public Economy: Minimize capital and O&M expenditures Security: Protection from potential saboteurs, natural phenomena Non-intmsiveness: Minimize required modifications to existing plant systems !
1 Maturity: Extent of industry experience with the technology 1 ALARA: Minimize cumulative dose !
Schedule: Minimize time to regain full-core offload capability Risk Management: Maximize probability of completing the expansion to support fuel storage needs Holtec Report HI-981933 11-2 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
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Rod Consolidation i L
Rod consolidation involves disassembly of spent fuel, followed by the storage of the fuel rods j i
from two assemblies into the volume of one, and the disposal of the fuel assembly sk.eleton . I i I l
outside of the pool (this is considered a 2:1 compaction ratio). The rods are stored in a stainless l 1 steel can that has the outer dimensions of a fuel assembly. The can is stored in the spent fuel
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racks. This technology is still in its developmental infancy and thus, based on the
. aforementioned DBNPS schedule, is not a viable option based on the time frame. l On-Site Dry Cask Storane Dry cask storage is a method of storing spent nuclear fuel in a high capacity container. The cask l provides radiation shielding and passive heat dissipation. Typical storage system capacities for
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PWR fuel range from 21 to 37 assemblies that have been removed from the reactor for at least five years.
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I In the early 1990s, Toledo Edison made the decision to reclaim some of the DBNPS SFP storage using a dry fuel storage system. - In January 1996, seventy-two spent fuel assemblies were loaded into three Dry Shielded Canisters and were placed in dry fuel storage utilizing the certified
- NUHOMS' system, in accordance with 10CFR72.214, Certificate Number 1004. Changes i i
within the dry spent fuel storage industry have caused cost increases. The contracted supplier of j the NUHOMS system voluntarily stopped fabrication activities and was unable to provide additional storage systems within an acceptable schedule. Further use of this technology was re-evaluated and determined not to be the best choice for future storage expansion at the DBNPS. ,
l This decision was based on economics, schedule, and risk management. j t
i Holtec Report HI-981933 11-3 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION
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Other Storare Options I
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Other options such as Modular Vault Dry Storage, Horizontal Silo Storage, and a new Fuel Storage Pool are overly expensive as compared to placing racks in the Cask Pit. Due to the
! - complexity of implementation, these options could not meet the required schedule for regaining and maintaining full-core offload capability. l 11.3.1 Alternative Option Summarv An estimate of relative costs in 1998 dollars for the aforementioned options is provided in the
] followingi Cask Pit Rack Expansion: $1-3 million
- Horizontal Silc: $35-45 million j J
~ Rod consolidation: $25 milhon a Metal cask (MPC): $68-100 million Modular vault: $56 million i New fuel pool: $150 million The above estimates are consistent with estimates by EPRI and others [11.2,11.3].
To summarize, based ~on the required short time schedule, the status of the dry spent fuel storage
. industry, and the storage expansion costs, the most acceptable alternative for increasing the on-
. site spent fuel storage capacity at the DBNPS is expansion of the wet storage capacity. First,
~t here are no commercial independent spent fuel storage facilities operating in the United States.
'Second, the adoption of the Nuclear Waste Policy Act (NWPA) created a de facto throw-away nuclear fuel cycle. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium, reprocessing represents an added cost for the nuclear fuel cycle which already includes the NWPA Nuclear Waste Fund fees. In any event, there are no domestic Holtec Report HI-981933 11-4 80284
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reprocessing facilities. Third, at over $% million per day replacement power cost, shutting down the DBNPS is many times more expensive than addition of high density racks to the existing SFP Cask Pit and the future re-racking of the SFP.
I1.4 Cost Estimate The plant modification proposed for the DBNPS fuel storage expansion utilizes freestanding, high density, poisoned spent fuel racks in'the Cask Pit. The engineering and design is completed for full racking of the Cask Pit. 'As stated in section 11.2, the first two racks placed in the Cask Pit will provide full-core offload capability through Fuel Cycle 13. This will allow time for the complete re-racking of the SFP during Cycle 13.
The total capital cost is estimated to be approximately $1.5 million as detailed below.
Engineering, design, project management: $1/2 million Rack fabrication: $1/2 million Rack installation: $1/2 million As described in the preceding section, other fuel storage expansion technologies were evaluated prior to deciding on the use of Cask Pit racks. Storage rack capacity expansion provides a cost advantage over other technologies.
Holtec Report HI-981933 11-5 80284
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l 11.5 Resource Commitment i
The expansion of the DBNPS Spent Fuel Pool capacity via the Cask Pit is expected to require the
- following primary resources:
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l Stainless steel: 18 tons l
Boral neutron absorber: 2 tons, of which 1.5 tons is Boron Carbide powder and 0.5 tons are aluminum.
The requirements for stainless steel and aluminum represent a small fraction of total world output of these metals (less than 0.001%). Although the fraction of world production of Boron Carbide required for the fabrication is somewhat higher than that of stainless steel or aluminum, it is unlikely that the commitment of Boron Carbide to this project will affect other alternatives.
Experience has shown that the production of Boron Carbide is highly variable, depends upon need, and can easily be expanded to accommodate worldwide needs.
I1.6 Environmental Considerations Due to the additional heat-load arising from increased Spent Fuel Pool' inventory, the anticipated maximum hulk pool temperature will increase by about 4"F. t at the time when the pool's capacity is exhausted. The increased bulk pool temperature will result in an increase in the pool water evaporation rate. This increase has been determined to increase the relative humidity of the Fuel Building atmosphere by less than 25 percent relative humidityt . This increase is within the capacity of both normal and the ESF Ventilation System. The net result of the increased heat loss and water vapor emission to the environment is negligible.
t These numbers are based on more than doubling the amount of fuel in the Spent Fuel Pool by re-racking the entire pool. This will be very conservative for the heat load added by placing fuel in Cask Pit racks Holtec Report HI-981933 - 11-6 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION l
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11,7 References l
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[11.1] OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, USNRC (April 1978).
[l1.2] Electric Power Research Institute, Report No. NF-3580, May 1984.
[11.3] " Spent Fuel Storage Options: A Critical Appraisal", Power Generation Technology, Sterling Publishers, pp. 137-140, U.K. (November 1990).
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i Holtec Report HI-981933 11-7 80284 SHADED REGIONS DESIGNATE PROPRIETARY INFORMATION