ML20126F411

Summary of 921215 Meeting W/Omaha Public Power District in Rockville,Md Re Spent Fuel Pool Rerack for Fort Calhoun Station

ML20126F411
Person / Time
Site:	Fort Calhoun
Issue date:	12/23/1992
From:	Steven Bloom Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9212300298
Download: ML20126F411 (46)

Text

__ _ _ _ _ . _ . _ _ _ _ _ . _ . _ _ _ __ _ . . . , _ _ _ ._

s- o, December 23, 1992-Docket 50-285 LICENSEE:- Omaha Public Power District FACILITY: -Fort Calhoun Station, Unit 1

SUMMARY

OF MEETING HELD ON DECEMBER 15, 1992, WITH OMAHA PUBLIC POWER DISTRICT TO DISCUSS SPENT FUEL POOL RERACK FOR_ FORT CALHOUN STATION The staff met with Omaha Public Power District (0 PPD) on December 15, 1992, at.

One White Flint-North-to discuss-the submittal for NRC approval of OPPD's Spent Fuel Pool Rerack for Fort Calhoun Station. Enclosure 1 contains a list of meeting attendees.

The meeting had two parts. In the first part, OPPD presented the background of the reracking, the present and future rack layout, installation plans, project millstones, waste disposal, and license changes. The second part was >

presented by Holtec and included a technical discussion on the rack design and ,

fabrication, heavy loads considerations, criticality considerations, thermal-hydraulic considerations, seismic / structural considerations, and a review of analyses performed. Enclosure 2 consists of the slides that were discussed in OPPD's presentation. ,

OPPD requested the final safety evaluation (SE) by May 1993.

ORIGINAL SIGNED BY:-

Steven D.-Bloom,_ Project Manager Project Directorate IV-1 Division of Reactor Project - III/IV/V -

Office of Nuclear Reactor Regulation

Enclosures:

1. List of attendees

2. OPPD's slides-cc' w/ enclosures:

See next page:

DISTRIBUTION:

1 Docket File -NRC & Local PDRs PD4-1 Reading l TMurley/FMiraglia JPartlow JRoe MVirgilio' JLarkins SBloom PNoonan OGC -EJordan DJeng Skim . HRathbun-VBeaston KParczewski 30005 u DCarter JMedoff.

JMinns LKopp.

SJones . .YKim; _ ACRS (10)-

JMitchell ABBeach . ETrottier- (14D1) 0FC LA:PD4 'DD PM FD4-d - D:PD4 '

PNoon$n M SdEoYm:$k JL[Jf!n[s NAME - P,N ~ ~

.y tm DATE )QM2/92- /2./2J/92 /L/2J/92 S- '

0FFICIAL RECORD COPY Document Name: FCDec15.mts

(

9212300298 92'1323

• L 1

fDR ADOCK 05000285 I 1 PDR 6L-

. -_ - - .~ _

1 *

1. 'o UNITED STATES

, g 8 n NUCLEAR REGULATORY COMMISSION

- W ASHINGTON, D. C. 20655 k*...+,o,I 8 December 23, 1992 t

a

Docket 50-285 1

LICENSEE: Omaha Public Power District FACILITY: Fort Calhoun Station, Unit 1

SUMMARY

OF MEETING HELD ON DECEMBER 15, 1992, WITH OMAHA PUBLIC POWER DISTRICT TO DISCUSS SPENT FUEL POOL RERACK FOR FORT CALHOUN STATION The staff met with Omaha Public Power District (OPPD) on December 15, 1992, at i One White Flint North to discuss the submittal for NRC approval of OPPD's i Spent Fuel Pool Rerack for Fort Calhoun Station. Enclosure 1 contains:a list l of meeting attendees.

The meeting had two parts. In the first part, OPPD presented the background of the reracking, the present and future rack layout, installation plans, i project millstones, waste disposal, and license _ changes. The second part was presented by Holtec and included a technical discussion on the rack design and

fabrication, heavy loads considerations, criticality considerations, thermal-hydraulic considerations, seismic / structural considerations, and a review of _

analyses performed. Enclosure 2 consists of the slides that were discussed in

OPPD's presentation.

OPPD requested the final safety evaluation-(SE) by May 1993.

.Y7 -

/

j Steven D. Bloom, Project Manager Project Directorate IV-1 Division of Reactor Project - III/IV/V

~

Office of Nuclear Reactor Regulation i

Enclosures:

1. List of attendees

2. OPPD's' slides

~

cc w/ enclosures:

See next-page 4

c l- . .

Omaha Public Power District Fort Calhoun Station, Unit No.1 i

l CC:

i

Harry H. Voigt, Esg.

LeBoeuf, Lamb, Leiby & MacRae

- 1875 Connecticut Avenue, NW Washington, D.C. 20009-5728 i Mr. Jack Jensen, Chairman l Washington County Board i of Supervisors

! Blair, Nebraska 68008

} Mr. Raymond P. Mullikin, Resident Inspector l U.S. Nuclear Regulatory Commission i Post Office Box 309 Fort Calhoun, Nebraska 68023

l. Mr. Charles B. Brinkman, Manager i Washington Nuclear Operations

[

I Combustion Engineering, Inc.

12300 Twinbrook Parkway, Suite 330 l Rockville, MD 20852

i. .. .

i Regional Administrator, Region IV i U.S. Nuclear Regulatory Commission i 611 Ryan Plaza Drive, Suite 1000 i Arlington, Texas 76011-

Harold Borchert, Director j Division of Radiological Health

~ Nebraska Department of- Health j 301 Centennial Hall, South.

i Post Office Box 95007 i Lincoln, Nebraska 68509 i Mr, T, L. Patterson, Manager 4

Fort Calhoun Station i Post Office Box 399 .

j Fort-Calhoun, Nebraska 68023 Mr. W. Gary Gates .

Division Manager - Nuclear Operations 4

Omaha Public Power District 444 South 16th Street Mall

- Mail Stop 8E/EP4' Omaha, Nebraska 68102-2247 a

n i

i i

Enclosure 1 MEETING BETWEEN NRC STAFF AND'0MAHA PUBLIC POWER DISTRICT

- FORT CALHOUN STATION, UNIT 1 SPENT FUEL-POOL RERACK December 15, 1992-MEETING PARTICIPANTS Magi ORGANIZATION D. Jeng NRR/DET/ESGB S. Kim NRR/DET/ESGB <

Y. Kim NRR/DET/ESGB J. Larkins NRR/PDIV-1 S. Bloom NRR/PDIV-1 H. Rathbun- NRR/PDIV-1 V. Beaston NRR/DRSS/PRPB D. Carter NRR/DRSS/PRPB/FRPS J. Minns -

-NRR/DRSS/PRPB K. Parczewski NRR/DE/EMCB J. Medoff NRR/DE/EMCB L. Kopp NRR/DSSA/SRXB S. Jones NRR/DSSA/SPLB .

S. Gambhir OPPD/ Division Manager - Production Engineering R. Phelps OPPD/ Manager - Design Engineering Nuclear R. Short OPPD/ Manager - Nuclear Licensing / Industry Affairs J. Herman OPPD/ Supervisor - Nuclear Licensing R. Lewis OPPD - Engineer-J. Bostelman OPPD - Engineer S. Bottum OPPD - Engineer K. Singh -Holtec Y. Wang Holtec C. Corbin -TU Electric - Comanche Peak D. Teague Winston & Strawn

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Enclosure 2 FORT CALHOEN STATION

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OMAHA PUBLIC POWER DISTRICT SPENT FUEL POOL RERACKING DECEMBER 15, 1992 W

Omaha Public Power District

4

Presentation to l The United States Nuclear Regulatory Commission l

! for i

i l Fort Calhoun Station

Omaha Public Power District l

[~ Spent Fuel Pool

[ Reracking i

l December 15,1992

~_ i

SPENT FUEL POOL RERACKING PRESENTATION AGENDA

1. Introduction - S. Gambhir/R. Phelps

11. OPPD Presentation

+ Background J. Bostelman

+ Rack Layout S. Bottum

+1nstallation Plans - S. Bottum

+ Milestones - S. Bottum

+ Waste Disposal - J. Bostelman

+ License Changes - J. Bostelman III. Holtec Presentation - Kris Singh

+ Technical Discussion on Rack Design and Fabrication

+ Heavy Loads Considerations

+ Criticality Considerations

+ Thermal Hydraulic Considerations

+ Seismic / Structural Considerations

+ Review of Analyses Performed IV. Closing - S. Gambhir i

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Fort Calhoun Station l

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Reracking I

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Background

u Approximately 44 assemblies are discharged every refueling outage a Discharged assemblies are stored onsite a Space available for 728 assemblies

, e Original pool storage 177 spaces e 1976 Rerack 483 spaces e 1984 Rerack 728 usable spaces a- Full Core Reserve Requirement is 133 assemblies

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! Storage Reauirements d

i m Spaces required for assembly storage

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j End of 1992 Refueling Outage 529+133=662 3

End of1993 Refueling Outage 570+133=703 t

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includes provisions for Full Core Off Load i a Additional space required prior to the j 1995 Refueling Outage

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! Ontions Evaluated i

m Rcrack l

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e Dry Cask Storage l

a Fuel Rod Consolidation i a Rerack with Future Capability for l

1 Fuel Rod Consolidation i

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- 5

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i Selection Criteria-Fuel Storage _

l m Accommodate average fuel enrichment of

4.2 w/o U235 t

! u Accommodate possibility of Federal 4

Repository available for fuel storage in near

term a Maximize onsite fuel storage capacity i

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Decision to Rerack u The best economical option a Expands storage capacity to 1083 storage cell spaces a Alternative onsite storage required after the year 2006 m Racks analyzed for possible future storage of consolidated fuel l

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! Cost Benefit Analysis 4

! Performed in 1990 t

i u Evaluated Technical and Economic Merits a Calculated Net Present Values of Options l

l m Utilized a probability of success weighting I decision tree

m A scenario which considered 1050 fuel l

l storage cells, net value rerack 21.8 million versus 23.9 million dry storage 4

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BASE CASE EXPECTED NET PRESENT VALUE j (S MILLIONS:1990)

J MRS

$4.3 10%

I Rerack l Rod Consolidate

". $26.4 i $?'.1 40%

Dry Storage 50%

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$12.8 10% --

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l Dry Storage

! $27.0 Dry Storage

$28.6 90%

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SPENT FUEL RACKS.

Current Racks Proposed Racks (GCA/ par) (Holtec)

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12 Free Standing 11 Free Standing Rack Modules Rack Modules Bearing Pads Same l

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) Between Rack

) Pedestal Base i & Pool Slab / Liner i

} Seismic Analyses Same i Based on USAR i

Appendix F l

j Criteria i

l Increased Loads Pool Structure

. Reviewed against Requalified using Existing Design S.R.P. 3.8.4

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l Calculations 3

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N - PROPOSED LAYOUT 399-

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-FICURE 2.1.2. PROPOSEB N0DULAR LAYOUT N

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e Fort Calhoun Station  !

Spent Fuel Pool Reracking I

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Installation Plans

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Installation Plans s Single Failure Proof Auxiliary Building Crane to be used for Heavy Load handling uSingle Failure Prooflift rigs to_be used for old-and new racks mSame physical attributes to lift rigs used at Millstone, Vogtle, Indian Point, Hope Creek, Zion, TMI, and FitzPatrick mWritten Procedures and administrative controls for heavy lifting will be used nFuel Shuffles and rack changeout scheme. l to preclude rack movement directly over spent I fuel.

- -15

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I Installation Plans (cont.)

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4 ALARA

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l eProcedures with full consideration of I ALARA principles i

i i uPrimary approach will be remote tooling I from pool perimeter and spent fuel handling machine f

i mDiver contingency for the removal of

! underwater interferences i

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-l l Milestones l

! l l 12/92- Submittal to NRC i

! 12/15/92-NRC/OPPD Meeting i

1/93 - Preliminary Design Change i

Packageissued 4

05/93- Requested SER from l NRC L

j 06/93- Rack Fabrication Start i

l 7/93 - Final Design Change Package issued I

l 02/94- Site Mobilization l

03/94- Rack Receipt Start 03/94- Rack Removal / Installation Start 08/94- Rack Installation Complete 08/94-95 Rack Disposal

.F*

- 18 l

Fort Calhoun Station .

Spent Fuel Pool Reracking

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! Waste Disposal i

l m Disposal of 160,000 lbs. of l

l Material (Stainless Steel) e l s Previous Experience with l .

Rack Decontamination-l Disposal

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.e j}(p3w A, O, l VI. License Changes

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! Proposed License Changes s Upgrade to 4.2 w/o U235 i

l m Require at least 100 ppm Boron i

m. Install Boral poison material, .

l and incorporate surveillance

! requirements l

f J

i

-, 22

i PROJECT OUTLINE _

u Discrete zone two region storage 8 All design & manufacturing technologies have over a decade of background experience and application in over two dozen plants e All computer codes utilized in the safety analysis effort have a long history of usage Min. # of Dockets Where Code Name Application Applied

1. CASMO Criticality Eval. 25

2. KENO Criticality Eval. 25 3.THERPOOL Thermal-Hydraulics 25 ,

4. ONEPOOL (BULKTEM) Thermal-Hydraulics 25

5. TBOIL Thermal-Hydraulics 12 6.DYNARACK Rack Dynamics 25 (DYNAMO) 7.GENEQ Time-History Generation 25

8. ORIGEN Radiological 25 9.ANSYS Pool Structural- 12 23 ieps

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I i i i Table 1.1.3 j i

i j RACK MODULE DATA, EXISTING AND PROPOSED RACKS i

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PROPOSED MAXIMUM g EXISTING PACKS DENSITY PACKS

} '

1 Humber of cells 729 1083 l

Humber of modules 12 11 i Neutron Absorber Boraflex Boral k

Approximate cell 9.935 8.652 (non-flux trap) pitch, inches 9.821 (flux trap in

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E-W directions) i 10.363 (flux trap in-

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Na B O 00 O z i km s* t, 4W FIGURE 2.1.1 LAYOUT OF THE EXISTING RACKS IN THE FCS POOL sh

Table 1.1.1B PROJECTED DISCl!ARGE SCHEDULE Assumed rull Month / Power Number of Cumulative Cycle Year of operation Assemblies Assemblies Number Discharce_ Qay1 Discharced Discharaed 14 8/1993 1575 45 574 15 2/1995 1575 45 619' 1575 45 664 16 8/1996 17 2/1998 1575 45 709 754

^

18 8/1999 1575 19 2/2001 1575 45 799 20 8/2002 1575 45 844 21 2/2004 1575 45 889 22 8/2005 1575 45 934 23 2/2007 1575 45 979 24 8/2008 1575 45 1024 i 25 2/2010 1575 45 1069 26 8/2011 1575 45 1114 27 2/2013 1575 45 1159

28 8/2014 1575 45 1204 i

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Table 1.1.2 i AVAILABLE STORAGE IN -THE TCS Pool WITH PRESENT TUEL Eoc LICENSED CAPACITY AFTER RERACXING CYCLE g (729 CM T ti (1081 Crf.T n)  !

13 1992 200 559  ;

14 1993 155 514 15 1995 110 469-16 1996 424-17 1998- 379 la 1999 334 -

19 2001 239 20 2002 244 21 2004 199 22 2005 154 23 2007 109' i

l Loss of full core reserves.

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tI: 0 '

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1

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' !O On t t t t i i O I i liI 6 i i I ,

I l- i I {

O ' 1 j 0 0 0 1;

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t

- t U

~ O 00 O l _

- I L l ,

~

,- 3 L

M j E FIGURE 2.1.2 4

i PROPOSED MODULAR LAYOUT

.m 28-

s a _1 -

• Table 2.1.1 MODULE DATA FOR FCS MAXIMUM DENSITY RACKS NUMBER MODULE ENVELOPE SHIPPING OF CELLS SIZE FINCHES) WEIGHT MODULE MODULE N-S E-W DIRECTION PER RACK H-S E-W (LBS) TYPE I.D. DIRECTION 8 80 102.06 77.4 16000 Reg. I Al 10 8 80 102.06 77.4 16000 Reg. I A2 10 108 104.16 78.12 15200 Reg. II B1 12 9 12 .9 108 104.16 78.12 15200 Reg. II B2 9 90 86.8 78.12 12600 Reg. II G1 10 G2 10 9 90 86.8 78.12 12600 Reg. II 11 9 99 95.48 69.44 13900 Reg. II C

11 8 88 95.48 69.44 12400 Reg. II D

10 10 100 86.8 86.8 14000 Reg. II E

12 10 120 104.16 86.8 16800 Reg. II F1 12 10 120 104.16 86.8 16800 Reg. II F2 U-

i 1

j  !

i i

i l

l j

i Table 2.2.1 l

TOTAL CELL DATA TOTAL NUMBER 1 RECTON NUMBER OF RAcPS or CELLS I (Flux-Trap) 2 160 II (Non-Flux Trap) 9 923 4

4 CRAND TOTAL: 1083 l i r

)

2-12 e-

' - 30

-v w-

Table 2.2.2 COMMON MODtJLE DATA Storage cell inside dimension (nominal): 8.46 inch Storage cell height 161 inch

, (above the baseplate):

Baseplate thickness: 0.75 inch support leg height: 5-1/4 inch (nominal)

Support leg type Remotely adjustable legs with lateral gussets.

Number of support legs: 4 (minimum)

Remote lifting and handling provision: Yes Poison material: Boral Poison length: 128 inch Poison width: 7.25 inch call Pitch: 8.652 inch (Region II) 9.821 inch (Region I, in E-W direction) 10.363 inch (Region I, in N-S direction) 2-13

(; 31

l i

i 4

i l

4 l MATERIALS OF CONSTRUCTION i

4 f CELLULAR SA240-304 i STRUCTURE:

! INTERNALLY 4 THREADED l MEMBERS: SA240-304 i EXTERNALLY SA564-630

THREADED (Precipitation

SPINDLE: Hardened)

I NEUTRON Boral ABSORBER: (Used in over 30-rerack and cask projects) 1

-Il - 32

!  ? ,

l l BORAL EXPERIENCE LIST I

U.- S. PLANTS 1

! PRESSURIZED WATER REACTORS ,

UTILITY HFG. YEAR f,LMil l

Bellefonte 1,2 Tennessee Valley Authority 1981 l; Indiana & Michigan Electric 1979 Donald C. Cook Indian Point 3 NY Power Authority 1987 i 1977 Maine Yankee Maine Yankee Atomic Power Salem 1,2 Public Service Elec. & Gas 1980 l Tennessee Valley Authority 1979 i Sequoyah 1,2 Yankee Rowe Yankee Atomic Power 1964/1983 i 1980 Commonwealth Edison Co.

1988 Zion Byron 1}2

,2 Commonwealth Edison Co.

1988 Comonwealth Edison Co.

i Braidwood 1,2

) Yankee Rowe Yankee Atomic Electric 1988 Three Mile Island I GPU Nuclear 1990 Shearon Harris Pool B Carolina Power & Light 1991 BOILING WATER REACTORS h.

UTILITY MFG. YEAR ELASI l

Browns Ferry 1,2,3 Tennessee Valley Authority 1980 Brunswick 1,2 Carolina Power & Light 1981 Clinton Illinois Power 1981 Cooper Nebraska Public Power 1979 4 Dresden 2,3 Comonwealth Edison Co. 1981

! Daane Arnold Iowa Elec. Light & Power - 1979

J. A. Fitz)atrick NY Power Authority 1978 E.1 Hatc1 1,2 Georgia Power 1981

Hope Creek Public Service Elec & Gas 1985

Humboldt Bay Pacific Gas & Electric 1986 i Lacrosse Dairyland Power 1976 Limerick 1,2 Philadelphia Electric 1980 Monticello Northern States Power 1978 Peachbottom 2,3 Philadelphia Electric 1980 s

2 1

Cleveland Elec. Illuminating 1979 Perryim Pilgr ,2 - -

Boston Edison 1978 3 Susquehanna 1,2 Pennsylvania Power & Light 1979

' Vermont Yankee- Vermont Yankee Atomic Power- 1978/1986 Hope Creek Public Service Elec. & Gas 1989 33 r

I

k j BORAL EXPERIENCE LIST 1

l OVERSEAS j l

l 1

I FRANCE i

j 12 PWR Plants Electricite De France 1

i

! SOUTH AFRICA i

i Koeberg 1,2 Escom l

2 SWITZERLAND i

j Beznau 1,2 Nordostschweizerische Kraftwerke AG

]

j Gosgen Kernkraftwerk Gosgen-Daniken AG i

i TAIWAN i

j

Chin-Shan 1,2 Taiwan Power Company

! Kuosheng 1,2 Taiwan Power Company l

! MEXICO Laguna Verde Comision Federal De Electricidad Units 1 & 2 b

2 I

34 i

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4 e

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N y -l WELD SEAM

/

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=:

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El.EVATION VIET OF A REC 10N 1- RACI l SH0ilNG TY0 STORAGE CELLS 1.

", s-is 39

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4 t.10N-FLUX TRAP CONSTRUCTION) 1'

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. 42 J

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~

1 l

i l HEAVY LOAD CONSIDERATIONS 1

AUXILIARY BUILDING CRANE (75 TON MAIN HOOK) IS SINGLE FAILURE PROOF.

10 TON AUXILIARY HOOK IS NOT SINGLE FAILURE PROOF.

1 LIFT RIG IS ALSO SINGLE FAILURE PROOF. RIG IS IDENTICAL IN DESIGN TO THOSE USED IN OVER A DOZEN HOLTEC RERACK PROJECTS. ,

DEFENSE-IN-DEPTH PROCEDURES TO  ;

FURTHER ENHANCE SAFETY OF THE RERACK OPERATION.

e

". 434 L ._ _ _ _. . _ _ . . . ._ .._ __.__ _, . . _ _ _ ._

i . * *.

l j

Table 2.5.1 I

HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG 0612) criterien ceme11ance j 1. Are safe lead paths defined for- Yes

< the movement of heavy loads to j minimize the potential of impact,

if dropped on irradiated fuel?

i 2. Wil2 procedures be developed to - Yes  ;

} cover identification of required

equipment, inspection and acceptance 1 criteria required before movement

of load, steps and proper sequence i for handling the load, defining the j safe load paths, and special j precautions?

! 3. Will crane operators be trained Yes j and qualified?

i

4. Will special lifting devices meet. Yes .

j the guideline of ANOI 14.6-19787

5. Will non-custom lifting devices Yes

' be installed and used in accordance with ANSI B30.9-19717

6. Will the cranes be inspected and- Yes

! tested prior to use in rarack?

c

7. Does the crane meet the intent of Yes

ANSI B30.2-1976 and CMMA-707 i s. Is-the crane and its main hook-single failure proof? Yes i

I a.

i 2-19

-44

I CRITICALITY CONSIDERATIONS BASIS: 4.2% ENRICHMENT FOR REGION 1 4.2% ENRICHMENT @ 32000 MWD /rON BURNUP FOR REGION 2 ASSEMBLY TYPES CONSIDERED:

WESTINGHOUSE, CE, AND ANF (14X14) .

ar

". 45

i

.. .. j 1

PROPOSED FIGURE 2-10 '

i 37500

35000

i 32500 :

)

30000

/ i 27500 i ACCEPTABLE /

1 .BJRNUP DOMAIN / ,

3 / '

, x 25000 : <

'B  :

! g,/ '

:s 22500 : .

j -

20000

I

$ 17500 :

i f7 '#

1. /'#,e a

. g 15000 . f ,

12500 ~

ef,1 fd j 10000 :

l .Y, '

4 i /, lAMCCEPTaBLE ,

7500 :

, BURNUF 009AIN 5000 :

( Re ouires Region i Storage) i 2500 '

l',',

I O' .l...,,,, ,,,, ..., ,,,, ,,,, ...,

)

4

. 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0

! INITIAL ENRICHMENT, wty. U235 i

LIMITING BURNUP- CRITERIA FOR ACCEPTABLE STORAGE IN REGION 2 NOTE: 1. Any fuel assembly (24.2% U235 average) mechanically coupled with e full length CEA may be located anywhere in Region 2.

' 2. Peripheral cells are those adjacent to the Spent Fuel Pool well 3 or the cask lagdown ereo. g

1 b .

l- ,

! l l l THERMAL-HYDRAULIC CONSIDERATIONS ,

i l ANALYSIS MENU:

I i

!, 1. BULK POOL TEMPERATURE PROFILE FOR VARIOUS DISCHARGE SCENARIOS.

l 2. M AXIM AX LOCAL POOL W ATER '

i TEMPERATURE.

I

3. TEMPERATURE INCREMENT DUE TO PARTIAL l FLOW BLOCKAGE.

i

! 4. FUEL CLADDING TEMPERATURE t

I DISCHARGE SCENARIOS.

l

CASE 1

FULL CORE OFFLOADED AT 28TH REFUELING (EOC DISCHARGE) i

CASE 2: CASE 1 OUTAGE IS. OF 56- DAY DURATION, EMERGENCY SHUTDOWN i 30 DATS AFTER RESTART (BOC DISCHARGE)-

j I

a. .

. 47'

, '; se a j

t HOLTEC INTERNATIONAL 4

ONE NORMAL REFUELING BATCH  :

' POOL FELED EXCEPT I w STORAGE CAPACITY FOR d 1 NORMAL 8ATCH E AND 1 FULL CORE REMAINING RELOAD 88 FAS BACM TO REACIOR ,

! $ bb -

/ ,

/ i rasouRS

= =  !

d -

- i2

( (

? ) J FULL CORE (133 FAS) OFFLOAD AT 3 FASAtR

! E

• t

___.._ _ ~ . p j REACTOR StARDOWN END OF OUTAGE l

FIGURE 5 4.1 FT. CALHOUN FUEL POOL DISCHARGE SCENARIO ONE i i

i =

4 l'

q r . -. - .--

_ i

.; o .

HOLTEC 'NTERNATIONAL NORMAL BATCH IN SCENARIO ONE . GRGENCY FULL k. ORE OFFLOAD ,

1. "YO REL'OAD 88 FAS BACK TO REACTOR f  : _

E /

)) / REACTOR OPERATING 144 TOURS n nOURs _

t d ('(

y FULL CORE OFFLOAD FULL CORE (t33 FAS) OFFLOAD AT 3 FASaiR

?

N

• se MYS 30 DAYS e -. ,

= * - * . n .,..e.*s . _ - .e- . .m o m 4

REACTOR SHUTDOWN N OF OUTAGE REACTOR StRRDOWN FIGURE 5 4.2 .

FT. CALHOUN FUEL POOL DISCHARGE SCENARIO TWO N b 1 -D ,

I

i i

i J

i i

EVAPORATION HEATLOSS

)

i

__s____

i i

}

1 SPENT FUEL POOL i 4 4

i i

HEATEXCHANGER i

P 1 ~ Tg r ,

g COOLANT tg ' R J

FIGURE 5.5.1  !

1 BULK POOL: ANALYSIS MODEL 5-25

. '50-

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

a

, r-,

6 B

1 4

{ 1 i

i

! Table 5.8.1 I

FUEL SPECIFIC POWER AND POOL CAPACITY DATA I

4 i

l Net water volume of pool, gal. 164,000 1'

l- Fuel pool thermal capacity, 10 8 Btu /*F 1.34 8

j' Average operating power of a fuel assembly, 10' Btu /hr

~

38.47 l SFP Cooler Coolant inlet temperature, 'F 90

, SFP Cooler Coolant flow rate, 108 'lb/hr 0.58 I

i i

i I

d 4

+

j 4

e s

[

1.

3

~

5-19

. . -51

i _

,

• s .

1 l

.. q 1

i

?

I a

i Table S.8.2 l SFP BULK POOL TEMPERATURE i

4 Coincident Coincident Coincident Heat Load

.; Maximum Time After. Evaporation

! Pool . Reactor Shut- to SFP Hxs Heat Losses

Temp., 'F down, hrs. '.10' Btu /hr 10' Btu /hr i Discharge 134.88- 122 20.72 0.253 l in Case 1 4

i i Discharge 128.03 195 17.58 0.162 f

in Case 2 i

i 1

S

.j .

t d

k.

4 4

5-20

-- 52

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

1

, 6 - .-

i '

1 4

j 1

I i ,

i 4

4 x

b

• 4 i

1 1

f Table-5.8.3-4 l TIME-To-BOIL RESULTS-I a

Time-to-Boil (Hours) Maximum

l (Without Make-Up Evaporation case No. Water) Rate (GPM) l

? 1 9.90 33.4 i 2 10.82 32.6-i 1

4 I.

f 4

i C

, 'W I

i 5-21 53-

.~.

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

1 _

i -

. .-.., 4 .

I

i. .

b 0

4 h' i i .. .

1 s

i i I

l 4

I I

1 i

i Table 5.7.1 i

PEAKING FACTORS Factor value-I

! Radial 1.955 i

Axial 1.4605 Total ~3.067 1,

4 I

5-18

'~

.y.

i, i

i-t 5

1 4

1 1

) ' Table 5.8.4 f MAXIMUM LOCAL POOL WATER AND FUEL CLADDING TEMPERATURE j - FOR THE LIMITING CASE ' (Case 1) l 1

l.

4 Maximum - -Maximum Local Local i- Pool Fuel Water- .

Cladding Teme.. 'F Teme.. 'F No Blockage- 207.9 254.9 l 50% Blockage 221.2 264.9 t

i I

4 f

5-22 i - 55

o4 a,

l l

i i

i-i 4

i SEISMIC / STRUCTURAL CONSIDERATIONS

! FUfiL RACKS i

i l 1. GENERATE FOUR SETS OF STATISTICALLY l INDEPENDENT SYNTHETIC TIME-HISTORIES, i

2. PREPARE NON-LINEAR DYNAMIC MODEL OF i . RACKS.

i

! 3. PERFORM 3-D SINGLE RACK AND WHOLE l POOL MULTI-RACK ANALYSES.

i

4. DEMONSTRATE STRESS AND KINEMATIC l . COMPLIANCE.

l a

l

't. 56

).,,

i i ATTRIBUTES OF SEISMIC ANALYSIS n

a The rack is modelled as a 3-D structure capable of simulating all j motions (sliding, bending, twisting, rocking, etc.)

4

! e Effect of liner-to-rack pedestal friction treated by using bounding i

values, and also gaussian distribution d

l u Four time-history sets and a large number of parametric studies i performed I

i

?

.i

~

7 M. g 57 D

d'*

4 e

I

,'l O

in- FT. CALHOUN STATION, OMAHA PUBLIC POWER DISTRICT, 6: MAXIMUM HYPOTHETICAL SEISMIC Synthetical Acceleration Time-History: ' SET-1-H1

File name: o-tmhe.h11

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Tim e, sec. .

_ . _ ~ . - ____-._ _ _. _ __ _ ..._ _. _ _ _ _ _ . _ _ _ .. _ _ . _

i l . .s ,

1 1

g Table 6.1.1 j LISTING OF PLANTS WHERE DYNARACK WAS APPLIED i

.i PLANT DocTET NO.

l ,

Enrico Fermi. Unit 2 USNRC 50-341 i

i Quad Cities 1 and 2 USNRC 50-254, 50-265 i Rancho Seco USNRC 50-312 i

a Grand Gulf Unit 1 . USNRC .50-416 Oyster Creek USNRC- 50-219 i

2 Pilgrim Unit I USNRC '50-293 V.C. Summer USNRC 50-395 -

t Diablo Canyon Units-1 and 2 .USNRC 50-275, 50-323 f 1 Byron Units 1 &L2 USNRC 50-454,-50-455-l Braidwood Units 1 & 2 USNRC 50-456, 50-457

Vogtle Unit 2 USNRC 50-425 St. Lucie Unit 1 USNRC 50-335 Millstone Unit 1 USNRC 50-24'i D.C. Cook Units 1 & 2 .USNRC 50-33.5, 50-316 l Indian-Point Unit 2 USNRC. 50-247

I

! Three Mile Island: Unit 1 USNRC 50-289

- J.A. FitzPP. trick USNRC 50-333

?

L Shearon Harris Unit-2 USNRC 50-401 Kuosheng Units 1 & 2' Taiwan Power company.

l l

Chin Shan Units 1 & 2 Taiwan Power Company.

Ulchin Unit 2 Korea Electric. Power Laguna Verde Units:1 &:2 Comision Federal de

-Electricidad .

Zion Station-Units 1 & 2 ,_ USNRC-' 50-295, 50-304 96 e -

t- s-35. 59

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

1 i

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. FIGURE 6.4.1 PICTORIAL VIEW OF. RACK STRUCTURE J

d 6-116

=

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g 11 xf .1- g 52 nru erauw-r FIG. S.4.2 500%IIC EB. F2 0??>.PJG 6-117-61-

_ ____ _ _ _ _ _ _ _ - _ . _ _ _ . . - _ _ _ . _ _ _ . . _ . _ __,_.;-_._._...-__._._._..___......_.__.-__.__,_. -___.m.~. - . _ - . . . _ . . . . _ . . . . _ . . . . . _

s .

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' chm / Sff0 5 h al 7 i

TYPICAL TOP 2

> IMPACT ELEMENT i o -

7 .

y

-e H RACK STRUCTURE O

e N

> 3

-o =o 7 C

- W K m  !

5 m 4

-4 TYPICAL BOTTOM _

IMPACT ELEMENT Yh0 -,

~

' @ \ Aw A+y 8 Av Am Aw A+y c

du.4 E555

~

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1 6-119 7

. 63

.--p.r #" "*ME'

LISTING OF ANALYSES PERFORMED ,

a 24 separate 3-D single rack siniulations under the controlling time-history acceleration' set. Different : cases: correspond-- to various modules, various conditions of fuel loading and limiting values of the interface friction coefficient a Two- discrete whole pool multi-rack simulations -(all- 11 racks'

' modelled simultaneously)'

E Large number of 3-D parametric runs ensures that upper bound of all' stresses are obtained.

m Whole pool multi-rack analysis corrects the unconservatism in displacement prediction of the 3-D~ single rack models i

i l

b 64 r

L .

.Eo ,

Table 6.7.2

SUMMARY

OF WORST.RISULTS FROM 24 RUNS OF SINGLE RACK ANALYSIS .

( LOADED WITH 1380# -REGULAR ITTIL ASSEMBLIES; MAXIMUM HYPOTHETICAL EARTHQUAKI TIME-HISTORY SET-3 x1.25 )

Item value Run I.D.

1. Maxi =um total: vertical pedestal load: 4 2 6,890 lbs. - drfimheo.rf2

2. Maximum vertical-load

~ 124,680 lbs.

in any single pedestal: drfimheo.rf8

3. Maximum shear-. load-in.any single pedestal: 48,758 lbs. dralmheo.rf8

4. Maximum fuel assembly-to-cell vall

. impact load-at'one local position:- 774 lbs. drfinhoo.re2

5. Maximum rack-to-wall impact load _at baseplate level: o lbs.

6. Maximum rack-to-wall impact load at the top of-rack: 0ilbs.

7. Maximum rack-to> rack impact load at baseplate level: 0 lbs.

8. Maximum rack-to-rack impact load at the top of. rack: 0 lbs.

9. Maximum corner displacements Top corner in x direction: .0.1566 in. drdahoo.hy8 in y direction: .. . 0.2481- in.- dralmbeo.rf8 Baseplate corner in_x direction: 0.0216.'in.: .

drdahoo.re2.

in y direction: -0.0490 'in. . -

dralmhoo.rf2

10. Maximum stress factors Above baseplate: 0.321 (R6). .dralmheo.rf8 support. pedestals: 0.192 (R6)- dre.1mheo.rf8 l

6-46 l

l e '

, , ,~ , - - - , , - . - , - , ,,-,.,-c. ,- ,an~+ --,- m , ,

1

! .-.6 ,

i -

4 i

i 1

1

Table 6.7.27 i

1

' COMPARISON OF CALCULATED AND ALLOWABLE LOADS / STRESSES

! AT IMPACT LOCATIONS AND AT. WELDS i

i Y.A L E

! Item / Location calculated Allowable t

i i

t .

! Fuel assembly / call wall 852 (multi-rack) 5409.

! impact, lbs.

Rack /basaplate Vald, psi 9876 -29820 l

i Padestal/basaplata-vald l

Cimensionless limit load .529 (multi-rack) 1.0 ratio - conservatively i neglects gussets in calculation)

! Call / call walds (lbs) 3079- 7906 i

l-Where-indicated', calculated results are based on bounding loads obtained.

from Whole Pool Multi-Rack analyses.

i I

4 i

t

~

6 I I

}.'

. T. .

Table 6.8.5 COMPARISON OF RESULTS FROM SINGLE RACK ANALYSIS AND WHOLE POOL MULTI-RACK ANALYSIS (WPMR)

(MHE; Fully Loaded with.1380 lb fuel)

Sincle Pack Analysis WPMR Analysis Maximum absolute 0.2481. 0.9939 displacement, in.

1 at top of rack Maximum pedestal 124,680 171,100 vertical load, ib.

S w

6-78 E 67

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O O-rd - - GAP TIME lilSTORY ' FORT CAlllOUN STATION, OPPD.

- . Gop between Rock-91 and East Wall, South Side. Top.

_ MHE-SEr4, Fully Loaded with 1380# Reg. Fuel Assernblies.

Friction coefficient = rondom . ( mean = 0.5' ). File:g1-w-4.dat.

o-9-

m-

.- .E "

~

[/ l o .t J

q_ -

.O' o - i . - . ......i i . .. . . . . .. i. .- i. , , .. . . . . i . .- .. . . . , . . i.

g> o0 -

5 10 15 20:

Time, sec.

FIGUHE 6.8.1

u - 2 a.u.a - . > .x a s - ..u an- .x-, ,a - u ... - . - - .. . .x..

i .. .

i i

l 4

1

, MECHANICAL ACCIDENT ANALYSES PERFORMED i

l I

1. DEEP DROP OF 2780 LBS (ASSEMBLY PLUS

TOOL) 4 l 2. SHALLOW DROP OF 2780 LBS i

3. GATE DROP i

HEIGHT

15" ABOVE TOP OF RACK FOR FAs 4.5' ABOVE TOP OF RACK FOR

! GATE i

i 4

e i

.; s .

9 POOL STRUCTURAL EVALUATION SEISMIC CATEGORY I STRUCTURE USE SRP 3.8.4 LOAD COMBINATION (ULTIMATE STRENGTH DESIGN METHOD)

POOL SLAB IS ON PILINGS.

TOP OF SFP FLOOR 995'-6" (EL.).

3D FINITE ELEMENT MODEL---OF THE POOL FLOOR MAT AND.THE WALLS N

l

- 71

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

,s ,

i -

i i

i-

! LOADINGS i

DEAD WEIGHT OF RACKS i

DEAD WEIGHT OF STORED FUEL WEIGHT OF POOL STRUCTURE AND WEIGHT.

i OF WATER L

)-

l VARIABLE VERTICAL LOAD FROM RACK

PEDESTALS

[

INERTIA LOADINGS FROM SLAB, POOL WALLS

AND CONTAINED WATER MASS l

HYDRODY.NAMIC LOAD l

I l

i

}

l p*O

, . 72-

,  ? *s .

J Table 8.5.1 CALCULATION OF EQt1IVALENT SLAB VERTICAL PRESSURES Slab Weight 12.5 psi f (average i thicknese 12')

Hydrostatic Load 18.63 psi (43' Water)

Rack + Fuel 26.2 psi Dynamic Adder (Watsr, Racks-and Fuels) (MEE) 5.42 psi Dynamic Adder a (Water, Racks and Fuels) (DE) 4.79 psi d

73

__ . .. I

~

as ,

h 4

Table,8.5.2 BENDING STRENGTH EVALUATION critical Limiting Load Combination Location Safety Marcin (see Section 8.4.3)

- East vall 1.61 2 West vall 1.57 2 North wall 1.79 2 Intermediate wall.

between fuel pool 1.61 2 and fuel transfer canal South wall 2.48 2 Slab 2.31 1 9

8-11

, . n ,

u

?

Table 8.5.3-SHEAR STRENGTH EVALUATION Critical Limiting Load Combination.

Location- Safety Marcin (see Section 8.4.3)

East wall 1.14 2 West wall 2.01 2 North wall 1.74 2 Intermediata vall- 1.03- 1 South wall 2.46- 2 Slab 2.86 1 a

O 8-12

.. 75

4 4

Table 8.5.4 RESULTS FOR PILE-ANALYSIS Limi. ting-Safety-Marain Average strengt.h 1.17 Local strength 1.07 M

E 9

8-13

[' . 76

. c. ,

t A COMPARISON OF THE OPPD LICENSING SUBMITTAL TO THE NRC'S OT POSITION FOR REVIEW AND ACCEPTANCE OF SPENT FUEL POOL STORAGE AND HANDLING APPLICATIONS APRIL 14,1978, REVISION J ANU ARY 18,1979 OPPD Ucensing Section Submittal Ref.

OT Position Descriotion Section 11 Overall Description 2.0 Nuclear and Thermal Hydraulic 4.0, 5.0 <

Ill Considerations Ill.1 Neutron Multiplication Considerations 4.1 Ill.1.1 Normal Storage 4.1 1.1.a 4.2.1, 4.3.1 1.1.b 4.1 1.1.c 4.1 1.1.d 4.5, 4.6 1.1.e 4.1, 4.2,1 1.2 Postulated Accidents 4.8 4.8.3, 7.2 1.2 (1)(a) 1.2.(1)(b) 4.8.5 1.2.(2) 9.1.2 .

1.2.(3) 4.8.4 4.8.1, Table 4.8.1 1.2.(4) 1.3 Calculation Methods 4.4, 4.5, 4.6 1.4 Rack Modification 1.4.(a)

Table 4.3.1,4.5,4.6 4.5, 4.5.2.4 1.4.(b) 1.4.(c) 4.5 1.4.d.(1) 4.3.2, 4.3.3, 4.5 1.4.d.(2)a 4.5.2.5, 4.6.3.5 1.4.d.(2)b 4.5.2.3, 4.6.3.3 1.4.d.(2)c 4.5.2.1, 4.6.3.1 1.5 Acceptance Criteria for Criticality 4.0 1.5.(1) Neutron AbsorberVerification 10.0 1.5.(2) Decay Heat Calculation for the 5.3 Spent Fuel 1.5.(3) Thermal Hydraulic Analysis of Spent 5.2,5.4-5.8 Fuel Cooling 1.5.(4) Potential Fuel and Rack Handling 2.5, 7.2.1, 7.2.2 Accidents Technical Specification T. S. 2.8, 4.4.2 1.5.(5) 1

.+-

OPPD Ucensing Section Submittal Ref.

OT Position Descriotion Section Mechanical, Material and Structural 3.2, 3.3 -

IV Considerations Description of Spent Fuel Pool and 7.1, 7.2, 8.1, 8.2 (1)

Racks Support of Spent Fuel Racks 3.2, 6.2 (1)a Fuel Handling 7.1, 7.2 (1)b Applicable Codes, Standards and 3.4 (2)

Specifications Seismic and Impact Loads 6.3 (3)

Loads, Loads Combination and 6.5, 7.4 (4)

Structural Acceptance Criteria Design and Analysis Procedures 6.4, 6.6 (5)

Structural Acceptance Criteria 6.0, 8.5 (6)

(7) Materials, Quality Control and 2.4,2.5,3.0,10.0 Special Construction Techniques (8) Testing and Inservice inspection 10.0 V Cost / Benefit Assessment 11.0 V. 1 Environmental Assessment 1.1 Special Needs for increased Storage Capacity 1.0-1.1.(a) Reprocessing - 11.1 1.1(b) Discharge Schedule 1.0, Table 1.1.1 1.1.(c) Fuel Assemblies Stored 4.7, Table 1.1.18 1.1(d) Control Rod Assemblies 46 CEA's 1.1.(e) Storage Capacity 1.0 1.1.(f) Full Core Discharge Capacity :1.0 1.2 Construction Costs 11.2 1.3 Alternatives to increased Storage 11.0 Capacity _

1.3 (a) Reprocessing Facilities -11.0 1 1.3.(b) Independent Spent Fuel 11.0

-Storage Facility 1.3.(c) Away From Reactor 11.1 1.3.(d) Replacement Power Shutting down the FortCalhoun Station is many times more expensive than simply reracking the

- existing pool.

2

.o ,

i <

e OPPD Ucensing 4

Submittal Ref.

Section Sectlert OT Position Descriotion Resources The expansionof the 1.4 spent fuel pool i

capacityis expected

, to require less than

.0001% of totalworld output of stainless steel and Boron Carbide. Experience has shown that the .

production of Boron CYcide is highly variable and depends upon need and can easily be exrMed to ac.vynodate wortowide needs.

Maximum Water Temperature 5.8,11.3 1.5 of Spent Fuel Pool V2 Radiological Evaluation 2.1 Radioactive Wastes 9.2

, 2.2 Releases of Kr-85 Documented in OPPD Annual Report

' Radionuclides Table 9.4.1 2.3.(a)

Dose Rates 9.5, OPPD Radiation 2.3.(b)

Protection Program Airbome Concentrations 9.4, and Measured 2.3.(c) in accordance with OPPD Radiation Monitoring Program increase in Airborne Releases 9.4,To be Monitored 2.3(d) in Accordance with OPPD Radiation Manitoring Program Rad Waste 9.4 -

2.3.(e)

Dose Rates-Increase 9.4 2.3.(f) 2.3(g) Emission Directly from Fuel 9.5

• 2.4 Rack Decontamination 9.2, excluding wts.

Accident Evaluation *

• 9.1.2, 2.5 V. 3 3.1.(a) Cask Drop **9.1.2 3

x __ . _ - - _

.a .

4 OPPD Ucensing Section Submittal Ref.

OT Position Descriotion Section Overhead Cranes, R. G.1.104 overhead system 3.1.(b) meets NUREG-0612 3.2 Cask Movement 9.1.2 Accident Aspects of Review 9.1.2 3.3.1(1)(a)

Accident Aspects of Review 9.1.2 3.3.1(1)(b)

Accident Aspects of Review 9.1.2 3.3.(2)

Cask Drop Analysis *

• 9.1.2 3.4 Maximum Weight of Loads 7.1,7.2 3.5 3.6 Radiological Consequences of Cask Drop "9.1.2

• Table 1 Existing Fort Calhoun Station SFP Racks to be Removed Module Size Capacity Quantity Est. wt. (Ibs) Total Cavities 8x9 72 1 15,800 72 7x9 63 7 13,825 441 6x9 54 4 11,850 216 Totals 12 159,975 729

• • The licensee does not presently utilize or own a spent fuel sample cask or spentfuel ship-ping cask, if such casks were to be utilized, the single failure-proof main hook of the Auxil-iary Building Crane would be used to transport these casks.This characteristic, in conjunc-tion with the presence of electricalinterlocks which preclude travel of the main hook and auxiliary hook over the spent fuel pool, would greatly reduce the likelihood of occurrence of a cask drop, obviating the need for consideration of radiological release impact of such an accident. Therefore, we conclude that an analysis of the radiological consequences of a cask drop accident is not required.

)

I 4