ML20080P841

Application for Amend to Licenses NPF-9 & NPF-17,allowing Spent Fuel Pool Storage Capacity Expansion from 500 to 1,463 Spaces for Each Spent Fuel Pool.Supporting Documentation Encl

ML20080P841
Person / Time
Site:	McGuire, Mcguire
Issue date:	02/17/1984
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20080P843	List: ML20080P841 ML20080P847
References
TAC-53531, TAC-53532, TAC-57316, TAC-57317, NUDOCS 8402230146
Download: ML20080P841 (65)

Text

. ..

N== c..mg DUKE POWER GOMPANY P.O. BOX 33180 CHARLOTTE. N.C. 28242 HAL B. TUCKER tes.c=non

== recomm (704) 3/3-4531 a=="== February 17, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S.. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: McGuire Nuclear Station Docket Nos. 50-369, 50-370

~

Dear Mr. Denton:

'Ihis letter is a proposed amendment to Facility Operating Licenses NPF*9 and NPF-17 for McGuire Nuclear Station Units 1 and 2. The proposed amendments would allow spent fuel pool storage capacity expansion from 500 to 1463 spaces for each spent fuel pool. The proposed expansion is to be achieved by reracking each spent fuel pool with two region, poison racks.

The rerack modification for McGuire's spent fuel pools was described to members of the NRC staff on January 31, 1984 in a meeting with Duke Power Company.

Attachment 1 is an analysis summary of the proposed amendment request and contains copies of revised overhead slides which were discussed at the meeting.

Duke Power's current schedule calls for reracking of the Unit 2 spent fuel pool to begin on August 1, 1984 and to be coupleted by December 31, 1984.

Under this schedule the reracking would be accomplished prior to the first refueling of Unit 2. Consequently, no water or fual would be in the pool which would make the reracking operation much simpler, safer, and less costly.

Pursuant to 10 CFR 550.92, Attachment 2 provides an cnalysis which concludes that the proposed amendments do not involve a significant hazards consideration.

-In order to provide for timely review and approval, Duke Power requests that the NRC perform its preliminary No Significant Hazards Consideration evaluation upon receipt of this formal amendment request.

Proposed changes to the Technical Specifications are contained in Attachment 3.

As discussed with members of the staff, details of the safety and environmental implications will be submitted to the NRC by March 15, 1984. This will enable the detailed safety evaluation to be performed.

These proposed changes are considered to consist of one Class III and one Class I license amendment; therefore, please find attached a check in the amount of $4,400.

Y, 8402230146 840217 PDR ADUCK 05000369 I 7Dl0

, P PDR

re- : ;.

.g "Mr. Harold R.' Denton, Director

' February 17,.1984 Page 2 Very truly yours, c-.

~

Y/ r s 1 B. Tucker-WHM/php Attachment-

=cc: Mr. James'P. O'Reilly, Regional Administrator

U. S. Nuclear Regulatory Commission l~ . Region II

'101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303

+

Mr. W.1T. Orders NRC. Senior Resident Inspector McGuire Nuclear Station Mr. Dayne H. Brown, Chief Radiation Protection Branch Division of racility Services Department of Humca Resources P. O. Box 12200 Raleigh, North Carolina 27605 a

m _

Mr. Harold R. Denton, Director February 17, 1984 Page 3 HAL B. TUCKER, being duly sworn, states that he is Vice President of Duke Power Coropany; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Cummission this revision to the McGuire Nuclear Station Technical Specifications, Appendix A to License Nos.

NPF-9 and NPF-17; and that all statemeats and matters set forth therein at true and correct to the best of his knowledge.

k ~ <<

Hal B. Tucker Vice President Subscribed and sworn to before me this 17th day of February, 1984.

Notary Public

^

My <ommission Expires Septe.no e 20, 1984

e DUKE POWER COMPANY MCGUIRE NUCLEAR STATION Attachment 1

~

Units 1 and 2 Spent Fuel Pools Two Region Rerack Analysis Summary k

DUKE POWER COMPANY McGUIRE NUCLEAR STATION UNITS 1 AND 2. SPENT FUEL POOL RERACK MEETING AGENDA INTRODUCTION DESCRIPTION OF SPENT FUEL POOL RERACK e RACK DESIGN

1. DESIGN DESCRIPTION

2. DESIGN EVALUATION - INCLUDES STRUCTURAL, CRITICALITY, AND THERMAL HYDRAULICS e SPENT FUEL POOL INTERFACE

1. STRUCTURAL

2. THERMAL

3. ADMINISTRATIVE CONTROLS e RACK INSTALLATION e RADIATION PROTECTION e SAFETY ANALYSIS

1. CONSTRUCTION ACCIDENT

2. CASK / HEAVY LOAD ACCIDENT

3. NATURAL DISASTERS

4. LOSS OF FORCED COOLING

5. FUEL HANDLING ACCIDENT LICENSING SCHEDULE

. DISCUSSION - COMMENTS 1

l PRESENT PROPOSED FULL CORE DISCHARGE 1990 2010 CAPABILITY RACK TYPE HIGH 2 - REGION / POISON DENSITY /

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N0 BOILING WHEN. COOLING SYSTEM OPE 2)

NO STRUCTURAL FAILURES AND SAFETY CON WHEN COOLING SYSTEM IS POSTULATED TO INOPERABLE 3)

MAXIMUM POOL TEMPERATURE NOT GREATER 150*F FOR NORMAL OPERATION 4)

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MCGUIRE SPENT-FUEL POOLS ARE DESIGNED TO

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- NORMAL DEAD AND EQUIPMENT LOADS PLUS DESIGN SEISMIC LOADS

--ALL NORMAL-DEAD, EQUIPMENT, AND LIVE LOADS

- NORMAL DEAD AND EQUIPMENT LOADS PLUS TORNADO WIND LOAD

- THERMAL STRESSES

- CASK DROP ACCIDENT CONCLUSION: PROPOSED RERACKING WITH ASSOCIATED LOADINGS, DESIGN CRITERIA, AND-ALLOWABLE STRESSES ARE IN COMPLIANCE WITH APPLICABLE FcAR REQUIREMENTS FOR CLASS I STRUCTURES b

20-

SPENT FUEL P00L INTERFACE - THERMAL Design Basis - Spent Fuel Pool Cooling System Standard Review Plan 9.1.3 method and criteria adopted:

. Maintain Spent Fuel Pool coolant temperature below 140*F under Normal Maximum Heat Load conditions assuming a single failure.

. Maintain Spent Fuel Pool coolant below s'aturation temperatures during periods following full core offloads (Abnormal Maximum Heat Load).

Additional assumptions:

. Fully loaded pool.

. Two Oconee units discharging 5 years decayed fuel to one McGuire fuel pool.

. Freshest irradiated spent fuel available at Oconee meeting 5 year minimum decay criteria

~

is transhipped. -

. No credit for heat loss throligh pool walls or from pool surface.

4 21

. ~

TABLE 3.2-3 Peak' Heat loads and Pool Temperatures for the McGuire Units 1 & 2 Spent Fuel Pools Followina Rerack 0perating Condition- Pool Temperatures ( F)

Heat Load Coolina Trains-Case- . (106 BTU /HR)_ Operating Design Basis Calculated Normal Maximum 18 2 120 113 18 1 140 133 Abnormal Maximum 41.6 12 140 137 41.6 1 <212 178 l'

e A

O 0

22

. TABLE.3.2-1 -

Normal Maximum Heat Load for McGuire ' Units 1 & .

Spent Fuel Pools J

1. of Spaces Irradiation Decay Hea Station- (Assemblies) (EFPD) Time (*10gOutput BTU /HR)

~

McGuire- L61 863 150 hrs. 11.4 McGuire 61 863 1 yr. 1.07 McGuire: 61- 863 2 yrs. .558 44 McGuire 61 863 3 yrs. .398 McGuire. ~~

61 863 4 yrs. .333 ,

~~ ~

1McGuire '61 863 5 yrs. .303 McGuire 61 863 6 yrs. .287 McGuire- 61 863 7 yrs. .277 McGuire 61 863 8 yrs. .269 McGuire 61 863 9 yrs. .262 McGuire 61- 863 10 yrs. .256 Oconee 72 1263 5 yrs. .339 Oconee 72 1263 6.5 yrs. .316 Oconee. 72 1263 8 yrs. .302 Oconee 72- 1263 9.5 yrs. .291 Oconee 72 1263 5 yrs. .339 Oconee 72- 1263 6.5 yrs. .316 0conee 72- 1263 8 yrs. .302 OconeeL 72 1263 9.5 yrs. .291 Oconee- 23 1263 11 yrs. .089 1270 18.00

' Note: This analysis conservatively nssumes the. freshest fuel meeting the-minimum 5 year decay criteria at two Oconee units will be'transhipped in a manner ~ as to maximize the ' total calculeted heat load for a fully loaded McGuire spent fuel pool . To the extent that Oconee to

'McGuire' transhipments will . involve fewer. assemblies with higher decay l times., actual peak heat loads will be lower.

23

TABLE 3.2-2 -

Abnormal Maximum Heat load for.

McGuire Units 1 & 2 Spent Fuel Pools

1. of Spaces Irradiation Decay Heat Output-Station ( Assemblies) (EFPD) Time (*106 BTU /HR)

McGuire 64 23.5 150 hrs. 6.53 McGuire 64 311.5 150 hrs. 11.2 McGuire 65 599.5 150 hrs. 11.9 McGuire 61. -863 36 days 5.38 McGuire 61 863 1 yr. 1.07 McGuire .61 863 2 yrs. .558 McGuire 61 863 3 yrs. .398

McGuire _61 863' 4 yrs. .333 McGu' ire 61 863 5 yrs. .303 McGuire 61 863 6 yrs. .287 McGuire 61 863 7 yrs. .277 McGuire '61 863 8 yrs. .269 McGuire 61 863 9 yrs. .262 McGuire 61 863 10 yrs. .256 Oconee 72 1263- 5 yrs. .339 Oconee 72 1263 6.5 yrs. .316 Oconee 72 1263 8 yrs. .302 Oconee 72 1263 9.5 yrs. .291 Oconee -

72 1263 5 yrs. .339

- Oconee -72 1263 6;.5 yrs. .316 Oconee 72 1263 8 yrs. .302 Oconee 72 1263 9.5 yrs. .291 i

Oconee 23 .1263 11 yrs. .089 1463 ,

41.61

~

O 9

4 24 m_

~~~

~ FUEL PLACEMENT ADMINISTRATIVE CONTROLS FUEL STORAGE " DIRECTORY"

' REGION 1: NO RESTRICTIONS + 4.0 w/o DESIGN 286 STORAGE CELLS RESERVED FOR NEW FUEL STORAGE RESERVED FOR TEMPORARY CORE OFF-LOADING RESERVED FOR " UNQUALIFIED" IRRADIATED FUEL

-REGION 2: RESTRICTED STORAGE 1177 STORAGE CELLS EXCLUSIVELY FOR " QUALIFIED" IRRADIATED FUEL NO RESTRICTIONS WITH CHECKERBOARDING 9

25

~.. .

f FUEL PLACEMENT ADMINISTRATION CONTROLS

~

SPENT FUEL CLASSIFICATION STEPS BURNUP CALCULATIONS AT END OF CYCLE CALCULATED BURNUP COMPARED WITH PROJECTED BURNUP VALUE COMPARED WITH BURRUP CURVE (TECH SPEC)

• . CLASSIFICATION AS "OUALIFIED" OR "UNGUALIFIED" DOCUMENTATION OF "GUALIFIED" ASSEMBLIES 9

26

,.-.-7 . , 7--, , -- , ,, .~-----c, , . ,- . - -- - - - . - .-,-m - -

,-~- - .--, ,

r-

, FUEL PLACEMENT ADMINSTRATIVE CONTROLS FUEL MOVEMENT SAFEGUARDS - REFUELING ALL-FUEL MOVEMENTS FROM REACTOR CORE WILL BE DIRECTLY INTO REGION I CORE TO REGION 1:

NO LOADING RESTRICTIONS N0 VERIFICATION REQUIREMENTS REGION 1 TO REGION 2:

-FUEL MUST HAVE DECAYED 16 DAYS REVIEW DOCUMENTATION DOUBLE VERIFICATION OF ID/ LOCATION REGION-2 PLACEMENT 27

r FUEL PLACEMENT ADMINISTRATIVE CONTROLS FUEL MOVEMENT SAFEGUARDS - GENERAL-ON-SITE TRANSFERS:.

GENERALLY WITHIN SAME REGION e'

DOCUMENTATION REVIEW AND DOUBLE VERIFICATION REQUIRED FOR REGION 1 TO REGION 2 MOVES OCONEE SHIPMENTS:

BURNUP VERIFICATION / DOCUMENTATION PRIOR TO SHIPMENT

. SHIPMENT'0F QUALIFIED ASSEMBLIES ONLY DOCUMENTATION REVIEW PRIOR TO PLACEMENT CHECKERBOARD STORAGE:

PROCEDURAL CONTROL ON EMPTY LOCATIONS BOUNDARY DEFINED BY ONE VACANT R0W N0fE:' PRE-RERACKINVENTORYWILLBEVERIFIEDANDDOC TO COMMENCEMENT 0F RERACKING OPERATIONS (McGUIRE 1) ,

28

s r-NEUTRON POISON SURVEILLANCE PROGRAM '

TWO SEPARATE SURVEILLANCE SPECIMENS PER LOT INDEPENDENT PROGRAMS FOR EACH REGION, POOL PROCEDURAL. CONTROLS ON FUEL PLACEMENT TO INSURE MAXIMUM SPECIMEN EXPOSURE PERIODIC TESTING OF MECHANICAL INTEGRITY AND ABSORPTION CAPABILITIES

• - INITIAL INSPECTION SCHEDULED FOR 5-YEARS FROM INSTALLATION t

6 s

29

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

c .

RACK IllSTALLATI0fl

.THE INSTALLATION PLAN IS BASED ON THE FOLLOWING OBJECTIVES:

- MAINTAINING INSTALLATION EXPOSURE LEVELS AS LOW AS REASONABLY ACHIEVABLE (ALARA)

- HAVE A MAXIMUM 0F 151 FUEL ASSEMBLIES TOTAL (INCLUDES OCONEE AtlD MCGUIRE FUEL) IN THE POOL AT THE COMMEtiCEf1EllT AllD DURING Tile RERACK OPERATION FOR UNIT 1

- UTILIZE AN Uf1 FLOODED POOL WITH NO FUEL ASSEf1BLIES STORED-FOR UNIT 2 TO ACHIEVE SIGilIFICAtlT EXPOSURE (ALARA) AND COST SAVINGS

- ACHIEVING ACCEPTABLE TOLERANCES ON MODULE VERTICALITY, LEVELilESS, AND POSITI0 FLING 30 t-- - , _ . . . _ . . _

. IllSTALLATION PLAN UNIT 1 A. J R g

Il il i

I M

8

-1 i -

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O O:O:0:0:0:01 -

O OIO O.0:O:0! - STEP 1 O o;O:0 cro:Oi -

O O!O!Oio;Olo' -

Move all fuel to

.O O Oio'O.O!O south end of pool IO O OiO CIOiG (151 assy's) 1 O OIO;O OlOD(i -

O O !O O.O:00<i O OC:O OIOD<i _

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-O O 0:0 010.0 -

O O O O1010.0 -

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II illi l l: 4 HJ R 31

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t I i 19 Niii .,

\ $ i l iD i i :f 10'-0" - # Ni /

; iNi / ;

i ; i y O 25 t ! yi CO 26 STEP 2

1. io!O!O O- -27

-'iolo!O80'C O:01 28 Remove existing grid 0100 0:O:O:01 OIO!O O10:O:0! ._ and cells in north-O C;C c:OiO:Oi west area of pool O OlC 01010!O' O 0:O:0f0:O:0 0 O Ci0.0.O:0 (O 0:Ofo'OlO X ~

joiO!ClO:COtX

, , iO;OctC:CO.k lOIO!OIC CiOtX 101OCIClOiC W O!O1CioiciO;O

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_ __OlO(ClO:OiOFx ClOIC!C:OlON C OCOOO'O e o cie OTO'G ~

- 49 lIll l

HJ R 32

• t _ __

UNIT 1 J R I

_1 <

i

}

f I i Rack #1 '

(13 x 11) , .

I t I

!. 8 {

i ! I ie i i Rack #2 (13 x 11) , ,

! 1

- 19

\-  !!

\ !f

'N i,

/ s ,

N e 10'-0" f '

r O - 25 o c O!O 00_ 26 27 O O O!ctO c!OI 28 O O O10:0 010 STEP 3 O c O!O!O OIO O O! con O10 o 0;O:O O Oto Install racks #2 and #1 O OC O!O OlO O O O 0;O OIO _

010 0:0 0 OD(

OiOlO;O;O O B<

0;O.010:0 Oik JOIO!OiO c O p<

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~

[ 151 assy's O Oc OCTD' O OO OO O X C OX 0000 o OiOTOTOUV-O O c10:0 0 C -

0 0 'OIDID 0h o o_c OEDT-

. !O OcOc O O C'6bUQ UQZ 49

~

l HJ R 33

• O

. UNIT 1 J R N

-1 i i -

I 1 Rack #1  !

8 34 g' O assy's ooo I 000 5" R,ack @

s '

I

1. 2 g , , , l ll 117 8 l o assy's , , ,

cacccc .s 1 i - 39 i

ll 1

Two Empty l

Rows STEP 4 Move fuel to Racks

1. 1 and #2 l

6 i it I t i e i i

S

' ~

- 49 l l 34

.x  ;

. .. =

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-u

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• I-assy's

• l t .. .

• * . I 1 1-ElCk #2 l ,

117. '

1 .

assy's-se s l . . .

t ,

e ii i'; i i

= ' '

T i,' , . ;e STEP 5

~

o I 'I' Ii

~ Remove existing rack grid and

',,' , , , cells in southend of pool

'e i: i . ,.

27 9

e-J O G e

b

,- ;'. .. =

.- 35

UNIT 1 J R i

ir-l

___1 i

, i ~ i i . s i , i i i i Rack #1 * '

• i

- ]

34 assy's

• 1I I t t t  !

s g g g i t ! i .

. a . l i i t o

• l .

Rack #2 i . ,

i ,

117 assy's - l : -

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i,, i

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• 4

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= .

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hi iI } } ! l l 1 I --

i : . , i

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4 I i

• l [ i y

Rack #6

. (12x16) STEP 6 Install racks #8, #7, and #6.

Rack #7 (12 x 16) t Rack #8 (12x16)

G 36

4 -,

UNIT 1 J R

,, I

! M. i i ,

-1 1 l 1 0 h l .

i .  : .

Rack #1 . ,

e i i

, , , i

< . . i i s s n Rack #2 S

e t j g I

i . i i 4 .

t ! l

• I . 1 i 1 e i 27 I

. Rack #6

. STEP 7 tiove fuel to rack #8 Rack #7 l

l i Rack #8

:::-3:-

mb 1

vooooococcooocco o

D  :' -

O-5- ;

c: i 151 assy's $

i r**

r {e c . , - - . -  ;

. .,,i, 16 rows i

i h

I 37 L

UNIT 1 q.

e i

4 Rack #1 ]

4 Rack #2 STEP 8 Remove remaining rack grid and cells.

l l

. Rack #6 h

N Rack #7 l-i Rack #8 EESEE8ccccccccc

! d C l

5 151 assy's t

$ D l dccccccccccccccb .

l 38

UNIT 1 nvN i Rack #1 Rack #3 (15 x 14)

Rack #2 Rack #4 (16 x 14) STEP 9 Install racks #5, #4, and #3.

Rack #5 (12x16)

. N 6.

Rack #6 ,

I Rack #7 F

Rack #8 151 assy's t

39

a .g-INSTALLATION PLAN UNIT 2 O I' ' N i O O C RACK =3 RACK =1 Rack Installation Sequence

=8, #7, FG, =5, =4, 52, =3, =1 15 X 14 13 X 11 9.125 CTC 10.4 CIC c c o c C C C C RACK =4 RACK =2 16714 13 X ll 9.125 CTC 10.4 CTC o c C .C O O RACK =5 12 X 16 9.125 CTC N O O 1

- RACK =6 12 X 16 9.125 CTC

! c c l c O RACK #7 12 X 16 9.125 CTC 0 C o o RACK =8 12 X 16 9.125 CTC C C 40

.. 3 RADIOLOGICAL CONSIDERATIONS

' EXPOSURE CONTROL FOR DIVERS SPENT FUEL PLACEMENT MAXIMIZES WORKING DISTANCES INSTALLATION PLAN MINIMIZES FUEL MOVEMENT

. 10 FT. DISTANCE FROM SPENT FUEL MAINTAINED

=-

EXTRA PERSONNEL MONITORING - EXTREMITIES / MULTIPLE WHOLE BODY UNDERWATER RADIATION SURVEYS CLOSE H.P. OBSERVATION DURING DIVING ACTIVITIES (DIRECT DECONTAMINATION) CONTAMINATION CONTROL - PROTECTIVE CLOTHING, LOW PRESSURE HOSEDOWN

.- EXTRA DOSIMETRY RECORD KEEPING 41 k

~

RADIOLOGICAL ~ CONSIDERATIONS PROJECTED RERACKING EXPOSURES PROJECTIONS BASED ON PREVIOUS OCONEE EXPERIENCE AND '

FOLLOWING ASSUMPTIONS:

1, 151 SPENT FUEL ASSEMBLIES PRESENT

2. ~302' SPENT FUEL MOVES i

3. 8 RACK MODULES INSTALLED.

ESTIMATED ALARA DOSES - PERSON REM:

H0IST AND CRANE INSTALLATION / REMOVAL 4.03 POOL CLEANUP OPERATIONS. 1.27-

.IN-POOL. FUEL TRANSFERS 1.01

-RACK REMOVAL-AND INSTALLATION 7.30 t

RACK DECON'AND DISPOSAL 0.69 TOTAL 14.30 9

I

,~..

, 42 4 r

,L" ,.

RADIOLOGICAL CONSIDERATIONS SPENT FUEL POOL CLEANUP SPENT FUEL P0OL COOLING SYSTEM DEMINERALIZER AND

, FILTERS

• PORTABLE, AUXILIARY SYSTEM USED AS BACKUP

• WEEKLY -SAMPLING OF POOL WATER

• DIVER WORKING AREAS CLEANED WITH PORTABLE VACUUM SYSTEM

• FLOATING SKIMMER USED T0 flINIMIZE FLOATING CRUD 3

,  ?

?

c l

l I

43 I

o RADIOLOGICAL CONSIDERATIONS INCREASES IN OCCUPATIONAL DOSE OVERALL EXPOSURE INCREASE DUE TO ADDITIONAL STORED FUEL ESTIMATED TO BE LESS THAN 1% OF THE ANNUAL ~ STATION DOSE, EXPOSURE INCREASE FROM DIRECT RADIATION NEGLIGIBLE DUE

, TO WATER DEPTH EXPOSURE INCREASE FROM CONTAMINATED WATER f1EGLIGIBLE DUE TO~ CIRCULATION' SYSTEM EXPOSURE INCREASE DUE TO RESIN / FILTER CHANGES NEGLIGIBLE DUE TO N0; INCREASE IN FUEL MOVEMENT ACTIVITY 9

,c' er-I

~

g. <

44

. )f 4

RADIOLOGICAL CONSIDERATIONS FUEL RACK DISPOSAL OPTIONS A. DECONTAMINATION FOR SALE AS SCRAP B. SHIPMENT TO LOW LEVEL WASTE BURIAL SITE C. LONG TERM ON-SITE STORAGE UNIT 1 RACKS -14,000 FT3 STAINLESS STEEL 45

• s- .

CONSTRUCTION ACCIDENTS

- UNIT 2 RERACKING PERFORMED IN DRY POOL WITH tl0 FUEL IN P0OL

- AS DISCUSSED IN THE IllSTALLATION PLAN FOR UtilT 1, FUEL WILL BE POSITIONED SUCH THAT NO LIFTS WILL BE f1ADE OVER FUEL

- LIFT RIGGING, H0ISTS, AND CRANES SHALL BE DESIGilED AtlD OPERATED IN ACCORDANCE WITH ANSI B30.9, B30.11 AND B30.20 RESPECTIVELY

- IN ADDITIO!1, A DROP RESULTIllG IN BREACHIrlG 151 FUEL ASSEMBLIES WAS POSTULATED. THE EXCLUSION AREA BOUNDARY DOSE, TAKIllG NO CREDIT FOR VEllTILATION SYSTEM (FILTRATION) IS HELL BELOW 10CFR100 LIMITS, a

46

e McGuire Units 1 & 2 Fuel Pool Rerack Construction Accident Analysis METEOROLOGY 2 Hour Accident X/Q 9.0E-4 s/m3 SPENT FUEL RADI0 ACTIVITY BASES

. Conservative case maximum assembly inventory FSAR Table 15.5-3

. Peaking factor 1.2

. Decay Periods 61(1)/65 days 90/1 year

. Gas Gap Fractions for Kr85 0.3 for other Nobles and Iodines 0.1 RADI0 ACTIVITY RELEASE BASES

. Number of fuel assemblies damaged 151

. Release quantity of gap activity 100%

. Spent Fuel Pool water iodine removal 100

. Filtration None DOSE BASES

. Receptor breathing rate 3 3.47E-4 m /sec

- DOSE CONSEQUENCES

. Whole Body 0.3 Rem

. Thyroid 70 Rem (1) Corresponds to one refueling batch.

47

U CASK / HEAVY LOAD ACCIDENT 1

- NO CHAllGES WILL BE MADE TO THE PHYSICAL STRUCTURE AllD LAYOUT OF THE POOLS AS A RESULT OF THE PROPOSED RACK REPLACEME.'lT

' CONCLUSION: THE AllALYSIS AilD CRITERIA DESCRIBED IN THE FSAR REMAIN VALID.

i-t e

b 4

I i

48

n NATURAL DISASTERS

- AN ANALYSIS HAS BEEN PERFORMED TO DETERMINE THE POTENTIAL CONSEQUENCES IN TERMS OF-THE MAXIMUM HUMBER OF FUEL ASSEMBLIES WHICH MAY BE DAMAGED UNDER THE MOST CONSERVATIVE COMBINATION OF MISSILE TYPE AND TRAJECTORY, SHOULD A TORNADO DROPELLED MISSILE. ENTER THE SPENT FUEL BUILDING .

-- A TOTAL OF 38 FUEL ASSEMBLIES IN REGION II (CONTROLLING CASE)

CAN POTENTIALLY SUFFER A TOTAL LOSS OF INTFGRITY DURING A TORNADO MISSILE IMPACT

- THE' RADIOLOGICAL CONSEQUENCES CF A TORNADO MISSILE IMPACT WILL BE MITIGATED BY LIMITING THE AGE OF FUEL DISCHARGED TO REGION II TO A MINIMUM 16 DAY DECAY PERIOD v

49

O h h b I

LL et< h l.* , ,' . < = .

.j . .

l.

i. .J i* j t

I 6

it l .

I g.

a e i. r N

N N .

N

\, i c.7.c cesriu na. N i 77$

• V ') x i

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s ,

e. . I.

N neuc. att a.c. i

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, .r -,q . ,

, ,' \ g Amut? F'*kW{ cf u,Mr.L N i N

i N  !,

N .

N  :

N

~

N I 38i FUEL

\(I \ N ELEMENTS # 5.4" i N s 9.125" 0.c. l s l

< P 4 (t'll E40 \J,  !

.! ,.]

l it s i i .

N l

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N i

h ' !

I

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1 i

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g I

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,a * **

g a

,--l

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• f . ..f .}

l  : '

TORilADO MISSILE STUDY i

(w) McGUIRE NUCLEAR 50

,s .c McGuire Units 1 & 2 Tornado Missile Impact on Spent Fuel Pool METEOROLOGY

. Atmospheric dilution for tornado conditions 8.1E-5 s/m3 I

SPENT FUEL RADI0 ACTIVITY BASES

. Conservative case maximum assembly inventory FSAR Table 15.5-3

. Peaking factor 1.2

. Decay Period 16 days

.- Gas Gap Fractions for Kr85 0.3 for other Nobles and Iodines 0.1

'RADI0 ACTIVITY RELEASE BASES

.. -Number of fuel assemblies damaged 38

. Release quantity of gap activity 1007,.

. Spent Fuel Pool water iodine removal 100

. Filtration None

-DOSE-BASES

.. Receptor breathing rate 3.47E-4 m3 /sec DOSE CONSEQUENCES

. Whole Body. 9.54E-1 Rem

. Thyroid 2.67E+2 Rem h

51

SAFETY ANALYSIS - LOSS OF FORCED COOLING Design Basis - Spent Fuel Pool Cooling System is designed to Seismic Category 1 Quality Group C requirements and is redur, dant in active components such that:

. Cooling system complies with Standard Review Plan 9.1.3 design guidance.

. Complete loss of forced cooling highly unlikely.

Loss of Cooling Analysis assumptions:

. Adiabatic treatment of fuel pool coolant mass.

. Maximum design basis heat loads.

. Design basis initial pool temperatures.

Additional Considerations:

. Redundant Seismic Category 1, Quality Group C makeup systems are provided.

~

52

. l TABLE 6.3-1 Time to Boiling Following loss of Forced Cooling Under Design uusis Conditions for McGuire Units 1 & 2 Spent Fuel Pools

. Heat Load Initial Pool Temperature Heat Up Time (100 BTU /HR) (*F) (HRS) 18 120 13.8 18 140 10.8 41.6 140 4.7 53

l

+f l

l SAFETY ANALYSIS - FUEL HAfiDLING ACCIDEtlT Design Basis - Radiological consequences of a fuel handling accident calculated in accordance with Regulatory Guide 1.25:

. Analysis unchanged from that presented in ficGuire FSAR Section 15.5.9.

e 54

s l .

MCGulRE RERACK PROJECT SCHEDULE BEGIN UNIT 2 RACK REMOVAL /lNSTALLATION 08-01-84

' COMPLETE UNIT 2 RACK INSTALLATION 12-31-84

• - PROJECTED UNIT 2 REFUELING OUTAGE- 02-01-85 '

BEGIN UNIT 1 RACK REMOVAL /lf1STALLATION 08-01-85 COMPLETE UNIT 1 RACK INSTALLATION 03-01-86 PROJECTED UNIT 1 REFUELING OUTAGE 86 1

55

McGUIRE RERACK LICENSING SCHEDULE FORMAL APPLICATION 2/14/84 NRC COMPLETE PRELIMINARY NO SIGNIFICANT 3/14/84 HAZARDS CONSIDERATIONS REVIEW FINAL NRC SUBMITTAL 3/15/84 PUBLIC NOTICE PUBLISHED IN FEDERAL 4/01/84 REGISTER LICENSE AMENDMENT APPROVED BY NRC 8/01/S4 56

DUKE POWER COMPANY MCGUIRE NUCLEAR STATION Attachment 2 No Significant Hazards Consideration Evaluation I

Attachment 2 NO SIGNIFICANT IIAZARDS CONSIDERATION EVALUATION Duke Power Company (Duke) has made the determination that this amendment request involves a No Significant Hazards Consideration by applying the guidance presented in 10 CFR 50.92. This ensures that operation of the facility in accordance with the proposed amendment would not:

1) Inyc]ve a significant increase in the probability or consequences of an accident evaluated; or

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) Involve a signific:nt reduction in a margin of safety.

To put the matter in perspective, necessary background is first provided.

Thereafter, a discussion of each of the significant hazards considerations is provided.

McGuire Nuclear Station was designed and constructed with two spent fuel storage pools--one associated with each unit. The design capacity of each pool is 500 spaces (approximately 2 1/2 cores). The McGuice Final Safety Analysis Report addresses the safety implications of these pools to include relevant parameters associated with criticality, structural integrity, and cooling (Safety Evaluation, Docket Nos. 50-369/370). The evaluation found the environmental and safety impacts of such storage to be acceptable.

On April 17, 1977, President Carter issued a policy statement on commercial reprocessing of spent nuclear fuel which effectively eliminated reprocessing as part of the relatively near term nuclear fuel cycle. On October 18, 1977, the GESMO proceedings were deferred indefinitely. The combined effect of this national policy was to leave operating nuclear plants, like McGuire, without a repository for the spent fuel previously generated or being generated. Thus, Duke is forced to do additional reracking of the McGuire spent fuel pools to further increase its storage capacity.

With this application, Duke Power is requesting approval to use Westinghouse designed / constructed two region poison racks to increase each McGuire spent fuel pool capacity to 1463 spaces - 286 spaces in Region 1 and 1177 spaces in Region 2. This modification would extend the McGuire fuel storage capability from the current 1990 date to the year 2010.

The increase in McGuire spent fuel storage capacity would be accomplished by replacing the existing 15.50 inch center-to-center high density non-poison racks with 10.40 and 9.125 inch center-to-center neutron absorbing racks in Region 1 and Region 2 respectively.

Duke's analysis summary of the proposed amendment request is set forth in Attachment 1. Such analysis addresses the areas addressed in the NRC's Guidance on Spent Fuel Pool Modifications dated April 14,1978 (revised January 18, 1979).

The following evaluation demonstrates by reference to the analysis summary contained in Attachment 1 that not one of the three significant safety hazards consideration guidelines are met.

below. Each of the three standards is discussed First Standard Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequence of an accident previously evaluated.

The analysis of this proposed reracking has been accomplished using current NRC Staff accepted Codes and Standards. The results of the analysis meet the specified acceptance criteria set forth in these standards. In addition Duke has reviewed NRC Staff Safety Evaluation Reports for prior PWR rerackings involving two region poison racks to ensure that there are no identified concerns not fully addressed in this submittal.

From our analyses and SER reviews, Duke has identified the following potential accident scenarios: 1) spent fuel cask drop; 2) loss of spent fuel pool forced cooling; 3) seismic event; 4) spent fuel assembly drop; 5) natural disaster; and

6) construction accident. The probability of any of the first five accidents is not affected by the racks themselves; thus, reracking cannot increase the probability of these accidents. As for the construction accident, the proposed McGuire reracking will not involve an increase in probability of any previously evaluated construction accident as accepted construction standards and procedures will be employed.

The consequences of the spent fuel cask drop accident have been evaluated with conclusions on page 48 of Attachment 1. The cask handling crana stops and administrative controls will continue to be used to prevent heavy loads and casks from being moved into the fuel pool area. Thus, the consequences of this type accident will not be significantly increased from previous analyses as described in the McGuire FSAR Section 9.1.2.3.2.

The consequences of the loss of spent fuel pool forced cooling accident have been evaluated (page 52 of Attachment 1). As indicated in Table 6.3-1 (page 53 of Attachment

1) there is ample time to effect repairs to the cooling system or to establish a makeup flow before boiling occurs. The consequences of this type accident will not be significantly increased from previously evaluated accidents by this proposed reracking.

The consequences of a seismic event have been evaluated. The racks were evaluated against the appropriate NRC Standards. The results of the seismic and structural analysis show that the proposed racks meet all of the NRC structural acceptance criteria and are consistent with results found acceptable by the NRC Staff in all previous two region poison rerack SERs. Thus, the consequences of seismic events will not significantly increase from previously evaluated seismic events.

The consequences of a spent fuel assembly drop accident are described on page 54 of Attachment 1. The radiological consequences for this type accident are unchanged from previous analyses presented in the McGuire FSAR Section 15.5.9, and Keff is shown to be always less than the NRC acceptance criteria of 0.95.

Thus, the consequences of this type accident will not be significantly increased from previously evaluated spent fuel assembly drop accidents.

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o The consequences of a construction accident are described on page 46 of Attachment 1. Since there will be no fuel assemblies in the Unit 2 fuel pool during rack installation, there will be no radiological consequence of any construction accident. The likelyhood of a construction accident is minimized through use of accepted construction practices. The consequences of a postulated construction accident in the Unit I spent fuel pool are well below 10 CFR 100 limits, and are bounded by the natural disaster analysis results.

The consequences of a natural disaster are described on page 49 of Attachment 1.

Analyses have shown that the maximum doses due to tornado missile damage in Region 1 are below 10 CFR 100 limits. Radiological consequences in Region 2 will be kept below 10 CFR 100 limits by administrative controls on fuel placement and decay. The consequences of this type accident will not significantly increase from previously evaluated natural disaster analyses.

It is shown that the proposed McGuire spent fuel pool reracks will not involve a significant increase in the probability or consequences of an accident previously evaluated.

Second Standard Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

Duke has evaluated the proposed reracking in accordance with "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate Industry Codes and Standards. In addition, Duke has reviewed previous NRC Safety Evaluation Reports for two region poison rerack applications. In Duke's analysis and review of NRC evaluations and Industry Standards and Codes. Duke finds that the proposed reracking does not in any way create the possibility of a new or different kind of accident previously evaluated.

Third Standard Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The issue of margin of safety when applied to a reracking modification will need to address the following areas (as established by the NRC Staff Safety Evaluation review process):

1. Nuclear criticality considerations

2. Thermal-hydraulic considerations

3. Mechanical, material, and structural considerations The margin of safety that has been established for nuclear criticality considerations is that the neutron multiplication factor in the spent fuel pool remains less than or equal to 0.95, including all uncertainties, under all conditions. The criticality analysis for the proposed modification is discussed on pages 14-16 of Attachment 1.

The methods utilized in the analysis conform with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants"; ANSI N210-1976, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations"; ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety"; NRC Standard Review Plan, Section 9.1.2, " Spent Fuel Storage", and the NRC guidance, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications".

The results of this analysis indicate that K gg e is always less than 0.95 including all uncertainties at a 95/95 probability / confidence level, thus meeting the acceptance cr1teria for criticality. The proposed rcrack therefore does not involve a significant reduction in the margin of safety for nuclear criticality.

For consideration of thermal-hydraulics, the areas of concern when evaluating if there is a significant reduction in margin of safety are: 1) maximum fuel temperature, and 2) the increase in temperature of the water in the pool. The thermal-hydraulic evaluation is described on pages 17-19 of Attachment 1.

Results of these analyses show that fuel cladding temperatures under abnormal conditions are sufficiently low to preclude failures and that boiling does not occur in the water channels between the fuel assemblies nor within the storage cells. Additionally, the existing spent fuel cooling system will provide the capacity to maintain an acceptable temperature range for normal and abnormal heat loads. The cooling system is described in the McGuire FSAR Section 9.1.3.2. Thus, there is no significant reduction in the margin of safety from a thermal-hydraulic standpoint or from a spent fuel cooling standpoint.

The mechanical,' material, and structural considerations of the proposed rerack are described. The racks are designed in accordance with "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and revised January 18, 1979. The racks are designed to Sciamic Category 1 requirementa and are classified as ANS Safety Class 3 and ASME Code Class 3 Component Support Structures. In addition, the racks are designed to withstand the loads which may result from fuel handling accidents and from the maximum uplift force of the fuel handling crane. The materials utilized are compatible with the spent fuel pool enrichment and the spent fuel assemblies. The structural considerations of the racks provide for a sufficient margin of safety against tilting that the racks do not impact each other nor impact the pool walls, and that sufficient clearance is provided to prevent the racks from sliding into pool floor obstructions. Structural integrity of the pool structure is maintained with additional dead load, live I

load, thermal load, wind load, and seismic load considerations; thus, the margin of safety is not significantly reduced by the proposed rerack.

It has been shown that the proposed McGuire spent fuel pool rerack modification does not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) Involve a significant reduction in a margin of safety.

As such. Duke has determined and submits that the proposed rerack described herein does not involve a significant hazard.