ML20204K097

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Proposed Tech Specs Supporting Spent Fuel Pool Rerack to Maintain Full Core Reserve Capability Approaching End of OL
ML20204K097
Person / Time
Site: Millstone Dominion icon.png
Issue date: 03/19/1999
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20137D872 List:
References
NUDOCS 9903300234
Download: ML20204K097 (65)


Text

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03/24/94 R#

um LIMITING CONDITIONS FOR' OPERATION AND_ SURVEILLANCE REOUIREMENTS.

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3/4.9.6 REFUELING MACHINE........................................

3/4.9.7 CRANE TRAVEL SPENT FUEL STORAGE AREAS..................

3/4 9-6' RESIDUAL HEAT REMOVAL AND COOLA 3/4 9-7 3/4.9.8 H igh Wate r Level.............. NT C I RCULATION Low Water Leve1...............

3/4 9-b 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM......

3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL.............................

3/4.9.11 WATER LEVEL - STORAGE POOL..............................

3/4 9-11 3/4.9.12. FUEL BUILDING EXHAUST FILTER SYSTEM......................

3/4 9-12 3/4 9-13 3/4.9.13 SPENT FUEL POOL - REACTIVITY.............................

3/4 9-16

.14 SPENT FUEL POOL - STORAGE PATTERN......J.................

3/4 9-17 F

RL3.9-1 FUEL ASSEMBLY MINIMUM BURNUP VtK5US INITIAL UZ35 k i M cHMENT FOR STORAG NEIGH II " T FUEL RACKS..

3/4'9-18 FIGURE 3.9-2 RESIOT 1 inREE CF F0iin FUEL ASSEMBLY LOAD SCHEMATIC FOR A TYPICAL 6X6 STORAGE MODULE............

3/4 9-19 3/4.10 - SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN..........................................

3/4 10-1

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3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS Four Loops 0parating..........................

Three Loops Operating......J.............................

3/4 10-2 3/4 10-3 3/4.10.3 PHYSICS TESTS...............................

3/4 10-4 3/4.10.4 REACTOR COOLANT L00PS....................................

3/4 10-5 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-6 3/4.11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration........................

Dose - Liquids.......................

3/4 11-1 3/4 11-2 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................

3/4 11-3 Dose - Noble' Gases.......................

3/4 11-4 Dose - Radiciodines, Radioactive Material in Particulate Form and Radionuclides Other Than Nob 3/4.11.3. TOTAL D0SE........................... l e G a s es..

3/4 11-5 i

3/4 11-6

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9903300234 990319 7 PDR ADOCK 05000423/

,.P PDR j MILLSTONE - UNIT 3 xii Amendment No. y, 89, es4t J

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INSERT A 1

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PAGE FIGURE 3.9-1 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL 3/4 9-18 INITIAL ENRICHMENT FOR REGION 1 4-OUT-OF-4 STORAGE CONFIGURATION -

FIGURE 3.9-2 REGION 1 3-OUT-OF-4 STORAGE FUEL ASSEMBLY 3/4 9-19 LOADING SCHEMATIC FIGURE 3.9-3 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL 3/4 9-20 INITIAL ENRICHMENT FOR REGION 2 STORAGE CONFIGURATION FIGURE 3.9-4 MINIMUM FUEL ASSEMBLY BURNUP AND DECAY TIME 3/4 9-21 VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 3 STORAGE CONFIGURATION 4

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.SECTION-E6S1 3/4.7.11 SEALED SOURCE CONTAMINATION

...... B 3/4 7-25 l

3/4.7.12 DELETED l

3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONIT00 4............... B 3/4 7-25 3/4.8 ELECTRICAL POWER SYSTEMS l

3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND i

.ONSITE POWER DISTRIBUTION B 3/4 8-1 3/4.8.4-

. ELECTRICAL EQUIPMENT PROTECTIVE DEVICES.........

B 3/4 B-3 3/4.9' REFUELING OPERATIONS l

l 3/4.9.1 BORON CONCENTRATION...................

B 3/4 9-1

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3/4.9.2 INSTRUMENTATION..................... B 3/4 9-1 3/4.9.3 DECAY TIME B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.............

B 3/4 9-1

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3/4.9.5 COMMUNICATIONS B 3/4 9-1 l

3/4.9.6 REFUELING MACHINE......

B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS.........B 3/4 9-2 i

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION......

B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM B 3/4 9-7 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL B 3/4 9-8 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM...........

B 3/4 9-8 3/4.9.13 SPENT FUEL POOL - REACTIVITY B 3/4 9-8 3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN............ B3/49[

l 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN..................... B 3/4 10-1 j

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS

. B 3/4 10-1 l

3/4.10.3

-PHYSICS TESTS

...................... B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS.................. B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN.......... B 3/4 10-1

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MILLSTONE - UNIT 3 xy Amendment No. #, 77, J99,197, JJP,13 osw

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January 3,1995 VENTING l

1.39 VENTING shall be the controlled process of discharging air or gas from a confinemer.t to maintain temperature, pressure, humidity, concentration other operating condition, in such a manner that replacement air or gas, or is not provided or required during VENTING. Vent, used in systen namese does not imply a VENTING process..

j[ h c4q,M. AseM SPENT FUEL POOL STORAGE PATTERNS:

1.40 e ion I spent fuel racks contain a cell blocking device in every 4th location iticality control. This 4th location will be refe s

the blocked locat STORAGE PATTERN refers to the ocation and all adjacent and diagona cell loca surrounding the blo'cked location. Boundary configuration betw _

on I and Region II must have cell blockers positioned in est row o Regiot. I perimeter, as shown in Figure 3.9-2 1.

gion II contains no cell blockers.

CORE OPERATING LIMITS REPORT (C0lR) 1.42 The CORE OPERATING LIMITS REPORT COLR) is the unit-specific document

-) that provides core operating limits fo(r the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6.

Unit Operation within these operating limits is addressed in individual specifications.

l Att0WED POWER LEVEL HD 1.43 APL is the minimum allowable nuclear design power level for base load operation and is specified in the COLR.

Il 1.44 APL is the maximum allowable power level when transitioning into base load operation.

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MILLSTONE - UNIT 3 1-7 Amendment No. U, S. M. 71,100

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i STQRAGE PATTERN 1.40 STORAGE PATTERN refers to the blocked location in a Region 1 fuel storage rack and all adjacent and diagonal Region 1 (or Region 2) cell locations surrounding the blocked location. The blocked location is for criticality control.

3-OUT-OF-4 and 4-OUT-OF-4 1.41 Region 1 spent fuel racks can stord fuel in either of 2 ways:

(a)

Areas of the Region 1 spent fuel racks with fuel allowed in every storage location are referred to as the 4-OUT-OF-4 Region 1 storage area.

(b)

Areas of the Region 1 spent fuel racks which contain a cell blocking device in every 4th location for criticality c.ontrol, are referred to as the 3-OUT-OF-4 Region 1 storage area. A STORAGE PATTERN is a subset of the 3-OUT-OF-4 j

Region 1 storage area.

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4/9/90 REFUELING OPERATIONS-j BORON CONCENTRATION

' LIMITING ' CONDITION FOR OPERATION m

x The bNon concentration of the Spent Fuel Poci.shall be 3.9.1.

uniform and sufficient to. ensure that the boron concen ater than or equal to 1750 ppm.

9 Apolicability l

Whenever fuel;assembl are in the spent fuel pool.

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Action With the boron concentrati a.

the boron concentration in thless than 1750 ppm, initiate actiun' to bring fuel pool to at least 1750 ppm within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and

.b.

With the boron concentration less th all fuel assemblies within the spent fu1750 ppm, suspend ths movement of fuel racks.

pool and loads over the spent SURVEILLANCE REQUIREMENTS 1

Verify that the boron concentration in the fue\\

4.9.1.2 or equal to 1750 ppm every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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HILLSTONE'- UNIT 3 osso 3/4 9-la Amendment No. M.

158' 4.

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l-3.9.1.2 The soluble boron concentration of the Spent Fuel Pool shall be maintained l

i uniform, and greater than or equal to 800 ppm.

l Anoticability During all fuel assembly movements within the spent fuel pool.

Action With the spent fuel pool soluble boron concentration less than 800 ppm, suspend the movement of all fuel assemblies within the spent fuel pool.

Surveillance Reauirements 4.9.1.2 Verify that the soluble boron concentration is greater than or equal to 800 ppm prior to any movement of a fuel assembly into or within the spent fuel pool, and every 7 days thereafter during fuel movement.

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October 25, 1990 i

REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2200 pounds shall be prohibited from travel over fuel assemblies in. the storage pool.

APPLICABILITY: With fuel assemblies in the storage pool.

ACTION:

a.

With the requirements of the above specification not satisfied, place the crane load in a safe condition, b.

The provisions of Specification 3.0.3 are not applicible.

SURVEILLANCE REOUIREMENTS

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Crane interlocks and physical st%opsL)shn v poo/

4.9.7 ic preven awTravel wit loads-in excess of 2200 pounds overtid...eius shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.

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MILLSTONE - UNIT 3 3/4 9-7 AMENDMENT NO. 57

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SPENT FUEL POOL - REACTIVITY LINITING.Cor0ITION FOR OPERATION 3.9. 3 The Reactivity condition of the Spent Fuel Pool shall be such that keff is less than or equal to 0.95 at all times.

APEL_CABILIM: Whenever fuel assemb' lies are in the spent fuel pool.

ACT10%

vith.keff greater than 0.95:

1 Borate the Spent Fuel Pool until k a.

isless'thanorequalto/ 4 ad eff

& S e5 b.

Initiate action to correct the cause of the misplaced / dropped

'_Lt>S fy1Lassembly, if required, and -

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f Following the drop of a ioao on the speht fuel racks, with fuel c.

in th'e fuel rack location, close and administratively control opening of. dilution pathways to the Spent Fuel Pool until Bora-in the Spent Fuel Pool is determined to be within design, an

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Following a seismic.

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magnitude or greater: nt of Operating Basis Earthquake (OBE) 1)

Close and administratively trol the opening of dilution '

pathways to the Spent Fuel Poo til Boraflex in the 1

Spent Fuel Pool is determined to be in design.

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2)

Notify the Commission of the action taken fnr nt Fuel Reactivity control as part of the report required

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- Specification 4.3.3.3.2.

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SURVEILLANCE REQUIREMENTS 4.9.133 Ensure that all fuel assemblies to be placed in Region II of the T,. 9 by checkinguel assembly's design and burn-up documentation.

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.13.2 Following a seismic event or operating Basis Earthquake (OBE)

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magni k rf is lesor greater, perform an engineering evaluation to determine that an or equal to 0.95 and that soluble boron is not required e

for control of k

'n the Spent Fuel Pool-. Pending completion of e

i engineering evalua ion, ke action as required for k

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0.95.

eff being greater than 4.9.13.3 ' Following the drop of a loa the fuel rack location, perform an engineerinhe Spent Fuel Racks, with fuel in eff is less than or equal to 0.95 and that solubleluation to determine that k

on is not required

,;, for control of keff in the Spent Fuel Pool.

Pending comp

'on of engineering evaluation, take action as required for k ry being 0.95.

e ater than, J

~ MILLSTONE - UNIT 3

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0561-3/4 9-16 Amendment No. 77, 1.58

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4.9.13.14.1 Ensure that all fuel assemblies to be placed in Region 1 "4-OUT-OF-4" fuel storage are within the enrichment and burnup limits of Figure 3.9-1 by checking the fuel assembly's design and burn-up documentation.

4.9.13.1h.L Ensure that all fuel assemblies to be placed in Region 2 fuel storage are t

within the enrichment and bumup limits of Figure 3.9-3 by checking the fuel assemtiy's design and burn-up documentation.

4.9.13.14 3 Ensure that all fuel assemblies to be placed in Region 3 fuel storage are within the enrichment, decay time, and burnup limits of Figure 3.9-4 by checking the fuel assembly's design, decay time, and burn-up documentation.

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b. Initiate immediate action to move any fuel' assembly which does not meet the requirements of Figures 3.9-1,3.9-3 or 3.9-4, to a location for which that fuel -

assembly is allowed.

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August 29, 1989 REFUELTNG OPERATIONS q

SPENT FUEL POOL - STORAGE PATTERN LIMITING CONDITION FOR OPERATION 3.9.14 > Each STORAGE PATTERN of the Region 1 spent fuel pool racks chall' 4

require that:

a.

Prior to storing fuel assemblies in the STORAGE PATTERN. per Figure 3.9-2, the cell blocking device for the cell location must be installed.

b.

Prior to removal of a cell blocking device from the cell location per Figure 3.9-2, the STORAGE PATTERN must be vacant of all stored fuel assemblies.

APPLICABILITY: Whenever fuel assemblies are in the spent fuel pool.

ACTION: Take immediate action to comply with 3.9.14(a), (b).

SURVEILLANCE RE0VIREMENT

. )

4.9.14 Verify that'3.9.14 is satisfied with no fuel assemblies stored in the STORAGE PATTERN prior to installing and removing a cell blocking device in the spent fuel racks.

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1 MILLSTONE - UNIT 3 3/4 9-17 Amendment No. 39

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Figure 3.9-1 l

MILLSTONE UNIT 3 FUEL ASSEMBLY MINIMUM BURNUP VS INITIAL U 5

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ENRICHMENT FOR STORAGE IN REGION 11 SPENT FUEL RACKS 1

Mill. STONE - UNIT 3 3/4 9-18 Amendment No. 39 i

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l4 FIGURE 3.91 Minimum Fu:1 Accambly Burnup V:rsus N min 21Initici Enrichm:nt for Region 1 4-OUT-OF-4 Fuel Storage Configuration 8

7 ACCEPTABLE DOMAIN 6

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3.50 3.75 4.00 4.25 4.50 4.75 5.00 Inital Fuel Enrichment ( w/o U-235)

Page 3/4 9-18

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1g!l August 29. 1989 la at

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is face must be along the vall the spent fuel pool, or other

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R ion 1 modules.

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Region II fuel may be 7his face must be along the placed along this face wall of the spent fuel, pool, or other Region I modules.

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Region II fuel may be placed along this face.

Fue Assembly Location Cell B eker Location Figure 3.9-2 MILLSTONE UNIT 3 REGION I THREE OF FOUR FUEL ASSEMB g

LOADING SCHEMATIC FOR A TYPICAL 6 X 6 STORAGE MODULE MILLSTONE - UNIT 3 3/4 9-19 Amendment No. 39 I

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/6 Region 2 or Region 14-OUT-OF-4 may be placed along this face X

X X

This face must be along j

the wallof the spent fuel pool, or other Region 1 Region 2 or Region 14-OUT-OF-4 3-OUT-OF-4 storage may be placed along this face X

X X

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This face must be along'the wall of the spent fuel pool, or other Region 1 3-OUT-OF-4 storage Cell Blockerlocation Fuel Assembly Storage location FIGURE 3.9-2

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REGION 13-OUT-OF-4 STORAGE FUEL ASSEMBLY LOADING SCHEMATIC 3/4 9-19 e

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FIGURE 3.9-3 Minimum Fuel A:sembly Burnup Vsraus N: min:lIniti:1 I

Enri:hment for Region 2 Storage Configuration 40 f

35 ACCEPTABLE DOMAIN 4

30 s

25

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q 20 e

15 10 5

0 2.0 2.5 3.0 3.5 4.0 4,5 5.0 Initial Fuel Enrichment ( w/o U-235 }

Page 3/4 9-20

- FIGURE 3.9-4 Minimum Fu;l Aemmbly Curnup cnd Decry Tims Vrrsua N:minil Initial Enrichment for Region 3 Storage Configuration 60-50 7

ACCEPTABLE DOMAJN 40

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-o-5 year decay time g

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-*- 10 year decay time E

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0 2.00 2.50 3.00 3.50 4.00 4.50 5.00 Initial Fuel Enrichment ( w/o U 235 )

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Page 3/4 9-21 2

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4/9/98 3/4.9 REFUELING OPERATIONS q

BASES' i

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gg,g 3/4b BORON CONCENTRATION

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v (1) the reactor willThe limitations on reactivity conditions during REFUE I

remain suberitical during CORE ALTERATIONS, uniform boron concentration is maintained for reactivity control in the w(2) and volume having direct access to the reactor vessel.

ater

'for K rr includes a e

1%

Ak/k conservative allowanceThe value of O'99 or les Similarly, the boron concentration value of 2600 for uncertainties.

conservative uncertainty allowance of 50 ppm boron. ppa ot greater includes a boron concentration measurement uncertainty between the spent fue The 2600 ppm provides for the RWST.

'The locking closed 'of the required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portion of the RCS.

water by closing flow paths from sources of unborated water.This action 3/4.9.1.2 Boron Concentration in Soent Fuel Pool

{k sad R maintaining KOuring nomal Spent Fuel Pool operation, the spent tuel r env% ment due to the geometry of the rack spacing and the neutron is a pessibi iber in the spent fuel racks.

Seismic analysis has shown that there event greater in mag

  • ude than an Operating Basis Earthq 1500 ppm boron in Spent At least k

.)

event could cause a loss of B Pool is required in anticipation that a seismic Boraflex,. a single misplaced fuel ex integrity.

If, in addition to a loss of 1750 ppm boron is required. The 1750 ppe21y is postulated, then a minimum of conditions for a loss of all Boraflex in the fueconcentration requirement bounds i

eks.

j The boron requirement in the spent fuel pool also ensur of a fuel assembly handling accident involving either a dropped or hat in the event assembly, the K

  • ' laced fuel-eff of the spent fuel storage rack will remain less than 0.95.

equal 3/4.9.2 INSTRUMENTATION

/

The OPERABILITY of the Source Range Neutron Flux Monitors ensure redundant monitoring capabilit is available to detect changes in the reactivity condition of the core.y 3/4.9.3 DECAY TIME

- The minimum requirement for reactor subcriticality prior to movement o irradiated fuel assemblies in the reactor vessel ensures that sufficien

'Fas el apsed to allow-the radioactive decay of the short-lived fission products. This decay time safety analyses.

is consistent with the assumptions used in the i-L

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3/4 9.1.2 Boron Concentration in Soent Fuel Pool During normal spent fuel pool operation, the spent fuei eacks cre capable of maintaining K,n t less than or equal to 0.95 in an unborated water environment. This is a

accomplished in Region 1,2, and 3 storage racks by the combination of geometry of the rack spacing, the use of fixed neutron absorbers in some fuel storage regions, the limits on fuel bumup, fuel enrichment and minimum fuel decay time, and the use of blocking devicas in certain fuel storage locations.

The boron requirement in the spent fuel pool specified in 3.9.1.2 ensures that in the -

event of a fuel assembly handling accident involving either a single dropped or misplaced fuel assembly, the Ken f the spent fuel storage racks will remain less than or o

equal to 0.95.

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I 4/9/98 REFUELING OPERATIONS g

BASES ~

) 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL The restrictions on minimum water level ensure that sufficient wate

.is available to remove 99% of the assumed 10% iodine gap activity re rupture of an irradiated fuel assembly. The minimum water depth is cons the assumptions of the safety analysis.-

3/4.9.12 FUEL Bull 0TNG EXHAUST FITTER SYSTEM i

The limitations on the Fuel Building Exhaust Filter System ensure t radioactive iodine released from an irradiated fuel assembly and storag will be filtered through the HEPA filters and charcoal adsorber prior to the atmosph'ere.

Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the bu moisture on the adsorbers and HEPA filters.

of the safety analyses.and the resulting iodine removal capacity are ANSI N510-1980 The heater W measured mustwill be used as a procedura for surveillance testing.

nameplate rating.

be corrected to its Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the' proportional to the square of the voltage. The filtration system removes output kW is radioiodine following a fuel handin would not be removed by the system. g or heavy load drop accident.

Noble gases Iodine-131 has the longest half-life:Other radionuclides would storage pool water.

decay time, there is essentially negligible iodind and filtration is unne

~8 days. After 60 days 4.9.lUPEIif FUEL POOL M CTIVITY_

~

h The limitations described by ' Figure 3.9-1 ensure that the reactivity o fuel a emblies introduced into Region II are conservatively with.in the assumpti of the safety analysis.

Administ tive controls have been developed and instituted to verify that the enrichment an burn-up limits of Figure 3.9-1 have been maintained for the k6 fuel assembly.

W During normal Spent maintaining k pt at less than1 Pool operation, the spent fu'dl racks are capable of geometry of the rack spacing 3d.95 in an unborated water environment due to the e

spent fuel racks. Oue to radiatione presence of Boraflex neutron absorber in the 6j that the Boraflex absorber could degrduced embrittlement, there is a possibility j

1500 ppm boron in the Spent Fuel Pool is following a seismic event. ~ At least event could cause a complete loss of all Borautred in anticipation that a seismic

Boraflex, ex.

If, in addition to a loss of.

a single misplaced fuel assembly is 1750 ppm boron is required. The 1750 ppm boron conostulated, then a minimum of i

conditions for a loss of all Boraflex in the fuel rackntration requirement bounds The action requirements of this specificatier. recogni a seismic event which could degrade the Boraflex neutron absorber 'n the sp the possibility of racks.'

Seismic analysis has shown that there is a possibility tha sorber could degrade following a seismic event greater in magnitude than an 1

he Boraflex

~

x.

a MILLSTONE - UNIT 3 0663 B 3/4 9-8 Amendment No. M. JpE. JP7. DO

l' BEFUELING OPERATIONS 4/9/98 3)

BASES

' W C 9.13 SPENT FUEL POOL - REACTIVITY fcontinueh ig Opera a4s Earthquake (OBE).

(

the OBE level or greater, which is approximately on M

a seismic even Safe Shutdown Earth condition of the Boraflex. (SSE). level, action will be taken to determine the p

has o'ccurred, then the boron ine a seismic event of greater than or equal to an OBE S eat Fuel Pool will Le credited to maintain k rf less than or eg' al to 0.95. The, e

u to the Spent Fuel Pool be closed and 'adminiification requires that dilution paths can be inspected and the condition of the Botively controlled until the racks specification also assumes that piping systems externaex can be determined.

The mounted such that they remain leak tight following an earo the Spent Fuel Pool are of an SSE, or will not direct water into the Spent Fuel Pool sake up to the level have been isolated from flow to prevent leakage into the Spent Fued they leak, ol.

SPENT FUEL POOL - STORAGE PATTERN f-J 3/4.9.14 J

,i; The limitations offtTtis specification e[isure.that.the conditions of the Region % storage racks and spent fuel pool k rp will re reactivity less than or equal to 0.95.

e I

3-00T-0P-9 The cell Blocking Devices in the 4th location of the Re racks are designed to prevent inadvertent placement and/or storage of fuel assemblies in the blocked locations. The blocked location remains sprovide the jadjacent and diagonal locations of the STORAGE PATTERN. flux tra empty to expanded from the walls of theSTORAGE PATTERN for the Region [st rag and control of the R*;Mpent_ _ fuel pool per Fioure 3.9-2 to ensure definition Iyegica -11 Laun63 and ainimize the number of boundaries where a fue ispiacement incident can occur.

h 'o,g 00T-0F-9 boviada3 oko-S4=raf 106 x

/'T V

HILLSTONE - UNIT 3 B 3/4 9-9 Amendment No. 79. J95. Jp3 158

nn w yve in k;

cd i

(

INSERT G j

i 3/4 9.13 Soent Fuel Pool Reactivity During normal spent fuel pool operation, the spent fuel racks are capac!e of maintaining i

K.s t less than or equal to 0.95 in an unborated water environment.

a l

Maintaining K n at less than or equal to 0.95 is accomplished in Region 1 3-OUT-OF-4 4

storage racks by the combination of geometry of the rack spacing, the use of fixed neutron absorbers in the racks, a maximum nominal 5 weight percent fuel enrichment, i

and the use of blocking devices in certain fuel storage locations, as specified by the i

interface requirements shown in Figure 3.9-2.

Maintaining K.,at less than or equal to 0.95 is accomplished in Region 1 4-OUT-OF-4 storage racks by the combination of geometry of the rack spacing, the use of fixed neutrcn absorbers in the racks, and the limits on fuel enrichment / fuel bumup specified in Figure 3.9-1.

Maintaining Ken t less than or equal to 0.95 is accomplished in Region 2 storage racks a

by the combination of geometry of the rack spacing, the use of fixed neutron absorbers i

in the racks, and the limits on fuel enrichment / fuel bumup specified in Figure 3.9-3.

Maintaining K.n t less than or equal to 0.95 is accomplished in Region 3 storage racks i

a

.g 1

by the combination of geometry of the rack spacing, and the limits on fuel enrichment / fuel burnup and fuel decay time specified in Figure 3.9-4. Fixed neutron i

absorbers are not credited in the Region 3 fuel storage racks.

The limitations described by Figures 3.9-1,3.9-2,3.9-3 and 3.9-4 ensure that the reactivity of the fuel assemblies stored in the spent fuel pool are conservatively within the assumptions of the safety analysis.

Administrative controis have been deve!oped and instituted to verify that the fuel enrichment, fuel bumup, fuel decay times, and fuel interface restrictions specified in Figures 3.9-1, 3.9.2, 3.9-3 and 3.9-4 are complied with.

)

March 11. 1991 OESIGN FEATURE' 3

5.6 FUEL STORAGE LRITICALITY 5.6.1.'1 T e spent fuel storage racks are designed and shall be maintaine with:

A k [ted uivalent to less than or equal to 0.95 when flooded with a.

unbM ter.

b.

A nominal 1. 5-inch center-to-center' distance between fuel assemblies placed ~ the storage racks.

c.

Fuel assemblies stored maximum nominal. fuel enricRegior I of the. spent fuel pool may have a ent of up to 5.0 we'ight percent U Region I is designed to pe storage of fuel in a 3-out $N4 array with the 4th storage loc

'on blocked as shown in Figure 3.9-2.

d.

Fuel assemblies stored in Region II of the nt fuel pool may have a maximum nominal fuel enrichment of up to conditional upon compliance with Figure 3.9-1 to weight percent, 3

sure that the f

design burnup of the fuel has been sustained.

) DRAINAGE 5.6.2' The spent fuel storage pool is 2esigned and shall be maintained to prevent inadvertent draining of the pool below elevation 45 feet.

j CAPACITY

~

5

.3 The e nt M 1 stcr:g p001 conrains 755 stcrage. l@eatttf65~BT whrc f 100 le::tient wol be bhcked.

a a

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified maintained within the cyclic or transient limits of Table 5.7-1.in Table 5.7-1 a g(

Si~E i

MILLSTONE - UNIT 3 5-6

..)

Amendment No. R 60

mmm,y INSERT H 5.6.1.1 The spent fuel surage racks are made up of 3 Regions which are designed and shall be maintained to ensure a Konless than or equal to 0.95 when flooded with unborated water. The 3 storage rack Regions are:

}

a. Region 1, a nominal 10.0 inch (North / South) and a nominal 10.455 inch (East / West) center to center distance, credits a fixed neutron absorber (BORAL) within the rack, and can store fuel in 2 storage configurations:

(1) With credit for fuel burnup as shown in Figure 3.9-1, fuel may be stored in a "4-OUT-OF-4" storage configuration.

(2) With credit for every 4th location blocked and empty of fuel, fuel up to 5 weight percent nominal enrichment, regardless of fuel burnup, may be 3

stored in a "3-OUT-OF-4" storage configuration. Fuel storage in this configuration is subject to the interface restrictions specified in Figure 3.9-2.

b. Region 2, a nominal 9.017 inch center to center distance, credits a fixed neutron absorber (BORAL) within the rack, and with credit for fuel burnup as shown in Figure 3.9-3, fuel may be stored in all available Region 2 storage locations.
c. Region 3, a nominal 10.35 inch center to center distance, with credit for fuel

)

bumup and fuel decay time as shown in Figure 3.9-4, fuel may be stored in all wailable Region 3 storage locations. The Boraflex contained inside these storage racks is not credited.

5.6.3 The spent fuel storage pool contains 350 Region 1 storage locations,673 Region 2 storage locations and 756 Region 3 storage locations, for a total of 1779 total available fuel storage locations. An additional Region 2 rack with 81 storage locations may be placed in the spent fuel pool, if needed. With this additional rack installed, the Region 2 storage capacity is 754 storage locations, for a total of 1860 total available fuel storage locations I

(

O h

Cko

Docket No. 50-423 B17343 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Spent Fuel Poo; Rerack (TSCR 3-22-98)

Retyped Pages March 1999

U.S. Nucliar Regulatory Commission B17343/ Attachment 2/Page 1 RETYPE OF PROPOSED REVISION Refer to the attached retype of the proposed revision to the Technical Specifications.

The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed retype.

The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

INDEX

-LIMITING CONDITIONS FOR OPERATION AM SURVEILLANCE REQUIREMENTS SECTION PME 3/4.9.6 REFUELING MACHINE.................... 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS 3/4 9-7 3/4.9.8 RESIDUAL-HEAT REMOVAL AND COOLANT CIRCULATION High Water Level 3/4 9-8 Low Water level 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION. SYSTEM 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL 3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE POOL 3/4 9-12 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM 3/4 9-13 3/4.9.13 SPENT FUEL POOL - REACTIVITY 3/4 9-16 3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN 3/4 9-17 FIGURE 3.9-1 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 1 4-0VT-0F-4 STORAGE CONFIGURATION....................

3/4 9-18 FIGURE 3.9-2 REGION 1 3-0VT-0F-4 STORAGE FUEL ASSEMBLY LOADING SCHEMATIC......................

3/4 9-19 FIGURE 3.9-3 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 2 STORAGE CONFIGURATION,...

3/4 9-20 FIGURE 3.9-4 MINIMUM FUEL ASSEMBLY BURNUP AND DECAY TIME VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 3 STORAGE CONFIGURATION....................

3/4 9-21 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN 3/4 10-1 3/4.10.2 GROUP HEIGHT,-INSERTION, AND POWER DISTRIBUTION LIMITS j

Four Loops Operating 3/4 10-2 Three Loops Operating 3/4 10-3 l

3/4.10.3 PHYSICS TESTS 3/4 10-4 3/4.10.4 REACTOR COOLANT LOOPS 3/410-5 3/4.10.5 POSITION INDICATION SYSTEM 3/410-6 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration 3/4 11-1 Dose - Liquids 3/411-2 3/4.11.2 GASEOUS EFFLUENTS Dose Rate 3/4 11-3 Dose - Noble Gases 3/411-4 Dose - Radiciodines, Radioactive Material in Particulate Form and Radionuclides Other Than Noble Gases 3/4 11-5 3/4.11.3 TOTAL DOSE 3/411-6 MILLSTONE - UNIT 3 xii Amendment JJ, JJ, osa

l!lDil BASES SECTION 3/4.7.11 SEALED SOURCE CONTAMINATION

........... B 3/4 7-25 3/4.7.12 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING............... B 3/4 7-25 3/4.8 ELECTRICAL POWER SYSTEMS.

3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................

B 3/4 9-1 3/4.9.2 INSTRUMENTATION.....................

B 3/4 9-1 3/4.9.3 DECAY TIME B 3/4 9-1 j

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS............

B 3/4 9-1 3/4.9.5 COMMUNICATIONS B 3/4 9-1 3/4.9.6 REFUELING MACHINE....................

B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS.........

B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION......

B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM B 3/4 9-7 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL B 3/4 9-8 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM...........

B 3/4 9-8 3/4.9.13 SPENT FUEL P00L - REACTIVITY B 3/4 9-8 3/4.9.14 SPENT FUEL P0OL - STORAGE PATTERN............

B 3/4 9-9 1 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN..................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS

, B 3/4 10-1 3/4.10.3 PHYSICS TESTS...................... B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS.................. B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN.......... B 3/4 10-1 MILLSTONE - UNIT 3 xy Amendment No. pp, pp, Jpp, JP/, Up, un yp,

p DEFINITIONS I M U Fa 1.39 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

SPENT FUEL POOL STORAGE PATTERNS:

STORAGE PATTERN 1.40 STORAGE PATTERN refers to the blocked location in a Region 1 fuel storage rack and all adjacent and diagonal Region 1 (or Region 2) cell locations surrounding the blocked location. The blocked location is for criticality control, 3-0VT-0F-4 and 4-0VT-0F-4 1.41 Region 1 spent fuel racks can store fuel in either of 2 ways:

(a)

Arcas of the Region 1 spent fuel racks with fuel allowed in every storage location are referred to as the 4-0VT-0F-4 Region 1 storage area.

(b)

Areas of the Region 1 spent fuel racks which contain a cell blocking device in every 4th location for criticality control, are referred to as the 3-0VT-0F-4 Region 1 storage area. A STORAGE PATTERN is a subset of the 3-0VT-0F-4 Region 1 storage area.

CORE OPERATING LIMITS REPORT (COLR) 1.42 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6.

Unit Operation within these operating limits is addressed in individual specifications.

ALLOWED POWER LEVEL ND 1.43 APL is the minimum allowable nuclear design power level for base load operation and is specified in the COLR.

1,44 APL is the maximum allowable power level when transitioning into base load operation.

NILLSTONE - UNIT 3 1-7 Amendment No. 77, pp, pp 77, Jpp, 0629

r-l-

f-REFUELING.0PERATIONS B0RON CONCEN' RATION LIMITING COMITION FOR OPERATION l

3.9.1.2 The soluble boron concentration of the Spent Fuel Pool shall be maintained uniform, and greater than or equal to 800 ppm.

Aeolicability l

i During all fuel assembly movements within the spent fuel pool.

Action i

With the spent fuel pool soluble boron concentration less than 800 ppm, suspend the movement of all fuel assemblies within the spent fuel pool.

i l

SURVEILLANCE REQUIRENENTS 4.9.1.2 Verify that the soluble boron concentration is greater than or equal l

to 800 ppm prior to any movement of a fuel assembly into or within l

the spent fuel pool, and every 7 days thereafter during fuel l g, movement.

l l

\\

l i

MILLSTONE - UNIT 3 3/4 9-la Amendment No. Jg JJp, 0028 l

4 REFUELING _0PIRATIONS 3/4.'9.7 CRANE TRAVEL - SPENT FUEL STORAGE' AREAS LINITING COMITION FOR OPERATION l

3.9.7 Loads in excess of 2200 pounds shall be prohibited from travel over

. fuel assemblies in the storage pool.

APPLICABILITY: With fuel assemblies in the storage pool.

ACTION:

a.

With the requirements of the above specification not satisfied, 4

place the crane load in a safe condition.

b.

The provisions of Specification 3.0.3 are not applicable.

~ SURVEILLANCE REQUIRENENTS i

4.9.7 Crane interlocks and physical stops which prevent crane travel with loads in excess of 2200 pounds over the fuel storage pool shall be demonstrated 1 OPERABLE within 7 days prior to crane use and at least - once per 7 days thereafter during crane operation. Administrative controls may be used in lieu of crane interlocks and physical stops for handling fuel racks, spent fuel pool gates, or loads less than 2200 pounds.

l NILLSTONE - UNIT 3 3/4 9-7 Amendment No. 77, 0621 i

r-i-

REFUELING OPERATIONS I

l l

3/4.9.13 SPENT FUEL P00L - REACTIVITY LINITING COMITION FOR OPERATION

,l' 3.9.13 The Reactivity-Condition of the Spent Fuel Pool shall be such that keff is less than or equal to 0.95 at all times.

l AEPLICABILITY: Whenever fuel assemblies are in the spent fuel pool.

ACTION: With keff greater than 0.95:

l-a.

Borate the Spent Fuel. Pool until keff is less than or equal to 0.95, and b.

Initiate immediate action to move any fuel assembly which does not meet the requirements of Figures 3.9-1, 3.9-3 or 3.9-4, to a location for which that fuel assembly is allowed.

SURVEILLANCE REQUIRENENTS i

4.9.13.1.1. Ensure that all fuel assemblies to be placed in Region 1 "4-0VT-0F-4" fuel storage are within the enrichment and burnup l

limits of Figure 3.9-1 by checking the fuel assembly's design l

and barn-up documentation.

4.9.13.1.2. Ensure that all fuel assemblies to be placed in Region 2 fuel storage are within the enrichment and burnup limits of Figure 3.9-3 by checking the fuel assembly's design and burn-up documentation.

4.9.13.1.3. Ensure that all fuel assemblies to be placed in Region 3 fuel storage are within the enrichment, decay time, and burnup limits of Figure 3.9-4 by checking the fuel asssembly's design, decay l

time, and burn-up documentation.

i 1

1 NILLSTONE - UNIT 3 3/4 9-16 Amendment No. 77. Jpp.

0022

REFUELING OPERATIONS SPENT FUEL POOL - STORAGE PATTERJ -

LIMITING CONDITION FOR OPERATIQM 3.9.14 Each STORAGE PATTERN of the Region I spent fuel pool racks shall I

require that:

a.

Prior to storing fuel assemblies in the STORAGE PATTERN per Figure'3.9-2, the cell blocking device for the cell location must be installed.

b.

Prior to removal of a cell blocking device from the cell location per Figure 3.9-2, the STORAGE PATTERN must be vacant of all stored fuel assemblies APPLICABILITY: Whenever fuel assemblies are in the spent fuel pool.

ACTION: Take immediate action to comply with 3.9.14(a), (b).

SURVEILLANCE REQUIREMENTS 4.9.14 Verify that 3.9.14 is satisfied with no fuel assemblies stored in the

-STORAGE PATTERN prior to installing and removing a cell blocking device in the spent fuel racks.

i MILLSTONE - UNIT 3 3/4 9-17 Amendment No. Jjl, 0622

FIGURE 3.9-1 Minimum Fuel Arcmbly Burnup V;rsu3 N min:1initi:1 Enrichmsnt for Region 1 4-OUT-OF-4 Fuel Storage Configuration 8

7 ACCEPTABLE DOMAIN 6

5

)

5 E

OO.

5 5

Tif 3

2 1

0 3.50 3.75 4.00 4.25 4.50 4.75 5.00 inital Fuel Enrichment ( w/o U-235)

Page 3/4 9-18

Region 2 or Region 14-OUT-OF-4 may be placed along this face

>0 X

X This face must be along the wall of the spent fuel

\\

pool, or other Region 1 f

Region 2 or Region 14-OUT-OF-4 3-OUT-OF-4 storage may be placed along this face X

X X

This face must be along the wall of the spent fuel pool, or other Region 1 3-OUT-OF-4 storage Cell Blockerlocation Fuel Assembly Storage location FIGURE 3.9-2 REGION 13-OUT-OF-4 STORAGE FUEL ASSEMBLY LOADING SCHEMATIC 3/4 9-19

p h

FIGURE 3.9-3 Minimurn Fu:1 Assembly Burnup Versua NominalInitial Enrichment for Region 2 Storage Configuration 40 i

I 35 ACCEPTABLE DOMAIN 30 25 a

1 g

s 20 s

1

?

E u.

/

15 10 5

0 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial Fuel Enrichment ( w/o U-235 )

Page 3/4 9-20

p FIGURE 3.9-4 Minimum Fu:1 A221mbly Burnup cnd Decry Time Varsus Nominal Initial Enrichment for Region 3 Storage Configuration 60 3

l l

ACCEPTABLE DOMAIN

[

40 i

/

/

S E

g j

-+-0 year decay time 0

/

/;

-o-5 year decay time g-

[

-*- 10 year decay time

[

-x-20 year decay time

)

s u.

20

/

c 10 'W

(

m l

l 1

0 2.00

-2.50 3.00 3.50 4.00 4.50 5.00 Initial Fuel Enrichment ( w/o U-235 )

Page 3/4 9-21 L.

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1.1 BORON CONCENTRATION l

The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor.will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.

The value of 0.95 or less for K ra includes a 1%

Ak/k ' conservative allowance for uncertainties.

e Similarly, the boron concentration value of 2600 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. The 2600 ppm provides for:

boron concentration measurement uncertainty between the spent fuel pool and the RWST.

The -locking closed of the required valves during refueling operations precludes.. the possibility of uncontrolled boron dilution of the filled portion of the RCS.

This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.

3/4.9.1.2 Boron Concentration in Soent Fuel Pool During normal spent fuel pool operation, the spent fuel racks are capable of maintaining K,. at less than or equal to 0.95 in an unborated water environment.

This is accomplished in Region 1, 2, and 3 storage racks by the combination of geometry of the rack spacing, the use of fixed neutron absorbers in some fuel storage regions, the limits on fuel burnup, fuel enrichment and.

minimum fuel decay time, and the use of blocking devices in certain fuel storage locations.

The boron requirement in the spent fuel pool specified in 3.9.1.2 ensures that in the event of a fuel assembly handling accident involving either a single dropped or misplaced fuel assembly, the K, of the spent fuel storage racks will remain less than or equal to 0.95.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability - is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is. consistent with the assumptions used in the safety analyses.

i i

l MILLSTONE - UNIT 3 8 3/4 9-1 Amendment No. #, pp, Jpp.

0623

REFUELING OPERATIONS i

BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.

3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM The limitations on the Fuel Building Exhaust. Filter System ensure that all radioactive iodine released from an irradiated fuel assembly and storage pool water will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses.

ANSI N510-1980 will be used as a procedural guide for surveillance testing.

The heater kW measured must be corrected to its nameplate rating.

Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional-to the square of the voltage. The filtration system removes radioiodine following a fuel handing or heavy load drop accident.

Noble gases would not be removed by the system. Other radionuclides would be scrubbed by the storage pool water. Iodine-131 has the longest half-life:

~8 days. After 60 days decay time, there is essentially negligible iodine and filtration is unnecessary.

3/4.9.13 SPENT FUEL POOL - REACTIVITY During normal spent fuel pool operation, the spent fuel racks are capable of maintaining K,n at less than or equal to 0.95 in an unborated water environment.

Maintaining K,n at less than or equal to 0.95 is accomplished in Region 1 3-0VT-0F-4 storage racks by the combination of geometry of the rack spacing, the use of-fixed neutron absorbers in the racks, a maximum nominal 5 weight percent fuel enrichment, and the use of blocking devices in certain fuel storage locations, as specified by the interface requirements shown in Figure 3.9-2.

Maintaining K,n at less than or equal to 0.95 is accomplished in Region 1 4-0VT-0F-4 storage. racks by the combination of geometry of the rack spacing, the use of fixed neutron absorbers in the racks, and the limits on fuel enrichment / fuel burnup specified in Figure 3.9-1.

Maintaining K,n at less than or. equal to 0.95 is accomplished in Region 2 storage racks by the combination of geometry of the rack spacing, the use of fixed neutron absorbers in the racks, and the limits on fuel enrichment / fuel burnup specified in Figure 3.9-3.

Maintaining K,n at less than or equal to 0.95 is accomplished in Region 3 storage racks by the combination of geometry of the rack spacing, and the limits on fuel enrichment / fuel burnup and fuel decay time specified in Figure 3.9-4.

Fixed neutron absorbers are not credited in the Region 3 fuel storage racks.

NILLSTONE - UNIT 3 B 3/4 9-8 Amendment No. 77, Jpp, Jpf, Jpp, 0824

n REFUELING OPERATIONS 1

BASES i

l 3/4.9.13 SPENT FUEL P0OL - REACTIVITY (continued) l The limitations described by Figures 3.9-1, 3.9-2, 3.9-3 and 3.9-4 cnsure that the reactivity of the fuel assemblies stored-in the spent fuel pool are conservatively within the assumptions of the safety analysis.

Administrative controls have been developed and instituted to verify that the fuel enrichment, fuel burnup, fuel decay times, and fuel interface restrictions specified in Figures 3.9-1, 3.9-2, 3.9-3 and 3.9-4 are complied with.

)

3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN l

. The limitations of this specification ensure that the reactivity conditions of the Region 1.3-0UT-0F-4 storage racks and spent fuel pool keff will I l

remain less than or equal to 0.95.

The Cell Blocking Devices in the 4th location of the Region 13-0VT-0F-4 I storage racks are designed to prevent inadvertent placement and/or storage of fuel assemblies in the blocked locations. The blocked location remains empty to provide the flux trap to maintain reactivity control for fuel assemblies in adjacent and diagonal locations of the STORAGE PATTERN.

STORAGE PATTERN for the Region 1 storage racks will be established and l l_

expanded from the walls of the spent fuel pool per Figure 3.9-2 to ensure l

definition and control of the Region 13-0VT-0F-4 Boundary to other Storage Regions l l

and minimize the number of boundaries where a fuel misplacement incident can occur.

NILLSTONE - UNIT 3 8 3/4 9-9 Amendment No. #, JpJ, 197 J M,

002,

DESIGN FEATURES 5.6 FUEL STORAGE l

CRITICALITY 5.6.1.1 The spent fuel storage racks are made up of 3 Regions which are designed and shall be maintained to ensure a K.,, less than or equal l

to 0.95 when flooded with unborated water.

The storage rack Regions are:

a.

Region 1, a nominal 10.0 inch (North / South) and a nominal 10.455 inch (East / West) center to center distance, credits a fixed neutron absorber (BORAL) within the rack, and can store fuel in 2 storage configurations:

(1)

With credit for fuel burnup as shown in Figure 3.9-1, fuel may be stored in a "4-0VT-0F-4" storage configuration.

1 (2)

With credit for every 4th location blocked and - empty of fuel, fuel up to 5 weight percent nominal enrichment, regardless of fuel burnup, may be stored in a "3-00T-0F-4" storage configuration.

Fuel storage in this configuration is subject to the interface restrictions specified in Figure 3.9-2.

b.

Region 2, a nominal 9.017 inch center to center distance, Fredits a fixed neutron absorber (BORAL) within the rack, and with credit for fuel burnup as shown in Figure 3.9-3, fuel may be stored in all available Region 2 storage locations.

c.

Region 3, a nominal 10.35 inch center to center distance, with credit for fuel burnup and fuel decay time as shown in Figure 3.9-4, fuel may be stored in all available Region 3 storage locations.

The Boraflex contained inside these storage racks is not credited.

DRAINAGE 5.6.2 The spent-fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 45 feet.

NILLSTONE - UNIT 3 5-6 Amendment No. 49, pp, 0626

p DESIGN FEATURES CAPACITY 5.6.3 The spent fuel storage poc1 contains 350 Region I storage locations, 673 Region 2 storage locations and 756 Region 3 storage locations, for a total of 1779 total available fuel storage locations.

An additional

' Region 2 rack with 81 storage locations may be placed in the spent fuel pool, if. needed.

With this. additional rack installed, the Region 2 storage capacity is 754 storage. locations, for a total of 1860 total available fuel storage locations.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT l

5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

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l MILLSTONE - UNIT 3 5-6a Amendment No. 49, pp, 0626 l.

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l Docket No. 50-423 l

B17343 l

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.1 Millstone Nuclear Power Station, Unit No. 3 l

Proposed Revision to Technical Specification Spent Fuel Pool Rerack (TSCR 3-22-98)

Background and Safety Summary l

l Ma;.n 1999 l

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U.S. Nucirr Regulatory Commission E17028\\ Attachment 3\\Page 1 Backoround Millstone Unit No. 3 received its low power operating licensing in November,1985.

The plant began operations with spent fuel pool racks in their precent configuration, which is 21 free standing spent fuel racks with a total storage capacity of 756 fuel assemblies. These racks use the silicone polymer Boraflex as the neutron absorption material.

At present, NNECO is contracted to the U. S. Department of Energy (DOE) to take Millstone Unit No. 3 spent fuel. However, the DOE has not yet begun taking spent fuel from reactor sites. When the DOE begins accepting spent fuel, they plan to accept the oldest spent fuel first. Because Millstone Unit No. 3 was licenned relatively recently, it will be among the last reactor sites to Degin its spent fuel shipments to the DOE.

Because Millstone Unit No. 3 will lose full core reserve capsbility in about two years, the plant must increase onsite fuel storage capacity.

NNECO has evaluated spent fuel storage attematives that have been licensed by the NRC and could be feasible for use at Millstone Unit No. 3. The result of the evaluation is that reracking the Millstone Unit No. 3 spent fuel pool is currently the most cost effective alternative. This increase in spent fuel stcrage capacity would preserve full core reserve discharge capability approaching the end of its current operating license in the year 2025.

Summarv Millstone Unit No. 3 must rerack its spent fuel pool to maintain full core reserve capability. NNECO proposes to achieve this goal by installing two types of additional higher density spent fuel racks into the spent fuel pool. Existing spent fuc.I racks will remain in the pool, but are reanalyzed to only accept fuel lower in reactivity than they are licensed to accept at.present. The proposed additional racks will have a closer assembly to assernbly spacing to help maximize fuel storage capacity.

The planned spent fuel pool storage expansion involves licensing 15 new rack modules for insertion into the Millstone Unit No. 3 spent fuel pool. The expansion will leave in place all of the existing 21 spent fuel racks that are in the Millstone Unit No. 3 spent fuel pool. After the expansion, the pool will contain three distinct administratively controlled storage regions as shown in attached Figure 1. Each region is characterized by a nominal center-to-center spacing of the cells. The new cells will contain a fixed neutron absorber for primary reactivity control. The new racks will be grouped in Regions 1 and 2. The existing racks that will remain in place will be designated as Region 3.

Region 1 and Region 2 racks will contain Boral as the neutron absorbing material. The Boral absorbers are to be sized to fully shadow the assembly total active fuel length.

U.S. Nucirer Regulatory Commission B17343/ Attachment 3/Page 2 l

The existing Region 3 racks contain the silicon rubber polymer, Boraflex, as the neutron absorbing material. But no credit is taken for Boraflex in the criticality analysis for Region 3.

. Region 1 racks have the capacity to store up to 350 fuel assemblies. Region 1 can store assemblies with a nominal 5.0 w/o U-235 enrichment in a 3-out-of-4 configuration without restriction on bumup. The 3-out-of 4 configuratMa utilizes a fuel cell blocker for critically control.

Region 1 can also store assemblies in an 4-out-of-4 storage configuration with bumup/ enrichment restrictions. Region 1 is sized to accommodate an emergency core offload.

Region 2 racks will be licensed to store 754 assemblies. The storage in Region 2 racks will have more restrictive bumup/ enrichment restrictions than Region 1 racks and use a 4-out-of-4 storage configuration.

Region 3 racks can store 756 assemblies. The storage in Region 3 racks will have

.more redrictive bumup/ enrichment restrictions than Region 2 racks. Region 3 racks will allow credit for decay of fissile plutonium and buildup of americium, which reduce reactivity, as a function of decay time.

Other domestic nuclear plants have been licensed for decay time credit.

The proposed Millstone Unit No. 3 rerack project will increase the licensed storage capacity from 756 to 1,860 fuel assemblies, which will provide sufficient licensed capacity to allow operation approaching the end of the current plant operating license in the year 2025. As shown in Figure 1, Millstone Unit No. 3 does not plan to install the southern most Region 2 rack at this time; it will be installed if and when necessary.

The structural analyses, seismic analyses, rerack analyses and the Significant Hazards Consideration assume that this rack is installed, which bounds the pool configuration of the rack not bdng installed.

All rack modules in the Millstone Unit No. 3 pool will be free-standing and self-supporting. This includes the existing racks that will comprise Region 3 after the transition phase. After installation, rack locations will be surveyed to ensure proper positioning. Attachments 5 an" 6 detail the proposed rack configuration in the reracked pool.

With the expanded capacity, the spent fuel pool cooling system will be required to remove an increased heat load while maintaining the pool water temperature within the design limit. The maximum heat load typically develops from the residual heat in the pool after the last core offload at the end of plant life. NNECO has reanalyzed spent fuel pool thermal performance. The fuel pool thermal performance analysis, as it applies to bulk pool temperature and equipment under higher heat loads, is under a separate NNECO letter dated January.18,1999 (B17004). However, this proposed amendment request dotes analyze local temperature peaks.

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U.S. Nuclecr Rsgulatory Commission B17343/ Attachment 3/Page 3 Seismic and structural analyses were performed for the racks and pool structure. The racks and pool structure will maintain their function and ensure the integrity, subcriticality margin, and coolability of fuel assemblies under postulated raismic events and mechanical accidenis.

The following addresses the safety issues arising from the raracking and proposed revisions to the Technical Specifications of Millstone Unit No. 3.

The scope of the technical analysis supporting this evaluation focused mainly on the final licensed configuration of the expanded spent fuel pool storage space, including all Region 2 racks.

Mechanical Desian Evaluation The new fuel rack design has been evaluated with respect to the mechanical and material qualifications, neutron poison and poison surveillance requirements, fuel handling qualifications, fuel interfaces, and accident considerations.

The proposed additional spent fuel racks are free standing and self supporting. The principal construction materials are ASME SA240-304L for stainless steel sheet and plate stock, and internally threaded support legs. The externally threaded support spindle is SA564-630 precipitation hardened stainless steel (heat treated to 1,100 F).

The only non-stainless steel material in the racks is the Boral which is a composite of boron carbide and type 1100 alloy aluminum, within a layer of type 1100 aluminum.

The governing quality assurance requirements for fabrication of the racks meet the quality assurance and quality control of 10CFR50, Appendix B requirements.

For primary nuclear criticality control in the new racks, the racks will integrate a fixed neutron absorber into its structure. The absorber, trade name Boral, is a boron carbide and aluminum-composite sandwich. It is chemically inert and has a long history of applications in the spent fuel pool environments where it has maintained its neutron attenuation capability under thermal loads. Boral is manufactured under the control of a quality assurance program which conforms to the requirements of 10CFR50, Appendix B. Region 3 racks contain Boraflex as the fixed neutron absorber. However, Boraflex will no longer be credited per this request.

The support legs on tha racks will allow for remote leveling and alignment of the rack modules to accommodate variations in the floor flatness. A thick bearing pad will be interposed between the rack pedestals and the floor to distribute the dead load over a wider support area.

The rack structural performance with respect to the impact and seismic loads, as well j

as the suberitical configuration, has been analyzed.

The analysis included an accidental drop of a fuel assembly during movement to a storage location, and induced tensile loads on the rack arising from a stuck assembly in the storage cell. It has been

U.S. Nuclur Rzgulatory Commission B17343/ Attachment 3/Page 4 shown that these accidents will not invalidate the mechanical design and material selection criteria to safely store spent fuel in a coolable and suberitical configuration in any region. The fuel will maintain its structural integrity and remain subcritical.

Testing procedures will be developed to periodically verify acceptable performance of the Boral. The testing will use Boral coupons to verify the quality and presence of a sufficient amount of neutron absorber in the racks to assure subcriticality margin. The testing will not extend to the Boraflex absorber in Region 3 since the Boraflex is not credited in the criticality analysis.

Criticality Consideratioris j

The proposed additional spent fuel racks are designed to maintain the required subcriticality margin when fully loaded with fuel of the maximum permissible reactivity for a given storage region, and in unborated water at a temperature within the normal operating range corresponding to the highest reactivity. For reactivity control in Region 1 and 2 racks, Boral panels will be used. The panels are sized to fully shadow the active fuel height of all assembly designs stored in the pool. The panels will be held in place and protected against damage by a stainless steel jacket that is welded to the cell walls. In Region 1, the panels will be mounted on the outside faces of each cell. In Region 2, the panels will be mounted either on the exterior or on the interior of the cells, in an alternating pattem. The existing racks, in what will become Region 3, contain Boraflex as the neutron absorber. However, no credit is taken for Boraflex in the criticality analysis.

The storage of spent fuel in each region will be controlled by the criteria defining the maximum permissible reactivity. Region 1 can store fuel assemblies of up to 5.0 w/o nominal enrichment. regardless of burnup, in a 3-out-of-4 storage array subject to a blocking / interface restriction. Region 1 can store fuel in a 4-out-of-4 array subject to proposed burnup! enrichment limits.

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Region 2 can store fuel in a 4-out-of-4 array subject to proposed burnup/ enrichment j

limits which are more restrictive than those in Region 1.

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Region 3 can store fuel in a 4-out-of-4 array subject to the bumup/ enrichment / decay time limits. Region 3 has the most restrictive bumup/ enrichment limits of the 3 regions.

l Also, Region 3 bumup limits decrease with increased fuel decay time.

l If a fuel assembly does not meet the requirements for storage in either Region 2 or 3, j

then it must be stored in Region 1.

l 1

The USNRC guidelines and the ANSI standards specify that the margin of safety for criticality be determined by the maximum neutron multiplication factor, k.n less than or equal to 0.95, including uncertainties, for all normal and accident conditions. The analysis has shown that this criterion is always maintained under all postulated 1

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U.S. Nuclerr Rrguhory Commission I

B17343/ Attachment 3/Page 5 accidents.

The accidents and malfunctions evaluated included a dropped fuel assembly onto fuel racks, impact on criticality of water temperature and density effects, impact on criticality of eccentric positioning of a fuel assembly within the rack, and misloading of the mort reactive assembly in a Region 1, Region 2, or Region 3 rack (highest reetivity erroi).

The proposed Technical Specifications will require a minimum concentration of 800 ppm of soluble boron in the pool water during fuel movement to assure k,n will remain less than or equal to 0.95 assuming a dropped or misloaded fuel assembly. The surveillance interval for this soluble boron concentration in the proposed Technica' Specifications is consistent with Westinghouse improved STS 3.7.16.

For spent fuel pool water temperature effects, the most reactive spent fuel pool water temperature in the normal operating range was used in the criticality calculations. The criticality analysis uses a range of 32*F to 160*F to bound the fuel pool normal operating water temperature span. For Regions 1 and 2, fuel pool water temperatures in excess of 160*F are less reactive. For Region 3, the most reactive temperature is boiling. However, fuel pool water temperatures in excess of 160*F are outside of the design basis of the fuel pool cooling system. The fuel pool cooling system is capable of meir'taining the fuel pool temperature less than 160*F.

Thermal Hvdraulics and Pool Coolina A comprehensive thermal-hydraulic evaluation of the expanded spent fuel pool has been done to eq;yze its thermal performance to support a separate licensing amendment request dated January 18,1999 (B17004). This comprehensive analysis supports treating full core offloads as a normal evolution. The submittal's assumed heat load bounds the heat load associated with this rerack licensing amendment request and NRC approval of the January 18, 1999, submittal is required prior to approval of this rerack licensing amendrnent.

However, this rerack licensing amendment request calculated the local peak water temperature and local peak clad temperature which is based on the January 18,1999 (B17004), submittal heat load.

The peak local water and fuel clad temperatures were computed for the rerack license amendment for the partially blocked hottest channel. The peak local water temperature was well below the boiling temperature at the top of fuel with fuel pool water level at its low level alarm. This analysis assures that flow will remain subcooled which minimizes the potential for fuel damage. Also, the peak clad temperature is well below the temperature where clad damage or a zirconium-water reaction would occur.

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U.S. Nucl=r Regulatory Commissian 817343/ Attachment 3/Page 6 Seismic and Structural Evaluation i

NNECO has re-evaluated the mechanical and civil structures to address the structural i

issues resulting from the Millstone Unit No. 3 rarack. The analysis considered the loads from seismic, thermal, and mechanical forces to determine the margin of safety in the structural integrity of the fuel racks, the spent fuel pool, and the pool liner. The loads, load combinations, and acceptance criteria were based on ASME Boiler and Pressurs Vessel Code,1995 Edition, Section lil, Subsection NF and NUREG-800, Standard Review Plan (SRP) Section 3.8.4.

a. The storage rack evaluation The final configuration of the pooi will consist of free standing fuel rack modules in all three regions. The seismic analysis has separately evaluated a single free-standing rack as well as the whole pool multi-rack structure in 3-dimensions. The analyses were based on the,imulation of the Safe Shutdown Earthquake (SSE) and the Operating Basis Earthquake (OBE) in accordance with SRP 3.7.1 requirements.

The following computed stress loadings were compared against the allowable stress loadings in' ASME Boiler and Pressure Vessel Code,1995 Edition, Section lil, Subsection NF:

Maximum Fuel Storage Cell Region Stress Factor - The maximum stress factor for i

every rack was computed to be within allowable limits.

Maximum Pedestal Thread Shear Stress - The maximum pedestal thread shear stress was computed to be within allowable limits.

impset Load Between Fuel Assembly and Fuel Storage Wall - The assembly is e

postulated to rattle against the cell wall during an SSE creating a load between the assembly and wall. The maximum load on the cell wall was computed to be well within allowable limits.

Impacts Between Adjacent Racks - The analysis shows that rack movement during the postulated SSE will not lead to impacts between rack cell walls of proposed additional racks, and between proposed additional racks and existing racks. The analysis only predicts rack-to-rack impacts between proposed additional racks at the 3/4 inch baseplate which extends m" of the bottom of the these racks. The I

highest computed impact stress woulu cause very little or no deformation of the baseplates. Rack storage cells, fuel, and Boral would be undamaged.

Baseplate to Fuel Rack Storage Cell Weld Stresses - The maximum stress on a weld between a base plate and a fuel storage cell is computed to be within allowable limits.

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U.S. Nucl=r Regul tory Commission B17343/ Attachment 3/Page 7 Baseplate to Fuel Rack Pedestal Weld Stresses - The maximum stress on a weld between a base plate and a pedestal is computed to be within allowable limits.

Fuel Rack Storage Cell to Fuel Rack Storage Cell Weld Stresses - The maximum stress on a weld between fuel storage cells is computed to be within allowable limits.

Rack Fatigue - The cumulative damage factor due to rack stress fatigue is computed to be with5 allowable limits.

The analyses results show that eah of the above factors are within their allowable limits. Thus, it is concluded that the racks will maintain their integrity, protect the fuel

-and Boral from damage, and maintain subcriticality margin and coolability under all postulated design conditions.

b. Pool structural evaluation Thepool structure has been analyzed using a 3-D finite element model seismically accelended with a synthetic time history motion applied just below the base mat level.

The analyses used the individual dead, live, thermal, and seismic loads and load combinations required by NUREG-800, SRP Section 3.8.4. The analyses show that the pool structure satisfies these required load combinations, and will maintain its integrity and protect racks and fuel for all postulated scenarios.

The following loadings were compared against allowable loadings:

Pool Walls - The analysis computed the limiting safe'y margin for the fuel pool for both bending strength and shear strength on the four fuel pool walls, the transfer canal wall, and the cask pit north and west walls. The smallest limiting safety margin for both bending strength and shear strength occurred on the cask pit west wall, and were well within allowable limits. All other computed safety margins were greater.

Thus, it is concluded that the structural capacity of the fuel pool is maintained under all required load combinations.

. Base Slab - This massive structural slab supporting the pool structure, is heavily reinforced, continuous throughout the Fuel Building area of concem, and supports the whole building. The load additions to the base slab due to the rerack are primarily compressive loads that are supported on bedrock grade. These load increases are very small in comparison to the base slab capacity. Therefore, a simplifying assumption is that the base mat remains adequate in total.

Local stresses on the basemat from fuel rack bear ng pads due to mechanical accidents and seismic loadings are discussed subsequently.

Pool Liner - The pool liner will maintain its integrity during a postulated seismic event. During the postulated seismic occurrence, the fuel rack pedestals will impart

U.S. Nucisar Rrgulatory Commission B17343/ Attachment 3/Page 8 loads onto the pool floor. The analysis found that these loads will not tear or cause fatigue failure of the fuel pool's stainless steel liner and welds.

Bearing Pads - Bearing pad pressun m the fuel pool slab meets the required limits after a postulated seismic occurrenc,e and for all loading conditions. Bearing pads are placed between the pedestal base and fuel pool liner to protect the liner from high localized dynamic loadings, and to distribute the load imparted to the slab.

During a seismic event, fuel rack pedestals impact the bearing pads transferring pedestal loads to the liner. Bearing pad dimensions are set to assure that the average pressure on the pool slab surface due to static and dynamic loads does not exceed allowable limits on bearing pressures. Two stress factors were computed, the average pressure at the slab / liner interface, and the maximum bending stress at the bearing pad. Both of these stress factors were found within allowable limits.

Therefore, the bearing pad design is adequate for all design basis loadings.

c. Mechanical Accidents in addition to the seismic loads, the racks and the pool liner were also analyzed for mechanical loads under accident conditions. The following accident scenarios were analyzed:

Fuel Assembly with Control Rod a id Handling Tool Drop Onto Racks - The analysis shows that if a fuel assembly with control rod and handling tool drop from above the maximum lift height of the spent fuel bridge hoist onto a rack, only the upper region of the impacted storage cell is damaged, thus protecting the Boral and stored faal assembly from damage. Also, local thermal hydraulic requirements continue to be met since only minor distortion of the fuel cell geometry will occur.

Fuel Assembly with Control Rod and Handling Tool Drop Through an Empty Rack Storage Cell Over a Pedestal Location - This scenario assumed the spent fuel bridge hoist drops a fuel assernbly with control rod and handling tool from above the maximum lift height into an empty fuel storage cell over a pedestal location. This scenario maximizes the load imparted to the pool liner. The analysis concluded that this scenario would cause negligible rack baseplate deformation and insignificant plastic strain in the liner. Thus, the liner would maintain its integrity.

Fuel Assembly with Control Rod and Hadling Tool Drop Through an Empty Interior Rack Storage Cell - This scenario assumeo lhe spent fuel bridge hoist drops a fuel assembly with control rod and handling tool from above the maximum lift height into an empty interior fuel storage cell. The fuel assembly falls unimpeded through the storage cell until it strikes the rack baseplate at the bottom of the storage cell. This impact is postulated to occur at an interior storage cell location to maximize the predicted baseplate deformation, and produces localized severing of the baseplate / storage cell welds. However, the baseplate still maintains its integrity

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o U.S. Nucl:ar Regulatory Commission B17343/ Attachment 3/Page 9 and prevents the fuel assembly from impacting the liner. Thus, no liner damage occurs.

Spent Fuel Rack is Dropped onto Fuel Pool Liner During Installation - The analysis concludes that if a rack drops 40 feet onto the liner during installation, liner puncture and small indentations in the pool floor concrete surface would occur. A small rate of water seepage, which is well within makeup capability, could occur.

Such seepage is considered minor and procedures exist that direct the operators to initiate emergency make-up to the pool, if necessary.

There will also be a contingency procedure to repair liner damage, should it occur, during the rack installation.

Fuel Assembly Becomes Stuck When Being Removed from Fuel Storage Rack -

The analysis shows that the rack structural integrity will not be compromised if a fuel assembly becomes stuck during removal from a rack.

Fuel Pool Gate Drops onto a Fuel Storage Rack - The transfer canal fuel pool gate will now be moved over fuel racks because Region 1 racks will be installed within several inches of the fuel pool west wall. The cask pit storage gate when being moved does go over existing fuel racks. Both the proposed additional racks and the existing racks were analyzed for a fuel pool gate drop. This analysis demonstrates that a gate drop would not damage a stored fuel assembly (provided the fuel assembly does not contain a control rod assembly or other insert) or cause damage j

to the neutron absorber material or to the pool liner. In addition, although the upper portion of the impacted rack suffers local deformation, the overall structural integrity of the rack is not compromised; thus the storage array configuration is maintained, and there are no resulting criticality concerns. Nevertheless, the requirements of Technical Specification 3.9.7 will continue to prohibit fuel pool gate movement over fuel assemblies since a gate weighs more than the imposed 2,200 lb. load limit.

Cask Drop - The consequences of dropping a fully loaded fuel shipping cask into e

the cask pit or on the fuel pool floor are not discussed in this Licensing Amendment Request since Millstone Unit No. 3 is not currently licensed to transport a cask into the spent fuel building. Therefore, this event has not been included in the list of analyzed accidents associated with this licensing amendment request.

10CFR55a(a)(3)(i) Reauest in accordance with 10CFR50.55a(a)(3)(i), NNECO is infonning the NRC of the use of the 1995 Edition of ASME Section ill Subsection NF for the design, materials, fabrication, and examination of the proposed new spent fuel storage racks, to be installed 'in the Millstone Point Unit 3 spent fuel pool, as an alternative to the requirements of 10CFR50.55a(b)(l).

i U.S. Nucle r Regulatory Commission B17343/ Attachment 3/Page 10 1

Both the original spent fuel storage racks and the proposed new racks meet the requirements of USNRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978, and as amended January 18,1979.

NNECO's code reconciliation evaluation confirmed that the technical requirements for the design, materials, fabrication, and examination of the proposed new spent fuel storage racks meet and exceed the original Owner requirements and the applicable Code of Construction requirements. Therefore, it is concluded that the proposed alternative will provide an acceptable level of quality and safety for the proposed new spent fuel storage racks. The design bases for the original spent fuel storage racks j

remain unchanged.

Proposed Technical Specification Chanoes j

Technical Specification Definitions 1.40 and 1.41 are reworded to provide the definitions for the new spent fuel rack configurations.

i Technical Specification 3.9.1.2 and its associated Bases Section were revised in Amendment 158 to require a boron concentration of 1,750 ppm. This change was requested by NNECO in a letter dated November 11,1997, which identified that a seismic event of a magnitude equal to or greater than an OBE could degrade the Boraflex in the spent fuel racks. To address this situation the required boron concentratio,1 in the spent fuel pool was increased from 800 ppm to 1,750 ppm. As discussed above, the Boraflex in the existing spent fuel pool racks will not be credited for critically control when the existing racks are designated Region 3 racks.

This design requirement was committed to by NNECO in the November 11,1997, letter. The boron concentration in the spent fuel pool will only be required during fuel movements for a dropped or misplaced assembly event. Therefore, the spent fuel pool boron concentration is being re' vised from requiring 1,750 to 300 ppm.

Technical Specification Surveillance 4.9.7 is being revised to clarify that the crane interlocks and stops prevent a crane from carrying a load in excess of 2,200 lbs over the spent fuel pool versus being carried over fuel assemblies as stated in the existing surveillance. This clarification more accurately describes the present crane interlocks and stops at Millstone Unit No. 3. This proposed change continues to l

prohibit loads in excess of 2,200 lbs from being carried over fuel in the fuel pool.

l Additiorally, Technical Specification Surveillance 4.9.7 is being expanded to allow fuel pool gates and spent fuel racks to be moved by crane under administrative controls, in lieu of crane interlocks and physical stops. The administrative controls will prevent the crane from carrying the load above fuel assemblies. NNECO in a response to NUREG-0612 dated March 14,1985, stated that when placing spent fuel racks into the spent fuel pool (which weigh more than 2,200 lbs), Millstone Unit No. 3 will utilize the new fuel handling crane and bypass its interlocks so that the I

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U.S. Nuclur Regulatory Commission B17343/ Attachment 3/Page 11 crane can move over the fuel pool. Additionally, in the March 14,1985, submittal, NNECO also stated that when moving fuel pool gates (which also weigh more than 2,200 lbs), Millstone Unit No. 3 will utilize the spent fuel bridge crane and bypass its interlocks so that the crane can move over the fuel pool. However, these kinds of evolutions will require written procedures and Shift Supervisor approval. The NRC in NUREG-1031, Supplement 2 dated September 1985, referenced the March 14,1985, submittal and stated that the overhead heavy load handling system meets the acceptance criteria of SRP Section 9.1.5.

When using administrative controls, improper operator action could lead to a crane carrying a lead greater than 2,200 lbs over fuel. However, when utilizing interlocks or physical stops to prevent movement of a load greater than 2,200 lbs over fuel, improper setting of the interlocks under administrative controls, or physical failure of an interlock, could also lead to a crane canying a load greater than 2,200 lbs over fuel. Thus, when bypassing interiocks so that a crane can carry a load greater than 2,200 lbs over the spent fuel pool, required administrative controls shall be adequate such that the probability of carrying the load over fuel is not greater than the probability of carrying the load over fuel when depending on interlocks or stops.

To drop the load onto fuel requires a double malfunction, operator error or interlock failure to bring the load over fuel, and then a crane malfunction to drop the load. If Technical Specification 3.9.7 is violated, the Technical Specification requirec that the load be placed into a safe condiQn.

Oso, proposed Tecnnica' W.9cification Surveillance 4.9.7 clarifies that loads tt #

aigh less than 2,200 lbs can be moved by crane under administrative controls, ir.

neu of crane interlocks and physical stops. This change cannot lead to violation of Technical Specification 3.9.7 because this Technical Specification only places restrictions on loads in excess of 2,200 lbs, and dor not place any requirements on loads less than 2,200 lbs.

Thus, the proposed change continues to meet the requirements of Technical Specification 3.9.7, that is it prohibits a crane from carrying a load greater than 2,200 lbs over fuel in the spent fuel pool.

Technical Specification 3.9.13 and its associated Bases Section were revised in Amendment 158 to require actions for an Operating Basis Earthquake. This change was requested by NNECO in a letter dated November 11, 1997, which identified that a seismic event of a magnitude equal to or greater than an OBE could degrade the Boraflex in the spent fuel racks. To address this situation the actions and surveillances were included in Technical Specification 3.9.13. As discussed above, the Boraflex in the existing spent fuel pool racks will not be credited for critically control when the existing racks are designated Region 3 racks.

This design requirement was committed to by NNECO in the November 11,1997, letter. The changes to the Technical Specification include; Action b will require that immediate action be initiated to move any misplaced fuel assembly into a location for which the

U.S. Nuciser Regulatory Commission B17343/ Attachment 3/Page 12 assembly is qualified, renumber Section 4.9.13.1 to 4.9.13.1.1 which requires appropriate documentation be reviewed to assure that fue! assemblies stored in a 4-out-of-4 storage pattem in Region 1 fuel racks meet the bumup/en*ichment requirements of Figure 3.9-1 (replaces old figure), add Section 4.9.13.1.2 which

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requires appropriate documentation be reviewed to assure that fuel assemblies stored in Region 2 fuel racks meet the bumup/ enrichment requirements of Figure 3.9-3 (new figure) and add Section 4.9.13.1.3, which requires appropriate documentation be reviewed to assure that fuel assemblies stored in Region 3 fuel racks meet the bumup/ enrichment / decay time requirements of Figure 3.9-4 (new figure).

Technical Specification 3.9.14 is revised to replace the roman numeral I with the number i for Region 1 designation. Note, for simplicity and clarity the fuel storage region designation is being changed from roman numerals to standard numbers. This change is editorial in nature, and does not impact the rerack project design or safety.

Technical Specification Figures 3.9-1 and 3.9-2 are replaced with new figures 3.9-1, 3.9-2, 3.9-3 and 3.9-4 indicating storage requirements for the proposed Regions 1, j

2 and 3 fuel racks.

Technical Specification Bases Section 3/4.9.1.1: BASES is revised to correct the section designator from 3/4.9.1 to 3/4.9.1.1.

Technical Specification Bases Section 3/4.9.14: BASES is revised to recognize that i

Region 1 can now be either in a 3-OUT-OF-4, or 4-OUT-OF-4 storage configuration.

Techn...

,'ecification Section 5.6.1.1: DESIGN FEATURES - CRITICALITY, is revised to describe the pitch, neutron absorber, storage pattern, and burnup".1richent/ decay time limits for each region of proposed fuel racks.

Technical Specification Section 5 6.3: DESIGN FEATURES - CAPACITY, is revised to list the storage capacity of each proposed region of fuel racks.

Revise INDEX pages xii and xv for new figures and page numbers.

Radioloaical Consecuences Radiological consequences of accidents in the spent fuel pool building have been evaluated. The existing design basis fuel drop accident in the fuel building described in FSAR Chapter 15.7.4 (fuel assembly drop onto another fuel assembly) is not affected by the rerack. Thus, potential radiological consequences from a fuel drop accident are not affected by the rerack.

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l U.S. Nucl=r Regulatory Commission

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B17343/ Attachment 3/Page 13 A rack drop accident with radiological consequences is unlikely since all rack movement during installation will follow safe load paths that prevent heavy loads from being transported over the stored spent fuel.

Thus, there are no radiological consequences from this accident.

Special Circumstance Reaardina Transitionina to Revised Technical Specifications A special circumstance will exist regarding transitioning to the proposed Technical Specifications after NRC approval of this licensing amendment request, except for Technical Specification 3/4.9.7 which will take immediate effect since it does not directly deal with criticality requirements. The existing Technical Specifications credit Boraflex in the existing spent fuel racks, which reduces fuel burnup requirements. The proposed Technical Specifications eliminate Boraflex credit in the existing fuel storage racks, which causes a significant step increase in the fuel bumup requirements to store fuel in the these racks. At the time of the rerack it is anticipated that about 120 fuel assemblies stored in the Boraflex racks would not meet fuel burnup requirements ci the proposed Technical Specifications. These 120 or so fuel assemblies will need to be transferred from the existing racks (called Region 3 under the proposed Technical Specifications) to the proposed additional storage racks (called Region 1 or 2 under proposed Technical Specifications) to comply with ths new proposed Technical Specifications fuel burnup requirements. This r ieans that Boraflex must be credited and existing surveillance requirements maintain ed until the rerack is complete, and these approximately 120 fuel assemblies can be transferred to Region 1 or Region 2 storage racks. If the proposed Technical Specifications, which do not credit Boraflex, are made fully effective before NNECO can transfer these fuel assemblies out of the existing racks, the plant would not be in compliance with the revised Technical Specifications.

To address this situation, NNECO proposes the following:

When the NRC issues the rerack license amendment, NNECO would rerack the fuel pml. After rack installation and survey are complete, and as the last step of the re-rack, NNECO would transfer the approximately 120 fuel assemblies discussed above to the new Region 1 or Region 2 fuel storage racks. NNECO would then fully implement the revised Technical Specifications from the rerack license amendment.

During the interim period from NRC approval of the proposed Technical Specifications to completion of the rerack, including assembly transfer out of existing racks, NNECO will continue to comply with the existing rack Technical Specifications requirements (except for Technical Specification 3/4.9.7 which will take immediate effect). Thus, all existing Boraflex related Technical Specification requirements would remain in place until all of these approximately 120 fuel assemblies are transferred from the existing racks.

U.S. Nuclerr Reguintory Commission B17343/ Attachment 3/Page 14 Wnan these approximately 120 fuel assemblies are in the process of being I

transferred to new rackr, NNECO will administratively comply with the fuel burnup/ enrichment requirements for the new racks (Regions 1 and 2) while simultaneously complying with the soluble boron requirements and Boraflex related surveillances of the existing Technical Specifications. The existing soluble boron requirements and Boraflex related surveillances are more restrictive than the proposed Technical Specifications. These actions will ensure that k, remains less than or equal to 0.95 for fuel in existing racks during the rerack, and for fuel in all racks during fuel transfer.

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4 Docket No. 50-423 B17343 Millstone Nuclear Power Station, Unit No. 3 Spent Fuel Pool Rerack (TSCR 3-22-98)

Significant Hazards Consideration and Environmental Considerations i

March 1999 l

U S. Nucinr R:gul tory Commission B17343/ Attachment 4/Page 1 Sionificant Hazards Consideration in accordance with 10CFR50.92, NNECO has reviewed the p oposed changes and has concluded that they do not involve a Significant Hazards Consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazard because they would not; 2.1 Involve a significant increase in the probability or conseque.

of an accident previously evaluated.

k in the analysis of the safety issues conceming the expanded pwi

=,1 age capacity, NNECO has considered the following potential accident scenarios:

i a.

A spent fuel assembly drop with control rod and handling tool b.

A fuel pool gate drop J

c.

Potential damage due to a seismic event d.

Fuel assembly mistoading/ drop or pool temperature exceeding 160*F e.

An accidental drop of a rack module during installation activity in the pool The probability that any of the first four accidents in the above list can occur is not significantly increased by the modification itself. All work in the pool area will be controlled and performed in strict accordance with the specific written procedures. As for an installation accident, safe load paths will be established that will prevent heavy loads from being transported over the spent fuel. Proper functioning of the cranes will be checked and verified before rack installation, and appropriate administrative controls imposed. All lift rigging and the crane / hoist system will be verified to comply with applicable plant and site procedures.

All heavy lifts will be performed in accordance with established station procedures, which will comply with NUREG-0612,

" Control of Heavy 1.csos at Nuclear Power Plants." These actions will minimize the possibility of a heavy ?oad drop accident. Fuel assembly handling procedures and techniques are not affectect by adding spent fuel racks, and the probability of a fuel handling accident or misloading is not increased.

Accordingly, the proposed modification does not involve a significant increase in the probability of an accident previously evaluated.

NNECO has evaluated the consequences of an accidental drop of a fuel assembly in the spent fuel pool. T')e results show that such an accident will not distort the racks sufficiently to impair t'ieir functionality. The minimum subcriticality margin, k.n less than or equal to 0.95, will 9e maintained. The radiological consequences of a fuel assembly

U.S. Nuclecr Regulatory Commission B17343/ Attachment 4/Page 2 drop are not increased from the existing postulated fuel drop accident in Millstone Unit No. 3 FSAR Section 15.7.4. Thus, the consequences of such an accident remain acceptable, and are not different from any previously evaluated accidents that the NRC has reviewed and accepted.

The consequences of an eccidental drop of a fuel pool gate onto racks has been evaluated. The results show that such an accident will not distort the racks sufficiently to impair their functionality. The' minimum subcriticality margin, k, less than or equal to 0.95, will be maintained in addition,'the Technical Specifications do not allow fuel to be under a fuel pool gate when one is moved. The analysis indicates no radiological consequences from this postulated accident. Thus, the consequences of such an accident remain acceptable, and are not different from any previously evaluated accidents that the NRC has reviewed and accepted.

The consequences of a design basis seismic event have been evaluated and found acceptable. The proposed additional racc and existing racks have been analyzed in their new configuration and found safe and impact-free during seismic motion, save for the baseplate-to-baseplate impacts of the proposed additional racks which are shown to cause no damage to the racks cells or Boral. The structural capability of the pool walls and basemat will not be exceeded under the loads. Thus, the consequences of a seismic event are not significantly increased.

The criticality consequences of a misloading/ drop of a fuel assembly during fuel movement have been evaluated. The minimum subcriticality margin, k.,less than or equal to 0.95, will continue to be maintained because of the propcsed pool water soluble boron related requirements. Thus, the consequences of such an accident remain acceptable, and are not different from any previously evaluated accidents that the NRC has reviewed and accepted.

The consequences of an accidental drop of a rack module into the pool during placement have been evaluated. The analysis confirmed that very limited damage to the liner could occur, which is repairable. Any small seepage occurring is well within makeup capability, and is mitigated by emergency operating procedures.

All movements of racks over the pool will comply with the applicable guidelinee. Therefore, the consequences of an installation accident are not increased from any previously evaluated accident.

The consequences of a spent fuel cask drop into the pool have not been considered in this submittal since NNECO is not currently licensed to move a fuel cask into the Millstone Unit No. 3 cask pit area.

Therefore, it is concluded that the proposed changes to the Technical Specifications and licensing basis of Millstone Unit No. 3 do not significantly increase the probability or consequences of any accident previously evaluated.

U.S. Nuclstr Regulatory Commission B17343/ Attachment 4/Page 3 j

2.2 Create the possibility of a'new or different kind of accident from any previously analyzed.

The proposed change does not alter the operating requirements of the plant or of the equipment credited in the mitigation of the design basis accidents. Therefore, the potential for an unanalyzed accident is not created. The postulated failure modes

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-associated with the change do not significantly decrease the coolability, criticality margin, or structural integrity of the spent fuel in the pool. The resulting structural, thermal, and seismic loads are acceptable.

]

Therefore, the change does not create the possibility of a new or different kind of accident from any previously analyzed.

2.3 Involve a significant reduction in the margin of safety.

The function of the spent fuel pool is to store the fuel assemblies in a subcritical and coolable configuration through all environmental and abnormal loadings, such as an earthquake, fuel assembly drop, fuel pool gate drop, or drop of another heavy object.

The new rack design must meet all applicable requirements for safe storage and be functionally compatible with the other rack design in the spent fuel pool.

NNECO has addressed the safety issues related to the expanded pool storage capacity in the following areas:

1.

Material, mechanical, and structural considerations 2.

Nuclear criticality 3.

Thermal-hydraulic and pool cooling The mechanical, material, and structural designs of the new racks have been reviewed in accordance with the applicable provisions of NRC "OT Position for the Review and Acceptance of Spent Fuel Storage and Handling Applications", April 14,1978, as amended January 18,1979. The rack materials used are compatible with the spent fuel assemblies and the spent fuel pool environment. The design of the new racks preserves the proper margin of safety during a'.; normal loads such as a dropped fuel assembly, a postulated seismic event, a dropped fuel pool gate, and tensile loads from a stuck fuel assembly. It has been shown that such loads will not invalidate the mechanical design and material selection to safely store fuel in a coolable and subcritical configuration. Also, it has been shown that the pool structure will maintain its integrity and function during normal operation, all postulated accident sequences, 2

and postulated seismic events.

_ U.S. Nucler Reguktory Commission-B17343/ Attachment 4/Page 4 -

LThe methodology used in the criticality analysis of the expanded spent fuel pool storage capacity meets the appropriate NRC guidelines and the ANSI standards. The margin of safety for subcriticality is determined by a neutron multiplication factor less than or equal to.0.95 under all accident conditions, including uncertainties. This criterion has been preserved in all analyzed accidents and seismic events.

The special circumstance regarding transitioning to the revised technical specifications was discussed. At present, NNECO estimates that there will be approximately 120 fuel assemblies stored in existing racks that will not meet the bumup/ enrichment requirements for storage in these racks under the proposed Technical Specifications.

During the actual raracking effort, including transfer of these assemblies from existing racks to Region 1 and 2 racks, existing soluble boron and Boraflex related requirements and'surveillances will continue to be enforced. Also, when transferring

-these assemblies to Region 1 and 2 racks, the burnup/ enrichment requirements of these racks will be enforced. After fuel transfer is complete, the revised Technical Specifications will be fully implemented. These requirements ensure that the neutron multiplication factor will remain less than or eque' +o 0.95 during the whole period of the rerack.

The rarack. thermal hydraulic analysis is based on NNECO's January 18, 1999, submittal analysis which bounds the heat load of this licensing amendment request.

The rerack thermal hydraulic analysis found that, in the blocked hottest stored assembly, the local peak water temperature will remain below boiling, and the fuel clad will not experience high temperatures.

Regarding. Technical Specification Surveillance 4.9.7, since the proposed change continues to meet the requirements of Technical Specification 3.9.7, that is it prohibits a crane from carrying a load greater than 2,200 lbs over fuel in the spent fuel pool to preclude fuel damage, the margin of safety is maintained.

Thus, it is concluded that the proposed changes to the Technical Specifications and licensing basis of Millstone Unit No. 3 do not involve a significant reduction in the margin of safety at Millstone Unit No. 3.

1 1

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U.S. Nucl;rr Reguirtory Commission B17343/ Attachment 4/Page 5 Environmental Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed revision does not involve 1

a significant hazard, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusion from the requirements for environmental review.

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U.S. Nuclear Regulatory Commission B17343/ Attachment 4/Page 6

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