ML20044E019

Proposed TS 1.39 Re Storage pattern,4.9.16.1 Re Decay Time of All Fuel & TS Figure 3.9, Min Required Fuel Assembly Exposure as Function of Initial Enrichment to Permit Storage in Region C

ML20044E019
Person / Time
Site:	Millstone
Issue date:	05/14/1993
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20044E017	List: B14470, Application for Amend to License DPR-65,changing Listed TS to Introduce Neutron Absorbing (Poison) Rodlets Into Stored Fuel & Increase Required Burnup in Region C to Permit Removal of Cell Blockers & Reclaim 234 Storage Locations ML20044E019
References
14470, NUDOCS 9305210192
Download: ML20044E019 (42)

Text

4 5

i Docket No. 50-336 B14470 l

3 f

Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Spent fuel Pool Modifications Marked Up Pages of Technical Specifications t

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P May 1993 9305210192 930514 PDR ADDCK 05000336 P

PDR

q n J Juns 4, 1992 DEFINITIONS I

VENTING

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1.35 VENTING is the controlled process of discharging air or gas from a confinement o maintain temperature, pressure, humidity, concentration or l

other operat.ng 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.

MEMBER (S) 0F THE PUBLIC 1.36 MEMBER (S) 0F THE PUBLIC shall include all persons who are not 1

occupationally associated with-the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make This category does include persons who use portions of the site deliveries.

for recreational, occupational or other purposes not associated with 'the plant.

The term 'REAL MEMBER OF THE PUBLIC* means an individ " aho is exposed to l

existing dose pathways at one particular location.

SITE BOUNDARY 1.37 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.

,,, q UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area i

within the site boundary used for residential quarters or industrial, commercial institutional and/or recreational purposes.

STORAGE PATTERN 1.39 The Region B and spent fuel racks contain a cell blocking device in every 4th rack loca on for administrative control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked l

location and all adjacent and diagonal cell locations surrounding the blocked location within the respective region.

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MILLSTONE - UNIT 2 1-8 Amendment No. JM JJ7,158 COED i

January 15, 1986 REFUELING OPERATIONS SHIELDED CASK LIMITING CONDITION FOR OPERATION 3.9.16.1 All fuel within a distance L from the center of the spent fuel l

pool cask set-down area shall have decayed for at leastll20 days;) The distance L equals the major dimension of the shielded cask.

'(- 1 Ea r' APPLICABILITY: Whenever a shielded cask is on the refueling floor.

f ACTION:

I With the requirements of the above specification not satisfied, do not move a shielded cask to the refueling floor. The provisions of Specifi-cation 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.16.1 The decay time of all fuel within a distance L from the center of j

th spent fuel pool cask set-down area shall be determined to be > [I20 l

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I da s within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to moving a shielded cask to the refueTing j i oor and at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter, i yca 4

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L MILLSTONE - UNIT 2 3/4 9-19 Amendment No. 30, 109

l June 4,1992 REFUELING O'PEIOLTIONS p

SPENT FUEL POOL--REACTIVITY CONDITION LIMITING CONDTTION FOR OPERATION 3.9.18 The Reactivity Condition of the spent fuel pool shall be such that K,77 is less-than-or-equal-to 0.95 at all times.

APPLICABILITY: Whenever fuel is in the spent fuel pool.

ACTION:

Berate until K,ff 5 95 is reached.

SURVElltANCE REOUIREMENT 4.9.18.1 Ensure that all fuel assemblies to be aced _in Region C (as shown

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in Figure 3.9-2) of the ent fuel p are w h~e enric7imed an rn-up ls T Figure E!C Kthe_ assembly's design and burn-up documen-

=

fa_tiongs,5 c;+J,he contents of each consolidated fuel storag w,:

Ensure that t 4.9.18.2 placed in Region C (as shown in Figure 3.9-2) of the spent fuel pool are i

within the enrichment and burn-up limits of Figure 3.9-3 by checking the design and burn-up documentation for storage box contents.

Ensure that all fuel assemblies to be placed in Region A (as shown 4.9.18.3 in Figure 3.9-2) of the spent fuel pool are within the enrichment and burnup of Figure 3.9-4 by checking the assembly's design and burnup r

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N REOIUN[d-SPENT FUEL PO0t. ARRA*8".EMENT UNIT #2 FIGURE ;

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June 4,1992

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REFUELING OPEPATIONS.

SPE FUEL POOL - STORAGE PATTERN

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tIMITY CONDTTION FOR OPERATION Egch STORAGE PATTERN of the Region C spent fue

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3.9.19.1 require either that:

(l') A ceir@ locking device is installed in those celldocations shown in i

Figure 3.9-2; or

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If a cell blocking device has been removed all cells of the STOPAGE (2)

PATTERN must'have consolidated fuel in them, including the formerly blocked locati'on; or (3) Meet both (a) and b):

If a cell block ~pg device as been removed, all cells of the STORAGE PATTERN must havp/ consolidated fuel in them ek (a) formerly blocked lo'c on.

(b) The formerly blockey location is vacant and a consolidated fuel box or cell blocking dhvice is imediately being placed into the formerly bio ked cell APPLICABILITY:

Fuel in the pent Fuel Pool-ACTION:

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Take imediate action'ko comply with either 3.9.19.1(1), (2) or (3).

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SURVE1ltANCE REOUTREMENTS

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4.9.19.1 Verify that 3.9.19.1 is satisfied at the following times.

I' P7 or to removing a cell blocking device (1)

(2)/ Prior to removing a consolidated fuel storage box. from its Region C

/ storage location.

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MIU. STONE - UNIT 2 3/4 9-26 Amendment No. JJ7, #J,158 cc:a

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June 4,1992 REFUELING OPERATIONS SPENT FUEL POOL - STORAGE PATTERN LTHTTING CONDITION FOR OPERATION 3.9.19 Each STORAGE PATTEPR of the Region B spent fuel pool racks shall require that:

(1) A cell blocking device is installed in those cell locations shown in Figure 3.9-2; or (2)

If a cell blocking device has been removed, all cells in the STORAGE PATTERN must be vacant of stored fuel assemblies.

APPLICABILITY:

Fuel in the spent fuel pool.

ACTION:

Take imediate action to comply with either 3.9.19fl) or (2).

SURVEILLANCE REOUTREMENTS e

4.9.19 rify that 3.9.1912 is satisfied prior to removing a cell blocking device.

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HILLSTONE - UNIT 2 3/4 9-26%

Amendment No. 158 ces2

June 4,1992 REFUELING OPERATIONS 4

BASES 3/4.9.13 STORAGE POOL RADIAT' ION HONITORING The OPERABILITY of the storage pool radiation monitors ensures that sufficient radiation monitoring capability is available to detect excessive radiation levels resulting from 1) the inadvertent lowering of the storage pool water level or 2) the release of activity from an irradiated fuel assembly.

3/4.9.14 & 3/4.9.15 STORAGE POOL AREA VENTILATION SYSTEM The limitations on the storage pool area ventilation system ensures that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

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

3/4.9.16 SHIELDED CASK The limitations of this specification ensure that in an event of a cask tilt accident 1) the doses from ruptured fuel assemblies will be within the will remain s.95.

assumptions of the safety analyses, 2) Keff 3/4.9.17 MOVEMENT OF FUEL IN SPENT FUEL POOL The limitations of this specification ensure that, in the event of a fuel assembly or a consolidated fuel storage box drop accident into a Region B or C will remain rack location completing a 4-out-of-4 fuel assembly geometry, Keff 1 0.95.

3/4.9.18 SPENT FUEL POOL - REACTIVITY CONDITION 33,g The limitations described by Figures 3.9-1[and 3.9-3 ensure that the reactivity of fuel assemblies and consolidated fuel storage boxes, intrdduced into the Region C spent fuel rteks, are conservatively within the assumptions of the safety analysis.

The limitations described by Figure 3.9-4 ensure that' the reactivity of the fuel assemblies, introducted into the Region A spent fuel racks, are conservatively within the assumptions of the safety analysis.

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MILLSTONE - UNIT 2 B 3/4 9-3 Amendment No. 7E, JEf, JJ7, JEJ,158 l

June 4 1992 0

REFUELING OPERATIONS BASES 3/4.9.19 SPENT FUEL' POOL - ST'OR/GE PATTERN The limitations of this specification ensure that the reactivity conditio of the Region B and C storage racks and spent fuel pool K,77 will remain less than or equal to 633.

The Cell Blocking Devices in the 4th location of the Region C storage l

racks are designed to prevent inadvertent placement and/or storage of fuel The blocked location remains empty to g-asserblies in the blocked locations.

provide the flux trap to maintain reactivity control for fuel assembly storage (

in any adjacent locations. Only loaded consolidated fuel storage boxes may be

)

placed and/or stored in the 4th location, completing the STORAGE PATTERN, after all

adjacent, and diagonal, locations are occupied by loaded i

v consolidated fuel storage boxes.

The Cell Blocking Devices if the 4th location of the Region B storage racks are designed to prevent inadvertent placement and/or storage in the blocked The blocked location remains empty to provide the flui trap to locations.

control for fuel asserbly storage in any. adjacent reactivity maintain Region B is designed for the storage of new assemblies in the locations.

spent fuel pool, and for fuel assemblies which have not sustained sufficient burnup to be stored in Region A or Region C.

3/4.9.20 SPENT FUEL POOL - CONSOL10ATION The limitations of these specifications ensure that the decay heat rates and radioactive inventory of the candidate fuel assemblies for consolidation are conservatively within the assumptions of the safety analysis.

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MILLSTONE - UNIT 2 B 3/4 9-4 Amendment No. JJ7, J S 158 CCE3

Juae 4, 1992 DESIGN FEATURES f

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 10,060 + 700/-0 cubic feet.

5.5 EMERGENCY CORE COOLING SYSTEMS The emergency core cooling systems are designed and shall be maintained 5.5.1 in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.6 FUEL STORAGE CRITICAtITY 5.6.1 a)

The new fuel (dry) storage racks are designed and shall be maintained with sufficient center to center distance between assehblies to

<.95.

The maximum nominal fuel enrichment to be stored in ensure a k'$ I.50 weight percent of U-235.

these racks b)

Region A of the spent fuel storage pool is designed and shall be maintained with a nominal 9.8 inch center to center distance between storage s.95 with the storage pool filled with unborated locations to ensure a K'sYored in this region must comply with Figure 3.9-4 to water. Fuel assemblies ensure that the design burnup has been sustained.

c)

Region. B of the spent fuel storage pool is designed and shall be maintained with a nominal 9.8 inch center-to-center distance between storage 1 95 with a storage pool filled with unborated locations to ensure K 77 water.

Fuel assembli8s stored in this region may have a maximum nominal enrichment of 4.5 weight percent U-235. Fuel assemblies stored in this region are placed in a 3 out of 4 STORAGE PATTERN for reactivity control.

d) Region C of the spent fuel storage pool is designed and shall be i

8 maintained with a 9.0 inch center to center distance between storage locations to ensure a K

<.95 with the storage pool filled _ with unborated water.

(Tu~el assembl%f4tTred'irrthIs region must cWmply with Figure 3.9-1 to ensure.

i that the design burn-up has been sustained.

Fuel assemblies stored in this The

\\ region are placed in a 3 out of 4 STORAGE PATTERN for reactivity control.

, contents of consolidated fuel storage boxes to be stored in this region must j

g. ply with Figure 3.9-3f emit storage of consolidated fuel (pent fuel storage pool is designed to_

e)

Region C of the s i

~ he storage rack nd Tri~lhe 4th location ot t

< 0.95./ Placement of consolidated fuel in the 4th locauNi~1s ensure a K (lon FTeWINd"IT Ell surrounding cells of the STORAGE PATTERN are occupied by 1 consolidated fuel.

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MILtSTONE - UNIT 2 5-5 Amendment No. 7E, EE, JEE, JJJ, Jff>,158 ccs2 i

INSERT "A" Fuel assemblies stored in this region must comply with Figures 3.9-la or 3.9-lb to ensure that the design burn-up has been sustained.

Additionally, fuel assemblics utilizing Figure 3.9-lb require that borated stainless steel poison pins are installed in the fuel assembly's center guide tube and in two diagonally opposite guide tubes.

The poison pins are solid 0.87 inch O.D. borated stainless steel, with a boron content of 2 weight percent boron.

i

June 4, 1994 l

DESIGN FEATURES DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 22'6".

CAPACITY l

5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 224 storage locations in Region A, 160 storage locations in Region B and 962 storage locations in Region C for a total of 1346 storage locations.*

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• This translates into 1237' storage locations to receive spent fuel and i

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,109' storage locations to remain blocked.

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. MILLSTONE - UNIT 2 5-Sa Amendment No. /E, EE, JE9, JJ7,Jff,158

l I

Docket No. 50-336

[

B14470 Millstone Nuclear Power Station, Unit No. 2 -

Proposed Revision to Technical Specifications Spent Fuel Pool Modifications

(

I Retyped Pages of Technical Specifications f

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I May 1993 i

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DEFINITIONS VENTING 1.35 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or i

other operating condition, in such a manner that replacement air-or gas is not provided or required dui ng venting. Vent, used in system names, does not i

imply a VENTING process.

MEMBERfS) 0F THE PUBLIC 1.36 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associtted with the plant. This category does not include i

employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the pl ant.

The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location.

SITE B0UNDARY 1.37 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.

UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial institutional and/or recreational purposes.

STORAGE PATTERN 1.39 The Region B spent fuel racks contain a cell blocking device in every 4th rack location for admir.istrative control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked location and all adjacent and diagonal cell locations surrounding the blocked location within the respective region.

MILLSTONE - UNIT 2 1-8 Amendment No. Jpf, JJ7, JEE 0085

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i REFUELING OPERATIONS SHIELDED CASK l

LIMITING CONDITION FOR OPERATION 3.9.16.1 All fuel within a distance L from the center of the spent fuel-pool cask set-down area shall have decayed for at least 1 year. The distance L equals the major dimension of the shielded cask.

l APPLICABILITY: Whenever a shielded cask is on the refueling floor.

ACTION:

With the requirements of the above specification not satisfied, do not move a shielded cask to the refueling floor. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.16.1 The decay time of all fuel within a distance L from the center of the spent fuel pool cask set-down area shall be determined to be 21 year witnin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to moving a shielded cask to the refueling floor and at 7'

l',ast once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter.

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i MILLSTONE - UNIT 2 3/4 9-19 Amendment No. 7#, JEJ 0086

i REFUELING OPERATIONS SPENT FUEL POOL--REACTIVITY CONDITION

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t

. LIMITING CONDITION FOR OPERATION r

3.9.18 The Reactivity Condition of the spent fuel pool shall be such that i

K,77 is less-than-or-equal-to 0.95 at all times.

APPLICABILITY: Whenever fuel is in the spent fuel pool.

ACTION:

[

Borate until K 1 95 is reached.

eff f

SURVEILLANCE REQUIREMENT 4.9.18.1 Ensure that all fuel assemblies to be placed in Region C (as shown in Figure 3.9-2) of the spent fuel pool satisfy either:

(a)

Fuel assembly enrichment and burnup are within the limits of Figure 3.9-la by checking the assembly's design and burnup documentation; or (b)

Fuel assembly enrichment and burnup are within the limits of Figure 3.9-lb by checking the assembly's design and burnup documentation, and borated stainless steel poison pins are installed in the assembly's center guide tube and in two diagonally opposite guide i

tubes.

l 4.9.18.2 Ensure that the contents of each consolidated fuel storage box to be placed in Region C (as shown in Figure 3.9-2) of the spent fuel pool are within the enrichment and burn-up limits of Figure 3.9-3 by checking the design and burn-up documentation for storage box contents.

e 4.9.18.3 Ensure that all fuel assemblies to be placed in Region A (as shown in Figure 3.9-2) of the spent fuel pool are within the enrichment and burnup limits of Figure 3.9-4 by checking the assembly's design and burnup documentation.

i MILLSTONE - UNIT 2 3/4 9-22 Amendment No. J M, JJ7 f

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FIGURE 3.9-1A WINIWUW REQUIRED FUEL ASSEWBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHWENT TO PERWii STORAGE IN REGION C l

MILLSTONE - UNIT 2 3/4 9-23 Amendment No. 197, JJJ JM,

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FIGURE 3.9-18 WINIWUW REQUIRED FUEL ASSEWBLY EXPOSURE AS A TUNCTION OF INiilAL ENRICHWENT TO PERWIT STORACE IN RECl0N C WITH POISON PINS INSTALLED MILLSTONE - UNIT 2 3/4 9-23a Amendment No.

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REFUELING OPERATIONS SPENT FUEL POOL - STORAGE PATTERN I

L;HiTING CONDITION FOR OPERATION 3.9.19 Each STORAGE PATTERN of the Region B spent fuel pool racks shall require that:

(1) A cell blocking device is installed in those cell locations l

shown in Figure 3.9-2; or (2)

If a cell blocking device has been removed, all cells in the STORAGE PATTERN must be vacant of stored fuel assemblies.

l APPLICABILITY:

Fual in the spent fuel pool.

ACTION:

j Take.immediate action to comply with either 3.9.19(1) or (2).

SURVEILLANCE RE0VIREMENTS 4.9.19 Verify that 3.9.19 is satisfied prior to removing a cell blocking

device, j

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MILLSTONE - UNIT 2 3/4 9-26 Amendment No. JEE 0088 r

REFUELING OPERATf0NS BASES 3/4.9.13 STORAGE POOL RADIATION MONITORING The OPERABILITY of the storage pool radiation monitors ensures that sufficient radiation monitoring capability is available to detect excessive radiation levels resulting from 1) the inadvertent lowering of the storage pool water level or 2) the release of activity from an irradiated fuel assembly.

3/4.9.14 & 3/4.9.15 STORAGE POOL AREA VENTILATION SYSTEM The limitations on the storage pool area ventilation system ensures that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

3/4.9.16 SHIELDED CASK The limitations of this specification ensure that in an event of a cask tilt accident 1) the doses from ruptured fuel assemblies will be within the assumptions of the safety analyses, 2) K,77 will remain s.95.

3/4.9.17 MOVEMENT OF FUEL IN SPENT FUEL POOL The limitations of this specification ensure that, in the event of a fuel assembly or a consolidated fuel storage box drop accident into a Region B or C rack location completing a 4-out-of-4 fuel assembly geometry, K will remain eff 1 0.95.

3/4.9.18 SPENT FUEL POOL - REACTIVITY CONDITION The limitations described by Figures 3.9-la, 3.9-lb, and 3.9-3 ensure that the reactivity of fuel assemblies and consolidated fuel storage boxes, introduced into the Region C spent fuel racks, are conservatively within the assumptions of the safety analysis.

The limitations described by figure 3.9-4 ensure that the reactivity of the fuel assemblies, introducted into the Region A spent fuel racks, are conservatively within the assumptions of the safety analysis.

1 MILLSTONE - UNIT 2 B 3/4 9-3 Amendment No. 79, JE9, JJ7, 0089 jf),jff,

REFUELING OPERATIONS l

BASES l

3/4.9.19 SPENT FUEL POOL - STORAGE PATTERN The limitations of this specification ensure that the reactivity will remain l

condition of the Region B storage racks and spent fuel pool Keff less than or equal to 0.95.

The Cell Blocking Devices in the 4th location of the Region B storage 1

racks are designed to prevent inadvertent placement and/or storage in the l

blocked locations. The blocked location remains empty to prcvide the flux trap to maintain reactivity control for fuel assembly storage in any adjacent i

locations.

Region B is designed for the storage of new assemblies in the spent fuel pool, and for fuel assemblies which have not sustained sufficient burnup to be stored in Region A or Region C.

3/4.9.20 SPENT FUEL POOL - CONSOLIDATION The limitations of these specifications ensure that the decay heat rates and radioactive inventory of the candidate fuel assemblies for consolidation are conservatively within the assumptions of the safety analysis.

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DESIGN FEATURES VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 10,060 + 700/-0 cubic feet, j

i 5.5 EMERGENCY CORE COOLING SYSTEMS f

5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

6 5.6 FUEL STORAGE r

CRITICALITY I

i 5.6.1 a)

The new fuel (dry) storage racks are designed and shall be

[

maintained with sufficient center to center distance between assemblies to

<.95.

The maximum nominal fuel enrichment to be stored in ensure a k*N 4.50 weight percent of U-235.

these racks b)

Region A of the spent fuel storage pool is designed and shall be maintained with a nominal 9.8 inch center to center distance between storage locations to ensure a K

$.95 with the storage pool filled with unborated water.

Fuel assemblies $fftored in this region must comply with Figure 3.9-4 to i

ensure that the design burnup has been sustained.

c)

Region B of the spent fuel storage pool is designed and shall be maintained with a nominal 9.8 inch center-to-center distance between storage locations to ensure K 5

95 with a storage pool filled with. unborated I7 assemblie$ stored in this region may have a maximum nominal water.

Fuel enrichment of 4.5 weight percent U-235.

Fuel assemblies stored in this region are placed in a 3 out of 4 STORAGE PATTERN for reactivity control.

d) Region C of the spent fuel storage pool is designed and shall be maintained with a 9.0 inch center to center distance between storage locations to ensure a K ff s.95 with the storage pool filled with unborated water.

Fuel assemblie$ stored in this region must comply with Figures 3.9-la or 3.9-lb to ensure that the design burn-up has been sustained.

Additionally, fuel assemblies utilizing Figure 3.9-lb require that borated stainless steel poison pins are installed in the fuel assembly's center guide tube and in two diagonally opposite guide tubes.

The poison pins are solid 0.87 inch 0.D.

borated stainless steel, with a boron content of 2 weight percent boren.

e)

Region C of the spent fuel storage pool is designed to permit

[

storage of consolidated fuel and ensure aK 0

The contents of I

consolidated fuel storage boxes to be stored 16 f <'sr.95.

f thi egion must comply with l

Figure 3.9-3.

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MILLSTONE - UNIT 2 5-5 Amendment No. 79, EE, 199, 117,159,115,.

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DESIGN FEATURES DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 22'6".

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 224 storage locations in Region A, 160 storage locations in Region B and 962 storage locations in Region C for a tetal of 1346 storage locations.*

• This translates into 1306 storage locations to receive spent fuel and 40 storage locations to remain blocked.

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t MILLSTONE - UNIT 2 5-Sa Amendment No. JE, JE, Jp),

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Millstone Nuclear Power Station, Unit No. 2 Criticality SLfety Analysis. Summary Spent Fuel Pool Modifications i

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U.S. Nuclear Regulatory Commission B14470/ Attachment 3/Page 1 May 14, 1993 7

Millstone Nuclear Power Station, Unit No. 2 Criticality Safety Analysis Summary Spent Fuel Pool Modifications This attachment summarizes the results of criticality safety analyses of the Hillstone Unit No. 2 spent fuel pool (SFP) storage racks.

These calculations utilized the CASM0-3 computer code augmented by the three-dimensional NITAWL-KENO-Sa model with the 27-group SCALE cross-section library.

Sections of the storage cells are identified in 3 regions.

Each of these sections is addressed below:

Recion A Region A,'(utilizing all of the cells in a 4-of-4 cell arrangement with credit for fuel burnup) is not changed from the previous analysis.

None of the rearrangements affect the results of the prior analysis and the burnup-limit curve for Region A remains valid.

L Recion B Region B, (utilizing three cells containing fuel with the fourth cell empty of fuel bearing material) also is not changed from the previous analysis.

None j

of the rearrangements affect the results of the prior analysis and the l

previous analysis for Region B remains valid.

Temperature Effects In Regions A and B, the temperature coefficient of reactivity is negative.

However,-for Region C; the temperature coefficient of reactivity is positive.

i for Region C, a normal maximum temperature was taken as 150*F, and temperatures above 150*F, should they occur, were considered accident conditions where credit for soluble boron is permitted.

For the highest i

accident SFP temperature, 0.95 K,n was not exceeded and therefore no credit for soluble boron is necessary.

Recion C (with Poison Pins) l Region C is designed to use borated stainless steel (B-SS) rodlets or pins I

inserted into the control rod thimbles (3 rodlets per assembly). Analysis for this arrangement resulted in a maximum K,n of 0.940 including calculational and manufacturing uncertainties. The analysis also includes the effect of axially distributed burnup in the assembly.

Table 1 summarizes the calculations for Region C.

Calculations were aise made to define the burnup limit curve as a function of the initial enrichment.

Figure 1 is the limiting burnup curve, yielding at each enrichment the same maximum K,n (0.940).

This burnup-limit

U.S. Nuclear Regulatory Commission B14470/ Attachment 3/Page 2 May 14, 1993 curve identifies the burnups required for acceptable storage of spent fuel in Region C with poison pins.

The B-SS rodlets are aligned diagonally within the fuel assembly.

Calculations showed that there is no difference (within the normal statistical variation of KENO-Sa) for the various possible orientations within the storage rack.

Therefore, the racks can safely accept assemblies with the specified rodlets, regardless of orientation, provided they meet the burnup requirements defined in Figure 1.

Recion C (No poison pins) f Region C is designed to safely accommodate spent fuel without any poison

?

rodlets and utilizing all cells (4-of-4 arrangement).

This requires a higher burnup and the burnup limit curve is shown in figure 1.

The maximum calculated K, in Region C with no poison pins is 0.9459, as indicated in Table l

1, including uncertainties and the estimated reactivity consequence of axial i

burnup distributions.

The reactivity of Region C with poison pins and Region C with no poison pin i

storage cells are very nearly the same.

Therefore, these storage cells are interchangeable and may be intermixed or utilized in any desired configuration in Region C.

Consolidated Fuel Calculations were also made for consolidated fuel (2-to-1 ratio) at one enrichment.

This calculation reasonably confirmed the Combustion Engineering (CE) calculation and lends credibility to the CE calculations for consolidated fuel at other enrichments.

Consolidated fuel bundles would therefore remain i

acceptable for storage in any Region C location, within the burnup-limits reported by CE (current Technical Specification Figure 3.9-3).

t Interface Calculations j

Calculations confirm that there are no adverse reactivity effects at the l

interfaces between any of the rack regions.

New Fuel Elevator The new fuel elevator is located along the east wall of the SFP. Calculations l

(KENO-Sa) of this area of the pool showed that there is virtually no i

interaction between the spent fuel and fresh fuel of 4.5% enrichment in the new fuel elevator.

Consequently, the new fuel elevator may be used without l

any restrictions other than the enrichment limit of 4.5%.

l l

i 4

i U.S. Nuclear Regulatory Commission B14470/ Attachment 3/Page 3 i

Hay 14, 1993

^

Accident conditions i

No accident conditions have been identified that would result in exceeding the regulatory limit on reactivity (K,n of 0.95), with the exception of the accidental misloading of a fresh fuel assembly of the highest permissible reactivity into a cell in Region C intended to receive spent fuel.

l Calculations, however, show that the soluble boron in the pool water (800 ppm 4

per Technical Specification Limit) is more than adequate to compensate for i

this accident condition.

Credit for soluble boron is permissible under accident conditions (single failure criterion) and will assure the reactivity i

is maintained within the regulatory limit.

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t U.S. Nuclear Regulatory Commission B14470/ Table 1/Page 1 l

May 14, 1993.

Table 1 I

Summary of Criticality Safety Calculations for Alternative Storage Arrangements in Millstone Unit 2 Item Recion C Recion C (With Poison Pins)

(No Poison Pins)

Calculated K,n 0.9252 0.9243 Temperature 150*F 150*F Calculational Method CASMO-3 CASM0-3 Bias 0.0000 0.0000 Uncertainties:

Uncertainty in Bias 1 0.0024 1 0.0024 f

Rodlet Diameter 1 0.0012 NA i

Rodlet B-10 loading 1 0.0008 NA Enrichment 1 0.0035 1 0.0054 U0 Density 1 0.0054 1 0.0039 2

Lattice Spacing 0.0052 1 0.0065 SS Wall Thickness 1 0.0024 1 0.0025 Uncertainty in Depletion 1 0.0119 1 0.0175 Calculations"'

Statistical Average 1 0.0150 1 0.0201 Axial Burnup Distribution"'

O.0 1 0.0015 Calculated Reactivity 0.9252 0.9258 i

1 0.0150 1 0.0201 l

f Maximum reactivity 0.9402 0.9459 L

i (1) Evaluated for 3% enriched Westinghouse fuel at 25 MWD /KgU for Region C with poison pins and 35 MWD /KgU for Region C with no poison pins.

Other enrichments and burnups evaluated for appropriate values, all yielding the same maximum reactivity (K n)-

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FIGURE 1 WINtWUW REQUIRED FUEL ASSEWBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHWENT TO PERWii STORACE IN RECl0N C 50 55

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Docket No. 50-336 B14470 i

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- Millstone Nuclear Power Station, Unit No. 2 Safety Assessment Summary for Criticality and Fuel Handling

~i Spent Fuel Pool Modifications s

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U.S. Nuclear Regulatory Commission B14470/ Attachment 4/Page 1 May 14, 1993 Millstone Nuclear Power Station, Unit No. 2 i

Safety Assessment Summary for Criticality and Fuel Handling Spent Fuel Pool Modifications 1.0 Safety Assessment 1.1 Safety Assessment Conclusions The proposed change does not constitute an Unreviewed Safety Question (USQ) and is safe.

1.2 Description of the Change t

L The proposed changes are to Region C of the Millstone Unit No. 2 spent fuel pool (SFP).

The 234 cell blockers will be removed to t

allow an additional 234 storage locations for spent fuel.

i To allow the removal of the blocking devices, borated stainless i

steel rodlets must be placed in fuel assemblies to be stored in Region C, unless the fuel assemblies have enough burnup that the poison rodlets are not needed.

To determine this, 1 curve of assembly average enrichment versus assembly average burnup is 1

provided to determine whether an assembly qualifies for Region C storage without the need for any poison rodlets.

A separate curve of assembly average enrichment versus assembly average burnup is provided to determine whether an assembly qualifies for-Region C storage with the poison rodlets.

If an assembly does not meet one of the two curves, then storage is not allowed in Region C under any ci rcumsta..ces.

An assembly that needs poison rodlets must have three poison rodlets installed in three guide tubes that are directly.in line; that is, the center guide tube and any two diagonally opposite guide-tubes.

Orientation of the fuel assembly in the Region C racks does not matter.

The rodlets are 0.87 inch OD-borated stainless steel, with a boron content of 2 weight percent.

Each poison rodlet is inserted.

into the fuel assembly guide tube and does not protrude above the top of the fuel assembly guide tube.

The poison rodlet shadows all of the active fuel height except for a small portion of the active fuel at the bottom. Under worst case conditions, less than 3 inches of the active fuel height at the bottom of the fuel stack would not be shadowed by the poison pins.

The OD of the poison rodlets is less than that of a CEA.

The length of the poison finger is about l

0.75 inch longer than'a CEA.

The weight of three poison rodlets is less than that of a CEA.

i I

U.S. Nuclear Regulatory Commission B14470/ Attachment 4/Page 2 May 14, 1993 1.3 Aspects of the Chance Evaluated i

This evaluation addresses the criticality aspects. and the fuel handling aspects of the proposed changes.

1.4 Malfunctions Evaluated The criticality analysis evaluates the following malfunctions:

(a) criticality impact of a dropped fuel assembly (b) criticality impact of a dropped cask e

(c) criticality impact of a

fresh 4.5 w/o fuel-assembly accidentally placed in Region C with no poison pins 1.5 References (1) Criticality Safety Analyses of the Millstone Unit No. 2 SFP with alternate arrangements and postulated gaps.

(2) Evaluation of the new ANF fuel design for storage in.the SFP.

(3) Millstone Unit No. 2 FSAR Section 14.7.4 and 14.7.5.

2.0 Unreviewed Safety Question 2.1 Unreviewed Safety Question Determination 2.1.1 List of Accidents Evaluated Per Reference (3), the fuel handling drop accident and the cask drop accident are the two accidents in the SFP that

-l could be affected.

2.1.2 Effect on the Probability of Occurrence of Previously Evaluated Accidents The removal of the cell blockers and the installation of I

poison pins will have no effect on the probability of occurrence of a fuel handling accident-or a cask drop.

2.1.3 Effect on the Probability of Occurrence of Previously Evaluated Malfunction of Eouioment Important to Safety

~

I Previously analyzed malfunctions from the criticality analysis are:

the cask drop accident, fuel assembly drop i

accident, and the accidental misloading of a fresh 4.5 w/o fuel assembly into Region C.

The removal of the cell blockers and the installation of-poison pins will have no effect on the probability of occurrence of previously I

evaluated malfunctions of equipment-important to safety.

l

U.S. Nuclear Regulatory Commission B14470/ Attachment 4/Page 3 May 14, 1993 2.1.4 Effect on the Conseauences of the Previously Evaluated Accidents There is no change in the consequences of the dropped fuel assembly accident since the installation of poison pins will not change the damage caused by the fuel assembly drop.

Therefore, there is no effect on the consequences of the previously evaluated accidents.

2.1.5 Effect on the Conseauences of the Previously Evaluated Malfunctions There is no effect on the consequences of previously evaluated malfunctions. Reference (1) shows that s-.95 K,n is maintained with credit for soluble boron, for the cask drop accident, fuel assembly drop accident, and the accidental misloading of a fresh 4.5 w/o fuel assembly into Region C.

2.2 Potential for a New Unanalyzed Accident 2.2.1 Possibility of an Accident of a Different Tyne than Previously Evaluated r

The removal of the cell blockers and the installation of poison pins does not create the possibility of an accident of a different type than previously evaluated.

l l

2.2.2 Possibility of a Malfunction of a Different Type than Previously Evaluated The only possible new malfunction would be the inadvertent removal of poison pins or not placing poison pins in an assembly that requires them.

Inadvertent removal of the poison pins is not credible since special tooling is required to remove them, effectively making them no different than a fixed poison.

Further, not placing poison pins in an assembly that requires -them is l

effectively a malfunction that is already analyzed, since a fresh 4.5 w/o fuel assembly (with no poison pins) is placed in Region C to verify the K,n is <.95 with soluble boron credit.

Therefore, there is no possibility of a malfunction of a different type than previously evaluated.

2.3 Impact on Marain of Safety The margin of safety is the.95 K,u limit on the SFP during normal and accident conditions.

References (1) and (2) show that K,n will be less than

.95 under all normal and accident conditions; I

i

I 1

U.S. Nuclear Regulatory Commission B14470/ Attachment 4/Page 4 May 14, 1993 therefore, there is no impact on the margin of safety for criticality.

3.0 Safety Determination Based on the fact that no USQ exists and based on the scope of review, the proposed change is safe.

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i Millstone Nuclear Power Station, Unit No. 2 Safety Assessment Summary for the Mechanical, Material, Structural Thermal / Pool Cooling, and Accident Considerations Spent Fuel Pool Modifications t

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U.S. Nuclear Regulatory Commission B14470/ Attachment 5/Page 1 May 14, 1993 Millstone Nuclear Power Station, Unit No. 2 i

Safety Assessment Summary for the Mechanical, Material, Structural Thermal / Pool Cooling, and Accident Considerations Spent Fuel Pool Modifications 1.0

SUMMARY

INFORMATION This assessment will address the mechanical, materi al s, structural, thermal / pool cooling and accident considerations resulting from the proposed change.

1.1 Safety Assessment Conclusions l

Per Northeast Utilities procedures and in accordance with the t

provisions of 10CFR50.59 and 50.92, respectively, the proposed technical specification changes are safe and are not an Unreviewed Safety Question and do not constitute a significant hazards consideration, respectively.

1.2 Description of the Chance The Millstone Unit No. 2 Region C unpoisoned spent fuel racks were originally licensed for 75 percent storage occupancy of intact spent fuel in 728 of the 962 cells (234 cells were blocked off).

This change proposes to introduce poisoned rodlets into the stored fuel t

in Region C to permit removal of the cell blockers and provide 234.

additional spent fuel storage locations.

1.3 Aspects of the Chance Evaluated Mechanical / Material design of the rodlets Seismic / Structural, Thermal Hydraulic consideration of Fuel /

Fuel Rack and Fuel Pool Structure Pool Cooling consideration associated with the increase in a

intact fuel storage capacity l

Radiological Consideration associated with the increase in intact fuel storage capacity l

1.4 Malfunction Evaluated Fuel / Fuel Rack / Fuel Pool Qualifications:

Mechanical Design Configuration of rodlet Thermal Consideration of fuel, fuel rack, and pool cooling Structural Effect of the weight of the rodlets on the fuel / fuel rack / fuel pool interfaces and drop qualifications

i U.S. Nuclear Regulatory Commission B14470/ Attachment 5/Page 2 May 14, 1993 1.5 References CE Licensing Report of Millstone Unit No. 2 Millstone Unit No. 2 PTSCR 2/8/85 Millstone Unit No. 2 Amendment Submittal NRC Amendment Approval No.128 Spent Fuel Seismic Report Rev. 01 Spent Fuel Rack Structural Report Rev. 01 Fuel Pool Cooling Report Rev. 02

[ Vendor] Report Criticality Safety Evaluation for Rodlets Millstone Unit No. 2 FSAR Chapter 14 NUSCO RAB Calculation

[ Vendor] Manufacturing Program / Engineering Package / Certification of Tests NUSCO Specification ASTM A 887-89 Standard Specification for Borated Stainless Steel for Nuclear Applications ASTM A 484-91 Standard Specification for General Requirements for Stainless Steel and Heat Resisting Bars, Billets, Forgings and Strip 2.0 Unreviewed Safety Question Determination 2.1 Impact on Previously Evaluated Accidents No impact to any previously evaluated accidents or malfunctions of equipment important to safety results from this change.

2.1.1 List of Accidents Evaluated t

Design Bases (FSAR)

Fuel Handling Accident FSAR Section 14.7.4 Cask Drop Accident FSAR Section 14.7.5 2.1.2 Effect on the Probability of-Occurrence of Previousiv Evaluated Accidents Design Bases Accidents have been evaluated and determined to be unaffected by the change.

The initiation of either the fuel handling and/or cask drop accidents are unrelated to this proposed change.

Therefore, this change does not i

involve a significant increase in the probability of an accident previously evaluated.

1 2.1.3 Effect on the Probability of Occurrence of a Previously Evaluated Malfunction of Eouioment important to Safety Fuel / fuel rack and fuel pool qualifications have been evaluated for malfunctions and determined to be unaffected by the change.

The initiation of any of the malfunctions o

U.S. Nuclear Regulatory Commission B14470/ Attachment 5/Page 3 May 14, 1993 identified and evaluated are unrelated to this proposed change.

2.1.4 Effect on the Conseauences of the Previously Evaluated Accidents i

Design Bases Accidents have been evaluated and determined to be unaffected by this change.

The referenced Fuel Handling Accident (2.1.1) relates to the radiological consequences of freshly discharged fuel that is dropped to the pool floor and results in the rupture of 14 fuel rods.

The referenced Cask Drop Accident (2.1.1) has been reanalyzed for the radiological consequences of the increased fuel storage capacity in the " targeted footprint area" and determined to be bounded by the previously evaluated accident parameters.

Therefore, this change does not involve a

significant increase in the consequences of an accident previously evaluated.

2.1.5 Effect on the Consecuences of a Previously Evaluated Malfunction of Eauipment Important to Safety Fuel / fuel rack and fuel pool qualifications have been evaluated and determined to be unaffected by this change.

The mechanical design configuration of the rodlets is t

consistent with the shape, size, and weight of a CEA. The material (borated stainless steel) is ASTM approved and licensed by the NRC for use in spent fuel storage technologies and spent fuel pools.

The thermal considerations of fuel are unaffected by the presence of the rodlet because the guide tube is designed for the

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presence of a CEA; therefore, it is not a primary coolant flow area.

The fuel rack normal thermal cooling and malfunction blocked cooling scenarios are unaffected by the presence of the rodlet in the fuel assembly. The fuel pool cooling scenarios of normal, abnormal, single-active failure, and loss of forced cooling are unaffected by the i

increase in intact fuel storage resulting from the rodlets because License Amendment No.128, dated March 31, 1988, and the Fuel Pool Cooling Report, dated February 28, 1985, accounted for an intact spent fuel inventory decay heat history to a maximum of 1965 fuel assemblies.

Therefore, i

the pool cooling scenarios are bounded by previous licensed analysis. The structural effect of the weight of the -rodlet on the fuel / fuel rack / fuel pool structural interfaces and drop qualifications are unaffected because with respect to the fuel, the combined weight of three rodlets is less than the weight of a CEA. With respect to the fuel rack and fuel pool I

i l

U.S. Nuclear Regulatory Commission B14470/ Attachment 5/Page 4 i

May 14, 1993 i

structural interfaces, they are bounded by the weight of a consolidated fuel storage box (-2500 lbs.) in every one of the 1346 storage locations per License Amendment No. 128, dated March 31, 1988.

l 2.2 Potential for a New Unanalyzed Accident i

No potential for any new unanalyzed accidents results from this change.

l 2.2.1 Possibility of an Accident of a Different Tvoe than l

Previous 1v Evaluated All failure modes that cause an accident have been evaluated (design bases, fuel handling, and cask drop accidents). A new failure mode that could represent a new unanalyzed accident is not apparent.

Therefore, this change does not create the possibility of a new or i'

different kind of accident from any accident previously evaluated.

2.2.2 Possibi] tv of a Malfunction of a Different Tvoe than Previousiv Evaluated All conditions that constitute a malfunction have been evaluated (fuel / fuel rack / fuel pool structural interface qualifications).

A new condition that represents a malfunction is not apparent.

Therefore, no new malfunction has been created.

1 2.3 Impact on the Marcin of Safety The mechanical properties and weight of the fuel assemblies remain 1

essentially unchanged. The fuel racks are freestanding and with the inclusion of the weight of the three (3) rodlets per assembly, the

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original mechanical and thermal analyses of the fuel assembly / fuel i

rack and fuel pool building interfaces currently licensed by License Amendment No.

128, dated March 31,

1988, remain valid and conservative. Therefore, this change does not involve a significant reduction in a margin of safety.

3.0 Safety Determination This change is SAFE and is not an Unreviewed Safety Question or a i

significant hazards consideration.

All of the mechanical design qualifications, attributes, and parameters of the fuel racks and fuel pool to store nuclear spent fuel, maintain the fuel assemblies coolable and in a safe subcritical configuration of K.,,

s.95 remain valid, unaffected, and unchanged.