ML20082L521

Proposed Tech Specs,Revising TS 3.9.4 That Addresses Containment Bldg Penetrations to Allow Use of Alternate Containment Bldg Penetration Closure Methodologies for Postulated Accident Scenarios During Core Alterations

ML20082L521
Person / Time
Site:	Seabrook
Issue date:	04/16/1995
From:	NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:
Shared Package
ML20082L492	List: ML20082L490 ML20082L500 ML20082L521
References
NUDOCS 9504210196
Download: ML20082L521 (16)

Text

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REFUElfNG OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIbiTINGCONDITIONFOROPERATION 3.9.4 Tne containment building penetrations shall be in the. following status:

a. The equipment door closed and held in place y_.a minimum o.f d ur_, e -

bolts, bewever T o r s coae. mr\ock m A minimum of one door in each airlock OPbqg .is close b.

c. Each penetration providing direct access from the containment atmosphere to sideatm here shall be either-y<a or avtomahc. givdwt,

1) Closed by A 1so ation va ve, blind flange, o = nuel vahst or .

2) Be capable of being closed by an ABLE automatic containment MurgeJ,ncLgxhausLilola_t ' o es; gded iuM4b\ avadle d%he.-

3 Ge up & oV big. cbscA  %

APPLICABILITY:

the containment.

Dur1 W E O or - vemen iifTr?sdilitedTtieT MiiiTr\fudtAv

%_r ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building.

SURVEILLANCE REOUIREMENTS repitec(

4.9.4 Each of the above requir alnment building penetrations shall be determined to be either in its -le: _, ice.;ted- ondition or capable of being closed by an OPERABLE automatic con ainmen -

rge and exhaust isolation valve within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:

qdred

a. Verifying the penetrations are in theirI condition,

b. Testing the containment purge and exhaust isolation valves per the

\ applicable portions of Specification 4.6.3.2.

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SEABROOK - UNIT 1 3/4 9-4 g TI 9504210196 950416 PDR ADOCK 05000443 P PDR

- ,. . t 3/4.9 REFUELING OPERATf0NS BAS'ES '

3/4.9.1 BORON CONCENTRATION- ,

The limitations on reactivity conditions during REFUELING ensure that:  :

(1) the reactor will remain subcritical during CORE ALTIRATIONS and (2) a-l uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are l

consistent with the initial conditions assumed for the boron dilution incident in the safety. analyses. The value of 0.95 or less for k,ff includes a 1% ak/k conservative allowance for uncertainties. Similarly, the boron ,

concentration value of' 2000 ppm or greater includes a conservative uncertainty '

allowance of 50 ppm boron.

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 i 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. ,

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS

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y' i The requirements on containment building penetration closure and OPERABILITY d h5Cf.T ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure  ;

b restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack.of containment pressurization i potential while in the REFUELING MODE.

3/4.9.5 COMMUNICATf0N The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

3/4.9.6 REFUELING MACHINE -

The OPERABILITY requirements for the refueling ma: hine ensure that:

(1) refueling machine will be used for movement of drive rods and fuel assem-blies, (2) each hoist has sufficient load capacity to lif t a drive rod or fuel l assembly, and (3) the core internals and reactor vessel are protected from l excessive lifting force in the event they are inadvertently engaged during lifting operations.

SEABROOK - UNIT 1 B 3/4 9-1 k i IWD s@ j

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... ,, INSERT A The 't uiting Condition for Operation (LCO) 1;mits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and exhaust penetrations, the approved alternate closure methods and the containment personnel airlock.

For the approved alternate closure methods, the LCO requires that a designated individual must be available to close or direct the remote closure of the penetration in the event of a fuel handling accident. "Available" means stationed at the penetration or performing activities controlled by a procedure on equipment associated with the penetration.

For the per;onnel airlocks (containment or equipment hatch), the LCO ensures that the airlock can be closed after containment evacuation in the event of a fuel handling accident. The requirement that the airlock door is capable of being closed requires that the door can be closed and is not blocked by objects that cannot be easily and quickly removed. As an example, the use of removable protective covers for the door seals and scaling surfaces is permitted. The requirement for a designated individual located outside of the airlock area available to close the door fdlowing evacuation f the containment will minimize the release of radioactive material.

The fuel handling accident analysis inside containment assumes both of the personnel airlock doors are open and an additional 12" diameter penetration (or equivalent area) is open. The analysis is bounded by these assumptions since all of the available activity is released within a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> priod.

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. ..diI. . , Retype of Proposed Chances See attached retype of proposed changes to Technical Specifications. The attached retype reflects the currently issued version of Technical Specifications. Pending Technical Specification changes or Technical Specification changes 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.

Revision bars are provided in the right hand margin to designate a change in the text.

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a REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION [

3.9.4 The containment building penetrations shall be in the following status:

a. The equipment door closed and held in place by a minimum of four bolts. ,

b. A uir S um of one door in each airlock is closed, however both doors of one personnel airlock may be open if:

1) One personnel airlock door is capable of being closed, and

2) A designated individual is available outside the personnel airlock to close the door. i

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1) Closed by a manual or automatic isolation valve, blind flange, or equivalent: or

2) Be capable of being closed by an OPERABLE automatic containment purge and exhaust isolation valve; or

3) Be capable of being closed by a designated individual available at the penetration.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the i containment.

ACTION:

With the requirements of the above specification not satisfied, immediately ,

suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel I in the containment building.

SURVEILLANCE REOUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by an OPERABLE automatic containment purge and exhaust isolation valve within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:

a. Verifying the penetrations are in their required condition, or

b. Testing the containment purge and exhaust isolation valves per the applicable portions of Specification 4.6.3.2. 1 l

SEABROOK - UNIT 1 3/4 9-4 Amendment No.

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1 3/4.9 REFUELING OPERATIONS

,;BASESI 3/4 9.1 BORON CONCENTRATION ,

i The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subtritical during CORE ALTERATIONS and (2) a uniform r boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The value of 0.95 or less for k includes a 1% Ak/k conservative allowance for uncertainties. Similarly,t$boronconcentrationvalueof2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron.

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 01{ the core.

3/4.9.3 DEhYTIME 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. ,

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The Limiting Condition for Operation (LCO) limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LC0 requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and exhaust enetrations, the approved alternate closure methods and the containment personne airlock .

For the approved alternate closure methods, the LCO requires that a l designated individual must be available to close or direct the remote closure of l the penetration in the event of a fuel handling accident. "Available" means stationed at the penetration or performing activities controlled by a procedure '

on equipment associated with the penetration.

For the Jersonnel airlocks (containment or equipment hatch). the LC0 ensures that t1e airlock can be closed after containment evacuation in the event of a fuel handling accident. The requirement that the airlock door is ca)able of being closed requires that the door can be closed and is not blocked by o)jects that cannot be easily and cuickly removed. As an example. the use of removable protective covers for the coor seals and sealing surfaces is permitted. The requirement for a designated individual locateo outside of the airlock area available to close the door following evacuation of the containment will minimize the release of radioactive material.

SEABROOK - UNIT 1 B 3/4 9-1 Amendment No.

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i REFUELING OPERATIONS

, #ASES' 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS (Continued)

The fuel handling accident analysis inside containment assumes both of the

, personnel airlock doors are open and an additional 12" diameter penetration (or .

equivalent area) is open. The analysis is bounded by these assumptions since all of the available activity is released within a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period. .

3/4.9.5 COMMUNICATIONS  !

The requirement for communications capability ensures that refueling

• station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. i 3/_4. 7 6 REFUELING MACHINE i i

The OPERABILITY requirements for the refueling machine ensure that:

(1) refueling machine will be used for movement of drive rods and fuel assem- 4 blies. (2) each hoist has sufficient load capacity to lift a drive rod or fuel  :

assembly. and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS >

The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel  :

assemblies in the storage pool ensures that in the event this load is dropped:

(1) the activity release will be limited to that contained in a single fuel assembly and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 RESIDUAL PEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140 F as required during the REFUELING MODE. and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23  !

feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loo) will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is available  :

for core cooling. Thus, in the event of a failure of the operating RHR loop. j adequate time is provided to initiate emergency procedures to cool the core.

SEABROOK - UNIT 1 B 3/4 9-2 Amendment No.

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REFUELING OPERATIONS BASES 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

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 consis-tent with the assumptions of the safety analysis.

3/4.9.12 FUEL STORAGE BUILDING EMERCENCY AIR CLEANING SYSTEM The limitations on the Fuel Storage Building Emergency Air Cleaning System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to dis-charge to the atmosphere. Operation of the system with the heaters operating for at least 10 continuous haurs in a 31-day 3eriod is sufficient to reduce the buildup of moisture on tne adsorbers and 4 EPA filters. The OPERABILITY of this ,

system and the resulting iodine removal ca)acity are consistent with the assumptions of the safety analyses. ANSI 1510-1980 will be used as a procedural guide for surveillance testing.

1 3/4.9.13 SPENT FUEL ASSEMBLY STORAGE Restrictions on placement of fuel assemblies of certain enrichments within the Spent Fuel Pool is dictated by Figure 3.9-1. These restrictions ensure that the K ,, of the Spent Fuel Pool will always remain less than 0.95 assuming the pool lo be flooded with unborated water. The restrictions delineated in Figure 3.9-1 and the action statement are consistent with the criticality safety analysis performed for the Spent Fuel Pool as documented in the FSAR. ,

3/4 9.14 NEW FUEL ASSEMBLY STORAGE Restrictions on placement of fuel assemblies of certain enrichments within the New Fuel Storage Vault is dictated by Specification 3/4.9.14. These restrictions ensure that the K,,, of the New Fuel Storage Vault will always remain less than 0.95 assuming the area to be flooded with unborated water. In addition, these restrictions ensure that the K,,, of the New Fuel Storage Vault will always remain less than 0.98 when aqueous foam moderation is assumed. The restrictions delineated in Specification 3/4.9.14 and the action statement are consistent with the criticality safety analysis performea for the New Fuel Storage Vault as documented in the FSAR.

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SEABROOK - UNIT 1 B 3/4 9-3 Amendment No. 6.

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IV. Determination of Sienificant Hazards for License Amendment Reauest 94-06 Proposed I

' ' Changes i Chance 1 - Containment Penetrations  ;

The proposed changes have been evaluated against the standards in 10 CFR 50.92 and have been determined to not involve a significant hazards consideration, in that operation of the facility in accordance with the proposed amendment:

1. Il'ould not involve a sigm'ficant increase in the probability or consequences of an accident previously evaluated The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The Seabrook Station Updated Final Safety Analysis Report (UFS AR) Section 15.7.4, Fuel llandling Accident describes the dropping of a fuel assemblywithin containment as a postulated accident. Using alternate closure methodology to achieve containment closure has no affect on events or conditions which could possibly lead to a Fuel llandling Accident. The use of an alternative closure methodology does not affect the equipment which is used for fuel handling. Therefore, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

The addition of a third option to Technical Specification 3.9.4.c describes a method other than the explicit use of isolation valves, blind flanges, or manual valves that is acceptable to ensure that containment closure capability is achieved for postulated accident scenarios during core alterations or movement ofirradiated fuel within containment. The alterr; ate closure methodology will ensure that penetrations providing direct access from inside containment atmosphere to outside containment are capable of restricting a release of radioactive material to the environment. The alternate closure methodology consists of a designated individual available to close the penetration for openings up to an equivalent 12" in diameter. The proposed closure method is substantiated by the attached radiological evaluation and ensures acceptable isolation for postulated accident scenarios during core alterations or movement of irradiated fuel within containment.

This change represents the potential for an increased dose at the site boundary due to a Fuel llandling Accident. The radiological evaluation was performed for a 12" equivalent diameter opening. With no credit taken for closing the 12" penetration, the calculated offsite and control room doses are well within the acceptance limits of 10 CFR Part 100 and within the acceptance limits of GDC 19 as presented in the evaluation results in Attachment 1. Therefore, the pioposed change will not significantly increase the consequences of an accident previously evaluated.

2. Il'ould not create the possibility of a new or different type of accident from any accident previously evaluated The proposed change afTects a previously evaluated accident, e.g., a Fuel 11andling Accident. It does not represent a significant change in the configuration or operation of the plant and therefore, does not create the possibility of a new or different type of accident from any accident previously evaluated. The alternate closure methodology will ensure that penetrations providing direct access from inside containment atmosphere to outside containment are capable of ustricting a release of radioactive material to the environment. The attached supporting radiological evaluation for the Fuel llandling Accident inside containment supports this change. No changes are being made to fuel handling equipment which could potentially introdace new failure mechanisms. Therefore, Page 9

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" .' because the alternate closure methodology will ensure acceptable isolation of the containment and because no changes are being made to fuel handling equipment, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated in the UFSAR.

3. IVould not involve a significant reduction in a margin of safety.

The proposed changes do not result in a signi0 cant reduction in the margin of safety. It is being proposed that an alternate closure methodology be utilized to provide containment closure for postulated accident scenarios during core alterations or movement of irradiated fuel within containment. The alternate closure methodology is substantiated by the attached radiological evaluation and assures that the calculated offsite and control room doses are well within the acceptance limits of 10 CFR Part 200 and within the acceptance limits of GDC 19. Containment closure means that all potential release paths from the containment are closed or capable of being closed. Containment closure will be provided by the use of a designated individual available to close the penetration in the event of a Fuel llandling Accident.

The alternate closure methodology will ensure that penetmtions providing direct access from inside containment atmosphere to outside containment are capable of restricting a release of radioactive material to the environment and limiting offsite doses to well within acceptance limits of 10 CFR Pan 100 and within the acceptance limits of GDC 19. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Chance 2 - Personnel Airlock Doors The proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to not involve a significant hazards consideration, in that operation of the facility in accordance with the proposed amendment:

1. Would not involve a sigm*ficant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to Technical Speci0 cation 3.9.4 and its Bases would allow the containment personnel airlock (PAL) doors to be open during core alterations or movement of irradiated fuel within containment. The PAL is not an initiator to any accident. Therefore, the position of the PAL doors (open or closed) during core alterations or movement of irradiated fuel within containment has no affect on the probability of any accident previously evaluated.

Allowing the PAL doors to be open during core alterations or movement of irradiated fuel within containment does increase the dose consequences of a Fuel Handling Accident in the containment.

The requirements of the Limiting Condition for Operation (LCO) 3.9.10, " Water Level - Reactor Vessel", and the minimum decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to core alterations (LCO 3.9.3, Decay Time) ensure that the release of fission product radioactivity, subsequent to a fuel handling ,

accident, results in offsite doses that are well within the acceptance limits specified in 10 CFR Part 100 and within the acceptance limits of the control room dose criteria of 5 Rem whole body or equivalent organ dose of GDC 19. Therefore, the results of the analysis supporting the proposed change do not represent an unacceptable increase in offsite and control room doses. In addition, the calculated doses are larger than the expected doses because the calculation does not incorporate the closing of the PAL door after the containment is evacuated. The proposed change will significantly reduce the dose to workers in the containment in the event of a Fuel Handling Page 10

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' , Accident by speeding the containment evacuation process. The proposed change will also significantly decrease the wear on the PAL doors and, consequently, increase the availability of the PAL doors in the event of an accident.

The proposed change does result in increased dose projections for the Fuel liandling Accident occurring inside the containment, however, the conservatively calculated results are well below the acceptance limits of 10 CFR Part 100 and GDC 19. Therefore, the proposed change does not result in a significant increase in the consequences of an accident previously evaluated in the UFSAR. The benefits provided to the refueling personnel and the undue wear and tear on the PAL door seals support the proposed change. ,

2. Would not create the possibility of a new or different type of accident from any accident previously evaluated.

The proposed change does affect a previously evaluated accident; the Fuel llandling Accident.

h does not represent a significant change in the configuration or operation of the plant and, therefore does not create the possibility of a new or different type of accident from any accident previously evaluated.

The alternate manual closure of the containment PAL doors will ensure that the containment is capable of restricting a release of radioactive material to the environment. The attached supporting radiological evaluation for the Fuel llandling Accident inside containment supports this change.

No changes are being made to fuel handling equipment which could potentially introduce new failure mechanisms. Therefore, the proposed change does not create the possibility of a new or difTerent accident from any accident previously evaluated in the UFSAR.

3. Would not involve a sigmficant reduction in a margin of safety.

Allowing the PAL doors to be open during core alterations or movement of irradiated fuel within containment does increase the dose consequences of a Fuel liandling Accident in the containment.

The increase in calculated offsite and control room doses resulting from a Fuel llandling Accident are well within the acceptance limits of 10 CFR Part 100 and within the acceptance limits of GDC

19. Actual offsite and control room doses in the event of a Fuel llandling Accident will be less than analyzed because the PAL door will be closed following evacuation of the containment.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The Commission has provided guidance concerning the application of standards in 10 CFR 50.92 by providing certain examples (March 6,1986 SIFR7751) of amendments that are considered not likely to involve a significant hazards consideration. This proposed change is very similar to example (vi) in the Federal Register Notice, in that this change results in an increase in the consequences of a previously analyzed accident, but the results of this change are clearly within all acceptance criteria. The criteria for Fuel llandling Accident results are found in Standard Review Plan 15.7.4, Revision 1, Section 11. " Acceptance Criteria", provides exposure guidelines fbr offsite dose calculations. The guidelines given are 75 Rem to the thyroid and 6 Rem to the whole body. The results calculated for the EAB,62.74 Rem the thyroid and 1.95 Rem to the whole body, and control room doses of 6.69 Rem to the thyroid and 0.29 Rem to the whole body, clearly fall within those acceptance criteria of the SRP and GDC 19 respectisely. Therefore, based on the information contained in this submittal, this change to the Technical Specifications does not result in a reduction to the margin of safety as defined in the basis for any technical specification.

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l V. . Ptsoosed Schedule for License Amendaient issuance and Effectiveness  !

North Atlantic requests NRC review of License Amendment Request 94-06 and issuance of a license amendment having immediate effectiveness by October 1,1995. North Atlantic is ,

scheduled to begin refueling outage 4 in November 1995. The proposed changes utilize alternate -

closure methods to provide containment closure during core alterations or movement ofirradiated {

fuel within co.3tainment. This would allow flexibility during times when containment closure is required whici' will simplify outage scheduling and assist in reducing the length of refueling outages. ,

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,- I., . . Environmental Impact Assessment North Atlantic has revie' ' the proposed license amendment against the criteria of 10CFR51.22 for environmental const ions. The proposed changes do not involve a significant hazards consideration, nor increax .he types and amounts of efiluent's that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, North Atlantic concludes that the proposed change meets the criteria delineated in 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement.

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i A TTACilMENTI Results and Summary ofinput and Assumptions Used to Analyze tire Containment Fuelllandline Accident .

He analysis follows the guidance of Standard Review Plan 15.7.4 and Regulatory Guide 1.25. Specific analysis input and assumptions are discussed below.

PARAMETERS USED TO EVAI.UATE CIIANGE 1

1. All 264 fuel rods contained in the most active fuel assembly are assumed to rupture and release the contained noble gas and volatile iodine fission gap fission product inventory. This source term is consistent with Regulatory Guide 1.25.

2. Fuel rod gap activities consist of 10% iodine, and 10% noble gases (except 30% Kr-85) of the maximum rated assembly gaseous activity inventory. These release fractions are consistent with Regulatory Guide 1.25,

3. Maximum rated fuel assembly gaseous iodine and noble gas isotopic inventories are taken from Seabrook Station UFSAR, Chapter 15, Section 15.7.4, Table 15.7-20. Nuclear characteristics of the highest rated discharge assembly are given in UFSAR, Table 15.7-19.

4. Equilibrium core fission product inventories are based on extended operation at 3654 MWt which is approximately 107% of the rated thermal power (3411 MWt).

5. The single damaged assembly is assumed to have experienced a power level (3654 MWt/193)

• 1.65. This 1.65 peaking factor is consistent with Regulatory Guide 1.25.

6. Consistent with Regulatory Guide 1.25, assembly gap iodine inventory is composed of 99.75%

inorganic and 0.25% organic species of iodine.

7. The fuel handling accident is assumed to occur 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> aller shutdown, consistent with Seabrook Station Technical Specification 3/4.9.3.

8. The fuel handling accident does not result in containment pressurization. A conservative containment driving force has been assumed using the following conservative assumptior.s; I

a) Containment pressure is derived from the containment on-line purge fan peak pressure, sustained throughout the release period, b) Containment temperature is assumed to be 50 F.

c) The containment exhaust fans are assumed to be out of service.

d) No credit is taken for containment depressurization as a result of the vent path opening.

e) The containment atmosphere vent flow rate is based on the area the 12-inch ,

containment s udge lance penetration. This conservatively bounds the four sludge l lance fluid system openings.

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f) Based on the above, the resulting containment atmospheric vent flow rate for the fuel handling accident inside the containment with potential releases through the containment sludge lance penetrations is conservatively calculated to be 7.6E+05 scfh, (approximately 2.7 containment volumes per day). Offsite (EAB) doses are evaluated for two cases,1) containment penetration isolation in 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, and no isolation for two hours.

9. The control room doses are evaluated for a 30 day time period using the containment vent flow rate release described above, continuing for 30 days. The control room emergency filtration system is activated by the redundant safety grade intake radiation monitors (2), with a 2 second bypass period for the filters. Control room intake and recirculation filter efficiencies are 95% for inorganic,95% for organic and 99% for particulate iodine species. The control room emergency filtration system operates at 2000 cfm (1200 cfm from 2 fresh air intakes (600 cfm each), one of which is assumed to be intaking clean air, and 800 scfm of recirculation air). The control room volume is 2.46E+05 ft'. A discussion of control room operability during a design bases event is given in Chapter 15, Section 15.6.5.4.e of the UFSAR.

10. No credit is taken for atmospheric cleanup systems in the containment (carbon recirculation filters are available).

11. Accident atmospheric dispersion factors assume a ground level release as given in the UFSAR Tables 15B-4 for the EAD and UFSAR Table 15B-6 for the control room. Control room intake X/Q's are divided by two to reflect the clean air.

12. No credit is 13 ken for deposition of the plume on the ground or decay of isotopes in transit to receptor locations.

13. Dose conversion factors for evaluating the thyroid and efTective dose equivalent whole body doses are consistent with ICRP 30 dose conversion factors.

14. The decontamination factor for noble gases in the pool is taken as 1.0. This assumption is consistent with Regulatory Guide 1.25

15. The total effective decontamination factor of iodine is taken as 100. This assumption is conservative and consistent with Regulatory Guide 1.25.

16. Buildup of daughters is taken into account as the source term nuclide's decay.

17. No credit is taken for iodine plateout in the containment.

18. The breathing rate of 3.47E-04 m'/sec is taken from Regulatory Guide 1.25.

PARAMETERS USED TO EVAL.UATE CIIANGE 2

1. All parameters are the same as used to evaluate change 1 above, except for the assumed containment leak rate. Radioactive gases that escape from the refueling pool to the containment building are released to the atmosphere exponentially over a two hour period. This is equivalent to releasing 77% of the available activity in 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,95% in I hour and 99.75% in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The site boundary doses are calculated over a 0-2 hour time period, while the control room doses are evaluated over a 0-30 day release period. This release rate assumption is bounding for the occupance of both the sludge lance penetration and the personnel airlock doors being open at the same time, since all of the available activity is released within a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period.

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',,, Radiolonical Dose Evaluation Results* i The total 0 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Exclusion Area Boundary (EAB) thyroid and whole body doses associated with ,

a fuel handling accident occurring inside containment and open PAL' doors are: }

Thyroid Dose = 63 Rem (inhalation and immersion contributions);  !

Whole Body Dose = 2 Rem -j i

The total 0-30 day Control Room Dose thyroid and whole body doses associated with a fuel l handling accident occurring inside containment and open PAL doors are:

Thyroid Dose = 6.7 Rem; ,

Whole Body Dose = 0.29 Rem The total 0-2 hour EAB thyroid dose and effective dose equivalent whole body dose associated with a fuel handling accident occurring inside the containment with open Steam Generator Sludge  !

Lance penetrations is: ,

Thyroid Dose = 8.4 Rem (0.5 hr isolation), and 24 Rem, with no credit for isolation. p Whole Body Dose = 0.26 Rem (0.5 hr isolation) and 0.74 Rem with no credit for isolation.

The total 0-30 day Control Room Dose thyroid and whole body doses associated with a fuel r handling accident occurring inside containment and an open Sludge Lance Containment Penetrations are:

0-30 Day Thyroid Dose = 3.8 Rem; 0-30 Day Whole Body Dose = 0.8 Rem  :

• REFERENCE = SBC-669 ,

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