ML20236F154

Application for Amends to Licenses NPF-68 & NPF-81,revising TS LCO 3.3.6 Re CVI Instrumentation & LCO 3.9.4 Re Containment Penetrations to Facilitate Outage Planning

ML20236F154
Person / Time
Site:	Vogtle
Issue date:	06/26/1998
From:	Woodard J SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20236F157	List: ML20236F154 ML20236F163
References
LCV-1149, NUDOCS 9807020047
Download: ML20236F154 (10)

Text

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J. D. Woodard Southern NitclIar

. Executive Vice President Operating Company,1:c.

40 inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205 992.5086 SOUTHERNk L COMPANY Energy to Serve hurWorld" LCV-1149 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Ladies and Gentlemen:

VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS CONTAINMENT PENETRATIONS AND CONTAINMENT -

VENTILATION 1 SOLATION INSTRUMENTATION in accordance with the requirements of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) proposes to revise the Vogtle Electric Generating Plant (VEGP) Unit I and Unit 2 Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.3.6, Containment Ventilation Isolation Instrumentation, and LCO a.9.4, Containment Penetrations, in addition, this submittal includes changes to the Bases appropriate to the proposed changes to the TS.

The proposed changes are as follows:

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e The Applicability of LCO 3.3.6 would be revised to refer to Table 3.3.6-1, and Table 3.3.6-1 would be revised to add a column entitled " APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS". Then, the applicabh modes for Manual Initiation, Automatic Actuation Logic and Actuation Relays, and Safety injection i ))f wot..d be revised to include only Modes 1,2,3, and 4. This has the effect of j n eliminating the requirement that system level manual initiation, automatic actuation '

'J 3 logic and actuation relays, and safety injection initiation be operable during core

. *" alterations and/or during movement ofirradiated fuel assemblies within containment.

Consistent with this proposed change, LCO 3.3.6, Condition C and Required Action C.2 would be revised to reflect that system level manual initiation and automatic l actuation would not be required during core alterations and/or during movement of 1 I

irradiated fuel assemblies within containment. Appropriate Bases changes are 1 included to reflect the proposed changes. l l

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i U. S. Nuclear Regulatory Commission LCV-1149 Page 2

e LCO 3.9.4 would be revised to allow the equipment hatch and the emergency airlock to be open during core alterations and/or during movement ofirradiated fuel

. assemblies within containment. In addition, the LCO statement would be revised to

' reflect that containment ventilation isolation (CVI) would be accomplished by manually closing the individual CVI valves as opposed to a system level manual or automatic initiation, consistent with the proposed change to LCO 3.3.6. The Surveillance Requirements (SRs) would be revised to reflect the proposed change to CVI and to reflect that the equipment hatch would be allowed to be open.

Appropriate Bases changes are included to reflect the proposed changes.

The proposed changes will facilitate outage work planning by: (1) allowing equipment and materials to be moved into containment in parallel with fuel movement, and (2) facilitating maintenance on the containment ventilation system during outages. The existing VEGP TS allow the personnel air lock to be open during core alterations and movement ofirradiated fuel assemblies inside' containment. As described in detail herein, the proposed changes can be made without adversely affecting the health and safety of the public based on the analyses previously performed by both SNC and the NRC and reviewed by the NRC in support of allowing the personnel air lock to be open.

As an administrative matter, SNC notes that existing LCO 3.7.6a, Condensate Storage Tank l (CST)- (Non-redundant CSTs) was created to address a design condition that rendered the i CSTs non-redundant. A note was included with LCO 3.7.6a that stated that this LCO was

! only applicable to.the Unit (s) which have not completed a design modification required for redundant CSTs. The necessary design modifications are now complete for both units.

I Therefore, in accordance with the note to the LCO, LCO 3.7.6a is no longer applicable to l , 'either unit, and SNC is proposing to delete LCO 3.7.6a, and LCO 3.7.6 would be revised to delete the words " Redundant CSTs" from the title. Appropriate Bases changes are included to reflect the proposed change.

- SNC would like to implement the proposed changes during the Spring 1999 Unit I refueling _

outage. To facilitate outage planning, SNC requests approval of the proposed changes by December 15,1998.'

l- The basis for the proposed changes is provided in Enclosure 1. Pursuant to 10 CFR 50.92, an evaluation that demonstrates that the proposed changes do not involve a significant hazard consideration is provided in Enclosure 2. The proposed changes are marked on the affected TS and Bases pages and provided in Enclosure 3. In addition, clean typed TS and Bases pages are provided in Enclosure 4.

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U. S. Nucleir Regulatory Commission

. LCV-1149 Page 3 1'

Mr. J. D. Woodard states that he is Executive Vice President of Southern Nuclear Operating Company and is authorized to execute this oath on behalf of Southern Nuclear Operating Company and that, to the best of his knowledge and belief, the facts set forth in this letter are true. l SOUTHERN NUCLEAR OPERATING COMPANY l

By: .lt

. D. WoodaiT N

' Sworn to andsubscribed before me thidYday o oy ,1998. .

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L, _( tu g Nc ay P ic  ;

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My ommission expires:  ?$78 l

JDW/NJS Enclosures xc: Southern Nuclear Operating Company l

' Mr. J. B. Beasley, Jr.

Mr. M. Sheibani '

NORMS-U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator l Mr. D. H. Jaffe, Senior Project Manager, NRR Mr. John Zeiler, Senior Resident luspector, Vogtle State of Georgia Mr. L. C. Barrett, Commissioner, Department of Natural Resources l

S:\VOOTLE\VOGTLE KVE\TECliSPEC\CNMTISO\lfVil49. DOC

Etciosure 1 Vogtle Electric Generating Plant Request to Revise Technical Specifications Containment Penetrations and Containment Ventilation Isolation Instrumentation Basis for Proposed Changes Proposed Changes The Applicability of LCO 3.3.6 would be revised to refer to Table 3.3.6-1, and Table 3.3.6-1 would be revised to add a column entitled " APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS".

Then, the applicable modes for Manual Initiation, Automatic Actuation Logic and Actuation Relays, and Safety injection would be revised to include only Modes 1,2,3, and 4. This has the effect of eliminating the requirement that system level manual initiation, automatic actuation logic and actuation relays, and i safety injection initiation be opercble during core alterations and/or during movement ofirradiated fuel assemblies within containment. Consistent with this proposed change, LCO 3.3.6, Condition C and Required Action C.2 would be revised to reDect that system level manual initiation and automatic actuation would not be required during core alterations and/or during movement ofirradiated fuel assemblies within containment. Appropriate Bases changes are included to reHect the proposed changes.

LCO 3.9.4 would be revised to allow the equipment hatch and the emergency airlock to be open during core alterations and/or during movement ofirradiated fuel assemblies within containment. In addition, the LCO statement would be revised to reflect that containment ventilation isolation (CVI) would be accomplished by manually closing the individual CVI valves as opposed to a system level manual or automatic initiation, consistent with the proposed change to LCO 3.3.6. The Surveillance Requirements (SRs) would be revised to reflect the proposed change to CVI and to reflect that the equipment hatch would be allowed to be open. Appropriate Bases changes are included to reHect the proposed changes.

As an administrative matter, SNC notes that existing LCO 3.7.6a, Condensate Storage Tank (CST)-

(Non-redundant CSTs) was created to address a design condition that rendered the CSTs non-redundant.

A note was included with LCO 3.7.6a that stated that this LCO was only applicable to the Unit (s) which  !

have not completed a design modification required for redundant CSTs. The necessary design i modifications are now complete for both units. Therefore, in accordance with the note to the LCO, LCO 3.7.6a is no longer applicable to either unit, and SNC is proposing to delete LCO 3.7.6a, and LCO 3.7.6 >

would be revised to delete the words " Redundant CSTs" from the title.

Basis By letter dated November 30,1995, the NRC issued Amendments 92 (Unit 1) and 70 (Unit 2) to the VEGP TS to allow both personnel air lock doors to be open during core alterations and movement of irradiated fuel assemblies inside containment. Given a fuel handling accident inside containment, the resulting dose consequences with both personnel air lock doors open were calculated to be 65.6 rem thyroid and 0.28 rem whole body. These results were within 25% of the 10 CFR 100 limits (75 rem to the thyroid and 6 rem to the whole body) and they are bounded by the current fuel handling accident analysis for the spent fuel pool. A fuel handling accident in the spent fuel pool results in offsite doses of 73 rem to the thyroid and 0.29 rem to the whole body, since no credit is taken for the fuel handling l building emergency filtration system charcoal Ulters. The control room dose associated with a fuel handling accident inside containment with the personnel air lock doors open was found to remain below ,

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4 Vogtle Electric Generating Plant l Request to Revise Technical Specifications Containment Penetrations and Containment Ventilation Isolation Instrumentation Basis for Proposed Changes 30 rem thyroid if me of the four emergency control room filtration units is operating within 7 minutes of the accident. These cesults are within the guidelines of General Design Criteria (GDC) 19 of Appendix A to 10 CFR 50 as definct by Standard Review Plan (SRP) Section 6.4.

The NRC staff pc rformed an independent analysis to determine conformance with the requirements of 10 CFR 100 and GDC 19. The staff's analysis used the accident source term given in Regulatory Guide 1.4, the assumptions contained in Regulatory Guide 1.25, and the review procedures specified in SRP Sections 15.7.4 and 6.4. The staff assumed an instantaneous puff release of noble gases and radioiodines from the gap and plenum of the broken fuel rods. These gas bubbles will then pass through at least 23 feet of water covering the fuel prior to reaching the containment atmosphere. All airborne activity reaching the containment atmosphere is assumed to exhaust to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The gap activity was assumed to have decayed for a period of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The offsite doses calculated by the NRC staff were 37.0 rem to the thyroid and 0.18 rem whole body. The control room operator doses calculated by the NRC staff were 1.38 rem to the thyroid and 0.29 rem to the whole body. The NRC stafi's independent analysis confirmed that the consequences of a fuel handling accident inside containment with the personnel air lock doors open are within the acceptance criteria given in SRP Section 15.7.4 and GDC 19.

The proposed change would allow the equipment hatch and the emergency air lock doors to be open during core alterations and/or movement ofirradiated fuel assemblies inside containment in addition to the personnel air lock doors, which are already allowed to be open. Furthermore, the proposed change would revise the TS such that automatic (or system level manual) containment ventilation isolation would no longer be required during core alterations and/or movement ofirradiated fuel assemblies inside containment. Given that the supporting analysis for the open personnel air lock doors assumed all airborne activity released to the containment atmosphere is released to the environment in a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

period, the presence of additional release paths (i.e., emergency air lock doors and equipment hatch) has no impact on the resulting offsite doses. The release path through the personnel air lock doors to the control room is shorter than the release path from the emergency air lock doors or equipment hatch to the control room. Automatic actuation of the control room emergency filtration system on intake radiogas will coatinue to be required with either or both units in Modes 1,2,3, or 4 and/or during movement of irradiated fuel and core alterations.

In addition, LCO 3.3.6, Table 3.3.6-1 would continue to require the radiation monitors (gaseous, particulate, iodine, and area low range) to be operable to the extent that they would provide alarms in the control room in the event of a fuel handling accident inside containment. Operators could then effect containment ventilation isolation by closing each individual purge and exhaust isolation valve from the control room. This is reflected in the enclosed mark-ups of Table 3.3.6-1, and Bases pages B 3.3-146, 1 147,148,149,and 150.

Even though the dose analysis assumes all airborne activity reaching the containment atmosphere is exhausted to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the personnel air lock must be isolable by at least one air El-2 l- 1 l

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Erciosure 1 Vogtle Electrie Generating Plant Request to Revise Technical Specifications Containment Penetrations and Containment Ventilation Isolation Instrumentation Basis for Proposed Changes lock door with a designated individual available to close the door. Similarly, the changes proposed herein would require the emergency air lock to be isolable by at least one air lock door with a designated individual available to close the door. If the equipment hatch is open, it must be capable of being closed, and SR 3.9.4.2 would be revised to require that the capability to install the equipment hatch be verified at 7 day intervals. The Bases for LCO 3.9.4 would be revised (in part) to state the following:

• " Equipment hatch closure capability is provided by a designated trained hatch closure crew and the necessary equipment. Personnel air lock closure capability is provided by the availability of l

at least one door and a designated individual to close it. Emergency air lock closure capability is provided by the availability of at least one door and a designated individual to close it."

Note that best estimate thyroid doses to the personnel installing the equipment hatch were calculated to be 22.3 rem. This assumes that it takes the personnel I hour to install the equipment hatch.

. "The equipment hatch is considered isolable when the following criteria are satisfied:

1. The necessary equipment required to close the equipment hatch is available.

2. At least 23 feet of water is maintained over the top of the reactor vessel flange in accordance with Specification 3.9.7.

3. A designated trained hatch closure crew is available.

Similar to the air locks, the equipment hatch cpening must be capable of being cleared of any obstruction so that closure can be achieved as soon as possible."

e The Bases for revised SR 3.9.4.2 state that the equipment hatch is provided with a set of hardware, tools, and equipment for moving the hatch from its storage location and installing it in l the opening. The required set of hardware, tools, and equipment shall be inspected to ensure that they can perform the required functions.

e The Bases would also be revised to state that the emergency air lock would be considered isolable under the same conditions as the personnel air lock.

The containment ventilation penetrations must be capable of being closed by redundant operable containment ventilation isolation valves, and new SR 3.9.4.3 would require verification that two containment ventilation isolation valves in each open containment ventilation penetration are capable of being closed from the control room.

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E; closure 1 Vogtle Electric Generating Plant Request to Revise Technical Specifications Containment Penetrations and Containment Ventilation Isolation Instrumentation Basis for Proposed Changes Conclusion As described in detail above, the proposed changes are consistent with TS provisions that allow the personnel air lock to be open during core alterations and movement ofirradiated fuel assemblies inside containment. The proposed changes can be made within the envelope of the dose analysis for the personnel air lock being open during a fuel handling accident. The calculated offsite and control room operator doses remain within the acceptance criteria of SRP 15.7.4 and GDC 19.

Finally, as an administrative matter, SNC notes that existing LCO 3.7.6a, Condensate Storage Tank (CST)-(Non-redundant CSTs) was created to address a design condition that rendered the CSTs non-redundant. A note was included with LCO 3.7.6a that stated that this LCO was only applicable to the Unit (s) which have not completed a design modification required for redundant CSTs. The necessary design modifications are now complete for both units. Therefore, in accordance with the note to the LCO, LCO 3.7.6a is no longer applicable to either unit, and SNC is proposing to delete LCO 3.7.6a, and LCO 3.7.6 would be revised to delete the words " Redundant CSTs" from the title.

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Vogtle Electric Generating Plant l Request to Revise Technical Specifications Containment Penetrations and Containment Ventilation Isolation Instrumentation Significant Hazards Consideration Evaluation Proposed Change In accordance with the requirements of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) proposes to revise the Vogtle Electric Generating Plant (VEGP) Unit I and Unit 2 Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.3.6, Containment Ventilation Isolation Instrumentation, and LCO 3.9.4, Containment Penetrations. In addition, this submittal includes changes f to the Bases appropriate to the proposed changes to the TS.

The proposed changes are as follows:

• The Applicability of LCO 3.3.6 would be revised to refer to Table 3.3.61, and Table 3.3.6-1 would be revised to add a column entitled " APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS". Then, the applicable modes for Manual laitiation, Automatic Actuation Logic and Actuation Relays, and Safety injection would be revised to include only Modes 1,2,3, and 4.

This has the effect of eliminating the requirement that system level manual initiation, automatic actuation logic and actuation relays, and safety injection initiation be operable during core alterations and/or during movement ofirradiated fuel assemblies within containment. Consistent with this proposed change, LCO 3.3.6, Condition C and Required Action C.2 would be revised to reflect that system level manual initiation and automatic actuation would not be required during core alterations and/or during movement ofirradiated fuel assemblies within containment.

Appropriate Bases changes are included to reDect the proposed changes.

i e LCO 3.9.4 would be revised to allow the equipment hatch and the emergency airlock to be open during core alterations and/or during movement ofirradiated fuel assemblies within containment.

In addition, the LCO statement would be revised to reflect that containment ventilation isolation

(CVI) would be accomplished by manually closing the individual CVI valves as opposed to a system level manual or automatic initiation, consistent with the proposed change to LCO 3.3.6.

The Surveillance Requirements (SRs) would be revised to reflect the proposed change to CVI and to reflect that the equipment hatch would be allowed to be open. Appropriate Bases changes are included to reflect the proposed changes.

• As an administrative matter, SNC notes that existing LCO 3.7.6a, Condensate Storage Tank (CST)-(Non-redundant CSTs) was created to address a design condition that rendered the CSTs non-redundant. A note was included with LCO 3.7.6a that stated that this LCO was only applicable to the Unit (s) which have not completed a design modification required for redundant CSTs. The necessary design modifications are now complete for both units. Therefore, in accordance with the note to the LCO, .LCO 3.7.6a is no longer applicable to either unit, and SNC is proposing to delete LCO 3.7.6a, and LCO 3.7.6 would be revised to delete the words

" Redundant CSTs" from the title.

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Enclosure 2 Vogtle Electric Generating Plant Request to Revise Technical Specifications Containment Penetrations and Containment Ventilation Isolation Instrumentation Significant flazards Consideration Evaluation Evaluation The proposed changes have been evaluated against the criteria of 10 CFR 50.92 as follows:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed changes would revise the VEGP Unit I and Unit 2 TS by semoving requirements for automatic and system level manual containment ventilation isolation, and allow the emergency air lock and the equipment hatch to be open during core alterations and movement ofirradiated fuel assemblies inside containment. The containment penetrations affected by the proposed changes are not initiators for any accident previously evaluated. Allowing these penetrations to be open under the ;

conditions specified will not afTect the probability of any accident previously evaluated.

1 The existing VEGP TS allow the personnel air lock doors to be open during core alterations and I movement ofirradiated fuel assemblies inside containment. The radiological consequences of a fuel handling accident inside containment have been determined to be below the Standard Review Plan (SRP) section 15.7.4 criteria and General Design Criteria (GDC) 19 criteria with the personnel air lock doors open. The proposed changes will not alter these previously determined consequences.

The existing dose analysis bounds the proposed changes. Therefore, the proposed changes will not increase the consequences of any accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

No. The proposed change does not create any new failure modes for any system or component, nor does it adversely affect plant operation. The previausly determined radiological consequences of a fuel handling accident inside containment with the personnel air lock doors open remain bounding for operation under the proposed changes. No new single failure scenarios are created, and the proposed changes do not introduce any new challenges to components and systems that could result in a new or different kind of accident from any previously evaluated.

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3. Do the proposed changes involve a significant reduction in a margin of safety?

No. The margin of safety for fission product release is 300 rem thyroid and 25 rem whole body as defined by 10 CFR 100. The previously determined radiological dose consequences for a fuel handling accident inside containment with the personnel air lock doors open remain bounding for operation under the proposed changes. These previously determined dose consequences were determined to be well within the limits of 10 CFR 100 by virtue of the fact that they meet SRP section 15.7.4 and GDC 19 acceptance criteria. Therefore, the proposed changes do not involve a significant j reduction in a margin of safety.

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Eaclos:re 2 Vogtle Electric Generating Plant Request to Revise Technical Specifications Containment Penetrations and Containment Ventilation Isolation Instrumentation Significant Hazards Consideration Evaluation Conclusion Based on the preceding evaluations, the proposed changes do not involve a significant increase in the probability or consequences of any accident previously evaluated, do not create the possibility of a new or different kind of accident from any accident, previously evaluated, and they do not involve a significant reduction in a margin of safety. Therefore, the proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.

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