ML20236H822

Revised Application for Amend to License DPR-73,consisting of Tech Spec Change Request 53 & Recovery Operations Plan Change Request 38,incorporating Responses to Five of Six Comments in NRC

ML20236H822
Person / Time
Site:	Crane
Issue date:	10/26/1987
From:	Standerfer F GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20236H565	List: ML20236H562 ML20236H576 ML20236H822
References
NUDOCS 8711040370
Download: ML20236H822 (17)

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~e, METROPOLITAN EDISON CCNPANY IJERSEY CENTRAL:POWERJAND LIGHT COMPANY i

PENNSYLVANIA ELECTRIC COMPANY 3,

r GPU NUCLEAR THREE MILE ISLAND NUCLEAR STATION UNIT II Operating License No. OPR-73 Docket'No. 50-320.

. Technical Specification Change Request No. 53 and Recovery Operations Plan Change Request No."38 - Revised i

This3 Technical Specification. Change Request and Recovery Operations Plan

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Change Request-is submitted in supporti>of Licensee's request to change Operating ' License No. DPR-73 for Three Mile Island Nuclear Station Unit 2.

As'a part.of this request, proposed replacement pages for i

Appendix A are also included.

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GPU NUCLEAR 4

By ~ D1' rector, TMI-2

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. Sworn and subscribed to me this AC* day of bcixdh

, 1987.

\\t kk Nota'ry 3ublic-tale til&M(LLE LIB 0, MOTARY PUBLIC i

gggY TWP., S.M)PHIN COMTY l

M C0ligt30el EllPlats MPT,11.1989 2 -

Neuber,Pennsyivants Association of Notaries i

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UNITED STATES 0F AMERICA l

NUCLEAR REGULATORY CCMMISSION IN THE MATTER OF DOCKET NO. 50-320 LICENSE NO. DPR-73 GPU NUCLEAR This is to certify that a copy of a revision to Technical. Specification Change Request No. 53 and Recovery Operations Plan Change Request No. 38 to Operating License DPR-73 for Three Mile Island Nuclear Station Unit 2 has been filed with the U.S. Nuclear Regulatory Commission and served to the chief

. executives of 1) Londonderry Township, Dauphin County, Pennsylvania, 2)

Dauphin County, Pennsylvania, and 3) the designated official of the Commonwealth of Pennsylvania by deposit in the United States mail, addressed as follows:

f Mr. Jay H. Kopp, Chairman Mr. Fred Rice, Chairman Board of Supervisors of Board of County Comndssioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection PA Dept of Environmental Resources P.O. Box 2063 Harrisburg, PA 17120 I

i GPU NUCLEAR

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Director, TMI-2

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u Three Mile Island Muclear Station, Unit-2 (TMI-2)

Operating License No. OPR-73 Docket No. 50-320 l-1 l

L Technical Specification Change Request (TSCR) No. 53 and Recovery Operations Plan Change Request (ROPCR) No. 38 - Revised The licensee requests that the attached Technical Specification and Recovery Operation Plan Change Requests replace, in their entirety, those previously submitted for NRC review and approval via GPU Nuclear letters 4410-87-L-0042 dated April 23, 1987, and 4410-87-L-0041 dated April 14, 1987.

l-Purpose of Change l

l-The above referenced CPU Nuclear letters sutxnitted TSCR 53 and R0PCR 38, respectively, for NRC approval. Comments from the NRC TMICPD on these submittals were received via letter NRC/TMI-87-073 dated September 25, 1987.

L The attachment provides GPU Nuclear's response to five (5) of the six -(6) NRC comments and the Safety Evaluation Justifying Change for each revision. The L

response to NRC Comment 6 will be submitted at a later date.

l Additionally, the attached revised pages reflect minor corrections to reflect current Technical Specification requirements ano to reflect references to current regulations.

L No Significant Haza_rds Determination 1

The proposed changes are within the scope of the No Significant Hazards Determination analysis originally submitted in TSCR 53.

The proposed changas are primarily administrative in nature to 1) delete the requirement for NRC-approved procedure pursuant to Specification 6.8.2, and

2) clarify the requirements of proposed specifications. For example, the proposed revision modifies the definition of Containment Integrity (i.e.,

1 Specification 1.7) to more accurately reflect the current conditions at THI-2. Also, in response to NRC comments, administrative clarification has been made to Table 1.1, " Facility Modes;" Specification 3.7.6.1, " Flood Protection;" and Specification 3.8.1.1, "A.C. Sources." The deletion of the

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reference to Specification 6.8.2 is based on an analysis which demonstrates j

that suitable controls are in place such that an activity that could result in l

an off-site dose greater than 10 CFR 50 Appendix I limits is highly unlikely, i

Furthermore, the reference to Specification 6.8.2 in other specifications has l

been replaced, where appropriate, with limitations currently imposed in NRC-approved procedures.

Thus, per the criteria of 10 CFR 50.92, the proposed revision does not:

1.

Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.

Create the possibility of a new or different kind of accident from any I

accident previously evaluated; or 3.

Involve a significant reduction in a margin of safety.

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' Based on the,No Significant Hazards Determination in the original submittal of TSCR 53 and from the above revlee, GPU Nuclear concludes that the changes

. proposed in revised TSCR 53,and ROPCR 38 do tiot involve Significant' Hazards.

J Amendment Class An application fee, per the requirements of 10 CFR 170, was previously i

submitted for TSCR 53 and R0PCR 38.

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1 NRC COMMENT 1 f

On page' 1-5 of the proposed TS in item 2.9, if you do not intend to complete

.def ueling of the reactor coolant system prior to entering mode 2 provide a l

justification.

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- GPU NUCLEAR RESPONSE l

- The initially proposed definition of Mode 2 in Table 1.1 of Technica3 Specification Change Request (TSCR) 53 requires that both in-vessel and ex-vessel defueling be j

completed print to transition to Mode 2.

This is supported by proposed criterion a

"a" of-Mode 2 which states: "There are no canisters containing core material in the Reactor Building." Additionally, the No Significant Hazards Determination in TSCR 53 states, "The transition from Mode 1 tu Mode 2 is based oc defueling the RV i

and Reactor Coolant System and precluding criticality."

For purposes of clarity, criterion "a" of Mode'2 has been revised to reference both

- the Reactor Vessel and the Reactor Coolant System.

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Additionally, based on discussion with the NRC TMICPD Staff, the following sentence has been added to Specification 1.3, " Mode,":

" Prior to the transition to the

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followlog facility Mode, a report shall be submitted to the NRC no later than 30 days prior to transition which provides the basis for the transition."

The proposed change to Specification 1.3 is administrative in nature to ensure that the NRC is notified within sufficient time prior to the transition to the following mode.

i NRC COMMENT 2 1

Please clarify your intent on containment integrity and containment isolation on j

pages 11, 1-2 and.1-4.

Will two operable valves or an otherwise closed penetration i

be required for containment integrity? If not>, what penetrations other than those j

1 listed in Table 3.6 2 will be affected, how, and why?

GPU NfJCLEAR RESPONSE

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Specification 1.7 currently requires that each Reactor Building penetration be l

closed by at least two (2) barriers to ensure that no releases to the public will be made via the Reactor Building through these penetrations. This section does 1

recognize that there are occasions where it is necessary to operate containment I

isolation valves. Therefore, in the original February 1980 order issuing the Recovery Technical Specifications, the NRC recognized that such operations would be i

required from time to time and established specific provision for submittal to the l

NRC for approval, pursuant to Specification 6.8.2, to ensure that no inadvertent

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n release to the environs occurs.

it is also noteworthy that anytime a containment

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isolation valve or barrier is opened for any purpose, it must be performed in I

accordance with an NRC-approved procedure, pursuant to Specification 6.8.2, specifically written for the purpose and function of the specific activity involved.

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. Containment Integrity, as currently defined in Technical Specification 1.7, refers to having the capability to maintain double barrier isolation of each penetration and is applicable during Mode 1.

Containment Isolation, as defined in proposed

. Technical Specification 1.21, refers to single barrier isolation of each penetration and is applicable during Modes 2 and 3.

Containment Integrity can be defined as follows:-

o With the exception of those penetrations identified in Technical Specification 4

Table 3.6-2, penetrations are:

Closed by manual valves, blind flanges, or deactivated automatic valves

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secured in their closed positions on each side of the penetration, j

Isolation valves inside the Reactor Building shall be capable of remote i

operation from a control station outside of the Reactor Building, or; Closed double valve isolation outside the Reactor Building when the f

inside barrier is inaccessible due to ALARA considerations or the penetration has been modified as part of the cleanup effort (i.e.,

Mode 1), or; Open per an approved procedure but can be closed pursuant to either of the above criteria.

o The Equipment Hatch is closed and sealed.

o Each airlock is OPERABLE pursuant to Specification 3.6.1.3.

.o' The sealing mechanism associated with each penetration (e.g., welds, bellows j

or 0-rings) is OPERABLE.

o Water level in Spent Fuel Storage Pool "A"/ Fuel Transfer Canal 1s maintained pursuant to Specifications 3.9.2/3.9.4 when the Fuel Transfers tubes are open.

This definition of containment integrity is consistent with the current definition, but has been modified to define specific criteria under which double valve l

isolation external to containment would be allowed due tc the unique circumstances of TMI-2.

These criteria are similar to those which fall under the present provision of allowing double valve isolation outside containment in accordance with j

Technical Specification 6.8.2 (i.e., NRC approval).

j A special case involving the Spent Fuel Storage Pool " A"/ Fuel Transfer Canal has J

also been added to the containment integrity definition which ensures containment j

integrity for the Fuel Transfer tubes by maintaining the water level in accordance j

with Technical Specification 3.9.2/3.9.4.

Evolutions which require both containment isolation valves to be open (e.g.,

Reactor Building purge, water processing / transfer) are permitted in accordance with approved procedures provided the capability exists to re-establish containment integrity in accordance with Technical Specification 3.6.1.1.

Therefore, GPU Nuclear has proposed that the following be incorporated in Specification 1.7 for i

those penetrations that are permitted open per an approved procedure:

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" Controls shall be implemented to minimize the tima the penetration is open and to specifj the conditions for d11ch the penetration is open. Penetrations shall be expeditiously' closed upon completion of the conditions specified in the approved procedures."

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The above wording has also been incorporated into the definition for Containment Isolation (i.e., Specification 1.21).

1 The current definition of Containment Integrity, as does the proposed revision, contains a requirement that:

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" Isolation valves in the Reactor Building shall be capable of remote operation frem a' control station outside of the Reactor Building."

The following manually-operated isolation valves located inside the Reactor

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Bnlding, per the plant design, currently do not satisfy this requirement:

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Valve No.

Penetration 1

SF-V104 R-524 j

DH-V161 R-525

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DW-V139 R-535 FS-V640 R-548 SV-V17 R-569 j

Per discussions with the NRC TMICPD staff, the aoove penetrations have oeen added to Table 3.6-2.

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Additionally, a provision has been added (i.e.., Specification 3.6.1.1.b) to allow 4

relaxation of the containment integrity recus.rements for those instances when a penetration is not capable of being isolateo Dy a double barrier (e.g., during a penetration modification) provided certain controls are taken to ensure isolation is maintained. These controls are specifieo in a revision to the Recovery Operations Plan (i.e., Section 4.6.1.1.b) which requires that the single barrier be 1

verified closed at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If, for some reason, the single

- barrier becomes inoperative, certain operations (i.e., core alterations, heavy load 1

handling in the Reactor Building, welding and burning) will be suspended until the i

capability of providing single barrier isolation is again provioed.

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operations are the same as those currently prohibited when both doors of the containment lock airlock are open.

Based on the proposed revision to Specification 1.7, the reference to Specification 6.8.2 has been deleted from Specification 3.6.1.a and Recovery Operations Plan Section 4.6.1.1.a.

Specifically, the LCO for Specification 3.6.1.1.a has been revised to state, " Primary Containment Integrity shall be maintained." Recovery Operations Plan Section 4.6.1.1.a has been revised to state, " Containment '.ntegrity shall be demonstrated at lease once per 31 days." Additionally, the LCO for Specification 3.6.1.2 has been revised to state, " Primary Containment Isolation l

shall be maintained."

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NRC COMMENT __3 It is. the NRC staff's intent that, other than section 3.9.13, the TS no longer take credit for NRC review of procedures; please update those sections that reference or imply NRC review of procedures. Approp2iate limitations should be substituted for procedure revie*,

(Sections 3.1.1.1, 3.6.1.1, 3.6.1.3)

GPU NUCLEAR RESPONSE Other than Specification 3.9.13, " Accident Generated Water,", the remaining specifications which currently require or imply NRC review of procedures are:

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Specifications 1.7, 3.6.1.1, and Recovery Operations Plan Section 4.6.1.1, l

" Containment Integrity."

o Specification 3.1.1.1.c, "Boration Control - Boration Cooling Water Injection."

o Specification 3.6.1.3 and Recovery Operations Plan Section 4.6.1.3,

" Containment Airlocks (Mode 1)."

o Specification 6.8.2, " Procedures."

o Specification 6.8.3.1, " Procedures."

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Recovery Operations Plan Section 4.6.1.4.b, " Internal Pressure."

The above specifications, with the exception of Specifications 1. 7, 3. 6.1.1, and 4.6.1.1 are addressed below.

The response to NRC Comment 2 addresses the change to Specifications 1.7, 3.6.1.1, and 4.6.1.1 including substituting the reference to 1

Specification 6.8 ? with appropriate limitations.

i Specification 3.1.1.1.c, "Boration Control - Barated Cooling Water Injection" The phrase "...except as changed per procedure approved pursuant to Specification 6.8.2..." has been deleted.

l The SWST water level was established by NRC Amendment of Order dated August 8, 1985, which approved TSCR 46. This water level was established based on an unisolable leak from the Reactor Vessel and feeding that leak from the BWST during the interim period before a recirculation flow path could be established to recirculate the water in the Reactor Building basement through the Reactor Vessel.

At the time TSCR 46 was prepared, it was planned to fill the deep end of the refueling canal with BWST water. Currently, there is no foreseeable need during l

Mode 1 to reduce the BWST volume below the minimum Technical Specification limit.

l Specification 3.6.1.3 and Recovery Operations Plan Section 4.6.1.3, " Containment Airlocks (Mode 1)"

Specification 3.6.1.3 has been revised to change the phrase "per procedures approved pursuant to Specification 6.8.2" to "per the criteria of Recovery Operation Plan Section 4.6.1.3.1."

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Tha' reference to procedures approved to Specification 6.8.2 has been deleted from 14.6.1.3.1.b and the section has been revised as follows:

"When both Equipment Hatch Personnel Airlock doors are opened simultaneously, verify the following conditions:

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The Airlock doors and containment purge are configured to restrict the outflow of air. During periods when both Airlock doors are open, the following activities shall be prohibited:

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Core alterations.

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Heavy load lifts in the Reactor Building.*

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Activities which could result in a fire in the Reactor Building (e.g.,

welding, burning)."

• This does preclude completion of the activities necessary to place the plant in a safe configuration.

The proposed limitations when both Airlock doors are open are currently identified in THI-2 Procedure 4210-OPS-3240.01, " Reactor Building Entry," Revision 2-00.

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procedure has received NRC review and approval. These limitations serve to minimize the potential for.significant releases to the environment during periods when both Airlock doors are open. Thus, since the proposed change is consistent with current NRC-approved limitations, this change is primarily administrative in nature and no further safety evaluation is required.

l GPU Nuclear believes that restrictions similar to the sbove are not required for Specification 4.6.1.6.1, " Containment Airlocks (Modes 2 and 3)."

During Modes 2 and 3, the Reactor Coolant System will be in a defueled, sub-critical condition, thus, the restriction in Core Alterations will not apply.

Additionally, during Modes 2 and 3, there will be no credible accident scenario in the Reactor Building (RB) that could result in an off-site dose greater than 10 CFR 50 Appendix I limits. Thus, the proposed restrictions on movements of heavy loads and activities which could result in a fire in the Reactor Building should not apply during Modes 2 and 3.

Proposed Specification 4.6.1.6.1 will require that during periods when both airlocks doors are open, the capability exists to close at least one (1) door and that the doors and containment purge are configured to restr.ict the outflow of air in accordance with site-approved procedures.

Additionally, Bases 3/4.6.1.3 has been revised to delete to reference to Specification 6.8.2.

Specification 3.9.13, " Accident Generated Water" Per the current Technical Specification requirements, the sentence, " Discharge of Accident Generated Water shall be prohibited until approved by the NRC," has been inserted into the LCO for this specification. Additionally, the reference to Specification 6.8.2 has been changed to "NRC-approved procedures."

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i THI-2 TSCR $6 as revised by GPU Nuclear letter 4410-87-L-0057, dated April 13 -

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1987, proposed to delete the above referenced sentence from Specification 3.9.13.

l Thus, the proposed change is administrative in nature to reflect the current

-language of the specification pending NRC approval of revised TSCR 56.

The proposed change to revise the phrase " procedures approved pursuant to Specification 6.8.2" to "NRC-approved procedures" is administrative in nature to retain the requirement for NRC approval of procedures for discharge of Accident Generated Water.

Thus, no further safety evaluation is required.

Recovery Operations Plan Section 4.6.1.4.b, " Internal Pressure" The reference to procedures approved pursuant to Specification 6.8.2 has been deleted. The following sentence has been added to this section: "The restrictions of 4.6.1.3.1.b, ' Containment Airlocks (Mode 1),' and 4.6.1.6.1.b, ' Containment l

Airlocks (Modes 2 and 3),' shall also apply."

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The reference to Specification 6.8.2 was added to Recovery Operations Plan Section l

I 4.6.1.4 via ROPC 28, dated April 12, 1985. This change was necessitated to recognize the possibility for the Control Room instruments to indicate a positive l

Reactor Building pressure (with respect to the Auxiliary Building) when, in fact,

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the Reactor Building pressure is negative with respect to the environment. The requirement for Specification 6.8.2 approval was to ensure the proper l

implementation of procedural requirements to maintain negative flow in the Reactor l

Building when it is opened to the environment. Presently, such procedural l

requirements are in effect (e.g., 4210-OPS-3240.01, " Reactor Building Entry") and are approved by the NRC.

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Additionally, the proposed change adds c reference to the proposed restrictions of 4.6.1.3.1.b, which applies during Mode 1, and 4.6.1.6.1.b, which applies during j

Modes 2 and 3, which prohibit certain activities when the Containment Building is 1

open to the atmosphere which have also been incorporated into Basis 3/4.6.1.4.

Thus, this proposed change will not reduce any safety margins or result in increased effluents to the environment.

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Specification 6.8.2.2 i

l Currently, Specification 6.8.2.2, as revised by License Amendment 28, dated June 25, 1987, states: " Procedures of 6.8.1.a and changes thereto which alter the distribution or processing of a quantity of radioactive raterial, the release of which could cause the magnitude of radiological releases to exceed 10 CFR 50 Appendix I design objectives, shall be subject to approval by the NRC prior to implementation."

Table 1 lists those activities, currently identified in NRC-approved Licensing Basis Documents, that have the potent 131 to exceed 10 CFR 50 Appendix I limits.

Each activity is addressed below:

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A.

Peavy Load ' Handling Inside Containment and Over Spent Fuel ' Storage Pool"A" -

The Safety Evaluation Report (SER) for Heavy Load Handling Inside Containment and Over Fuel Pool "A"-(Reference 1) and the SER for Heavy Lead Handling Over the TMI-2 Reactor Vessel (Reference 2) both address the drop of a heavy load

.into the Reactor Vessel (RV). As noted in Table 1, both SERs state that if such a drop occurred resulting in an instantaneous total release of the unaccounted for Kr-85 inventory, the resultant off-site dose release could exceed 10 CFR 50 Appendix I limits.. Based on the following analyses, GPU Nuclear believes that this accident scenario is no longer credible.

An analysis was performed (Reference 3) to determine a more realistic source term for the Kr-85 release postulated to occur due to a heavy load drop into the RV.,.

The source term was reevaluated based on the considerable experience gained from defueling activities and on the current conditions of the core.

4 The maximum amount of Kr-85 available for release due to a heavy load drop was determined to be 335 C1. The resultant. maximum off-site doses were calculated to be 0.1 mR (whole body) and 8.8 mR (skin). Both are below 10 CFR 50 Appendix.I limits.

Additionally, the Defueling Work Platform positioned over the RV is designed to withstand the impact of a heavy loao based on the load and lift height limits in the SER for Heavy Load Handling over the Reactor Vestal, i

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Canister Orop - The NRC-approved Defueling SER (Reference 4) identifies that the worst-case scenario (i.e., entire canister contents spilled out on dry surface) would result in a maximum off-site dose greater than 10 CFR 50 Appendix I limits. However, based on the following analysis, GPU Nuclear believes that this worst-case accident scenario is highly unlikely.

Section 4.4.2.2, '" Canister Drop Accident," of Rcfarence 4 states that the a

Canister Transfer Shield la " designed with dive m means for preventing a canister drop accident while the canister is being transported from the Reactor Vessel to the deep end of the Fuel Transfer Canal. Since multiple

, failures are required for a canister drop accident to occur over the dry portion of the Fuel Transfer Canal, such an event is considered extremely J

unlikely. The NRC-approved SER for Canister Handling and Preparation for

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Shipment (Reference 5), Section 8.0, "Unreviewed Safety Question Evaluation (10 CFR 50.59)," states that the Fuel Transfer Cask is " designed to meet NUREG-0612 and to provide single failure proof protection against the drop of

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a canister during transport." Similarly, as discussed in the NRC-approved Technical Evaluation Report (TER) for Defueling Canisters (Reference 6),

j leakage of canister contents is not expected for the drop heights tested based on tests performed by Babcock and Wilcox.

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Additionally, the experience gained to date with canister handling coupled i

I with the controls established in current NRC-approved procedures (e.g.,

4200-OPS-3255.01, "Defueling Operations") provide a high degree of confidence

.that a canister drop accident is highly unlikely.

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I' Fire in the Waste Handling and Packaging Facility (!!HPF) or the Containment C,

, Air Control Envelope (CACE) - Tne NRC-approved WPHF TER (Reference 7) and CACE TER (Reference 8) identify that a fire in either of these facilities could result in a maximum off-site dose release greater than 10 CFR 50 Appendix I l

limits. Based on the following analyses, GPU Nuclear believes that the fire l

accident' scenario in either the WHPF TER or the CACE TER is highly unlikely.

The WHPF TER does not take credit for the fire protection system in the WHPF for a contaminated fire.

In actuality, this systems would serve to mitigate the off-site dose resulting from a fire in the facility.

Table 4.3-3 of the Recovery Operations Plan requires that the WHPF Exhaust Monitor be operable during operation of the WHPF. As noted in the NRC-approved Recovery Operations Plan Change 34, dated July 10, 1986, "This monitor will provide for the rapid detection of a release of radiation to the environment thereby allowing sufficient time for the licensee to initiate mitigating actions to protect the public."

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The WHPF contains a fire suppression deluge system and portable fire extinguishers.. Additionally, fire alarms are provided which annunciate both locally and in the TMI-2 Control Room.

Also, those procedures which currently I

authorize burning or cutting in the WHPF (i..e, 4230-IMP-7270.01, " Operation of the Plasma Arc Cutting Out#1t Inside the Waste Handling and Packaging Facility," and 4230-IMP-7220.02, " Operation of Oxygen / Acetylene Cutting Torch Inside the Waste Handling and Packaging Facility") include controls to minimize the potential and mitigate the consequences of a fire in this I

facility. These controls include:

o Verification that the fire protection system is operable prior to performing activities that have the potential to generate unintended combustion and that portable fire protection is readily available, o

Enclosing the work area with welding screens / curtains, as necessary.

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o Removing all combustibles from the area, if possible.

Concerning a fire in the CACE, Section 3.3.3.2, " Fire Protection," of the CACE TER states:

"Should a fire occur in the containment when both of the airlock doors are open, one of the doors will be closed by the individual recuired by the procedure, and the control room notified, thereby reestablishing the containment boundary and preventing an uncontrolled release of radioactivity to the environment.

The addition of the CACE will not change the reactor j

building fire boundary since the equipment hatch will remain installed."

Additionally, Table 4.3-3 of the TMI-2 Recovery Operations Plan requires that l

the CACE Vent Monitor be operable when the CACE is in operation.

NRC-approved 1

Recovery Operations Plan Change 26, dated April 3,1985, states, "The monitor will provide confirmation that releases from the CACE are a small fraction of 10 CFR 50 Appendix I and plant Tecnnical Specification limits during normal l

operaon.

The monitor will also be able to detect a degradation of filter l

efficiency."

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Thus, based on the rationale provided above, the need to require NRC-approved

' procedures in the current safe-shutdoen condition of THI-2, eiPh the single exception of those procedures regarding disposal of Accident Generated Water, is no longer existent and this requirement an be deleted.

Specification 6.8.3.1 i

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Item "d" currently requires that temporary changes to procedures described by i

Specification 6.8.2 are to be submitted to the NRC within 72-hours following site approval. The reference to Specification 6.8.2 has been changed to reference i

d Specification 3.9.13, " Accident Generated Water."

The proposed change is administrative in nature.

The reference to Specification 6.8.2 will no longer apply based on the proposal to delete Specification 6.8.2.

However, NRC approval will still be required, per Specification 3.9.13, for those procedures which discharge Accident Generated Water.

Therefore, no further safety evaluation is required.

1 NRC COMMENT 4 Sections 3.7.6.1 and 3.7.11 refer to safety related systems, components and structures. How will the list of these systems, components and structures be defined and maintained.

GPU NUCLEAR RESPONSE The Limiting Condition for Operations (LCO) for Specification 3.7.6.1, " Flood Protection," states, " Flood protection shall be provided for all Safety Related systems, components, and struc,tures..."

The LCO for Specification 3.7.11, " Penetration Fire Barriers," states, "All Penetrations Fire Barriers protecting Safety Related areas shall be functional."

Concerning Specification 3.7.6.1, the action statement specifically identifies actions to be taken in the event of a flood.

These actions include installation of all flood panels and door seals, checking all water tight doors to ensure proper operation and checking all containers in the Southeast Storage Facility. GPU Nuclear has proposed to retain the current requirements of this specification through Modes 1, 2, and 3.

Additionally, GPU Nuclear letter 4410-87-L-0093 dated June 23, 1987, informed the NRC that, as a minimum, the requirement for the current flood protection doors will remain in effect during Post-Defueling Monitored Storage (PDMS).

Thus, the term " safety-related" is not germane to Specification 3.7.6.1 and this term may be deleted from the LCO and Basis 3/4.7.6.

Concerning Specification 3.7.11, GPU Nuclear is currently preparing for NRC review and approval a revision to the TMI-2 Fire Protection Program Evaluation (FPPE) to reduce the number of fire areas based on the current safe shutdown condition of TMI-2.

In conjunction with this submittal, GPU Nuclear will also submit a TSCR to

.the current fire protection requirements, including Specification 3.7.11, based on the proposed revision to the FPPE.

Thus, there is no need to modify the term

" safety-related" in Specification 3.7.11 as part of this revision to TSCR 53.

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.NRC COMMENT 5 The proposed section 3.8.1.1 would allow heavy load lifts in the reactor building with both circuits between offsite transmission and the onsite class lE distribution system inoperable. Please update the request or justify an alternate method of reactor building sump recirculation in the event of a load drop in the reactor vessel area.

l GPU NUCLEAR RESPONSE Action Statement "B" of Specification 3.8.1.1, "A.C. Sources," was revised via License Amendment 27, dated April 17, 1987, to prohibit Core Alterations and movements of heavy loads in the Reactor Building in the event of a loss of off-site power. Specification 3.8.1.1, as originally submitted in TSCR 53, has been revised to reflect this restriction.

MISCELLANEOUS CHANGES Specification 3.9.2, " Spent Fuel Storage Pool

'A' Water Level" and 3.9.4, " Fuel Transfer Canal (Deep End) Water Level" Description of Change The action statements of the above specifications have been revised to comply with the current requirements as revised by License Amendment 28, dated June 25, 1987.

Reason for Change To comply with the current requirements for the referenced specifications.

Safety Evaluation Justifying Change The proposed change is administrative in nature to comply with current rechnical Specification requirements as revised by License Amendment 28 which was issued af ter the submittal of TSCR 53.

Specification 6.4.1.1, " Training" Description of Change The reference to Appendix "A" of 10 CFR Part 55, " Operators' Licenses," has been replaced with a reference to 10 CFR 55.59, "Requalification."

Reason for Change This proposed change complies with current regulations.

10 CFR 55, Appendix A formerly titled, "Requalification Programs for Licensed Operators of Production and utilization Facilities," has been deleted effective May 26, 1987 (reference 52 FR 9453 published March 25, 1987).

The correct refercnce is 10 CFR 55.59, "Requalification."

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i Safety Evaluation Justifying Change l-L The proposed change is administrative in nature to comply with current L

' regulations. Therefore, no further safety evaluation is required.

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TABLE l':.

SUMMARY

OF OFF-SITE DOSES SER OR TER SCENARIO ANNUAL MAX. OFF-SITE DOSE i

10 CFR 50,' Appendix I Limits 5 mrem-(total body) 15 mrem (skin) 15 mrem (other organs) 1 Heavy Load Handling Drop in Reactor Vessel that 9.7 mrem (total body)

Inside Containment /

releases all unaccountable Heavy Load Handling Kr-85 r

Over The Reactor.

i Vessel SERs Defueling SER Canister drop that releases 3,000 mrem (bone)-

canister contents i

Waste Handling and Fire 119 mrem (equivalent y

YI Packaging Facility TER whole body dose) i]

Containment Air Fire 119 mrem (equivalent

/

-Control Envelope TER whole body dose) d

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REFERENCES

~

31,'1987,"HeavyLoutlf ' ~ p. fling 1. ' rPU Nuclear ' letter 4410-87-L-0123, dated August Inside Containment and Over Fuel Pool 'A'." 2. GPU Nuclear letter 4410-85-L-0089, dated April 19,1985, " Heavy Load Handling Otst the TMI-2 Reactint Vest;el." R ll V, L 3. Design Engineering memorandum DESE-0479, dated August'18, 1987, "Offsite Oose Due to a Heavy Load drop in the Resi: tor Vessel." 4. GPU Nuclear letter 4410-86-L-0049, dated May 15,1986,"DhfuelingSafety Evaluation Report - Revision 10." 1 e/ 5. GPU Nuclear letter 4410-86-L-0099, dated June 11,1986, " Canister Handling and ' Preparation for Shipment Safety Evaluation Report." p b 6. CPU Nuclear letter 4410-87-L-0052, dated May 7, 1987, " Annual Update of the f Defueling Canister Technical Evaluation Report." l I 7. GPU Nuclear. letter 4410-87-L-C016, datsd.Februady'2p(1987, " Wast.e Handling and Packaging Facility Technical Evaluation' Report.F1 8. GPU Nuclear letter 4410-86-L-0206, dated Occ,eiber 24, 1986, " Containment Air Control Envelope Technical Evaluation Rei[ ort." I, 4 I J I r ~\\ -f r f

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