ML20195B699

Safety Evaluation Supporting Amend 30 to License DPR-73

ML20195B699
Person / Time
Site:	Three Mile Island
Issue date:	05/27/1988
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20195B673	List: ML20195B666 ML20195B688 ML20195B699
References
NUDOCS 8806210488
Download: ML20195B699 (16)

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\ 1 NUCLE AR REGULATORY COMMISSION UNITED STATES

[ - W ASHINGTON. O. C. 20333 r , ;f t 2 gv.....f SAFETY EVALUATION BY TFE OFFICE OF NL' CLEAR REACTGR REG'JLA*:09 RELATED TO AMENCHENT NO. 30 TO FACILITY OPERATING LICENSE NO. CPR ~3 GPU NUCLEAR CORPORATION THREE MILE ISLAND NUCLEAR STATICN, UNIT NO. 2 00CKET NO. 50-320

1.0 INTRODUCTION

By letter dated April 23, 1987, GPU Nuclear Corporation (GPUN or the licensee) reouested the approval of changes to the Technical Specifications -

of Facility Operating License No. DPR-73 for Three Mile Island Nuclear Station, Unit No. 2. GPUN provided supporting information for the propcsed modifications. The request was revised by letters dated October 25, 1987, November 9, 1987, and December 4, 1987. The proposed amendment includes revisions to the Apoendix A Technical Specifications Sections 1-Oefinitions, 2-Safety limits, 3-Limiting Conditions for Operation, 3/4-Basis for Limiting Conditions for Operation, and 6-Administrative Controls. The proposed amendment is an extensive revision intended to appropriately align the license requirements to the current, as well as, the projected future plant conditions as the plant progresses in stages through the remainder of the cleanup operations. The revision defines three modes of operation which correspond to key milestones in the cleanup operatien. Only those license conditions applicable to the facility conditions in.each mode would require implementation. Phase out of specific license conditions as the plant proceeds to successive modes would be predicated on the submission by the licensee of a report which would describe the plant conditions and give the basis for progression to the next mode.

2.0 OISCUSSION AND EVALUATION Section 1.3 Technical Specification 1.3 defines Recovery Mode as the plant condition in which the reactor is suberitical and the average reactor coolant system temperature is less than 200*F. No other modes are defined.

The licensee procoses to modify the Technical Specification by defining three distinct facility modes corresponding to the projected plant conditiens as facility cleanuo progresses. The definition of Mode 1 is the same as the present definition of Recovery Mode and corresponds to tre presently existing plant conditions during which defueling and other major tasks are in progress. Mode 2 will exist subsequent to the defueling of the reacter and reactor coolant system when the possibility of criticality 8806210488 PDR 880527 P

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y in the reactor buildirg is precluded and no canisters containing ccre material remain in the reactor building. Mode 3 will be the period following shipment of all canisters containing core material to an offsite location. Thirty days prior to an anticipated ecde change, the licensee proposes to submit a report to the NRC staff providing the basis for the mode change. The staff, however, requires, and the licensee agrees, that the reports providing the basis for a mode change be submitted a minimum of sixty (60) days in advance of the anticipated transition date. Additionally the staff requires that at least 90 days orier to transition to Facility Mcde 2 the licensee submit a repcrt which evaluates the applicability of the technical provisions of 10 CFR 50 and the Appendices thereto, to TMI-2 in Facility Pode 2 and thereafter. This report shall demonstrate that necessary design features required for protection of the public health and safety exist. Although not part cf their original preposal the licensee has agreed to ccmply with the staff's requirement.

The staff has concluded that the modes as defined in the licensee's oroposal provide for suitable transition points for describing general plant ccnditions through the remaining phases of the cleanup program. T h e-mode changes are planned to occur coincident with significant reductions in the levels of risk associated with the facility.

Section 1.7 Technical Specification 1.7 cefines the conditions that must be met to provide Containment Integrity. It requires (11 the capability of dcuble isolation of containment penetrations in accordance with NP.C approved procedures, (2) that isolation valves inside of the contairment be capable of remote cceration, and (3) that penetrations without NRC aporoved precedures for isolation be closed with double isolation of the penetration.

The licensee's proposed revision incorporates the same requirements on isolation capability of penetrations with double barriers. However, it adds clarification on the controls that must be implemented to assure the capability to isolate those penetrations that are open for operational needs. In addition, the revision adds a requirement that 5:ent Fuel Fcol

'A' and Fuel Transfer Canal water level be within the specified range to meet the definition of Containment Integrity.

The staff has concluded that the proposed definition adecuately describes the requirements to assure a containment structure capable of preventing the release of radioactive material to the environment. Since NRC staff review of licensee procedures will no longer be required, the proposec Technical Specification has incorporated requirerrents for centrol of penetrations open for operational needs. These requirements are ecuivalent to those ;reviously imposed through the NPC staff review process. In addition, it includes the added conservatism of maintaining a sufficient water level in the Spent Fuel Fool ' A' and Fuel Transfer Canal to ensure a nater seal over the Fuel Transfer Tubes.

Section 1.21 The 'echnical 5:ecificatiers do not define any level of c:ntainment closure capability otner than Centainment Integrity, i

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The licensee has proposed a new Technical Specification definiticn defining Containment ! solation. This would prescribe a contairment functienal capability less than full containment integrity and will involve a condition similar to contair.rert integrity but with cnly single barrier isolation of contaircent penetrations.

The staff.has concluded that this is an acceptable level of containment performarce for the olant conditions that will exist after cefueling when no potential for criticality in the reactor building exists. The containment will serve as a barrier against the spread of residual contamination where perfontance levels ccmparable to that of auxiliary building and fuel handling buildings are adequate. This is discussed further under the evaluation of sections 3.6.1.1 and 3.6.1.2.

Section 2.0 The current Technical Specifications state under Safety Limits, Reactor Ccolant System Pressure, that it is "not applicable," This is the only Technical Specification under "Safety Limits." -

The proposed revision would modify Technical Specification 2.0 to read, "There are no safety limits which apply to TMI-2."

The staff has concluded that this is an acceptable change since it accurately reflects the fact that there are no safety limits applicable to this facility under 10 CFR 50.36(c)(11 10 CFR 50.36(c)(1) requires safety limits necessary to reascnably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity. Essential parameters typically limited in plants that are operational are inapplicable at TMI-2. The integrity of physical barriers is not challenged at TMI-2 to the extent that it is at operational plants since there is no source of energy to widely dispense radioactivity should the physical barriers be breached.

This represents an administrative change, since the current specification already indicates that there are no applicable safety limits.

Section 3.0.1 Technical Specification 3.0.1 states that the Limiting Conditions for Operation and Action requirements are applicable during the Recovery Mode.

The licensee proposes to modify Technical Specification 3.0.1 to state that the Limiting Conditions for Operation and Action recuirements are applicable during the Facility Mede specified fer each LCO and associated Action requirement.

This is consistent with the realignment of licer.se conditions and the definitions of three dif'erent Facility Modes as previously discussed.

This proposed change is acceptable to the staff.

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Section 3.1.1.1 Technical Specification 3.1.1.1 prescribes the operational recuirerents for available sources of Borated Cooling Water Injection and the requirec acticrs when one or more sources become unavailable. It specifies applicability of the requirement during the Recovery Mode. :t allows deviation frem the borated water s:crage tank minimum volume recuirements if the deviation is controlled in accordance with an NRC acproved procedure.

The licensee proposes to modify this section by specifying its applicability to only Mode 1, deleting the allowance to deviate from the specification if procedurally approved, and to add the requirement to invoke the action statement if borated water storage tank temperature is outside of the specified range.

The staff has concluded that removal of the allowance to deviate from the reouirement without NRC prior approval is consistent with the intent to remove the require: rent for NRC staff approval of licensee procedures. -

This change will require the licensee to seek fonnal license amendment to deviate from the requirement during the mode in which it is applicable.

Addition of the minimum temperature action requirement is consistent with the Limiting Conditions for Operation as previously written. Modifying the applicability to Mode 1 is consistent with new definitions of Facility Modes. Mode 1 is the only Facility Mode in which criticality or core cooling are issues. Following defueling and assuring that no potential for a criticality in the reactor building exists, there is no longer a need for borated cooling water injection capability. The staff finds the proposed changes acceptable.

Sec tic _ns 3.1.1.2, 3.1.1.3, and 3.1.1.4 Technical Specifications 3.1.1.2 and 3.1.1.3 prescribe the required boron concentrations and temperature in the water in the Reactor Coolant System, Fuel Transfer Canal, and the Spent Fuel Storage Pool ' A' . T>ese parameters must be maintained within the specified limits during the Recovery Mode.

The licensee proposes a modification that would prescribe the same limits on boron concentration and temperature in the Reactor Coolant

! System and Fuel Transfer Canal and would change the applicability to Mode 1. The modification would place the requirements on boron concentration and temperature for the Spent Fuel Storage Pcol ' A' in a new section 3.1.1.a with applicability to Modes 1 and 2.

l l The staff has concluded that the specified baron concentration and temperature in the Reactor Coolant System, Fuel Transfer Canal, and Spent Fuel Storage Pcol ' A' need only be maintained when there is sufficient

! core material present to pose a potential threat of inadvertent i criticality. Af ter transition frem Mode 1, there will be no cessibility l

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< of a criticality in the reactor building per the definitions in Technical Specification 1.3. Therefore, apolicability of the prescsed Technical Specifications 3,1,1.2 and 3.1.1.3 in only Mode 1 is appropriate. Since Mode 2 will still permit storage of filleo fuel canisters in the Spent Fuel Storage Pool ' A', it is appropriate to provide boration and temperature control in the pool. However, upon transition to Mode 3, all fuel containing canisters will have been removed from the site and Specification 3.1.1.4 may appropriately be relaxed. The staff finds tne proposed changes acceptable.

Section 3.3.1.1 Technical Specification 3.3.1.1 requires operability of the specified nuclear instrumentation during Recovery Mode. It prescribes the remedial action to be taken in the event of inoperability of the nuclear instrumentation and recuires special reports on these events to the Cemission.

The licensee proposes to modify the specification by changing the -

apolicability to Mode 1 and by eliminating the requirement in subsequent modes.

In conjunction with the issuance of this amendment, the Commissicn has granted an exemption to 10 CFR 70.24, Criticality accident requirements to the licensee.

The staff has concluded that since the purpose of the nuclear instrumentation is to provide information on the shutdown status of the core, it is not needed for this purpose when defueling has progressed to the point of having no possibility of a criticality in the reactor building. Therefore, changing the applicability of this specification to Mode 1 as defined in Specification 1.3 is acceptable.

Sections 3.3.2 and 3.3.2.1 Technical Specifications 3.3.2 and 3.3.2.1 prescribe operability requirements for the Engineered Safety Actuation System Instrumentation.

The licensee proposed to delete these specifications.

The instrumentation affected by this specification is that instrumentation that causes an automatic start of the emergency diesel generators in the event of a loss of off-site electrical power. The staff has concluded, based on our previous approval of deletion of operability requirements of the emergency diesel generators, that all loads served by the diesel generators may be safely interrupted until off-site pcwer is restered or can be supplied with back-up pcwer frem the existing station batteries. Therefore, since the diesel generator operability requirements have been previously deleted, the requirement for autcmatic start instrumentation may also be safely deleted. The staff firds the proposed e.hanges acceptable, i

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er Secticn 3;3.3.4 Technica1' Specification 3.3.3.4 recuires ocerability of the meteorological instrumentation when in the Recovery Yode and recuires restoring incperable instrumentation to operable status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The licensee proposes to modify this specification by changing the applicability to Modes 1 and 2 and by requiring restoratien of inocerable instrumentation within 7 days.

The basis for requiring monitoring meteorological ata is to evaluate the need to initiate measures to protect the health and safety of the public during accidents involving off site radiological releases. The worst case accident during Mode 3 is a fire in the reactor building. Analysis of this accident shows that the offsite dose from the facility would be less than the 10 CFR 50 Appendix I numerical design objectives for offsite doses from releases due to normal operations. Thus no protective measures are required. Therefore, there would be no recuirement to maintain operable meteorological monitoring instrumentation for TMI-2 in Pode 3.

The proposed action statement time period of 7 days is consistent with BfN Standard Technical Specifications. The more conservative 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period was imposed during the phases of cleanup when high levels of Krypton-85 and other radionuclides existed in the containment atmosphere. The high levels of radioactivity presented a higher level of potential threat to the public health and safety if released, thus a more significant need for the ability to assess the threat. As cleanup has progressed the magnitude of potentially airborne radioactivity that could be released has diminished significantly. Therefore, the original need to rapidly restore meteorolcgical instrueentation no longer exists and the staff finds the proposed changes acceptable.

Section 3.3.3.5 Technical Specification 3.3.3.5 requires operability during Recovery "ode of Essential Parameters Monitoring Instrumentation used to monitor Reactor Building Pressure, Reactor Vessel Water Level, Incore Themoccuoles, Reactor Building Water Level, Borated Water Storage Tank Level, Steam Generator Level, Spent Fuel Storage Pool ' A' Water Level, and Fuel Transfer Canal Water Level.

The licensee proposes to modify this specification by requiring operability of the Incore Thermocouples, Reactor Building Water Level, Borated Water Storage Tank Level, and Steam Generator Water Level during Mode 1 only.

These parameters are monitored to verify the core condition and assure availability of core cooling and criticality control. Following core removal and the subsequent transition to Modes 2 and 3 the need to maintain these parameters will no longer exist. The recuirements for l

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7 Reactor Building Pressure has been retained but transferred to the Recovery Operations Plan requirement associated with Technical Specification 3.6.1.4 The recuirement for Reactor Vessel Water t.evel has been retained but transferred to Technical Specificatien 3.4.2 and the associated Recovery Operattens Plan requirement. The requirement for Spent Fuel Storage Pool ' A' and the Fuel Transfer Canal water levels have been retained but transferred to Technical Specificaticns 3.9.1 and 3.9.3 respectively. The staff finds the proposed changes acceptable.

Section 3.3.3.3.7 Technical Specification 3.3.3.7 requires operability of the Chlorine Cetection System during the Recovery Mode.

The licensee proposes to require operability only in Mode 1. ~he basis for operability of the system is to provide automatic isolation of the control room ventilation system to ensure habitability of the control room in the event of a chlorine leak. Since the manning of the control -

room will not be required after core removal and the subsecuent transition to Modes 2 and 3, ensuring a habitable control rcom will no longer be necessary to protect the public health and safety. The staff finds the proposed change acceptable.

Section 3.3.3.8 Technical Specification 3.3.3.8 reouires operability of the fire detection instrumentation during the Recovery Mode.

The licensee has proposed an administrative change to reouire operability in Modes 1, 2 and 3.

This is consistent with the propesed definitions of Facility Yodes and enccmpasses all plant conditions oreviously included in the Recovery Mode. The staff finds the proposed change acceptable.

Section 3.4.2 Technical Specification 3.4.2 requires operability of reactor vessel water level instrumentation during Recovery Mode when the reactor vessel head is removed.

The licensee proposes a revisien to this specification that would require operability only during Mode 1. This instrumentation provides 'or indication of loss of reactor coolant system inventory or water inleekage to the reactor coolant system that could result in e boron dilution.

Once defueling of the reactor vessel is comolete, it will no longer M necessary to maintain water in the reactor coolant system. Therefore, this instrumentation will not be required in Medes 2 and 3. The staf#

finds the proposed change acceptable.

y I-Section 3.4.9

! Technical Specification 3.4.9 prescribes the reactor coolant system temperature and pressure limits that must be maintained during the

'ecovery Mode.

The licensee proposes to modify the applicability of this specification

to Mode 1.

Temperature and pressure limits are prescribed to prevent non-ductile failure of the reactor vessel, to prevent precipitation of the bcron used to control criticality, and to prevent boiling in the reactor coolant system. Following defueling of the reactor coolant system, the reactor vessel will no longer be reouired to contain the core, the potential source of heat that could cause boiling will be removed, and there will be no further need to maintain soluble boron to control criticality. Therefore, there is no need to monitor these paraceters following transition to Modes 2 and 3. The staff finds the proposed -

change acceptable.

Section 3.5.1 Technical Specification 3.5.1 requires direct ccmunication between the Control Room or Comand Center and personnel in the Reactor Guilding.

It is applicable during Core Alternations.

The licensee proposes to modify this specification by making it applicable in Mode 1 during Core Alternations.

This change is administrative in nature since after transition to Modes 2 and 3, core alternations will no longer be possible. The change is consistent with the licensee's definition of Facility Modes. The staff finds the proposed change acceptable.

Section 3.6.1.1 Technical Specification 3.6.1.1 requires that Containment Integrity be maintained unless it is not required by procedures approved by the NRC staff. The specification is applicable during the Recovery Mode, i

The licensee proposes to tredify this specification by making it applicatie to Mode 1 and by removing the statement that permits relaxing integrity pursuant to NRC approved procedures.

l Provisions for procedural controls over open containment penetrations has been incorporated into the definition of Containment Integrity in Specification 1.7 and is therefore no longer required in this section.

Following Mede 1 defueling, the maxirum radionuclide release postulated for the worst case accident, a fire inside containment, is less than the 10 CFR 50 Appendix ! numerical guidelines for normal releases. Thus, tne requirement for double barrier isclation of centainment penetrations no longer exists.

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The licensee has proposed an additional soecificatien, Section 3.6.1.1.b, applicable during Mode 1 only, that would allow mocifications to containment penetrations provided that a closed single barrier is maintained. If no barrier exists, the Action Statement would require terminatirn of certain activities inside containment that could pose a thrtat of radiological releases.

Allowing penetration mcdification activities pFovided a single isciation barrier is maintained is consistent with the existing Action Statemert which requires closing a penetration with one manual valve, blind flange, or one deactivated autcmatic valve when one of the double barriers is incapable of closing. The sti'f finds the proposeo changes acceptable, Section 3.6.1.2 The licensee has proposed a new specification requiring maintaining Containment Isolation during Modes 2 and 3.

This proposed specification provides for an acceptable level of containment performance to assure a single barrier agairst the spread of the remaining radioactive contanination after Mode 1 defueling is complete.

Section 3.6.1.3 Technical Specification 3.6.1.3 requires that the containment airlocks be operable with both doors closed except during transit through the airlock when only one door must be closed. Both dcors may be open if allowed per procedures approved by the NRC staff. This specification is applicable during the Recovery Mode.

The licensee has proposed changes to this specification that would rake it applicable during Mode 1 only and would refer to Recovery Operations Plan section 4.6.1.3.1 for the criteria for allowing both doors to be open.

The change of applicability to Mode 1 is consistent with the change of applicability of Containment Integrity specifications to Mode 1 as previously discussed. The procedural requirements previously imposed by the NRC staff prior to allowing opening of both airlock doors are fully incorporated into the Recovery Operations Plan and are thus still applicable. NRC approval of the procedure would no longer be required.

. The staff finds the proposed changes acceptable, l

Secticn 3.6.1.6 ihe licensee has proposed a new technical specification 3.6.1.6 that wculd require the containment airlocks to be operable with at least one door closed unless the criteria of the Recovery Operations Plan l Section 4.6.1.6.1 are met. This specification would be applicable in Medes 2 and 3.

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e. The staff has concluded that this proposed specification is acceptable and consistent with previcusly discussed changes tc the Containment integrity and Containment Isolation tecnnical specifications.

'Section 3.6.1.4 Technical Specification 3.6.1.4 prescribes limits for prirary containment pressure. It is applicable during the Recovery Mode.

The licensee proposes to change the applicability to Modes 1, 2 and 3.

This change is acceptable and consistent with the licensee's definitien of Facility Modes as previously discussed. It is administrative in nature since Modes 1, 2 and 3 encompass all facility conditions previcusly included in the Recovery Mode.

"action 3.6.1.5 ._,

Technical Specification 3.6.1.5 prescribes the limits within which primary containment air temperatures must be maintained. It is applicable during the Recovery Mode.

The licensee proposes to change the applicability of this specificatien to Mode 1.

The basis for the primary containment air temperature limits is to assure r6 actor coolant system temperature remains high enough to prevent precipitatien of the soluble boron used for criticality

, control. In addition, the temperature band was established to assure maximum service life of equipment and instrumentation installed in the reactor building. Following completion of defueling and transition to Modes 2 and 3, the need for criticality control and for operability of most of the equipment in the reactor building will no longer exist.

Thus, there will be no need to impose these temperature restrictions after Modt 1. The staff finds the proposed change acceptable.

Section 3.6.3.1 Technical Specification 3.6.3.1 requires that one train of the Containment Purge Exhaust system be operable during purge operatiens.

When no trains are operable, purge operations are prohibited, and one train must be restored to operable status within 7 days.

The licensee proposes to change he applicability of the specification to Modes 1, 2, 3; and to change the 7 day time clock of the action statement to require restoration of an operable train prior to resuming ourge operations.

,* The revision of the acclicability is consistent with the licensee's proposed definitions of Facility Modes. It will still reauire operability during all purge operations during all three modes.

Revisino the 7 day time clock to "prior to resuming ourge ocerations" alicws the licensee to aopropriately prioritize facility activities and to best utilize available manpower while still assuring that no purge operations will be performeo without an available purge exhaust train.

The staff finds the proposed changes acceptable.

Secticn 3.7.6.1 Technical Specification 3.7.6.1 specifies the required Flood Protecticn measures and is applicable "at all times".

The licensee proposes to change the applicability to Modes 1, ? and 3.

This change is acministrative in '1ature since the defined Modes 1, 2 and 3, 3re the same as "at all times" for the duration of the existing cleanuo program.

Section 3.7.7.1 Technical Specification 3.7.7.1 requires operability of the Control Room Ventilation and Emergency Air Cleanup System during the Recovery Mode.

The licensee proposes to change the applicability of the specification to Mode 1. The Control Room Emergency Air Cleanup System is required to be maintained ccerable to orotect control room operators in the event of an accident and to maintain centrol room habitability in the event of chemical releases. Once Mode 1 defueling is completed there will be ro recuirement to man the control room (see proposed changes to Section 6.2.21, censequently maintenance of control room habitability is not required. The staff finds the proposed changes acceptable.

Section 3.7.9.1

'achnical Specification 3.7.9.1 prescribes limits for removable contamination on sealed sources. It is applicable during the Recovery Mode.

The licensee proposes to change the applicability to Modes 1, 2 anc 3.

This is an administrative change since Wodes 1, 2 and 3 collectively include all facility conditions previously included in the Recovery Mode. The staff finds the oropcsed change acceptable, i

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Sections 3.7.10.1. 3.7.10.2. 3.7.10.4, and 3.7.11 Technicar Specifications 3.7.10.1, 3.7.10.2, 3.7.10.4 and 3.7.1' orescribe the operability requirements and action statements 'or the plant fire protection systems. These specifications are aoplicable during-the Recovery Mode.

The licensee proposes to change the applicability of these specifications to Modes 1, 2 and 3. In addition, 3.7.10.2 will be modified to reouire operability of certain portions of the systen cnly if the asscciated charccal filters are installed.

The change of applicability is administrative in nature as previously discussed. In addition, the designated portions of the system crovide orotection due to the potential for a fire in the charcoal filters.

When the filters have been removed, no fire hazard exists and the coerable sprinkler /deluoe systems are no longer needed. The staff finds the oroposed changes acceptable.

Section 3.7.10.3 Technical Specification 3.7.10.3 prescribes operability requirements for Halon fire suppression systems in the Cable and Transformer Rocm and the Air Intake Tunnel. The specification is applicable during the Pecoverv Mode.

The licensee proposes to change the applicability to Mode 1. The wa lon System protects circuits and equipment required for safe shutdown and core protection in specific areas of the plant from the propagation of a fire. Once Mode 1 defueling is completed there will be no circuits or equipment necessary for the protection of the core. The staff finds the proposed change acceptable.

Sections 3.8.1 and 3.8.2 Technical Specifications 3.8.1 and 3.8.2 prescribe the operability requirement for A.C. and 0.C. electrical distribution systems. The specifications are applicable in the Recovery Mode.

The licensee has proposed changing the applicability of these operability requirements to Mode 1 and has proposed administrative changes involving renumbering of paragraphs.

The purpose of these specifications is to assure that the power sources and associated distribution systems are available to supply the safety related eouiement eequired to maintain the unit in a stable condition.

Once Mode 1 defueling is completed no safety related equipment will be required to maintain the unit in a safe and stable condition, consecuently, the ocwer sources and distribution systems would not be recuired. The staff finds the proposed changes acceptable.

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Sections 3.9.1, 3.9.2. 3.9.3, and 3.9.4 Technicat Specifications 3.9.1, 3.9.2, 3.9.3, and 3.9.4 orescribe recuirerents for maintaining and menitoring water levels in me Scent Fuel Storage Pool 'A' and the Fuel Transfer Canal. The rec. ements of 3.9.1 and 3.9.2 are applicable whenever mere are canisters containing core material in the Spent Fuel Storage Pool 'A'. The recuirement of 3.9.3 and 3.9.4 are applicable whenever canisters containing core material and/or the plenum are stored in the Fuel- -

Transfer Canal.

The licensee proposes to establish the applicability of scecifications

,'t 3.9.1 and 3.9.2 during Modes 1 and 2 when canisters containing core material are stored in the ecol.

This change is administrative since it is only during Modes 1 and 2 that canisters containing core material C&Y be stored in the Spent Fuel -

Storage Pool 'A'.

The licensee proposes to change the applicability of specifications 3.9.3 and 3.9.4 to Mode 1.

The basis for water level control in the Fuel Transfer Canal is radiation erotection for personnel when there are fuel canisters and/or the plenum assembly stored in the Fuel Transfer Canal. During Modes 2 and 3 there will be no canisters stored in the Fuel Transfer Canal.

There' ore, water level control will not be required for purposes of personnel radiation protection from canisters. During Modes 2 and 3 the plenum may be stored in the Fuel Transfer Canal; however, there will be a significant reduction of cleanup activity in the building.

Therefore, with fewer people in the RB requiring radiation rotection and no canisters stored in the Fuel Transfer Canal, mainterince of water in the canal will no longer be required. The staff finds the procosed changes acceptable.

-Sections 3.9.12.1, and 3.9. M Technical Specifications 3.9.12.' and 3.9.12.2 require operability of the Fuel Handling Building Air Cleanup Exhaust System and the Auxiliary Building Air Cleanup Exhaust Systems respectively. The scecifications i

are applicable in the Recovery Mode and require suspension of ccerations involving movement of liquid and gaseous radioactive wastes as the respective areas of the system become inoperable.

The licensee proposes to modify these specifications by changing the

! acolicability to Modes 1, 2 and 3. In addition, the action statement l

will be changed to require suspension of crerations involving movement of liquid and solid radioactive wastes the release of which eculd exceed 50 cercent of the Apoendix B Technical Specification instantaneous release rate for gaseous effluents. The change of applicability is i administrative in nature and consistent with the orevious discussions of L

, . ;4 the newly defined facility modes. Deletion of gaseous radioactive wastes from the Action' Statement is appropriate since the gaseous radioactive source term has been essentially eliminated as a resalt of.the orogress of the defueling and decontamination crogram. The ramaining potential source terms are the result of still existing licuid and solid radioactive waste material and. contamination with!n the facility. The quantitative restrictions in the proposed Action Statement will ensure that suspension of radioactive material from any coeration will not result in releases in excess of these allcwed by 10 CFR 50 Appendix I. The staff finds the proposed changes acceptable.

Section 3.9.13 Technical Specification 3.9.13 prohibits discharge of Accident Generated Water until approved by the NRC and requires that its discharge, once approved, be in accordance with crocedures approved pursuant to specification 6.8.7. It is applicable during the Recovery -

Mode.

The licensee proposes administrative changes to this specification that would make it applicable in Modes 1, 2 and 3. The proposed change is administrative only and does not change any presently imposed license restrictions. The staff finds the proposed change acceptable.

Section 3.10.1 Technical Specification 3.10.1 imposes restrictions on the handling of heavy leads in the containment building during the Recovery Mode.

The licensee proposes to modify the applicability of this specification

' to Moce 1.

The purpose of this specification was to limit load handling to preclude heavy load drops that might result in a core reconfiguration or a less of reactor vessel integrity. Once Mode 1 defueling is completed, the basis for controlling heavy loads inside the containment will be eliminated and the specification will not be required. The staff finds the proposed change acceptable.

Secticn 3.10.2 Technical Specification 3.10.2 imposes restrictions on the handling of heavy loads in the Fuel Handling Building during the Recovery Mode.

The licensee proposes to modify the applicability of this specification to Modes 1 and ?. The basis for this specification is to prevent a load drop in the Fuel Handling Building causing damace to canisters containing core material. Subsequent to Modes 1 and 2 all core material will have been shipped off-site. Thus, the basis for controlling neavy loads inside the Fuel Handling Building is stiminated and the scecification is not recuired. The staff finds the proposed chance acceptable.

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1. l Section 6.2.2 Technical' Specification 6.2.2 in part, lists the recuired minimum shift crew composition, requires a licensed operator in the control reem when fuel is in the reactor, and requires an individual cualified in radiation protection procecures on site when fuel is in the reacter.

The licensee proposes to modify these requirements by specifying that they apply only during Mede 1.

The requirements for licensed operators are specified in 10 CFR 50.Sa for fueled nuclear power plants. Upon transition from Mode 1, the facility will no longer be considered fueled and these requirements will no longer apply. The staff finds the proposed change acceptable.

Section 6.8.2.2 Technical Specification 6.8.2.2 specifies the scope of licensee precedures that require NRC approval prior to implementation.

The licensee proposss to delete the recuirement for NRC approval of procedures excep'c for those involving disposal of Accident Generated Water.

Since the accident, the licensee has implemented a number of major management and organization changes designed to more effectively manage the unicus challenge of the post-accident cleanuo. These changes have resulted in an organization structure which places an acceptable emphasis on the safe conduct of cleanuo activities with adecuate provisions for management review and oversite of facility activities.

The staff's ongoing assessment of the licensee's procedure development and review program indicates that it is working effectively to assure implementation of the Technical Specifications and compliance with regulatory requirements. Therefore, the staff concludes that the cleanup activities at the Three Mile Island Unit 2 facility no longer recuires the unique administrative controls that have been recuired in the past. The staff will, however, require the licensee to submit for approval procedures involving the disposal of the accident generated water.

Section 6.11 Technical Specification 6.11 requires that personnel radiation erotection be consistent with the requirements of 10 CFR 20 and the NPC approved Radiation Protection Plan.

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The licensee proposes to modify this specification by deleting the recuirement for NRC approval of the Cadiation Protection Plan. Removal of this requirement is based on the acgeotable past performance of the licensee'in the area of radiation protection. It is also consistent with the Standard Technical Scecifications for Babcock and Wilcox plants.

Auditing by the NRC of the Radiation Protection Plan and licensee compliarce with the plan would continue consistent with practice at the TMI site for the operating unit, TMI-1, 3.0 CONTACT WITH STATE OFFICIALS, On March 31, 1988, the NRC staf' contacted by telephone, the Bureau of Radiation Protection, Department of Environmental Resources, Comenwealth of Pennsylvania, on the proposed determination of no significant ha: arcs consideration. No objection to the proposed action was voiced.

4.0 ENVIRONMENTAL CONSIDERATION

This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined -

in 10 CFR Part 20. We have determined that the amendment involves no significant increase in the amounts, and no significant chance in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly,-this amerdment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c H9). Pursuant to 10 CFR 51.22(b), no environmental inpact statement or environmental assessment need be prepared in ccnnection with the issuance of this amendment.

5.0 CONCLUSION

We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Comissien's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.

Dated: MAY 2 71988 Principal Contributor: John A. Thomas