ML20079M791

Safety Evaluation and Environmental Assessment,Three Mile Island Nuclear Station,Unit No. 2.Docket No. 50-320. (Metropolitan Edison Company,Jersey Central Power and Light Company and Pennsylvania Electric Company)

ML20079M791
Person / Time
Site:	Crane
Issue date:	02/28/1980
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0647, NUREG-647, NUDOCS 8401270427
Download: ML20079M791 (24)

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NUREG-0647 Safety Evaluation and Environmental Assessment Metropolitan Edison Company Jersey Central Power and Light Company Pennsylvania Electr:c Company Docket No. 50-320 Three Mile Island Nuclear Station, Unit No. 2 By the Office of Nuclear Reactor Regulation U.s. Nuclear Regulatory Commission February 1980 I I 170 %5078fiTI GI] IUh p-PDH NURE p-V

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8401270427 800228 PDR ADOCK 05000320 E

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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

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SAFETY EVALUATION AND ENVIRONMENTAL ASSESSMENT BY THE OFFICE OF NUCL' EAR REACTOR REGULATION METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCXET NO. 50-320 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 2 a

INTRODUCTION On March 28, 1979 an accident at the Three Mile Island Nuclear Station Unit

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2 resulted in substantial damage to the reactor core and to certain reactor systems and ccmponents. The facility is not capable of operation and is in a shutdown condition with damaged fuel in the core. Although some systems were damaged or have subsequently failed, the facility is being maintained in a safe and stable cooling condition utilizing a substantial number of systenis an'd components., ' ~Some of the ' systems' an'd ' components currently beitig used to maintain the facility in its present mode.of operation were not origin'aily in'cluded ir.' the faci 1ity'5 Nchnical spec'ifications because these systems were not required for safe operation of the facility under l

pre-accident conditions.

t Since these additional systems and components are now being used to remove decay heat from the core, revised technical specifications to encompass the additional systems and components should be included in the facility license.

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and other technical specifications for equipment not required during the j'

present mode of operatipn should be deleted.

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The systems and comoonents available to provide plant safety, including icng 4

term cooling of the core, under the present conditiens with the facility in, cold shutdown and while cleanup and recovery of the facility proceed, have been reviewed. The reactor is presently being maintained in a stable, long tenn cooling mode with decay heat'being removed by natural convection circulation of primary coolant through the core with heat rejection through the "A" steam generator. The "A" steam generator is producir1g steam which is condensed in the condenser and recirculated to the "A" steam generator. An alternate means of removing decay heat frcm the primary coolant is through the "B" steam

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generator. The steam side of the "B" steam generator has been modifin:d to provide a water solid, closed loop cooling system which is in turn cooled by the secondary services closed cooling water system.

Either steam generator 1

cooling mode is adequate to remove decay heat from the primary coolant.

If

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natural circulation cooling of the core should be lost, contingency plans and procedures have been prepared and approved for alternate means.of providing

'lon'g term core cooling. These' alternate core cooling means iriclude forced '

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circulation of the primary cooling using the reactor coolant pumps or. decay heat remova'l pumps.' OperatiorY of va'rious systems to control the release of radioactive materials will also be required during the cleanup of radioactive

' materials'. released within the facility and the recove.ry of the facility from the effects of the accident. Appropriate Appendix A Technical Specifications governing this period (long tenn cooling of the core and during cleanup and recovery of the facility) have been established through conferences between the staff and the licensee.

This safety evaluation describes the protection y

required to provide adequate safety during present conditions.- It does not j

authorize removal of fuel from the reactor pressure vessel.

Such authorization l

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must be cbcained prior to any such removal and will be accressed in a subsequent safety evaluation.

This amendment does not include any changes in Appendix B (which remains in effect and includes effluent release limits) to the facility operating license, except that Appendix B Technical Specifications 5.1, 5.2 and 5.3, which identify the licensee's pre-accident management organization for activities addressed by the Appendix B Technical Specif.ications, are deleted since those requirements will new be performed in accordance with proposed Technical Specifications 6.1, 6.2 and 6.5 'which will set forth the current requirements for the licensee's management organization for all licensed activities. This evaluation does not encompass operation of the EPICOR-II decontamination system currently being utilized at the facility pursuant to the tems of the Comission's Memorandum and Order of October 16,*

1979 to process decontaminated intermediate-level radioactive waste water accumulatedin'theauxiliary'b'uilding.'TheIm~pactofusingEPICOR-II'was

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evaluated in an Environmental Assessment (NUREG-0591) prepared by this Office on Octobdr 3,1979'.' 'See ajso Drde'r'for Modification'oflicense and Negative Declaration issued by the Director of this Office on October 18 1979. The Comission's decision of October 16, 1979 does not address the s...

I subject of disposal of the decontaminated water processed by EPICOR-II.

Pursuant to the Comission's Statement of May,25,1979, discharge of EPICOR-II processed waste water is not permitted until completion of an environmental review of such discharges.

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Curing the prccess of preparing revised technical specificaticns, a new coerational mcde was defined.

This new operational mode (designated the

" Recovery Mode" and defined in Technical Specification 1.3) is 1ntended to apply throughout the long-tem ecoling of.the core and facility cleanup and recovery operations.

This change in mode applicabil'.ty is reflected in the revised technical specifications. This amend:r nt deletes other operating modes and thereby precludes operation in other than the shutdown conditions defined for the Recovery Mode.

The March 28, 1979 accident resulted in excessively high radiation areas in certain portions of the facility; therefore, provisions have been included in the surveillance requirements for the revised technical specifications

.j which relieve the licensee from the requirement to perfom certain surveillance requirements when access to the equipment would result in excessive' occupational exposures.

It is expected that the areas in which this relief is necessary will.. be reduced as cleanup.of the,faciJity progresses.

. Minor changes have been made in Technical Specifications.3.3.3.1, 3.3.3.3, 3.3.3.4, 3.3 f 3.5 7 5.3.'.6', 3.3.3.7', 3.3.3.$,1.'4.3, 3.6.1.3, 3.6.1.4, 3.6.1.5, 3

3.6.4.1, 3.7.3.1, 3.7.3.2, 3.7.4.1, 3.7.6.1, 3.7.7.1, 3.7.10.1, 3.7.10.2,

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3.7.10.3, 3.7.10.4, 3.7.11 and 3.8.2.3.

These minor changes consist of changes l

in applicability requirements, changes to existing action statements which require reactor shutdown or prohibit plant startup with inoperable equipment, and c'eletion of operability requirements for equipment which has failed and cannot be repaired or equipment which is not required in the plant's present condition.

These changes do not significantly increase the probability or consequences of an accident or significantly decrease a safety margin and, 1

in fact, are of no safety significance.

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i The following Technical Specifications are being deleted since they are applicaole only during operation in Modes 1, 2, 3, 4 and 6:

2.1.1, 2.1.2, 2.2.1, 3.0.4, 4.0.4,i 4.0. 5, 3.1.1.3, 3.1.1.4, 3.1.2.1 - 3.1.2.9, 3.1.3.2 - 3.1.3.9,. 3.2.1 - 3.2. 5, 3.3.3.2, 3.4.2, 3.4.4 - 3.4.8, 3.4.9.2, 3.4. 10.1, 3.5 3.5.4, 3.6.1.2, 3.6.1.6, 3.'S.1.7, 3.5.2.1 - 3.6.2.3, 3.6.3.1, 3.6.4.2, 3.6.4.4, 3.6.5, 3.7.1.2 - 3.7.1.6, 3.7.2.1, 3.7.5.1, 3.7.8.1, 3.7.9.1, 3.8.1.2, 3.8.2.2, 3.8.2.4, 3.9.1 - 3.9.11 and 3.10.1 - 3.10.4.

Operation in Modes 1, 2, 3, 4 and 6 is no lon.ger autnor: red; deletion of these Technical Specifications, therefore, does not significantly

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increase the probability or consequences of an accident or significantly decrease a safety margin. Therefore, these deletions do not involve a significant hazards consideration and in fact are'of no safety significance.

.. x E'/ALUATION 1,.

Nuclear Safety The full length control rods (safety, and regulating) were fully inserted into the, core during.the reactor. trip which occurred at.the beginning of the March 28, 1979 accident.

To provide assurance that control rod motion will not cause a change in core reactivity; Technical Specification 3.1.3.1 requires 'that the' cont'rdi rod'drhe b'reakers b'e maintained oped.

Sini:e the integrity of the control rods and the fuel rods is uninown, the staff has performed analyses which show that with a reactor coolant boron concentra-tion of about 3000 ppm, the core will be maintained subcritical in all possible configurations (Reference 1).

Consequently, revised Technical Specifications j

3.1.1.1 and 3.1.1.2 have'been prepared requiring two operable systems for injecting borated cooling water into,the reactor coolant system and requiring L-

6-the reactor coolant baron concentration th be maintair.ed between 200i ace 4500 ppm.

The maximum bcron concentration has been specified to assure that baron precipitation will not oc:ur.

A concentration of 4j00 ppm boron in water has a precipitation temperaturs of approximately 450F.

Therefore, a requirement has been added to maintain the reactor coolant minimum temperature above 50 F thereby assuring that bcron" precipitation will not 0

occur.

2.

Core Cooline. Water Inventory and Reactor Coclant Systat. Pressure Centrol The core is presently being maintained in a stable cold shutdown condition and is being cooled by the reactor coolant systam operating i

in natural circulation. Heat removal from the reactor coolant system y

is 'through the "A" steam generator which is producing steam. The steam is being routed to the condenser where it is being condensed and then recirculated to the "A" steam ' generator.

An alternate means of removing decay. heat from the primary coolant is available through the S" steam gene'rator.

The'stecm side of the S" steam generator has 'been moBified to provide a water solid, closed loop cooling system..-

which is in turned cooled by the secondary services closed cooling water system.(Reference 2).

Operability of the steam generators and associated

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cooling water system is required by Technical Specifications 3.7.1 and 3.7.2.1.

Either steam generator cooling mode is adequate to remove the decky heat from the prin:ary coolant (Reference 1).

Technical Specification 3.4.1 requires that the recctor coolant pumps be maintained operable for e

i possible forced circulation of reactor coolant in the event forced cir-

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culation cooling is required.

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f, standby react:r coolant system pressure control systa.a has been adced to the facility to maintain the reactor coolant system level and pressure for normal operation in the " Recovery Mode" and over a wide range of anticipated transient events which would cause shrinkage of the reactor coolant (R.eference 2).

These anticipated transients include loss of natural circulation :coling due to a loss of all secondary side cooling i

with ' restart of one secondary cooling loop following a hot leg temperature 1

rise of 500F.

More severe transients which this system is not designed to accomodate would be handled by the high pressure injection pumps, the operability of which is required by Technical Specification 3.1.1.1.

Appropriate surveillance requirements which demonstrate the operability

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of these systems have been incorporated.

The operability of barated.

water rources which are sufficient to acccmmodate all possible transients-is assured by. appropriate surveillance requirements.

Technical. Specification 3.4'.9.1 has been' modified to restrict th'e reactor',

coolant system temperature'and, p'ressure to 2800F ami #20 psig.

This provides assurance that the reactor pressure vessel wC) not be subjected to conditions which could result in its. brittle fracture.

3.

Instrumentation Since the reactor will not be operated duri'ng this time period, the only portions of, the reactor protection instrumentation required to be maintained in an operable condition are the source range and intennediate range neutron monitoring channels.

Although the reactor will be e

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i maintained suberitical via baron in the reactor c:ciant (Reference 1),

tnese instruments,ars required to be maintained in an operable concition.

per Technical Specification 3.3.1.1 to provide the capability for monitoring the aeutron level in the core.

l The only Engineered Safety Feature Actuation System (ESFAS) instrumentation required to be maintained operable during this period is that provided to start the Class-IE diesel generators upon detection.of a foss of offsite electrical power.

This instrumentation is required operable per Technical Specification 3.3.2.1.

Other ESFAS instrumentation is not required due to the low decay heat loads and the ample time available for manual initiation of systems available to acccmmodate possible transients.

This is acceptable based upon the present plant conditions (Reference 2).

Since the reactor coolant system pressure instrumenta. tion, reactor

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building water level instrumentation and the incere thermocouples are bei.ng used.to assur,e cor.e coo,li,ng.an,d ts provide assurance that vital equipment in the containment is not flooded, their operability is required and operability requirements for this instrumentation have been added to-Technical Specification '3.3.3.6i -

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4 Containment Systems L

Significant quantities of radioactive materials have been released into t

the containment.

Containment integrity is required to.be maintained

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by Technical Specification 3.5.1.1 to ensure that these materials are

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not inacvertently released to tne envirens.

This Technical 5;:eef-fientien also prohibits venting or purging or other treatment of the react:r building atmosphere until such activity has been accreved by the.'tRC.

Since the, licensee has pre:csed that the containment atmosphere be removedTy purging through the hydrogen purge cleanup system (?.eference 3), Technical Specificat 3.5.a.3 is being retained to ensure the coerability of this system in the evenc purging of the centainment' is approved and authorized 5.

Fire Detection and Fire Sucaressien As part of the facility modifications made for long term c:oling o the core, additional fire detection instrumentation and deluge / s systems.were installed.

These additions included fire detection instrumentation to protect the self-contained skid mounted " Gre and " White" Balance of Plant (BOP) diesel generators and a delu

'sprink1' r 'systeni to protect the auxiliary building exhaust filter.

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' Operability requirements for this added equ'ipmerit ha

' into Tehhrifcal' Sp~ e'c'ificat' ions 3.3.3.8 and 3.7.10.2.

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The operability of these fire suppression systems ensures that adequate fire su capability is available to confina.and extinguish. fires.

De s'urve11 lance requirements provide assurance that the minimum operab.ility req of the fire ' uppressten systems are met..

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U s::. t:1' M..er The electrical energy to opera:e the sys ams seing usec o remove cacay heat frem the core is provided by redundant circuits from t.ke.offsite transmission ne:wcrk and by onsite power su; plies.

The presen: c:ciing mode requires the use of electrical power to operate equipment which oreviously did not require protection against loss-of-of' site ;cwer.

Therefore, an additional 13.2 kv circuit frcm the 'Uddlet:wn' Juncticn Substation and two redundant balance cf plant diesel generators have been installed to increase the reliability of the o'ffsite and onsite electrical pcwer supplies (P.eference 2).

The new 13.2 kv circuit prcvides a backup offsite electrical power supply for two circulating water pumps (one of these pumps provides adequate cooling for removing decay heat).

In the event of a total loss of offsite power the core can be cooled using only the onsite diesel generators as a power supply (Reference 2).

The redundant self-contai.ned skid, mounted " Gray" and " White" diesel generators.have been installed to provide backup protection to all electrical.. loads which are requ' red for core cooling and which were i

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.not previously protected against loss-of-offsite pcwer. Therefore, Technical Specification 3.8.1.l'has been modified to require the operability.

o'f the backup 13.2' ky' circuit und the two additional, redundant, balance of plant (" Gray" and " White") diesel generators.

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7 Cen:rol of Ract: active Materials in Casecus Effluents The auxil.iary building air. cleanup system has been installed to fil:er gaseous effluents frem the auxiliary building.

Operation cf this systam in conjunction with the fuel handling building air cleanup ' system, ensures that any radioactive materials in effluents frem these buildings will be processed through HE:A filters prior to release to the environs. The c:erability requirements for the auxiliary building air cleanup sys:am have been added to Technical Specification 3.9.12 which previously contained the operability requirements for only the fuel handling building air cleanup system.

The Surveillance Requirements for the charcoal adsorbers in the fuel handling building air cleanup

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system have been deleted shnce.the radioactive iodine is no longer present; it has decayed away.

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Control of Radioactive Materials. in L'iouid Effluents The discharge of, water' processed by the EPICOR-II system and the

' processing and discharge of highly contaminated water contained i'n.

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the Reactor Building. sump and Reactor Coolant System is prohibited in accordance with the Commission's Statement of May 25, 1979 pending

, evaluation 'off these actions.

Furthermore, on November 21, 1979, the Commission announced its decision to prepare a programatic environmental inpact statement to address, among other things, the decontamination and disposal of radioactive waste water resulting from the accident.

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Of F:licy and Notice of Intant to Fracare a Fr:grs=atic nyir:nmental Im:act Statement",(ad F.?.

57738).

The Cermissien ecserved that as the cec:ntamination of TMI-2 progresses the Comission will make available 9

any new inf:rmation to the public and to the extant necessary will also prepara separate environmental statements or as essments for individual

ortions of the overall cleanup effort.

The Comission also indicated that in the event it.should decide before c:mpletien of its programatic sta ament that it is in the best interest of the public health and safety to dec:ntaminate the high-level waste water new in the containri.ent building

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or to purge the building of its radioactive gases, such action wculd not be taken until it had undergone an environmental review consistent with a

i its May 25, 1979 Statement.

The Comission has further recogni:ed, however,

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the possibility-that an emergency sit 0ation, now unforeseen, may arise which could require rapid action.

Accordingly, Technical Specifications 3.9.13 and 3.9.1'4.have been added.to..

implement' thes'a requirements.

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Review and Audit Functions The accident of March 28,1979 has resulted in the generation of large l

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. ua,n,tities.,9,f radioactive, pastes. Therefore, the licensee has augmented ',, ;

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i the membership of his Plant Operations Review Comnittee.and Generation Review Committee to provide additional expertise in the area of radioactive waste management. We have added requirements to sections s

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6.5.1 and 6.5.2 in the Technical Specifications to implapent these l

additional functions. We consider the addition of this expertise in oJ radioactive wasta management to these comittees to be appropriate since l

the licensee will.be handling and processing significant quantities of

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'1 ridicactive wastes.

These committees will assure : hat sucn activi:ies are procer'ly reviewed and controlled by licensee personnel with appropriate and adequate expertise.

In addition,' Appendix B Technical Specifications 5.1, 5.2 and 5.b, which identify the licensee's pre-accident management orgaai:ation for activities addressed by the Appendix B Technical Specifi. cations

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(wnica were imposed for' the protection of the environment) are deleted since those requirements will now be perfor ned in accordance with procesad Technical Specifications 6.1, 6.2 and 6.5 which will set forth the current requirements for the licensee's..anagement organization for all licensed activities. The deletion of these Appendix 3

...d Technical Specifications does not affect any existing limits on effluent releases and discharges and does not authorize a change in effluent types or amounts nor does it affect the power level of the

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facility. Furthermore, their. deletion would not. result in any

. increase in the probability or Ionsequences of an accident nor wIll it result in a decrease in 'a margin of safety since the requirements will in any event be continued in an updated requirement of proposed

. Technical Specifications 6.1, 6.2 and 6,.5 which reflects the curre,nt,.

post-accSen't' requiremand f'or t'he facilit'f's inaintenance'

Thus, deletion of Appendix 3 Technical. Specifications 5.1, 5.2 and 5.3 will have no environmental impact or effect on plant' safety, and is purely administrative in nature.

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Su=ary The technical specification changes associated with this amendment reflect the changas that are necessary to account for the present condition of the facility and to assure the continued maintenance of the safe, stable condition of the facility in the " Recovery Mode".

Certain additional controls and equipment requirements, not required'in 'the pre-accident technical specifications, have been added to provide additional assurance that the facility will be maintained in a' safe and stable cold shutdown condition during the present and planned

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accivities for facility recovery frcm the accident. The technical specifications associated with this amendment include these adced centrols and equipment requirements.

Except as necessitated by the physicil realities that exist due to

. damage caused by or as a result of the accident,. no safety limit, limiting condifion for operation or surveillance' requirement in the pre-accident technical. specifications that is pertinent to the present cold shutdown candition of the facility has been modified, relaxed, or deleted by-this amendment.

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Although the facility is presently being maintained in accordance with NRC approved procedures, the present plant conditions were not expressly contemplated nor provided for in the facility operating J

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license, consequently, the current facility c:erating license does not include any orovisiens or tachnical specificaticns for assuring

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he continued maintenance of the plant in a safe, stable condition or for providing for foreseeable off-normal conditions.

These revisad technical specifications explicitly impose such license requirements and thereby provide an increased assurance of plant safety.

In addition, by deletten of coerating..cdes other than the.'tecovery Mode, and by the changes to existing Technical Soecificatiens discussed herein, planned o;eration of the faciit:y in other :han the stable shu::own conditien cf the Recovery Mode is precluded.

Based on the aYeve, the public health, safety and interest required that the requirements imposed

'by the proposed Technical Specifications set forth in Attachment 1

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to the Director's Order of this date beccme effective inmediately.

ENVIRONMENTAL ASSESSMENT The environmental impacts resulting from nomal op'eration of the facility were evaluated by the Staff as set forth in the Final Environmental State-ment issued in December'1972 and in the Final Susiplement to the Final Environmental Statement issued in December 1976. Although the licensee's-authority to operate the facili.ty was suspended by Order for Modification of License dated July 20,1979, and is now limited to maintenance of the.

reactor in its current mode, the limits on effluent releases and discharges

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previously established are not changed by virtue of revised and/or new Technical Specifications being' imposed, nor do they authorize a change in

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effluent types or total amounts nor an increase in power level.

Thus, any environmental impacts which are attributable to maintenance of the facility

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in acc:rdance with the revised'and/or new Technical 5:ecifications will be within,'indeed likely substantially less thanlthe impacts previously evaluated and found acceptable.

Furthermore, those licanse conditions and Technical Scecifications (Appendix B) imposed for the protection of the environment upon issuance of the TbI-2 operating, license are not being relaxed in any way by these Technical Specifications.

The eight areas affected by the revised and/or new Technical Specifications:

.luclear Safety; Core Cooling, Water Inventory and Reactor Coolant System Pressure Control; Instr.: mentation; Containment Systeas; Fire Detection and Fire Sappressici ; Electrical Power; Control of Radioactive Materials in '

W Liquid and Gaseous Effluents; and Review and Audit Functions have been revised frem the standpoint of safety considerations, as discussed above.

From the environmental standpoint, no reasonable or meaningful alternatives to the provisiens of the Technical Specifications have been identified.

liowever, the staff is including. Technical Specifications which specifically proh'i, bit certa'in act.ivities 'which would otherwise~ be ' authorized at a normally operating facility.

In particular, the Technical Specifications include prohibitions against the purging or other treatment of the reactor building atmosphem,.the dischirge or other'disposaf pf water decontaminated by the EPICOR-II system and the treatment and discha'rge or other disposal of the high-level radioactively contaminated water now in the reactor building, even though such activities might be conducted in full compliance with

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effluent limitations or Comission regulations currently in effect and applicable to TMI-2.

It is possible, as an ' alternative, that these activities could have been allowed under the same effluent limitations as would apply

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,g in the case ;f a ner-ally operating facility.

Hcwever, the C:missicn has

. determined that the public interest warrants prchibiting these uncertakings '

pending ccmpletion of an environmental review.

See Comissien's Statament of May 25,1979 respecting decontamination of radicactively contaminated water, and Ccmission's Statement of PoIfcy and Notice of Intent to Prepara a Programatic Eny'ironmental Impact Statement, dated November 21,1979 (4 F.R. 67738 A variety of longer range alternatives associated with the overall decontamination t

and cleanup of the facility will be addressed in the programmatic environmental impact statement.

The Technical Scacifications do not authorize any new releases external to the facility.

Consequently, no off-site environmental impacts are anticipiated.

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Onsite maintenance of the facility pursuant to these Technical Speciff-cations similarly does not entail any new releases of effluents nor the exposure of any workers to a radiological environment except as previously evaluated..and.found acceptable, and, as a result, no. change in on-site,

impacts will result.

' ' NO t'h's i'irego'ing ieasod,'It has bee'n di.'teNiined Shat this' actic$ % '

insignificant frem the standpoint of environmental impact and that an

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environmental statement need not be prepared.. Accordingly, pursuant to 10 CFR $51.5(c)(1), a negative declaration will be issued.

i CONCt.USION

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The changes in technical specifications authorized in connection with this

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evaluation result in enhancement of safety under present conditions, as i

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discussed above.

Based :n nese c:nsiderations, we have concludec that:

(1) :nere is reasonable assurance tha the health and safety of the puolic will not be endangsred by cperation 'n the proposed manner, and (2) such i

activities will be conducted in ecmpliance with the C:missien's regulations and the issuance of this amendment will not be inimical to the comon

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defense and security or to t'he health and safety of the public.

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References:

1.

.luREG-0557, " Evaluation of Long-Term Post-Accident Core Cooling of Three Mile Island Unit 2," NRC Staff Report, May 1979.

2.

Memorandum for R. Vollmer from A. Ignatonis, "TMI-2 Plant Macifications for Cold Shutdown, Revision 2," June 8,1979.

3.

Letter to R. Vollmer, NRC, from' R'. C. Arnold, Metropolitan Edison Company,

" Reactor Containment Suilding Atmosphere CleantTp", November 13, 1979. '

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i t 3 TITLE AND SU8TsTLE e RECIPatNI 5 ACCE S$10N NUMSER Safety Evaluation and Environmental Assessment i oA'e aiPOar CO-P't't o Three Mile Island Nuclear Station, Unit No. 2

j. TAR MONY-February 1980 6 AvYMOHisi 7 OATE REPOR T 155uf D MONTM vtAR February 1980 9 PROJECTIT ASE/ WORK uNif Nuwet R 8 Pt R$ORUING ORG ANilaTION NAME AND MAILING ADORE 55 fievde le CodrJ Division of Operating Reactors Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D.C.

20555 11 SPON50R4NG ORG AN#2 ATION N AME AND MAeLaNG ADDRESS ifac8pde l@ CodrA I ?s TYPE OF REPOR T Technical Same as 9. above.

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13 $LPPLEVtNT ARY NOTE S to ABSTR ACT (100 woras or JersJ This report contains an order for the Three Mile Island Nuclear Station, Unit 2, issued by the NRC. The order (1) requires that effective immediately, the' facility be maintained in accordance with the requirements of the attached proposed Technical Specifications and (2) proposes to fomally amend the Facility Operating License to include the proposed Technical Specifications, taking into account the present condition of plant systems, so as to ensure that the unit will remain in a safe posture during the Recovery Mode.

15e KE Y WORDS AND OOCOVE NT ANALv51$

150 Ot SCRIPTOR S 16 AV AIL ARittiv ST Af t ut NT 17 $ECuRIT Y CLA55e# ICA b TON lu NUM8tH OF PAGES "OhTT8ssified Unlimited

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e NUR EG-0647 SAFETY EVALUATION AND ENVIRONMENTAL ASSESSMENT FEBRUARY 1980 FOR THREE MILE ISLAND NUCLEAR STATION, UNIT NO.2 4

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