ML20148S052

Order That 800211 Proposed Tech Specs Re Use of mini-decay Heat Removal Sys Be Modified in Accordance w/801114 Amend of Order.Amend of Order,Safety Evaluation & Tech Specs Encl. W/Certificate of Svc

ML20148S052
Person / Time
Site:	Three Mile Island
Issue date:	01/26/1981
From:	Wolf J Atomic Safety and Licensing Board Panel
To:	Office of Nuclear Reactor Regulation
References
ISSUANCES-OLA, NUDOCS 8102020535
Download: ML20148S052 (5)

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UNITED STATES OF AMERICA \

NUCLEAR REGULATORY COMMISSION 4

/ o@ e ATOMIC SAFETY AND LICENSING BOARD k

Before Administrative Judges: Nc John F. Wol f, Chai rman 4 Mg7 Or. Oscar H. Paris Frederick J. Shon d In the Matter of:

METROPOLITAN EDISON COMPANY, ET AL. Docket No. 50-320-OLA

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(Three Mile Island Nuclear Station, Unit 2) January 26,1981 ORDER On November 14, 1980, the Director, Office of Nuclear Reactor Regulation issued an Amendment of Order providing for the use of the Mini-Decay Heat Removal System (M0 HRS) as a means of cooling the reactor core in lieu of natural convection circulation. Natural convection circulation has been the primary method for maintaining the reactor in a stable shutdown cooling mode since April 1979. It was the method adopted in the Director's February 11, 1980 order for TMI-2.

The NRC Staff in its notice of issuance'of Amendment j of Order and motion to conform stated:

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- " Pursuant to an Order of the Commission dated May 12,- 1980, this Licensing Board has been designated to conduct the proceeding with respect to the proposed formal license amendment incorporating Technical Specifi-cations proposed in the February 11,1980 order. In order ,

to conform the proposed Technical Specifications which are i

the subject matter of the present proceeding, with the actions taken by the Director. The NRC Staff hereby moves that the proposed Technical Specifications be formally >

modified in accordance with the Amendment of Order of November 14, 1980, subject, of course, to the authority which 10 CFR f 2.717(b) explicitly preserves to the ...

[presidingofficer]."

The Licensee through its counsel states:

"The formal license amendment incorporating the proposed Technical Specifications, as amended, would await the outcome of the prospective hearing, with this understanding, Licensee does not oppose the motion."

No opposition to the motion was voiced by any party.

Upon~ an analysis of the effect of the revision of the proposed Technical Specification in accordance with the November 14, 1980 Amendment of Order and' pursuant to 10 CFR S2.717(b) it is this 26th day of January 1981 ORDERED That the Technical Specifications proposed in the I February 11, 1980 order be modified in accordance with the Amendment of Order, dated November 14, 1980 subject to the authority preserved in the presiding officer pursuant to 10 CFR 52.717(b). Attached and made a part here of are Amendment of Order dated November 14, 1980; Safety Evaluation by Office of Nuclear Reactor Regulation, (TMI Nuclear Station, Unit No. 2); Appendix "A" Technical Specifications Facility Operating License No. OPR-73, Docket No. 50-320.

FOR THE ATOMIC SAFETY AND LICENSING BOARD k 4 L<~4 ,

MINISTRATIVE JUDGE /

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

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METROPOLITAN EDISON COMPANY, _et _al. ) Occket No. 50-320 OLA

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(Three Mile Island Nuclear Station, )

Unit 2) )

AMEN 0 MENT OF ORDER

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Metropolitan Edison Company, Jersey Central Power and Lignt Company and

Pennsylvania Electric Company (collectively, the Licensee) are the holders of Facility Operating License No. DPR-73, which had authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts themal. The f acility, which is located in Londonderry Township, Dauphin County, Dennsylvania, is a pressured water reactor pre-viously used for the commercial generation of electricity.

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Sy Order for Modification of License, dated July 20, 1979, tne Licensee's authority to coerate the facility was susoended and the Licensee's autnerity was limited to maintenance of the facility in the present shutdown cooling mode (aa F.R. 45271). By further Order of tne Director, Office of Nuclear

. Reactor Regulation, datec February 11, 1980, a new set of femal license recuirements was imposed to reflect the post-accident condition of the y V () b G y

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.l facility and to assure the continued maintenance of the current safe, ' stable, long-tem cooling condition of the. facility '(45 F.R.11282). These re:;uire-ments, in the form of proposed Technical Specifications, would modify the

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facility operating license so as to

(1) define operating parameters for the current ' safe, stable,

, long-tem cooling mode for the facility (defined as the recovery mode), and-delete all other pemissible operating modes so as to assure that operation of the facility in other than the stable shutdown condition of the recovery mode is precluded; (2) impose functional, operability, redundancy and surveillance requirements as well as safety limits ano limiting concitions with regard to those structures, systems, equipment and components necessary to maintain the facility in the current safe, stable shutdown condition and to cope with foreseeable off-nomal conditions; (3) prohibit venting or purging or other treatment of [the approxi-mately 57,000 curies of krypton-85 in)-the reactor building '

atmosphere, the discharge of water decontaminated by EPICOR-II system, and the treatment and disposal of high-level. radio-actively -#

contaminated water in the reactor building, until eacn these activities has been approved by the NRO,1/

consistent with the Commissicn's State ent of Policy aiid Notice of Intent to Prepare a Programatic Environmental Impact Statement (44 F.R. 67738).

On the basis of the public health, safety, and interest, the requirements of the proposed. Technical Specifications, attached to the February 11, 1980 Order, were made effective immediately. Under the tems of the Order, since requests for a hearing- are pending before an Atomic Safety and Licensing Boarc, the  ;

i proposed femal license amencment incorporating these proposed Technical 1

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Sy Memorancum and Order, dated June 12, 1980, the Commission gave the approval contemplated by this restriction insofar as necessary for the Licensee to conduct a purging of the TMI-2 containment in accordance with procedures approved by the NRC. CL I 25. This activity was completed on July 11, 1980, l

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1 Specifications will beccee effective, in the event a hearing is granted, on the date specified in an order made following the hearing or, upon other

' final disposition of such proceeding, i

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1 Following the March 28, 1979 accident at TM!-2, it became necessary in late

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April 1979 to alter the preferred cooling mode for the reactor by a transition fg from use of forced circulation by the reactor coolant system pumps to natural convection circulation.

By le tters da ted July 31,1980 (Reference 1) and August 5,1980 (Ref'erence 2),

e o the Metropolitan Edison Company (licensee) proposed changes to the Recovery i

Moce technical specifications for Three Mile Island Unit 2 (THI-2) providing for ,ne implementation of the Mini Decay Heat Removal System (u.0 HRS) for len;-tem c:*e cooling. Altneugh several neces for removing decay heat would be available, the MOHRS woulc provide a forced flow system for removing decay heat fr:n the TMI-2 reactor fuel. Accordingly, the proposed changes I, would impose operability requirements for the MDHRS and would also delete tne operability recuirements for certain Balance of Plant (30P) systems wnien nave teen used since the March 28, 1979, accicent, but would no longer be recuirec, for removing tne decay heat. The operacility requirements for these SOF systems had been imposed by the Order of the Director of the i

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i Office of Nuclear Reactor Regulation on February 11,1980, (45 F.R.11282) in the form of proposed Technical Specifications.

The proposed change would make available a newly installed MDHRS to remove decay heat rather than the present method which acccm;'ishes tnis function by using the "A" steam generator in a steaming mode to the condenser. The licensee's proposal required certain modifications to meet our recuirements.

With the incorporation of these staff required modifications, we have found a

the proposal to be an acceptable method for removing the cecay heat and have therefore granted the licensee's recuest to modify tne method used for long term core cooling.

The TM!-2 Reactor Coolant System (RCS) is currently operating in a natural circulation heat removal mode with heat rejection from it being accomplisned by botn loss to ambient (reactor at.mosphere and sum: water) and through the "A" steam generator. The reactor building is in turn' befog cocle: cy the reactor building fan coolers while the "A" steam generator is steaming to the condenser through the turbine bypass valve. This mode of core cooling has been in effect since late April 1979. With passage of time and the associated reduction of decay heat generation rate (presently approximately 75 kw), the natural circulation flow has changed fron continuous to cyclic with increasing intervals between the cyclic flow " burps".

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1' Feat rejection through the "A" steam generator by steaming to the condenser  !

i requires the operation of several major B0p systens including: circulating l

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5-water syste, main steam system and the "A" steam generator, condensate and feedwater systee.s, main condenser and package boiler. Conversion to and use of the MDHRS for core cooling would simplify the plant operations since use of the MOHR$ would eliminate the need for operating the previously noted BOP systes.

The MDHRS is classified as a nonsafety grade system but it is designed and installed to seismic Category I requirements up to and including the second isolation valve in its supply and discharge lines. The balance of the syste is cesigned and installed to Operating Basis Earthquake requirements.

The CHRS takes suction from the "B" loop of the Decay Heat Syste (DH) cutlet from tne reactor vessel via a connection to tne Alternate Decay Heat Removal Syst s (ADH). Af ter passing through one of the MDHRS's parallel heat ex: hangers and pumps, the reactor coolant is returned to the reactor coolant systr ta. rough the "B" Core Flooding injection no::le via a connection to tne O H an: 04 systa s. Conne: tion of the CH to the DH was evaluated and accroved in NUREG 0557. The MDHRS is si:ed such that one pumo and one heat exchanger (two of each are installed) could remove up to approximately 1 MW of decay heat. Therefore the MDHRS has more than adequate cooling capacity for re.oving t tne present and future decay heat loacs. The MOHRS would be cooled by the Nuclear Services Closec Cycle Cooling Syste wnicn is reovired to be operable by proposed Technical Specification 3.7.3.1. The power sup:1y for the M HRS pumps and motor operated valves is from reduncant Class 1E busses which would be manually loaded on the Class 1E diesel ;enerators in the event of a less of off-site pcwer.

In the event _ the MDHRS is not used or becomes inoperable, backup cooling modes are available for removing the decay heat from the RCS. These backup cooling modes include the long term "B" steam generator cooling system and "Less-to Ambient." The long term "B" steam generator cooling system has

-been previously evaluated and its operability would continue to be required by proposed Technical Specification 3.7.1. The NRC staff has reviewed the licensee's results of the " Loss to Ambient" cooling mode and has performed an independent analysis, the results of which are in agreement with the l licensee's conclusion. Therefore, we have concluded that any one of these cooling methods provides an acceptable means for long term cooling of the reactor core. The staff's overall' evaluation of the MDHRS is presented in

ne concurrently issued Safety Evaluation Report (SER).

The MDHRS has a design pressure of 235 psig. Therefore, consideration was

iven to possible sources of overoressuri

ation of the MD"8.S. Three ooten-tial sources of MDHR5 overpressurization aere identified. These sources

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= (1) Makeup pump operation with MDHRS in operation, (2) Pressuri:er heater coeration and (3) Malfunction of the Standby pressure Control (SPC)

System.

To preclude operation of a makeup pump during operation of the MCHRS, the licensee proposed to delete the recuirement for an operable makeus pump from the proposed Technical Specifications but tJ retain the option to Ocerate the pump for certain operations (e.g., degassing). The licensee further

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' l stated that the electrical power supply cir:uit breakers for the makeup pumps would be "ra:ked out" when valve OH-VI or DH-V171 is open. Since operation of a makeup pump may be required in one or more of the backW f cooling modes or for degassing operations, we will retain a recuirement for its operability in pr0 posed Technical Specification 3.1.1.1. However, l.

I to provide assurance that the MDHRS will not be overpressuri:ed due to l

- operation of a makeup pump, we propose to add a recuirement to proposed Tecnnical Specifications 3.1.1.1 that all makeup pumps be made inoperable when valve DH-Vi or DH-Y171 is open by " racking out" their ele:trical power supply circuit breakers. We would also add a surveillance requirement to the Re::very Operations Plan to periodically verify that these breakers are 1 i

"ra:ked out." These actions pr:vice assurance that the MCHRS would not be  ;

l over;ressuri:ed due to coeration of the makeup pumps.

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Operation of the pressuri:er heaters while operating the ROS in a water solid moce with the MOHR$ in operation creates the potential for ovegres-suri:ation of the MDHRS due to volumetric expansion of the reactor coolant

• as a result of heat input to the reactor coolant. The licensee has cal-  !

culated that the electrical energi:ation of all tne pressuri:er heaters (1638 kw) would result in a volumetric expansicn on the reacter coolant which would recuire a compensating relief capa:ity Of 8.5 gpm. The MOHR5 has an installed relief valve capacity of 53.5 gpm. The NRO staff has reviewed '.he licensee's result of this potential overpressuri:sti:n event and has per#crmed an independent check, the results cf which are in agree-ment with the licensee's conclusion. Therefore, we agree that operation Of i

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l' the pressurizer heaters while operating the MOHR$ with a water solid RCS would not-result in overpressuri:ation of the MDHRS.

Overpressurization of the MDHRS due to a malfunction of the SPC system has 9

been precluded by reducing the SPC in-service nitrogen bank pressure to a new operating range of 225 to 400 psig and by the installation of a SPC systems pressure relief valve (SPC-R14) which has been set to provide e/erpressure relief if the SPC system pressure exceeds 125 psig which is substantially below the MOHRS design pressure of 235 psig. The change in the nitrogen bank pressure was approved on July 25,1980 (Reference 3).

The licensee also proposed deleting from proposed Technical Specifica-tion 3.1.1.1 the requirements for a boric acid storage system and an asso-ciated flow path to the RCS, We have reviewed this proposed change and since redundant boren injection flow paths from the BWST to the RCS via the nakeup pun; and decay heat removal . pump exist, we find the proposed change ac cep table .

The licensee's proposed Technical Specification for t.he MDHRS would require only one operable M0 HRS pump and heat exchanger with an action statement providing ins tructions to be taken in the event of their inoperability. Ou r position is that this Technical Specification should require the operacility of both MDHRS pumps and heat exchangers and the associated flow path anc that apolicaole action statenents should be supplied dealing with the inoperability of the various components in the MOHRS. The staff's position

9 is consistent with the operability requirements for similar systems (e.g.,

proposed Technical Specifications 3.7.2.1, 3.7.3.1 and 3.7.4.1) in which the redundant systems are required operable and action statements are pro.

vided for when one or both systems are inoperable. Unless both MDHRS pumps are required operable and periodically demonstrated so in accordance with applicaole surveillance requirements, there is no assurance of the operability of the redundant pump should its use be required for any reason. We have

. therefore modified the Technical Specification proposed for the MDHRS to be in accordance with our position. The licensee has agreed with our position on tnis matter. We have also added appropriate surveillance recuirements I to the Operations Recovery Plan to periodically demonstrate the operability of the N HRS.

l Ine Licensee also proposed to delete from proposed Technical Specifica-tien 2.8.2.1, the ope-ability requirements for several electrical power l busses. Its basis for proposing to delete these requirements was tnat tney supplied electrical power to the various 80P systems which were proposed for deletion from the proposed Technical Specifications uoen incorporation of the MDHRS. However, in our review of these proposed changes, we deter.

minec that four of the busses proposed for deletion (480 volt Susses 2-25, I '-36, 2 45 and 2 46) also supplied electrical power to the auxiliary tuild-ing and the fuel nandling building air cleanup systems which are in turn

! required cperable per proposed Technical Seccification 3.9.12. Therefore, to ensure the electrical power supply for these air cleanup systems, we I

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have retained the operability requirements for these four busses in proposed Technical Specification 3.8.2.1. The licensee has agreed that the opera-

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bility requirements for these frur busses should be retained in proposed-

] Technical Specification 3.8.2.1.. We agree that the operability rt;uirements for:the other busses can be deleted from the. proposed Technical Specifica-tions as proposed by the licensee since the other busses proposed for dele-tion do net supply electrical power to any- systems' required for maintain-ing the plant in its safe shutdown condition.

Based on tne staff's review of accident considerations, as presented in the.SER, the staff has concluded that use of C.e MOHRS does not increase tne probablity or consequences of an accident or malfunction previously considered or reduce a margin of safety, and, thus, does not involve a signi'icant ha:ards consideration. Indeed, as described abot'?, the staff considers that the use and availability of the MOHRS will enhance the licensee's ability to maintain the reactor in a safe shutdown cooling mode by providing a simplified and appropriately si:ed decay heat removal system.

We have also determined that the modification does not authori:e a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental imcact. Havin; made this deter-i l

mination, we have further concluded tnat the modification involves an action which is insignificant from the standpoint of environmental imcact and, pursuant to 10 CFR Section 51.5(d)(a), that an environmental imoact state.

ment or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the modification. R

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Accorcingly, pursuant to the Atcrnic Energy Act of 1954, as amended, the

. Director's Order of February 11, 1980 is hereby revised to incorporate the I

celetions, a:ditions and modifications set fortn in A::achment A here:3.

i For furtner details witn respe:: to this a icn, see (1) Letter to B. Snyder, ,

USNRC, from R. C. Arnold, Met. Ed/GPU, Technical Specifica:icn Change Recuest No. 2a, cated July 31,1980, (TLL 372); (2) Letter to 3. Snyder, USNRC, frr.

R. C. Arnold, Met. Ed/GPU, Te:hnical Spe:ification Change Reques: No. 24, dated August 5,1980, (TLL 382); (3) Letter to R. C. Arnold, Met. Ed/GPU, fr::- Jchn T. Collins, USNRC, TM!-2 Re:cvery Operations Dlan Change Re;ues:

i No. 4, da ted July 25.,1980, (NRC/TMI-SC-115); and (4) the Director's Order of Fecruary 11, 1980.

All of :ne at:ye documents are available for ins;e::1on at :ne Commissien's Public Document Room,1717 H Street, N.W. , Washington. 0.C., and at the Commission's Lc:51 Public Document Rocm at the State Library of Pennsylvania, Governmen; Publications Section, Education Building, Cor=onwealth and Walnut j Streets, Harrism;q, Pennsylvania 17126.

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r0R THE NUCLEAR REGULATORY COMMISS:0N l

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/ Wb Harold R. Denton, Director i Office of Nuclear Reactor Regula:icn  !

l Effective date: Neve_ber la. 1980 Da ted at 3etnesda , Marylanc l

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e SAFETY EVALUATION BY THE OFFICE OF HUCLEAR REACTOR REGULATION METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-320 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 2 Introduction By letters cated July 31,1980 (Reference 1) and Augus: 5,1980 (Reference 2),

tne Metropolitan Edison Company (licensee) proposed changes to the Recovery Moce technical specifications for Three Mile Islanc Unit 2 (TMI-2) dealing with the implementation of the Mini Decay Heat Removal System (MDHRS). Althougn several modes for removing decay heat would be available, the MDHRS would provice a forced flow system for removing dccay heat from the TMI-2 reactor fuel. Accordingly, tne proposeo changes would impose operacility requirements for tne MDHR$ anc would also

, celete One operacility requirements for certain Balance of Plant (50P) systems wnicn

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have been used since the March 28, 1979, accident, but woule no longer ce requireo, for removing tne decay heat. The operability requirements for tnese 80P systems nac ceen imposec ey tne Order of the Director of the Office of Nuclear Reactor i Regulation on Feoruary 11,1980, (F.R.11282) in the form of proposed Tecnnical l

Specifications.

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r The licensee has recuestec flRC staff accroval of an additional long er l i

l  : ore cooling me: nod. The proposed cnanje would make available a newly installec j

't0 HRS to remove decay heat rather nan ne present me nod wnien accorelisnes :nis function by using :ne "A" steam genera or in a steaming mode to ,ne concenser. l j

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l 2-ine licensee's proposal required certain modifications to meet our requirements wi:n wnich :ne licensee has agreed. Llith tne incor oration of tnese staff re-cuired modifications, we have found the proposal to be an acceptable me: nod for removing the decay neat and have tnerefore gran:ed no licensee's request

modify the method used for long term core cooling.

Evaluation The TMI-2 Reactor Coolant System (RCS) is currently operating in a natural cir-culation neat removal moce with heat rejection from it oeing accomplishec ey coth loss to amoient (reactor suilcing atmosphere and sump water) and through tne "A" steam generator. The reactor ouilcing is in turn being coolec by tne reactor cuilcing

.f an coolers while the "A" steam generator is steaming to the concenser througn the turoine Dy-pass valve. This mode of core cooling has been in effect since late April 1979. With tne passage of time and the associatec recuction of decay heat generation rate (presently approximately 75 kw), the natural circulation flow has enangec from con-inuous to cyclic with increasing intervals between the cyclic flow "ourps".

Hea rejection through the "A" steam generator ey steaming to the concenser requires the operation of several major BCP systems including: circulating water system, main steam system and the "A" steam generator, concensate and feecwater systems, main concenser and package boiler. Conversion to anc use of the MDnRS for core cooling woulc simplify the plant operations since use of :ne M0 HRS woule l I

el'iminate the need for operating :ne previously noted 30P systems.

The MDnRS is classifiec as a nonsafety grace system out it is cesignec anc installed to seismic Category I requirements up to anc inclucing the seconc isolation valve in its supply anc cischarge lines. The calance of :ne system is cesignec anc insta11ec to Operating Basis Earthquake requirements. The MOHR5 takes suction from :ne "

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loop of the Decay Heat System (DH) outlet from the reactor vessel via a connection to the Alternate Decay Heat Removal System (ADH). Af ter passing through one of the MDHRS's parallei heat exchangers and pumps, the reactor coolant is returnec to tne reactor coolant system through the "3" Core Floocing injection noz:le via a connection to the ADH anc DH systems. Connection of the ADH to tne DH was evaluatec and approvec in NUREG-0557. The MDHRS is sizec such tnat one pump anc one neat exchanger (two of each are installed) could remove up to approximately 1 N of' cecay heat. Therefore the MDHRS has more than acequate cooling capacity for removing tne present anc future cecay heat loacs. ine MDkRS woulo ce coolec Dy :ne Nuclear Services Closec Cycle Cooling System wnicn is requirec to se operaole ey propesec Tecn-nical Specification 3.7.3.1'. The power supply fcr :ne MDHRS pumps anc motor operatec valves is from recundant Class 1E ouses wnicn woule ce manually loacec on :ne Class 1E ciesel generators in tne event of a icss of off-site power. A cetailec cescription of I the MDnRS anc of its principal moces of operation is provicec in the enclosure to Reference 3. i.

In tne event the MDHRS is'not used or secomes inopersole, oackup cooling moces are availaole for removing tne cecay neat from the RCS. These oackup cooling moces incluce tne long term "B" steam generator cooling system anc " Loss to Amcient". ine long term "B" steam generator cooling system has oeen previously evaluatec anc its operaoility woule continue to se requirec cy proposec Tecnnical Specification 3.7.1.

In its analysis of tne " Loss of Amcient" cooling moce (discussec in References 1 anc 21, tBe licensee calculated tnat witn the present cecay neat generation rate, :ne reaccor coolant oulk temoerature would initially increase at a rate of approximately U.4 U F/ncur l

anc that this heatup rate would gracually cecrease suen tnat at a RCS temperature of l

E approximately 190 0 F, the heatup rate would De approximately zero. At this RCS tempera-ture, tne heat loss from the RCS to amoient (reactor building atmosphere anc sumo water)' would De equal to the cecay heat generation rate. The RC5 operating pressure curing MDHR5 operation would be maintained at !9010 psig; therefore a substantial margin to the satura*f on temperature exists and the plant concitions would reacn a stacle, equilibrium condition. The NRC staff has reviewed the licensee's results 'of tne

" Loss to Amoient" cooling moce and has performed an incepencent analysis, the results of wnich are in agreement with the licensee's conclusion. Therefore, we nave conciucec tnat any one of tnese cooling methocs provides an acceptacle means for long term cooling of tne reactor core.

Written procedures for operating the MORR5 and for. operation in the " Loss to Amoient cooling moce will ce required for operation in eitner of these cooling moces. Tnese pro-cecures will De preparea and submitted to the NRC staff in accorcance witn the retu1re-ments of proposeo Technical Specification 6.8.1 and ti.8.2 prior to implementation.

The MDHR$ has a design pressure of 235 psig. Therefore, consideration was given to possiele sources of everpressurization of tne MDhRS. Three potential sources of MOHRS overpressurization were identified. These sources were: (1)

Maxeuo pump operation witn MOHRS in operation, (2) Pressuri:er heater operatien, anc (3) Malfunction of the Stancby Pressure Control (SPC) System.

1 To erecluce operation of a makeup pump curing operation of tne MDHRS, tne l

licensee proposed to delet. tne requirement for an operacie maceup pump f rom tne orcoosea Tecnnical Specifications out to retain tne option to operate the :t.mo for certain operations (e.g. cegassing). The licensee further stateo tnat tna electrical ower supply circuit t.reakers for the makeup pumps woule ce "racxec out" wnen valve

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")H-V1 or DH-V171 is open. Since operation of a makeus pump may be recuired in one or more of :ne backup cooling moces or for degassing operations, we will retain a reouf rement for its operability in proposed Tecnnical Specification 3.1.1.1. However, ,

to provide assurance that the MDHRS will not be overoressuri:ed due to opera: ion of a makeup pumo, we propose to add a requirement to proposed Tecnnical Soecification 3.1.1.1 that all makeuo pumps be made inocerable when valve DH-Vi or CH-V171 is open ,

by " racking out" tneir electrical power supply circui- breakers. We would also acd a surveillance requiremen to :ne Recovery Ooerations Plan to periccically verify na:

nese creakers are " racked out". These actions provice assurance :na: :ne MOHRS wouic not be overpressuri:ed due to operation of :ne makeup pumps.

Operation of tne pressuri:er heaters while operating the RCS in a water solic mode with the MDHRS in operation creates tne potential for overpressuri:ation of tne MDHRS cue to volumetric expansion of the reactor coolant as a result of hea; input to :ne reactor coolant. The license,e has calculated that the electrical energi:aticn of all the pressuri:er heaters (1638 kw) would result in a volumetric expansion o' the reactor coolant whien would require a compensating relief capacity of S.6 gpm.

The MDHRS nas an installed relief valve capacity of 52.5 gpm. ' ne NRC staff nas reviewed the licensee's result of tnis potential overpressuri:ation event anc nas performec an incependent check, tne results of wnich are in agreement with the licensee's' conclusion. Therefore, we agree that operation of tne pressuri:er neaters wnile operating the MDHRS with a water solic RCS would not result in overpressuri:ation J of the MDHRS. l 1

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I Overpressurization of the MDHRS due to a malfunction of tne SPC system has been precluded by reducing the SPC in-service nitrogen Dank pressure to a new operating range of 225 to 400 psig and by .tne installation of a SPC systems pressure relief valve (SPC-R14) which has been set to provide overpressure relief if the SPC system pressure exceecs 125 psig whien is substantially celow the MOHRS cesign pressure of 235 psig. The change in the nitrogen cank pressure was approved on July 25, 1980 (Reference 4).

The licensee also proposed deleting from proposec Technical Specification 3.1.1.1 the requirements for a boric acid storage system and an associatec flow patn to the RCS. We have reviewee tnis prop ~osed enange anc since recundant Doron injection flow patns from sne SWST to the RCS via the makeup pump anc decay heat removal pump exist, we fina the proposed change acceptacle.

The licensee's proposed Technical Specification for the M0kRS woulc require only one operaole M0KRS pump and heat exchanger with an action statement provicing instruc-tions to De taken in the event of their inoperaoility. Our positien is tnat snis Tecnnical Specification shoulc require the operacility of both MOHR$ pumps anc neat exchangers and the associated flow path and that applicaole action statements shoulc be suppliec cealing witn the inoperacility of the various components in tne M0nRS.

The staff's position is consistent with the operacility requirements for similar sys-tems (e.g. proposec Tecnnical Specifica 1 css 3.7.2.1, 3.7.3.1 anc 3.7.4.1) i n wnica ne recuncant systems are requirec operable and action statements are provicea for wnen one or moth systems are inoperable. Unless ooth MDkRS pumps are recuirec oper-aole and periodically cemonstrated so in accordance witn applicaole surveillance recuirements, there is no assurance of the operacility of tne recuncant pumo shoule its use ce required for any reason. We have therefore modifiec tne Technical Speci-l 1

~ . .- - - . . .

fication proposed for the MOHRS'to oe in accorcance with our position. The licensee has agreec with our position on this' matter. We have also accec appropriate surveillance l

requirements to the Operations Recovery Plan to periccically demonstrate the operability

)

of the MOHRS.

The Licensee also proposec to celete from proposec Technical Specification 3.6.2.1, the operaoility requirements for several electrical power ousses. Its  ;

basis for proposing to celete tnese requirements was tnat tney supplied electrical ooner :: the various 30P systems wnien were proposed for celetion from the pro-posec Tecnnical Specifications upon incorporation of tne MDnRS. However, in our review of these proposec cnanges, we cetermined tnat four of the ousses proposec _ ~

for celetion (480 voit Busses 2-35, 2-36, 2-45 and 2 a6) also suppliec electrical pcwer to ne auxiliary builcing anc tne fuel hancling ouilcing air cleanup systems which are in turn required operaole per proposed Technical Specification 3.9.12.

Therefore, to ensure tne electrical poner supply for these air cleanup systems, we have retained the operability requir'ements for tnese four Dusses in proposec Tecnnical Specification 3.8.2.1. The licensee has agreec tnat the operaoility re-

~

quirements for tnese four busses should De retainec in proposec Tecnnical Specificatien 3.8.2.1. We agree that the operaoility requirements for tne otner cusses can ce celeted from tne proposec Technical Specifications as proposec ey .ne licensee since the other cusses proposed for celetion Co not supply electrical pcwer to any systems recuirec for maintaining the plant in its safe snutcown concition. 1 Tne licensee has postulatec an accicent involving isolation. of tne M0nRS at inlet anc outlet isolation valves followec ey approximately 1200 gallons of coolant water inventory being cumped on the floor of the auxiliary cuilcing. Since the i MOHRS operating temoerature is snown oy the licensee's tnennal analysis never to exceec I

8 l

I i

190 F, even when the system neat removal capacity is lost, dumping of tne coolant

'is no excected to result in any flashing. In the a":ysis, the licensee assumed nat the airborne source term will not exceed 1S of tne :ctal coolan inventory of radionuclides censisting primarily of cesium and strontium. The airoorne scurce term in :ne auxiliary building is then released to the a:noschere as a puff of radionuclices. As a result of tnis analysis, the licensee calculated a site boundary cose of 1.le x 10'# rem.

l We also considered the possioility of recriticality of .tne reactor core curing operatien of tne M0 HRS. The potential for recriticality with tne reactor coolan:

system operating in cyclic natural circulation was evaluated in Reference 5. Tne only metnod identified therein that coule lead to recriticality was by boron ci-lution of :ne reactor coolant. In Reference 5 it was conclucec tnat the timing of suen an accicent was very long anc that several weeks of continuous coran cilu-

1cn a :ne then present makeue rate woule ce requirec to reacn criticality. Tne present makeup rate is lower by a f actor of at least two : nan the makeup rate usec in that analysis. Due to the mixing action provicec ey the M0nRS pumps, use of tne MDkRS will ensure uniform coron concentrations in :ne core and, as tescrioec in Reference 3, will provice a more representative samole of tne reactor coolant anc tnerefore a potential :cron cilution accident woulc ce cetec ec even more reacily nan as cescrioec in Reference 5. Therefore, we concur :na: :ne accicent pcstulatec cy ne licensee is the severest accicent. We performec a consequences analysis uti-li:ing tne source term calculatec from the racionuclice concentrations measurec in TMI-2 coolant samples by Oak Ridge National Lacoratory. The cata snowed aosence of any nocle gases, presence of extremely small concen: rations of iocine, anc :resence c' cesium ano strontium. The source term was therefore calculatec ey neglecting noole gases anc iocines.

.g.

I l

Based upon the review of the vapor pressure of cesium anc strontium at the 0

maximum coolant temperature of 190 F calculated oy the licensee, we celieve that ,

l tne licensee's assumption of it of the coolant racionuclices being airoorne is j l

conservative, and have used this value in our evaluation of off-site coses (Exclusion Area Bouncary) shown in Table 1.

I For the. postulated accident, the staff has used a "small fraction" (acout 1 percent) l of the 10 CFR 100 dose guicelines as a criterion for the raciological consequences which would not present an undue risk to tne public.

Since the coses calculated for a f ailure of tne MDhas, as snown in Tacle 1, are suostantially less tnan a "small fracticn cf Part 100," it is conclucec tnat tne postulated accicent would not cause an uncue risk to public healtn anc safety.

TABLE 1 Offsite Ooses For Mini Decay Heat Removal System Failure nnele Bocy Coolant Samples ~

X/Q Dese Nucli ce Concentration

1. Ci/cm sec/m mrem CS-134 20 1.1x10~3 2.4

~

Cs-136 0.5 1.1x10 1.5x10

~3 Cs-137 100 1.1x10 6.8

~

Sr-89 300 1.1x10 0.55 Sr-90 20 1.1x10" 29.2 TCTAL 39.3

1 i

l Environmental Consideration We have determined that the modification coes not authorize a change in effluent j types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having mace this cetermination, we nave furtner concluced that the modification involves an action which is insignificant from tne standpoint of environmental impact and, pursuant to 10 CFR Section 51.5(d) (4), tnat an environmental impact statement or negative ceclaration ano environmental impact appraisal neec not be prepared in connection with the issuance of the mocification.

Conclusien ,

Basec upon our review of M0 HRS mocification, one attencant Tecnnical Specifica-tions, and our findings that the proposec mode provice cooling options wnich are reliaole anc less corplex means for long term core cooling, we find the licensee's request to oe acceptacle and grant the request to make said mocifications. The measures authorized in connection with tnis evaluation will assure the continuec maintenance of the f acility in a safe, stacle, long-term cooling condition, as ciscussec aoove. Basec on these considerations, we have concluced tnat: (1) tne mocification e coes not involve a significant increase in the probacility or consequences of accicents  ;

previously consicerec or a significant reduction in a margin of safety anc coes no  :

involve a significant hazarcs consiceration, (2) there is reasonaole assurance tnat i the healtn and safety of the public will not be encangereo ey operation in tne mocifiec manner, anc (3) such activities will De concucted in compliance witn tne Commission's  ;

regulations and the issuance of this mocification will not ce inimical to the common cefense and security or to the health anc safety of the puolic.  !

f e

h

i REFERENCES l

1. Letter to B. Snyder, USNRC, from R. C. Arnold, Met. Ed/GPU, Technical Specification i Change Request No. 24, datec July 31, 1980, (TLL 372).

2. Letter to B. Snycer, USNRC, from R. C. Arnold, Met. Ed/GPU, Technical Specification Change Request No. 24, cateo August 5,1980, (TLL 362).

3. Letter to John I. Collins, USNRC, from G. K. Hovey, Met. Ec/GPU, MOKR System Description, Revision 3, catec Septemoer 8,198u, (TLL 438).

4 Letter to R. C. Arno10, Met. Ed/GPU, from John T. Collins, USNRC, THI-2 Recovery Operations Plan Cnange Request No. 4, detec July 25, 1960, (NRC/TH1-60-115).

5. Memorancum for William J. DircKs, from Normal M. Haller, " Report of Special Task Force on inree Mile Islanc Cleanup", catec Fecruary 25, 1980.

I a

l l

e Attachment A 1

FACILITY OPERATING LICENSE NO. OPR-73 DOCKET NO. 50-320 Re: lace the following pages of the proposed Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages contain vartical lines incicating the area of change. The ccrresponding overleaf pages are also provided to rnaintain document cogleteness.

Paces 3.1-1 3.1-2

3. -1 3.7-1 3.7-2 3.B 4 5 J/4 4-1 B 3/4 7-1 I

THREE MILE ISLAND - UNIT 2 i 1

l _ , . . , . - _

.(-

i LIMITING CONDITIONS-FOR OPERATION 1 l

\

l 3.1 WATER INJECTION COOLING AND REACTIVITY CONTROL SYSTEMS , , I 3.1.1 BORATION CONTROL l

BORON INJECTION l 3.1.1.1 At least two systems cacable of injecting berated cooling water ir.to I tne Reactor Coolant System shall be OPERABLE

• with: '

l

a. One system comprisec of: l l

.1. One OPERABLE makeup pump.#

2. One OPERABLE decay heat removal pump.

3. An OPERABLE flow path from the BWST. Tne BWST shall con'tain at least 100,000 gallons of berated water at a minimv:n temoerature of 50*F and at a boron concentration of between 3000 anc 4500 ppm.

b. The second system comorised of the Standby Reactor Cc.olant System Pressure Control System.

APDLICAEILITY: When fuel is in the reactor pressure vessel.

ACTION:

With one of the above requi ed systems inoperable, *estore the inoperable system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

BORON CONCEN RA~ICN 3.1.1.2 The reactor coolant shall b~e maintained at a boron concentration of between 3000 and 4500 pcm and at a temperature atose 50*F.

APPLICABILITY: When fuel is in the reactor pressJre vessel.

jCTION None exceot as proviced in Scecification 3.0.3.

l l

"sotn systems sna11 be consicered OPERAS.E wnen alignec per procedures approved i cursuant to Specification 6.8.2. l

1. All makeuc pumps shall be mace inoperrole wnen valve CH-V1 or OH-V171 is open by racking out their electrical power'succly circuit creakers.

THREE MILE ISLAND - UNIT 2 '3.1-1

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c Celete::

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l THREE MILE ISLANO - UNIT 2 3.1-2

LIMITING CONDITIONS FOR OPERATION 3.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS 3.4.1 Tne Reactor Coolant System shall be operatec in accordance_with procedures approved pursuant to Specification 6.8.2.

APPLICABI.ITY: RECOVERY M00E.

ACTION:

None except as provided in Specification 3.0.3.

SAFETY VALVES 3.4.3 All pressuricer code safety valves snali ce OPERABLE witn a lift setting of 2435 PSIG 1%."

AcoLICAEILITs: RECOVERY M00E.

ACTION:

hone except as provided in Specification 3.0.3.

3.4.9 PRESSURE /TEMDERATURE LIMITS REACTOR COOLANT SYSTEM ,

3.4.9.1 ine Reactor Coolant System'sna11 be maintained at a T avg of less tnan 280*F and at a pressure of less than 600 psig.

APDLICABILITY: When fuel is in the reactor pressure vessel.

ACTION:

Ncne exceot as provicec in Specification 3.0.3.

I "ine lift setting pressure snali correscond to amoient concitions of tne valve at riominal operating temoerature and pressure.

THREE MILE ISLAND - UNIT 2 3.4-1 l

k' LIMITING CONDITIONS FOR OPERATION 3.7 PLANT SYSTEMS 3.7.1 FEEDWATER SYSTEM 3.7.1 The long term "B" steam generator cooling system shall be maintained in an OPERABLE status.

APDLICAEILI'v: RECOVERY MCOE.

ACTICN:

Witn the long term cooling "B" steam generator cooling system inoperable, restore the inoperable system to OPERABLE status -ithin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3.T 2 SECONCACV SEcVICES CLOSED CCOL:NO WATER SYSTEM 3.7.2.1 At least two seconcary services closeo cooling wate- system pumos and j heat excnangers anc the associated flo- patn snail ce OPERABLE witn eacn pum:

ca'paele of being powered from separate Ousses.

_ J APPLICAEILITY: RECOVERY MODE.

ACT 0N:

With only one secono, , se* vices closed cooling water pum: or only cne seconcary services heat exchanger dPERABLE.,_festore the inoperable pump or heat exchanger to 00ERAE'E statas witnin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3.7.3 CLCSED CvCLE C00 LING WATER SYSTEM NUCLEAR SERVICES CLOSED CvCLE COOLING SYSTEM 3.7.3.1 At least two independent nuclear services closed cycle cooling water pumps anc heat exchangers and the associated flow path shall De OPERABLE with eacn pump C3Dable of being powerec from separate emergency busses.

A;oLICAE:.TY: RECCVERv MCDE.

I ACT:CN; I With only one nuclear services closed cycle cooling -ater sumo or only one  ;

nuclear services heat excnanger OPERAELE, restore tne incoeracle pue or nea; i excnanger to OPERAELE status witnin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

THREE MILE ISLAND - UN:T2 3.7-1 l 1

LIMITING CON:!TIONS FOR OPERATION DECAY HEAT CLOSED COOLING WATER SYSTEM 3.7.3.2 At least one decay heat closed cooling water loop snall be OPERA 5LE.

A:DL: CAB:::TY: RECOVERY MODE.

ACTION:

Witn he Oe:a) neat cicsed cooling water loop OPERAELE, restore tne inoperable loco t: GEERAE E status wintin 24 nours.

MIN: DECAv HEAT REMOVAL SYSTEM 3.7.3.3 Two mini de:ay neat removal pumps and heat ex:nangers and tne asse:i-4 ate:. flow at", snail be OPERAELE.

ADO.:"AE:.:T' : When fuel is in tne reactor pressure vessel.

a ', . ; b 's ,

s. With cne mini decay heat removal pum* and/or heat exchanger inoceracle, restore one inopera:le pumo and/or neat ex:nanger : OPERABLE'statas witnin 72 nours.

. With t'wo mini decay heat removal cumps and/or heat exenangers or tne asso:iated flow patn inoperacle, restore at least one pum: and heat ex:nanger ano the asse:iatec flow path to CPERABLE status witmin 24 now s Or itnin tne next 48 nours make a ca:ku: cooling syste- -

(eitner LTC "B" or " Loss to Amoient") OPERABLE.

2.7.4 NU:'. EAR SERVICE RIVER WATER SYSTEM 2.7.4.1 Tw: independent nuclear service river water loops snall be OPERABLE.

APOLICAE!LITY: RECOVERY MODE. ,

ACTI:V

'ad t :ne ucles' service river water looc inoperacle, rest: e tne inoce-atie

: 00ERAE.'E status witnin 72 nours. I 1

THREE MILE ISLANO - UNIT 2 3.7-2

- - - .. . . . . = _ . .

1 l

l I

l TABLE'3.8-2 RESTORATION TIME MATRIX t

Restore One Restore Other i Component Component  !

(Hours) (Hours) l l

e l aa 24 l 72 ,

.5 a =

ifi

.: =

i 222 g ab 12 72 i i ? -$ $  !

3 .5 5 bb 12 72 I

Nete: a and acove cc respenc t: ::m:ene.9ts cescribe in 5:ecificat i on

a. . .1 ite.ts a ar.c b respectively. l I

O i

i l

i t

I i

l THREE MILE ISLAND - UNIT 2 3.8-3

J 4

LIM **ING CON 0!TIONS FOR OPERATION  ;

4 3.2.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTICN 3.B.2.1 _ine following A.C. electrical busses shall ce OPERABLE anc ene gi:ed witn tie creakers open (unless closed in accordance with procecures approved pursuant tc Specification 6.8.2) between reduncant busses:

a*50

. ve't Emergency Bus # 2-1E and 2-2E als:- volt Emergency Bus # 2-2E and 2-4E 4150 volt Busses # 2-3 and 2-4 A S .' velt Emergency Bus # 2-11E, 2-12E anc 2-31E 45; vclt Emergency Bus # 2-21E, 2-22E anc 2-41E 45: vel Busses # 2-3'., 2-41, 2-32, 2-42, 2-35, 2-35. 2-45, a-d 2-45 l

12: velt A.C. Vital Bus # 2-1V 120 vcit A.C. Vital Bus # 2-2V

• 20

. voit A.O. Vital Bus # 2-3V 120 volt A.C. Vital Bus # 2-4V

-- ...-:... ' s:.sv:s' Mu....:.

m.

.....N:

Wi t- less tr.aq ine above comolement of A.C. busses CPERABLE, restere the ircoeracie tus to OPERAELE status witrin 8 nours.

i inREE MILE ISLANO - UNIT 2 3.8-4

3/4.4 REACTOR COOLANT SYSTEM BASES ,

3/4.4.1 REACTOR COOLANT LOOPS Several alternative methods are available for removal of reactor decay heat. These methods include use of the Mini Decay Heat Removal System, the

" Loss to Amoient" cooling mode, and operation of tne Reactor Coolant System in the natural circulation moce with heat rejection via the long term "B" 6 team generator cooling mode. Any one of these cooling methods provides adecuate cooling of the reactor and each method is available for decay heat removal. Procedures have been prepared and approved for use of these various cooling methods. I 3/a.4.3 SAFETY VALVES 1

The cressuri:er coce safety valves operate to prevent the RCS from be:ng cressurizec above its Safety Limit of 2750 psig. Eacn safety valve is cesigned to relieve 348,072 lbs per hour of saturated steam at tne valve's setpoint. 1 1

3/4.4.9 PRESSURE / TEMPERATURE LIMIT ine RCS pressure anc temperature will be controllec in accordance with accrevec procecures to prevent a nonductile failure of the RCS wnile at the same time cermitting the RCS pressure to be maintained at a sufficiently hign value to cermit operation of the reactor Cociant pumos.

l THREE MILE ISLANO - UNIT 2 B 3/4 4-1

3/4.7 PLANT SYSTEMS BASES '- -

t 3/4.7.1 FEE 0 WATER SYSTEM The long term "B" steam generator cooling system is required to be main-tained in an OPERABLE status since it is an alternative method for removing decay neat from the reactor coolant system.

3/4.7.2 SECONCARY SERVICES CLOSED COOLING WATER SYSTEM The secondary. services closed cooling water system is required to be maintained in an OPERABLE condition since it is used to cool the "B" steam generator closed loop cooling system.

3/4.7.3 CLOSED CYCLE COOLING WATER SYSTEM -

3/a.7.3.1 NUCLEAR SERVICES CLOSED CYCLE COOLING SYSTEM OPERABILITY of the nuclear services closed cycle cooling system is recuired during operation of the MDHRS since this system provides the heat sink for the MOHRS.

3/4.7.3.2 DECAY HEAT CLOSED COOLING WATER SYSTEM The decay heat closed cooling water system is required to be maintained

- in an OPERABLE status since it is provided to remove heat from the OHR system-which is being maintained OPERABLE in a backup status for possible core cooling.

3/4.7.3.3 MINI DECAY HEAT REMOVAL SYSTEM (MDHRS)  ;

OPERABILITY of the M0 HRS is required since it is an alternative method for removing decay heat from tne reactor. The M0 HRS is provided with two pumos i

and two neat exchangers; one pump and one heat exchanger have adequate capacity for removing tne present level of decay heat from the core. ,

j 3/4.7.4 NUCLEAR SERVICE RIVER WATER SYSTEM The nuclear service river wate* system uses river water to coc1 tne nuclear services closed cycle cooling system, the secondary services closee cooling water system, and cecay heat closec cooling water system; therefore, it must be OPERABLE too. This system rejects its heat to the river as the ultimate heat sink.

THREE MILE ISLAND - UNIT 2 B 3/4 7-1

CONTAINMENT SYSTEMS

.B A.S E S

~ ,.,,

3/4.7.6 FLOOD PROTECTION The limitation on flood pectection ensures that f acility protective actions will be taken in tne event of flood conditions. Tne limit of elevation 302 Mean Sea Level is basec on :ne maximum elevation at which facility flood control measures provice protection to safety related equipment.

3/4.7.7 CONTROL ROCM EMERGENCY AIR CLEANUP SYSTEu The OPERABILITY of the control room emergency air cleanup system ensures tnat 1) the ameient air temperature coes not exceed the allowable temoerature for continuous duty rating for the ecuipment and instrumentation cooled by nis system and 2) the control room will remain nabitable for operations pers:nnei curing and following all credible accicent conditions. The OPERABILITY

, of nis system in conjunction witn control roce cesign provisions is based on limiting tne raciation exposure to personnel occupying tne conteci room to 5 rem or less anole cecy, or its ecuivalent. This limitation is consistent with ne recuirements of General Design Criterion 15 of Appencix "A", 10 CFR 50.

3/4.7.10 FIRE SUPPRESSION SYSTEMS The OPERAEILITY of the fire suppression systems ensures tnat adecuate fire superession capacility is availatie to confine anc extinguish fires occuring in any portion of the f acility where safety related ecuipment is locatec. The fire suppre5sion system consists of the water sys*em, spray anc/or serinkleas, Halon anc fire nose stations. The collective capatility of the fire suppression systems is adecuate to minimi:e potential camage to saf ety related ecuipment and is a major element in the f acility fire protec-tion prog am. Any two of the four main fire pum:s previce com ined capacity greater inan 3575 gpm.

In tne event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is recuirec to be made availacle in the affected areas until the affected ecuipment can be restored to service.

In the event that the fire sue:ression water system Decomes inoperable, immediate corrective measures must be taken since inis system provides the major fire suppression capacility of the plant. The recuirement for a Special Report to tne Commission provices for timely evaluation of the acceptacility of the corrective measures to provice adecuate fire suppression capa:ility for tne continued operation of tne nuclear plant.

THREE MILE ISLAND - UNIT 2 B 3/4 7-2

Attachment A FACILITY CPERATING LICENSE NO. DPR-73 '

00Cr.ET NO. 50-320 Replace the follow ng pages of the proposed Appencix 'A" Technical Specifications d

with the enclosed pages as indicated. The revised pages contain vartical lines incicating the area of change. Tne corresponding overleaf pages are also provided to mainta* n document completeness.

Paces 3.1-1 3.1-2 3.5-1 3.7 1 3.7-2 3.5 4 3 3/4 A-1 B 3/4 7-1 e

1 i

l l THREE MILE ISLAND - UNIT 2 i l

l

LIMITING CONDITIONS FOR OPERATION ,,

3.1 WATER INJECTION COOLING AND REACTIVITY CONTROL SYSTEMS ,

3.1.1 BORATION CONTROL BORON I_NJECTION 3.1.1.1 At least two systems capable of injecting berated cooling water into the Reactor Coolant System shall be OPERABLE

• with:

a. One system comprised of:

1. One OPERABLE makeup pump.#

2. One OPERABLE decay heat removal pump.

3. An OPERABLE flow patn from the BWST. The BWST shall contain at least 100,000 gallons of beratad water at a minimum temperature of 50*F and at a boren concentration of between 3000 anc A500 ppm.

b. The seconc system com:-ised of the Stancby Reactor Coolant System Pressure Control System. l APDLICAEILITY: When fuel is in the reactor pressure vessel, ACTION: l 1

With one of the above required systems inoperable, restore the inoperable I system to OPERABLE status within 72 hc;rs.

I l

BORON CONCENTRATION 3.1.1.2 The reactor coolant shall be maintained at a $oron concentration cf between 3000 and 4500 pom and at a temperature above 50*F.

APPLICABILITY: When fuel is in the reactor pressure vessel.

ACTION None except as provided in Soecification 3.0.3.

• Sotn systems snall be considerec OPERABLE when aligned per procecures aoprovec pursuant to Specification 6,8.2.

1. All makeup pumps shall be made inoperacle when valve OH-V1 or OH-V171 is open i by racking out their electrical power supply circuit breakers. l THREE MILE ISLAND - UNIT 2 3.1-1

Oelete:

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1 1

1 1

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THREE MILE ISLAND - UNIT 2 3,1-2

LIMITING' CONDITIONS FOR OPERATION 3.4 ' REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS

' 3.4.1 The Reactor Coolant System shall be operatec in accordance with procedures approved pursuant to.5pecification 6.6.2.

APDLICAEILITY: RECCVERY MODE. ,

ACTION:

None except as p-ovided in Specification 3.0.3.

SAFETY VALVES 3.4.3 All cressuri:er code safety valves shall be OPERABLE with a lif t setting of 2435 PSIG - 1%."

APPLICAE'LITY: RECOVERY MODE.

ACTICN:

None except as provided in Specification 3.0.3.

3.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM 3.4.9.1 The Reactor Coolant System shall be maintainec at a T avg of less than ,

280*F and at a pressure of less than 600 psig,

. ApeLICAEILITY: When fuel is in the reactor pressure vessel. 1 ACTION:

None except as provicec in Specification 3.0.3.

d P

"Tne lift setting pressure shall correspond te ambient concitions of the valve at nominal operating temperature and pressure. ,

THREE MILE ISLAND - UNIT 2 3.A-1

1 l

l

'_LI W: TING CONDITIONS-FOR OPERATION 3.7 PLANT SYSTEMS 3.7.1 FEE 0 WATER SYSTEM. I 1

3.7.1 ine long term "E" steam gene-ator cooling system shall be maintaine: in an OPERAELE status. I i

APDL:CAE: L! Y: RECOVERY MOCE. ,

i ACTICN:

Witn the.long term cooling "B" steam generator cooling system inoperable, restore tne inoperacle. system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.  ;

I 3.~.2 SE:ONCaD* SERVICES CLOSED COOLING WATER SYSTEM

3. 7. C.1 A: least two seconca y services closed cooli'ng water system pumps and l nest ee:nangers an: the associatec flow patn snail be OPERAELE witv ea:n um; ca:a:'e cf ceing powerec from separate busses. g A8:LICAE:.ITY: RECOVERY MODE.

ACT:CN:

With only one se:encary services closed cooling water cum: or only one seconcary services neat ex: banger CPERAELE, restore the inoperable pum: or neat ex nanger to 03ERAE.E status witnin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3. 7. 3 CL:5E: CYCLE COOLING WATER SYSTEM NUCLEat SERVICE 5 CLOSED CYCLE COOLING SYSTEM 3.7.3.1 At least two independent nuclear services closed cycle cooling wate-pum:s anc neat ex:nangers and the associated flow. path shall ce OPERAELE with eacn pum: ca:able of being powered from separate emergency busses.

A:DLICAE: L:~v: RECOVERY MCOE.

...uN:

W't only one nuclear services closed cycle cooling water pum: or ori) cae  ;

nu:' ear servi:es neat exenanger OPERAELE, restore tne inoce acle pum: Or neat i ex:nanger to OPERAE.E status witnin-72 nours.

1 THREE MILE ISLAND - UNIT 2 3.7-1 i l

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A LIhIYINGCONDITIONSFOROPERATION

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'OE"AY-HEAT CLOSED COOLING WATER'5YSTEM-3.7.3.2 At least one decay heat closed cooling water. loop shall be OPERABLE.

A::u!" AE:'.:7v: RECOVERY MODE.

ACTION:

W'tn ne ce:a3 heat closec cooling water loop OPERABLE, restore,;he inoperacle loc: :: OPERAE.E status intin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

MIN

• DEC*v WEAT REMOVAL SYSTEM

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3.7,3.2 T*s mini decay neat removal pumps anc neat excnangers anc the asscci-a:e: ' c* :s:s sna11 be OPERABLE, A::.::AE:.: - When fuel is in tne reacter pressure vesse'

a. Wf n ene mini decay heat removal pum: and/or neat ex:r.ange* incoerable,

-res re the incoerable cum: and/or neat exchange- to OPERAELE status witnin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

O. w't.n two mini decay heat removal pumos anc/or heat excnangers o- the-associatec flow path inoce-a:1e, restore at least one pump anc hea; ex:. anger anc the associatec flew patn to CPERABLE status witr.in 2a nee s or witnin the next AE hours make a tacku: cec 14ng syste?.

(e :ner LTC "B" or " Loss to Am ient") OPERABLE, d

3.7.4 NJ:. EAR SERVICE R:VER WATER SYSTEM i

3.7.A 1 T*c independent nuclear service river water locos shall be OPE;AE'E. .

1 A":LICAEIL:TY: RECOVERY MCDE.

a tr o*e nucles" service rivea water' loco.inece acle, rest 0re ;ne ine:eaa:'e ic:: :: OPERAE.E status witnin 72 neurs.

THREE MILE ISLANO - UN:T2 3,7-2 i

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TABLE 3.8-2 RESTORATION TIME MATRIX ., .

Restore One Restore Other i Component Component (Hours) (Hours) .

_ aa 24 72

_5 .o. e ,

-s e - =

2 ~* s E ao 12 72

.:: c 5; e 5: E- E-j

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j! $ bb 12 72 a an: 0 above correspond to components descri:ed in Specifi:ation

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Note: ,

3.8.1.1 items a and o respectively.

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THREE MILE ISLAND - UNIT 2 3.8-3

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4 LIMITING CONDITIONS FOR OoERATION 4

2.S.2- CNSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION 3.9.2.1 The following A.C. electrical busses snali be CPERABLE and energi:ec with tie Dreakers q;en (unless closed in accordance with procedures.approvec pursuant to Scecification 5.5.2) between redundant busses:

4150 volt Emergency Bus # 2-1E and 2-3E A160 voit Emergency Sus # 2-2E anc 2-4E a;6' volt Busses # 2-3 and 2-4 450- volt Emergency Bus # 2-11E, 2-12E anc 2-3'.E ABC velt Eme gency Bus # 2-21E, 2-22E anc 2-41E 450 volt Susses # 2-31, 2-41. 2-32, 2-12, 2-25, 2-35, 2-4E,'anc 2-46 l 120 volt A.C. Vital Bus e 2-1V 12 volt A.C. Vitai Bus # 2-2V 120 voit A.C. Vital Bus # 2-3V

'2C

_ velt A.C. Vital Bus # 2-4'.

. g-n .e . .e g:s- s.

AC*:0N- .

Wits.less tnan the above com:lement of A.C. busses OPERAE. E, restore tne

. incoera:1e Ous to OPERAE'.E status witnin 8 nours.

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THREE MILE ISLAND - UNIT 2 3.8-4

3/4.4 REACTOR COOLANT SYSTEM BASES .

3/4.4.1 REACTOR COOLANT LOOPS Several alternative methods are available for removal of. reactor decay heat. These methods include use of the Mini Decay Heat Removal. System, the

" Loss to Amoient" cooling mode, and operation of the Reactor Coolant System in tne natural circulation mode with heat rejection via the long term "5" steam generator cooling mode. Any one of these cooling methods provides adecuate cooling of the reactor and each method is available for cecay (

heat removal. Procedures have been prepared and approved for use of tnese various cooling methods.

3/4.4.3 SAFETY VALVES ine pressurizer code safety valves operate to prevent the RC5 from being pressuri:ec acove its Safety Limit of 27E0 psig. Eacn safety valve is cesignec to relieve 348,072 lbs per hour of saturatec steam at tne valve's setpoint.

3/4.a.9 PRESSURE /TEMDERATURE LIMIT ine RCS pressure and temperature will be controlled in accorcance with accrevec procedures to prevent a nonductile failure of the RCS wnile at the same time permitting the RCS pressure to be maintained at a sufficiently hign value to :ermit operation of the reactor coolant pumps.

1 THREE MILE ISLAND - UNIT 2 B 3/4 4-1

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• l 3/4.7 : PLANT SYSTEMS.

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BASES -

3/4.7.1- FEE 0 WATER SYSTEM

'The long term "B" steam generator cooling system is required.to be main-tained in an OPERABLE status since.it is an alternative method for removing  :

cecay heat from the reactor coolant system, 3/4.7.2 SECONDARY SERVICES CLOSED COOLING WATER SYSTEM

-Tne secondary services closed cooling water system is required to be maintained in an OPERABLE condition since it is usec to cool the "B" steam.

generator closed loop cooling system.

3/a.7.3 CLOSED CYCLE COOLING WATER SYSTEM 3/a.7.3.1 NUCLEAR SERVICES CLOSE0 CYCLE COOLING SYSTEM OPERABILITY of the nuclear services closed cycle cooling system is recuired during operation of the MOHRS since this. system orovides the-heat sinx for the M0 HRS.

!3/4.7.3.2 OECAY HEAT CLOSED COOLING WATER' SYSTEM Ine decay heat closed cooling water system is recuired to ce maintained in an OPERABLE status since it fs provided to remove heat.from tne OHR system _

, wnien is being maintained OPERABLE in'a backup status for pessible core cooling.

3/4.7.3.3 MINI DECAY HEAT REMOVAL SYSTEM (MOHRS)

OPERABILITY of the MOHRS is required since it is an alternative method for removing decay heat from the reactor. The MOKRS is provided with two pumos anc two heat exchangers; one pumc and one heat excnanger have adecuate ca:acity for removing the present level of decay heat from the core. '

3/4.7.4 NUCLEAR SERVICE RIVER WATER SYSTEM The nuclear service river water system uses river water to cool tne nuclear services closed cycle cooling system, the seconeary services ciesec cooling water system, and decay heat closec cociing water system; nerefore, it must be OPERABLE too. This system rejects its heat to the river as tne ultimate heat sink.

THREE MILE ISLANO UNIT 2 8 3/4 7-1

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B CONTAINWENT SYSTEMS BASES-3/4.7.6 FLOOD PROTECTION The limitation on flood protection ensures that facility protective actions ~ will be taken in the event of flood conditions. The limit of elevation 302 Mean Sea Level is based on the maximum elevation at which facility flood control measures provice protection to safety related ecuipment.

3/4.7.7 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM The OPERABILITY of the control room emergency air cleanup system ensures that 1)_the amOient air temperature coes not exceec the allowable temperature for continuous duty rating for the equipment and instrumentation coelec Dy this system anc 2) the control room will remain haDitable for operations pe-sonnel curing anc following all crecible accident concisions. The OPERAB!LITY of this system in conjunction with control room design provisions is basec on limiting tne raciation exposure te personnel occupying the control room to 5 rem or less .nole body, or its equivalent. This limitation is consistent witn the recuirements ef General Design Criterion 19 of Appendix "A",10 CFR 50.

3/4.7.10 FIRE SUPPRESSION SYSTEMS i

The OPERASILITY of the fire suceression systems ensures thrt adecuate l fire suppression cacability is available to confine and extinguish fires occuring in any portion of the facili*y where safety related ecuipment is l

locatec. The fire suppression system consists of the water system, spray i anc/or scrinKlers, Halen and fire hose stations. The collective capacili y of l the fire s'uceression systems is adequate to minimi:e potential damage to safety related equipment and is a mafer element in the facility fire protec-tion program. Any two of the four main fire pumps provice com:inec capaci*y greater tnan 3575 gpm.

In tne event that portions of the fire suopression systems are inoperable, alternate backup fire fighting equipment is required to be mace available in the af fectec areas until the affected equipment can be restored to service.

In the event tnat tne fire suppression water system becomes inoceracle, immediate corrective measures must be taken since this system provices tne major fire suceression capability of the plant. The recuirement for a Scecial Report to the Commission provides for timely evaluation of the acce:tacility of the corrective measures to provice adecuate fire suopression ca:acility for the continued operati0n of the nuclear plant.

THREE MILE 15LANO - U'iIT 2 8 3/4 7-2

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. Replace the following pages of the Recovery Operations Plan with the enclosed pages, ,

as.incicated. The revised pages contain vertical lines indicating,the area of enange. The corresponding overleaf pages are also provided to maintain document completeness.

Paces 4.1-1 4.1-2 4.7-1 4.7-2 i

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THREE MILE ISLAND - UNIT 2 i

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SURVEILLANCE REOUIREMENTS 4.1 WATER INJECTION COOLING AND REACTIVITY CONTROL SYSTEMS 1

4.1.1 BORATION CONTROL I BORON INJECTION -

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4.1.1.1 Two systems capable of injecting borated cooling water into the l Reactor Coolant System sna11 be demonstrated CPERASLE: '

a. Deletec.

b. At least once per 31 days by verifying that each accessible (per occucational exposure considerations) valve (manual, power operated or automatic) in eacn flow path that is not locked, sealed, or otnerwise secured in position, is in its correct position.

1

c. At least once per 31 days (wnen makeuo pump is recuirec OPERABLE) by l verifying (per occupational exposure consicerations), that on recircu- l lation flow, tne makeup pump required by Technical Specification  !

3.1.1.1 develops a discharge pressure of greater than or equal to I 1125 psig and that each pump operates for at least 15 minutes. l l

d. At least once per 31 days by wifying (per occupational exposure 3 considerations), that on recirculation flow, the decay heat removal l pump required by Technical Specification 3.1.1.1 develoos a distnarge i cressure of. greater than or equal to 151 psig and tna eacn pumo I operates for at least 15 minutes.

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e. Deleted. - '

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f. At least once per 7 days when valve OH-V1 or OH-V171 is open oy verifying tnat the makeup pump electrical power supoly circuit breakers are " racked out."

g. At least once per 7 days by: '

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1. Deleted.  !

2. verifying tne coron concentration in the SwST is between 3000 ,

and 4500 ppm.

3. Deleted.

4. Verifying the contained borated water volume of the SWST is at least 100,000 gallons.  :

5. Deletec.

THREE MILE ISLAND - UNIT 2 4.1-1 Change No. 4

f.

SURVEILLANCE REOUIREMENTS

' BORON INJECTION (Continued)-

h. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying tne BWST temperature is at least 50*F when the outside air temperature is less than 50*F.-

i. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (when system is in operation) by l verifying.that the standby reactor coolant system pressure control system:

1 Surge tank water volume is filled to between 55% and 80% of ';

tank capacity and the tank is pressurized to tne operating RCS pressure : 25 psig but not higher than 600 psig. '

2. Isolation valves on the discharge side of the water filled tank nearest the reactor coolant system are open.

^

.3. The in-service nitrogen supply bank is pressuri:ed to betweer 225 anc 400 psi,.

f. At least once per 7 days by verifying that the standby reactor l coolant syste a pressure control system watet filled tanks, the surge '

tank, and the degassed water supply tank contain borated water witni

. A boren concentration of petween 3000 and 4500 ppm, ,

2. A disselved gas concentration of less than 15 sec/kg of water.

(. At least once per 31 days by verifying that the standby reactor j coolant systemetressure contaol system isolation valve on the cis-  ;

charge side of the water filled tank nearest the reactor coolan*

system closes automatically on a tank low level test signal.  !

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THREE MILE ISLAND - UNIT 2 4.1-2 Change No. 4 l

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SURVEILLANCE REOUIREMENTS l

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4.7 PLANT SYSTEMS (

4.7.1 FEEDWATER SYSTEM 4.7.1.1 Deleted.

4.7.1.2' Deleted.

4.7.1.3 --The "B". steam generator closed loop cooling system shall be demons-tratec OPERABLE at'least once per 31. cays by starting (unless alreacy operating) the pump and verifying a flow rate of at least 2000 gpm when operating in the recirculation mode.

4.7.2 SECONDARY SERVICES CLOSED COOLING WATER SYSTEM j 4.7.2.1 ine secondary services closec cooling water system sna11 be demons-trated OPERABLE at least once per 31-days by verifying snat eacn of tne tnree pumps start and operate (unless already operating) for at least 15 minutes.

4.7.3 CLOSEC CYCLE COOLING WATER SYSTEM l

NUCLEAR SERVICES CLOSED CYCLE COOLING SYSTEM I i

4.7.3.1 Eacn nuclear services closed cycle cooling water loop snail be oemons . I tratec OPERABLE:

a. At least once per 31 days by:

1-. Verifying that each pumo starts and operates (unless alreacy operating) for at least 15 minutes and tnat during pumo operation:

the "A" pump develops'a differential pressure of at least 62.1 psic, the "B" pump develops a differential pressure of at least.

63.1 psid, and the "C" pump develops a differential pressure of at least 64.1 psic.

2. Verifyi,ng that each accessible (per occupational ex;osure considerations) valve (manual, power operatec or automatic) servicing safety relatec ecuipment that is not lockec, sea'et or otnerwise secured in position, is in its correct position.

b. At least once per 92 cays by cycling each testacle valve in'tne fic.

path tnrough at least one complete cycle of full travel.

THREE MILE ISLAND - UNIT 2 4.7-1 Change No. 4

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$URVEILLANCE REOUIREMENTS I

. l DECAY HEAT CLO5ED COOLING WATER SYSTEM (

)

A.7.3.2 The ' decay neat closed cooling water loop required by Technical l Soecification 3.7.3.2 shall ce cemenstratec CPERABLE:

a. At least once per 31 cays by: )

1. Verifying the required cumo starts and operates (unless already operating) for at least 15 minutes and that during camp operction:

(a) the "A" pump develo:s a differential pressure of at least 30.6 psic at a flow of 249; gpm and (b) the "B" pump develops a differential pressure of at least 30.6 psic at a flow of 2527 gpm.

2. Verifying that each accessible (per occupational exposure consicerations) valve (manual, power operated or automatic) servicing safety related equipment tnat is not locked, sealed or otherwise securec in posi tien, is in its c:rrect position.

. At least once per 92 days by cycling each testable valve in tne flew patn througn at least one complete cycle of full travel.

M!NI DECAY HEAT REHCVAL SYS7EM /40HDS) 4.7.3.3 The MCHRS shall be demonstratec OPERABLE by performing inservice

-tests cf eacn WCHR$ pump and eacn MDnRS valve in tne flew path in accercance with Section x: of the ASME Soiler and Dressure Vessel Code and apelicacie Accenca as recui ed cy 10 CFR 50, Section 50.55a(g), except where doecific written relief has een granted by the Ccmmission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

THREE MILE ISLAND - UNIT 2 4.7-2 Change No. A

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./ -:u - i UNITED STATES OF AMERICA . I

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.- . NUCLEAR REGULATORY..COMMISSIQN \

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=.;q. . . :. . . ,

.T... In the Matter of ' -

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?.,:.. METROPOLITAN ~ EDISON COMPANY, )* . Docket No. (s) 5 0- 3 2 0 -OLA ,.,.

37 Ag, - ~ . - *

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CERTIFICATE '0F SERVICE

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i I hereby certify that I bIve this day served the forego 9g docu=ent(s) upon

, each person designated on the official service list co=ptied by the Office l

- of the Secretary of the Cc-4ssion in this proceeding in accordance with the require =ents of Section 2.712 of 10 CFR Part 2 - Rules of Practice, of the

' Nucle.ar Regulatory rc,.cnission's Rules and Regulations.

Dated at Washington, D.C. this

[ day of bM 19 8 / .

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h. @.

Office of the Sderetary of the Co ission .

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! UNITED STATES OF AMERICA  !

NUCLEAR-REGULATORY COMMISSION In the Matter of )'

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METROPOLITAN EDISON COMPANY, ET AL. ) Dccket No.(s) 50-320-OLA

) .

(Three Mile Island, Unit 2) ,

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a SERVICE LIST' s

, John F. Wolf,.Esq. George F. Trowbridge, Esq.

3409 Shepherd Street Shaw, Pittman, Potts & Torvbridge Chevy Chase, Maryland 20015 1800 M Street, N.W.

Washington, D.C. 20036 Dr. Oscar H. Paris Atomic Safety and Liceasing Board Mr. Steven C. Sholly U.S. Nuclear Regulatory Commission 304 South Market Street Washington, D.C. 20555 Mechar.icsburg, Pennsylvania 17055 Mr. Frederick J. Shon Mr. William A. Lochstet Atomic Safety and Licensing Board 119 East Aaron Drive U.S. Nuclear Regulatory Commission State College, Pennsylvania 16801 Washington, D.C. 20555 Dr. Judith H. Johnsrud Counsel for NRC Staff Environmental Coalition on Nuclear Power.

Office of the Executive Legal Director 433 Orlando Avenue U.S. Nuclear Regulatory Commission State College, Pennsylvania 16801 Washington, D.C. 20555 Karin W. Carter, Esq.

Metropolitan Edison Company Assistant Attorney General ATTN: Mr. J.B. Herbein Dept. of Environmental Resources Vice President P.O. Box 2357 P.O. Box 542 Harrisburg, Pennsylvania 17120 Reading, Pennsylvania 19603 O

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