ML19290A349

Amend 12 to DPR-50,clarifying Tech Specs Re Reactor Sys Pressure Safety Limits

ML19290A349
Person / Time
Site:	Crane
Issue date:	02/17/1976
From:	Reid R Office of Nuclear Reactor Regulation
To:
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ML19290A348	List: ML19290A347 ML19290A349 ML19290A350
References
NUDOCS 7911060586
Download: ML19290A349 (9)

Text

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.. m Anendment No. 12 License No. OPR-50 1.

Tne Nuclear Regulatory Corcissicn (the Cor-.issien) has fcund that:

A.

The application for a endcent by "ctrcyclitan Edison C:rcany, Jersev. Central Fcxer and h:.n: to:perf, and Tennsylvenia Elcetric Cc Jany (the licensee) da:cd ':c;cnber 17, 10 5, complies with the standards and recnircrents of the Atcr.ic o

Fnergy Act cf 10,E ;, as a.cnded (the Act),. si the Cu:: :ission's rules and regulations set forth in 10 CF2 Chapter I; 3.

The facility will cperate in ecnferni:y.i:S the application, the provisions of the Act, and the rules 2nd regulaticas of the Cercissien; C.

Tacre is reasonable assurance (i) th: the activities authori:cd by this r.endmen: can be cendue:ed withcut endan;ering :he health and safety of the public, and (ii) that such activities will be ccnducted in conpliance with Cercissien's rer.:1ations; D.

The issuance of this atendment will nc: be ini,ical to the corren defense and security or to the health 2nd safety cf the public; and E.

An ensironmental statenent or negative declaration need not be prepared in ccnnection with the issuance cf this amend.nent.

1556 074 POCR (Rum..

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Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment.

3.

This license anendment is effective as of the date of its issuance.

FOR THE NUCLEAR RECULUCRY CC.'O:ISSIC'N m

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Robert W. Reid, Chief Cperating Reactors Eranch f4 Division of Cperating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

February 17, 1976 1556 075

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n ATTACINENT TO LICENSE MIEND:ENT NO. 12 FACILITY CPERATING LICENSE No. DPR-50 DCCKET NO. 50-239 Replace pages 2-3, 2-4, 3-1, 3-2, 4-7, and 4-3 of the Appendix A Teci: ical Specifications 1:ith the attached pages. Changes on these pages are shown by carginal lines.

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... e The maxi =u ther.al pcVer for other reactor ecclant pu=p conditions is pro-duced in a si ilar =anner.

For each curve of Figure 2.1-3, a pressure-te perature point above and. to the left of the curve vould result in a 0:T3R g-eate "a

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in Red Sundles Sy=pesiu=:

(a) 'Jilson, et. al., Critical Heat Flux in Ncn-Unifer= Heater Red SunAles.

(b) Gellerstedt, et. al., Cerrelation of a Critical Heat Flu.x in a Sundle Ccoled by Fressurized 'n'ater.

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Acclicahility Applies to the limit en reacter ecolant systes pressure.

Cb3ective To =aintain the integrity of the reactor ecolant syste= and to prevent the release of significant encunts of fissien product activity.

Scecification 2.2.1 The reacter ecclant systen pressure shall net exceed 2750 psig when there are fuel assenblies in the reacter vessel.

Bases The reactor coolant system (1) serves as a barrier to prevent radienuclides in the reacter ecclant fres reaching the at csphere.

In the event of a fuel cladding failure, the reactor ceclant syste= is a barrier against the release Of fission products. Establishing a systen press"-= ~i-helps to assure the integrity of the reacter ecclant system. The maxi =u= transient pressure allevable in the reacter ecolant system pressury essel under the ASSE Code,Section III, is 110 percent of design pressure.(2 The maxi =un transient pressure allovable in the reactor coolant system piping, valves, and fittings under A:iSI Section 331.7 is llo percent of design pressure. Thus, the safety Ibnit of 275{,I g (110 percent of the 25C0 psig design pressure) has heen si and the pressurizer ccde safety valves (2h35 psis)(t)" essure trip (2355 psig) established. '

The settings for the reacter high 2 have been established in accordance with ASME 2ciler and Pressurizer Vessel Code,Section III, Article 0, 'Jinter,1968 to assure that the reactor ecclant systen pressure safety

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'* *s not exceeded. The initial hydrostatic test was cenducted at 3125 psig (125 percent of design pressure) to verify the integrity of the reacter ecclant system. Additional assurance that the reactor ecclant systen pressure dces not exceed the safety linit is provided by settir:g the pressuriner electrecatic relief valve at 2255 psig.(h)

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FSAR, Section k (2)

FSAR, Section h.3.10.1 1556 078 (3)

FSAR, Section 4.2.u (h)

FSAR, Table L-1

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.a-3 1.1 CPERATIC:iAL CCMPC:iE'iTS Acclicabilitf Applies to the operating status of reacter c clant syste= conpenents.

Objective To specify those li=iting conditiens for cperation of reactor c0clant syste=

cc=ponents which =ust be =et to_ ensure safe reactor operati:ns.

S:ecificatien 3.1.1.1 Reacter Coolant Pu=ps Pump ce=binations per=issible for given power levels shall a.

be as shcvn in Specification Table 2.3-1.

b.

The boren concentratien in the reactor ecolant syste=

shall not be reduced unless at least one reacter ecciant pu=p or one decay heat re= oval pu=p is circulating reactor coolant.

3.1.1.2 Stea= Generator One stea= generator shall be Operable whenever the reactor a.

ecclant average te=perature is above 250 7.

3.1.1.3 Pressurizer Safety Valves a.

The reactor shall not remain critical unless both pressurizer code safety valves are operable with a lift setting of 243S PSIG 21's.

I b.

When the reactor is suberitical, at least one pressuri:er code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III.

Bases

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A reactor coolant pump or decay heat re=cval pu=p is required to be in cperation before the borca concentration is reduced by dilution with =akeup water. Either pu=p vill provide =ixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. Cne decay

'a. eat re= oval pu=p vill circulate the equivalent of the reactor ecolant syste: volu=e in One half hour or less.,

1556 079 Amendment No. 12 3-1

The decay heat re=cval sys.e= 7uction piping is des;3..ed for 3C0 F and syste= is tel:v this temperature.lgvg) decay heat when the rea::::

370 psig; thus, the syste= can re ccolant Cne pressurizer code safety valve is capable of preventing overpressurizati:n when the reacter is not critical since its relieving capacity is greater than that required by the su= of the available heat sources which are pu=p energy,

pressurizer heaters, and rea:ter decay heat.(')

Seth pressurizer cede safety valves are required to be in service prior to criticality to confer = to the syste= design relief capabilities. The ecde safety valves prevent overpres-sure for a red withdrawal accident.(5)

."he pressurizer code safety valve lift set point shall be set at 2h35 psig :1 percent all:vance for error and each valve shall be capable of relieving 311,7CO lb/h of saturated stea at a pressure not greater than three percent above the set pressure.

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.SAR, Tables 9-10 and h-3 through h-7 (2)

FSAR, Secticns L.2 51 and 9 5 2.3 (3)

FSAR, Section h.2.5.h (k)

FS AR, Secticas h.3 10.h and L.2.h (5)

FS AR,' Se ction L. 3 7 1556 080 3-2

TABI.E 1.1-1 (Continued) 4 CilANNEI, DESCRII'fION CI!ECK TEST CAI.I DRATE REMARKS 38.

Stera Generator Water Iavel W

NA R

39 Turbine Overspeed Trip NA R

NA 14 0. Soditan Thiosulfate Tank Ievel NA NA R

Indicator 141.

Sodium Ilydroxide Tank Ievel NA NA R

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Indicator 12.

Diesel Generator Protective NA NA R

4 Relaying 13.

1 KV ES Bus Undervoltage Relays NA M(1)

R (1) llelay operation will be checked

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4 (Diesel Start) by local test pushbuttons.

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Reactor Coolant Pressure S(1)

M R

(1) When reactor coolant system is Dil Valve Interlock Bistable preusurized above 300 psig or Taves is greater than 2000F.

O 3 - Each Shift, T/W - Twice per week 11 - Each Re fueling Period D - Daily B/M -- Every 2 months NA - Not Applicable W - Weekly Q - Quarterly B/W - Every two weeks tn Ln D

M - 14mt.hly P - Prior to each startup if not, done previous week CD

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== ' 1-2 MI.'i! MUM EQUI?'EZ 1TST ??.EQUE"CY Item Test Frec_uen cy 1.

Control Rods Rod dr p ti=es of all Each refueling shutdevn Pall length rods 2.

Control Rod Fbvement of each red Eve:/ two veeks, when reactor Move =ent is critical 3

Pressurizer Safety Setpoint

• 50% each refueling period j

Valves Main Steam Safety Setpoint 25% each rePaeling period Valves 5.

RePaeling Syste=

Functional Start of each rePaeling period Inter 10cks 6.

Main Steam (See Section 4.8)

Isolation Valves 7

Renctor Ccolant Evaluate Daily, when reactor coolant System Leakage system te=perature is greater than 5250F 8.

Charecal and high DCP test en EEPA filters, Each refueling peried and at efficiency filters freon test on charecal any ti=e verk on filters for Contrcl Rec =,

filter units could alter their integrity and R3 Purge Filters 9

Spent Fuel Cooling Functional Each rePaeling period pricr to System fuel handling 10.

Intake Pu=p House (a) Silt Accu =ulatien-Each rePaeling period Floor Visual inspection of Intake (Elevatien 262 Ft Pu=p House Flcer 6 in.)

(b) Silt Accu =ulatica Quarterly Measure =ent of Fu=p Ecuse Flow 11.

Hydraulic Shock Inspecticn of Hydraulic Refaeling Interval Suppressors en Shcek Suppresscrs (Snubbers)

Safety Related en Safety Related Syste=s.

Syste=s

• The setpoint of the pressurizer code safety valves shall be in necordance with ASME Boiler and Pressurizer Vessel Code,Section III, Article 9, Winter,1968.

Amendment No. 12 h-8 1556 082