Text

Amend 57 to License NPF-3,modifying Specs to Provide Addl Protection to Minimize Potential for Low Temp Overpressure Transients
	ML20023C852
Person / Time
Site:	Davis Besse
Issue date:	05/05/1983
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Toledo Edison Co, Cleveland Electric Illuminating Co
Shared Package
ML20023C853	List: ML20023C852; ML20023C862; ML20023C877;
References
	NPF-03-A-057 NUDOCS 8305180114
	Download: ML20023C852 (10)
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, UNITED STATES 3

P' NUCLEAR REGULATORY COMMISSION n

- [a wasmNGTON, D. C. 20655

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THE TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNI'T NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment N6. 57 License No. NPF-3 1.

The Nuclear Regulatory Comission (the Commission) has found that:

A.

The application for amendment by the Toledo Edison Company and The Cleveland Electric Illuminating Cogany (the licensees) dated December 26, 1980, coglies with the standards and require-men'ts of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such ac.tivities will be conducted in compliance with the Comi'ssion's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

e 8305180114 830505 PDR ADOCK 05000346 P

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2.

Accordingly, Facility Operating License flo. flPF-3 is hereby amended as indicated below and by changes to the Technical Specifications as indicated in the attachment to this license amendment:

Revise paragraph 2.C.(2) to read as follows:

O Technical Specifications

~

The Technical Specificati6ns contained in Appendices A and 8, as revised through Amendment No.57, are hereby incorporated in the license.

The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its iss uar.ce.

FOR.THE NUCLEAR REGULATORY COMMISSION V

4 bl( / '

J6 F. Stolz, Chief, (Op' vision oferating Reactors Branch #4

. 'Di Licensing

Attachment:

Changes to the Technical Specifications Date of issuance:

May 5,1983 e

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jfTACHMENT TO LICENSE AMENDMENT NO. 57 FACILITY OPERATING' LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with.the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

Pages 3/4 4-3 3/4 5-6 3/4 4-4a (new page) 3/4 4-4b (new page)

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' REACTOR COOLANT SYSTE SAFETT VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 Decay Heat Removal System relief va2.ve,DH-4849 shall be OPERA 31J with a lift. setting of ( 330 PSIG* and isolation valve's DE-11 and DE-12 open and control power to their valve operators removed.

APPLICABILITY: MODES 4 and 5.

ACTION:

C' A.

With DH-4849 not OPERABLE:

s 1.

Maka the valve OPERABLE within eight hours; or-2.

a.

Within next one hour, disable the capability of both high pressure injectio'n (EPI) pumps to inject water into the l

reactor coolant system; and b.

Within next eight hours:

1.

Disable the automatio transfer of makeup pump ' suction to the borated wate' _ storage tank on low makeup tank -

r level; and i

2.

Reducemakeuptank'levefto473inchesandreduce reactor coolant system pressure and pressurizar level within the acceptable reg!.on on Figures 3.4.2-a (in MODE 4)and 3.4.2-b (in MODE'S).

3.

With DH-11 or DE-12 closed, open DE-21 and DE-23 within one hour.

C.

With the control power not removed from DE-11 and DE-12, remove the power to the valve. operators at the Motor Control Centers within one hour.

SURVEILLANCE REQUIRDENT!i l

4.4.2.

Decay Heat Removal System relief valve CH-4849 shall be deter-mined OPERABLE:

a.,

per the' surveillance requirements of Specification 4.0.5..

b.

at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying either:

1.

isolation valves DE-11 and DB-12 open with control power removed from their valve operators; or 2.

valves DH-21 and DE-23 open.

c s

- The lif t setting pressure shall correspond to a=bient conditions of the. valve at nominal operating temperature and pressure.

l DAVIS-BESSE, UNIT 1 3/4 4-3 Amendment No.57

4 REACTOR COOLANT SYSTEM SAFETY. VALVE! AND ELECTROMATIC RELIEF VALVE - OPERATING 1

LIMITING CONDITION FOR OPERATION i

1 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2435 PSIG 1 15.*

When not isolated, the pressurizer land an allowable value ofelectromatic relief valve shall have a trip setpoint of 1 2385.5 PSIG.**

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APDLICABILITY: MODES 1, 2 and 3.

ACTION:

With one prssrurizar ccde safety valve incpe'rable, either restore the inocerable valve to OPE.VBLE status within 15 minutes or be in HOT Shut 00W within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 SURVEILLANCE RECUIREMENTS 4.4.3 For the pressurizer code safety valves, there are no additional Surveillance Requirements other than those required by Specification 4.0.5.

For the pressurizer electrematic relief valve a channel cali-i bration check shall be performed every 18 months.

The l1ft setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

." Albable /alue for channel calibration check.

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, DAVIS-BESSEoUNIT 1: 17 M4Lf66

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T;Y_ < 280'F b

LIMITING CORDITION FOR OPERATION 3.5.3 As a minimum, one ECC5 subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE decay heat (DH) pump, b.

One 0.PERABLE DH cooler, and c.

An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) and manually transferring suction to the containment emergency sump during the recirculation phase of operation.

APPLICABILITY: MODE 4.

ACTION:

a.

With no ECCS subsystem OPERABLE because of the inoperability of the DH pump, the DH cooler or the flow path from the BWST, restore at least one ECCS subsystem to OPERABLE status within or.e heu or..rittain the Reacter Coclart System Tavg less than 280*F by use of alternate heat removal methods, b.

In the event the ECCS is actuated and injects water into the

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reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

5URVEILLANCE REQUIREMENTS 4.5.3~ The ECC5 subsystems shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.,

3 DAVI5-BESSE, UNIT 1 3/4 5-6 Amendment No.JWM 57

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EMERGENCY CORE COOLING SYSTEMS 1

BORATED WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The borated water storage tank (BWST) shall be OPERABLE with:

a.

A contained bora'ted water volume of between 482,778 and l

550,000 gallons, b.

Between 1800 and 2200 ppm of boron, and c.

A minimum water temperature of 35'F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the bo' rated water storage tank inoperable, restore the tank to L

OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.5.4 The BWST shall be demonstrated OPERABLE:

I a.

At least once per 7 days by:

1.

Verifying the contained borated. water volume in the tank, 2.

Verifying the boron concentration of the water.

Atleastonceper24hoursbyverifyingthewaterteSerature b.

when outside air temperature <35'F.

L DAVIS-BESSE, UNIT 1 3/4 5-7 Amendnent No. 36

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3/4.4 REACTOR CCOLA'.*T SYSTEM SASES 3/4.4.1 REACTOR COCLANT LOOPS The plant is designed to operate with both reactor ceolant loops in operation, and maintain DN3R above 1.30 during all nomal operations and ar.ticipated transients. With one reactor coolant pump not in operation in one loop, TEEFF.AL POWER is restricted by the Nuclear Overpower Based on RCS 71ov and AIIAL POWER IMBALANCE, ensuring that the DNBR will be maintained above 1.30 at the maximum possible THERMAL POWER for the number of reactor coolant pumps in operation or the local quality at the point of inimum DNBR equal to 222, whichever is more restrictive.

i In MODES 3, 4 had 5, a single reactor coolant loop or DER loop provides sufficient heat renoval capability for removing decay heat; but singis l

failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires two DER loops to be OPERABLE.

Natural circulation flow or the operation of one DER pump provides

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adequate flov'to ensure r.ixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated wi'th boren reduction vill, therefore, be within the capacity of operator recognition and control.

3/A.4.2 an: 3/4.4.3 SAFETf VALVES The ;tessuri:er c:de safety valves cperate to prevent the RCS from teing ;ressuri: d a:cve its Safety Limit of 2750 psig. Each safety valve is casignec :o relieve 336,000 lbs per hour of saturated steam at the valve's set:cin:.

The relief capacity of a single safe:y valve is adequatn to relieve any everpressure c:nditten whien could cc:ur during shut:own.

In the event that ne safety valves are OPERABLE, an c;erating CHR loop, c:n-ne::ed :: the RCS, pr:vides overpressure relief' capability and will

reven: RCS cverpressuri:ation.

Ouring ::eration, all pressuri:er code safety valves must be CPERABLE

preven the RCS from being pressuri:ed above its safety limit of 2750 psig. The c:msined relief capacity of all of these valves is greater than the maximum surge rate resulting 'from any tran'sient.

The relief capacity of the decay heat removal system relief valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that this relief valve is not OPERABLE, reactor.

coolant system pressure, pressurizer level and.make up water inventory is limited and the capability of the high pressure injection system to ir. ject water into the reactor coolant system is disabled to ensure l

o:eratior within reactor coolant system pressure - temperature limits.

l Demonstra:icn of the safety valves' lift se::ings wi,11 cc:ur only curing snute:wn and will be perfcrmed in ac:criance with the provisions cf See:ica XI of :3e ASME 5cilar and Pressure Code.

4 l

B 3/4 4-1 Amendment No. JJ,fE, 57 DA7IS-BESSE UNIT 1

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