ML19283B787

Forwards Proposed Changes to Operating License NPF-2 Tech Specs Involving Overpressurization Mitigating Sys Spec.Ser Supporting Change Encl

ML19283B787
Person / Time
Site:	Farley
Issue date:	03/21/1979
From:	Clayton F ALABAMA POWER CO.
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 7903260230
Download: ML19283B787 (12)

Text

Natuma Power Comptny e

GW.om t e, S ree:

PE C*rce b 26:1 O rv719 A'MJ3%.d 36231 Turmr-2J5 323 5341 A

F. L CLAYTON, JR.

Alabama Power Sen.or Vce President the sou hem e! ctoc sys!cm L

March 21, 1979 Docket No. 50-348 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commissten Washington, D.C.

20555

~

Attn:

Mr. A. Schwencer

Dear Mr. Schwencer:

Re: Changes to Operating License No. NPF-2 Technical Specifications Alabama Power Company proposes the attached changes to Joseph M.

Farley Nuclear Plant Operating License No. NFF-2 Technical Specifications involving overpressurization mitigating system specifications in addition to those sent in my letter dated January 4, 1979.

The Plant Operations Review Committee and the Nuclear Operations Review Board have reviewed the proposed changes and have determined that the changes do not involve an unreviewed safety question.as shown in the attached safety evaluation.

The overpressuri7ation micigating system was required by the NRC in a letter from Mr. Jol a F. Stolz ta Mr. Alan R. Barton dated December 29, 1976. Since this we. prior to the Technical Specification change fee re-quirement in 10 CFR '70, no fee should be necessary.

In accordance with 10 CFR 50.30(c)(1)(i), three (3) signed originals and thirty-seven (37) additional copies of these proposed changes are en-closed.

If you have any questions, please advise.

Yours very truly, (2.c k '

F.

L. Clayton Jr.

SWORN TO AND SUBSCRIBED BEFORF, FLCJr/TNE:bhj ME THIS ll.sh)AY OF MARCH,1979.

Attachment h)ym

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iA ~

$NO y

cc:

Mr. R. A. Thomas NOIARYGPUBLIC Mr. G.

F. Trowbridge My Commission Expires: b. Ih 7003260 A30

SAFETY EVALUATION FOR REACTOR COOLANT SYSTEM OVERPRESSURIZATION MITIGATING SYSTEM PROPOSED TECHNICAL SPECIFICATIONS

Background:

In a letter dated December 29, 1976, the NRC stated that Alabama Power Company (APCo) must provide a long-term protection system for the Farley Nuclear Plant (FNP) against low temperature overpressurization of the Reactor Coolant System such that the pressure limitations imposed by Appendix G to 10CFR50 are not exceeded. APCo submitted its response to the NRC letter on September 6,1978.

This submittal proposed an Overpressuri 2ation Mitigating System (OMS) utilizing the two existing RHR suction line relief valves (8708A and 8708B).

In letters dated November 3, November 9, and November 17, 1978 APCo provided additional informa-tion.and responded to the subsequent NRC concerns.

By letter dated January 4,1979 APCo proposed technical specifications associated with the RHR relief valves.

Recent discussions with the '!RC have revealed that they are in the process of completing a review of the Farley Nuclear Plant's proposed OMS and that committing to additional technical specifications is imperative to a favorable review.

References:

(1)

NRC's letter to APCo dated December 29, 1976 (2) APCo's submittal to the NRC dated September 6,1978 (3) APCo's submittal to the NRC dated November 3,1978 (4) APCo's submittal to the NRC dated November 9,1978 y

(5) APCo's submittal to the NRC dated November 17, 1978 (6) APCo's submittal to the NRC dated January 4,1979 (7) Proposed Technical Specification 3.1.2.3 and 3.4.1 (8) Technical Specification Bases Sections 3/4.1.2, 3/4.4.1 Bases:

Attached is a copy of the additional Reactor Coolant System Overpressurization Mitigating System's proposed technical specifications for the Farley Nuclear Plant.

The NRC's standard technical specification format was appropriately modified commensurate with FNP's proposed OMS. The following provides the bases for the proposed technical specification changes.

The proposed OMS is required to be operational whenever the RCS temperature s 3100F. The 3100F temperature limitation was imposed by the NRC since the limiting Appendix G pressure for a 1000F/hr heatup rate (worst case) at 3100F RCS temperature corresponds to 2,500 psia and the RCS overpressure protection in this pressure vicinity (2,500 psia) is provided by the pressurizer safety relief valves.

Specification 3.1.2.3:

The proposed ONS for the Farley Nuclear Plant is designed to mitigate the consequences of overpressurization transients initiated by inadvertent mass and heat additions whenever the RCS temperature is 1 3100F.

Since the Appendix G pressure limit is a function of the RCS tempercture, the OMS design basis mass input transients considered corresponded to specific RCS tenperature bands.

Inadvertent operation of two ( ) or more charging pumps was considered as the design basis event between RCS temperature of 3100F and 1800F whereas the design basis event was the operation of a single charging pump for RCS temperatures below 180 F.

The peak RCS pressure resulting from the design basis mass input transients in relation to the liniting Appendix G pressure values as functions of RCS temperature is shown in Revision 2 of Figures 5 and 6 of Reference (2) which are attached.

The proposed Technical Specification 3.1.2.3 requires that whenever the RCS temperature is 1 180 F only one centrifugal charging pump be operatble and the other two charging pumps be rendered inoperable by removing power to their motor circuit breakers in order to assure remaining within bounds of the OFG design whenever the RCS temperature is 1 180 F.

The proposed Technical Specification also assures adequate protection against the consequences of a RHR suction line break in Mode 5 by allowing one operable charging pump.

This satisfies the Emergency Core Cooling and Reactivity Control System requiremen:s on the basis of the stable reactivity condition of the reactor and the limited core cooling requirement in accordance with the existing Technical Specification Bases for ECCS subsystems and the proposed bases for the Boration Systems in Sections 3/4. 5. 2, 3/4. 5. 3 and 3/4.1. 2.

Adequate protection ap,ninst the consequences of a RHR suction line break in Mode 4 is assured by the existing Technical Specification which requires at least two operable charging pumps in this mode of operation.

Reactor coolant level would be maintained above the cere at the hot leg level since 2* charging pumps would be available for manual starting in the control room when the low pressurizer level alarm was initiated.

Specification 3.4.1:

~

The worst heat input transient considered in the 0:iS design is the inadvertent operation of a single reactor coolant pump with a maximum of 50 F temperature differential between the RCS and the S.G.

In order to assure that the severity of the heat input transient does not exceed the OMS design basis heat input transient, the proposed technical specification imposes one of the two following conditions before a RCP is started when the RCS temperature is 13100F.

Con-dition one requies that the AT between the RCS and the S.G. be measured to ensure it is below 500F befroe a RCP is started.

The other condition requires that pressurizer water volume be less than 770 ft3 This provides a steam bubble in the pressurizer suf ficient to mitigate a heat input transient initiated by the operation of a RCP when a 100 F AT exists between the RCS and the S.G. without actuating the RHR relief valves.

==

Conclusion:==

The technical specifications for the RCS overpressurization protection do not involve an unreviewed safety question as defined by 10CFR50.59.

sev.

//

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s.

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION J

3.1.2.3 At least one charging pump in the baron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE c'.Tergency bus.

APPLICABILITY:

MODES 5*and 6.

ACTION:

With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive teactivity changes until one charging pump is restored to OPERABLE status.

'~

SURVEILLANCE REQUIREMENTS 4.1.2.3.1 At least the above required charging pump s' hall be demonstrated l

OPERABLE by verifying, that on recirculation flow, the pump develops a discharge pressure of > 2458 psig when tested pursuant to Specification 4.0.5.

4.1.2.3.2.All chargina pumps, except the above recuired OPERABLE pump, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verifying that the motor circuit breakers have been renoved from their electr.ical power supply circuits.

• A maximum of one centrifugal charging pump shall be OPERABLE whenever g

the temperature of one or more of the RCS cold legs is < 180 F.

'T __

s FARLEY - UNIT 1 3/4 1-9

m.__.

i REACTOR COOLA::T SYSTE!i ACT!0:1(Continued) b)

. Place the following reactor trip system and ESF instrumentation channcis, associated with the loop not in operation, in their tripped conditions:

1)

Overpower AT channel.

2)

Overtemperature AT channel.

3)

T

-- Low-Low channel used in the coinci-d8.De circuit with Steam Flow - High for Steam Line Isolation.

4)

Steam Line Pressure - Low channel used for Safety Injection.

5)

Steam Flow-High channel used for MSIV Isolation.

6)

Differential Pressure Between Steam Lines - High

(

channel used for Safety ' Injection (trip all i

bistables which indicate low active loop steam pressure with respect to -the idle loop steam pressure).

c)

Change the P-8 interlock setpoint from the value

-~

specified in Table 3.3-1 to 1 66% of PATED THEPJGL POWER.

2.

THERIML POWER is restricted to 1 61% of PALED THERIGL POWER.

JZ Below P-7:

With K 1.0, operation may proceed provided at least two a.

reactofboo>lantlocpsandassociatedpumpsareinoperation.

f b.

With K

< l.0, operation may proceed provided at least one reactof[oolantloopisinoperationwithanassociatedreactor" f

coolant or residual heat removal pump.*

I c.

.The provisions of Specifications 3.0.3 and 3.0.4 are not applicabl e.

'All reactor coolant pumps and residual heat removal pumps may l

be deenergized for up to one (1) hour provided no operations are perm.itted which could cause dilution pf reactor coolant system boron concentration.

cT l FAR'..EY - U:HT 1 3/4 4-2 I endment Ib.

7

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1. With one reactor coolant loop and associated pump not in operation, at least once per 31 days determine that:

a.

The applicable reactor trip system and/or ESF actuation system instrumentation channels specified in the ACTION statements above have been placed in their tripped conditions, and b.

The P-8 interlock setpoint i's within the following limits if the P-8 interlock was reset for 2 loop operation 1 66% of RATED THERMAL POWER.

1. A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures 2 3100F unless gp e S,^ g e g S % %*[

1) the pressurizer water volume is less than'770 cub 4e-4est or

2) the secondary water temperature of each steam generator is less than 500F above each of the RCS cold leg temperatures.

FARLEY - UNIT 1 3/4 4-3

/

v' REACTIVITY C0t1 TROL SYSTEMS BASES 3/4.1.2 BORATI0tl SYSTEMS (Continued)

With the RCS temperature below 200 F, one injection system is

~

acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATI0 tis and positive reactivity change in the event the single injection system becomes inoperable.

---

Insert

• The boron capability required below 200 F is sufficient to provide a SHUTDOWft MARGIt! of 1% ak/k af ter xenon decay and cooldown from 200"F to 140 F.

This co.idition, requires either 2000. gallons of 7000 ppm borated water.from the boric acid storage tanks or 9,000 gallons of 2000 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The OPERABILITY of one boron injection system during REFUELItiG ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE C0iiTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWri MARGIt1 is maintained, and (3) limit the potential effects of rod misalignment and associated accident analysis.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The ACTI0tl statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design' criteria are met.

Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER; either of these restrictions ~ ovide assurance of fuel rod integrity during continued operation.

I n a,dd '..

those accidents analyses affected by a misaligned rod are re-eve

_ced to confirm that the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with T

> 541 F and with all reactor coolant pumps operating ensures that the men 0 rid drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

.FARLEY - UtlIT 1 B 3/4 1-3

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 180 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single RHRRV.

0

?

(,

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal opera With one reactor coolant loop not in operation, THERMAL POWER is restricted to < 36 percent of RATED THE anticipated transients.

the Overtemperature AT trip is reset.A loss of. flow in two loops will cause a will be maintained above 1.30. reactor trip if operating above P-7 (11 while a loss of flow in one loop will cause a reactor trip if operating above P-8 (36 percent of RATED THERMAL POWER).

A single reactor coolant loop provides sufficient heat removal capability for removing core decay heat while in HO in the shutdown cooling mo'de if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

- ------ ---Insert

• 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from Each safety being pressurized above its Safety Limit of 2735 psig.lbs per hour of sa 345,000 valve is designed to relieveThe relief capacity of a single safety valve is the valve set point.

adequate to relieve any overpressure condition which co shutdown.

RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 the combined relief capacity of all of these valves is greater psig.

than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is 'taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves lift setting will occur only during shutdown and will be performed in accordance will the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

I e

D.

FARLEY-UNIT 1 8 3/4 4-1 mme

e The restrictions on starting a Reactor Coolant Pump below P-7 with one or more RCS cold legs 1 3100F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which would exceed the limits of Appendix G to 10CFR Part 50. The RCS will be protected against overpressurization transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and therd7j providing a volume for the primary coolant to expand into or (2) by restricting from starting the RCP's to when the secondary water temperature of each steam generator is less than 500F above each of the RCS cold leg temperatures. -

e