ML19327A397

Amend 49 to License DPR-40,changing Tech Specs to Remove Requirement to Cycle Nonredundant Pumps & Valves & Adding Requirement for Ensuring Auxiliary Feedwater Availability

ML19327A397
Person / Time
Site:	Fort Calhoun
Issue date:	07/25/1980
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML19327A395	List: ML19327A394 ML19327A397 ML19327A400 ML19327A402
References
NUDOCS 8008060141
Download: ML19327A397 (10)

Text

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UNITED STATES y

NUCLEAR REGULATORY COMMISSION g,

j WASHINGTON, D. C. 20585

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OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FO'RT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 49 License No. DPR-40 1.

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

A-The applications for amendment by Omaha Public Power District (the licensee) dated July 5,1979 and May 12, 1980, as. supplemented May 14, 1980 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the pmvisions.of the Act, and the rules and re,gulations of the Comission; 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) tnat such activities will be conducted in compliance with the Comission's regulation

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D.

The issuance of this amendment will not be inimical to the comon i

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 Comission's regulations and all applicable requirements have been satisfied.

8 00806 0/Y/

2-2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B... 'of Facility Operating License No. DPR-40 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendice~s.

A and B, as revised through Amendment No. 49,'are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSI0ft b

h Robert A.' Clark, Chief Operating Reactors Branch # 3 Division of Licensing -

Attachment:

Changes to the Technical Specifications Date of Issuance:

uly 25, 1980 D

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ATTACHMENT TO CICENSE AMENDMENT NO. do-

' FACILITY OPERATING LICENSE NO. OPR-40_

DOCKET NO. 50-285 Replace the folloking pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.,

Pages 2-51 2-23 2-23a*

2-25 2-28 3-62 3-62a (added)

• Overleaf page provided for your convenience.

2.0

!.IMITING CCNDITIONS FOR OPERATION 2.3 Imeraency Core Coolise System (Continu$d)

(2)

Modification of Misimum Reouirements During power cperation, the Fin 4m Requirements =ay be modi-fied to allev ene of the following conditions to be true at any ene time. If the system is not restored to meet the misi-mum requirements within the time period specified belov, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If the dn4-um requirements are not met within an addi.

tional k8 hours the reactor shall be placed in a cold shutdown condition vithis 2h hours.

a..

One low-pressure safety injection pu=p may be inoperable provided the pump is rest.ored to cperable status withd4 i

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I b.

0:e high-pressure safety injecti= ;u=p =ay be incperable provided the pu=p is restcred to cperable status withis 2k hours.

l c.

C e shutdevn heat exchanger and two cf four ecmpenent d

coo 11=g vater heat' exchangers =ay te incperable for a period of :o rare than 2h hcurs.

d.

A:y valves, interlocks er pipi:g directly associated with c e of the above ec=ponents a=d required to functics duri g accident ecediticus shall be deemed to be part of that ec=pe:ent and shall =ee: the same require =e: s as listed for th' t cesponent.

a A:7 valve,1:tericek' er pipi:g associated with the safety e.

i=jectics and shutdev: cooli g syste= vhich is net cevered u der d. above but which is required to fu ction during accident ec d'tions =ay be inoperable for a pericd of no ore ths: 2h hours.

l f.

One safety injection' tank may be inoperable for a period of no more than one hour.

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g.

Level and pressure instrumentatic: en one safety injectica tart =ay be inoperable for a period cf c=e hcur.

Amendment No. 49 2-21 w.m.

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2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Dnergency Core Cooling System (Continued) used for. shut down cooling, the valving vill be changed and must be properly aligned prior to start-up of the reactor.

i The operable status of the various systems and co=penents is to be demonstrated by periodic tests. A large fraction of these tests vill be perfomed while the reactor is operating 1

in the power range.

i If a component is found to be inoperable, it irill be possible.

in most cases to effect repairs and restore the system to full i

operability within a relatively short time. For a single com-pouent to be inoperable does not negate the ability.cf the system to perforn its function. If it develops that t,he. inoperable component is not repaired within the specified allevable tine period, or a second cc=penent in the same or related system is found to be inoperable, the reactor vill initially be put in the hot shutdown condi-i tien to provide for reduction of cooling requirements after a postulated loss-of-coolant accident. This vill also permit i= proved access for repairs in some cases. After a limited time in hot shutdown, if the malfunctien(s) is not corrected, the reactor vill be placed in the cold shutdova condition utilizing ner=al shutdown and cooldovn procedures.

If the cold shutdown condition, release of fission products or j

~ da= age of the fuel elements is not considered possible.

I The plant operating procedures vill require i= mediate action l

ta effect repairs of an inoperable ec=penent and therefore l

in =ost cases repairs vill be ccepleted in less -'- " "-

specified allevable repair times.

The limiting ti=es to re-

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pair are intended to assure that operability of the ccaponent vill be restored pro =ptly and yet allev sufficient time to effect repairs using safe and proper procedures.

The requirement for core cooling in case of postulated loss-of-coolant accident while in the hot shutdown condition is significantly reduced below the requirements for a postulated

loss-of-coolant accident during power operation. Putting the /

reactor in the hot shutdovn condition reduces the ccusequences of a loss-of-coolant accident and also allows more free access to some of the engineered safeguards components in order to effact repairs.

Failure to complete repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to the hot shutdown condition is considered indiri t.Ava of a require-ment for major =aintenance and, thereft i, A retch a case, the reactor is to be put into the ed J !gu+(ovn condition. '

a Amend =ent No. K X, jf, 49 g_g3

2.0 LIMITING CONDITIONS FOR OPERATION 8

2.3 Emergency Core Cooling Syste= (Continued)

With respect to the core cooling function, there $s *ipnctional

• redunfincy over = cst of the range of break air.es.t3) s)

The LOCA analysis confirms adequate core cooling for the break spectru up to and including the 32 inch double-ended break assuming the safety injection capability which most adversely affects accident consequences and are defined as follows. The entire contents of all four safety injection tanks are assumed to be available for e=ergency core cooling, but the contents of one of the tanks is assu=ed to be lost through the reactor coolant syste=.

In addition, of the three high-pressurei ssfety injection pu=ps and tne two lov-pressure safety injection pu=ps, for large break analysis it is assu=ed that two high pressure and one low pressure operate while only one of each type is assu=ed to operate in the small break analysis (5); and also that 25% of their cc=bined discharge rate is lost from the reactor coolant system out of the break. The transient hot spot fuel cle.1 tenperatures for the break sines considered a.re shown on FSAR Figures 1-19 (A=end=ent No. 3h).

Inadvertent actuation of three (3) EPSI pu=ps and three (3) charging pu=ps, coincident with the opening of one of the two PORV's, vould result in a peak pri=ary syste= pressure of 1190 psia.

1190 psia corresponds vith a mini =u= per=issible te=perature of 320cF on Figure 2-13.

Thus, at least one HPSI punp is disabled at 3200F.

Inadvertent actuation of two (2) EPSI pu=ps and three (3) charging pu=ps, coincident vith the cpening of cne of the two PORV's, vould result in a peak primary syste= pressure of 10!+0 psia. 10h0 psia corresponds vith a mini =u= permissible te=perature of 310cF on Figure 2-13.

.Thus, at least two HPSI pumps vill be disabled at 310cF.

Inadvertent actuation of cne (1) EPSI and three (3) charging pu=ps, coincident with opening of one of the two PORV's, vould

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result in a peak primary syste= pressure of 685 psia.

685 psia corresponds with a mini =us allovable te=perature of 2760F on Figure 2-13.

Thus, all three EPSI pu=ps vill be disabled at, 2760F.

Inadvertent actuation of three (3) charging pu.sps, coincident with opening of one of the two PORV's, vould result in a peak -

prisaiy syste= pressure of 160 psia. 160 psia corresponds with a =inimum allevable te=perature of 780F '(approximately the boltup te=perature of 820F) on Figure 2-13.

Thus, dis-abling of the charging pu=ps is not required.

Re= oval of the reactor vessel head, one pressurirer safety

~ valve, or one PORV provides sufficient expansion volu=e to limit any of the design basis pressure transients. Thus, no additional relief capacity is required.

Amend =en No. 39, 47 2-23a

2. O LIMITING CONDITIONS FOR OPERATICN 2.k Containment Cooling (Continued)

During power operation one of the ec=pements listed above (in tidition to one rav vater pu=p) =ay be incperable. If the. inoperable co=ponent is not restored to operability vithin seven days, the reactor shall be placed in a het shutdown conditien within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the ineper-able ec=penent is not restored to operability vithin an additional k8 hours, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(2)

Modificatics of Minimu= Recuirements During pcwer operation, the

~d 4- ' requirements may be =cdified to allow a total of two of the ec=ponents listed in a. and b. to be incperable at any one time (in additica to one rav vater pu=p) provided that the e=ergency diesel-generator connected to the other engineered safeguards h.16-kV bus (1A4 or 1A3) is started to de=custrate cperability.

Only tvo rav vater pt=:ps may be out of service.

If the cperability of both ec=penents is not l

restored within 2h hours, the reacter shall be placed in a hot shutdevn condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the operability of both ec=penents is not restored within an additicnal kB bours, the reactor shall be placed in a ecid shutd. min ecndition vithin 2k hours.

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Any valves, interlocks and piping directly asscciated with ene of the above ec=penents and required to functics during accident eenditicas sha be deened to be pa.rt of that ec=penent and sk='1

-- ' - sane require =ents e.s.fer that cenpenent.

Any valve, interlock or piping associated vith the cent ='--ant cooling systen which is not included in the above paragraph and which is required to function during accident cenditions =ay be inoperable for a period of no =cre than 2k hours. If cperatica g

is not restored within 2h. hours, the reactor sF*"

be placed in a hot shutdeva ccndition within 12 hcurs.

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3 asis 5 requ:.rements of Section 2.3, E= erg _ncy Core Ceal' g syste=, apply to the specifications above with respect to the operability of the I

l Amendment No. 49 2-25

2.0 LIMITING CONDITIONS FOR O'PERATION

• 25 Steam and Feedvater Systems Applicability Applies to the operating status of the steam and feedvater systems.

Objective To define certain conditions of the steam and feedwater system necessary to assure adequate decay heat re= oval.

Specifications The reactor coolant shall not be heated about 3000F unless the following conditions are met:

(1)

Both auxiliary feedwater pu=ps are operable.

One of the auxiliary feedvater pu=ps may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided that the redundant co=ponent shall be tested to demonstrate operability.

(2)

A minimum of 55,000 gallons of water in the emergency feedwater storage tank and a backup water supply to the emergency feedvater storage tank from the Missouri River by the fire vater system.

(3)

All valves, interlocks and piping associated with the above components require:i to function during accident conditions are operable. Mr.ual valves that could inter-rupt auxiliary feedvater flow to the stear.. generators shall be locked in the required position to ensure a flow path to the steam generators.

(h)

The main steam stop valves are operable and capable of

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closing in four seconds or less under no-flow conditions.

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Basis A reactor shutdown from power requires a removal of core decay heat.

I==ediate decay heat removal requirements are no mally satisfied by the steam bypass to the condenser. Therefore, core decay heat can be continuously dissipated via the steam bypass to the condenser as long as feedvater to the steam generator is available. Normally, the capability to supply feedvater to the steam generators is provided by operation

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of the turbine cycle feedvater system.

In the unlikely event of complete loss.of electrical power to the station, decay heat removal is by steam discharge to the atmosphere via the

' main steam safety and atmospheric du=p valves. Either auxi-liary feedvater pump can supply sufficient feedvater for e-moval of decay heat from the plant. The minimum amount of water in the emergency feedwater storage tank is the amount needed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of such operation.

'"he tank can 1pe re-supplied with water from the fire protection system.tl) 2-28 Amendment No. 49 x.

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3.0 SURVEILLANCE REQUIRDIENTS 39 Auxiliary Feedvater System Applicability _

Applies to periodic testing iequirements of the turbine-driven and motor-driven auxiliary feedvater pumps..

Objective To verify the operability of the' auxiliary feedvater (AW)-

system and its ability to respond properly when required Scecifications (1)

The position of valves necessary to ensure auxiliary.

feedvater flov to the steam generators shall be verified by a =onthly inspection. Anytime caintenance is per-formed on the auxiliary feedvater system which alters valve alignments, an operator shall check that the A W systes valves are properly aligned, to ensure AFW flow to the steam generators, and a second operator shall independently verify proper valve align =ent.

(2)

The operability of the motor-driven auxiliary feedvater pump, the steam turbine-driven auxiliary feedvater pump, and the auxiliary feedvater pumps' steam generator level regulating valves HCV-110TA, HCV-llO73, HCV-1108A, HCV-11083, and auxiliary feedvater cross-tie valve HCV-1384 shall be confirmed at least every three =enths.

(3)

The capabilities of the motor-driven and turbine-driven auxiliary feedvater pumps shall be verified by using local pressure indicators and.flov indicators in the control room. The discharge pressure vill be verified

.to be LO psig above the steam generator pressure at rated steas flov.

(h)

Following cold shutdown and prior to raising the reactor coolant temperature above 3000F, the =otor-driven auxi-liary feedvater pump shall be tested to verify the nor-

=al flow path for auxiliary feedvater to the steam gener-ators.

Basis The valve position verifications perfor=ed monthly and follow-ing auxiliary feedvater system maintenance vill confir9 the

. availability of an auxiliary feedvater flow path to the steam generators.

W endment No. 49 3-62 i

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3.0 SURvzILLAIicz RI;UIaneNTs 39 Auxiliary Feedvater Systes The testing every three =enths and after cold shutdowns of.

the auxiliary feedvater pu=ps vill verify their operability' by recirculating vater to the emergency feedvater storage -

f tank and cperating, one at a time, the regulating valves.

(HCV-11073 and Ec7-11083) to confirs a flow path to the stean generaters and operability of the valves.

Proper functioning of the steam turbine =A-4 ssion valve a;d starting of the feedvater pump vill demonstrate the inte-grity of the steam driven pu=p.

Verification of correct operatics vil.1 be made both fres instrumentation within the main centrol rocs and iirect visual observation of the pu=ps.

Beterences

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(1)

FSAR, Sec-ica 9.k (2)

Technical Specificatica 2 5 l

Amendment No. 49 3-c2a m

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