ML20067E723

Amend 168 to License DPR-59,clarifying & Defining ECCS Requirements When Plant in Cold Shutdown Condition

ML20067E723
Person / Time
Site:	FitzPatrick
Issue date:	02/13/1991
From:	Capra R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20067E724	List: ML20067E723 ML20067E725
References
NUDOCS 9102180074
Download: ML20067E723 (11)

Text

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[pn Cloy *[o, UNITED STATES g

NUCLEAR REGULATORY COMMISSION a

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y WASHINGTON D. C. 20555

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DOWER AUTHORITY-OF.THE STATE OF NEW-YORK DOCKET NO.-50-333

.1AMES.A, FITIPATRICK-NUCLEAR: POWER _ PLANT AMENDMENT.TO. FACILITY OPERATING; LICENSE-Amendment No.168 License No. DPR-59 1.

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

A.

The application for amendment by Power Authority of the State of New York (the licensee) dated April 2, 1990, complies with the standards and requirements ~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 appilcation, tne omvlsions of the Act, and the rules and regulations of the Co mVn fcn; C.

There is reasonable assurance (1) that the activities authorized by this uniennunt can be conducted without endangering the health and safety of the public, and (ii) that such activities will lie conducted.in con!pliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defenu and cocurity or to the health and safety of the public; l

and l

l E.

The issuance of this amendment is in accordance-with 10 CFR Part l

51 of the Commission's regulations and all applicable requirements-have been satisfied.

2.

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

l amendrrent, and paragraph 2.C.(2) of Facility Operating License l

No. DPR-59 is-hereby amended to read as follows:

9102180074 910213 i

PDR ADOCK 05000333 p

PDR.

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(2) Technical -Speci fica tio_ns -

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.168, 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 to be implemented within 30 days..

FOR THE NUCLEAR REGULATORY COMMISSION l

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SY 0.fg

. Robert.A. Capra, Director-Project Directorate I-1

. Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation-

Attachment:

Changes to the Technical Specifications Date of Issuance: February 13,1991l l

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i ATTACHNE_NT T0- L ICENSE. AMENDMENT. NO.168 FACIllTY2 0PERATING: LICENSE =:NO. DPR '

' DOCKET NO. 50-333 Revise. Appendix A as follows:

Remove.Pages in_ sert Pages ii 11 122 122^

l22a 122a 129

- 130'

. 129

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130-i 133 133' 165 165

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TABLE OF CONTENTS (Cont'd) o F.

ECCS-Cold Condition

' F.

l 1

G.

Maintenance of Filled Discharge Pipe G. -

122a

.l H.

Average Planar Unear Heat Generation Rate (APLHGR)

H.

l.

Unear Heat Generation Rate (UiGR) 1.

. 123.

124 J.

Thermal Hydraulic Stability J.

124a SURVEILLANCE UMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.6 Reactor Cooict Systa>;.

4.6 136 A.

Pressurization ext Thermal Umits A.

136:

B.

DELETED.

C.

Coolant Chemistry-C.

139-D.

Coolant Leakage

. D.

141 E.

Safety and Safety / Relief Valves E.

142a F.

StructuralIntegrity F.

144 G.

Jet Pumps G.

144 H.

DELETED 1.

Shock Suppressors (Snubbers) 1.

145b 3.7 Conta!nment Systems -

4.7 165 A.

Primary Containment A.

165 B.

Standby Gas Treatment System B.

- 181 C.

Secondary Containment C.

- 184 D.

Primary Containment is,olation Valves -

D.

185 3.8 Miscellaneous Radioactive Material Sources 4.8 214 3.9 Auxiliary Electrical Systems 4.9 215 A.

Normal and Reserve AC Power Systems A.

215 B.

Emergency AC Power System B.

216 C.

Diesel Fuel '

- C.

218' O.

Diesel Generator Operability D.

220-E.

Station Batteries E.

221 F.

LPCI MOVIndependent Power Supplies F.

222a G.

Reactor Protection System Electrical Protection Assemblies '

. G.

222c e

i 3.10 Core Alterations 4.10 227 A.

Refueling Interlocks

- A.

227; B.

Core Monitoring B.

- 230-C.

Spent Fuel Storage PoolWater Level C.

231 D.

Control Rod and Control Rod Drive Maintenance

. D.

231 3.11 Additional Safety Related Plant Capabiliiles 4.11 237 A.

. Main Control Room Ventilation A.

' 237 B.

Crescent Area Ventilation B.

239-C.

Battery Room Ventilation C.

239 i

Amendment No. f, f,%,1[ 168 v

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35 (cont'd) 4.5 (cord'd) i-i.

F.

. ECCS-Cold Condibon F.~

ECCS-Cold Condition i

1.

A minimurn of two low pressure Emagercy Core Cooling d the k pressure M systems r@

subsystems shaN be operab

• h irradised fuel is in by 3lJ.F.1 and 3.5.F.2 shall be as follows:

' the reactw, the reactor is in he cold condihon, and work is 1.

a ham ah once m 36 on h-I~

being performed with the potential for drainog the reactor g

-required Core Spray pump (s) and/or the RHR pump (s).

t e

Each Core Spray pump shall deliver at least 4,625 gpm h

2.

' A minimum of one low pressure Emergency Core Cooling a W W car % to a reactw W subsystem shaR be operable whenever irradised fuel is in pressure greater than or equal to 113 psi above primary I'

the reactor, the reactor is in the cold condibon, and no cwth passwe. M M pump M N at

. least 9900 gpm against a system head corresponding to a j

work is bemg performed W the potenkd fu dr% the reacta W.

reactor vessel to pnmary cordainment differenhal pressore of > 20 psid.

3.

Emergency Core Coolog subsystems are not required to 2.

Perform a monthly operability test - on the required Cao

' be operable provided. that the reactor vessel head is,

Spray and/or LPCI motor operated valves.

. removed, the cavdy is flooded, the spent fuel pool gates 3.

Once each shift verify the suppression pool water level is.

are removeo, and the water level above the fuel is in greater than oriequal to 10.33;ft...whenever the low.

I 2

3.10.C.

pressure ECCS subsystems are aligned to the suppression l

With tho'~ requhements of 3.5.F.1, 3.5.F.2, or 3.5.F.3 not pool' 4.

sabslied, suspend core alterations and aR operahons with -

4.

Once each shift verify a minimum of 324 inches of water is the potenhal for draming the reactor, vessel Restore at available.In the Condensate Storage Tanks- (CST) least one system to operable =taha wdhin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or whenever the Core Spray System (s) is ' aligned to the actahrah Secondary Contamment integnty wdhm the next tanks.

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8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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3.5 (cont'd) 4.5 (cont'd) l

'G.

Maintenance of Filled Discharge Pipe G.

Mantenance of Filled Dischinya Pipe Whenever core spray subsystems, LPCI subsystems, HPCI, or The followmg sunellance requrements shall be adhered to, in RCIC are required to be operable, the discharge piping from the order to' assure that the discharge piping of the core spray pump discharge of these systems to the last block valve shall be subsystem, LPCI stbsystem, HPCI, and RCIC are filled:

- filled.

1.

Every month prior to the tasting of the LPCI subsystem i

a.

From and after the time that the pump discharge piping of the and core spray subsystem, the discharge piping of thesu i

HPCI, RCIC, -LPCI, or Core Spray Systems cannot be

, systems shall be vented from the high point, and water i

mamtamed in a filled

. flow observed.

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t Amendment No-168 122a

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3.5 BASES (cont'd) i vessel head off the LPCI and Core Spray Systems will p,1u in F.

ECCS-Shutdown Mode ths designed safety funcbon without the help of the ADS.

Low pressure Emerm Core W Systems (ECCS) are l

E.-

Reactor Core ledation Cooling (RCIC) System required when the reactor is in. a:ooid conc 9 tion to ensure -

b The RCIC is designed to provide makeup to the Reactor irwentory h case d an Went

-drandown of the reactor-vessel.jTwo low pressure ECCS j

Coolant System as a planned ' operation for periods when the -

i mJ W sink is W h Nh m ' as subsy._,ioim are required operable to meet the single-failure j

M enon.

l redundant makeup system on total loss of au offsite power in ;

j the event that HPCI is unavalable. In' aN other postulated The low pressure ECCS subsystems consist of two CSJ j

acciderts and transients, f.he ADS provides redundancy for the -

systems,Ltwo LPCI. subsystems, or a' combination thereof.;

l

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HPCI. Based on this and judgements on the reliabdity of the Each ~ CS : system consists of. one motor-driven pump,1 HPCI system, an allowable repair time of 7. days is speedied.

associated piping, and valves. Each CS system is cas, Glo of -

Inwrweiaea and daily verificahons of HPCI operabdity dunng transiemng water to the raartnr vessel from the suppression RCIC outage is considered adequde based on judgement and poolLor,, when the suppression pool. is unavailable, the.

pchA4ty.

condensate storage tank.L in the cold conc 2 tion, each LPCI subsystem consists of.one motor-driven pump, associated-

'. Low power physics testing and raarts operata training with inoperable way06ents will be conducted only wtien the RCIC piping, and valvas. Each LPCI s:W is capable of -

transferring water from the suppression pool to the reactor-System is not required, (reactor coolant scr.iperature $2127 5

m RHR m k W p W Wsth and coolant pressure 1150 psig). If the plant parameters are of 'its h ee~

'to a Core Spray io below the point where the RCIC System is required

• physics ala WW d W > the plant in an System. A LPCI subsystem operating in the shutdown cooling f

mode of RHR is considered operable for the ECCS function if it '

can be realigned manuaNy (either remote or local) to the LPCI-i-

Operabdity of the RCIC System is required only when reactor mode and is not olherwise inoperable. In the cold contNtion,

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pressure is greater than.150 ' psig. and: reactor coolant the RHR system cross-tie valves are not required to be closed.

{

temperature is greaser than 2127 haranca Core spray and low ECCS @ @ N W

'nject

.can protect the core for any size r

an Went M i

draindown. However, with on!y one low pressure system 1 operable, the overaN system reliability is reduced because a 3

single-failure could render the ECCS incapable of po1viming J,

itsintended AmendmentNo. f,If,If 168 129

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JAFNPP 3.5 BASES (cont'd) functot Therefore, operation with the potential for draining generation rate of a!! the rods of a fuel assembly at any axial the reactor vessei is not a!! owed with only one low pressure location and is only dependent secondarily on the rod to rod ECCS subsystem operable.

power distribution within an assembly. Since expected local ECCS systems are not required to be operable during refueling variations in power distribution within a fuel assemL; affect the conditions. Sufficient coolant inventory is available above the calculated peak clad temperature by less than + 207 relative fuel to allow operator action to terminate the inventory loss to the peak temperature for a typical fuel desigEthe limit on prior to fuel uncovery in case of an inadvertent draindown.

the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 G.

Maintenance of FiDed Discharge Pipe Appendix K limit. The limiting values for APLHGR are specified if the discharge piping of the core spray, LPCI, RCIC, and HPCI in the @abng Umns RW Mg We %

are not f!!!ed, a water hammer can develop in this piping when

@atim a Wier is appned to Me valmsA demon tne pump (s) are started. To minimize damage to the discharge

. f this multiplier can be found in Bases 3.5.K, Reference 1.

piping and to ensure added margin in the operation of these 1.

Linear Heat Generation Rate (LHGR) systems, this technical specification requires the discharge M

MH W

atim rate in lines to be filled whenever th6 system is required to be operable. If a discharge pipe is not filled, the pumps the supply ny sless MWesign h W gmerah that the must be assumed to be inoperable for technical The LHGR shall be checked daily during reactor operation at specification purposes. However,if a water hammer were to 25% rated thermal power to determine if fuel burnup, or control occur, the system would still perform i+s design function.

rod movement, has caused changes in power distribution. For H.

Average Planar Linear Heat Generation Rate (APLHGR)

R to M a Wng valm Mow 25% raMW power, the ratio of local LHGR to average LHGR would have to be This specification assures that the peak cladding temperature greater than 10 which is precluded by a considerable margin following the postulated design basis loss-of-coolant accident when employing any permissible control rod pattem.

will not exceed the limit specified in 10 CFR 50 Appendix K.

The peak cladding temperature following a postulated loss-of-coolant accident is primaril a function of the average heat

/

l Amendment No. f4,XMM Ig 1,W, JEf 168 130

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i JAFNPP 45 BASES (cont'd)

. the line is in a full condition. Between the rnonthly intervals at which the lines are vented, instrumentation has been provided in the Core Spray System and LPCI System to rnonitor the presence of water in the dischange piping. This instsumentation will be calibrated on the same frequency as the safety system instrumentahon. This penod of periodic testing ensures that dunng the interval between the rnonthly checks the status of the discharge piping is monitored on a conhnuous bases.

Normally the low pressure ECCS subsystems required by Speedication 3.5.F.1 are demonstrated operable' by the surveillance tests in Sp6cA.ations 45.A.1 and 4.5.A.3. Section 4.5.F cpaeidiae periodic sunellance tests for the low pressure ECCS subsystems which are eppik.abi6 when the reactor is in L

the cold condshon. These tests in cosjunchon with the

~

requirements on filled discharge piping (SpecA i.on 3.5.G),

and the requirements on ECCS actuation instrumentation-(Speedication 32.B), assure W_mie ECCS capability in the cold concthon. The water level in tre suppression pool, or the Condensate Storage Tanks (CST) when the suppression pool is t

inoperable, is checked once each shift to ensure that sufficient wateris available for core cooling.

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f Amendment No [ 168 133

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JAFNPP 3.7 LIMITING CONDITIONS FOR OPERATION

' 4.7 SURVEILLANCE REQUIREMENTS

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3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Applicability:

Applicability-Applies to the operating status' of ' the pnrnary and secondary Applies to the pnmary and secondary containment integrity.

containment systerns.

Objective:

Objective:

To assure the integnty of the primary. and secondary containment To verify the integnty of the primary, and secondary containment

' systems.

systems.

Specificatiort Specification:

A."

Primary Containment A..

Prirnary Containment 1.

The volume and temperature of the water in the pressure 1.

The pressure suppression denber water levei and suppmss'on chamber shali be maintained within ' the temperature shall be checked once per day. The follomng linwts whenever the reactor is entical or.whenever acessbie interior surfaces of the drywell and above the the reactor coolant temperature is greater than 212 F and water line of the pressure suppression derrber shall be-Irradiated fuel is in the reactor vessel:

inspected at each refueling outage for evidence of deterioration. Whenever there is indication of relief valve a.

Maximum vent submergence level of 53 inches.

operation or testing which adds heat to the supp&#00 W,

t6e M M WW MwM b.

Minimum vent submergence level of 51.5 'mches.

and also observed and logged every 5 minutes until the The suppression chamber water level may be heat addition is terminated. Whenever there is indication outside the above limits for a maximum of four (4) of' relief valve operation with the temperature of the hours dunng required operability testing of HPCI, supprermn pool reaching 160'F or more and the primari RCIC, RHR, CS, and the Suppression Chamber -

coolant system pressure greater than 200 psig, an exterreJ DrywellVacuum System.

visual examination of the suppression chamber shall N Wwe rMng poww opwath c.

Maximum water temperature (1)

During normal power operation maximum water temperature shall be 95*F.

Amendment No. f168 165 L______E___

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' 3.7 BASES _ (cont'd) l q

Usmg w muumum or mauamum downcomer submergence Using a 40'F rise (Sechon 5.2 FSAR) in'the ?vppression i

levels rpven in the specscation, containment pressure dunng chamber water temperature and a medmum initi; amperature i

the design asis acririarit is approximately 45 psig wtuch is'

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of 95*F, a temperature of 145"F is achieved, whic, is wee below l

below the. design 'of 56 psig.D The minimum downcomer

!the 170"F _ temperature which 'is. : used. for-complete

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i-submergence of 51.5 in. results in a mirumum suppression condensation.

E chamber water volume of 10G,000 ft. The majonty of the 3

h an L W wh @ h e Bodega tests (9) were run with a submerged length of 4 ft. and

,,,gg,F a M NWWd g

with complete condensasion. Thus,' with respect to downcomer i

containment cooling pumps (two LPCI pumps and two RHR

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3 4 p u p p,",,'cnic g '

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• • ** ** P""PS) **"'""**"' P'***" i* " ' '*4""*d '*

j' Containment-Suppression Ct. amber M Program indicate maintain ariaryuma nat positwo suction head (HPSH) for the-z ll the adequacy of the vuw*wi range of submergence to ensure f a spmy W W W N

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that dynamic forces smarianari with pool swell do not.'esult in Umding suppression pool temperature to 130'F during RCIC, _

t overstress.'of. - the i sappression chamber or. associatari HPCI, or relief valve operabon, when decay hast and stored j

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structures.

energy are removed form the primary system ty (Ascharging i

reactor steam directly to the. suppression charmer assures

.The maximum temperature e the end of h W dunng the Humboldt Bay (10)'and Bodega Bay tests was ariaryuma margin for.a potendal blowdown any. time during

- 170'F, and this is ces is _ ~d, taken to be the limit for :

. RCIC, HPCI, or relief valve operabon.

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complete condensa:fon of the limit for complete condermhrn Experimental data irweiratan that excesolve steem corwiensing l

j1 of the raar*r ent*mt, abhough condepenNnQ would occur for loads JCan be avoided if, the peek temperature of the f

p temperatures above 170'F.

suppression pool is maintained below 100'F during any period of relief valve operation with sonic contStions e the discharge g

exit.. Speciih eks6 have been placed-on the envelope-of i

reactor operating conc $tions so that.the reactor can. be E

depressurized in a timely manner to avoid the. regime of i

potentially high suppression chamber loarings.

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