ML20003F556

Amend 53 to License DPR-59,revising Tech Specs to Include Typographical & Editorial Corrections & Administrative Changes

ML20003F556
Person / Time
Site:	FitzPatrick  (DPR-59-A-053, DPR-59-A-53)
Issue date:	04/13/1981
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20003F557	List: ML20003F556 ML20003F561
References
NUDOCS 8104230016
Download: ML20003F556 (10)

Text

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a rec oq[C UNITED STATES g

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NUCLEAR REGULATORY COMMISSION L-l0 7. gl WASHINGTON. D. C. 20555

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POWER Al'THORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 53 License No. DPR-59 1.

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

A.

The applications for amendment by Power Authority of the State of New York (the licensee) dated January 6,1981 and February 20, 1981 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions's r'.es and regulations set forth in 10 CFR Chapter I;

~

i B

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of l

the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comissions'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 Comissions's regulations and all applicable requirements have been satisfied.

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re104 n o&

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. 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. OPR-59 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 53, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This ifcense amendment is effective as of the'date of its issuance.

FOR THE NUCLEAR REGULATORY COMISSION J

'/

Thomas Ippolito, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: April 13,1981

ATTACHMENT TO LICENSE AMENDMENT NO. 53 FACILITY OPERATING LICENSE NO.

DPR-59 00CKET NO. 50-333 4

A Revise Appendix A as follows:

Remove Pages Insert Pages 32 32 56 56 74 74 79 79 1 01 101 1 02 102 159 159 l

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.,,.---.,-..-.,,-.,.. ~,.,...,..-,- -.....,.

JArtfPP 3.1 BASE;E The outputs of the subchannels are 2

combined in a 1 out of 2 logict 1.e.,

an l

The reactor protection syntem input.

signal on either one or luth of i

autwatically initiates a re. actor neram the subchannels will cause a trip system tot trip.

The out put s of the trip systems trip on both age.etr.inqud so that a 1.

Preserve the integrity of the systens is rcepiired to produce a reactor fuel cladding.

scram.

2.

Preserve the integrit.y of the This system moets the intent of IEEE - 3 Reactor th)lant System.

279 (1971) for Nuc1 car Powur Plant Protection Synte*ms.

The system h.is a 3.

Minimize the energy which must.

ro11 ability greater th.in that of a 2 out be abrxirbed f ollowing a loss of of 3 system aint somewhat less t han t hat coolant. accident, avul prevent of a 1 out. of 2 system.

inadvertent criticality.

With t he exception of t he average gewer This specification provides the limit ing rango sm>ni t o:

(APHM)

channels, the conditions for operation uncessary to intermediate tanqu emm i t or (IHM) preserve the ability of the system to channuls, the main sten ino14 tion v.ilvo perfotu its intended function even closuro and tho tutbino stop valve i

during periods when instrument. channels

closuto, each subchannot has onn may be out of serrvico because of instrument channol.

When the minimum condition for operation on the number of maintenanew.

When necussary, one channel may be made inopetable for briot operable ins t rument.

channels por intervals to conduct reipaired f unctional untripped protection trip system is met l

tests and calibrations.

or it it cannot. be met and the affected I

ptotect. ion trip system is placed in a The Reactor Protection System is of the tripped condition, the effectivennas of dual channel type (Heterence subsection the ptotect. ion system is prenotved.

7.2 FSAH).

'Ihn Syst em is made up of two independent trip s y s t. ems,

each having

'll*h r e e APMM inst rument channels ate two subchannels of tripping devices.

provided for cach ptotection trip Each subchannot has an input ftom at system. APitit's A and E operat u tuntacts l 1 east one instaumont channel which in one subchannel and APHH'n C and E numitors a critical parameter.

opotato co'itacts in the other 32 AmendmentNo,y.53

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is.

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1 closure group. The water level steam !!ne isolation valves, main steam instrumentation initiates protection fo'r l

drain valves, recirc. sample valves the full npectrum of loss-of-coolant J

(Group 11, initiates the itPCI and RCIC and trips the recirculation pumps. The accidents.

Iow-low-low reactor water level instru-venturis are provided in the main steam mentation is set to trip when the water lines as a maans of measuring steam flow i

level is 18 in, aLove the top of the and also limiting the loss of mass I

active fuel. This trip activates the inventory from the vessel during a stata remainder of the ECCS subsystems, and line break accident. The pr ima ry,

starts the emerger.cy diesel generators.

function of the instrumentation is to 3

These trip level settings were chosan to detect a break in the main steam line.

be high enough to prevent spurious actu-For the worst case accident, main steam j

atton but low enough to initiate ECCS line break outside the drywell, a trip l

operation and prlh.ary system isolation setting of 140 percent of rated steam i

so that post-accident cooling can he ac-flow in conjunction with the flow j

comp 11shed and the guidelines of limiters and main steam line valve 10CrR100 will not be exceeded, ror closure, limits the mase inventory loss

]

large breaks up to the complete auch that fual is not uncovered, fuel circumferential break of a 24 in.

temperature peak tt approximately j

recirculation lino and with the trip 1,000"r and releaua of radtuactivity to setting given above, ECCS initiation ind the environs is below 10CFR100 guide-I primary system isolation are initiated lines. I:eference Section 14.6.5 FSAR.

i in time to meet the above criteria, Reference paragraph 6.5.3.1 FSAR.

d The high drywell pressure instru-mentation is a diverse signal for mal-functione to the water level instru-i mentation and in addition to initiating 4

j ECCS, it causes isolation of Groups a and 3 isolation valves. For the breaks Wiscussed above, this instrumentation will generally initiate ECCS operation l

f before the low-lou-low water level instrumentations thus the results given above are applicable here also. See Specification 3.1 for tsalation valve l

re.iendment No. W. y, 53 5 f, i

l 1

d TA012 3.2-6 I

suavElt1ANCO INSTRtMDfrATION j

ntnimum see.

Iso. of channele l

of operable i

Instrument Type Indication Provided Channe1e Inotrument and mange by Deeign Actton_

l (suppreselon Chamber Indicator

)

j (Water IAvel Recorder

)

l t (Wide Range)

-72 to 472 inches) 1

(

)

2 (2) l (suppression Chamber Indicator

)

(Water Level Recorder

)

l

( (Harrow Range)

-6 to +6 inches )

}

N/A Control Rod Indicator 1

(7)

Position Indication Post,lon 00 to 4B i

2 Source Range Indicator 4

(8) 9 Honitore Recordeg l

_)

I to 10 cpe i

3 Intermediate Indicator 8

(t) (9)

Range Monitor R

rder l

i 10~

to 404 Itated Power 2

Average Power Indicator 6

(S) (9) j Recorder i

I Range Monitor 0-125t Rated Power l

i 1

Drywell-suppreselon Recorder 2

(2)

Chamber Differential O to 5 poi i

Pressure Computer O to 5 pel i

i esuits Poe Tasta 3.2-6 I

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1. From and after the date that the minimum number of operable instrument channele is one less than the f

minimum number specified for each parameter, continued operation le permiselble during the succeeding l

30 days untees the minimum number specified is made operable sooner.

2. In the

• vent that all indications of this parameter le disabled and such indication cannot be restored j

in six (6) hours, an orderly shutdown shall im initiated and the reactor shall be in a Hot Shutdown condLtion in six (6) hours and a Cold Shutdown condition in the following eighteen (18) bours.

l Amendment flo. Alf, 53 76a 1

5 o

)

JAlllPP Tant.c 4.2-2 HIttlltitti TEST AHD CAI.IDRATIOil QFQHfilCY FOR COpr. At3D CollTAlllHEt4T COOLitJG SYSTDiS Inst riunent Channel Instrument Punctional Test Calibration l'requency Instrument Check

1) Beactor Water level (1)

Once/3 months once/ day i

21 Dryvell Pressure (1)

Once/3 monthe Hone

3) Reactor Pressure (1)

Once/3 months Hone

4) Auto Sequencing Timers liA Once/ operating cycle Hone 51 ADS - LPCI or CS Pump Disch.

(1)

Once/3 months None Pressure Interlock

6) Trip System Dus Power Honitors (1)

N/A Hone l

0) Core Spray Sparger d/p (1)

Once/ 3 monthe once/ day s.

9) Steam Line High I' low (HPCI & RCIC)

(1)

Once/3 monthe Hone

10) Steam Line/ Area High Temp.(llPCI & RCIC)

(1)

Once/ operating cycle Once/ day

12) IIPCI & RCIC Steam Line low Pressure (1)

Once/3 months Hone j

i

11) HPCI Suction Source levels (1)

Once/3 months' None j

e

14) 4KV Dnergency Power Under-Voltago Relaya Once/ operating cycle Once/ operating cycle None and tir. ors 1

4

15) IIPCI & RCIC Exhaust olaphraga Pressure (1)

Once/3 monthe Hon's liigh

17) LPCI/ Cross connect valva Position Onco / operating cycle HA HA Hote: See listing of notes following Table 4.2-6 fdr the notes s'eferred to herein.

1). /

, 53 d

19 Amendment No.

3

3 d end 4.3 BASES (cont'd)

JAFNPP At power Icvels below 20% of rated, rod drop accident consequences are almormal cont rol rod patterns could acceptable. Control rod pattern produce rod worths high enough to be of constraints above 20*, of rated power are concern reintive to the 2.'l0 calories per imposed by power distribution requirements I

s efined in Sect ion 3.~,.3.5 of these

."l.

gram drop limit.

In this range, the Ittel ecluiscal Speci fic t ions.

Power level and HSCS constrain the cont rol rod or autom tic cutout of the HSCS function sequence and patterns to those which is 8 nseil y Gnt stage turliine pressure.

involvo only acceptable mod wortles, Hecause the inst rument has an instrument crror of a 2*. of full power, the nominal 1

1hc Itoil hoi t h Plinimizer and the Hod inst ru:nent set t ing i s 22*. of rat ed power.

Sequence. Control System provide Power level for automat ic cutout of i

automatic supervision to assuro that the Hl01 function is sensed by feedwater out-of-sequence control rods will not and steam flow and is set manually at he withdrawn or inserted; i.e., it 30*. of rated power to be consistent with i

limit s operator deviance f run planned the HSCS setting.

}

wi t hilrawa t sequences. They serve as a hackup to proccitural control of I?unct ional testing of the RhH prior control rod sequences which limit to the start of control rod withdrawal the maximal reactivity worth of at startup, and prior to attaining 20%

cont rol ro.ls, in tho event that the rateil thermal power during rod insertion Hail Worth Minimelzer is out of service, while shutting Jewn, will ensure eclialle when required, a second licensed operat ion and minini ze the proliabili ty operator or at her spiali fled t echnical of the rod.trup accident.

j plant employee lho HSCS can be functionally tested I

can manually fulfill the control rod prior to cont rol rod wit hdrawal for pattern conformance functions of this reactor startup.

Ily selecting, for system.

In this case, t he it';CS i s exampic Ag2 and attempting to withdraw, l

leacked up by independent procedural by one notch, a ro1 or all rods in cont rol to assure conformance.

cach other group, it can be determined 4

j that the Al2 group is exclusive. By bypassing to full-out all' Ag y md s,

The fimetions of the H101 and HSCS make it unnecessary to speelfy n selecting A 4 and attempting to withdraw, license linit on rod worth to preclude l'Y anc notc i, a md or' all mis in group 8, tlu A34 gmup is determined exclusive.

unacceptahic consequences in the event

.Hu' same procetlum can lac relutated for of a cont rol roit ilrop. At low powers.

]

helow 20".,

these devices force adherence tlie H gmups. Af t er 50*. of the cont rol to acceptalsic rod patterns. Above 20*.

of rated power, em const raint on rod 10 1 I

pattern is requireil to assure that Amendment No. y(, 53 i

e 1

JAt'HPP l

l 3.1 and 4. 3 BAssa (cont'd) sede have been withdrawn (e.g., groupe A32 asut liais system backs up the operator who l

l l

Ap), it le demonstrated that the Group natch withdraws control rods according to l

made for the cositrol drives le enforced. 11al e written sequences. The specified re-l demosist ration is made by performing the hardware strictiosse with one channel out of functional test me<pience.

11 e Group match re-maavice conservatively assure that j

stralists are automation 11y removed above 204 power.

f uel damage will not occur due to rod wittutrawat errors when this condition exists.

Ikaring reactor shutdown, almilar surveillance

c. hacks shall be made with regard to rod group availability me soon as automatto initiation of A limiting control rod pattern le a pattern the RSCH ocoiare and subsequently at appropriate which results in the core being on a thermal stages of the control rod insertion.

hydraulio limit. (i.e., HCPR limite as shown y

in speoirication 3.1.D),

puring use of l

• 4 We source Range Monitor (anH) Syste e performe no much patterns, it le jaulged that testing j

automatlo safety sysptam functions 1.w.,

it has no of the RNt Systeam prior to willutrawal of i

moram function. It does provide the operator witig much rode to ammiere its operability will a visual indication of neutron level. 1ine con-aamiere that improper withdraw does not occur.

It is the responath111ty of the j

moquences of reactivity accidente are fusictionse of the initial neutrog flum. We requirement of at Henctor Analyst to identify these Ilm!L-i i

least 3 cotante per soo amourou that any transient, ing patterne and the designate,d roda either when the patterns are initially established j

abound it ocpr, begine at. or above the initial or as they develop due to the occurrence valise of 10' of rated power used in the analyses j

of trasimient. cold cosulttions. One operable Slot of inoperable control rode in other than 1

channel would be adequate to monitor the approach limiting patterne. Other qualified to criticality using homogeneous patterne of personnol may perform this function.

agattered control rod withdrawal. A minimine of two operable Slot's are provided as an added I

C.

Scram Inmartion Times conservatism.

5.

W e mod alock Monitor (steH) la designed to auto-

%e Control Mod System la designed to bring the reactor aut> critical at a rate fast l

matloally prevent fuel damage in the event of erroneous rod willutrawal from locations of enough to prevent fuel damages i.e.,

to high power density during high power level prevent the HCPR frose becoseing less than operation. Two channels are provided, and the safety Limit. Scram insertion time one of these may be hypemmed from the consolo and scram reactivity curvos shown in Ht3M)-

for malentenance and/or testing. Tripping of 24242, Figures 2a, 2h asul 20 were used one of the cleannels will block erroneous rod in analyses of power transiente to determine wittuirmwat soon enoujh to prevent fuel damage.

HCPR limits. 11:e screa insertion time tesit criteria of Section 3.3.c.1 conform to the scram insertion times of Ht3X)-24242.

Therefore, the rc< paired protection is provided.

Amendment No. 49',53 102

TABLF 4.6-1 (cont'd)

As Psr ASME Code Section XI-IS-200 FitzPatrick Pro [ m ed Program Steet 6 of 6 a

1.xamination Extent of Extent of Examination,4 Item Category Components and Parts Examination Examination Intesvals No.

Table 15-251 t o im Examined

__Mettni

% in 10 Years to Years 5 Years Accessibility C(meents and I xamination Met hods E-2 1

1.6 Pressure-Welds in vessel at Visual 25 25 0

Access is provided Visual examination will be perfogued l

containing control rod drive by observation using optical equiteent capable of welds in penetrations and in-ports in tettom psoviding a cumplete viewing of the a

vessel pene-core monitor housing head insulation.

0.D. of the housing external to t!un trations (Staal tube-to-housing vessel for signs of leakage.

azul vessel)

F 1.7 Pressure-Primary nozzles to

visual, 100 100 33 Heactor pressure boote or local visual examination will containin9 safe-end welds: He-

surface, vessel safe-emi be Ierformed. The extent to which dissimilar circulation inlet and welds are made surface examination is perfonned is metal welds.

Recirculation outlet volumetric accessible by re-determined by radiological considerations.

Closuro nozzles moving tlwrmal in-Manual ultrasonic examination will be Core spray sulation and performed, where possible, until auto-sacrificial shield mated equipment is available but the

plugs, extent of the examination will be deter-mined by radiological considerations.

4.1 Pressure-Pipiry, pressure

Visual, 100 100 33 Safe ends in Remote or local visual examination will containin9 boumlary safe-

surface, branch welds and be performed. Ttw extent to which sur-dissimilar emis in branch amt dissimilar metals face examination is performed is deter-metal welds.

piping welds volumetric welds are male mined by radiological considerations.

accessible by re-Although radiation dosage is high, manual Piping pressure moving piping ultrasonic examination will be performed, boundary welds thermal insulation. wtwre gossible, until automated equilment between dis-becomme available, but the extent of the similar metals.

examination will be determined by radio-logical considerations.

Exception is taken to volumetric examina-tion of welds which require drainage of the reactor vessel.

{

5.3 Pressure-Itop pressure Visual 100 0

0 Hot applicable.

Not app!! cable. %ere are no nossles to i

containing boundary nozzles-and Safe-erns welds on pumps, i

iddfM"wdT5s to-safe erut welds volumetric 6.3 Psessure-Valve pressure Visual ams 100 100 33 Safe emis in vature b oote or local visual examination will be gntagng boumlary valve to volumetric are made access-terformed. Although radiation dosage is melekwei5s safe-ent welds able ley removing high, manual ultrasonic examination will thesual insulation. be serformed, wtwre possitale, until auto-Except iequaissentis taken to volumetric ex.usina-matic leecymes.avallebte.

on Ameminent No. J 7,53 ti-or welam which reautre dr.inage of 15g reactor vessel.

L 6