ML20033B307

Amend 8 to License DPR-22,changing Tech Specs to Reflect Plant Mods as Part of Mark I Containment long-term Program

ML20033B307
Person / Time
Site:	Monticello
Issue date:	11/05/1981
From:	Eccleston K Office of Nuclear Reactor Regulation
To:
Shared Package
ML20033B308	List: ML20033B307 ML20033B312 ML20033B316
References
NUDOCS 8112010214
Download: ML20033B307 (7)

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UNITED STATES y

g, NUCLEAR REGULATORY COMMISSION g'

. E WASHINGTON. D. C. 20555

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NORTHERN STATES POWER COMPANY DOCKET NO. 50-263

.MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 8 License No. DPR-22.

1.

The Nuclear' Regulatory Commission Me Commission) has found that:

A.

The application for amendment by Northern States Power Company (the.

licensee) dated June 4, 1981 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 c.hapter 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 andangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the commor.

derense 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 hav'e been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-22 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendices A and B as revised through Amendment No.

8 are hereby incorporated in the license.

The licensee shall operate the facility in acco.rdance with the Technical Specifications.

8112010214 811105 PDR ADOCK 05000263 P

PDR

2 N._ThislIcenseamendmentiseffectiveasofthe'dateofitsissuance.

FOR THE NUCLEAR REGULATORY COMMISSION h

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

Attachment:

Changes to the. Technical Specificatio'ns fateofIs'uance:

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e ATTACHMENT.TO LICENSE AMENDMENT NO. 8 FACILITY OPERATING LICENSE N0. DPR-22

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DOCKET NO. 50-263 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

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

Remove Insert 164

.164 175 175

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4.0 SUltVEll.1,AtlCE 11EQUlitEllEt3TS 1.1tilTillG C0t3DIT10 tis F0lt OPEllAT! ort Pressure, Suppression, Chamber-Drywell 3.0

.*: 4.

Pressure Suppression, Chamb"r -Drywel i

,Va c u um !!ireake r s Vacuum 4.

lir cake rs Operability and full closure of th,c

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a.

drywell-suppression chamber vacuuln is iequired, all a.

Wh en pr ima ry c ont ai nme nt cight drywell-suppression chamber vacuum breaker,s shall be verified by performance breakers shall be opei able and posit ioned in,

'of the following:

indicated by the the closed pos it ion as (1) Monthly each operable drywell-except during posit ion indicat ion syst em, sparilied in 3./.A.

suppression chamber vacuum tenting and except as g

breriker shall be exercised through o

4.h and c below, i

an opening-closing cycle.

cbamber vacunm b.

Any drywell-suppression breaker may be nonfully closed (2) Once cach operating cycle,~d,ry-an indicated by the posit inn ind icat inn and well to suppression chamber lentage less alarm systems provided that d i y we l l t o shall be, demonstrated to be suppression chamber di f ferent ial pressurc than that equivalent to a one-inch en decay does not exceed t hat shown on Figure diameter orifice and cach vacuum treaker shall be visually inspected.

3. 7.1.

(Containment access required) chamber two drywel1-supprension c.

Up to vacuum breakers may be inoperable

'(3) Once cach operating cycle, vacuum provided that: (1) the vacuum breakers breaker position indication and are determined to be f ully c losed and at alarm systems shall be calibrated and 1 cast one posit ion al ai m c i rcu i t" is funct ionally tested. (Containment operable or (2) the vacuum becaker in acce.ss required) secured in thc closed pos it ion or

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replaced by a blank Ilange.

(4) Once each operating cycle, the vacuum breakers shall be tested to determine that the force required to open each valve from fully closed to fully open does not exceed that-J-

equivalent to 0.5 psi act ing' on the suppression chamber face of the valve disc.

(Containment access required)

, e 164 o

i I

3.7/4.7 Amendment No. 8

\\b nases:

3.7 A.

Primary Containment

..i in The integrity of the primary containment and operation of the emergency core cooling system comb ina t ion, limit the off-uite doses to values less than 10 CFit 100 guideline values in the event of a break in the primary system piping.

Thus, containment integrity is specified whenever the s

potential for violat ion of t he prima ry react or system integrity exists.

Concern about such a violation exists whenever the react or is critical and above atmospheric pressure. An exception is made to this requirement during inirial core loading and while the low. power test program is being ~

conducted and ready acceus to the reactor vousel is required. There will be no pressure on the system at this time which wilI great ly reduce the chances of a pipe break.

The reactor may be taken critical during thin period; however, res t rict ive operating procedures will be in ef fect again t o minimize the probability at an accident occurring.

procedures and the Itod Worth Minimizer would' limit increment al cont rol wort h to less than 1.3% delta k.

A drop of a 1.3% delta k increment of a rod does not result in any fuel damage.

In addition, in the unlikely event that un excursion did occur, the s cact or building and s tandby gas treatment system, which shall be operational during this time, offers a sufticient barrier to ke'ep of f-s ite doses well within 10 CFit 100 guide line values.

The pressure suppression pool water provides the hajat sink for the reactor primary system energy release following a postul,ated rupture of the system.

The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1000 psig.

Since all of the gases in the drywell are purged int o the pressure' suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the exceed 62 psig, Lhe maximum allowable primary containment vapor pressure of the liquid must notsuppression chamber (water and air) was obtained by considering pressure.

The design volume of the that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is' purged to the' suppression chamber.

Iteference 5.2.3 FSAR.

IIsing the minimum or maximum water volumes given in the specification, containment pressure during the design basis accident is approximately 41 psig which is below the. allowahle pressure of 62 psig.

175 3.7 !!ASES I

e Amendment No. 8

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Hasen Cont inued :

Vent system downccmcr nnhmergence in three feet below,the minimum specified suppression pool water icvel. This lennth han been shown to result in reduced posttil'ated accident loading of the torus f

while at the same time annuring Ihe downcomers remain subme sged under all seisinic and accident j

conditions and possess.nlequat e condenn at s on e f fect iveness.

The maximum temperature at the end of blowdown tested during the ilumboldt Bay (t) and Bodega Bay (2) i tests was 170 F and this is connervatively take'n to be the limit for complete condensat' ion of the o

I reactor coolant, althoughcondennationpouldoccur for temperatures above 170 F.

S Experimental data indicate that excennihe steam condensing loads can be avoided if the peak temperature of the suppression pool i n ma i nt ai ned hkil ow 160 F during any period of relief valve operation with sonic conditions at the discharge exit ! Specifications have been placed on the envelope of reactor r

no that tini react olr can be depressurized in a timely manner to avoid the regime e

operating conditions of potentially high supprennion chamhed loadings.

i In addition to the limit s on t emperat uye of the suppccasion chamber pool water, operating procedures define the action to be eaken in the eqent a relief valve inadvertently opens or ticks open. This l'

act ion would include: (1) use of all available means to close'the valve, (2) initiate suppression i

pool water cooling heat exchangers, (3) initiate reactor shutdown, and*(4) if other relief valves are uned to depressurize the react or, their discharge shall be separated from that of the stuck-open relief i

valve to assure mixing and uniformity of energy insertion to the pool.

8 initial maximum nuppression chamber water temperature of 90 F and assuming the normal com-

.l For an plement of cont ai nment cooling pumps (21,pCl pumps and 2 containment cooling service water pumps)

'f containment pressure is n o t. icquired to maintain adequate net positive suction head (NpSil) for the Ilowever, during an approximately one-day. period starting a few l

core spray, I,pCI and lipCI pumps.

hours after a loss-of-coolant accident, should one RilR loop be inoperable and should the contain-ment pressure be reduced to at mospheric pressure through any means, adequate NpSil would not be avail,

.j able.

Since an ext remely degraded condit ion must exist, the period of vulnerability to this event is

't restricted by Speci f icat ion ').7.A.1.h by limiting the suppression pool initial temperature and the period of operat ion.wi t h one inoperable HilR loop.

ll Tl)

Robbins, C ll. 'Wn' tin' o f Ful1 Scale 1/4'8 Segment of the llumboldt Bay pressure f

Suppress ica Cont ai nment," GFAp-3596, November 17, 1960.

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l (2) nodega Bay prcliminary llazards Summary Report, hppendix 1, Docket 50-205, December 28, 1962.

.l (3) General Electric HEDE-21885-p, "Ita s k I' Containment program Downcomer Reduced Submcrgence

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t Functional Assessment Report",.inne, 1978.

t' 176 l

3.7 nases Amendment No. 8

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