ML20141D453

Amend 100 to License DPR-16,changing Tech Specs Re Mechanical & Hydraulic Snubbers

ML20141D453
Person / Time
Site:	Oyster Creek
Issue date:	03/31/1986
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20141D458	List: ML20141D453 ML20141D461
References
NUDOCS 8604080083
Download: ML20141D453 (14)

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UNITED STATES y

'y g NUCLEAR REGULATORY COMMISSION C

p WASHINGTON, D. C. 20555

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GPU NilCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET N0. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.100 License No. DPR-16 1.

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

A.

The acclication for amendment by GPU Nuclear Corporation and Jersey Central Power and Light Company (the licensees) dated November 13, 1985, complies with the standards and requirements of the Atonic Enerav Act of 1954, as amended (the Actl, and the Commission's rules and reaulations set forth in 10 CFR Chapter I; R.

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 endangering 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 common defense and security or to the health and safety of the public; a r.d E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

l 8604080093 860331 DR ADOCK 05000219 p

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, 2.

Accordingly, the license is amended by changes to the Technical Specifications es indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Provisional Operating License No. DDR-16 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.100, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR TH NtCLEARREGlqtAT0EYCOMMISSION

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v John. Zwolinski, Director BWR D oject Directorate #1 Division of BWR Licensina

Attachment:

Chances to the Technical Specifications Date of Issuance: March 31,1986

L ATTACHMENT TO LICENSE AMENDMENT N0.100 PROVISIONAL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 4

Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT i -

3.5-3 3.5-3 3.5-5 3.5-5 3.5-6 3.5-6 3.5-8 to 3.5-8 3.5-13 j

3.5-13a 3.5-9 4.5-6a 4.5-6a 4.5-6a-1 4.5-6a-1 to 4.5-6a-3 4.5-9b 4.5-9b 6-23 6-23 4

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3.5-3 b.

Two of the fourteen suppression chamber - drywell vacuum breakers may be inoperable provided that they are secured in the closed position.

c.

One position alarm circuit for each operable vacuum breaker may be inoperable for up to 15 days provided that each operable sup-pression chamber - drywell vacuum breaker with one defective alarm circuit is physically verified to be close? immediately and daily during this period.

6.

After completion of the startup test program and deinonstration of plant electrical output, tne primary containicient atmospnere shall be reduced to less than 4.M 02 witn nitrogen gas within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor mode selector switch is placed in the run mode. Primary containment deinerting may corrnence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.

7.

If specifications 3.5. A l.a b, c(1) and 3.5. A.2 tnrough 3.5. A.5 cannot be icet, reactor shutdown shall be initiated and the reactor shall be in tne cold shutdown condition witnin 24 nours.

8.

Shock Suppressors (Snubbers) a.

All safety related snubbers are required to be operable whenever the systems they protect are required to be operaDie except as noted in 3.5. A.8.b and c below.

b.

With one or more snubbers inoperable, witnin 72 nours replace or restore the inoperable snobber(s) to operable status, c.

If the requirements of 3.5.A.8.a and 3.5.A.d.b cannot be met, ceclare the protected system inoperaole and folldw the appropriate action statement for that syste:n.

d.

An engineering evaluation shall be performed to determine if the components protected by tne snubber (s) were adversely affected by the inoperability of the snubber prior to returning tne system to operable status.

Amendment No. 32, 86, 87,100

J.5-5 importantly, the accessioility of tne valve lever ara and position reference external to tne valve. The fail-safe feature of the alare circuits assures operator attention if a line fault occurs.

Conservative estimates of the hydrogen produced, consistent witn toe core cooling system provided, snow that the hydrogen air mixture resulting from a loss-of-coolant accident is consideraoly belos tne flammaoility limit and hence it cannot burn, and inerting would not be needed. However, inerting of the primary containment was included in the proposed design and operation. The 5% oxygen limit is the oxygen concentration limit stated by the American Gas Association for (4) hydrogen-oxygen mixtures below which combustion will not occur.

The 4% oxygen limit was estaolished by analysis of tne Generation ar.d Mitigation of {gmyustiole Gas Mixtures in Inerted BWR Mark I Containments.

2i To preclude the possioility of starting up the reactor and operating a long period of time witn a significant leak in the primary system, leak cnecks must be made wnen the system is ay9H10, that an or ngar rated temperature and pressure.

It has been snown.

acceptable margin witn respect to flammability exists witnout containment inerting.

Inerting the primary containment provides additional margin to that already considered acceptaale. Inerefore, permitting access to the drywell for the purpose of leak checking would not reduce the margin of safety celow that consioered adequate and is judged prudent in terms of tne addeo plant safety offereo by tne opportunity for leak inspection. Tne 24-nour time to provide inerting is judged to be a reasonable time to perform tne cperation and establish the requirea 02 limit.

I Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as mignt occur during an earthquake or severe transien.,

while allowing normal thermal motion during startup and shutdown.

Tne consequence of an inoperable snubber is an increase in tne procaDility of structural damage to piping as a result of a seismic or other event, initiating dynamic loads.

It is, therefore, required tnat all snubbers required to protect the primary coolant system or any other safe'.y system or component be operaole wnenever tne systems tney protect are required to oe operable.

Amendment 30. 73, 86,100 t

a

1 3.5-6 The purpose of an engineering evaluation is to determine if the components protected by tne snubber were adversely affected by the inoperability of the snubber. This ensures that the protected component remains capaole of meeting tne designed service. A documented visual field inspection will usually be sufficient to determir.e system operability.

Because snubber protection is required only during low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replacemehts.

Secondary containmentN is designed to minimize any ground level release of radioactive materials wnich mignt result from a serious accident. The reactor building providet secondary containment during reactor operation wnen the drywell is sealed and in service and provides primary cuntainner.t when tne reactor is shutdown arid the drydell is open, as dJring refueling. Because the secondary containment is an integral part of the overail contaircent system, it is required at all times tnat prireary containment is required.

Moreover, secondary centaf'icent is requirea during fuel handling operaticns and wnanever work is being cerformed on tne reactor or its connected systets in the reactor building since their operation could result in inadvertent release of radioactive material.

Amendment No. U, D, 75,100

3.5-8 TaDlc J.6-1 (Deleted) i n

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TABl.E 3.5.2 CONTAINHeiNT ISul.ATION VA1.VES VAI.VE Fl!NLTION/ VALVE DESIGNAT!GN ISCL ATitW SiCNALS Main Steam Isolation Valves (NS03A, NS03ft, NSO4A, NSO4B) 1 Hsin Steam Condensate Drain Valves (V-1-106, V-1-107, V-1-llo, V-1-111) 1 Reactor Building Closed Ceoling Valves (V-5-147, V-5-166, V-5-167) 2 Instrument Air Valve (V-6-395)

I 1:mergency Condenser Vent Valves (V-14-1, V-14-5, V-14-19, V-14-20) 1 Reactor Cleanup Valves (V-16-1, V-16-2, V-16-14, V-16-61) 3 Shutdown Cooling Valves (V-17-19, V-17-54) 3 Drywell Equipment Drain Tank Valves (V-22-1, V-22-2) 3 lirywell Sump Valves (V-22-28, V-22-29) 3 b~y Drywell G Torus Atmosphere Cont rol Valves (V-27-1, V-27-2, V-27-3, V-27-4, 3

rr V-28-17. V-28-18 V-23-21, V-23-22, g

V '8-4 7, V-23-13, V-23-14, V-23-15, V-23-16, V-23-17, V-23-18, V-23-19, y

V-23-20)

Reactor Recirculation Loop Siunple Valves (V-24-29, V-24-30)

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Torus to Reactor ihailding Vacuum Relief Valves (V-26-16, V-26-18) 3*

Traversing In-Core Probe Systene (Tip machine ball valve No. 1, No. 2, No. 3, No. 4) 3 w

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1) Reactor Isolation Signals as shown in Table 3.1.1

2) Low-Low Reactor Water I.evel and liigh Drywell Pressure; or low-Low-Low Reactor Water Level.

3)l'rimary Containment I*,lation Signals as shown in Table 3.1.1

• Valves automatically reset to provide vacuum relief

4.5-fa l

P.

Suppression Cnamb,er Surveillance 1.

At least or.ce per day the suppression chamber water level and temperature and pressure suppression system pressure shall oe checked.

2.

A visual inspection cf the suppression enamber interior, including water line regions, shall be made at each major refueling outage.

3.

Whenever heat from relief valve operation is being added to tne suppression pool, the pool temperature shall be continually monitored and also observed until tne heat addition is terminated.

4.

Whenever operation of a reitef valve is indicated and the suppression pool temperature reacnes 160*F or above wnile tne reactor primary coolant system pressure is greater than 180 psig, an external visual examination of the suppression enamner shall be made before resuming normal power operation.

4 v

Amendment No. I8, 32, 87,100

4.5-6a-1 Q.

Shock Suppressors (Snabbers) 1.

Each snubber snail be demonstrated operaDie by perfonaance of tne fol-lowing inspection program a.

Visual Inspections All snubbers shall De visually inspected in accordance with the fol-lowing schedule:

No. Inoperable Snubbers Subsequent Visual Per Inspection Period Inspection Period

• 0 18 mentns + 26%

1 12 months T 25%

2 6 months T 25L 3,4 124 days 7 25%

5,6,7 62 days T 25%

8 or more 31 days 1251

• The provisions of Tecnnical Specification 1.24 are not applicable.

The required inspection interval shall not be lengthened more than one Step at a time.

Tne snubbers may be categorized into two groups: tnose access-ible and those inaccessible during reactor operation, nacn group may be inspected independently in accordance witn tne above senedule.

b.

Visual Inspection Acceptance Criteria Visual inspections snall verify (1) that there are no visible indications c' damage cm impaired OPERABILITY, (2) attachments to tne foundation or sup-porting structure are secure, and (3) in tnose locations wnere snubber move-ment can De manually induced without disconnecting the snubber, that tne snubber nas freedom of movement and is not frozen up. Snubbers einicn appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose cf establishing the next visual inspection interval, providing tnat tne affected snubber is functionally tested in the as found condition and determined operable per Specification 4.5.Q.d or 4.5.Q.e as applicable and that tne cause for the rejection has been clearly established and remedied for trat particular snubber.

c.

Functional Tests At least once eacn refueling cycle, a representative sample (101, of the total of each type of snubber in use in tne plant) shall be functionally tested either in place or in a bench test.

For each snuober tnat does not meet the functional test acceptance criteria of Specification 4.5.Q.d or 4.5.Q.e, an additional 107, of tnat type of snubber shall oe functionally tested. As used in tnis specification, type of snubber shall mean snubbers of the same design and manufacturer, mecnanical or hydraulic.

Amendment No. IB, 32, 87,100

4.5-6a-2 The representative sample selected for functional testing shall include the various configurations, operating environments and tne range of size and capacity of snubbers. At least 25% of the snubbers in the representative sdmple shall include snubbers from the following three categories:

1.

The first snubber away from eacn reactor vessel nozzle.

2.

Snubbers witnin 5 feet of heavy equipment (valve, pump, motor, etc.),

3.

Snuncers within 10 feet of the discharge from a safety relief valve.

In addition to the regular sample, snubbers which failed the previous func-tional test shall be retested during the next test period.

If a spare snuD-Der has been installed in place of a failed snuocer, tnen botn tne failed (if it is repaired and installed in another position) and the replacement snubber shall be retested. Tne results from testing of tnese snuocers are not to be included for determining additional sampling requirements.

For any snubber tnat fails to lockup or fails to move, i.e., frozen in place, the cause will be evaluated.

If caused by manufacturer or design deficiency, actions shall be taken to ensure tnat all snubbers of the same design are not subject to the same defect.

d.

Hydraulic Snubbers Functional Test Acceptance Criteria The hycraulic snuboer functional test snall verify tnat:

1.

Activation (restraining action) is acnieved witnin the specified range of velocity or acceleration in botn tension and compression.

2.

Snubber bleed, or release rate, wnere required, is within the speci-fled range in con 1pression or tension. For snubDers specifically required to not displace under continuous load, tne aoility of the snubber to withstand loaa witnout displacement shall be verified, e.

Mechanical Snubbers Functional Test Acceptance Criteria The mechanical snaboer functional test shall verify tnat:

1 1.

The force that initiates free movement of the snubber rod in either I

tension or compression is less tnan tne specified maximum drag force.

l l

l Amendment No.100 l

4.5-6a-3 2.

Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

3.

Snubber release rate, where required, is within the specified range in compression or tension. For snuDbers specifically required not to displace under continuous load, the aDility of the snubber to withstand load without displacement shall be verified.

f.

Snubber Service Life Monitoring A record of the service life of each snubber, the date at wnicn the desig-nated service life comences and tne installation and maintenance records on which tne designated service life is based shall be maintained as required by Specification 6.10.2.1.

Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, tne installation and maintenance records for each snubDer shall be reviewed to verify that t.1e indicated service life nas not been exceeded or will not be exceeded prior to the next scheduled snubber service life review.

If the indicated service life will be exceeded prior to tne next scheduled snuboer service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review.

This reevaluation, replacement or reconditioning shall oe in-dicated in the records. Service life shall not at any time affect reactor operations.

Amendment No.100

4.5-9b of the system. Althougn this is basically a leak test, since the filters have cnarcoal of known efficiency and holding capacity for elemental iodine and/or methyl iodide, the test also gives an indication of the relative ef-ficiency of the installed system.

The test procedure is an adaptation of test procedures developed at the Savannan River Laboratory whicn were des-i cribed in tne Ninth AEC Air Cleaning Conference.*

High efficiency particulate filters are installed before and after the char-coal filters to minimize potentia release of particulates to tne environ-ment and to prevent clogging of the iodine filters. An efficiency of 995 is adequate to retain particulates tnat may be released to the reactor building following an accident. This will be demonstrated by testing witn DOP as testing medium.

If laboratory tests for tne adsorber material in one circuit of the Standby Gas Treatment System are unacceptable, all adsorber material in that circuit shall be replaced with adsorbent qualified according to Regulatory Guide 1.S2.

Any HEPA filters found defective shall be replaced with tnose quali-fied with Regulatory Position C.3.d of Regulatory Guide 1.62.

The snubber inspection frequency is based upon maintaining a constant level of snubber protection. Thus, the required inspection interv61 varies in-versely witn the observed snubber failures.

Tne number of inoperable snub-i bers found during a required inspection detennines tne time interval for tne next required inspection.

Visual Inspections perfonned before an inspection l

interval has elapsed may oe used as a new reference point to determine tne next inspection. However, tne results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not De used to lengthen tne required inspection interval.

Any inspection whose results require a shorter inspection interval will overrice the previous schedule.

To furtner increase the assurance of snuober reliability, functional tests snould be performed at least once each refueling outage. Tnese tests will include stroking of the snubbers to verify proper piston movement, lock-up and oleed. Ten percent of tne snuboers represents an adequate sample for such tests.

Ooserved failures of tnese samples require testing of additional units.

• D. R. Munbaier, "In Place Hondestructive Leak Test for Iodine Adsorbers", Proceedings of tne Ninth AEC Air Cleaning Conference, USAEC Report CONF-660904,1966.

Amendment No. 18,100

'6 10 2

• The following records snall be retained for the duration of tne Facility Operating License:

a.

Record and drawing changes reflecting facility design modifica-tions made to systems and equipment described in tne Final Safety Analysis Report, b.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup nistories.

c.

Records of facility radiation.and contamination surveys, d.

Records of radiation exposure for all individuals entering radi-ation control areas.

e.

Records of gaseous and liquid radioactive material released to the environs, f.

Records of transient or operational cycles for tnose facility components designed for a limited number of transients or cycles.

g.

Records of training and qualification for current members of tne plant staff.

h.

Records of inservice inspections performed pursuant to tnese technical specificatons.

i.

Records of reviews perfonned for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

j.

Records of reviews by the Independent Onsite Safety Review Group.

k.

Records for Environmental Qualification wnicn are covered under the provisions of paragraph 6.14.

1.

Records of tne service lives of all snuboers, including tne date at which the service life cormaences, and associated installation and maintenance records.

6.10.3 Quality Assurance Records shall oe retained as specified by the Quality Assurance Plan.

6.11 RADIATI0fl PROTECTI0ri PROGRAM Procedures for personnel radiation protection shall be prepared consistent witn the requirements of 10 CFR 20 and shall De ap-proved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 (Deleted) 6-23 Amendment No. O, 78,100 4

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