ML20101F803

Application for Amend to License NPF-3,revising Tech Specs to Incorporate Revised Reporting Requirements of 10CFR50.72 & 10CFR50.73.Marked-up Tech Specs Encl.Fee Paid

ML20101F803
Person / Time
Site:	Davis Besse
Issue date:	12/16/1984
From:	Crouse R TOLEDO EDISON CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
1108, TAC-55704, NUDOCS 8412270305
Download: ML20101F803 (18)

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%ms EDISON Docket'No.-50-346 Rom P. CAOUSE License.No. NPF-3 vc= P=~t

.Seria1 No'. 1108 N.,3.m, December 16, 1984' Director of Nuclear Reactor Regulation Attention:

Mr. John F. Stolz Operating Reactor Branch No. 4 Division of Licensing United States Nuclear Regulatory Commission Washington, DC 20555

Dear Mr..Stolz:

Under separate cover, we are transmitting three (3) original and forty (40) conformed copies of an application for Amendment to Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Power Station Unit No. 1.

This application requests that the Davis-Besse Nuclear Power Station Unit 1 Technical Specifications, Appendix A, be revised to reflect the changes attached. The proposed changes involve Sections 1.7, 3.3.3.8, 3.4.8,.3.7.9.1, 3.7.9.2, 3.7.10, 6.5, 6.6, 6.9 and 6.10.

'The attachments identify the changes, Safety Evaluation and Significant

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Hazard Considerations for the proposed amendment. The amendment request

. concerns the reporting requirements as amended by 10 CFR 50.72 and 50.73.

The request revises the Technical Specification to comply with the rule as outlined in Generic Letter 83-43 dated December 19, 1983 (Log 1423).

Enclosed is a check for $150 as required by 10CFR170.12(C) for license application.

Very truly yours,

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RPC: GAB rf Attachment cc: DB-1 NRC Resident Inspector d

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ADOCK 05000346 p

PDR THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OH10 43652

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' APPLICATION FOR AMENDMENT-TO

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FACILITY OPERATING LICENSE NO. NPF y FOR

DAVIS-BESSE NUCLEAR POWER STATION 1_

UNIT NO. 1 Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit No. 1 Facility Operating License No.- NPF-3, together with the Safety Evaluation for the requested change.

The proposed changes include Sections 1.7, 3.3.3.8, 3.4.8, 3.7.9.1, 3.7.9.2, 3.7.10, 6.5,;6.6, 6.9 and 6.10.

By /s/ R.,P. Crouse Vice President,, Nuclear Sworn and subscribed before me this-16th day of December, 1984.

/s/ Laurie A. Hinkle, nee (Brudzinski)

Notary Public State of Ohio My Commission Expires May 16, 1986.

SEAL

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't-S, Docket'No. 50-346:

. License No.'NPF-3 Serial No. 1108 December 16, 1984

Attachment:

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Changes to Davis-Besse. Nuclear Power Station Unit-1, Appendix A-

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Technical Specifications 1.7,.3.3.3.8, 3.4.8, 3.7.9.1, 3.7.9.2,,

- 3.7.10, 6.5, 6.6, 6.9 and 6.10.

.A.

Time required.to Implement. This change is to be effective-

.upon NRC approval.

B.

Reason for Change (Facility Change R' quest 84-112 Rev.' A).

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.To comply with the revised reporting rule 10 CFR 50.72 and 50.73 as contained in Generic Letter 83-43 dated December 19, 1983

'(Log No. 1423).

C.

Safety Evaluation

-(See. Attached) i

.D.

Significant-Hazard ~ Consideration (See Attached) l f

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SAFETY EVALUATION

. This amendment request'is to revise the Davis-Besse Technical Specifications to. incorporate 'the revised'" Reporting Requirements" of 10 CFR 50, Section 50.72 and Section 50.73 (new),- incorporated in the NRC Generic Letter

, 43..The' sections of the technical specifications affected are.l.7, 13.3.3.8, 3.4.8, 3.7.9.1, 3.7.9.2, 3.7.10, 6.5, 6.6, 6.9 and 6.10.

The safety function of the revised technical specification sections on

" Reporting Requirements" is to provide timely.and. accurate-documentation of _ plant events with safety significance and outline the actions to prevent reoccurrence..These. reports are submitted to the NRC to aid them

- in exercising their responsibilities for protecting the public health' and safety.

The revised NRC reporting requirements, as outlined ~in Generic Letter 83-43, will accomplish the following:

.1.

The reporting requirements are now uniformly applied to all power reactors, since technical specifications from different plants will all reference the same rules.

(10 CFR 50.72 and 50.73.)

2.

. Eliminate-reporting of minor incidents with the anticipation that they will be covered by the INPO Nuclear Plant Reliability Data (N2RD) program.

3.

Add to the list of reportable events certain events which were

-considered not reportable before, such as reactor trip and actuation of ESF Systems.

The revised' requirements do not degrade the safety function of the affected Technical Specification Sections. Therefore, the changes as proposed do not involve an unreviewed safety question.

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r-SIGNIFICANT HAZARD CONSIDERATION i

4This~ amendment request'to revise the reporting requirements contained in

'the Technical ~. Specifications'does'not represent a Significant Hazard.

In-May_1982 the NRC proposed rule making that would modify and codify the

' existing Licensing Event Report.(LER) System.

A. revised rule was approved 1with the addition-of a new section concerning LER's..On December 19, 1984 1(Log ~No; 1403) the'NRC; issued Generic Letter (84-43) requesting licensees nto revise-their Technical Specifications to reflect the LER rule change.

This amendment request is a result of that request.

The amendment request revises the nomenclature and defines reporting requirement to be consistent with the rule. 'Some of these requirements

' constitute an additional limitation not previously; required of the licensee.

The' Commission has provided examples of amendments which are not likely to involve a.significant hazards. consideration (48 FR 14870), such as a change that constitutes an additional _ limitation, restriction,' or control not presently included in the technical specifications:

for example, a more stringent surveillance requirement (example 11). The revised rule-(10 CFR 50.72 and 50.73) add an additional limitation not previously required by the licensee. The additional items to be reported will aid the commission in identifying types of reactor events and problems that are believed to be significant and useful in aiding the NRC in its efforts

to identify and resolve potential safety significant to the public health-and safety.

. Based on_the above.information, this amendment request would not 1)

~ involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3)

-involve a significant reduction in a margin of safety.

'Therefore, based on the above, the requested licerise amendement does not present a Significant Hazard.

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's DEFINITIONS i

EW4r REPORTABLE 4CC"""~'!CE A REPORTABLE C_:"""$:C" shall be any of those conditions specified 1.7 SCTlM 50t 3 to /0dFR Att50.

7 in CONTAINMENT INTEGRITY 1.8 CONTAINMENT INTEGRITY shall exist when.

All penetrations required to be closed during accident con-a.

ditions are either:

21..- Capable of being closed by the Sa etyleatures ' Actuation U~

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System, or Closed by manual valves, blind flanges, or deactivated 2.

automatic valves secured in their closed positions, except as provided in Table 3.5-2 of Specification 3.6.3.1.

b.

All equioment hatches are closed and sealed, Each airlock is OPERABLE pursuant to Specification 3.6.1,3, c.

The containment leakage rates are within the limits of Specificatter.

d.

3.6.1.2, and The sealing mechanism associated with each penetration (e.g.,

e.

welds, bellows or 0-rings) is OPERABLE.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the 1.9 channel outout such that it responds with necessary range and accuracyThe CHANNEL to known values of the parameter wnich the channel monitors.

CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TE CHANNEL CALIBRATION may be perfonned by any series of sequential, over-lapping or total channel steps such that the entire channel is calibrat CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel

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This determination shall 1.10 behavior during operation by observation.

include, wnere cossible, comoarison of the channel indication and/or status with otner indica-icos and/or status derived frcm ir.decencent instrument :hannels measuring tne same paramstar.

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1-2 DAVIS-BESSE, UNIT 1

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INSTRL?.ENTATION

- FIRE DETECTION INSTRU?fENTATION_

1,IMITDIC CONDITION FOR OPERATION As s'ainimum, the fire detection instrumentation for each fire 3.3.3.8 detection some shova in Table 3.3-14 shall be OPERA 3LE.

Whenever equipment in that fire detection some is required l

APPLICA51LITT:

- to be OPEMBLE.

With the number of OPERABLE fire detection instrument (s) l h

the miniawn number OPERABLE reguirement c. s Table 3.3-14:

ACTION:

Wkthin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a fire vacch patrol to inspect the least accessible zone (s) with the inoperable instrument (s) at a.

i once per hour, and Restore the inoperable instrument (s) to OPERA 3LE status withim 2 L, ^r-.*f' b.

14 days orf ' '

f - _. --'-- ;,' Q prepara and submit a Special Report to the Connaission pursuant to specification 6.9.2 within the next 30the cau days outlining the action taken,and the plans and schedule for rest CPERA3LE status.

The provisions of specifications 3.0.3 and 3.0.4 are not k

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ayylispble.

sutvsguNCg RavGIREMENT5

,s 4.3.3.8.1 Each of the above required accessible fire detection instruments shall be demonstrated CFERABLE at leas $ once per 6 months by performan Each of the above required inaccessible i-of a CHANNEL FUNCTIONAL TEST.

least once fire detection instrument shall be demonstrat'ed CPERA3tE at per 18 months by performance of a CHANNEL TUNCTIONAL TEST.

4.3.3.8.2 The NFPA Code 72D Class A supervised circuits supervision i

d' fire associated with the detector alarms of each of the above requ re least once per detecties instruments shall,be demonstrated CPERABLE at i

6 months.

4.3.3.8.3 The poa-supervised'elrevits between the loce.1 panels la specification 4.3.3.3.3 and the sogtyet room sh411 ha demonsstated OPERAaLE at least once per 35 days, '

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Nepdment No. h7 f 3/4 3-52

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i REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY t

LIMITING CONDITION FOR OPERATION 1

3.4,8 The specific activity of the primary coolant shall be limited to:.

a.

< 1.0 uCi/ gram DOSE EQUIVALENT I-131, and i

...b.

. < 100/E UCi/ gram

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I APPLICABILITY: MODES 1, 2, 3, 4 and 5.

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ACTION:.., e m.-

MODES 1, 2 and 3*.

With the scacific activity of the primary coolant > 1.0 uCi/ gram '

a.

00SE.EQu1 VALENT I-131 but within the allowable limit.(below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided that operaticn under these

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circumstanceTsha11 not exceed 10". of the unit's total yearly operating time. The provisions of Specificat16n 3.0.4 are not applicable.

b.

With the specific activity of the primary coolant > 1.0 vC1/ gram DOSE EQUIVALENT I-131 for more.than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T,yg < 530*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With the specific activity of the primary coolant > 100/?

c.

< 530*F uC1/ gram, be in at least HOT STANDBY with T"V9 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

MODES 1, 2, 3, 4 and 5:

With the specific activity of the primary coolant > 1.0 uCt/ gram a.

DOSE EQUIVALENT I-131 or > 100/E uCi/ gram, per#orm the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within

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of ut vees n. estmb) analyses along wth ye,fu/kayD,

Witn i,yg > 530* F.

DAVIS-BESSE, UNIT 1 3/4 4-20 e

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PLANT SYSTEMS 3/4.7.9 FIRE SUPPRESSION SYSTEMS __

FIRE SUPRESSION WATER SYSTEM-LIMITING COND'ITION FOR OPERATIOil 3.7.9.1 The fire supression water system shall be OPERABLE with; Two high pressure pumps, each with a capacity of 2500 gpm,

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a.

with their discharge aligned to.the, fire suppression header,

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b.

Separate water supplies, e'ach with a minimum contained volume of 250,000 gallons, and An OPERABLE flow path capable of taking suction from the Intake c.

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Forabay and the Fire Water Storage Tank and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves and the first valve head of the water flow alarm device on i

i each sprinkler hose standpipe or spray system riser required

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to be OPERABLE per Specifications 3.7.9.2 and 3.7.9.3.

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APPLICABILITY: At all times.

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ACTION:

J With one pump and/or one water supply fpoperable, restore the

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a.

inoperable equipment to OPERAGLE status within 7 days o% 4e--

l pre-pare and submit a Special Report

• to the Comission pursuant to lI Specification 6.9.2 within the next 30 days outlining the piens 1

and procedures to be used to provide for the loss of redundancy in this sytem. The provisions of Specifications 3.0.3 and

's 3.0.4 are not applicable.

With the fire suppression water system otherwise inoperable:

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b.

1.-

Establish a backup fire suppression water system within k) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 2.

Submit a Special Report in accordance with Specification 6.9.2; a)

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

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Confirmed by talcorsch, mailgram or facsimile trans-mission no later than the first working day fcilowing the event, and wn c.n eur mifT 3 3/4 7-38 Amendment flo. I

PLA:iT SYSTD'S SPRAY A!:0?0R 57RI::KLER SYSTEP.S LIP.!TI!!G C0'*0!TIO!! FO?. 0?EP.ATION 3.7.9.2. "The fo11cwing spray and/or sprinkler systems shall be OPEP.ASLE:

a.

Diesel Generater Day Tank Rooms 320A and 321A b.

' Diesel Generator Roc:ss 318 and 319 APPLICA3ILITY: Nherever equipment in the spray / sprinkler protected areas,

is requires to be OPERA 3LE.

. ACTIO:1:

a.-Wi-th-one-or-core of-the above requires soray and/or sprinkler syste-.s ir.: parable, establish a continu:us fire wat:h with backup fire suppressica equiprent for the unprotected ares (s) within 1 hourt restere the systs:1 to C?ERAELE status within IA days or. "

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4rev W prepara and sub=it c Special Rc;cr: to tne C:.=icsf on i

pursuant to Specificatica 6.9.2 within the next 30 days cut-t lining the actica tai en, the cause of the incperability ar.d the t

plans ans schedule for restoring the systen to GPERABLE status.

b.

The provisions of Specification 3.0.3 and 3.0.4 are not applicab'le.

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SURY! ALLA*:CE P.E0'JIP.E'T:TS l

l 4.7.9.2 Each of the above required spray and/or sprinkler systems'shall be demonstrated CPERASLE:

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a.

At least once per 12 months by cycling each testable valve in the l

flow path through at least one complete cycle of full travel.

1 b.

At least once per 18 months:

, 1.

By perfortiing a sysf em functional test which includes simulated automatic actuation of the system, and:

l a)-

Verifying theat the autor.stic valves in the fica path actuate to, their correct positions en a test signal, and DAVIS-GCSSE, C::lT 1 3l4742 Amendment 1:o.f

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' ' 3/4.7.10 FIRE BARRIER PEHETRATIONS LIMITING CON 0iTION FOR OPERATTON r

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i 3.7.jo All, fire barrier penetrations (including cable penetration barrieri

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fire onors and fire dancers) in. fire zone boundaries protecting safety

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related areas shall be functional.

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l APPt.!CA8ICITY: At all tisas.

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With one or more of the above required fire barrier penetrations non-functional, within one hour either, establish a continuous fire a.

watch on at least one side of the.affected penetration, or verify the OPERABILITY of fire, detectors on at least one side of the non-

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functional. fire barrier and estatritsh an ' hourly fire watch pa' trol.,.

Restore the non-functional fire barrier penetration (s) to func:fonal

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' i, status vithin 7 days or L:.":___._.. ;.;. L prepara and submit a Specia Report to the

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1 Consission pursuant to Specification E.9.2 within the next 30 days s

I outlining the action taken, the cause of the non-functional pene-penetration (plans and schedule for restoring the fire trati6n and s) to functional status.

The provisions of Specificaduos 3.0.3 and 3.0.4 are not applicable.

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JURVEI!. LANCE RE01.lIRD4fMT5 I

P.7.10 The above required penetration fire barriers shall be verf fled l

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s be functicna

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. At least once par 18 months by a visual inspection.,

a.

b.

• Prior to. returning a penetration fire barrier to functional -

status following repairs or maintenance by performance of a visual inspection,of the affected penstration fire barrier (s),

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".AY!S-iiESSE, UNIT 1 3/4 7-47 Amendment No.

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3 I ADMINISTRATIVE CONTROLS

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c.

Review o'f all proposed changes to Appendix "A" Technical Spccifications.

d.

Review of all-proposed changes or modifications to plant systems or equipment that affect nuclear safety.

e.

Investigation of all violations of the Technical Specificationt.

4 including preparation and fonvarding of reports covering evalua-4 f

tion and recomendations to prevent recurrence to the Vice 4:*-

President - Nuclear and to the Chainnan of the Company Nuclear Review Board.

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g.

Review of facility operations to detect potential safety hazards.

h.

Performance of special reviews, investigations and analyses and reports thereon as recuested by the Chairman of the g

Company Nuclear Review Board..

i.

Review of the Plant Security Plan and implementing procedures and shall submit reconnended changes to the Chairman of the Company Nuclear Review Board.

j.

. Review of the Emergency Plan and implementing procedures and

'shall ' submit recensended Shanges to the Chairsan of the Company Nuclear lteview Boprd.

AUTHORITY-6.5.1.7 The Station Review Board shall:

s.

Recomend to the Station Superintendent written approval or disapproval of items considered under 6.5.1.6(a) through (d) i above.

A b.

Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

ri c.-

Provide written notification withjg 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> la the Vice 7

President, Nuclear and the Company Nuclear Review loard

',l of disagreement between the SRB and the. Station Superintendent;

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however, the Station Superintendent shall have responsibility

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for resolution of such disagreements pursuant to 6.1.1 abcve.

3 AVIS-BESSE, UNIT 1 6-7

/aend:ent No.

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ADMINISTRATIVE CONTROLS REVIEW' l

6.5.2.7 The Company Muclear Review Board shall review:

a.

The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such' actions did not constitute an unreviewed safety question.

b.

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59,10 CFR.

' c.

Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

d.

' Proposed changes in Technical Specifications or this Operating License.

e.

Violations of codes, regulations, orders. Technical SpecifRatibns, license requirements, or of internal procedures or instructions having nuclear safety significante.

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f.. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

Eeg Raamtms Eens.-_..c..,...::; 21 h;g..:.....,'ir. tic: R O.; i ci::f g.

h.

All recognized indications pf an unanticipated deficiency in some aspect of design or operation of safety related structures.

h.,.;, systems, or,. components,

1.

Reports' and meetings minutes of the Station Review Board.

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DAVIS-BESSE, UNIT 1 6-10 I

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ADMINISTRATIVE CONTROLS RECORDS 6.5.2.10 Records of Company Nuclear Review Board activities shall be prepared, approved and distributed as indicated below:

.a.. Minutes of each CNRS meeting shall be prepared, approved and forwarded to the President and Chief Operating Officer and CNR8 members within 14 days following each meeting.

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b.

Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared.-approved and forwarded to the President and Chief Operating Officer and CNRB members within 14 days following completion of the review.

c.

. Audit reports encompassed by Section 6.5.2.8 above, shall be

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forwarded to the President and Chief Operating Officer and CNRS members and to the management positions responsible

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for the areas audited within 30 days after completion of the 1

audit.

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6.6 REPORTABLE-666ttRReMee ACTION

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6M 6.6.1 The following actions shall be take,n for REPORTA8LE 966WARf!fHiE9.

a.

The Comnission shall be notified and/or' a report submitted

. pursuant to the requi nts of CF..;1.... -SEAT 18W40.73 16 LFR. RNCT50; e

..' b.

Each REPORTABLE 000',;;.',0.";Q f; ' '.y t:"$ C:--.

shall be revi d by the SR8 a N

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'1 ADtINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6 6.9.1 In addition to the applicable reporting requirements of Title 10 Code of Federal Regulations, the following reports shall be submitted to the appropriate Regional Office unless othemise noted.

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p,ARTUPREPORT 6.9.1.1 A sucr.tary report of plant startup and power escalation testin shall be submitted.following (1) receipt of an operating license. (2) g amendment to the license involving a planned increase in power level.

(3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered tne nuclear, thermal, or hydraulic perfor-mance of the plant.

6.9.1.2 The report shall address each of the tests identified in the FSAA and shall include a description of the measured values of the operating conditions or characteristics obtained during the tast program and a comparison of these values with design predictions and specifica-tions. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details g

required in licensa conditions based on other comitments shall be in-cluded in this report.

6.9.1.3 Startup repcets shall be submitted within (1) 90 days following completion of the startup test program (2) 90 days following resumption or commencement of cosmerical power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e.

initial criticality, completion of startup test program, and resumption or concencement of commercial e

4 QAVIS-BESSE UNIT 1 6-144 Anendment No, 9

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ADMINISTRATIVE CONTROLS y--ih law bELET.E 6.9.1.

The REPORTABLE QCCURREN'CEs of Specifications 6.9.1 and 6.9.1.9, includi corrective actions and measures to preventc rocu nee, shall be repo to the NRC. Supp.lemental reports may be re red to fully describe-f I resolution of occurrence.

In case of cc ected or suppl emental sports, a licensee event report shall b completad and reference sha be made to the original report date.

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6.9.

5.The types events listed below shall e - reported wi thin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> telephone an confinned by talegraph, 11 gram, or facsi le transai ion to the Di ctor of the Regional fice, or his desi ate no later an the first orking day followi the event, with a cittan followup rt within tw weeks. The wri n followup report hall include,as minimum, a e plated c:py.o a licensee event r pcrt ferm.

Information p vided on the icensas eye. report form shal' be su;pla-mented, as nee i, by additio al narra* ve material to pro-de c:=pleta explanation of t circumstanc surr nding the event.

a.

Failure o the reactor taction system or 5er systams

~

sub,fect to initing saf systam settings initiata the i

required pro. ctive f et n by the time monitored para-l-

~matar reaches a se,, int pacified as he limiting safety.

1 system setting techni i specif ations or failure to complete the r uired pro' tive action.

b.

Operation of th uni or affect ystems when any paramatar or operation s ject a limiti condition for operation' is less consa ative th the as, conservative aspect of the limiting con ition for o a on est lished in the technical r

specificati ns.

c.

Abno'rmal agradation di ver in fuel adding, reester

. coolant ressurs. bound

, or p mary con inment.

t d..

React ity anc=ali' involving dis reement th the predictad valu of reactivi balance under s ady stat conditions du'r g power op ation greatar than equal

• 1% ak/k; a ca ulated rea.ivity balance indicat a SHUT,,'4M MARGIN 1

s conserv ive than specified in the achnical pacifica-ons; shor+ term, reactivity increases th t corres nd to a reactor.riod of less than 5 sac:nds or, if subc 'tical, an unplan ed reactivity insertion of more th 0.5: a k; er L

occurren a of any unplanned criticality.

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Ar,.ene=entac.,.K.

mvts-! Esse,untTi s-1s t

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ADMIM1STRATIVE CCMTM01.S'

'e.

Failure or selfunction of one or more components = ch prevents 1

r could prevent. by itse]f. the fulfillment of functional irements of system (s) used to cope with acc ents analyze in e SAR.

- Person el error or procedural inadequacy w ch )revents could prevent, by itself, the fulfillasnt of th functional uire-ments of taas required to cope with a idents anal in the SAA.

g,, C ditions a sing from natural or n-made events at, as a di t result f the event require lagt shutdown operation

~

of s ety sys L. or other prote ive measures utred by tachni i specif tions.

l h.

Errors di vered i the trant ent or actiden analyses or in the mathe used for uch an~ yses as descri d in the safety analysis re or in is b es for the te ical specifications that have or uld hav p istad reactor erstion in a manner less conservati e than used in the ana es.

~

1.

Performance of st ctu ystans, or onents that requires remedial action or ecti asasu to prevent operation in a manner less ce rvativ than used in the accident analyses in the sa t a.ialys rep or technical specifica-tions bases; or scovo duri p at life of conditions not specifically co idered the ety analysis report or tecnnical spec icatiotts tr tre remedial 4ction or cor-rective meas" es to prevent 5 ex tance or development of an unsafe cond' ion.

h-

_ N 8 6.9.1.9 The ty s of events its below sha b the subject of writtan reports to the tractor of the R ional Office th n thirty days of occurrence of a event. The w ttan report sha i clude, as a sinimum, a ccepleted py of a license vent report form, n rmation provided on the lic see event report ora shall be supplane e as needed, by additiona narrative matari to provide completa ex a tion of the circuss aces surrounding e event.

9 Reactor prote ion system or engineered safaty ea e instru-ment settingr which are found to be less conse ti than those estab shed by the technical specifications aut hich do not preven the fulfillment of the func 14nal requ ts of affected stans.

Amendment No..&,7,

QAVIS-SE3SE.

6-17

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' Routine surveillance instrument calibrat maintenance which require sys nf as described in Section 6.9.1.9.a and 6.9.1.9.b ne~

=d except where test results themselve.s reveal aced condition requi rective action.

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ActfrNI57RATTVE CD'NTRc! 5 b.

Conditions Ina -

to operstian.in a d unds peraf a limiting con on for operati plant shutd qu by a 11st' ting' ttia operatina.

I c.

Observed daquacia:

lanantat acministratiYt cause reduction of or procadura

,s which

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degras of re provided 1 protsetion systazs ture cans.

d.safat,.

or engt spgstfissi]

d.

1 degradatio a

other thin a;spjg

, in E.S.I.B.c ab igned 1g tajpyp-ing rusniting the ffssisa p p 4 SPECIAt. REPORT 3-6.3.2 special nports.shall be suhzittad to the Director ef the Office of Inspection.and Enforement Regional:0ffica within tne time period Thess rsportz shall be sutmittad covsr:ing specified for each r= port.

the activities identified.below pursuant to the requirements.cf tha.se -

plicable rsfersace specification:

s.

EC:5 Actuation specifications 3.E2 and 32,:3.

.. h.

Insperable.3sismic finnitoring huntion. 3pscification 1

3.3.3.3.

Inoperable Mstmorological Planitoring Instnmentation. Speciff-c.

cation 3.3.3.4.

d.

3aismic event analysis. Specification 4.3.3.3.3.

Firs.Detsetion hw astion 5ps.ification.3.3.3.3.

a.

Fire supp'rsssion Systmas. 3pecifications 3.7.5.1 and 3.7J.2.

f.

Fire Barrier Penetrations, Sascification 3.7.10 f

g.

E.10 RECDRD b.dTTON 6.10.1 The following r_-c=rds shall be reuined fer at least fits yters:

Recdds and logs of-facility operation. covering tins interval a.

at each. power level.

Records and logstof prf.zeipai maintsnanca. activities. inspeci:fons, b.

repair and rsplacmant of: principal itms of equipment relatzd tm nucisar safety.

EVEARS c.

ALI. RUwATABLE Records of surveillancs activities inspections and eslibrations d.

rui;uired by these Technical.3pssffications.

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