IR 05000237/1990022

Insp Repts 50-237/90-22 & 50-249/90-22 on 900628-0920. Violation Noted.Major Areas Inspected:Practice of Utilizing Temporary Sample Pump to Obtain Daily Drywell Air Sample

ML17202U844
Person / Time
Site:	Dresden
Issue date:	10/04/1990
From:	Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML17202U845	List: IR 05000237/1990022
References
50-237-90-22, 50-249-90-22, NUDOCS 9010170156
Download: ML17202U844 (11)

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• U. S.' NUCLEAR.REGULATORY COMMISSION

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REGION I II Repd~t Nos: 50-237/90-022(DRP);* ~6~249/90-022(DRP)

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Docket Nos. 50-237;.59-249

-~ice. n s e" Nos. DP'R-19; DP.R-25 licerise*e:

Commonwealth Edison *Company (Jpus. West I I I 1400 Opus Place Do~ners Grove, IL 60515 Facility Name:

Dresden Nuclear *Power Station, Units 2 and 3 *.

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• * Inspecti'on..At:.

Dresden '.'site.;. Mo~ris,, *ri"li~ciis fnspeciion* Cond~cted: June 28* thro0gh September 20, 1990

.Inspector:

0..E. Hi!ls*.

Approved Section 18

• Inspection Summary

O ~A.-- 9-o Date *

.. Inspection during the period of June 28 through September 20, 1990.

• **(Report Nos. 50-237/90-022(DRP); 50~249/90-022(DRP))'

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Area~ Inspected:

Special, ~nnounced safety inspection of.the licensee's

'.previous practice of utilizing a temporary sample*pump to obtain the daily drywell air sampl (Modi,Jle 92701)

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Results:

The inspection r~sulted in the identification. of one apparent

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10 CFR 50.59 ~iolation in that th~ litensee'~ pra~tice effectively constitut~d a change in r:echnical Specifications and an unrev]ewed safety question existed in regard to.the tempo.rary* sample pump. *Prior NRC approval was not sought or obtained..

Th.is temporary 'alteration'reduced the margin of safety as defined in ~he basis of Technical Specific~tions in regard to the ma~imum allowable

• pr.imary containment accident leak rat Primary containment was effectively de~raded to unacceptable levels whenever the daily drywell air samples were being obtained with the temporary sample pum (50-237/90-022-0l(DRP);.

249/90~022-0l(DRP), paragraph 4).

. *The apparent_ violation reflects adversely on :the safety assessment/qual.it verification.and engineering/technical support functional area It represents a sign~ficant failure to meet the requirements of_lO CFR 50.59 requirements in that a required license amendment was riot sought prior to implementing a facility chang This effectively circumvented the NRC's role in the regulatory proces Analyses can be performed which may show, using more realistic assumptions than the more conservative assumptions contained in the plant licensing basis,.that offsite and control room dose.projections are 9010170156 901004

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within acc:eµfablle. criteri However, the determination of the safety

. significance caf the change (whether the change is safe or ui'isafe) for an*.

unreviewed ~s:a*-f.BiM question *;s clearly an NRC fu,ncti9n and not within* the authority 'CO:l :thE 1 i cen se :.

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• DETAILS Persons Contacted Commonwealth Edi.son Company

• E. Eenigenburg, Station Manager

• L. Gerner, Technical Superintendeht D. Van Pelt, Assistant Superintendent -*Mainteriance J. K6t6wski~ Production Superintendent J. Achterberg, Assistant Superintendent :--Work Planning G. Smith, Assistarit Super1ntendent-Operations

• K. Peterman, Regulatory Assurance* Supervi ~or M. *Korchynsky, *Operating Engineer

B. Zank, Opera:t i ng Engineer

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J. Williams, Operating 'Enginee * Strai~. T~chnical Staff ~upervisor

~*L. Johrisbn, -O:C. Supervisor

• -o. Morey, Chemistry Services Supervisor Sacto~ando, Health Physics.Services Supervisor

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The insp;~ctor-~lSo talked-w';th and*interviewed*several other liceris-ee employees,: inc]uding members of the technical and engineering staff *

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.**Denotes those attending one or more exit interviews.conduc.ted informally ai vafiou~ times throughout the inspection period:

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.. Licens-ee Actions on. Previously Id~?"ntified Items* (92701) * *

• • -(Closed) Unresolved Item'(50-237/90017-04(DRP)).

This.item concerned the licensee's past practice of utilizing a temporary sample pump to obtain daily drywell air sample This action. created an unattended and

. unmonltored vent path from the drywell (pr1mary containment) through the sample line to the reactor building (secondary containment).* This item was open pending completion of a licen.see 10 CFR 50.59 safety evaluation re*gardi,ng this practic The-*lice.nsee's safety evaluation was completed and ~s-discussed in paragraph 5 of this*.repo~t. This item ~lso conterned the adequacy *of the drywe,.l man,ifold sample system containment isolation prox1s1ons.. Further review indicated that the conta.inment isolation provisions-for this system were approved by the NRC*in a sa*fety.

. -,Evaluation Report (SER) dated March 5, 1980, in regard to N-UREG-0578 Category A Item 2.1.4 (NUREG-0737 Item II.E.4.2) "Containment Isolation."

As such, the inspector has no further concerns regarding*this portion of the ite The revfew of licensee actions in regard to. the temporary sample pump usage and the resulting affect upon primary containment indicated an apparent violation of 10 CFR 50.59 as discussed in paragraph 4 of this report. *Since the apparent violation involving the temporarj sample pump will be.tracked as a separate item.and the NRC previously approved the containment isolation provisions for the drywell manifold sample system, this unresolved item is considered closed.

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Background Drywe1r*Manifold SampleSystem Description

.The purpose of the drywell' manifol~ sample system is *to provide air

.. samp 1 es tci i de.nt ify the 1 ocat ion of reactor coo 1 ant pressu*re **.

boundary leaks inside of the drywel 1.,The drywell manifold sample system (one for each unit) is desi'gned to take a.suction from...

~2 s~mple points in primary containment with each half inch sample line having its own two manual primary containment isolation valves

'(both located outside *of primary containment) and a filter cartridg Flow 'then 'passes through a.common header from which the sample pump

• takes a suet i or:i.. Retur*n back to the primary containment is provided

. ~hrough.a* *connection to the cont 1 nuous oxygen monitori'ng system

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. which discharges fo.the drywell through tw9 automa~ic contafnmen.t :

isolati'on 'valv~s whi.ch ~lo.se* on a Group II isolation: s.ignal....Thus, the drywell mani.fold. sampling system.has automatic isolatio.n only on its djscharge;. The* co~tainment 'isolation: provisions for thissystem were approv~d by~~he NRC in i.SER d~ted.. March 5, 19§0*in rega~d to

. ~UREG-0578 Cat'egory A Item *2.1.4 (NUREG-0737 Item II.E.4.2),

• '11 Containment Isolii.tion. 11

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.. *Daily Orywel'l Air.Samples* Utiliiing.Temporary Sample. Pump.'

Since 1978 and possibl.y before,.. th~ license~ used a.t~mporarY sam.ple

. pump. as a back4p method 'to obtain the Jechnical Specification.-...

. required "da i 1 y drywe 11.. a i r sampl use of 'the. tempera ry 'samp 1 e' *~~i'ump was frequen.t, especially in the last couple_ oLyears.due.to..

Tecurrin~ pr~blems with the.p~rmanent pumps; The licensee tndic~ted *

that the perma~ent pum~s,were operable *only a. few weeks ~hrough the major portion of 1988 th.rough.199 Jhe licens_ee,,p:lso indi-cated that the reliabilitY:Pf these ~umps*was prior pri~~ ~o 198 Use of

. the temporary sample pump involved brea~ing the... cJosed *loop on the drywell manifold sample system below the sample filter on one of the samp 1 e. 1 i nes, attaching a rubber hose 'with a qui ck disconnect

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• .. fi.tting, *connecting the hose to t'he temporary sample pum*p and

. discharging the pump exhaust to the reactor bu.ildi The system was left unattended whil~ a *sam¢le was being takep although iutomatic isolation ~as noi *provide Obtaining a representative

• .sample required r~nning the-sys~em in this conf~guration for at least* 50 minutes. '.(A subsequent* procedure specified a minimum of one hour.)' This all-0wed an unatt~nded and unmoni~ored path from

~he drywell (primary containment) through the sample line to *the reactor buil~i~g (secondary containment).

A proced~re was written on May 25, 1989 to cover.this operation due to a riondocumented third party reviewer commen This procedure contained a* prerequisite to

• notify the control room prior to sampling and a* precaution that the two valves *upstream of each filter :holder must be closed when drywell isolation is require No analysii was done by the l*icense~ td determine the effect on the offsite and control room doses in

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consideration of. manuar reaction time and accessibility durfog',

• design basis:-accidents. ::The licensee's technical staff system engine~r identified the pr~blem on June 28, 19~ '

. Ramifications of Temporary Sample Pump Usage This use of the temporary samp 1 e pump i.n this con.figuration was contrary to Techn'i cal Specification. 3*. 7.A. 2 which r~qutred primary containment integrity when the.reactor was*critical of.the reactor water temperature was 'above 212 degrees F.: (The definition* of *

primary containment integrHy requires that all manual isolation, v~lves on lines connecting to c~htainment which are not 'required lo be open during accident conditions ~re_closed.)* *In this

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configuration,.a:t a postulated 'design _basis* loss of coolant;:accideht

• (LOCA) value of 48 psi~, the licen~ee-determined that a'leaka[e of 4: 73 percent p'er day would oc_cur through thfs 1 i rie. **.This exceeded

• the TeChnical ~Spec1ficati9n. al-lowed primary, containment leakage test

. value of*l.6. percent per da (The Technical Sp'ecification limit *

would actual l.y be exceeded by a greater amount when 1 eakage fro this line is.added to other.leakage.sources.. )

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Technical° Speci.fication. *action statement *3.0.A*required hot shutdown

':within 12 fiburs and cold shutdown within the. foi-lowing 24 hou*rs.. *

Since the isolation vaives were open for'sampling'***for a sufficiently.*

short ¢~ratio~~ this action statemeht*wa.~ not.exceede _

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4..... ;compariso*n of ~Practice to Requi.remerit

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i O CFR 50.'59 sta:t,~*s that a*holder. of-~ lic~n.se may (i) mak_e.:chan~es-in.:":'.

.. **the* facility as described in the safety ana*lysi.s -report ('SAR),*_.(ii) make

... ~ **chan*ges in the procedure:S as describe.d fo *the safet,y analysis report,, and (iii): conduct tests or. experiments' not described.in. the safety, analysis

• report,. without Commission approval,. unless the proposed chahge, test or experiment involves a change in the t'echn'ical. specifications incbrpbrated*

irr.the lice~se or ~n ~nreViewed.safety ~Gesiion. *The l{censee 6ri nu~erous occasions since at lea~t 1978 ~nd without NRC approval made-changes in

• th.e facility as described*in* the safe.ty analys*is.report, which involved a*

• change in Technical Specifications -and c;onstifoted an *unrevi~wed safety

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, question, by per.forming a temporary-alteration utilizing a tempqrary sample

• .pump to-obtain the daily drywell air sampl This temporary alteration reduced the margin of safety as defi.ned in the basis 'of Technical

Specifications in regard to the maximum allowable* p_r.imary cci.ntafoment accident Veak rat Jhfs is *an *appar*ent violqtion (50-237/90-022-0l(DRP);

50-249/90022-0l(DRP)).

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The tempor.ary' alteration represe.nted a change *_in. the facility as

• .described in various portions of the SAR in* regards to the *drywell manifold sample system design, the primary containment leak rate and

• its affect on the accident a~alysis: * *

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'.I Technical Specificatio~ 3.7:A.2.a.(3) prescribes a maximum allowable test leakage rate of 1.6 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 48 psi Usage of the temporary sample pump with the single line represented an ad~itional leakage of 4.73 percent per day beyond riormal

• contain~ent leakage..

Th~s number exc~eds the technical specification allowable leakage and th~refore effectively constitutes~ ch?nge in the.*

Technical Spec1fications incorpor~ted in the license,_.*

._*10 CFR 50.59 states that "a proposed ch~nge, te?t*,. or' experiment shall be

. deemed to invoive an unreviewed safety question (i) if the probability of occurrence or.the consequences of an acciderit *or m_alfunction of equipment

. important to. safety previously* evaluated in the sa.fety analysiS report*

.. may be-inc.rea*sed; or (ii) if a possib'ility fo'r an accident or malfunction

. of a:different t~pe than any ~valuated ~revi6usly in the safety.analysis report may be cneated; or (iil) if the ma*rgin of safety as defined... in the basis.for.any technical specification is reduc~d". *.

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Techniqi.l Sp_ecificati6n 4.7.A ba_sis ind_icates *that the des*ign_basis loss '-,:

of coolant ~c~ident ~as evaluated at the primary containment maximum allowable q.ccident. leak rate of 2*.0 percent per day at 48 psig_..This was the basis for determining q.'max.imum allowable test leak rat~* of 1~6*

percent' per day at a pressure *of 48~ psj (The difference wa{to ac*coun for the.. effects* of* con ta i nmen't en vi ronhient under accident and* test****

co,n.~itions*by applying a**o.8 correCtionfactor.)' Technical SP,eCification 4.7.A basis states that "the specHi,ed primary containment:leak rate and

• ... *fil.ter ~ffi~ienc::y are coriservativ~*an:q providemargin be:tween exp~cted.

. _offsite dose*s and 10 CFR 100 guidel ines.

.It further states that 1

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• *11 althoug.h. the do'se calcu_lations *sugg.est 'that: _the accident lea'k rate could.:*

• b'e al*lowed -tofncrease*fo. apout.3.? percen.t per day before the guideline thyroid dose*s give~n*,in 10CJ:R*100 would be exceeded,. est'ablishing the..

test limit of* l. 6* per.cent.per.day provides*.adequate ma.rgi n of safety to

"a$sure the_.hea*Jthand safety of the gehera*l public.' 11 * Usage.of.the

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' *temporary sample pump represented *a.n additional leakage of 4.73 pe'rcent per day during accident ~ondition~. This ad~itional leakage rat nullified the ~argin of ~afety as defined in th~ Technical Specificatio basis and also exceeded the** value the Technical '_Specification basis

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indicates as an acceptable cohsequence to public heal*th and safety under 10 CFR 100 if_a design *-basis LOCA occ.urre~ during samplin Alth~ugh there have been subsequent l~censihg*actions regarding the 10 CFR 100 analysis.for the d~sign basi.s loss of-coolant accident, th l~cense~ was*not able to provide ahy documentation indicating that th margtri*of safety defined i_n the Techhical Specification*basis was

,specifically change * Compa.'rison of.Li.censee Safety Evaluation and Supporting Analysis With Licensing Bases and Technical Specifications Following identification *Of this issue,.the licensee performed a 10 CFR 50.59 safety evaluation for unattended usage of the temporary

• sample-pump to~determine whether an unreviewed safety question existe The 11c~nsee 1 s analysis indicated that requirements were met and the safety evaluation indicated that an unreviewed saf~ty question regarding pa~t usage.of the temporary sample ~ump did not exis. "

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NRC review of the licensee's analysis r~garding this issue in comp~rison to licensing basis a~sumptions indicates questionable rationale wtth the'

licensee's* tonclusions~ (The NRC review considered only a~sumptions used in the analysis and not the methodology and computer *codes for the calculations themselves).

In -particular, in order to show acceptable~*.

values, this analysis used assumptions that were contrary to more

tiohservative assump~ions specifically stated in the licensing basis~ and,

"in* some cases, reflected i"n Technical Specifications.. Even with,:the open**

_lin~ (an additional 4.73 percent ¢er day leakage), the licensee's analysis showed significantly smaller offsit~ ddses'~han the December 7,rl981 SER for Sys~ematic Evaluation Pro.gram (SEP)' topic XV-19 11Loss of Coolant

• .Accidents Resulting From: Spectrum of Postulated Piping Breaks With in the Reactor.Coolant Pressure Bounqary.

These reduction.s.maY have been in part accomplished by applying specific:: assumptfons.used.in the** control*

~oom dos~ analysis (and s~bsequent]~ apprb~ed i~ an' SER* for that arialy~is)

to the offsite dose.*analysi Although the permissibility of this_. is

.~ncle~~.~there may be some merit to this approach, as long as.the

appTi"cability is essentially the same a.n'd there are *not technical reasons.

to preven~ -0~ a specific caie (such as inconsistenc~es wiih Techntca.

Specifications), the usage of less conservathie* assumptions* approv"ed*..

for anot~er*type of lnalysis~

However, some of the licensee 1*s assumptions were less conservative than th assum¢ti-0n~ used in the ficensing basis fQr the ~orit~ol room~dQse analysis.(SER o.n the control'"roo'm habitabiliti:study) or Tec-hnical

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• .specification *values. The control rqom hab~tab~lity SER was ~ssued ~n May 11, 1983 for NUREG-0737, Item III.0.. 3.4*, *control ~6o"m Habitability,* * * ~- *

~hich ~i:cepted t~e licensee 1 s;confrol.room habitabiliiy study as:..

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. indicating an acceptable ~ontrol* room ventilation desig In order to

• ._achieve acceptaole.results for the temporary sample pump analysiS, the*,

licensee had to assume at leas*t a Standby Gas. Treatment :system (SGTS).

organic efficiency of 96.3 percen\\.. However~ the Technical Spe~ifitafion ACCeptahce crit~ria is 90 percen Although the hi~her efficiency could

• probably be justified based upon historical test1ng re*su1ts, this is

• still less conservatlve in regard*to Technical Specifications and thus is not justifiable to use in a 10 CFR 50.. 59 safety evaluation.* The liCensee.implicitly. recogni~ed this ihconsistency when it re-performed

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~oom habitability study to 'incorporate the 90 percent value on April' *19, 198 The licensee.'s safety evaluation for the temporary sainp*le pump issue assumed normal"contr'ol room venti.lation operation for only 40 minutes as opposed to the eight'hours specified in the licensing bas The licensee's reanalysis*for*the control.room habitability study made the same 40-minute assumption*.

The licensee,* cohtrary"'to the Technical Specification basis; assumed a normal containment 'leakage rate of 1.6 percent per day instead of the apparentiy required 2.0*percent per day rat Although Technical Specifications* prescribed a 1.6 percent per day maximum all-0wable test -leak rat~, the Technical

.Specification basts indicated the actual maximum allowable*accident leak

• rate,was 2.0 percent *pei: *da (The tes.ting acceptance friteria were derived from this value by appl~ing a correcticn factor to account for uncertainties from the effetts of the testing environment compared to the accident environment.)

Therefore, usage of the smaller value in offsite*

and control room dose calculations was n6n-conservativ Previous NRC *

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10 CFR 100 caiculations for offsite doses reflected* in the o~igin~l licensing of the plant and the SER for SEP topic XV-19 used 2.0;pe~cent per day leakage:.

Both the control room habitability.analysis and reanalys~s used 1.6 percent per day leakag The acceptability of theie last two ~ssumptions, specifically for control ~com habitability, a~e an unresblved item (237/90-022-02 (D~P); 24§/90~022-02 (DRP).. Finally,.

the analysis as~umed a constani maximum desi~n accideht pressure of 48 psig ove~ the entire cou~se of _the accident in accordance with the

• licensing basis but assumed a decreasing actident pr~ssure for the extra

. 4~73 percent per day* leakage porti'o This last assumption_. was not only contrary to the l:ic'ensing basis but also contrary to the current Standard*

Review Pla*n* (SRP) provisions which pr.escribe a_ constant maximum pressure**

• for built-in conservatism to* the.calculation *

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following NRC and licensee discussions of these.. i.ssues with respect to,*

the safety evaluation,. the licensee provided adc;!itional clar.ification of the 1ntenCof the safety. evaluation.-

This c'larification acknowledged

• that assu~ptions used ~n. inalyses supporting-forward looking*lO CFR 50.59.'

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• safety evaluations should be consistent-with.. those previously approved by.

.the NR How~ver*, si nee thi-s analysis was performed for a past practice- *

that wa.s* d,i?cohtiriue*d following discovery',.the liC:ensee beli.eved thes ': assumptions*were 1appropriate to indic~te*wheth~r an unrevie~ed safety

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In su.mmary, the licensee's**~nalysis.'co_ncerni'ng_.this Jssue:resorted to

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non-~ohservative *assumption.s with respect to the plant -licensing basis or*..

• Technical *Specifi.cations in order.to achieve acceptable results.*.. Review:

• • of the licensee'sanalysisi_ndicates* that the calculated consequences *qf-.*

an accident may have been increased if the in~ti.al assumptions were i~. ~~

• • accordance with Techn-ica1* Specifications and the Hcensing basi All.*

accepta*nce cr-iteria e*xceeded Cannot.*,be, expl 1citly *ldenti.f-i'ed-'Wi"thout a

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• re~nalysis usind the mor~ coriservative *~ssumptions.

.,. The li'censee's safe'ty evaluation for this issue i'ndicated that the rnargin of safety as d~firied.in T~chni~al Specifications was not.reduced bised

upon the definition of margi~ of safety as defined.in Nuclear Safe~y

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Analysis Center (NSAC)-125 "Guidelines~.for 10 CFR 50.59 Safety Evaluat*ion.

This was. due *to the r'esulting'doses being less than the acceptance limit (10 CFR 100 'and.*General De.sign Criterion 19) _in the licens1ng basi Thjs evaluation indicated a lack of understanding of NSAC-125 provisions.* NSAC-1_25, *sect*i_on 3'.6 states that "changes in *.

barrier performance that.do.not result in increased radiolog'ical dqse'to

• the public are addressed under margin of safety."

NSAC-125 indicates an increase iri consequinc~s of accidents must* involve an increase in d6~es above the licensing~ limit; however, the* margin of safety as defined in

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the basis of any Technical Specification does not.rely on this* prov'isi'o.n.

. Therefore, the' dose to J,he pub 1 i c -1 s not the determining factor in -the

• margin of s~fety port~bn of the definition of an unreviewed saf~ty *.

question. rin actuality,:t_he margin.of.safety was reduced' whether a comparison against 10 CFR 50.. 59 wordi'ng is used or NSAC-125 provisfons are relied upon..

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In addition, the liceniee 1 s safety evaluation also stated that the dose

• .analysis methodology was based upon the original FSAR and* SER dated

• August 31; 196 However, the SERs whic.h cover~d*the final licen?ing o the plant dated October 17, 1969 for Unit 2 and November 18, 1970 for Unit 3 granted approval based on differerit calculations than the previous SE The licensee.could not locate any copies of these other SERs indicatirig a failure t~ maintain kn6wledge of the licensing basi The inspectoi*subsequeritly obtained copies of these SERs and ~rovided them.to

. *the license Licensee deficiencies in incorporating SERs.in the Updated

• Final.Safety Ana*ly?is Report.is an unresolyed item (237/90""'.0?2-03; 249/90-022-03 (DRP))~

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Corrective Actions As*a re:sult of this.problem, tne licensee **completed or is planni~g the

• following actions:

. A prellmi'nary analysis was performed to quantify the -amount of leakage th.rough *a one.half inch primary containment pe,netration a design *accident pressur After. finding that *this greatly exceeded allowa'ble limits_ the licensee reported th~ problem, in accordance with 10 CFR 50.72 and 50.7 A temporary change to the procedure regarding usage of the temporary sample pumps' was* iSsued to require an individual in c:ontinual attendance arid *in contact with. the control.rooni by radio while the.*

manual isolation valves* are ope' A temporary alteration wa ~ubseguently perfoimed that moved the sample poirit for the Technical Specification required dailysample fo a line thafhad automatic*

isolatio The temporary p'rocedure change was di*scontinued following the temporary alteration:

' All* Radiation Prote.ction shift perso'nnel we're briefed as to.the problem tb preclude*-improper usage of the sYste * A deviation report-was i.nitiated to track the licens*ee 1 s investigation of the problem.. A potentially signif.icant everit r~port was also initiated for corpor~te manageme~ *~Th~ licensee performed a saf~t~ evaluation on the unatt~nded usage 9f the temporary sample pump which indicated that an.unreviewed safety question did not exis The licensee is currently reviewing the problems with the pe~manent sample pumps and assessing what is needed to complete repair ~:

A review of the design.basis and the need for any system design improvements is being conducted.* Since the design basis could not..

be identified, the licensee decided to re~onstitute the design *

basis. The licensee is reviewing whether the system will be repaired and used or whether it is to. be abandoned, dismantled and the lines cappe **

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.. A.re.view is being conducte,d to determine possible methods whereby a*

temporary return line to the drywell could* be established for ~se with the temporary sample *pump~ *(Although *automatic isolation is now provided, the temporary sample pump still exhausts to the reactor building which presents ALARA consider:-ations.) The licensee _is reviewing other prac~ices, procedures and surveillances for any other items.that could violate con~ainment integrity or system operabilit Root Cause The root' cause is attributed to *~* m9,nagerilent deficiency in th~t' safety evaluation ad!Tlini'strative requirements were inadequat~ or. were not*

adequately applie As administrative requirements were upgraded ov~r the years~ no actions were taken to ensure past standing practices ~ere.

. reviewe A 10 CFR 50.59 safety:evaluation* *was never done on this a*lteration* (use of the temporary s,ample pump) since the original administrative'...

requiremE?rits for *temporary alterations only appli_ed to lifte.d leads and jumpers~. When the admi.ni strati-ve requirements exp*anded to mechanical.'

. equipment, previously_ existing a_l terat ions were not eva l uat"ed for * *

applicabill.ty to the administrative requir.eme.nts:

As. such.~ in recent:.

years each time this *alterati'on,was performed i.t was*done confrary to the-l:fren.see' s admini strati ve pr.o.ced.ures. * (Dresden: Admi ni strati ve Procedure.

(OAP} 7-4 !'Control of Jumpers or Lifted Leads,"- Revisfon 8, *was is'sue.d *

. on December 24, 19S5,. w_hi ch ad_ded the **requirement fqr * 1b CFR * 59. 59 safety*

evaluations to be performed* ori jumpers :and*lifted l~ad~.* Revision 11.o d this procedure was issu'ed on *August 15, 1988 to expand ther def.inition of,*.*

temporary alteratfon to *additional 'items *such as mechanical. equipment.).

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A procedure :covering the us.e of the tempo~ary sample punip. did not exi.st.,

until 1989 and thus the problem was not previ~usly disc~vered through*~

procE!dure safety evaluatio Due to a non~docutnented third party reviewer's' comment concerning use of the temporary sample pump without a..

procedure, Dre'sden Radi at fon Prote*ct fon (DRP) Procedure* 1350-3, "Sampling the Drywell Man.ifold System Using the Radeco Air_. Sampler" was first issued in May 1989.. *This*was a missed ch.ance to.detect the problem sfnc a 10 CFR 50.59 safety.evaluat,iori should;'have been performed.* The screening

• criteria in effect at the-time*~llowed enti~e categories of proc~dures *

(such as DRPs not related to effluent monitoring) to be*automati'cally ruled out for a safety e~aluation as long as they.were not new or changed procedures or administrative controls described in the Final Safety Analysis Report (FSAR) or Technical Specification In this* particular case, since it wa~ a.new procedure, the criteria required a safety

evaluation to be performe However, the reviewers mistakenly used the ~rong administrat~ve path_ as if it were a revision to this type of procedure instead of a new procedure'.. Therefore, a safety eval.uation was not performed due to a failure to follo~ the administrative requirement Additionally, the crite~ia themselves were still inappropriate since the licensee could have instead just made a revision to DRP 1350-7, "Operation of the Unit 2(3) Drywell Afr Sampling Manifold 1....

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Sys~em 11 to _al low usage of the _temporary sample pump. *In that case, the

.licensee 1 s administrative requirements would not. ha_ve requ:ired a safety evaluation to be performed with the same result (usage of the temporary sample pump without a 5afety ev~luation). The screening criteria were

• revised on January 25, 1990; such that this is no longer a concern for recently issued piocedures and revision.

Unresolved Items Unresolved items are matters about which more information is required in 6rder to ascertain whether they are acceptable items,- violations, or *

. _'.dev~ations. The unre~olved items di~closed dufing the ~nspec~ion*are discussed in paragraph '5:

~xit Interview (i0703)'.

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• Th~ inspeetors summarized the scope and findings o~f the inspection-by

.,"telephone with the 'licensee 1s repre~entatives. (denoted in paragraph '1) ori

,September 27, 199 The licensee acknowled~ed this informatio The inspector also discussed the likely informational content of 'the

'inspection report wi-th regard to 'documents or processes reviewed by_the fospector during the inspectio The licensee did ncit identify any. such- --

. documents/processes as pro'pri'etar *

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