IR 05000237/1990019

Insp Repts 50-237/90-19 & 50-249/90-19 on 900801-0928.No Violations Noted.Major Areas Inspected:Licensee Event Repts Followup,Plant Operations,Maintenance & Surveillances,Insp of Check Vlave Programs Effectiveness & Rept Review

ML17202U857
Person / Time
Site:	Dresden
Issue date:	10/23/1990
From:	Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML17202U856	List: IR 05000237/1990019 ML17202U855
References
50-237-90-19, 50-249-90-19, NUDOCS 9010310130
Download: ML17202U857 (18)

Text

.U. S. NUCLEAR REGULATORY COMMISSION*

REGION Ill Report No /90019(bRP); 50-249/90019(DRP)-

Docket No ; 50-249 License No~.

D~R-19~ DPR-25 Licerisee:

Commoriwe~lth"Edisori Compan * ~

P.. 0. Box *767.

Chi.~ago, TL 60690 *

Fac.i l i ty Name:

Dresden Nuclear Power Station, Units 2.and 3 Inspecti~n At:* * Dresden Site, Morris, I Inspection Conducted.:

August.1.through September 28, 1990

.

.

.

Inspectors:

~: G. Du Pont D. E. Hills M. Approv.ed By:

J~ M. Hi Projects Inspection *summary Inspection. during the period of Au.gust 1 through Septembe.r 28, 1990 *

(Report Nos. 50-237/90019{DRP); 50-249/90019(DRP)).

.

..

Areas Inspected:* Routine unannounced resident inspettion of prev~ously identifi~d inspection items, licensee event reports*followup, pl~nt operations, maintenance. and surveillances,**inspection of check val;ve progra effectiveness and report revie *

Result.s:

On~ ~on~cited violation was identified in paragraph 5.a(l) in rega~d to inadequate*cont~ol m~a~~res to prevent the use of incorrect component In particular a light bulb wittl an incorrect vol fage rating was used to.replace a burned out

.indicating bulb on a main.control room panel which resulted in the.automat1c isblation of the isolation tondense This was not ~ited as this was an i~olated event and appropriate*

cofrettive actions were initiated prior to the end -0f th ins~ection in accordance with 10. CFR 2; Appendix C, Sec.ti on. Two unresolved items were identified during the inspection.*

. The first ih~olved a lack of adherence to administrativ * procedures regarding usage of the.control rod drive accumulator alarm log and documentation of procedural adherence 9010310130 901023 PDR ADOCK 05000237 G

PNU

~*

'

.....

.*

deficiencie This item as di~cussed in paragraph 4 was awaiting.further root cause r~view by both the licensee ~nd the NR The other unresolved item discussed in paragraph 5.a(l)

regarded review of Dresden specific tequirements for separation of control and protection circuitr.

• Plant~Operations During the ir:ispection period (August 1 through September 28,.

• 1990) both Units 2 and 3 operated at or near full *power *until Unit 2 was shuidown ori September 23, * 1990 for a refue 1 i ng outag.The shutdown was successfully comp 1 eted with minima 1

.~roblems. Review of the inspection results as they apply to

• th~ operatioris functional area is indeterminate at this ~ime. *

The operators' response to th~ spurious opening of the T~rge~.

Rock Safety Relief Va~~e qn Unit 2 as discu~sed in.paragraph 5.a (4°) was rev.iewed by the NRC Office of Analysis and

Evalua~ion of Operational Data (AEOD) fo~ the.NRC ~rogram t study factors which affect human performance.* Therefore, thes actions.are to be evaluated through that progra Maintenance/Surveillance

• The l.icensee;s faiiure to establish a program to monitor the Tar9et Rock Safety Reli~f Val~e tailpipe temperature tb identify potential pilot valve leakage problems *prior to the

• everit in accci~danc~*with General Electric Service Informatibri

• Letier recommendation.reflect~ adversel~ on the

.. *

. cinainten~nce/surveillance*funttiorial.are Such a pr6gram was.

inc1ud~d fn the licensee's*planned corrective action The*

failure to properly control main control pa.nel light bulbs with

.the corre.ctive voltage* ratings as discussed in paragraph 4, also reflects adversely on the maintenance/surveillan~e fun~tionaT are However, a Notice of Violation wa~ not issued

- in accordance with 10 CFR 2~ Apperidix C, Section **Engineering/Technical Support

• A revie~ of the licensee's check valve in~pection program dis~ussed in para~raph 6 ~ndicated tha~ both NRC and licens~e corporate requirements were. greatly exceeded; The extent and deta i1 of this program reflects favorably on this fun ct i ona l are Safety Assessments/Quality Verification Management involvement and control were highly ~vident in the evaluation of ev~nts and dete~minatio~ of corrective actions for the above item Only one scram occurred since the last Systematic Assessment of Licensee Performance (SALP) perio However, this area will be closely observ~d following completion of the current Unit 2 refueli~g outage since historically more scrams have occurred within the first couple of months following refueling outage DETAILS* Persons Contacted *

.'.

-*. ~. '~'

. 2 :-

.Com~onwealth Edtson Company*

• E: Eenigenburg,. Station *Manager*

_,,:

• L. Gerner.,, Technical Superintendent:

E. Mantel, Services Director

.

• D. Van Pelt, Assistant Superintendent - M*airitenanc J. Kotowski, Productio~ Superinte~d~nt~

J. Achterberg; Assistant Superintendent - Work Planriing G. Sf!lith, Assistant Superintendent-Operations

. K. Peterman, Regulatory Assurance Supervisor*

• M. Korchynsky,"Operating Engineer B. Zank, Operating Engineer

. J. Williams,* Operating Engine~r

.R. Stobert, Operiting* Engineet

.

• M~- Strait, Techni~al Staff Supervisor

~- John.s.on, Q.C*. Supervisor*

J. Mayer, Statibn Security Admin~strator D. Morey, Chemistry Services S~pervisor

..

D. Saccomando, Health* Physics Services Supervisor K. Kociuba, Quality Assurance Superintenden *

J

"*

,,,...

The ii'1sp.ectors also talked with and intetviewed s~ve~~1:other *1*icensee

.. *erilployees, inclucHng members of t~e technical and engineering *staffs,

react~r and auxil,iary operators,.shift.engineers and forenien; ele¢trica mechariical and instrument persohnel, and ~on~~act security personnel.*

• Denotes th_o?e attend1ng *one or more" exit interviews -conducted informally at various times -throughout the_ inspectiqn_ perio '

~reviously Ideriti-fied Inspectio~ I~e~s (91702) *

(Closed) Unresolved Item (249/86009-03):*.Evaluat.e* the use of leak befor break concept and reliance on non-safety leak. detection systems for.

meeting requirement This iss.ue was forwarded to.the Office of Nuclear Reactor Regulation (NRR) _for. reso.lutio NRR determined: that the lea before* break concept ~a~ not acceptabl This issue ~s ~~nsidered to be closed. *

• *

'

• *

• *

(Open) Unresolved Item (50~*237/90'o22-02(DRP);249/90022-02(DRP)):

Evaluate concerns involving the licensee control room habitability analysi This analysis was conducted.in response to NUREG-073i, *

"Clarificatio_n of Three Mile Island (TMI) Action Plan Requireinen.ts 11 *:

Item III.D.3.4, "Control Room Habitability" to assure that control room operators were adequately protected against the effects of* accidental release o.f.toxic and radioactive gases and the plant could be safely 6perated or shutdown under design basis accident conditions in accordance with 10 CFR 50, Appendix A, General Design Criterion (GDC) 1 The TM!

item specifically indicated that NUREG-0800 "Standard Review Plan (SRP),

Section 6.4 requirements_were to be met and that the design basis

• .

accident (OBA) radiation source term was to be. for the 16~s of coolant accident (LOCA} containment leakage described in Appendix_A and B of SRP Section 15.6: GDC 19 limited exposures to control* room personnel to less than 5 rein whole body or its equivalent to any part of the body for the durati~n of the acciden This was further clarified in SRP

• S~ction 6.4 ~ith specific acceptance criteri~ of less than 5 rem whol~

• body gamma, 30 rem-thyroid and 30 rem beta skin dose SRP

Section 15.6"5 Appendix A stipulated that the primary containment. leakage should correspond to that given in Technical Spicificatioris:

The licensee iubmittal dated September 1, 1981 and subsequent revisions t~ritaining the control room habitability analysis were approved by the NRC in a safety evaluation report (SER) dated May 11', 1983.. * The SER found the,control room habitability systems to b~ acceptable with specific modifications-the licensee

• committed to perform. 'This was *based*'upon'the ability to.meet the guidelines of_ GDC 19 in regard~ to whole* body, *thyroid and.skin dose In. 1987 the licensee identified a discrepancy_,between the control room

. habitability an~l-ysis and Techn1cal Specifications. **This was first*

identified at Quad Cities and was the subject of Information

..

Notice 88-6 The absorpt fon efficiency* of the Standby Gas Treatment System (SGTS) was* assumed as 99 percent for 6rganic iodide removal in the control room habitability analysis while the Technical Spetificatidn

• testing acceptance cr*iteria only*requi.red 90 percen A review *of the actual values obtained from testing.as far back as 1977 showed only.o'n tiine when a value of less tha:n 99 percent was cib.taine *(The 1980 tes.1 showed an organic iodine efticiency 6f.97.85~pe.rcent:) Although this provided~a justificat~on f6r using a higher effici~n~i.in analy~~ *"Technical Specifications would not require ariy act"ion. unless' the value*.

was found to be less**t.han 90 pe_rc*ent:

The Technical Specification value,.

• . therefore, _became the ove~r.i ding factcfr* for *analyses assumpt 1 on As a *

• result, the. license'e repea*ted the control.room habitability analysiS assuming the correct 90 percent effici:enc A licensee 10.CFR 50.59.

safety evaluation dated April *4, 1988 i nd1cated.that these new calculations did not represent an 'unreviewed safety question since the SRP guidelines were still me The licensee, therefore, updated _the Final Safety-Analysis Repqrt (FSAR) with._this *new evaluation withqut submitting the change in the analysfs to the NRC forapprov~l. *.

~h~ inspec~or's re~iew of these calc~lations 1~d~cated that in order to\\

meet the SRP gu_idelines a certain assumption had to::be utilized which was less conser~ative than the corresponding ~ssumption'stated~in the SER *on the original control room habitability* analysi.s. -In particular, ::the SER, which is the licensing basis for this analysis, specifically documented the assumptfon that normal control room ventilation ~emained in ~peration for eight hours, following the initiation of the radioactive releas Assuming the 90.percent SGTS efficiency, the licensee*determined ~hat the 30 day control room thyroid dose rate would be 73.2 rem as comparedc..to * *

• the 30 rem allowable in the SRP and plant licensing basis.* Therefore, the licensee reduced *the eight _hours to only 40 minutes for ~alculatiori *

• purposes which* reduced the 30 day control room thyroid dose rate* to a barely.atceptable 29.33 re The licensee provided justification for this assumption in the 10 CFR 50.59 safety evaluatio The acceptability of utilizing an assumption during a licensee 10 CFR 50.59 safety

.... ~

evaluation that is contrary to a specific assumption state-din the l*icensing basis 1s being* evaluated by the NR Since the NRC based approval on a ~pecific assumption, this evaluation would reflect upon whether the licensee should.have resubmitted the analysis for NRC review of the justification and approval of the changed assumption prior to changing the FSA.

-

Further review o*f the original control room habi-tability analysis _a_nd the

.licensee's reanaiysis indicated that an additional discrepancy existed between Technical Specifications and ihe analyses which was not

- previously i"dentified or addressed by the license In both cases, *a containment leak ~aie.of 1.60 p~rcent per day corresponding to the Technical Specification t~sting acceptance criteria was use.Ho~ever, as explained in the T~chnical Spe~ificatioh base~,-~he co~responding maximu~ allo~ible accident leak rate was 2.0 p*rcent per da The diff~rence w~s cbnservatism added foi uncertaintie~ inherent in th~ use of a prim~ry containment environment for testing, air "or nitrogen,- befog

- different than. the ac-tua l accident environment, steam and-hot ai Therefore, the't:ontrol room**habitability anaJys_is used. a leak rate that was non-conservative with respect_ to Technical Specifications.. Other. -

licen.sing actions in regard to offsite do_se analyses exemplify usage of*

the higher va*lue-.

An SER dated January 5, 1982 on Systematic Evaluation Program (SEP) Topic XV-19 "Radiological Coriseq-uenc_es of a Loss of Coola!'.lt Accident From~ Pi pi.ng Breaks Within the Reactor Cool ant Pressure Boundary" as well as the SERs for the original licensing of.the plant, dated -

October 17~ 1969 for Unit 2 ahd November 18, 1970 fo~ Unit 3,,specified a

- 2.p percent per day value:

The TMI item speci-fically prescribed a value.*

in accordance with the same reference which guides NRC review of the

_.

offsite.dose analysi This issue is similar in*nature to the Technical Specifi cat i cin versus analysis discrepancies discussed.above regarding

• sGTS efficiency for which the licensee performe_d a.reanalysi The NRC is currently' eva 1 uat i ng whet.her an analysis can be.considered acceptab 1 e*

if it had been previously approved by the NRC but apparently conflicted *

with Technical Specifications (and, therefore, the TMI item *for which it was to address.)* Since the licensee's reanalysis indicated just barely acceptable results in regard to control* room thyroid doses, an adverse findi~g in regard ~o ihis assumption could indicate an unacceptable control room ventilation syste~'.desig It was also noted that th~ SER which approved.the licensee's original_

analysis* indicated that the NRC_believed automati~*1s6lation of the control.-rooril ventilation system on high radia_tiqn-was desirable (but riot required based on the computed dos*es) but _the~*Hcensee had declined the

- opportunity extended to further revise the proposed modifications t include. this featur The TMI item also indicated that technical

Specification changes were require Technical Specificatfons have not yet been _changed to provide for requirements regardi_ng the control room ventilation syste The licensee has however, -s~bm,itted proposed changes to the NR (Open) Unresolved Item (50-237/900~2~03(DRP); -~0-249/90022-03(DRP)):

Review licensee's incorporation of SERs into the Updated Final Safety Analysis Report (UFSAR).

The inspectors noted that the offsite dose analysis for the design basis LOCA reflected in the SERs dated

• .. *~

..

Oct6b~r 17, 1969 and Novemb~r 18, 1970, for the original plant licensing were* nt:ver*incorporated into the UFSA In addition, the same type of analysis in the SER dated December 7, 1981 for SEP Topic XV-.19 Loss of Coolant Accidents R~sulting Fro~ Spectrum of Postuldted Piping Breaks Within the Reactor Coolant Pressure Boundary, 11 was also not incorporated

. into the UFSA Instead, the offsite dose analysis for the 'design ba~is**

LOCA.reflected in th.e UFSAR was the same* one as in the original.FSAR. *In

• addition, the licensee could not locate copies -of the 1969 and 1970 *sER The i.nspector subsequently provided copies of these SERs.to the _Hcensee.

Failure ~6 incorporate SERs into the fSAR results in the UFSAR no longer reflecting the licensing basi..

• Utilization of an FSAR which does not reflect the licensing basis can

~ausE specifi~ problem~. For'example, reiiew of these previou~ SERs ~n an earlier SER dat~d August 31~ 1966, tndi~at~d.the NRC b~lieved the *

licensee's calcul~tions refletted in th~ origirial FSA~ utiliz~d *. :

nonconservative assumption A~.a result, the licensing of the plant ~~s approved ba_sed upon other calculdtions which incorporated more

. * *

• cons~rvative *assumptions but still indicated 10.CFR 100 off site dose* * *

acc:epta:nce criteria. w*ere met., A compari so.n of the results of these two

sets of calculations indi~~t~d ~ large diffe~ence. For example, the

.. * thyroid site boundary two hour: do*se reflectt!d in the 1969 and 1970 -SERs

• was*185 rem wh.ile that in the or'i~inal FSAR was on the orde~_of

.. :5 E -6 rem depending upon specific meteorlogical *cond'itions~* *one

. assumption which greatly contributed to this diTft:rence-.was a SGT *\\

• .***

~organic.iodjne efficiency* of 99 percent in th~" original FSAR and 90 percent 'in the SERs. :Technital Spec_ifications also reflected the * ~*

90* perct;!nt efficiency value.iri regard to SGTS *filter* efficiency testing **

• • .acceptant~ iriteri~. Although the 1ic~nsing* bi~is and*Technica Specifications reflec~ed t.he 90 p'ercent value, cases were identi_fied..

wh~re the licen~ee improperly.used the 99 percent value from the FSAR i~

subsequent analyses.. A case in point wis the licensee's control room

..

habitab.ility analysis on which approval of the control room ventilation design in a May 11, 1983 SER was base This analysis used the 99 ~ercent ~~lu~ from.the-UFSAR inste~d of the more ccinserv~tive 90 percent from the licensing basis and Technical Specificatioh The li.censee implicitly recognized this inc.onsistency when.it reperformed.tht!

control room habitabi llty. analysis to incorporate the 90 percent value on*

April 19, 1988~ (The or:iginal discovery of this _9iscrepancy' was made at *

.the Quad Cities p~lant *and is one of the subjects--of Tnforma'tion Notite 88-61.) * Revision 6 *to the UFSAR added. a note to the 10 CFR 100 analysis description t_o *indicate the existence of the SER dated August 31, 1966 and. that it prescribed ~ 90 percent valu However~ *the actu~l discussion itself was not changed*although several *other assumptions and the results differed from the licensing ba~i Iri addition, subsequent SERs with different calculat1ons regarding this area were flOt ~eflected *

1n the UFSAR.. There ~ere inditat.ions that the UFSAR valu'e mai *h~ve been utilized in other analyses as wel For example, a litensee submittal dated August 18, 1980 for SEP Topic XV-16, "Radiological Consequences of Small Lines Carrying Primary Coolant Qutside Containment," indicated tha~

the corresponding analyses assumed a 99.99 percent value "based on the

. FSAR.

The corres~onding SER indicated approval was actually based ~pon independent calculations but did not list the SGTS efficiency assumed in the independent calculation The licensee indicated that. they were

"

aware-that there was a problem. with past incorporation of SERs into the U~SAR and had already planned o~ performing an UFSAR.reconstitutio This item is ope~ pending revie~ or the licensee's projected tim~ frame for comp1eting this activity and any actions being taken in the interim to ensure 10 CFR 50.59 safety evaluations reflett the lic~nsing ba~is.

. - N~ violations or deviations were. identified in this area.* Licensee. Event Reports Follciwup (90712 arid 92700)

• Through direct observations, discussions* with. 1 icense*e. personnel, and*

• review bf records-, the fo 11 owi rig event reports were reviewed to determine *

that reportabi hty requi.renients were fulf i 11 ed; immediate corrective

action was accomplished, and corrective action to prevent recurrence had beeh acc6mplished in accorda~ce with Technical Specification *

• (Cl()sed) LER (237/90005(DRP)):

Unpfanned*primary containment.:Group V..

isolation due to procedural deficiency in voltag~ specificati~n for*.

. ~:.. feplacement light bulb This event and c~rresponding to~rectfve.actions

• • are discuss~d in paragraph 5.a(l)..

'

.

~o vi6lations or deviations were identified in this *ar~a *eicept as delin"eated in.paragraph 5.a(l).. * P.lant Operations (71707) * * * *

The inspectors~ observed control. room* operatiOns, reviewed applicable. logs. -*

, and ~ondu~ted discussions; with ~o~troJ room ~perators ~uri~g th1~ peri6, * The tnspectors verified the operability of selected emergency systems~:.

.. reviewed tagout fecord~ and verifjed proper retu~n to service of affected

• -co~ponents. Tours of Units 2 and 3 reactor buildings and tu~bihe

buildings were*conducted to observe: plant ~quipment conditions, inc.luding potential fire hazards,- fluid leaks, and excessive vibrations and to

• verify that maintenance requests had been initiated for equipment in need of maintenanc *

-

Each week duri.ng routine activities or tours, the. inspe.ctor monitored the licensee'.s security program to ensure that observed.actions were being implemented according to _thei.r approved secur.ity pla The.inspector

• noted that persons within the protected area d1s-p-layed proper**

phbto-identific~tion -badges and those individuals requiring escort~ were*

properly e~corte~. The ihspector also verifi~d that ch~cked* vital areas*

were locked and alarme Additionally, the inspector a1~6 ~erified that observed personne 1 and packages entering the protected area were searched,.

• by appropriate equipment or by hand.,*

.

.

.

The inspectors verified that th* licensee's radiological ~rotection

. program was implemented in accordance with facility policies and programs and was in compliance with regulatory requirement The inspectors reviewed new procedures and chan~es to pr6cedures that were implemented d~ring the inspection p~riod. The review consisted of a verification for accuracy, correctness, and complia~ce with regulatory requi re*ment ' 7

..

~

. ;.

. These reviews and observ.at ions.were conducted to verify that facility operations were in conformance with the requirements established under

• technical specifications, 10,CFR, and administrative procedures:.

Various operaiion~l occurrences were also re~iewed as f6llows~

-

!..

-*

On two *occasions during the inspection perfod, the *inspect.ors* observed~

• the Unit 2 Control Rod Driv~ (CRD) AtcumJlator High Level/Low Pressure (AHLLP) control room.annuriciator (902-5 G-2) in the alarm ~ondition. *o both occasions the Nuclear Station Ope~at6rs (NSO) indicated the *

annunciator was in the alarm condit~on at sh~ft change *nd the Equipmen *Attendant would check thi local accumulator panel, located in the teactor building, during t~e performance of the shift routin The inspectors inquired i.f any record or 1 og entry would be made if the source of the AHLLP alarm was.confir~ed to be 'ither high* water level or low pressure

• in the CRD accum~lator. The NSOs indicated CRD AHLLP ala~~s were no*

• longer logged:*

The AHLLP annunciator response procedure, Dr~_sden Oper.ati ng Abnormal.

(DOA) 902-5 G-2, Revision ~; requtred ~*determination to.be made cif.the source of the alarm, either high water.lev.el or low pr~ssure, at the *

local CRD accumulator alarm* panel. *Step 4 of DOA* 9.02-5 G.:.2-further requfred operations personnel to review the Accumulator High Water/Low

• Pressure Alarm Log.(AHWLPAL), maintained a.t the Unit operators'desk, to

• dete~mine if a pirticula~ accumulator ~ad ~ reo~cur~ing 'proble Ste~ 7 of DOA 902-5 G-2 required operation~ personnel to. make~~ applicable.

. >. -entry i ri.the* AHW~PAL for ea'ch high wate.r 1 eve l or 1 ow pressure a 1 arm-received in the unit control ro6 Additionally I Dresden Admfoi strati ve Procedure (OAP) 7-5; Operating Logs and Records, Revision 8; ~rovided detailed instructions for the maintenance *Of reco.rds and* logs that were administratively required to be:

maintained for the life of the plan Step B.8 of OAP 7.5 required *a AHWLPAL to be maintained for each_ unit as an on going record ~f CRD ~

c accumulator. alarms. :.The AHWLPAL served as a mechanism for the

. identification, tracking and mitigation of repeating high. water.level or low pressure conditions on individual CRD accumulators.* Upon NRC

.

request, tKe licensee ~reduced the Unit 3 CRD.AHWLPAL for the inspect6r's rev.iew but fndicated the Unit *2 AHWLPAL had been Jost.. The. iast entry in the Unit 3 AHWLPAL was-.made in April 1990. *The. average"frequency of. *

AHLLP alarms was approximately dnce per ~hift per uni The inspectors ascertained through interviews that the CRD*system*

engineer was aware of the programmatic failure of op,rations personnel to maintain the.AHWLPAL, p,er the administrative requirements of OAP 7-5 and DOA 902-5 G-2, sinc*e approximately May '1990. * OAP 9-12, "Procedural Adherence Deficiencies," Revision 0, required procedural adherence deficiencies identified, including *failure to meet the procedural intent or to per.form steps and activities containedwithin a procedure, to be documented on Form 9-12A; When questioned by the inspectors, it became apparent that the system engineer was not cognizant of.the*

requirements of OAP 9-1 Additionally, the. systems engineer did not recognize that the failure of operations personnel to maintain the

--

-

,._:T"

~.-.

• .

~.

• • • .. *.

'*'"

AHWLPAL constitutea*a condition that was adverse to qualit The inspection revealed t~at no OAP 9-12A, or other mec;:hanisni documenting the condition had been initiated by the license * On August 30, 1990, Dresden Station transferred the. requirements for the AHWLPAL f~om OAP 7~5 to the Unit Operator's Daily Surveillance* Log, Append1x A;* The transfer was the result of a c6mmitfuent to a third party

• re.Vie\\i organization to include -documentation of independent verification_*

of valve manipulations*during CRD accumulator charging, or.wat~r re~oval activit_ies, fnitiating from AHLLP alarms. *Following the transfer *of the AHWLPAL requirements to Appendix A, the. i n*spectors verified* the proper maintenance of the log-by the shift NSO *

The -failu*re to complete *the AHWLPAL, between Aprii 1990 and *

Augu~t 30~ 1990,.is of.concern: because:

.*

Plant operatiOns~personnel,-without proper direct.ion, stopped performance of the CRD accumulator record keepihg activities.*

.This log was administratively required to.be maintained for the life of-the plant:

• *

.

.

...

. At. least one system.engineer wa~ unaware of the mechani srri.and import~rice of documenting ~rocedural adherehce deficiencie This fssue is considered an unresolved 1te*(Tl (50-237/90019-01: (DRP)).

pending a revi_ew by the licensee to determine the root cause* and to the extent plant staff is.unaware:.of the me_chanism of reporting procedural:*

-. non-adherenc Further review. of the matter is al so required by: the.. NRC....

.

..

..

,

.

'

~..

No violations or deviations *were identified in.this are Maintenance and Surveillances (62703, 61726, and.g3702) Maintenance Activities

,*

• $tati~n niaintenan~e activities of syitems a~d compone~tl listed belo~ were observed*or review~d to ascertain t~at they wer~

. conducted in accordance with *approv"ed procedures, regulatory guides_

ahd ~ndustry codes or standards and in conformance wi.thTechnical

,. Specificatfons. *

The. following items ~ere considered d~ring ~his r!~iew:,

The Limiting Conditions for Operation (LCOs) w~re met while.

componer:its or _systems were removed from service; approvals were *

ribtained prior to initiating the work; activities were accomplished

, using a~proved procedures and were _inspected as applicable;.

functional testing and/or calibratfons were performed priOr to returning components or systems to service'; qua 11 ty control records.

were maintained; activ.ities were accomplished by qualified

personnel; parts and materials used were properly certified;

• radiol_ogical controls were implemented; and, fi_re prevention

-.~*------=.

• .

. '.*

...,,

• .

• ,..

controls were implemente Work requests were *reviewed to de:termirie status of outstanding jobs and to assure that priority is assigned.

to safety""reiated equipment maintenance which.may affect system performanc :(l) Oh July 30, )990, while replaCing a burned out light bulb.

on the control room indication for inboard manual isolation*

valve 2-JS01.:..26A on the. low pressure coolant injection (LP.CI).system, a Group V isolation signal* w~s. generated.*.

• when a power supply fuse opene As a result,.the isolation.condenser automatically i~ola~ed. The licensee

. i01mediately. replaced the failed fuse and reset the isolation signa Fuse 595-714B was the supply *power fuse for both the -LP.CI valve indication and the.isolation condenser primary containment group V isolation circuitry.. The fuse.failure was attributed to. the reduced r~sistari~e a~sociated wit~ the

• replacement bul Two different voltage rattrig~ were ~sed

• in indicating light buJbs tn ~h~ cont~ol room, iSS and** *

.'* 30 ~olts. Dresden bperations Procedure (bop) 640-4 11Control Panel Light Bulb Replacement 11 did not provide instru~tions to deter~ine the correct yoltage ratin~ fo~ use in the various application The isolation signal affected only th.e isolation condenser and did not result in any

violation of Technical Specifications. The licensee revised*

. the pro~edure to include in~tructions limiting the use of'

30 volt bulbs to,th~ neutron monitoring.~ystem. A*separate

'storage area w~s provided in the* control room for 30 volt..

bulb Operations Departffient Memorandum No.* 19 ~as* also Jssued

• to explain the proper ~se of the 30.volt bulb~.* As this was.*

considered an *isolated event and approptiate corrective actions.*

.were initiated prior to the end of the inspection, a Notice of Violatiq~ is no~ being issued in.accordance with 10 tFR 2,

Appendix C,.Section V.A:

(50-237/90019-02(DRP))~ The inspector:

has no other concerns in this are.

.

The system engineer il')itiated a drawing change request to c;:larify the circuitry interaction associated with LPCI valVes and primary containment isol.atio In addition, the licensee.

initiated a circuitry desigri review to determ.ine.if separation of the affected circuitry is *appropriate!-_.. j:~Sep*a~ation of the control and protection circuitry and rev*iew of *speC:if1t:

requirements as they apply to* Dresden are cory_side.~.~d. *to be an

.unresolved i:tem (SQ-237/90019-03(DRP)).

.-~:.r>*.:

• *

(2)

On August 2,* 1990, while Unit 2 was in cold shutdown, a spurious primary containment group V isolation occurre Based on the. history of previous isoiation events, *the..::UCe!lsee believed. the differential pressure instrumentation, which*

initiated the isolation, ~as vibration sensitive; Immediately following the event, operations personnel were dispatched to the area of the differential pressure instrumentation. A stati.on labor supe*r.visor was found to be-within direct eyesight of the instr~mentation during the time*of the even Based on

.

~-.

• "

.

,.,.. *

• • ... r.,

.'..

the ob.servations of the station labor supervisor, *no personnel were identified as having inadvertently jarred or otherwise pert~rbated the isolation initiation in~truments.. The licensee's investigation did not identify any abnormalities, open fuses 6r other electrical proble~s fn the group V.

.

• .isolation circuitr As part of the corrective action, the

.1 icensee pi aced* a placard on th.e instrument rack to caution*

~erson~el against inadvertent jarri~g. Additionally, the

• licensee planned to perfo~m vibration monitoring on the*

isolation initiation components prior to the. completion of the current Unit~ refue1ing o~tage..

(3)

On ~ug~st 20\\ 1990, while Unit 2*was at 82 percent power, th high pressure toolant injection (HPCI) steam line flow differential pressure transmitter (PT), 2-2352, was found ou~side of tolerance during t~e perfor~ante of Dresden*

. *Instrument Survei Hance (DIS) 2300-1 The PT was one of two

• transmitters which {nitiated an iso1ation to the HPCt steam supply following detection of high steam flow in the ~PCI steam

. supply. line.*. The* plant Instrument Maintenance Department (IMO)

was unable to successfully recalibrate the PT during

performance. of the. survei 11 an*ce procedure." : The 1 i censee declared the PT.inoperable and in accordance with Technical Specific~tion (TS) Table 3.2.1, Attion*o, closed the HPCi steam

.supply ~solation _valve The HPCI system was also subsequently declared _inoperable per)S 3.5.C. The licensee fullfilled the. immediate reporting* requirements per 1,0 CFR 50.'72 fo 11 owing the ~eclaration of the HPCI sys~em inoperabl A replacement *

PT was *ins_talled on August 22, 199 Foll.owing the calibration

• and testing of the replace~ent PT, the HPCI system was returned fo operable status and TS Limiting Condition for Operation

(LCD) Actions 3.5.C.2.a and Table 3.2.1 were terminated~

.

.

Diagnostic testing was conducted*by the plant IMO staf following the removal of-th~ PT in an effort to determine the cause of the failur During bench testing, the,PT was successf~lly recalibrated, however, the causi of the fail~~e could not be-.determined by the. license Following*

decontamination, plant personnel plan to return the inst~ument to the vendor, Rosemount. Inc.,- for further failure analysi The licensee's response dat.ed July 20,."1990, <to NRC -

Bulletin 90-01, "Loss of Fill Oil Transmitters Manufactured By Rosemount,~ included a list of all Model 1153 Ser1es*B,

. Model 1153, Series D and Model 1154 differential pressure*

transmitters utilized in safety related systems and systems installed in accordance with 10 CFR.50. 6 A 1 though. PT.* 2-2352 was included on this list, it was not shown as frcim the suspect manufacturing lots identified by Rosemoun However, because of the unavailability of an improved model, the replacement Rosemount transmitter was one fro~ a suspect manufacturing lo The licensee planned to change out the replacement transmitter prior to the restart of Unit 2 following the ~ompletion of the current refueling outag *

'.

~~...

(4)

On August 2, 1990~ with Unit 2 at 87 percent rated thermal power, Target Rock,Saf~ty Relief Valve (TRSRV) 203-3A *

spuriously opene The first indication noted by the NSO was th~ ~caustic monitcir actuated alar A check of the backpanels

.indicated that the TRSRV tailpipe temperatur~ was 310 degrees F compared to a normal. 120 to 140 degrees A. dro'p of

-600, 000 lb/hour steam* fl ow was al so note The crew fo 11 owed

.. the actions prescribed i.n Dresden _Operating Abnormal (DOA) -

procedure 250-1; "Relief -Valve Failure".to attempt closing the

• ,TRSRV by cycling its contro.l switch and Automatic. -

. Depressurization.System (ADS) Inhibit.Switch" and by removin fuses to de~enetgize the valve's control circui Due to increasing torus water temperature the emerge_ncy operating procedures were e,ntered and torus cooling was initia-te The reactor was manuaJly scrammed about eleven minutes following first ind{cat1o~ of the.valve opening prior to exfeeding *

1io degrees r. torus *water temperatu're 1ri accordance'* with

• .Technical Specification 3.7.A.Lc(3).. Three control *rods which

.

.stopped at posltion 02 were manual)y inser.te (Problems *\\.iith *

....

... control.. rods* stopping at position-.Q2 during a scram and

. correspqnping licensee actions w.ere discussed i*n inspection

. - report 50-237 /90018(DRP)"; 50-249/90018(DRP)). **Approximately

. one. minute after the scram,- the operat~rs opened the ttirbine

. bypass valves tq limit heat *input to the: torus a,nd to lowe *.*.reactor *pressure in an attempt to get the*TRSRV closed. *Th turbine bypass valves were clos~d*within two minutes to. limit the cooldown rate-.. An Urusual.Event _was. declared and*

.

appropriate notifications were accomplished. _The TRSRV wa determined to be closed' by acoustic monitor indicati,on and an

i ncrea*sed rate of torus cooling about two hours *and 45 minutes after the event began when reactor pressure had decreased to less-than lOO-psig~

T~e Unusual Event was terminated following a*t6rus temperature decrease to below 110 degrees Technical Specifications limited the rate of reactor coolant temperature c~ange t6.100 ~egrees ~per h6u Although the

• cooldo~n rate over shor.ter periods was higher,' the *Cooldown

. rate when averaged*.over*one hour was a maximum of 129.3 degrees

• per hou A review of the Dresde~ FSAR and General Electric's *

-Safety.Relief Valve Slowdown (SRVB)"'analysis indicated.that the

'design allowed for twehe SRVB events over 40-.. years of*_

.

6peration:

This anal~~is assumed a maximum'25f.3 degrees-per*

hour coolddwn rate when averaged over*one hour. *The cooldown

.rate achieved du~i~g this event was much l~ss than that indicated in the design analysi Since only two previous SRVB events have occurred on Unit 2, this event was within the design basis analysi The maximum bulk torus water temperature reached du~ing the event was 122 degree~ F ~hich was determined to be within.the Mark I containment design basis*

analysis. *The licensee~s visual examination of the re1ief lines and.su~~orts indicated that* no relief line damage h~d occurred.

..

'...

, '

....

,The TRSRV'was disassembled, in~pected and:tested to determine the cause of the spurious openin The licensee determined that the failure was most likely caused, by a severely steam cut pilot valve disc which allowed pressure to be.transferred to *.

the second stage pis~-0n. This cau~ed the valve to open..

~hen the reactor pres~u~e reached 100 psig, the main valve preload spring force over.came, the reactor. pressure force and the -main *

valve disk closed. This exact failure methanism wa~ the *

subject of :Gene~~l Electric (GE) Service Irifor~atio~ Letter *

(SIL) 196 dated September 30; 197 The TRSRV were overhauled each refueling out~ge in atcordance with T~chnicaJ

,.

Specificition~.

Thi~ particular TRSRV had been installed on February 11, 1989 during the previous Unit 2 refueling outage'.

'As. a result.of this ~vent, *the 'licensee placed.the TRSRV pilot assemblies on the short outage list s6 that any TRSRV in service.for longer than eight months would. be h~pl aced if the

• unit was placed into cold shutdown and the drywell was.,. *

accessibl The licensee a.lso plan.ned to establish a program

• _to monitor the* TR$RV tailpipe temperatures to identify

. potential. pilot valv,e leakage problem A similar program was

, a recommendation of~the GE

~I *

,

While investigating the failed TRSRV:in the drywell, the licensee discov~red that an electrical junction*box (2PB~2020)

. related *to. the TRSRV circuitry,had fallen from its mounting and a conduit had separated from the TRSRV bellows s*eal pressure"* *

switch.. These* were. determined not to have ca.used the event....

• .

The fallen junction bo:X had been attached to* another. junction*

box (2PB-202l) and both junction.boxes were-replaced_ with a*.

• single juncti.on box.with seismically designed supp_orts.. A new,.

support for.the pressure s~itch was installed to replace the old i riadequate support.. The 1 i censee perform.ed a wa l kdown. to i~spect the mounting of electri~al j~n~tion boxes in the Unit 2 drywel Although no other.damage was identified, two junction*

boxes were found to be supported only by their respective

. conduit The licensee determined that the conduits were solidly connectetj to the junction bo,xes and were adequately supported* but planned further anJl~sis ~o determine long term

• acceptabilit The license~ also planned.to do~ similar *

review for Unit 3 ~rywell.junction boxes during *the next outage with sufficient time for drywell access.. *

The operators 1 opening *of the turbine bypass va 1 ves was a judgement call involving two opposing concerns (limiting increasing torus temperature versus limiting th*e reactor cooldown rate).* While the procedures*wer~ not written to specifically call for thts acti~n, they did not specificall, exclude this ~ction. A review of -Operator actions during the event was performed by theNRC Office for Analysis and Evaluation of Operational Data (AEOD) throtigh the NRC program to study factors which affect human performarice~ Operator actions are therefore, to be evaluated through 'that program and reflected through the associated report:

13...

.

.

Confirmatory Act.ion Letter (CAL)-RIII-90-014 was issued on

~ugust 3i 1990; to addre~s the variou~ issues regarding this event:

The CAL was terminated in.a letter dated August 14, 1990; following verification.of.licensee*actions in accordance with the CA The l.icensee submitted a.wriiten response to the 'CAL dated September 13, 1990,. to provide th result,~ of the various evaluations. *These were.reviewed by the inspectors as discussed above..

Surveillance Act1vities The inspeciors observed s~rveillance testi~g. including required Technical Spicification surveillante testing, and verified for*

actual activi~1es observed that testing ~as performed in accordance

. wi~h adequate pr6cetjures. * The i~spectors also verified ~h~t test instrumentation w*s calibrated, that Ltmiiing*Conditions for Operation were met, that removal~ and ~estoration of the affected components were atcomplished and that test results conformed ~ith *

Technical. Specification and procedure requirement Additionally,

"the i nspect6rs ensure.d that the test results were reviewed by peJ'.'SOnnel other than the. individual Oirecting the test, and.that any deficienci~s identified-during ~he testing were properly reviewed

and resolved by appropriate management per~onne The inspectors.witnessed *or reviewed portions of th'e following test activity:

Units 2 and 3 Nucle~r Instrumentation ~alibration

' '

.

.

.

No* violations.or-deviations were.identified in this are.

Inspection o('Check Valve Programs Effectiveness (73756)

The inspector reviewed the licensee's programs and procedu~es pertaining to inspection and testing of check valve The following is a general*

description of the licensee's programs and the v~rious check *valve..

designs used at Dresde *

The Dresden *rnservice Testing (IST) Program contained ~heck

. _.valves; stop check and excess flow check valves meeting the

"requirements of the American Society of Mechanical Engineers*

(ASME) Class 1, 2 and 3 valve In addition, the !ST program contained both ASME classified valves of sizes not required by.*

Class 1, 2 and 3, and thereby exceeded the requirements of ASME Sett ion *x The licensee also incorporated non-ASME required check valves in.the !ST progra The basis for addition of these.

va 1 ves, although on non-ASME portions of the systems,.was th,e potential of these valves.to affect the systems important to safet This action on the pa rt of the licensee exceeded the.

requirements of 10 CFR 50.SSa(g).

The augmented check valve program, in addition t6 the requirements of 10CFRS0.55a(g),

contained about 100 check valves (50 per Units 2 and 3 each and 6 common to both units).

The 10 CFR 50.55a(g) required portion of

.the !ST program contained about 300 check valves (147 on Unit 2,

-,

,.....

• ,

• ,,

. '

149 on Unit 3 and 3 com~on to both unit~): The IST pr6gra~ tested check valves ranging from a size of.25 inch to 24 inche Valves less than 2 inches*were not required to be tested by the CECo corporate program and therefore the progr~m e~ceeded the licensee's

• corporate program requirements. Approximately.100 of the 400 check valves tested.at Dresden were classified as either non-ASME

. Code Class -1: 2, or 3,~ but safety related; or non-ASME Code

• Class 1, 2, or 3,-and not safety relate *The~ types of IST te~ts applied to check v~ives~inc1ude~ 10CFR5o,*

Appendix J valve seat leakage, and/or check valve*exercise full open, and/or check valve exercise f~11 clos The testi~g.schedule.

. for check. valves was primarily on a refuel.tng outage basis not to._. - *

• • exceed.two years: Several.also had test frequencies of quarterly during normal operatio Ninety-nine of the ASME *Code. Class 1, 2',

~nd 3 check va1ves were tested.quarterly.dur~ng nd~mal*operation (47 on Unit 2;'49 on Unit 3 ~rid 3 common to both units)..

.

..

....

• .

"

.*

.

.

'*.

• In addition *to.the IST program, Dr~sden developed arid implemented a**

check valve inspection ptogta Dresden Administrative Procedure,

.OAP u..:25, Revision 02~ "General Check Valve Irispection Program,".

controlled.the inspection' actiVftie DAP.11-25 incorpor.ated industry recommendations, CECo Corporate Directive (Directive NOD~TS.9), Dresden IST program, ASME Section XI and NRC Generic Letter 89~0 OAP *11..:25 contained about 200* check val~es in 29 categories. * About 150 of these.chec.k va i ves were beyond the scope of the* I-Sl progra The. 29 categories were _based upori industry. -

recommendati6ns'for ex~ahdin~ the IST progiams and to incorporate i*ndustrY lessons* lea rn~d *as*soci ated with check valve *fa i1 ures. :*The*

. 29 ca_tegori es encompassed the fo 11 owing s.yste!Tls/subsyst._eins:

..

..

._ * Main.. Steam Isoiation Valve (MSIV)

• *Diesel Generator Coolin~.-Water Diesel Air Start Main Feedwater*

._ Cor.e Spray_. _

Pressure Suppression Reactor Water Cleanup Standby Liquid* Control Reactor Head Cooling

.

Low Pres~~re Coolant Injection Containment Closed Cooling.

._Hi~h Pressure Co6lant lnjectio~ *

-Condensate Demineralizer Service Water Extracti6n Steam

The inspection program contained safety related, non-safety related

• but important to operability, and balance of plant check valve *OAP 11-25 prescribed an inspection frequency (1) for at least one

• • valve in each category to be inspected prior to the 'end of the Unit 3 Cycle 11 refueling outage (completed in February 1990) and (2) of not less *than once every 10 years for all check valve The

,.

'

~-

.'.t *.

'.

. inspection also rec"orded the conditions of erosion/corrosion,

.

physical da~age, fr~cturesi cracking~ missing pa~t~, loose or br6ken parts, thread damage, impact damage, local Yield, wear surfac~

• indication, out-of-alignments, out-of-plane, ex.isting forefgn materials and. any"possible indication of material sketching~**

.

\\

'*

.

The procedure also c6ni~ined criteria for deierminin[ fai~ure and

• cor~ective re~airs. The inspector found th~t-the criteria was

. conse~vative in that~ degraded condition ~f check valve iriternals :

. represented a failure.. Additionally, if any significant degradation

. was identified such ttiat the check_ valve was* inop~.rable during the previous operating cycle or would have became inoperable during the subsequ~nt tycle,.then'* failure ~~alysfs would be performed and.ari additiona"l check valve chosen froni the same category would*also be" disassembled and -Inspected.. *

-

The. inspector reviewed.the* check valve inspection* data.base and d_etermi.ned that 27 check valves had been.identified as fa11ures s.ince the check valve inspection program had been implementec Of these 27 check valve fa.ilures, 25 failures were predictive in.that*

the degrading conditions had been di.scovered prior to inopera:btlity of the check valve *

The ins~ector reviewed these prbgram~-a~d verif~ed completeness veri fyi rig that the check va 1 ves of the fo l1 owing systems were

_iMcluded in the IST tes~ing and/o~;the check valve inspection programs:

..

High Press~re Coola~t Injection Low Press~re Coolant lnje~tion Main Steam and Main :Steam Drain Systems

.

.

by The *.inspector. also verified that check valves on systems *.important to safety were _encompassed by the licensee's programs, including

• *

verification of the safety functions associated with the check valves.*

. L

'.

-*--***~

\\1,

'..

l

',,

" "

.. The.inspector selected sav~ral check va)ves and v~rified throug review of the Dresden IST Program (July 1, 1989) and record review that the licensee's testing*methodol~gies demonstrated that the valves were capable of per.forming their required safety function The. testing methodologies included1eak r~t~ testi~g and_

. verificatiot') of full open or ful.l close function _or both. *

In addition; 207 check valves wefe periodically insp~cted for*

internal degrading conditions. These degradi ng*.components were.**

either *~eplaced or evaluated for continued operation~

_The in~p~c~or al.so review~d the licensee's ~r~v~~tive and predictive maintenance activitie~ associated~ith check valv~s ~nd determine that the activities were effective in identifying degradation before *

failure.. The replacement of the internals associated with 25 of th theck valve failures was an example of the ~ffectiveneis of these predictive and preventive m~inten~rice activitie *

.

.

.

.I

• .

.

• • The.~nspectbr reviewed the.licensee'i onsit~ tra~ni~g progr~m

• associated with.the check valve inspec:tion. program *for maintenan.ce and engineering personne Th~ training,program*was detailed and cohtain~d ihformation b~sed upon the different d~signs of check*

._valve types, advantages and disadvantages of different maintenance practi~es, failure modes of thevar'ious check. valve internals, an lessons learned from NRC.and industry studies~ as well as

• ':

experiences*at.the six CECo sites. **.The.training program also i nc.l uded_ a v i.deo. frorri.actual inspect i ori s per.formed. at Dresden per OAP 11~2s: This Video *demonstrated the vari~us findings and ex~mpl~s.of failutes that coul¢ occu Th~ insp~ction found~~~

training'pr6gram to be an effective enhanceme~t to~the che~k valve inspection progra Unr~solved Items

. *.*

Unresolved items are matters about which more information is required in order to ascertain whether.it is an acceptable item, an. open item, a dev1at:i.on or a violatio Unresolved items disclosed during this inspectio~' are discussed in pa~agraphs 4 ahd s.~(1)~ Report Rev~ew During the inspection period, the inspector reviewed the licensee's Monthly Operating Report for Augus The inspector confirmed that*the information provided met the requirements rif Technical. Specification 6.6.A.3 and Regulatory.Guide 1..1..

..

.

\\.

'9.

.!..

.

Exit In~erview (30703)

The inspe~tors.met.wi~h licensee representatives (denoted tn Paragraph 1)

on September 2s; 1990, and informally throughout the inspection peri6d,. *

and. summari.zed the* scope and findings of the inspectiOn activities;

.Th.e inspectors also discus~edthe likely inf~rmationa*1*co~tent of the

• inspection report.with regard to documents or processes reviewed by th inspector dur.ing the inspectio The *licensee did. not.identlfy any such documents/processes *as proprietar The l i cense.e *acknowledged the

. f~ndings bf.the inspectio : ~

18