RS-21-066, Amendment 23 to Fire Protection Report, Volume 3, Cross Reference from Old to New Configurations of Fppdp

Amendment 23 to Fire Protection Report, Volume 3, Cross Reference from Old to New Configurations of Fppdp

ML21179A055
Person / Time
Site:	Dresden
Issue date:	06/21/2021
From:	Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML21179A042	List: ML21179A027 ML21179A028 ML21179A031 ML21179A032 ML21179A033 ML21179A034 ML21179A035 ML21179A036 ML21179A037 ML21179A038 ML21179A039 ML21179A040 ML21179A041 ML21179A043 ML21179A044 ML21179A045 ML21179A046 ML21179A047 ML21179A048 ML21179A049 ML21179A050 ML21179A051 ML21179A052 ML21179A053 ML21196A008 ML21196A009 RS-21-066, Amendment 23 to Fire Protection Report, Volume 3, Cross Reference from Old to New Configurations of Fppdp
References
RS-21-066
Download: ML21179A055 (762)

Text

FPR VOL 3

CROSS REFERENCE FROM OLD TO NEW CONFIGURATIONS OF FPPDP DRESDEN 2 &3 HISTORICAL VOLUME - LICENSING BASIS SUPPORT DOCUMENTATION Book 1 I Correspondence Referenced in Fire Protection Safety Evaluation Reports (see Vol. 1 and 2 of FPPDP)

Book 2 II Major CECo Submittals Referenced in Fire Protection Safety Evaluation Reports (see Vol. 2 of FPPDP)

III General Fire Protection Correspondence (see Vol. 2 of FPPDP)

( Book 3 Fire Protection Drawings (For Reference Only) (see Vol. 3 of FPPDP)

VOLUME 1 - LICENSING BASIS I Regu1atory Regulatory Documents (see Vol. 4 of FPPDP)

II Safety Evaluation Reports for Appendix A to BTP APCSB 9.5-1 and Functional Responsibilities (see Vol. 5 of FPR)

III Safety Evaluation Reports for 10 CFR 50 Appendix R, Sections III.G and III.L and 10 CFR 50.48 (see Vol. 5 of FPR)

IV Fire Protection Technical Specifications and Related Safety Evaluation Reports (see Vol. 3 of FPR)

V NRC Inspection Reports (see Vol. 3 of FPR)

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(~- Cross Reference (cont'd)

VOLUME 2 - UPDATED FIRE HAZARDS ANALYSIS I Fire Hazards Analysis Report (see Vol. 1 of FPR)

II Fire Hazards Analysis Preparation Summary (see Vol. 4 of FPPDP)

III list List of Fire Protection Drawings (see Vol. 4 of FPPDP)

VOLUME 33 -- APPENDIX R SECTIONS III.G. III.J, III.J. AND III.l III.L CONFORMANCE Book 11 I Safe Shutdown Report (see Vol. 2 of FPR)

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II Safe Shutdown Report Preparation Summary and Division of Responsibility, January 27, 1987 (see Vol. 4 of FPPDP)

Book 22 Interim Measures/Exemption Requests I Executive Summary (see Vol. 4 of FPR)

II Interim Compensatory Measures (see Vol. 4 of FPR)

III Appendix R Exemption Requests and AnalysiS (see Vol. 4 of FPR)

IV Letters (see Appendix R Exemption Requests and Analysis Transmittal letters Vol. 4 of FPPDP) vV Generic letter Letter 86-10 Evaluations (see Vol. 4 of FPR) 2

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Cross Reference (cont'd)

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VOLUME 4 - FIRE PROTECTION REPORTS/SUPPLEMENTARY GUIDANCE Book 1 I Combustible Loading (see Vol. 5 of FPPDP)

II Structural Steel Analysis (see Vol. 5 of FPPDP)

III Fire Rated Barrier Design Support Data (see Vol. 66 of FPPDP)

Book 22 IV Supplementary Guidance Review (see Vol. 7 of FPPDP)

( V Emergency Lighting Report (see Vol. 7 of FPPDP)

VI Rated Barrier Maintenance and Surveillance Program (see Vol. 7 of FPPDP)

VII Communications (see Vol. 7 of FPPDP)

VIII Safe Shutdown Equipment Access (see Vol. 7 of FPPDP)

IX Volume 4 Microfiche (see Vol. 7 of FPPDP)

X OSHA Fire Protection Requirements (see Vol. 7 of FPPDP)

VOLUME 55 - NFPA CODE CONFORMANCE (HYDRAULIC CALCULATIONS)

Book 1 NFPA Code Review (see Vols. 8 and 9 of FPPDP)

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3

Cross Reference (cont/d)

Book 2 I Fire Suppression System Hydraulic Verification Study (see Vol. 10 of FPPDP) .

II Drawings in Support of the Hydraulic Verification Study (see Vol. 10 of FPPDP)

III Hydraulic Calculations (see Vol. 10 of FPPDP)

Book 3 IV C-Factor Verification Data (C = 80) (see Vol. 11 of FPPDP)

V V Supplement One - Hydraulic Verification Study (see Vol. 11 of FPPDP)

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VI Supplement Two - Hydraulic Verification study Study (see Vol. 11 of FPPDP)

VII Supplement Three - Hydraulic Verification Study (see Vol. 11 of FPPDP)

VIII Supplement Four - Hydraulic Verification Study (see Vol. II 11 of FPPDP)

VOLUME 6 - REFERENCE Book 1 I Fire Protection Program Audit and Open Item Closure (see Vol. 12 of FPPDP)

II Previous Commitment Review and Open Item Closure (see Vol. 12 of FPPDP)

III PLC Review of Procedures in Support of Technical Specifications (see Vol. 12 of FPPDP)

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t( . Cross Reference (cant/d)

(cont'd)

IV Fire Protection Procedures (see Vol. 12 of FPPDP)

V Pre-Fire Plans (see Vol. 12 of FPPDP)

VI Suppression Effects Analysis (see Vol. 12 of FPPDP)

VII Fire Protection Reevaluation Project Plan (see Vol. 12 of FPPDP)

VIII Fire Protection Evaluation Plan (see Vol. 12 of FPPDP)

IX Audit/Inspection Reports, Responses, and Closure (see Vol. 12 of FPPDP)

X Fire Protection Supporting Calculations (see Vol. 13 of FPPDP)

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XI Miscellaneous Letters, Memos and Meeting Notes (see Vol. 13 of FPPDP)

XII Volume 6 Microfiche (see Vol. 13 of FPPDP) 5

VOLUME INDEX

( DRESDEN 2 &3 FIRE PROTECTION REPORTS Volume Index VOLUME 1 - Updated Fire Hazards Analysis VOLUME 2 - Appendix R Conformance (Sections III.G, III.J, and III.L) - Safe Shutdown Report VOLUME 3 - Technical Specifications, Technical Requirements, and Inspection Reports VOLUME 4 - Interim Measures/Exemption Requests

(. VOLUME 5 - Safety Evaluation Reports

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Dresden 2 &3 FIRE PROTECTION REPORTS Volume Index VOLUME 3 - TECHNICAL SPECIFICATIONS. TECHNICAL REQUIREMENTS. AND INSPECTION REPORTS I Fire Protection Technical Specifications and Related Safety Evaluation Reports 1 February 22, 1989 Proposed Amendment for Removal of Fire Protection Technical Specifications 2 June 30, 1989 Technical Specification Amendment which removed the Fire Protection Technical Specifications 3 August 9, 1989 Technical Specification Amendment corrected index pages.

II Dresden Administrative Technical Requirements (DATR's) for Fire Protection III NRC In~pection Reports 1 Inspection Report No. 50-010/84-01, 50-237/84-06, 50-249/84-05 2 Inspection Report No. 50-010/84-09, 50-237/84-11, 50-249/84-10 3 Inspection Report No. 50-237/85033, 50-249/85-029 44 Inspection Report No. 50-249/86006 5 Inspection Report No. 50-237/87035, 50-249/87034 6 Inspection Report No. 50-237/87037, 50-249/87036 7 Inspection Report No. 50-237/88010, 50-249/88012, 8 Inspection Report No. 50-237/88030, 50-240/88031, 9 Inspection Report No. 50-249/89044, 10 Inspection Report No. 50-237/89008, 50-249/89009 11 Inspection Report No. 50-237/89013, 50-249/89012

":~l/ . .. 12 Inspection Report No. 50-010/89002, 50-537/89017 50-537/89017,, 50-249/89016 1

AMENDMENT 161

/ DRESDEN 2 & 3 I

FIRE PROTECTION REPORT Volume Index Volume 3 - (cont'd) 13 Inspection Report No. 50-237/89022, 50-249/89021 14 Inspection Report No. 50-237/90017, 50-249/90017 15 Inspection Report No. 50-237/90023, 50-249/90023 16 Inspection Report No. 50-237/90027, 50-249/90026 17 Inspection Report No. 50-237/91004, 50-249/91004 18 Inspection Report No. 50-237/93002, 50-249/93002 19 Inspection Report No. 50-237/96002, 50-249/96002 C*

C:-,) 20 21 Inspection Report No. 50-237/96012, 50-249/96012 Inspection Report No. 50-237/97021, 50-249/97021 22 Inspection Report No. 50-237/98029, 50-249/98029 23 Inspection Report No. 50-237/02-06 (DRS) , 50-249/02-06 50-237/02-06(DRS) I (DRS) 50-249/02-06(DRS) 24 Inspection Report No. 0500023712005002 05000237/2005002 (DRS) ,

I 05000249/2005002 (DRS) 2

Tab I Revision 8B April1 1992 Apri c DRESDEN 2&3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Fire Protection Technical Specifications and License Condition fire Fire protection technical specifications have been removed per Generic Letter 86-10, as discussed 1n 86-10. the following letters which are included in this in ~he section:

1 February 22.

february 22, 1989 CECa CECo letter from J. A. Sllady (NLA) to T. E. Murley Sl1ady (NlA)

(NRC) transmitting a proposed amendment to replace the Fire Protection Technical TeChnical SpeCifications Specifications with a standard license condition and appropriate administrative procedures.

2 June 30, B. L. Siegel to T. J. Kovach (CECo) 3D, 1989 NRC letter from 8. (CECa) issuing Technical TeChnical Specification amendments to replace the existing license conditions on fire protection with the standard condition noted in Generic Letter 86-10.

3 August 9, 1989 NRC letter from 8. B. L. Siegel to T. J. Kovach (CECo) issuing corrected index pages for the Technical Specifications amendments provided by letter dated June 30, 3D, 1989.

I .0- i

Tab 1 Commonwealth Commonwe.lth Edleon EclllOn One FirS! National Plaza. Chic:l~. INinois Address A.Ply to: Post Office x 767 Revision 8 Cha.go, IMil'oQis 606iO

• 0767 April 1992 February 22,

22. 1989 Hurley, Director Dr. Thomas E. Hurley.

Office of Nuclear .eactor aeactor ** ;ulatlon

.equlation ATtN. Doc~ent ATTN: Document Control De.k De.t U.S. Nuclear Regulatory C~i ** ion Commia.ion Washington.

Hashing ton, DC 20555

Subject:

Dre.den Nuclear Power StatioD UDit. Unit_ 2 and 3 Propo.ed AMendment

~endment to the rire Fire Protection License Condition and:Technica1 and:Technical Specifications NRC Dociet Ho,. Noc, 50-231 apO $Q-240 sO-Z4V

Reference:

Generic Letters 86-10 and 88-12 dated April 24, 24.

1986 and August 2. 2, 1988 1988,. *. respectively.

Dear Dr. Hurley:

Pursuant to 10 CrR CFR 50.90, Commonwealth Edi.on proposes to amend Provilional Operating License DPR-19 and Facility pperatin9 Provisional pperating Licenae Licen.e DP8-25 DPR-25 for Dresden Nuclear Power Station and their respective Appendix A Technical

'--~ Specifications. The proposed amendment revi.es revises the Units 2 and 3 Licenses and Technical Specifications in response to the referenced Generic Letters and-As aftd-As p&.rt pf:..rt of the Dresden Improvenlent Program's Technical Specification Action Plan. Ple-n.

The referenced Generic Letters Letter. auggected auggested replaCeMent replaceMent of Fire Protection Technical Specifications with It a .tandard atandard licens.

licen.e condition and appropriate administrative procedures, after updating tbe rSAR FSAR to reflect the approved fire protection program. Similar amen~ents amendments have bave been previously approved, such as Amendment approved. AMendment 10 to the Byron Station Operating Licen.es (NPF-37 and NPr-66)

NPF-66) issued September 9. 9, 1987. The changes are summarizedaummari.ed in Attachment 1 and further described in Attachment 3. The affected pages page. of the Licenses Licen.es and Technical Specification. are contained containad in Attac~.nt Attachment 2.

The proposed propo.ed changeschangel have b.en been reviewed and approved by both On-Site and Off-Site Review in accordance with witb Coamouw.alth Commoawealth Edison Ediaon procedures.

procedure.. He bave reviewed these proposed ~eDdmentl heve amendment. 1n accordance with 10 erR in accordaDce CrR SO.92(c) 50.92(c) and

.ignificant hazards determined that no significant hazard. consideration con.ideration ezilts.

exi.t.. Thi.

Tbi. evaluation

i. documented in Attachment 4.

Enclosed as Attachment 5 are the propo_ed propo.ed Dresden Dre.den Administrative Admini.trative Technical Requirements (DATRs) for Fire Protection. Tbey They are .ubmitted submitted as

,upporting information for this ~.ndmeDt supporting amendment but have not ,et yet ~eD been approved for implementation by On-Site and Off-Site Revie~. Review. Since lomeaome DATR provisions are different from existing Technical Specifications, they cannot be fully

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1-11 I. 1-1.

Revision 8 April 1992 Dr. T.E.

T.t. Murley re~ruary 22. 1989 implemented until the ~endm.nt

~en6ment has been issued. Altbough .OMe

~een i.sued. _inor changes SOMe .inor change.

should not affect their technical content **

which .houlO .ay y be required prior to final off-site review and approval. CECo on-site and off-.ite CECa believe.

believes the enclo.ed anclosed verlion to be technically adequate preliminary version aupport the Staff'. review ade~u.te to .upport of the propo.ed proposed amendment.

amen6ment.

Commonwealth Edison iis Illinois of our

. notifying the State of Illiaoi.

application for this ameD~ent amendment by traaamittia; tranamlttiD9 a copy of this and its thi. letter aad attachmentl to tbe attachments the desiqnated deliqnatea State Official.

Pleale direct any questions you **

Plea.e y have regarding this .atter to

.ay this office.

thh truly yours.

Very trUly

~q~

9..Cf.f:!::j-Licensing Administrator Nuclear ~icensin9 A6ministrator 1m Attachments 1:

Attachment.s l: Change.

Summary of Changes

( 2::

2 Proposed Changes to Appendis Appendix A 7echnical Technical Specifications for Dresden Units 22 and 3 3: Description and Bases for Amendment Request 4: Significant Hazards Evaluation 5: A6ministrative Technical Proposed Dresden Administrative Requirements for Fire Protection cc: A.B. Davis - R@gional Regional Administrator.

Administrator, RIll P.R.

D.R. Muller - Project Director. NRR S.G. DuPont - NRC Senior Resident Inspector. Dresden B.L. Siegel - Project Manager. NRR P.R.

D.R. Hoffman - Escel Eacel Services Office of Nuclear Facility Services - IONS SW0ll!:, to

~day r, ~~~~~~_. 1989 5554K I .1-2

ATUCI!!SENI 1 Revision 8 SUMMARy Of SUHMAIX or ClANQES plANCES April 1992 chan, ** hive follo~inc chanl.'

The followinc have been identified for Ore.den and 3:

Dre.den Unit. 2 end (1

(1)) Pa,e 4 of licen..

Pare (OPR~19 licen.. (DPJl ...19 and -25)

• 25)

Chan,e the licen.e condition Section 3.B for Unit 2 and Section Chan,.

3.G for Unit 3 to tbe .taadard nauclarcl fbe fire prot_cU_

protectioo lie liceD.e coodition

• *e condition identified in Generic LetterLatter 1'-10.

.6*10.

(2) 'a,e. 3/4.12-1 throu,h

'ale. throuah 3/4.12-21 (nIl-I' and 25)

(D,a-19 .ncl Delete all .action.

.ection. of tbe fire ,rotectlO1l proteetioo Tecbnical Tecbaieal Specification Spec ifica tiona***

(3) la,e 6-1 (DP.

(DPR 19 and 25)

Delete canninl cannina requir ...nt.;for requirement.; for fire brllade.

bri,ade.

-(4)

(4) Pa,e 6-7 (DPR-19 Pase (O'R-19 and 25)

Add new 6.1.G.l.a.l1 whicb ne~ Section 6.1.G.l.a.ll which .tate. that the relponsibilities of the Off-Site aeview and lavllti.ative re.ponlibilitie. InvI.ticative

.ball include the review of chance.

Function .hall than,e. to the tbe lire Fire Proaram and !mpl Protection Pro,ram iepl ..entin.

entinc procedurel.

procldure**

(5) Pa,e 6-13 (DPR-19 and 25)

Paae Add new Section 6.1.G.2.a.13 which Which .tate. that the

...-- re.pon.ibilities of th~ On-Site Review Revilw and lnve.ti,ative Inve.ticative c-("-'..,---,'

Function shall include the review of ehanles Prolram and implementinc Protection Program chances to the Fire implementinl procedures.

3055&

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1.1-3

Revision 8 ArtAC!II!l!l 22 ArtACIIm!I April 1992

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PROPOSEP CHANGES 10 PROPOSED LIC!HSt AND 1Q LJC~$t ANP TECHNICAL SPICIFrCAIIQNS SPECIFICATIgNS A((I:!:Ir;1l rAGEli AFFECTEP UNIT 2 PAGES VIII PlIT 3 PAGIGPAGES AFFI~£n A((!!:%EP (OP8-l5!

(PPR-19)) (prR- U )

'oPR-25)

Liceas.

Lie.Dle Paqe ..4 3/4.12-1 Lic....

Lie....

3/f.12-1 314.12-1

..9- 4 P.9. 4 3/f.12-~

3/4.12-2 3/4.12-2

,-~

3/4.12-3 3/4.12-3 31'4.12-4 3/4.12-4 3/4.12-4 3/4.12-5 3/4.12-5 3/&.12-5 3/4.12-6 '3/4.12-6

• ]/4.12-6 3/4.12-7 314.12-7 3/4.12-8 3/4.12-8 3/4.12-9 3/4.12-9 3/4.12-10 3/4.12-10 3/4.12-11 3/4.12-11 3/4.12-12 3/4.12-12 3/4.12-13 3/4.12-13 3/4.12-14 3/4.12-14 B 3/4.12-15 BII 3/4.12-15 8II 3/4.12-16 II 3/4.12-16 BII 3/4.12-17 8II 3/4.12-17 B 3/4.12-18 8B 3/4.12-18 II B ~ 1(..12-19

'U~ 12-19 B II 3/4.12-19 B 3/4.12-20 8B 3/4.12-20 B 3/4.12-21 B 3/4.12-21 II 6-1 6-1 6-7 6-7 6-13 6-13 f.~--

I .1-4

Tab 2 O~*3-4 UNITED STATES NUCLEAR REGULATORY COMMISSION

~/:;'~/J.3,U41''

d c.C:/.

M.

M,

J~II.3.M~"",,

tJ,t~

tl.~"".. ,"," . t WASHINGTON. D. C. 20555 WASHINGTON, 'j....

'i~ ..." . ':.:' ..

-"~ ....

M~~~./J*1~

M.~~.I.J**:t.

June 30,30. 1989

( Revision 8 April 1992 Docket Nos.: 50-237 and 50-249 Mr. Thomas J. Kovich Nuclear Ll,ensfng Hucl~lr L1censing Manager (OIIIIOnwea COMmOnwealth 1th Ed t son Compa Edison Company ny PGst Post Office Box 767 Chicago.

Chicago, 11i1nofs Ili1nois 60690 Delr

SUBJECT:

REPLACE FIRE PROTECTION LICENSE*'ONDITIO~ LICErlSE eONDITIOtl ~ND REMOVAL OF FIRE PROTECTION TECHNICAL SPECIFICATtONS AS PER GENERIC LETTERS 86-10 AND 88-12 (TAC 88*12 (TAe NOS. 71256 AND 71257)

Re: DresdH NuclecU Dresder. Nucleu Power Fower Stat1orl, Statior., Unit Hos~

Nos~ and 3 2 anc!

COIMlhsicr. has issued the enc1u~ed Amendment 11".

The Comlhisicr.

TI14: Uc.. 106 to Provisiona Provfsiona 1 Operatins OperatfllS li Facllit)

License /i".

cense ~o. DPR-19 Operating Licehse F." Iity Oper.l!t1ng liPR-19 for Dresden Ur,'it licehse No. DPR-25 for Dresdt!JI Ur.'it 2 a.nd and AMendment Nu. 101 1.0 AlilendlllEnt Ho.

Dresdt:r. linit 3. Thue llIendrnt!nts IlIendml!nts are 1n arf in respons~

respollst: 1.0~o your applicotiorl ciilttd February 22, 1989.

appl1cotior. ciilUd The afl)rt:lllentiullcd afurt,mentiuncd l!mendments

~lIlendments replac~ httnse conditions elfisting ht~n$e replace the existfng cundftfons on fire ffre prut~ct1on prutection with the w1th the stdnU4rd stdnUard condition noted nuted in GEn~r1c Gen"ric Lett!r Letter 66-10

&6-10 andInd rllDKJv~ r~qufrements r4HIKJvt: requirements for fir~fire cittect10n ottection SystE:lIlS, systeills. flrC suppress1"h $,)'stems, fIre suppreSS1t1h $,)'stems.

c. .

• fire barriers Ind in ir, ten~rfc ter,Hic L~ttEr nre br1gadp.

Inc t'1re LettEr 86-10 and stloffing re~uirernents brigade staffing anc 66-12.

requirements ftS liS pt.r gu1C1llice conta fned guidllice contained A copy of our reli:lt~a relilte.c Sdft=ty Sclfety Eva luation hi~ aliO Ev~luation 111$0 enclosed. The Notice of IssuiJr,ce Issuilrlce

~j II bE:

"'ill uded in the Commission iro~luded bt ir.(.l Commission's FedHal1 Register "otices.

I s biweekly FedHa I,otfces.

Sincere ly, Sfncere ly.

~~~©~D~~\f J\ll \ OzS9

\*,'1*'

• J\0\ ~~<.~~j.ct

~~~~j"t HI.lgor HOM,,,

Project Directorate 111-2

._---- ------~

Division of Reactur Projects III.

IV. V.

V, Ind and Speci61 Spech.1 Project~

III,

Enclosures:

1. ArriencR.£nt No. 106 to Amenalltnt L;cen$~

License ~o. No. DPR-19

<:. Amendment Ijo.

~. I~Ci. 101 to ic.IHlse No. DPR-25 Lic..twse

3. Safety Evaluatic,rlEvaluati"r.

ct cc w/enclosures:

c. See liU I;tl<. t pag~

I .2-

.2-11

Revision B 8

April 1992 Mr. Thomas J. Kovach Dresden Nuclear Power Station DreSden

(/ Commonwealth Edison Company Units 2 and 3 cc:

Michael I. Miller. Esq.

Michlel Sidley Ind Austin Hltional Pllza One First Hational Chtcego.*Illtnois 60603 Ch1C1go,*Jll1no1s

~.

~. Eeli.enburg

_. Eelllenburg Plant Superintendent Nuclear Power Stlt10n Dresden Nucl'lr Statlon Ru ral Route 11 Rural II

':"Norr15,

'fIorrls. 1111no1s Il11no15 60450 Commisslon U. S. Nuclear Rfgulatory Commission Resident Inspectors Office Dresden Station Rur.l RoutE:

Rural Ruute 11 til Morris. Illinois 60450 Horris, Chalnnan Chatman Board of Supervisors of Grundy County Grundy County Courthouse Grund~ C~unty Il11nois 60450 Morri~. Illinois "orrf~,

c. Regional Administrator Regulatory Nuclear Regu Commission.

1iltory Comm1 Roosevelt Road, Bldg. 14 799 Roost!velt Regitlll ss lon, Reg it/(! IIII II Illinois 60137 Glen Ellyn, Illino;s Michael E. Parker, Chief Mr. Michie' DiVIsion of En9in~er1ng D1~ls1on Engin~ering I11irlois III Departl:1t!lIt of Nuclear Safety irlois Departr:lt!llt Cuter Park Driv~. 5th Floor 1035 Outer Springfleld, Il1in01s 62704 Springfield. Illinois

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I .2-2

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. O. C.

C, 20555 Revision 8 April 1992

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(-

COMMONWEALTH EDISON COMPANY DOCKET ~O. 50-237 50*237 DRESDEN NUCLEAR POWER STATION, UNIT NO. 2 AMENDHEHT TO PROVISIONAL OPERATING LICENSE AMENmtEHT Amendmellt No. 106 Amendme"t license L1cense No. OPR*19 OPR-19

l. The Nuclear fl.t:glJlatory COIIIII;551011 (the Rt'gulatory Corrrnissfoll Connission) hIS four.d that:

(tile Conn1ssion)

A. The application for amendment by the Commonwealth Commollwealth £d1~on Edi~on Compar~

eompal~

(the l1ccnsee) licensee) dated February 22, ~2, 1989 c~pl1es w1th th~

complies with the standards requirements of the AtomiC and requfremeflts Atom1c Energy Act of 1954, 6S 'IIIend~d u .mended (the Act). anQ the C<H\'II\issicm's Act), ana regulatlolls set forth COIIIIlissicm's rules and regulat10rls in 10 CFR Chapter Ij I:

B. The facility will op~rate operate in conformlty conformIty with the application,

~ ... the provisions of the Act alld a~d the rules and regulations of the COlMliss1onj Commission; C. Th~re 1s ThHt is reasonable assurance (i) (1) that th~ thli activities author1zea authorlzea

. by thth11s arner.dment cal, b~

amendment CIrJ btl conducteo conducted t/i thoLit enc.inger1ng ttithout enc.angering thE' hti:llth htillth iwd sahty cHiO of the public.

safety (If public, ilid6lid (11) th~tthat such acthltles will be cc.nducted in co~p';ance cvnducted co~pliance withw1th the C~ission's regulations; regulations:

D. Tht issuance of thlS lmenQment Th~ is~uance amenclment will not be .inimical the comorl to ttle

.inimlcal tG COlTlllor, defer;sE:

defens~ and securl securi ty or to thr; thE: health Ina and nfety of thE:tht' pub 11 11c; cndcno E. The issuance iuuance of thisth1s IJnendmE:JIt alnl!ndmel,t ;s 15 in accordance with 10 CfR efR Part Sl 51 Commiss1on's regulations ano of the Commission's and .11all applicable requirements have been satisfied.

satisHed.

c I .2-3

Revision 8 April 1992

( 2. aJ:lended by changes to the Technical Accordingly, the license is amended Specifications as indicat~Q in the attachment to this license amendmenta~ndment and 3.H. of Provisional Operating License No. DPR-19 and paragraphs 3.B. anci DPR*19 ire are hereby amended to read IS as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A. A, as revised through Amendment No. 106, Ire are hereby incorporated licensee Shill in the license. The Hcensee shall operate the facility facility*.

Specifications.

in accordance with the Technical Specificatfons.

H. Company shaTl Commonwealth Edison Complny shall impleaent and ..

i~pleJent Ind maintain inta1n in effect

.11 all provisions of the approved fire protections program as described in the Updated final Analysis Report for the Final Safety Analysfs facility and as approved in the SERs dated Harch focility March 22.

22, 1978 with supplements dated December 2. 2, 1980 and February 12, 1981; January 19, 1983; July 17, 1987; September 28. 28, 1987.

1987; and January 5, 1989, subject to the follo~ing provisions:

1985, changes to the approved fire protection The licensee may make change~

program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and Inc maintain safe shutdown in the event of a fire.

3. This licer,se amendment is effective as of the date of its issuance to be irllplemented ililplemented within 60 days. .-

FOR THE ~UClEAR NUCLEAR REGULATORY COMMISSION COH~~ISSION

?~c.s~

Paul C. Shemanski, ShemanSki, Acting Director Project Directorate 111-2 Division of Reactor Projects - III. III, IV, V and Special Projects AttachltlE!nt:

Attachment:

Changes to the Technical Specifications Date of Issuance: June 30, 1989

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,~--

1.2-4

Revision B ATTACHNENT TO LICENSE AMENDMENT NO. 106 April 1992 PROVISIONAL PReVl OPERATING LICENSE DPR-19 SIONAL OPERAlING DOCKET NO. 50-237 Revise the Appendix A Technical Specifications by removing the pages identified below Ind and inserting the attached pages. The revised pages Ire are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Irea REMOVE INSERT iv 1v iv 3/4.12-1 3/4.12-2 3/4.12-3 3/4.12-4 3/4.12-5 3/4.12-6 3/4.12-7 3/4.12-B 3/4.12-9 3/4.12-10 3/4.12-11 3/4.12-12 3/4.12-13 3/4.12-14 B 3/4.12-15 B 3/4.12-16 B 3/4.12-17 B 3/4.12-18 B 3/4.12-19 B 3/4.12-20 B 3/4.12-£:1 3/4.12-<1 6-1 6-1 6-7 6-5 6-13 6-11

(

"'-- I .2-5

Revision 8 April 1992 DRESDEN II DPR-19 OPR-19 Amendinent Amendment No. 82,83, 82 "83 106 t t (Table of Contents.

Contents, tont'd.)

Cont'd.)

Page 3.9.C Diesel Fuel 3/4.9 - 4 3.9.0 Diesel Generator Operability 3/4.9 - 4 limiting Conditions for Operation Bases (1.9)

Limiting (3.9) B 3/4.9 - 7 B

Surveillance Requirement Bases (4.9) B 3/4.9 - a8 B

3.10 Refueling 3/4.10- 1 3.10.A Refueling Interlocks 3/4.10- 1 3.10.B 3.10.8 Core Monitoring 3/4.10- 1 3.10.C Fuel Storage Pool Water Level level 3/4.10- 2 1/4.10-3.10.0 Control Rod and Control Rod Drive Maintenance 3/4.10- 3 3.10.E Extended Core Maintenance 3/4.10- 4 3.10.F Spent Fuel Cask Handling 3/4.10- 5 limiting Conditions for Operation Bases (3.10)

Limiting B 3/4.10- 8 Surveillance Requirement Bases (4.10) B 3/4.10-11 3.11 High Energy Piping Integrity (outside containment) 3/4.11- 1 (3.11)

Limiting Conditions for Operation Bases (3.l1) limiting B 3/4.11- 4 B

Surveillance Requirement Bases (4.11) B 3/4.11- 4 3.12 Fire Protection Systems - Sections 3.12.A through Letters 86-10 3.12.H - Deleted per Generic letters B6-10 and 88-12 (Amendment 106)

C"

( 4.0 SURVEILLANCE RE UIREMENTS 4.1 eactor rotectlon ystem 3/4.1 - 1 3/4.1-1 4.2 Protective Instrumentation 3/4.2 - 1 4.2.A Primary Containment Isolation Functions 3/4.2 - 1 4.2.B Core and Containment Cooling Systems --

Initiation and Control 3/4.2 - 1 4.2.C Control Rod Block Actuation 3/4.2 - 2 4.2.D Refueling Floor Radiation Monitors 3/4.2 - 2 4.2. E 4.2.E Post Accident Instrumentation 3/4.2 - 3 4.2.F Radioactive Liquid Effluent Instrumentation 3/4.2 - 4 4.2.G Radioactive Gaseous Effluent Instrumentation 3/4.2 - 5 4.3 Reactivity Control 3/4.3 - 1 4.3.A Limitations Reactivity limitations 3/4.3 ..- 1 4.3.B Contro 1 Rods Control 3/4.3 - 4 4.3.C Scram Insertion Times" Times 3/4.3 -10

v iv AMENDMENT NO. 106 I .2-6

Revision 8 April 1992 DRESDEN II DPR-19

(,( '

Amendment No. 82,86,97.105,106 82,86,97,105,106 6.0 ADMINISTRATIVE CONTROLS Organization, Review, Investigation and Audit-6.1 Organization.

A. shall-be established for the unit Onsite and offsite organizations shall'be management, respectively. The onsite and operation and corporate management.

offsite organizations shall include, include. the positions for Ictivities affecti ng the safety of the nuclear power pllnt.

affecting plant. _

'. ,_.~._ "'-::?,'\.,

~ .~~~:( .,,: ,,>-:".. '.,-'

J. * ., *

1. lines of luthority.

Unes luthority, responsfbf1itY~-lnd responsibility; and cOMunication co-unicltion shall be established and defined for the highest ..nlgelent levels intermedilte levels to .nd through the intermediate end including 111 operating operlting organization positions. These relationships shall be docuaented docu.ented and updated, as appropriate, in the form of organization charts, functional descriptions of depart.ent responsibilities and Ind relationships, and job descriptions for key personnel positions, or in the equivalent forms of documentation. The requirements shall be documented in the Quality Assurance Manual or the Management Plan for Nuclear Operations, Section 3 Organiza-tional Authority, Authority. Activity; Section 6 Interdepartmental Relationships.

2. The Station Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of plant.

3. The Senior Vice President-Nuclear Operations shall have the corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, operating. maintaining, maintaining. and providing teChnical support to the plant to ensure nuclear safety.

4. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however. they shall have sufficient organizational freedom to ensure their independence from operational pressures.

B.

8. DELETED C. The shift manning for the station shall be as shown in Table 6.1.1.

The Operating Assistant Superintendent, Operating Engineers, Engineers. Shift Engineers. and Shift Foremen shall have a Senior Operating license. License.

The Fuel Handling Foreman shall have a limited Senior Operating License. The Vice President BWR Operations, license. Operations on the corporate level has responsibility for the Fire Protection Program. An Operating Engineer at the station will be responsible for implementation of the Fire Protection Program.

Fir~ .,-,I

( AMENDMENT NO. 106 6-1

1. 2 - 7 1.2-7

Revision 8 April 1992 DRESDEN II ]PR-19

.DPR-19

(, Amendment No. 82,86,97,105,106 6.0 ADMINISTRATIVE CONTROLS (Cont'd.)

(5)

(S) Noncompliance with NRC requirements, or of internal procedures or instructions having nucleir safety significance. -

(6) Significant operating*lbnor.alities operating* abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety as referred to it by the On-site Review and Investigative Function.

(7) Reportable events under 10,CFR 10.CFR 50.73.

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

structures, systems or components.

(9) Review and report findings and recommendations regarding all changes to the Generating Stations Emergency Plan prior to implementation of such changes.

(10) Review and report findings and recommendations

( regarding all items referred by the Technical Staff Supervisor, Station Manager, Vice President BWR Operations and AVP Qual ity Programs -and and Assessment.

(11) Review changes to the Fire Protection Program and implementing procedures.

b. Station Audit Function The Station Audit Function shall be the responsibility of the AVP Quality Programs and Assessment independent of BWR Operations. Such responsibility is delegated to the Nuclear Quality Programs Manager.

Either of the above, or designated Corporate Staff or Supervisor approved by AVP Quality Programs and Assessment shall approve the audit agenda and checklists, the findings and the report of each audit. Audits shall be performed in accordance with the Company Quality Assurance Program and Procedures. Audits shall be performed to assure that safety-related functions are covered within the period designated below:

(1) Audit of the Conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per year.

6-5 AMENDMENT NO. 106 1.2-8

Revision 8 April 1992 DRESDEN II DPR-19 Amendment No. 82,83,86,97,105,106 6.0 ADMINISTRATIVE CONTROLS (Contld.)

(Cont'd.)

(7) Pertor.ance Perfor.ance of speci.l special reviews and investigations and Ind reports thereon as IS requested by the Superintendent of the Off-site Revi~,.nd Revi~,and Investigative Function.

(8) Review of the Station Security Plan and SecurityPlln Ind shall submit recommended changes to the Director of Corporlte Security and the AVP Quality Programs and Corporate Ind Assessment in lieu of distribution in accordance Iccordance with 6.1.G.2.t.(1).

6.1.G.2.c.(1).

(9) Review of the Emergency PllnPlan and Ind station implementing procedures and identific~tion identification of ~ecommended recommended changes.

(10) Review of reportable events and Ind actions taken to prevent recurrence.

(11) Review of any Iny unplanned on-site release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation recommendations and Ind disposition of the corrective action to prevent recurrence to the Vice President BWR Operations and to the Superintendent of" of the Off*site Off-site Review and Ind Investigative Function.

(12) Review of changes to the PCP and ODeMDDCM and major changes to the radwaste treatment systems.

(13) Review changes to the Fire Protection Program.and implementing procedures.

b. Authority The Technical Staff Supervisor is responsible to the Station Manager (or designee) and Ind shall maKe make recommenda-tions in a timely.manner in allIII areas Ireas of review, investiga-tion, and Ind quality control phases of plant maintenance.

maintenance, operation and administrative procedures relating rellting to facility operations. The Technical Staff Supervisor shall have the authority to request the action necessary to ensure compliance with rules, regulations, and procedures when in his opinion such action is necessary. The Station Manager (or designee) shall follow such recommendations or select a course of action Iction that is more conservative regarding safe operation of the facility. All such disagreements shall be reported immediately to the Vice President BWR Operations and the Superintendent of the Off-site Review and Investigative Function.

( 6-11 AMENDMENT NO. 106 I .2-9 1.2-9

UNITED STATES NUCLEAR REGULATORY COMMISSION Revision 8B W"~HINGTON, D. C. 20555 WA:;HINGTON. April 1992

('

COMMONWEALTH EDISON COMPANY . '..

. ~. .

DOCKET NO. 50-2495D-249

.~.

DRESDEN NUCLEAP POWER STATloH.UNIT' STATloH.UNIT ** 110. 3 AMErm~lEHT AHEIlD~lENT TO FACILITY OPERATUIG OPERATING LICENSE Arcn_tit '.0.

AII".endllent r.o. 101 Licenst No. DPR-25 Llcens~

1. Th~

Tht Commission (the Commission) has found thlt:

Nuclear Regulatory Commlssion that:

A. The application app11cetion for l~ndment ComIOnwealth Edfson Cor,plny a~ndment by th~ CoaDOnwealth (the licensee) Gated dated February 22, 22. 1989 complies with the standards stlndards illd rt!qufrements of the AtODlic arid Atomic Erlergy Er,ergy Act of 1954, as amended (the Act),

Act). and the Commission Commission's ' s ~les nules and regulations set forth Chapter' 11;i in 10 CFR Chapt~r' r--

(-- B. The fati 11ty will facf1ity .conformfty wi wfll operate ir, ,conformfty th the application, with appl1cation, (C_ the provisions of the Act glid Commissfor,;

Commfssior,;

IIlid the rules and regu16tions regulations of the c.

C. There is reasonable assurance (i) (1) that the activities iuthorized iuthorfzed by tf'lis Uf s amendmerlt amendment can be conducted without endangering the ht=alth hti lth Ind and safety of the public, pub 11c, and (11) that such activfties activities wi 11 be will conducteo in compliance with the Co~iss1o~'S Coanission's regulltions; regulations; D.

O. The issuance of this ancnQrrent ancndnent will not be 1n1~ic.l ini~ic.1 to the common co~n defellse ano defense lllci security or to the hulth htalth and sifety safety of the public; publici and E. The issuance of this amendment is in accordance with 10 CFR Part 51 Commission's of the Commission ' s regulations and .11 app1iclble requirements have 111 applicable bt!tn sat1sfied.

b~~n satisfied.

2- 10 I .2-10 1.

Revision 8 April 1992

2. Accordingly, the license is amended by changes to the Technical alendlent Specifications as indicated in the attachment to this license amend.ent Facility Operating license No. DPR-25 are and paragraphs 3.B. and 3.G. of Facili~

hereby amended to read as follows:

B. Technical Specifications contained in Appendix A, IS The Technical Specifications contained. IS revised through Amendment Ho.

No. 101

• Ire hereby incorporated incorporlted licensee shall operate in the license. The lfcensee operlte the facflity flcility SpecifiCitions.

in accordance with the Technical Specifications.

S.

6. ilplement and Commonwealth Edison Company shall i~lement Ind .. intain fntain in effect provisions of the approved-fire all proviSions approved'fire protections prograM IS as AnalYSis Report for the described in the Updated Final Safety Analysis as approved 1n facility and IS in the SERs dated March 22, 1978 with supplements dated December 2, 1980 and February 12, 1981; 19, 1983; July 17, 1987; September 28, 1987. and January 5, January 19.

1989, subject to the following provisions:

1989.

The licensee may make changes to the approved fire protection pr09ram without prior approval of the C~1ssion program Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

c. 3. This license amendment is effective as of the date of its issuance to be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION ry~ C . S:~"~jO~..:

Paul C. Shemanski, Acting Director Project Directorate 111-2 Division of Reactor Projects - III, IV.

IV, V and Special Specfal Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: June 30, 1989

/"'

~.--

~ ...

I .2-11

B Revision 8 April 1992 ATTACHMENT TO LICENSE AAENDMENT 110.101 rm.l01 (c_

FACILITY OPERATING LICENSE DPR-25

(

DOCKET NO. 50-249 Revise the Appendix A Technical Specifications by removing-removing the pages identified below aDd arId inserting the attached pages. The revised pages are identified by the captioned amendment number IIIrginal ltnes nunter and contain marginal lines indicating the 6 rea of cha nge

• REMOVE INSERT tv tv 3/4.12-1 3/4.12-2 3/4.12-3 3/4.12-4 3/4.12-5 3/4.12-6 3/4.12-7 3/4.12-8 3/4.12-9 3/4.12-10

( 3/4.12-11 3/4.12-12

( 3/4.12-13 3/4.12-14 B 3/4.12-15 B

B 3/4.12-16 B 3/4.12-17 B 3/4.12-18 B 3/4.12-19 B 3/4.12-20 B 3/4.12-21 6-1 6-1 6-7 66-5

.. 5 6-13 6-11 lc-I .2-12

Revision 8 April 1992 III DRESDEN 111 DPR-25 75~77~ 101 Amendment No. 75,77, tont'd.)

(Table of Contents, Cont'd.)

Page Paee 3.9.C Fuel Diesel fuel 3/4.9-4 3.9.0 3.9.D Diesel Generator Operability ., 3/4.9-4 limiting Conditions for Operation Bases (3.9) B 3/4.9-7 S

Surveillance Requirement Bases (4.9) B 3/4.9-8 B

3.10 Refueling 3/4.10-1 3.10.A Refueling Interlocks 3/4.10-1 3.10.B Core Monitoring 3/4.10-1 3.10.C Fuel Storage Pool Water level fuel 3/4.10-2 3.10.0 Orive Maintenance Control Rod and Control Rod Drive 3/4.10-3 3.10.E Extended Core Maintenance 3/4.10-4, 3/4.10-4.

3.10.F Spent Fuel Cask Handling 3/4.10-5 limiting Conditions for Operation Bases (3.10)

Limiting B 3/4.10-8 Surveillance Requirement Bases (4.10) B 3/4.10-11 S

3.11 High Energy Piping Integrity (outside containment) 3/4.11-1 limiting Conditions for Operation Bases (3.11)

Limiting B 3/4.11-4 Surveillance Requirement Bases (4.11) B 3/4.11-4 3.12 Fire Protection Systems - Sections 3.12.A through 3.12.H -

Letters 86-10 and 88-12 (Amendment 101)

Deleted per Generic letters

/- 4.0 SURVEILLANCE REOUIREMENTS

" 4.1 Reactor Protectl0n System 3/4.1-1

(,( 4.2 Protective Instrumentation 3/4.2-1 4.2.A Primary Containment Isolation Functions 3/4.2-1 4.2.B Core and Containment Cooling Systems --

Initiation and Control 3/4.2-1 4.2.C Control Rod Block Actuation 3/4.2-2 4.2.0 Refueling Floor Radiation Monitors 3/4.2-2 4.Z.E 4.2.E Post Accident Instrumentation 3/4.2-3 4.Z.F 4.2.F Liquid Effluent Instrumentation Radioactive liquid 3/4.2-4 4.2.G Radioactive Gaseous Effluent Instrumentation 3/4.2-5 4.3 Reactivity Control 3/4.3-1 4.3.A Reactivity Limitations 3/4.3-1 4.3.B Control Rods 3/4.3-4 4.3.C Scram Insertion Times 3/4.3-10 iv AMENDMENT NO. 101

(

I .2-13 1.2-13

Revision 8 April 1992 DRESDEN III OPR-25 Amendment No. 75,79 92,100,101 75,79,92,100,101 J

6.0 ADMINISTRATIVE CONTROLS 6.1 Organization. Review, Investigation and Audit A. Onsite and offsite organizations shall be established for the unit operation and corporaie corporate aanagement,

.anagement, respectively. The onsite and offsite organizations shall include.the positions for activities off51te affecting the safety of the nuclear power .plant.,-", .plant *. ~ ..

..." ***<:*:<'~i::'l;**.

• -*<:":<c'h'i::";~***

1. lines responsibility, 'and Lines of authority, responsibility. 'and co.unication ca.unication shall be established and defined for the highest aanagement levels inte!"lllediate levels to and including all operating through the inteJ'lllediate organization positions. These relationships shall shill be documented and updated, as appropriate, in the form or organization charts, functional descriptions of department departlent responsibilities and relationships, and job descriptions for key personnel positions, forms of documentation. The requirements or in the equivalent fonms shall be documented in the Quality Assurance Manual or the Management Plan for Nuclear Operations, Section 3 Organizational Authority, Activity; Section 6 Interdepartmental Relationships.

2. The Station Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of plant.

( 3. The Senior Vice President-Nuclear Operations shall have the corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

4. The individuals who train the operating staff and those who carry out health physics and qua1ity quality assurance functions may report to the appropriate onsite manager; however, however. they shall have sufficient organizational freedom to ensure their independence from opera-tional pressures.

B.

8. DELETED c.C. The shift ~anning manning for the station shall be as shown in Table 6.1.1. The Operating Assistant Superintendent, Operating Engineers, Shift Engineers, and Shift Foremen shall have a Senior Operating license. The Fuel Handling Foreman shall have a limited Senior Operating License. The Vice President BWR Operations on the corporate level has responsibility for the Fire Protection Program. An Operating Engineer at It the station stltion will be responsible for implementation of the Fire Protection Program.

( 6-1 Amendment No. 101 I .2-14

Revision 8 April 1992 DRESDEN III DPR-25 Amendment No. 75,79,92,100,101 6.0 ADMINlSTRATIVE ADMINISTRATIVE CONTROLS (Cont'd.)

(4) Proposed changes in Technical Specifications or NRC operating licenses. . :

(5) Noncompliance with NRC requirellents, requfrellents, or of inter-nal procedures or instructions having nuclear significance; safety significance~ .

(6) Significant operating abnor.alities abnon.alities or deviations' from normal and expected perfor.ance perfoMlince of plant equipment that affect nuclear safety IS as referred to it by the Onsite Review and ind Investigative Function..

Function., .' ..

(7) Reportable Events reported under 10 CFReFR 50.73.

(8)

(S) All recognized indications of Inan unanticiplted unanticipated deficiency in some aspect of design or operation of safety related structures, systems or components.

(9) Review and report findings and recommendations regarding all changes to the Generating Stations

/ Emergency Plan prior to implementation of such changes.

C, C. (10) Review and report findings and recommendations regarding all items referred by the Technical Staff Supervisor, Station Manager, Vice President BWR 8WR Operations and AVP Quality Program and .

Assessment.

(11) Review changes to the Fire Protection Program and implementing procedures.

b. Station Audit Function The Station Audit Function sha" shall be the responsibility of the AVP Quality Programs and Assessment independent of the Production Department. Such responsibility is delegated to the Nuclear Quality Programs Manager.

Hanager.

Either of the above, or designated Corporate Staff or Super-vision approved by AVP Quality Programs and Assessment sballshall approve the audit agenda and checklists, tbethe findings and the report of each audit. Audits shall be performed in accordance with the Company Quality Assurance Program and Procedures.

Audits shall be performed to assure that safety-related functions are covered within the period designated below.

6-5 Amendment No. 101 1.2-15

Revision 8 April 1992 DRESDEN III DPR-25 r(

\

\.

Amendment No. 75,77,79,92,100,101 (Cont'd.) ..

6.0 ADMINISTRATIVE CONTROLS CCont'd.)

(6) Review of facility operations to detect potential safety hazards.

(7) Perfo~ance Perforaance of special reviews and investigations and, and* reports thereon as requested by the Super-Off-site* Review and Investigative intendent of the Off-lite*

Function.

(8) Review the Station Security Plan and shall submit recommended changes to the Director of Corporate Security and the AVP Quality prograas Prograas and A5sess~nt Assesllent in lieu Of0.1 distribution in accordance with 6.1.G.2.cCl}.

6.l.G.2.c(l).

(9) Review the Emergency Plan and "station station ilplementing changes.

procedures and identification of recommended chlnges.

(10) Review of reportable events and actions taken to prevent recurrence.

(11) Review of any unplanned on-site release of radioactive material to the environs including the preparation and fowarding of reports covering evaluation recommendations and

(( disposition of the corrective action to prevent recurrence to the Vice President BWR Operations and to the Superintendent of the Off-site Review and Investigative Function.

(12) Review of changes to the PCP and aDCMODeM and major changes to the radwaste treatment systems. .

(13) Review changes to the Fire Protection Program and implementing procedures.

b. Author; ty Authority The Technical Staff Supervisor is responsible to the Station Manager (or designee) and shall make recom* recom-mendations in a timely aanner in all areas of review.review, investigation, and quality control phases of plant maintenance, operation and administrative procedures relating to facility operations. The Technical Staff Supervisor shall have the authority to request the action necessary to ensure compliance with rules, regulations, and procedures when in his opinion such luch action is necessary. The Station Manager (or designee) shall follow such recommendations or select a course of action that is more conservative regarding safe operation of the facility. All such disagreements shall be reported immediately to the Vice President BWR Operations and the

(( Superintendent of the Off-site Review and Investigative Function.

6-11 Amendment No. 101 I .2-16

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. O. C. 20555 Revision 8 April 1992

(

SAFETY EVAlUATION EVALUATION BY 6Y THE OFFICE Of OF ~UCLEAR REACTOR REGULATION SUPPORTIIIG SUPPORTING AMEHDMHT AMENDHEtiT NO. 106 TO PR(lVISIONAL.

IDS TO* PIII!VISIONAL OPERATING UCEI4SE LICEIISE HO.

NO. OPR-19 DPR-19 AND AMENDMENT 110.101 HO.IOI TO FACILITY OPERATInG OPERATING LICENSE NO. DPR-25 COf4MONWEAL COIoIMOHWEAl TH EDISON *COMPANY COMPANY DRESDEN NUCLEAR POWER STATION, STATION. UNIT NOS. 2 AND 3 DOCKET NOS. 50-237 AND 50-249

1.0 INTRODUCTION

F~bruary 22, 1989, COII~nwealth By letter dated Fibruary Ey COID~nwealth Edison Co~plnyCo~pany (th~

(the litensee) licensee) proposed that the f:xisting cond1tions on txisting license conditions protect1on be ,,-plated 011 fire pl"Otecti()n replued with the standard condition conoition noted fn Generic Letter 86-10 and 11so also pr~posed chonges tu to the Appendfx A Technical Specfffcations (TS) tor Tor Dresden Units 2 and 3. The proposed changes wouldr~ve would r~ve requirements require~nts for flre fire detection detectfon suppression systems,

~stems, fire supprtssfon systems, syste~s, fire barriers, and ffre brfglGt brigaat $taffing itaffing requirtllllents as recommended r£qutr~nts recOlllllended by Generfc Letter 86-10. The proposed changes would .1so mod1fy the aominl~trMtive

.150 modify requir"ments of n,.

admini~tr~tive control r.qu1r~ments the TS to add requirements for the F1~ Fire Protection Protectfon Progr~ that Ir! are s1mflar simflar to requirements requir_nts implementea by 1licellse for other programs fmplementea icelose conditfon. Guid.nce Guidance on these change$ to TS was prov1a~a proposed th4nges proviaea to 111 all power po~er reactor iicensee and epplfcants by G~ner1' eppl1cants Generi, L~tter 88-12, dat~d dated August Z, 2, 1988.

2.0 BACKGROUND

Following the fire at the Browns Ferry Nuclt:u Nucltlf' Power Plant on March Mlrch 22. 1975, 22, 1975.

COliJIlissiol, undertook aI number of actions to .nsurt the COliJllissioJI illprovellll!nts wert:

ensurt that improvements 1~1~nted t~lemtnted 1n Fire Prot~Ctl0n in the Ffre ProteCtlOn Prugrbms for .11 all powEr power r~lctor facilities.

r~actor facilftitS.

Be'iuse of the ext~nsive BetiUSe modificatfon of Fire Protection Progrlms exttnsive mOdificatfon Programs and the nUlDer of open 1ssuts number issues resulting frOM fr~ stiff ~vlluat10ns, evaluations. Ia number of revisions Ind and alterat10ns alterations occurred in fn these programs over the years. Consequently, l1cense~s lfcensees were requested by Gen~rjc Gen~ric Letter 86*10 tncorporat~ the ffnll 86-10 to 1ncorporlt~ final NRC Ffre Protettiun approved Fire Protection Progra~ in their Final Safety Safe~ Analysis Reports (FSARs). In this mann~r.manner, the Ffre Protection Protectton Program -- tncluding tncludlng the

$Ystt~S,

~sttms. the Id~fnistrltive controls, the organizot10n.

technfcal control~~

ad~inlstrative and technical organfzotfon, and plant featurts other pl.nt features associated wfth with fire protection -- would ha~e have a stltus status consisttnt with th~t of other plant features descrtbed cons1stknt described intn the FSAR. In addftion, Itddftfon. the Coumfss ion concluded that I.a standird Coumfssion standard Heense Hcense eonclit1on, COlldit1on, requiring eor.pliance re~u1rlng eor.plilnce with wfth the provi.ions provisions of the Ffrt Ffre Protectfon Protection Progrl~

Program ISas descrfbeo in the FSAR, should be used to e'lsure descr1beci uniform ~riforc:ement ellsure unfform ~I;forcement of fire protection requlr~nents.

protectfon requir~~nts. Finally.

Fin~lly, the C~~fssion Co.~isslon stated thit that Wlth wlth the requested .ct1ons.

requ~sted 4ctioos, l1cens~es licenSees may request In an lmendu~nt amendu~nt to del~tE the fire protectiun protectiull TS that would now br be unnttcessar:y.

unnt'cessaty.

(

1.2-17

Revision 8 April 1992

~, The licensees for the Callaway and Wolf Creek plants submitted ledd-plant lead-plant proposa1s proposals to remove fire protection requirements from their TS. This action was an industry effort to obtain NRC guidance 9uidance on an acceptable fo~t format for license amendment requests to remove fire protection requtrrequir...nts from 1S.

TS.

Additionally, in the licensing review of new' Additionally. plant~.'thl new plants. 'the staff has approved applicant requests to remove fire protection requirements from TS issued with the operating license. Thus, on the basis of the lead-plant proposals and the staff's experience w1thwith TS for new licenses.

licenses, Generic letter Letter 88-12 was issued to provide guidance on removing fire protection req~trrequir...nts from TS.

3.0 EVALUATION Generic letter Letter 86-10 recommended the removal of fire protection requirements from the TS. Although a comprehensive Fire Protection Program is essential to plant safety, the basis for this recommendat10n recommendation is*that ~any details of this program that are currently addressed in TS can be modified without .ffecting affecting nuclear safety. Such modifications can be made provided that there Ire are suitlble suitable administrative controls over these changes. These details, that are presently included in TS and which are removed by' this amendment.

amendment, do not constitute performance requirements necessary to ensure safe operation of the facility and, therefore, do not warrant being 1ncluded included in TS. At the slme same tillll!, suitable administrative controls ensure that there will be careful t1~.

review and analysis by competent individuals of any changes 1n in the Fire Protection Program including those technical and administrative requirements removed from the TS to ensure that nuclear safety is not adversely affected.

( These controls include: (1) the TS administrative controls that are applicable to the Fire Protection Program, (2) the license condition on

.i~ lementation of, and subsequent changes to, the Fire Protection Program, and i~lementation (3}

(3) the 10 CFR 50.59 criteria for .evaluating changes to the Fire Protection Program as described in the FSAR.

The specific details relating to fire protection requirements removed from 1S TS by this amenament include those specifications for fire detection systems, fire suppression suppressioll systems, fire barriers, and fire brigade staffing requirements. The administrative control requirements have been modified to include Fire Protection Program implementation as an element for which written procedures must be established.

established, implemented, and maintained. In addition, the audit responsibilities of the On-Site Review and Investigative Function were expanded to include the review of the Fire Protection Program and implementing procedures and submittal of recommended changes to the Off-Site Review and Investigative Function.

The TS changes proposed by the licensee are in accordance with the' the guidance provided by Generic Letter 88-12.

88-12, as addressed in the items below.

(1) Specification 6.1.G.2.a.13.

6.1.G.2.a.13, On-Site Review and Investigative Function.

Function, was revised to add the review of the fire protection program implementation dnd Specification 6.1.G.l.a 6.1.G.1.a was revised to include the review of recommended Changes chAnges by the Off-Site Review and Investigative Function.

I .2-18

Revision 8 April 1992 6.1.G.l.a(11). Fire Prct~ct;on (2) With the inclusion of Specification 6.1.G.1.a(11), Prct~ction Pr~gram implementation has been added to those programs for which writteh writte~ procedures are required. Specification 6.1.H.

Ire 6.1.H, which was approved in Ia previous amendment, amendDent.

contains an 'nspection inspection and audit requireMent for the Fire Protection ProgrUl.

Progrllil.

(3) Specifications 3.12 and 4.12,Ffre 4.12,Fire Suppression SysteltS, S1stellS, their .associated associated Surveillance Requirements, Requirements. and 8ases Bases (including Fire Barriers and Fire InstrUJnentation, and their associated Surveillance Requfrements Detectiol, lnstrwnentation.

Detectior. Requirements and Sises)

Blses) were removed.

rewoved.

(4) Specification 6.2.C on fire brigade staffing requfreMents requireMents was relOved.

relaved.

As required by Gellerfc letter 86-10, the licensee by letter dated dune Gelleric Letter June 20, 1989 confirR~d confirv~d that the N~C NPC .pprov~d Fire Protection Program has been incorporated approved fire into the fSAR FSAR and stated that a~ other references determined to be appropriate would be included in the 1989 calendar year fSAR FSAR update. Also, the licensee has hiS proposed that the existing licenSing conditions en pruposed on the f1re Fire Protectiorl Protectioll Program ProgT'11I be rep laced with the standard condition replaced in Generic letter 86-10.

noted.in c~r.dition noted.

The licensee confirmed that the operational conditions, r~dial remedial Ict10ns.

actions, and test requir~ments requirements associated with the firt fire protection TS will be approved pr10r prior to removal from the 1S TS and within 60 d~s from the approval date of this saftty safety evaluation Indand included in the Fire Protection Progrlll Progrilll fncorporated incorporated into the UFSAR. This Satisfies the guidance of Generic Letter letter 88-12.

en the ba3is On bi;is of its review of the above 1tems, items, the staff concludes that the

( licensee has met the guidance of Generic Letter lice.lsee letter 88-12. Therefore.

Therefore, the staff finds the proposed thanges fino$ changes acceptable.

4.0 ENVIRONME~TAl ENVIRONME~TAL CONSIDERATION aRiendments iinvolve These iIIltndments changes to the use of the fac nvo lve charfges facii li Htyty components located within the r~stricted restricted area as defined in 10 CFR Part 20. The staff has detennined that the amendments 1m'o involve significant increase in the allOunts lve no sign1fiClnt no sigflificant aud ne sigl,ificant change in th~

the types of any effluents that NY be released and that there is no significant increase 1n off site, Ind in individual or cumulative occupational exposure. The staff has determined detennined that the consideration, Ind

~ndn~nts involve no significant hazards conSideration, and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 eFR eligibilit~ CFR 51.22{c)(9). Pursuant to 10 CFR Sl.22(b) 51.22(c)(9). 51.22(b) no environmental impact stateMent or environmental assessment need be prepared in connection with the issuance of these amendments. .

5.0 CONCLUSION

amendments involve no .

The Commission made proposed determinations that the amer.dments significant hazard~ consideration, which were published in the Federal Register (54 FR 13762) on or. April 5, 1989. The Conrnission COllll1ission consulttd with the

(

I .2-19

Revision 8 April 1992 received, and the state of State of Illinois. No public comments were rec~fvedt Illinois did not have any coanentS.

cOi£ents. ".

baSis of the considerations discussed above, the staff concludes that On the basfs (1) there is re.sonable and slf~ty reason.ble assurance that the health Ind safety of the public wtll will not be endangered by operation in the proposed .anner, 5Inner. (2) such activfties fn c~plfance with the Confttssfon's wfll be conducted in activftfes wiTl Connfssfon's regulltions, regulations, and (3) the fssulnc~ ~n inimical to the COImOn fssuanc~ of these amendments will hOt be fnimical s.fe~ of the public.

dlflnse and security or to the health and $Ife~

d1f.nse Prfncipal Prfncfpal Contrfbutors: Kubicki, SPlB/DEST Dennfs J. Kubfckf, COntributors: Dennis SPLB/DEST Thomas G. Dunning, OTSB/DOEA Byron L. Stegel, S1egel, NRRlDRSP tiRR/DRSP Dlted: June 30, 3D, 1989

(

("

(

\.

I .2-20

Tab 3 UNITED STATES Revision 8 NUCLEAR REGULATORY COMMISSION April 1992 D. "".20555 WASHINGTON, O. r.20555 August 9, 9. 1989 AUG -'.14 4 &is

$3

(

DocKet Nos.

DOcKet 50-237 and 50-249

. Mr. Thomas J. Kovach Nuclear Lfcensing Licensing Manager Commonwealth Edison Company Post Off;ce Office Box 767 Chicago, Illinois 60690 Chicago.

SUBJECT:

INOEX PAGES FOR DRESDEN LINITS NO'S 2 AND 3 CORRECTION TO THE INDEX AI1ENDMENT NOS. 106 and 101 (TAC NOS. 71256 AND 71257)

FOR AHENDMENT The purpose of this letter is to notify you' you" that the index pages for these amendn~nts contain two errors. The pages identified as 6~7 6-7 and 6-13 to be removed should be changed to pages 6*56-5 and 6*11 6-11 as shown in the enclosed index pages. These changes are necessary to correct the original page numbers whfch which were superseded by a subsequent amendment.

( Sincerely Sincerely,t

~:~~<Ojett N.",,,

~::(,-&(j"t 111-2 Proj~ct Directorate" 111*2 M.n.~.r Division of Reactor Projects III, IV, V. V, and Special Projects

Enclosures:

As stated cc w/enclosures: See next page

(

1. 3- 1

Revision 8 April 1992 Mr. Thomas J. Kovach Dresden Nuclear Power Station COlM,onwealth Edison Compan,},

COITIII,onwealth Company Units 2 anc and 3

(

cc:

Michael I. Mfll~r.

Michlel Mill~r, Esq.

Sidley and Austin National Plaza

. One First N~tional Chicago, Illinois 60603 Mr. J. Eentgenburg Eenigenburg Plant Superintendent Dr~sden Nuclear Power Station Rural Route II Morris, Illinois 60450 U. S. Nuclear Regulatory Commi~sion COl11Tli~sion ..

Resident Inspectors Office DresdEm Station Dresd~n R('ute II Rural Pt'ute #1 Morris, Illinois 60450 Chairman Board of Supervis~rs of Grundy County Grundy County Courthouse Morris, Illinois 6045C

( Regional Administrator Regulatury Commission, Region III Nuclear Regulatory 799 Roosevelt Road, Bldg. #4 Glen Ellyn, Illinois 6013i Mr. Michael E. Parker, Chief Division of Engineering Illinois Department of Nuclear Safety I11in01s 1035 Outer Park Drive, 5th floorFloor Springfield, Illinois 62704 1.3-2

ATTACHr,;n.;T ATTACHME~T TO LICENSE AMENDMENT fiO.106 NO. 106 Revision 8B April 1992 PROVISIONAL OPERATING LICENSE DPR-19

( DOCKET NO. 50-237 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attachtd attach~d pages. The revised pages Ire tdentifted identified by the captioned amendmer,t amendment number and contain llargfnal IIIIrginal lines indfclttng indicating the area of change.

REMOVE INSERT tv iv tv 1v 3/4.12-1 3/4.12-2

,3/4.12-3 "3/4.12-3 3/4.12 3/4.12-4.. 4 3/4.12 3/4.12-5.. 5 3/4.12-6 3/4.12 3/4.12-7.. 7 3/4.12 3/4.12-8.. 8 3/4.12*9 3/4.12-9 3/4.12-lO 3/4.12-lCJ 3/4.12-11 3/4.12-12

( 3/4.12-13 3/4.12-14 B 3/4.12-15 B 3/4.12-16 3/4.12-17 Eib 3/~.12-17 B 3/4.12-18 B 3/4.12-19 B 3/4.1Z-20 S 3/4.12-20 E 3/4.12-21 E

6-1 6-1 6-5 6*5 6-5 6-11 6-11 C

(

I .3-3

~TTACHMENT TO LICENSE AfolENDMENT AfoIENDMENT NC,.

NC*. 101 Revision 8 April 1992

( FACIllTY FACILiTY OPERATING LICENSE DPR-25 DOCKET NO. 50-249 Revise the Appendix A Technical Specifications by removing the plges pages identified-identified.

below And inserting the attached pages. The revised pages are idtntffted and 1nsertinQ identfffed by the captioned amendment number -and *and contain marginal lines indicating fndicating the area of change. .

REMOVE INSERT i~ fv 3/4.12-1 3/4.1Z-1 3/4.12-2 3/4.12-3 3/4.12-4 3/4.12-5 3/4.12 3/4.12-6.. 6 3/4.12-7 3/4.12-8 3/4.12-9 3/4.12-10 3/4.12-11

( --.-

.--.,... 3/4.12-12 3/4.12-13 3/4.12-14 BB 3/4.12-15 B 3/4.12-16 E 3/4.12-17 B 3/4.12-18

£E 3/4.12-19 B

E 3/4.12-20 E 3/4.12-21 6-1 6-1 6-5 6-5 6-11 6-11 C ~/

I .

I .3-4

Tab II DRESDEN 2&3 AMENDMENT 11

,,( JUNE 1998

, '\"-'

Fire Protection Report Dresden Administrative Teclmical Technical Requirements (DATR's) for Fire Protection The Fire Protection and Safe Shutdown DATR's are available through Central Files.

/

(

l 11.1-1 II. 1-1

Tab III Revision 8 April 1992 f

c

( DRESDEN 2&3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE NRC Inspection Reports Ii!!.

IiQ Titl e 1 Hay 25, 1984 Inspection Reports No. 50-010/84-01, 50-237/84-06; 50-249/84-05 2 July 3, 1984 Inspection Report No. 50-010/84-09, 50-237/84-11; 50-249/84-10 July 30, 1984 CECo letter from D. L. Farrar to J. "J. G. Keppler (NRC);

Response to Inspection Report No. 50-010/84-09, 50-237/84-11, 50-249/84-10 3 November 14, 1985 Inspection Report No. 50-237/85033, 50-249/85029 December 26, 1985 Notice of Violation Concerning Inspection Reports No. 50-237/85033, 50-249/85029 January 24, 1986 CECo letter from D. t. l. Farrar to J. G. Keppler (NRC) transmitting response to Inspection Report No. 50-237/85033

( and 50-249/85029 4 February 26, 1986 Inspection Report. No. 50-249/86006 May 6, 1986 (CECo) letter from D. L. l. Farrar to J. G. Keppler (NRC) transmitting the response to Inspection Report 50-249/86006 .

July 17, 1986 (CECo) letter from D. l. Farrar to J. G. Keppler (NRC) discussing Inspection Report No. 50-249/86006 5 December 21, 1987 Inspection Report No. 50-237/87035 and 50-249/87034 6 December 14, 1987 Inspection Report No. 50-237/87037 and 50-249/87036 7 January 3, 1989 Inspection Report No. 50-237/88010 and 50-249/88012 to assess compliance with 10 eFRCFR 50 Appendix R Appendix R Audit Questions April 18-22, 1988 February 1, 1989 CECo letter from H. E. Bliss to A. Bert Davis transmitting the response to Inspection Report No. 50-237/88010 and 50-249/88012

.O-i III ,0-;

Revision 8 April 1992 DRESDEN 2&3

( FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE NRC Inspection Report (Cont/d)

(Cont'd) 8 January 23, 1989 Inspection Report No. 50-237/88030 and 50-249/88031 to review deficiencies in fire wrap installations and training to installers .

April 14, 1989 CECo letter from E. D. Eenigenberg to R. J. Israelson E-50 Fire Wrap Removable Covers (3M) on review of installed £-50 May 3, 1989 letter from R. J. Israelson (3M) to £. E. D. £enigenberg Eenigenberg (CECo), response to April 14, 1989 letter 9 February 28, 1989 Inspection Report 50-249/89004 concerning the June Drywell Expansion Gap 4, 1988 fire in the Drywel1 10 April 14, 1989 Inspection Report 50-237/89008 and 50-249/89009 reviewing allegations regarding unsealed openings inside conduits in wallss fire wall

( 11 II June 9, 1989 Inspection Report 50-237/89013 and 50-249/89012 to review implementation of the licensee's fire protection program 12 July 31, 1989 Inspection Report 50-010/89002, 50-237/89017 and 50-249/89016 13 December 26. 1989 Inspection Report No. 50-237/89022 and 50-249/89021 January 25,

25. 1990 CECo letter from T. J. Kovach to A. Bert Davis (NRC), Response to Notice of Violation and Inspection Report No. 50-237/89022 and 50-249/89021 14 24. 1990 Inspection Report No. 50-237/90017 and 50-249/90017 August 24, September 24,24. 1990 CECo letter from T. J. Kovach to A. Bert Davis (NRC). Response to Notice of Violation and Inspection Report No. 50-(NRC),

237/90017 and 50-249/90017 November 28,

28. 1990 NRC letter from A. Bert Davis to C. Reed (CECo) proposed Imposition of Civil Penalty

(

II1.0-ii III.O-ii , >

AMENDMENT 12 DRESDEN 2&3 FIRE PROTECTION REPORT NRC Inspection Reports (Contld)

(Cont'd)

Tab*

Tab Title 15 December 7, 1990 Inspection Report No. 50-237/90023 and 50-249/90003.

December 14, 1990 CECo letter from T.J. Kovach to A. Bert Davis (NRC) discussing unresolved Item 50-237/9002-06 in Inspection Report No. 50-237/90023 and 50-249/90023.

January 7, 1991 CECo letter form T.J. Kovach to A. Bert Davis (NRC), response to Notice of Violation contained in Inspection Report No. 50-237/90023 and 50-249/90023.

February 6, 1991 NRC letter from H.J. Miller to C. Reed (CECo) responding to CECOIS CECo's letter of January 7, 1991.

16 January 7, 1991 Inspection Report No. 50-237/90027 and 50-249/90026.

( February 15, 1991 CECo letter from T.J. Kovach and A. Bert Davis (NRC). Response to Notice of Violation Associated with with" Inspection Report No. 50-237/90027 and 50-237/90026.

17 March 15, 1991 Inspection Report No. 50-237/91004 and 50-249/91004 to review implementation of the routine fire protection program.

March 27, 1991 CECo letter from T.J. Kovach to A. Bert Davis (NRC). Response to Notice of Violation Associ.ated Associated with Inspection Report No. 50-237/91004 and 50-249/91004.

18 March 2, 1993 Inspection Report No. 50-237/93002 and 50-249/93002.

19 Nos. 50-010/96002, May 20, 1996 Inspection Report Nos'.

50-237/96002 and 50/249/96002.

20 November 14, 1996 Inspection Report Nos. 50-237/96012, 50-249/96012, 50-254/96016, and 50-265/96016 December 12, 1996 CornEd letter from E.S. Kraft to NRC, response to apparent viblation contained in Inspection Report Nos. 50-237/96012, 50-249/96012, 50-254/96016, and 50-265/96016.

III. o-iii III.O-iii

Amendment 16 DRESDEN 2&3

(

FIRE PROTECTION REPORT NRC Inspection Reports (Cant' d)

(Cont'd)

Title December 20, 1996 CornEd letter from 1. 1- B. Hosmer to NRC, supplemental Nos, 50-237/96012, response to apparent violation contained in Inspection Report Nos.

50-265/96016, 50-249/96012,50-254/96016 and 50-265/96016.

March 6, 1997 CornEd letter from J.J, B. Hosmer to NRC, regarding shorts, protection of motor operated valves during postulated hot shorts.

21 March 6, 1998 Inspection Report Nos. 50-237/97021 and 50-249/97021.

April 6, 1998 CornEd letter from J .M. Heffley to NRC, response to Notice of Violation contained in Inspection Report Nos. 50-237/97021 and 50-249/97021.

22 December 18, 1998 Inspection Report Nos. 50-237/98029 and 50-249/98029.

( 23 June 19,2002 Inspection Report Nos. 50-237/02-06(DRS) and 50-249102-50-249/02-06(DRS).

24 May 5, 2005 Inspection Report Nos. 0500023712005002(DRS) and 0500024912005002(DRS)

(

III.O-iv

Tab 1

.. UNITED STATES Revision 8 April 1992 NUCLEAR REGULATORY COMMISSION REGION III 799 R005EVEL ROOSEVELT T ROAD GLEN ELLVN, ELLVN. ILI..INOIS ILLINOIS 60U7 60137 MAY 2 5 1984 HAY 3 Ct(V

~.~Y J ( r'D Docket No. 50-10 Docket No. 50-237 Docket No. 50-249 Coamonwealth Edison Company Ca.monwealth ATTN: Mr. Cordell Reed Vice President Post Office Box 767 Chicago, IL JL 60690 Gentlemen:

This refers to the rout.iDe routi.!)e safety inspection conducted by T. M. Tongue, Tongue.

~. Stasek, C. D: ..Anc(e_rg)n-,-~~d D,.Ande..rgln .... ..!!1d R. lanksbury Lanksbury of thls this office on March 27. 1984 1;hrough

~hrough May 21, 1984, !Jf pf ~ctivities

~.ctivities at Dresden Nuclear Power Stat jon. Unjts 1.

_~. and and}..~ .authorized

_authorized by NRC Operating Dperating licenses Licenses No. DPR-02, DPR-19, and DPR-25.

DPR-25, and to the discussion of our findings with Mr. D. Scott and others of your staff at the conclusion of the in~pection.

The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel.

~~

~o items of noncompliance with NRC requirements w~re were identified during the course of this inspection. .

In accordance with 10 CFR 2. 790(a) a copy of thi 2.790(a). J this5 letter and the enclos'ure(s) enclosure(s) will be placeq placed in the NRC Public Document Room unless you notify this office, by telephone, within ten days of the date of this letter and submit written application to withhold information contained therein within thirty days of the date of this letter. Such application must be consistent with the re-quirements of 2.790(b)(1). If we do not hear from you in this regard within the specified periods noted above, a copy of this letter and the enclosed inspection report will be placed in the Public Document Room.

/'

(

III.]*]

III.1-1

Revision 8 April 1992 Commonwealth Edison Company 2 MAY 2 5 1~c4 l~o4 We will gladly discuss any questions you have concerning this inspection.

i~/,

~~cerel~

1//

~

J!:_~~~~fer,

. Shafer,

.~..,:jt~~~ .It Chief

/

./

Projects Branch 2

Enclosure:

Inspection Reports 50-010/84-01(DPRP);

No. 50-010/84-01{DPRP);

No. 50-237/84-06(DPRP);

No. SO-249/84-05(DPRP) 50-249/84-05(DPRP) cc w/encl:

D. l. Farrar, Director of Nuclear licensing D. J. Scott, Station Superintendent OMS/Document Control Desk (RIDS)

OMB/Document Resident Inspector, RIll Phyllis Dunton, Attorney GeneralIs General's Office, Environmental Control Division

(

(

III.I*2 III.1-2

Revision 8 April 1992 U. S. NUCLEAR REGULATORY COMMISSION

(

I, REGION IIII II

,Reports No. 50-010/84-01(DPRP);

50-010/84~Ol(DPRP); 50-237/84-06(DPRP); 50-249/84-0S(DPRP) 50-249/84-05(DPRP)

Docket Nos.50-010; 50-010~ 50-237; 50-249 l;censes licenses No. DPR-02; DPR-19; DPR-25 licensee: Commonwealth Edison Company P. O. Box 767 IL 60690 Chicago, Il Facility Name: Dresden Nuclear Power Station, Units 1, I, 2, and 3 Inspection At: Dresden Site, Morris, IL Inspection Conducted: March 27, 1984 through May 21, 1984 Inspectors: T. M. Tongue S. Stasek C. D. Anderson R. Lanksbury ,

Approved By: ~*"i6timos.

t,'~£;timos, P'rojects Branch 2C Chief Date Inspection Summary Ins ection durin the 21 1984 Re orts 50-237/84-06(DPRP)j 50-249 84-05(DPRP>>

No. 50-10 84-01(DPRP)j 50-237/84-06(DPRP);

Areas Inspected: Routine unannounced resident inspection of action on previous inspection findings, regional requests, 10 CFR 21 notifications, operational surveillance, maintenance, IE Bulletins, safety, events, fire protection program, surveillance.

reports, spent fuel shipments, Three Mile Island modifications, licensee event reports.

regulatory performance improvement plan, Unit 1 chemical cleaning, independent inspection, report review, and meeting with local municipal officials. The inspection involved a total of 398 inspector-hours onsite by 4 NRC inspectors inspector-hours onsite during off-shifts.

including 78 inspector~hours .

Results: Of the 16 areas inspected, no items of noncompliance or deviations were identified.

III.1-3 III .1-3

Revision 8 April 1992 DETAILS (0'

( SECTION I

1. Persons Contacted Commonwealth Edison - Station Personnel

• 0. Scott, Station Superintendent R. Ragan, Operations Assistant Superintendent J. Eenigenburg, Maintenance Assistant Superintendent

• J. Wujciga, Administrative and Support Services Assistant Superintendent J. Brunner, Technical Staff Supervisor R. Christensen, Unit 1 Operat;ng Operating Engineer

• J. Almer, Unit 2 Operating Engineer T. Ciesla, Unit 3 Operating Engineer D. Sharper, Waste Systems Engineer

• G. Myrick, Radiation Chemistry Supervisor B. Saunders.

Saunders, Station Security Administrator

.. Williams, Quality Assurance Coordinator ll.Williams,

• R. Stobert, Quality Assurance Inspector R. Stols, Quality Assurance Engineer

• T. Gilman, Chemistry Supervisor

• S. McDonald, Radiation Protection Supervisor M. Dillon, Fire Marshal T. Ziak;s, Ziakis, Emergency Preparedness Coordinator D. Ambler, Health Physicist

( Commonwealth Edison - Corporate Personnel D. A. Adam, Lead lead Health Physicist Field Services Engineer Contractors Home Transportation Company K. Jones, Driver Coyne Industrial Indus t ri a1 Laundry laundry - Joliet Jo 1i et E. Kasmark, General Manager NRC Personnel Region III R. Paul, Health Physicist A. Januska, Health Physicist G. France, Health Physicist R. lickus, Lickus, Director, State and Government Affairs 2

IlI.J-4 111.1-4

Revision 8 April 1992 The inspectors also talked with and interviewed several other licensee

( employees, including members of the teChnical technical and engineering staffs, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel, and contract security personnel.

• Denotes those attending one or more exit interviews conducted on April 19, April 23, May 4, May 8, and May 21, 1984, and informally at various times throughout the inspection period.

2. Action on Previous Inspection Findings Findinqs (Closed) Open Item (237/75-01(DPRP>>;

(237/75-01(DPRP>>: Torus (suppression pool) baffle removed. The baffles were removed in accordance with the Mark II con-tainment modifications program.

(Closed) Open Item (237/76-06(DPRP>>: Traversing Intore Incore Probe (TIP) isolation ball valve. This item is being reviewed generically. If any action is required, it will be forthcoming in appropriate NRC correspondence.

(Closed) Open Item (249/77-01(DPRP>>: High Pressure Coolant Injection (HPCI) system motor operated valve wiring change to prevent cycling after closing. The inspector reviewed modification packages and, through interviews with station personnel, verified that Unit 3 HPCI valves MO 2301-3, 5, 8 and 9 had wiring and motor operator modifications to prevent hammering and damage.

c.( (Closed) Open Item (237/77-145(DPRP>>: Control of Licensee Offsite Work.

Licensee offsite work has been reviewed on numerous occasions directly and indirectly. This was done through Division of Engineering routine inspections, management appraisal team audit and a performance appraisal team inspection. In each case all inspection findings have been appro-priately addressed by the NRC and the licensee.

(Closed) Open Item (237/78-07-02(DPRP) and 249/78-07-02(DPRP>>: Long term corrective action on improper 125 volt D.C. cable separation. The licensee has replaced the affected cables and has installed an additional battery charger.

(Closed) Open Item (249/78-25-04(DPRP>>;

(249/78-25-04(DPRP>>: Small leakage identified in the 3A, 8, B, and C reactor feedpump minimum flow lines. The licensee completed the long term corrective actions by replacing the flow control valves, installing smaller orifices, and replacing the eroded pipe elbOWS elbows with harder chrome alloy steel.

(Closed) Open Item (249/79-01-01(DPRP>>: Leak on main steam line control valve above seat drain line. The licensee replaced the affected 3000 psi steel line components with 6000 psi .stee1 components.

(

3 III .1-5 111.1-5

Revision 8 April 1992 (Closed) Open Item (237/79-13-02(DPRP) and 249/79-11-02(DPRP>>: Torus drain isolation to inhibit inadvertent torus draining to main condenser.

The licensee has completed installation of extra valving in the ljne used to pump excess water inventory in the torus to the condenser hotwel1.

hotwell.

Valves 2(3)-1599-61 and 2(3)-1599-62 are located downstream of the Low low Pressure Coolant Injection (LPCI) System Cross-Tie line on both UnltsUnits 2

• associated Modification and 3. The inspector reviewed the "associated modification package and afore.entioned valves and verified completion of installation of the aforementioned found all to be acceptable.

(Closed) Open Item (249/79-18-01(OPRP>>: Replacement of G.E. CR-120A relays. The relays were replaced as stated and in accordance with the licensee's modification program.

(Closed) Unresolved Item (lO/79-19-03(DPRP),

(10/79-19-03{DPRP). 237/79-23-03(OPRP).

237/79-23-03(DPRP). and 249/79-21-01(OPRP)}:

249/79-21-01(DPRP>>: licensee organizational changes were not shown

n in Technical Specifications. This matter was also identif;ed identified during reviews of Three Mile Island modifications (HUREG-0737 Item I.A.l.3.2.A)

I.A.1.3.2.A) 4, 19B2, the licensee stated that modifications" and by letter dated June 4,1982, modifications were implemented as required and a change request to the Technical sUbmitted.

Specifications had been SUbmitted.

(Closed) Noncompliance (10/80-18-01 (10/80-1B-01 (Region V Inspection>>: Radwaste drum lid improperly secured; identified at the Richland, Washington disposal site. The licensee has improved procedures for closing radwaste containers, required a management person to verify the condition of all shipments, and is showing more quality assurance and quality control attention to thes~

these shipments. There have been no additional findings

( citatio.n.

since this citatio~.

(Open) Unresolved Items (237/81-09-06(DE)

(237/B1-09-06(DE) and 249/81*06-06(OE>>:

249/BI-06-06(OE>>: Not all fire brigade members participate in at least two drills per year; (237/B1-09-07(DE) and 249/81-06-07(OE>>.

(237/81-09-07(0£) 249/B1-06-07(OE>>. Fire brigade training does not" not*

include the use of preplans or strategies for specific specHic instruction ~nd and reference during an emergency. By memo dated December 28, 2B, 1983, a request was resubmitted to NRC headquarters for guidance.

(Closed) Open Item (237IBl-20-01(OPRP)

(237/Bl-20-01(OPRP) and 249/81-14-01(DPRP>>:

249/81-14-01(OPRP>>: Safety and relief valve acoustical monitors - environmental and seismic certification. The licensee has submitted information to NRR concerning the environmental and seismic qualifications of the installed system in accordance with NUREG-0737 Task Action Item 11.0.3. NRR currently has the licenseels licensee's submittals under review. This item will therefore be followed under Task Action Item 11.0.3.

(237/Bl-24-02(DPRP>>: High Pressure Coolant (Closed) Noncompliance (237/81-24-02(DPRP>>:

Injection (HPCI) inoperable due to maintenance on the steam stop valve without a proper procedure. The licensee has developed a More .ore detailed Maintenance

.aintenance manual for the steam stop valve.

vahe. In addition.

addition, the licensee developing a permanent maintenance procedure for this activity.

is deve10ping

(

4 IIl.l-6

Revision 8S April 1992 (237/S1-35-01(DE) and 249181-27-01(0£>>:

(Closed) Unresolved Item (237/81-35-01(OE) 249/S1-27-01(DE: Con-flicting statement by equipment operator related. to event of November 25, 25 , 19S1, regarding starts, stops and associated alarms of the emergency 1981, diesel-generator circulating water pumps. This was resolved through investigation by the licensee and reported to the NRC via letter-dated letter*dated December 4, 1 9 S 81.* . (237/S1-35-02(DPRP) and 249/81-27-02(DPRP>>: - (Closed) Unresolved Item (237/81-35-02(DPRP) Z49/S1-Z7-DZ(DPRP>>: Possible violation of General Design Criteria 44 of 10CFR50 Appendix A in regard to common suction modes for the diesel generator cooling water pumps (OGCWPs). (DGCWPs). Via a telephone conversation with HRR NRR (as 1982), the c~on documented in a memo dated June 18, 1982). ca.mon suction .ade of DGCWPs was deemed to be acceptable as long as it was operation for the OGCWPs used only on a temporary basis (such as for required ** .aintenance), intenance), and procedures were in place to ensure that a water source was available to the pumps. The licensee currently has Dresden Operating Procedure (OOP) (DOP) 4400-5 in place that addresses the implementation of these requirements. Moreover, following discussions with the inspector, the licensee has initiated further changes to the procedure to better clarify the afore- . mentioned requirements. (Closed) Open Item (237/Bl-37-07(OPRP) (237/S1-37-07(DPRP) and-249/81-29-07(OPRP>>: and*Z49/S1-Z9-07(DPRP>>: Inspect pilot va1ve valve junction box interiors on electromatic relief valves to assure there is no interference with electrical contacts. This was completed on Units 2Z and 3 and no further problems were identified. (Closed) Open Item (237IBl-38-02(DPRP) (Z37/S1-3S-02(DPRP) and 249/81-31-02(DPRP}): Z49/81-31-02(DPRP)}: Access covers left open on components following maintenance. The licensee has issued a memo to all departments at Dresden stating the importance of restoring equipment following maintenance and surveillance. (237/82-10-02(DPRP) and 249/82-11-02(OPRP>>: (Closed) Noncompliance (237/82-10-02(OPRP) 249/S2-11-02(DPRP>>: Improper usage of general purpose hoses. The licensee has modified Dresden Administrative Procedure (DAP) 3-7 to reflect requirements* for proper hose usage onsite. Also, modifications have been made so that only breathing air hoses use special snap-tite connectors. All other general purpose hoses utilize Chicago type fittings. (237/S2-18-01(DE) and 249182-19-01(0£>>: (Closed) Noncompliance (237/82-18-01(0£) Z49/S2-19-01(DE>>: Inadequate fire protection surveillances. The 1icensee licensee has added the Cardox manual valve to the surveillance list and corrected the control room smoke detector surveillance lists. (237/S2-20-01(DPRP) and 249/82-21-01(OPRP>>: (Closed) Noncompliance (237/82-20-01(DPRP) 249/S2-21-01(DPRP>>: Measures were not established to control nonconforming parts in order to prevent their inadvertent installation or use. The licensee identified where all of the nonconforming valve guides were used and replaced the only nonconforming component used on Unit 3 during an October 1982 19S2 outage. Dresden Administrative Procedures, DAP 11-4 and 11-5, were implemented for classification and evaluation of spare parts used in safety related applications. app 1i cat ions. 5 111.]-7 IILl-?

Revision 8

                                                                         .April April 1992 (Closed) Noncompliance (237/82-20-02(DPRP) and 249/82-21-02(DPRP>>:

Measures were not established to distribute safety classifications to ( appropriate corporate or onsite personnel to assure the procurement, installation, and use of quality parts. Station Stati~n Nuclear Engineering Department (SNED) revised procedure Q.12 to assure correspondence on classification and listings of safety-related items is *distributed to the _Site,

     - site, corporate quality assurance and respective proje~t. groups.

project (237/82-22-D1(DPRP>>: 'Faflure (Closed) Noncompliance (237/82-22-01(DPRP>>: *Failure to .Iintlin

                                                                  .aintain primary containment integrity per Technical Specification requirements.

The licensee has completed all modifications, procedure changes, and Ind operator retraining concerning both Units 2 and 3 torus sightgllss sightg1ass operation per commitments. (Closed) Noncompliance (237/82-22-02(DPRP>>: Failure to report con-tainment integrity violation in a timely .. nner. All clerks responsible

                                                 ~nner_

for operating the telefax machines have been retrained concerning the importance of NRC notification requirements and the licensee's procedures concerning the telefaxing of these notifications. (Closed) Open Item (237/82-23-01(DPRP) and 249/82-23-01(DPRP)}: 249/82-23-01(DPRP>>: Process computer alarm disabled, bypasses operator acknowledgement. The licensee has installed new computer panels within reasonable reach of the operator, reviewed and removed some computer alarms, and implemented a shift surveillance to identify bypassed alarms. (Closed) Open Item (10/82-17-02(EPS); 237/82-24-02(EPS); and 249/82-24-02(EPS)}: Inaoequate hydrological forecasting exists at the 249/82-24-02(EPS>>: ( site and load dispatcherJs dispatcher's office. The licensee has an Army Corps of Engineers letter dated June 21, 1983, confirming a procedure to inform Lock and Dam. Dresden of best estimates on crest forecasts at the Dresden lock Darn. In addition, by letter dated June 9, 1983, the licensee has expanded weather forecasting agency verifying the willingness their contract with a Weather to respond to the station needs during an emergency_ emergency. (Closed) Noncompliance (249/82-28-01(OE>>: (249/82-28-01(DE>>: Failure of the Dresden Onsite Review Committee to provide an adequate review of a procedure. The inspector reviewed the licensee's response to this item, discussed it with the licensee and concluded that their contentions of no addi-appears to be valid. Based on tional corrective measures being required Ippears this and the currently inplace measures to ensure that the station procedures for calculating core thermal power are properly reviewed prior to their use, the inspector has no further concerns in this area. (Closed) Open Item (249/82-28-02(DE>>: (249/82-28-02(DE}): lack Lack of signoff/dating blocks on checklists in some startup physics test procedures. The inspector reviewed the startup physics test procedures and noted that the licensee had added additional signoff/dating blocks in the procedure where they deemed Idded it appropriate to do so. The inspector did not note any additional areas where it would appear appropriate to have signoff/dating.blocks Ind signoff/dating.b10cks and therefore has no further concerns in this area. ( 6 III.1-8 111.1*8

Revision 8 April 1992 (Closed) Open Item (237/83-11-09(DPRP) (Z37/83-11-09(OPRP) and 249/83-09-09(DPRP>>: Develop a training module and train maintenance mechanics on the proper use of sealants and lubricants. Maintenance mechanics were trained within the prescribed commitment and a permanent training module was developed by the Production Training Center for training new employees. _ (237/83-11-12(DPRP) and 249/83-09-12(DPRP}}: _ (Closed) Open Item (Z37/83-11-12(DPRP) 249/83-09-12(DPRP>>: -Torus (suppression pool) internals where modifications had no paint (preserva-tion) on newly welded areas. The licensee has a plan for periodic re-painting torus internals during subsequent outages. 249/83-13-01(DPRP>>: (Closed) Noncompliance (237/83-14-01(DPRP) and 249/83-13-01{DPRP>>: Failure to restore systems to normal following maintenance and/or surveillance. The licensee has issued a memo to all departments at Dresden to assure the adequacy of housekeeping practices for restoration of systems and components to prevent the intrusion of substances that could result in equipment failures. (Closed) Noncompliance (237/83-14-02(DPRP) and 249/83-13-02(DPRP>>: Inadequate corrective action from a previous event resulted in one train of an emergency core cooling system being inoperable. The licensee modified maintenance procedure DMP 040-6, "Safety Related Motor Operated Settings", given specific instructions to appropriate Valves - Data and Settingsll. personnel and distributed copies of deviation reports to Maintenance Departments for greater awareness. (Closed) Open Item (lO/83-12-01(DPRP). (lO/83-12-01(DPRP), 237/83-20-01(DPRP). 237/83-20-01(DPRP), and 249/83-18-01(OPRP>>): 18-01(DPRP>>: Licensee on-the-job training for maintenance personnel ( needs to be upgraded and documented. The licensee is implementing a more formal, four to five year program for mechanics. This was implemented in mid-1983 for mechanical and electrical mechanics, and is SCheduled to be implemented for instrument mechanics by September 1984. (Closed) Open Item (lO/83-12-02(DPRP), 237/83-20-D2{DPRP). 237/83-20-D2{DPRP), and 249/83-18-02(DPRP>>: New procedure and procedure modification backlog needs to be reduced, maintenance procedures need greater detail, and development of maintenance manuals should be stepped up. The licensee has developed a formal review process to help reduce the backlog, maintenance procedures are being developed with more attention to details, and more maintenance manuals are being developed or modified to help assure maintenance with better control over the work. In addition, the licensee is converting a number of maintenance manuals to approved maintenance procedures. (lO/83-12-03(DPRP), 237/83-20-03(DPRP), and 249/83-(Closed) Open Item (lO/83-12-03(DPRP). 18-03(DPRP>>: Work packages on safety related valves needs a -generic set 18-03(DPRP)): of Quality Assurance/Quality Control (QA/QC) hold points. In addition. addition, work could start without appropriate QA or QC approval. The licensee has made the following changes: (/\- 8 III.1-9

Revision 8 April 1992

3. Modified Dresden Administrative Procedure OAP DAP 15-1 "Work Requests Requests",

ll witness points and generated by adding a set of guidelines on QC hold and ~itness DAP 15-3, "Preparation of Safety Related or Reliability Related Work Packages on Off-shifts", which whi ch requires QC and/or andlor QA approval to start safety related work. .'

            '(Closed) Open Item (lO/83-12-04(DPRP),

(10/83-12-04(DPRP), 237/83-20-04(DPRP), 237/83-20-04(OPRP), and 249/83-18-D4(DPRP>>: Equipment and parts obtained as nonsafety related .ust b~

       .- 18-04(DPRP>>):                                                                     be upgraded prior to use in a safety related application. The licensee has modified Dresden Administrative Procedures DAP    OAP 11-4, uControl
                                                                        *Control of the Classification list of Safety Related (SR).(SR), Non-safety Related (HSR) and American Society of MeChanical Engineering (ASHE) Code-Related Systems, Structures and Components"; OAPDAP 11-5 Classification of Non-Safety Related (NSR) Subcomponents/Parts Used on/in onlin Safety Related (SR) Systems, Struc-tures and Components"; DAP 11-6, "Request for Purchase and Receiving Inspection Guidelines"; and OAPDAP 11-7, "Technical Evaluation of Parts Used in Safety Related Components.

Components."II These were reviewed by the' the inspector and appear to be acceptable to correct the problem identified. More recently problems with code related drywel1- drywell-to*torus to-torus vacuum breaker shaft seals is being addressed under a separate special inspection. . Dpen Item (lO/B3-12-0S(DPRP), (Closed) Open (lO/83-12-05(DPRP), 237/83-20-0S(DPRP), 237/83-20-05(DPRP), and 249/83-18-0S(DPRP>>: Improved communications needed between Maintenance 18-0S(DPRP)}: maintenance operations and radiation protection. In addition to previous corrective actions identified, an ALARA review is required on work request forms. Review by

,                                            corrective actions have been successful.

the inspector shows that the corre~tive ( (Closed) Open Items (249/83-19-01(DPRP), 249/83-19-02(OPRP), 249/83-19-02(DPRP), 249/83-19-03(DPRP), and 249/83-19-04(DPRP)): 249/83-19-04(DPRP>>: NRC Order of August 26. 26, 1983, cracKs identified in BWR large diameter piping. The licensee related to cracks implemented leakage control requirements, shutdown and examined all piping as required, and completed the remainder of the items as required .. By NRR letter dated March 15, IS, 1984, the order has been rescinded to allow continued operation within some constraints. (237/83-21-01(DPRP>>: Insufficient corrective (Closed) Noncompliance (237/83-21~Ol(OPRP>>: action relative to holes in torus-to*drywell torus-to-drywell vacuum breaker lines. Following the second incident on August 11, II, 1983, the licensee instituted

            ~ore intensive corrective actions (as outlined in a letter dated November 10, 1983) to ensure further incidents of this type would not occur at Dresden. Modifications to the piping supports for the torus-to-drywell vacuum breaker lines have since been completed for both Units 2 and 3 without further incident.

(Closed) Noncompliance (237/83-21-03(DPRP}): (237/83-21-03(DPRP>>: Valve misalignment

                                                                       ~isalignment due to inadequate procedures. The licensee has modified the identified procedures (DOS 1500-1 and DOS                         ~ore specific instructions for DDS 1600-1) to include more correctly draining between containment spray valves 1501-27A(B) lSOl*27A(B} and
         ,. 1501-28A(B) during valve operability surveillances.

No further items of noncompliance or deviations ~ere were identified. ( 9 1II.l-10

Revision 8* 8 April 1992 for those components. In addition, the inspector verified that the licensee and the supplier were aware of this issue for identification of other components that may be supplied in the future. No items of noncompliance or deviations were identified.

5. Operational Safety Verification The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the period from March 27 to May 21, 1984. The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of Unit 2 reactor building and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance.

During the inspection period while Unit 3 was in an outage to repair the. main turbine, the inspectors verified that surveillance tests were conducted, containment integrity requirements were met, and emergency systems were available as necessary. Throughout the entire inspection period. period, Unit 1 remained in a longterm shutdown condition with all fuel removed from the vessel. The inspectors verified that all applicable requirements for Unit 1 were met during this period. ( The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security plan. The inspectors observed plant housekeeping/cleanliness conditions and verified implementation of radiation p~otection pr:otection controls. During the inspection, the inspector walked down the accessible portions of the following systems to verify operability by comparing system lineup with plant drawings, as-built configuration or present valve lineup lists; observing equipment conditions that could degrade performance; and verified that instrumentation was properly val ve.d , functioning, and valved, calibrated.

a. Unit 2 low Pressure Coolant Injection System (both loops), Core Spray System (both loops), Isolation Condenser, Unit 2 Emergency Diesel Generator, and portions of the Control Rod Drive System.

b. Unit 3 Unit 3 Emergency Diesel Generator 11 111.1-11

Revision 8 April 1992 an upper bound. (NRC 10 eFRCFR 50.72 reporting requirements sets ( 2 MPC as the lower limit that need be reported.) Followup review of this incident has been assigned to the Region III Facilities Radiation Protection Section (FRPS) to be looked at during thei-r their next inspection at Dresden. No items of noncompliance or deviations were identified.

7. Fire Protection Program During the inspection period period, the inspector reviewed the licensee's t

fire protection program against Technical Specification requireaents require~nts and licensee commitments. Walkdowns were conducted of the accessible portions of the cardox system, halon system, water suppression system, and portable fire protection equipment. The following surveillances were reviewed for adequacy and completeness: OFPP DFPP 4114-2 Reactor Building Monthly Fire Equipment Inspection OFPP DFPP 4114-3 Turbine Buildino Monthlv Fire Eouioment Insoection OFPP DFPP 4145-1 Cardox System Semi-Annual Maintenance Test OFPP DFPP 4153-2 Emergency lighting Monthly Inspection DFPP 4185-2 Smoke Detector Semi-Annual Maintenance Test No items of noncompliance or deviations were identified in this area.

8. Surveillance Observation

( The inspectors observed Technical Specifications required surveillance survei llance

  '. testing on the Unit 2 Emergency Diesel Generator and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.

The inspectors also witnessed/reviewed portions of the following test activities:

a. Unit 2 local Power Range Monitors Calibration, Core Spray System Pump Test with Torus Available, Low low Pressure Coolant Injection System Valve Operability Test, and High Pressure Coolant Injection System Motor Operated Valves and Pump Operability Test.

b. Unit 3 Low low Pressure Coolant Injection System Valve Operability Test No items of noncompliance or deviations were identified in this area.

13 III .1-12 III.1-12

Revision 8 April 1992 ( 9. Maintenance Observation Station maintenance activities of safety related systems and components listed below were observed/reviewed to ascert~in ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications. The following items were considered during this review: the 11.;t1ng limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemented. Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance. The following maintenance activities were observed/reviewed:

a. Unit 2 lei Traversing lncore

               'C'             Incore Probe Machine Maintenance and Emergency Diesel Generator Quarterly Maintenance

b. Unit 3

               'A' Low IAI low Pressure Coolant Injection Heat Exchanger Cleaning No items of noncompliance or deviations were identified in this are~.

area.

10. IE Bulletin Followup Each of the following IE Bulletins were reviewed by the resident inspector to determine if: (1) the licenseels licensee's written response was submitted within the time limitations stated in the bulletin, (2) the written response reported, (3) the written response included all information required to be reported.

included adequate corrective action commitments based on information presented in the bulletin and the licensee's response, (4) licensee management forwarded copies of the written response to the required

        ~anagement ansite onsite management representatives.

representatives, (5) information discussed in the licenseels licensee's response was accurate, and (6) the corrective action taken was as described

        ;inn the response.

(Closed) Miller, Inc. IEB 83-07 Apparent Fraudulent Products Sold by Ray Miller. (Closed) IEB 83-08 Electrical Circuit Breakers With an Under-Voltage lEB Trip Feature in Use in Safety-Related Applications Other Than the Reactor Trip Syste~.

 \.      No items of noncompliance or deviations were identified in this area.

1A III.1-13 rrI.l*13

Revision -B8 April 1992 of Transportation (the responsible agency) was informed of the event. event, The licensee examined and repaired the trailer in accordance with applicable ( codes per DOT and placed it back in service. No items of noncompliance or deviations were identified in this area.

13. Three Mile Island Modifications The inspector reviewed the following TMI items for development and implementation per NUREG-0737 requirements and licensee commitments.

II.F.l.3: Accident Monitoring - Contain.ent Containaent High Range In response to this task action item, the licensee~ as dOCUMented licensee, IS docUMented via a letter dated April 15, IS, 1982, installed contai~ent contairwent radiation .onitors IIOnitors on both Units 2 and 3. Further licensee commitments specified that associated procedures, when finalized, would include appropriate correction factors to modify instrument readings to correspond with actual contain.ent containaent radiation levels. On June 1, I, 1982, the info~ation information concerning these correction factors and how they were arrived at were submitted for review. Verification of licensee actions concerning this task item was aSSignedassigned to the Region III Facilities Radiation Protection Section. As documented in Inspection Report (237/82-30; 249/82-31(DRMSP>>, the licensee appears to meet the intent of the NUREG-0737 requirements for this item. Therefore II.F.l.3 is considered closed at this time. III. A. 2. 4: Installation of Emergency Response Facility (ERF) Meteoro-logical Hardware and Software. ( IILA.2.S: IILA.2.5: III. A. 2. 6: IILA.2.6: Full Operability of III A.2.4. Review of Dose Calculation Methodology (DeM) (DCM) By The Licensee. IILA.2.8: IILA.2.B: Full Operation of Class B Model. These task action items, as currently outlined, reflect requirements as issued in NUREG*0660 NUREG-0660 and NUREG-0737. However, since their thei r issuance*, issuance', Secy 82-111 "Requirements "Requi rements for Emergency Response Capability" Capabil i ty" has been issued which significantly modified the original requirements. As documented in a memo dated March 1, 1984 from C. Paperiello to C. Norelius, the Emergency Response Facility (ERF) Appraisal Program is the current program proposed for the review of these items. Since these task action items will be reviewed using the Secy 82-111 criteria by the ERF Appraisal Program, items III.A.2.4, III.A.2.S; III.A.2.S, III.A.2.6, and III.A.2.B III.A.2.8 are considered closed because the criteria specific to NUREG-0737 no longer fully apply. No items of noncampl,ance noncompliance or deviations were identified.

14. Regulatory.Performance Regulatory ,Performance Improvement Plan Commonwealth Edison Company implemented a Regulatory Performance Improvement Plan (RPIP) in February 1984. The plan concept was a tonaal fOnlal effort to improve safety and error-free operations. This was developed in response

( 17 III .1-14

• Revision aB April 1992 Hl92

( to NRC concerns over recent errors and escalated enforcement actions. During the inspection period, the inspectors have observed licensee actions such as followup on potentially significant events, conduct of operations, cleaning, painting and improvements in appearance of the plant. In addition, the inspectors have had discussions with shift overview (50S), the station superintendent, superi ntendents (50S). superintendents superi ntendent, and various vari ous corp*orate _ personnel. It appears that there is an improving trend. However.However, further observations are necessary to form a conclusion. This Matter aatter will be observed and addressed in future inspections. No items of noncompliance or deviations were identified in this area.

15. Unit 1 Chemical Cleaning The inspectors observed the licensee's preparation for the chemical cleaning of Unit 1 currently planned for July 1984. The inspectors reviewed new and modified procedures, interviewed personnel involved in project, the proj facilities.

ect, and toured the fac Thiss wi i1 it i es. Thi will 11 continue to be observed as part of the routine inspection program until the project is complete .. No items of noncompliance or deviations were identified in this area.

16. Independent Inspection Unit 3, Main Turbine Repair

   /         During the inspection period, the inspectors followed the repair activities

( on the Unit 3 high pressure turbine and verified that adequate radiation protection precautions were being implemented. No items of noncompliance or deviations were identified in this area.

17. Report Review During the inspection period, the inspectors reviewed the licensee's Monthly Operating Reports for March and April 1984. The inspectors confirmed that the information provided met the requirements of Technical Specification 6.6.A.3 and Regulatory Guide 1.16.

No items of noncompliance or deviations were identified in this area.

18. Local Municipal Officials Meeting with local On May 10, la, 1984, at 7:00 pm, a meeting was held for local public officials in the board meeting room of the Grundy County Court House. The purpose of the meeting was to provide an opportunity for local public officials to meet the resident inspectors and associated Region III personnel, and discuss to the Resident Inspection Program for Dresden. The .eeting aeeting was attended by approximately 26 officials, their guests, and several indi-viduals from State of Illinois agencies. The major areas of interest were 18 III.l-1S III.1-15

        ..*                                                                           Revision 8 April 1992 in emergency preparedness and its demand on local resources with minimal compensation, the resident inspectors' roles in daily plant activities, and the NRC enforcement program. At"~he At.~he conclusion, the Grundy County Sheriff provided a tour of the emergency preparedness communications and notification facilities for the NRC personnel.

19._ Unresolved Items 19.-

                                                                      .. ,.0_:... ,,/

Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items. items, items of noncompliance, or deviations. Unresolved items addressed during the inspection are discussed in Paragraph 2.

20. Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. Open items addressed during the inspection are discussed in Paragraph 2.

21. Exit Interview The inspectors met with lieensee representatives (denoted in Paragraph 1) throughout the inspection period and at the conclusion of the inspection on May 21, 1984, and summarized the scope and findings of the inspection activities. The licensee acknowledged the findings of the inspection.

(

 ,/"

t

   /
   \\ .
                                                    ]9 III.1-16

Tab 2 Revision"S April 1992 DRESDEN 2 & &3 ( FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspect jon Report No. 50-01Q/84-09, 50-010/84-09. 50-237/84-11. 50-249/84-10

       ~
       .flgg                       ~

111.2-1 III. 2-1 Inspection Report No, No. 50-010/84-09, 50-237/84-11, 50-249/84-05 dated HayMay 25, 1984. I1f.2-13 IIf.2-13 July 30, 1984 CECo letter from D. l. L. Farrar to J. G. Keppler (NRC); Response to Inspection Report No. 50-010/84-09, 50-237/84-11, 50-249/84-10. ( c(~ III .2-; Ill.2-i

 .,                                      UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III Revision 8 April 1992 "t

7U ROO$EVEL T ROAO ROAD GLEN ELLVN. ELL VN. ILLINOIS 60137

                                                         '0137 tJUL 3 "UL JUl66 JUL        It'D It'll Docket No. 50-10 Docket No. 50-237 Docket No. 50-249 Commonwealth Edison Company ATTN:*

ATTN:- Mr. Cordell Reed Vice President Post Office Box 767 Chicago, Il 60690 Gentlemen: This refers to the routine safet~ inspection conducted by Messrs. T. M. Tongue and s. Sta~eJ~~f stase_k~f thjs this office dYrlog dYrlng thg the perjod period of Ma,y "22 through .Il1ne 15, 1984 throllgh ')II"e

    ~.~~~!.~_ties
    ~_~~~i~_ties at Dresden Nuclear power Statio",                      Ind 3, authorized by Station, Units "1, 2 and NRC Operating licenses No. DPR-02, DPR-19, and DPR-25, and to the discussion of our findings with Mr. D. Scott and others of your staff at the conclusion of the inspection.

The enclosed copy of our inspection report identifies areas examined during the areas, th~ inspection. Within these areas. the inspection consisted of a selective ( examination of procedures and representative records, observations, and inter-views with personnel. During this inspection, inspection. certain of your activities appeared apoeared to be in noncom-

    ~liance
    ~]iance  with NRC requjrements, requirements, as specifjed specified jn the enclosed Appendj¥ Append;x        A written response reS~Qnse is reguire9.

In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure(s) will be placed in the NRC Public Document Room unless you notify this office, by telephone, within ten days of the date of this letter and submit written application to withhold information contained therein within thirty days of the date of this letter. Such application must be consistent with the requirements 2.790(b)(I). If we do not hear from you in this regard within the specified of 2.790(b){I). periods noted above, a copy of this letter. letter, the enclosure(s}, and your response to this letter will be placed in the Public Document Room. The responses directed by this letter (and the accompanying Notice) are not subject to the clearance procedures of the Office of Management and Ind Budget as required by the Paperwork Reduction Act of 1980, Pl 96-511. II1.2*1

Revision 8 April 1992 ( Commonwealth Edison Company 2

3 klUL 3 krUL We will gladly discuss any questions you have concerning this inspection.

Sincerely, If)1&'/ 1.- w££/ I. . W. D. Sha~

w. D.

Reactor Porj~ts Branch 2

Enclosures:

1. Appendix, Notice of Violation

2. Inspection Report No. SO-OlO/84-09(DRP);

SO-010/84-09(DRP); No. 50-237/84-11(DRP); SO-237/84-11(DRP); No. SO-249/84-10(DRP) cc w/encl: D. L. Farrar, Director Nuclear Licensing of Nuclea~

   , D. J. Scott, Station O.

Superintendent ('" DMB/Document Control Desk (RIDS) Resident Inspector, RIll Phyllis Dunton, Attorney General*s General's Office, Environmental

      ,Control Division

( IIL2-2 II L 2-2

Revision 8 April 1992 ( Appendix NOTICE OF VIOLATION Commonwealth Co~onwealth Edison Company Docket No. 50-237 Docket No. 50-249 As a result of the inspection conducted on May 22 through June 15. 15, 1984, and in accordance with the General Policy and Procedures for NRC Enforce.entActions, Enforte.entActions, (10 CFR Part 2, 2. Appendix e), C), the following violation was identified: Dresden Technical Specification 4.12.0.3 states in part ItAt "At least once per cycle. the (Cardox) system valves and associated dampers will be operating cycle, veri fi ed to actuate automatically verified automat i ca lly and manually."

                   .:th!Labqye. the Cardox s~stem Contrary to .1.h~.aI;>QY~_.                 system discharge .aster yalve valve was not tested
    \0 verify actuation in the automatic mode and it appears that testing in the suryeillaDce program.

automatic mode was not included in thg surveillance Pursuant to the provisions of 10 CFR 2.201, 2.201. you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply. including for each ;tem item of noncompliance: (I) (1) corrective ( (. action taken and the results achievedj achieved; (2) corrective action to be taken to avoid further noncompliance; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown. tJUL

                'JUL 3     ~19b4 Dated

( (- Il1.2-3

Revision 8 Apri1 April 1992 ( U. S. NUCLEAR REGULATORY COMMISSION REGION HI 50-010/84-09(oPRP); 50-237/84-11(OPRP); Reports No. 50-010/84-09(DPRP); 50-237/84-11(oPRP); 50-249/84-10(OPRP) 50-249/84-10(oPRP) Docket Nos. 50-010; 50-237; 50-249 Licenses No. DPR-02; DPR-19; DPR-25 Licensee: Commonwealth Edison Company P. O. Box 767 Chicago, IL 60690 Name: Facility Harne: Dresden Nuclear Power Station, Units 1, 2, and 3 Inspection At: Site, Morris. Dresden Site. Horris, Il IL Inspection Conducted: May 22 through June 15, 1984 Inspectors: T. M. Tongue oJ" Stasek.~'/ S. Stasek ( Approved By: ~;(11

                    ~~~   ~~tj;os,
                          ~~~~or;os, Chief Re~t~n~rOjects Re~t~n~rojects Branch 2C                        Date Inspection Summary Inspection during the period of May 22 through June 15.15, 1984 (Report Nos. 50-10/84-09(DPRP)j 50-237/84-11(DPRP);

Hos. 50-237/84-11(DPRP)j 50-249/84-10(DPRP>> 50-249/84-10(oPRP>> . 0 Areas Inspected: Routine unannounced resident inspection of action on previous findings, headquarters request, regional request, operational inspection findings. safety, fire protection program, surveillance program, .aintenance, licensee event reports, I.E. Information Notices, Unit 1 chemical cleaning, spent fuel shipments, and report review. The inspection involved a total of 122 inspector-hours onsite by 2 NRC inspectors including 22 inspector*hours inspector-hours onsite during off-shifts. Results: Of the 12 areas inspected no items of noncompliance or deviations were identified in 11 areas; one item of noncompliance was identified in one area (inadequate surveillance testing of cardox system - paragraph 6). c.c IlI.2-4

Revision 8 April 1992 DETAILS

1. Persons Contacted

      *D.
      *0. Scott, Station Superintendent R. Ragan, Operations Assistant Superintendent J. Eenigenburg, Maintenance Assistant Superintendent J. Wujciga.

Wujciga, Administrative and Support Services Assistant Superintendent J. Brunner, Technical Staff Supervisor R. Christensen, Unit 1 Operating Engineer J. Almer, Unit 2 Operating Engineer T. Ciesla, Unit 3 Operating Engineer D. Sharper, Waste Systems Engineer G. ~rick, Radiation Chemistry Supervisor B. Saunders, Station Security Administrator M. Dillon, Station Fire Marshall S. McDonald, Radiation Protection Supervisor J. Achterberg. Achterberg, Assistant Technical Staff Supervisor D. Ringo, Surveillance Coordinator

      *R. Stobert, Quality Assurance Inspector Cont ractor:

Contractor: K. Jones, Driver, Home Transportation Corporation (c. State of Illinois:

v. Muzzallup*o, V. Muzzallupo, Illinois Department of Nuclear Safety R. Reese, Hazardous Materials Officer, Illinois State Police The inspector also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel, and contract security personnel.

      *Denotes those attending one or more exit interviews conducted on May 25 and June 15, 1984, and informally at various times throughout the inspec-tion period.

2. Action on Previous Inspection Findings (Closed) Open item (237/79-20-01(DPRP>>: Loss of secondary containment blow-off bolts for the panels and integrity. The licensee redesigned the bloW""off has permanently blocked the supply fan vortex dampers to assure proper containment integrity.

2 III.2-5 111.2-5

RevisionB Revision 8 April 1992 ( TechnicalSpecifi-(Closed) Unresolved item (237/83-21-02(DPRP>>: Dresden Technical .Specifi-cations do not adequately address the limits on allowable primary contain-ment leakage during plant operation. By memo dated April 4, 1984, an interpretation was made by NRR on this Matter matter which essentially agreed licensee's interpretation. That is, La is considered to be the with the licensee*s leakage limit during plant operation and 0.75 La is 11.iting li.iting ~nly during the performance of a Type A test. . . No further items of noncompliance or deviations were identified.

3. Headquarters Request By memorandum dated May 24, 1984, the Director of the Office for Analysis and Evaluation of Operational Data requested the resident inspectors to review certain occurrences with the licensee for reportability under 10 CFR 50.73. The events in question were reported through the NRC's -Morn-ing report" system but were not reported as licensee events (lERs)

(LERs) pursu-ant to 10 CFR 50.73. The specific event at Dresden was when the licensee reported, per 10 CFR 50.72, "Immediate Notifications" of being in an unusual event due to outages on redundant ECCS equipment. Further evalua-tion revealed that personnel on earlier shifts had considered the circum-stances and that the situation was perMitted permitted by Technical Specifications. The resident inspectors discussed this with the licensee at the ti.e time and agreed that the licensee had not been in an unusual event., event.. The MOrning morning report was written as a fol1owup followup to the ENS phone call. The event oc-curred on January 26, 1984. No further action is considered necessary on ( this issue. No items of noncompliance or deviations were identified in this area.

4. Regional Request A regional request was reviewed by the resident inspectors for applicabil-ity at Dresden based on an event identified at the Quad Cities (QC) nuclear station. (Reference inspection report 50-254/83-04(DPRP);

50-26S/84-03(DPRP>> 50-265/84-03(DPRP>> During a 125 volt D.C. battery discharge test at QC Unit 1, it was found that the discharge rate was at B5 85 amperes steady state. The Final Safety Analysis Report (FSAR) stated that the battery discharge rate should be 62.3 amperes for 8 hours. Investigation revealed that modifications (additional loads) have been added to the battery and it appears that the added loads were not reviewed pursuant to 10 CFR 50.59 for its effect on the battery capability. conditions, Subsequent evaluation revealed that the battery, under present conditions. would have insufficient capacity under certain accident conditions. ( 3 III J1I .2-6

Revision B8 April 1992 ( The 10 CFR 50.59 reviews were conducted within the Station Nuclear Engi-neering Department (SNED) for both QC and Dresden and it appears that the same omission occurred for Dresden. However, review of records at Dresden since 1981 show no evidence of exceeding a level of 60 aMperes (nOMinally (nOlinally loads were about 52 amps). In addition, ~nlike unlike QC. QC, the Dresden FSAR shows a specific list of battery loads. Personnel at Dresden are also submit-ting an Action Item to SNED for a battery load profile review.

                                                " -: '~'. '-:~~ '"'--- ;-;-. ' ... -

The information on the QC battery has been submitted to NRR for review and evaluation. The outcome of that review will be used to deter.ine enforce-Mnt ment action at QC as well as Dresden. This issue is presently considered an unresolved inspection item. (50-237/84-11-01(DPRP); (50-237/84-11-D1(DPRP); SO-249/84-10-01{DPRP>>. 50-249/84-10-01(DPRP>>. No items of noncompliance or deviations were identified in this area.

5. Operational Safety Verification The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the period from Hay 22 to June 15, IS, 1984. The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of Unit 2 reactor build~ build-ing and turbine building were conducted to observe plant equipment condi-tions, including potential fire hazards, fluid leaks, and 'excessive and*excessive maintenance requests had been initiated for vibrations and to verify that maJntenance

( equipment in need of maintenance. During the inspection period while Unit 3 was in an outage to repair the main turbine, the inspectors verified that surveillance tests were con-ducted, containment integrity requirements were met, and emergency systems were available as necessary. Throughout the entire inspection period, Unit 1 remained in a 10ngterm longterm shutdown condition with all fuel removed from the vessel. The inspectors verified that all applicable requirements for Unit 1 were .et during this period. The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security plan. The inspectors observed plant housekeeping/cleanliness conditions and verified implementation of radiation protection controls. During the inspection, the inspectors walked down the Iccessible accessible portions of the following systems to verify operability by comparing system lineup with plant drawings, as~built as-built configuration or present valve lineup lists; observing equipment conditions that could degrade perforaance; perfor.ance; and veri-fied that instrumentation was properly valved, functioning, and calibrated. (. 4 III .2-7 III.2-7

Revision 8 April1 1992" Apr; 1992. ( a. Unit 2 Low Pressure Coolant Injection ~ystem (both loops), Core Spray System (both loops), and Isolation Condenser.

b. Unit 3 Low Pressure Coolant Injection Syste.

Systetl ("B"'.loop), Core Spray System ("B" (118" loop) loop), Isolation Condenser, and Unit 3 ERrgency t Ellergency Diesel '. . Generator.

c. Unit 2/3 (Common)

Standby Gas Treatment System, Cardox Syste.. Systetl, Halon System, and Fire Water System. The inspectors reviewed new procedures and changes to procedures that were i~p1emented during the inspection period. i~plemented The review consisted of a verification for accuracy, correctness, and compliance c~liance with regulatory requirements. The inspectors also witnessed portions of the radioactive waste system controls associated with radwaste shipaents and barreling. These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedures. C" ( No items of noncompliance or deviations were identified in this area.

6. Fire Protection Program During the inspection period, the inspector reviewed the licensee's fire protection program against Technical Specification requirements require~ents and licensee commitments. The inspector verified that welding and cutting operations along with other activities involving open fla.e flaae ignition sources in safety related areas were properly perfor.ed perforaed in confor.ance confoMlance with appropriate station procedures. Walkdowns were conducted of the accessible portions of the cardox system, halon system, water suppreSSion suppression system, and portable fire protection equipment. Proper housekeeping in safety related areas was also verified during plant tours. Training sessions and periodic drills for fire brigade .embers members were found to be inplace and acceptable. The following surveillances were reviewed for adequacy and completeness!

completeness: DFPP OFPP 4123-5 "Diesel Fire Pump Week Operabilityll Operability" OFPP DFPP 4132-1 IIVerification of U-2/3 Sprinkler SystHls "Verification Systems Integrity" OFPP 4132-2 "Verification of Unit 2 Sprinkler Systells Systems Integrity" DFPP OFPP 4132-:3 4132~3 IIVerification of Unit 3 Sprinkler Systells "Verification Syste~s Integrity" OFPP DFPP 4145-1 uCardox System Semi-Annual Maintenance Test" "Cardox DFPP OFPP 4153-2 IIEmergency Ughting "Emergency Lighting Monthly Inspection" OFPP 4175-1 "FireStop "Fire Stop Integrity and Maintenance Maintenance"ll DFPP OFPP 4175-2 "Operati "Operatingng Fi Surveillance"ll re Stop/Barrier Survei1lance Fire OFPP 4175-4175-33 Shutdown Fire "Shutdown II Fi re Stop Survei 11 ance" llance" OFPP 4185-3 "Fire Detection System Operation" OFPP 4195-1 "Halon Systems Semi-Annual Maintenance" 5 III.2-8

Revision 8 April 1992 j While reviewing surveillance DFPP 4145-1. Whil~ 4145-1, the inspector noted that the master discharge valve, in the Cardox system was not verified to open on an automatic initiation of the system. This valve is located on the discharge piping downstream of the CO2 storage tank, and is used to pressurize the system's main header in the event that a fire is sensed in systea anyone of the areas protected by the Cardox 5yst .. (including all three eMergency diesel generator rooms). When station .. nagelent was .. aanagaent de .ware aade aware of the deficiency in the surveillance, the Cardox systea systaa was declared inoperable and a special procedure (SP 84-5-35) was wrftten to test the

              .aster discharge valve. The valve was subsequently tested the SAMesa.! d~

dey and verified to operate correctly. The inspector reviewed the procedure and witnessed the test and found all aspects of the surveillance to be acceptable. Following the successful completion of the test. test, the licensee again declared the system operable. ,. 4.12.0.3 requires that all valves in the Dresden Technical Specification 4.12.D.3 Cardox system be tested at least once per operating cycle to verify each will actuate manually and automatically. Because the .. ster discharge aaster valve was not tested for automatic actuation in accordance with the aforementioned requirement, this is considered an item of noncompliance (237/84-11-02(DRP); (237/84-11-D2(DRP); 249/84-10-02(DRP}). 249/84-10-02(DRP)). One item of noncompliance was identified in this area.

7. Surveillance Observation The inspectors observed Technical Specifications required surveillance

(---- testing on the Unit 2/3 (Swing) Emergency Diesel Generator Indand verified

     ... "    that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated.

calibrated, that limiting conditions for opera-tion were met. met, that removal and restoration of the affected components were accomplished, that test results conformed with TeChnical Technical Specifica-tions and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel. The inspectors also witnessed/reviewed portions of the following test activities: Unit 2 Core Spray System Pump Test With Torus Available, Core Spray System Valve Operability Check, low Pressure Coolant Injection (lPCI) System Valve Test, and LPCI Operability Test. lPCI System Pump Operability Test With Torus Available. Unit 3 Dai1yIWeekly Storage Battery CheCK. Check. Unit 2/3 (Common) i Cardox System Master Valve Operability Test. 6 III.2-9 III .2-9

                                                                             ....... _ _ .u.>oIt.£'!'Lo. .. """'-.

Revision 8 AprH 1992 April ( The inspector also reviewed the master surveillance program for testing and calibration as required by technical specifications. This involved a verification of frequencies, responsiple plant groups and test status. The inspector tested the system to verify that recent technical specifi-cations had been appropriately addressed and, that fo~al fOnlal ~thods and responsibilities had been defined for review of test data and reporting deficiencies, etc. ,. No items of noncompliance or deviations were identified in this area ..

8. Maintenance Observation St.tion Station .ainten.nce maintenance activities of safety related systels syste.s and cOlponents coaponents listed below were observed/reviewed to ascertain that they were conducted regulato~ guides and industry in accordance with approved procedures, regul.to~

codes or standards and in conformance with Technical Specifications. The following items were considered during this review: the li.iting limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to ,initiating the work; to.initiating activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were perfor.ed perfo~d prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, and. fire prevention controls were implemented. ( Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance. The following maintenance activity was observed/reviewed: Unit 3 Emergency Diesel Generator Bi-Annual Maintenance No items of noncompliance or deviations were identified in this area.

9. licensee Event Reports Followup Through direct observations, discussions with licensee personnel, and review of records, records. the following event reports were reviewed to determine that reportability requirements were fulfilled, i.mediate immediate corrective accomplished. and corrective action to prevent recurrence had action was accomplished, been accomplished in accordance with technical specifications.

( 7 III.2-10 I1I.2-10

Revision 8B April 1992 ( Unit 2 (Closed) 237/82-49 Torus Inert and Purge Valve 2-1601-56 Did Not Auto-Close During Surveillance. (Closed) 237/83-08 Indications Discovered During Inservice Inspection. .. (Closed) 237/83-12 Mechanical Snubber Failure on Main Steam Une. SteU! Line. (Closed) 237/83-29 Excessive leakage Leakage Found During Integrated Leak Rate Test. Unit 3 (Closed) 249/82-06 Batte~ Quarterly Surveillance. Missed Batter;y (Closed) 249/82-14 Discovery of Crack in Unit 3 Head Seal Leak-Off Line. line. (Closed) 249/84-04 Reactor Scram While Performing Surveillance. The preceding lERs have been reviewed against the criteria of 10 CFR 2, Appendix C, and when the incidents described .eet all of the following requirements, no Notice of Violation is normally issued for that item.

a. The event was identified by the licensee,

b. The event was an incident that, according to the current enforcement policy, met* the criteria for severity levels IV or V violations,

c. The event was appropriately reported,

( d. The event was orDr will be corrected (including .easures to prevent recurrence within a reasonable amount of time), and

e. The event was not a violation that could have been prevented by the t

1icensee licensee's s corrective actions for a previous violation. No items of noncompliance or Dr deviations were identified in this area.

10. I.E. Information Notice Followup Each of the following I.E. Information Notices (lEN) was reviewed by the Resident Inspector to verify 1) that the information notice was received by licensee management, 2) that Ia review for applicability was performed.

performed, and 3) that if the information notice was applicable to the facility, appropriate actions were taken or were scheduled to be taken. (Closed) lEN 83-01: Ray Miller, Inc. I.E. Bulletin subsequently issued addressing this matter. (Closed) lEN 83-02: limitorque HOBC, HIBC, H1BC, H2BC, and H3BC Gearheads. Not applicable to Dresden. (Closed) lEN 83-03: Calibration of Liquidliquid Level level InstrUMents. Instru.ents. Density and temperature compensation is accounted for during calibration operations. (Closed) lEN 83-04: Failure of ELMA ElMA Power Supply Units. Not applicable to Dresden. ( 8 111.2-11 II1.2-11

Revision 8 April 1992 ( 16. Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. An open item addressed during the inspection is discussed in Paragraph 2. Z.

17. Exit Interview The inspectors met with licensee representatives (denoted in Paragraph 1) throughout the inspection period and at the conclusion of the inspection on June 15, 1984, and summarized the scope and findings of the inspection activities. The licensee acknowledged the findings of the inspection.

( ( 12 III.2-12 111.2-12

e Commonwealth Edison One ~llst Firs! National Address Reply 10 Post Chicago lIhnOI~ Nallonal Plaza ChIcago Posl Office I/I,no,s 60690 Ch,cago. /IIIOOIS Chicago. illinOIS OIJoee Box 767 July 30, 1984 Revision 8 April 1992 C. Keppler Mr. James G. Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Clen Ellyn, IL 60137 Glen

                                                                                              . .".. .:1" .'-:.-
                                                                                                     ~
                                                                                                          -'-~

Subject:

Dresden Station Units 1, l~ 2, .nd and 3~ '~: ......

                                                                                        .. .. --. ~ ';.;:""'"- .'

Response to Inspection Report Nos *. *..' 50-010/84-09, 50-237/84-11 and 50-249/84-10 NRC Docket Nos. 50.010 50-010,t 50-237 & 50-249 Reference (a): w. W. D. Shafer letter to Cordell Reed dated July 3, 1984.

Dear Mr. Keppler:

This letter is in response to the inspection conducted by Messrs. T. M. Tongue and S. Stasek during the period of May 22 thru June 15, 1984, of activities at Dresden Station. Reference (a) indicated that certain activities appeared to be in In noncompliance with NRC requirements. The Commonwealth Edison Company response to the Notice of Violation is provided in*the in the enclosure. If you have any further questions on this matter, please ~lease direct them to this office. uly yours,

                                                          ~. . . . -.. . 0'-

d.,~

                                                                          '?_.       ..--

D. L. Farrar Director of Nuclear Licensing 1m Attachment Cc: cc: NRC Resident Inspector - Dresden ( 8996N III.2-13 II1.2-13

Revision 8 April 1992 ( ATTACHMENT A COMMONWEALTH EDISON COMPANY RESPONSE TO NOTICE OF VIOLATION "Dresden

     . Dresden Technical Specification 4.12.0.3 states In part:
           "At least once per operating cycle, the (Cardox) system valves and associated dampers will be verified to.actuate automatically and manually.1I manually."

Contrary to the above, the Cardox system discharge master valve was not tested to verify actuation in the automatic mode and it appears that testing in the automatic mode was not included in the surveillance program. DISCUSSION During a routine NRC inspection from May 22 through June 15, 1984 of Dresden's Fire Protection Program, it was discovered that no documentatIon existed to verify that the Cardox system electro-mechanical master pilot valve was operable in the automatic ( mode. This master" master* pilot valve controls the position of the selector valves which control the flow of C02 into each of the diesel generator rooms. In reviewing Procedure DFPP 4145-1, Cardox System Semi-Annual Maintenance Test, Revision 1, it was found that the master pilot valve was tested only for manual actuation i.e., the procedure did not address a test for verifying automatic operation. Since the surveillance interval outlined in Technical specification 4.12.0.3 was exceeded, the Cardox system was immediately declared inoperable and an hourly fire inspection was established per Technical SpeCification 3.12.0.4. Also, backup fire suppression equipment was provided for these areas. A Special Procedure was written, on-site reviewed and performed to verify automatic actuation of the master pilot valve. upon Upon completion of the Special Procedure the Cardox system was returned to service and the hourly fire inspection was discontinued. CORRECTIVE ACTION TAKEN TO AVOID FURTHER NONCOMPLIANCE The corrective action taken to avoid further non-compliance was to incorporate the SpeCial Procedure for testing the automatic function of the master pilot valve into the existIng existing Cardox System Semi-Annual Maintenance Test, DFPP 4145-1. Also, a review of Dresden's TeChnical Specification Section 4.12 has been initiated to verify that all surveillance items are ( performed within their specified time intervals using approved station procedures. III.2-14

Revision 8 April 1992 ( DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED DFPP 4145-1 will be revised by August 31, 1984 Procedure Drpp and the Technical Specification review will be completed by September 28, 1984. 8996N ( III.2-15 ....

Tab 3 Revision 8 April 1992 c ( DRESDEN 2 &3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspection Report No. 50-237/85033, 50-237/85033. 50-249/85029 Page Title IlI.3-1 111.3-1 Inspection Report No. 50-237/85033, 50-249/85029 dated November 14, 1985. III.3-20 II I.3-20 December 26, 1985 Notice of Violation Concerning Inspection Report No. 50-237/85033, 50-249/85029. III.3-45 IlI.3-45 January 24, 1986 CECo letter from D. L. Farrar to J. G. Keppler (NRC) transmitting response to Inspection Report No. 50-237/85033 and 50-249/85029. '(, III.3-i

ENCLOSURE 3 UNITED STATU Revision 8

                   ...*              NUCLEAR REGULATORY COMMISSION                              April 1992 cc
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                                                     .OOUVEL T IIIIOAO OL.E'" ~L.LV"',

III0AD ELLV"-, IL.L.INOIS ILLINOIS MIJ7

                                                                        &01)7
                               ~/~?q Docket No. 50-23f'-b ,,;}

Docket No. 50-249 t\~ Commonwealth Edison Company ATTN: Mr. Cordell Reed Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen: This refers to the routine safety inspection conducted by Messrs. J. Holmes and C. Ramsey of this office on September 30 through October 21, 1985, of activities at Dresden Nuclear Power Station, Units 2 and 3. 3, authorized by NRC Overating Licenses No. OPR-19 O~rating licenses DPR-19 Ind and No. OPR-2S DPR-25 and to the discussion of our findings with Mr. D. Scott It at the conclusion of the inspection. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations. observations, and .( interviews with personnel. During this inspection, certain of your activities appeared to be in violation with NRC requirements. These issues, identified 1n in paragraphs 3 and 7.a of the enclosed inspection report, are being reviewed for potential escalated enforcement action. The results of that review will be forwarded to you by separate correspondence which will identify the nature of expected fo~l response. In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter and the enclosures will be placed in the NRC's Public Document Room. The responses directed by this letter (and the accompanying Notice) Ire are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. III.3-1

Revision 88 April 1992 ((' vommonwealth Edison Company

     ¥ommonwealth                           2 1I0V 11.* (185 IOV       *S We will gladly discuss any questions you have concerning this inspection.

Sincerely,

                                            ~~
                                            ~!1::        ,.11., Director i.ll0, "",,,,

Division of eactor Safety

Enclosure:

Inspection Reports No. 50-237/85033(DRS); and 50-249/85029(DRS} No. 50-249/85029{DRS) cc w/enclosure: D. L. Farrar, Director Licensing of Nuclear licensing Scott, Plant Manager D. J. Scott. DCS/RSB (RIDS) Licensing Fee Management Branch licensing Resident Inspector, RIll (c._. Phyllis Dunton, Attorney General's Office. Generalis Office, Environmental Control Division RI~I,J RI~~fJ. ~

      ~jt
      £;!t 11/14/85 III.3-2 I1I.3-2

Revision 8 Apr; April1 1992 (

\.                          U.S. NUCLEAR REGULATORY COMMISSION REGION III II I Reports No. SO-237/85033(DRS);

50-237/85033(DRS); SO-249/85029(DRS) 50-249/85029(DRS) 50-237; 50-249 Docket Nos. 50-237: Licenses No. DPR-19; DPR-25 licenses licensee: Licensee: Co.monwealth Edison Company P. O. Box 767 Chicago, Il 60690 facility Facility Name: Dresden Nuclear Power Station, Units 2 and 3 Inspection At: Morris, Il Inspection Conducted: September 30 through October 21, 1985 l\J.J~ CUd ~ Inspectors: PHol~es-P.Hol~es- It- ,:.;-9S-1I-/3-SS-Date bate 1(- ,.4r 114r Date bate Approved By: V. W. G. ul mond, Chief Operational Programs Section II

• Date

                                                                                '.-.1.-

Inspection Summary Ins ection on Se tember 30 throu h October 21 1985 Re orts

o. Ow reas nspected: outlne, ne, unannounced safety inspection conducted to verify the adequacy of the facility's fire protection program iIP1ementation implementation and to detenaine the status of LERs and previous open items. The inspection involved deteraine 71 inspector-hours by two NRC inspectors including 2 inspector-hours onsite during off-shifts and 11 inspector-hours conducting in-office review It at the Region III office.

Results: Of the 6 areas inspected, no violations or deviations were identified in four areas. Two violations were identified in the remaining two lreas 1n areas Paragraph 3; failure to (failure to adhere to program staffing requirements - Plragraph comply with a-license a *license condition to install an automatic lutomatic fire detection system in the Reactor Building refueling floor area - Paragraph 7a).

• c III.3-3

Revision 8 April 1992 DETAILS

1. Persons Contacted DNPS D. Adam.

Adam, Compliance' Compliance* Adlinistrator

       -J. Brunner, Assistant Superintendent, Technical Services T. Ciesla.

Ciesla, Assistant Superintendent, Operations

       -M. Dillon, Fire Marshall
       -R. Flissner, Service Superintendent
       -T. Hausheer.

Hausheer, Nuclear Services, Technical

       -P. Lau, QA Supervisor
       -J. McDonald, Station Nuclear Engineering
       -S.
       -B. Rybak, Station Nuclear Engineering
       -D. Scott, Station Manager
       -R. Whalen, Technical Staff J. Wujciga, Production Superintendent US NRC E. Hare, Resident Inspector
       -L. McGregor, Senior Resident Inspector S. Stasek, Resident Inspector

.C"" -Denotes those in attendance It at the exit leeting of October 4, 1985.

2. licensee Licensee Actions on Previous Inspection Findings a.

I. (Open) LER (237/85029) and Violation (237/85028-01): Auxiliary electric equip.ent equipilent room roOil halon systel systell declared inoperable due to ventilation dampers failing to close. Fire watch was not established per Technical Specification No. 3.12.H.2. 3.l2.H.2. Region IIIls Ill's followup of this event *is dOCUMented docUMented in Inspection Report No. 50-237/85028(OR5). 50-237/85028(DRS). As a result of this followup violation No. 237/85028-01 was issued. No response to this violation was required because the licensee's interi. and long ter. ten. corrective actions were dete~ined detenlined satisfactory., The interi. corrective actions i.p1e.ented prior to or during the followup were i.plemented fo11owup inspection. The proposed long te,.. terw corrective actions have not been 11!p1Htented. illP1aented. Therefore. Therefore, this event report reains reaains open.

b. (Closed) LER (249/85014): Wet pipe sprinkler system in Unit 3 turbine trackway had to be rerouted to all~ allow for overhead clearance for new turbine rotors. The sprinkler system was out of service 5 112 hours beyond the 14 day l;.;t li.it peraitted penlitted by Technical Specification 3.12.C.3.

3.l2.C.3. Z 2 III.3-4

Revision 8 April 1992 (\/' The event Mlport report is closed based on the licenseels licensee's corrective actions taken ~ich included restoration of the syste~ to service Ictions Ind and functional testing prior to declaring the systemsyste~ operable.

c. (Closed) LER (237/85010): Fire door for the Unit 2 12SV 125V DC battery ro~ found open. A fire watch was not established within one hour per Technical Specification 3.12.F.2.

This event report is closed based on the licensee's corrective actions taken which Ictions ~ich included ii ..ediate closure of the fire door Ind and require.ents to keep training/instruction of plant operators on the requireaents fire doors closed at all ti.es when ~en not in use.

d. (Open) LER (237/84*20):

(237/84-20): Two of seven root valves that were installed on fire hose stations were found to be in the closed position rendering the fire hose stations inoperable.

        ~lthough  the licensee's corrective actions for this event included valves. the inspectors deter.ined prOMpt opening of the closed root valves,                       detenlined that the licensee's program for administratively controlling valves that are not electrically supervised using wire seals to secure these valves in the open position and perforaing perfonling IOnthly inspections to verify valve positions does not appear to be working. During the inspection the inspectors observed several non-electrically supervised valves in the fire protection system with .issing or ~ged -nre       .nre

,( seals. To correct this problem, proble~. the licensee stated that the program for adainistratively a~inistratively controlling these valves is being upgraded to include locking these valves (chain and lock) in the open position IOnthly inspections to verify each valve position in in addition to aonthly accordance with NFPA Standard 26. This event report re.ains open pending Region III verification of the licensee's upgraded corrective actions.

e. (Open) LER (237/84-17; 237/84-05): Failure to establish continuous or hourly fire watch patrols due to inoperability of all or portions of fire detection and sprinkler alarm systems in the control room.

During these events and at the present ti.e. the fire detection and alann systel sprinkler alarm systeR printer indications are interlocked (dependent) into the plant security system ccaputer. cOlputer. Indication of fire detection allnls in the control roo. was and sprinkler allnas WIS lost in two events either because of aI loss of ~r power to the plant security systlm_cOIPuter systemcOlPuter or because of -adifications IOdifications being aade

                                            .ade to upgrade the plant security syst.. computer. Apparently. any failure of the plant security syste.

system computer can cause the 1055loss of .11 all or portions of fire alann annunciation in the control detection and sprinkler alarm roo.. ton-trol rooa. This installation does not coaply co.ply with the licensee's ca.mitaent co.mit.ent to NFPA 720 as stated in the licensee's April 1977 response (point-by-

• c
• c: point comparison) to Appendix A to NRC Branch Technical Tec~nical Position 3

II I.3-5 III.3-5

Revision 8 April 1992 rI-I.I . (SlP) (STP) APCSB APCSS 9.S-1. 9.5-1. The licensee's scheduled plant .odification No. Ml2-2/3~84-109 Ml2-2/3-84-109 identifies corrective action for this probl~ IS the installation of independent circuits for fire detection and Ind sprinkler system alarms which alarm Ilarm and annunciate in the control Iccordance with HFPA room in accordance NFPA 720. This .edification

                                                      .adification is scheduled to be ca-pleted c~leted in December 1985.

19S5. These lERs LERs will rela;n relain open ~~nding

                                                                          ~Inding Region III verification OTor the licensee's corrective actions.

f. (Closed) lER LER (237/84-11): Fire wall penetrations to Unit 2/3 diesel generator rooms were not sealed.

This event report is closed based on the licensee's corrective taken, which include establishment actions taken. establishlient of a fire watch within 3.12.F.2 and sealing the one hour per Technical Specification No. l.12.F.2 penetrations per drawing No. 12E-6058. .

g. .(Closed) LER (237/84-08): NRC inspection of the licensee's compliance with fire protection Technical Specification surveillance requirements identified that a cardox system .aster valve was not

                                           .ade. The .aster valve test procedure being tested in the automatic .ode.

was written to test the valve in the .anual .ade. This event report is closed based on the licensee's corrective actions taken which included prompt removal of the .aster valve from service, revision of the surveillance test procedure. procedure, and c h. satisfactory testing of the valve in the autoaatic (Closed) lER LER (249/83-34/03L): auto.atic .ade. (249/S3-34/03L): Unit 3 trackway sprinkler system out of service due to damage by .abile crane boo..bODe. This event is closed based on the licenseels licensee's corrective actions taken, which included .. tlken, .aking necessary system repairs, PrGlpt king the nece5sa~ proapt restoration of the system to service and instructions to plant

                                    .ave.ent of .obile personnel regarding the 80veaent        .abile cranes and the fragility of systems and components in their path.

i. (Closed) LER (249/83-17/03L):

(249/S3-17/03L): HPCI deluge system solenoid valve taken out of service because the valve would not reset. This event report is closed based on the licensee's corrective actions taken, which included prompt removal of the deluge system

                          .aking from service, .. king the necessary repairs to the solenoid valve and restoration of the syste.

system to service.

j. (Closed) LER (237/81-1S/03l):

(237/S1-15/03L): Unit 2/3 diesel generator rooerooa CO 2 system heat detector surveillances not perfor.ed per Technical 4.12.A.1. Specification 4.12.A.l. This event report is closed based on the licensee's corrective actions taken, which included instruction to plant personnel to perform the required heat detector surveillances and satisfactory performance of the surveillance. 4 111.3-6 III.3-6

Revision 8 April 1992

k. (Closed) Violations (237/81-09-01; 249/81-06-01):

249181-06-01): Four penetration seals identified IS as being defective were inoperable for an excessive period of time. Neither prompt nor timely corrective action was taken. This item is closed based on the licensee's June 29, 1981 response to Region III ~ich which discussed the licensee's corrective actions taken to avoid fut~re taten futwre violations in this area. The inspector's review of Procedure No. DFPP-417S-2. DFPP-4175-2, Revision 4, indicated that appropriate instructions are provided to plant personnel which refer to detail drawings for proper installation of penetration fire seals.

1. (Open) Violations (237/81*09*03; (237/81-09-03; 249/81-06-03):

249181-06-03): Ca) (a) Fifty percent of fire extinguishers sampled did not have 1981 .anthly inspection tags attached; (b) 5 year hydrostatic test for portable CO 2 extinguisher cylinders were overdue; and (e) (c) nuaerous nUierous cOIpressed cOlpressed

      .gas cyl   inders were illproperly cylinders          improperly stored.

The licensee's corrective actions identified in their June 29. 29, 1981 response to items (a) and (b) of this violation were ineffective. Subsequent QA audits and surveillance by the licensee's onsite QA department have revealed that these deficiencies are continuing. for For example. example, deficiencies identified in QA surveillance No. QAS 12*85~236 for the period September 23 through 29. 12-85-236 29, 1985 include the following: wrong date on extinguisher tags; extinguishers past due for 5 year hydro testing; no service date on extinguisher tags, tags; no ,( seal on extinguisher pull pin; partially discharged extinguisher. During plant tours by the inspectors, identical deficiencies were Observed. observed. In one instance, a CO 2 portable extinguisher hose was d.-aged to the extent that the webbing in the hose was exposed .* A hole existed in the webbing that .. aayy have allowed the extinguishing agent to escape through the hose prior to reaching the CO 2 di$charge discharge nozzle. This extinguisher was located on fire cart No.2. In addition, the inspectors observed that wheeled d~ dry chemical extinguishing units Nos. PK 21 and PK 22 had tags which indicated that surveillances were .issedmissed the .onths months of May and Septe.ber 1985. The continuing existence of this type of deficiency is indicative of a lack of .anagement management attention in this area. Manage.ent Management attention Ittention and Ind staffing is the subject of aI violation documented in paragraph parlgraph 3 of this report. Your response to that violation should address your corrective actions for fliling failing to properly aaintain

                                                                 .. intain fire extinguishers. These itells items will ralin relain open pending the further review of the licensee's corrective actions by Region 111.      III.

Ite. (c) of this violation is closed based on the licensee's Item corrective action Iction taken which included the installation installition of 8etal storlge racks for compressed gas cylinders, securing the cylinders storage with .. tal chains and 8etal Ind revision of Procedure No. DAP CAP 3-11. M. (237/81-09~06; 249/81*06-06): (Open) Unresolved Item (237/81-09-06; 249181-06-06): Fire brigade drills and training do not appear to leet 8eet the intent of NRC requirements. requi rements. ' t 5 II1'.3-7

Revision 8 April 1992 c Section 6.0 of the original fire protection SER, SER. dated March 1978, reco.mended that the licensee's administrative controls follow the recOMmended 1978. guidelines set forth in the NRC Guidance Docu.ent entitled -Nuclear

                                                            ~inistrative Responsibilities. Administrative Plant Fire Protection Functional Responsibilities, Controls and Quality Assurance." A supplaent supplnent to the original SER suppl~ntal SER was issued December 2. 1980. Section 3.1 of t~is suppleaental closes out the issue of administrative controls with the NRC staff's adiinistrative controls acceptance of the licensee's discussion of adlinistrative provided in letters dated January 24,       February 24,

24. Februa~ 24. March 20 and July 27,

27. 1978, 1978. January 31,

31. and April 30,

30. 1979. Therefore. Section III.l.3.b of Appendix R is not applicable to administrative controls for fire protection at Dresden.

Based on the licensee's submittals discussed above. the NRC staff concluded that the licensee's administrative controls for fire protection aet let NRC guidelines and, and. the applicable regulatory

 -requirement for fire protection administrative controls at Dresden iaplementation of General is the Commission's guidance issued on the i~lementation Design Criterion 3 of Appendix A to 10 eFR   CFR Part 50 for existing power plants.

Section 2.0 and 3.0 of Attachment No.2 to NRC Guidance DocUient "Nuclear Plant Fire IINuclear Fhe Protection Functional Responsibilities, Admi ni strat i ve Controls and Quality Assurance" requires Administrative requhes practice sessions be held for fire brigade lembers to provide each brigade

 .ember with experience in actual fire extinguishlent and the use of emergency breathing a~aratus under strenuous conditions. Fire perforled 50 brigade drills are required to be perfonled       so that the fire brigade can practice as a team. The drills are to be perforled at regular intervals but not to exceed three .onths for each fire brigade. The drills are required to be critiqued to Issessassess each brigade .elber's leiber's knowledge of his role in fire fighting strategy.

The licensee is not leeting these requirements for the following reasons: (1) By attempting to .eet the requirements contained in Section III.I.3.b III.l.3.b of Appendix R to 10 CFR 50, the licensee has been conducting one fire drill per .anth with the intent of getting all designated fire brigade .embers lembers involved in at least two drills per year. - (2) Practice sessions that provide each brigade .ember lelber with actual Ictual fire extinguishment experience and the use of emergency eeergency breathing apparatus under strenuous conditions (full fire fighting gear) have not been conducted due to a breakdown in contractual arrangements with an In independent firm. fir.. (3) Fire brigade drills have not been critiqued at three year intervals by qualified individuals independent of the licensee's staff. 6 III .3-8

Revision 8 April 1992 ( To resolve this concern, the licensee is requested to aa~e aake Iyaillble available

                *a detailed .ssess~ent assessment of fire brigade drills, practice sessions and individuals three year audits of fire brigade drills by qualified 1ndividuals independent of the licensee's staff. This assessaent should establish whether the licensee is in cOlplilnce c~pliance with cOImit.ents cOimitaents aade to the NRC which resulted in the NRC staff conclusions that
                      ~icensee'5 administrative controls for fire protection were the 1icensee's acceptable.                                           .

remains open pending region review of the licensee's This item re~ains assessment.

n. (Open) Unresolved Item (237/81-09-07; 249/81-06-07): Specific*

pre-fire fighting plans or strategies for all safety-related areas areas presenting a hazard to safety-related equipment were not and Ireas Ind developed and iaplemented.

               "As
               *As discussed in item 237/S1-09-06; 237/81-09-06; 249/81-06-06 above, the III.K.ll and 12) requirements of 10 CFR 50, Appendix R (Sections JII.K.11 are not applicable in this case. The applicable requirements are contained in Attachment No. 5 of NRC Guidance Doc.-ent Document -Nuclear
                                                                              -Nuclea; Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance.

Assurance."1I During the inspection, the licensee provided the inspectors .nth with a copy of pre-fire plans that contained specific fire fighting

( ~" .

strategies for fighting fires in all safety-related Ireasareas and areas that present a hazard to safety-related equipment. The pre-fire plans appear to provide adequate fire fighting procedures iap1e.ented. and instructions. However, these plans have not been i~leaented. iap1emented and According to the licensee, the plans will be iaplelented incorporated into fire brigade training lesson plans by the end of the first quarter of 1986. This item will remain open pending iapl elltntati on. said i~lementation. .

o. (237/81-09-08; 2~9/S1-06-0S):

(Open) Unresolved Item (237/S1-09-0S, 249/81-06-08): Fire brigade practice sessions have not been conducted in accordance with cOimitMents aade to the NRC. A hands-on practice session was held co.mitMents in 1979 with full brigade attendance, but no practice session was held in 19S0. 1980. This item will remain open pending Region III review of the licensee's response to Item b of Unresolved Item No. 237/81-09-06; 249/81-06-06 as discussed in this report.

3. Fire Protection Program Organization and Personnel Staffing 10 CFR 50.48 requires that each operating nuclear power plant have a fire protection plan that satisfies Criterion 3 of Appendix A to 10 CFR eFR 50.

Except for the requirements of Section III.G, III.J, and 111.0 of Appendix R to 10 CFR SO, sltisfies 50, the approved fire protection plan that sitisfies Criterion 3 of Appendix A to 10 CFR 50 is discussed in the original fire ,( SER, dated March 1978. protection SER. 1978, a fire protection SER S~pl S~plement,

                                                                                ..ent. dated Dectlber 2, 1980, and th~ licensee's Fire Hazard Analysis submittals entitled Deceeber 7

III.3-9 III .3-9

Revision 8 April 1992 ( "Inforaation Relevant to Fire Protection Systellls and Programs" Systems Ind Programs" dated October 1976, January 1977, and April 1977. Furthenaore, the licensee ca.i tted to follow ca.itted fo 11 ow certain certa i n NRC Supp.lellental Supp.l ellenta 1 Guidance Gui dance Docl.Iftents Docl.IIIents as discussed in letters to the NRC, dated January 24, Februa~ 24, March 20 and July 27.27, 1978; January Janua~ 31 and April 3D, 30, 1979. Th~ require.ents for overall responsibility for the Fire Protection require-ents Program are discussed in Sections IV.A and l.I.A.I 3.1.A.I of Parts 1 and 3 of the licensee's Fire Hazard Analysis submittal, dated October 1, 1976 and April 1977. The NRC's position, IS as restated stated in Section l.I.A.13.I.A.l of this document establishes guidance on i.ple.entation of basic criteria for fire protection program organization Ind and personnel staffing. In response to the NRC's position discussed in Section A.l A.I of Appendix A to NRC Branch Technical Position APCSB 9.5~1 9.5-1 concerning the qUllification qualification requirelents for the Fire Protection Engineer who will assist in various require.ents aspects of fire protection program development for the operating plant, licensee states "comply" in Section 3:I.A.l the Hcensee 3:I.A.I of the Fire Hlzard Hazard Analysis submittal. The licensee further states, in part, "CECo *CECo has 1a Fire Protection Coordinator who reports to the Supervisor of Safety *** ** Responsibilities of the Fire Protection Coordinator are: coordination of activities; procurement of equipment, resolve questions on standards recOimendations for iaprov~nts; and technical issues; .ake reca.mendations i~rov~ntsi coordinate. coordinate, plan. plan, and conduct inspections (make inspections of Dresden, Units 2 and 3. 3, once a .anth); IIOnth); ensure that adequate fire fighting equipment is provided c( . and that such equipeent equipment is .aintained in good operating condition. coordinate with off5ite offsite fire depart.ent; conduct nor.al testing; provide forms publicatiqns condition, noraal and preoperational forss and instruc~ions for reporting fires; issue eBployee policy and procedures in fire protection; publicati~ns outlining e.ployee assist and supervise training of personnel; assist and advise departMents concerned with established rules and standards; coordinate with the staff

         ..tters of autual all .atters       .utual concern and .ake IIIke final reca.aendations recOlllendations for specific Ictions actions to be taken on fire protection issues."

The inspectors identified the following examples of the licensee's failure to consistently and effectively comply with the staffing requirellents for fire protection program iaplementation: require.ents implementation: I.

a. Fire Protection Engineer A qualified Fire Protection Engineer was not involved in the development of certain aspects of the fire protection program developaent for the operating plant IS as required by Section 3.l.A.l 3.1.A.I of the Hazard Analysis SUbmittal.

licensee's Fire Hazlrd sUbmittal. The qUllifications qualifications for this individual were not stated in any docuaent. docUient. The resuae resUie of the individual performing the original Fire Hlzard Hazard Analysis is contained in Attachment 2A of the Fire Hazard Analysis submittal, but this individual is no longer ~loyedemployed by the licensee. According to the licensee, there was WlS 1a contract with M&M Protection Consultants which included services that would satisfy SOMe some of the responsibilities of the Fire Protection Engineer, but this contract 'C expired in December 1984. 8 III.3-10

Revision 8 April 1992 Although the licensee has employed another qualified Fire Protection Consultant firm to do some specific fire protection work relative to

• upgrading the fire protection p~ogram, this firl firm was not retained to fulfill all of the responsibilities of the Fire Protection Engineer.

b. Fire Protection Coordinatnr perfonaing all the duties The Fire Protection Coordinator was not perfor.ing at the site that are delineated in Section 3.1.A.1 of the licensee's sUbmittal. According to the licensee's staff.

Fire Hazard Analysis submittal. staff, the individual that was originally assigned these duties was trans-ferred to Corporate Quality Assurance sa.e vacated, SOle ti~ ago. Once vacated. this position was not filled. The duties and responsibilities of the position were delegated to the Fire Marshal and other individuals CECo organization. within the CECa Through Amendment No. 86 to Facility Operating License No. DPR-19 (Unit 2) and Amendment No. 79 to Facility Operating License No. DPR-2S DPR-25 (Unit 3), the NRC accepted a proposed licensee staffing change. Figure 6.1-1 (Corporate and Station Organization Chart) shows a Fire Ins~ector reporting to the Corporate Dlrector of Quality Protection Inspector Assurance Operations. The inspectors requested, but the licensee did docu.entation to verify that the NRC not provide the inspectors with docUientation was aware that the same individual who was the site Fire Protection Coordinator was filling the position entitled -Fire "Fire Protection Inspector"ll for Corporate Quality Assurance. Inspector

• (

The licensee's failure to adhere to the staffing require.ents requi~nts prograMmatic breakdowns that have discussed above resulted in prograMmatiC decreased the level of fire protection that was intended to satisfy Criterion 3 of Appendix A to 10 CFR SO. 50. For exa.ple: exaaple: (1) A fire detection system was not installed on the refueling floor as required by Aaendment No. 33 to Facility Operating License No. DPR-25. (This is discussed in Paragraph 7.a of the report.) (2) Installed fire protection hardware and equipment was not being properly .. intained. (This is discussed in Paragraphs alintained. 2.d, 2.e, 4, 5, and 7 of the report.) (3) Technical specification surveillance procedures did not incorporate appropriate testing of quality affecting par~ters para.eters in accordance with design and governing code requi~nts. require.ents. (Thi sis (This is" di scussed in Paragraph 4 of the report.) ". discussed (4) Administrative controls did not adequately control fire protection features. (This is discussed in Paragraph 5S of the report.) (5) Many deficiencies that were identified in LERs, NRC inspections, QA audits, and QA surveillances did not receive ProlPtpro.pt or effec-( tive corrective action. (This is discussed in Paragraph 2 and 6 (This;s of the report.) 9 III .3-11

Revision a8 Apri 1 1992 c ( (6) Weaknesses in the scheduling of fire drills were identified. (This is discussed in Paragraphs 22.1, .* , and 2.0 of this report). Failure to comply with the staffing requireaents require.ents for development and i.ple.entation ilpl~ntation of the fife fire protection program progrlll is considered a violation of 10 CFR 50.48 and Criterion 3 of Appendix A to 10 CFR 50 249/85029-01(DRS>>. (237/85033-01; 249/85029-01(ORS>>. The Station Fire Marshal's qualifications include 58 junior college credits in fire science; an In associates degree in electronics engineering and 15 years experience as a volunteer firefighter. He has held the position of station fire aarshal for seven years. At the present ti.e. time, the fire aarshal larshal is assigned the following responsibilities:

a. Coordinate and assist in fire systems systees periodic testing.

b. Plan, coordinate, conduct, and critique fire drills.

c. Fire Brigade 8rigade classroom training.

d. Review, revise, and write new administrative procedures.

e. Review, revise and write new surveillance procedures. Make Review.

work requests to repair deficiencies, verify surveillances are completed as IS required and .aintain laintain files on coapleted cDepleted (~: C surveillances.

f. aodifications, assist in training, testing.

Review plant aodifications. testing, and development of procedures.

g. Maintain fire equipment, verify availability of spare parts and procurement of parts.

h. Technical Specification Participates in insurance inspections, Technica' Reviews, QA, IMPO, Ind NRC audits.

INPO, and

i. Assure Technical Specification compliance.

j. Review work requests.

k. Verify fire watch and insurance notification.

1. depart8ent.

Coordinate activities with the offsite fire departaent. II. Make reports on deviations and fire daaage experiences.

n. Perfo~ plant cleanliness inspections.

o. Correspond with other agencies on fire protection issues.

  .. ,  pp.. Assure that the fire protection program .eets  aeets NRC and other rellents.

requi rements. {C. 10 III .3-12

Revision 8 April 1992 ( q. Explain fire protection requirements to the licensee's staff when requ required. 1red. According to the licensee's staff and Station Nuclear Engineering Department (SNED) (SHED) procedure number PE Q.44, a qualified corporate fire protection engineer reviews new plant IOdificatfons

                                                          ~dificatfons prior prfor to implementation i~plementation by the Architect-Engineering fi~. This appears to be the extent of the corporate fire protection engineer's involvement.

The qualifications of the Station Fire Marshal do not Ippearappear to be commensurate with the list of responsibilities Issigned aSSigned to that position. This lengthy list of responsibilities constitute Ia workload that

           ~y not be achievable by Ia single individual. regardless of the individual's qualification and experience.

To resolve this concern, concern. the licensee is requested to provide It at site, a written evaluation (complete work study) of the responsi-the site.

          .bilities assigned to the station fire ~rshal. This evaluation should make a determination of the fire marshal's ability to effectively achieve each delegated responsibility based on his qualifications and time constraints.

This is considered an Unresolved Item (237/85033-02; (237/85033-02. 249/85029-02(DRS>> pending Region Ill's review of this evaluation.

4. Technical Specification Surveillance Review

. ( Technical Specification 6.2.A.ll 6.2.A.11 requires that detailed deta.11ed written wrnten procedures be developed, approved and adhered to for implementation of the Fire Protection Program. The fnspector's inspector's review of the licensee's surveillance procedures and test results for fire protection Technical Specification surveillance requirements resulted in identification of the following discrepancies:

a. Testing of Diesel Fire Pump at least Least Once Per Operating Cycle 4.12.B.l.(e) of Technical Specification No. 3.12.8 Section 4.12.B.I.(e) 3.12.B requires that the station diesel fire pumps be demonstrated operable by per-fonning forming a system functional test which includes simulated si~lated lutomatic automatic actuation of the pumps throughout their operating sequence. The licensee's commitment in Section 3.5.E.2 of the Fire Hazard Analysis Report dated April. 1977. requires the fire pump installations installatfons to conform to NFPA standard No. 20. This camtftllent confonn commitnent states that a plant .adiffcation modification would provide an adequate flow ilgegage for full flow testing of the pumps in accordance with NFPA standard 20. The licensee's surveillance procedure Nos. DFPP 4124-3 and DFPP 4124-4 were deficient in that:

(1) The procedure required ~nual throttling of the pumps to achieve the specific flows contained in Technical Specification 3.12.8. 3.12.B. and did not address automatic activation. activatfon. (2) The procedures required testing the pumps to the specific head and flow contained in the Technical Technfcal Specification No. 3.12.B, 3.12.B. 11 III .3-13

Revision 8 April 1992 c( but failed to require testing for head and flow as specified in NFPA 20. (3) Measurement of quality affecting para.eters such IS as PUlP vibration under full flow conditions were not included in the test procedure or the test results. (4) The test results were not coapared cDipared to the original aanuflcturer's aanufacturer's shop test curve or field acceptance test for the PUlps PUlPs because neither of these curves were available to the licensee's staff.

b. Testing of Water Suppression Systems It at least Once Per Operating Cycle Section 4.12.8.1.(e) of Technical Specification No. 3.12.8 requires that fire suppression water systems be demonstrated operable by

        'performing a system functional test which includes siaulated lutomatic automatic actuation of the systems throughout their operating sequence. The licensee's commitment in Section 3.S.E.3 of the Fire Hazard Analysis Report requires that automatic sprinkler systels systeas confora to NFPA Standard No. 13.

The licensee's surveillance procedure No. SP 84-6-39 failed to incorporate appropriate test requirements to demonstrate the sprinkler system is operable in accordance with NFPA 13 in that: ( (,,' (1) The procedure did not require flow from the inspector's test valve of wet sprinkler syste.s. systems. Instead, the alar. alara bypass valve was used for this test. (2) The procedure did not require flow fro. the two ;nchinch drain valve of wet or d~ dry syste.s. systems. Instead. Instead, the alara bypass valve was used for this test.

c. Semiannual Testing of Fire Detectors Section 4.12.A of Technical Specification No. 3.12.A requires that delOnstrated operable by perfor.ing the fire detection system be deaonstrated perforaing a channel functional test every six aonths. The licensee ca.mit.ent coamitaent in Section 3.S.E.1 3.S.E.l of the Fire Hazard Analysis Report requires that the fire detector system conform confo~ to the requireMents requirelents of NFPA Standard 720.

72D. The licensee* licensee'ss surveillance procedure No. DFPP 4185*2 4185-2 (Revision 4) failed to incorporate the following quality affecting para.eters paraaeters as re'qui requi red by NFPA 72D: (1) Periodic cleaning of detector units. adjustaent for sensitivity (Section 3.1.2 of the (2) Periodic adjust.ent original SER required this test to be conducted). ,( 12 III .3-14

Revision 8 April 1992 ( According to the licensee's staff. staff, an independent fire protection consultant has been employed to review .11 III technical specification adeqUAcy in accordance procedures and test results to evaluate their adequacy with NFPA standards and design requirements. This Issess.ent assess~nt was in progress at the ti.e of the inspection and is expected to be cOIpleted cOlpleted by the end of 1985. According to the licensee, where necessa~. necessa~, the procedures will be revised to coincide with the governing code and design requirements. This is considered an Open Ite~ Item (237/85033-03i (237/85033-03; 249/85-029-03(DRS>> pending Region Ill's review of the licensee's actions. No violations or deviations were identified.

5. AdministratIve Controls - Control of Welding.

Administrati.ve Welding, Cutting. Cutting, and Ignition Sources Licensee procedure No. DAP 3-11 (Revision 4) contained what appears to be acceptable instructions for controlling storage of flalmable and Ind liquids, storage of compressed gas cylinders, and accUiulation combustible liquids. accuaulation roabustibles such a5 of rubbish and transient rOibustibles as wood scaffolding, etc. The procedure specifies housekeeping and cleaning responsibilities to be followed by all employees and contractors. No violations or deviations were identified in this area, however; the inspectors cautioned the licensee on a proposed revision to welding and cutting procedure No. DMP 4100-1 that would include a provision to facilitate ALARA concerns in high radiation areas. The inspectors infor.ed the licensee that any relief fro. the require.ents requi~nts for a firewatch to remain in the i~ediate ilmediate area thirty .inutes ainutes after cutting and welding has been completed would have to be discussed with NRR.

6. Quality Assurance Program The licensee's commitment to Quality Assurance for fire protection is documented in Section 3.3 of "Inforaation Nlnfor.ation Relevant to Fire Protection Programs"U and in letters to the NRC on this subject dated Systems and Programs Systells Janua~ 24, February 24, March 20, and March 27. 27, 1978, Janua~ 31 andInd 3D, 1979.

April 30. The inspectors review of the licensee's Quality Assurance Program for Fire Protection included review of the following:

a. Eleven criteria applicable to fire protection that satisfy Technical Position 9.5-1 and suppleaent Appendix A to Branch Technica' suppllaent guidance "Nuclear Plant Functional Responsibilities, Adlinistrative Assurance."11 Controls and Quality Assurance.

b. Septeaber 3-6, 1985, Quality Assurance Surveillance Reports dated Septeiber Septelber 16-30, 1985.

September 5-9, 1985, September 9-13, 1985, and Septeaber

c. Annual Quality Assurance Audits Nos. QAA 12-84-1 dated April 17,

{( ,.,./ .. 1984, and Q~~ Q~A 12-83-1 dated April 15, 1983.

  \\.

13 III.3-15

Revision 8 April 1992 c" d. Triennial Audit by M&M Protection Consultants dated December"C, No violations or deviations were identified. December 4, 1984. identified, however, the inspectors suggested to the licensee that for clarification, the stltelents made fn statenents ~de in Section 3.3 of the *Infonmation

                            *Information Relevant to Fire Protection Systems and Programs* should be IOdif1ed IOdified to indicate their specific commitment conmitlent to *a QA prograll to fire protection. As written, this statement cln be interpreted to mean Ilean that the licensee committed to .pply Ipply all III of the criteria of Appendix B in 10 CFR 50 to fire protection.

The inspectors detenmined determined that the licensee's practice of considering fire protection as reliability-related is acceptable because this practice ensures that ,11 all of the eleven criteria contained in the NRC's Guidance are included in the program. In addition, this practice allows for the nonmal normal QA program for safety-related systems to be applied to fire it's entirety. Only one QA ~nual protection in 1t's manual exists for re'1ability~ reliability-related systems and fire protection systems. Although the licensee's Quality Assurance Program appears to be effectively identifying issues that Ire are contributing to hardware and programmatiC programmatic weaknesses. weaknesses, the licensee does not appear to be taking prompt and effective corrective .etions. actions. This is exemplified by the remaining open items that have been identified in QA audits and surveillances. surveillances, LERs. LERs, and NRC inspections. (This is further discussed in 3.b.(S) 3.b.(5) of the report.) report. )

• c 7. Plant Tours During tours of the plant, the inspectors observed the following deficient conditions:

a. Failure to comflY Com~'Y with License Condition No. 2.8.2.B. of Amendment No. 33 to Fae; o~erating License No. DPR-ZS Faci ity 0ferating DPR-z5 and Amenament Amendment No. 36 to Provisiona Operat1ng License No. DPR-19.

Operating license Section 5.1.6.6 of the original Fire Protection SER for Dresden Units 2/3 dated March 22, 1978 states that the lfcensee licensee proposed the installation of an automatic fire detection system to provide early warning of a fire in tn the Refueling Floor Area in order to satisfy the objectives of Criterion 3 of Appendix A to 10 CRF 50. Amendment No. 36 to Provisional Operating license License No. DPR-19 (Unit 2) and AMendment No. 33 to'to* Facility Oper.tfng Operating license License No. DPR-25 (Unit 3) dated October 1. I, 1980, require that the early warning automatic fire detection system for the refueling floor lrea area be installed by start up following the 1979 Un1tUnit 3 refueling outage. As of the date of this inspection (approximately stx six years after

                                                                               .fter start up following the Unit 3 1979 refueling outage) the licensee

• has failed to comply with the provisions of Amendment No. 36 to License No. DPR-19 and Amendment No. 33 to Provisional Operating license Facility Operating License No. DPR-25. An early warning lutomltic F.cility automatic fire detection system fire detection system has not been installed 14 III .3-16 III.3-16

Revision 8 April 1992 ('

\,

in the Refueling Floor Area and no compensatory IeASUreS ~asures have been taken as a result of this decre~sed decreased effectiveness of the plant's fire protection features. The installation of an In automatic early warning fire detection system in the refueling floor area was not discussed in any of the licensee's correspondence to the NRC that thlt requested amendments llendments to

      .odify
      ~dify the plant's fire protection Technic.'

Technfcil Specifications Specfffcatfons to incorporate Li~iting Limiting Conditions for Operation Operatfon and Ind Surveill.nce Survefllince Requirements for the fire ffre protection ~iflcatfons modifications required requfred by the original SER for Dresden Units 2/3. None of the proposed Tables 3.12.1 to Technical Specification Specificatfon 3.12 listed fire ffre detect10n detectfon instruments 1n in the refueling floor area. However, sufficient sufffcient 1nfonmation infonnation existed which should have alerted the licensee lfcensee that he was 1nin violation vfolation of a license condition. For example:

     '(1)   By  letter dated February 25, 1980 (R. F. Janecek-CECO to T. A. Ippolito-NRC) the licensee noted that they did not believe installation of an    In automatic early warning fire detection system in the refueling floor area ~s d~tection                                             was warranted based on low fire loading and the ability Ibilfty to ~kemake up wlter water and Ind  cool  the   spent  fuel  pools in the  event  of  *a loss of e1ther either Unit's spent Fuel pool cooling equipment due to fire. This letter did not request relief from the installatfon installation of a refueling floor fire detection system. Ho       No officill official NRC response was issued for ~his  this letter. ,

By letter dated March 18. 18, 1980 (L. Derderiln-NRC Derderian-NRC to M. Antonetti - Gage Glge Babcock and Associates - Consultants to the Licensee) the NRC referenced a March 17. 17, 1980 telecon record with T. Pickens (CECO) in which the followfng was agreed following WIS Igreed to concerning Reactor Buflding Building Refueling Refuelfng floor fire dete~tion detection systems for Dresden Units 2/3 and Quads Cities Units 1 and 2: (a) The license was to confirm to the NRC that in the .ast ~st heavy fire loading situations (f.e.(i.e. refueling periods). periods), the loading would not exceed that necessary to cluse cause structural failures. (b) The licensee was to confirm that structural concrete protection extends from the floor to some specified lessening the likelihood of structural faflure. height, lessenfng failure. (c) The licensee lfcensee was to recalcullte recalculate Iverage average combustible loading subtracting out the pool areas.lreas. The lfcensee licensee could not provide the inspectors with documented evidence that these issues were ad~ressed. This failure to fol1owup followup on implementation of a license condition is indicative of a programmatic breakdown which has resulted in , Co' a reduced level of fire protection than was intended to sitisfy satisfy 1~ III.3-17

Revision 8 April 1992 c criterion 3 of Appendix A to 10 CFR SO 50 and is considered Ia violation of Amendment No. 36 to Provisional Operating license License DPR-19, Amendment No. 33 to Facility Operating license No. OPR-19. License DPR-2S, 10 CFR SO No. DPR-2S. 50 (237/85-033-04; 249/85-029-04)(ORS). 249/85-029-04)(DRS).

b. ~riparati.9.!l~

         ~rvparati2~J     for    th~EOOling th~~Eoming               ~nit 3 Outage Separation of Extended !Jnlt n t-rTrom t""TTrom lint lin its t5  2G .

2~ - During plant tours and in meetings with the licensee during the inspection, the licensee agreed to update their response to the NRC inspection. and describe the administrative controls Ind and the Ictions actions that will be necessary to isolate Unit 1 from Units 2 Ind and 3 s1nce since Unit 1 is no longer operational but shares common areas with Units 2 and 3. The inspectors also requested that the licensee describe those administrative controls Ind

       .*administrative                 and Ictions actions that will be necessary to separate common areas in Units 2/3 while Unit 2 is operating Ind            and
       *Unit 3 is in an extended outage.

This is considered an Open Item (237/85-033-05; 249/85-029-05)(ORS) 249/85-029-05)(DRS) pending further review by Region Ill.

c. Self Contained Breathing Air Supply for the Fire Brigade Section 3.4.D.4{h) 3.4.D.4(h) of the document entitled *Infonmation

                                                                 -Information Relevant to Fire Protection Systems and Programs*,Programs", requires that breathing c         apparatus using full flee   face piece positive pressure .asks approved by NIOSH be provided for the fire brigade.

Masks that Ire The inspectors examined the fire brigade Scott *Air Pak breathing lir air cylinders that were provided on Fire Chart No.2. Four out of four of these cylinders contained 1800 pounds of air pressure. According to the licensee's staff, a ~lnimum ~inimum of 2200 pounds of air lir pressure should be contained in each cylinder. 2400 pounds of afr lir pressure would indicate the cylinder is full and Ind ~y provide a 30 ~inute air supply for the average fire brigade .ember. lember. The cylinder gauges have a range of up to 3000 pounds of air pressure. A December 1984 three year audit reconnended rKOIIIIIended that a set of written instructions be provided It at the breathing lir cylinder fillfngfilling station to assure that the cylinders Ire are filled properly. Filling of the cylinders fs is the responsibility of Health Physics. Due to time constraints. constraints, the inspectors were unable to contact Health Physics to follow up this concern. Therefore. Therefore, thelfcensee the licensee is requested to provide at It the sfte site the appropriate acceptance Icceptance criteria for f1111n9 fillin9 breathing Ifr lir supply cylinders. This is considered In Open Item (237/85-033-06; 249/85-049-06)(DRS) pending Region III licensee's breathing afr review of the licenseels lir cylinder ffl11ng filling procedures.

d. 300 Pound Fixe9 Fixe~ Cardox System Supply Tank First Floor, Turbine

         'BiiTIiIing TulTding

• c . During plant tours, the inspectors observed the following deficiencies on the ~in Co, CO? system storage tank located on the first floor of the turb1ne buildings.

16 III .3-18

Revision 8 April 1992 ( (1) The access door to the tan~ cu.pres50r cuaprtssor .ator was .fsling.

                                                                          .issing.

(2) The glass cover to the tan~'s tan~ls aercoid

                                                      ~rcoid switch located inside the access door was .;ss;ng.
                                          .issing.

The licensee had no explanation for these deficiencies, ~ut agreed to take ta~e i.-ediate corrective actions. This is considered an Open Item (237/85-033-07; 2.9/83-029-07)(ORS) 249/83-029-07)(DRS) pending fUrther further verification of the licensee's corrective actions by Region III.

8. Open Items Open items Ire aatters which have been discussed with the licensee, which are .atters will be reviewed revi~d further by the inspector, and which involve SOle SQIe action on the part of the NRC of licensee of both. Open iteas disclosed during the inspection Ire are discussed in Paragraphs 4.c, 5.1.

S.a, 7.b, 7.c, 7.d.

9. Unresolved Items Unresolved items are .atters aatters about which .are

                                                       ~re infonaation infor.ation is required in order to ascertain whether they are acceptable ite.s, itels, ite.s itels of noncoapliance, or deviations. An unresolved itel nonca.pliance,                                  itt. disclosed during the inspection is discussed in Paragraph 3.c.

e--- (

\.'L'"
\."'"

10. Exit Interview The inspectors .et let with the licensee representatives at the conclusion of the inspection on October 4.

4, 1985, and su..arized the scope and findings of the inspection. The licensee acknowledged the statelents

           .ade by the inspectors. The inspectors also1150 discussed the likely infor.ational content of the inspection report ~thwith regard to docUients docuaents reviewed revi~d by the inspector during the inspect;on.

inspection. The licensee did not identify any such docu.ents docuaents as proprietary. On October 21, 1985, in a telephone conversation with the licensee additional concerns regarding the lack lick of fire detectors on the refueling floor were discussed with the licensee. c. 17 III.3-19

B~033 / PoC:;,-,7C, Revision 8 UNITED STATES April 1992 NUCL.EAR REGULATORY COMMISSION REGION III 711 ItOOSEVIELT "OOSEVELT !tOAD "OAO eLlEN CLEH ELLYN, ELL. VH, ILLINOIS 50U7

                                                            '0127 DEC 22SS 1985 J9S5 Docket Dccket No. 50-237 Docket No. 50-249 Commonwealth Edison Company                                     854943 ATTN: Mr. Cordell Reed Vice President Post Office Box 767 Chicago, Il IL 60690 R£l!tD R£I!'D DEC 27198S 271985 Gentlemen: .

14, 1985, (cop~ enclosed) Region III transmitted to By letter dated November 14. you Inspection Reports 50-237/85033(DRS) and 50-249/85029(DRS). These reports documented the results of an unannounced safety inspection conducted between September 30 and October 21, 1985. 1985, to establish the adequacy of fire protection program implementation Itat Dresden Nuclear Power Station. Station, Units 2 and 3. As discussed in the November 14 letter. letter, two issues were identified during the inspection which were under consideration for escalated enforcement actions. c These issues were failure to install fire detectors on the refueling floor as required by license condition and failure to effectively implement the fire protection program in accordance ~th with 10 CFR 50.48 as evidenced by numerous cases,.recurring and, in some cases *. recurring deficiencies. On November 19, 1985. 1985, an enforcement conference was held in the Region III office with you and members of your staff to review these issues and obtain additional information regarding their significance. A list of attendees is contained in Enclosure 2 to this letter. During this conference, Region III management expressed concerns relative to your failure to satisfy an explicit license condition requirement and your apparent failure to provide sufficient resources to effectively implement the Dresden Fire Protection Program. In response to the concern expressed over your failure to comply with an explicit license condition, you presented information demonstrating that (1) failure to install fire detectors on the refueling floor was of minor technical significance based on the low fire loading and the lack of safe shutdown equipment in that area; (2) the Office of Nuclear Reactor Regulation had been informed in a letter dated February 25, 1980. 1980, that fire detectors were not necessary on the refueling floor; and (3) you had undertaken a0 review of.regulato~, commitment, and code compliance at your operating stations which had identified other issues requiring resolution and would likely have identified the failure to install the subject fire detectors. It was your contention that item (2) above demonstrated that you were aware of and sensitive to the license condition requiring the installation in~tallation of refueling floor fire detectors and that item (3) demonstrated your commitmerf't to you'/' commitme~t to*. ensuring that all required fire protection features had been implemented. implemented * . ( III.3-20

(

  -                                                                              Revision 8 April 1992 Commonwealth Edison Company                2 DEC 26 2 G 1S8S In response to the concern expressed over your apparent failure to effectively implement the fire protection program at Dresden.

Dresden, you presented information on the existing ffre fire protection staffing and experience at Dresden but indicated that the matter was under review and that additional fire protection expertise vay be indicated. Your response to the issues discussed in the November 19 Enforcement Conference was supplemented in a Tetter letter dated December 2, 1985, submitted to the Office of Nuclear Reactor Regulation discussing the prelimin~ry results of your review of the status of compliance with fire protection requirements at Dresden and Quad Cities. This letter identified several outstanding deficiencies. deficiencies, proposed methods for resolution, established completion dates. dates, and requested approval of the proposed resolutions. Region III reviewed the information presented at the November 19 Enforcement Conference and contained in your December 2 letter and has reached the following conclusions: .

1. Failure to install fire detectors on the refueling floor and failure to effectively implement your fire protection program at Dresden are violations of NRC regulations.

• c 2. Failure to satisfy a license condition is Qf of significant regulatory concern; however, you have demonstrated that, in the case of the refueling concernj floor fire detectors, the safety Significance 1s is low and that you were actively pursuing a program to ensure that compliance would have been achieved. Additionally, your December 2. 2, 1985, letter provides us assurances that this and similar issues are being aggressively pursued in a timely fashion. .

3. With regard to the failure to effectively implement the fire protection program at Dresden, you demonstrated that you had previously identified concerns in that area and were pursuing resolution of those concerns.

During the enforcement conference. conference, you verbally committed to bring additional resources to bear in this area. Based on the above, it is concluded that while escalated enforcement action is not warranted for your failure to install fire detectors on the refueling floor and your failure to effectively implement the Dresden Fire Protection Program, issuance of a Notice of Violation is appropriate. Accordingly. Accordingly, Enclosure 1 to this letter transmits to you a Notice of Violation for which a written response is required. In accordance with 10 CFR 2.790(a), 2.7g0(a), a copy of this letter and the enclosures will be placed in the NRC Public Document Room. II r. 3-21 rII'.3-21

Revision 8 April 1992 ( Commonwealth Edison Company 3 The responses directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of ~~nagement and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. Sincerely,

                                         ~~i('¥~Qa..
                                         ~~i("'1fl1"Q.,....

tIOames

                                         .0ames G. Keppler Regional Administrator

Enclosures:

1. Notice of Violation

2. November 19, 1985 Enforcement Conference Attendance List

3. Letter dtd 11/14/85, NRC to Commonwealth Edison Co.

cc w/enclosures: D. L. Farrar, Director of Nuclear Licensing D. J. Scott, Plant Manager DCS/RSB (RIDS) Licensing Fee Management Branch Resident Inspector, RIll Phyllis Dunton, Attorney General's Office. Office, Environmental Control Division c-( rrI'.3-22

Revision 8 April 1992 ( ENCLOSURE 1 NOTICE OF VIOLATION Commonwealth Edison Company Docket No. 50*237 50-237 Docket No. 50-249 As a result of the inspection conducted on Septe~~er 30 through October 21, 1985, and in accordance with the -General 'Policy and Procedures for NRC "General'Policy Enforcement Actions,- Actions," 10 CFR Part 2, Appendix C (1985). (1985), the following violations were identified:

1. Amendment No 36 to Provisional Operating License No. DPR*19 DPR-19 and ~~ndrnent

                                                                                 ~~ndment No. 33 to Facility Operating License No. DPR-25 require the licensee to complete the modifications identified in Paragraphs 3.1.1 through 3.1.23 of the NRC's Fire Protection Safety Evaluation dated March 1978 by startup following the 1979 Unit 3 refueling refuelfng outage. Paragraph 3.l.1 3.1.1 subparagraph (6) of the NRC's Fire Protection Safety Evaluation dated March 1978 states that early warning Fire Detection Systems will be provided for the Reactor Building refueling floor.

Contrary to the above, during the period September 30 through October 21. 21, 1985, it was identified that an early warning fire detection system was

  ,(         not installed on the Reactor Building 8uilding refueling floor. Further, it was determined that an early warning fire detection system had never been installed on the refu,elfng refueling floor.

This is a Severity Level IV violation (Supplement I).

2. 10 CFR 50.48(a) requires that each operating nuclear power plant have a to. 10 CFR fire protection plan that satisfies Criterion 3 of Appendix A to.lO '.

Part 50. It further requires that the plan shall describe specific features necessary to implement the program such as administrative controls anc and personnel requirements to limit'fire damage to structures, systems, or components important to safety so that the capability to safely shut down the plant is ensured. Section 3.1.A.l of the licensee's Fire Hazards Analysis Submittal. Submittal, which forms part of the licensee's approved fire protection program, states that the licensee has a Fire Protection Coordinator whose responsibilities include, in part, program coordination, equipment procurement, program include. enhancement. enhancement, conducting inspections, and supervising training of personnel. '. Contrary to the above, the licensee has failed to consistently and effectively staff the Fire Protection Coordinator position pOSition with the result that certain fire protect;n protectin equipment was not installed, hardware and equipment were not being properly maintained, required traini~g trainiQg was not completed, and prompt and effective corrective action was not taken for identified deficiencies. , (( This is a Severity Level IV violation (Supplement I). III .3-23

Revision 8 April1 1992 Apri ( With respect to Item 1, I, information provided after the inspection showed that action had been taken to resolve the fdentified identified violation and to prevent recurrence. Consequently, no reply to this violation is required and we have no further questions regarding this matter. With respect to Item 2, pursuant to the provisions of 10 eFR CFR 2.201, you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply for the violation: (1) corrective'action corrective"action taken and the achieved; (2) corrective action to be taken to avoid further results achievedi violations; and (3) the date when full compliance compl iance will be achieved. Consideration may be given to extending your response time for good cause shown. (~---

• c.

~ ~, .. - 2 III.3-24 111.3-24

Revision 8 Apri 1 1992 April ( ENCLOSURE 2 NOYI~~..E~. NOYI~g.R. )J_,_J~'?'?J~FCPCEf.~ENT

                                       )J_,_JJ'?'?_J~FCPCElmlT COt:fJF.E!i.C.t COt:H~_E!l_CJ:_ AnENOANCE AnENDANCE LIST A. Conmonwea 1th         Ed1~9!'-.£.OF:P~.r;(.P!!~.fl.r-!'sJ Edi~9!t£~:PE.r.Y_.E.!!!~.!'!l!'sJ B. B. Stephenson.

Stephenson, Division Vice President C. Reed, Vice President of Nuclear Operations L. Del George, Assistant Vice President, Engineering and licensing Licensing J. Reed, Quad Cities Fire Protection Coordinator L. Reed. G. Spedl, Assistant Superintendent of Technical Services, Services. Quad Cities J. J. McDcnald, Station Nuclear Engineering Department Fire Protection Coordinator Coordi nator J. Achterberg, Technical Staff Supervisor, Dresden J. Bitel, Operations Quality Assurance Manager

             *P.

P. F. Hart, Hart. Quality Assurance Fire Protection Engineer P. A. Lau, Quality Assurance Supervisor, Dresden J. Wojnarowski~ Wojnarowski. Nuclear LicenSing Administrator L. Davis, Supervisor of Station Support Services T. G. Hausheer, Support Services Fire Protection Engiener L. F. Gerner,

l. Gerner. Regulatory Assurance Superintendent D. JJ Scott, Scott. Station Manager, Dresden J. D. Brunner, AsSistant Assistant Superintendent of Technical Services, Dresden Turnback. Operating Plant Licensing Director M. Turnback, t(*

,( R. Rybak, Station Nuclear Engineering Department Fire Protection Supervisor J. W. Dingler, Senior Licensing Project Engineer - Sargent and Lundy B. !I..:..J.:-!'IjJ.fJ.!!~!...!egul

             ~.:..!l.!l.!=J.!!E!.1.egul at9!j'_C.fl,!l!Tlj~~

at9!y_C.!'!l!f1j~~ ion i on Personnel A. B. Davis, Deputy Regional Administrator, Region III C. J. Paperiell0, Paperiello. Director, Director. Division of Reactor Safety, Safety. Region HI III B. A~ Berson, Berson. Regional Counsel, Region III . L. A. Reyes, Chief,Chief. Operations Branch, Branch. Region III W. G. Guldemond, Guldemond. Chief, Chief. Operational Progr~s Programs Section. Region III C. B. Ramsey, Reactor Inspector, Region III A. Madison. Senior Resident Inspector. L. Madison, Inspector, Quad Cities, Region III L. McGregor, Senior Resident Inspector, Dresden, Region III G. McGregor. W. H. Schultz, Enforcement Coordinator, Region III R. A. Gilbert, Dresden Project Manager, Office of Nuclear Reactor Regulation c.( III .3-25

ENCLOSURE 3

    ,cY........ ,                                                                            Revision 8
 ~
  !¥... "\
  ~ '. '
   ~.
               ~~

I I UN,TED UNITED nATII.

                                             " "OO""'~T
                                        .~.H ITATU.

NUCLEAR REGULATORY COMMISSION RIOlo.. IIEGION III

                                                .OOsaWLT ..   .OAD OAD u,

April 1992

   \., ....... ..l                                     ILLINOIS . . , J7 GLaN eLLYN, 1L.L.INOIS 6/~~~

6'~~~ Docket No. 50.23~~f'~ 50-23~At'~ Docket No. 50-249 ~, ~7 Commonwealth Edison Company ATTN: Mr. Cordell Reed

              . Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:

This refers to the routine safety inspection conducted by Messrs. J. Holmes and C. Ramsey of this office on Sept_er SeptenCer 30 through October 21, 1985, of activities at Dresden Nuclear Power Station, Units 2 and 3, 3. authorized by NRC Operating Licenses No. DPR-19 Ind and No. DPR-25 Ind and to the discussion of our findings with Mr. O.D. Scott at the conclusion of the inspection. The enclosed copy of our inspection report identifies arels areas examined during the inspection. Within these areas, areas. the inspection consisted of a selective examination exal1ination of procedures and representathe representative records. observations. and interviews with personnel. During this inspection, inspettion. certain of your activities appeared to be t.n i.n violation with NRC requirements. These issues, issues. identified in paragraphs 3 and 7.a of the enclosed inspection report, Ire are being reviewed for potential escalated enforcement action. The results of that review will be forwarded to you by separate correspondence which will identify the nature of expected for.al fOnDal response. In accordance with 10 CFR 2.790 of the Commission's regulations, regulations. Ia copy of this letter and the enclosures will.be will be placed in the NRC's Public Document. Room. RoOlll. The responses directed by this Jetter (and the Iccompanyingaccompanying Notice) Ire are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980. 198O. PL 96-511. ( III.3-26

Revision a 8 April 1992 Commonwealth Edison Company 2 We will gladly 9ladly discuss any questions you have concerning this inspection. Sincerely, Sincerely. M~

                                        ~~          ,.11., "Director 1.110.

Division of elctor eactor Safety

Enclosure:

Inspection Reports No. SO-237/8S033(DRS); and No. 50-249/SS029(DRS) SO-249/85029(DRS) cc w/enclosure: D. L. Farrar, Director of Nuclear Licensing D. J. Scott. Scott, Plant Manager DCS/RSB (RIDS) Licensing Fee Management Branch Resident Inspector, RIll Phyllis Dunton. Dunton, Attorney Generalis General's Office, Environmental Control Division RI~I,J

~~

11/14/85 III .3-27

Revision 8 Apri 1 1992 April ( U.S. NUCLEAR REGULATORY COMMISSION REGION III Reports No. 50-237/85033(DRS)j 50-237/85033(DR5); 50-249/85029(DRS) Docket Nos. 50-237; 50-249 . Licenses licenses No. DPR-19j OPR-19; DPR-25 Ca.monwealth Edison Company Licensee: CQaIOnwealth P. O. Box 767 Chicago, IL 60690 Facility Name: Dresden Nuclear P~er Station, Units 2 and 3 Inspection At: Morris, IL Inspection Conducted: September 30 through October 21. 21, 1985 C\f~ l\f~ Inspectors: J:'Holaes 11- ,~-e5" Ir- 13-BS' bate 1(. If. '" ,__ 4, i bate Approved By: V. W. G. ul. mond. mond, Chief ,1.,.-.£-

                                                                              "   - ' * * *6.-

Operational Programs Section bate Inspection SummarY Summary 1985 Re orts

   ~~~i~;~~~~~~f:~!~~~i~n;sp~e~ct~i~o;n:c;O;ndUcted o.

reas nspected: out ne, unannounce safety inspection conducted to verify

   ~\hi adequacy of the facility'ss fire protection dete1"lline protect     progr .. i~leaentation prograa                 to and il!plllHntation detel"lline the status of LERs and previous open itas. The inspection involved verify to by two NRC inspectors. including 2 inspector-hours onsite 71 inspector-hours by*two during off-shifts and 11 inspector-hours conducting in-officI in-office review at the Region III office.

Results: Of the 6 areas inspected, no violations or deviations were identified in four areas. Two violations were identified in the reaaining raaining two arelsareas (failure to adhere to proprogru requil'llllnts - 'aragraph gr. staffing requirements Paragraph 3;3j failure to cOl))ly co.ply with a license condition to install an autOlitic autoeatic fire detection system in the Reactor Building refueling floor area *- Paragraph 7a).

C

C_

III.3-28

  ,

• Revision 8 April1 1992 Apri

( DETAILS

1. Persons Contacted DNPS D. Adam, Co.pliance CQlPliance Adlinistrltor

       *J. Brunner, Assistant Superintendent, Technical Technicil Services T. Ciesla, Ciesll, Assistant Superintendent, Operations Operltions
       *M. D1110n, Dillon, Fire Marshall
       *R. Flissner, Servic_

Service Superintendent

       *T. Hausheer, Nuclear Services, Technical
       *P. Lau, QA Supervisor
       *J. McDonald, Station Nuclear Engineering
       *8.
       *B. Rybak, Station Nuclear Engineering
       *0.
       *D. Scott, Station Manager
       *R. Whalen, Technical Staff J. Wujciga, Production Superintendent US NRC E. Hare, Resident Inspector

.( *L. S. McGregor, Senior Resident Inspector Stasek, Resident Inspector

       *Denotes those in attendance It   at the exit ..      eting of October 4, 1985.
                                                          ~eting

2. Licensee Actions on Previous Inspection Findings I.

a. (Open) LER (237/85029) and Violation (237/85028-01): Auxilia~ Auxiliary electric equipment rooa hllon halon syste. declared inoperable due* to ventilation dampers fliling failing to close. fire Fire watch was not established per Technical Specification No. 3.12.H.2.

3.l2.H.2. Region III's Ill's followup of this event is docUiented docUlinted in Inspection Report No. So-Z37/85028(DRS). 50-237/8502B(DR5). As II result of this fol1owup followup violation No. 237/85028-01 VIS "as issued. No response to this violation vas "as required because the 1fcensee's licensee's interi. Ind and long tel'll corrective actions were vere detel'llined satisfactory., satisfacto~.. The interi. corrective Ictions were vere i8pllllented iapl ...nted prior to or during the followup inspection. The proposed long tel'll corrective act10ns actions have not been i.llp1eeented. il!pleaented. Therefore, this Ivent event report reuins _ins open.

b. (Closed) LER (249/85014): Wet p1pe pipe sprinkler systel syste. in Unit 3 turbine trackway trlckway had to be rerouted to allow for overhead clearance clelrance for new nev turbine rotors. The sprinkler systa syste. was out of service S

5 112 hours beyond the 14 day 1i.it li.it peraitted pel'llitted by llthnical Technical Specification 3.l2.C.3. .- IIIIII .3- 29 2

Revision Revi 5 i on 8 April 1992 c ( The event report is closed based on the licensee's corrective actions taken which included restoration of the system to service and functional testing prior to declaring the system operable.

c. (Closed) LER (237/85010): Fire .door ,door for the Unit 2 125V DC batte~

battery roOil roOli found open. A fire watch was not established within one hour per Technical Specification 3.12.F.2. This event report is closed based on the licensee's corrective actions taken which included ii..ediate

                                                ...dilte closure of the fire door and training/instruction of plant operators on the requirelents requirellnts to keep fire doors closed at all tileS when.not in use.

d. (Open) LER (237/84"20):

(237/84-20): Two of seven root valves that were installed on fire hose stations were found to be in the closed position . rendering the fire hose stations inoperable.

             ~lthough  the licensee's corrective actions for this event included prompt opening of the closed root valves, the inspectors deterained that the licensee's progr.. for adlinistratively controlling valves that are not electrically supervised using wire seals to secure these valves in the open position and perfora;ng perforaing .anthly aonthly inspections to verify valve positions does not appear to be working. During the

• inspection the inspectors observed several non--,lectrically non-electrically supervised valves in the fire protection systelsystea with .issing or daaaged -nre wire seals. .

To correct this probl .. , the licensee stated that the progr.. for problea, adlinistratively controlling these valves is being upgraded to I~inistratively include locking these valves (chain and lock) in the open position in addition to aonthly inspections to verify each valve position in with NFPA Standard 26. This event report relains accordance ~th reaains open pending Region III verification of the licensee's upgraded corrective actions.

e. (Open) LER (237/84-17; 237/84"05):

237/84-05): Failure to establish continuous or hourly fire watch patrols due to inoperability of all or portions of fire detection and sprinkler alara 5ystHs systeas in the control roOil. roOli. During these events and at the present tile, the fire detection and sprinkler alar. alara systea printer indications art are interlocked (dependent) into the pllnt plant security 5yst .. cOIputer. systea coaputer. Indication of fire detection and sprinkler .'aras Ind alaras in the control rooa roOII was lost in two events either because of *a loss of power to the plant security 5ySteisystea ca.puter computer or because of aodif1cations aodifications being aade to upgrade the plant security systell systea computer. Apparently. Apparently, any failure of the plant security systea coaputer can cause the loss of all or portions of fire detection and sprinkler alara annunciation in the control roGa roOII

• This installation does not comply .nth with the licensee's coaaitaent coeaitlent to NFPA 720 as stated in the licensee's April 1977 response (point"by-(point-by-point comparison) to Appendix A to NRC Branch Technical Position III.3-30 II 1.3 -30 3

Revision 8 April 1992 (~-- ( (BTP) APCSB 9.5-1. The licensee's scheduled plant .edification IOdification Mo. No. Ml2-2/3-84-109 identifies corrective action for this probln IS the installation of independent circuits for fire detection Ind and sprinkler syste. syst.. allras alanls which alara alanl and annunciate in the control roo. in accordance with NFPA 72D. This aodification IOdification is scheduled to be c.-plated co.pleted in Decaber 1985. These LERs will ..... in open pending l'tIIIin Region III verifiCAtion verification aTOf the licensee*s licensee's corrective actions.

f. (Closed) LER (237/84-11): Fire wall penetrations to Unit 2/3 di.sel di est1 generator I"OOIIS I'OOIIS were not sealed.

This event report is closed based on the licensee's corrective actions taken, which include establishlent of a fire watch within Ictions one hour per Technical Specification No. 3.12.F.2 and sealing the penetrations per drawing No. 12E-6058. 12E-60SS. *

g. .(Closed) LER (237/84-08): NRC inspection of the licenseelslicensee's co.pliance coepliance with fire protection Technical Specification surveillance requiraents requil'tlllnts identified that a cardox system systH IIIster Rster valve was WIS not .'

being tested in the lutoutic autoaatic .ode. The .. asur ster valve test procedure was written to test the valve in the .. Mnual nual .ade.

                                                                .ode.

This event report is closed based on the licensee's corrective actions taken which included prompt retIOval removal of the .. Aster ster valve

(--

.( fro. service, revision of the surveillance test procedure, and satisfactory testing of the valve in the autoutic lOde. auto.atic .ode.

h. (Closed) LER (249/83-34/03L): Unit 3 trackway sprinkler systel syste. out of service due to damage duage by .abile IK)bile crane boOli.

boo.. .' This event is closed based on the licensee', licensee's corrective Ictions actions taken, which included .aking the necessary systell repairs, Pl'OIIpt PrDllPt restoration of the sysu. systH to service and instructions to plant personnel regarding the -avelent IOv...nt of IK)bile IObile cranes and the fragility of systels systHs and components coeponents in their path.

i. (Closed) LER (249/83-17/03L): HPCI deluge systell solenoid valve taken out of service because the valve would not reset.

This event report is closed based on the licensee's corrective actions taken, which included pro.ptproapt relaval rtIOval of the deluge system fro. service, uk_fng Rking the necessary repairs to the solenoid valve and Nstorati on of the systell restoration systH to ,ervi ceo service. ,

j. (C.losed)

(Closed) LER (237/81-15/03L): (237181-15/03L): Unit 2/3 diesel generator 1"00II 1'00II CO2 systea systH heat detector surveillances not perforled perforaed per Technical SpeCification Specification 4.12.A.l. This tvent event report is closed based on the licensee's corrective actions taken, which included instruction to plant personnel to perform the required heat detector surveillances and satisfactory performance of the surveillance. III.3-31 III .3-31 4

Revision 8 April 1992 ( t.

k. (Closed) Violations (237/81-09-01; 249/81-06-01): Four penetration seals identified IS as being defective were inoperable for an excessive period of ti.e.

ti~. Neither prompt nor tiae1y tiaely corrective Iction action WIS was taken. This item is closed based on the licensee's June 29, 1981 response to Region III which discussed the licensee's corrective Ictions actions taken to avoid futwre fut ..re violations in this arel.area. The inspector's DFPP-4175-2, Revision 4, review of Procedure No. DFPP-417S-2, 4. indicated thlt that appropriate instructions are provided to plant personnel which refer to detail .drawings for proper installation of penetration fire seals. fin

1. Violations (237/81-09-03; 249/81-06-03): Cal (Open) Vfolati~ns of fire extinguishers sampled did not have 1981 IOnthly (a) Fifty percent

                                                                 .onthly inspection tags attached; (b) 5 year hydrostatic test for portable CO2 extinguisher cylinders were overdue; and (c) nUieraus  nUierous ca.pressed
      .gas cyl inders i"ders were iI.properly il!properly stored.

The 1itens.e's licensee's corrective actions identified in their June 29. 29, 1981 response to ite.s items (a) and (b) of this violation were ineff.ctive. ineffective. Subsequent QA audits and surveillance by the 1iclnsee's onsite QA licensee's onsi1l depart.ent departient have revealed that these deficiencies are continuing. For example, deficiencies identified 1n in QA surveillance No. QAS Septellber 23 through 29. 12-85-236 for the period September 29, 1985 include the wrong date on extinguisher tags; extinguishers past due fonowing: WTDng following: for 5 year hydro testing; no service date on extinguisher tags; no seal on extinguisher pull pin; partially discharged extinguisher. Dur;ng plant tours by the inspectors, identical deficiencies were observed. In one instance, a CO 2 portable extinguisher hose was duaged to the extent that the webbing in the hose was exposed ** daaaged A hole existed in the webbing that ..yy have allowed the extinguishing A agent to escape through the hose prior to reaching the COl COz discharge nozzle. This extinguisher was located on fire cart No.2. In addition. the inspectors observed that wheeled d~ cne.ical addition, cne.;cal extinguishing units Nos. PI Pit n Pit 22 had tags which indicated 21 and PK surveillances that surveil lances were .i .issed ssed the .onths IOnths of May and Septellber 1985. The continuing existence of this type of deficienc,y deficiency is indicative of a lack of ..nagHent nagllllnt attention in this area. ManagIMnt JlanagHent attention and staffing is the subject of *a violation doc&lllnted docuaented in paragraph 3 of this report. Your response to that violation lhould should address your corrective actions for failing to properly ..intain fire extinguishers. These ftBs itas will reuin reaain open pending the further licensee's corrective actions by Region III. review of the licinsee'l Ite. (c) of this violation is closed based on the lfcensee's Ita ec) licensee's corrective action taken which included the installation of .. tal

                                                                              ~tal ca.pressed gas cylinders, lecuring storage racks for capressed                          securing the cylinders with _tal
              ~tal chains and revision of Procedure No. CAP       DAP 3-U:-

a. (Open) Unresolved Item (237/81-09-06; 249/81-06-06): Fire brigade

'c drills and training do not appear to .. requirelents. requirements. II I .3-32 III.3-32 et the intent of NRC aeet 5

Revision 8 Apri] 1992 April ( .. ( Section 6.0 of the original fire protection SER, dated March 1978, recOlDended that the licensee's administrative controls fol1~ recOMmended foll~ the guidelines set forth in the NRC Guidance DocuaentDocUient entitled -Nuclear Plant Fire Protection Functional Responsibilities. Responsibilities, ~inistrative Adlinistrative Controls and Quality Assurance.- A supple.ent suppl ... nt to the original SER was issued Declilber Oecelber 2, 1980. Section 3.1 of this suppl s~pl . .ntll

                                                                          ... ntal SER closes out the issue of a~infstrltive adiinistrative controls -nth            NRC staff's with the IRC acceptance of the 1fcensee's licensee's discussion of adlinfstrativ.

adiinistrative controls provided in letters dated January 24, February 24, Mlrch March 20 and July 27, 1978. 1978, January 31. 31, and April 30, 1979. Therefore, Section III.1.3.b III.l.3.b of Appendix R 1s is not applicable to adlinistrativ. adiinistrative controls for fire protec~ion protKUon at Dresden. aased Based on the licensee's submittals discussed above, the HRC NRC staff concluded that the licenseels* licensee's- administrative controls for fire protection 8et..t NRC guidelines and, the applicable regulatory

       -requirement for fire protection adainistrative controls at Dresden is the Commission's guidance issued on the i8pl..entation i!pl ...ntation of General Design Criterion 3 of Appendix A to 10 CFR Part 50       SO for existing p(Ner p~er plants.

Section 2.0 and 3.0 of Attachment No. 2 to NRC Guidance Docu.ent DocUient

        -Nuclear Pllnt Plant Fire Protection Functional Responsibilities.

Responsibilities, Administrative Controls Ind Quality Assurance* and QUllity Assurance- requires practice sessions be held for fire brigade .. lbers to provide each brigade Ie~ers

        -..ber with experience in actual fire extinguishient extinguishment Indand the use of emergency breathing a~aratus under strenuous conditions. Fire brigade drills are required to be perfor.ed so that the fire brigade can practice IS as a team. The drills are to be    ba perfor.ed It at regular intervals but not to exceed three .onths for each fire brigade. The drills are required to be critiqued to assess lach     each brigade -.ber's leiber's knowledge kn~ledge of his role in fire fighting strategy.

The licensee is not ..eting these require.ents require..nts for the following foll~ng reasons: (1) By attellpting attempting to .eet

                                  ..et the requiraents require..nts contained in Section III.1.3.b III.l.3.b of Appendix R to 10 CFR SO, the licensee bas has beln been cDnducting conducting one fire drill per .,nth -nth  with the intent of getting &11 all designated fire brigade "'e1"58111bers involved in at least two drills per year.

(2) Practice sessions that provide elch each brigade ~er with actual lelber ~th fire extingufst.ent IXtinguishitnt experience and the use of . .rgeney

                                                                     ... rgency breathing apparatus under strenuous conditions (full fire fighting gear) have not been conducted due to a breakdown in contractual arrangements with an independent fi~.

(3) Fire brigade drills hive have not been critiqued at threiyear intervals by qualified individuals independent of the

             *licensee's lfcensee's staff.

II 1. 3-33 III.3-33 6

Revision 8 April 1992 ( To resolve this concern. concern, the licensee is requested to aake Ivai1abl, IVlillbl, Issessment of fire brigade drills. I detailed assessment drills, practice sessions-and sessions -Ind three year audits ludits of fire brigade drills by qualified individuals independent of the licensee's staff. This assessaent assesslent should establish whether the licensee is in compliance complilnce with ca..itients Co.litients aade to the '-NRC NRC which resulted 1n in the NRC staff conclusions that the 1fcensee's 1icensee's Idlinistrative adlinistrltive controls for fire protection were acceptable. .' "- . This it.. it. ruains raains open pending "gionregion review of the licensee's licensH's assesslDent. IssesslDent. '

n. (Open) Unresolved It.. (237/81-09-07; 249/81-06-07): Specific pre-fire fighting plans pllns or strategies strltegies for 111 III IIfetyere11ted safety-rellted areaslrels and Ind areas lrels presenting aI hazard to safety-related slfety-related equip.ent were not developed and iiplemented.

             'As
             ~s discussed in item 237/81-09-06; 249/81-06-06 above, the requirements of 10 CFR 50,   SO, Appendix R (Sections III.l.llIII.K.ll and Ind 12) are Ire not applicable Ipplicable in this case.

cise. The applicable requirements requirelllnts Ire contained in Attachment No. 5 of NRC Guidance Guidlnce Docu.ent DocUllnt -Nuclear Plant Pllnt Fire Protection Functional Responsibilities, Adlinistrative Controls and Quality Assurance.- During the inspection. inspection, the licensee provided the inspectors with c ,(,., aI copy of pre-fire plans that Ireas areas that present aI hazard thlt contained specific fire fighting strategies for fighting fires in all safety-related safety-rellted arels hlzard to safetyerelated safety-related tquipaent. lreas Ind Iquip.ent. The pre-fire plans pllns appear Ippeir to provide Idequate fire fighting procedures and instructions. However. However, these plans have not been iiplellnted. i~le.ented. According to the licensee. licensee, the plans wi11 will be illplaented illplellnted and Ind incorporated into fire brigade trlining lesson plans pllns by the end of the first quarter of 1986. This ita it. will raain reaain open pending peilding iiplellntation. said illPlnentation.

o. (Open) Unresolved lte. Itt. (237/81-09-08; 249/81-06-08): Fire brigade practice sessions have not been bHn conducted in accordance with cDiDit.ents ..

cDlillit.ents aade HRC. A hands*on de to the NRC. hinds-on practice prlctice session was held-' held"- in 1979 with full brigade attendance, but noprlctice session less ion was held in 1980. 1980~ n... wi This it. will11 reui raainn open pendi pendingng Reg; Regionon III revi ew of the review 1fcensee'l licensee's response to Itt.It. b of Unresolved Ita. ltea No. 237/81-09-06. 237/81-09-06; 249/81-06-06 as IS discussed in this report.

3. Fire PrOtection Progr.. Organization and'Personnel and "Personnel Stiffing Staffing 10 CFR SO~48 requires that each operating nuclelr nuclear power pllnt have Ia fire protection plan that satisfies Criterion 3 of Appendix A to 10. 10.. CfR CFR SO.

Except for the requirements requireaents of Section III.G, III.J. III.J, and 111.0 of Appendix R to 10 CFR 50, the approved fire protection plan that sltisfies satisfies .(-. Criterion 3 of Appendix A to 10 CFR SO is discussed in the original fire protection SER, dated March 1978, aI fire protection SER Supplement, Suppllllent, dated December 2. 2, 1980, and th~thw licensee's Fire HazlrdHazard Analysis submittals entitled 7 III.3-34

Revision 8 April 1992 (

   "Infor.ation
   -Infor.ation Relevant to Fire Protection Systeas Syste.s and Programs" Programs- dated October 1976.

1976, Janulr,y January 1977. 1977, and April 1977. Further.ort, FurtherlOre, the licens.e licensee committed to follow certain certlin NRC Supple.ental Guidance DocUients DocUitnts ISas discussed in letters to the NRC, dated January 24, February 24, March 20 2D Ind and July 27, 1978; January 31 and April 3D, 1979. 1978, Janua~ n .. requireaents for overall responsibility for Tt~ the Fire Protection Prograa Program Ire are discussed in Sections IV.A and 3.1.A.1 of 'arts Parts 1 Ind and 3 of the licensee's Fire Hazard Analysis subllittal, dated October 1, I, 1976 and April 19n. The NRC* NRC'ss pos it ion. a, restated stated 1inn Section 3.1. position,as A. 1 3.1.A.1 of this docllHnt docUitnt establ:fshes establ.fshes guidance on 1!1pl ...ntltion of basic criteria i!lpl_ntation for fire protection progr .. organization and personnel staffing. program In response to the NRC's position discussed in Section A.l of Appendix A to NRC Branch Technical Position APCSB 9.5-1 concerning the qualification require.ents for the Fire Protection Engineer who ~11 will assist in various aspects of fire protection progrprogram.. deyelo~nt developilnt for the operating plant, the licensee states "comply"

                          *comply" in Section 3."1.A.1 3.1..A.l of the Fire Hazard Analysis submittal. The licensee further states, in part,       pert, *CECo has a Fire Protectic;m Protection Coordinator who reports to the Supervisor of Safety ***          **

Responsibilities of the Fire Protection Coordinator are: coordination of activities; activities, procureaent procure.ent of equiplent, equipllent, resolve questions on standards and technical issues; lAkeuke reca.endatfons reco.endations for i.ravMents; i!lprov_nts; coordinate. coordinate, plan, Ind and conduct inspections (aaka (uke inspections of Dresden. Dresden, Units 2 and 3, once Ia IOnth);

            ~nth); ensure that adequate fire fighting equiPllnt tquiPlint is provided and that such equiPilent equipllent is lAintained uintained in good operating condition, coordinate with offsite fire departaent; departllent; conduct norulnor.al and preoperational       rr-testing; provide foras for.s and instructions for reporting ffres;  fires; issue publications outlining .-ployee employee polf~

policy and ~rocedures procedures in fire protection; assist and supervise training of personnel; assist and advise departBents departaents concerned with established rules and standards; coordinate with the staff an .. 111 ttars of .utual concem utters concern and uke final rec . .ndations for specific rec..-endations actions to be talten taken on fire protection issues." issues. - . The inspectors identified the following foll~ng examples of the licensee licensee's's failure to consistently and effectively ca.ply co.ply with the staffing require.ents for fire protection prograa requiraents program illP'tllentation: i!lplllHntation:

a. Fire Protection Engineer A qualified Fire PrOtection Protection Engineer was not involved in the develoPllnt of certain aspects of the fire protection prograa developient program for the operating plant IS as required by Section 3.1.A.l 3.1.A.1 of the licensee's Fire Hazard Analysis SUD.ittal.

subllittal. The qualifications for this individual were not stated in any docIMnt.docUitnt. The res ... of the resUlt individual perfo1"8ing perfor.ing the original Fire Hazard Analysis is contained in Attact.ent 2A of the Fire Hazard Analysis saittal. subllittal. but this individual is no longer 1IIIP10yed tllploYld by the 11censH HcensH *. According to the licensee, there WlS was a contract with NUt MU4 ;ratection Protection COnSUltants which included services that would satisfy Consultants SODe of the satis~ SOle 'c. responsibilities of the Fire Protection Engineer. expired in Decelftber Decelllber 1984. III.3-35 Engineer, but this contract 8

Revision 8 c April 1992 .\ Although the licensee has .-played

                                            .-ployed another qualified Fire Protection Consultant fir.

finl to do scat SOle specific fire protection work relative to

• upgrading the fire protection program, progra., this fir. finl vas not retained to fulfill all of the responsibilities of the fire Fire Protection Engineer.

b. Fire Protection Coordinator The Fire Protection Coordinator was vas not perfor.ing perfonling all the duties at,the site that Ire It,the are delineated in Section 3.l.A.13.1.A.l of the licenseels licensee's fire Fire Hazard Analysis subllittal.

subaittal. According to the licensee*s licensee's staff, the individual that vas originally assigned these duties was vas trans* trans-ferred to Corporate Quality Assurance sa.. SOlI tiae ti .. ago. Once Onee vacated, this position VISvas not filled. The duties and responsibilities of the position were delegated to the fi Fire... Marshal and other individuals vithin the CECa within CECo organization. Jhrough Amendment No. 86 to Facility Operating License No. DPR-19 (Unit 2) and Amendment No. 79 to Facility Operating License No. DPR-25 (Unit 3), the N~C accepted a proposed licensee staffing change. Figure 6.1-1 (Corporate and Station Organization Chart) shows a Fire Protection Inspector reporting to the Corporate Director of Quality Assurance Operations. The inspectors "quested, requested, but the licensee did not provide the inspectors withvith docUlintation cIocU81ntation to verify that the NRC was vas aware hare that the suesa.. individual who lIIho was vas the site Fire Protection Coordinator was vas filling the position entitled -Fire Protection ,( Inspector- for Corporate Quality Assurance. The licensee l s failure to adhere to the staffing requireaeots licensee's require8lnts discussed above resulted in progrllllllatic progrlllllatic breakdowns breakdCMIs that have decreased the level of fire protection that was decreased'the vas intended to satisfy Criterion 3 of Appendix A to 10 CFR 50. For lX&lpl.: exa.ple: (1) AA fire detecti,on detection sysu. syst.. was not installed on the refueling floor as required by ~ndment Amendment No. 33 to Facility Operating License No. DPR-25. (This is discussed in Paragraph 7.a of the report.) (2) Insulled Installed fire protection hardware hardvare and Iquipillnt equipillnt was not being properly .. inuined. (This is discussed in Paragraphs aaintained. 2.d, 2.1, 2.e, 4, 5, and 7 of the report.) (3) Technical specification surveillance procedures did not incorporate appropriate testing of quality affecting parameters in accordance with design and governing cod~ code requi .....nts. require8lnts. (This is discussed in Paragraph 4 of the report.) (4) Administrative controls did not adequately control fire protection features. (This is discussed in Paragraph 5 of the report.) (5) Many deficiencies that verewere identified in LERs. LERs, NRC inspections, QA audits, and QA suryeillances surveillances did not receive pro.pt or effec-

 ~-,- '               tive corrective action. (This is discussed in Paragraph 2 and 6 of the report.)         II 1. 3-36 111.3-36 9

Revision 8 i:

t. Apri 1 1992 April

( (6) Weaknesses in the scheduling of fire drills were identified. (This is discussed in Paragraphs 2.* ***, Ind and 2.0 of this report) *. Failure to cOIply coaply vith the staffing require.ents for develop.ent developaent Ind and iaplaentation iapleaentation of the fire protection prograa progru is considered a violation of 10 CFR 50.48 Ind and Criterion 3 of Appendix A to 10 CFR 50 (237/85033-01; 249/85029-01(DRS>>. The Station Fire Marshal's qualifications include ,58 .58 junior college credits in fire science; In an Illociltes associates degree in electronics engineering and 15 years experience as IS Ia volunteer firefighter. He has held the position of station fire ..rshal rshll for seven years. At the present ti., tiM, the fire ..rshal is assigned the following responsibilities: a. I. Coordinate and Issistassist in fire syste8s systeas periodic testing.

        'b;
        'b~    Plan, coordinate, conduct.

conduct, Ind and critique fire drills.

c. Fire Brigade classrooa classroo. training.

d. Review.

Review, revise. revise, Ind and write vrite new Idllinistrative adiinistrative procedures. ee.* Review. Review, revise and ,"itt vrite new survei11ance surveillance procedures. Make ((~. . work requests to repair deficiencies. Ire are cOIPleted coapleted IS deficiencies, verify surveillances as required and .aintifn

                                                     .aintain files on coapleted surveillances.

f. Review plant .edifications, assist in training. training, testing, Ind and developaent develo~nt of procedures.

g. Iquipaent, verify availability Maintain fire lquipMnt. availibility of spare Plrts parts Ind and procureeent procure.ent of parts.

h. insurance inspections, Technical Specification Participates in insurlnce Reviews, QA, INPO. INPO, and NRC ludits.

audits.

i. Assure Technical Specification cOIpliance.

coapliance.

j. vork "quests.

Review work requests.

k. vatch Ind Verify fire watch and insurance notification.

1. Coordinlte Coordinate activities vith the offsite fire depart.ent.depertaent.

a.

          **    Make reports on deviations Ind    and fire daaige dalage experiencls.

experiencls *

n. Perfor.

Perform plant cleanliness inspections.

o. Correspond vith other agencies on fire protection issues.

C

C::. p. Assure that the fire protection progr....

requ i rements. requi rllll!nts. ets NRC and other

                                                               .. Mets III.3-37 III .3-37 10

 \ .                                                                                  Revision 8 April 1992

(

q. Explain fire protection requirements to the licensee's staff stiff when required

• According to the licensee's staff Stiff and Station Nuclear Engineering Department (SNED) procedure number PE Q.44, Ia qualified corporate fire protection engineer reY1~reviews new plant -adiffcations

                                                                 .adifications prior to implementation illlpletnentition by the Archftect-Engineering Architect-Engineering finl. ff~. This appears to be the extent of the corporate fire protection engineer's involvement.

The qualifications of the Station Stltion Fire Marshal do not appear to be commensurate with the list of responsibilities assigned to that position. This lengthy list of responsibilities constitute Ia workload that may not be achievable by a single individual, regardless of the individual's qualification Ind and experience. concern, the licensee 15 To resolve this concern. is requested to provide It at the site. site, Ia written evaluation (complete work study) of the responsi-

          ,bilities
          .bflities assigned to the station fire marshal. This evaluation should ~ke make Ia determination of the fire marshal's ability to effectively achieve each delegated responsibility based on his.

qualifications and time constraints. This is considered an Unresolved Item Jt~ (237/85033-02; 249/85029.02(DRS>> 249/85029-02(DRS>> pending Region Ill's review of this evaluation.

4. Technical Specification Surveillance Review Technical Specification 6.2.A.l1 6.2.A.ll requires that detailed written procedures developed, approved and adhered to for illPlementation be developed. illlplementation of th~

the Fire Protection Program. The inspector's review of the licensee's surveillance procedures and test results for fire protection Technical Specification surveillance requirements resulted in identification of the following discrepancies:

a. Testing of Diesel Fire Pump at Least Once Per Operating Cycle Section 4.12.8.1.(e) of Technical SpeCification Mo. No. 3.12.B 3.12.8 requires that the station diesel fire pumps be demonstrated operable by per-fOnling a system functional test which includes silUlated forming simulated automatic actuation of the pumps PUIIIPs throughout their operating sequence. The licensee's commitment in Section 3.5.E.2 of the Fire Hazard Analysis Report dated April. 1977, requires the fire IU'P punp installations to conform to NFPA standard No. 20. This cCJllllitllent confont commitllent states that a plant .edification modification would provide an adequate flow glge gage for'full flow testing of the pumps in accordance with HFPA NFPA standard 20. The licensee's surveillance procedure Nos. DFPP 4124-3 and DFPP 4124-4 were deficient in that:

(1) manual throttling of the PtlllpS The procedure required IIInul' pumps to achieve Specificat-1on 3.12.8. the specific flows contained in Technical Spec1f1cat-1on and did not address automatic activation. c (2) The procedures required test1ng testing the pumps to the speciftc specific head and flow contained in the Technical Specification No. 3.12.8. 3.12.8, 11 II I'. 3-38 III'.3-38

Revision 8

 , ,                                                                            April 1992

( but failed to require testing for head and flow as specifi.d in NFPA 20. paraaeters such as PUlp (3) Measurement of quality affecting para.eters PUlp vibration under full flow conditions wert were not included in the test procedure or the test results. (4) cOilpared to the original .. The test results were not cDipared nufacturer*s Mnufacturer's fi.ld acceptance test for the PUlpS shop test curve or field PUlps because neither Mither of these CU1"Yes curves were avaflable available to the licensHls licensee's staff.

b. Water Suppression Systeas Testing of Watlr Systells at least Once Per Operating CYcle Cycle Secti~4.l2.8.1.~of Secti~ 4. 12. 8. 1.J!j)of Technical Specification No. 3.l2.83.12.8 requires' requires*

flre suppiessl0n that f1rt syst..s be demonstrated operable by suppie~s10n water systels

        ~erfonling a syst.. functional test which includes si.ulated lutoaatic
        ~erfor.ing                                                             autoaatic systells throughout their operating sequence. The actuation of the systels commit8ent in Section 3.S.E.3 of the Fire Hazard Analysis licensee's commitaent licenseels autoaatic sprinkl.r Report requires that autolatic                  systells confo,. to NFPA -

sprinkler systeas Standard No. 13. The lic.nseels ~~P 84-6-39 fiNed licensee's surveillance procedure Nch-.,SP f~ed to requirelents to diiOnstrate the incorporate appropriate test requfrtlents systell is operable in accordance with NFPA 13 in that: sprinkler systea (1) The procedure did not require flow frca f~ the inspector's test systells. Instead, the ala,. valve of wet sprinkler systeas. ala~ bypass valve was used for this test. (2) The procedure did not require flow frca f~ thlthe two tw inch drain syst..S. valve of wet or dry 5yst s. Instead, the alar. ala,. bypass-bypass* valve was used for this test.

c. Semiannual Testing of Fire Detectors Section 4.12.A of Technical Specification No. 3.12.A requires that syst18 be deIonstrated the fire detection syste. dI80nstrated operable by perfor8ing a lV.ry six .anths.

channel functional test every 8Onths. The licensee lic.nsee ca.itltent c_itlient 3.S.E.l of the Fire Hazard Analysis Report in Section 3.S.E.1 R.port requires that syst18 confol'll the fire detector syst. confo,. to the nquirtlllnts requirtllents of NFPA Standard 720. - licensee's surveillance The licenseels surv.il1ance procedure No. DFPP 418S-2 (Revision 4) failed to incorporate the following quality affecting pirl8eters paraaeters as required by NFPA 720: cleaning of detectar (1) Periodic cl.aning detector units. adjustlent for sensitivity (Section 3.1.2 of the (2) Periodic adjust.ent original SER required this test to be conducted). III .3-39 12

Revision 8 A April 1992 According to the lfcensee's licensee's Itaff, staff, an independent fire protection eaployed to review all technical specification consultant has been .-ployed procedures and test results to evaluate their ldequacyadequacy in accordlnce accordance with NFPA standards Ind requireeents. This IISeSS8Int and design requireaents. assesslent was WlS in progress at the ti. tile of the inspection Ind and is expected to be cOIIpllted cOllpleted by the end of 1985. According to the ticensH. licensH, where lIICessary, necessary. the procedures will be revised to coincide w~th with the governing code and require..nts. design require.ents. . This is considered an Open Ita.Itt. (237/85033-03; 249/85-029-03(DRS>> 249/85-029-03(ORS>> pending Region Ill's review of the licensee's Ictions.actions. No viol.tions violations or deviations were identified.

5. Administrative Controls - Control of Welding. Cutting. and Ignition Sources Licensee procedure No. DAP licensee OAP 3-11 (Revision 4) contained what appears to be acceptable instructions for controlling storage of fl&lmable fla.mable Ind and combustible liquids, storage of cOlpressed gas c.ylinders, cylinders, Ind and accUiulation of rubbish and transient ca.bustibles cOilbustib*les such IS as wood scaffolding.

scaffolding, etc. The procedure specifies housekeeping and cleaning responsibilities to be

                            .-p10yees Ind followed by all taployees     and contractors.

No violations or deviations were identified in this area, however; the (~:: inspectors cautioned the licensee on Ia proposed revision to welding and cutting procedure No. DMP 4100 that would include Ia provision to ALARA concerns in high radiation areas. The inspectors facilitate AURA inforled the licensee that any rel1ef fnfor.ed relief fro. fro! the requireDents require..nts for'. for*a firewatch to 1"IIIIin reIIIin in the i.ediate i_diate area thirty .inutes after cutting

          .nd and welding has been ca.pleted would have to be discussed with NRR.

6. Progr..

Quality Assurance Proar" licensee's ca.mit.ent The licenseels cDIBit..nt to Quality Assurance for fire protection is documented in Section 3.3 of -Infor.ation Relevant to Fire Protection Systns and Progrus-Systells Progr. .- and in letters to the NRC on this subject dated January 24, February 24, March 20, and March 27, 1978, January 31 and April 30, 1979. The .inspectors review of the 11clnsee licensee'sl s Quality Qua1i~ Assurance Progr.. for Protection included review of the following: Fire Protec;tfon a.

          **     Ellven Eleven criteria applicable to fire protection that satisfy Appendix A to Branch tlchnical Technical Position 9.5-1 and suppl ...nt guidance -Nuclear Pllnt Plant Functional Responsibilitils.

Responsibilities, Adlinistrltive Adlinistrative Assurance.- Controls and Quality AssurancI.-

b. Quality Assurance Survlillance Surveillance Reports dated SeptHber 3-6, 1985.

Septe.ber 3-6. 1985, September 5-9, 1985, Sept_er September 9-13, 1985. 1985, and Septellber Sept.ber 16-30, 1985.

c. Annual Quality Assurance Audits Nos. QAA 12-84-1 dated April 17,

                .1984, and QAA 12-83-1 dated April lS, 1983.
                .1984.

13 III.3-40 III .3-40

Revision Revi sian 8 April 1992

d. Triennial Audit 'by M&M Protection Consultants dated DecemberDecenber 4.

4, 1984. identified. however, the inspectors No violations or deviations were identified, suggested to the licensee that for clarification, the statementsstateaents made in "Infonlltion Relevant to Fire Protection Systems and Section 3.3 of the *Inform.t10n ProgrllS" should be modified to indicate their specific coanftlent ProgrlR$- coanitlent to a statelnent can QA prograll to fire ,protection. As written, this statement tin be interpreted to llean mean thlt that the licensee comitted committed to Ipply apply an all of the criteria of Appendix 1B in 10 CfR SO criteril 50 to fire protection. The inspectors determined that the licensee's practice of considering fire protection as reliability-related is acceptable because,this practice' ensures that IIIall of,the eteven eleven criteria contained in the NRC's QuidAnce Guidance Ire are included in the program. In addition, this practice allows for the nonnal QA program for safety-related systems to be applied to fire nOrNl protection in it's entirety. Only one QA manull manual exists for rel'ability-reliability-related systems and fire protection systems. Although the licensee's Quality Assurance Program appears to be effectively identifying issues that are contributing to hardware and progrannatic progralllllltic weaknesses, the licensee does not appear to be taking prompt actions. This is exemplified by the remaining and effective corrective Ictions. open items that have been identified in QA audits and surveillances. surveillances, LERs, and NRC inspections.' (This is further discussed in 3.b.(5) of the LERs. report.)

7. Plant Tours plant, the inspectors observed the following During tours of the plant.

deficient conditions: a. I. Failure to C~lY C~ll with License Condition No. 2.1. 2.B. of Amendment NO. l3 33 to Fie Fac 1ty License No. bPR-i5 O~rating L1cense 1tl 0rerlt1ng DPR-z5 Ind and Amendment No. 36 provisiona Operating License No. bPR-I9. jti to provision. DPR-19. Section 5.1.6.6 of the original Fire Protection SER for Dresden Units 2/3 dated March 22, 1978 states that the 1licenseeicensee proposed the installation of an lutanatic instillation automatic fire detection system to provide early warning of *a fire in the Refueling Floor Area in order to satisfy the objectives of Criterion 3 of Appendix A to 10 CRF 50. Allendment Alnendllent No. 36 to Provisional Operating License No. OPR-19

10. DPR-19 (Unit 2) and Alnendment No. 33 to Facility Operating License No. DPR-25 (Unit 3)

Amendment dated October I, 1980, requi re that the early warning, require warning automatic f1 re fire detection system for the refueling floor area be installed by start u, up following*the following 'the 1979 Unit 3 refueling outage. (approximately six years after As of the date of this inspection (approxillltely start up following the Unit'Unit 3,. 3,.1979

                                             ,1979 refueling outage) the licensee

• cOllply .nth has failed to comply with the provisions of Amendment lo~ No~ 36 to Provisional Operating License No. DPR-19 and Amendment No. 33 *to 'to License No. DPR-25. An early warning automatic Facility Operating Lfcense fire detection system fire detection system hiS has not been installed IIL3-41 14

Revision 8 April 1992 ( in the Refueling Floor Area and no compensatory NlSu*res IlelSures have hIVe been taken as *a result of this decreased effectiveness of the plant's fire protection features. The installation of In autonatic elrl1 an lu~tic early wlrning warning fire detection systee systell in the refueling floor lrel was not area WlS IIQt discussed in Iny of the licensee's correspondence to* the NRC that requested uendments licensee'S aaendments to modify the pllnt's*fire plant's* fire protection Technical Specifications to incorporlte incorporate Li.tting Limiting Conditions for Operation and Surveillance Surveillince Requirements for the fire protection lW)difica.tions lIOCIific:a.tions required by the original SER for Dresden Units 2/3. Hone None of the proposed Tables Tlbles 3.12.1 to Technical Specification 3.12 listed fire detection instruments in the refueling floor lrea. However. However, sufficient infonnation existed which should have alerted the licensee that info~tion he was in violation of Ia license condition. For example:

    .(1)
    '(1) By letter dlteddated Februlry February 25, 1980 (R. F. Jlnecek-CECa Janecek-CECO to T. A. Ippolito-NRC) the licensee noted thlt     that they did not believe installation of an automatic early warning fire detection system systtlll in the refueling floor lrea was warranted based on low fire loading 10iding Ind the Ibility Ibf1ity to IIIke up water Ind cool the spent fuel pools in th, event of *a loss of either and Unit '5 spent Fuel pool cooling equipment due to fire. This Unit's letter did not request relief from the installation of a refueling floor fire detection systea.systt.. No official officill NRC was issued for this letter.

response WIS (~) By letter dated March 18. 18, 1980 (L. Derderian-NRC to ' M. Antonetti - &age Gage Babcock and Associates - Consul tants Consultants to the Licensee) the NRC referenced a March 17, 1980 telecon record with T. Pickens (CECO) (CECa) in which tilethe follOWing was agreed to concerning Reactor Building Refueling floor fire detection systems for Dresden Units 2/3 and Quads Cities Units 1 and 2: (_) The license was (a) canfina to the NRC that in the IIOst WIS to confirm IIOSt heavy fire loading situations (i.e. refueling periods), necessary to cause loading would not exceed that necesslry the 10iding Cluse structural flilures. failures. (b) The licensee was WlS to confil"ll confirm that structural concrete protection extends froll fl'Oll the floor to saneSOlllt specified height. lessening the likelihood of structural failure. height, (c) The licensee was to recalculate average cOll1bustible combustible . loading subtracting out the pool areas. The licensee could not provide the inspectors with docIIIII!nted documented evidence that these issues were ad~ressed.

• This failure to followup on implementation of a license condition is indicative of a programmatic breakdown which has resulted in

C . aI reduced level of fire protection than was intended to satisfy III.3-42 II 1.3-42 15

Revision 8 April 1992 (\. criterion 3 of Appendix A to 10 CFR 50 and is considered Ia violation of Amendment No. 36 to Provisional Operating License Ho. No. DPR-19. Amendment No. 33 to Facilit~ Facility Operating License No. DPR-25. 10 CFR 50 (237/85-033-04i (237/85-033-04. 249/85-029-04)(DRS).

b. Grvparations

         &rvparations for the Upcoming Extended Unit 3 Outage Separation of n fTTroii"n1ts fTTroiillnits 2/3                              -

During plant toUF'$ tours Ind and in .etings _eUngs with the licensee during the fnspection. icensee agreed to update inspection. the 1licensee upclate their response to the NRC and describe the administrative controls Ind Ind and the Ictions actions that will be necessary to isolate fsolate Unit 1 from fl'Cllll Units 2 and 3 since sfnce Unit 1 is no longer operational but shares COlllllOn COllll1On areas arelS with wfth Units Unfts 2 and 3. The inspectors fnspectors also requested that the licensee describe those administrative controls and actions that will be necessary to separate common areas in Units 2/3 while Unit 2 is operating and

        'Unit 3 is in Inan extended outage.

This is conSidered In Open Item (237/85-033-05; considered an (237/85-033-05. 249/85-029-05)(DRS) pending further review by Region III.

c. Self Contained Breathing Air Supply for the Fire Brigade Section 3.4.D.4(h) of the document entitled -Information Relevant to Fire Protection Systems and Programs-,

Programs-. requires that breathing apparatus using full face piece positive pressure mlsks ~sks that are NIOSH be provided for the fire brigade. approved by NlOSH The inspectors examined the fire brigade Scott Air Pak breathing air cyl inders that were provided on Fire Chart No.2. Four out of four cylinders of these cylinders contained 1800 pounds of lir air pressure. ~cord1ng According to the licensee's staff. Ia lIinillUm lIinilllllll of 2200 pounds of 11r afr pressure should be contained 1n in each cylinder. 2400 pounds of lir air pressure would indicate the cylinder is full Ind and .. mayy provide a 30 lIinute minute air supply for the average fireffre brigade member. The cylinder gauges have' have" range"of up to 3000 pounds of air pressure. a ringe-of Dec_er 1984 three year audit recannencfed A Deceumer reccnnended that a set of written instructions be provided It at the breathing air cylinder filling fillfng station to assure ISsure that the cylinders are filled f111ed properly. Fining Filling of the cylinders is the responsibility of Health PhYSics. Due to tille constraints. the inspectors were unable to contact Health t1_ follow up this concern. Physics to fo11ow concem. Therefore, Therefore. the Ueensee licensee is requested to provide at the site the appropriate acceptance criteria fillin9 breathing .1r for filling air supply cylinders. This is considered an (237/85-033-06. 249/85-049-06)(DRS) pending Regton Open Item (237/85-033-06i Region III .* review of the licensee's breathing air cylinder filling procedures.

d. 300 Pound Fixed Cardox System Supply Tank First Floor. Floor, Turbine

         !iiTIiling Tu'lTding

..Cc. During plant tours, tours. the inspectors observed the following deficiencies on the main

                                 ~in Co, system storage tank located on follOWing the first floor of the turblne buildings.

I II'" 3-43 III'.3-43 16

Revision 8

• April 1992

(

                                                                                                       .'0" (1) The access door to the tank coapressor cOlpressor IOtor was~s .i55ing.
                                                                                   .issing.

(2) The glass cover to the tank's aercoid switch located inside the access door was .;s5ing. The Hcensee licensee had no explanation for these deficiencies, but agreed to take fi_diate

                                ...diate corrective actions.

This fsis considered an Open Itt. It.. (237/85-033-07; 249/83-029-07)(DRS) pending further verification of the lfcensee licensee'sl s corrective actions by Region III.

8. Open It..s Itllls Open itos are utters which have been discussed with the licensH, licensH. which will be reviewed further by the inspector, and which involve SOle action on the.part of the NRC of licensee of both. Open itels itas disclosed during the inspection are discussed in Paragraphs 4.c, 5.a, 7.b, 7.e. 7.c. 7.d.

9. Unresolved Items Unresolved itells itas Ire are .. tters about which IW)re utters lOre fnforMtion inforution is required in order to ascertain whether they are acceptable ite.s, itas. itt.s itels of nonco.pliance.

nonco.pliance, or deviations. An unresolved itel itt. disclosed during ( ... the inspection is discussed in Paragraph 3.c. 3.c *

10. Exit Interview The inspectors ..t with the tlie licensee representatives It at the conclusion of the inspection on October 4. 4, 1985, 1985. and su..arized su.aarized the scope and findings of the inspection. The lfcensee licensee Icknowledged acknowledged the statelents stateHnts udede by the inspectors. The inspectors also discussed the likely*

inforutional content of the inspection report with regard to doclIHnts docu.ents reviewed by the inspector during the inspection. The licensH did not identify Iny any such doc~nts docunents ISas proprietary. proprieta~. On October 21, 1985, in a

21. 1985.

telephone conversation with the licensee additional concerns regarding the lack of fire detectors on the refueling floor were discussed with the licensee. ( II 1.3-44 17

Revision 8 e April 1992 Commonwealth Commonw.alth Edison One Firsl ~Iaza. Chicago. Illinois FirSI National Flla:a. ( Address Reply to: Post Office Box 767 Chicago. Illinois 60690 January 24.24, 1986 Mr. . . . . G. Keppler 1Ir. JJ_s hppler

    .esional aeSional Administrator u.s u.s.* .uclear IUclear aesulatory aeSulatory Commis.ion ReSion aaSion tIlIII aoo.evelt Road 799 Roosevelt       aoad Glen Ellyn, IL     lL 60137 SUbject:

Subject:

Dresden Dre.den Station Units Unit. 2 and 3 aesponse ae.ponse to Inspection Report Raport .os. 50-237/85-033 and 50-249/85-029 lie IRC Poeket Docket &08. .os. 50-237 and 50-249 Raference:

Reference:

Letter from f1"Olll J. G. ~eppler I:appler to Cordell .eed aeed dated December 26, 1985.

Dear 1Ir. I:

eppler: (

 '"               This letter i. in response    reapon.e to the inspection inapection condueted conducted by Xe.srs.

Xe ** rs. J. Holmes and C. Ilamsey Ramsey of your staff .taff betwenbetween September 30 and Oetober October 21. 21, 1985, of activities at Dre.den Dresden Station. !be The referenced letter indicated that certain activities appeared to be in noncompliance with .ac IRC require-

    .ants.
    .ant.. The Commonwealth Idison          Idi.on Company responae reaponse to the .otice of Violation

i. provided in the encloaure.enclosure.

In addition to the response reapon.. to the tba .otice of Violation which Which wwe have provided. bave provided, we bav. have al.o attached our curnnt currant plans plana for resolYinc re.olvins the remainins ramainins concern. concerna tbatthat the inspector identifi.d identified in hi. hia report. The.e Th*** plans plana are de.cribed in Attachment B. Iff ,"OU I you have any further furth.r que.tion. qu.stion. on tbb this atter. utter, pI.... pl.... direct them to this offic office. ** L...

• L. arrar rarrar Director Dir.ctor of IUcle.r IUcl.ar Licen.ins Licensins Attachment cc: .ac IRC R ** ident Inspector - Dresden Re.ident 11711:

1171K

• III.3-45

Revision 8 April 1992 ( ATTACHKEIiT A COIIMOIilWEALTH EDISOIl COKMONWEALTH EOISOIII COKPAllY COMPAIIY

                                               '!Sl'OaS! TO .OTIC! 01" VlOLArIOR PESCRIPTIO. OF     01" vtOLArIO.

VIOLATIO. en 50.48 Ca) requir 10 CPR requi1"** that each .ach operatins op.1"atina nuclear nucl.u powerPOW1" plant have hav. a fire fi1"& protection pl"Otaction plan that .at1.fie. .aU.fia. CriterionCdt.don 3 of Appendix A ... to 10 era en Part 50. It further P.l"t SO. require. fU1"th.1" 1".qui1" ** that the plan plen .hall cte.cribe d** c1"ibe apecific

                                                                                                                 &p.cific featur f ..tu1"** n.c **** 1"y       ry to implement t.plement the prolram    pl"Ol1"am such auch .. a~i.tr.tiv.

admini.t1"ativa control. cont1"ol. and parionnelp.1"80nn.l requirement. l"equil"8m8nt. to limit liait fir. fi1". dames. d8mal. to .tructure.,

                                                                                                                  .tl"UctU1"***

systams.

• y.t.... or 01" component. important t.pol"tant to .afaty .af.ty .0 that the capability to

            .afely
             .af.ly ahut.but down the plant is               i. ensured.

enaul"ed. Section S.ction 3.1.A.1 3.1 ..... 1 of the licans.e's licen*** ** Fira l"i1". Hazard. Hazal"d. Anal,sil An.ly.i. SUbmitt.l, SUbmittal. which fonu fOl"ma part pal"t of the licen** ***e'a

                                                           ** approv.d appl"Ov.d fire  fi1". protection pl"Ot.ction prolraa, Pl"Oll"am **  .tate.

tat** that the licensee licana** has ha. a Fire l"i1". Protection P1"ot.ction Coordinator who.. Coo1"dinato1" Who.e respon.ibili-l"eaponaibili-tie, ti ** include, includ** in part, pal"t. prosram pl"Ol1"am coordination, cool"dination. equipMntequiptMl\t procu~t, Pl"OCU~t. prolrq Pl"Ol1"UI enhancement, enbanc-..t. conductins conductina inspections. inap.ction.. and aupervisins aup.l"Viaina traininl tl"aininl of personnel. p.1".onn.l. c Contrary Cont1"at"y to the abov., abov.. the lican licen.e. *** halha. failed fail.d to conai.tentl, con.i.tently and effectiv.ly

            .ff.ctively ataff     .taff the Fire    l"i1"e Protection P1"ot.ction Coordinator Cool"dinato1" po.ition with the reault 1"&ault that certain  c.l"tain fire           p1"otection equipment va. not inltalled.

fi1". protection in.tall.d. hardware ha1"dva1"e and equipment were W81"a not bains heina prop.rly p1"op.1"ly maintainad, maintain.d. required requi1".d trainin& tl"&inina was va. not completed, compl.ted. and end prompt P1"ompt andend eff.ctive

                                                                                        .ff.ctiv. corrective COrTectiv. **.ction
                                                                                                                         'action vas not taken wal           tak.n for     f01" identified id.ntifi.d deficiencie deficienci ***     **

PISCUSSIOR OF PISCUSSIOH RESPOIIISE TO THE VlOLATIOV 01" RESPONSE yrOLATIOIII At

                    ...t the tm. tiae Section s.ction 3.1 *.....  .1..11 of the Fire l"i1"& Hazarda Hazal"d. AAaly.is ARely.i. .      ,

va. written, wdtten. a Fire l'i1". Protection Pl"Ot.ction Coordinator Cool"dinato1" reported l"&POl"ted to the S,..t. Syat_ S.faty Saf.ty Dep.ttment. Sub.. Depa~t. **quentl:r. quently. the Fire l"i1". Protection Pl"Ot.ction Coordinator Cool"dil'iato1" wasva. tran.ferred t1"an.f.l"l"8d to the qualit,. Quality Assurance

                                .... au1"anc. o.partDant.

Depa1"taent. Shortl,. Shol"tly thereafter, the1"..ft.1". the individual filliD& this po.itionposition retired.1".ti1"&d. CUrrently. CUl"l"lfttly. the Compan,. Company employ. three th1"&e Pire l"i1"& Protection Pl"Ot.ction EDlineer. Bnlin** l"& in the th. General Gen.l"&l Office. Offic.. ~ of the.e th*** Fire l"i1"& Protection Pl"Ot.ction

       'lnsineera Bn&ine.l"& are  a1". in the JUclear  Jlucl"1" Servic S.l"Vic.. rechnical T.chnical DapartMnt, Depa~t. the third        thil"d is ia in the Quality Assurance
                          .... au1"anc. Dapart.ent Depa1"taent and hal        ha. the title of QA Fire        l"i1". Protection P1"ot.ction Coordinator.

Cool"dinato1". Kan,. Kany of the Fira l"i1"& Protection Pl"Ot.ction coordinator'. Cool"dinato1"*. dutiesduti.. lilted li.ted in Section S.ction 3.1.A.1 3.1 ..... 1 area1". currently cUl"l'ently pe-rfol"lMd pe1"fol:"ll8d by the QA Q'" Fire l"i1". Protection P1"ot.ction Coordinator, Cool"dinato1". .ST aST Fire l'i1"& Protection P1"ot.ction EDlineera Bnline.l"& and Station perlonnel. p.1".onn.l. rhul, Thu ** aub_equant aub.equent to the th. initial submittalaubmittal of Section S.ction 3.1.A.1, 3.1 ..... 1. the th. Compan,. Company baa employed threethl"** Firel"i1". Protection P1"ot.ction Inlin.erl Bnlin**1". in the General Offic. in order Gen.1"al Office 0l"d.1" to improva t.p1"ov. the fire fi1". protection pl"Ot.ction prOlr&m. P1"OI1"UI. c. c.-.* I I: 3-46 IIIU-46

Revision 8 Apri 1 1992 (' ( CORRECTm COIU!ECTm ACTIO. TAlCD TAJCEII ABO AIID tHE THE IISULTS USULTS ACHIEVED An 1ST liST Fire Protection In&ineer i. 18 now at

                                                                    .t Dre.den Dreaden approxt-atel,
                                                                                  .pproxisatel:r one day d.y per week we.k to a.a.iat
                         ** i.t the Station.

St.tion. this Thia p.r.on p.raon will continue in this thia capacity c.p.city until the TalkT** k Fore. Forc. report, r.port, which is ia di,cus.ed diacua ** d below, balow, i.ia accepted

                                                                                            .ccepted impl_ted. IItt 18 and implemented.         b expected exp.cted that the 'fa.k    Teak Force will reco.end rec_d a* cour.ecoura.

of action

         .ction that will relieve r.lieve 1ST liST frca from the -.ekly we.kl:r requirement.

CORRECTIVE CORRECTlY! ACTIO. TO U TAXn TAJCEII '1'0 TO AVOID rtJ!THER, rtJIlT!!D VIOLATIO.S VIOLATIOIIS A task teak foree force has been a.aaembl.d

                                               ..embled to examine the vadoul  v.riou. fire prot.ction protection duties duti.. and
                          .nd tasks that have to b. performe~   performa~ on a

• company wide ba.il.

b.da. The taskteak force forc. hal has been baen instructed inatrUcted to report tbair thair recommendations rec_d.tiona for improvement improvementaI in the fire prot.ction protection pro&ram, includin& oraantzational oraeni**tional and staffins atafUns requirement.. requir_ta, to the Vice Vic. Pr . . ident of .... rraaidant IIucl...r Operatione c1. Operationa by b:r April, 1986. April. pAT! WD WEll rut.L FULL COHPLIAlIIC! COHPLIAIICE WILL BI BE ACHI!V!D ACHIEVED Pull compliance compli.nc. will be achievedachi.ved .t aueh time a' at such .a the ta.k t **k force recOMmendations recommendations have beenba.n reviewed, evaluated ev.luat.d and 1mple=ented implemented to the extent n.e**** ry. W. will provide deemed n.c provide. reapon.e addre"in& a follow-up relpon.e .ddre.ain& the talk teak recommend.tions by July I, force recommend.tiona 1, 1986. ('( III'.3-47 II I'.3-47

Revision 8 April 1992 ATTACHMEllT ATTACHI!ElI'r B COHIIOWEAL TIl EOISOIl COMMOHWEALTH EDISOIf COKPAllY COMPAllY

                         .LARS fLAIlS POR FOR KESOLVIlG RESOLVI!G PIKE FIRE PROTIClIOIl PROTICIIOIf ISSUES Thi. attacbment This  attachment responds respond. to the i.au..

i.8U.. identifi.d in the routine

   .afety insp.ction
   .af.ty   in.p.ction conduct.d conducted by Ke Kenr           Holme. and C......y at Dr** den
                                            ** r ** J. Holmes lluclear Power Station on September IUcle.r                        Ssptember 30 throush throu&h october October 21.

21, 1985. Kany llany of the item. identifi.d by the items th. insp.ctor. as a. exampl.s exampl.. of prosrammatic pro,rammatic brbreakdowns

                                                                                    ..kdowns
        .lready been identified durins a review of the had already bad                                                          th. r.sulatory.

r.,ulatory, commit.ent, cod. compliance in the fire prot.ction and code protection area at oUr oiIr op.raUns op.ntins .tation** We f.el that the r.view which we _ had undertaken hal ha. demon.trated our commitment comad en8Urili& that all fire prot.ction tment to enaurlli& protection ffeatur .. at our Itations

                                                                  ..tur..          atationa have been implemented.

bave pr**ent expectation. OUr pre.ent .xpectation. for addr** ,ins ,in, tho ** item. identified id.ntifi.d by the inspector insp.ctor ala. indicative of a prosrammatic pro,rammatic breakdown br.akdown are ala. follows: A. d.t.ction syst Th. fir. detection .y.t... in.talled on the refuelins

                                                  . not installed             refu.lin, floor (Para,raph 7.a of the th. insp.ction report) was     va. ldentified identified a. part SIR item. _1'.

Company'. Appendix. rN of the company'. items were pr**ented conference, no further rea.**** ent.d previously _ t project.

                                                            . . .amant  proj.ct. Sine.

previou.ly at the enforcement furth.r responi.e r.spon.e 18 Sinc. the enforcaaant requir.d at this is required thi. time. time * . ( B. Maintenanc. of fire llaintenance fir. prot.ction equipment follow.: corrected a. followa:

                                                           .quipment and hardware was   va.

(1) Work i. pro&re** to is in prolre chain and lock the bo.ebo *** tation

                                                                                    .tation root valve.. (Para,raph 2d                     insp.ction report) We of the inspection exp.ct that the valv expect               val v** will                      proc.dure revision.

be lock.d and procedure revi.ions compl.t.d by Jusu.t will be completed Au,u.t 31, 1986. (2) A .odification

                            .edification ...vas initiated in 1984 to tn.tallinstall fire d.t.ction and sprinkler .Ylt detection                       .y.t.. alaras alarae in accordance accordanc. with .rPA 72 D. (Para,raph 2e   2a of the inspection insp.ction report). rbi   Thi.*
                  .odification                      8Urv.illanca procedures
                  ~dification and relat.d aurveillanc.              proc.dur.. will be
                  .4 compl.ted and placed in .ervice complet.d the modification are pre.ently
                   ~d of the 'all (3) ItItem.

Fall 19.6

                                                  **rvic. in .ection..

preaently Icheduled 19S6 Unit 2 a.fu.lins

                                                                 **ction.. All portion. of
                                                           .cheduled to be
                                                       **fualins outase.

outq.. he complete compl.t. by the

                            ... identified by the insp.ctor'. review of Technical  T.chnical 8Urv.illanc.. (Parasraph sp.cification surv.illance.         (Para,raph 4 of the insp.ction report) ar. b.ins b.in, rr.,olved
                                          ** olved al a. follows.

a.) Di**el .ir. e.) Fir. Pump surv.illance 8Urv.illanc. procedur.s proc.dur** are in the

                             ** , of beins proc***       heine revi **d al a. a result r.8Ult of our ~A
                                                                               .rPA Cod.

Code

                        .eview. The a.view. Th***  revi **d proc.dure. al"8
                                       ** revi8ed                   are exp.ct.d to be Au,u.t 31, 1986.

implemented by Ausu.t IIL3-48

Revision 8 Apri 1 1992 April ( b.) Wat.r wuppression

                         .uppre.sion .ystem wurv.illanc.
                                                 .urv.illanc. procedure. and pipina ehana..

chana*** .r.

r. in prosreu **s

• result prosre. . . rewlt of our JlPrA IIPFA Cod.

Cod

• a.vi.w.

            ** vi.w. Thrh*** revi ravi**d proc.dures and n.c****ry        r" piplna pipina chanses .re expected chances        expect.d to be implement.d by Auluat   Ausu.t 31. 1986.

c.) ... A IIOdific.tion

                    .edific.tion bi. in prosre..         on our fire d.tection
                                                    ** 011              d.t.ction
            .y.t_. and surv.illanc.
            .y.t_,         wrv.iUanc. procedur.s procedur.. .re beinS bainS rev18e4 reviaed in accordanc.
            .ccordanc. with .,.,   ... 72D as IIFPA        ***a result rewlt of our JlPPA IIFPA CCHSe Cod.

review. Thi. rhi. modific.tion and r.l.t.d surv.illanc. wrv.illanc. proc.dure will be ba complet.d compl.t.d and plac.d pl.c.d in **rvicervic. in

            ..ctions.

ction.. All portions of the modific.tion are .re .chedul.d to be ba compl.t. by the end of the outase. out*S**

                                                     "raIl 1986 Unit 2 ..

the*F.ll fu.lins Rafu.lins (4) In the area (~) .rea of Admini.trativ. Admini.tr.tiv. Control. (P.rasrapb (P.rasr.ph 5 in the tha report) the inspector. cautioned the licen. lican8 ** on a* propo.ed propo**d

                     .,.ldina and cuttina procedure DIIP rev18ion to waldina revidon                                                             that would DIll' 4100-1 tbat inc  Iud. a* provi.ion to facilitat.

includ. f.cilit.t. ALAIA conc.rns conc.rna in blSh hiSh radiation r.di.tion area **

                   .reas.

DKP DIll' 4100-1 will be ba r.vi ** d to clearly require tb. the 30 ~nut. ainut. fir. watch within lin.-of-.isht of the work .rea. .re.. Thi. proc.dure ...was in the proc proc... revi.ion a***** result

                                       *** of revision            rewlt of our D'PA IIFPA Code Cod*****view. This rh18 revision rev18ion 18 pre.ently pre.antly .cheduled for i,C:~" compl.tion by Karcb completion        Karch 14,

14. 1986.

Durina tour. of tb. (5) Durins (5) the plant (Parasraph (P.r.sr.ph 7 of the insp.ctor. report) tb.the insp.ctor identifi.d d.ficient condition. Which which ar.

                                                                                   .re bein&

bains correct.d ** follows:

            ** ) Th.

rh. in.p.ctor identified identifi.d the lack l.ck of refu.linl refu.lins floor d.t.ction *** violation. viol.tion. !b. rh. violation viol.tion notic. indicated indic.ted no furtb.r furth.r respon.e 18 i. n.c.... r" for this it

                                             ****ry               it_.

b.) !h. rh. insp.ctor r.i **d conc.rna conc.rns about

                                                         .bout i.olation i.ol.tion of Unit 1 from unit.

Unit. 2 and 3. and admini.trativ.

                                             .dmini.trativ. control. and action.
            .ction. n.c ****ryr" to sep.rate
                                       .ep.r.t. coaaon comaon .rea.
                                                           .re.. in Unit. 2/3 while whil. Unit 2 is b op.raUns op.ratina and Unit 3 is     b in afteft extended out.se. A stricter out.S..      .trict.r tran.ient combustibl.

combu.tibl. control procedure proc.dur.

i. bainl 18 bainS d.velop.d, d.v.lop.d. and ia i. pre.ently pre**ntly .chedul.d

                                                           .chadul.d for implement.tion impl-.nt.tion by September 30,     30. 1986. ... A cOlnizant     fo~

cosnizant foraan baa has b**n d.sisn.t.d d** isn.t.d to *** isti.t th. fire ..r.hal in timely r.h.l .in correction corr.ction of bous.te.pins bou **k**pina deficienci... d.ficienci... %he rhe Unit 3

            **circul.tion Pipina aeplacement aepl.cement primarily involv.s involv.. the dr,well drywall of the th. shutdown
                              .hutdown unit and does  doe. not .ffect aff.ct coa.on CaaDOn barrien. How.ver.

fir. barriers. Howev.r. a d.taUed d.tail.d MmOrand\a _rand1a eUscuuins di.cu.aina the th. prop.r handlinl handlins of firefir. barrier. barri.r. halbaa been baen di,cus.ed di.cu ** ed with all per.onnel p.r.onnel at the Station St.tion .s** part of the th* w..k17

                                                                          .,..kly "tailsate" "t.ilsate" staff
                         .taff ... tin,.. Al.o procedure
                                 ..etinss.            proc.dure DrPP DFPP 4175 -1.

Fire Barrier III.3-49

Revision 8 April 1992 ( Intesrity Int.srity and Maintenance. llaintenance, hal been be.n revi.ed r.viled to further furth.r clarify the proper prop.r handlins Cd and .. intenanc. of f.ire maintenance f.1re b.rri.re, includins barriers. includinS fir. door door.,

                                                      ** fir. vall.,

valli, pen.tration pen.trltion

                 ** al. for ..

I.al. . .chanical and .lectrical

                                                 .l.ctrical component.,

componentl, and fire fir. damp.re. ne damp.rs. Th. .eparation lap.ration of Unit 1 ia iI beina beins covered COYllrel1 b,. b)" the Appendix App.ndix a review r.vi_ prosr... prosrUl. Thi. Thil infonaation information i. is beins added to th*. updated rir. Fire Hazard. £nal,..i. Anal)".il for Unit. Unitl 2 and 3. c.) A proc.dur. i. il heins beins d.v.loped d.velop.d b,. b)" the "diationl aadietionl Chemiltry Chemistry Depart:ment Dapartlllant which vbich will provide .tandareS. ltandardl for the prop.r refillinl refillins of the SCBA air pacle.. packs. !'hh This procedure wili will be po. ted at the air pacle pOlted pack refillina refillins .tation. ltation. The procedure ia is pr ..antIy .cheduled prelently Icheduled to be illplemented illpl_ted by Jun. 30, 1986. 30. d.) The Th. muins mil.ins door and 11a ** cover slall cov.r have been replaced on the carbon dioxide .y.tem IYlt.. .tora,. Itoras. tank. c. C. !'he Th. in.p.ctor insp.ctor identified technical t.chnical .,.cification ap.cification surv.illance surv.illanc. procedur.. proc.dural that did not incorporate incorporat. appropriate te.tins t.ltins of qualit,.-aff.ctinl qualit)"-aff.ctins param.ter. paraat.rl in accordanc. with de.ien d.lian and

           ,ovemine covemins codecod. requirement..

requirementl. (Para,raph (Paraeraph 4 of the inspector. insp.ctor. report)

• OUrOur re.olution to it.. i t _ in Parqraph Paracraph ..4 of the cc*. .
• D.

in.,ector. inap.ctor. report i. di.cu **ed above. ne Th. insp.ction inap.ction report not ad.quatel,. abov** raport .tat** that admin18trativ. admini.trativ. controls ad.quatal)" control fire prot.ction f ..tures a. di.cu ** control. did

                                                                                     ***edd in
           'arasraph Parasraph S5 of the report. AI indicat.d in our above           abova respon.e r.apon **

to para,raph paraeraph S,5, the weldins and cuttins procedure 18 is heins beins revi.ed to re.olv. ra.olv. the inspector'. insp.ctor'. concern. conc.m. B. E. The i~ection

           !be in LDs, LBaa, ftC inspection.,
                                                        .an,.

insp.ction report .tated that a&n)" d.ficiencie. d.ficienci.. identified insp.ctionl, QA audit. and QA surveillance. surv.illanc** did not receive rec.iv. prompt or .ffectiv

                                      .ff.ctiv*. correctiv. action. nTh*** items        it_

are identified a. identifi.d in 'ara,raphl Parasrapha 2 and 6 of the report. n.ir r ** olution i. a. foll0W8. followa. Th.ir (1) !be The lone lons term t.ra corrective actiona baYe bay. been cOapleted cOapl.ted for the Auxiliarr Auxiliary Blectric El.ctric Equipment aoo. aoam HVAC dAmFer damp.r** (Paraeraph 2. ('aralraph 2a of the report) raport) (2) 'ara,r.ph Paraer.ph 2d di.cu **** ho.e ho *** .tation tation root valv... val v... Our re.olution i.il di.cu ***d above. abov** (3) '.ra,r.ph Paracraph 2. di.cu.... the int.rconnection int.rconn.ction of tha the ** curity

                 .y.tem
                 .y.t.. comput.r with the plant fire det.ction  d.t.ction and sprinkler aprinkl.r
                 .y.tam
                 .y.t.. al.rms. OUr Our re.olution r ** olution ii. * *addr ddr**** d by the propo ** d implement.tion of the 1984 .edification implementation                     modification to inatell install fire fir.

detection d.t.ction and sprinkler Iprinkler alarm. al.rae for .rfAWFPA 72D. ( III.3-50

Revision 8 April 1992 ((' 4 -

                                                  - ot (4)

(ot) P.r.zr.ph Para,raph 21 discusses di.cu.... deficiencies daficienci.. in port.ble portable fire BXtin,uish.r.. .4 extinzuilh.ra. 4 d.dicat.d d.dic.t.d work crew has been .stablbhed

                                                                                      .stablished to .liminate         b.cklo, of fir. prot.ction work ~.st8.
                              .liminat. the backloz                                     requ.at**

U of January As Janu.ry 8, 1986. this

8. 1986, thia backloz b.cklo, has been eliainated.

                                                                                    .1Wnated.

Th. fir. extinzui.h.r

                          !h.          BXtin,ui.h.r discrepanci..

di.crepanci.. are .re tentativ.ly sch.dul.d

                          .cheduled for compl.tion completion of corr.ctiv.

corr.ctive action

                                                                             .ction by Januar,y Janu.ry 31,

31. 1986.

(5) P.razr.ph P.ra,r.ph 2m discuBses di.cu.ses fir. fire briz.d. bri,.d. drill. and traininz. tr.inin,. An ***** ~tamant will be .sad..d. of fire bri,.d. bri,.de drill ** tr.ininz. and practic tr.inins. pr.ctic*****ions.ione. and the tbre.-year three-y..r critiqu** of fire briz.d. independent critique. bri,.d. drill.. !he Th*

                          *****  amant is seB~t       i. pre.ently 8chedul.d
                                                         .chedul.d to be co.pl.ted compl.ted by Auzu.t Au,u.t

31. 1986. -.

(6) Par.zr.ph P.r.,r.ph 2n di.cu **** Pre-fire Pre-fir. plan.. Pre-fir. plan. bav. hay. been d.v.loped d.velop.d and .re in the proce.. proc... of beinsbeinz 1apl~t.d. implement.d. Pull Full impl...ntation implementation i. exp.cted by .. reb 14. 1986. Karch (7) P.r.,r.ph 20 20 dilcul.el di.cu **** hand.-on fire bri,ad. traininz. u bri,.d. trainins. As

                          .tated
                          .tat.d abov
                                   .bov** an .....
                                               *****amant will bb.     . .sad.
                                                                           .d. of fire bri,ade bri,.d.

trainins* traininz* P.r.,r.ph ,6 di.cuB.es (8) P.razr.ph di.cu.... tbe the .pparent

                                                            .pp.rant l.ck of prompt and effective
                          .ffectiv. corrective corr.ctiv** .ctions ctione to problem.

problema identified identifi.d by the pro&r8lll. .. Q4 pro,ram. QA u discu diacus****edd abov.,

                                                            .bov.. a* dedic.t.d work ~re" cr." hal h**

be.n e.t.bli.bed been ** t.bli.h.d to .limin.t. the backlo& b.cklo& of fire . protection prot.ction work r.qu.st r.qu ** t **

r. Th.

The inep.ction report identifi.. insp.ction w.tn.....

                                                               ....Ien.....           .ch.du!inz in the .cbedul1ns (P.r.&r.pha 2m and 20 of the report) ..

of fir. drill.. (Para,raphl u

                                  .bov.. an ***** ...nt of the di.cu ** ed above,                                bri,.d. trainins tbe brie.de     traininz proer pro,r..

will be .. vill sadde. 1171)( 11711t

   ,e III.3-Sl III.3-SI

Revision 8 April 1992 July 7, 1986 DJS LTR: 86-477 TO: J. R. Wojnarowski

SUBJECT:

Review of Commitments Made in Dresden Station Units 2 and 3 Response to Inspection Reports No. 50-237/85-033 and 50-249/85-029

REFERENCES:

1) Letter of January 24, 1986 from D. L. Farrar to J. G. Keppler, Response to Notice of Violation (NL-86-0131).

2) Letter from J. G. Keppler to Cordell Reed, dated December 26, 1985.

As you requested by phone July 1, 1986, the commitments associated with the above-referenced letters have been reviewed. The attached table provides a status update regarding the Dresden Action Items. If there are any questions, please contact R. Whalen at extension 665. Prepared by=-~~~~~~~_____________

*c                                                 R.

Approved by tlJLd-f)/iLfi-D. J:jfScott JOJ1Scott Statton Manager Dresden Nuclear Power Station DJS :RW:hjb DJS:RW:hjb Enclosure cc: J. Achterberg M. Dillon R. Christensen R. Whalen B. Zank D. Adam B. Rybak G. Smith J. McDona1d T. Hausheer R. Hunnicutt S. Becker File/Fire Protection File/Numerical c III.3-S2 .~ \' I

/j (" ~ r* FIRE PROTECTION AUDIT ACTION ITEMS Ac tt ion Item Per Attachment 8, B, Commitment Reference 1 Description Date Current Status Cognizant Person E.4 Fire extinguisher discrep- 01/31/86 Completed on schedule. M. Dillon ancies. 8.4 B.4 Revising DMP 4100-1, cutting 03/14/86 Complete. Procedure was B, B. Geier and welding procedure to approved 2/28/86. insure continuous fire watch 30 minutes after work stops. E.6 Implementation of pre-fire 03/14/86 Complete. On-Site Reivew M. Dillon plans

• was completed 3/13/86',

3/13/86.

   ......         B.S.C. Posted procedure for               06/30/86       Complete. Procedure DRP             L. Burczak w,

(.oJ 8.S.C . refilling SCBA air packs packs.* 1310-11 was approved 6/2/86. I

  <.n w

(.oJ B.

8. 1L Chain and lock hose station 08/31/86 Locks have been purchased and M. Dillon root valves; change valve work is proceeding on Bchedu~e.

schedu~e. checklist as appropriate. 8.3.a. B.3.a. Revise diesel fire pump 08/31/86 Complete. DFPP 4123-6 (2/3 R. Whalen surveillance procedures to Diesel Fire Pump Annual Capa-meet NFPA 20 requirements. city Test) and DFPP 4123-7 (Unit 1 Diesel Fire Pump Annual Capacity Test) were revised to include acceptance curves 6/30/86. These revisions incor-porate items 4.a.(I) through (4) as listed in Enclosure 3 of Reference 2, with the exception that automatic activation test- "';0 "Oro ing Is is covered under operability -s<

                                                                                                                               ~V>

surveillances surveillsnces DFPP 4123-5 and

                                                                                                                               ~o DFPP 4123-1.                                    <0 :::l NCO

-:;./

 . / ..

f (\ \ I,

                                                                   .~.                                              .....--..."
                                                                   ~-

Item Per Action Ite,m Attachment B, Commitment Heference l1 Description Date Current Status Cognizant PereonPerson B.3.b. 1l.3.b. Revising suppression system 03/31/86 DFPP 4114-2 and 4114-3, R. Hunnicutt! Hunnicutt/ surveillance tests to meet Reactor and Turbine Building T. Hausheer NFPA 13 requirements. Monthly Fire Equipment Inspection, will be revised to include a waterflow alarm check on the west pipe sys-tems from the remote inspec-tor's test location. This requires completion of certain modifications, some of which may not be completed until after 8/31/86. An evaluation of this approach is being performed by a quali-

        .....                                                            fied fire protection engineer
        ~
        ~
        ~

to insure that the requirements w, V) are met.

        <..n
        ~                                                                Note: Section 4.b in Enclosure       R. Whalen 3 of Reference 2 also refers to Technical Specification 4.12.B.l.(e),

4.12.B.1.(e), which addresses a triennial flow test of the under-ground mains. DFPP 4123-8 was approved for use 6/30/86, and will be used in place of SP-84 39. However, the inspector's concerns about alarm testing do not appear to apply in this case. E.S Fire Brigade drills and .. 08/31/86 Regarding the frequency of Fire M. Dillon/ training assessment. Brigade drills, it is believed T. Hausheer I ):>:;0)0>::0 that Dresden is committed only S S.*. Becker -0 m (1)

                                                                                                                                 -s<

to the following position from an August 8, 1977 letter from

                                                                                                                                 ~     V>
                                                                                                                                       ~
                                                                                                                                 ..... 0 M. Turbak (NLA) to Davis (NRC).                         \0 ::J
                                                                                                                                 \0 N     ex>

//                                             ;...-....."
                                                                                                       ..- ~,,

(',

                                                -~-

Action Item Per Attachment B, COlllJI1itment Commitment Reference 1 Description DescriPtion Date Current Status Cognizant Person E.S E.5 - (Cont'd) "Fire Drills are conducted monthly in accordance with approved station surveil-lance schedules. The designation of which shift will conduct a specific drill is the responsibility of the Fire Marshal. When a fire drill is conducted by NML the Fire Brigade Leader (Fire Chief) as well as the Fire Brigade. ..* "n Brigade, are evaluated *** Note: Currently the Fire Brigade Leader Isis evaluated during all drills.

       ~
       .w,
       ~
       ~

In process of issuing the W I December 2, 1980 fire protec-tTl tTl tion SER supplement the NRC seems to have accepted the existing drill program since it specifically references the August 8, 1977 Turbak letter as having been reviewed. However, an assessment is being performed of this position. Regarding the hands-on Fire training, implementa-Brigade training. tion plans are under review by

                                                           .,the the Training Department. An implementation plan is scheduled                   :>:>",
                                                                                                               "On>

for development by 8/31/86, ""<:

                                                                                                                - ' . -I.

including a timel1ne timeline for resolv- -In

                                                                                                               ~o ing this 1item.

tem. <0",

                                                                                                               <0
                                                                                                               "'co

("\ r*

                                                     \   ~
                                                                                                   ;'~"

Action Item Per Attachment H, B, Commitment Reference 1 Description Date Current Status Cognizant Person 8.S.b IL5.b Control of transient combus- 09/30/86 A transient combustible R. Whalen tibles. procedure 1s is being developed. and is scheduled for imple-mentation by 9/30/86. Also. the Unit 1 separation concerns are being incor-porated into the Unit 2/3 fire hazards analysis. 8.2, B.2, Installation of detection/ 03/01/87 Modification work is proceed- R. Hunnicutt B.3.c. 8.3. c. alarm system separate from (End of U-2 ing on schedule. the security system and outage) addressing cleaning/ sensitivity testing issues issues.* ~ ~ ~ Ww, I <n (]"I O'l (J)

>>;0 "0

                                                                                                          -s<    '"
                                                                                                          ~ I/)
                                                                                                          ~o In
                                                                                                                 ~
                                                                                                                 ~.

0 1.0 ::::I 1.0 NCO

Tab 4 Revision Revi sion 8 April 1992 1992. DRESDEN 2 &3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspection Report No. 50-249/86006 Title I1I'.4-1 III'.4-1 Inspection Report No. 50-249/86006 dated February 26, 1986. III.4-12 May 6, 1986 CECo letter from D. l. L. Farrar to J. G. Keppler (NRC) transmitting the response to Inspection Report 50-249/86006. II1.4-16 III.4-16 July 17, 1986 letter from D. l. L. Farrar to J. G. Keppler (NRC) discussing Inspection Report No. 50-249/86006.

                                .4-;

III.4-i III

UNITED IT ATEI STATEI Revision 8 NUCLEAR REGULATORY COMMISSION April 1992 REGION III ('"

                                                  '7tI G!.EN
                                                    " "OOSEVEL.

ROOSEVEL T "DAD!lOAD EL.L. YN. ILLINOIS GLEN ELLYN. II.LINOIlIO'37 10137 FEB 2 G r.:: f.?: Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Vice Vi ce President Pres i dent 86 696 Post Off;ce Office Box 767 ~HC'O FEb 27 '~.~

                                                                                 ~~C'O fEB Chicago, IL 60690 Gent1ellen:

Gentlemen: Th;S This refers to the special safety inspection conducted by NRC Personnel of this office on January 28, 29 and Februa~ February 7 and 13, 1986, of circUistancescircuastances associated with a fire in the Dresden Nuclear Power Station, Unit 3 d~ell drywell expansion gap on Janua~ January 20, 1986, and to the discussion of our findings with Mr. D. Scott at the conclusion of the inspection. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. e.:.:: ii (,":" No violations vi 0 lit ions of NRC requirements requi rements were identified i dent i fi ed during duri ng the course of this thi s inspection; however, the you are requested to for.ally fonaally respond to each of the issues identified in Paragraph 3 prior to Unit 3 restart. We will gladly discuss any questions you have concerning this inspection. Sincerely, Sincerely.

                                                                  '2!.!"
                                                                 /2 /'1 ">
                                                                ~ tII ~

W ~-11~;;'-~ I~;/"f- /

                                                                                           ,/.f f

Pape~ello, Director Carl J. pape;{el10, Division of Reactor Safety

Enclosure:

Inspection Report

           . No. 50-249/86006(DRSS) cc w/enclosure:

D. L. Farrar, Director of Nuclear Licensing D. J. Scott, Plant Manager DCS/RSB (RIDS) Licensing Fee Management Branch Resident Inspector, RIll Phyllis Dunton, Attorney General's Office, Environmental Generalis Control Division III.4-1 111.4-1

Revi si on 8 Revision April 1992 U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. SO-249/86006(DRSS) 50-249/86006(DRSS) Docket No. 50-249 License No. DPR-25 Licensee: Commonwealth Edison Company P.O. Box 767 Chicago, IL 60690 Facil ity Name: Facility Dresden Nuclear Power Station, Unit 3 Inspection Location: Morris, Il IL Inspection condUC~U.~ condUC~Uary 28, 29 and Febru.~ February 7 and 13, 1986 Inspectors: D~/' ~~~

                   ..10/

F'i'ii:: e:,*"ii:... I_----'D/ I_--/f¥

                                                                    .2,..,YOV
                                                                    .:lplS/()f.{
                 ~NA~    ./,,1 ~                                   bate
                 ~"~'~

Date

                                                                      ,L J. Patterson J. Patterson '                                     a.fi 2.£2 el,'

e/?6

                   ~

tV£M~,/?!. "~~- bate .(( By: Approved By:. t4

                   ~~t        .~",,I'?'#f ond, Chaf,    1'//J-
                                               -                      ".!i!:!!..07{i~
                                                                      ..L.-!t>J_
                                                                      ~;;It.J~

Operational Programs Section t::"Da""'lt~e"';""""'-=~-- Date Summaev Inspection Summa~ Ins ecton on Janua 28 29 and Februa 7 and 13 1986 Re ort No. 50-249 86006 DRS Areas nspected: Announced special safety inspection conducted to review potential damage to the facility originating from a fire in the drywell expansion gap on January 20, 1986. The inspection involved 60 inspector-hours by six NRC inspectors. Results: No violations or deviations were identified. III.4-2

Revision 8 April 1992 ( DETAILS

 \.

1. Persons Contacted

       £lli D. Scott, Station Manager R. Flessner, Services Superintendent J. Brunner, Assistant Services Superintendent T. Hauser, Fire Protection Engineer R. Mirochina, SHED SNED D. Wilgus, SHED SNED M. Dillion, Station Fire Marshal J. Schrange, Health Physicist Rolf Jensen and Associates J. Klien, Consultant E. Hare, Resident Inspector S. Stasek, Resident Inspector L.

l. McGregor, Senior Resident Inspector R. landsman, Region III Project Manager

2. JanuarY 20, 1986 Drywell January 20. Drywel1 Expansion Gap Fire

a. Apparent Origin of the Fire At approximately approxiMately 0830 hours on January 20, 1986, with Unit 3 shutdown and defueled, an air arc cutting activity began on containment pipe penetration Ho. ("B"11 reactor water cleanup system pipe) inside the No. 113 (118 reactor water cleanup system (RWCU) uB "B"Il .heat exchanger room.

roOGl. At 0905 hours workers in the area observed smoke in the vicinity of the pipe penetration. The shift engineer's office and the control room rOOM were notified at approximately 0916 hours.

b. Initial Response The fire watch for the air arc cutting activity apparently discharged a dry chemical extinguisher on or in the vicinity of pipe penetration No. 113. Subsequently, a fire brigade leader arrived to investigate the fire and determined that the fire had been extinguished.

At.approxi~ately At approximately 1000 hours, the station fire aarshal was notified by the shift engineer of smoke in the Unit 3 reactor building. At by*the 1004 hours the reactor building ventilation system, which had been turned off to support Standby Gas Treatment SysteaSystem testing, was turned drywell, all on to remove smoke from the Unit 3 reactor building and dr,ywel1, personnel were evacuated from the Unit 3 torus and drywell areas, and air samples were taken to verify the quality of air for personnel safety. . 2 III.4-3

Revision 8 April1 1992 Apr; engineer contacted the station fire marshal At 1030 hours. the shift en9ineer and informed him that the smoke was clearing from the Unit 3 reactor building. Apparently th~ fire brigade leader and station construction concluded that the problem was under und~r control because the fire watch had earlier discharged a dry chemical extinguisher on or in the vicinity of the pipe penetration in the RWCU heat exchanger room and smoke was being cleared from the reactor building and the drywell by ventilation system. the reactor* building ~entilation At approximately 1120 hours personnel were allowed to reenter the. drywell. At 1130 hours station technical staff personnel discovered drywel1. drywell in the viCinity a hot spot in the drywel1 vicinity of penetration No. 113. Workers complained of intense heat 4 to 5 feet lway away from the drywell hours. .11 steel liner. At 1155 hours, all personnel were again evacuated from liner. A construction the drywell because of the overheated drywell lfner. staff person took general use (not calibrated) pyrometer readings in the vicinity of pipe penetration No. 113 on the inside of the drywell liner (unexposed sfde) side) between 1230 and 1315 hours. The highest . 440-450*D F. reading recorded was 440-450 fire marshal to The heated drywell liner condition alerted the station ffre investigate what could be burning on the other side of the drywell liner. His review of Section 5 of the Dresden FSAR identified the drywell expansion presence of polyurethane foam installed inside the dr~ll gap between the steel liner and the concrete shell.

c. DryWell Expansion Gap Design Drywell The outer surface of the steel drywell liner 1s is enclosed in 8 feet of drywell structural and shielding concrete. Thermal expansion of the drywel1 liner as a result of normal reactor operations will cause the liner to expand both radially and vertically. To accommodate this expansion.

construction. an expansion gap was provided between the during construction, structural concrete and the drywell steel liner. The sizing Of the expansion gap was based on the maximum drywell steel liner temperature following a postulated loss of coolant accident.

d. DryWell Expansion Gap Materials Used to Fill the Drywel1 To maintain sufficient space for liner expansion, expansion. prefabricated polyurethane foam sheets were installed over the entire liner exterior surface. Epoxy impregnated fiberglass tape was applied over all joints in the foam and one-fourt~ inch to 3/8 inch fiberglass-epoxy prefabricated cover panels were installed over the foam panels. The fiberglass panels were made of fibrous glass in chopped fiber form isophatal1ic resin as a binder.

with an isophatallic . sitE on mock ups of the drywell steel Tests were conducted at the Site liner/polyurethane foam/fiberglass panels to determine their displacement from thE pour of structural and shielding concrete. The thar. test results showed that the fiberglass was displaced less thar, 1/4 inch from the pouring and curing of the concrete. Therefore. it

• c. . was assumed that the drywell expansion gap design space was maintained during construction.

3 III.4-4 II 1.4-4

Revision 8 April 1992 ( e. Determination Made that Polyurethane Foa~ Panels Were Burning Inside the Drywel1 Drywell Expans10n Gap As a result of the station fire marshal's review of Section 5.2.3.6 of the FSAR, he determined that hot sla9 slag (~lten metal) from the air arc cutting activity on pipe penetration No. 113 in the RWCU heat exchanger room had come in contact with and ignited the polyurethane foam material in the drywell expansion gap. The typical dr,ywel1 drywell pipe penetration dedetail. tan . (figure S.2.3.27 of the FSAR) shows a 2 inch gap penetration sleeve, which provides between the pipe sleeve and the penetratioll providu a di rect path to the polyurethane foam material. Furthermore, direct Furthennore, the drywell expansion gap is not air tight. The fiberglass panels installed over the polyurethane foam ~terfal material do not fo~ fona a barrier that will exclude air from coming in contact with the polyurethane foam material. The 45 degree angle thlt that pipe penetration No. 113 is installed through the drywell adds credibility to this hypothesis IS as to the origin of the fire.

f. Extinguishment of the Fire Since the fire was determined to be in a concealed space that was impossible for the fire brigade to reach.reach, the station fire marshal directed the fire brigade leader to start applying water from a 1 1/4 inch (3/4 inch inside diameter) rubber hose (supplied by the demineralized ~ater water system at 100 PSI) to the 2 inch gap between the pipe sleeve and the penetration sleeve on penetration No. 113. This action was initiated between 1230 and 1300 hours. As the fire marshal was not certain that water applied through this penetration

    ~rshal fire, additional hose streams supplied by the would extinguish the fire.

fire water system at 100 PSI were applied above and adjacent to the penetration (pipe penetration Nos. 133, 122. 122, 144 and 143). At 1330 hours, the licensee decided to monitor

                                                 ~nitor the drywell liner drywell. At 1700 hours.

temperature on the inside of the drywel'. hours, inside drywell liner temperatures were recorded at 140, 110 and 90° F. At 1730 hours, the licensee's corporate fire protection engineers and the station fire marshal considered the fire to be extinguished due to declining inside drywell liner temperatures. At 2100 hours. hours, inside drywell liner temperatures were determined to be normal and the dr~,ell application of water to the drywell expansion gap was discontinued. No offsite fire department assistance was requested and no emergency event was declared by the licensee at any point during this event.

g. Potential Damage Resulting From the Fire At the time of the inspection the licensee had not determined the extent of damage resul ting from the fire. In two principal areas resulting inside the drywell (approxi~tEly (approximately 10 feet in diameter and 10 feet apart), charred.

apart). charred, discolored, blistered or burned away paint was drywell liner. The drywell steel liner is visible on the drywel1 approximately 1 1/8 inch thick carbon steel. c 4 III.4-5

Revision 8 April 1992 ( Polyurethane foam materials are synthetically produced from glycols

 ... and diisocyanates. It has been established by actual fires and certified fire testing laboratories that urethane foam ~terials materials ignite easily and burn vigorously with the production of dense black smoke and a very black.

black, viscous melt product which can burn with the intensity of a flammable 11quid liquid (Reference Underwr1ters Underwriters Laboratories Inc. and Factory Mutual laboratories Laboratories Inc. 1969-74 studies on the Flammability of Cellular Plastics). Burning polyurethane ~terials Fla~bi'ity materials also produce corrosive and toxic oxides of nitrogen, together with other toxic gases and corrosive that are haMiful hanlful to Metals. letals. It appears that this fire began some time after 0830 hours, when the air arc cutting activity began on pipe penetration No. 113. It burned with some intenSity intensity and it is suspected that high temperatures were reached inside the drywell expansion gap. It is not known how much polyurethane foam material wasWlS consUied by the fire or how far the fire spread vertically or horizontally around the drywel,. drywell. The 41 hours burn time from 0830 hours to 1300 hours (when water was first applied through penetration No. 113) indicate that substantial burning may have occurred. Apparently, a substantial amount of water wasWlS applied to the drywell expansion gap to extinguish the fire (approximately 500 gallons per minute (GPM) for 8 hours or 240,000 gallons). However. However, according to the licensee, only 20,000 gallons of excess water

                                                       ~ter was removed from the torus basement by the radwaste system the day after the fire.

The licensee did provide the inspectors with a draft copy of proposed work to be performed by Sargent and Lundy (SIL (S&L Project No. 7368-30) to evaluate the integrity of the Unit 3 drywell for affects from the fire. This evaluation did not appear to consider some of the specific NRC concerns detailed in Paragraph 3 of this report and is not scheduled to be completed until March 3131,t 1986.

h. Emergency Preparedness Implications The inspecto~ reviewed records associated with the event; interviewed several available persons knowledgeable of the event; and reviewed the Station's Emergency Action Levels (EALs) and the notification requirements of 10 CFR 50.72 for applicability. The event was ~s not claSSifiable classifiable as an emergency per the current EALs for the Fire Condition (No.5) for the following reasons: offsite fire fighting assistance was not requested; eqUipment was not degraded such that a assi~tance Limiting Condition for Operation (LCO) required a reactor shutdown.

equipment was not degraded such that a cold shutdown or hot shutdown could not be achieved or maintained; maintained. and required safety systems were not potentially affected. Since all fuel had been removed from the reactor vessel for some months, there was no need to be able to achieve and maintain shutdown and no reactor safety systems were required to be in operation. The event was not classifiable cla~sifiable as an emergency per the current EALs for the "miscellaneous* Condition (No. 18) which was worded as follows:

        "any other conditions of equivalent magnitude to the criteria used to define the accident category as determined by the Station 5

III.4-6

Revision 8 April 1992 Director. Director."II The Unusual Event EAL for Condtion No. 18 listed a number of circumstances that warranted increased awareness on the part of State and/or local offsite offici.ls. officials. The Alert EAL for Condition No. 18 listed severa' several circUistances which warranted precautionary activation of the onsit.onsite Technical Support Center (TSC) and near site ~rgency Operations Facility (EOF). The Site Area Emergency EAL for Condition No. 18 addressed activation of these Emergency Response Facilities, radiological 8Onitoring IOnitoring teams, and precautionary notification of the public near the site. The Genera' General Emergency EAL for Condition No. 18 addressed an i.minent flminent core melt situation. No EAL associated with Condition No. 18 was applicable to the fir.fire incident. Since no EAL was applicabl., applicable, an ..ergeney

                                         ..ergency declaration and activation e.trgency Plan (GSEP) did not occur.

of the Generating Stations e.ergency Consequently, initial notifications of the Illinois o.part.ent of Nuclear Safety and Illinois e.ergency

                                    ~rgency Services and Disaster Disastlr Agency were neither required nor perfoMied.

perfonlld. SI.ilarly. Si.ilar1y, Initial fnitial notification of the NRC Operations center was not required per 10 CFR 50.72(a); however, the licensee did notifY the Station's Senior Resident Inspector between 4 p... p.** and 5 p... p*** on January 20. That individual informed his supervisor. Neither the licensee nor the aforementioned Region III personnel dee.ed de..ed it necessa~ necessary to promptly notify the NRC Operations Center per the require.ents of 10 CFR 50.72{b) 50.72(b) or (c). Due to the extensive nature of .. intenance being performed on the Unit 3 reactor coolant systea, systt., and the fact 'e 1(: that the vessel had been completely defueled for sa.. S08e .anths, 8Onths, regional emergency preparedness staff have also concluded that the requirements of 10 CFR 50.72(b) and (c) were not applicable to this situation. The wording of the Unusual Event EAL for Condition No. 18 was not in close agreement with regulatory guidance found in NUREG HUREG 0654, Revision 1. The licensee' licensee's5 EAL stated, in part, that ...-a condi~;on condi~ion that warrants increased awareness on the part of the State and/or local offsite officials." officials.- Relevant regulatory guidance for the Unusual Event classification-states, classification* states, in part, that "other

                                                              *other plant conditions exist ,that warrant increased awareness on the part of Ia plant operating staff (emphasis added) or State and/or local offsite authorities. IIIi During the course of the licensee's response to the fire incident, there were a nu.ber of ..       etings in the TSC involving
                                                .eetings teChnical staffi personnel were evacuated Station management and/or technical fro.

from the reactor building for a ti"i ti.e; the licensee's General Office was informed of the incident; and personnel .ade ..de repeated entries into the d~ll drywell to obtain temperature readings to help determine whether the fire still existed. There WIS was clearly increased awareness and activity by plant operating and other plant staffs in response to the fire. Had the Unusual Event EAL for Condition No. 18 included the phrase "plant operating staff,"staff,n per the regulatory guidance, there would be no question whether or not the NRC Operations Center and appropriate State agenCies agencies needed to be require8lnts of promptly informed of the fire incident, per the requireaents 10 CFR 50.72 and 10 CFR 50, Appendix E, Paragraph IV.D.3. ( 6 III .4-7

Revision 8 April 1992 Therefore, to prevent recurrence of any uncertainties regarding the need for the licensee to pro~ptly informinfo~ the NRC Operations Center and appropriate State agencies of significant responses by Station operations personnel to abnormal conditions onsite, the phrase staff"n should be "added "plant operating staff *added to the Unusual Event EAL for Condition Ho. No. 18.

3. NRC Request For Information To Be Provided By The Licensee Prior To Unit 3 Startup From The Current Outage tn In view of the damage that aaymay have occurred occu~red to the d~11 d~ll steel liner, the structural and shielding concrete.

concrete, electrical and pipe penetrations, or other structures and equiPlent required for safe operation of the Unit, the licensee is requested to provide to Region III a detailed assess.ent assess.. nt of this event that will include a confiraation of short ten. teMi and long term te~ cOlponents. This operability of the affected structures, syste.s and/or cOIponents. assess.ent assess.. nt .ust include an ,valuation evaluation of the following concerns for Region III and Office of Nuclelr Nuclear Reactor Regulation review prior to restart frOll the current outage: of Unit 3 fro.

a. Detailed Chronology of the Fire Event Provide a detailed chronology of the January 20, 1986 Unit 3 dr,ywel1 drywell expansion gap fire occurrence and describe the sequence of events depart.ent assistance was that led to the decision that offsite fire departant not needed.

c. b. Duration and Intensity of the Fire Determine the duration, physical extent, and intensity of the fire and include in this assess~ent the highest ..tal and concrete temperatures reached during the fire. If no systematic approach was taken to temperatures reached during the fire, deteraioe record actual teaperatures deteMiine the highest temperature that the steel and concrete structures .ay have been exposed to based on published (f.e. (i.e. Underwriters laboratories Laboratories Inc., Factory Mutual laboratories Laboratories Inc.) free burning polyurethane foam calorific heat values for a fire of this duration. Provide an estimate lIaterial properties of the esti~ate of what changes occurred in the material steel.

steel, concrete, electrical and pipe penetrations, drywell penetration wells and other affected equiPlent equi~nt or components. coaponents. For the nonaal nOMial operating and accident condition, dete~ine deteMiine the temperature profile through the d~ll drywell steel liner with and without polyurethane present in order to show any changes in d~11 d~ll expansion fro. the original design. PerformPerfo~ a structural analysis Ina1ysis d~ll steel Jiner which evaluates the state of stress of the dr.ywell liner during the fire and compare this with the yield strengths of the material. lIaterial. 'e 7 II I .4-8 III.4-8

Revision 8 April 1992 ( c. Drywell Expansion Gap Corrosive Species Introduced Into the OrXWe1l Determine the type and quantity of corrosives that were introduced into the drywell expansion gap as Ia result of the fire and its extinguishment. Determine the short and long tenn term effects of these drywell steel corrosive species on the structural integrity of the drywel1 liner structural and shielding concrete, electrical and pipe penetrations, drywell penetration welds and other affected equipment and components.

d. Effects of Spilling Spalling Concrete and Polyurethane Residue Remaining Inside On'we 11 Expansion Ins i de the Drrwe11 Expans i on Sap .

Determine the effects of polyurethane and fiberglass residue as well as -hard "hard spots* that may have been created by spalling sPI"ing concrete into the drywell expansion gap. Detenaine Determine the effects of potential *hard "hard spots" on the drywall drywell steel liner under pressure and temperature loads dUfing during nOMmal nonmal operating and accident conditions and determine the compressive strength these -hard"hard spots* spots" must have to be of concern.

e. Amount of Water Applied to the Drywell Expansion Gap to Extinguish Fire the flre Determine any thenmal thermal shock that rmy may have occurred to the drywell steel liner and determine the amount of water used to extinguish the

,( drywell expansion gapfirei how much of this water was removed; removedi how much remains unaccounted for and what actions will be taken to remove any remaining moisture

                       ~oisture in the drywell expansion gap or in the surrounding structural and shielding concrete.

f. Basic Sasic Drywell Liner and Structural and Shielding Concrete OesignDesign Furtctions Functions Determine to what extent (if any) the fire ~y may have otherwise degraded the drywell steel liner's ability to provide a barrier which controls the release of fission products to the secondary containment.

Determine to what extent if any, the fire may have otherwise degraded drywell electrical or pipe penetrations and the structural and shielding concrete design functions.

g. com~liance Com~'iance with the Safe Shutdown Requirements of Appendix R to 10 fR FR Part 50 Detenmine Determine the effects of a fire of this nature on safe shutdown capability as prescribed 1n in Section III S.2 G.2 of Appendix R to 10 CFR 50. During normal operation, this section requires redundant cables, including non-safety circuits that could adversely affect safe shutdown capability that afe are located in the same fire area outside of the primary containment, to be separated by a 3-hour fire barrier; be encased in a I-hour l-hour fire barrier with automatic fire detection Ind and suppression installed in the fire area; areai or be separated by a distance of more than 20 feet with no intervening combustible or fire hazards

( with automatic fire detection and suppression installed in the fire 8

III.4-9 IIL4-9

Revision 8 April 1992 area. For normal operation of both Dresden Un1tsUnits 2 and 3. explair, 3, explain how such electrical cables and circuits passing through the drywell expansion gap are in compliance ""ith lOith the requirements of Appendix R so thbt thet a fire of this nature will not affect safe shutdown capability during normal operations.

h. Potential Pot~ntial Repairs Needed Determine the need for repairs (if any) to the drywel1 drywell steel liner, concrete. electrical and pipe penetrations or structural and shielding concrete, other affected equipment a5as a result of the fire. Include in this assessment a time frame for completion and the impact of such repairs or. nora~l noru~l reactor operations.

i.

f. ~J~.s~.9!_!,!~!!r

         ~J~_S_.9!_!l!~~r   and Polyurethane Residue Samples Provide the results of any and all extinguishing water and fire residue samples collected as a result of the fire for NRC review.

j. Corrective Actions Taken to Prevent_~~pccurreps~

Prevent_~~pccurreEf~ Describe in detail the corrective actions that Will-be will*be taken to prevent fires involving polyurethane material in the drywell expansion gap, including interim measures currently in place.

k. Provide an assessment cfof the extent and results of the radiolytic

( and thermal decomposition of I.t~rials J~terials in the drywell d~ll expansion gap in Unit 2 and an estimate of the effects of such decomposition on fire potential and containment structural integrity.

1. Provide a list of other plant locations where polyurethane or other combustible foam materials are installed in concealed spaces.

Identify whether these materials were explicitly addressed as part of our fire hazards analysis. Items a through 1 above will be tracked as an open item (50-249/86006-01(DRS)).

m. Emergency Preparedness Concerns "plant operating staff*,

Add the phrase, *plant staff", to the Unusual Event Emergency Action level Level for Condition No. 18. This is an open item (50-249/86006-02(DRSS>>. (50-249/86006-02(DRSS)).

4. Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC.

NRC, the licensee. licensee, or both. Open items disclosed during the inspection are discussed in Paragraph 3. 9 III.4-10 III.4-IO

Revision 8 April 1992

5. Exit Interview

(

    '-. The inspectors met with licensee representatives at the conclusion of the inspection on February 7, 1986, and summarized the scope and findings of the inspection. The licensee achknow1edged achknowledged the statements made by the inspectors. The inspectors also discussed the likely informational content of the inspection report with regard to documents reviewed by the inspectors during the inspection. The licensee did not identify Iny any such documents as proprietary. On February 13, 1986, 1n in a telephone conversation with the licensee, additional concerns regarding compliance with the requirements of Appendix R to 10 CFR Part SO were discussed with the 11 licensee.

censee. I( ((~ 10 111.4-11 III.4-1I

de, -

                                                                                        '2 dt:'t     a L nAI
                                                                                                           /)A/~

Commonwealth Commonw.alth Edison Revision 8 One F,rSI FIrS! Nahonll Nallen.1 Pllza PIIZa ChlCI90 IIhnOIs illinOIS April 1992 Address AddreSS RePly Reply 10 POS! O.hce Box 767 POS' Office Chlca90. illinOIS 60690 ChIcago. IIhnOls May 6,

                                                            "-1y 6. 1986

( Kr. JUles Mr. James G. lCeppler leglonal Keppl.r _ Regional Administrator Nucl.ar Regulatory Commission U.S. Nuclear Region tIl III 799 Roosevelt Road Glen Ellyn, Ellyn. IL 60137

Subject:

Dresden Station Un1t Unit 3 Response to Inspect10n Inspection Report No. 50-249/86-006 NRC Docket No. 50-249 Reference (a): Letter from C. J. Paperiello to Cordell Reed dated February 26. 1986.

Dear Mr. Keppler:

C." C:.' This transmittal is in response ~o Jco the inspect~n conducted by your staff on January 28, 28. 29 and February 7 and 13, 13. 1986 of circumstances associated with the January 20, 20. 1986 fire in the Dresden unit 3 drywall drywell expansion gap. Although no violations of NRC requirements were identified during the inspection. reference (a) requested that we respond to the open items identified 1n in Section 3 of the Inspection Report. . The enclosed report provides an overall evaluation of the fire and its consequences. The Appendix to the report specifically addresses the open items from the Inspection Report with the exception of item 3a. 3a, Emergency Preparedness Concerns. We are currently reviewing this ita itfllll in the General Office and will provide a response at a later date. Section VII of the report provides a Fire Hazards Anllysis Analysis of the expansion gap and provides the basis for an exemption to Appendix I Section III.G.3. An exemption request 1s is currently being prepared for submittal to NRI. NR1l. 'c, . III.4-12

Revision 8

                                                                     -         April 1992 Rr.
   ~. J. G. Keppler                                                May 6,

6. 1986 If you have any further questions on this aatter,

                                                           .atter. ple.se please dlreet direct th.

thl!!l to this office. ~o

                            ~.                            -.

Very tr~ly yours yours.*

                                                               ~

(Parrar D. L. Parrar Dlreetor Director of Nuclear Lieensing Licensing 1m Attachment

   ~ttachment ec:

cc: H. R. Denton - NRR R. ~. A. Gilbert - NRR NRC Resident Inspector - Dresden (co l660K 1660K III.4-13 II 1. 4-13

Revision 8 199~ April1 1992 Apr; May 1, 1986 I \

Subject:

NLA Letters NL-86-0290 (R.B". (R.B*. Bevan Report on NRR Investigation of Dresden unit 3 Drywell Drywel1 Fire), NL-86-0324 (February 25, 1986 Letter from J.A. Zwolinski

    -,-.   ~"
           ~.

ad5 NL-86-032S to D.L. Farrar), a~ NL-86-0325 (February 26, 1986 Letter from C.J. Paeriello to C. Reed)

Attachment:

Evaluation for the effects of the Dresden Unit 3 Polyurethane Fire Mr. J.R. Wojnarowski: 861655 Attached is our response to the subject documents. This report was planned prior to-NRC requests for information and is submitted here as requested by R.B. Bevan at a site visit to Dresden Station on February 6, 1986. 6,1986. It has subsequently been expanded to address the concern of whether Mark I Containments meet the separation criteria of 10CFR 10CFR SO, 50, Appendix R R as requested in the February 25, 1986 letter from J.A. Zwolinski to D.L. Farrar. Sufficient bases is presented in this report to justify an exemption to Appendix R. An exemption request will be submitted separately. In the February 26, 1986 letter from C.J. Paeriello to C. Reed the NRC requested additional information outlined in paragraph 3 of the accompanying inspection report. An appendix to the attached report addresses all questions specifically except 3m., EmergencY EmerqencY Preparedness Concerns. This item is being addressed by Dresden Station separately. ai. dJ~~ Ot.. dJ~- D.L. Wilgus Approved: f;;f.lt:~ J~~~td:~ Dresden/Quad Cities Project Engineer DLW/rr 7563D cc: FF.A.

          .A. Palmer T.J. Harkabus D.J. Scott E.R. Zebus D.L. Sanderson               II 1.4-14 III.4-14

AMENDMENT 11 i JUNE 1998 I, The referenced report, "Evaluation of the Effects of the Dresden Unit 3 Polyurethane Fire," is found in FPPDP Volume 13, Section X.II. X.ll. ( 111.4-15

Revision 8 c e Commonwealth Commonwe.lth Edt.on Addr... Reply to: POll Address Chicago, Edison . 72 West Adams Street, Chicago, ChieaOO. Illinois Post Office Ottice Box 767 lliinoia 60690 - 0767 Chicago. Illinois July 17. 17, 1986 April 1992 1Ir. JUleS G. Keppler fir. JUleS legional Regional Administrator u.s. Nuclear Regulatory eo..l1s1on eo.aission Region III 799 Roosevelt ao&d Road Ellyn, IL 60137 Olen Ellyn.

Subject:

Dresden Station unit 3 Drywell Drywel1 Plre-COrrection Pire-OOrrection to avaluation leport Pire lValuation Report NRC Docket Mo. No. 50-249 References (a): NRC Inspection .eport Report No. 50-249/86-006. (b): Letter fra. fro. D. L. Parrar to J. o. Keppler dated Kay 6, 1986. Dear 1Ir.

           ~. Keppler:

c* 'file reference (a) InspecUon

               'fhe of the Dresden unit 3 drywe       drywell Inspection Report ctoa.ented 11 fire and identified in the report. our reference (b) trans.ittal
                                                                          ~nted your ataff's    staff's review arid requested we respond to c:cncems transaittal provided our concerna response in the fora of a report documenting       dOCUDenting our evaluation of the fire. we              We have recently beca.ebeCODe aware of a          e condition at Dreaden Unit      unit 3 Which which conflicts stat_nt ude in our report.

With a state.ent on page 83 of the reference (b) report, On report . .... a response to an waC aC question, the following atat question. stat_nt...nt linwas .. ude With regard to the use of

d. with polyurethane in other plant locations:

         *Both polyurethane and polyethylene have been used .
         -Both                                                                               a fl11er
                                                                                        . .. . fl11ar
         .. terial at the top of"'bloek uterlal                         of'bloek waU. valls and polyurethane 1s          111 used to
         .eal seal penetrations. None of the ,block               *block .,.Us
                                                                      -Us areara considered oonaldared rated fire barrier.

barriers and in those tho.. vallswalls that use polyurethane . ... . a penetration .eal. seal. either the wall vall 1s 111 not a rated barrier ct. 01'. if it ia, is. the polyurethane haa has been bean replaced witb With *a fire rated oc or noncaabuetlble .. noncoabustlble uterlal*. terial-. ftal

               'fIIis**stat_nt tat...nt is not caapletely  cc:.pletely KCUrate ac:c:urate .. tbethe polyuret.bane polyurethane ....                      .-.. - ..... '.,.
                                                                                                                                          ," ~

not been replaced in aU all Appendix JtR rated barriera. barriers. !his'f11111 ._ . 41scovered d1aooverad recently at Dreaden elevatlon 58" Dreacten at elevation 519'-0* in the unit unlt 3 reactor aul14lng building during ongoing Appendix. Appendlx R waltd-owns. valkdowna. As a re.ult result of this d1aoovery. _ are th1ll dhcovary. currently conducting additional walkdowns _lkdowna to identifyidentlfy an, any other ar ... ~re arau previous walkdowns .. prevlous "alkdowns uy y not have identified ldentified polyurethane -.chanical ..chanica! penetration seals in Appendix R rated barriera. berriers. ( (. ~.~ _. -~- !IS!

                                                                                                                            ?.            eiii:a

!CIS! :ii:o rrI.4-16 _.-,"_ .;.. . -

                                                                                                                        ~::   7_~-:'-==-"_"*

Revision 8

      ..                                                                               April 1992

( Mr. J. G. Keppler July 17 1986 All polyurethane llechanical penetraUon penetration seals discovered in AppencU.x Appendix R rated barriers will be replaced With appropriately rated ... !be., ls. !he.e

                                                                           ..al..

activities will be completed in accordance with our existing schedule activitles schedple for

         .ealing Appendix JtR penetrations at Dresden (cc.pleUon          and of the next (cx.pletion by end inteda. we will 1JIpl Unit 2. refuel outage). In the tnt.ria.               illpl_nt        _
                                                                    ...nt the AM co.pensatory cx.pensatory ..asures we've co.altted co.aitted to for other outstanding Appendix *R 1IOd1f1caUons.

lIOdif1cations

• Ve We will viII also be conducting addit10nal additional walkdOMnl valkdowns at Dresden Dreeden Unit 2 and OUad Quad Cities un1ts Units 1 and 2 to identify any .1ailar

                                                            .lailar applications of polyurethane in Appendix R  R rated barriers. Any prObI..s prObleas identified*will identified*Will be resolved in the same .anner aanner as described above.

We have discussed this situation

                                             .ituation with NRR (I.

(R. Gilbert. J. Stang) stang) during a telecon on July 14. 1986. During that call, call. IIRR NRR concurred with our proposed resolution of this issue. If you have any questions regarding this transaittal, pl...e contact transaittal. pleas. this of fice

• office.

. ( .*.. Yo<y '" y -"d.. ~ . very tru y yours.~ ......

                                                    . L. 'arrar Parrar Director of IUclear UUclear Licensing la cc:   H. R. Denton - DR R. A. Gilbert - NRR I.                     "'

R. B. Bevan - DR I. NRR NRC Resident Inspector - Dresden KRC XRC Resident Inspector - Quad Cities 1855K l855K ( III .4-17

Tab 5 Revision 8 UNITED STATES April 1992 NUCLEAR REGULATORY COMMISSION (" ( ,It REGION III

                                        ?It RDO$EVEL ROOSEVELT  T ROAD GLEN ELLYN, ILLINOIS 1013, 10137 t.\.

Docket No. 50-237 SO-237 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed 873633 Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen: This refers to the routine safety inspection conducted by S. G. Du Pont and P. D. Kaufman of this office on October 23 through December 8. 8, 1987. 1987, of activities at Dresden ~uclear Power Station, Units 2Z and 3 authorized by Operating License No. DPR-19 and No. DPR-25 DPR-2S and to the discussion of our findings with Mr. E. Een1genburg Eenigenburg and others at the conclusion of the inspection. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. No violations with NRC requirements were identified during the course of this inspection. In accordance with 10 CFR 2.790, of the Commission's regulations, a copy of this letter and the enclosure will be placed in the NRC Public Document Rpom. We will gladly discuss any questions you have concerning this inspection. Sincerely. Sincerely, (--x(~~/47 (71'(~'/<7 hi ef W. L. Forney, thief Reactor Projects Branch 1

Enclosure:

Inspection R~ort Report SO-237/8703S(DRP); No. 50-237/87035(DRP); No. 50-249/87034(DRP) See Attached Distribution III.S-l

Revision 8 ReVision April 1992 ( Commonwealth Edison Company t Distribution Distri but i on cc w/enclosure: L. O. D. Butterfield, Jr., Nuclear Licensing Manager E. O. D. Eenigenburg, Plant Manager DCS/RSB (RIDS) Licensing Fee Management Branch Resident Inspector, RIll Richard Hubbard J. W. McCaffrey. McCaffrey, Chief, Publfc Public Utilities Division III. II 1. 5-2

Revision 8 April 1992 ((

\
  \
    .....                           U. S. NUCLEAR REGULATORY COMMISSION REGION III 50-237/87035(DRP); 50-249/87034(DRP)

Report Nos. 50-237/87035(DRP)i Docket Nos. 50-237; 50-249 License Nos. DPR-19; DPR-25 Licensee: Commonwealth Edison Company P. O. 80x Box 767 Chicago, 1LIL 60690 Facility Name: Dresden Nuclear Power Station, Units 2 and 3 Inspection At: Dresden Site, Morris, IL Inspection Conducted: October 23 through December 8, 1987 Inspectors: S. G. Du Pont P. D. Kaufman Approved By: M. A. Ring, Chie~ <<~7 ((~7 Section'lC , Projects Section'1C ' -- I J Date' Date '

   'C      Inspection Summary Ins ection durin the eriod of October 23 throu h December 8, 1987 ee ort os.

as. . reas nspected: outine unannounced sa ety inspection by the resident inspectors on previous inspection items; operational safety verification; monthly surveillance observation; followup of events; licensee event report followup; management meeting; report review; I.E. Information Notices; maintenance; and commissioners tour. . Results: Of the 10 areas inspected, no violations or deviations of NRC requlrements were identified in 9 areas; one violation was identified in the remaining area; however, in accordance with 10 eFRCFR 2, Appendix e, C, Section V.A., a Notice of Violation was not issued (failure to perform Technical Specification fire barrier surveillance within required time period - . Paragraph 6.) I1I.S-3 II 1. 5-3

Revision 8 Apri 1 1992 April ( DETAILS

1. Persons Contacted Commonwealth Edison Company

           *E. Eenigenburg, Station Manager J. Wujciga, Production Superintendent
           *C. Schroeder, Services Superintendent
           *l.
           *L. Gerner, Superintendent of Performance Improvement T. Ciesla, Assistant Superintendent - Planning D. Van Pelt, Assistant Superintendent - Maintenance J. Brunner.

Brunner, Assistant Superintendent - Technical Services J. Kotowski. Kotowski, Assistant Superintendent - Operations R. Christensen, Unit 1 Operating Engineer G. Smith, Unit 2 Operating Engineer

           *E. Armstrong.

Armstrong, Regulatory Assurance Supervisor W. Pietryga, Unit 3 Operating Engineer J. Achterberg, Technical Staff Supervisor' Supervisor, R. Geier. Geier, Q.C. Supervisor D. Sharper, Waste Systems Engineer D. Adam, Radiation Chemistry Supervisor J. Mayer, Station Security Administrator D. Morey,- Morey .. Chemistry Chemi stry Supervisor Supervi sor D. Saccomando, Radiation Protection Supervisor E. Netzel, Q.A. Superintendent

( *C. Turley, Station Q.A. .'

   '"       R. Stols, Q.A. Engineer
           *R. Janecek, Senior Participant - Nuclear Safety The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel~

personnel, and contract security personnel.

           *Denotes those attending the exit interview conducted on December 8, 1987,~ and informally at various times throughout the .inspection 1987                                                          ,inspection period.

2. Review of Previous Inspection Items (92702)

(Clssed) Violation (249/87010-01): Reactor ~ater temperature exceeded 212 F with fuel in the reactor and without primary containment integrity, and low power physics tests 0were not in progress. l{rhe l(rhe maximum temperature reached was 223°F, 223 F, which is a violation of Technical .1,*. ,.' Specification 3.7 .A.2 LCO. The licensee's innediate 3.7.A.2 immediate corbective cor&ective 'actions taken were to reduce reactor water temperature below 212 F and establish primary containment. These actions were completed within 22 minutes from .....-' .-, .

                                                                                                          ~

the time of discovery. Plant cool down procedure DOP 1000-3 has been cooldown revised to provide instruction to the operator regarding proper computer ".' - . point pOint selection and monitoring, primary containment requirements, and reactor water cooling requirements. -Quiet ,Quiet hours have been instituted to iC 2 III. 5-4 II1.5-4

Revision 8 April 1992 \ provide an atmosphere more conducive to turnovers. The NSO and SCRE ( involved in this event received certain disciplinary action. In addition, the licensee implemented an Error Free Operation Plan designed to achieve error free startups after refuel outages and subsequent operations.

3. Operational Safety Verification (71710 and 71707)

The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the period from October 23 to December Oecember 8, 1987. The inspectors verified the systems, reviewed tagout records and operability of selected emergency systems. verified proper return to service of affected components. Tours of Units 2 and 3 reactor buildings and turbine buildings were conducted to observe plant equipment conditions, conditions. including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspectors, by observation and direct interview, verified that the physical security' security* plan was being implemented in accordance with the station security plan. The inspectors observed plant housekeeping/cleanliness conditions and verified implementation of radiation protection controls. During the inspection, the inspectors walked down the accessible portions of the systems listed l!sted below to verify operability by comparing system lineup p1ant'drawings, as-built configuration or present valve lineup with plant'drawings, lists; observlng observing equipment conditions that could degrade performance; and verified that instrumentation was properly valved, functioning, and ca 11 bra ted. While touring the Unit 2 and Unit 3 High Pressure Coolant Injection rooms, the inspectors observed the following conditions: o HPCI exhaust line drain pot level switch cover' removed and wire broken off. The licensee had issued work request '69207 on September 25, 1987, to repair the switch, however. however, no work was inprocess during the inspectors wa1kdown. o ,Local station HPCI motor control valves valve position indicating' lights not working for the following HPCI valves: MO-2-2301-9 HPCI pump discharge valve. MO-2-2301-14 HPCI main pump recirc to torus valve. MO-2-2301-10 HPCI pump discharge to CST valve. MO-2-2301-3 HPCI turbine steam supply valve. MO-3-2301-3 HPCI turbine steam supply valve. 3 III.S-S

Revision 8 April 1992 ( o HPCI motor control valve stem covers' missing on the following

 \.       limitorque valves:

MO-2-2301-3 'S8400 was written on 10/3/86 to Work request 158400 replace misSing missing stem cover. MO-2-2301-10 Stem cover is missing and no work request has-been written. MO-3-2301-10 Stem cover is missing and no work request has been written. MO-3-2301-lS MO-3-2301-15 is missing and no work request has Stem cover 1s been written. The licensee performs a valve position indicating light walkdown on a monthly basis per DOS OOS 040-4, Revision 2. Unit 2's HPCI valve walkdown, 1987, denoted no which was completed by the licensee on November 15, 1987. lamp problems. H~wever, the residents walkdown on December Oecember 4, 1987, found the above indicating lights not working. Work requests should be initiated to replace the missing limitorque valve stem covers and repair the indicating lights. The inspectors reviewed new procedures and changes to procedures that were implemented during the inspection period. The review consisted of a verification for accuracy, correctness, and compliance with regulatory requirements. ( The inspectors also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barreling. These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under Technical Specifications, 10 CFR, and administrative procedures. The following systems were inspected: Unit 2 High Pressure Coolant Injection System Core Spray System Unit 3 High Pressure Coolant Injection System Low Pressure Coolant Injection System Conmon Standby Gas Treatment System 4 III.5-6 II!. 5-6

Revision 8 April 1992 c

4. Followup of Events (92700)

During the inspection period, the licensee experienced several events, some of which required prompt notification of the NRC pursuant to 10 CFR 50.72. The inspectors pursued the events onsite with licensee and/or other NRC officials. In each case, the inspectors verified that the notification was correct and timely, if appropriate, that the . licensee was taking prompt and appropriate actions, that activities were conducted within regulatory requirements and that corrective actions would prevent future recurrence. The specific event was as follows: On December 4, 1987, the licensee reported to NRC Region III that an (Stationman) was charged on December 4, 1987, with "possession employee (Statfonman) intent to deliver" two ounces of cocaine to an undercover with fntent metropolitan area narcotics squad agent. The licensee pulled the individual's site security badge, and performed a search of the individual's personal locker onsite with negative results. On December 7, 1987. 1987, the licensee performed a drug dog search wfthio withio the protected area with negative results. No violations or deviations were identified in this area.

5. Monthly Maintenance Observation (62703. 71710)

Station maintenance activities of safety related systems and components listed below were observed/reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specifications. The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were. performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by personnel; parts and materials used were properly certified; qualified perso~nel; radiological controls were implemented; and, fire prevention controls were implemented. Work requests were reviewed to determine status of outstanding jobs and to assure that'priority that*priority is assigned to safety related equipment maintenance which may affect system performance. following maintenance activities were observed/reviewed: The follOWing Unit Z2 2C Condensate/Booster pump - inboard seal leaking - repair/replace per request 1069912. work requ~st #069912. ZB 2B Condensate Transfer pump - mechanical seal leaking --repair/replace "repair/replace per work request 1070309.

                       #070309.

• c 5 III .5-7 III.S-7

Revision 8 April 1992 ( " Unit 3 3-1601-62 Air operated valve - replace solenoid on December 7, 1987, per w work request '70591. Unit 3 SP-87-10-156, Monthly HPCI System Pump Test for the Inservice Test Program. This special test was performed to take vibration data on the HPCI pump. Impeller replacement on the booster pump is being scheduled for the March, 1988 refueling outage. No violations or deviations were identified in this area.

6. Monthly Surveillance Observation (61726)

The inspectors observed surveillance testing required by technical specifications for the items lfsted listed below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with technical specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel. (. The inspectors witnessed portions of-the of* the following test activities: SP-87-10-156, Monthly HPCI System Pump Test for the Inservice Test SP*87-10-156, Program. This special test was performed to take vibration data on the HPCI pump_ pump. Impeller replacement on the booster pump is being scheduled for the March 1988 refueling outage. No violations or deviations were identified in this area.

7. Licensee Event Reports Followup (93702) licensee £vent Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to detenmine determine that reportability requirements were fulfilled, illl11ediate immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications Unit 2 (Closed) 87029-00: High Pressure Coolant Injection System Inoperable Due to Steam Leak.

leak. 'While While performing surveillance testing of the HPCI system I, 1987, with the unit at 97% rated power, a steam leak in the on October 1, vicinity of the HPCI turbine shaft seal was observed. After discussions between shift supervision and the Operating Engineer, HPCI was detennined determined to be inoperable at 1450 hours on October 1, I, 1987. Exact cause of the 6 III.5-8 III .5-8

Revision 8 April 1992 steam leak is unknown. A possible cause was a momentary misalignment of the HPCI turbine spring sprins loaded labyrinth shaft seals due to the starting vibrations of the HPCI turbine. The HPCI system was run on October 1. O~ I, 1987 at 1845 hours and on October 2. 2, 1987 at 1345 hours hours,t with no leakage observed; HPCl HPCI was then declared operable. (Closed) 87030-00: ~iain Steam Safety Valve Setpoints F~und Fcund Outside Technical Specification limits Due to Mishandling and Setpoint Drifts. (Serial No. BK 6290 and BK S2EO} Two Main Steam Safety Valves {Serial 62EO) relloved re!lIoved during thf: th~ 1987 Unit 2 outage were tested to determine their as-found setpoint. Valve BK 6290, with a design setpoint of 1260 psig pSig,t opened cleanly at 1276 psig, thus exceeding the plus or minus one percent tolerance required by Technical Specification 4.6.E. Valve BK 6260 designed to open at 1260 psig, failed to open twice at 1300 psiS. en a pSig, which exceeded the third attempt, the valve opened cleanly at 1282 psig. tt,e T.S. 4.6.E tolerance. Mishandling of the valve during the transport transpcrt between the drywell and the test boiler caused contact between the shaft b~tween

  . and the internal adjustment guide which increased the valve's setpc.dnt.

setp~jnt. BK 6260 and BK 6290 were overhauled and setpoints adjusted within Valves BY. the one percent to'lerance. to'lt!rance. To prevent future damage in transport, a protective guard for the stem assembly has been fabricated. The guarci guard will be required during transit per procedural change. (Closed) 87031-00: Reactor Building Ventilation Isolation Start of SBGT System Cue to Irradiated Metal on Fuel Cask. Review of this event is. is documented in paragraph 4 of Region III Inspection Report 50-237/87026(DRP). 50-237/87026{DRP}. (Closee) 6iO~~-OO; (Closed) 6i032-00: Reactor Scram Due to $purious Spurious Main Steam line low Pressure Signal Caused by Vibration. Onsite followup of this Event event was conducted and documented under Followup of Events in Region III Ir.spection Report No. 50-237/87026 Paragraph 4.r. Inspection (Closed) 870034-00: Nonconservative Core Thermal Power (CTP) Ca lcula.ticn Calculaticn Due to Inadequate Calibration Procedure. Three Rosemount 1151 dp feedwater flow transmitters on Unit 2 and one transmitter on Unit 3 were fe~dwater incc,'rectly incol'rectly calibrated. The transmitters had not been calibrated tc compensate for the effects of static pressure span compressure which resulted in a nonconservative error of 0.44% in DP calculations. It is r~sulted exceedee the CTP estimated that Units 2 and 3 exceedec ClP limit for a total of 46 hours since September 1987 and 100 hours since October 1986, error, the transmitters respectively. Upon discovery of the calculation error. were recalibrated. Dresden Instrumentation Procedure (DIP) 600-1. 600-1, Feedwater Control Calibration and Maintenance and Dresden Technical Staff procedure (DTS) 8733, Unit Z/3)Z/3} Computer Ccmputer Feedwater Flow Calibration. have been revised to include the proper calibration curve for the Rosemount 1151 dp transmitter.

c. 7 III.S-9 III.5-9

Revision 8 April 1992 ( (Closed) 86019-01: Unit 2 Reactor Scram From Main Turbine Trip on High

 \. Water Level Due to Failure of Feedwater Regulating Valve and Personnel Error. This supplemental report was issued to provide additional root cause information found after the initial LER was submitted. The tcensee discovered that the "A" reactor feedwater 1licensee                                            feeciwater discharge check valve failure and a personnel error on the part of an HSO          NSO for failure to comply with Dresden Operating Procedure (OOP) 040-4, Control Panel Light Bulb.

Replacement resulted in a reactor scram. (Closed) 87001-01: UT Indications Found on Primary System Piping Due to Intergranular Stress Corrosion Cracking. This supplemental report was issued to provide the results of additional UT examinations performed on piping welds as a result of the indications reported 1n in the original LER. No additional indications were found upon completion of all UT testing. Unit 3 (Closed) 87016-00: Primary Containment Group I Isolation and Reactor Scram Due to Apparent Personnel Error. Review of this event is documented in Region III Inspection Report 50-249/87025. 50-249/87025, Paragraph 4. The preceding LERs have been reviewed against the criteria of 10 CFR 2, Appendix C, and the incidents described meet all of the following requirements. Thus no Notice of Violation is being issued for these items.

                 .-,-  ....~

a. The event was identified by the licensee,

'c* b. The event was an incident that, according to the current enforcement policy, met the criteria for Severity levels IV or V V violations,

c. The event was appropriately reported,

d. The event was or will be corrected (including measures to prevent recurrence within a reasonable amount of time), and

e. the event was not a violation that could have been prevented by the licensee's corrective actions for a previous violation.

No violations or" deviations were identified in this area. or'deviations (Clused) 50-249/87018-00: (fire (Clased) (Fire Stop 18 Month Surveillance Interval Exceeded Due to Procedural Deficiency.; On September 20, 1981. 1987, a review of upcoming surveillances was being bein9 performed when it was found that Dresden Fire Protection Procedure (DFPP) lDFPP) 4175-3. 4175-3, *Shutdown Fire Stop/Break Surveillance-, Surveillance", was incorrectly classified as due each refueling outage in the surveillance program. The critical surveillance date for this surveillance was November 1, I, 1985. This surveillance was completed on April 24, 1986, 5 months and 23 days after the critical date. The critical date was missed due to improper categorizing of the 18 month surveillance as a refuelin9 outage surveillance on the computer. Dresden Technical Specification (TS) 4.12.F.l requires that these fire* (stop/break) penetrations be inspected once every 18 months. barrier (stop/brea.k) Failure to perform this Technical Specification surveillance within the required time period is a violation of TS 4.12.F.l 4.12.F.1 (237/87035-01; 8 ..

                                                                                                   ~

1 III.5-10 III.S-IO

Revision 8 April 1992 249/87034-01)., However. However, since Unit 3 was in co1d cold shutdown throughout this period due to an extended refueling outage, the safety significance' of exceeding the surveillance date was minimal. Consequently, this violation meets the tests of 10 CFR 2, Appendix C. C, and no Notice of Violation will be issueG. issued. This item is considered closed. ~' One viclation vic1ation was identified in this area.

8. I.E. Information Notice Followup Fo1lowup (92701)

Each of th~ the following I.E. Infonmation Information Notices (IEK) was reviewed by the Resident Inspectors to verify (1) that the Information Notice was received by licensee management, (2) that a review for applicability was perforrr~d. perforrr~d, and (3) that if the Information Notice was applicable to the facility, applicable ~ctions were taken or were scheduled to be taken. applicabl~ actions (Clc.sed) lEN (elGsed) rEt: 87-23: Loss of Decay Heat Removal During low Low Reactor Coolant Level Operation. Dresden Units 2 &3 do not have RHR systems to remove decay heat,. remov~ heat,* so the licensee reviewed the Shutdown Cooling system with regard to the problem identified in this Infonmatfon Information Notice. The Shutdown CQoling Cooling system takes suction from the Recirculation system. The Recirculation system takes its suction from the reactor vessel annulus area and this piping is not isolatable. The Shutdown Cooling water is returned to the discharge side sioe of the recirculation pump via the LPCI line. Thus, the reactor water level woulcwould have to decrease to a point of uncovering the irradiated fuel before a loss of suction could ccc~r.occur. Dr~sd~n has five independent categories of reactor vessel water level Dresden C,' instrumentation to monitor reactor water level. (Closed) lEN 87-42: Dies~l Diesel Generator Fuse Contacts. The licensee had a similar, but lESS lESS serious. serious, incident occur during a test of the Unit 2 gEnerator in 1974 following a maintenance outage. Dresden diesel snerator utilizes a similar contact arrangement for the diesel generator genErator potential transformer (PT} (PrJ fuses and connecting cables. The 1licensee i censee issued DVR 12-2-74-16 and the ensuing inspection revealed burn marks on the B phase contacts and cabJes (cables charred), indicative of arcing. Poor mating of the fuse and stationary contacts were identified 2! a~ t~e cause of the arcing. In M~.y 197~, the 1 M~y !97~, icenses made modi licenses fications (M12-2-74-32; modifications (MI2-2-74-32; M12-2-74-33; M12-3-74-48) to the electrical contrel cabinet doors which enabled them to be screwed closed, thus, insuring a proper mating of the moveable contact finger connects with the stationery stationary contact. No violations viol~tions or deviations were identified in this area.

9. Commissioner's Visit On October 23, 1987, 1987~ Commissioner Kenneth C. Rogers, accompanied by the NRC Region'III Deputy Regional Administrator, visited the Dresden Nuclear Station. While on sitE,

                            ~ite, the Commissioner and Region III Management toured the facility with the licensee's corporate and plant management on a familiarization a~ce~c plant improvement tour. The Commissioner also held meetings with licens~e    plant management and supervisory personnel. The licensee p~ant Commissioner complimented Commonwealth Edison on Dresden's positive progress and direction.

tc. 9 III.S-ll IlLS-II

Revision 8 April 1992 /--- / \\. 10. Management Meetings The President, Executive Vice President and other members of Commonwealth Edison Company met with Mr. Victor Stello, Stello~ Jr., Mr. James M. Taylor, the Region Re9ion III Regional Administrator, and other NRC representatives in Headquarters on October 28, 1987. 1987, to discuss the company's plans to sustafned performance improvement effect sustained fmprovement at the Dresden Nuclear Power " Station.

11. Report Review During Durfng the inspection fnspectfon period, the inspectors fnspectors reviewed revfewed the licensee's lfcensee's Monthly Operating Operatfng Report for October 1987. The inspectors fnspectors confirmed conffrmed that the information fnformation provided met the requirements requfrements of Technical Speciffcatfon 6.6.A.3 and Regulatory Guide 1.16.

Specification The 1license~ icense~ announced announce"d -the "the following Dresden sitesfte management changes effective ~vember 25, 1987: "-

                             ~QA) Superintendent, M. Jeisy, will assume the position Qualfty Assurance 1QA)

Quality of INPO Coordinator. The new Q.A. Superintendent 1s fs E. Netzel, ~' transferring from the Braidwood Station. Statfon.

12. Exit Interview (30703)

The inspectors fnspectors met with wfth licensee representatives representatfves (denoted in fn Paragraph 1) informally fnformally throughout the inspection inspectfon period perfod and at the conclusion conclusfon of the inspection fnspection on December 8, 1987, and summarized the scope and findings of the inspection activities. The inspector also discussed the likely informational informatfonal content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. fnspection. The licensee did dfd not identify fdentify any such documents/processes as proprietary. The licenseelfcensee acknowledged the findings findfngs of the inspection. 10 III.S-12 111.5-12

Tab 6 Revision 8 UNITED STATES April 1992 NUCLEAR REGULATORY COMMISSION REGION III111 ( GLEN 7" ROOSEVELT ROOSEV[L T ROAO GL.EN ELL. ROAD YN. ILLINOIS 50137 ELL.YN, '0137 Docket No. 50-237 DEC 14 1987 rEC 21mi' [EC21~ Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Post Office Box 767 Chicago, Il 60690 Gentlemen: This refers to the routine safety inspection conducted by Mr. T. J. PlosKi Ploski and others of this office on November 16-19, 1987, of activities at the Dresden Nuclear Generating Station, Units 2 and 3, authorized by NRC Operatin9 Licenses licenses No. DPR-19 and No. DPR-25, and. to the discussion of our findlngs with Mr. E. Eenigenberg and others of your staff at the conclusion of the i on. inspect ion. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. ~- No yjolatjons of NRC requirements were identified during the course of this inspection. In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC Public Pub 1i c Document Room.

 /

We will gladly discuss any questions you have concerning this inspection. Si ncere ly, Sincerely, lJj.e. .s;. W*G s.. ifJ I~ LP 14"- W. D. Shafer, Chief Emergency Preparedness and' and* Radiological Protection Branch

Enclosure:

Inspection Reports No. 50-237/87037(DRSS); No. 50-249/87036(DRSS) See Attached Distribution III.6-1

Revision 88 April 1992 ( Commonwealth Edison Company 2 DEC 14 1987 Distribution cc w/enclosure: D. Butterfield, Nuclear Licensing Manager J. Eenigenburg, Plant Manager DCD/DeB (RIDS) OCD/DCB Licensing Fee Management Branch Resident Inspector, RIll Richard Hubbard J. W. McCaffreY McCaffrey!t Chief, Public Utilities Divlsion EPB, NRR D. Matthews, EPS, W. Weaver, FEMA, RV c:.c c-III.6-2 II I. 6-2

Revision 8 April 1992 c U. S. NUCLEAR REGULATORY COMMISSION REGION III Reports No. 50-237/87037(DRSS); 50-249/87036(DRSS) Docket Nos. 50-237; 50-249 Licenses No. DPR-19; No. DPR-25 licenses Licensee: licensee: Commonwealth Edison Company P. O. Box 767 Chicago, IL 60690 Facil i ty Name: Dresden Nuclear Generating Station, Units 2 and 3 Facility Inspection At: Dresden Site, Morris, Illinois Production Training Center, Braidwood, Illinois Inspected Conducted: November 16-19, '1987

                      /,"p, J?d,.A. '

7-p,!2tA.' Inspectors: T. vJ. Ploski. 1~~7 1Z/~7 Date

                     /!i         (!'*t{~j?;*/f*:/-

II.. (jlt{jj:;*Ij.>-

                     /j 11.,

G. M. Chri stoffer (i g.?: 9*~~-~ ;r:_~

                 ~.      E. Foste~

L;11. J.yJn~'-

                   '111,  J. 0.nu-d~

M. j. Smlth Approved By: ~~ fffef effef Emergency Preparedness Section Inspection Summary Ins ection on November 16-19 1987

o. .

reas nspec e: ou lne, unannounced inspection of the following areas of the Dresden Station's emergency preparedness program: licensee actions on previously-identified items; emergency plan activations; operational status of the program; emergency detection and classification; protective action IeIe decision-making; notification and communications provisionsj changes to the program; shift staffing and augmentation; training; and audlts. The inspection involved four NRC inspectors. Results: No violations of NRC requirements were identified during this 1 nsped 1 on. lnspectlon. III.6-3 II 1.6-3

Revision 8 April 1992 c* ( DETAILS

1. Persons Contacted

           *E.
           "'E. Eenigenberg, Station Mana~er ManaQer
           *C.
           "'C. Schroeder, Services Superlntendent
           *R.
           "'R. Holman, GSEP Coordinator
           *R.
           "'R. Jeisy, Quality Assurance Superintendent
           *T.
           "'T. Gallaher, Quality Assurance Engineer
           *S.
           "'S. Stiles, Training Supervisor
           *E.
           "'E. Armstrong, Regulatory Assurance Supervisor
           *T.
           "'T. Gilman, Emergency Planning Supervisor R. Mitzel, Shift Engineer (5E)

(SE) R. Sitts, SE T. Palanyk, Station Control Engineer (SCRE) R. Speroff, SeRE seRE J. Bowman, Corporate Emergency Planning Staff

l. DeCarlo, Drill Controller D. Marco, GSEP Training Instructor W. Reimers, Training Department Staff K. licari, Licari, Production Training Center

           *Indicates
           "'Indicates those persons who attended the November 19, 1987 exit meeting.

('-- -

2. licensee Licensee Action on Previously-Identified Items (Closed) Item Nos. 237/85013-01 and 249/85012-01: Revise Emergency

         . Action level Level (EAL) Condition No. 12 for General Emergency to indicate that this emergency class can also be declared based on environs measurements. Emergency Plan Implementing Procedure (EPIP) 200-T1~

200-T1. Classification of GSEP Conditions, has been revised so that a General Emergency canl be classified based on a source term derived from field survey teams teams' measurements. This item is closed. (Open) Item No. 249/86002-02: Add the phrase "plant staff"U to Uplant operating staff the Unusual Event EAL for Condition No. 18. This item resulted from the investigation of the January 20, 1986 drywell fire incident during a plant outage, which was not a classifiable emergency per the licensee's EAls. EALs. A review of the EALs listed in EPIP 200-T1, Revision 6, indicated that no EAL had been revised to satisfy the concern expressed .in in this Open Item. The Dresden Station's EAls st~tes of EALs were in the latter states revision at the time of this inspection. These proposed EAls EALs adequately addressed* the concern; however, their submittal for NRC review was 1988~ tentatively scheduled for sometime during the first quarter of 1988, This item remains open. (Closed) Item Nos. 237/87028-01 and 249/87027-01: Due to a backlog of filing controlled documents at the Mazon Emergency Operations Facility (EOF), a Severity Level IV violation was issued for not maintaining the facility in an adequate state of readiness. A tour of the Mazon EOF during this inspection revealed that all controlled documents were filed, IU.6-4 III.6-4 2

Revision 8 April 1992 ( and adequate permanent administrative support was available to ensure the maintenance of the operational readiness of the EOF. A controlled document room had been completed the week before this inspection and all controlled documents were filed in this new area. The 1988 audit schedule, issued November 18, 1987, i,ncluded included an item to verify document control at the EOF. This audit line item was part of the licensee's commitment in response to the Notice of Violation. This item is closed.

3. Emergency Plan Activations NRC and licensee records associated with all emergency plan activations that occurred between December 6, 1986 and October 4, 1987 were reviewed.

These records included: licensee Event Reports (LERs); (lERs); records generated by NRC Duty Officers; Control Room logs; Nuclear Accident Reportlng System (NARS) forms completed by onshift personnel following each emergency declaration; the licensee's Deviation Reports; and evaluations of each emergency plan activation that were performed by the GSEP Coordinator. During this time period, onshift personnel correctly classified thirteen Unusual Events. Based on the LERlER review, there were no other classifiable events through October 4, 1987. Initial notifications to State and NRC officials were completed within the regulatory time limits following each emergency declaration. ie Based on ,the

               .the above findings was acceptable.

findings, this portion of the licensee's program t

4. Operational Status of the Emergency Preparedness Program (82701)

a. Emergency Plan and Implementing Procedures (Also 82204)

The licensee's procedures for the preparation, review and . distribution of new and revised EPIPs were basically unchanged from the previous inspection and were adequate. A review of six EPIP revisions indicated that proper procedures were followed to incorporate these revisions into the program. Review of the EPIP Distribution Transmittal Log log showed that revisions were distributed to licensee personnel and the NRC within one week after approval. However, EplPs EPIPs sent to NRC Region III were not being tracked to ensure receipt of the revisions. When this was mentioned to the licensee, Region III was added to that portion of the EPIP Distribution Transmittal log that tracks receipt of documents. Current copies of the emergency plan and implementing procedures were readily available in the Control Room, TSC and EOF. Based on the above findings this portion of. the licensee's program is acceptable. III.6-5 II I. 6-5 3

Revision 8 April 1992 ( b. Readiness of Emergency Response Facilities and Supplies (Also 82204) A tour of Emergency Response Facilities (ERFs) indicated that the Technical Support Center (TSC), Operational Support Center (OSC), and Emergency Operations Facility (EOF) were maintained in an adequate state of operational readiness. Plans, procedures and drawings were filed; communications equipment was operational; and adequate supplies were available. A controlled document area had been constructed in the EOF to contain plans, procedures and drawings relevant to all of the licensee's nuclear stations. Upgraded equipment was scheduled for installation in the EOF to bring its layout and computer capabilities equal to that of the Zion Station's EOF, which was utilized in the 1987 Federal Field Exercise. Included in this upgrade will be electronic status boards, a PRIME computer system for administrative functions, and an upgraded plant data computer system. A dual purpose transportation facility had been constructed next to the EOF. Thi Thiss faei 1i ty wi facility 11 house a dedi will cated uGSEP dedicated "GSEP Van" for ofts i te survey offsite team use. The structure was also constructed to accommodate overflow EOF and Joint Public Information Center (JPIC) personnel. A limited number of telephone outlets will be available for temporary use by overflow personnel. The re-modeled JPIC was toured and appeared to be about 90 percent complete at the time of this inspection. A review of 1987 records for emergency equipment and supplies ( inventories was performed. All required inventories had been completed and adequately documented as required by EPIPs. Inventory records indicated that identified deficiencies had been promptly corrected. An "Emergency IIEmergency Response Telephone Directori Directory"' had been developed for use in the ERFs. This directory contained instructions on use of different communication systems and the business telephone numbers for offsite licensee personnel and offsite support agencies. This computerized directory was scheduled for quarterly review and update by Corporate emergency planning personnel. An inventory of the supplies in the decontamination and medical facility was requested per relevant EPIPs. The following discrepancies were found:

        **     EPIP 500-4 stated that there is one portable eye wash deyice device located in the medical and decontamination area. One portable eye wash was in that area; however, it was questionable whether its operability had been checked during the inventory process.

When this concern was brought to the attention of the Radiation Chemistry Foreman. Foreman, he immediately contacted the Operations Department to find out if they had conducted surveillances on this piece of equipment. The Operations Department reported that no monthly operability check had been conducted. They agreed to add this particular portable eye wash to the monthly 'c i(. . surveillance schedule. III,6-6 III.6-6 4

Revision 88 April 1992 c ( Several minor differences were noted between the actual contents of the No. 36 first aid kits as described in EPIP 500-1, IIInventory "Inventory Sheet for First Aid Kits. Kits."1I When told of the discrepancies, the GSEP Coordinator stated that he was in the process of upgrading the No. 36 first aid kits to the Corporate kit standards as stated in CECo General Procedure 826. Additionally, it was observed that the items listed in the Medical and Decontamination Area Inventory were stored in a disorganized manner in various locations in the room. Based on the above findings, this portion of the licensee's program 8ased was acceptable; however, the following item should be considered for improvement:

• The licensee should arrange supplies kept in the decontamination and medical facility in an organized manner.

c. Organization and Management Control (Also 82204)

The 1licensee's icensee ' s IIStrategic "Strategic Plan for Excellence in Nuclear .- Operations, 1988-1992" included the objective of lIaintaining an effective emergency preparedness program in terms of plans, procedures, personnel, and facilities. With respect to personnel at the Dresden Station, this corporate plan has been translated into a "Basic Expectations for Management Personnel. Personnel."II Supervisors would be held responsible for ensurin9 ensurinQ that their staffs complete all their . emergency preparedness trainlng as scheduled, and that routine activities, such as drills, surveillances, and inventories are adequately done .. The GSEP Coordinator was a Dresden Station employee. The GSEP Coordinato.r' Coordinator'ss reporting chain has changed since the last inspection. He formerly reported to the Services Superintendent throuQh the Regulatory Assurance Supervisor. He now reports to that indivldual throuQh the Rad Chem Supervisor. This change was made lIade in order to be conslstent with the licensees licensees other nuclear stations and to facilitate the coordinator's interface with Rad Chem Department staff, which includes the former GSEP Coordinator. During this inspection the corporate emergency planninQ staff received approval to *expand

                             -expand its scope of responsibilitles with the addition of a position titled GSEP Staff Coordinator, who will report to a corporate Emergency Planning Supervisor based at the Mazon EOF.

The GSEP Staff Coordinator's responsibilities will include the coordination of Corporate and Stations' GSEP programs includinQ including training. The coordination of Station and corporate GSEP trainlng efforts and interface with the nearby Production Training Center had been a responsibility of emergency planners based in the licensee's corporate offices in downtown Chicago. II 1.6-7 111.6-7 5

Revision 8B April 1992 ( Letters of Agreement with offsite aQencies a~encies were current. Annual radiological emergency response tralning for these agencies was conducted, as required by 10 CFR 50.47(b), on September 3, 1987. The agenda included a review of EALs, Emergency Classifications, and information on requesting QA Department assessments of the Station1s Station's interface with State and local response agencies. Agenda materials had been revised and upgraded since 1986. A "Nuclear Power Handbook" pamphlet, containing plant systems diagrams and fundamentals of radiation information, plus a booklet on EALs and ERFs were distributed to meeting attendees. Visual aids were also upgraded~ and inc 1uded a"a' fi included 1m of the plant site and photos of the TSC and tOF. film A tour of plant facilities was also offered. This trainin9 traininQ program improvement was the result of a coordinated effort by Statl0n Statlon and corporate personnel. Based on the above findings, this portion of the licensee1slicensee's program was acceptable.

d. Emergency Preparedness Training (Also 82206)

The annual training of onsite emergency response personnel had been completed and adequatel~ adequatel¥ documented, with the exception of two individuals, whose trainlng was scheduled before December 31. The 1987 training had been accomplished utilizing a combination of EPIPs and training modules relevant to specific positions in the onsite emergency organization. Examinations on training materials wefe were reviewed and found to be adequate. By memo dated February 3, 1987, the Dresden Station's Training Department was provided with a set of training modules that had been refined from an earlier version by staff at the licensee'slicensee1s Braidwood Station. Site-specific adjustments to these modules had then been made for the Dresden Station. Section 8.2 of the generic GSEP described.an described,an lIapproved "approved GSEP Training Matrix" which delineates the training applicable to specific emergenc~emergenc¥ organization positions. (The matrix of 1987 onsite training requlrements had not been formally approved. This was a finding of an October 1987 Qualit~ Qualit¥ Assurance (QA) Department Audit No. P-87-IV.) The station's station1s tralning matrix was formally approved by appropriate Dresden Station management on November 19, 1987. The offsite GSEP Training Matrix was approved on November 20, 1987. Nineteen of thirt~-five thirt¥-five training lIodulesmodules were reviewed for inconsistencies wlth the GSEP and relevant procedures. "A 'A discrepancy regarding the Station Director'sDirector1s undelegatable responsibilities is described "in in Paragraph 6 of this report. The only other problem identified was that another .adule indicated that a field survey team was to retreat if it encountered a radiation field of at least 100 mR/hour. This guidance was inconsistent with that found in Procedure EG-3, Revision 6, which indicated that a team must immediately inrnediately inform an Environs Director when encountering a radiation field of at I II. 6-8 III.6-8 6

Revision 88 April 1992 ( least 100 mR/hour. When the licensee was informed of these training module errors, both were adequately corrected prior to the exit interview. Interviews were conducted with five members of the onsite emergency organization. Personnel were adequately knowledgeable of their emergency responsibilities. Additional details regarding the walkthroughs of Control Room personnel are provided in Sections 5 and 6 of this report. . Records review indicated that all required drills had been conducted, critiqued, and adequately documented for the period October 1986 through September 1987. The licensee's evaluation of the September 1987 exercise had also been adequately documented. The final critique report for the November 1987 Medical Drill was not yet available for review. On November 17th, an inspector observed a semi-annual Inplant Health Physics drill and subsequent critique. A Rad Chern Foreman and two technicians participated in the drill; which was evaluated by two controllers. The response of the ;np inp ant team was realistic, with minimal simUlation simulation of protective clothing, special dosimetry, radiation survey devices, and communications equipment. An air sample and a number of smear samples were collected. An adequate critique was conducted after the drill. Player feedback was encouraged.

( Based on the above findings, this portion of the licensee's program was acceptable.

e. Independent Reviews/Aduits (Also 82210)

Records of the Quality quality Assurance (QA) Department audits and surveillances Slnce August 1986 were reviewed. All records were readily available and complete. Two audits and seven surveillances were conducted in 1987. Surveillance topics included: drill and exercise evaluations; evaluationsj document control at the Mazon EOF; EOFj and responses to two actual emergency emerQency plan activations. Audits and surveillances were adequate 1n ln scope and depth. The regulatory requirements of 10 CFR 50.54(t) SO.S4(t) were adequately addressed. The adequacy of interface between the Station and various governmental agencies was'also was* also assessed as adequate per Audit AA-87-23. The QA Department adequately ade9uately tracked corrective action taken on audit and surveillance flndings f1ndings and recommendations. A report of corrective action taken or planned is required within 30 days. The QA Department then conducts a followup audit in 90 days to evaluate the effectiveness of the action taken. th~ A review of the GSEP Coordinator's informal tracking system for corrective actions to be taken on identified drill and exercise improvement items was conducted. The tracking system was current up to the September 1987 exercise. The corrective actions taken on

C ....

l earlier drill items were adequately documented.

111.6-9 III.6-9 7

Revision 8B April 1992 ( The ~SEP GSEP Coordinator conducted thorough reviews of all internal documentation associated with all emerQency plan activations since the last inspection. The informal reVlew procedure has been upgraded, as the Coordinator also determined whether a declaration was appropriate, rather than only focusing on the ti.eliness and completeness of the various notifications. The review procedure included provisions for informing Station management and a corporate emergency planning supervisor of any identified problems. Based on the above findings, this position of the licensee's program was acceptable. acceptab 1e.

5. Emergency Detection and Classification (88201)

EALs contained in EPIP 200-T1 200-Tl were consistent with those listed in the current revision of the Dresden Annex to the GSEP. Personnel from the Dresden and Quad Cities Stations and corporate emergency planning staff have been meeting to substantially revise and standardize both Stations' EALs. Although a recent draft of the proposed EALs was available, the licensee did not expect the revised Dresden Station EALs to be submitted for NRC review and approval until the first quarter of 1988. Two walkthroughs were conducted with Control Room personnel. Each walkthrough involved a Shift Engineer (SE) and a Station Control Room '( Engineer (SeRE). (SCRE). Both SEs clearly understood that they had the undelegatable responsibility to declare an emergency. Both sets of personnel adequately demonstrated the capability to properly classify

     .abnormal situations in accordance with the Station1s Station's EALs. The individuals were adequately familiar with regulatory requirements and procedural guidance for informing State and NRC officials following any emergency declaration.

Based on the above findings, this portion of the licensee's program was acceptab 1e. acceptable. .

6. Protective Action Decisionmaking (82202)

Procedural qui dance regarding onsite and offsite protective action Quidance decisionmaklng was consistent with that found in the current GSEP and Dresden Annex. The locations of onsite assembly areas identified in EPIP 300-3 (Assembly and Evacuation of Personnel) were identical with those shown on an emergency information card made available ~o to personnel granted unescorted access *privileges. During observation of an inplant Health Physics drill, the inspector noted that the Unit 2 trackway area was adequately marked as an assembly area. Signs giving directions ~o to this assembly area were readily visible on the 570-foot elevation of the Unit 2 elevation.of portion of the Reactor Building, on building elevations between that level - and the ground level assembly area, and on stairways leading to the Unit 2 trackway. During the walkthroughs with both sets of SEs and SCREs, it became apparent that there was some uncertainty uncerta i nty regarding regardi ng whether the decision deci s i on to recommend offsite protective actions and/or the authorization of rrI.6-10 111.6-10 8

Revision 8 April 1992 ( emergency worker exposures were undelegatable responsibilities of the Acting Station Director, as stated i.n the GSEP. The interviewees were assured that both items were undelegatable responsibilities per the GSEP. A review of relevant procedures and lesson plans uncovered several inconsistencies with the GSEP regarding undelegatable responsibilities. EPIP 100-C1 IOO-Cl (Station Director) and EPIP 300-4 (Emergency Personnel Dose Limits) did not clearly indicate that authorization of exposures in excess of 10 CFR Part 20 limits was an undelegatable responsibility. The relevant lesson plan indicated that the declaration of an emergency was the only undelegatable responsibility of the ActinQ Acting Station Director. These procedural and training program inconsistencles were brought to Station management's attention during the inspection. Both procedures and the relevant lesson plan were revised, and approved per Station procedures, prior to the November 19th exit interview. Onshift personnel would be informed of the procedure changes through the required reading program. Based on the above findings, this portion of the licensee's program was acceptable.

7. Notifications and Communications (82203)

A review of test documentation for the period April-October 1987 indicated that the licensee has adequately maintained the Prompt Notification (siren) System utilized by offsite officials to alert the public within ( the 10-mile Emergency Plannin9 Planning Zone (EPZ) of a serious emergency situation at the Dresden Statlon. ~- The Station's annual emergency communications test was conducted on February 3, 1987. Documentation was adequate, including indications of prompt corrective actions taken on a few minor problems. Records of periodic communications tests for the period January-November 1987! 1987 1 plus a random testinQ testing of equipment during the inspection, indicated tha that portable communlcations equipment and fixed equipment in the emergency response facilities had been adequately maintained. EPIP 500-7 (Nuclear Accident Reporting System (NARS) Test Checklist) was used to document the monthly test of this dedicated system for notifying State and local officials. Completed copies of the checklist also contained handwritten, commercial telephone numbers for the various agencies. However, incorrect commercial numbers had been wri~ten written on the checklists for the Illinois Department of Nuclear Safety, Will County ESDA, and the Kendall County Emergency Operations Center. Although the procedure required the licensee's caller to test the NARS and to write backup telephone number information on the checklist, it did not require the caller to verify the backup telephone numbers during the NARS t~st. test. Consequently, over a period of time, several incorrect backup telephone numbers had appeared on the completed checklists. Correct backup telephone numbers for the locations in question were contained in appropriate EP1Ps EPIPs and the GSEP Telephone Directory, both of which would be available to emergency response personnel. III.6-11 111.6-11 9

Revision 8 April 1992 ( \'" A tour of the Unit 2/3 Control Room indicated that a copy of the NRC Duty Officer's Event Notification Worksheet had been placed in the "Red IIRed Phone Logbook" for use as a reference when onshift personnel would cOlIDunicate cOlIIDunicate with the Duty Officer. The licensee indicated that the intent was to improve the quality of communications with the NRC, as onshift personnel could better anticipate the agency's information needs. Based on the above findings, this portion of the licensee's program was acceptable; however, the following item should be considered for improvement:

• If completing EPIP 500-7 is not intended as a means of verifying backup telephone numbers, then the licensee should delete the procedural requirement to write-in such data on this checklist.

8. Shift Staffing and Augmentation (82205)

The numbers and types of persons required for augmentation of onshift personnel following declaration of a given emergency class were specified in Section 4 of the GSEP, EPIP 300-1, and in a prioritized callout list. Augmentation provisions met the criteria in Table B-1 of NUREG-0654, Revision 1. Augmentation of onshift personnel is initiated through an Operations Duty Supervisor. The callout lists have been updated on a c quarterly basis .

      . The licensee conducted quarterly drills during 1987 which successfully demonstrated the capability to adequately augment onshift personnel in a timely manner. The quarterly off-hours drills were done in accordance with a commitment in the Dresden Annex to the GSEP. The generic GSEP contained a semi-annual commitment to conduct such drills.

Based on the above findings, this portion of the licensee's program was acceptable. acceptab 1e. -,.-

9. Exit Interview On November 19, 1987 the inspectors met with those licensee representatives denoted in Paragraph 1 to present their preliminary inspection findings. The licensee indicated that none of the matters discussed were proprietary in nature. .

III.6-12 10

Tab 7 Revision 8 April 1992 DRESDEN 2 &3 ( FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspection Report No. 50-237/88QI0 50-237/88010 and 50-249/88012

  ~

III.7-1 II!.7-I Inspection Reports No. 50-237/88010 and 50-249/88012 dated January 3, 1989. II 1. 7-33 Appendix R Audit Questions April 18-22, 1988.

1. 7 -83 IIL7-83 II February 1, 1989 CECo letter from H. E. Bliss to A. Bert Davis transmitting the response to Inspection Report No. 50-237/88010 and 50-249/88012.

'( IIII.7-i II. 7-; {-

                                                                                             '~y/     ).

UNITED STATES N. A:.':U<. ~ .' NUCLEAR REGULATORY COMMISSION 1" IIOOSE RECION REGION III ROOSEVELT ROAD IIEI.1' "lOAD Revision 8 GLEN EL.LVN. ELLYN. IL.L.INOIS ILLINOIS 60lH 60131 April 1992

                                                       . JAN 3 1989    J989 Docket No. 50-237 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:

Gent lemen: This refers to the special safety inspection jnspection conducted by Messrs. J. Holmes. Holmes, R. Hodar, Hodor, and K. Parkinson of this office on April 18-22, May Hay 11-13; 11-13, August 15.15, 13, 1988. and December 13. 1988, of activities at Dresden Nuclear Power power Station IInits tz.. Stat jon Units and 33. authorized by NRC Operating Licenses No. ]£R-19 ]PR-19 and No. DPR~2S. DPR-25, and to the discussion of our findings with Mr. Hr. E. D. Eenigenburg at the conclusion of the inspection. inspection, This inspection was conducted to assess compliance wjth with 10 eFR implementatlon of certain Fire *PrQtectjPD CFR 50, Aependix R, and to review implementat10n *Protectipn Program regul*rements. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective !i\ examination of procedures and representative records, observations, and interviews with personnel. '.

                                                                                                                  'fwo During this inspection, ~~..;.:.:-~~*-.::.:::.:::.:.;..~~~:.:::.=.:~~~~~~~..:::..:.:..,

certain of our activities a eared to be in violatiun Levc..1 Levc.I of NRC reguirements, as descrlbed in the enclosed Notlce. Wlth respect to 'J!'

                                                                                                                   ~'$  's Item 1 of the Notice, the ins ection showed that actions had bee~taken to correct the 1 dent 1 f i ed V10 at lon and                  v                . ur understand i ng No o your correc lve actlons are described in Paragraph 3.g of the enclosed                                   yes~

inspection report. Consequently. no repl¥ to this violation is re~uired and we have no further guestions regardlng thlS matter at thlS time.e ardln th~10dA.) re~alnln Hem W~l h co~ rn the m ibl 1 "Ve>,,",, wrltten response lS regulred. . , w'th 10 CFR 2.790 of the Commission's regul~tions, a copy of In accordance w:th this letter, the enclosures, and your response to thi thiss letter will wi 11 be placed in the NRC Public Document Room. The responses directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511,' 96-511.' i(,,-_. III.7-1

Revision 8 April 1992 ( Commonwealth Edison Company 2 JAi... ~ JAi 1989 We will gladly discuss any questions you have concerning this inspection. Sincerely. Sincerely, . J1;lt'

                                              #It,)~;.v  (ll.v /(                                                                  /f'o-'

Hubert J. Director J, Miller, Director' Division of Reactor Safety

Enclosures:

1, Notice of Violation 1. 2, Inspection

2. Inspect ion Reports No. 50-237/88010(DRS);

SO-249/88012(DRS) No. 50-249/88012(DRS) cc w/enclosures: H, 81iss, H. Bliss, Nuclear licensing Manager Licensing J, Eenigenburg, Plant Manager J. oCO/oCB DCDIDCB (RIDS) Licensing Fee Management Branch 8ranch Resident Inspectur, RIll ( Richard Hubbard J, W. McCaffrey, Chief, Public J. Utilities Division

(

III. 7-2

Revision 8 April 1992 ( NOTICE OF VIOLATION Co"",onw~a 1th Edison COITJr.onw~alth Edi son Company Docket No. No, 50-237 Dr'esden Nuclear Station Dr~sden Docket No. 50-249 As dd r~sult result of the inspection insp~ction conducted during April 18-22. 18-22, May 11-i3,ll-i3, IS, and D~cember August 15. December 13, 1988, and in accordance with the "General Policy dnd PrOCt:dur~ Proc~dure fur NRC Enforc.:m~nt Actions," 10 CFR Part 2, Appendix e Enforcement Actions." C (1988). (1988), the th~ following violations were identified:

1. III.L of Appendi~

Section III.l Appendix R to 10 eFR CFR Part 50 requires that alternate shutdown capabil i ty provided for a specific fire ar~a shall be capable capability of m~intain;ng the reactor coolant level* above the top of the core. In r,',aintaining th~ addition. supportin'g functions shall be capable of providing the proc~ss addition, supporting process cou;ing lubrication, etc., necessary to permit the operation of the cOuiing th~ equipme:nt equiprn~nt used for saft!safe shutdown functions. Further, Section IILL requires that procedures shall be in effe:ct requir~s eff~ct to implement implem~nt this capability. Th~ licensee was r~quired The required to establish th~se these procedures by July 19. 19~5. Contrary to the above, during the inspection conducted on Apri 1 18-22, 19EL, an NRC ins ect r identified that no admini rati dures or controls were in eff~ct to insure that regy;r~d alternative shut own e~ui~ment (i .e., control rod drive pump, service water pump, 4Kv Bus, 4 0 US anc 480V MCC) WdS available for the operating unit when ( the 0 osite unit (which houses the alternative shutdown e ui in an outa e or S ut own an e re Ulre a erna e s ut own e ui ",ent nt was was remuved from serviCe for schedu ed maintenance or repalr. This is a Severity Sev~rity Level IV violation (Supplement 1).I) . ._-'

2. CFR 50.40(a) requires that each operating nuclear pow~r 10 eFR power plant hav~ have ad fire protection plan that satisfies Criterion 3 of Appendix A to .10 ,10 CFR Part 50. It further requires that the plan shall describe specific features necessary to implement the program such as administrative adm~nistrative controlS controls 'and the means to limit fire damage to structures, systems, or components important to safety so that the capability to saf~ly safely shutdown the plant is ensured.

Section B.2 of the lic~n5ee's licensee's response to the Guidelines of A~pendix A to APCSB 9.5-1 as acc~pted APCS8 accepted in the 1980 Supplemental Safety Evaludtion Report indicates that effective administrative measures will be impl~nent~d indiCates impl~nented totu prohibit bulk storage of combustible materials inside or adjacent to safety-related buildings or systems during operdtion saf~ty-related operation or:maintenance or,maintenance p~riods. periods. .

                ~~~l~'\1 Cortrary to the above, during a previous inspection conducted 9n April l~,

19S8, an NRC ins ector observed twent 55-gallon drums of lubricatin uil II stored in 0 safet -re eva 1n e sout west corner of the Unit 2 Reactor Building. condition existed ~ t~arch 31 to April 13, 1988. III.7-3

Revision 8 April 1992 c- Notice of Violation 2 This is a Severity Level IV violation (Supplement I). With respect to Item 1, the inspection showed that action had been taken to correct the identified violation and to prevent recurrence. CQnsequently. CQnsequently, n(1 n<' reply to this violation is required and we have no further questions regarding this matter. With respect to Item 2, pursuant to the provision of 10 CFR 2.201, 2.201. you are required to submit to this office wjthin wjthjn 30 days of the date of this Notice a wrltten wrHten statement or explanation in replYi reply}* including for this violation: (1) corrective correctlve action actlon taken and the results acnieved; (2) corrective action to be taken to avoid further, further violations; and (3) the

        -. date when full compliance will be achieved. Consideration may   lIIay be given to e~tending your response time for good cause shown.

extending shown . Dated Hubert J. Mlller, 01rector Dlrector Division of Reactor Safety ( ( r r

.C._-

III.7-4 III. 7-4

                                                                                                    ----.-~~~..............   "-

Revision 8 Apri April1 1992 (c U. S. NUCLEAR REGULATORY COMMISSION REGION III Rrports R~ports 5C-237/88010(DRS}; 50-249/88012(DRS) No. 5C*237/88010(DRS); 50-249/88012(DRS} Dock~t Nos. 50-237; 50-249 Licens~s No. DPR-19; DPR-2~ licenses DPR-2', L 1 Crnsee: Li(~llsee: CortlTlonwea 1th Edison COrmlonwealth Edi son Company Post Office Box 767 ChicdgO, Il Chicago, IL 60690 Fd~;lity Name: Fdcility ~ame: Dresden Dresd~n Nuclear Power Station. Station, Units 2 and 3 Icspectioo r~sp~ctio~ At: Dresden Sit~, Morris, Illinois Il1sprctioll Conducted: IIlSpcctioli Conduct~d: April 18-22, May 11-13, August 15. 15, and Decemb~r December 13, 1988 (\.~ Inspectors: Inspt!ctors: 1. Holmes /z./z7/88 _ l'LIZ7IB8 Date

                        ~~~
                        ~R. ~ Hodor (BNl)

Hador f,.rt-(BNL) dZ-/7..7 {SB._ v'*** ate

\                        ~,~~
                         ~.~~

K. Parkinson (BNL) ,z./1. 7 {BB Date Wl~J{1~

                         ~lw.JYJ~~

Approved Approv~d By: R. Gardner, Chief 2/. ). . .

                                                                                            , 21.

I :.~ ~'i: }..- ~!~, Plant System Section Date Dat~ SurtlTlary Inspectiol1 SUlT1TIary Inspection Ins ~ction on A ril 18-22. 18-22, Ma 11-13.11-13, Au ust 15 and December 13, 1988 {Re (Re arts orts No. 50*,50-, 7 88010  ; 50-249/88012 S redS nspecte: pecia, announced inspection conducted to asses's plant pecta, compliance with 10 CFR Part 50. 50, Appendix R and to review imple~ntation of certain F;reFire Protection Program requirements. Th~ The inspection was perfonned accordance with NRC Manual Chapter Procedures 30703, 64100. in accorcance 64100, and 64704. Results: Of th~ the areas inspected. inspected, two apparent violations were.identified. licensee has developed a safe shutdown methodology to prev~nt Tt.~ Hcensee Ttl!:: prevent fuel cldd damage. or rupture of any allY ~rimary j)rimary coolant boundary in the event of a disabling' fi r~ in the plant. "iolo'f!Ver, f;r~ Yo.'ever, the methodology chosen by thE: the 1licensee i censee does nut il,.:orporato! ClQ dt:aicdt~d illl:orporatt! deaicdted safe shutdown pan~l pano!l for a disabling fire requiring tht-' th~

      ~VdcUdtion of tht:

o!vdcudtion th" control room, but rt:l rel ies on many manudl actions to dchi~vt.* dchiev~ III.7-5 IIL7-S

Revision 8 April 1992 sof~ Saf~ shutdown conditions. The str~ngth strength or weakness of this program in achieving its goals in saf~ly safely shutting down the r~actors will b~ d~p~nd~nt be d~pendent training, prudent use of administrative controls and upon good op~rator trdining, maintailling th~the present fire protection systems. Weaknt!sses Weaknesses observed includt'd includ~d th~ following: (1) licensee did not provide administrative controls to insur~ thdt tht' th~ required oppusite opposite unit equipment was available availabl~ for the operating unit wh~n the r~quir~d requir~d opposite unit equi~nt equipment was down for repair {Paragraph 3.g:. 2.g:. ad,"inistrativ~ controls for combustibles were not effectively utilized (2) administrdtive in that ill tha t the 1licensee i censee permitted permi t ted the storage of twenty 55 gallon ga 11 on drums of 1luhe uhe oil in d safety-related area where an exemption from the installation of a sprinkler system has been submitted to NRR due to th~ the lack of combustibl~s in tht' area (Pdragraph 4.b); (3) Unit 1 is no longer operation~l th~ drea operational and does not adequately isolated from Units 2 and 3 (Paragraph Z.e.); appear adequ4tely (4) in th~ 2.e.); and {4} disabling fire, two hot shorts in multiple conductor cables 33674 ~vent of a disdbling dna dno 33934 could cause the spurious operation of the target rock and the el~ctromilt;c el~ctromatic rel relief Valves. While the safe shutdown analysis ief valves. analYSis addresses spuriuus uperation of one valve.valve, the simultaneous spurious opening of the TargetTdrg~t Ruc~ V~lv~ Valv~ and all of the Electromatic Relief Valves has not b~en analyzed J.f). Strengths were not~d (Paragraph 3.f). noted in the application applicatiun of salient fire pr0tection features between Units 2 and 3 (Paragraph 3.h) and also in the coordination and execution of the fire pump capacity test (Paragraph 2.c). III.7-6

Revision 8 April 1992 ( DETAILS DnAI LS

1. Persons Contacted Commonwealth Edison Company (CECo)

                   #*+E.
                   #*+E. Eenigenburg, Station Manager 10.
                      #0. Barnett, Quality Assurance
                     *+B.
                     *+8. Bartil, Technical Staff Engineer Sartil,
                       *W.
                       *W. Betourne, Quality Assurance
                     #*R.
                    '*R. Black, Assistant Fire Marshal
                    #+J.
                    '+J. Brunner, Assistant Superintendent Technical Services
                    '*R.
                    #*R. Christensen, Operations
                  #*+M. Dillon, Fire Marshal
                    #+T. Hausheer, Fire Protection Engineer
                  #*+R.
                  '*+R. Johnson, Technical Staff Group Leader
                    #'J.
                    #*J. Kotowski, Operations Assistant Superintendent
                    #+T.
                    '+T. Lewis, Regulatory Assurance
                      *G. Mauropoulos, Boiling Water Reactor Engineer
                      +E. Netzel, QA Superintendent
                    #+K.
                    '+K. Peter~an, Regulatory Assurance Supervisor
                      *w. Pierce, Engineering Support Service NO.
                      #0. Roberts, Fire Protection Engineer
                      *R. Roebert, BWR Engineering
                    *+C. Schroeder, Services Superintendent
                    #+J. Silady, Nuclear Licensing Administrator E. Skowron, Technical Staff Engineer

( *+R.

                      *J.
                      ~J.

Stachniak, Technical Staff Engineer Wajciga, Production Superintendent

                  #*+R. Whalen, Tech Staff Mechanical System Group leaderLeader
                      *J. Williams, Regulatory Assurance Sargent and Lundy (S&l)   (S&L)

R. Brown, Electrical Engineer F. Fisher, Electrical Engineer J. Ke lly';' Boiling Kelly~' Boil i ng Water Reactor Engineer Engi neer C. Ruth, Electrical Engineer Professional loss Loss Control (PLC) oM. Mowrer, Vice President

                   *M.
                   *C. Ksobiech, Senior Fire Protection Engineer U.S. Nuclear Regulatory Commission (NRC)
                   +S. Dupont, Senior Resident Inspector
                   *0
                   #D .. Jones, Project Inspector
                   *P. Kaufman, Resident Inspector
                   .Denotes those attending the April 22, 1988
                   *Denotes                                        19BB exit meeting.
                    +Denotes those attending the October 18, 1988 exit meeting.
                    #Denotes those attending the December 13, 1988 exit meeting.

NOenotes i (

j. ~ .... ,.-

3 111.7-7 Ill. 7-7

Revision 8 April 1992 . 2. Licensee Actions on Previous Inspection Findings a. In the 1licensee's icensee' s letter dated January 24, 1986, *to to J. Keppler, NRC, from D. Farrar, CECo, the licensee responded to the violation by indicating that a task force had been assembled to examin~examine the various fire protection duties and tasks that are required to be performed on a company wide basis. The licensee also indicated that, in the interim, a Nuclear Service Technical Fire Protection Engineer from the General Office would assist the station one day Engin~er per week until the Task Force Report is accepted and implemented. The licensee further indicated that full compliance would be achieved after the Task Force force recommendations had been reviewed, evaluated and implemented to the extent deemed necessary. On July 16, 1986, a followup meeting was held in Region III. In this meeting, the licensee presented several of the recommendations developed by the task force which included providing an Assistant fire Marshal to the Oresden Fire Dresden site and the formation of a fully fire Protection Group by late 1987. The licensep staffed Corporate Fire licensee indicated that an Assistant Fire Marshal at Dresden was hired as a ( result of the Task Force'recommendation. However, the Task Force recommendations had not been fully implemented and were being reviewed by upper management. The inspector subsequently requested the 1icensee licensee to provide a completion date as to when the Fire Protection Task Force . recommendations would be implemented. In a letter dated June 10, 10. 1988, from C. Reed, CECo, to A. B. Davis, NRC, the licensee stated that during March 1988 Executive Management had reviewed previous fire protection program assessments and the status of the Fire Protection Task Force Report. The review concluded that increasing the size of the Corporate Firefire Protection Group was a desirable enhancement however Executive Management concluded that at that tiM tt.e tt,e group did not need to be as large as recommended recolllllended by the Task Force Report. The June 10, 1988 1etterletter states, "In summary, summary. Connonwealth COMmonwealth Edison believes that the fire protection program deficiencies at Oresden appl icable corporat~ Dresden have been corrected and applicable corporate recommendations have been implemented. Further further implementation of the corporate Task Force force recommendations will be driven by the desire to achieve excellence in fire protection for all our planL~ planh including Dresden.H Dresden." iC-**' i 4 IIl.7-8 III. 7-8

Revision 8 April 1992 c Based on the licensee's actions of assembling a Fire Protection Task Force to examine various fire protection duties on a company wide basis, hiring an Assistant Fire Marshal and implementing applicable Task Force recommendations, this item is considered closed.

b. (Closed) Unresolved Item (237/85033-02(DRS)A 249/85029-02(DRS>>:

The qualifications quallflcatlons of the Station Statlon Fire Flre Mars al did not appear to be commensurate with the list of responsibilities assigned to that position. The lengthy list of responsibilities constituted a work load that may not have been achievable by a single individual, regardless of the individualls individual's qualifications and experience. 24, 1986 letter, the licensee indicated that a Task In the January 24. Force had been assembled to examine the various fire protection duties and tasks that are required to be performed on a company wide basis. The Task Force duties included review of the primary responsibilities of the Fire Marshal position. The Task Force recommended a proposed organizational structure for effectively performing fire protection duties in the company. The Task Force indicated_ indicated that given the numerous duties and responsibilities at the station level, all nuclear stations needed to provide a full-time Assistant Fire Marshal Ma,'shal and fire brigade instructor, in addition to the Fire Marshal. Consequently, in a June 10, 1988 letter from C. Reed, CECo to A. 8. Javis, NRC, the licensee stated that Commonwealth Edison believed that the fire protection program deficiencies at Dresden had been corrected and applicable corporate recommendations had been implemented. Based on the licensee's actions of hiring an Assistant Fire Marshal, Marshal. providing at least one qualified fire brigade instructor to assist the Fire Marshal, and the June 10, 1988 response, this item is considered closed.

c. o en) Unresolved Item (237/85033-03(DRS)' 249/85029-03(DRS>>:

evera 0 t e lcensee s lre protectlon ec nlca pecl lcation survei'lance procedures did not contain appropriate test requirement* and fail ed to incorporate qual i ty affecting parameters as de I ineatpu in NFPA standards. In the letter dated January 24, 1986 to J. Keppler, NRC, from D. Farrar. Farrar, CECo. CECo, the licensee responded to the items identified by the inspector. The licensee indicated that as a result o~ an NFPA code review the surveillance procedures would be reVIsed. During this inspection, the inspector reviewed the revised surveillance procedures that were previously identified as deficient. During this review. review, the inspector identified the following concerns: i ( 5 III.7-9

Revision 8 April 1992 ( (1) Oiesel Diesel Fire Pump Testing (a) The inspector reviewed the updated licensee's diesel fire pump annual annua I capacity capaci ty checK check and weekly week ly operability operab iii ty surveillance procedures to verify the automatic operation of the diesel fire pumps. The inspector noted that the annual capacity check procedure did not verify automatiL automatil operation ope rat i on of the fire fi re pump. The licensee contended that tha t the fire pump is automatically started at least once a month. The weekly surveillance procedures direct the testing personnel1 to automat personne automatically i ca 11 y start the fifire re pump by open i n~l openin~1 the test petcock which is on the side of the fire pump controller. In the 15th edition of the Fire Protection Handbook, Section 16. 16, Chapter 6, Paragraph 5, titled "Annual Pump Test,!! Test," it indicates that when testing the pumps it is not sufficient to initiate a pressure drop by the test cock on the controller to simulate automatic operation. On June 2, 1988, the inspector informed the licensee that at least once a year, preferably during th~ annual fire pump test, the automatic mode of the controller should be tested by opening a two inch drain valve (on a fire protection water system riser) or hydrant as illfened infened from the 15th Edition of the Fire Protection Handbook. The licensee agreed to incorporate into the annual pump test procedure a simulated pressure drop by opening a two inch drain on a fire protection system riser or opening a fire hydrant. (b) In the Unit 2/3 Diesel Fire Pump Check Surveillance Procedure, it indicates in Section C titled, JtPrerequisitf.'<,1I "Prerequisit@," that IIIf "If a vibration analysis machine ;s is to be used, the Fire Marshal should contact the cognizant Technical Staff Engineer." Engi neer." The inspector di scussed wi th the licensee with I icensee the es tab Ii shme'lt estab shme"t of vibration analysis baseline data for the diesel fire pum~5 pum~, and conducting the fire pump vibrational analysis test in conjunction with the annual fire pump test. The licenspp licensee indicated that the vibrational analysis will be performeu performed as part of the annual fire pump test. (c) (e) In the NFPA 20 Formal Interpretations, No. 83~2, it indicates that the results of the annual fire pump test should be compared to the manufacture's certified shop test characteristic curve and field acceptance characteristic curve to determine the pump's ability to continue to atta;n attain satisfactory performance at peak loads. 6

                              !IL7-IO III.7-10

Revision 8 April 1992 ( During the previous inspection, it was identified by the manufacturer's5 shop te5t NRC inspector that the original manufacturer' test curve or field acceptance test were not available to the licensee's staff. Since the previous inspection, the licensee has developed fire pump curves from the manufacturer data plates on the fire pump. The licensee developed has incorporated the deve the' ,. loped fire pump curves into tht-; procedures as part of upgrading the Fire Pump Capacity Check Procedure. During this inspection, at the request of the inspector, the licensee performed a capacity check for the Unit 2/3 Diesel Fire Pump. The licensee performed excellently in the coordination and execution of the test No discrepancies were noted from the test results. At th~ the request of the inspector, the licensee agreed to update the procedure to include certain pump parameters such as water jacket temperature and oil pressure. (2) Testing of Water Suppression Systems 4.12.8.1(e) of Technical Specifications requires that Section 4.12.B.l(e) fire suppression water systems be demonstrated operable by p~rforming a system functional test which includes simulated performing automatic actuation of the systems throughout their operating sequence. The licensee's commitment in Section 3.5.£.3 3.5.E.3 of the Fire Hazard Analysis (FHA) Report requires that automatic sprinkler systems conform to NFPA Standard No. 13. ( The licensee's Surveillance Procedure No. SP 84-6-39 failed til t" incorporate appropriate test requirements to demonstrate that the sprinkler system is operable in accordance with NFPA 13

     ;n in that the procedure did not require flow from the two inch drain valve of wet or dry pipe sprinkler systems.

The licensee indicated to the inspector that the two inch drain test is not conducted because the fire protection wate. wate, (river water) destroys the radwaste demineralizer beds. The licensee contends that the two inch drain test does not need to be conducted because the fire protection control valves ar~ are provided with tampers or locks that ensure an adequate water contend~ supply will be available. In addition, the licensee contend5 that water is available to the sprinkler system because the inspector test is conducted. The inspector discussed several methods of conducting the two inch drain test that would provide assurance that the system is operable and minimize impact to the demineralizer beds. The inspector informed the 1 icensee that the two inch licensee regardi"1j drain test should be performed and that any deviation regardir.q the two inch drain test should be adequately justified and documented in the NFPA Code Review Section of the Fire Hazanl Analysis. 7 III.7-11 III. 7-11

Revision 8 April 1992 ( The licensee indicated to the inspector that resolution to this issue will be provided tentatively by January 1, I, 1989. Therefore, the unresolved item will remain open pending review and acceptance of the licensee's resolution to this issue.

d. Closed) Violation (237/85033-04 DRS)* 249/85029-04 DRS : An ear y warnIng warnlng automa 1C Ire etectIon lC lre etectlon system was no 1nstalled lnstalled in the Refueling Floor Area as required by provisions of Amendment Alltendment No. 36 to Provisional Operating license License No. DPR-19 and Amendment No. 33 to Facility Operating license License No. OPR-25.

DPR-25. The l;censee has requested an exemption from the requireMents requirelents of providing a fire detection system on the Refueling Floor which is currently being reviewed by NRR. Based on the exemption request, this item is closed. .

e. (Closed) 0 en Item (237/85033-05 DRS)* 249/85029-05(DRS>>: The lcensee agreed to up ate t elr response to t e an escribe the administrative controls and the actions that will be necessary to isolate Unit 1 from Unit 2 and Unit 3 since Unit 1 is no longer operational, but shares common areas with Units 2 and 3. The inspectors also requested that the licensee describe those administrative controls and actions that will be necessary to sera rate common areas.

In a letter dated January 24, 1986, to J. Keppler, NRC, from fro~ D. Farrar, CECa CECo (Responding to the Open Item), the licensee indicated that a stricter transient combustible control procedure

• c.(~

would be developed. In addition, a cognizant foreman was designated to assist the Fire Marshal in timely correction of housekeeping deficiencies. The licensee indicated that a detailed memorandum discussing the proper handling of fire barriers had been discussed with all personnel at the station as part of the weekly "tailgate" "tai 19ate" staff meeting. Also, Procedure No. OFPP DFPP 4175-1. 4175-1, Fire Barrier clarify the Integrity and Maintenance, has been revised to further clar)fy proper handling and maintenance of fire barriers (including fire doors, fire dampers, fire walls, penetration seals) for mechanical and* electrical components. Based on the licensee's updated response, this Open Item is considered closed, although other specific concerns are being raised as d&scribed d~scribed below. In the letter dated January 24, 1986, to J. Keppler, NRC, from D. Farrar, CECo, the licensee indicated that the separation of Unit 1 was being covered by the Appendix R review program and that this information was being added to the updated FHA for Units 2 and 3. In Section 4.15.9 of the updated FHA the analysis describe~ describe, t/,e only portion of the Unit 1 structure ~hich that tI,e con~acts ~he. which contacts the . Unit 2/3 structures is the west wall of the Un1t Unit 1 Turblne Turbine BUIldIng. Building. The FHA also indicates that the wall separating the Unit 1 Turbine Building from Auxil iary Electric Equipme.nt Auxiliary Equipment Room (AEER) (Fire Zone (,.2) 6.2) c.~ . 8 III.7-12 III .7-12

Revision 8 April 1992 ( is a minimum mlnlmum 3 foot 3 inch reinforced concrete three hour fire batTier. ba'Tier. The remaining wall west of the Unit 1 Turbine Building has metal siding on unprotected structural steel with openings (non-fire rated doors) that expose Unit 2 Safety Related Areas. The Unit 2 side of this portion of the west wall is identified as Five Zone B.2.5.A. S.2.S.A. During this inspection, the inspectors noted large ~ounts amounts of RAD worker's clothing and flammable/combustible liquids stored in Unit i in an area where if a fire occurred the Unit 2 Fire Zone 8.2.S.AS.2.S.A (Safety Related Area) may have been exposed since there is no fire rated barrier between the two areas on Elevation 517'-6 11 517'-6".

• The licent'"ee 1 icen-ee indicated to the inspector that in the event of a fire fl'om from Unit 1 affecting the Unit 2 side (Fire Zone B.Z.S.A)

S.2.S.A) one Safe Shutdown Path would still be available. The licensee's response would allow a fire to migrate from Unit 1 to Unit 2 AEER and does not appear to be consistent with Section F.IB F.18 of the licensee's FHA which indicates that storage areas should be located such that a fire or effects of a fire including smoke will not adversely affect any safety-related systems or equipment. On December 13, 1988, 198B, during a site visit, the inspector was infonned informed by the licensee that the present abandoned equipment in Unit 1 restricts large amounts of combustible storage. The licensee 1icensee agreed to limit the combustible loading in the area to a low fire load (as defined by plant combustible load procedures) to at least 20 feet from Unit 1 control room walls and the metal wall between Unit 1 and 2. The inspector also observed two non-rated metal doors between Unit 1 and Unit 2 that were maintained in the open position. Unit 1 is currently being Decommissioned and it is expected that combustibles will be stored in Unit 1 and that cutting and welding operations will be performed. It is the inspector's concern that the Unit 2/3 Control Room and Unit 2 Safety Related areas may be exposed to Unit 1 fire since they are not separated by three hour fire walls or other recognized fire protection methods of protecting safety-related areas from adjacent exposures. This is considered an unresolved item (237/88010-01(ORS); (237/88010-01(DRS); 249/88012-01(ORS>> 249/88012-01(DRS>> pending resolution from NRR. The licensee indicated that a three hour fire wall is tentatively scheduled to b~ installed between Unit 1 and the Unit 2/3 Control Room by December 1989.

3. Assessment of Appendix R Compliance Compl iance basis, the inspectors examined measures that the licensee On a sample basis.

implemented to assure safe shutdown capability and compliance with 10 CFR Part 50, Appendix R. The inspection consisted of an assessment of the licensee's implementation of Appendix R requirements for physical components, plant conditions, required operator actions, systems, and components. operator training, supplemental procedures, and methodology employed to mitigate resultant adverse equipment operability due to plant exposure to fires. The results of the inspector's review are as follows: c 9 III:7-13 III .7-13

Revision 8 ( April 1992

a. Systems Required for Safe Shutdown The Appendix R goals req'u'ired req"u*ired to achieve post-fire safe shutdown are:

• Reactivity control capable of achieving and maintaining cold shutdown reactivity conditions (reactor coolant temperature less than or equal to 200°F).
• Reactor coolant makeup capable of maintaining water level above the top of the core at all times during shutdown operation.
• Reactor pressure control and decay heat removal.
• Process monitoring capable of providing direct readings to perform and control the above functions.
• Supporting functions capable of providing process cooling, lubrication, etc., necessary to permit operation of the equipment used for safe shutdown functions.

In accomplishing the goals outlined above, the equipment and systems used to achieve and maintain hot shutdown conditions should be free of fire damage and capable of maintaining such conditions for 72 hours, using offsite or onsite emergency power. The equipment and syctems sy~tems used to achieve and maintain cold shutdown conditions should be either free of fire damage or the damage to these system~ system'

• c,. should be limited such that repairs can be made and cold shutdown conditions achieved within 72 hours, using offsite or onsite emergency power.

During the post-fire shutdown, the reactor coolant system process variables shall be maintained within those predicted for a loss of nQrmal ac power, and the fission product integrity shall not be affected; i.e. i.e., there shall be no fuel clad damage, rupture of the I containment boundary. (1) Reactivity Control The licensee takes credit for a reactor trip even for a postulated fire that requires evacuation of the control room. Upon loss of power, or in case of fire damage to the logic circuitry, the system is designed to fail safe (rods fully inserted) . (2) Reactor Coolant Makeup The isolation condenser method of hot shutdown utilizes the Control Rod Drive (eRD) (CRD) Hydraulic System to provide makeup to the reactor vessel. One of the two eRD CRD pumps per unitprovide~. unitprovidl". all of the reactor makeup required due to leakage and shrinkagp. shrinkage. during cooldown. The eRD CRD pumps take suction from the condensate storage tank and the condenser hotwell. The eRD CRD pump*s pump's dischargp 10 III.7-14 III, 7-14

Revision 8 April 1992 ( pressure can be monitored locally on mechanical indicators P12(3)-302-73A PI2(3)-302-73A and PI2(J)-J02-7JB. PI2(3)-302-73B. Local control pushbutton stations have been installed for the CRD pumps. The CRD water headers for the two units are connected with a crosstie line which is normally isolated by manual valves. The valves are located on the mezzanine level of the Turbine Building in th~ thp area with wi th accessibility access i bil i ty to either ei ther set of pumps. Therefore. Therefore." <t fire in one unit will not prevent the other unit's pump from supplying makeup water to the affected unit. The CRO CRD pumps C~I. Cd' be cooled by the Service Water system if normal cooling from the Turbine Building Closed Cooling Water (TBCCW) system is lost. For those fires where the isolation condenser method of shutdown is unavailable, unavailable. the High Pressure Coolant Injectioll (HPCI) system is used. The HPCI system consists of a steam turbine driven pump that can take suction from either the suppression pool or the condensate storage tank and pump water to the reactor vessel. The steam that drives the turbine come~ come, from the reactqr reacto.r and is exhausted to the suppression pool. The HPCI system automatically initiates on low-low water lev~l HPCr level signal (-59 inches) or can be manually initiated from the contro 1 room. control The HPCr HPCI pump injects water from the condensate storage tank to the reactor vessel. The HPCI system pumps makeup water to the reactor at a rate of 5,600 gpm. The operator can manually { operate the flow controller in the control room. Condensate storage tank level is normally monitored in LI2/3-3341-3 and the control room using level indicators l12/J-3341-J L12/J-J341-4. Ll2/3-3341-4. Level can be monitored on mechanical indicators L12/J341-77A LI2/3341-77A and l12/J341-778 LI2/3341-77B located in the Turbine Building in the southeast corner of the Unit 2 reactol feed pump room. If long-term operation of the HPCI system depletes the condensate storage supply, the operator will align the HPCI suction with the suppression pool by opening Valves M02(J)2JOl-35

     -Valves   M02(3)2301-35 and M02(3)2301-J6.

M02(3)2301-36. The HPCI suction

      ;s is automatically shifted to the suppression pool when the condensate storage tank contains less than 10,000 gallons.

HPCI pump discharge pressure can be monitored in the control room on pressure indicator P12(J)2J40-2 PI2(3)2340-2 and locally on mechanical indicator P12(3)-2357. PI2(3)-2357. (3) Reactor Pressure Control and Decay Heat Removal Initial pressure control and decay heat removal for the reactul' react"r is supplied by the electromatic relief valves. However. the target rock valve (mechanical (meChanical mode) and mechanical safety valves on the steamlines will provide these functions if operation for the relief valves has been affected by a fire. Long term (up to 72 hours) reactor pressure control and decay 11 I1I.7-15 IIL7-IS

Revision 8 Apri April1 1992 ( heat removal remova lisis provided prov i ded by the isolation i so I at i on condenser system sys tem in that the system is sized to handle the total decay heat load five minutes after scram. The isolation condenser consists of two tube bundles in a large water-filled shell. The reactor fl ows through the tubes, is condensed, and returns to th~ steam flows the reactor vessel. The water in the shell is boiled off and vented to the atmosphere. The vent line to the mal mai n steaml steam I ii !'It* r,,* is isolated upon initiation of the isolation condenser system. If a fire has affected automatic operation of the accessible isolation condenser ~alvesvalves (H02(3)-1301-2 and H02(3)-1301-3; H02(3)-1301-2 1S Valve M02(3)-1301-2 is normally open), the operators can remove power from the appropriate motor control centers so that the valves may then be opened by use of handwheels. Normally open valves H02(3)-1301-1 and H02(3}-1301-4 H02(3)-1301-4 are located in the drywell and are therefore not accessible for manual operation. In the event a fire causes these valves to spuriously close, an alternate 480V power feed to each of these valves is provided along with a local control station. In addition, isolation switches have been installed for the normal control and power cables. If the valves spuriously close, the alternate feed is energized and the valves opened. The operator can then deenergize the valves in the open position. Valve M02(3)-1301-3 H02(3)-1301-3 is manually throttled to control the cool down. cc. . Initial makeup to the condenser will be supplied from the condensate storage tanks via the condensate transfer pump. Wi th no makeup, the water stored above the isolation With tubes is depleted in 20 minutes after initiation of the i so 1at i on condenseI' condense I' isolation condenser system. The isolation condenser level is normally monitored in the control room on level indicator Ll2(3)-1340-2. The operator can locally monitor the level in L12(3)-1340-2. the isolation condenser on an existing sight glass by opening two manual valves. Any of the four condensate transfer pumps (two per unit) can supply makeup water to either unit's isolation condenser through the norma1lynormally open tie line. Therefore, a fire in one unit will not prevent the other unit's pump from supplying makeup water. When the HPCI shutdown method is used, reactor pressure control and decay heat removal are accomplished by the HPCI turbine (driven by reactor vessel steam) in conjunction with electromatic 2(3)-0302-38 through 2(3)-0203-3E. The HPCI relief Valves 2(3}-0302-3B turbine steam supply line, the target rock valve, and the ee'ectromatic 1 ectromatic relief valves discharge to the suppression pool. Continued operation of the HPCI system r.esults in heatLJp heatllp of the suppression pool water. One division of Low Pressure (LPCI)/Component Cooling Service water (CCSw) Coolant Injection (lPCI)/Component is sufficient to remove decay heat from the suppression pool.

      ;s LPCIICCSW system into operation The operator manually places the LPCI/CCSW                               operat ion
      ;n in the torus cooling mode from the control room, thus lIaintitillill~  maintail\in~

12 IIU-16 IlL7 -16

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

Revision 8 1992 Apri 1 .1992 ( the water temperature within ac~eptable acceptable limits. The operator can also throttle flow as appropriate to obtain the desired cool iny. Each LPCI pump is capable of providing a flow of 5,000 gpm. Each CCWS pump is capable of providing a flow of 3,500 gpm. (4) Process Monitoring j*he

        ~he    operator requires a means to ascertain the values of various plant parameters in order to perform required system transitions and essential operator actions. Various proces~process monitoring functions are available to adequately support reactivity control.

control, reactor coolant makeup, pressure controf~ control*, and decay heat removal as follows:

• Reactor Vessel Level
• Reactor Vessel Pressure
• Suppression Pool Level
• Suppression Pool Temperature
• Condensate Storage Tank Level
• Isolation Condenser Level Additionally, discharg~ pressure indication is provided for the eRD CRD pumps, condensate transfer pumps and service water pumps.

Support Equipment The following equipment is available for post-fire shutdown: (. ** Emergency Diesel Generators

       **       4160V 4160V ac
        **      480V 480V ac
        **      125V de dc
        **      120V ae ac

• Communication System

        ***     Emergency Service Water System (ESW)

• Reactor Building Closed Cooling Water System (RBCCW)

        ***     Turbine Building Closed Cooling Water System (TBCCW)
         **     Containment Cooling Water System (CCSW)
         **     Service Water System (SWS)

• Fire Water System (FWS)

(5) Cold Shutdown Two systems are identified at Dresden Station to bring the p~ant piant to cold shutdown (reactor coolant equal to or less than 212°F). The preferred shutdown cooling path is the shutdown the Shutdown ~ooling cooling path which utilizes the,Shutdown Cooling System (SD~S). (SDCS). For those fires where the SOCS available. LPCI/CCSW I~ is not aval1able. SDCS 1S i~ used to achieve and maintain cold shutdown. i( 13 IIL7-I7 III.7-17

Revision 8 April 1992 ('( The SDCS pumps take suction from the reactor recirculation loops through motor-operated valves 1001-lA and 1001-lB. IDOl-lB. These valves are inside containment. They are powered from 480VAC MCC 28-1 (38-1) which can be supplied from the emergency diesel generators. They are closed until initiation requirements (reactor coolant system temperature less than 3S00F) 350°F) are met and operator action is taken. The two inlet lines join in one header outside of containment. This header feeds three separate loops. Each loop has a DC powered motor operated pump inlet isolation valve (IOOI-2A, (1001-2A, 1001-2B, or 1001-2C), lOOI-2C). a centrifugal pump rated at 6.750 6,750 gpm at "full operation,1I operation," a heat exchanger, and a DC powered motor . operated pump outlet isolation valve (lOOl-4A. (1001-4A, 1001-48, 1001-4B, or 1001-4C). lOOl-4C). Downstream of the pump outlet isolation valves. valves, and still st ill outside outs i de containment, conta i nment, the three branches again.feed agai n feed a common header. This common header divides into two return lines. lines, each containing an AC powered motor operated isolation valve (IDOl-SA (1001-5A and 1001-5B). Each return line penetrates the containment and rejoins the reactor coolant system through connections into one division of the lPCI LPCI system. Each lPCILPCI division connects to one of the reactor recirculation loops. Although the capability exists to permit flow from and to both recirculation loops, normally only one loop is selected for such service. Either Recirculation Loop Valve(s) 0202-5A(B) and 0202-7A{B) 0202-7A(B) or 0202-4A(B) must be closed to prevent back flow through the reactor recirculating pump. The heat exchangers of the SDeS SDCS are cooled by water from the RBCeW RBCCW system. system, with the heat eXChangers of the RBCCW system in turn coo 1led ed by the SW system. If the SDCS is not avail ava il ab 1e, th~ LPCI system can be used to inject cooling water into the COf'e cOl'e or.ce the injection initiation limits (350 psig) are met. The system is a low pressure, high volume system capable of providill~J providin~1 substantial volumes of cooling water to the core. The pump i~ i, powered from "emergency" buses. buses, and a 11 motor operate-d operated va 1yes ves are powered from "emergency" HCCs MCCs and are also outside containment, accessible for manual operation if needed. The reactor vessel is allowed to fill using lPCI, LPCI, overflowing hot water to the pressure suppression chamber (torus) through the relief valves. The continuous cycle of water through the core, through the relief valves to the torus and back again after cooling via the containment cooling heat exchangers, would only be limited by the design of the relief valves themselves. These valves incorporate a spring which must be overridden by system pressure to open the valve. The valve will reseat at approximately 50 psig and will be held shut until the core heats up again and raises pressure, pressure. or until the pressure is increased to 150 psig LPCI pumps (de~igl1 ps i g by the lPCI (des i gil head 114 psig at 0 psig reactor pressure to 245 psig at 200 psig Suppress i on pool reactor pressure). Suppression poo 1 water is pumped through cooling heat exchangers and then injected into lh~ containment co01ing lhp reactor vessel.

c. 14 II 1.77-18 III. -18

Revision 8 April 1992 (

b. Alternate Shutdown The licensee has chosen five different Appendix R shutdown path~

paths per unit. Two of the paths per unit have been designated as a. alternative shutdown paths as described below:

• Path Al utilizes the Unit 2 pumps and power train by llechan;c Illechan;(" ..d.1 crossties to shutdown Unit 3 for a fire in Fire Area RB3-II RB3-11 n'.1 0'1 Fire Area 18-III.

TB-III.

• Path A2 is used to shutdown Unit 2 for a fire in Fire Area TB~vTB-,

TB-II. or Fire Area T8-II.

• Path 81 B1 utilizes Unit 3 pumps and power trains via mechanical crossties to shutdown Unit 2 for a fire in Fire Area R82-11 RB2-II andanti Fire Area TB-1.
• Path 82 B2 is utilized to shutdown Unit 3 for a f*ire f.ire in the Fire Area TB-V or Fire Area 18-11.

TB-II. For a fire in the control room or auxiliary electric equipment room requiring control room evacuation, the licensee has developed Procedure EPIP 200-20 for post fire safe shutdown.

c. Pro'cedures for Alternate Safe Shutdown Ie The licensee has developed Procedure No. EPIP 200-20, Revision 4, 19B8, to be used in the event of a fire in the Control dated April 1988, Room or the AEER which requires evacuation of the Control Room. A staff of 13 licensee personnel is used to implement the procedure which provides for achieving stable hot shutdown for both Units 2 and 3. A two-column format is used with one column assigning responsibility, and the other column listing the actions required.

The procedures include Attachments 1 through 9. Each attachment summarizes the actions for an individual operator. After stable hot shutdown conditions have been achieved, Procedure DSSP 200'-5 is entered to bring the units to cold shutdown. . Once the decision to evacuate the control room is made, the reactor~ reactor, are tripped from the control room driving the control rods in for hot hut shutdown reactivity control. Several other immediate actions are attempted in the control room prior to evacuation; however, if unsuccessful, they are covered by procedure from outside the control room after the evacuation. The scope of the team review was to ascertain that post-fire ~afe safe shutdown using the steps in the procedure could be attained in a safe and orderly manner, while achieving the functional goals of Appendix R. No unacceptable items were found by the team review . of the procedure. 15 IIl.7-19

Revision 8 April 1992 ( A walkdown of Procedure No. EPIP 200-20, Revision 4, April 1988, :'"

    "Contro "Control1 Room Evacuation/Safe Shutdown      ,II was conducted on April Shutdown,"                              20.

1988, at 1300 hours. The purpose of the walkdown was to determine by simulation that alternate safe shutdown could be implemented in a safe and orderly manner for a fire in the Control Room or AEER. Four inspectors accompanied the operators during the walkdowns. The following conditions were specified for the simulated shutdowIl. shutdow".

• Reactor at 100% power with systems lined up in normal full powp!" power configurat ion.

configuration.

• Credit for one manual action prior to evacuating the Control Contl'ol Room.
• Loss of offsite power.

loss

• Manual start of emergency diesel generator.

The teJm paid particular attention to the feasibility of each manual action, ease of access, operator familiarity with procedural steps and equipment, communications, emergency lighting. lighting, and the direction of the shutdown by the shift engineer. The walkdown was halted when the licensee had adequately demonstrated the capability to achieve simulated stable hot shutdown conditions. No unacceptable items were identified by the team during the .( walkdown. However, in subsequent discussion with the shift engine~I' the licensee was informep. informed. that a visual aid. engineer aid, showing on a single page the flow of actions for each of the nine Individual Operator' Operatol' Attachments, would facilitate the shutdown training provided by th~ Attachments. the shift ~ngineer. Angineer. The licensee~agreed licensee'agreed to implement this recommendation.

• Hot Shutdown Repairs and Manual Actions The licensee has identified in Section 7.3 of the Dresden Fire Protection Documentation Package entitled, "Procedures Shutdown," hot shutdown repairs and manual Relevant to Hot Shutdown."

actions necessary to achieve hot shutdown. NRR has reviewed the identified hot shutdown repairs and manual actions in th~ the July 17, 1987 SER. Approval was granted contingent upon verification verifi cat i on by the inspection team. The team reviewrev i ew conduct~d condue ted during the April 18-22. 18-22, 1988 audit focused on verifying that the necessary actions can be completed within the specified times for assuring safe shutdown. Based on a detailed review of the Dresden safe shutdown procedures, inclUding 200~20 procedllf'p including a walkdown of the EPIP 200:"'20 procedlll'" for shutdown outside the Control Room. Room, the inspection team determined (taking into account the licensee's available manpower for post-fire safe shutdown - 11 personnel exclusive of fire brigade) that post-fire safe shutdown can be 16 III'. 7-20 IlL7-20

Revision 8 April 1992 ( accomplished. This included the initiation of makeup to isolation condenser shell within 20 minutes, and closure of d

           'puriously opened relief valve within 10 minutes.
           ~puriously

d. Operator Training on Safe Shutdown Procedures In addition to observing the operator's performance during the walkdown, training personnel were interviewed and lesson plans reviewed concerning operator training on Appendix R post-fire saf~

shutdown procedures and equipment. Training records for operating shift personnel were also reviewed. The areas reviewed were found to be satisfactory.

e. Protect i on for As Protection soci ated Ci Associated rcuits Circuits The licensee's associated circuits analysis was provided in Dresdell Dresden Station Units 2 and 3 Fire Protection Program Documentation Package. Package, Volume 3, Book 1, Section 3.3, Associated Circuits.

The following associated circuits were evaluated:

• Common Bus Concern The common bus associated circuit concern is found in circuits, either safety-related or non safety-related, where there is a common power source withwi th shutdown equipment equi pment and the power sourcf' sou)"o' is not electrically protected from the circuit of concern.

I(

• Spurious Signals The spurious signals concern is made up of two items: item.:

The false motor, control, and instrument readings such a~ a, Occurred at the 1975 Browns Ferry fire. The~e those which occurred The,e could be caused by fire initiated grounds. shorts, or open grounds, short~, circuits. Spurious operation of safety-related or non safety-related components that would adversely affect safe shutdown capability (e.g., RHR/RCS isolation valves).

• Common Enclosure The common enclosure associated circuit concern is found when redundant circuits are routed together in a raceway or enclo~o.Jrl:'

enclo>u)"e and they are not provided with adequat~ adequate el~ctrical electrical ls~latlon isolation protection, or fire can destroy both clrcults circuits que due to lnadequatp inadequatp fire protection methods. The inspection results were as fol1ows: follows: 17 111.7-21 III.7-21

Revision 8 April 1992 ( (1) Common Bus Concern The common bus concern consists of two items:

• Circuit Coordination
• High Impedance Fault Analysis (a) Circuit Coordination Breaker Coordination is audited by reviewing the time current curves developed during the licenseels licensee's bus coordination study. Licensee representatives stated thaI. that.

the original plant design provided circuit coordination. However, documentation demonstrating coordination of electrical devices was not provided to the electrical inspector. Additionally, licensee representatives stated that in the 480V 4BOV distribution systems circuit coordInation coordination does not exist for some circuits. The ~icenseels licensee's analysis identifies the lack of coordination between the 480V Switchgear Buses 18, 1B, 19, 28, and 29 main feeds to HCCs MCCs and the motor control branch circuits. Based on the existing lack of coordination for 480V MCCs. HeCs. the lack of readily available records, the lack of coordination curves demonstrating coordination, and the requirement to provide protection in the case of high faul ts and spurious impedance faults spuri ous operations, operat ions, the licensee ha~ ha*.

• c provided circuit coordination by manual operations. The manual operations specified in procedures include: circuit breaker. disconnect, and switch operations and fuse remo~al. removal.

The following circuits were randomly selected for review to verify that circuit coordination was provided procedura lly: CIRCUIT COr+tENT COMMENT 4kV Bus 23 Coordinated by procedures procedure~ 4kV Bus 24 Coordinated by procedures 480V Bus 38 Coordinated by procedures 480V Bus 39 Coordinated by procedures procedure~ 480V MCC 28-2 Coordinated by procedures procedure~ 125V DC Bus 3A Coordinated by procedures 125V DC Panel No. 2 Coordinated by procedures Manual credit for breaker coordination was found to be sat i sfactory. satisfactory. Control of fuse replacement is required to ensure maintenance of coordination for circuits protected by fuses. The 1licensee icensee does not have an establ i shed pro~r es tab I ished proljr ...ull UlI control I ing fuse replacement. By memo or procedure for controlling dated April 20, 198B, the licensee promulgated the

20. 1988, following policy on replacing blown fuses:

18 III.7-22 111.7-22

Revision 8 April 1992

• Compare the new fuse to the old fuse to verify that. that they are "like for like."

like." This comparison should include manufacturer, physical size, shape, voltage rating, current rating, and fuse type (quick-acting. slow-blow, etc.).

            *                       ~issing markings on the old fuse Jr.

If illegible or missing not permit a complete verification of voltage and current ratings or fuse type, the Shift Supervisor will obtain verification of such data from wiring manual s or by consultat ion diagrams and/or vendor manuals with the Technical Staff. The licensee's policy on replacing blown fuses will provide protection for fuse coordination. The effectiveness of this policy will be reviewed during subsequent inspections. (b) High Impedance Fault Analysis The high impedance fault concern is found in the case where multiple high impedance faults exist as loads on a safe shutdown power supply and cause the loss of the safe impedance shutdown power supply prior to clearing the high impedancp. fault. Since the licensee's procedures to .anually manually coordinate electrical circuits will provide protection for the high impedance fault concern, the licensee's (.( protection for high impedance faults was found to be satisfactory. (2) ~urious Signals (a) High/Low Pressure Interfaces High/low pressure interfaces are examined to determine it the licensee 1i censee has provided prov i ded measures to prevent fire fi re indUCE-ii i nduc~d spurious signals from producing a fire induced loss of coolant accident (LOCA). NRC guidance for protecting high/low pressure interfaces includes:

• Multiple (unlimited) hot short circuits, open circuits, and short circuits to ground are credible circuits.

(the single spurious signal criteria does not app apply). Iy).

• Three phase hot short circuits are credible.
• Hot short circuits in ungrounded DC circuits are credible.

The above guidance was employed in the review of the high/low pressure interface spurious signal concern at the Dresden Nuclear Station. 19 III.7-23

Revision 8 April 1992 ( The licensee identified the shutdown cooling system on Units 2 and 3 as being high/low pressure interfaces. The licensee's analysis 1i censee' sana lys i s demonstrated Un it"~ demons trated protection for the Unit and 3 shutdown cooling high/low pressure interfaces. Appendix C. C, shutdown cooli ng system high/low cooling high/low pressuf'e pressuI'e interface protection, provides the technical details pertaining to the review of the Dresden Units 2 and 3 shutdown cooling high/low pressure interfaces. (b) Isolation of Fire Instigated Spurious Signals The licensee has provided isolation for fire instigateu-instigateu' spurious signals by various methods, including:

• Administrative controls
• Isolation/transfer switches
• Fi re wrap Fire
• Cable relocation
• Manual component operation The licensee has requested exemption for hot shutdown repairs to accomplish the following:
• To allow the pulling of fuses in order to place the condensate transfer pumps into local control.
• To allow the pulling of fuses to defeat high impedance faults.
• To allow the pulling and replacement of fuses on selected control circuits in lieu of redundant fusing.

The licensee's methodology of pulling fuses is considered a hot shutdown repair which is not permitted by Appendix R. The licensee had previously submitted an exemption reque- reque' tI for fuse pUlling. This is considered an Unresolved Itew Ite"- (237/88010-02(DRS); 249/88012-02(ORS>> 249/88012-02(DRS>> pending disposition of the licensee1s licensee's exemption request. (3) Common Enclosure The common enclosure associated circuit concern is found when whell redundant circuits are routed together in a raceway or enclu~un~ enclu!>ure and they are not elec~rically protected, or fire can destruy destroy both circuits due to inadequate fire protection means. During the inspection, licensee representatives stated:

• Redundant safe shutdown cables are never routed in common enclosures.
• c 20 III.7-24

Revision 8 April 1992

• Non safety-related cables may be routed in common enclosures with safety-related cables, but non safety-related cables are never routed between redundant safety-related divisions or trains.
• All cables are electrically protected.

During the inspection, randomly selected non safety-related cables routed in common enclosure with safety-related cables were verified to be electrically protected.

f. Fire Instigated s~urious o~eration 0leration of Unit 3 Target Rock Valve and Electromatlc ellef Va ve The NRC electrical inspector identified that, in the event of a disabling fire two hot shorts in a multiple conductor cable would cause spurious opening of the target rock valve and the electromatic relief valves. The licensee indicated that, based e1ectromatic on Generic Letter 86-10 and discussion held with NRR, they had analyzed and provided protection for spurious operation of the target rock valve or one of the electromatic relief valves.

During the inspection the Appendix R inspection team consulted with NRR and were advised that the target rock valve and the electromatic relief valve~ valves. were not considered to be high/low pressure interfaces. The NRR Technical Reviewer was informed of the potential simultaneous spurious operation of the target rock valve and all of the electromatic relief valves by failure of control cable 33934. The NRR Technical Reviewer stated that if failure of a single cable could cause the simultaneous spurious operation of more than one single electromatic relief valve, than a safety concern may exist. Further discussions between the NRR Technical Reviewer and the electrical inspector identified control cable 33934 as being a potential cable separation or common enclosure concern. The inspector review of control circuits for Target Rock Valve 203-3A and Electromatic Relief Valves 203-38, C. C, 0, and E identified the following:

• The Target Rock Valve and Electromatic Relief Valves open when 12SVOC 125VDC power ;s is supplied to the Target Rock solenoid or the respective Electromatic Relief Valve pickup coil.
• 125VDC power is supplied to the Target Rock Valve or Electromatic Relief Valves via the following relay contacts (Note: the listed relay contacts are installed .in series wilh the respective solenoid or pickup coil):

VALVE RELAY CONTACTS RELAY CON1ACTS CON1ACIS 203-3A 2203-32/287-1068 7&8 2203-32/287-1078 8&7 6&7

( 21 III. 7-25 III.7-25

Rev;s;on Revision 8 Apr; April1 1992 ( .

                                                                                        .,"'-:~-
                                                                                          .~~

203-38 2203-32/287-1068 9&10 2203-321287-1078 10&9 or or 2203-321287-106A 5&6 2203-32/287-107A 2203-321287-107A 6&5 203-3C 2203-321287-1068 2203-321287-106B 3&4 2203-321287-1078 4&3 or or 2203-321287-106A 2203- 321287-106A 7&8 2203-32/287-107A 2203-321287-107A 8&7 203-3C 203-3[' 2203-321287-106A 2203- 32/287-106A 9&10 2203-321287-107A 10&9 or or 2203-321287-1068 5&6 2203-321287-1078 6&5 203-3E 2203-321287-106A 11&12 2203-32/287-107A 2203-321287-107A 12&11 or or 2203-321287-1068 11&12 2203-32/287-1078 2203-321287-1078 12&11

• 125VOC is applied to terminal 13 of relays When positive 125VDC 2203-321287-107A, and 2203-32/287-106A, 2203-321287-1068, 2203-321287-107A.

2203-321287-1078 the relays actuate to close the contacts listed above.

• Control Cable 33934, a 12-conductor 14 AWG,cable, has conductors connected to terminal 13 of the above listed relays. One of the conductors in cable 33934 has positive

              ,125 VDC
              .125  VOC applied from panel 903-32, terminal E£-21    EE-21 (Drawing 12E-3462 SH 2 refers).

,( ,( *

• Since Control Cable 33934 ;s is installed downstream of the Auto 8lowdown Inhibit Swjtch 903-31287-304 contacts, auto blowdown inhibit may be bypassed by fire induced hot shorts in control cable 33934.
• Control Cable 33674, a multiconductor cable, has conductors c~nnected cunnected to 2203-321287-1078 2203-32/287-1078 contact 11 via 2203-32 terminal 88-50 2203- 321287-1078 contact 5 vii a 2203-32 term; B8-50 and 2203-321287-1078 termi na 1 BB-88- ,19, 09, One of the following spurious operations may occur from fire induced failure of Control Cables 33674 and 33934:

Hot shorting one conductor to the positive 125 VOC conductor in cable 33674 may cause either valve 203-30 or 01' valve 203-3E to spuriously open. Hot shorting two conductors to the positive 125 VDC VOC conductor in cable 3367~ may cause both valve 203-30 and valve 203-3E to spuriously open. Hot shorting two conductors to the positive 125 VDC VOC conductor in cable 33934 may cause valves 203-3A. 203- 3A. 8, 8. C. D, 0 and E to spuriously open. The licensee's analysis indicated that the resolution for Control Cable 33674 discrepancies was: IITarget "Target Rock (manual function) ann sa f ety valves are ava safety il ab 1e for RPV pres available sure control pressure control".II. The sstated ta ted resolution does not demonstrate protection for simultaneous spurious opening of valve 203-30 and valve 203-3E. (

 "'.                                         22 111.7-26 III.7-26

Revision 8 April 1992 ( The licensee's analysis also indicated that the resolution for Control Cable 33934 discrepancies was:

         "While in hot shutdown, it is necessary to prevent the electromatic relief valves from spuriously opening to preserve the reactor vessel coolant inventory. For fires external to the main control room, an AUTO SLOWDOWN BLOWDOWN INHIBIT switch at the MCB will prevent spurious blowdown. If the fire is in the MCB.

it may be necessary to trip all of the power feeds to the blowdown logic. This is covered by procedures. Excessive reactor pressure will be controlled by the mechanically actuated target rock. rock or safety valves. valves."II This r~solutian r~solution does not appear correct since fire induced failure5 cable5 33674 and 33934 may bypass and defeat the function of of cables the Auto Slowdown Blowdown Inhibit Switch at the Main Control Board. The licensee's resolution to trip all of the power feeds to the . 1ow~own log; bblowdown logicc be; ng covered by p"rocedures being procedures ;iss correct; however. however, the lmplemented implemented procedures were developed to provide protection for the spurious opening of either the Target Rock. Rock Valve or one of the Electromatic Relief Valves. Since simultaneous spurious opening of the Target Rock Valve and all :of the Electromatic Relief Valves has ha5 not been analyzed, procedural protection has not been demonstrated. The simultaneous spurious opening of the Target Rock Valve and Electrnmatic Relief Valves has a tremendous impact on reactor coolant inventory based on the limited capacity of the CRD Hydraulic c System to restore or maintain reactor coolant inventory. Due to th~ significance of this issue and its generic implications, the spurious operation of the Target Rock Valve and Electromatic Relief Valves has been referred to NRR. This is considered an Unresolved Item (237/88010-03(DRS); (237!88010-03(DRS); 249/88012-03(ORS)} 249!88012-03(DRS>> pending resolution from NRR. On August 15, 1988, the inspector met with the licensee to discU55 discu5s appropriate fire protection features and measures to prevent or mitigate consequences or spurious operation of the Target Rock. Rock Va1ve Valve and Electromatic Relief Valves. In addition, the inspector walked down the areas of concern. As a result of the discussions and wa1k walk conduclect down of the areas on August 15, 1988, a conference call was conductect on August 17, 1988, between Dresden, Quad Cities, CECo Licensing and Region III to discuss the fire protection features and compensatory

  .,asures leasures that would be taken to prevent or mitigate the consequence of a disabling fire from causing a spurious operation of the Target  Targel Rock Valve and Electromatic Relief valves.

Attachment D of a September 16. 16, 1988 letter from J. Silady, CECo, to T. Murley, NRC, summarized the conference call of August 17, 1988. In this letter, the licensee indicated that in all area~ areas through which the subject cables are routed, there are automatic suppre~sion suppression and detection systems, except for the ~ezzanine mezzanine floor of the Dresden Unit 3 Reactor Building which only has a detection detectIon system. The following Interim Compensatory Measures were implemented implemenled for the affected areas of the reactor building mezzanine floor: 23 IH,7-27 IU.7-27

Revision 8 Apri1 1992 April c(

• Declared area combustible free fire zone Additional portable fire fighting equipment was brought into the area
• A combustible loading leading inspection was conducted per shift basis by station operators

                                                .operators The licensee indicated that the above actions would commence   cemmence immediately and be in effect until this issue is resolved.

g.

      .9* Availability of Opposite Unit Safe Shutdown Equipment Eguipment During Refuel1ng Outages The licensee has selected two primary systems for achieving hot shutdown in the event of a disabling fire concurrent with a loss 01            .01 effsite power.

offsite pewer. The systems are the isolation iselation condenser system and the HPCI system. As previously mentioned, five different Appendix R hot shutdown paths per unit are identified in the licensee's safe shutdown methodology. methodelogy. FourFeur of the paths per unit utilize the respective respec";ve unitls unit's isolation condenser, and differ only .only in that they employ different power pewer trains, diesel generators, CRD CRO pumps, and/or operating methods. The fifth path per unit is the HPCI/LPCI method methed of shutdown. shutdewn. The Dresden safe shutdown procedures utilize safe shutdown equipment frem the unaffected unit during certain fire scenarios. Included in from this equipment are condensate transfer pumps, CRO CRD pumps, service water bus ses, and 480V Mecs. pumps, 4KV and 480V busses, HCCs. It was identified i dent ifi ed by the

c. inspectors that the licensee had no administrative controls to inSIII'f>

that the required opposite unit equipment was available .or compensatory measures would be in place during a refueling .outage insure that at least one train of safe shutdown equipment was or that inslll'~ putage tt,*.. available in the event of a disabling fire in the .operatingoperating unit. The failure of the licensee to establish procedures or controls to ensure that required alternative shutdown shutdewn equipment was avai'lable for safe shutdown of the operating unit when the opposite unit (which houses the alternative shutdown equipment) was in an outage or shutdown and the required alternate shutdown equipment was removed from frem service for fer scheduled maintenance or .or repair is censidered a violation considered vielatien of.of Appendix R tote 10 CFR 50 (237/88010-05(DR~); (237/88010-05(OR~); 249/88012-05(ORS)} 249/88012-05(ORS>> as described ;n in the Notice Netice of

                                                                .of Violation.

Vielatien. At the request of .of the inspectors, inspecters, the licensee developed develeped draft adIIinistrative controls adlRinistrative centrols forfer safe shutdown equipment during refuelinq outages (letter dated April 21, 1988, from

          .outages                                      frem E. O. Eenigenburg, CECa, CECe, to J. Holmes, Helmes, NRC).

The draft administrative procedure precedure has been forwarded ferwarded to te NRR for review. C.' 24 III.7-28

Revision 8 April 1992

h. Fire Protection of Safe Shutdown Capability In the 1licensee's i censee' s safe shutdown report, the 1licensee i censee has identified i dent if i ecf several safe shutdown pathways in whiChwhich at least one pathway per unit will be available in the event of a disabling fire in either Unit 2 or Unit 3. The inspectors toured both units and observed fire walls and suppression and detection systems which appear~d to be well designed and installed as described in the Safp Saf~

Shutdown Report. There were no identified discrepancies, however. however, a concern has been identified regarding the adequacy of separation of Unit 1 from Unit 213 (Se~ Unresolved Item (237/88010~Ol(DRS); 2/3 (See (237/88010-01(DRS); 249/S8012-01(DRS>> 249/88012-01(DRS>> of this thlS report).

4. Fire Protection Features As part of the Appendix R compliance assessment, several fire protection features were also reviewed as listed below:

• Carbon Dioxide Systems
• Control of Combustibles

a. Carbon Dioxide System The licensee has provided total flooding carbon dioxide ((C0 C0 2) 2)

suppression systems for the AEER, three diesel generators and the Diesel Tank Rooms. (( .. , The inspector requested the original CO 2 concentration test results for the diesel generator rooms. The licensee indicated that the original tests were not available however CO 2 concentration tests were planned to be conducted by the end of the Unit 2 refueling outage. At the request of the inspector, the licensee performed puff t~sts tlsts on Diesel Generator No.3. As a result of the test, the licensee was informed of the following inspector observations: (1) Two employees entered the testing.area immediately after the CO 2 discharge test. Measures should be provided to ensure that personnel will not enter the test area until the appropriate personnel have tested the area to ensure it is safe to enter. enter'. (2) Procedures should inform test personnel of specific fire dampers and other equipment that are expected to function during the performance of the test. (3) During the CO 2 puff test, Damper 3-5772~102 3-5772-102 failed to close. (4) (4 ) The predischarge alarm for the diesel room CO 2 system is an audible alarm. There;s There is no visual alarm. The licensee wa~ requested to verify that the audible alarm is sufficient to warn personnel that may be ;n in the area with the diesel op~rating. operating. This is considered an Open Item (237/88010-05(DRS); 249/88012-05(DRS>> pending review of the licensee's action~. 249/88012-0S(DRS>> actions. c 25 III.7-29 111.7-29

Revision 8 April 1992 ( As a result of the CO 2 auxiliary equipment failure the licensee initiated a work request for HVAC Damper 3-5772-102. The licen~ee licen,ee acknowledged the inspector's concerns previously identiti~didentitied and indicated that actions regarding these concerns would be tentatively completed by November 3D, 30, 1988.

b. Control of Combustibles 50.48(a) requires that each operating nuclear power plant 10 CFR 50.4B(a) have a fire protection plan that satisfies Criterion 3 of Appendix A to 10 CFR Part 50. It further requires that the plan describe specific features necessary to implement the program such as tontrols to limit fire damage to structures, system,.

administrative controls system~. or components important to safety so that the capability to safely shutd~wn the plant is ensured. The licensee satisfied Criterion 33 by meeting the applicable requirements of Sections III.G, III.J, and IrI.O 111.0 of Appendix R and by meeting the fire protection requirements identified in the guidelines of Appendix A to Branch Technical Position B.T.P APCSB 9.5-1 as reflected in the staff fire protection safety evaluation issued prior to the effective date of the Appendix R rule. In Section B.2 of the licensee's re~ponse to the guidelines of Appendix A to APCSB 9.5-1 (Amendment 2-2/86) the licensee indicated that effective administrative measures have been implemented to prohibit bulk storage of combustible materials inside or adjacent to safety-related buildings or systems during operation or 'C maintenance periods. i censee has implemented The 1licensee imp 1emented Administrative Admi ni strat i ve Procedure "Control "Contro 1 of Transient Combustibles, Storage Areas and No Smoking Areas." The procedure establishes guidelines for storage and handling of transient combustible materials in certain areas of the plant which contain safety-related components and/or equipment . important to safe shutdown. These plant areas are identified belo~: below:

• Unit 2/3 Reactor Building
• Unit 2/3 Turbine Building
• U~it 2/3 Cribhouse
• Unit 1 Cribhouse The procedure is provided with a fire loading of COlllllon CO/llllon matedal mate";al chart that establishes a low, medium or high fire load based on thp the amount and type of material (such as wooden scaffolding, Class 1 01' Or 2 combustible liquids, etc).

The procedure indicates that for low transient fire loads, addilioll.,1 addiliol,.,1 fire protection equipment or a work permit is not required. Howevpr. the work area should be kept clean and all materials relllovedremoved a~ suon SUlIn as practicable. 26 III. 7-30 III.7-30

Revision 8 April 1992 ( For m~dium transient fire loads, the procedure r~quires requires the complellon* compl~llun* of ita transient combustible pennit (DAP Fonn 3-3A) and the placem~lIt (OAP Form p1acem~lIl of a supplemental fire extinguisher at the jobsite. The permit pennit is to tu be completed by the responsible work group supervisor or foreman foremdn ctlld dlld posted in th~ post~d the work area. One copy should be sent to the Fire Marshdl. The procedure further indicates that the Fire Marshall may specify Th~ requirements upon receipt of the pennit. additional require~nts For high transient fire loads the responsible work group supervisor must obtain th~the approval of the Fire Marshal on the permit pennit before posting it in the work area. At this time, the Fire Marshal will requirements on the permit specify appropriate requir~nts pennit such as supplemental suppl~ntal extinguishers, hoses or fire watch~s. fire extinguish~rs, watches. The procedure further specifies that during major outages, transient combustibles shall be controlled in accordance with th~ the c~nbustibl~ c~nbustible control procedure except that the accumulation of transient pennitted provided the accumulation of transient combustibles is permitted combustibles does not exceed that which could be removed by th~ the nonnal shift. end of the next normal On April l~.Ie, 1988. when Unit 2 was operating and Unit 3 was in an outage, an NRC inspector identified that approximately twenty 55-g~1~on drums of lubricating (lube) oil was stored in a 55-garon safety-related area on Elevation 517' - 66"11 (southwest corner - 1.1.2.2) of the Unit 2 Reactor Building. The licensee had Fir~ Area 1.1.2.2} { previously requested an exemption from Section 111.G.3 III.G.3 of Appendix R that required fixed fire suppression system in this area. The licensee indicated to the inspector that on March 31. 1988. the lube oil was transferred to the Unit 2 side for approximately discovered as a problem and transferred to 13 days before it was discov~red the Unit 3 trackway. The inspector requested the transient pennit for the lube oil observed at the Unit 2 R~actor combustible permit Elevation 517'-6". Southwest corner. The licensee was Building. Elt!vation unable to provide the inspector with the transient combustible p~rillit permit (short termtenn document) for the specified storage of the lUbe oil in the Unit 2 Reactor Building. However, 0;1 However. according to the infonned of the transfer of lube oil to the Marshal. he was informed Fire Marshal, Unit 2 side and concluded that the temporary storage of lub~ lube oil was acceptable and no additional fire protection features were required based on the following: (I) (1) Low traffic area ((22)) Fire detection was available (3) Lube oil flash point characteristics were such thdt that it was difficult to ignite (4) Safe shutdown could have been achieved in the event of a sab 1ing fire di sab' fi re util izing i zi ng the equipment equi pment for safe shutdown Path B-1 27 III.7-31 t

                                                                                     * .~. -----------.._.

Revision 8 April 1992 c ( The 1licensee establ ished a formal transient combustible proc~~ur~ icensee had established however it did not appear that the transient combustible permit wa~ proc@dure was effective for preventing the storage of lubricating 0;1 oil in the Unit 2 Reactor Building. The failure of the licensee to meet th~ the requi rements of their requirements thei r approved fire fi re protection protect i on prograll program by permittinq permi tt i n9 the storage of twenty 55-gallon drums of lube oil in the safety-related area is considered a violation (237/88010-06(ORS); (237/88010-06(DRS); 249/88012-06(DRS>> as described in the Notice of Violation. 249/88012-06(ORS>>)

5. Open Items Open items are matters that have been discussed with the licensee. that will wi 11 be reviewed revi ewed further by the inspector, and that involvei nvo 1ve some action act i on on the part of the NRC, the licensee, or both. Open items disclosed during the inspection are discussed in Paragraphs 3.e and 4.a.

6. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, or items of noncompliance, or deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 2.e, 3.e, 3.f, 3.g, 4.a and 4.b.

7. Exit Meeting The inspector met with licensee representatives on October 18 and December 13, 1988. The licensee indicated the likely content of this

( report and the information discussed during the inspection was not considered proprietary in nature.

ie\ i( 28 III.7-32 III. 7-32

Revision 8 April 1992 c( N=" t~4;C ~o~en~i:'

                 ~.D8en~i:'    ~. .

3LI,t.~e:,: 3ub.~ e: ":

                                      ~.U=lt
                                      ~.U:::lt ;~'-,e;tl0:1S    Hp~il
                                                           ":':::Led Bv

. ~

;S~. a.: lS:3S::9 h==~::,nsl:.12 ~:-Igr~

                                                                                                  ~~ "~1OI!f f?R.s"~

E;e_el.~

                                                                                                  ~e_e 1 t i!:'t1

_S"auT~ Comt Date Iiate 1 Operation of is~ sWlt=hes NA tour/demo oi action and fuse puiling 2 P~ovide of fire area! ,JH x NA pe. th dr-awi ngs 3 Why oil barrel without JH x NA permlt 4 Inventorv for DG :/3 NA SSD £lax S~rinkler oriEn~stian JH x NA 6 Inadvertent CC: d~s- x NA cha.rg: 7 Maintence procedures RW x NA and schedules for bkrs C' a

            ~nd relay:

Csic far re~ct~r water RH NA G Sign-off for ~PIP RH NA

O(j-2:)

10 55D ecuip ODS centrel: on Ltnaffectej ~tni-:: 11 Ccmmer x NA JH 1:: Cont~ol of fuse repl aCetTlEmt x KP x 1= Provide fire :top detail x NA 16 High/lo~ int9rfac~ ~a~ie F:E3 NA rQuti~g and c=n~rc! oree 17  :=:EB~ FW<=. x NA i

 \..                                                              *]H III. 7-33 III.7-33

Revision 8 April 1992 ( C~Eition5 c.'_'E::tion:: Aprli  ::~ 1'138 at 1938 ~5:38:~~

                                                                                                                          -~: 15:38:':::;-

NC'.  ::_'.:

                                      ;: -.'.: -'.1 ec     '?': t                         A5Ked?~

ASr,:ed F .., Re~p~n5ible

                                                                                                          ~:==p:.n=i b1 e =:::!ors  ~~gr5   ~esGlved i:::escl ved Comt DCI.te Date 19          H~.tchw.:y H.o:.tchw:y                     dr.~.it dr.~.tt             cLtrtai r, CLlr-tair,              JH              Cf<, MM CK,                                x X           lilA NA

22()1) Penetr-C'<.h 0', seal ial:.el::

Penetrc!<.1:.1o:-! label: JH BMB X X NA 21 Fir-e Fire PlI.mp Pump Tests TEst:: JH MD

     ,.,:-a
     ,.,.~
     -~          Light :.:;

Ligi;t ~l ::,*:k~ C;*:k5- detectc!"'sdetectc~= JH

                                                                                             .jH             :;8.
                                                                                                             ~:B.       E-~j SW~~    F:L
                                                                                                                                ~:L             X X           NA

3

     ..,:.,:"    Emerge1"'!:v 1I iight E:merger:=v                                    ght tests tEsts                 JH              1-1[,

1-\[' 24 CO2 Pl_!.":';  ;:u":';'  :.e=~

.==~ JH Mr, MD

     .-,c-2S
     -~

r"". _ ": f"""._

"'QJ.

1...Q.i.

                         ~

hj, =-:,:*ri h~  :::'*:ri E: E~ on brea.;.!?~-~ bre2.~:e~-5 KP f:::F  ;..11 h... ,

                                                                                                                      ~ ES F:S                     X           NA
                 ~.;:d Ci.;-:d         re:2~;s rei2fs 2t 2t::        t'-:2.i ro.: E-'

r<;';'.1"',: _l. ~- -'"'..':" =

.: '.1. llance

i 1 lance F'-e= <F' f<F' 0.;"

'

1..::' X r,A r'~A br-eake~-= brEakE~-5 27 87-50 concer-ns concerns on te~t- tE:t- KP KF' JFK~ C!'lL C~lL X X NA (:~: C able crie::k crie*:k valves 2~ 2S F'i Fire re t~.~-rl b=.~-rler er sw-yi s,w*vi el- JH r*D. M:~'\ ['K CK lj ';'riC.~

                      ':-'rl-:5 2;'

2;~ Pene:r a1: Penetr 2.:'1 ,:*n seal ll:*n 5231 JH r'~D. r'm~ U.: D:': 30 Ave,i Ave-i 1l.:,t,i 2.t,i 1 it..,: i ty of i2S-Vdc:: i2S-\/dc KP t:::F' CEF:~ FWF. FUIi.F:l'.i. X 5i3/Sa 5/3/88 Bf'lB Bt'1B 31 :iSO-v!:: 250-\/:: b.~ttEr-y b.;;t't.er\1 cn5.t-.~ E:-:: Ch5.t-i~E:-= KF CER X NA

       ~~
      .~~..:..
      ::  Functi :'1.

FLmcti cr. of device c1n c.n JH BF~1B Br'13 X NA side of 07 SW pumps-oil wa.ter :liea=Llr :1"iea::urement ement

       ..:.'.~.
       .";._.e           eIT,; t i ::1 requEst E:{Eiq:.

E:{ request for JH GMi-eM!... x X NA filA cribho~!.se cribhoL.;*E-E fin: fi t~e detec-:iQl1 detec-:i8i'i 34 Effect :If c.f w~.ter water sprayspra',,; JH CER X X NA elE::rical cables on elE:'::rical

       ~~
        ""!:~
       .~'.J
       "_'...J     Spur i CL:E- operation Spur-ic~~=                        Dperati en I:*f        cf                KP KF'              Bt-1B, ;EB, BI'1B,      ~EB, CC;RCE:R         :~

X. NA valvE 11,:~(j:-2:: vaivE  ;:',(11-2:: high/ high.! low Dre~Eu:-e nre:=Eure lnte("re,ce lntef"Te.ce C.C  :::6 36 E2' . .**1 e~*i ':'T Eevlo?i-i' c,*: ?= :: or-e*:e*::= [:11-

                                            ':T !i:c*d      !i:cd   rE=\tl~W reVl~W                    tG:'

r~:~' 2M3 X NA III. 7 -34

Revision 8 April 1992 ( NF\C Ap~endl:; NhC Ap2endl:; ~ ;';.1::: Hcr-l! :i~ ;.988 OLIE;tlon: Ht:il"l! h;J'::": Gi:"IEstlCn: 1.988 ~t 1~::9: 1:-:?9: 1:

        ~C.*
     ..~'k             Su:' J eo: t Subject                              ASf-:ed S,,'

ASf:ed F(o?sponslb:i.e ::ngr: S.,,* ;;:e:;ponslbJ.e ~ngr= F:e5c*lved F:esolved :omt Somt Date 37 Exemption E:{empt i on reQuest r-E~Llest ~~r.; ='~ CML x NA DG 2/3 fusefl..5e repls~ement repl acement 38 Commo'! Commo~ enclosure CEF:, FWF xx NA 39 3~' Teledyne Tel edyne info i nf~ on vesrly VE5.r 1 y JH BMB dischs.rge di schi'.rge test: 40 4(, Upgrade Upgr-,sde DFPP 4153-: to te, JH B~lB BNB include mfgr instructions on c:i s;:: 1le-'-'el cis:: e**-'el 41 Proviie Provi~e la:;t last mcntMly ffionthly inEpec-inspec- J~ BMB x NA 42 ERV-high/low lrterfsce r:. F JFi<~ JFK~ CML Ct"lL x 43 Pump curve CI..\l""ve for Unit 1 fire tire Jh 44 Provide copy of transient JH MD c.( ~5 combustible permit for

              ~nit 3 HPCI 011 dr~rrE unplug Emergen;:-y 1.mp!ug drJrrs Emergency ;...:qr-,ts
                                            ~:Qht5 xX         :'\.:. .
                                                                                                                               ....,~

47 :urrent tran5~~rrn=~=

                                                                                 ...,c.;-,                        X             .....,
                                                                                                                                .r!

r 48 Wrapping tr~i3 CEF; X [<A ['~A

                                                                                                                               .;1'.
                                                               ;{                                                  X           ~*J4
                                                                                                                               .-~-:

wi th V1 O!1 I

                     .Dl vi si OI'"l

,,( c. II 1.7 -35 III.7-35

Revision 8 April 1992 ( DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22~1988 22,1988 DATE: ~ -/9-

              .f -/1- 8"8' gg-co. / AE PERSON QUESTIONED:

CE CO./AE QUESTI ONED: ___ _ _t;_- ___._I<?_u_~~~~~f;~~~/_-S_~ C ___h_~_f'_ _t2._u._+f....:....:..f.d---.:'F,:.-:..h-.:/-.:-S~c.~It:.-.e_f'____ __ COGNII ZANT NRC PERSON, COGN PERSON _ _ _ J<_*-:.....~_l::<_fl_k_(_-n_S_O_Yl

                                              }(_' _ff._l::(_f't_k_,_n_s_O_Yl                 ________

LOCATION/AREA WHERE QUEST IION II-/J I t _ ON WAS ASKED: _...:..rr....:.-..;k.....;...().._f C ro_r_r....;.!J.Mt~-=V=-::. _....<...:...;.:/t..:...(J._lTO'-.-r_1

                                                                                                                 /.                       )

lUA{..:..:.;.>~V....::t.{).:;.S..:...t-l.{It...:.AJ......:'((._c._Q.J7U.t-=--~*>7"~_ tt1::.,:>:...:.t--s...:{A/......:.-f2.:..C_~-=--...,.>~=---_ QUESTION/ITEM' .L {I - r ..1- - / DISCUSSED: (){)e-ro.r-rrn1. Or 15610.. nO;!! SWlftlte5 d-

      ~                  I c

PERSONNEL INVOLVED IN RESOLUTION: . FOLLOW-UP ACTION IF ANY: _ _ _ _~_~ _ _ __ _ _ _ _ _ _ __

• c

.C.:. III.7-36 RR30

Revision 8 April 1992 c (' DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22~1988 22,1988 CE CO. / AE PERSON F'ERSON QUEST QUESTII ONED; ONED: .-i~~,.::M~fhJ'-£L;?fi)~~

                                                                            . ,~p,. 4.r. ....:.;WtJ::..L;L.:luaJ-..:::;.....,...

M ________ PERSON~\lL...l.(~~~(..a.'A..Ll~.:;,c:~;.....,... COGNI ZANT NRC PERSON ....... COGNIZANT ...J4"'-""t.5::-. ___________

                                                      ,--w!1;....p.&/,I'C                 _ _ _ _ _ _ _ _ _ _ ___

LOCATION/AREA WHERE , QUESTION WAS ASKED: --"U,-,' ~U:.....::Z~~~~;....J~(,-7+- 2."----'-'yt"'--~;..J('-7+-_ _ _ _ _ _ _ _ _ __ QUESTION/ITEM" DISCUSSED:

             £faf<       /1~

j

                             =", Z;-::= .9Wk / _

I 71lItv.[IJ. <; .( ,(

 ~-,. ----

FOLLOW-UP IONS: _.....:..~--::.....-_C)_0_\ ACTIONS: ACT \-<<,C)'-.)\O'::' O-=--=~=--_~-:.O_-l.t1--=c)_L.....;/):..-..;8~---.S..L--

                                                                    'TO tJc)l..{'lS                                                S _ _ _ _ __

PERSONNEL F'ERSONNEL INVOLVED IN RESOLUTION; RESOLUTION: FOLLOW-UP ACTION IF ANY: ____________________________ ( RR30 III. 7-37 I11.7-37

                                                                                              ~
                                                                                              ~

Revision 8 April 1992 c

 ~-     DRESDEN STATION APPENDIX J            APRIL 18 uRn AUDIT QUESTIONNAIRE FORM URn 22,1988 DATE:     l(t {Vh CE CO./AE CO. /AE PERSON    QUESTIONED:_M~(~;:;,.;1_L....;;.fA--l'IJ~

QUESTI ON ED : --'..111-,-,-,(~,-,I...;..L..:;.~-!.w-=-_ __ __ COGNIZANT COGN I ZANT NRC PERSON_1 ............I..t.frJ~LM-..;...~ PERSON_1"",-,-, .J...(,fv",--,()I--=-~_ _ _ _ _~ __ _ _ __ LOCATION/AREA WHERE ~ QUESTION QUEST I ON WAS ASKED; _f \-.1 ASKED: _'Y""':~

                                              ~1 (})A
                                      ' -'-'l119
                                                          ~ lI        tJ kO<..-.....K..;::O-O",.,t!I--_

(( Uv _ _ _ _ _ _ __

                                --~~=-~~~------------------------

QUESTION/ITEM i.' .f

                      ~~4_~~_~~L-~~~~~~~~~~~

DISCUSSED: __~W~~~~~~LUL-~~~~~~~~~ _ _-__ { FOLLOW-UP FOLLDW-UP ACTIONS: ACTIONS: ____

        -   'r-
                      ~~=T___________~~~~c-~~~\l-o-~----__~~--~~--

____~~~~~~~--~~~~~G~.~~'I~o_~~~~~~~~~..I...-

            'r-                                 'Q                 ~     ~ ...... 00t"\.

PERSONNEL INVOLVED IN RESOLUTION: n.0ILLJO n.0/LLJO D 10 FOLLOW-UP ACTION ACT! ON II F ANY: ______________________

• c(.~-

RR30 IU.7-38 III.7-38

Revision Revi si on 8 April 1992 ( DRESDEN STATION AF'PENDI AF'PENDIXX "R","R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22~1988 22,1988 DATE~~ -i..firl(-+~-'~~- DATE -i-llrl",,+i.'N=:.-- CE CO. / AE PERSON QUEST! QUESTI ONED: ONED ~ -...:~~I ..&.~-ittl!/~~--------

                                                       -..:~~I..Jl.~..:..ittJl/~oI---------_

COGN COGNII Z ANT NRC ZANT PERSON_--+=i:....l(_~-.:~~r5_oJ_-"'-/--:K~0_0-=-...:...~ PERSON_~f-,-,(~~...;.;~->--~_oJ_~/....:K~0_0:.=..-,-j--:>_ LOCATION/AREA WHERE QUESTION WAS ASKED: _ J/?ff J/

                                                ~
                                            ' ;,~~~-
                                              ;,"-"'~~_-

{(" PERSONNEL INVOLVED IN RESOLUTION~ RESOLUTION: FO~ACTION

          ~ ~ .A ¥ IF ANY: /J E: :!::f":t;
                                                -#-u___ ~~~;._
                                                                    ?       E ~
 ,(
   '-~-

RR30 III.7-39

(5) Revision 8 April 1992 (  !! DRESDEN STATION APPENDIX uRn

                 ~           APRIL 18 - 22,1988 "R" AUDIT QUESTIONNAIRE FORM DATE:     ti I~lit I{

I~ rJ CE CO. /AE PERSON QUESTIONED: QUESTI ONED: _....:;"_t_~_~...;,,.'fU# _-,~c.;.I_~...:.;~...:.;'!fI$ COGNI ZANT NRC PERSON _ _1 _, .,.;...110&_,_,_1'_M_ts

                                                           ..;..f/o&_.           __________

1_65__________ _ LOCATION/AREA WHERE fJP. f}f). 1.1 1).1 HAn HdTJ QUEST QUEST II ON ON WASWAS ---'d'-'-J._...o..xY"--.:.:I1I'-'~'__.:.J.I~--------- ASKED: ---..;d;;...:;;J._...;.K:.LY ASKED: P I--.....;J1I:..:....4._....lI=I:IIIf_ _ _ _ _ _ _ _ __ I QUESTrON/ITEM~~} QUESTION/ITEM .~ I.I. ~ DISCUSSED:,~~~~~;~~~~~~I~~~~~~~~~E~~~ DISCUSSED:~ I ~=l {sL

                   -L          (It                                            -       !5:t!f~

Ie RESOLUTION:_~~------------------ PERSONNEL IINVOLVED F'ERSONNEL NVOLVED IN RESOLUTION: RESOLUTI ON: FOLLOW-UP ACT ACTION F ANY: _ _ _ _ _ _ _ _ _ _ _ _ _"""'--_ ION IIF ~ __ ( II 1.7 1. 7-40 RR30

(~ Revision 8

• April 1992

( (, DRESDEN STATION APPENDIX "RII "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE~ i."{~9 DATE, 1/(9(99 CE CO./AE QUESTIONED:_~~,~ll_A_r_[_O ;V___________________ co. / AE PERSON QUESTI ONED, ..,:.1.4--,-,""7)o...A_r_L-_O_N __ __________ COGN COGN Ir ZANT ZANT NRC NRC PERSON-----l(a1lf~.&O--+-Pfa~u....((-( M_~. ;: dI'-'-!- _ _ _ _ _ _ __ PERSON~mm>_=J..;I.1l.-....Ptn+W'....((-.:.(-M....;5cr1""-'--------- LOCATION/AREA WHERE l~ ~,h/

                                                 ~,~/

QUESTION QUESTION WAS WAS ASKED:~~~~~r~_~~~~~ ASKED:~~~~~r~_~~eL~~-------------------- ________________________ QUESTION/ITEM DISCUSSED:~~~~~~=-~~~~=-~~~~~~~~~~~ DISCUSSED:~~~~~~~~~~~~~~~~~~~~~~~~ c.CJ Co ( FOLLOW-UP . ACTIONS: /~

      --lil7i:., a*~ ~ a, Iff?

S,.,c.. 0..<0 d j"4' t F CO?_ ,4p . /

                                                                                     )

RESOLUTION: Q,K Q.K - 1.(-/7-;8'(163 y-/'J-/?l?' (hG PERSONNEL INVOLVED IN RESOLUTION: FOLLOW-UP ACTION IF ANY: ________________________________ ( III,7-41 III.7-41 RR30

Revision 8 April 1992 ('-- DRESDEN STATION APPENDIX "R" ~R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE: 411t/ f%'" CE CO./AE QUESTI ONED: ~C.~.--,-R.;..JIUy,--+_~ CO. / AE PERSON QUESTIONED: +_t".-'--________

                                                                         --=C;;.;,-i".;R.;;ouv.... _________ _
                                              ~}(~!~~~A=r.k~r~~~f~&rl~____________________

COGNIZANT NRC PERSON __~f(~,~~~a~r.k~/~*~~f~~~ ______________________ __ LOCATION/AREA WHERE QUESTION QUEST I ON WAS ASt(ED: ~arlw,..,... ASKED: --,-A.:.:l'-.;::.::J::.;,..!.~ Ar.;..J, """=-:..;:'tM.,..='--_ _ _ _ _ _ _ _ _ _ _ __

                                      --~~~~~----------------------------

.(

(----

FOLLOW-UP ACT IONS: ACTIONS: .J

                     .JVf'Z-ft-1 I/f't..eit--I RESOLUTION:      ~ '=/9~??

PERSONNEL INVOLVED IN RESOLUTION; RESOLUTION: FOLLOW-UP ACTION IF ANY: __________________________________ I. 7-42 III.7-42 II RR30

Revision 8 April 1992 ( AF'PENDIX "R" DRESDEN STATION AF'F'ENDIX uRn AUDIT QUESTIONNAIRE FORM AF'RIL 18 - 22,1988 APRIL DATE: OI..f-le'-C3g QUE ST IIONED CE CO. / AE F'ERSON QUEST ONED 1: _ _ 03~G3~:.looc:+d\"'~+-.:..-l.,~______ _--=G3~glL..>.oi~s:.v=S"'...c;:+..L1.,.!.- ________ F'ERSON_~R~~~~~>>~~ COGNIZANT NRC PERSON _______________________ __~Rw-~~~>>~~ _ _ _ _ _ _ _ _ _ _ __ LOCATION/AREA WHERE QUESTION WAS ASKED: ~ ~ (L., '" - QUESTION/ITEM' _ r'\ _ Ii -,t""_,_ . . . . i DISCUSSED: eMAil

                      ~~

n =L~!s =t.c-t- =+/-O~:-Zrc. ~~ ~

• c RESOLUT ION: P QU"":>l 0 IS 0 0 tv V 0\ U r; ~ (5 AOa> I<. ez: (

F'ERSONNEL INVOLVED PERSONNEL IN RESOLUTION: FOLLOW-UF' ACTION IF ANY: ________________________________ _ III, III. 7-43 RR30

Revision 8 April 1992 ( DRESDEN STATION AF'PENDIX APPENDIX "R'" "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22~1989 22,1988 DATE: 'f~ tr-??f' CO. /AE PERSON QUESTIONED:_---"5;>c;.;;;;;......_~ CE CD./AE QUESTIONED I _---'.52>=-_~ __________ COGNIZANT NRC PERSON __~a~~~-==-~_ _ _ _ _ _ _ _ _ _ ___ PERSON_~~~_~~~~ _______________________ LOCATION/AREA WHERE ~ QUEST IbN ION WAS ASKED: ASKED I _ ~.o....;;;;.;;.=..;;'---..;.....-_:-... _.L~L!::==::::":-=:=";"-=- QUESTIONI ITEM DISCUSSED I ----'~".-""""'7I""":::........s1<iHt.::.C-'t:r:z.""'<-- .....£'-'(J'--'-'1P,--,d..:::..o.::..C:::.---,::I.~O_ _ c FOLLOW-UP ACTIONS: 7

                &?990~                     :tF          I?       ~                   .......

PERSONNEL INVOLVED IN RESOLUTION: RESOLUTI ON I ACTI ON IIF FOLLOW-UP ACTION F ANY: ANY 1 ________________________________ Ie III.7-44 RR30

Revision 8 Apr; April1 1992 (( DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE: 4/1g'/t'6 I CE CE CO. /AE PERSON QUESTIONED: __ __ t2_*_~_lI....:f4-.:. e_-_~_l(-,ft;~________ ________ _ COGNI ZANT NRC PERSON_......:...k~/...:..~_t:t:r.......;.....'--.-;.rn=5_(Jv1--'- PERSON_--,-k~,. :. .~. .; ;t:U';~k.: .;, ,-,Yt:;5. . :i1v1-,-,_ -__ LOCATION/AREA WHERE fi J I ON WAS ASKED: -.:.!T~U;..:::;.~.l:..,;(:""* QUEST ION fo~rt......:...iUft1=:J.--_

                                     -!.!T~u"",~,,-,r,-,*+O~r....!/,-,*  Uf!1~.:..J...-_ _ __

i (c. t;k./:

                                                                                                            //of 3 bk-/:

U!3 PERSONNEL INVOLVED IN RESOLUTION: FOLLOW-UP ACTION ACTI ON IF I F ANY: '"DDC~S4f?~V

                                                   --""D::....:D=:..::C:..:~:::*
                                                                              =':==1.....:->;.::-~7--=~:::::!-.:=..:~7LV--

7 III,7-45 III. 7-45 RR30

Revision 8 April 1992 DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1986 22,1988 DATE: ¥-/?-l'S'

                 -¥-If-TV CE CO. / AE PERSON QUEST QUESTII ONED:       _~R~._ItMt:::!..!~~l!!:::!!!.,:!!c

_..JR~'..1::/fM~~~<!!!'::!..!:!.:...!~'! ~~l::....... l!E:. . .____ COGN I Z ANT NRC PERSON CDGN J<..dfi ~, PERSON_...!.J<~4i~*~_-!8~ **!:!/!dL~(~,(;::!!';'~.,_,,- he e< ; , 1>'. _ _ _ _ _ __ LOCATION/AREA WHERE ~

                                     ¥l       I  ~

QUESTION WAS ASKED: r+~..s.J.,.Hcitr

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Revision 8 Apri J 1992 April ( 20,t 1988 April 20

     .1JC LTR:

JJC LTll: 188-019 To: Operations Department Personnel (( -

Subject:

Subjeet: policy Replaeins Blown ru poliey on leplacins ** s ruses A A procedure proeedure will be prepared to formalize the Operations Department policy poliey on replacins blown fuses. That policy poliey i. is as follows:

1. Compare the new fuse to the old fuse to verify that they are "like for like." This conrpariaon eompariaon should include inelude manufacturer, lIaIlufaeturer, physical physieal size, shape, shspe, voltase voltsse and current eurrent ratinss.

ratinss, and fuse type (quick-&ctins, (quick-aetins, slow-blow, etc.). ete.).

2. If ,illesible

                      .illesible or missins markin,s markinss for the old fuse do not permit
                 *a eomplete complete Verification verifieation of volta,e voltase and current eurrent ratinss or fuse type, the Shift Supervisor will obtain verifieation*of verification-of such sueh data from wiring wirins diagrams diasrams and/or and/~ vendor manuals or by consultation eonsultation with the Teehnical Technieal Staff.

c

                                                          .1oe(~otowski Joe ~otowsld Asst. Supt. - Operations
     ,JJC:R.1:rs J1C:RJ:r, ee: J.

cc: .1. Wujcisa Wujeisa Operatin, Operatins Engineers Ensineers c.( 13118 1311a III.7-47 III. 7-47

Revision 8 April 1992 c-( DRESDEN STATION APPENDIX "R" "Rn AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATEr DATEI 4- (2-~ CE COI/AE CO./AE PERSON QUESTIONED: ____ _____~]C~V~c??~~~S3~S>~c~~~~~~-----

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COGNIZANT NRC PERSON _____~~L=f~~~~~~2~fr~¥:~)L~l~~~~~C~\~)~

                                                             ~~C~f~~~~~~?~fr~SZ~~~c~~~5~C~'~)~________              ___________     ___

LOCATION/AREA WHERE QUESTION WAS ASKED: ______~~~~~~~\w~~~£T~~~~~.~ ___________________

                                                              ~~~~~~~\~~=_~fT~~~~~.~________________                                   ___

QUESTION/ITEM QUESTI ON / ITEM DISCUSSED: S\'\Z

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u 13-rc.oo-tS" Dr D o-fli" DE ""t+tS cQUM ~Mbrr:-

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'(.~. _.' FOLLOW-UP ACTIONS: ______ ~\J~~~~~c:~

                                       ~}J~e~U~c:~         ____________________________________

RESOLUTION: ________________________________________________ PERSONNEL INVOLVED IN RESOLUTION: FOLLOW-UP ACTION IF ANY: __________________________________ __ ( IlI.7-48 II 1. 7-48 RR30

                                                                                                        ..... ~

Revision 8 April 1992 c ( DRESDEN STATION APPENDIX uR" APRIL 18 - 22,1988 "R" AUDIT QUESTIONNAIRE FORM CE CO. IAE PERSON QUESTIONED CO./AE QUESTIONED:_-"f'-" I _...... fJW~=-_________

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    -                                              I II. 7 -49 RR30

Revision 8 April 1992 (( DRESDEN STATION APPENDIX "Rn "R" AUDIT QUESTIONNAIRE FORM F'QRM APRIL 18 - 22~1988 22,1988 DATE: ~ 4= -(

                -{ 't - 88' 88 CE CO./AE PERSON QUESTIONED:

COGN COGNII Z ANT NRC PERSON ZANT __ PERSON_-,-JL..._, ~A_'fJ72)~_~_A/-..,;..S_O_IJ IL.-_, ..!..A_'ff72j.~,-~_A/--=-5_0_N LOCATION/AREA WHERE QUESTION WAS ASKED: RESOLUTION:

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    -RR30                                           !IL7-50 III.7-50

Revision 8 April 1992 DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE: _0....~~~r____ CE CO./AE CO. / AE PERSON QUESTIONED: QUESTI ONED: __ ~b?~~~~~~~k~_~~:~/______ _--.l!.!Q,---,4t4!o.LL,8!;t';,f41<!_~<Q,:'/ COGNIZANT NRC PERSON~+(~~~~'a~~~~~~~~~'~4e~e~ PERSON~t(.~~~~'~~~~~~~~~~*1b~e~______________ ___________ LOCATION/AREA WHERE QUEST I ON WAS ASKED: QUESTION A

                              ~      ~

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                      ~5~~~~~~~~a~o~v~~=_________________

RESOLUTION: ____~5~~~~~~Ps~a~o~v~a:~~~ ___________________________ PERSONNEL INVOLVED IN RESOLUTION: '1:- f.- z:. z:.. (J> yc,- (]>ra v.J

                                                            ....... ....J c....." r=-

F- L....I, of Cl f:= (/.J" . . . . te.... C C_e..-rtIL FOLLOW-UP ACTION IF ANY: __________________________________ IIL7-51 III.7-51 RR30

Revision 8 Apri1 April 1992 ((

   '.         DRESDEN STATION APPENDIX lOR"                        "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 CE CO. / AE F'ERSON PERSON QUEST                             tt1. 7)/lli~

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COGNIZANT NRC PERSON __:l~._A1~o~~~~~______________________ PERSON~:s~.~A1~d~~=~~ _________________________ ___ LOCATION/AREA WHERE QUESTION WAS ASKED: "$

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QUESTION/ITEM . DISCUSSED: ~"Zj} ~ ~ 12rSC4!J .:.;g:m,A nllLI(A7?:II~1/

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',C"... FOLLOW-UP ACTIONS:

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RES LUTION: AL J' 11) ~

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fY'I \/:.9 Mowu;~ FOLLOW-UP ACTION IF ANY: __LN~o~N~E*- ~N~o~N~'__________________________ ( III.7-S2 III.7-52 RR3() RR30

Revision 8 April 1992 c DRESDEN STATION APPENDIX lOR" "R" AUDIT QUESTIONNAIRE FORM J APRIL 18 - 22,1988 DATE'~¢fI. DATE:1)/c CO. / AE PERSON QUESTIONED:~~~/~2)~y~~~~ CE CO./AE QUESTI ONED : .... 4"-'-'-,...,,72'""-<.y.... ________________

                                                               'lLt;.:::kO...<--_ _ _ _ _ _ __ __

I ZANT NRC PERSON _ COGNIZANT COGN ___S_*__M~t'.;..IICun~..<S"",5______________

                           ~5_*_~~~~~~.~~1~                      ______________________       _ _

LOCATION/AREA WHERE QUESTION WAS ASKED: UZ Uz 7Zc::elCZl>t?..~~ 7Zc:AqM.-ft.nu~ 'C RESOLUTION: ______________________________________ PERSONNEL INVOLVED IN ~~~ RESOLUTION: ~~ 1L2-~~

                                     \

FOLLOW-UP ACTION IF ANY: ______________________

  -                                               III. 7-53 RR30

__ ,_,__ .(20

                                                                                                                 ,~0 Revision 8 April 1992 c    DRESDEN STATION APPENDIX "R'" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE:

DATE. -¥~

        .,..~

I7

         /AE PERSON QUEST IONED:

CE CO. IAE .~j{~...,I;;>>~'IIldf . .~Ir.W....

                                        ~,d~',",-,>>:..<.£'1..U",~""",-    _ __ __  __ _ ________  _ _

COGNIZANT NRC PERSON-=:S~.~~~~~~~~________________________ LOCATION/AREA WHERE QUESTION WAS ASKED: _z._~",,--==&.....;;..IOIIC.4oR_w~>t¥=t-Z. ~~~ _________ QUESTION/ITEM . DISCUSSED:.Lit;#!" n.:rr4§S W DISCUSSED: i-It;#/ ~ 4/,11)'; 5/J7(#Lj)C)#d0,..(2'~/-z..2.s,)

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RESOLUTION: ______________________________________________ _ PERSONNEL INVOLVED IN ~

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RESOLUTION: ]2. &.4!k:-

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                               ~! 4=e. - Prr ACT! ON IF FOLLOW-UP ACTION      I F ANY: ______________________

c. III.7-54 III.7-S4 RR30

Revision 8 April 1992 ( DRESDEN STATION APPENDIX DRESDEhl uR" AUDIT QUESTIONNAIRE FOt,11 "R" FOr"1 Jj APRIL 18 - 22.1988 DATE: DATE: ~~~ ydyl~ CE CO. I/ AE PERSON DUESQUES T TII ONED; ONED: 4-#"'-L-' J.JII~,_j)<<. v.:...;;..'/._:L....:::~'-'-'--_

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• c PERSONNEL INVOLVED IN ItJ RESOLUTIDN:

RESOLUTION:

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1:...: .~*. oJ _/ _ _/ _ (

                                                                                       ---~--'--.

Revision 8 April 1992 ( DRESDEN STATtON STATION APPENDIX uR" "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE: 1-19-~¥' 1-/1-if' CE CO. IAE

         /AE PERSON QUEST       I ONEO.

QUESTIONED: ££; J.eJ.R Ma+, '~ lY1Q*~<C.. d RO) (01'/0) (0 COGNIZANT COGNI ZANT NRC PERSON K,:f£. r;,~ 1<.,:K 8,....kw.; .. .... LOCATION/AREA WHERE ASKED: QUESTION WAS ASJ<ED: A""(Tit,tc..*

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                                                       /

mu~ J>1u Oat!l.y I QUESTION/ITEM' QUESTI ON / ITEM' DISCUSSED: J ~\ :f.

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UJ-... :s UJtIk.:i.s /4 g] ,t j:;,;,.,... ( FOLLOW-UP .* A . /, ACTIONS: Z~ ~ oALJ ~ RESOLUTION: Wcs~ {?d)3 G<"'~f'~ (?>Q #~ #r PERSONNEL INVOLVED IN RESOLUTION: FOLLOW-UP ACTION IF ANY: _ _ _ _ _ __ __ III.7-56 III .7-56 RR30

Revision 8 April 1992 ( DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 OATE~ DATE: .I{-/9-Pf" 1-t1-Pf" CO./AE CE CO. QUESTIONED: I AE PERSON QUEST! ONED' ' it:t+.~ It "!j LJ..:

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COGNIZANT PERsoN __ ~t<~£4~11~___ COGNI ZANT NRC PERSON_-'-K",.,..£J'-7l ...... ~~~~='~s~~~ _-J7?u~ ...~!::I' SU<b-Ii!:Ioo.____________ It / '...L .. L -I-- n

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DISCUSSED: WtlM.k~ L..J "'" ~Cto' f't1':tH'1

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PERSONNEL INVOLVED IN RESOLUTION: FOLLOW-UP ACTION IF ANY: ________________________________ RR30 1II.7-S7 III. 7-57

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                 '14,;'

APRIL 18 - 22,1988

                                                                                                                                ~     "

DATE: DATE, ---'~,l-b=}.dr'l-"-X"J<...(_ I T-' CE CO./AE PERSON QUESTIONED:~~~~_~ QUESTIONED: __~~~_~~~~~' __ ~ COGNIZANT COGNI ZANT NRC PERSON __ __ ~K~~~~~ ____~____________________ KL.>---,,-R-==~-='-..:.= _________ __ LOCATION/AREA WHERE QUESTION QUEST I ON WAS ASKED: _--<At..,;lE:::"'::::':=:eP.:::::~:::::* WAS ASKED: ___L;At::':5!eP.::::~",,*

                                                             =~::::'                   ==---------"""7"-----------

..:::..::.....;;:.::;*.=;;.:.._ _ _ _ _ _ _ _ _ _ _ _ __

i(- i(

 ~. . FOLLOW-UP ACTIONS: ___          ~Y....JL~_____

_ _~~~~------------- ACTIONS: RESOLUTION: ~ 'f-PLB.trY ~O -rl-+PrT .2S0\l Dc- g"..-n-C HlQ.-&-bl".e..s ~e(} <'\¢T r2.L-<Kl..-,e.I~_Q FO~. srO e p...,cc '£..tp-n:"'a tIS:; (h & . PA@..k:1 t-J.s "",,-

      -----------~------------------------

PERSONNEL INVOLVED IN RESOLUTION: RESOLUTI ON: FOLLOW-UP ACTION IF ANY: __~bJ~~~AJ~C~- tJ oNe,___C C~_~d~'_~4C_L=-£_T~L~O cJ '--fLZ T L 0__ -=-_____

      -RR30                                            IIL7-68 II 1.7 -68

Revision 8 ~'lI

                                                                                                                                           ")'lI April 1992

(/' [, DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DA T E: _-Jyir.....="2-<>-'r/&""<'<"'=f'C--.- CE CO./AE PERSON QUEST IONED: __ \E-"--_s&.A....::..c:z:::lo!:=~/_________ COGN I ZANT NRC PERSON __  ;:t

                                                                         ;}
                                                                    ~W-~.~~~~

bke~ Le,n..v.,,-l

                                                            ----~--~.~~~-----------------------

LOCATION/AREA WHERE  ; I QUEST!I ON WAS ASKED: _.-...-:C~~~:.i::::n:~~"~*~Q....=- QUEST ____________ _--.JC....~~~:::Tl:!~~..~*~*~Q..,::.-_______________ QUESTION/ITEM ~ .. DISCUSeED: . ~ ~ ~ ~ ~ (fJ ....... ,i,

                 .....J-
                                    ~;~y! ~I.v' /J.J.1_~?

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                              ,ret,     4'~!:        ."........

i( ',-- FOLLOW-ACT  : _CL1J~

                                              ,~
                                                     ",..,.~:::t.. ,. . ,~         .... '6, ScJ    -&J~        t';t;:; evf   l' RESOLUTION: _____                                ~         _____________________________

PERSONNEL INVOLVED IN RESOLUTION: RESOLUTI ON: FOLLOW-UP ACT ACT!I ON IF I F ANY: _ _______________________________ ( RR30 II 1.7 -69 III.7-69

Revision 8 April 1992 c DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22~198g 22,1988 DATE: __________ CE CO. //AE QUESTIONED: AE PERSON QUEST I ONED: ...........---"=~=--~....;;.="";;"';:O.::::.:::L~J __---"~~~_=='""'""~I___________ _ _ _ _ _ _ _ _ _ ___ COGNIZANT NRC PERSON ___~~-~~~~~~~

                                *~~-_~~~~~~______________                  ____________________               __

LOCATION/AREA WHERE ,. GD ~ _ A QUESTION WAS AS~(ED: c""""-~~'-"--"=~=--=:::--_____________ ____"""CL~=o...lL.....",'-=="'-- ASKED: ___ _ _ _ _ _ _ _ _ __ _ c/ FOLLOW-UP ACTIONS: PERSONNEL INVOLVED IN RESOLUTION: ACTI ON IF FOLLOW-UP ACTION I F ANY: -!...~--,-tw.;:..=L=-______________________

                                 ~A_tJM.....:....;:~=-   ______________

c: RR30 RR31) IIU-70 II I. 7 -70

Revision 8 April 1992 ( DRESDEN STATION APPENDIX uR" "R" AUDIT QUESTIONNAIRE FORM APRIL AF'RIL 18 - 22,1988 22~1988

                ~~~7~~~
         -'-h1f--'L"""of-46C>o1-~

DATE: ___ 1 7 I. CO./AE CE CO.l AE PERSON ONED, __~!~~lrl3~~~*~~~1~/6~7r/_((:}f QUESTIONED: QUEST! (3 .tb-",,""iti;fJf f / I __-~s; bI ~ ~ ___ S~~~b~/~~'~h~ COGNIZANT NRC COGNIZANT NRC PERSON ____~:s-=-~~~=_~~) PERSON ____ ____________________

                                          ~:s-=-~~~~~ul~------------------

LOCATION/AREA WHERE n .,~  :~ QUESTION WAS ASKED: ASKED, ____..::~=~.~'-_.:..::~=:..._

                                           --:;~~..=..o.~'--_---"=-___________

QUESTION/ITEM DISCUS..S~ cufi::!d!i:i zS ~ C._ - /I ~ .....

                                                                                     -..uaJ
       /(.(~/

c* FOLLOW-UP ACTIONS: PERSONNEL INVOLVED IN RESOLUTION: RESOLUTION, FOLLOW-UP ACTION FOLLOW-UP ACTION IF ANY: ANY' _______________________________

• c III.7-71 RR30

Revision 8 April 1992 ( DRESDEN STATION APPENDIX "Rot, "R'" AUDIT QUESTIONNAIRE FORM

 \

APRIL 18 - 22,1988 DATE: A--{9 - SE' A--{9-S8' CE CO./AE PERSON QUESTIONED: COGN r ZANT NRC PERSON J< t r!lt2 I  !'lfl2;(

                                 ----------------~-----------------

K.!(J./ 1/ft Af 5"c AI LOCA TI ON I AREA WHERE LOCATION/AREA .../1} /'t

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                                     ,1,        /l ,/)""OA Ii '"'0 Pi               _ ~!~
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                                  -'-_K..;..:f.-=--...:. 17_~~~=--_.Lff!1:.JL...::.-.,;t:l:....;.
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• PERSONNEL INVOLVED IN f;:ESOLUTII ON:

Fi:ESOLUT ACTI ON IF ANY: _ FOLLOW-UP ACTION __________________________ RR30 RR3() III. II I. 7-72

Revision 8 April 1992 ( DRESDEN STATION APPENDIX "R" AUDIT AUD IT QUESTIONNAIRE DATE: 1i'O!/ 1.1'017 APRIL 18

                                      - 22~1988 22,1988 QUEST IONNA I RE FORM CE CO. / AE F'ERSON  QUESTII ONED : ---'f"---...:....(-'I=S_J-'--"'<=..;....;~~~,

PERSON OUEST _L[--'.(-,-,:ISo..;J::!-:..:~::::.""CL..,~"""_c...,_£...:d1~U..!-t.....l~

                                                                                      --=c:".,~£~B...w.:.ut.L:J.~__

COGNI ZANT NRC F'ERSON PERSON _ _ _ --+K-=-----'p'--~~.:......___.:=_______

                                     --+K-"-_;.",p..::~_=_"____'=_~                       ______                   __

LOCATION/AREA WHERE ~~ QUESTION WAS ASKED: ASkED: _ _ ____ _-'-'~===--"-'::.=

                                       ~~....l....:=:..==-----;:,_~     __    __    __   __    __   __   __    _ _ ___

QUESTION/ITEM _ 0 DISCUSSED: tu~~ .A..2.IlH".,.J ~v J?f-kR. ~

       ;~           ~A1AH'q~  ;                                                             .

,e ,( RESOLUTI ON:

       ~4ccecDV~e~5~N~~~~-~~~~~~~d~~--~~'~                 ~-------------

PERSONNEL INVOLVED IN RESOLUTION: RESOLUTI ON: FOLLOW-UF' ACT I ON IIF ACTION F ANY: _ _ _._ __ __ __ __ _~ -'----_ (

    --                                           III.7-73 111.7-73 RR30

Revision B April 1992 DRESDEN STATION AF'PENDIX APPENDIX "R'I "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE: ?l-zo-B6 4-2D -B6 CE CO./AE CO. /AE F'ERSON QUESTIONED: -;::-, ~,~~ ~I <::... G=:~"'TH I COGN I Z ANT NRC F'ERSON_--""-""",,,"'--l.\~~~:::.L_"""=:t:C>V""",""'.!:=>::o;o",IJ,,,,-_ _ _ _ _ __ LOCATION/AREA WHERE QUEST -l.'R.......... QUESTII ON WAS ASKED: _ _....I'6:I....O?~'H:U.l\\. __

                                                                        ....)c--_..!fr~v::...!>"""--

l?:;..),..(.I:l...\l-a\.""">_ _.:.Jft"-\J.::::....Il> __ __ ____ __ ____ _ __ _ QUESTION/ITEM QUESTI ON/ ITEM 01 SCUSSED: DISCUSSED: sz.s=ro...:l IZ.SP0"" " ? Sit> HiS:> ft cwe:t CgnP'*')' ~ t c j~ l? E\>><;;:" E'->><SI'"

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                                                                                                       ~~'S'""""'T           ~
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RESOLUTION: p(\OV\CQ "Tv PC\oVI'OEO dV '-;)$ rn-+

        ~?t-2'-Bg
        &24f--2.\ - egg PERSONNEL INVOLVED IN RESOLUTION:

FOLLOW-UP ACTION IF ANY~ ANY, _ _ __ __ __ __ _ _ _ __ __ ____ __ _ __ ____ __ _ __ , C. RR31) RR30 III. III.7-747-74

Revision 8 April 1992 ( DRESDEN STATION AF'F'ENDIX APF'ENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22~1988 22,1988

                                                                           /l
                                                                           /1 CE CO./AE CO. / AE F'ERSON PERSON QUESTIONED:

QUESTI ON ED : _C=)_~ C-(, .:.;;..£_",,_~_v_n_~ k V n-{_______

                                               ~~~~~----------------

COGN II ZANT NRC F'ERSON_--,f._Z,--~(_/1 ~~!<-~..,;. ./1_:1 PERSON_-'f.'-"Z=----1(_,/14i~~/!._'_( ( >_O_M

                                                                 /.,-V-'-.>,-O IJ~______  _______

LOCATION/AREA WHERE /!{ lie ~ / . J. .1 J~ . - ) I ON WAS AS~<ED: QUEST ION i V'- tU~G:..<"I..d ASf<ED: --,-I_v'_"_rl/-,@=~ (rkrl UAI11....._ _ _ _ _ __

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                                                    ....(U_G:-<.!_*

7 ((. RES

   ~r~~~~~~~~~~~~~~~~~~~~~~~C~'~~~LC PERSONNEL INVOLVED IN f;:ESOLUTiON; f';:ESOLUTI ON:

FOLLOW-UP ACTION IF ANY: _ _ _ _ __ __ __ __ __ __ ____

   -RR30                                      III.7-75

Revision 8 April 1992 ( DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22~198e 22,1988 DATE: 4: 4-. 'lo-81;

                    'W-st
                                                                                              /)

co. // AE PERSON CE CO. QUESTII ONEO: QUEST _..::C::::.-,.=6_/ ON ED : _-=C::...-, e=--/_!_c~*{/....,;,(_H~ f,-(...:.V_(~H~______ _ _ _ _ _ ___ PERSON_.:-k~,--..:..;J._/f7.~2_~;.....~--...;.I-,-;1/_)_O_N COGN I ZANT NRC F'ERSON _ _ k-,--,-, !. ?. . :/f7.~<:. . . :.'?.!....g~(#..::..:..)-.:O:..:N-=--_ LOCATION/AREAWHERE LOCATION/AREA WHERE tJf~r

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                                       -'-'-'--'-'-'-_'--'/_-~--=-----"~-==':':::"::==l~----

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     ~     S;  fA: v' rtJ-t4J tf'~ l'~ ,!H>~                                                 ££1/.

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c. DA 4-"./U..C.(~::.".j'p~

                      +-.ru.e.0,J,P~ /                       # J7.J
                                                         /#.:ru               t.,y f..,'/                              II I.77-76 III.   -76 Uc. 's'.s 1'6>,7/.,,;

UCo  !'HI f! ,,,,J

Commonw** lth Edllon Edl,on .. ~-- ........ -- ... One 0". Fil'lt National P1.n. First NatiOl'll1 Plan. Chicago. nhnois illinois R,PlY to; pSi! Addr... A,PlY pOi! Office box 7e7

                                                          !lox 767 C"ago,     lIIinDil &oeQO Chicago. IIUnci,    e<leQO

• 0767 Revision 8 April 1992

( July JUly 23~23, 1981 1987 Mr. ~I

         ~s E. B. Murl,y, Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory commission Wa.hington. DC 20555 subject:     OUld ouad Cities Station Unit.               units 1 and 2 "10 CI"R erR 50, so, Appendix R ReqUirementsR.qUir.m.nts Por High-LOW High-Low Pnllure Interfact,"      Intufacu" NRC pocket Noft.       Nos. 'O-2~4 50-254 endi~o-365 and!5D-365 i

Dear Mr. Murley:

preparatiOn for the 10 CFR 50. In prepar.t1on 50, Appendix R audit, commonwealth Idilon Company (Clco) is performing a review of Idison company Qf the OUld ouad Cities Citiea Station', Appendix R safe Shutdown An.ly~il. Analysis, one of thl the issue. raised by this review concerns the app11¢~bl11ty applicability of the guidance prov1d'~ prpvid.d in Generic ~ett'r~etter 86-10 Inclolure 2, Section 5.3.1, reglrding regarding the anelY'is analyats ot of multiple -hot "hot shorts" in electrical circuits involving high-low pressure interfaces. interfaces, to the lolenoid solen01d op.r.te~ operated reactor relief valves, 1(2)-203-3A 1(2)-203-3A,** .,, C, D and I, the active .( cCilponents of the ~utomat cCIIlpon.nts Automat ic B Blowdown lowdown systemsysum (iP.BS). (lABS)

• It 1s is the position of Cleo CICO that the relief valves 40 do not con.titute constitute a high-low pressure interface for the purpos.s purposes *of Appendix R Inaly,1s. anelYlis. Thus, ana~ys1s of th***

   .na~ysis         .. e vaIv valves   .. is not lubject  IUbject to the consirleraUonconsideration of multiple "hot
   .hort." 1n
   .horts"   in the individual v.lve       valve control c1rc~1try  circuitry or ADS circuitry **      al required
                ~etter 86-10.

by Generic ~etttr 86-10, The basi. basis for th1.,pos1t1on th11:posltion 1s is conta1ned 1n the _ provid.d by NlC guidance provided I/IIC in veneric oeneric Letter 81-12 and it. ~1Iriflcat10n. c*larification *.*Th. Th. Staff'. concern with hlgh-l~ Btaff*. high-low pressure interfaces 1~ pre,sure interflces i. th.t a aingle

                                                                                               ,ingle fire could Cluse CIU" redundant reactor r.actor coolant bound.ry      boundary va         lye. to open, resulting val.e.                relulting in a fire-initiated ~OCA fir.-initiated    LOCA throUih through the subject     lubject intetface.

intetflce. This concern does not r.gard to the rellef exi.t 1n regard relier valves valve. for two rel.onl. reasons, Pirst. Pirst, the relief redundant coolant .yste. valves are not redundlnt .ystem i.olation i.olltion valveD. vllve,. The opening of any indivIdual relief valve will create a flow p~th for reactor individual reaetor coolant through suppres.ion pool located 1n: the valve to the suppression in, the pre,sure-Iuppr pr.ssure-suppr *** lon ion chamber (torus) partion primary containment. ~he: pertion of prtmary The: second rea.on the valves are not high-low pre,sure prellsure tntirfacil inurfacII il is this flow of: reactor coolant doe. does not constitute

   *I LOCA since the coollnt 1,         i . ..a1ntained
                                                . intained in a recoverablerecoverlble location locltion (i (i.e.,
                                                                                                       .*** the torus which i. is expressly designed de.igned for this Purpose)       purpose) within priMary containment.

containment, Thu., no fire-induced LOCA i, i. po ** ibl. due ~ po.,ibl. .purious oper'tion to .purioul operation of the relief valves and therefore the valves are not consid.red considered to b~ be high-low hIgh-low pre.sure interfaces for the purpo 1nterf.c.s purpol' ** of Appendix IR an~lYI1a. anlly.is. III.7-77

Revision 8 April 1992 PfUrl.y

f. I. KurllY July 23. 1987 Jl.Ily c

The response to Generic Letter 81-12 wa. wal provided by Cleo for the puad Cities Stltion st.tion by letter dated dat.d July 1,1. 1982. AI Aa Itat.d th.re. "the atat.d there, "The only id.ntifi.d high-eo-low identifi.d high-to-lON presaure puuun interhce interface with: dual du.l IIOtor opent,d op.ut.d isolation isOlation v.lv v.1v**..*** re r. located on the Residual Residu.l H,at H.at RemOval R.mOvel System Syst.m shutdown

                                                                                    .hutdown cooling PUlllP luction p\ap   suction lines ..*..... " In order to prevent I.    ** tire-induced fire-induced spurious IIpuriOl.ls operation from causing a LOCA through this     thia intlrf.ce.

intlrfae** it w.s propos.d v.s propo **d thlt that norm.lly clOled the normally elo** d ftHR v.1ve be locked in a deenergized RKR shutdown cooling valve polition position at Ippropriate motor control cent.r. It the appropriate c.nt,r. Th. ~h. NRC .tlff st.ff r.vi ...d r.viewed the r.sponse and found it acceptable. acceptabl ***ass docum,nted documented in the December D.c.mb.r 30. 1982

    .If.ty evaluation
    .aflty      Ivllu.tion Report, Section 3.4.3.               ii it conclullion, it is CICc's position th.t In conclusion,                                  ~h.t the eon.ideration con.iderltion of Ellt1pl.

IlUltipl' "hot _horts" electric.1 circuit:! shorts" in electrical circuit, involving high-low high-ION pressure pr.. sure int.rflc,. to the A!S interfAce. .nd related ABS And operatld reli.f rel.ted solenoid operat.d valves is outside r.li.f VAlves the requirements of Generic Letters 86-10 and Ind 81-12. We further believe beli.ve that your staff has accepted this position in ~he past thet p.st as evidenced evidenc.d in your December Dec~.r 30, 30. 1982. safety Slfety !vAluation. Ivllu.tion. dir.ct any questions you may have regarding Pl **** ~1rect r.garding this .att.r to th1, -att.r this offic:e. office. V.ry trulf Very truly your., your **

• c

.( 1. I. M. J h on Nucle.r Nucl.ar Licensing Lic.naing A 1m 1111 cc: R.glonal Administrator NRC Ilg10nal Adlilinistrator NRC Resident Inspector - QUad QU.d Cities T. ROil

           '1'. RO.. - WR I.

c. 3387K

1. 7-78 III.7-78 II

Revision 8 April 1992 ( DRESDEN STATION APPENDIX "RIO "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE: DATE : ---'~f+Y;...<::7-4-/-"-',(R"---

           -:~f+-; . .; ;. 7- . .,. .........~ff{u..--

CE CO./AE PERSON QUESTIONED: ______ ____ J[S~*~__(~~~G~-_~~~~

                                                                        -J1:\~    (~t$~G~-~~~'"'~______

COGNIZANT NRC COGNIZANT NRC PERSON PERSON _____ f\~ ____~f\~ __ __~~~~~~ ______________________

                                                                    ~~~~~~--------------------             __

LOCATION/AREA LOCATION/AREA WHERE QUESTION WAS ASKED: ----4~.........,,= WHERE

                                                             ~
                                                             ~~-T
                                                             ~u...,,=:.c::..:: _ _______ ___ _______ ______ __

c ( RESDLUTION: RESOLUTION: _____________________________________ PERSONNEL INVOLVED IN RESOLUTION: RESOLUT ION: FOLLOW-UP ACT ACTION I ON IF ANY; ANY: ________________________ RR30 III. 7-79

Revision 8 April 1992 ( DRESDEN STATION AF'PENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22~1988 22,1988 LOCATION/AREA WHERE N/) Nil QUESTION WAS ASKED:I<-L. ASKED:/<..L.

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                                           /01)/0/2/(//1 PERSONNEL INVOLVED IN RESOLUTION:

RESOLUT ION: FOLLOW-UP ACTION IF ANY: ______________________________ III.7-80 RR30

Revision 8 April 1992 C' ( DRESDEN STATION APPENDIX "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE: f- 2J-f:Z-e-g CE CO./AE CO. / AE PERSON QUESTIONED: QUESTI ONED: __ C:_-_~ __~~~ __G/__~~______________

                                                  --:;C'----'E~-....:£--..:u::......;TH'~

COGNIZANT NRC PERSON I.e'. .~ ;/,4/2K!7.Iftf'#

                                        /.!      //1-t'2K#.Ifd"H
                                    --~~~~~~~~---------------

LOCATION/AREA WHERE QUESTION WAS ASKED: QUESTION/ITEM, QUESTION/ ITEM . ~ DISCUSSED: Jt/f;;;.!fJ;1;jf

                         /{/~"f!~ ~J<<';

12.G7> Wt_ (?qllit2G7:>

     !?t<'"@l/i                 J/'/tC" I',{._ JL'i&    ?

z:;$4-~: 7, WHitT W/fJ'1 r (s (AI~ Is IN ~ ~ FOLLOW-UP ACTIONS: PERSONNEL INVOLVED IN RESOLUTION: FOLLOW-UP ACTION IF ANY: ________________________________ c (. RR30 II 1. 7-81 7 -81

Revision 8 April 1992 DRESDEN STATION AF'PENDIX "R"" "R" AUDIT QUESTIONNAIRE FORM APRIL 18 - 22,1988 DATE: 4- '2/-

                      '2.1- g 8 CE CO./AE PERSON QUESTIONED: ___                                __C_'_~

__ __-_A? C ._~ ____ __~_A?lt/_~ /~

                                                                '11

_)_~_,_f_yh.....t.f_'_r.....:;7 f...-:.'G<..:.(J/~~_6...:..,N J/"-'-~_6.:c~=--- 7 COGN I Z ANT NRC PERSON _ ___ ""t:::..cf.....;;'G..:...;(

                                              }_i::::_*....;,f...;,yh-J..,                                __     __

LOCATION/AREA WHERE QUEST QUESTII ON WAS ASI<ED:

                                          !2 f)

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                            'J);V/S(utV                ~ (72/l-y           r72A Y JUA/C.!16tV               J(,! A/c. vet!                w i T1I DtV UJiT1!     Drv.:r:  . .:r.~
    .M 12,
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IN<e,UI/CG"b Wtftti.$-atfr ci-= i-= CMklf> CMktfS WO(lt..b WDi.)£..p P/'.55* rtltgo jJf':SS' T1f&O ViiI-! Uft! T%{15 "f'U15 JLltv'C-7U'/\/JUalC-7T(71\/. FOllOW-UP FOLLOW-UP ACTIONS: ______ .. ___._ _ __ __ __ __ __ __ __ ___ __ __ ____ __ __ __ __ _ _ ___ RESOLUTION: .s;:1,4~t> mAr c~4-:£'t.t;F /dJUt7/~6';IrG"~~;1-77'~A/ ()fJ..../'tE;?,IA 11£1-04) BOP CA-8L,.GS" Z"4I CRC.~~ rill'> '/U&/kUOt0 PERSONNEL INVOLVED IN RESOLUTION: RESOLUTI ON: FOLLOW-UP ACTION IF ANY: _______________________________ ______________________ ~ c. RR3C) RR30 III.7-82 II 1. 7-82

1 Revision 8 Commonwealth Edison April 1992 c*c On~' F"~I On,> Ftf51 Ntl~ll1nal N~'\lona' Plnzn, Chlc<lgo. IlImnl~ Prill;,!, ChiCilqO. IIhnOIt. Addres~ Ad'dress Reply !O"'P0510Iflce" 80x767 RODr], IO***POGtOlflce* 80:<767 ChlC:<lgo. illinoIs 60690 ChICJgO. IllinoIS february February 1, 1Qng lQIl9 Mr. A. Bert Davis Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt R9ad Road Glen Ellyn. IL 60137

Subject:

Dresden Nuclear Power Station Units 2 and 3 Response to Notice of Violation Nos. 50-237/88010-06 and 50-249/88012-06 50-249/B8012-06 NRC pocket Docket Nos. 50-237 and !?~_,._. __.__ _ 5~__________

Reference:

J.J. Harrison letter to Cordell Reed dated January 3. 3, 1989 including Notice oC Violation concerning improper storage of transip.nt combustible liquids. -(

Dear Mr. Davis:

The referenced letter provided the results of. special safet.y inspections conducted by Messrs. J. Holmes. Holmes, R. Hodor, Hodar, and K. Parkinson. on' April 18-22, 18-22. May 11-3,11-3. August 15 artd and December 13. 1988. of Fire PcotectjQn Protection activities at Dresden Oresden Nuclear Power Station. During the course of these inspections, certain activities appeared to be in noncompliance nOllcompliance with NRC requested, a response to Item 2 of the Notice of Violation requirements. As requested. is provided in the Attachment. Commonwealth Edison understands the significance of the issues identified in the Notice of Violation and has implemented corrective actions to prevent recurrence. Although several other Aspects of the Inspection Report warrant clarification, our comments will be provided via a separate transmittal. II I. 77-83

                                                                     -83

                                                                                           .... ~.-- ......... ,.. - . .-

Revision 8 April 1992 ( A.B. DQvis [lovis regardiJlq this response, plQil~~ If there are any !urther questions reg"rdiJJ~1 plp'o'\!lf'!' contact this office. Very truly yours, G)j, ~--~ 0t F.-I?l~ H. ~" H.~"S Nuclear Licensing Manager 1m Attachment ( cc: B.L. Siegel - Project Manager, NRR S.G. DuPont - Senior Resident Inspector, Dresdell 5.G. Dresdcll 552SK 552BK c( . II I. 7-84 IIl.7-84

Revision 8 April 1992 c REJ~.!.3---I..Q...1!Q.TJ..Cf,;_ Qf _'!..I OL hT I 011 10 CFR eFR 50 .40( a) requires that each operatinq nuclear power plant. IJiWf"! IlilVP. a Cire fire protection plan that satisfies General Desiqn r=riterion r:riteciQn 3 (If ( I f AppellcUx AppcucUx II" to 10 CFR eFR Part 50. It further requires that the pl.,n sh"ll Sh,,"ll de:'Ocrihe de:;crihe :;pndCie fipnr.iCir: features necessary to implement the proqrarn proqraJn sueh such as adlninlsl.rative adJninisl.rat.ive cont.rol:; cont.rol!; and the means to limit fire damage to structure:=;, structucp.s, systems. systems, or component:; component5 important to safety so that the capability to safely :.hutdowu plaut is r.hutdowll t.he plant ensured. Section B.2 of tIle the licensee's responsE! re5pOnSE'! to the GlIhlp.line5 or I.he Guhlp.lines of AppP-lulix Appelulix A to Branch Technical Position APCSB 9.5-1 as acceptpd ncceptpd in the 1981} 1980 Supplemental Safety Evaluation Report. illdicat~s iudic"!'.!,;!5 that p.((p.clivp. i'\c1rnlllist.l"i'\t.jv~ the"'lt p.[fp.clivp. nc1minisf.,..'lUvP mei\sures will be implemented to prohibit blllk 5tc)ra9~ of cornl)u:o;t.ible bulk storaq" comhu::>t.ible mntpl in!!; mntpi inl:; inside or adjacent adjC\cent to safety-related buildi buildings ngs or sysl".emr: sysl".em:-: during operal.ioll operal.ion or maintenance periods. ( Contrary to the above, during a previr)IJs April 12, 1988. previous iJl:;pF!ction 1988, an NRC inspector observed twenty 55 *<:jilllon il1r;[u'!ction conduct.eel conduct.ed 011 "r.JClllon drwns orof lubdc:flUIl'J Iubl"icnlilv] oil stored in a safety-related area on Elevation 517' - 6" (in (ill t.he south",p.~t. southwe~t corner) of the Unit 2 Reactor Building. This c:oncliLlon (:ondiLlon exist.ed [rolO Crom Mill M"I r.h r-:h :11 31 to April 13, 1988. This is a Severity Level IV violation (Supplement 1). I). As part of the Unit 3 refuel outage, outage. the Unit Unit. 3 HPCI HPC! turbine W~5 Wi1S inspected and required maintenance performed. This Thjs inspection iuspection included draining the lube oil reservoir and either c:leanjng r.leanjng or replacing the thft oil. Thp. Unit 3 HPCI HPC! lube oil reservoir was drained into twenty 55 S5 gallon drwns local~el located ou Elevation 511 517 of the Unit 3 Reactor Building (souL/least (souLheast corned.corner). Shut*tty Shut"tty after this job was completed, other outage work required the "311." "'3A" LPCl heatheat. eXChanger be accessible. To obtain access to the ..fI)A" 3""' LPCI heat exr.hangp.r-. exr.hangp.I"; tltf' thfl relocated~ oil drwns had to be relocated. The Fire Marshal was "ontllcted r'ontilcted and detE-lln i lied detE'lInined that the barrels could be moved to Unit 2 Reactor Building Elevatiull ElevatiuJl S11 wj!.h 517 ."jl.h no additional fire protection, based on: (1) beinq a low traffic area, it being (2) the fire detection system was available, available. (3) safe shutdown could still be achieved, ( (. Ill. III. 7-85

  *..'                                                                                       Revision 8a April 1992 c                                                - 2 *.

(4) the characteristics of the lube oil (beill'J (beiu9 difficult dirCjcult to ignit:~ ignit~ <lJlr.! ilud having a relatively high flash point), (5) no other work being performed or combustibles beiug beiu9 stored in tha area, ar~ar and (6) the increase in fire loading did not exc~edexceed a low fire eire loa~ing load lng (i.p.., (i. Po. , less than 100,000 BTU/it 2 ), based 011 the storage st.orage are~ areA. involveu, A~ involved, l\'!; deCined de! illed in *the

                              ,the NFPA handbook (15lh (15th edition).

ed i tion)

• Despite the above, CECo understands the inspector's concern (';oncern regArding rp.gi'\rding the storage of significant quantities of transient combustible liguids liquids in safety safet.y related areas of the plant without additional co'"pensnlory compeu5nlory measures.

CQ!!.K!;'cTIVE ACTION AN~.!.&TS ACHIEVEQ On April 13, 1988, the barrels of lubricating oil were removed from the Unit 2 Reactor Building, thus eliminating the ~oncern~ r.oncern-.

C Dresden Administrative Procedure (DAP)(OA.P) 3-3, Transi~llt.

3-3. "Control of Transip.llt. Combustible Storage Areas and No Smoking Areas," will be revised by the fit(~ fitp. Marshal and a Fire Protection Engineer from Support Support. Sprvic~s. S/!IIrvicl?s. The Thp. revi51un revi5iun will ensure that routine bulk storage of transient combustibles will not exi:>t. eKi~t. in the plaIlt except in designated areas. The revision will establish temP?f<uy temP?1"iuy bulk storage guidelines defining the time a specified amount of tr~nsi~nl transi~nl combustibles will be allowed in a non-designated area and compensatory measur~s measur~$ to be taken during that time. In addition. re',liews will be basp.~ addition, future reviews basAd strictly on the total amount of combustibles allowed ill an area. The proceourn procedl1r~ revision will be completed by March 31, 1989.

    ~LL                  COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on April 13, 19A8     19R8 c    5528K
                                                   !IL7-a6 II I. 7-86

Tab 8 Revision 8 April Apri 1 1992 ( DRESDEN 2 &3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspectl0n Inspection Report No. 50-237/88010 and 50-249/88012 Title rrI,8-1 III.8-1 Inspection Report No. 50-237/88030 and 50-249/88031 dated January 23, 1989. rrI,8-8 II I. 8-8 April 14, 1989 CECa CECo letter from E. D. Eenigenberg to R. J. Israelson (3M) on review of installed E-50 Fire Wrap Removable Covers. III.8-12 May 3, 19890 letter from R. J. Israelson (3M) to E. D. Eenigenberg, response to April 14, 1989 letter. letter . . (. ((~- III,8-i III.8-i

• c.\.~~R 'H;C,,<..t UNITED STATES Revision 8

         ~,        *           ~~$"r.:;:r:;-;~-';-:~~!Iif\RA REGULATORY COMMISSION
                               /I Apri1 1992 April
        ~.

t;, .

                                                      !~          REGION III REGION III 199 ROOSEVELT ROAO
                                                            '"               ROAD
        ~                                                 LEN ELLYN, EL.LYN. ILLINOIS 60137 "1-"

I.I'----_-.JL:/! JAN 231989 2:3 1989 Docket No. 50-237 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen: This refers to the special safety inspection conduct~d by Mr. J. Holmes of this office on October 22-23, 1987; May 5-6, 11, 12'~nd December 21-23,1988; II, 12'and 21-23, 1988; and January 20, 1989, of activities at Dresden Nuclear Power Station, Units 2 and 3 authorized by NRC Operating Licenses No. DPR-19 and No. DPR-2S DPR-25 and to the discussion of our findings with Mr. E. D. Eenigenburg at the conclusion of the inspection. This inspection was conducted to review allegations regarding deficiencies in fire wrap installations and the training provided to new installers. installers * ,((.. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. No violations of NRC requirements were identified during the course of this inspection. i nspecti on. In'accordance In*accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC Public Document Room. We will gladly discuss any questions you have concerning this inspection.

                                                                      ~~

J. J. Harrison, Chief Engineering Branch

Enclosure:

Inspection Reports 50-237/88030(DRS); No. 50-237/88030{DRS); No. 50*249/88031(DRS) 50-249/88031(DRS) See Attached Distribution C'

                           ,                                   III.8-1

Revision 8 April 1992 c Commonwealth Edison Company 2 JAN? "' 1989 Oistrubtion Distrubtion cc w/enclosure: H. Bliss, Nuclear Licensing Manager J. Eenigenbur9, Plant Manager DCD/DCB (RIDS) licensing Licensing Fee Management Branch Resident Inspector. Inspector, RIll Richard Hubbard J. W. McCaffrey, Chief, Public Utilities Division .( c III.8-2

Revision 8 April 1992 ( U.S. NUCLEAR REGULATORY COMMISSION REGION II I Reports No. 50-237/88030(DRS); 50-249/88031(DRS) Docket Nos. 50-237; 50-249 Licenses No. DPR-19; DPR-25 Licensee: licensee: Commonwealth Edison Company Post Office Box 767 Chicago, Il IL 60690 Facility Name: Dresden Nuclear Power Station, Units 2 and 3 Inspection At: IL 60450 Morris, Il Inspection Conducted: October 22-23, 1987; May 5-6, 11, II, 12 and December 21-23, 1988; and January 20, 1989. Inspector: inspector: J~m~ Vt.o/89 oDV1.O/89 te

                     ~~}f~;
                     ~~Jf-'

.( . (~: Approved 8y: By: \t. VN:-~~~~~r,

                     \!. VN:'~-r~~~r, Chief Plant System Section Inspection Summary Inspection on October 22-23, 1987; May 5-6, II, 12 and December 21-23, 1988; and January 20, 1989 (Report Nos. SO-237/88030(DRS);

50-237/88030(DRS); 50-249/88031(DRS) Areas Inspected: Special safety inspection into allegations of deficiencies in the fire wrap installations and deficiencies in the training provided to new installers. Results: No violations or deviations were identified. o The inspection concluded that while two of the three alleger's concerns were substantiated, no violations of NRC regulatory requirements were identified. With regard to the alleger's third concern, there was no evidence found to support the allegations that there was a lack of independence between Quality Control and Production activities. (.- II 1. 8-3 III.8-3

Revision 8 April 1992 ( DETAILS

1. Persons Contacted Commonwealth Edison (CECo)

           *E. D. Eenigenburg, Station Manager E. Armstrong, Regulatory Assurance Supervisor
           **8.
           *B. Barth, Technical Staff Engineer R. Black, Assistant Fire Marshal
           *M. Dillon, Fire Marshal T. G. Hausheer, Fire Protection Engineer, Production Services
           *K. Peterman, Regulatory Assurance Supervisor C. W. Schroeder, Services Superintendent Transco G. Jarose, Engineering Manager L. Anderson, General Foreman W. Saar, Baar, Installer B. Fatt, Division Quality Assurance Manager P. Greaney, Installer B. Leone, Quality Control D. Marz, Installer S. Pearson, Quality Control D. Sisk, Quality Control U.S. Nuclear Regulatory Commission (U.S. NRC)

S. DuPont, Senior Resident Inspector

           *Denotes these person participating in the telecon te1econ exit meeting on January 20, 1989.

2. Allegation RIII-87-A-0074 Region III received a telephone call on May 21, 1987.

1987, from a former contractor employee at Dresden who contended that deficiencies existed in fire wrap installations and in the training provided to new fire wrap installers. The individual also indicated that there was a lack of independence between Quality Control and Production Activities. Each of the individual's concerns are addressed below: Concern 1: The training program provided to new installers consisted of requiring the installer to read the procedure and sign a document that indicated that the installers had read and understood the procedure. The training program did not contain any practical demonstrations and new installers were expected to obtain their training on the job. NRC Review: The allegation was substantiated in that training provided to new installers consisted of having new installers read the procedure and then sign a document showing that the installers had read and ((-

   .. -    understood the procedure. The allegation was also correct in that the 2

111.8-4 III.8-4

Revision 8 April 1992 ( training did not contain any practical demonstration and the new employees were expected to obtain their training on the job. The Transco procedure for qualification of site craft personnel (PSQAP 2.1) indicates that the indoctrination period varies in length length, and scope, and t is totally dependent upon the complexity of the functions involved and past experience *of

                        'of the individual. In addition addition, the procedure indicates t

that indoctrination is administered either on-the-job or within a classroom environment and is recorded on the "Site Personnel Certification Forml! Form" as attestment to qualification by the Transco Field Superintendent. In discussions with the licensee and Transco, Transco indicated that the individuals who are hired as installers must have a union card which is obtained by apprenticeship with an experienced installer for at least two years. Transco indicated that if the individual installer can follow directions installing insulation, then the individual can follow Transco procedures. Transco indicated that the procedures are required to be read and this takes approximately 15-30 minutes. Afterwards, the Superintendent reviews the procedures with the installers and discusses key points using the specific details and pertinent documents. The installer is then transferred to a Foreman or leadman. The Foreman or leadman is respoO*ible respon&ible for the crew and usually determines the duties of the new installer (the aSSigned to a member of the crew). new installer is normally assigned The inspector conducted field walkdowns and reviewed the training records and the installation procedures. The inspector also discussed the Transco .(. ,C.. ' training program with several installers, and Quality Control personnel. The Transco employees indicated a mixed opinion regarding the training from excellent to additional training is required. The general consensus was that the General Foreman and Quality Control personnel would insure that an adequate fire wrap was installed. Cone 1us; on: Based on a deta

Conclusion:

il ed revi detailed ew of the fi review e 1d "take-off" field ntake-off" records, installation drawings, nonconformance reports, field walkdowns, and interviews with Transco employees, no discrepancies or violations of regulatory requirements were identified. Although the training provided by Transco to new installers may have been weak in certain cases, it appeared that the Transco General Foreman and Quality Control personnel insured that the installation was done according to design criteria. Concern 2: On-the-job training was given by new employees and therefore untrained new employees were providing on-the-job training to newly hired employees. NRC Review: ._ This allegation was substantiated. In discussions with Transco and the licensee, they acknowledged that new employees may have been in a position to provide on-the-job training to new employees, but that the General Foreman and Quality Control personnel observed the key parameters in the installation and would have identified an incorrect installation.

Conclusion:

Based on detailed review of the field IItake-offu "take-off" records, installation drawings, non-conformance reports, field walkdown, and (" interviews with Trancso employees, no discrepancies or violations of 3 III.8-5 III .8-5

Revision 8 April 1992 regulatory requirements were identified. Although on-the-job training may have been given by new employees, it appeared that the Transco General Foreman and Quality Control personnel insured that the installation was done correctly. Concern 3: There was a lack of independence between Quality Control and Production Activities in that the Production Superintendent (or General Foreman) was contacting the Quality Assurance Manager and complaining that Quality Control was delaying production. Also, the Production Superintendent controlled the company telephone and truck and prevented Quality Control from using the telephone or truck unless pennission permission was granted from the Production Superintendent or General Foreman. NRC Review: In discussions with the Quality Assurance Manager, the Manager indicated that telephone calls were received from the field superintendent (or General Foreman) regarding design and installation of the Fire Wrap. The Quality Assurance Manager further indicated that no calls were received regarding Quality Control Inspectors or Quality Control Managers delaying Production. Also, the Quality Assurance Manager indicated that during the exit interviews of the Quality Control Inspectors and Quality Control Managers, no safety issues or issues regarding Production Superintendents contacting the Quality Assurance Manager was discussed.

      ~Qnager cc:~_/ In addition, the Quality Assurance Manager indicated that Quality Control Inspectors and Qual ity Control r'tanagers f4anagers were allowed to use the office telephone for business and not for personal reasons. The Quality Assurance Manager also indicated that the Transco truck was strictly used piCK-Up mail and that permission from the to transport material and pic~-up Production Superintendent was required to utilize the company truck.

In discussions with Transco management personnel, Transco indicated that the Quality Control Group was under the direction of the Quality Assurance organization which reported directly to the President of the company and that if any disagreement between production and Quality Control personnel did occur and could not be resolved thru the management organization then it would be resolved by the President of the company.

Conclusion:

Based on discussions with the Quality Assurance Manager there was no evidence that the production superintendent (or General Foreman) contacting* the Quality Assurance Manager to report a Quality Control was contacting-Inspector or Quality Control Manager for delaying production. addition, based on discussions with Transco_management In addition. Transco.management personnel, the telephone was available for Quality Control, however, the company truck (which was used to transport material) was not available to the Quality Control Group unless permission was granted from the Production Superin-tendent. The company truck was considered part of the equipment utilized by production and it is not considered unreasonable that the Quality Control Group requested permission to use the company truck. c Based on the above, there was no indication that a lack of independence existed between the Quality Control and Production Activities. 4 III.8-6 IIL8-G

Revision 8 April 1992 Unit 2 Trackway Fire Wrap Details The licensee has fire wrapped risers on elevation 517' and 534' consisting R38D which interconnect two large sheet metal of cable tray risers R379 and R380 pull boxes. Transco developed a fire wrap access cover to these pull boxes by using criteria from Transco Detail J6 and Special Drawing EJ 44 (dated 3D, 1987). Due to the numerous physical configurations that may January 30, be encountered in the field, 3M allows variances in its application of the material as long as it meets its design criteria. The observed access cover developed by Transco for the licensee appeared to meet the critical criteria such as number of layers, bands, caulking, etc., however, due to its unique design, it was requested that 3M review the installation of this design to ensure that its unique design had not invalidated its fire rating. This is considered an Open Item (237/88030-01{DRS)i (237/8803D-Ol(DRS); 249/88031-01(DRS)) pending review of the 3M response.

4. Crib House During an inspector walkdown, it was observed that a small portion of the fire wrap installation on a junction box did not contain caulk. After the licensee was infonmed informed of this concern, the fire wrap was declared -

partially degraded. In discussion with the licensee, the licensee indicated that work had been performed on the junction box and the original fire wrap removed. c After work was completed, the wrap was replaced and the caulk not replaced in the lefthand corner of the barrier. The licensee indicated to the inspector that 3M will be conducting training sessions for the installation of the fire wrap for workers and Quality Control personnel at the end of January 1989. The licensee also indicated that the small opening will be recaulked by the end of January 1989.

5. Open Items Open items are matters that have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. An open item disclosed during this inspection is discussed in Paragraph 3.

6. Exit Interview The inspector conducted a telecon meeting with licensee representatives at the conclusion of the inspection and summarized sunmarized the scope' and findings of the inspection. The licensee acknowledged the inspector's comments.

The inspector also discussed the likely informational content of the inspection inspec,tion report with regard to documents or processes reviewed during the inspection. The licensee did not identify any such documents or processes as proprietary. l 5 III~8-7 rIl.8-7

Commonwealth Edison Dresden Nuclear Power Station Revision 8 R.A. RR,#l#1 .' April 1992 c Morris, Illinois 6().450 so.cSO Telephone 815/S.2-2920 815/9-42-2920 April 14, 1989 EDE LTR: #89-311 Mr. Ronald J. Israelson 3M Ceramic Materials Department Building 207-1SC-12, 3M Center St. Paul, MN 55144-1000

Subject:

Review of Installed E-50 Fire Wrap Removable Covers

Dear Mr. Israelson:

As part of the E-50 fire wrap systems installed at Dresden Station during 1987, several configurations were installed that did not follow a standard 3M detail. Deviations from details are permitted as according to 3M drawing 5300-QA, provided that critical design requirements are met. Several of the deviations included installation of removable covers on electrical junction boxes and pull boxes (see attachments). During the installation process, 3M representatives assisted the installers in proper installation procedures and techniques, though few standard detail drawings were formally prepared. The Nuclear Regulatory Commission (NRC), during a review of Dresden Station's 3M fire wrap installations, questioned the practice of installing non-standard designs without the development of special "site-specific" details. The NRC requested that CECa CECo have the installed designs reviewed by 3M Corporation to ensure adherence to E-SO E-50 fire wrap system requirements. _ The attachments list the standard details which are believed to have been followed during the design and installation of the removable covers. During a 3M E-SO E-50 system training session held at Dresden Station on January 25, 1989, you were questioned by the Technical Staff Fire Protection System Engineer regarding the installed configuration of removable fire wrap covers at Dresden Station. At that time, you indicated that the design appeared to meet E-50 fire wrap system critical design requirements. Dresden Station is requesting 3M Corporation to review the as-built sketches for compliance to E-50 fire wrap system requirements. The Station understands that 3M will provide technical support for it's E-50 fire wrap product at no additional cost to the purchaser. If you require additional design information, please contact Eric Skowron, Technical Staff Fire Protection System Engineer at extension 2353. EDE:EJS:jmt EDE:EJS:jrnt Attachments Dresden Power Station c C cc: M. Strait R. Whalen M. Dillon E. Skowron File/T.S. File (4100) File/Mise Fi1e/Misc File/Numerical 3427a II I.8-8 III.8-8

ATrACHMENI' 1 Revision 8 April 1992

                                                                   '-       STEEL PLAn:

PLATE 53()()-J6 53OO-J5 2-50A E-SOA LATER REQUI!!EHEHTS REQUlREHENTS

                £-50A      7 CS-195 INTERFACE DETAIL 10 TO WALL VALL 5~.

5:JOO..<11l. MA'l'EIIIM.S MATERIALS AND GEHEAAL GENERAL INSTALLATION I\EQUI.RDiENTS; REQUIRDiENTS: (Pl-8l (Pl-8) 1 ... 50 HAT E~50 CS-195 COHPOSITE COKPOSlTE SHEET CP-25 CAULK T-"9 ALUHlHUM ALUMINUM TIU'E TAPE ST!EL BANDING S'l'!EL STEEL S'l'EEL COVEJI COVER (SUBSTITUTED FOR S'l'!!L STEEL WOVEN MESHHESH I

   ~l!J59
   ~&J59 STEEL BAND ANGLE BRACKET ANCHOR DETAIL
   ~B.J41 6000-&1'41 STEEL BANDING TRANsvtRSESLY TRANSVERSESLY CONNECTED 10'1'0 OTHER BANOS
   ~B.J"5 6000-BJ45 OVERLAP llEQUlREMEHTS REQUIREMENTS Co' CABlE 'mAY           RISE'Ri R379 Am
                         ':mAY RISFR5             AR> R380 R3SD III.8-9 IIL8-9

Revision 8 April 1992 (

                                                                    ~\O:

ru'

 -10 N

53OO-J6 £-50A I-SOA LAYER REQUIREMENTS E-50A

                   £-501. , CS-195 INTERFACE DETAIL TO ~ALL WALL 5JOO--QA 5.JOO..-.O\   MATERIALS AND GENERAl.

GENERAL INSTALU.TION INSTALIJI,TION REQUIREMENTS: (Pl-8} (PI-B) £-50A E-SOA HAT KAT CS-195 COHPOSlTE SHEET CS-19S CP-25 CAULK 0-25 T-49 ALUMINUM TAPE STEEL BANDING 6000-&141 STEEL BAND ANCHOR DETAIL fiOOI).-IIJ41 6000-BJ45 OVERLAP REQUlREHENTS 6000-&145 REQUIREMENTS JlH:TICN 001 3CB-9 III.8-IO III.8-10

A.'ITAClIME:Nr 3 Revision 8 April 1992 ( Ol/~ n

                      '/,       .
    ~1f2,
 ~

o or 5lOO-J6 53OO-J6  !-50A E-50A LAYER Ll.YER REQUI:REKElfl'S REQtJIIU!KEIfl'S E-50A , CS-195 INTERFACE DETAIL TO WALL 53QO...gA 5JOO...ga MATERIALS AND GlHERAL INSTALLATION REQUIREMENTS: GENERAL INSTALI.ATION (Pl-8) (Pl-a) £-50A MAT CS-19~ CS-19!\. CQUIOSID SHEET CQ!POSITE CP-25 CP ...25 CAUU: CAULIC T-49 ALUMINUM TAPE TAP! STEEL IlAllDI!<<: S'l'!EL BAHDIHC IOOO-&.JU S'1'EEL IOOO-JrJU S'1'E!L BAND ANCHOR DETAIL D!'V.IL 6000-&145 OVERLAP RBQUIRDiEN'l'S 1OCJO....U45 RlQUIRDmN'l'S PAl£[. 2223-109 III.8-11

              .p,.... ". ,.
                    .  ~\..;~

• j;".

                                     .,j'_ *
  ",         ~~~L-'  .".              ,.

Revision 8 April 1992 ( 3M MAY 3, 1989 Mr. E.O. Eenigenburq Eenigenburg Station Manaqer Manager JUt Mil .1 Dre,den Nuclear Power station Dreaden Nuclea~ Morria, Ill. 60450 Horris, DeAr

Dear Mr. Eeniqenburg,

Eenigenburg, I would like you know that I received your letter datea dated April 14, 1989. I am familiar with the installationsinatallationa that are drawings labelled "Cable Tray ailer8 detailed on the d~awin9s Risers ~379 R379 and R380", "Junction Sox Box 3CS-9 3CS-9", f1

                                                                                   , and "Panel 2223-109".

an.d 2223-109", Although I cannot verify that the installation. installationa were performed to meet the drawings, I can state that the drawings represent suitable applications of the 3M Fire Protection requirements. ASluming that the installations were partorme~ Aalumin~ performed as described .(( on the drawings, eaeh each installation representl & inst~llation represents a full 1 hour of fire protection. please Please eall call me at (612}736~38l6 (612)736~3816 if you need any additcnial additonial assistance. Sineerely, Sincerely, Ronald J. Israel.on Israellon 3M Technical ServiceS.~vice F?OCC INITIAL. INITIAl.. cc; Eric S~owrQn Skowron *- Technical Staff Fire protection Systam. Systems c Engineer' III,8-12 III.8-12

Tab 9 UNITED STATES Revision 8 NUCLEAR REGULATORY COMMISSION April 1992 c*( , , REGION III 799 ROOSEVELT ROAO ROAD GLEN ELLYN, ILLINOIS 60137 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen: This refers to conducted by lor .*leff Holmes of this office on ~~~~~~~~~~~~~~~~~of circumstances associated with~fi~~~~~~~~~~~~~~~~~~~~1*~~~~~~~~L.. our The enclosed copy of our inspection report id~ntif;es examin~d during areas examined id~ntifies area~ inspecticn. Withlll the inspection. withlll these areas, tile inspe~tion consisted of a selective areas~ the examination of proced~res examin~tion procedures and representativ~ representative records, observations, and .. ( interviews with personnel. intervi~ws During this inspection, Certain certain of your activities appeared to be in violation of NRC reqUirements, as described in the entlosed enclosed Notice. The inspection showed that actions had been taken to correct the identified violation and to revent recurrer:cf;'. recurrenc!:. Our unders tandi n9 of your correcti understanding corrective actions is ve actlorlS Paragrap 2.e. described in Paraqrap Z.e. of the enclosed inspection report. Consequently. repl)' to the violation is required and we have no further guesticn$ no reply guesticns regurding this matter at this time. reg~rding Corr~ission's regulations, a copy of In accordance with 10 CFR 2.79G of the Corr~issionls letter' arid the e~closed this letter erlclosed inspection report will be placed in thf the NRC Public Document Room. We will gladly discuss any questions you have concerning this inspection. Sincerely, Si neere ly, Ct17f~

                                                          )z, )1/ U17fJIL
                                                          ;t R. W. Cooper: Chief Engineering Branch

Enclosures:

  ./

Notice of Violation

1. Noti~e

t( , 2. Inspectiun Inspection Report

       "          No. 50-249/69004(DRS) 50-249/&9004(DkS)

See Attach~d Attac~cd Distribution III.9-1

Revision 8B April1 1992 Apri c( Commonwedlth Edison Company Distribution cc w/enclosures: H. Bliss, Nuclear Licensing Manager J. Eenigenbur9, Manager Eenigenbur~, Plant ~anager DCD/DCB (kIDS) DeD/DCB (KIDS) Licensing Fee Management Branch Resident Inspector, RIll Richard Hubbard

w. ftlcCaftrey, J. W. ~lcCaff .. ey, Chief, Pub1ic Public Utilities Division

( ( III.9-2

Revision 8 April 1992 NOTICE OF VIOLATION Con1110nw~a l1..h Con1l10nw~a lth Ed; Edi son Company Docket No. 50-249 As a result of the inspection conducted on June 4, 1988 through February 8, 1989, and in accordance with 10 CFR Part 2, Appendix C - General Statement of Policy and Procedure for NRC Enforcement Actions (1988), the following violation was identified: Dresden Technical Specification Section 6.2, entitled, "Plant nPlant Operating Procedures ,'I Procedures," requires that deta;l~d detail~d Fire Protection Program Proc~dures be prepared, approved and adhered to. Th~ license~'s Fire Preventive Procedure 3-2 re uires weldin activities art: in ro rt:ss care shall be taken rom cutting operiltions t rous *near y openings. Contrary to th& the above, on June 4, 1988, the license& licensEe did not adequately protect penetration x-i1a, penetrati~n x-114, ~hich I<'hich leac5 to the combustible drywell liner, from cutting and welding uctiviti~. ilctivitieL As a result, u fire WaS WilS initiated in th~ drywell liner. This is a Severity Level IV violationvlulation (Supplement :1.). I). insp~ction showed that actions had bt:en The inspt:ction beer. taken to correct the icientified violbtion violation and to prevent recurrence. Consequently, ~~ reply to the violation nc; r~ply is required and we have no further qu~stionsqUtostions regarding this matter. Datec Dated R. w. W. Cooper~ Cooper, Chief Engineering Branch III,9-3 II L 9-3

Revision 8 April1 1992 Apri ( U. S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-249/89004(DRS) Docket No. 50-249 License No. DPR-25 licensee: Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Dresden Nuclear Power Station, Unit 3 Inspection At: Dresden Station, Morris. Morris, Illinois Inspection Conducted: June 4, 1988 through February 8, 1989 Inspector: ~!~ 2/28/89 2./28/89 Date t.;:~~~ \\ L",~ \\ ~~~'J-.....

                              ~,,\,.,J-......

Approved By: Ronald N. Gardner, Chief Plant Systems Section Date Inspection Summary ectiu~ on June 4, 1988 throu h Februar' Ins ectiGIi Februa~ 8, 1989 (Re ort No. 50-249/89004(DRS ) reas nspected: nnounced specia r~as sp~cia lnspectl0n con ucte to reVlew sa ety lnspectlon licensee actions with regard to a fire in the Unit 3 Dry..lell Dr~/ell Expansion Gap which occurred on June 4, 1988. This inspectiorrinspectioll was performed in accordance with NRC Manual Chapter Procedures 64704, and 93702. Results: Of the areas inspected, one violation was identified in that h Jicensee failed to adhere to fire prevention procedures Paragraph 2.e.). ( (. III.9-4

Revision 8 April 1992 DETAILS

1. Persons Contacted Commonwealth Edison Company (CECo)

    +E. Eenig~nburg, Station Manager
    +8.
    +B. Barth, Technical Staff Engineer
    +M. Dillon, Fire Mijrshal Marshal
    +K. Peterman, Regulatory Assurance Supervisor U. S. Nuclear Regulatory Commission (NRC)
    +S. DuPont, Senior Resident Inspector
    +Denotes those participating in the tel~con telecon exit meeting on February 8, 1989.

2. June 4, 1988 Drywell Expansion Gap Fire

a. Apparent Origin of the Fire At 0600 hours on June 4, 1988, with Unit 3 shutdown and in the*

the refueling mode, an air arc cutting activity on the dryw~11 dryw~ll flued head anchor by contractor personnel was

                                               ~Ias in progress in the Unit 33

( Reactor Water Cleanup (R~JCU) (R.JCU) pipeway. During this time the helper/fire watch observed black smoke, but no flames by the welder's legs l~gs near penetration X-114. Initially, the fire watch thought the welder's rubber boots were on fire but after observing that the boots were not on fire, the contractors unhooked their air hoses and climbed up to the adjacent landing to obtain the fire extinguisher. At approximately 0612, two smoke alarms were received in the control room identifying smoke in the Unit 3 Transversing Incore Probe (TIP) room. At 0615, the control room was notified and at 0619 a second call to the shift engineer was placed.

b. Initial Response and Extinguishment Activities As previously indicated, the welder climbed to the adjacent landing and obtained a fire extinguisher (which had been placed prior to the start of the welding activity on the drywell flued head anchor). When the welder returned to penetration X-1l4, he discharged the entire contents of the dry chemical extinguisher into the penetration sleeve.

At approximately 0630, the fire brigade reported that the fire appeared "out," however, the fire brigade personnel discharged a second dry chemical extinguisher into the penetration sleeve. 2 III.9-5 III .9-5

Revision 8 April 1992 ( At approximately 0645, the day shift foreman arrived for duty. At 0705, the day shift foreman, while enrout~ to the Unit 3 RWCU area (fire scene) noticed a haze on the ground floor of the Unit 3 Reactor Building. At 0720, an alarm was received at the Center Desk which indicated smoke above Unit 3, Reactor Building, East Accumulator Bank. The day shift foreman returned to the Unit 3 Ground Floor and recognized that the symptoms were similar to those which occurred during the 1986 Unit 33 drywell liner fire. The day shift foreman telephoned the shift engineer and recommended sounding the fire Siren, siren, evacuating the Reactor Building and informing the fire brigade to spray water into the penetration sleeve.

c. Extinguishment of the Fire At 0730, control room personnel sounded the fire alarm and announced over the PA system for personnel to evacuate the Reactor Building.

A ftcr the aarrival After rr; va 1 of the fi re bri fire gade aatt the fi brigade re scene, fire scene,a.a wa 1kdown walkdown and size-up of the fire was performed. At approximately 0745 the fire brigade applied water to the X-114 penetration. The water was applied to the X-114 penetration until 0800. At 0940 the Fire Marshal, Mechanical Maintenance Foreman and the Rad Chern Technician drywell and determined, by use of a heat sun, entered the drywe11 gun, that no hot spots existed in the drywel1drywell liner. The licensee estimates (' ( ttlat that approximately 500 gallons of water were used to cool and extinguish the drywe1l drywell liner fire.

d. Licensee's Followup Actions licensee's Actior;s to the 1988 Drywell Gap Fire o

The station manager issued a welding and cutting stop work order on June 4, 1988 at approximately 0800. The release of the welding and cutting stop work order would be allowed after a Projects and Construction (PACS) walkdown with subsequent Fire Marshal or designee approval prior to work resumption. o Daily station overview of construction jobs would be provided for the r~mainder of the outage. o

                    'The licensee photographed und video taped the affected area on June 4, 1988.

o On June 6, 1988, a boroscopic examination of the penetration was performed which wh i ch revealed no discernable revea 1ed 110 di scernab 1e damage but identified i denti fi ed debris in the annulus. An attempt was made to remove the debris. A similar examination of the annulus could not be performed due to equipment limitations. o The fire proof wrapping utilized on penetration X-113 was quarantined area for further inspection. removed and stored in a quarantilled ( ... 3 II 1. 9-6 III.9-6

Revision 8 April 1992 c*( c The station manager and the site PACS superintendent conducted meetings on June 6, 1988, to discuss the drywell gap fire event with all craft personnel and to emphasize the need to adhere to the station procedures. o The drywel1 drywell sand pocket drains were checked for accumulation of water leakage on June 4, 1988. The licensee indicated only minor dripping was present from the sand pocket drains. e,

e. Cutting and Welding Procedure Following the January 1986 drywell gap fire, the licensee upgraded cutting and welding Procedure DAP 3-2 to include the following statement.

liThe "ThE exterior steel skin of both the Unit 2 and Unit 3 drywel1s drywells are covered with a polyurethane foam used during initial construction activities. Although procured as self-extinguishing, the foam h~shas previously been ignited through contact with hot slag from cutting operations on a drywell penetration (see Reference 3;.3). Exercise caution when working around upenings openings that lead to the exterior drywell skin. skin."II The procedure also indicated that when employing a process that geflf'ratl;!s generates sparks or slag (cutting, brazing, grinding, etc.) above grating decks, or near floor or wall openings, the deck or op~ning gloating d~cks, optming ( operation shall be covered with suitable noncombustible below the op~ration noncomuustible material. Care shall be taken not to direct the slag stream from the cutting operation through nearby openings. th~ During the cutting arid welding operation that was being performed on June 4, 1988, the contractor did not provide a suitable noncombustible cover for unprotected penetration X-114 which was located only'a only "a few feet away_away. The failure to protect the opening in penetration X-114 X-1l4 during dul"ing cutting or welding activities was contrary to the licensee's approved fire protl;!ction/prevention protection/prevention administrative procedure anel and is considered 0Q violation (249/89004-01(DRS)). As part of the licensee's corrective action, the licensee has revised the cutting and welding procedure to require an initial inspection by the station Fire Marshal or designee prior to the start of any cutting and welding activity. This is in addition to the area being inspected by the work group supervisor. The inspector activities, all informed the licensee that prior to welding or cutting activities. drywell penetrations within 35 feet should be packed and then covered material. The inspector also requested with suitable noncombustible materia1. that the fire watch illspect inspect the outer covering of the noncombustible materials to ensure that rips, tears and/or openings in the outer covering are repaired should these conditions exist. c ( 4 III.9-7

Revision 8 Apri 1 1992 ( f. Evaluation of the 1988 Drywell Gap Fire JI On July 20,

20. 1986, defueled. an air arc 1986. with Unit 3 shutdown and defueled, cutting activity on containment pipe penetration No. X-113 resulted in a fire in the Unit 3 Drywell Expansion Gap. The licensee was requested to address several concerns as presented in NRC Inspection Report No. 50-249/86006{DRS).

50-249/86006(DRS). During this inspection, inspection. the licensee was requested to readdress those concerns described in the 1986 NRC Inspecti~n Inspection Report. The licensee provided the inspector with the Sargent and Lundy June 4, 4. 1988 Fire Report that indicated the fire occurred in the same location as the January 1986 event and therefor~, therefore. eValuation of the effects of the 1986 fire were used as a basis the evaluation for the current assessment. The report indicated that the 1986 fire burned for a much longer period of time and involved a much wider burn~d area and thus the analysis presented in the 1986 report bounded al~ any effects that resulted from the 1988 fire. The report readdressed each of the concerns as presented in the licensee's original response dated May 6, Farrar. CECa,

6. 1986 from D. Farrar, CECo. to J. Keppler, NRC.

Based on the premise that NRR accepted the licensee's 1986 response and that this fire was bounded by the original 1986 flre fire.* .ri!!

                                                                            ~

II

                                                                                          \1 licensee has addressed the concerns for the 1988 drrwell gap fire.
 ~

Meeting Exit Me~ting On Februal~Y Februal'Y 8, 1989. a conference cal1

8. 1989, call was held "lith

                                                          ~Iith the inspector and the licerlsee's licensee's representativE:!s.

representatives. The inspector discussed the 1likelyfke1.v ( content of this report and the licensee did not indicate that ar~ any information discussed during the inspection could be considered proprietary in nature. 5 I. 9-8 III.9-8 II

Tab 10

         .......                                     UNITED STATES NUCLEAR REGULATORY COMMISSION               APR 2 fI HEC' HEL'OD REGION REGION III 711 ROOSEVELT ROAD                  Revision 8 c.***                                          GLEN ELLYN, ILLINOIS 50' II 50137            April 1992 APR 141989 i 4 1989 Docket No. 50-237 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:

This refers to the special safety inspection conducted by Mr. J. Holmes of this office on March 16-28, 1989, of activities at Dresden Nuclear Power Station, Units 2 and 3, authorized by NRC Operating Licenses No. DPR-19 and No. DPR-25, and to the discussion of our findings with Mr. E. D. Eenigenburg at the conclusion of the inspection. The inspection was conducted to review allegations regardin$ regardin~ unsealed openings inside conduits in fire walls and the use of polyurethane 1n ln fire wails. walls. The enclosed copy of our inspection report identifies areas examined during ((...... the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. No yiolatjons of NRC requirements were identified during the course of this inspection. In accordance with 10 CFR 2.790 of the Commissionls Commission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC Public Document Room. We will gladly discuss any questions you have concerning this inspection. Sincerely,

                                                             /Z..ff: ~/~
                                                             /Z.,)Y;       ~/42.

R. W. Cooper, II, Chief Engineering Branch

Enclosure:

Inspection Reports No. 50-237/89008(DRS)i 50-237/B9008(DRS); No. 50-249/89009(DRS) See Attached Distribution .( III.IO-l

Revision 8 Apri1 1992 April ( Commonwealth Edison Company 2 141989 APR 1" 1989 Distribution cc w/encTosure: w/enclosure: T. Kovach, Nuclear licensing Licensing Manager E. D. Eenigenburg, Plant Manager DCD/DCB (RIDS) Licensing Fee Management Branch licensing Resident Inspector, RIll Richard Hubbard J. W. McCaffrey, Chief, Public Utilities Division ( II I.lO-2 111.10-2

Revision 8 April 1992 U.S. NUCLEAR REGULATORY COMMISSION REGION III Reports No. 50-237/89008(DRS); 50-249/89009(ORS) 50-249/89009(DRS) Docket Nos. 50-237; 50-249 Licenses No. DPR-19; DPR-25 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago. Chicago, IL 60690 Facility Name: Dresden Nuclear Power Station, Units 2 and 3 Inspection At: Morris, IL 60450 Inspection Conducted: March 16-28, 1989 Inspector: ~~~

               ~~J;~~

It\.~

                "'i\.~

Approved By: R. N. Gardner, Chief 1-(3-89 Plant System Section Date Inspection Summary esu ts: NQNo yiglat1gns Yiglatigns were identified. The inspection concluded that the one oDe allegation allegatjgn was sybstantiated. hpwever, hgwever, no violations yiolations of NRC regylatgry. regulatory. requirements reqUirements were identifjed. jdentifjed. ( III .10-3

Revision 88 Apri 11 1992 DETAILS

1. Persons Contacted Commonwealth Edison (CECo)

Eenigenburg. Station f'anager

     *E. D. Eenigenburg,            11anager
     *K. Deck. Quality Assurance Deck,
     *M.
     *H. Dillon. Fire Marshal Dillon,
     *R. Falbo. Regulatory Assurance Falbo,
     *L. Kline. Regulatory Assurance Kline,
     *D.
     *0. Roberts. Fire Protection Engineer Roberts, Sargent and Lundy (S&L)
     *Brian Barth, Barth. Technical Staff Engineer
     *Denotes those attending   ~1arch  17. 1989 exit meeting.

17, L. Allegation RIII-88-A-180 On December 16, 1989, 1989. Region III received an allegation that there were firewal1s at the Dresden Nuclear unsealed openings inside conduits in the firewalls Power Station. In addition, the alleger indicated that pyrocrete masked ( the presence of polyurethane in the f1rewal1s. firewalls. Each of the individual concerns are addressed below: Concern 1: The firewalls at Dresden contain unsealed openings inside conduit penetrations. This allegation was general for all 1l jt D~~O~' firewalls and no specific areas were received from the alleger. Jrr,". jr~".

                                                                     .                VI
                                                                                      'II    ..
                                                                                             "IJ NRC Review:    The requirement for sealing conduits wh1ch which penetrate       ~.~c~
                                                                                     ~.~t~

firewalls is contained in the licensee's updated Fire F1re ~-.ra

                                                                                     ~~~

Analysis, Section 5.0. Hazards Analysis. 5.0, entitled "Guidelines

                                                                  -Guidelines of      fo"'" *
                                                                                     ;'olWl Appendix A to APCSB 9.5-1". This document 1ndicates indicates that        ~S condu1t and piping should be sealed or closed to provide conduit                                               prov1de~f'      Jr' a fire resistance rating at least equal to that of the barrier. In discussions with the cognizant cogn1zant NRR reviewer on Harch 28.

28, 1989. the inspector determined that the document only required the licensee to install seals between firewalls f1rewalls and conduits which penetrate the firewall. The inspector discussed this matter with licensee personnel t1arshal. The licensee was aware of the including the Fire 11arshal. condyit seal requirements And indicated that seo1sseals hid been installed between firewalls and all conduits at the points where the conduits enter or exit the firewalls. f1rewal's. During this inspection, the inspector reviewed a sample of the licensee's completed surveillances of conduits which 2 III.I0-4 IILI0-4

Revision 8 April 1992 ( penetrate firewa11s. These surveillances did not identify. an instances of im ro er conduit seal installations and were eterm ne to The inspector also selected several representative firewalls for walkdown wa1kdown to determine whether the licensee was complying with the fire seal requirements. Durin Durin¥ the walkdown, the r seals were inspector determined that all required f1re fire insta 11ed. lled.

Conclusion:

This allegation concerned a perceived need to install seals inside conduit openings for all conduits which penetrate firewalls at the Dresden Station. However, since the licensee was not required to seal these conduit openings and since the inspector determined that the licensee was installing all required fire seals, this allegation was not substantiated. Concern 2: Pyrocrete covers polyurethane in firewalls. firewa11s. NRC Review: The licensee's Fire Protection Program includes the Guidelines of Appendix A to APCSB 9.5-1. This document requires the licensee to provide 3 hour rated floors. floors, walls, and ceilings enclosing the separate fire areas identified in the Safe Shutdown Analysis. Deviations in the fire barriers were justified in Exemption Requests and have been reviewed and accepted as identified in the NRC Safety Evaluation Report dated January 5, 1989. Based on review of the pertinent documents, the inspector determined olvurethane the licensee was re uired to remove the oivurethane from the fire walls grpr demonstrate tt at t e ~o yuret ane in the firewall did not affect the 3 hour ra 1ngin9 of the fire barrier. During this inspection, the licensee indicated that polyurethane was commonly installed in firewalls in the past to prevent air leaks. The plant had preViOUSly preViously realized the potential hazard of utilizing polyurethane firewa'ls and had hired outside contractors to remeve in f1rewalJs remove the polyurethane from the f1rewal1s. firewalls. The licensee indicated

               . to the inspector that the majority of the polyurethane had been removed. However, the lfcensee licensee indicated that polyurethane covered by pyrocrete remained in a firewall between the turbine building and Unit 2 on elevation 545'-6" at coordinates Hand 43 through 44. The licensee 545'                   had elected to cover the polyurethane with pyrocrete due to high radiation exposure and the possibility of breaching secondary containment.

The licensee also indicated to the inspector that polyurethane without a pyrocrete covering was located .(: II 1.10-5

Revision 8 April 1992 ( around a 12 inch pipe penetration located between the. the Units 2 and 3 reactor building on elevation 545 -6 w at 545'-6" 1 coordinates 44 and H through J. The licensee indicated that due to radiation concerns the polyurethane had not yet been removed. I The licensee also indicated to the inspector that an outside fire protection engineering firm has conducted two fire barrier surveillances which did not identify other instances of installed polyurethane.

Conclusion:

This all~gat1on all~gation was substantiated in that pyrocrete does cover po yurethane installed in one plant location and polyurethane without a pyrocrete covering exists in another location. However, However. prior to the allegation the licensee removed and replaced the major1ty majority of the polyurethane with an appropriate f1re fire rated barrier or seal. Where the licensee ynable to remove tbe l1censee was gnable the polyurethane due to high radiation and concerns regarding the breaching of secondar~ secondar!' containment, containment. the licensee was performing the required assessment of the effect of the

                   ¥olyurethane on the 3 hour ratingratinq of the fire barrier.

herefore no violations or deviations of NRC requirements herefore! iden t lfled. were tden ,f1ed.

3. Exit Interview w*

The inspector met with licensee representatives on March 28, 28. 1989. The inspector discussed the likely content of this report and the licensee did not indicate that any information discussed during the inspection could be considered propr1etary1n proprietary in nature. 4 rrr.l0-6 rrF.I0-6

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III.IO-7 III.I0-7 ,"

Revision 8 April 1992 ALLEGATION'ACTION ALLEGATION' ACTION PLAN ALLEGATION NO. RIII-88-A-lSO RIII-88-A-180 Licensee: Commonwealth Edlson Edison Company Docket/License No: 50-237 and 50-249/0PR-19 and DPR-25 Assigned Division: DRS Attached Pertinent Documents: Allegation Do~umentation I. Division Action Allegations regarding unsealed conduit penetrations and the adequacy of of fire resistive material covering polyurethane in fire walls at the Dresden Station were received by Mr. Jeff Holmes on December 16, 1988. All ega ti on No.1 Allegation X" cfR"nllol *GJAC,AC0't6

r"'CfRTIII.J *eJAt, MD'Q

( .. a11eger alleged that there are unsealed conduit penetrationsLthru The alleger the fire wall at Dresden. In addition, the alleger a11eger alleged that pyrocrete the polyurethane in the fire walls. covers t~e NRC Action

1. Request the licensee to provide documentation addressing conduit and pyrocrete penetration fire barrier commitments/requirements.

2. Review plant surveillance/maintenance procedures that cover conduit and pyrocrete penetration fire barrier configurations.

3. Interview the plant manager and the fire marshal regarding knowledge of concerns regarding the unsealed conduit and pyrocrete over the polyurethane in fire walls. r I '"

I

                                                                 ~1\..ae
                                                                 ~"'" ""t  "'t Al\~,;""

Al\~.,.....

                                                               ~
                                                               ~

4. Conduct an inplant inp1ant review of selected installed conduit and pyrocrete fire barrier configurations.

                                                 .,..n.~
                                       ~
                                                 .,....,.~

A. .Prepared by:

                     ,Prepared              seph M. UHe                                            3-Cf-'&~

3-Cf-~tt ec:hnic:al S~ff echnical Date bate

                                         ~f\~

R-..uf\~ B. Reviewed by: Ronald N. Gardner J-'-i9

                                         &t;o~2ief.JP
                                         &tio~2ief~                                                  Date bate C. Approved by:        R1Chr.~

R'C:hrr~ Branch Chief

      ~--

II I. 10-8 III.

ReviSion Revision 8 Apri April1 1992 Allegation Action Plan 2 Allegation No. RIII-88-A-180 II. Allegation Review Board Action Allegation Review Board Membership tt I7 Approved As Is t I7 Approved with Modifications as Documented in Plan. tt 7 Disapproved for Following Reasons: Allegation Review Board Chairman Date ( .(" III. II 10-9 1.10-9

Revision 8 April 1992 ALLEGATION/PERIPHERAL ISSUE ACTION PLAN Concerns and any peripheral issues associated with a concern should be documented on a separate page. Each concern and peripheral issue.issue, if any, should be documented in the followup report as is stated in this plan. If there are several concerns in one area, one page can be used. Otherwise. Otherwise, a separate page should be used for each concern. I/ 7 Concern No. /I 7 Peripheral Issues Associated with Concern No. I. Action Evaluation: The following method of resolution is recommended (circie): (circle): A. Send to Licensee Requesting Response in Days with RIll Followup* B. Priority RIll Followup ----- C. Followup During Routine Inspection Within 60 Days n: 0: Followup with Assistance from 01 E. No Action - Outside NRC's Charter (describe basis below) F. No Action - Without Merit (describe basis below) G. Refer to H. Other (specify)

• If the proposal is to send to the licensee, the Action Plan should describe the general areas we expect the licensee to address.

II. Inspector's Actions: The following areas at a minimum will be reviewed during the inspection into the above mentioned concern and/or peripheral issue. A. Objective B. Methods

1. Persons to be contacted:

a. Plant Manager Station *Fire b*. Statfon Fire Marshal

c. Other personnel as necessary

2. Documents and/or activities to be reviewed:

I( '( IIl.lD-ID IILIO-10

Revision 8 April 1992 ( I~ Allegation/Peripheral Issue 2 Action Plan

            .)    3. Time period to be covered:

4. Locations/specific areas to visit:

5. Other areas (specify):

Allegation No. RIII-88-A-180 (- iC. II 1.10*11 III. 10-11

Tab 11 81013 ig1"1olz UNITED STATES NUCLEAR REGULATORY ,COMMISSION Revision 8 REGION III April 1992 1" ROOSEVEL.T ROAD 7" GL.EN EL.LYN. ELLYN, ILLINOIS 'Ofl7

                                                              '01l7 No, 50-237 Docket No.

Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen: This refers to the routine'safety inspection conducted by Mr. J. Holmes of this office on April 3-7 and May 24, 1989, of activities at Dresden Nuclear Power Station, Units 2 and 3, authorized by NRC Operating Licenses No. DPR-19 DPR-25, and to the discussion of our findings with Mr. C. Schroeder and No. OPR-25, at the conclusion of the inspection. The purpose of this inspection was to review the implementation of the licensee's fire protection program. The enclosed copy of our inspection report identifies areas examined during i,nspection. Within these areas, the inspection consisted of a selective the ;*nspection. examination of procedures and representative records, observations, and ( interviews with personnel. No violations of NRC requirements were identified during the course of this inspection. In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC Public Document Room. We will gladly discuss any questions you have concerning this inspection. Sincerely, R. W. Cooper, Chief Engineering Branch Enc 1osur.e-:-- -Inspection Enclosure{--- -Inspect i on "Reports N~. SO-237/89013(DRS); . Ho. No. 50-249/89012(ORS) 50-249/89012(DRS) See Attached Distribution

'((

IILl1-1

                                              !ILl!-!

Revision 8 Apri 11 1992 c( Commonwealth Edison Company 2 Jur\l JUf\! 99 1989 Distribution cc w/enclosure: T. Kovach, Nuclear Licensing Manager E. D. Eenigenburg, Station Manager DCD/DCB (RIDS) DCo/DCB Licensing Fee Management Branch licensing RILL Resident Inspector, RIll Richard Hubbard  : J. W. McCaffrey, Chief, Public Utilities Division A. Datta, NMSS/IMSB C. McCracken, NRR/ECEB A. Krasopoulos, RI/oRS RI/DRS RII/DRS G. Wiseman, RII/oRS A. Singh, RIV/DRS C. Ramsey, RV/DRS (

(

i( III. 11-2

Revision 8 April Apri 1 1992 ( u.s. NUCLEAR REGULATORY COMMISSION REGION IIII II Reports No. 50-237/89013(DRS); 50-249/890l2(DRS) 50-249/89012(DRS) Docket Nos. 50-237; 50-249 Licenses No. DPR-19; DPR-25 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Dresden Nuclear'Power Nuclear Power Station, Units 2 and.3* Inspection At: Morris, Illinois Inspection Conducted: April 3-7 and May 24, 1989 Jm ~ (\flAM-Inspector: 9sIJ, Holmes S,1989 JUNE.. S,I989 Date

                   ~\.~

Approved By: Ronald N. Gardner, Chief Plant System Section Date

'(

Inspection Summary Ins~ection Ins ection on A~ril Aril 3-7 and t*1ay Hay 24, 1989 (Reports No. 50-237/890l3(DRS); 50-237/89013(DRS); 50- 249/8901Z(DR 49!8901z(DR ))II Areas Inspected: Routine, unannounced safety inspection conducted to review the implementation of the licensee's fire protection program including a fo110wup of licensee action on previous inspection findings. This inspection followup was conducted in accordance with inspection procedures 64704 and 92701. Results: Of the areas inspected, no violations were identified. One unresolved item and one open item was identified. The unresolved item concerned the need for the licensee to verify the shelf life of two types of fire fighting foam concentrate utilized for flammable liquid fires (Paragraph 3.e). The open item concerned ~hethe need to develop a six month functional test for linear detection (Paragraph 3.a). In addition, the inspector identified and discussed with the licensee the need to develop pre-fire plans for areas such as the main power unit transformer area and hydrogen storage tank areas (Paragraph 3.e) 3.c) and a need to develop a maintenance program for the hydrogen storage equipment

                ~Paragraph 3.d).

and piping (Paragraph III. 11-3 III.I1-3

Revision 8 Apri 11 1992 DETAILS

1. Persons Contacted Commonwealth Edison Company (CECo)

       *C.
       *c. Schroeder, Services Superintendent
       *c.
       *C. Allen, Performance Improvement Supervisor
       #D.
       #0. Barnett, Senior Quality Assurance Supervisor
       *R. Black, Assistant Fire Marshal
       *K. Deck, Quality Assurance Engineer
    *,#M. Dillon, Fire Marshal
       *R. Falbo, Regulatory Assurance Assistant
    *,#L. Gerner, Production Superintendent
       #T. Lewis, Regulatory Assurance
    *,#K.
    * ,#K. Peterman, Regulatory Assurance Supervis,or
       *0.
       *D. Roberts, Fire Protection Engineer
       *J. Silady, Nuclear Licensing Administrator
       *E. Skoron, Technical Staff Engineer
       *S.
       *5. Stiles, Training Supervisor Sargent and Lundy (S&L)
    *Brian Barth, Technical Staff Engineer Nuclear Regulatory Commission (NRC)

( #5. DuPont, Senior Resident Inspector

    *Denotes those attending the April 7, 1989 exit meeting.
    #Denotes those participating in the May 26, 1989 exit meeting (telecon).

2. Action on Previous Inspection Findings (92701)

a. (Closed) Unresolved Item (237/85033-03(ORS);

(237/85033-03(DRS); 249/85029-03(ORS>>: 249/85029-03(DRS>>: surveillance procedure (SP 84-6-39) failed to The licensee's survelllance incorporate appropriate test requirements to demonstrate that the sprinkler system was operable in accordance with NFPA 13 in that the procedure did not require flow from the two inch drain valve for wet pipe or dry pipe sprinkler systems. During the week of January 2, 1989, 19B9, the licensee provided the inspector with an internal memorandum to E. Eenigenburg from R. Black that stated the following: liThe test flow discharges will be handled through pre-installed drain piping to the station storm drains. On systems where pre-installed piping'is nonexistent, a 1-1/2 inch maintenance hose will be drain piping-is used. We believe a 1-1/2 inch hose provides adequate flow at our system pressure (125 psi) to ensure water supplies and connections are in order. 2 III .11-4 III. 11-4

Revision 8 April 1992 The reactor building fire hose risers will be tested by connecting a ( 1-1/2 inch maintenance hose to the upper most hose station locations and test flowed to station storm drains. It is our plan to complete the writing and get approvals of the surveillance by April 30, 1989 and perform the surveillance for the first time by June 30, 1989. Additional surveillances will be done on an annual basis.~ basis." Based on the licensee's commitment to conduct the drain test, this item is considered closed.

b. (Closed) Open Item (237/85033-06(DRS); 249/85029-06(DRS>>: The licensee was requested to provide appropriate ac~eptance acceptance criteria for filling breathing air supply cylinders.

The licensee provided the inspector with. the procedure titled "Use IIUse of the Cascade Recharging System for Filling Self Contained Breathing Apparatus Bottles," DRP 1310-11, Revision O. The inspector reviewed the procedure and no discrepancies were identified. Based on the inspector's review of this procedure, this item is considered closed.

c. 249/85029-07(DRS>>: The (Closed) Open Item (237/85033-07(DRS); 249/85029-07(DRS):

deficiencies on the maln inspectors observed deflc1encles main carbon dl0x;de dioxide storage tank located on the first floor of the turbine building. The included the following: deficiencies inclUded ( (1) The access door to the tank compressor motor was missing. (2) The glass cover to the tanks mercoid switch located inside the access door was missing. During the Appendix R inspection that ended in December 1988, the inspector had observed that the licensee had taken corrective actions to replace the access door and glass cover to the mercoid switch. Based on the licensee's actions, this item is closed.

d. (Closed) Open Item (237/88010-05(DRS); 249/88012-05(DRS>>: The pre-dlscharge pre-discharge alarm for the diese1 room carbon dloxlde dioxide system is only an audible alarm. The licensee was requested to verify that the audible alarm is sufficient to warn personnel that may be in the area with the diesel operating.

The licensee provided the inspector with an internal document dated January 31, 1989, from R. Black, Assistant Fire Marshal to-the Fire Protection File. The document indicates that on December 23, 1988, during the performance of DFPP 4145-1, KCardox

                                                                  *Cardox System Semiannual Maintenance Test," which was modified for a running diesel, the audible alarm was heard by the occupants in the room.

3 111.11-5

Revision 8 April 1992 Based on the licensee 1s actions, this item ;s licensee's is considered closed. <.( e. 249/88012-06(DRS)): The (Closed) Violation (237/88010-06(DRS); 249/aa012-06(DRS>>: their approved fire licensee failed to meet the requirements of the1r protection program by permitting the storage of twenty 55-gallon drums of lube oil in a safety-related area. The licensee provided the inspector with the "Control QControl of Transient Combustibles" procedure that has been revised to include the Combustibles>> following: For medium fire loading of an area (5-25 gallons for flammable liquid, 55-120 gallons for combustible liquid) and high fire loading of an area (25 gallon for flammable liquid, 120-240 gallons for combustible liquid) a transient combustible permit signed by the fire marshal or his designee is required. In addition, compensatory measures are combustibl,es into the plant. The fire required prior to introducing combustibles marshal is also required by the procedure to review the fire hazards analysis for the fire area of concern. The basis for acceptance of a high fire load includes the consideration of equipment and combustibles presently in the area and any suppression or detection systems. Compensatory measures are then established. Based on the review of the updated procedures, this item is considered closed.

3. Routine Fire Protection Program The Dresden fire protection program utilizes the defense-in-depth concept

( against hostile fires to ensure that safe shutdown capability is not impaired. The Dresden fire protection program philosophy of defense-in-depth consists of:

a. Fire Prevention

b. Detection and Suppression

c. Mitigation of Fire Damage reviewed, on a sample basis, the licensee's administrative The inspector reViewed, procedures and fire protection surveillances. The inspector also walked down several fire protection systems. The results of the inspector's review are as follows:

a. Fire Protection System Surveillances licensee's fire protection program requires that the licensee The licensee1s test fire protection equipment and systems that are included in regularly scheduled station operating surveillance procedures. The inspector selected a sample of the licensee's completed surveillance procedures for review. Our; pr.ocedures During ng the review, the inspector determined the following:

( 1) (1) Weekly Unit 1 Diesel Fire Pump Operability Surveillance The licensee's Unit 1 diesel fire pump weekly operability 4123-1,t Revision a, surveillance, DFPP 4l23-l 8, includes the ( 4 III. 11-6 III.1l-6

Revision 8 April 1992 verification that the fire pump batteries are provided ( \~ith proper electrolyte level and specific gravity. with In addition, the procedure verifies that the battery charger is operating and that proper oil level is provided in the engine case and right angle gear drive. The inspector reviewed the Unit 1 diesel fire pump weekly operability data sheet dated March 28, 1989, and found the results to be satisfactory. (2) Quarterly Auxiliary Electric Equipment Room Halon Damper Test The auxiliary electric equipment room halon damper test, DFPP 4195-3, Rev~sion Revision 2, verifies that dampers required to close, prior to discharge are operating as designed. I The auxiliary electric equipment room halon damper test results t9 be satisfactory. dated March 14, 1989, were found tq (3) 18 Month Operating Fire Stop/Break Surveillance The operating fire stop/break surveillance, OFPP DFPP 4175-2, Revision 5, verifies by visual observation that the fire stop/break is intact. The fire stop/break surveillance dated February 29, 1988, was found to be satisfactory. ( (4) (4 ) Annual Auxiliar AUXi1iar{ Electrical Equipment Room Fire Resistive t Structural Stee and Cable Coating Surveillance The auxiliary electrical equipment room fire resistive DFPP 4175-4, structural steel and cable coating surveillance, OFPP requires visual verification of the integrity of the Revision 1, r~quires auxiliary electrical equipment room fire resistive structural steel and cable coating. In the surveillance dated January 18, 1989, the licensee identified several areas where structural steel fire proofing was found degraded and initiated a OVR. DVR. Based on the licensee1s licensee's actions to correct the degraded fire proofing, the surveillance was found to be acceptable. (5) Monthly Fire System Yard Loop Inspection The fire system yard loop monthly inspection, DAP 11-2, Revision 15, checks equipment such as fire hydrants, hose houses, fire hose reels, fire main valves and other fire equipment. The fire system yard loop monthly inspection dated March 10, 1989, was reviewed and found to be satisfactory. 5 111.11-7 III.!1-7

Revision 8 Apri 11 1992 Apr; (6) Monthly Fire System Inspection For Unit 2 The Unit 2 monthly fire system inspection, OFPP DFPP 4114-2, revision 9, visually inspects equipment such as hose reels, fire main valves, fire equipment carts, carbon dioxide systems and other fire equipment. The Unit 2 monthly inspection results dated March 27, 1989, were reviewed and found to be satisfactory. {7} (7) Six Month Fire Detection Test licensee's smoke detector semiannual maintenance test, The 1icensee"s DFPP 4185-2, rev~sion OFPP rev'ision 6, verifies the response of the fire detection system. The inspector requested the last sjx month channel functional tests conducted for Unit 2 fire zones 1.1.2.1 (e1ev. 476'-6"), 1.1.2.2 (elev. (e1ev. 517'-6"), 1.1.2.3 (e1ev. 545'-6"), and Unit 3 fire zones 1.1.1.1 (e1ev. 476'-6"), 1.1.1.2 (elev. 5171-6"), 517'-6"), and 1.1.1.3 (elev. (e1ev. 545-6 545-6"). Fire zones 1.1.2.1 and 1.1.1.1 are 11

                                        ).

provided with linear thermal detection. The inspector requested the six month functional test for these areas. However, the licensee had not yet developed a six month channel functional test for the linear thermal detectors in these areas. In discussion with the licensee, the licensee indicated that a recent audit had identified the same concern and that the ( surveillance was in th.e the process of being developed. The licensee indicated to the inspector that the surveillance will be completed by July 21, 1989. This is considered an open item (237/89013-01(DRS); (237/89013-0l{DRS); 249/89012-01(DRS>>) 249/890l2-01{DRS>> pending NRC's review of the surveillance procedure. The inspector reviewed the last six month channel functional test dated January 1989 for Fire Zones 1.1.2.2, 1.1.2.3, 1.1.1.2, and 1.1.1.3. The functional test performed did in some cases identify minor problems. The licensee wrote a work request to address those concerns. Based on review of the surveillance test results and the 1icensee licensee's 's actions, the surveillance was found to be acceptable.

b. Personnel Required for Safe Shutdown and Fire Fighting Activities In the event of a disabling fire which requires evacuation of the Unit 2/3 Common Control Room when both units are operating, it would be necessary to provide sufficient personnel to shutdown the operating reactors and provide manual fire fighting capabilities.

(1) (l) Safe Shutdown Personnel The licensee has developed alternative safe shutdown procedure EPIP 200-20, titled "Control Room Evacuation/Safe Shutdown," ReVision Revision 6, dated February 1989. 6 III. 11-8 II-8

Revision 8 April 1992 The licensee's staff required to implement the alternative. safe contro1'-'room shutdown procedure requiring the evacuation of the control~croom is as follows: . Shift Engineer (SE) Station Control Room Engineer (SCRE) Unit 2 Shift Foreman (SF) Unit 3 Shift Foreman (SF) Engineer Assistant (EA) Center Desk Nuclear Station Operator (NSO) Unit 2 Nuclear Station Operator (NSO) Unit 3 Nuclear Station Operator (NSO) Unit 1 Levell Operator/Equipment attendant. Unit 2 Level 1 Operator/Equipment attendant Unit 3 Level 1 Operator/Equipment attendant Utility Levell Operator/Equipment attendant (2) Fire Brigade Personnel The licensee also has developed "Fire Fighting" procedure EPIP 200-4, Revision 5, dated December 1987, which the organization of the fire fighting brigade and describes t~e delineates the duties of the fire brigade. This procedure indicates that the composition of the fire brigade for all shifts is as follows: Shift Foreman - Fire Chief High Voltage Operator - Fire Fighter Radwaste Roving Operator - Fire Fighter Unit 2/3 Max Recycle Operator - Fire Fighter Rover - Fire Fighter (3) Operations Department Organization The licensee has developed operations department organization procedure DAP 7-1 which identifies the staffing normally required for operating shifts 1, 2, and 3. The inspector verified that the minimum number of personnel for safe shutdown and fire fighting was included in the procedure. (4) ConcluSion The inspector requested records to demonstrate that the 12 personnel required to implement the control room evacuation procedure and the 5 personnel required for fire fighting activities were available for three shifts on April 13, 1989, and April 26, 1989. The inspector was provided with copies of the appropriate sections of the shift's engineers and center desk books. 7 III.11-9

Revision 8 April 1992 In cases where names were inadvertently left out of the logs, ( the licensee provided backup documentation to demonstrate that these personnel were available. The inspector verified, based on the licensee's documentation provided, that the appropriate personnel for the 12 positions were available to implement the control room evacuation safe shutdown procedure. The inspector also verified that in addition to the staff required for safe shutdown, the licensee provided a 5 member fire brigade consisting of a fire chief and four fire fighters. Based on the inspector's review of the licensee's documentation, the inspector determined that on April 13 and 26, 1989 (all shifts), the licensee provided sufficient personnel for safely shutting do~n*the reactors and to support any req"uired required fire fighting activities.

c. Pre-Fire Plans The licensee has developed pre-fire plans for fire in safety-related areas as described in the fire hazard analysis. The pre-fire plans indicated important parameters for each fire area such as access, hazards, fire protection equipment, ventilation, communications, exposures (safety-related equipment), construction, guidelines for attack, etc. In addition, the licensee has provided a schematic for each fire area which also indicates location of fire fighting equip-ment, communication, access points, etc. It appears that the licensee has developed good fire pre-plans for fighting fires in safety-related areas within the plant. However, the inspector identified that pre-fire plans did not exist for areas such as the hydrogen storage area and main power transformers for Unit 2 and Unit 3. Both of these are non safety-related areas.

The licensee was informed that it would be prudent to develop pre-fire plans for all areas with high combustible loading and/or where special precautions may be required to prevent injury to fire fighting personnel or damage to the plant. The licensee acknowledged the inspector's concern and indicated that plant areas not addressed in the fire hazard analysis such as the main power unit transformers and hydrogen storage areas will be reviewed and pre-fire plans developed by December 31, 1989.

d. Hydrogen Storage The tank farm and the hydrogen injection storage are two areas at the Dresden site that currently store hydrogen for normal plant operation to provide hydrogen cooling for the turbine generator and also to prevent intergranular intergranu1ar stress corrosion cracking in primary piping and equipment. Both of these hydrogen systems are non safety-related.

(ll (1) Tank Farm According to the licensee, the tank farm was installed in ( 1968 and is provided with fifty fixed storage vessels capable ( 8 III. 11-10 II 1.11-10

Revision 8 April Apri 1 1992 of storing 35,000 standard cubic feet (scf) of hydrogen at ( 1250 pounds per square inch (psi). The extra heavy red brass piping from the tank farm to the regulator is pressurized to approximately 1250 psi. After the hydrogen is stepped down by the regulators to a line pressure of 70 psi, the hydrogen then enters into 150 carbon steel pipe. The underground pipe for this system is provided with cathodic protection. The system has been designed for automatic operation and is provided with a high flow supply line trip. This hydrogen system has also been provided with alarms such as gas purity, gas pressure, high flow, low flow, low main bank pressure and hydrogen storage ,reserve bank low pressure. The inspector toured the hydrogen tank farm'and observed that the piping from the relief valves was rusty and was not provided with plastic caps. (2) (2 ) Hydrogen Injection Storage There are two hydrogen supply trailers, each with a total capacity of 125,000 scf. This system has been designed for automatic operation and is provided with trips resulting from reactor scram, low feedwater, low offgas, hydrogen high area alarm, local panel shutdown switch and control room shutdown switch. The piping installed from the hydrogen storage trailers to the plant is 304 stainless steel pipe and is ( provided with cathodic protection. ( The pipe installed inside the building is 300 carbon steel pipe. The inspector toured the hydrogen storage trailers and noted that the pressure regulator cabinet that steps down the pressure from the trailer tanks to the system piping was not securely mounted or protected from trucks that deliver hydrogen. In addition, the inspector observed that the trailers were not provided with chocks to secure the wheels to prevent movement. The lice~see acknowledged the inspector1s inspector's concerns and indicated that the pressure regulator cabinet would be secured, barrier protection for the pressure regulation cabinet would be installed and that chockS chocks would be provided for the Wheels wheels to prevent movement. (3) Conclusion In discussions with the licensee, it was identified that no regular maintenance had been performed on the hydrogen tank farm since it was installed in 1968. The licensee indicated that there is no regular inspection or maintenance for pressure regulators, relief valves, interlocks, etc. For the hydrogen injection storage area which is a relatively new addition, the licensee also indicated that no periodic (

   '(      inspection or maintenance has been established for interlocks, 9

III. II-II III.11-11

Revision 8 April 1992 ( pressure regulators, relief valves, etc. The licensee concurred that an evaluation should be performed for the hydrogen tank farm and hydrogen injection system to develop an appropriate maintenance program.

e. Plant Tour The inspector toured several areas of the Unit 2 and Unit 3 reactor building and turbine building. During this tour, the inspector visually observed several hose stations, extinguishers, sprinkler valves, carbon dioxide valves, emergency lights, and housekeeping.

The inspector observed that the equipment was in an apparently well maintained condition.' Housekeeping, in general, ,was good. The inspector informed the licensee that the placement of reflective tags identifying appropriate switches for Appendix MR""R" safe shutdown equipment (for example, at the 250 vdc bus) would be beneficial to th'at the station is currently the operator. The licensee indicated that assessing the use of reflective tags for identifying appropriate switches for Appendix uRn "R" safe shutdown equipment. The inspector also indicated to the licensee that the overall outside housekeeping needed to be improved. The licensee concurred with the inspector and indicated that housekeeping will be improved in conjunction with the decontamination efforts. During the tour, the inspector also observed the 750 gallons per minute deluge gun (located in the 2/3 cribhouse) which may be used to assist in fighting ( a main power unit transformer or hydrogen tank fire. The deluge gun (monitor nozzle) is provided with mechanical gears which allow the operator to change the nozzle elevation. The inspector identified that it was difficult to change the elevation of the deluge gun. The inspector suggested that the deluge gun be included in a preventive maintenance program. The licensee concurred with the inspector and indicated that the two deluge guns at the plant would be disassembled and inspected. After the results of the inspection are known, long term continuing maintenance will be established. The licensee indicated that the inspection for the deluge gun would be completed by April 30, 1989 *. Also during the tour, the inspector observed that the licensee stored Rockwood 6% foam concentrate (1981) and Ansul AFFF 3% foam concentrate (1981) in five gallon cans in the waste water treatment facility. The licensee maintains approximately 50 gallons of foam concentrate at the waste water treatment facility. The inspector questioned the licensee regarding the shelf life of the foam concentrates. The licensee was not aware of the shelf life of the foam concentrate and the licensee indicated that the foam c. (.- 10 IILll-12 III.1l-12

Revision 8 April 1992 concentrate is not sent out for testing to determine if it will ( perform as intended. The licensee indicated that fire fighting foam concentrate shelf life will be verified and if testing is required, it will be scheduled, or the foam concentrates will be replaced by May 31, 1989. The shelf life of the foam concentrate is considered an unresolved item (237/89013-02(DRS); 249/890l2-02(DRS>> pending review of the licensee's actions. The inspector informed the licensee that it would be prudent to use one type of foam concentrate and that the foam concentrate should be rotated such that the older foam concentrate, if needed, can be used during fire fighting training. The licensee acknowledged and inspector's comments. concurred with the in~pector's

4. Open Items Open items are matters that have been discussed with the licensee, that will be reviewed further by the inspector, arid that involve some action on the part of the NRC, the licensee, or both. Open items disclosed during the inspection are discussed in Paragraph 3.a.

5. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, or items of noncompliance or deviations. An unresolved item disclosed during the inspection is discussed in Paragraph 3.e.

;(

6. Exit Meeting Heeting The inspector met with the licensee representative on April 7, 1989 and also held a conference call with the licensee on May 26, 1989.

The inspector discussed the likely content of this report and the licensee did not indicate that any information discussed during the inspection could be proprietary in nature.

('

 ,(

11 111.11-13 IlI.l1-13

Tab 12 UNITED STATES o/!/? ~ ~/ fA.- t r .. _ .

• f1 .. .I?'

Ce.* NUCLEAR REGULATORY COMMISSION Revision 8 REGION III Apr; 1 1992

                                    '99 "9 ROOSEVELT AOOSEVEL T ROAD               e,f) ....

GLEN ELLYN. ELLYN, ILLINOIS 601 J7

                                                       &0137 cv< /{. f£{iMWIJ! Il~

L-~I6JV ~ 4 :__ JUL 3 1 1989 Docket No. 50-010 Docket No. 50-237 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen: This refers to the routine safety inspection conducted by S. G. Du Pont, K. R. Ridgway, D. E. Hills and D. E. Miller, of this office on May 30 through July 14, 1989, of activities at Dresden Nuclear Power Station, Units 1, 2 and 3 authorized by Operating Licenses No. DPR-02, No. DPR-19 and No. DPR-25 and to the discussion of our findings with Mr. E. Eenigenburg and others at c( . the conclusion of the inspection. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records~ records. observations, and interviews with personnel. During this inspection, certain of your activities appeared to be in violation of NRC requirements, as described in the enclosed Notice. The inspection showed that action had been taken to correct the identified violation and to prevent recurrence. Our understanding of your corrective actions are described in Paragraph 11.b of the enclosed report. Consequently, no reply to the violation is required and we have no further questions regarding this matter at this time. In accordance with 10 CFR 2.790, of the Commissionls Commission's Regulations, a copy of this letter and the enclosure will be placed in the NRC Public Document Room. ( III.l2-1 III.12-1

Revision 8 April1 1992 Apri ( IJUl Commonwealth Edison Company 2 'JUl 31 fi89 We will gladly discuss any questions you have concerning this inspection. Sincerely, tJvSU. ~1/. uJvSu. ~1/

w. D. Sh:~Ch;ef Sh:~Chief Reactor Projects Branch 1

Enclosures:

1. Notice of Violation

2. Inspection Reports 50-010/89002(DRP);

No. 50-010/89002(DRP);. No. 50-237/89017(DRP) and No. 50-249/89016(DRP) cc w/enclosures: T. Kovach, Nuclear Licensing Manager E. D. Eenigenburg, Station Manager DCD/DCB (RIDS) ( Licensing Fee Management Branch Resident Inspector, RIll Richard Hubbard J. W. McCaffrey, Chief, Public Utilities Division ( III. 12-2 111.12-2

Revision 8 April 1992 NOTICE OF VIOLATION As a result of the inspection conducted on May 30, 1989 through July 14, 1989, and in accordance with the General Policy and Procedures for NRC Enforcement Actions~ Actions; (10 CFR Part 2, Appendix C), the following violation was identified: Dresden Technical Specification 6.2.A states that detailed written procedures covering preventative and corrective maintenance operations, which could have an effect on the safety of the facility . . . and. testing and surveillance requirements shall be prepared, approved and adhered to. Contrary to the above, ventilation hatches in the Unit"2 Unit 2 drywell left c in an improper closed position resulting in excessive upper elevation temperatures during Cycle 11 were due to inadequate maintenance and surveillance procedures. This is a Severity Level IV violation (Supplement I). The inspection showed that action had been taken to correct the identified violation and to prevent occurrence. Consequently, no reply to the violation is required and we have no further questions regarding this matter. Dated 1 ( 111.12-3

Revision 8 April 1992 ( U. S. NUCLEAR REGULATORY COMMISSION I II REGION III Reports No. 50-010/89002(DRP)j 50-010/89002(DRP); 50-237/89017(DRP); 50-249/89016(DRP) Docket Nos. 50-010; 50-237; 50-249 Licenses No. DPR-02; DPR-19; DPR-25 Licensee: Commonwealth Edison Company P. O. Box 767 Chicago, IL 60690 Facility Name: Dresden Nuclear Power Station, Units 1, 2 and 3 Inspection At: Dresden Site, Morris, Illinois Inspection Conducted: May 30 through July 14, 1989 Inspectors: S. G. Du Pont K. R. Ridgway D. E. Hills O. co. D. :j JMiller

i /Miller (f1r+IJ/r~J.4IW (jt+/J/rtJ.4JW Approved By: J. J. Harrison, Chief Reactor Projects Date Section 1B Inspection Summary Ins ection durln durin the No. 0-010 9002 DRP . No. 89 1 . No. - 49 9 Areas nspected: Routlne unannounced resldent lnspectlon of prevlously identified inspection items, license events reports followup, allegations followup, plant operations, maintenance and surveillances, safety assessment/

quality verification, radiological controls, engineering/technical support, Dresden Station management organization and report review. Results: One violation was identified during this inspection period concerning the Unit 2 excessive drywell temperature event of October 29, 1988 (Paragraph 11).

• During this inspection period, one reactor scram occurred from steamline power. This one scram was attributed to drifting main steaMline temperature switches swftches during duri ng a surveillance survei 11 ance test.

( III.12-4

Revision 8 April 1992 I \ DETAILS

1. Persons Contacted Commonwealth Edison Company (CECo)

      *E. Eenigenburg, Station Manager
      *J. Kotowski, Production Superintendent
      *L. Gerner, Technical Superintendent C. Allen, Administrative Service Superintendent D. Van Pelt, Assistant Superintendent - Maintenance G. Smith, Assistant Superintendent - Operations B. Zank, Operating Engineer K. Peterman, Regulatory Assurance Supervisor

w. Pietryga, Operating Engineer J. Achterberg, Technical Staff Supervisor L. Johnson, Q.C. Supervisor J. Mayer, Station Security Administrator D. Morey, Chemistry Services Supervisor D. Saccomando, Health Physics Services Supervisor

      *K. Kociuba, Q.A. Superintendent
      *R. Falbo, Regulatory Assurance Group Leader The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs,

.( reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical mechani~al and instrument person~el, personnel, and contract security personnel.

      *Denotes those attending one or more exit interviews conducted
      *Oenotes informally and formally at various times throughout the fnspection period.
  *2.

2. Previously Identified Inspection Items (92701 and 92702)

(Closed) Open Item ~237/88012-01 and 249/88014-01): Review calculations to valldate drywell spray initiation pressure limit curve. During the review of Dresden Emergency Operating Procedures, (EOP) the inspectors requested and could not be provided the calculations to validate the 5 psid differential pressure limit between torus and drywell for initiation of drywell sprays. The licensee found an evaluation for Pilgrim Station Mark I Containment, that is similar to Dresden's, which verified a safe torus to drywell differential pressure capability of 8 psid. The licensee later calculated a site specific limit of 8.3 psid as the maximum allowable negative pressure differential between the drywell and torus with a positlve positive torus pressure. These items are considered to be closed. (Closed) Open Item (237/88012-02 and 249/88014-02): Review justification for using 200 degrees as an entry condition for primary containment high temperature. The use of a 200 degree F temperature limit for entry into the Primary Containment Control EOP, which is

(

.( greater than the maximum normal operating average temperature 2 111.12-5

Revision 8 April 1992 specified by the Boiling Water Reactor Owners Group (BWROG) Emergency Procedure Guidelines (EPG), is justified because of the location of some of the thermocouples close to equipment with high temperatures during operation. The entry condition for Primary Containment Control remained after the licensee's engineering evaluation, at 200 degrees F as indicated on anyone of the five thermocouples specified. Closed) 0 en Item 237/88012-03 and 249/88014-03: Review ca cu atl0ns s oWlng transltlon transltl0n rom torus to rywell pressure used to create nomograph showing allowable pump Net Positive Suction Hea9 (NPSH). A review of the data used to develop the Emergency Core Cooling System (ECCS) suction nomographs, which are provided~n Dresden Emergency Operating Procedure (DEOP) 010 to protect the ECCS pumps from cavitation, indicated that the correlation between Drywell and Torus pressure could be incorrect when torus water level was above 11 feet. The ECCS suction nomograph was revised on October 27, 1988, to use the newly installed Torus Bottom Pressure Indication which indicates from 0 to 100 psig. These items are considered to be closed because of the use of the bottom pressure indication.

   ~C10sed)
   ~Closed)   Violation (237/87040-01l:

(237/87040-01~: Previous corrective actions ailed to prevent a repetitlve vlolation. vl01ation. This violation involved the by-passing of more Average Power Range Monitor (APRM) channels than permitted by Technical Specifications and was similar to a previous violation (50-237/87026-01) where the number of Reactor ( Protection System (RPS) Channel B APRM/ Intermediate Range Monitor (IRM) c~mpan;on c~mpanion trip functions ~ad been reduced to only one. The root cause of this violation was attributed to a personnel error. The inspector reviewed the following corrective actions taken to prevent recurrence: .

• Precautions were added to procedures along with a table illustrating the IRM/APRM companion relationship.
• A placard was added to the panel board listing the IRM/APRM complements.
• A procedure change to Appendix A, Shift Turnover, requires a check of the IRM/APRM configuration each shift turn over.
• Operator training has been completed.
• A Technical Specification Amendment was requested and issued, Amendments 237/100 and 249/96 on August 24, 1988,

         .to eliminate the APRM downscale trip requirements.

• A cover has been placed over the IRM/APRM bypass joysticks as a reminder to assure proper configuration prior to IRM .

bypassing an APRM or IRM. . ( 3 II 1.12-6

Revision 8 April 1992

• A letter to all licensed personnel reviewed the event and emphasized the importance of joystick configuration.

Closed) Unresolved Item (249/86012-30: Safety System Outage Modlflcatlon Inspection SSOMI Unresolved Item 2.4-2, Seismic Qualification of LPCI Room Cooler Motors. This unresolved item concerned the adequacy of seismic qualifications for Westinghouse motors used for "operation operation of the lPCI LPCI room coolers. In question was the application of the rigid mount criteria used in the original seismic qualification to the flexible motor mount that is used in the field installation. Also see Inspection Report 50-249/88200. The licensee had obtained calculations for the LPCI room cooler fan motor mounting configuration which confirmed that the lPCI LPCI fan motors are seismically qualified as installed. The unresolved item is considered to be closed because of the licensee obtaining a recent seismic qualification for the actually installed fan motor.

3. Licensee Event Reports (lER)

(LER) Followup (92700) Through direct observations, discussions with licensee personnel~ personnel,. and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished or scheduled in accordance with Technical Specifications. (Closed) LER 249/89003: Spurious Group V Primary Containment (Closed} Isolation While Shutdown Due to a Design Deficiency. With Unit 3 shutdown for a scheduled maintenance outage, an unexpected Group V isolation occurred resulting in the isolation of the isolation condenser. This event was attributed to differential pressure spikes and/or noise generated by an annubar flow instrument. This instrument was installed on the isolation condenser condensate return line to replace the previous elbow type instrument in 1985. Due to three previous occurrences, the last of which occurred on August 7, 1987, time delays were installed in the isolation circuitry. Because of a problem with setpoint drift on these time delay relays, a modification has been initiated to install relays with a shorter time delay during the next refuel outage. Closed LER 249/89007: Primary Containment Personnel Access Hatch local Local leak Leak Rate Test LlRT) LLRT) Failure. With Unit 3 at 21% rated thermal power following a scheduled maintenance outage.*a outage,'a drywell personnel access airlock failed its local leak rate test. The reactor was shutdown and primary containment de-inerted to facilitate further inspection. The licensee investigation revealed that the airlock inner door gasket seal had not seated evenly in the gasket grove causing the seal to be pinched through repeated usage of the airlock during the outage. This resulted in a six inch longitudinal tear in the gasket seal. The local leak rate test was successfully completed following replacement of the gasket seal. A revision is to ( 4 111.12-7

Revision 8 April 1992 (; be submitted for Dresden Operating Surveillance (DOS) 1600-10, . Pre-Startup Drywell Inspection Plan, to include a detailed inspection checklist to ensure proper seating of the gasket seal prior to final closing. The preventative maintenance program is also to be revised to require the gasket to be replaced every refueling outage.

   *~Closed) LER 237/89016:   High Pressure Coolant Injection (HPCI)

P1ping Found 1n Vlolatlon V1o1at1on of Final Safety Analysis Report (FSAR) Design Criteria due to Management Deficiency. Through a licensee HPCI Safety System Functional Inspection (SSFI) and subsequent analysis, it was determined that the Unit 2 and 3 HPCI turbine steam supply valves 2(3}-2301-3 2(3)-2301-3 drain pot piping did not meet FSAR design-design* criteria. Further analysis showed that the piping would, however, events*. The licensee remain operable under all design basis events~ attributed this event to modifications performed in 1982 without the benefit of a formal thermal and seismic analysis_ analysis. Although the Boiling Water Reactor Engineering Department (SWRED) (BWRED) had previously become aware of the design discrepancies as early as September 1984, Station personnel were not notified since it was believed the problem would be corrected through a pending modification. However, the pending modification was subsequently cancelled, and BWRED was not notified of the cancellation. Because of this, the design discrepancies remained until they were recently re-identified by the licensee's HPCI SSFI. Additional supports were subsequently added to Unit 3 and similar work is currently ongoing for Unit 2. The modification program was previously upgraded through a revision to ( Dresden Administrative Procedure (DAP) 5-1 Plant Modification Program. This included modification cancellation instructions requiring notification of affected station departments, Nuclear Licensing, BWRED and the designer. More detailed administrative controls on modifications were also delineated, as well as a design walkdown checklist used to confirm conceptual design and to provide input into the detailed design. In addition, BWRED is currently developing a procedure to give guidance when an analysis finds equipment in conflict with the FSAR. The previous inadequate design controls have been identified as an licensee identified violation (237/89017-01) and is considered closed in this report due to adequately completed or planned corrective actions meeting the criteria of 10 CFR 2 listed below.

• Closed LER 249/89008:

249189008: Fire Damper Discovered Obstructed by Weldlng Weld1ng qUlpment qU1pment Due to Management deficiency. The description of this event, including licensee investigation and corrective actions are described in Paragraph 6.c.3 of this report.

   *Note:   The preceding LERs have been reviewed against the criteria of
   ~R IOirFR  2, Appendix C, and the incidents described meet all of the .

following requirements. Thus no Notice of Violation is being issued for these items.

a. The event was identified by the licensee,

(( 5 111.12-8

Revision 8 April 1992 ( b. The event was an incident that, according to the current enforcement policy, met the criteria for Severity Levels IV or V violations,

c. The event was appropriately reported,

d. The event was or will be corrected (including measures to prevent recurrence within a reasonable amount of time), and

e. the event was not a violation that could have been prevented by the licensee1s licensee's corrective actions for a previous violation.

No violations or deviations, other than the noted licensee identified, were identified in this area.

4. Allegations Followup (AMS No. RIII-89-A-0044) (Closed)

On March 20,_ 20,. 1989, the Region III duty officer received a telephone call from an individual who expressed concerns related to leaks in the Unit 2 offgas system during late February and early March 1989. The caller would not provide his name. During this inspection, the inspector reviewed licensee records and reports and interviewed licensee and contractor personnel to determine the validity and consequences of the concerns expressed by ( (~ the alleger. The allegations are described and discussed below. Allegation: Plant management was not very concerned about a leak in the offgas rooms which resulted in several personal contaminations. Discussion: According to licensee personnel and records, on February 25, 1989, shortly after startup of Unit 2 after a refueling and maintenance outage, the clothing of some personnel on the 549-foot level of the turbine building was becoming contaminated with short-lived particulate daughters of noble gases. On February 25 and licensee*found some problems with offgas recombiner fans and 26, the licensee-found fan doors; these possible sources of the offgas leaks were repaired. After the problem began on February 25, the licensee collected particulate air samples near the steam jet air ejector rooms on a shift and/or dally basis dependent on air activity levels. The particulate air activity was found to increase periodically but at no set frequency. The air activity was always a small fraction of isotopic maximum permissible concentrations and displayed a half-life of about thirty minutes. I, 1989, licensee radiation protection personnel performed Dn March 1, On radiation surveys on the hydrogen addition systems and found a valve which was leaking. The Unit 2 shift foreman was informed of the leak. The leaking valve, however, was not repaired until March 8, 1989. After repair of the leak, air samples no longer identified elevated short-lived particulate activity. 6 111.12-9 I11.12-9

Revision 8 April 1992 ( 1989, the licensee again began to experience increased On March 16, 1989) airborne particulate activity in the same general area. On March Harch 20, 1989, the licensee again performed radiation surveys on the hydrogen addition system; no leaks were found. During review of airborne particulate air sample results, the Unit 2 Radiation Protection (RP) Foreman noted that the airborne particulate activity was elevated when the hydrogen addition system was on, and low or nonexistent when the hydrogen system was off. The RP foreman reported this information to the shift foreman for operations who had an operator check the valving lineup between the hydrogen addition and hydrogen monitoring system. The operator found and corrected a valve alignment problem. No further problem with airborne particulate activity was experienced. During review of this matter, the inspector learned that work was intermittently in progress to perform a modification of the hydrogen monitoring system. It appears that there was more than one source of offgas leakage during the period February 25 through March Harch 20, 1989, and work on the hydrogen monitor contributed to the offgas leaking problem. Finding: The allegation/concern was partially correct in that an offgas leak was identified on March 1,I, 1989, Which which was not repaired until March 8, 1989. However, no licensee procedure or policy, or regulatory requirement, was violated. The leak did not pose a significant radiological hazard to station personnel. c Allegation: unknown. Lung dose to workers from airborne radioactivity is lung Discussion: 10 CFR,20.103(a)(I)Note CFR*20.103(a)(I)Note 2 allows individual exposures to noble gases and their daughters to be accounted for as part of the limitation on individual external doses because the Maximum Permissible Concentrations (MPCs) listed in Table 1 Column 1 are based'on based on exposure to the material as an external radiation source. Therefore, it is not necessary to make an additional determination of lung dose for exposure to these nuclides. Finding: sUbstantiated. The allegation/concern was not substantiated.

5. Plant Operations (71710, 71707 and 93702)

a. Enforcement History During this inspection period, no violations or deviations were identified in the plant operations functional area. However, one item which occurred in a previous inspection period dealing with the high drywell temperature event of October 28, 1988, was II.b of determined to be a violation as described in Paragraph 11.b this report.

7 111.12-10

Revision 8 April 1992 c**( b.

h. Operational Events (1) On July 10, 1989, with Unit 2 at 63% rated thermal power, Recirculation Pump A speed unexpectedly drifted downward causing about a 3% decrease in both total core flow and reactor power. This caused the plant to enter the instability region of the power to flow map. The operators manually locked up the Recirculation Pump scoop tube to stop the speed drift and inserted CRAM arrays to exit the instability region. Recirculation Pump speeds were then matched by manual hand cranking of the scoop tube for Recirculation Pump A. Specific maintenance activities associated with this event are described in Paragraph 6.c.S 6.c.5 of this report.

(2) On July 12, 1989, Unit 2 received a reactor scram on a spurious Main Steam Line (MSL) High temperature signal RPS Channel B while Channel A was in a half scram condition during surveillance testing. The operators were able to achieve pressure control with the reactor water cleanup system (due to low decay heat) and vessel level was maintained within the normal operating range. Specific maintenance activities associated with this event are described in Paragraph 6.c.4 of this report. . (c cc.. Observation of Operations The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during this period. The inspectors verifie.d the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the Unit 2 and 3 reactor buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security plan. The inspectors reviewed new procedures and changes to procedures that were implemented during the inspection period. The review consisted of a verification for accuracy, correctness, and compliance with regulatory requirements. The inspectors also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barreling.

( (

8 III.12-11 111.12-11

Revision 8 April 1992 ( These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedures.

d. Engineered Safety Features (ESF) System Walkdown The inspector walked down the accessible portions of the Units 2 and 3 Standby Gas Treatment System (SGTS) to verify operability operabil i ty by comparing system lineup with plant drawings, as-built configuration, and operations checklists; observing equipment that could degrade performance; and verifying that instrumentation was properly valved, functioning, and calibrated. The inspectors also observed plant housekeeping/cleanliness conditions and radiation protection practices.

The inspector noted several discrepancies between the as-built configuration and plant drawing M-49. This included differences in the locations of specific temperature indicators, incorrect numbering of temperature indicators on the drawing and a pressure indicator installed in place of a temperature indicator shown on the drawing. All of these discrepancies were discussed with the System Engineer for resolution. No violations or deviations were identified in this area except as II.b of this report. described in Paragraph 11.b (

6. Maintenance and Surveillances (62703, 61726, and 93702)

The inspectors performed the following:

a. Unit 1 As general background, Unit 1 was shutdown in the late seventies and was never restarted after Three Mile Island (TMI) because of the costs associated with bringing the facility into conformance with post TM!

TMI safety requirements. All fuel elements and control rods were removed and stored in the fuel storage pool. The primary system was thoroughly chemically cleaned. The licensee proposed that Unit 1 would remain in this SAFSTOR condition until Units 2 and 3 are shutdown for decommissioning and submitted a SAFSTOR Decommissioning Plan and associated Technical Specifications (T5) (TS) for this period. These proposals are presently under review by NRC. In the course of the review of the proposed TS surveillance program for Unit 1, I, an inspection of the present surveillance program required by the existing TS was conducted. Since Unit 1 is in the shutdown defueled condition described above, operational surveillance requirements are no longer necessary such as safety limits, limiting 9 III.12-12

Revision 8 April 1992 ( safety system settings and most of the limiting conditions for operation (LCO). However, the licensee still conducts lCO LCO TS surveillances on radiological materials (airborne and liquid effluents, waste storage and environmental monitoring), storage fuel pool water level, fire protection systems and auxiliary electrical system batteries. The inspector reviewed surveillance procedures, check-sheets and schedules to verify that all TS required surveillances for Unit 1 were being conducted at the required frequencies. In addition, the inspection also reviewed the other safety and' and preventive maintenance (PM) checks contained in the Unit 1 General Surveillance System Master File and required by TS to ascertain that these checks were scheduled and completed. These surveillances and PM items numbered 190 and included such areas as instrument, Area Radiation Monitor (ARM) and gauge calibrations, routine radiation and contamination surveys, boiler and pressurized vessel inspections, fuel pool structure and fuel assembly conditions, fuel inventory, crane and hoist inspections, inspection and lubrication of pumps, valves, blowers, compressors, traveling screens, and emergency lighting. The inspector concluded that the instrumentation and equipment necessary to safely maintain Unit 1 in the SAFSTOR mode were listed in the Master File and the surveillances and services were properly being conducted as scheduled. ( b. Units 2 and 3 The inspectors observed surveillance testing required by technical specifications for the items listed below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated and that limiting conditions for operation were met. The inspectors also verified that removal and restoration of the affected components were accomplished, that test results conformed with technical specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel. The inspectors witnessed portions of the following test activities: Unit 2 Average Power Range Monitor/Rod Block Monitor Flow Converter to Total Core Flow Adjustment Standby Liquid Control System Pump Test Quarterly Standby liquid Liquid Control Pump Test for Inservice Testing Program ( 10 111.12-13

Revision 8 April 1992 Unit 3 Core Spray System Pump Test Core Spray System Valve Operability check Low Pressure Coolant Injection System Pump Operability Test Suppression Chamber to Orywell Drywell Vacuum Breaker Full Stroke Exercises Rod Block Monitor Functional Testing Units 2 and 3 HPCI System Operability Verification Reactor Low Water Level Scram and Low Low Water Level Isolation Trip HPCI Steam Line High Flow Isolation Trip HPCI Turbine Permissive Station maintenance activities of safety related systems and components listed below were observed/reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specifications. The following items were considered during this review: The limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to ((~ initiating the work; and activities were accomplished using approved procedures and were inspected as applicable. Functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; and parts and materials used were properly certified. Radiological and fire protection controls were properly implemented. Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance. Various maintenance activities associated with the following events were observed/reviewed.

c. Operational Events (1) On May 31, 1989, during the Unit 3 startup following the main transformer replacement outage, a controlled shutdown was conducted to perform repairs on a primary containment inner interlock door. This event is discussed in more detail in Paragraph 3 of this report.

(2) On June 13, 1989, the Unit 2 Reactor Feedwater Pump (RFP) C Outboard Seal was discovered to have failed. RFP B was already out of service due to a previous leak in it's discharge check valve. RFP A although operating, had a ( 11 111.12-14

Revision 8 April 1992 ( sma11 small leak in it's suction valve. Removal of RFP C from service necessitated a power reduction from about 500 MWe HWe to 280 MWe HWe to stay well within the capabilities of the remaining RFP A. (3) On June 17, 1989, while performing rounds, a Shift Foreman found two vertical fire dampers in a Unit 3 HPCI room non-ducted ventilation opening blocked open with an air hose and a welding lead. These obstacles were immediately removed such that the fire dampers were returned to operability. The licensee determined that these obstructions were routed through the fire barrier on June 14, 1989, while maintenance was being performed in the HPCI room. Dresden Technical Specification 3.12.F.2 requires that a continuous fire watch be established within one hour when a penetration fire barrier protecting safety related areas is not intact and equipment on either side of the barrier is required to be operable. Since Unit 3 was at power during that time and thus the HPCI pump on one side and various low pressure ECCS pumps on the other side of the fire barrier were required to be operable, the failure to establish a continuous fire watch constitutes a violation of Technical Specifications. This event has been reviewed against the criteria of 10 CFR 2, Appendix C, and the incident described meets all ( the requirements described in the note in Paragraph 3 of this report. Thus, no Notice of Violation is being issued for this item (249/89016-01) and this item is considered closed. (4) On June 23, 1989, with Unit 3 at 98% rated thermal power, a small flash fire occurred in the Main Generator Core Moni tor; The 1licensee Monitor~ i censee believed bel i eved that a small sma 11 hydrogen 1leak eak in the Core Monitor .led to an excessive hydrogen concentration and subsequent detonation While while an Instrument Technician was preforming maintenance. The fire was only momentary and a fire extinguisher was immediately used to further ensure that the fire was out. No damage was visually apparent and no injuries occurred. (5) (S) On July 7, 1989, with Unit 2 at 72% rated thermal power, the Reactor Building Ventilation System isolated and the Standby Gas Treatment System actuated during a Reactor Building Ventilation Radiation Monitor functional surveillance test. When the controller for Radiation Monitor B was pulled from its panel to conduct the surveillance a nicked wire shorted against the chassis causing a spike on the channel. The exposed wire was temporarily taped pending scheduling of permanent repairs. ( 12 IlI.12-15 III.12-15

Revision 8 April 1992 ( (6) On July 12, 1989, Unit 2 received a reactor scram on Main Steam Isolation Valve (MSIV) closure during a surveillance on the RPS BB power supply. The RPS 8B motor generator supplies the A channels of RPS and Group 1 Primary Containment Isolation System (PCIS). During the surveillance, a half scram and half isolation was received on RPS Channel A and pelS PCIS Channel A per the procedure. Line (MSL) High Temperature However, a spurious Main Steam line Signal was received on Channel B prior to resetting the surveillance induced half scram and half isolation signals. This resulted in a full Group 1 isolation (MSIVs closing) and a resulting reactor scram. All systems responded as expected and no safety systems actuated. (7) The inspectors observed completion of a work request involving Unit 22 APRM flow-biased scram, rod block and downscale calibrations. This work conducted was the result of failure of an APRM rod block functional surveillance test.

d. Approach to the Identification and Resolution of Technical Issues From a Safety Standpoint (1) A power reduction on Unit 2 to 300 MWe was held on June 9, 1989.

This was to facilitate a drywell entry in order to complete repairs on a Traversing Incore Probe (TIP) machine. The unit ( was also placed in single loop operation to facilitate repair of Recirculation Pump A Motor Generator Set outboard bearings. Other Unit 2 activities that occurred during the power reduction included investigation of spurious oscillations received on Turbine Control Valve #1, Main Steam Isolation Valve (MSIV) timing and replacement of cards in the Feedwater Level Control System panels. (2) Following the removal of Unit 2 RFP C from service on June 13, 1989, RFP B was restarted later in the day with the discharge check valve leaking 2-3 gpm and reactor power was increased as requested by the load dispatcher. The leak in the RFP A suction valve and the failed RFP C outboard seal were subsequently repaired. RFP B discharge check valve still leaked as of the end of the inspection period. (3) Following the Unit 2 scram of July 12, 1989, the temperature switches (which had experienced drift) associated with MSL Channel B detectors were replaced and the plant commenced startup on July 13, 1989. (5) Following the drifting of Recirculation Pump 2A speed on July 10, la, 1989 (see Paragraph S.b of this report), the licensee replaced the Recirculation Pump Motor-Generator (M-G) set tachometer which was sending incorrect signals to the velocity feedback circuitry. In addition, a mil1i milli volt/amp converter was

(

• . ( replaced in the velocity feedback circuitry.

13 111.12-16 III.12-16

Revision 8 April 1992 ( e. Responsiveness to NRC Initiatives The inspector expressed a concern to the licensee about an upward trend in control room work requests as early as February 1989. Although initial licensee action was delayed as to this concern, the licensee began to address this issue in May 1989. The licensee's investigation found that the work analysts were not recognizing these as affecting the control room and, as such, the priorities assigned to these work activities were too low.

7. Radiological Controls (92702)

Operational Events

• On June 8, 1989, the May 1989 Batch Waste Release Tank composite sample for tritium and gross alpha was inadvertently discarded before it could be sent offsite for analysis. Further review by the NRC is required, this is considered an Unresolved Item (237/89017-02) .

(237/89017-02).

• On June 15, 1989, while performing a Quality Assurance walkdown of owner controlled property outside the protected area,the area, the licensee found 145 55-gallon drums, some bearing visible low specific activity markings, in an old dumpsite. It appeared that the markings had at one time been painted or taped over to obscure them. The licensee surveyed the drums as they were removed from

( the dumpsite. Three of these drums and a concrete liner also di~covered in the dumpsite-were found to have low levels of radioactive contamination. These levels included 1.2k, 30k and 60k disintegrations per minute (DPM) on each of the three drums, respectively, and 300k OPM DPM on the concrete liner. The licensee indicated that the materials were placed in the dumpsite prior to 1981 and that these contamination levels were too low to detect with instrumentation available at that time.

The licensee removed the empty noncontaminated drums for general disposal. Drums with contamination were removed and stored in the radwaste area. Approximately 23 drums containing liquid or solid residue were also stored pending chemical analysis. An NRC inspector was dispatched to the site on June 22, 1989 to verify the licensee's findings and observe some of the licensee's radiological surveys. The inspectors agreed with the licensee's findings and corrective actions. No vio)ations vio'lations or deviations were identified in this area. ( 14 III .12-17 III.12-17

Revision 8 April 1992

8. Safety Assessment/Quality Verification (40500)

a. The inspector observed a licensee training session pertaining to the history of the counterf~it counterfeit molded-case circuit breaker (MCCB) issue and Nuclear Management and Resources Council (NUMARC) visual inspection guidelines. This training, as described in the letter, M. H. Richter to U. S. NRC, dated July 7, 1989, was a result of licensee participation in a NUMARC industry initiative to ensure reliable performance of MeCBs MCCBs used in non-safety related applications. The training was conducted prior to performing a visual inspection of the non-safety related MCC8 MCCB inventory. .

b. The inspector attended the licensee 1 licensee'ss June 1989 monthly performance review meeting. In addition to discussions involving the plant plant's status and activities for the previous month, each of the plantl.s top ten technical issues as determined by the licensee were reviewed.

A summary of performance during the Unit 3 transformer outage and activities of the Scram/Engineered Safety Feature Actuation Reduction Committee were also discussed. Particular management concern relating to an increase in Control Room work requests was expressed. The licensee conducted an evaluation to determine the cause of this increase. The results of the evaluation are discussed in Paragraph 6.d.2 of this report. No violations or deviations were identified in this area. ( 9. Engineering/Technical Support (37700)

a. Approach to the Identification and Resolution of Technical Issues From a Safety Standpolnt The inspector reviewed Partial Modification Design Package, M12-2/3-87-05C, Control Room Modifications, one of 10 packages concerning the consolidation of the Unit 1 Control room into the Unit 2/3 Control room and the utilization of the control room as additional office space. All unnecessary Unit 1 Control room panels and instrumentation will be removed and a new seismically designed three-hour rated fire wall and security barrier will be installed to separate the Unit 2/3 Control room from the new office space. The subject partial modification concerns the installation and testing of a new Process/Meteorology/Radiation Panel, 901-2. Except for existing Panels 18 and 18C, IBC, Electrical Switchyard Control and Instrumentation, and Panel 8-1, Station Auxiliary Power Control and Indication, which will be retained intact; all necessary Unit 1 instrumentation and controls will De be consolidated into the new panel. All of the above panels, along with new kitchen-eating facilities and locker room-toliet facilities, will be located within the Unit 2/3 Control room.

15 III~12-18 rII.12-1B

Revision 8 April 1992 Instrumentation and controls for the new 901-2 Panel will include:

1. A new ARM recorder.

2. A new service water discharge monitor.

3. A new annunciator panel for Unit 1 systems.

4. Connecting existing ARMs to the new recorder-annunciator.

5. Relocation of controls, indications and trouble annunciations from the old Unit 1 panels for service water system, bearing lube water system, turbine building closed cooling water system, fire pump discharge pressure, screen wash pumps, condenser circulating water pumps, well water system, clean demineralized water tank, contaminated water makeup, instrument air system, service air system, meterological data (2 recorders), and other trouble annunciators such as sphere and turbine building ventilation, radwaste building, instrument air dryer, sphere drain tank high level, heating system boiler and fuel pool high and low level.

Many of the above system indication relocations will include new pressure transmitters and transmitter power supplies. The modification will also require relocation of existing facilities such as breathing air piping, control room penetrations, HVAC system ducts, and electrical-telephone systems. ' modification package to , The inspector reviewed the partial mOdification ( verify that all systems depicted in the Unit 1 Decommissioning Pl~n were included, that all new and relocated instruments, annunciators, and controls would be calibrated and tested following the modification and before use, and that a 10 CFR 50.59 review had been completed and approved.

b. Responsiveness to NRC Initiatives The licensee was particularly responsive to providing answers to questions on various technical issues requested by the NRC regional office. These areas included plant specific testing of diesel generator trips and bypasses and the source of RPS response times used in reload safety evaluations.

No violations o*r or deviations were identified in this area.

10. Dresden Station Management Organization During this inspection period, CECa CECo announced several key management changes including the following:

C. Schroeder, Technical Superintendent to Corporate Outage Planning L. Gerner, Production Superintendent to Technical Superintendent J. Kotowski, Assistant Superintendent-Operations to Production

(( Superintendent G. Smith, Operating Engineer to Assistant Superintendent-Operations 16 III.12-19

Revision 8 April 1992 ( 11. Report Review

a. During the inspection period, the inspectors reviewed the licensee's Monthly Operating Report for June. The inspectors confirmed that the information provided met the requirements of Technical Specification 6.6.A.3 and Regulatory Guide 1.16.

b. The inspectors completed the review of the Dresden Unit 2 Drywell Temperature Event Evaluation Report prepared by Commonwealth Edison Company and Sargent and Lundy for the October 29, 1988 event. This event was previously discussed in Inspection Reports 50/237/88026; 50/249/88026 and 50-237/89011; 50-249/89010. The licensee .

attributed this event to the absence of cooling airflow to the reactor head area due to the ventilation hatches, provided in the bulkhead plate, being left in the closed position. The licensee determined the primary root cause to be inadequacies in procedures which direct operations and maintenance personnel to open the hatches and perform an inspection prior to startup. Although Dresden Maintenance Procedure (DMP) 1600-5, Drywell Head Replacement and Installation of Shield Blocks, Revision 2, contained a step to open all required ventilation openings, it did not clearly identify which hatches were required to be open. Only the manway hatches were found open. In addition, Dresden Operating Surveillance Procedure COOS) (DOS) 1600-10, Pre-Startup Drywell Inspection Plan, Revision 4, which contains a step to verify that the h~tch hatch doors to the reactor head area are open; open, ( was misinterpreted by the shift supervisors who made the ins~ection as applying only to the manway hatches. Consequently, ventilation hatches were not checked. The inadequate procedures to which this event was attributed are considered to be a violation of Technical Specification 6.2.A (237/89017-03) . (237/89017-03). The licensee identified and implemented extensive corrective actions in response to this event and, as such, this item is considered closed. These corrective actions included the following:

• Revisions of the inadequate procedures.
• Evaluation of the remaining life of environmentally qualified equipment.
• Repair and replacement of electrical and mechanical equipment and cables as required.
• Installation of an upgraded drywell temperature monitoring system.
• Repair of the drywell cooling system and conduct of a performance test.

( 17 III. 12-20 111.12-20

Revision 8 Apri 1 1992 April

• Repair of thermal insulation as required.
• Repaint of the drywe" drywell dome and scraping of other drywell surfaces to remove loose paint.
• Implementation of a procedure for monitoring and elevating drywell thermocouple data.
• Performance of a drywel1 drywell insulation system evaluation.
• Review and update of equipment qualification binders as necessary.

No other violations or deviations were identified.

12. Exit Interview (30703)

The inspectors met with licensee representatives (denoted in Paragraph 1) on July 13, 1989, formally and informally throughout the inspection period, and summarized the scope and findings of the inspection activities. The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any such documents/processes as proprietary. The licensee

• c

.( acknowledged the findings of the inspection. 18 I II .12-21 II1.12-21

Tab 13 Revision 8 April 1992 ( DRESDEN 2 &3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspection Report No. 50-237/89022 and 50-249/89021

  ~
  £igg                  Title IIL13-1 III. 13-1             Inspection Report No. 50-237/89022 and 50-249/89021 dated December 26, 1989.

11I.13-23 II I.13-23 January 25, 1990 CECo letter from T. J. Kovach to A. Bert Davis (NRC), Response to Notice of Violation and Inspection Report No. 50-237/89022 and 50-249/89021. ( III.13-i I11.13-i

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UNITED STATES ~J_ Rev; s ; on 8 _,S'~ April 19 92 0 NUCLEA'R REGULATORY COMMISSIO ,.wi, s;&--.-l., A.......P-co ( GLEN I:LL GLI:N REGION III 711 ROOSEVELT ROAD

                                                     " ' ROOSEVELT ROAD ELLYN.             .0137 YN. ILLINOIS 10137 f'l * ...,

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( ('.*;-1..,,-/t!J*~*6~ Docket No. 50-237 Dock~t No. 50-249

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Commonwealth Edison Company 11- pJ>/j ....:. .,PC) V ATTN: Mr. Cordell Reed ') .,;t

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• Seni"or Sen for Vice Vi ce President Post Office Box 767 Pres i dent

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Chicago, IL 60690 .;.:t...-

                                                                                                            ~./A/f            .-/ A--7 Gentlemen:                                                                                  ~.a;-~
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                                                                                                            ~ F..J-......i.

This refers to the routine safety inspection conducted by S. G. Du Pont ~ ~ and D. E. Hills of this office on October 11 through December 1, I, 1989, of ~~ ~~ ~u activities at Dresden Nuclear Power Station, Units 2 and 3, authorized by ._ ~; ~ NRC Operating Licenses No. OPR-19DPR-19 and No. DPR-25 and to the discussion of ~¥~ ~~~ ** our findings with* Mr. E. Eenigenburg and others at the conclusion of the ~~~ I'~~ inspection. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and c ( interviews with personnel. During this inspection, certain of your activities appeared to be in violation of NRC requirements, as described in the enclosed Notice. A written response is required. In accordance with 10 CFR 2.790, of the Commission1s Commission's regulations, a copy of this letter and the enclosures will be placed in the NRC Public Document Room. The responses directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. We will gladly discuss any questions you have concerning this inspection. Sincerely, ow. rSDJih~v D. ~er, Chief Reactor Projects Branch 1

Enclosures:

Violation

1. Notice of Vio1ation

2. Inspection Report No. SO-237IB9022(DRP) 50-237/89022(DRP)

No. 50-249/89021(DRP) (, See Attached Distribution IIL13-1 III.13-1

Revision 8 April 1992 NOTICE OF VIOLATION Commonwealth Edison Company Docket No. 50-237 Dresden Nuclear Station Doc~et No. 50-249 Ooc~et As a result of the inspection conducted on October 11 thru December 1.1, 1989, and in accordance with the General Policy and Procedures for NRC Enforcement Actions, (10 CFR Part 2, Appendix C), (1989) the following violation was identified: identi fied:

1. 10 CFR 50.48(a) requires that each operating nuclear power plant have

a. fire protection plan that satisfies Criterion 3 of Appendix A to 10 CFR Part 50. It further requires that the plan p1an shall describe specific features necessary to implement the plan such as administrative controls to limit fire damage to structures, systems or components important to safety so that the capability to safely shut down the plant is assured.

Section C.l of the licensee's response to the Guidelines of Appendix A to Branch Technical Position APCSB 9.5-1 as accepted in the 1980 Supplemental Safety Evaluation Report indicates' indicates that administrative measures are established to ensure that guidelines of the Branch Technical Position are included in design and procurement documents and that deviations therefrom are controlled. Contrary to the above, a penetration in a three hour fire rated wall ( located in a safety related area of the 570 feet elevation of the reactor building, as prescribed by Section D.l.j D.1.j of the Branch Technical Position, was not included in design documents and deviations were not controlled. The fire rated wall was degraded in 1985 by replacement of the original piping with non-approved polyvinyl chloride plastic piping and was further degraded on October 25, 1989 when the piping was completely removed and the penetration left unsealed. This is a Severity Level IV violation (Supplement I) (No. 237/89022-02(DRP)). 237/89022-02(DRP>>. Pursuant to the provisions of 10 CFR 2.201, you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply, including for each violation: (1) corrective action taken and the results achieved; aChieved; (2) corrective action to be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown. on: OEr. 2 6 1989 1999 Date 1 ( .. III.13-2

Revision 8 April 1992

( u.s. NUCLEAR REGULATORY COMMISSION REGION III 50-237/89022(DRP); No. 50-249/89021(ORP)

Reports No. 50-237/89022{DRP); 50-249/8902I(DRP) Docket Nos. 50-237; 50-249 licenses No. DPR-19; DPR-25 Licensee: Commonwealth Edison Company P. O. Box 767 Chicago, Il 60690 Facility'Hame: Facility"Name: Dresden' Dresden" Nuclear Power Station, Units 2 and 3 Inspection At: Dresden Site, Morris, Illinois Inspection Conducted: October 11 through December 1, I, 1989 Inspectors: s. S. S. G. Du Pont

                      ~)~;y.~ a~ ".
                      ~)~;Y.~?io1--

Approved By: H. Hin~r., Chief J. M. Reactor Projects Section IS IB C'( Inspection Summary eriod of October 11 throu h December 1 *1989

                                                                          °1989 (Re ort
     ~~~~~~~~~~~~

o.  ; o. ..

Areas Inspected: Routine unannounced safe~ safety inspection by resident inspectors ot of prevlously prevl0usiy ldentified inspection items, licensee event reports, plant operations, ma intenance and survei maintenance llance, safety assessment/qua surveillance, lify verification, assessment/qualify engineering/technical support and report review. Results: o Specific events demonstrating management involvement and a regard for correctly meeting requirements as well as for minimizing unplanned transients were noted. o One violation was identified during the inspection period as described in Paragraph 5.b.B. S.b.B. This involved the failure to properly control the design of a penetration through a fire barrier such that maintenance personnel degraded that barrier on two separate occasions. This specific event was considered to be of minimum safety Significance although a previous de9rad~tion degrad~tion of a fire barrier by maintenance personnel was documer.ted in a previous inspection report. This was not considered to documented b~ indicative of what are usually thorough and effective corrective actions by the licensee. III. 13-3

Revision 8 April 1992 ( COlTlllonwealth COllJllonwealth Edison Company 2 ~EC 2 , 1939

                                                   ~EC:'  1989 Distribution cc w/encTosures:

w/enclosures: T. kovach, Kovach, Nuclear Licensing Poanager

                         ~anager E. D. Eenigenburg, Station Manager DCD/DCB (RIDS)

Licensing Fee Management Branch Richard Hubbard J. W. McCaffrey, Chief, Public

         . Utilities Division LaSa lle RIO laSa Quad Cities RIO Dresden RIO

( ( IlI.13-4 III.13-4

Revision 8 April 1992 co( o Two unresolved items were identified in Paragraphs S.b.S and 7.b.3.i;One 7.b.3.:,One involved whether adequate correct.ive actions were taken in response to previously identified HPCI piping support discrepancies. The other involved installation of main steamline leak detection temperature switches without the appropriate environmental qualification documentation. (,,--. cc- 2 III.13-5 .- ---_... -. ---

Revision 8 April 1992 DETAILS

1. Persons Contacted Commonwealth Edison Company

         *E. Eenigenburg, Station Manager
        *L. Gerner, Technical Superintendent E. Mantel, Services Director
        *J. Kotowski, Production Superintendent O.

D. Van Pelt, Assistant Superintendent, Maintenance J. Achterberg, Assistant Superintendent, Work Planning

        *G. Smith, Assistant Superintendent.

Superintendent, Operations

        *K. Peterman, Regulato~

Regulatory Assurance Supervisor

        *C. Allen, Performance Improvement Supervisor W. Pietryga, Operating Engineer
        *R. Stobert, Operating Engineer M. KorchynskY, Korchynsky, Operating Engineer B. Zank, Operating Engineer J. Williams, Operating Engineer
        *M. Strait,.

Strait" Technical Staff Supervisor L.

l. Johnson, Q.C. Supervisor J. Mayer, Station Security Administrator

        *D.
        *0. Morey, Chemistry Services Supervisor Horey,
        *D.
        *0. Saccomando, Health Physics Services Supervisor E. Netzel, Q.A. Superintendent c        *R.
        *K.

K. Falbo, Regulatory Assurance Group Leader Yates, Nuclear Safety Supervisor Kociuba, Quality Assurance Superintendent Kociuba. The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs,

        ..reactor eactor and auxiliary operators, shift engineers and foremen, electrical, personnel, and contract security personnel.

mechanical and instrument personnel.

        *Denotes those attending one or more exit interviews conducted informally at various times throughout the inspection period.

2. Previously Identified Inspection Items (92701 and 92702)

(Open) Open Item (No. 249/89011-02): The licensee was to provide a written response describlng describing planned corrective actions to ensure that usage of the isolation condenser for extended time periods without offsite power would not result tn ion radioactive releases. The latest response to this cootained in the letter from J. A. Silady to issue by the licensee was cOCltained A. B. Davis dated November 15, 1989. A tentative schedule was established for. the resp:ect for ,the i ve un:ft ~e'fue res~'E!cti\'e ling outages at the end of eyc refueling 1e 13 ;inn 1992 Cycle to install diesel driven p~mps pumps for supply of clean demineralized water to shell side of the isolation condensers from the clean demineralized the sheil water storage tank. A proposed design improvement to supply 480 VAC power to the isolation condenser shell side motor-operated clean demineralized water fill valves was being reviewed with respect to impact on the (( 3 III.13-6

Revision 8 B April 1992 ( Appendix R safe shutdown analysis. The licensee committed to providing a final update concerning this part of the design within two months of the date of the letter. No violations or deviations were identified in this area.

3. Licensee Event Reports (lER) licensee (LER) Followup (90712 and 92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine fulfilled" immediate corrective that reportability requirements were fulfilled"immediate action was accomplished, and corrective action to prevent recurrence had been accomplished or planned in accordance with Technical Specifications.

            ~Closed)  LER No. 237/89025: Inadvertent Automatic Isolation of the High lER ressure Coolant Injection (HPCI) System Due to Design Deficiency. The activities resulting in this occurrence were discussed in inspection report No. 50-237/89019; No. 50-249/89018. The licensee attributed the root cause of this event to a design deficiency within the Analog Trip System (ATS) panel such that the master trip unit (MTU) mounting configuration can result in spurious trips when adjacent MTUs are removed.

licensee, determined that Dresden Instrument Surveillance (DIS) 2300-11, The licensee* System Isolation-Reactor Pressure Transmitter Calibration and Maintenance' Inspection, was the only HPCI instrument surveillance procedure that Therefore, the licensee planned to required removal of adjacent MTUs. Therefore. incorporate precautions in this procedure to exhibit care when removing c( and replacing MTUs and to require prior notification to the Operations Shift Supervisor that MTU replacement may result in an isolation signal. The licensee also planned to post signs on the ATS panels to indicate the same caution and requirement. The licensee did not plan to change the MTU mounting configuration since they considered this to be an isolated event and MTU removal was a rare occurrence due to a high reliability of the component.

                                                                 'Treatment System Due to (Closed) LER No. 237/89026: Start of Standby Gas "Treatment Loose Reactor BUlldlng Ventilation System Radiation Monitor Connection.

This event including initial licensee actions was described in inspection report No. 50-237/89019; No. 50-249/89018. In addition, the licensee planned to revise DIS 1700-7, Reactor Building Ventilation (RBV) Radiation Monitor Functional Test, to require checking RBV radiation monitors for loose connections and exposed wiring during the surveillance. The licensee also planned to evaluate possible methods to improve instrument department response time ti~e to this type of event and to evaluate a generic radiation monitor troubleshooting procedure. Closed) LER No. 237/89027: Postulated Low Pressure Coolant Injection wlng Bus Loss Resulting From Diesel Generator Voltage Regulator Failure Due to Design Deficiency. This item and corresponding licensee 7.b.1 and 7.c of this report. actions are described in Paragraphs 7.b.l (' (" 4 III.13-7 - -._ .... -

                                                                               -~-".'-    . - - -,------.

Revision 8 P.pril

                                                                        !-pril 1992

( Closed lER LER No. 237/89028: Containment Cooling Service Water (CCSW) Pump uctlon ay Water Leve . Reduction. This event was discussed in inspection report No. 50-237/89019; No. 50-249/89018. As a long term corrective . action, the licensee planned to review methods to proceduralize a program that was initiated to measure water level drop across the trash bars. This would contribute to earlier recognize of CCSW suction level bay decreases. ' Closed LER No. 237/89029: Elevated HPCI Discharge Piping Temperature Due to eactor eedwater ystem Back Le.akage. This item and corresponding licensee actions are described in Paragraphs 5.b.4, 5.b.5 and 7.c of this report and report No. 50-237/89023; and No. SO-249/89022. Closed LER No. 237/89030: Reactor Building Fire Wall Degraded By An nau horlzed enetratlon pening Due to Management Deficiency. This item and corresponding short term licensee actions are described in Paragraphs S.b.8 and S.c of this report. S.b.S Closed) LER No. 249/89004: HPCI System Declared Inoperable Due to Failed oom Cooler Fan rlve elts. This item and corresponding licensee actions are described in Paragraph S.b.3 of this report. No violations or deviations were identified in this area except as described in Paragraph S.h.8 S.b.8 of this report.

4. Plant Operations (71707, 71710 and 93702)

a. Enforcement History During this inspection period, no violations or deviations were identified in the plant operations functional area.

b. Operational Events On October 10, 1989, the Unit 2/3 Cribhouse Basement Cable Tray Fire Suppression Deluge System was inadvertently actuated during performance of Dresden Fire Protection Procedure (DFPP) 4114-6, Fire System Yard loop Loop Monthly Inspection, Revision 10. While inspecting the protectowire fire alarm control panel and power supply for the cribhouse basement cable tray fire detection system, fire panel 2223-112.

2223-112, the operator attempted to replace burned out light bulbs as required by the procedure. In order to identify the burned out bulbs, the operator depressed a panel button labeled Alarm Devices-Push to Test, which whiCh he thought would just illuminate the panel lights. However. However, this button instead tested the fire panel relays which actuated the deluge system spraying water into the Unit 2/3 cribhouse bas~ent. basenent. The operator immediately isolated flow by breaking the locking device on cribhouse cable tray isolation valve 2/3-4199-176 and closing the valve. A second initiation occurred later that same day due to grounds on the protectowire located in the cable trays which were caused by water from the first initiation. The area was allowed to dry out and inspections revealed no other equipment damage. c ( 5S III .13-8 III.13-8 -_._----

Revision B 8 April 1992

c. Approach to the Identification and Resolution of Technical Issues From a Safety Standpoint The licensee exhibited regard toward ensuring-operators were aware of adverse conditions, their affect on the plant and mitigation techniques. This was exemplified by informing operators of an alternate method to determine if Electrohydraulic Control (EHC)

DC power were lost as described in Paragraph S.b.l of this report.- Due to a relay failure at that time, a loss of EHC DC power would have rendered various main turbine trips inoperable without-a corresponding alarm to warn the operator of this condition. ._ Questioning of the operators by the inspectors indicated that they were aware of the alternate method. The licensee's investigation into "the inadvertent deluge system _ into-the actuation represented a thorough and comprehensive root cause analysis and corresponding corrective actions. The -licensee attributed the cause to inaccurate labeling which did not make the function of the pushbutton apparent. In addition, OFPP DFPP 4114-6 was deficient in that it did not caution the operator concerning this pushbutton. Finally, the licensee determined that operator training was deficient in that the fire system lesson plan also did not provide this information. As a result, the licensee installed an additional label below the pushbutton that read Push to Initiate Deluge. The licensee also proposed the following corrective actions to ensure this event would not be repeated with respect to other fire protection panels: (--- (I) (1) Discuss the event in Operations and Maintenance tailgate sessions such that personnel are aware of this pushbutton in protectowire fire panels. (2) Identify all protectowire fire panels that have an equivalent pushbutton and provide the additional warning labels below each of the pushbuttons. - (3) Revise DFPP 4114-6 to identify protectowire fire panels which do not contain a light test button. (4) Revise the fire system training lesson plan to include this event and to stress the existence of this pushbutton. (5) Determine the requirements for having the pushbutton in protectcwire fire panels and remove those not required.

d. Responsiveness to NRC Concerns Issuance of Dresden Operating Abnormal (OOA)

Issuanc~ (DOA) Procedure 0500-02, Partial Half or Full Scram Actuation, in November 1989 was in response to NRC concerns and indicated the ability to apply lessons learned from other plants. This procedure prescribed mitigating operator actions upon a half or full scram for which Reactor c- 6 III. 13-9 III.13-9 .-.-- ..,._- ......-.---. ----- ..------...---

                                                              . ._.-_. _- ...                             ~----'----

Revision 8 April 1992 Protection System scram solenoid indicating lights do not extj~guish as they should. This procedure was developed as a result of*~* of~~~* commitments made to the NRC following such an event at Commonwealth Edison's LaSalle plant. .~

e. Assurance of Quality, Including Management Involvement and Control .

The licensee's decision involving when to initiate a Unit 2 shutdown due to the HPCI piping support damage as discussed in Paragraph S.b.4 of this report demonstrated management involvement and a desire to ensure that technical specification requirements were met. Previous licensee guidance had concerned the case in which a 24 hour shutdown (LCD) was immediately entered. In Limiting Condition for Operation (LCO) that case, the licensee's interpretation did not require immediately reducing power if it was legitimately felt that the problem could be rectified and the lCO LCO exited in sufficient time such that an orderly shutdown could still be completed within the original 24 hours if needed. However, the case in question differed from previous guidance LCD was entered prior to ent~ in that a seven day lCO entry into the 24 hour shutdown LCO verses being immediately placed into the 24 hour shutdown LCD. LCO. Thus, the guidance was unclear as it applied to this situation. To ensure compliance with the requirements, the licensee consulted with NRC regional upper management as to the applicability of previous guidance to this situation. As the licensee felt that actions to consider the system operable could be completed within 12 hours, the decision was made to actually begin the shutdown 12 hours after entry into the 24 hour LCD. LCO. This left enough time for completion of an orderly shutdown within the original 24 hours in case the actions did ( not get completed on time. When the actions were not completed on time, the licensee initiated the shutdown at the time agreed to with the NRC. The inspectors also noted during discussions with licensed operators regarding the incident that they possessed a genuine desire to ensure conservative compliance with technical specifications and, in fact, were concerned as to what appeared to several of them to be actions possibly contrary to previous guidance that they had received in this area. The inspectors regarded this concern to be indicative of a professional attitude of the licensed operators toward their individual licensed responsibilities.

f. Observation of Operations The inspectors observed control room operations, reviewed applicable discussions with control room operators during logs and conducted dis<:.ussions this period. The inspectors verified the operability of selected emergency systems, reviewed tag~ut tagout records and verified proper of affected co~ponents. Tours of Units 2 and 3 return to service af reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance leaks.

requests had been initiated for equipment in need of maintenance. The inspectors also walked down various HPCI piping supports to ascertain damage and verify repairs as described in Paragraphs S.b.4 and S.b.S of this report. ( 7 III.13-10 ---

                                                                                       . .-_4_ ."---

__ ~

Revision 8B April 1992 ( The the inspectors, by observation and direct interview, verified that physical security plan was being implemented in accordance with the station security plan. The inspectors reviewed new procedures and changes to procedures that were implemented during the inspection period. The review consisted of a verification for accuracy, correctness, and compliance with regulatory regUlatory requirements. The inspectors also witnessed portions of the radioactive waste system controls associated with radwaste shipments and bar.reling. barreling. These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under TeChnical Specifications, 10 CFR, and administrative procedures.

5. Maintenance and Surveillance (62703, 61726 and 93702)

a. Enforcement History During this inspection period, one violation was identified in the maintenance/surveillance functional area. This concerned a failure to properly control the design of a penetration through a fire barrier such that maintenance personnel degraded that barrier on two separate occasions.

b. Operational EVents Events

( Various maintenance activities associated with the following events were observed or reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guid~s and industry codes or standards and in conformance with technical specifications. The following items were considered during this review: The LeOs were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or $ystems systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemented. Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance. (-( 8 III. 13-11 III.13-11

Revision 8B April 1992 (1) On October 12, 1989, alarms for EHC DC Power Failure and EHC Electrical Malfunction were received on Unit 3. This was of particular concern since loss of EHC DC power would render many of the main turbine trips inoperable. Troubleshooting activities maintenance ana witnessed by the conducted by instrument .aintenance inspectors indicated that DC power was still available and that the alarm relay itself was malfunctioning. However, it was decided not to replace the relay since such an action would be main turbine trip. The relay highly susceptible to causing a .ain in question was located on a circuit card which also contained several other trip relays. These relays were of a mercury type such that inappropriate movement when replacing the card could cause a trip. Thus, the licensee intended to wait until the next time power was reduced to less than 4~ 45% to repair the problem so that a turbine trip wculd not also result in a reactor scram. (2) On October 15, 1989, the breaker for the Unit 2 LPCI Room Cooler B was found to have been damaged when operators investigated a report that smoke was seen coming from the breaker. The inoperability of the room cooler also required Core Spray Loop Band LPCI loopLoop B to be declared inoperable, placing Unit 2 into a 24 hour required shutdown LCO. The breaker was, however, repaired later that same day such that the shutdown did not have to commence. (3) On October 22, 1989, the Unit 3 HPCI System was declared c ( inoperable due to discovery of broken fan belts on the HPCI room cooler. This placed the unit into a seven day LCD. belts were replaced and the system declared operable on LCO. The October 23, 1989. The licensee had previously planned to take the room cooler out-of-service on October 23, 1989, for bearing work. Thus, ThUS, this activity was also completed. A previous event

      ., concerning Unit 2 HPCI room cooler broken fan belts was discussed in inspection report No. 50-237/89019; No. 50-249/89018.

The root cause of that event was determined to be excessive use of the room cooler due to elevated HPCI room temperatures caused by feedwater system bac.k 1eakage into the feedwater 1iines. backleakilge nes. Increased HPCI line temperatures eventually led to inoperability of the Unit 2 HPCl HPCI system as discussed in Paragraph S.b.4 5.b.4 of this report. However, the licensee indicated that the Unit 3 HPCI room and the HPCI line te~peratures temperatures were much less than on Unit 2. Thus, the licensee initially indicated that these events were biickleakage was not a problem on Unit 3 HPCI. unrelated and back.lea'Kage The licensee attri~uted attributed the cause of the Unit 3 HPCI room cooler belt failure to be shaft .isalignment misalignment due to the worn bearing. Although the exact cause of the worn bearing was unknown, the most probable cause was inappropriate drive belt tensioning. Dresden Electrical Procedure (OEP)(DEP) 5700-4, Electrical Maintenance and Surveillance of HPCI Room Fan Motors, ( 9 111.13-12 III.13-12

Revision 8B April 1992 instructed the user to ensure proper belt tension was achieved but gave no additional guidance as to what this tension should be. Therefore, the licensee planned to revise DEP 5700-4 to . ncl ude proper belt tension information .. iinclude (4) an On October 23, 1989, the licensee found the Unit 2 HPCI system discharge piping water temperature to be sufficiently high to potentially cause voids to form within the piping. Piping temperatures were discovered to have increased to 275 degrees F between HPCI pump discharge outboard valve 2-2301-8 and HPCI pump discharge inboard valve 2-2301-9, 246 degrees F at the HPCI pump and 135 degrees F near the condensate storage tank (CST). The corresponding static pressure in the HPCI discharge piping at the pump was 32 psig (47 psia). Thus, temperature and pressure in particular areas of the system, represented possible saturated conditions which the licensee believed provided the potential for a waterhammer event. Therefore, the licensee declared the Unit 2 HPCI system inoperable and entered a seven day LCO. On October 27-28, 1989, the licensee

     . discovered numerous signs of damage to various Unit 2 HPCI discharge piping supports. An unusual event (UE) was declared on October 31, 1989, when the licensee initiated a technical specification required shutdown due to a failure to return the HPCl HPCI system to operability within the seven day LCO. The system was returned to operability that same day prior to completion of the shutdown. This event, including licensee corrective actions, was discussed in detail in inspection report No. 50-237/89023; No. 50-249/89022. A clamp on Unit 2

( HPCl HPCI piping support M-1151D-154 located on top of the torus was identified to be rotated on the pipe and a work request initiated during the last Unit 2 refueling outage. However, this work request was not completed during that outage. This is considered part of an unresolved item (No. 237/890ZZ-01(DRP>>, 237/89022-01(DRP>>, together with the item in Paragraph S.b.S 5.b.5 of this report, pending NRC review and determination of why this work was deferred. (5) On October 29, 1989, the licensee found the Unit 3 HPCI system discharge piping temperature at an elbow of the piping near its emergence from the X-area (steam tunnel) to be 256 degrees F. Additional measurements o~tained obtained on October 31, 1989, indicated piping temperature just upstream on the other side of the elbow measured between 163 and 133 degrees F depending on the circumference location. The corresponding static pressure in circ~mference the HPCI dfscharge discharge piping at the pump was about 45 psig (60 psia). The licensee beH@ved* belf,eved temperature and pressure conditions near the elbow could potentially cause steam pocket formation. Thus, the licensee declared the Unit 3 HPCI system inoperable. On November 1, 1989, Ig89, the licensee also discovered signs of damage Untt 3 HPCI piping supports. The Unit 3 HPCI system was to Unit returned to service on November 7, 1989. This event including c 10 III.13-13 -- - _.. " .... -. .... ---.-

Revision 8 April 1992 " ('"

  /

( correc~ive actions, was discussed in detail in licensee correc:tive inspection report No. 50-237/89023; No. 50-249/89022. Unit 3 H-11870-110 was found to have the baseplate HPCl piping support M-11870-110 and all four concrete expansion anchors "pulled *pu11ed from the wall. Evidence also showed that a licensee walkdown conducted in 1979 Evidence" wallmount pulling away. This is considered part of an noted a wal1mount unresolved item (No. 237/89022-01(DRP>>, 237/89022-01(ORP>>, together with the item in Paragraph S.b.45.b.4 of this report, pending NRC determination of whether this is the same damage as originally identified. (6) On November 6, 1989, the Unit 2 HPCl Motor Hotor Gear Unit (MGU) high speed stop (H55) indicating light was discovered to be blinking on and off. However, the MGU HGU was still functional since it automatically returned to it's H55 from it's low speed stop (L55). A large amount of noise was discovered in the DC output signal and, thus, the HPCl MGU was taken out of servtce servTce to repair it on November 8, 1989. The MGU HGU H55 indication fluctuations were eliminated by replacement of a circuit capacitor and HPCl was declared operable on November 10, 1989. (7) Throughout much of the inspection period, Unit 2 operated at sol i ght ly reduced power due to repeated spur; slightly spurious ous primary containment half isolation signals received at full power conditions. These half isolations were caused by failure of steam1ine low pressure switch PS-261-30B. main steaml;ne P5-261-30B. The licensee believed that rapid pressure fluctuations within the pressure line caused by vibration was prematurely degrading the bourdon ( tube within the switch. This had been a recurring problem in the past with previous actions involving vibration testing of steam1ine low pressure switches and installation of the main steaml;ne a pressure snubber in the sensing line. The switch sw~tch had been replaced several times but would typically fail after approximately one month. Load was reduced to 65 percent on November 18. 18, 1989, in order to allow entry to th, the heater bay to conduct a walkdown wa1kdown of the sensing line. This walkdown wa1kdown did p"ob1ems with the 1line. not identify any pl*oblems ine. On November 22, 1989, P5-261-30B was replaced and a new portion of sensing line on PS-261-30B the instrument rack was installed in a looped configuration in hopes of dampening any pressure fluctuations to the switch. The licensee was* also evaluating possible future replacement with a different and less susceptible type switch. (8) On October 26, 1989.1989, the Station Manager discovered a three penetration stuffed with rags in a three hour fire inch open pene~ration separating the Units 2 and 3 reactor buildings at rated wall separati"ng elevation 570 feet. The mechanical maintenance department was dislllarrtling and cleaning an area on the in the process of disnrarttling Unit 2 side ~f of the wall which was formerly a control rod drive (CRD) maintena~ce area. The work being performed under a blanket work request for general plant cleanup was not intended to disrupt or alter plant components or systems. A drain line connected to a CRD flush tank had previously been routed ( 11 III.13-14

Revision 8 April 1992 ( through the penetration to a floor drai drainn on the opposite-.~ide opposite,,$ide of the wall. Due to high radiation levels from the drain'line drain"line and the fact that the CRD flush tank was to be removed during

   ~he cleanup, removal of the drain line was also added to the scope of the work. Maintenance personnel did not realize that the wall was a rated fire barrjer barrier or tnat it would be degraded by the open penetration, although a nearby fire door in the same wall was present and easily identifiable. Under a normal work request, a determination by the working department would have been required as to whether a fire hazard review by the fire marshall should be accomplished during the work planning stage. This would have included a review to determine the applicability of DFPP 4175-1, Fire Barrier In~egrity Integrity and Maintenance, and DFPP 4175-2, Operating Fire Stop/Break StoplBreak Surveillance. However, a blanket work request bypassed these types of controls. Approximately 24 hours elapsed between the time the piping was removed from the penetration and discovery by the Station Manager. During this period of time, an hourly fire watCh, watch, although required as a result of the inoperable penetration by Dresden Administrative Technical Requirement (DATR) 3.1.6.1, did not exist. The DATRs were first implemented (OATR) on August 29, 1989, to incorporate fire protection requirements that were deleted from technical specifications as described in Paragraph 7.b.2 of this report.

A previous event also involving degradation of a fire barrier by maintenance personnel occurred on June 14, 1989. Failing to ( recognize a fire barrier, workers routed a welding cable and air hose through an unducted ventilation opening in the fire wall separating the Unit 3 east LPCI room and the Unit 3 HPCI room. This prevented closure of -an automatic vertical fire damper in the ventilation opening. The technical specification requirement in effect at that time required a continuous fire watch to be inoperable fire barrier established within one hour due to the inoperabl~ penetration. This was not established until the degradation was discovered three days after it occurred. This event was described in inspection report No. 50-237/89017; No. 50-249/89016. NRC review indicated that this previous event met the criteria of 10 CFR 2, Appendix C, and thus no notice of violation was issued at that time. Corrective action to prevent recurrence involved marking of unducted ventilation openings in fire barriers to make them more recognizable and, therefore, was very specific to that event. This corrective action also was not complete at the time of this latest occurrence in that of five identified unducted ventilation openings in fire barriers marked. The remaining only one had already been appropriately markee. were to be comp1eted completed during the December 1989 Unit 3 refueling outage. This action did not address the broader aspects of maintenance personnel recognition of fire barriers in general and, therefore, could not have prevented this latest occurrence and) even if it had been completed. ( 12 III. II 1. 13-15

Revision 8 April 1992 Further review by the licensee determined that the rated .fire ( the "piping assembly penetration had been degraded even prior to the~piping removal. The penetration was originally installed in 1982 1982.. .. However, at some date between April. 1,I, 1985 and July 8, 1985 sections of the piping including the portion going through the penetration were replaced with polyvinyl chloride (PVC) plastic piping, a non-approved material for fire barrier penetrations. Plastic materials will burn with an intensity and heat production in a range similar to that of ordinary hydrocarbons. In addition, when burning, they produce heavy smoke that obscures visibility and can plug air filters. The halogenated plastics also release free chlorine and hydrogen chloride when burning. burning, which are toxic to humans and corrosive to equipment. The work request under which this change was completed indicated that no fire hazards review was necessary. The design drawing, fire barrier location drawing F-88, failed to identify the penetration. Drawing F-88 was inspected by the architect-engineer (AE) for fire barrier drawing development on February 14, 1985. This inspection was to identify all Februa~ penetrations in the fire wall including both mechanical and electrical penetrations. In addition, performance of surveillance DFPP 4175-2 failed to identify the existence of the penetration. This surveillance, required to be performed on an 18 month cycle, contained specific instructions to enter data on the Operating Fire Stop Surveillance Log and initiate a drawing change request for the ( pr.otection drawing if a fire barrier penetration appropriate fire pr-otection was found that was not on the drawings. Instructions for review of mechanical penetration seals were incorporated into the procedure on December 29, 1986 with Revision 5 of 'of the procedure. Previous revisions required inspection only with respect to electrical fire seal penetration configurations. Inspections per this procedure including those pertaining to mechanical penetration se'als seals were accomplished on February I, 1988 and again on February 1,I, 1989, each time failing to identify the penetration in question. This is considered to be a violation of 10 CFR 50.48(a) (No. 237/89022-02(ORP)} 237/89022-02(ORP)) in that the licensee failed to control the design for this fire rated assembly (fire wall). The penetration was not identified during Appendix R walkdowns, was included on fire protection drawings. not im:.luded drawings, and was not identified through the fire protection surveillances on the fire barrier. Furthermore, the fire rated wall was degraded in 1985 by installation of combustible PVC piping and again recently with remO'iaf of the piping. Each time~ complete removaf time, the effect on the fire barrier was not properly analyzed or considered. The cause of the more recent degradation of the fire barrier was, in fact, similar in nature to a fire barrier degradation which occurred earlier this year. In both, maintenance personnel failed to recognize a fire barrier and, therefore, the effect their actions would have on it.it~ ( 13 l3 IlI.13-16

                                                                                                    --~-   ...-....--.- ... -.,.--.- .... - .... ~-.

Revision 8 April 1992

c. Approach to the Identification and Resolution of Technical Issues From a Safety Standpoint licensee's The licensee 1 s approach to resolution of technical issues in the maintenance area was mixed as demonstrated by the violation associated with the fire barrier degradation as opposed to the actions associated with the EHC DC power failure alarm relay.

Licensee corrective actions to the June 1989 fire barrier degradation. degradation by maintenance personnel,. in retrospect retrospect, proved to be too narrow in t scope to prevent another fire barrier degradation. Upon discovery of the later degraded fire barrier described in Paragraph S.b.8 of this report, the licensee initiated an hourly fire watch. A temporary fire seal was installed on October 26, 1989, 19B9 t and a permanent seal was installed on November 17, 19B9, 1989, when proper materials were available. The decision to wait for better conditions prior to replacing the EHC DC power failure alarm relay as described in Paragraph S.b.l of this report was an example of a regard for minimization of unplanned transients. In this way, if a main turbine trip would result from the activity, it would not also cause a reactor scram. The inspectors also noted that instrument maintenance personnel troubleshooting the problem were highly knowledgeable of detailed EHC system circuitry design. Licensee actions taken in response to the main steamline low pressure switch failures was regarded by the inspectors to be a good attempt to identify the specific problem and resolve it.

d. Responsiveness to NRC Initiatives

( The licensee's timeliness of control room work request completions continued to be in response to NRC concerns. To ensure prompt resolution of such problems the licensee revised Op~rat;ons Operations Department Policy Statement Number 16. This statement established a white work request sticker for the control room to be used in addition to the existing salmon colored stickers. A salmon sticker was to be used to identify problems with control room indications such that the operator could no longer believe the indication or the indication was no longer available. A white sticker was used to identify problems that required corrective maintenance but control room indications were not affected. Salmon stickers were to receive a 8-1 priority which required work to start within 24 hours if parts were available.

e. Observation of Surveillance Activities The inspectors observed surveillance testing required by Technical for tne Specifications fo~ the items listed below and verified that testing was performed in accordance with adequate procedures, that test instrumentation ~~s calibrated, that LeOs LCOs were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne 1.

personnel. ( 14 III.13-17 _- -_

                                                                                                                               --.-.----    .~---

Revision 8B April 1992 The inspectors witnessed portions of the following test activities pertaining to Units 2 and/or 3: Local Power Range Monitor (LPRM) Amplifier Gain Calibration Average Power Range Monitor (APRM) Gain Adjustment Individual LPRM Recove~ Recovery Quarterly Primary Containment Isolation Valve Timing APRM Rod Block and Scram Functional Test Intermediate Range Monitor Downscale Rod Block Functional Test Int*ermediate HPCI Valve Operability.Test Operabi1ity.Test

6. Safety Assessment/Quality Verification (40500)

a. Enforcement History During this inspection period, no violations or deviations were identified in the safety assessment/quality verification functional area.

b. Assurance of Quality, Including Management Involvement and Control Management involvement in assuring quality was evident when the plant manager discovered the degraded fire wall as described in Paragraph S.b.8 S.b.B of this report. The inspectors continued to note frequent and effective tours of the plant by management.

The inspectors observed the monthly performance review meeting c conducted on October 13) 13, 1989. Plant management reviewed items of interest which occurred since the last meeting including engineered safety feature actuations, specific Technical Specification limiting entered, continuous or occurring control conditions for operation entered) room alarms, degraded or out of service equipment and potentially significant events. In addition, the status of the top technical issues was discussed. In order to facilitate greater sharing of information with similar facilities, a representative from the Quad Cities plant was also present. In addition, the meeting was attended by a licensed plant operator who presented his own areas of concern. The inspectors considered attendance by both these individuals to be beneficial toward maintaining management awareness and involvement in relevant issues both internal and external to the plant. Attendance by plant operators also tended to promote greater professionalism and a sense of responsibility among that group. The inspectors also reviewed the monthly status report for the month of October. The inspectors found this to be an excellent management tool for remaining cognizant and identifying trends in various departmental indicators. 15 III.13-18

Revision 8 April 1992

7. Engineering/Technical Support (37828 and 93702)

( B. I. Enforcement History During this inspection period, no violations or deviations were id~ntified in the engineering/technical support functional area.

b. Operational Events (1) The licensee informed the reSident inspectors on October 12, 1989, that they had confirmed a possible single failure that could occur during a loss of coolant accident (LOCA) following a loss of offsite power that could prevent the low pressure (LPCI) swing bus, MeC coolant injection (lPCI) MCC 28-7/29-7 on Unit 2 (MCC 38-7/39.7 3), from pe~orming 38-7/39-7 on Unit 3). perfo~ing its intended LPCI swing bus could be supplied power from .

function. The lPCI either bus 28 or bus 29 on Unit 2 (bus 38 or bus 39 on Unit

3) which in turn were supplied power from OPPOSite oppOSite engineered divisional buses. A low voltage condition safety feature (ESF) div;sional on the lPCI LPCI swing bus was designed to cause an automatic transfer of the bus to the bus supplied from the other div.ision.

di~ision. However, a diesel generator could suffer a voltage regulator failure such that voltage would be too low to properly operate bus leads but not low enough to cause the LPCI swing bus to automatically transfer to the division lPCI supplied by the other diesel generator. The lPCI LPCI injection valves were supplied power from the lPCJ LPCT swing bus. Thus, the LPCI system and one division of core spray would be ( incapable of auto~ti~ automati~ injection in this scenario. This would leave only one core spray pump for automatic low pressure emergency core cooling system (feeS)(ECCS) injection. (2) 16, 1989, the licensee discovered that a DATR On November 16. involving a fire detection instrument had been inadvertently missed. Technical Specification amendment numbers 106 for Unit 2 and 101 for Unit 3 removed the fire protection requirements from technical specifications in accordance with Generic letters guidance presented in Gener;c Letters 86*10 86-10 and 88-12. The DATRs incorporated these technical speCification requirements while also including the fire protection features added during whi1e Appendix R fire protection modifications. This the 10 CFR 50 Appenal~ luded the additic,n included inc additic*n of lCD LCD actions to reflpct reflect the added fire feiOtures. These. protection fea-tures. The!e. technica 1I specification amendments were approved by the NRC an on June 29. 29, 1989. 1989, with 60 days given to implement implEment the change. In preparation for implementation, work reouests were revjewed reviewed by the system engineer and the fire marshall to see if inoperable equipment was affected by the marsna}l DATRs. A total 07 of 25 26 wcrk work requests were identified including one involving the Unit 3 lPCl LPCI room/torus fire detection device which was written on July 26, 1989. The (protectowire) devtce associated DATR 3.1.1.1 lCO LCD action statement required a once per hour fire inspection to be established within one hour. However, the work request review inappropriately identified ( 16 III.13-19

Revision 8 April 1992 ( another action statement which was applicable to the other-work other work requests as also applicable to this work request. This other action statement allowed 14 days to restore the device prior to establishing the fire watch. Thus, when the DATRs became effective on August 29, 1989, the fire watch was not established. On September 12, 1989, when the 14 days expired, the fire watch was established and a deviation report written. The device was repaired and considered operable on September 23, 1989. While reviewing the deviation report on November 16, 1989, the system engineer discovered the error. The inspectors regarded this incident as an isolated occurrence induced by implementation of the new program requirements and a review process which differed from normal practices. The inspectors had not noted any further problems with DATR compliance under normal practices since their implementation, except as described in Paragraph S.b.8 of this report. This exception,however, exception, however, was attributable to a different root cause. (3) -While

              *While assembling work packages to install and calibrate United Electric Temperature switches for main steamline and HPCI steamline leak detection and automatic isolation, the licensee discovered that the model FIOO FI00 switches to be installed were not referenced in the environmental qualification (EQ) binder.

Further review by the licensee on November 14, 1989, indicated that five of the 16 Unit 2 main steaml;ne steamline temperature switches c( were already installed without the proper EQ documentation. One of these was installed in February 1989 and the other four in July 1989. The other Unit 2 main steamline, as well as all steamline Unit 3 main steamline and Units 2 and 3 HPCI steaml;ne temperature switches were properly EQ qualified*model F7 switches. Although the suitability of application previously FI00 switch indicated that it

             . completed by the licensee for the FIOO was EQ qualified, this determination was based on a vendor test report and not on the required EQ binder. This is considered 237/89022-03(DRP)) pending further to be an unresolved item (No. 237/89022-03{DRP))

NRC review of this matter.

c. Approach to The Identification and Resolution of Technical Issues From a Safety Standpoint The licensee's determination of the lPCI swing bus design problem indicated a commitment toward remaining cognizant of industry issues and problems that could be relevant to Dresden. The review that identified this problem was implemented in response to similar deficiencies discovered at other nuclear power plants. licensee subsequent actions included evaluating possible design changes and

       . contacting the facilities with similar identified deficiencies to ascertain their respective courses of action. Two possibilities that were under review included additional protective relays or powering the involved motor control centers with an uninterrupted power supply. The licensee also issued Dresden General Abnormal

( 17 III .13-20 II 1.13-20

Revision 8 April 1992 ( (DGA) Procedure 5, Degraded Voltage on MeC MCC 29-7/28-7 (39-7/38-7) Due to a Failure of the Unit 2(3) Diesel Generator Voltage Regulator During a LOCA/Loss of Offsite Power Event. This procedure required the operator to trip the diesel generator if adequate voltage could not be restored such that the LPCI swirig bus would automatically transfer. trans thiss attempt fa fer. If thi failed, iled, .the operator was instructed to

           .anually transfer the lPCl
           .anual1y                   LPCI swing bus.               .

The inspectors regarded the missed DATR concerning the fire protection protectowire device to be an excellent example of a commitment to self-identification of problems by not only the licensee but also the individual who discovered and reported his Ifcensee own error. The licensee planned to include a *discussion of the incident in station personnel tailgate sessions and in the licensed operator requa Hf1catio" continuing training program. The 11 requalification censee licensee also identified the EQ problem regarding five of the Unit 2 main steamli"e steamline temperature switches. As a result, the licensee completed* equipment qualification variation form 89-023 including a justification for continued operation. An EQ binder was also belng being developed to rectify the problem. The inspectors regarded the lfcenseelicensee investigation, root cause analysis and corrective actions concerning *the HPCl HPCI system backleakage and damaged piping supports, as described 1n in Paragraphs 5.h.4 S.b.4 and S.b.S of this report as an example of aggressive self identification S.b.5 and resolution of problems. The review of elevated room temperatures c and corresponding actions which led to discovery of the feedwater backleakage into the HPClHPCI system was particularly insightful. The system walkdowns used to identify the HPCl HPCI support damage were ve~ detailed and comprehensive. In addition, safety evaluations performed to support alternate HPCr HPCI system standby lineups addressed all relevant issues. Planned litensee licensee actions to determine the root HPCI system valve leakage, to access the effectiveness of cause of HPCl the Inservice Inspection (lSI) program as it applied.to structural supports and to perform similar walkdowns on other systems indicated an excellent attitude toward self-identification and assessment.

d. Responsiveness to NRC Concerns responsive to a regional NRC request pla~t technical staff was responsfve The plant regarding maintenance of shutdown margin requirements for information regardfng during refueling.

8. (90113)

Report Review (901]3) perfod, the inspectors reviewed the l1censee During the inspection perTod, licensee'sl s Monthly Operatin~ Report for October. lhe The inspectors confirmed that the pro¥idec met tbe information pro~ided t~e require~ents of Technical Specification 6.6.A.:; and Regulatory Gui~e 6.6.A.J Gui<lie 1.16. The.The inspectors also reviewed the Test Report Summary and confirmed that it met Unit 2 Cycle 12 Startup' T.est the requirements of Technical Specification 6.6.A.l. ( 18 IIIl.13-21 Il.13-21

Revision 8 April 1992 ( 9. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or deviations. Unresolved items disclosed during the inspection are discussed in Paragrapns Paragrap~s S.b.4, S.b.S and 7.b.3 of this report.

10. Exit Interview (30703)

The inspectors met with licensee representatives (denoted in Paragraph 1) on December 1, 1989, and informally throughout the inspection period, and summarized the scope and findings of the inspection activities. The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the not tdentify any such inspector during the inspection. The licensee did hot

                 ,      documents/processes as proprietary. The licensee acknowledged the findings of the inspection .
              ** 1t c(~.

(.......-.- 19 III.13-22 III. 13-22

Commonwealth Edison One Firs! NatIOnal Plaza. ChIcago. IllinOIS FIrst Nallonal Revision 8B Address Reply to: Post Of1ice ORice Box 767 April 1992 Chicago. illinOIs IllinoIs 60690

• 0767

( January 25, 1990 Mr. A. Bert Davis Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137

Subject:

Dresden Nuclear Power Station Units 2 and 3 Response to Notice of Violation and Inspection Report Nos. 50-237/89022 and 50-249/89021 NRC Docket Nos. 50-237 and 50-249

Reference:

Letter from W.D. Shafer to Cordell Reed dated December 26, 1989, transmitting the subject Inspection Report and Notice of Violation. ( Mr. Davis: Enclosed is the Commonwealth Edison Company (CECo) response to the subject Notice of Violation (NOV) and Inspection Report (rR) (IR) which identified deficiencies in the control of a fire barrier penetration. . CECo understands the significance of the issues involved. Corrective actions have been taken or have been initiated to prevent similar non-compliances from recurring in the future. Please contact this office should further information be required. Very truly yours, ach Nuclear Licensing Manager cc: B.L. Siegel - Project Manager, NRR S.G. DuPont - Senior Resident Inspector, Dresden ( lw/0593T II 1.13-23

Revision 8 April 1992 ATTACHMENT ATTACIIMENT ( COMMONWEALTH EDISON COMPANY Response to Notice of Violation 50-237/89022-02 (DRP) Severity Level IV VIOLATION YIOLATION 10CFR50.48(a) requires that each operating nuclear power plant have a fire protection plan that satisfies Criterion 3 of Appendix A to IOCFR 10CFR Part 50. It further requires that the plan shall describe specific features necessary to implement the plan such as administrative controls to limit fire damage to structures, systems or components important to safety 80 so that the capability to safely shutdown the plant is assured *. Section C.l of the licensee's response to the Guidelines of Appendix A to Branch Technical Position APCSB 9.5-1 as accepted in the 1980 Supplemental Safety Evaluation Report indicates that administrative measures are established to ensure that guidelines of the Branch Technical Position are included in design and procurement documents and that deviations therefrom are con trolled. controlled. Contrary to the above, a penetration in a three hour fire rated wall located in a safety related area of the 570 foot elevation of the reactor building, as prescribed by Section D.l.j of the Branch Technical Position, was not included in design documents and deviations were not controlled. The fire rated wall ( was degraded in 1985 by replacement of the original piping with non-approved polyvinyl chloride plastic piping and was further degraded on October 25, 1989 when the piping was completely removed and the penetration left unsealed. This is a Severity Level IV violation (Supplement I) [No. 237/89022-02 (DRP)]. DISCUSSION The Station's Technical Specifications include a license amendment that requires adherence to the approved fire protection program. This amendment is implemented through the Dresden Administrative Technical Requirements (DATRs) for fire protection. DATR 3.1.6.1.8 3.1.6.1.a requires that a fire watch be posted within one hour whenever a fire rated sealing device is inoperable. Because the investigation into this event established that the fire watch time constraint was exceeded, this event was reported under 10CFR50.73(a)(2)(i)(B) for a condition that is prohibited by the Technical Specifications (LER 89-30/050237).* 89-301050237) The Mechanical Maintenance Department was in the process of dismantling and cleaning an area which was formerly a Control Rod Drive (CRD) maintenance area. This work was being performed under Blanket Work Request No. 208 for general plant cleanup. The work that was to be performed was not intended to disrupt or alter plant components or systems. The work described on the ( IIII II .13-24

Revision 8

                                                                           . April 1992 Blanket Work Request form must be approved by a Maintenance Department.

Supervisor. Prior to commencing on the cleanup work, the Radiation Protection Department surveyed the work area and identified a drain line as a source of high radiation. The drain line was connected to a eRD CRD flush tank and routed through the Unit 2 and Unit 3 Reactor building common wall directly to *a floor drain. Because the CRD flush tank was to be removed per the blanket work request, removal of the drain line was improperly added to the blanket work request job scope. In order to reduce personnel exposure, the drain line was hydrolazed and removed before other work in the area resumed. A substantial portion of the line was hydrolazed hydro lazed and removed between October 24 and *October 26, 1989. On October 26, 1989, the final portion of pipe remaining in the Reactor Building common wall was removed. At approximately 1100 hours, the Maintenance Mechanics stuffed the penetration with rags and left the area. Further investigation* into this incident ~evealed revealed that the drain line penetration was originally installed in 1982 per fire protection requirements for a three hour barrier. However, subsequent to its initial installation, sections of th~ the piping were replaced with Polyvinyl Chloride (PVC) plastic piping including the portion that went through the common wall penetration. Further investigation revealed that the PVC pipe alteration occurred in 1985 when iitt was insufficiently described in the associated Work Request to be identified as involving a fire barrier penetration. The scope of the Work Request as written was to improve the drainline flow by changing the angularity of the pipe. Consequently, it was not identified as Reliability or Regulatory Related. Since that time, the quality of work instructions has been upgraded and all fire protection related work is classified as Reg~latory Regulatory ( Related which requires review by the Fire Marshall as well as Quality Control. The most recently performed Technical Staff Fire Protection Procedure (DFPP) 4175-2, "Operating Fire Stop/Break Surveillance," failed to identify the drain barrier location drawings, which were first issued line penetration. The fire harrier in 1985 following a detailed fire barrier survey, also failed to show the penetration. The DFPP 4175-2 surveillance, which is performed on an 18 month cycle, includes instructions to inspect Appendix R wall and floor fire barriers for evidence of new penetrations or breaches. If an unrated penetration seal or breach in an Appendix R fire barrier is identified, the Operations Department Shift Supervisor is to be notified to implement

 ~ediate immediate corrective actions. The penetration would then be documented in the surveillance procedure, and in the fire barrier location drawings. It is believed that performance of the penetration surveillance was hampered due to the continuing maintenance work in the areas on either side of the wall. The surveillance technicians's line of sight was most likely obscured or obstructed in each case while inspecting the third floor Unit 2/3 Reactor Building wall, thus preventing detection of the drain line penetration.

CORRECTIVE ACTIONS TAKEN AND RESULTS ACHIEVED The immediate corrective actions were notification of Operations Department SuperviSion, and the initiation of an hourly fire watch pursuant to DATR Shift Supervision, 3/4.1.6. The penetration was then sealed with a temporary fire seal in accordance with Dresden Fire Protection Procedure (DFPP) 4175-1, "Fire Barrier ( Integrity and Maintenance. Maintenance." Once the temporary fire seal was inspected and II approved, the fire watch was terminated. Contrary to DFPP 4175-1, however, a IIl.13-25 111.13-25

Revision 8

                                                                              .>April
                                                                              )\pril 1992 permanent seal was not installed within the prescribed seven days. Materials

((

\

to make the repair were not available in time to complete the repair. 'The The Station Fire Marshall, at his discretion, permitted the seven day administrative limit to expire provided that the temporary barrier was intact, and that the permanent barrier was installed as soon a8 practicable. Mechanical Maintenance installed the permanent seal under Work Request 88289 on November 17, 1989. CORRECTIVE ACTIONS TAKEN TO AYOID FURTHER NON-CQMPLIANCES NQN-CQMPLIANCES

1. DFPP 4175-2, will be revised by the Technical Staff to include this fire seal on the surveillance checklist. Abo Also.t to aid in performing the next fire barrier surveillance, surveillance. a Drawing Change Request (DeR) (DCR) will be initiated to identify the fire seal location on fire barrier drawings F-88 sheets 1 and 2. This will be completed by February 28, 1990.

2. In order to make rated fire walls in the plant more easily identifiable, the Technical Staff system engineer will prepare a fire barrier reference guide including plan views of all the fire areas for use by all working departments. A revision to Dresden Administrative Procedure (DAP) 3-1, "Fire Protection Program," will also be implemented to control preparation and updating of the reference guide. This will be completed by July 31, 1990.

3. The Fire Marshall will provide the Training Department with additional training material on fire barriers by February 12, 1990.

(

4. Additional training on fire barriers will be given to the Mechanical Maintenance Department during an upcoming continuing training session. A review of this event shall be included in the material to be presented. Emphasis will be placed on the conservative practice of assuming that all walls, floors, and ceilings in the Reactor and Turbine Buildings are fire barriers unless otherwise specified. This will be completed by May 25, 1990.

5. This event was reviewed in a tailgate meeting for all station personnel on December 21, 1989. The conservative practice described in Item 4 will also be emphasized in additional tailgate meetings for all station work groups, substation construction, and ENC to be completed by February 23, 1990 *.1 J It will be included in entrance training for contractor personnel by May 25, 1990.

6. This event will be reviewed with the Mechanical Maintenance Supervisor and Crew who were directly involved by January 31, 1990.

7. A statement on the appropriate use of the Blanket Work Request system was added to DAP 15-1 by the Maintenance Staff on January 12, 1990.

c.

                                           -   1 l -

III ,13-26 III.13-26

Revision 8 April 1992

8. Precautionary statements will be added to fire barrier surveillance procedures DFPP4175-2 DFPP 4175-2 and DFPP 4175-3 ("Shutdown Fire Stop/Break Surveillance") concerning:

a) improperly modified penetrations, penetrations. and b) removal of obstructions., obstructions.* as appropriate, appropriate. in order to assure that the entire barrier is properly inspected. These procedure changes will be implemented by June 29,

29. 1990, 1990. i.e; i.e.

prior to the next 18 month surveillance.

9. Changes have also been implemented in DFPP 4175-1 to clarify the process by which temporary seals may be approved for longer than seven days. It now provides more detailed installation instructions and inspection frequency requirements to ensure that temporary fire seals provide adequate barrier protection for periods exceeding seven days. This was completed on January 12,12. 1990.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED As described previously, previously. a fire watch was promptly established upon discovery of the degraded fire barrier. The fire barrier penetration opening was then sealed with an approved temporary configuration in accordance with DFPP 4175-1. Once the temporary fire seal was inspected satisfactorily, satisfactorily. the fire watch was terminated. Mechanical Maintenance then installed a permanent seal under Work Request 88289. The permanent seal was then inspected satisfactorily on November 17,

17. 1989, 1989. at which time all actions to achieve full compliance were complete.

( 0593T III .13-27

TAB 14 Revision 8

                                                                          . Apri 1 1992 c                                     DRESDEN 2 &3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspection Report No. 50-237/90017 and 50-249/90017 fm fill                   Title III.14-1              Inspection Report No. 50-237/90017 and 50-249/90017 dated August 24, 1990.

II L 14-26 III. September 24, 1990 CECo letter from T. J. Kovach to Nptice of Violation A. Bert Davis (NRC), Response to Notice and Inspection report No. 50-237/90017 and 50-249/90017. IIL14-30 II1.l4-30 November 28, 1990 NRC letter from A. Bert Davis to (CECo) proposed Imposition of Civil Penalty. C. Reed (CECa) ( ('-- III.14-i

                                                                     '10017  Itto~/7
                                                                     '1001/ l,?ot!>/7 UNITED STATES                           Revision 8
                   ---,,~~-NUCLEAR
                   -  - - _ -NUCL.EAR REGULATORY COMMISSION April 1992
                            .'                 REGION III "t

7" ROOSEVELT ROAD GLEN ELLYN, ILLINOIS '0137 101:17 6'1e 28 -:---

                                ~ ..

AUG 2 4 1900 Docket No. 50-237 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President

• Opus West IIIIII 1400 Opus Place Downers Grove, Il 60515 Gentlemen:

This refers to the routine safety inspection conducted by S. G. Ou Du Pont, o. D. E. Hills and M. S. Peck of this office on June 13 through July 31, 1990, of activities at Dresden Nuclear Power Station, Units 2 and 3, authorized by Operating Licenses licenses No. DPR-19 and No. DPR-25 and to the discussion of our findings with Mr. J. Kotowski and others at the conclusion of the inspection. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. During this inspection, certain of your activities appeared to 'be -be in violation of NRC requirements, as described in the enclosed Notice. These activities were reviewed against the criteria of 10 CFR 2 Appendix C, Section V.G.1 V.G.l for exercise of discretion, but were not deemed applicable due to the similarity of root causes and adequate length of time for effective corrective action between two of the examples. A written response is required. An unresolved item described in the enclosed inspection report is awaiting completion of licensee 10 CFR 50.59 safety evaluations regarding specific past practices of drywell manifold sampling system usage. The NRC plans to review these safety evaluations upon their completion prior to resolution of this item. In accordance with 10 CFR 2.790~ of the Commission's Regulations, a copy of enclosure(sJ will be placed in the NRC Public Document this letter and the enclosure(s} Room. The responses directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Pl96-511. PL96-511. c_ ( III.14-1

Revision 8 April 1992 Commonwealth Edison Company 2 AUG 2 4 1990 We will gladly discuss anY any questions you have concerning this inspection. Sincere 1y, ly, tf. F!~-&k.t 0/. f.~-&:kt Reactor Projects Branch 1 .

Enclosures:

1. Notice of Violation

2. Inspection Report No. 50-237/90017(DRP);

No. 50-249/90017(DRP) w/enc10sures: cc w/enclosures: D. Galle, Vice President - BWR Operations T. Kovach, Nuclear Hanager Licensing Manager E. D. Eenigenburg, Station Manager Hanager DCD/DCB (RIDS) DeD/DCB Licensing Fee Management Branch licensing Resident Inspectors LaSalle, Dresden, Quad Cities Ri chard Hubbard Richard f1cCaffrey, Ch ii ef, Pub 1ii c J. W. ficCaffrey, Utilities Division Robert Newmann, Office Offi ce of Pub lie 1i c Counsel, State of Illinois Center Counsel.

                                          !II .14-2 III.14-2

                                                                               *o Revision 8 April 1992 Appendix

( '.~ ....-. NOTICE OF VIOLATION Commonwealth Edison Company Docket No. 50-237 Dresden Nuclear Station Docket No. 50-249 As a result of the inspection conducted on June 13, through July 31, 1990, and in accordance with the General Policy and Procedures for NRC Enforcement Actions, (10 CFR Part 2, Appendix C), (1990) the following violation was identified:

• 10 CFR 50, Appendix B, Criterion V, as implemented by Commonwealth Edison Company's Quality Assurance Program, requires that activities.affecting quality be prescribed by documented instructions, procedures or drawings of a type appropriate to the circumstances.

Contrary to the above, documented instructions for activities affecting quality prescribed in equipment outage checklists were inappropriate to the circumstances in the following cases:

a. Outage number 111-460 implemented on February 4, 1990 failed to recognize all consequences of a fuse removal, resulting in an unexpected group II primary containment isolation, standby gas treatment system automatic initiation and reactor building ventilation system isolation.

  /. b. Outage number 11-412 implemented on June 11, 1990 prescribed the closure

(( of incorrect valves, resulting in an unexpected recirculation pump trip.

c. Outage number 11-421 II-421 implemented on June 13, 1990 failed to*too recognize all consequences of opening a breaker, resulting in an unexpected half group II primary containment isolation signal.

This is a Severity level Level IV violation (Supplement I). (237/90017-02(DRP)) Pursuant to the provisions of 10 CFR 2.201, you are requirec requireOd to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply, including for each violation: (1) the corrective steps that have been taken and the results achieved; aChieved; (2) the corrective steps that will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown. Reactor Proj cts Branch 1 III.14-3 II I.14-3

Revision 8 April 1992 ( COMI1ISSION U. S. NUCLEAR REGULATORY COM11ISSION REGION II I Reports No. SO-237/90017(DRP); 50-237/90017(DRP); 50-249/90017(DRP) Docket Nos. 50-237; 50-249 license Nos. DPR-19; DPR-25 DPR-25' Licensee: licensee: Commonwealth Edison Company P. O. Box 767 Chicago, Il 60690 Facility Name: Dresden Nuclear Power Station, Units 2 and 3 Inspection At: Dresden Site, Morris, IL Il Inspection Conducted: June 13 through July 31, 1990 Inspectors: S. G. Du Pont D. E. Hills H. S. Peck

                    ~1.

Approved By: IB Date ( Inspection Summary eriod of June 13 throu h Jul 31, 1990 (Re arts orts Nos. 9 DRP ; No. - 49 9 1 P Areas Insaected: Routine unannounced resident inspection of previously identifle Inspection items, licensee event reports, plant operations, maintenance/surveillances, engineering/technical support and report review. Results: . o One violation was identified involving three examples of inadequate equipment outage checklists. Two of these examples had similar root causes although an adequate length of time to implement effective corrective actions had occurred between these two examples. Therefore, this item was determined not to fit the criteria for exercise of discretion under 10 CFR 2, Appendix C, Section V.G.l. Although the results of the individual examples were of minimal safety significance, taken in aggregate the inspectors considered them to be indicative of prob*em prob,em in control of this area and thus possible precursors to a more serious event. o Three unresolved items were identified. The issue involving the drywelldrywel1 manifold sampling system as described in paragraph 6.b was awaiting licensee completion of 10 eFRCFR 50.59 safety evaluations to address specific past practices in the usage of this system .. *..~he The issue involving components from three systems not appropriately included in the primary c IIl.14-4

Revision 8 April 1992 containment local leak rate testing-program as described in paragraph 6.c further review by NRC regional specialists. The issue was awaiting fUrther involving the facility's compliance with 10 CFR 50.62, anticipated rule, as described in paragraph 6.d was awaiting transient without scram rule. further NRC technical review of design calculations and post-modification testing. o Two non-cited violations were identified which both involved missed fire watches occurring approximately one month apart as described in paragraphs 4 and S.a.4. 5.a.4. However, root causes were sufficiently dissimilar such that corrective actions from the first event could not reasonably had been expected to prevent the second eyent. Therefore, these violations were not cited in accordance with 10 CFR 2, Appendix C, Section V.G.!. V.G.1. o A loss of condenser vacuum event which nearly resulted in a reactor scram is described in paragraph S.a.5. 5.a.6. Although operator actions were sufficient to mitigate the event, it was noteworthy that this event, was precipitated by balance of plant equipment failures. The licensee initiated actions to prevent similar failures in related equipment. The inspectors are continuing to follow the balance of plant equipment maintenance area to ascertain the potential for significant events and the affect upon safety-related equ ipment. equipment. o Operations continued to be good as indicated by the operator response to events exhibited during the loss of condenser vacuum event. Additional

...... concerns regarding the adequacy of equipment outage checklists was viewed as a weakness in the maintenance program. Until resolution of the

\. unresolved items in the engineering/technical support area, this area is considered indeterminate. 2

                                               .14-5 III.14-5 III

Revision 8 Apri April1 1992 DETAILS ( , (

1. Persons Contacted Commonwealth Edison Company E. Eenigenburg, Station Manager

            *L. Gerner, Technical Superintendent E. Mantel, Services Director O. Van Pelt, Assistant Superintendent - Maintenance
            *J. Kotowski, Production Superintendent                  *r J. Achterberg, Assistant Superintendent - Work Planning
            *G. Smith, Assistant Superintendent - Operations
            *K. Peterman, Regulatory Assurance Supervisor W. Pietryga, Operating Engineer M. Korchynsky, Operating Engineer Korchyns~,

B. Zank, Operating Engineer Zank. J. Williams, Operating Engineer R. Stobert, Operating Engineer

              ~1. Strait, Technical Staff Supervisor l.

L. Johnson, Quality Control Supervisor J. Mayer, Station Security Administrator D. Morey, Chemistry Services Supervisor D. Saccomando, Health Physics Services Supervisor

            *K. Kociuba, Quality Assurance Superintendent
            *R. Falbo, Regulatory Assurance Assistant

('(. *0.

            *l.
            *L.

Lowenstein, Regulatory Assurance Assistant Sebby. Sebby, Station Maintenance Supervision

            *R. Whalen, Assistant Technical Staff Supervisor Whalen.

The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shift engineers and foremen; foremen~ electrical, mechanical and instrument personnel, and contract security personnel.

            *Oenotes
            *Denotes those attending one or more exit interviews eonducted informally at various times throughout the inspection period. (£.;r  (c:. ..;r 7-3 / -'10) 7-3I-fC)

2. Previously Identified Inspection Items (92701 and 92702)

(Closed) Unresolved Item (50-237/89018-03): licenseeLicensee to resolve atmospheric containment atmosphere dilution/containment atmosphere monitoring (ACAO/CAM) (ACAD/CAM) power supply design deficiency. The ACAD/CAM design is part of the larger hydrogen generation issue currently being handled by the Office of Nuclear Reactor Regulation (NRR) under TAe TAC number 56579/56580. This item is considered closed since the issue is being. reviewed and tracked by other means. (Closed) Unresolved Item (50-237/89005~03): (50-237/89005-03): Evaluate effectiveness of engineered safety features (ESF) actuation reduction program due to the number nUmber of events involving undervoltage testing. During the December 1988 through February Februa ry 1990 Unit Un it 2 refueling refue ling outage, a *tota~otal1 of 12 (. 3 II1.14-6 rrI.14-6

Revision 8 April 1992 ( unplanned ESF actuations occurred. Primarily due to the efforts of,the program, this number was reduced to only three during scram/ESF reduction program. the more recent December 1989 through February 1990 Unit 3 refueling outage. In particular, the licensee investigation of near misses, and'ha1f isolations, resulted in numerous actions including half scrams and'half to address this issue., The inspectors have no further concerns in this area. ' (Closed) Open Item (50.237/90003-01): (50-237/90003-01): licensee to complete a 10 CFR 50.59 safety evaluation to determine whether an unreviewed safety question exists in regard to the single failure analysis for a turbine pressure regulator failure. Section 11.2.3.2 of th~Final th~Fina1 Safety Ana lysis Report (FSAR) indicated that a pressure regulator failure ,in in the wide open direction would result in a 100 psi vessel pressure drop in the first 'ten seconds resulting in a Main Steam Isolation Valve (MSIV) closure at less than 850 psi reactor pressure. A scram would result from the MSIV closure and depressurization would be stopped due to the isolation. However, with reactor water level initially near the top of the range allowed by the operating procedures, the reactor water level swell due to the single failure could cause a turbine trip on high reactor water level prior to reaching 850 psi reactor pressure. In the condition where reactor power was greater than 40 percent, the reactor would scram due to the turbine trip. The MSIV automatic closure was bypassed when the mode switch was not in the RUN position. If the control room operator immediately placed the mode switch to the shutdown position following the scram in accordance with instructions in the abnormal operating procedures, the MSIV closure would not occur at 850 psi. The FSAR analysis analysiS did not account for the possible turbine ( trip if reactor water level were assumed to be near the top of the allowed operating range. The licensee completed a safety evaluation dated May 10. 10, 1990, regarding the FSAR discrepancy. This evaluation concluded that the pressure regulator failure at high reactor water level was bounded by existing design, the licensee plant failure analyses. Because of plant specific deSign, concluded that vessel overfill was not a credible event and that vessel cooldown i~ the plant1s coo1down would not exceed the limitations addressed iii plant's design basis. The inspectors no longer have a concern as to whether this failure at high reactor water level constitutes an unreviewed safety question. The licensee planned to incorporate the results of the safety evaluation into the next FSAR update. No violations or deviations were identified in this area.

3. licensee Event Reports (lER) Followup Fo11owup (90712 and 92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications. ( 4 III.14-7 III .14-7

   .                                                                       Revision 8 April 1992 (Closed) lER 50-237/90003: Partial Group II Primary Containment

( Isolation and Standby Gas Treatment Initiation Due to Personnel Error. This event and corresponding corrective actions are discussed in report.*.. paragraph S.a.l of this report No violations or deviations were identified in this area except as S.a.l of this report_. identified in paragraph S.a.1 report.

4. Plant Operations (71707, 60710 and 93702)

The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operator$ during this period. The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of Units 2 and 3 reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. Each week during routine activities or tours, the inspector monitored the licensee's licensee IS se.curity program to ensure that observed actions were being implemented according to their approved security plan. The inspector noted that persons within the protected area displayed proper photo-identification badges and those individuals requiring escorts were escorted. The inspector also verified that checked vital areas properly escorted; were locked and alarmed. Additionally, the inspector also verified that observed personnel .and packages entering the protected area were searched ( by appropriate equipment or by hand. The inspectors verified that the licensee's radiological protection program was implemented in accordance with facility policies and programs and was in compliance with regulatory requirements. The inspectors also observed new fuel receipt and inspection for the upcoming Unit 2 refueling outage. The inspectors reviewed new procedures and changes to procedures that were implemented during the inspection period. The review consisted of a verification for accuracy, correctness, and compliance with regulatory requirements. req ui reme nts

• These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedures.

In addition, the following operational occurrence was reviewed: On May 14, 1990 the Unit 3 reactor building low pressure coolant injection (lPCI~ (lPCIl rooms/pressure suppression chamber fire alarm light actuated on local fire panel 2223-114 and the device 34-29 (Unit 3 reactor building lower elevation protectowire) was shown in the alarm operators attempted condition on the control room fire alarm typer. The ~perators unsuccessfully to reset the alarm and performed an inspection of the unsuccessfu11y 5 III.l4-8 III.14-8

Revision 8 April 1992 area to ensure that no fire actually existed. When the alarm would;not

 , reset the operators assumed equipment failure was preventing the reset 1 and a work request was submitted for repairs. In actuality, the how to reset this particular alarm and operators did not understand how'to the protectowire device could have functioned if it had been correctly reset. The* alarm response portion of Dresden Fire Protection Procedure (DFPP) 4185-1, "Xl-3 (OFPP)           *Xl-3 Fire Detection System Operation Operation"'t was referenced for required actions. However, this procedure had not been updated to indicate the requirements of the Dresden Administrative Technical Requirements (DATR). The DATRs were developed and went into effect in August 1989 to contain the previous fire protection required actions upon their removal from Technical Specifications and.other and. other 10 CFR 50 Appendix R requirements.

These requirements were removed from Technical Specifications in accordance with Generic letters 86-10 and 88-12. The DATRs were in many cases more extensive and stringent than the previous Technical Specification requirements. DFPP 4185-1 still contained the previous Technical Specification requirements which did not address this device. Therefore, no further actions were taken. Approximately eight hours later an equipment operator on the next shift while performing rounds noted the local light in the alarm condition and notified the control room. An inspection of the area was performed and the alarm was correctly reset. As SUCh, such, a period of approximately eight hours existed in which the alarm was not reset and would not have been able to provide notice of an actual fire if one occurred. DATR Section 3.1.1.1.a required an hourly fire ( watch to be established in the lPCI rooms and a once per shift fire watch to be established in the pressure suppression area within one hour of accomplished during finding this device inoperable. This action was not accompliShed the eight hours. Further review indicated that DFPP 4185-1 was not among the fire protection procedures that had been updated when the DATRs were instituted. At that time, the fire protection procedures were reviewed to determine the effect of the changed requirements and 24 procedures were revised as a result. pro*tection procedures However, it was determined that the remaining fire protection could be revised at later dates in accordance with the procedure upgrade program. The majority of these procedures were surveillances with references to the previous applicable Technical Specifications. However, DFPP 4185-1 also contained the alarm response procedures for the XL-3 Xl-3 fire detection system, contrary to what the procedure title would seem to imply as to scope limits of the procedure content. Therefore, this review did not identify that DFPP 4185~1 4185-1 should also have been changed prior to implementation of the DATRs. In addition, DFPP 4185-1 did not contain specific directions on how to locally reset this particular alarm. Since the operators could not reset the alarm, they incorrectly assumed that the alarm.was inoperable. Failure to perform the required fire watches was considered to be a violation of Technical Specification 6.2.A.ll 6.2.A.11 which required adherence to the fire protection program implementing procedures (50~237/90017-01{DRP)). (50-237/90017-01(DRP)). However, the criteria of 10 CFR 2, Appendix C, Section V.G.1 V.G.l for discretionary enforcement was determined to be applicable and therefore no notice of violation is being issued *. 6 II I.l4-9 111.14-9

Revision 8 April 1992

                                                                             .April

( As a result of this event, the licensee instituted a temporary change to DFPP 4185-1 to ensure proper reference to the DATR requirements and appropriate local reset methods. A permanent revision was planned after the Operational Oper.ational Analysis Division completed reviewing alarms on the Xl-3XL-3 computer for identification. The licensee also reviewed the remaining fire protection procedures to ensure that they did not req~ire require immediate changes. Although training had been given to the operators regarding th~ the DATRs when they were first instituted, the licensee determined that further training was advisable in light of deficiencies in operator knowledge exhibited by this event. Therefore, the licensee counseled the involved individuals to ensure their awareness of the requirements, wrote daily orders to operations personnel to address this.issue this. issue and planned to include further training in the operator requalification program. The licensee was also reviewing possible causes of the spurious linear heat detection alarm and the system engineer was monitoring tbe performance of the linear heat detection equipment. Due to a subsequent spurious alarm, a work request was written for maintenance to troubleshoot the problem if it should reoccur. A temporary change was made to DFPP 4185*1 4185-1 to instruct the operators to contact electrical maintenance to perform this activity prior to resetting the alarm. No violations'or violations*or deviations were identified in this area except for the non-cited violation described above.

5. Maintenance and Surveillances (62703, 61726, and 93702)

a. Maintenance Activities

( Station maintenance activities of systems and components listed below were observed or reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications. The following items were considered during this review: The limiting Conditions for Operation (leOs) (lCOs) were'met were"met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemented. Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may affect system performance. (1) On February 4, 1990, while performing eqUipment equipment outage number r1r-460, a Unit 3 partial group II primary containment 111-460, isolation unexpectedly occurred initiating a*standby a'standby gas treatment system (SGTS) automatic start and reactor building ( 7 III.14*10 111.14-10

Revision 8 April 1992 ( ventilation (RBV) system isolation. The fuse removed during the equipment outage was replaced and the isolation reset. SGTS and the RBV system were returned to normal. norma 1. Further review indicated that the outage was being performed in accordance with work request 090128 to allow replacement of a broken terminal point on control room panel' 903-4. The fuse was removed in accordance with the outage checklist. The equipment outage checklist for outage number 111-460 was inappropriate in that it described removing a fuse which caused the event. The review of the outage by maintenance and operations personnel (including two Senio~Reactor Senior, Reactor Operators) was inadequate in that it failed to identify all effects of removing the fuse. The incorrect equipment outage checklist is considered to be an example of a violation 150-237/90017-02A (ORP>> (ORP)) regarding inappropriate instructions. Safety significance of the resulting action was minimal since the system failed in the safe direction. A review of the drawings and interviews with involved personnel indicated that although the electrical drawings were correct and were reviewed, these individuals did not identify the detailed information on the drawings regarding the pur.pose purpose of the relays which caused the event. Individuals clearly understood how to read the drawings. As a result, all SROs received additional training in the continuing training program on the importance of reviewing the detailed information supplied on drawings for individual components. This was accomplished during the 6 week Cycle 4 IS, 1990. This event training which was completed on June 15~ was also reviewed with the work analysts as part of a reading package completed on May 30, 1990, to stress the* importance of reading all information supplied on drawings with respect to individual components and allowing an adequate amount of time to review the drawings. In addition, the licensee planned on providing additional training to licensed operators stressing the importance of taking adequate time to review the drawings. The licensee also planned to review the SGTS initiation logic to determine possible improvements to circuits with single fuse initiation capability. These last two actions had not been eqUipment completed prior to the two events involving inadequate equipment outage checklists discussed below. In retrospect, these actions were not adequate or timely enough to prevent two other examples of inadequate equipment outage checklists approximately four months later as described in the following paragraphs. Only one of these other examples, however, was related to the same root cause as this event. (2) II, 1990, Unit 2 recirculation pump A tripped while On June 11, performing outage number 11-412 for the recirculation pump B motor-generator (MG) oil cooler temperature control valve 2-3905-B. This was caused by an MG set trip on high (TCV) 2-3905-8. coupling temperature when recirculation pump A MG oil cooler TeV 2-3905-A was mistakenly taken out of ser.vice instead. TCV 2-390S-A TCV bypass in preparation for removing The throttling of the TeV ( 8 III. 14-11 111.14*11

Revision 8 April 1992 ventilation (RBV) system isolation. The fuse removed during ( the equipment outage was replaced and the isolation reset. SGTS and the RBV system were returned to normal. Further review indicated that the outage was being performed in accordance with work request 090128 to allow replacement of a broken terminal point on control room panel' 903-4. The fuse was removed in accordance with the outage checklist. The equipment outage checklist for outage number 111-460 was inappropriate in that it described removing a fuse which caused the event. The review of the outage by maintenance and operations personnel (including two Senior. Reactor Operators) was inadequate in that it failed to identify all effects of removing the fuse. The incorrect equipment outage checklist is considered to be an example of a violation 150-237/90017-02A (ORP)) regarding inappropriate instructions. Safety significance (ORP>> of the resulting action was minimal since the system failed in the safe direction. A review of the drawings and interviews with involved personnel indicated that although the electrical drawings were correct and were reviewed, these individuals did not identify the detailed information on the drawings regarding the purpose of the relays which caused the event. Individuals clearly understood how to read the drawings. As a result, all SROs received additional training in the continuing training program on the importance of reviewing the detailed information supplied on drawings for individual components. This was accomplished during the 6 week Cycle 4 ( {' IS, 1990. This event training which was completed on June 15, was also reviewed with the work analysts as part of a reading package completed on May 30, 1990, to stress the importance of reading all information supplied on drawings with respect to individual components and allowing an adequate amount of time to review the drawings. In addition, the licensee planned on providing additional training to licensed operators stressing the importance of taking adequate time to review the drawings. The licensee also planned to review the SGTS initiation logic to determine possible improvements to circuits with single fuse initiation capability. These last two actions had not been completed prior to the two events involving inadequate equipment outage checklists discussed below. In retrospect, these actions were not adequate or timely enough to prevent two other examples of inadequate equipment outage checklists approximately four months later as described in the following paragraphs. Only one of these other examples, however, was related to the same root cause as this event. (2) II, 1990, Unit 2 recirculation pump A tripped while On June 11, performing outage number 11-412 for the recirculation pump B motor-generator (MG) oil cooler temperature control valve 2-3905-B. This was caused by an MG set trip on high (TCV) 2-3905-8. coupling temperature when recirculation pump A MG oil cooler TeV 2-3905-A was mistakenly taken out of service instead. 8 II1.14-12 JII.14-12

Revision 8 Apri)1 1992 Apri The throttling of the TeV bypass in preparation for removing Tev TeV 2-3905-B had been accomplished prior to this activity. Further review indicated that the equipment outage checklist for outage number 1I-412 11-412 was incorrect in that it listed the iSolation valve numbers (2-3909~501 isolation (2-3909-501 and 500) for the recircu lation pump B ~IG set TCV TeV instead of the iSolation va lve numbers (2-3940-501 and 500) for the intended recirculation pump A MG set TeV. The incorrect eqUipment equipment outage checklist is considered to be an example of a violation (50-237/90017-02B(ORP) (50-237/90017-02B(ORP)) regarding inappropriate instructions. Safety significance of the resulting action was minimal since the sys~em failed in the safe direction. The applicable critical drawing (M-22) in the"control room, indicating the correct configuration found in the plant, had been corrected to reflect drawing change request (OCR) 89-106. The change request was submitted on August 29, 1989, and was still outstanding. The critical drawing in the shift engineer's office, which was not updated to OCR 89-106, was used in preparation of the outage. This drawing incorrectly showed the TeV for the recirculation pump B MG set oil coolers to be TCV TCV 2-3905-A. Dresden Administrative Procedure (DAP) (OAP) 2-9, "As-Built Critica 1 Drawings," Orawings ," covered only the hard copy up-to-date as-built drawings in the control rbom. These were provided for operating shift and maintenance personnel for shift decisions, deciSions, outage management and trouble-shooting. The critical drawings in the shift engineer's office were not "as-built" critical drawings and, as such, should not have been used to prepare or review the ( outage without reference to the control room drawings. Control room drawings were updated by hand when drawing change requests were received by the station. The revised drawings for the shift engineer's office were issued through engineering and could take up to six months or more after the change request was issued. DAP OAP 3-5, "Out-of-Service and Personnel Protection Cards", prescribed that "only the controlled crftica critical1 plant piping and instrumentation diagrams, electrical prints card file or Central File shall be utilized for reference to accurately identify the pOints of isolation." This was misleading since although the points drawings in the shift engineers satellite file were controlled, they did not in fact, directly reflect pending drawing change requests. DAPOAP 2-3 "Operation and Control of the Central and Satellite Sate 11 ite Files," required the appropriate satellite file aperture card to be marked "Revision Pending." This would signify that additional information was needed which could be obta i ned on the "a obtained s-bu i1 t" contro "as-built" control1 room copy or in Cen tra 1 Central File. In this case, the outage was prepared from a set of drawings which were not up-to-date and the additional information was not obtained from the Control Room or Central File. Interviews with operating personnel indicated that there 9 III. 14-13

Revision 8 Apri April1 1992 ( was confusion as to which set of drawings could be used for each type of drawing. In addition, the equipment' attendant (EA) (fA) knew that rev TCV 2-3905-B was to be taken out-of-service but did not question the isolation valves listed on the equipment outage checklist., Upon noticing that the isolation valves listed on the outage matched the "A" rev TCV instead of the "8"8" rev, 11 TCV, the EA "A"n fA hung the outage on the IIA TeV TCV isolation valves. The NSO observed the rapidly increasing temperatures on the computer display and the Shift Supervisor fA returned to the MG sets. There was insufficient time and EA for these individuals to take action since.only since.on1y ten minutes elapsed from the beginning of the increasing temperatures to the pump trip. tri p. As a result of this event, Operations Department Oepartment memorandum No. 18 was issued on June 26, 1990, which described this event. Specific guidance was included to assist in performing the self check process. It also stressed that if a question or uncertainty exists that the Shift Supervisor should be contacted for assistance. Finally, it gave specific guidance as t*o to whi ch set of drawings to use for outage preparation. which (3) On June 13, 1990, a half group II isolation signal was received on Unit 2 while performing outage number 11-421 for work request D89780. 089780. This work request involved replacement of non-environmentally qualified terminal blocks with environ-( mentally qualified splices in junction boxes which provided electrical continuity for torus wide range level transmitter 2-1641-58. The half group II isolation signal was caused by a loss of power to drywell monitor B on the drywe11 high radiation monito~ ACAO/CAM panel when a breaker was opened main control room ACAD/CAM during the performance of the out-of-service. The equipment attendant was contacted, the breaker was reclosed rec10sed and the half group II isolation signal was reset. Further review indicated that the equipment outage checklist for outage number 11-421 II-421 was inappropriate in that it prescribed opening 480 volt motor control center 29-3 120 volt distribution panel circuit number 6. Review of the outage by maintenance and operations personnel was inadequate in that it failed to identify all effects of opening this breaker. The incorrect eqUipment outage checklist is (SO-237/90017-02C considered to be an example of a violation (50-237/90017-02e (ORP)) regarding inappropriate instructions. Safety (DRP)) significance of the resulting actions was minimal since the system failed in the safe direction. A review of the drawings and interviews with involved personnel indicated that although the electrical drawings were correct and reviewed, these individuals did not identify the detailed information on the drawings dealing with this function. (The function of an additional wire leading from this breaker on electrical drawing 12E2679A 12f2679A was not determined.) Individuals clear1y clearly understood how to read the drawings. Therefore, the root 10 III.14-14 II 1.14-14

Revision 8

                                                                    .Apri 1 1992
                                                                    .April

( cause of this event involving inattention to detai1 detail, was 7 wast!1e

                                                                     *th.e same as that of the February 4, 1990 event described in "~

paragraph S.a.I. 5.a.1. As a result of this event and its similarity to the previous. event, the licensee planned to develop a self~ch~ck self-ch~ck program consisting of a committee to promote attention to detail and self-checking while performing the task. This committee was to include individuals who were directly involved in these events. (4) On June 17, 1990, the Unit 3 reactor building lPCI LPCI rooms/pressure chamber fire alarm light actuated on local fire panel suppression chamber*fire 2223-114 and device 34-29 (Unit 3 reactor building lower elevation protectowire) was shown in the alarm condition on the control room fire alarm typer. The Center Desk Nuclear Station Operator (NSO) acknowledged the alarm and noted work request sticker 82074 on the typer plexiglass for this alarm. Incorrectly assuming, due to the work request sticker, that the device was known to be inoperable and therefore alreaqy alrea~ handled, the NSO took no other actions. Approximately 17 hours later, another fire protection device alarmed in the trouble condition. While Wh ile resetting this other device, the NSO noticed that device 34-29 was in the alarm condition. An inspection of the affected area was performed to ensure that an actual fire did not exist. Appropriate fire watches were established in accordance with 3.1.1.1.a and the fire marshal was contacted for instructions DATR 3.1.I.1.a on how to reset the local alarm. Although a temporary procedure change to DFPP 4185-1 had been instituted, as a result of the ( previous event discus'sed discu~sed in paragraph 4, to provide these instructions, operating personnel were still unsure of which button to depress in the local fire protection panel. The local panel alarm was reset which allowed the alarm condition to be cleared on the XL-3 computer. At that time, the fire watch was terminated. The crew that discovered this problem and took appropriate action was the same crew that missed the fire watch described in paragraph 4. Therefore, these jndividuals, in particular, had heightened interest to ensure compliance with fire protection requirements. As such, a period of approximately 17 hours existed in which the alarm was not reset and thus would not have been able to provide notice of an actual fire if one occurred. DATR 3.1.1.1.a required an hourly fire watch to be established in 3.1.I.I.a the LPCI rooms and a once per shift fire watch to be established in the pressure suppression area within one hour of finding this device inoperable. This action was not accomplished during those 17 hours. Failure to perform the required fire watches was considered to be a violation of Technical Specification 6.2.A.ll 6.2.A.11 which required adherence to

    ;, __ !he fire protection program implementing procedures
   .      (50~23$/90017-03(DRPJ).

(50-23$/90017-03(DRP)). However, the criteria of 10 eFR CFR 2, Appendix C, Section V.G.1 V.G.! for discretionary enforcement was determined to be applicable and therefore no' notice of violation is being issued. This determination recognized that 11 III.14-15 III.14-!5

Revision 8 April 1992 ( the root cause of this event as discussed below and the event discussed in paragraph 4 were sufficiently dissimilar such that corrective actions from the first event could not reasonably had been expected to prevent the second event. F-urther Further review of this event indicated that the ,root root cause was regard<ing work due to inadequate administrative controls regarding request processing. The work request sticker for this device had been written during the May 14.14, 1990 event described in paragraph 4. Once the device was determined to be operable and event, the work request the alarm was reset during the previous event. was cancelled. However, the corresponding. work request sticker was never removed. This incorrectly led the NSO to believe that there was an outstanding work request against the device. IS-I, "Initiating and Dresden Administrative Procedure (DAP) 15-1, Processing ProceSSing a Work Request, Request,"II placed responsibility for remova removal1 of work request stickers with the originator of the work request. However, no dependable method existed to ensure that the originator was informed of this need in a timely manner. In fact, the licensee found that seven of the 18 work request stickers on the typer plexiglass were no longer valid. These were removed. In addition, DAP 15-5, "Supplemental Maintenance RequestN Request" did not address cancellation of work requests and removal of stickers at all. Supplemental work requests were written for equipment maintained on a routine or repetitive basis which already had outstanding base work requests. As a result, the licensee planned to revise DAP 15-1 and DAP 15-5 to ( require that the work group which requested cancellation of a

       ,          work request remove the corresponding work request sticker.

In addition, a set of daily orders was issued between June 19 and July 2, 1990, to emphasis the importance of DATR compliance and that any new alarm or trouble alarm on the Xl-3 fire system was to be treated as a valid alarm (regardless of work request stickers). It also contained a list of the fire detection devices requiring a fire watch if only the one device were inoperable. As described in paragraph 4, a temporary procedure DFPP 4185-1 was issued to ensure electrical maintenance change to OFPP performed troubleshooting of this alarm upon recurrence prior to resetting. The licensee also planned to conduct a tailgate session covering this event with the operators to stress that there were eight devices listed in the DATRs which alone would require fire watches if inoperable. The establishment of a log for the Xl-3 fire system, similar to the degraded equipment log was planned. This would provide more information than that available on the work request stickers. The log is expected to be established by the end of September 1990. Finally, the licensee was in the process of setting up a committee to assess various problems encountered with the XL-3 Xl-3 fire detection system. This committee was to specifically address concerns of the operators who had been critical of the system. (5) 3D, 1990, Unit 3 was shutdown for a*a< maintenance outage. On June 30 t The shutdown was initiated due to high temperatures between 230

   "'<<                                     12 III.14-16 III .14-16

B Revision 8 April 1992 plate; On and 240 degrees F on the main turbine thrust bearing plate~ June 28, 1990, the licensee reduced power to about 40 percent in an attempt to reduce the thrust bearing plate temperature. The vendor (General Electric) recommended a shutdown on temperatures above 250 degrees F. Since the temperatures could not be reduced with load reduction, the licensee. initiated a maintenance outage. Other major activities completed during the outage include replacement of one control rod drive, replacement of a main transformer bushing, and repairs to recirculation pump seal leakoff line flow instrumentation. Approximately 70 items on the unscheduled outage list were also addressed. Upon investigation of the. main turbine thrust bearing high temperatures, the licensee found damage to the thrust bearing plate. This was replaced. The licensee did not conclusively determine the root cause of the dimage but suspected an improperly placed thermocouple. The unit was restarted on July 4, 1990. (6) On July 1, I, 1990, while attempting to reverse circulating water flow on Unit 2 in accordance with Dresden Operating Procedure (OOP) (DOP) 4400-8 "Circulating Water System Flow Reversal," circulating water flow reversal valves 2-4402A and 2-44038 breakers tripped and the offgas east suction valve 2-54018 result, condenser vacuum decreased to failed to open. As a result. about 24 inches and a half scram on reactor protection system channel B8 was received. The scram setpoint was 23 inches. The operator noted the vacuum decrease and immediately reduced ( recirculation flow to try to maintain condenser vacuum in \\ . accordance with Dresden Operating Abnormal (DOA) Procedure 3300-2 "Loss of Condenser Vacuum." In addition, the flow reversal was changed back to the original direction such that condenser vacuum recovered. The inspectors considered the actions of the control room operators as exhibiting high attentiveness and quick response to changing conditions to prevent a reactor scram. The ASCO solenoid valve body for offgas east suction valve 2-54018 was subsequently changed out after it was determined not to operate. Testing of the molded case circuit breakers for valves 2-4402A and 2-44038 determined that their trip setpoints were too low. The licensee had not conclusively determined the cause for the low trip settings by the end of the inspection period. The trip setting for the breaker for valve 2-44038 could not be adjusted to within acceptable tolerances and so it was replaced. No maintenance history was found on these nonsafety-related breakers. The trip settings on both breakers were reset and returned to service on July 15, 1990. Due to the failure of two of the eight flow reversal valves on Unit 2, the licensee wrote work requests on the remaining flow reversal valves on both units and planned to enter them into the surveillance tracking system for periodic preventative maintenance. Problem analysis data sheets were also initiated to track root cause analysis' of the breaker failures.

 \'-.. ....

13 III.l4*17 IIL14-17

Revision 8 April 1992

b. Surveillance Activities

( The inspectors observed surveillance testing, including required Technical Specification surveillance testing, and verified for actual activities observed that testing was performed in accordance verified that test with adequate procedures. The inspectors also verifi~d instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were accomplished and that test results conformed with Technical Specification and procedure requirements. Additionally, the inspectors ensured that the test results were reviewed by personnel other than the -individual individual directing tbe test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel. The inspectors witnessed or reviewed portions of the following test activities: Control Rod Drive Hydraulic Withdrawal Stall Flow Testing Standby liquid Liquid Control (SlC) (SLC) System Pump Test Quarterly SLC System Pump Test for the Inservice Test Program One violation as described above and no deviations were identified in this area. In addition, one non-cited violation was identified as described above.

6. Engineering/Technical Support (93702)

( a. The inspectors reviewed concerns with control rod drives going to position "02" during scrams. The subject was discussed in length in inspection report 50-237/87007;50-249/87006, and in* a letter to Mr. A. Bert Davis from I. M. Johnson (CECo Nuclear Licensing) licensing) dated July 14, 1987. The original initiator of the NRC concerns was the August 11, 1986 Dresden Unit 2 scram which resulted in 56 control rods stopping at position "02". As noted in the licensee letter and the inspection report, this phenomenon had occurr~d at Dresden since 1971 as well as other BWRs, although to a much lesser extent. This phenomenon was also the object of an NRC safety evaluation issued June 15, 1981. The NRC safety evaluation identified the apparent cause as leakage past worn stop and drive piston seals internal to the drive which allowed scram water to act as a buffer on the drive. This was described as a hydraulic lock occurring because of worn seals and the design of the drive. The design of these drives, associated with BWR classes 3 and 4, had a relative large buffer area and small vent path to slow drives during a scram to prevent internal damage. later Later models did not have this apparent problem because of increased vent paths and reduced buffer area size. General Electric (GE) recommended a revised CRD venting procedure to remove trapped air which could also aid in developing the the drives to prevent phenomenon. GE also recommended cleaning of th~ build up of crud that could also result in drive seal deterioration. c 14 III.14-18

Revision 8 April 1992 ( The safety significance of the phenomenon was nonexistent since.*both the 1987 NRC inspection and 1981 NRC safety evaluation determined that sufficient shutdown margin exists even ~ith all rods inserted only to the "02" position. The licensee began a series of correction actions in 19B7 1987 to reduce or eliminate the 110211 "02" phenomenon. These included incorporating the GE revised venting procedure, cleaning drive tubes during refueling outages, overhauling drives demonstrating the "02 "02"11 phenomenon (indication of seal deterioration) and, if needed, replacing drives with newer models (BWR/6 drives). As a result, Cycle 11 for both units demonstrated a significant reduction. The licensee had replaced or overhauled all of the "0211 "02" drives during Cycle 10 and initiated cleaning of guide tubes. The licensee also replaced all 14 drives in Unit 3 during the last refueling outage. These drives had the following history: C-09, C-12, H-14 and K-12 occurred once. F-05, F-10, L-02and l-OS L-05 occurred twice. G-03 occurred on four occasions. 6-03 The following is a tab,e tab'E of "02'1 "02" occurrence on Unit 2 during Cycle 11. Date "02" Rods "0211 7/12/89 7/12/B9 C-B, C-8, D-10 and K-I0 0-10 K-10 c rj 04/89 01/05/90 01/16/90 C-6, C-6, C-6, 0-10 and K-10 D-10, E-5, E-8 and F-s 0-10, 0-10, E-5, E-B, E-8, E-I0, F-5 E-10, F-S F-5 and F-l1 F-11 As noted in this tab,e, tab'E, the NRC safety evaluation and NRC inspection report, when "02" phenomenon once occurred, the phenomenon wwld wOJ ld more than likely repeat within a cycle. These drives were scheduled to_be to.be replaced during the next scheduled refueling outage on Unit 2. The licensee has also reviewed the status of all CROs in Unit 3 and determined that only 14 of the original 1971 CROs remain installed in Unit 3. These were also scheduled to be replaced with overhauled BWR/6 drives during the next refueling outage in 1991. The licensee was continuing with their efforts to resolve the "02" phenomenon. Although a final resolution had not yet been found, these efforts had significantly reduced the occurrence of the phenomenon. Since the licensee was continuing to place efforts on reducing the occurrence of the phenomenon and these efforts did appear to be effective, the inspector has no remaining concerns iinn th i s aarea. this rea.

b. On June 28, 1990, the licensee informed the resident inspectors of an alteration to the drywell manifold sample systems on both Units 2 and 3 which affected primary containment integrity. The purpose of c the drywell manifold sample system was to provide air samples to 15 I1I.14-19 III.14-19

Revision 8 April 1992 ( identify the location of reactor coolant pressure boundary leaks inside of the drywell. The drywel1 drywell manifold sample system (one for each unit) was designed to ~ake a suction from 22 sample points in the drywell with each half inch sample line having its own two manual primary containment isolation valves (both located outside of primary containment) and a filter-cartridge. filter*cartridge. Flow then passed through a common header from which the sample pump took a suction. Return back to the drywell was provided through a connection to the continuous oxygen monitoring system which discharged to the drywell through two automatic containment isolation valves which closed on a Group II isolation signal. Thus, the drywell manifold sampling . system had automatic isolation only on its discharge. dis~arge. Piping downstream of the manual isolation valves was nonsafety-related (A portion of this passed through a braided flexible hose as opposed to the rest of the system which was hard piped.). 1here There were also four additional lines which actually took a suction from the continuous oxygen monitoring system, as opposed to directly from containment, and therefore had automatic isolation on both, primary containment. both. the suction and discharge (The continuous oxygen monitoring system had automatic isolation on its suction as well as its discharge.) The drywell manifold sample system had been in place since the plant was built. Technical Specification surveillance requirement 4.6.0.1 required drywell air sampling to be performed once per day to detect reactor coolant system leakage. This sample was originally obtained through a continuous atmosphere monitoring system which was replaced by ( another continuous atmosph~re monitoring system in the early 1980s. Automatic containment isolation was provided with these systems. As a backup to these systems the drywell manifold sample system as described above was used. As a secondary backup (in case the permanent pump was inoperable) a temporary sample pump was used as far back in time as 1978 and possibly before. The temporary sample pump was readily available since it was already used to obtain samples from the X-area (steam tunnel) at the same sample rack. The second continuous atmosphere monitoring ~stemsystem was abandoned in 1987 due to problems with moisture intrusion, therefore the drywell manifold sampling system and the temporary sample pump became the primary and secondary methods, respectively, of obtaining the Technical Specification required sample. Use of the temporary sample pump involved breaking the closed loop on the drywell manifold sample system below the sample filter on one of the sample lines, attaching a rubber hose with a quick disconnect fitting, running the hose to the temporary sample pump and discharging the pump exhaust to the reactor building. The setup was typically left unattended while a sample was being taken although automatic isolation was not provided. Obtaining a representative sample required running the system in this configuration for at least 50 minutes but in many cases probably went much longer than this (A subsequent procedure specified a minimum of one hour.). This allowed an unattended and unmonitored path from the drywe11 drywell (primary containment) through the sample line to the reactor building bu il ding (secondary (seconda ry containment). contai nment) . ( 16 III.14-20 III. 14-20

Revision 8 Apri J 1992 April This use of the temporary sample pump in that configuration w~t~ wa~_, contrary to Technica1 Technical Specification 3.7.A.2 which required ,"'.

                                                                        ,~:

maintaining of primary containment integrity when the reactor was critical or the reactor water temperature was above 212 degrees F. (The definition of primary containment integrity required that all manual-isolation manual' isolation valves on lines connecting to containment which were not required to be open during accident conditions be closed.) Therefore, each time the licensee used the temporary sample pump to sample the drywell, the applicable Technical Specification action statement 3.0.A was unknowingly entered. However, due to the length of time thfs this condition would have existed, this action statement would have been exited prior to any anY actual shutQown. Calculations performed by the licensee assuming one open half inch sample line at design accident containment pressure, Pa (48 psig), indicated that the leak rate would be 4.73 percent per day. When 4dded to the Technical Specification 3.7.A.2a(3) allowed leakage of 1.6 percent per day, a total leakage of 6.33 percent per day was obtained. This was compared to the design basis leakage of 2.0 percent per day prescribed in the bases of Technical Specifications *. A 10 CFR 50.59 safety evaluation was never done on this alteration (use of the temporary sample pump) since the original administrative requirements jumpers. When the administrative only applied to lifted leads and jUmpers. requirements expanded to mechanical equipment, no thought was given to an alteration that had been routinely used for years. As such, in recent years each time this temporary alteration was performed it was done contrary to the licensee's administrative procedures. A procedure covering the use of the temporary sample pump did not exist (until 1989 as described below) and thus the problem was not ( caught early on through a-procedure safety evaluation. Use of the temporary sample pump was frequent, especially espeCially in the last couple of years due to recurring problems with the permanent

 . pumps. (The permanent pumps were estimated by the licensee to have been operable only a few weeks over the last year or two and were troublesome even before that.) Due to a non-documented reviewer comment concerning use of the temporary sample pu~p without a procedure, Dresden Radiation Protection (DRP) procedure 1350-3, "Sampling the Drywell Manifold System USingUsing the Radeco Air Sampler" was first issued in May 1989. This was a missed chance to detect the problem since a 10 CFR 50.59 safety evaluation should have been performed; however a safety evaluation was not performed. The screening criteria in effect at the time allowed entire categories of procedures (such as DRPs not related to effluent monitoring) to be automatically ruled out for a safety evaluation as long as they were not new or changed "procedures or administrative controls" described in the FSAR or Technical Specifications. In this particular case, since it was a new procedure, the criteria required a safety evaluation to be performed. However, the reviewers mistakenly used the wrong administrative path as if it were a revision to this type of procedure instead of a new procedure. Therefore, a safety evaluation was not performed due to a failure to follow administrative requirements. However, the criteria themselves were still inappropriate since the licensee could*coul~*have have instead 17 IIII 1.14-21 I.l4-21

Revision 8 April 1992 just made a reVls10n reVlslon to DRP 1350-7, 1l0 peration of the Unit 2(3) .

                                            "Operation

( Drywell Air Sampling Manifold System" to allow usage of the . temporary sample pump. In that case, the licensee's administrative requirements would not have required a safety evaluation to be performed and the same result would have been the same (usage of the temporary sample pump without a safety evaluation). The screening criteria had since been revised such that this was no longer a concern for recent procedures and revisions. In addition to the Technical Specification required drywell air sample, the drywell manifold sampling system had been used since the plant was built to obtain weekly samples from all the sampling points. This consisted of using the permanent pump to obtain samples from half the sampling points at one time. (Thus, sampling was done with half the sampling lines in simultaneo~s use twice a week.) This sampling was not done when the permanent sampling pump was inoperable. The design of the drywell manifold sampling system provided for two manual isolation valves both of which were located outside of primary containment. The portion of the drywell manifold system located outboard of the manual containment isolation valves was nonsafety~related. nonsafety-related. Thus, eleven sample lines with no automatic isolation. were routinely and simultaneously opened and left unattended for at least one hour twice a week, providing a path from the drywell, through nonsafety~related nonsafety-related piping, back to the drywell. The licensee took the following actions regarding this issue: o An assistant technical staff supervisor identified the original problem while revieWing reviewing a revision to ORP DRP 1350-3. 1350~3. During this review the individual felt it was confUSing confusing as to which valves were being addressed and therefore discussed with the author the possibility of including a diagram in the procedure. During this discussion the individual became aware that the temporary sample pump discharge was into the reactor building. This was not entirely obvious from just reading the procedure. o Upon discovering the problem, the licensee performed a preliminary analysis to quantify the amount of leakage through a one half inch penetration through primary containment at design accident pressure. After finding that this greatly exceeded allowable limits the licensee informed the NRC. o The licensee issued a temporary change to the procedure regarding usage of the temporary sample pumps to require an individual in continual attendance and in contact with the control room by radio while the manual isolation valves are open. The licensee subsequently performed a temporary alteration that moved the sample point for the Technical Specification required daily sample to a line that had automatic isolation. o All incoming Radiation Protection shift personnel were briefed as to the problem to preclude improper usag-of usage* of the system. (c. 18 II 1.14-22 111.14-22

Revision 8 April 1992 /' o \ The licensee initiated a deviation report to track the licensee investigation of the problem. The licensee also initiated a potentially significant event report for corporate management. o The licensee infonmed informed Quad Cities of the problem. In addition, the licensee has initiated or planned the following actions: o Due to questions regarding the original system design the licensee was reviewing the design basis and the need for any system design improvements. The licensee bad not made a decision whether the system would be repaired and used or whether it was to be abandoned, dismantled and the lines capped. o The licensee was reviewing methods whereby a temporary return line to the drywell could be established for use with the temporary sample pump. (Although automatic isolation was now provided, the temporary sample pump still exhausted to the reactor building which presented ALARA considerations.) o Due to the problem with the previous 10 eFR CFR 50.59 safety evaluation screening criteria, the licensee was attempting to determine the population of previous procedures and revisions that would need to be rescreened under the current criteria. o The licensee was performing a 10 eFR CFR 50.59 safety evaluation addressing two past practices: (l) (I) Use of the temporary sample pump exhausting 'to the reactor building atmosphere with the manual isolation valves left open and unattended. (2) Usage of the permanent as-designed system with eleven sampling lines left simultaneously open, open. and unattended. These safety evaluations were to include a 10 CFR 100 analysis for offsite doses and a 10 eFR CFR 50, Appendix A, General DeSign Design Criterion 19 analysis for control room doses. This issue is considered an unresolved item (50-237/90017-04(DRP)) (50-237/90017-04(ORP>> pending completion of the licensee's safety evaluations and NRC review of these documents.

c. On July 20, 1990, a dual unit shutdown began from 92 percent and 99 percent rated thermal power on both Units 2 and 3, respectively, in accordance with Technical SpeCification Specification action statement 3.0.A requiring hot shutdown within 12 hours and cold shutdown within the following 24 hours. A corresponding Unusual Event was declared due to initiation of a shutdown required by Technical Specifications.

The shutdown was due to the identification by the licensee of specific components, applied a~plied to both units, which-had which* had not been local leak rate tested (llRT) (LLRT) in accordance with 10 eFRCFR 50 Appendix J 19 III.14-23 III .14-23

Revision 8 April 1992 ( requirements. These included a check valve which had not been*.~ tested at all and two manual isolation valves whose testing been* _

                                                                                     ~

methodology was in question in the reactor building closed cooling water (RBCCW) system inlet to the drywell.drywe11. In *addition, addition, both the inboard and outboard manual isolation valves on a control rod drive line to the recirculation pump seals had not received llRTs. LLRTs. Finally, a flexitallic gasket on a torus water level transmitter had finally, LLRT. This last item was only a concern for Unit 22 not received an llRT. since the one on Unit 3 had been subjected to Integrated leak Leak Rate Testing (ILRT) pressure within the past 24 months. The problem with RBCCW had been identified earlier at Quad Cities, but was not initia 11y corrected at Dresden. This was because the problem initially prob lem at . Quad Cities involved total absence of llRTs LLRTs on the RBCCW system and the Dresden problem only involved partial llRT LLRT of this system *. Thus, communication only involved whether llRTs LLRTs were wera done on RBCCW and not the total extent of the llRTs.LLRTs. The absence of these

          . components in these three systems from the llRT    LLRT program and the licensee's corrective actions are considered an unresolved item licensee1s (50-237j90017-05(DRP))

(50-237/90017-05(DRP>> pending fUrtherfurther review by regional NRC specialists. . shutdown was stopped and the Unusua 1 Event terminated with the The shutd.own units at 73 and 80 percent power, respectively, later that same evening upon receipt of a verbal waiver of compliance from the NRC. The waiver wa i ver of compliance comp 1rance allowed 48 hours to conduct appropriate testing on the control rod drive system and torus water level transmitter line components and until the next refueling outage for each unit on the RBCCW line components. The licensee submitted the ( formal documentation to support this action on July 23, 1990 and also submitted an emergency Technical Specification amendment request on July 31, 1990, regarding the RBCCW line components. All actions regarding the control rod drive system and torus water level transmitter line components including modifications needed to conduct testing and the testing itself were completed on July 22, 1990. The licensee also issued an operating order describing actions to be taken regarding RBCCW in .the event of a LOCA. -.

d. During 1987, the licensee completed modifications to the Dresden Liquid Control System eSleS)

Station Standby liquid (SLCS) suction piping to facilitate dual pump operation. The modification was performed in pursuit of compliance with the Anticipated Transient Without Scram (ATWS) rule (10 CFR 50.62). At BWRs, the ATWS rule required the SLCS negative reactivity injection rate be increased to the equivalent of 86 gallons per minute of 13 wtj% eqUivalent wt/% sodium pentaborate solution. The rule further required the SLCS system to be "designed to perform its function in a reliable manner." The licensee 1s SlCS licensee's SLCS ATWS modification safety evaluation (10 CFR 50.59) stated, in part, "the lithe suction piping has been designed to assure two pump net positive suction head (NPSH) and eliminate concerns of mutua lly reinforcing pulsations." The inspectors reviewed the SlCS mutually SLCS ATWS modification NPSH design calculation. The review indicated the calculation did not include an analytical demonstration of adequate (.~ 20 II I.l4-24 111.14-24

Revision B

                                                                                 . April 1992

./ i NPSH but was built upon an assumed plant history of satisfactory single SLCS SlCS pump operation. The calculation incorporated the philosophy that minimum available NPSH for two pump operation could be maintained by the addition of a second section of pipin9~piping, of piping, connecting the SlCS storage similar design to the original piping. tank to the SlCS pump suction header *. The calculatfon calculation indicated that a strict analytical approach to the computation of available wQ/ld be overly conservative and placed a reliance on post NPSH WQlld modification testing to demonstrate satisfactory performance IErformance with both pumps in operation. The inspectors also reviewed the Unit 2 i SLCS SlCS ATI(S post modification test. The test consisted of the monthly single pump operational

          . surveillance test and the single pump reactor vessel injection addition, both pumps were run sill14,ltane survei llance. In addition~                           simqltane O1s1y Q/sly for a 64 second period to verify the dual pump flow rate. During the dual pump test, NPSH was verified by "absence of large noises associated with pump cavitation." The single SlCS pump in-service test program required each SlCS "pump to be run (individually) at least five minutes prior to obtaining data to allow each pump to reach hydrau lie stabil hydraulic         1ty. II In light of the design ca stability."                                lculations I reliance calculations' on the s;-te sfte testing to ensure SlCS HPSH, NPSH, the post modification testing was critical to the acceptance of the modification to meet 10 eFR CFR 50.62 criteria. This is considered an unresolved item (50-237/90017-06(DRP>> pending further NRC review to determine adequacy of the design calculations and the post modification testing.

( No violations or deviations were identified in this area.

7. Report Review During the inspection period, the inspector reviewed the licensee's Monthly Operating Report for June 1990. The inspector confirmed that the information provided met the requirements of Technical Specification 6.6.A.3 and Regulatory Guide 1.16.

B. Unresolved Items An unresolved item is a matter about which more information is required in order to ascertain whether it is an acceptable item, an open item, a deviation, or a violation. Unresolved items disclosed during this inspection are discussed in Paragraphs 6.b, 6.c and 6.d.

9. Exit Interview (30703)

The inspectors met with licensee representatives (denoted in Paragraph 1) on July 31, 1990 and informally throughout the inspection period, and summarized the scope and findings of the inspection activities. The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not* identify any such documents/processes as proprietary. The licensee acknow1edged acknowledged the findings of the inspection. 21 III.14*25 III .14-25

( e Commonw .. 1400 OPUS Edison

                           ** lth Edllon Opus Place IIlinoi.*60S1S Downers Grove, IIlinois'60515
                                                                               -Rev;
                                                                               *Revi sion si on 8
                                                                               *;Aoril1 1992
                                                                               -;ADri September 24, 1990 Mr. A. Bert Davis*

Davis Administrator, Region III Regional Administrator. U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137

Subject:

Dresden Station Units 2 and 3 Response to Notice of Violation Contained in Inspection Report 50-237/90017 and 50-249/90017 NRC Docket Nos. 50-237 and 50-249 NBC

Reference:

W. Shafer (NRC) letter to C. Reed (CECo), dated August 24, 1990. Mr. Davis: ( The referenced letter transmitted Inspection Report 50-237/90017 and jO-249/90017 for Dresden Station. The Inspection Report contained one (1) Notice of

   >0-249/90017 Violation regarding inappropriate equipment outage checklists. Commonwealth Edison Company (CECo) has reviewed the Notice of Violation and agrees that the violation occurred as described. Attachment 'A'to this letter preSents JlreSents CECo's response to the violation, and describes corrective actions which are being taken to prevent similar occurrences.

Please direct any questions or comments on this response to this office. Respectfully, Respectfully. flJ T. J. vach Nuclear Licensing Manager Attachment A: Commonwealth Edison Company Response to Notice of Violation 50-237/9)017-02. cc: B. Siegel - NRR Project Manager NRR Document Control Desk S. DuPont - Senior Resident Inspect r -, Dresden ( MR/TKIlmw MRITKIlmw ZNLD304 IIL14-26 IIl.14-26

C<H1ONW!ALTB C<HIOIIW!.ALIB EDISON C<!n'ANY C<m'ANY Revision 8

                                                                                  .Revision RESPONSE TO NOTICE OF VIOLATION                       April 1992 SO-237/900l7-{}2 50-237/90017-02

( VIOLATIQN (50-237/90017-02) (SO-237/900l7-02) cn 50, Appendix B, Criterion V, as implemented *by Coanonwealth 10 CFR COIIIDOnwealth Edison Company's Quality Assurance Program, requires that -activities *activities affecting quality be prescribed by documented instructions, procedures or drawings of a type appropriate to the circumstances. Contrary to the above, documented instructions for activities affecting quality prescribed in equipment outage checklists were inappropriate to the circumstances in the following ca~e8: cases:

a. Outage number 111-460 III-460 implemented on February 4, 1990 failed to recognize all consequences of a fuse removal, resulting in an unexpected Group II primary containment isolation, ilolation, .tandby standby gas treatment system automatic initiation and reactor building ventilation system isolation.

ilolation.

b. Outage number II-412 II-4l2 implemented on June 11, 1990 prescribed the closure of. incorrect valves, resulting in an unexpected recirculation pump trip.

c. Outage number 11-421 implemented on June 13, 1990 failed to recognize all consequences of opening a breaker, resulting relul ting in an mlexpected unexpected half Group II primary containment i.olation iaolation signal.

This is a Severity Level IV violation (Supplement I) Commonwealth Edison Company agrees with the violation as stated in the Notice of Violation. Although the three cases cited involved inappropriate equipment outage checklists for existing plant conditions, there tbere is a fundamental difference between the recirculation pump trip event that occurred on June 11, 1990 and the other two cases cited. The February 4, 1990 and June 13, 1990 events resulted from inattention to detail during the tbe preparation and review of the equipment outage checklists. In the tbe recirculation pump trip event, the equipment outage checklist checklilt was prepared with witb a drawing which wbicb did not indicate a requested drawing change to reflect the tbe in-plant labeling of the tbe temperature control valves. Additionally, the operator hanging the tbe outage failed to question activities that did not seem appropriate for the tbe work in progress progrels and plan t condi tions tionl

• Dresden Station has been conducting plant walkdowna walkdownl to upgrade plant labeling. Items which are found not to conform with witb COlllDon common labeling convention are corrected and drawing changes submitted. In the conventio~ tbe recirculation pump trip event event,t the as-built drawing in the tbe control room had been updated to reflect the tbe correct labeling of the tbe temperature control valves a8 as identified during the tbe plant walkdown; however, that drawing was not used in preparation of the equipm~nt outage checklist.

A-l I III. 14-27 1.14-27

Revision 8 April 1992 ( As a result of these events, Dresden Station has taken actions to emphasize: 1) the need to contact appropriate supervisory personnel if questions or uncertainties arise during any plant activity; 2) the joint responsibility of Operating Department personnel and Maintenance Department work analysts to perform a thorough review to determine the ilnpac impactt of all equipment outages; and 3) the use of the ItOlt lICIt up-to-date information available when preparing and reviewing equipment outages. COWCTIVE ACTION W\EN CQWCTIYE TAQN AND 'mULl'S US!1LTS ACIIIEUJ) AClUEVtJ) Immediate actions to restore the plant to no~l normal conditions were:

1. February 44,t 1990 event .:.. The subject fuse wa.

event':' was iJlaediately iDaediately replaced, the isolation reset, and the Standby Gas Treatment and Reactor Building Ventilation Systems were returned to normal.

  -2. June 11, 1990 event - The subject           valves were reopened, but not in
       . time to prevent the trip of the            recirculation pump. Control room operators correctly carried out             the requirements of DOA 202-1, "Recirculation Pump Trip - One             or Both, Both* Pumps."      The plant was returned to two* loop operation.

3. June 13, 1990 event - The subject breaker Was racked back in, and the half Group II isolation signal was reset.

Immediately following each event, an investigation was conducted to determine the root cause of each event, and to formulate and implement corrective ac~ion8. ae. ~ions. The events in June 1990 prompted additional corrective actions regarding the deVelopment, development, review and implementation of equipment outages. CORRECTIVE ACTIONS TAD'.N CORRECtIVE TAQN TO AVOID lUUlIEIl mT!IEIl Kmf...QOOLWfCES KOH--cooLIANCES Following the June 1990 events, the following corrective actions were taken.

1. Operations Department Memorandum #18 was issued to reaffirm with all Operations Department shift personnel the need to Ule use the most up-to-date available critical drawings when preparing end and verifying equipment outages, and to contact supervisory personnel When when activ! ties - do activities not seem appropriate for current plant conditions/evolutions prior to performing the activity.

2. A letter discussing the causes of the three events, the similarities of the events, and the corrective actions taken talcen to prevent reoccurrence has been sent to all Operating Depart.ent Department shift ahift penonnel and Maintenance Department work ana~y8ts.

per,onnel ana~ysts. A further det~.iled det"iled review of these events with .hift shift pusonne1 Ensonnel and work analysts will be conducted by December 14. 1990. c IIl.14-28 III. 14-28 A-2

Revision 8B April 1992

3. In order to provide readily available, accurate information '*"for '**for personnel involved with equipment outage preparation :~d

                                                                                ..~d verification, an additional Bet set of as-built critical drawings wil1~be will'be placed in the Operations Department Scheduler!s Scheduler's office. Dreaden Dresden Administrative Procedure 2-9, liAs-Built "As-Built Critical Drawin I .."isoil bein revue       0  contro   t cse ese drawings.      This let set of drewiDSs drawings will be copies of those used in the control room and will reflect the tlu-built" "as-built" condition of the plant, including inCluding any outstandinl outstandinc drawing change requests. These actions will be completed by September 28.

1990.

4. Dresden Station has formed a cOtllDittee cosmittee to develop a "Self-Check" policy for personnel to follow while While performing work in the plant.

The policy includes verifying all equipment, labeling and procedures prior to starting a job, anticipating expected plant responses, stopping if any response is not received, and ob.erving observing that all anticipated responses occur. A draft of these guidelines has been developed and will be implemented by October 1. 1990. DATE W1!EN :rnu. W1IEN FUJ.L CWLIAlfCE COOLIAlfCE WILL BE ACHIEVED ACBIEVEP Full compliance was achieved on June 13, 1990 wben when the half Group II isolation was reset. Co A-3 II 1.14-29 III .14-29

Revision 8 UNITED STATES April 1992 c( NUCl.EAR NUCLEAR REGULATORV REGULATORY COMMISSION REGION III 711 ItDOSIE:VELT IItOOSEVELT !tOAD

                                                          !!tOAD GLEN ELLVN.

ELLVN, ILLINOIS ILL.INOIS IOU7 lOU7 November 28, 1990 Docket Nos. 50-237 and 50-249 license Nos. DPR-19 and DPR-25

      £A EA 90-168 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Opus West III 1400 Opus Place Downers Grove, Illinois 60515 Gentlemen:

SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY - $37,500 (NRC INSPECTION REPORT NOs. 50-237/90017(DRP}i 50-237/90017(DRP); 50-249/90017(DRP); 50-249/90017(DRP}; 50-237/90022(DRP); 50-237/90022(DRP}; 50-249/90022(DRP)) 50-249/90022(DRP}} This refers to th~ special safety inspections conducted during the period of June 13 through July 31, 1990 and during the period of June 28 through September 20, 1990 at the Dresden Nuclear Power Station. During these c inspections a violation of NRC requirements was identified by your staff, and on October 12, 1990, an enforcement conference was held in the Region III office between Mr. D. Galle, Galle. and othtr members of your staff, and Dr. C. J. Paperie110, and other members of the NRC staff. Copies of the inspection reports were Paperiello, mailed to you on August 24, 1990 and October 4, 1990, and a co~y copy of the enforcement conference report was sent on October 24, 1990. On June 28, 1990, with Units 2 and 3 operating at 99% and 48~ 48: power respectively, during the review of a proposed revision to DRP 1350-3 tlSampling "Sampling the Drywell Manifold System Using the RaDeco Air Sampler Sampler", one of your employees, an ll Assistant Technical Staff Supervisor, discovered that obtaining the required daily air sample using this procedure both challenged the integrity of primary containment and potentially violated Technical Specification (TS) 3.7.A.a.{3) 3.7.A.a.(3) primary containment leakage requirements. Specifically, this procedure addressed obtaining the required air sample by breaking the closed loop on the drywell manifold air sample system and using a temporary sample pump in lieu of the normal air sample pump. In this configuration and under this procedure, the temporary sample pump would run unattended for approximately one hour daily and exhaust into the secondary containment with no automatic isolation capability. In addition, this represented a condition that could, could. by your own calculations, increase primary containment leakage beyond the allowed leakage of 1.6% 1.6~ per day (TS {TS 3.7.A.a.(3}) 3.7.A.a.{3}} by an additional 4.73% per day for a total leakage of 6.33% per day. It is my understanding that this method of air sampling using the temporary sampling pump has been used as a secondary backup method to obtain the required air sample since approximately 1978. We also understand that the required air samples were originally obtained through the use of a continuous air monitor (CAM) with the drywell manifold air sample system as the primary ( backup. CERTIFIED MAIL RETURN RECEIPT REQUESTED III.14-30

   ,                                                                          Revision 8
   *                                                                          ~pril April 1992

( Commonwealth Edison Company November 28, 1990 The root cause of this event was your failure to recognize that use of the temporary air pump constituted a design change that required the performance of a proper engineering review and the establishment of proper procedural controls prior to its implementation. Consequently, this resulted in a significant failure to meet the requirements of 10 CFR 50.59. Specifically, each time that the temporary sample pump was used, you failed to perfonm perform the evaluatton necessary to determine whether the activity constituted a change in the TSs and/or an unreviewed safety question. In this case, the use of the temporary sample pump effectively constituted a change in the TSs' allowable leakage rate and represented an unreviewed safety question in that the additional leakage rate (4.73%) nullified the margin of safety as defined in the basis to the TSs. This violation is significant in that (based on your calculations using design basis methodologies) both limits for the thyroid dose for control room habitability and for the 30 day thyroid dose at the low population zone would have been exceeded. Although you performed additional analyses that indicated that acceptable offsite and control room doses would have been obtained, those analyses, tnatthat were based on assumptions that were less conservative than those used in the plant licensing basis, still would nave have required changes to the Final Safety Analysis Report (FSAR), TSs, and TS bases. However, the determination of the acceptability of such analyses is an NRC function, and requires NRC approval prior to implementation of the change. Therefore, in accordance with the "General Statement of Policy and Procedure for (- NRC Enforcement Actions,U Actions," (Enforcement Policy) 10 CFR Part 2, Appendix C (1990), this violation has been categorized as a Severity level Level III violation. The NRC recognizes that immediate corrective action was taken when the violation was identified. In addition, the NRC was informed of your subsequent corrective actions during the October 12, 1990 enforcement conference. During this discus-Sion, sion, you informed us that as part of your corrective action for this event, that you had identified that a 10 CFR 50.59 review had not been completed prior to disconnecting the CAM in the early 1980's~ despite the fact that it was an FSAR requirement. I understand that you have reinstalled the CAM on Unit 3 and will reinstall it on Unit 2 prior to its startup from its current refueling outage. To emphasize the need for recognizing design changes and for performing the necessary evaluations in accordance with the provisions of 10 CFR 50.59, I have been authorized, after consultation with the Director, Office of Enforcement~ Enforcement; and the Deputy Executive Director for Nuclear Reactor Regulation, Regional Operations and Research to issue the enclosed Notice of Violation and Pro osed 1m osition of Civil Penalt Notice in the amount of 37 500 for the Severit level III" violation. The base value of a civil penalty for a Severity Level III Level III-violation is $50.000.

                    $50,000. The escalation and mitigation factors in the Enforcement Policy were considered.

I recognize that your employee went beyond his normal duties in identifying the violation and wish to encourage you to continue such aggressive reviews. The fact that this employee took the time to look into and question the process ( instead of routinely approving a procedure revision is to be commended. 111.14-31 III.!4-3!

   ,                                                                                  Revision 8 April 1992

( Commonwealth Edison Company November 28, 1990 However, this violation might have been identified earlier if an aggressive review had taken place on several prior occasions. First, in 1986, the unreviewed safety question might have been identified if your revisions to the temporary alteration program had extended to cover use of the temporary sample pump, either at that time or when use of the pump was reinstated in 1987. Second, in August 1988, when your temporary alteration program was extended to mech-anical equipment, the unreviewed safety question might have been cover use of mechanical identified if you had recognized the use-of the temporary pump as a temporary alteration. Finally, in May 1989, when the procedure governing use of the 1989,when temporary sample pump was created (in response to a third party reviewer's recommendation made in 1988), the unreviewed safety question might have been identified if you had properly performed a safety evaluation as required by your own procedure. Therefore. Therefore, only partial mitigation (25%) was deemed warranted for the identification factor. Fifty percent mitigation was applied due to the extensiveness of your corrective actions, once you recognized that an unrevlewed safety guestlon eXlsted. !,Ilth Wlth respect to your past perfonmance, performance, the NRC notes that you received two previous Severity level Level IV violations involving changes to the facility without prior evaluation and authorization in the past two years. I recognize that the corrective action for those violations would not necessarily have prevented the subject violation. In addition, the NRC has noted a significant improvement in the performance of your technical staff organization as evidenced by your latest SALP rating in E&T5, as well as the more aggressive scrutiny that your employees the area of E&TS, ( are giving to routine reviews. Therefore, 50% mitigation was applied for past performance. However, I am especial1y especially concerned in this case due to the number of years that the temporary sample pump was regularly used on a daily basis and the potential for a significant offsite release should a design basis lOCA LOCA have occurred during those times. In addition, the NRC is concerned that, for a substantial number of years, it appears that you failed to properly understand and evaluate eva 1ua te the intent .and _and requirements requi rements of the containment contai nment air ai r sample samp 1e system such that the proper corrective actions for the system requirements could have been implemented. Therefore, the base civil penalty was escalated by 100% based on the duration factor. The other factors of the POllCY Policy were considered dnd and no

     'further adjust~Ent adjustffiEnt to the base civil penalty was considered appropriat~.

Therefore. Therefore, based on the above, a civil penalty in thE the final amount of S37,500 537,500 is proposed. You are required to respond to this letter and should follow the instructions enclosed Notice when preparing your response. In your response, specified in the nclosed you should document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements. In accordance with 10 CFR 2.790 of the NRCts NRC's ItRules "Rules of Practice," a copy of this letter and its enclosure will be placed ;n in the NRC Public Document Room. c( I II.14-32

Revision 8

                                                                           . April 1992 Commonwealth Edison Company                                      28,s November 28   '1990 1990 The responses directed by this letter and the enclosed Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Pl PL 96-511.

Sincerely, Regional Administrator Regional

Enclosures:

1. Notice of Violation and Proposed Imposition of Civil Penalty

2. Inspection Report Nos. 50-237/90022(DRP);

50-249/90022(DRP) 50-249/90022(ORP) cc w/enclosures: BWR D. Galle, Vice President - B~R Operations Opera t ions (( T. Kovach, Nuclear

\   Licensing Manager E. D. Eenigenburg, Station ~anager Manager OeD/DeB DCD/DCB (RIDS)

(R IDS) OC/lFDCB OC/LFDCB Resident Inspectors LaSalle, Dresden, Quad Cities Richard Hubbard J. W. McCaffrey, Chief, Public Utilities Division Robert Newmann, Office of Public Illinois Center Counsel, State of Illino;s (

                                            .14-33 I II1.14-33 II

Revision 8 April 1992 NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY Commonwealth COlll1l0nl'Iea lth Edison Edi son Docket Nos. 50*237 50-237 and 50-249 Dresden Nuclear Power Station license Nos. OPR-19 and DPR-25 EA 90-168 During NRC inspections ir.spections conducted on June 13 through July 31, 1990 and June 28 through Septer.:ber Septe~:ber 20, 1990, a violation of NRC requirements was identified. In accordanCE accordance with the ~General "G~neral Statement of Policy and Procedure for NRC Enforcement .!Ict"ions,"

                       ';ctions," 10 CFR Part 2, Appendix C (l990L      the Nuclear Regulatory (1990), thE:

Commission proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954. 1954, as aw.ended arr.ended (Act), 42 U.S.C. 2282, and 10 CFR 2.205.2.205 ..- The particular violation and associated civil penalty is set forth belo~ .: belo~': 10 CFR 50.59(a) states, in part, that a holder of a license may make changes in the facility as described in the thc safety analysis report without prior Commission approval unless the proposed change involves a change in th~ the technical specifi-( cations inccrporated incorporated in the license or an unreviewed safety question. It also

"-  states, in part, that a proposed change shall be deemed to involve an unreviewed safety question if the margin of safety as defined in thE         the basis for any technical specification is reduced.

Section 14.2.6.4.1 of the Final Safety Analysis Report (FSAR) states, in ir, part, that the Air Sample System be configured such that the flair "air sample will be drawn through the tubing, out through a drywe11 drywell penetration, auto-isolation valves, and then to a continuous air monitor." Section 14.2.4.2.C of the Updated Safety Analysis Report (USAR), which discusses offsite dose releases following a Loss of Coolant Accident (lOCA), (LOCA), states, in part, that the primary containment leaks 0.5 percent of the contained free volume per 24 hours at 25 psig. Section 14.2.4.3 1~.2.4.3 of the USAR, which discusses post-LOCA control room dose rates, states, in part, that activity releases are based on a containment leakage rate of 1.6 percent per day. Technical Specification 3.7.A.2.a(3) 3.1.A.2.a(3) states that the maximum allowable leakage rate at a pressure of Pa, la, La, is equal to 1.6 percent by weight of the contain-ment air per 24 hours at 48 psig. The bases for the surveillance requirements for S~ction Section 3.7.A.2 explain that the maximum allowable test leak rate (1.6% was derived from the maximum allowable accident leak rate of about 2 percent/day, when corrected for the effects of containment environment under accident and test conditions. The bases additionally state that the accident leak rate could be allowed tu increase to about 3.2 percent/day before the guideline thyroid doses value given in 10 CFR 100 would be exceeded, so that establishing the ( test limit of 1.6 percent/day provides an adequate margin of safety to assure the health and safety of the general public. III.14-34

Revision 8 April 1992 ( Notice I of V;oldtion Violation Contrary to the above, the licensee, without prior Commission approval, on a sporadic basis since 1978 and on an almost daily basis from 1987 up to discovery in June 1990, made changes to the facility as described above in the safety analysis report (automatic isolation was not provided during containment air sampling) that involved a change to the Technical Specifications (TSs) and constituted an unreviewed safety questiorL. question.. Specifically, use of a temporary sample pump to obtain the required daily drywell air sample would have involved a change to the TSs in thdt that the maximum allowable leakage rate (1.6 percent/day) would have been increased by 4.73 percent/day for a total leakage of approxi-mately 6.33 percent/day. Use of the temporary sample pump constituted an unreviewed safety question in that this amcuntamount exceeded the leakage specified in the bases for the above TS section, such that the margin of safety defined therein was eliminated. This is a Severity level III violation (Supplement I) Civil Penalty - 537,500.

                        $37,500.

Pursuant to the provisions of 10 eFR CFR 2.201, Commonwealth Edison Company (Licensee) ;sis hereby required to submit a written statem~nt of explanation to (co the Director, Dir~ctor, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice of Violation and Proposed Imposition of Civil Penalty (Notice). This reply should be clearlycle:arly marked as a "Reply to a Notice Violation" and should include for each alleged violation: (1) admission or of Violation Jl denial of the alleged violation, (2) the reasons for the violation if admitted, and if denied, the reasons why, (3) the corrective steps that have been taken and the results achieved, (4) the corrective steps that will be taken to avoid achieved.~IIMlf~ further violations, and (5) the date when full compliance will be ach;eved.~II~f~ an adequate reply 1S lS not recewea recelVea wltnln tnc tne tlme specnled ln thlS SpeClTlea 1n tnlS Notlce, an order may b~ issued to show cause why the license should not be modified, suspended, or revoked or why such other actions as may be proper should not be taken. Consid~ration Considr:ration may be given to extending the response time for good cause shown. Under the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affinmation. affirmation. Within the same time as provided for the response required under 10 CFR 2.201, the licensee Licensee may pay the civil penalty by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, CommiSSion, with a check, check. draft, draft. money order. or electrordc electror"c transfer payable to the Treasurer of the United States in the amount of the civil penalty proposed above, above. or the cumulative perlaltles 1+ amount of the civil per,oltles H more than one C1V11 penalty is proposed. or may protest imposition of the civil penalty in whole or in part, by a written answer addressed to the Director, Office of Enforcement, Enforcement. U. S. Nuclear Regulatory Commission. Should the licensee Licensee fail to answer within the time specified, an order imposing the civil penalty will be issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalty,penalty. in cco whole or in part, such answer should be clearly marked as an "Answer to a Notice of Violation Violation"ll and may: (1) deny the violation listed in this Notice in whole or in part, part. (2) demonstrate extenuating circumstances, (3) show error in this Notice, or (4) show other reasons why the penalty should not be imposed. In addition to protesting the civil penalty ;n in whole or in part, part. such answer may request remission or mitigation of the penalty.

                                                ].14-35 II 1.14-35

Revision 8 April 1992 Notice of Violation In requesting mitigation of the proposed penalty, the -factors "factors addressed in Section V.B of 10 CFR Part 2, Appendix C (1990), should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. The attention of the Licensee

    ;s is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing a civil penalty.

Upon failure to pay any civil penalty due which subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty, ur.less compromised, remitted, mitigated, may be collected by civil action pursuant to Section 234c of the or mitigatEd, Act, 42 U.S.C. 2282c. The response noted above (Reply to Notice of Violation, letter with payment of

  .civil and Answer to a Notice of Violation) should be addressed to: Director, Office of Enforcement, U.S. Nuclear Reaulatory Commission, ATTN: Document I-:ashington, D.C. 20555 with a copy to the Regional Administrator, Control Desk, \-.'ashington, c   Po,,'er Po"       Station.
       .. er Sta t i on.

Commission, Region III, 799 Roosevelt Road, Glen Ellyn, U.S. Nuclear Regulatory CommisSion, Illinois 60137, and a copy to the NRC Resident Inspector at the Dresden Nuclear FOR THE NUCLEAR REGULATORY COMMISSION UJ~iuw1tt UJ~iX~j1Jt-t A. Bert DaviT- - ~- - ) Regional Administrator Dated at Glen Ellyn, Illinois this 28th day of November 1990 (- I.l4-36 II 1.14-36

1 1 av ab 5

Revision 8 April 1992 (, DRESDEN 2 &

                                            &3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspection Report No. 50-237/90023 and 50-249/90023 Title 15-l IIL15-1 III.                   Inspection Reports No. 50-237/90023 and 50-249/90023 dated December 7, 1990.

1.15-31 III. II 15-31 December 14, 1990 CECo letter from T. J. Kovach to A. Bert Davis (NRC) discussing unresolved Item 50-237/ 90023 and 50-249/90023. II 1.15-41 I.l5-41 January 7, 1991 CECo letter from T. J. Kovach to A. Bert Davis (NRC), Response to Notice of Violation contained in Inspection Report No. 50-237/90023 and 50-249/90023. 111.15-56 II I.lS-56 February 6, 1991 NRC letter from H. J. Miller to C. Reed (CECo) responding to CECo's letter of January 7, { 1991. ( IlL15-i III.lS-i

UNITED STATES NUCLEAR REGULATORY COMMISSION Revision 8 REGION III April 1992 ( '"

                                      '   IItOOSEVE:L T
                                        " 1II00SlEVlEI.. T IIIOAO ROAD
,                                        ELLYN.

GL.EN EL.L. IL.LINOIS 10137 YN. IL.L.INOIS DEC O£c 0 7 1990 Docket Do~ket No. 50-237 Do~ket No. 50-249 Docket Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice Vi~e President Opus West III 1400 Opus Place Pla~e Downers Grove, IL 60515 Gentlemen: This refers to the routine safety inspection inspe~tion conducted ~ondu~ted by O. D. E. Hills, M. S. Peck. Pe~k, J. D. Monninger, D. E. Jones and J. A. Holmes of this office offi~e on September 29 through November 16, 1990 of activities at Dresden Nuclear Nu~lear Power Station, Units 2 and 3 authorized by Operating License Nos. DPR-19 and DPR-25 and to the discussion of our findings with Mr. E. Eenigenburg and others at the conclusion of the inspection. The enclosed en~losed copy of our inspection inspe~tion report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and ,( ( interviews with personnel. During this inspection, certain of your activities appeared to be in violation requirements, as specified in the enclosed Notice. A written response of NRC requirements. is required. One licensee identified non-cited non-~ited violation is identified within this report. This issue involved an inadequate out-of-service chec~list checklist which resulted in an inadvertent automatic start of the swing diesel generator. We have chosen not to issue a notice of violation because be~ause this violation met the criteria delineated in 10 CFR Part 2. a~cordance with 10 CFR 2.790, of the Commission's Regulations, a copy of In accordance this letter and the enclosure(s) will be placed in the NRC Public Document Room. i( III.15-1

Revision 8 April 1992 Me OfC 0 7 1990 Commonwealth Edison Company 2 The responses directed by this letter and the accompanying Notice are not subject to the clearance procedures of th~th.e Office of Man~gement Man.agement and Budget as required by the Paperwork Reduction Redu:tion Act of 1980, PL 96-511. We will gladly discuss_a.ny discuss.any questions you have concerning this inspection. 1~~.~1:A~ (~~.~y(~ Reactor Projects Branch 1

Enclosure:

Inspection Reports No. SO-237/90023(DRP); 50-237/90023(DRP); No. 50-249/90023(DRP) cc w/enclosure: D. Galle, Vice President - BWR Operations T. Kovach, Nuclear licensing Licensing Manager ( D. Eenigenburg. E. O. Eenigenburg, Station Manager DCD/DCB (RIDS) OCiLFOCB OClLFDCB Resident Inspectors LaSalle, Dresden, Quad Cities Richard Hubbard J. W. McCaffrey. McCaffrey, Chief, Public Utilities Division PubliC Robert Newmann, Office of Public Counsel, State of Illinois Center -c( III.15-2 III .15-2

Revision 8 April 1992 ( APPENDIX NOTICE OF VIOLATION As a result of the inspection conducted on September 24 through November 16, 1990, and in accordance with the "General Policy Po lfcy and Procedures for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1990), the following violations were identified:

1. 10 CFR 50, Appendix S,B, Criterion II, as implemented by Commonwealth Edison's "Quality

                    ~Quality Assurance Program" requires indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained.

Contrary to the above, indoctrination and training of personnel performing activities affecting quality was inadequate in assuring proficiency was achieved and maintained as to administrative requirements as indicated in the following examples:

a. Lack of operations personnel knowledge of Dresden Administrative Procedure (DAP) 7-5, "Operating Logs and Records," Revision 8, and Dresden Operating Abnormal (DOA) Procedure 902-5 G-2, Revision 3, requirements for maintaining the Control Rod Drive Accumulator High Water 1/ Low Pressure Alarm Log (AHWLPAl)

(AHWLPAL) resulted in the AHWLPALs ( for both units not being maintained between April 1990 and ( August 3, 1990. As such the licensee's program to identify repeat failures of accumulator alarms was not effective during that time period. {50-237/90023-01a (50-237/90023-01a (DRP) (DRP>>

b. Lack of technical staff personnel knowledge regarding recognizing and processing conditions adverse to quality resulted in a failure to properly identify a procedural nonadherence involving maintenance of the AWHLPAL when discovered in May 1990. Because of this, corrective actions to prevent recurrence were not taken at that time. (50-237/90023-0lb (50-237/90023-01b (DRP>>

This is a Severity Level IV violation (Supplement 1).

2. 10 CFR 50, Appendix B, Criterion V, as implemented by Commonwealth Edison Company's Quality Assurance Program, requires that activities affecting quality be accomplished in accordance with documented instructions, procedures or drawings.

Contrary to the above, activities affecting quality were not accomplished in accordance with documented instructions, procedures, or drawings in the following examples: .

a. Dresden Operating Procedure (DOP) 1900-3, -Reactor "Reactor Cavity-Dryer Separator Storage Pit Fill and Operation of the Fuel '001 Fool Cooling and Cleanup System During Refueling, Refueling," Revision 8, requires constant II communication between the refueling floor and the control room while Ie filling the reactor vessel. Constant communication between the refueling floor and the control room was not maintained while IIL15-3 III.15-3

Revision 8 April 1992 Notice of Violation 2 filling the Unit 2 reactor vessel on October 14, 14. 1990, 1990. resulting in the overfilling of the vessel into the ventilation ducts and contamination of various areas of the third and fourth floors of the (50-237/90023-02a (DRP) reactor building. {50-237/90023-02a (DRP})

b. Specific practices required by DAP OAP 3-5, 3-5. *Out of Service and Personnel Cards. Revision 22, Protection Cards, 22. were not followed as to preparation, preparation.

review, review. approval, approval. documentation and independent Verification verification in the removal and return to service of the Unit 2 diesel fuel oil day tank drain valve on October 29,

29. 1990. This resulted in the inadvertent draining of the day tank when the drain valve was placed in the incorrect position. (50-237/90023-02b (DRP)) (DRP})

c. OAP DAP 7-14, 7-14. "Control and Criteria For Locked EqUipment Equipment and Valves,"

Valves." Revision 2,

2. requires manual valves in the flowpath of systems required for plant shutdown during post-accident situations o~ or which provide a controlled path to the environs, environs. including primary and secondary containment isolation valves to be locked. Prior to November 1990, 1990. manual valves including the Units 2, 2. 3 and 2/3 diesel generator service water three-way valves and the Units 2 and 3 drywel1 drywe11 manifold sampling system containment isolation valves were not locked or deSignated designated to be locked. (50-237/90023-02c (DRP})

d. DAP 15-6, 15-6. "Preparation and Control of Work Requests," Revision 0, O.

requires work to be performed per repair manua1(s), manual(s). travelers/ procedures, procedures. or work instructions provided in the work package. On October 15,

15. 1990, work prescribed for disassembly of the Outboard Containment Isolation Feedwater Check Valve 220-62B was performed instead on Outboard Containment Isolation Feedwater Check Valve 220-62A. (SO-237/90023-02d (50-237/90023-02d (DRP>>

(DRP})

e. OAP DAP 15-6, "Preparation and Control of Work Requests,"

Requests.* Revision 0, requirements were violated on August 8, 8. 1990, 1990. when work prescribed for calibration of Unit 3 Torus to Reactor Building Vacuum Breaker A Pressure Transmitter DPT-1622A was performed instead on Pressure Transmitter DPT-1622B. This resulted in advertant opening of the Unit 3 Reactor Building Vacuum Breaker .S. .B. (50-237/90023-02e (DRP>> (DRP)) This is a Severity level Level IV violation (Supplement 1).

3. 10 CFR SO, 50, Appendix B, B. Criterion XVI, XVI. as implemented by Commonwealth Edison's "Quality HQuality Assurance Program,"

Program." requires that conditions adverse to quality be promptly identified and corrected and, and. in the clse case of . significant conditions, conditions. the measures assure the cause is determined and corrective action taken to prevent repetition. Contrary to the above, above. following the fuel bundle mispos1tion1ng mispositioning events of January 10 and 12. 12, 1989, 1989. corrective actions were insufficient to prevent repetition in that similar events occurred on October 1, 1. 1990 and October 2,

2. 1990. (50-237/90023-08 (DRP>>(DRP})

III.1S-4 111.15-4

Revision 8 April1 1992 Apri ( Notice of Violation 3 ( This is a Severity Level IV violation (Supplement 1). Pursuant to the provisions of 10 CFR 2.201, you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply, including for each violation: (1) ~orrective action taken and the results achieved; {2) (2) corrective action to be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown.

w. D. ,e~

pr~BranCh pr~Branch Shafer, let Reactor 1

(( {/

( III.15-5 III.lS-S

Revision 8 April 1992 ( U. S. NUCLEAR REGULATORY COMMISSION REGION III Report Nos. 50-237/90023(DRP); 50-249/90023(DRP) 50-237/90023(DRP)i Docket Nos. 50-237; 50-249 License Nos. DPR-19; DPR-Z5 DPR-l9; DPR-25 Licensee: Commonwealth Edison Company P. O. Box 767 Chicago, IL 606QO 60690 Facility Name: Dresden'Nuclear Power Station, Units 2 and 3 Inspection At: Dresden Site. Site, Morris, IL Inspection Conducted: September 29 through November 16, 1990 Inspectors: D. E. Hills Peck M. S. Peck. J. D. Monninger D. E. Jones

                    ~

J. A. Ho}mes ./ .( '\ .~ /.

                             /      ve...l
                                      . ../

Urg~Chie"f urg~Chief '\. ;f

   ;            By:,/.iY Approved By:  ....-1Y Projects Section 18      IB Inspection Summary Inspection during the period of September 29 through November 16, 1990 (Reports No. SO-237/90023(DRP)j 50-237/90023(DRP)j SO-249/90023(DRP>>.

Areas Inspected: Routine unannounced resident inspection of previously identified inspection items, licensee event reports followup, p1ant plant operations, maintenance and surveillances, engineering and technical support, safety assessment/quality verification and report review. Results: Resul ts: Three violations were identified with numerous examples. One involved the failure to follow procedures and instructions and included five examples. These examples permeated different disciplines and involved failing to uti-lize utilize or ignoring procedures and instructions or Dr inattention to detail in implementing these requirements. Specifics are described in paragraphs 4.a, 4.c, 4.e. 4.e, 5.a.2 S.a.2 and S.b.l. The second violation involved ,( III.1S-6 III.lS-6

Revision 8 April 1992 ( (\ inadequate corrective actions in regard to fuel bundle misPositioning events with two examples. Specifics are described in paragraph 7.a. The third violation involved inadequate training to assure adequate knowledge of plant administrative requirements with two examples. Specifics are described in paragraph 2. One violation was identified which concerned an inadequate out of service checklist. However, a Notice of Violation was not issued in accordance with the discretionary enforcement policy described in 10 eFR CFR 2, Appendix C, Section V.A. Specifics are described in paragraph 4.b. Five unresolved items were identified. An unresolved item involving a possibly inoperable source range monitor while moving fuel 1n in that core quadrant is pending further NRC review of the event (paragraph 4.f). An unresolved item involving the licensee's policy of not declaring equipment inoperable and not entering corresponding limiting conditions for operation when equipment was purposely rendered inoperable for surveillance testing is pending further clarification of requirements 4.g). An unresolved item involving licensee maintenance (paragraph 4.9), practices on Appendix R fire protection emergency lighting 1s is pending completion of a licensee investigation report (paragraph 5.b.3). S.b.3). An unresolved item involving the licensee's discovery that the filter media in the Unit 3 Reactor Building Ventilation Air Particulate Sampler had been misalligned ;s is pending further review by NRC regional specialists (paragraph 5.b.2). S.b.2). Finally, an unresolved item involving the licensee's usage of Quality Control Inspection Feedback Sheets is pending further. NRC review of that area (paragraph 7.c). ,( Plant Operations A number of events occurred during the current Unit 2 refueling outage indicative of personnel performance problems such as communications and inattention to detail. Although they were spread across several disciplines, noteworthy events involving the plant operat1ons operations functional area included two fuel bundle mispositioning events, a reactor cavity overflow event, inadvertent draining of a diesel generator fuel oil day tank and an inadvertent diesel generator automatic start. Although the safety significance in all cases was minimal, the number of events represent an adverse trend. Maintenance/Surveillance In addition to the events above, other adverse events occurred in the Maintenance/Surveillance functional area. Noteworthy among these were an inadvertent automatic start of a core spray pump, disassembly of the wrong feedwater containment isolation check valve and calibration adjustments to the wrong torus to reactor building vacuum breaker pressure transmitter. These were indicative of personnel performance problems such as communications and attention to detail. detail .

.(,

2 111.15-7

Revision 8B April 1992 Encinee"ino/Te:hnical Enc~nee~inQ/Te:hn;cal Suooort SUDDort Review of a modifica~ion and associated field work did not identify any problems. One of the violations described in the report involved the Tack lack of a formal training program to assure appropriate technical staff personnel were trained on applicable administrative requirements. Safety Assessment/Quality Verification i~ the number of events Licensee management recognized the adverse trend iri indicative of personnel performance problems. Management involvement was highly evident in the review of these ~vents and the determination of corrective actions. In addition, generic corrective actions were implemented as described in paragraph 7.b. However, one violation concerned inadequate corrective actions in regard to fuel bundle mispositioning events. Another involved failure of technical staff personnel to recognize procedural nonadherence as a condition adverse to quality such that corrective actions to address the root cause was not taken. This was indicative of a personnel training deficiency. It must be noted however that the inspectors regard licensee corrective actions to r.ormal1y normally be thorough and comprehensive. ( '(( 3 111.15-8

Revision 8 Apri] 1992 April DETAILS (

1. Persons Contacted Commonwealth Edision Comoany

        *E. Eenigenburg, Station Manager
        *L. Gerner, Technical Superintendent E. Mantel, Services Director
        *0. Van Pelt, Assistant Superintendent - Maintenance
        *J. Kotowski, Production Superintendent J. Achterberg, Assistant Superintendent - Work Planning
        *G. Smith, Assistant Superintendent-Operations K. Peterman, Regulatory Assurance Supervisor M. Korchynsky, Operating Engineer

8. Zank, Operating Engineer J.

J, Williams, Operating Engineer R. Stobert, Operating Engineer M. Strait, Technical Staff Supervisor L. Johnson, Q.C. Supervisor J. Mayer. Mayer, Station Security Administrator D. Morey. Morey, Chemistry Services Supervisor D. Saccomando, Health Physics Services Supervisor

        *K. Kociuba, Quality Assurance Superintendent
        *0. Wheeler, Engineering and Construction
        *8. Viehl, Engineering and Construction" Construction'
        *G. Kusnik, Quality Control

( WK.

        *K. Yates, Onsite Nuclear S~fety Group Administrator The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engin!ering engin@ering staffs, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel.

personnel, and contract security personnel.

        *Denotes those attending one or more exit interviews conducted informally at various times throughout the inspection period.

2. Previously Identified Inspection Items (92701 and 92702)

(Closed) Violation SO-237/89019-01(DRP): Failure to place isolated emergency core cooling system (ECCS) level switch in tripped condition resulting in Technical Specification (TS) violation. In addition to interim actions taken by the licensee, the inspector verified that the licensee had developed and placed in the control room a Technical Specification Instrumentation Operability Manual. This provided guidance on the preferred method of placing Technical T@chnical Specification instrumentation in the tripped condition and assistance in locating the proper controlled documents to be used in this regard. Operations Policy Statement No. 23 was issued on July 31, 1990, to provide instructions regarding usage of this manual. The. The, inspector has no other concerns in this area. ( 4 IlI.15-9 II L15-9

Revision 8 April 1992 ( ( \. (Closed) Unresolved Item SO-237/90019-01(DRP): Review shift operations failure to maintain the Control Rod Drive (CRD) Accumulator High ,* Water/Low Pressure Alarm Log (AHWLPAL) for the period between April 1990 and August 30, 1990. The AHWLPAL was used to document eRDCRD accumulators that become degraded due to either a low pressure or high water level condition and facilitated as a tracking tool to determine if a particular accumulator exhibited a recurring problem. During the period in question, no record of CRD accumulators degraded by a low pressure or Or high water level condition could be located by the licensee. The average frequency of accumulator alarms was approximately once per shift per unit. Dresden Administrative Procedure (DAP) 7-5, "Operating Logs and Records", Revision 8, provided detailed instructions for the maintenance of records and logs which were administratively required to be maintained for the life of the plant. Step B.8 of DAP 7-5 required a AHWLPAl AHWLPAL to be maintained for each unit as an ongoing record of CRD accumulator alarms. Additionally, the Accumulator High Water/Low Pressure annunciator response procedure, Dresden Operating Abnormal (DOA) 902-5 G-2, Revision 3, directed the Nuclear Station Operator (NSO) to review past entries in the AHWLPAL following a new alarm, and to initiate a maintenance work request if a particular accumulator was exhibiting a recurring problem. DOA 902-5 902-S G-2 also required the NSO to document the new accumulator alarms in the AHWLPAL. The requirements for the AHWLPAL were transferred into OAP DAP 7-5 on December 8, 1989, from the Unit Operator's Daily Surveillance log, Log, Appendix A. The failure of shift personnel to complete the AHWLPAL during the period between April 1990 and August 3D, 30, 1990, was related, in part, to inadequate training of operations personnel at the time of the transfer such that some individuals were not aware of the administrative requirement. Review of the Unit 3 AHWLPAL (the Unit 2 AHWLPAL had been lost) indicated at least seven NSOs had followed the CRD logging requirements until April 1990. Interviews indicated that inadequate training also contributed to these NSOs ceasing performance of the logging requirements in that they were not aware that this was a continuing official requirement. However, the source document, OAP DAP 7-5 was identified on each AHWLPAL page. Additionally, copies of the Source source document, sheathed in a clear plastic document protector and defining the requirements for the log, were found at the beginning of the log book. This is of concern because plant operations personnel. personnel, without proper direction from management, stopped the performance of documentation activities for records. Inadequate training of appropriate personnel is as to administrative requirements concerning the AHWLPAl AHWLPAL was considered to be an example of a violation (50-237/90023-01a CORP>> (DRP)) of 10 CFR 50. SO, Appendix B. B, Criterion II. The inspectors found through interviews, that the technical staff eROCRD system engineer knew through independent review of the programmatic main.tain the AHWLPAL. failure to maintain AHWLPAL, per the administrative requirements of DAP 7-5 and DOA 902-5 902-S G-2, since approximately May 1990. The system engineer was not cognizant of and had not been trained on the requirements of DAP 9-12, "Procedural Adherence DefiCiencies," Revision 0, to document failures to meet the procedural intent or to 5 IILIS-IO lILIS-IO

Revision 8B April 1992 ( perform steps and activities contained within a procedure. Through \ additional interviews, the inspectors found that the problem of unfamiliarity and laCK lack of training for the documentation of procedural adherence deficiencies was not limited to this single individual. This was significant in that the use of DAPOAP 9-12 facilitates the identification, management review of, and resolution tracking including corrective actions of conditions adverse to quality associated with procedural inadherence. Although the system engineer knew a change in the method of documenting CRD accumulator alarms was planned and, as such, was not concerned, this did not correct the immediate problem nor did it address why the NSOs were not following an administrative requirement. Although other plant reporting and corrective action mechanisms existed that could have also provided these functions, these other plant deviation reporting programs were also not used. Inadequate training of appropriate personnel in regard to recognizing and processing this procedural inadherence as a condition adVerse adverse to quality such that adequate corrective action could be taken ;s is considered an example of a violation (50-237/90023-01b CORP>> of 10 CFR 50, Appendix e, (SO-237/90023-01b (ORP>> B, Cri teri on II. Criterion 80th Both of these examples of violations would appear to be indicative of an overall problem involving personnel knowledge of plant administrative requirements and the significance of these requirements. Although some training on administrative requirements is given to personnel, there is an absence of an overall program to control and ensure appropriate personnel are trained on administrative requirements that they need to know to perform their duties. (Closed) Unresolved Item SO-237/90022-03(DRP); SO-237/90022-03(ORP); SO-249/90022-03(DRP): SO-249/90022-03(ORP): Review licensee's incorporation of safety evaluation reports into the Updated Final Safety Analysis Report (UFSAR). In an Enforcement October 12. Conference conducted in the NRC Region III Office on Oc*tober 12, 1990, the licensee described the schedule for reconstitution of the UFSAR and measures to ensure adequate 10 CFR 50.59 evaluations in the interim. The Enforcement Conference is documented in Inspection Report 50-237/90025; 50-249/90024. The inspector has no further concerns in this area. 50-249/86012-48: Observation 2.5.4 from Safety System (Closed) Open Item 50-249/86012-48; Outage Modification Inspection (SSOMI). Concern regarding use of silicone grease on valve gaskets, seals and seats versus leak tightness. This item was reviewed in Inspection Report 50-237/89026; 50-249/89025, in response to the licensee's discovery of grease on the internals of the Unit 3 reactor building to torus vacuum breaker check valves. It was concluded that the grease discovered on the cheCk check valves was applied prior to the corrective actions to prevent greasing of valve seats to pass local leak rate tests. These corrective actions were described in package for that report. The inspector also reviewed the work request paCKage feedwater outboard check valve 220-62e 220-62B which contained specific prohibitions against use of lubricant on valve seats including a quality control hold point to verify this. The inspector has no other concerns in thi thiss area. Allegation AMS No. RIII-90-A-OI02 (Part B): Falsification of (Closed) A'legation Training Records. An allegation was made to the NRC concerning

 !(

6 111.15-11 IILIS-ll

Revision 8 April 1992 falsification of training records by "whiting-out "whiting-out"ll and backdating to show that training was received prior to performing work. According to the alleger, training was given on grinding and flapping of welds for generic use on October 10, 1990. The craft workers were told to backdate the training tralnlng records to September 20, 1990, to show that training was given prior to starting the task. The alleger and two otner other workers refused to backdate the training record and entered October 10, 1990. These three entries were "whited-out" IIwhited-out ll and changed to September 20, 1990. The inspector interviewed employees of Fluor Contractors International, Inc., (FCll), (FCII), and reviewed FCllFCII Site Procedure SP-lI-02. SP-II-02. Revision 0, "Orientation, Indoctrination and Training." Fell FCII Procedure SP-II-02 FCI! training matrix for required training. Grinding and referenced the Fell flapping are craft skills that would be performed either by a p~pefit~er or boilermaker. The required training for these crafts was F~II F~!I orientation and OAPs 1-4. Only the pipefitter and bOilermaker foremen were requi~ed, requi-ed, by the Fell FCII training matrix, to receive training in job specific proc;dures. procedures. In order to reduce job errors, the foremen performed a walkdown of thr, the, job and reviewed the task to be performed with the craft prior to starting the work. To give the craft a sense of personal responsibil :y, this informal training was documented using the Training Report Form found in Fell FCII training procedure SP-II-02. This work review and tra; trai ;ogi~g documentation was not procedurally required. The inspector reviewed work areas found in the "Outage IIOutage Package :7atus

c3tUS Report.

Report."1I Three areas were identified that would~ould include grindin~ nd flapping as part of the work. These were Inservice Inspection (1:,'), (I:",

(

( Erosion/Corrosion, and the Reactor Vessel Level Instrumentation Sy.c!m (RVLIS) Modification. The inspector reviewed the training report r~ S).~!m rc ~;~.

~~.

associated with the following work packages: 151 lSI Work Package Nos. 093346-1 through 21 Erosion Corrosion Work Package Nos. 093350-1 through 7 RVLIS Work Package Nos. 094094-1 through 10 The allegation was partially substantiated, in that there were training report entries where the date had been altered by writing over the original date. In one instance, the training report was dated September 21, 1990, and the first three entries were originally dated October 10 or 20, 1990 and then written Over over to reflect September 20, 1990. No white*out white-out was used to alter the entry. However, the training was not procedurally required and the training record was not a document required by the quality program. The contractor has indicated that a new form may be used in the future to document the work review. No further action is considered necessary in this area. Duplicate ~ following Unit 3 items are being closed because they are duplicates The fo11owing of corresponding Unit 2 items. These issues are still open and being tracked through the Unit 2 tracking numbers. {( i(

     \.\                                          7 III III.15-12
                                               .15-12

Revision 8 April 1992 50-249/90022-01 50-249/90022-02 Two examples of a violation and no deviations were identified in this area.

3. Licensee Event Renorts Reoorts Followup (90712 and 92700)

Through direct observations. discussions with licensee personnel,personnel. and review of records, records. the following event report was reviewed to determine that reportability requirements were fulfilled, fulfilled. immediate corrective action was accomplished. accomplished, and corrective action to prevent recurrence had been accomplished accompl i shed in i*n accordance with Technical Specifications. (Closed) LER 237/90010: Core Spray Pump 28 2B Automatic Start. This event including licensee corrective actions is discussed in paragraph S.a.l. 5.a.1. No violations or deviations were identified in this area.

4. Plant Operations (60705, 60710, 71707, 71710, 71714 and 93702)

The inspectors observed control room operations, operations. reviewed applicable less lc~s and conducted discussions di scussi on s with contro.l room operators during this wi th contr~l thi s peri~:. peri"=. The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affectlC affect ,,: components. Tours of Units 2 and 3 reactor buildings and turbine conditions. ir~:udin buildings were conducted to observe plant equipment conditions, irc:udin' potential fire hazards, hazards. fluid leaks, leaks. and excessive vibrations and :c ~c verify that maintenance requests had been initiated for equipment in nee". nee'.. of maintenance. The inspectors reviewed new procedures and changes to procedures that were implemented during the inspection period. The revi;".* revi~.* consisted of a verification for accuracy, and correctness. These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under Technical Specification' 10 CFR, and administrative procedures. Each week during routine activities or tours, the inspector monitored the licensee's security program to ensure that observed actions were being implemented according to their approved security plan. The inspector noted that persons within the protected area displayed proper photo-identification badges and those individuals requiring escorts were properly escorted. The inspector also verified that chec~ed ~ftal areas check.edvit&l &reas were locked and alarmed. Additionally. Additionally, the inspector also verified that observed personnel and packages entering the protected area were searched by appropriate equipment or by hand. In addition, a general plant walkthrough walk.through inspection was performed by NRC, Region III, Division of Reactor Projects, Branch Z, on October 16, 1990. i(: 8 I1I.15-13 III.15-13

Revision 8 April 1992 Comments from that inspection includlng including those concerning radiation practices were provided to the licensee for resolution. Unit 2 was shutdown for refueling on September 23, 1990. The inspectors reviewed the technical adequacy of approved procedures and establishment of administrative controls for refueling activities through Dresden Fuel"Fuel* Procedure (DFP) 800-1, ~Master "Master Refueling Procedure.~ Procedure," and other associated refueling and operating surveillance procedures. The inspector also verified implementation of these administrative controls prior to and during fuel movements by review of ~ppropriate rppropriate com~leted cheCklists, checklists, logs and surveillances, s'Jrveillances, direct observation, personnel interviews, and verification that Technical Specification requirements for refueling were met. Observation of new fuel receipt and licensee inspection was documented in inspection report 50-237/90017; 50-249/90017. Activities prior to fuel movement were also observed including reactor shutdown and various aspects of removal of the shielding blocks. blocks, drywell drywe11 head, reactor vessel head and dryer/separator. The inspectors verified that key personnel possessed an adequate understanding of their individual responsibilities and admini-strative requirements through direct observation and personnel interviews. Adequate staffing for refueling activities and adequate plant cleanliness conditions were also verified by the inspectors. Appropriate radiation protection controls were verified to have been implemented in conjunction with these activities. The inspectors also verified that steps were being taken for the fuel handling foremen to activate their senior reactor operator licenses in accordance with 10 CFR 55.53(f)(2). Specific incidents involving fuel handling activities are discussed in paragraph 7.a. The inspectors performed a detailed walkdown wa1kdown of the accessible portions of the Unit 2 high pressure coolant injection (HPCI) system and the Unit 3 core spray (C5) (CS) system. At the time of the walkdown. walkdown, the Unit 2 HPCI system was out of service for maintenance and modifications. Several minor defiCiencies deficiencies regarding the HPCI and CS systems were noted by the inspectors which were quickly resolved by the plant staff to the inspectors' satisfaction. The inspector reviewed the licensee's program and procedures re1ating relating to preventative measures taken for extreme cold weather. In response to IE Bulletin 79-24, the licensee stated that safety-related process, instrument and sampling lines had not experienced freezing and that the above ground ECCS lines entering the Dresden Unit 2/3 contaminated condensate storage tanks were well insulated, heat traced and contained in an insulated permanent enclosure. In addition, all other safety-related instrument and sampling lines were indoors and not exposed to sub-freezing temperatures. The inspector verified the material condition of the insulation on the ECCS lines, the presence of heat tracing and the adequacy of the insulated enclosure. The inspector verified the completion of Dresden Operating Surveillance (OOS) 010-9, Revision 2, which outlined equipment manipulations and inspections to be performed in preparation for seasonal weather changes. This surveillance specified the seasonal requirements for energizing tank heaters. heaters, heat ,( 9 III .15-14 III.15-14

Revision 8 April 1992 tracing and space heaters, and for inspecting steam heating coils and ( pipe insulation for signs of degradation. . '"~ Various operational occurrences were also reviewed as follows:

a. On October 14, 1990, while Unit 2 was defueled. defueled, approx1~ate1y Ipproxi~ltely 1,300 gallons of contaminated condensate water were spilled onto the third and fourth floors of the reactor building. The spill was the result of overflow of water through the reactor cavity ventilation duct openings. The reactor cavity was being flooded to support reactor vessel internal inspection but level should not have been raised past the bottom of the duct openings. Cavity fill was accomplished with condensate flow from the condenser hotwell with makeup from the condensate storage tank. The fuel handlers were in it i a 11 y mon initially i tori ng ca monitoring vi ty 1level cavity eve 1 from the refue refuel1 floor but 1later ater left, and informed the NSO of their departure. The change in level from that last reported by the fuel handlers and that later reported by an Equipment Attendant (EA) was noted to differ from the change reflected on the control room indication. In addition. addition, the NSO realized that control room indicated level had risen to where it had been maintained a week earlier. As such. such, the Shift Engineer and Shift Supervisor verified level to be below the ducts from the refuel floor. However, they did not approach close enough for positive verification since this would have necessitated changing into anti-contamination clothing. Therefore, they verified that the EA had gotten closer on his earlier check. Although the EA was later dispatched to again check level. level, the overflow occurred prior to the EA reaching the refuel floor.

Further review indicated that a precaution in Dresden Operating Procedure (DOP) (OOP) 1900-3, IIReactor "Reactor Cavity-Dryer Separator Storage Pit Fill and Operation of the Fuel Pool Cooling and Cleanup System During Refueling,lI Refueling," ReviSion 8, required constant communication between the refueling floor and the control room while filling the reactor vessel to prevent overflow into the ventilation ducting. However, neither of the two operating crews involved in the vessel filling actually utilized the procedure nor was the precaution followed. Failure to maintain constant communication between the refueling floor and control room while filling the reactor vessel in accordance with DOP 1900-3. 1900-3, is considered to be an example of a violation (50-237/90023-02a CORP>> (DRP)) of 10 CFR 50, Appendix 8, B, Criterion V. The operating crews were counselled in the significance of the event. event, the need for attention to detail and procedural adherence. All Operating Engineers wer. were. instructed to reference procedures when possible in Daily Orders. (The Daily Orders which prescribed filling the reactor vessel had not done this.) In addition. addition, a misleading operator aid being used in the control room was revised as to ventilation opening level. The Shift Engineers were also instructed to ensure procedures were out and adhered to for all complex. complex, unique or infrequent evolutions. Further corrective actions to address general concerns about events during the refueling outage are discussed in paragrap.h 7.b. ( 10 IILI5-IS III.IS-IS

Revision 8 April 1992 ( Additional longer term event specific corrective actions were were. being developed by the licensee.

b. On October 27.

27, 1990, the Swing Unit 2/3 Diesel Generator (OG) received an unplanned automatic start and tied to Unit 2 ESF Bus 23-1. At the time of the event, Unit 2 was in a refueling outage and Unit 3 was in power operation. The event occurred while removing Busses 23 and 23-1 from service in accordance with out-of-service (OOS) (O~S) request 11-1549 to facilitate breaker and cubicle preventative maintenance work. The intent was to remove these buses from service while still allowing the swing DG to supply* supply' Unit 3 if required. Further review indicated that actions were accomplished with 005 ODS 11-1549; however, the ODS was incorrect. The individual who wrote the ~OS. ODS, who held an inactive Senior Reactor Operator (SRO) license, correctly summarized by reviewing the applicable electrical schematic drawing that four knife switches had to be opened to accomplish the desired action. As this individual believed the drawing to be unclear as to the precise designation and location of the knife switches such as to make identification of the actual corresponding switches in the plant difficult, Dresden Operating Surveillance Survei llance (DOS) (~OS) 6600-6, "Bus Undervoltage and Emergency Core Cooling System Test for the Unit 2/3 DG" was referred to for clarification. Unfortunately, one of the switches in the procedure was not the same as to what that individual thought was the corresponding switch on the drawing. While the COrrect correct SWitch switch deSignated on the drawing was actually located on Bus 23-1, the one

 / -    in the procedure was located on a small panel about 3 feet behind Bus 23-1. It was incorrect to use the procedure in this respect

~ since it was designed for a different function. (In fact, in this test, the diesel generator was supposed to start.) DOP 6500-11, "De-energizing 4KV Bus 23-1 for Maintenance," referenced the proper knife switches but was also not utilized in prepar'ng the 005.~OS. The OOS ODS was reviewed in accordance with the licensee's administrative program by a Shift Foreman (SF) with an active SRO license. The first individual had attached a copy of the relevant page from the procedure to the 005 ODS which keyed the SF into using it 1n in his review.

Therefore, the OOS ODS was incorrect due to referencing of inappropriate documents for clarification of the electrical schematics during its preparation. As such, the 005 ODS was not appropriate to the (DRP>> of 10 CFR 50. circumstances in violation (50-237/90023-03 (ORP>> 50, Appendix B, Criterion V. The inspectors reviewed a recent previous violation involving incorrect 005 DOS checklists with three examples and determined the root causes to be sufficiently dissimilar. Therefore. Therefore, this event could not have reasonably been expected to have been prevented by the licensee's corrective action for the previous violation. The licensee initiated improvements to the undervoltage knife switches for all the Unit 2 and Unit 3 4 KV busses which had the potential for an unplanned DG start. The licensee also planned to develop specific procedures for de-energization of all Unit 2Z and Unit 3 4 kv bus combinations which have the potential for an unplanned DG start. Additional plans were initiated for issuance of a policy statement clarifying types of situations in which Operations should

(,

11 111.15-16 III.IS-I6

Revision 8 April 1992 request assistance from other departments during OOS preparation and verification. As this was considered to be an isolated occurrence and appropriate corrective actions were initiated, a Notice of Violation is not being issued in accordance with 10 CFR 2, Appendix C, Section V.A. Safety significance was also minimal since all loads had already been removed from Bus 23-1. Opening of the incorrect switch defeated some interlocks for ECCS equipment that were already ODS OOS for the outage.

c. On October 20, 1990, a fuel oil spill occurred in the Unit 2 diesel generator room. This was discovered by two members of the Technical Staff about the same time Unit 2 DG fuel oil day tank level alarm was received in the control room. Diesel fuel oil 011 day tank drain valve 2-5212-500 was found partially open and was immediately closed. A fire watch was posted until the spill was cleaned up.

Approximately 500 gallons of fuel was spilled to the oil separator tank with some drain funnel overflow onto the Unit 2 DG room floor. Safety significance was minimal since the DG was OOS for maintenance at the time. Further review indicated that this valve and diesel fuel oil transfer pump suction valve 2-52018-500 were checked to be shut by a non-licensed Operations Supervisor on October 8, 1990, in preparation for cleaning the main fuel oil storage tank. "Do Not Operate" tags supplied by the cleaning vendor were placed on the valves. However, no Dresden DOS ODS was written for this activity. On October 20, 1990, the Operations Supervisor opened both these valves to restore them to what he believed to be their previous pOSitions positions and, thereby creating the drain path. The Operations Supervisor was aware of DOS administrative requirements but failed to follow them to expedite the process. These administrative requirements contained in DAP 3-5, "Out-of-Service and Personnel Protection Cards," prescribe specific practices for removing and returning equipment to and from service servi~e including preparation. preparation, review, approval, documentation and independent verification methodologies. Failing to follow DAP 3-5 in regards to ODS requirements is . (50-237/90023-02b (ORP)) considered to be an example of a violation (SQ-237/90023-0Zb (DRP)) of 10 CFR 50, Appendix B, Criterion V. The Operations Supervisor was counseled as to the importance of interacting with Operations Department shift personnel and the necessity of following OOS ODS administrative requirements. In addition, the day tank valves on all emergency DGs were locked shut.

d. During observation of the repair of the Unit 2 diesel generator service water (OG (DG SW) Dezurik three-way valves (2-3905-525 and inspectors 2-3931-525) per Work Requests 090498 and 090499, the 1nspectors developed concerns regarding previous operations of the OG. In February, 1990, both valve stems were found sheered through at It the bonnet separating the valve operators from the plugs. The valves are used for flow reversal through the OGDG cooling water heat exchangers (HX). If either one of the two valve positions were changed without the other, then cooling water flow would completely bypass the DG cooling HX.

When the Shift Supervisor ($5) (5S) was notified of the degraded DG SW valves on February 9, 1990 a determination of the Unit 2 DG I( 12 III.15-17 III .15-17

Revision 8 April 1992 ( operability was appropriate. Although it was not clear tnrougn through interviews with associated individuals what the licensee considered in the operability determination, through review of additional documentation the inspectors agree that the DG was operable. However, as the determination of operability was not easily discernible, the inspectors were concerned that the justification for the operability determination was not documented. DAP 7-9, "Malfunction of Safety Related Equipment" discuHed discussed logging in the Shift Supervisor's Log significant information surrounding the circumstances so that a reasonable judgement can be made of the cause of the problem and its significance. However, DAP 7-9 was ambiguous as to the threshold for safety-related equipment problems for which this would apply. Review of the Shift Supervisor's log and interviews with licensee personnel indicated that documentation of the justification for operability calls was not a current practice at Dresden. As a result of a Corporate Nuclear Operations Directive issued prior to the inspector's concern, the licensee already had plans to address this as part of an equipment operability program. Specifically, the licensee planned to have a procedure that would prescribe documentation by December 31, 1990. The inspector has no further concerns in this area. A review of past performances of Dresden Operating Surveillance (~OS) (DOS) 6600-2, "Reversal of Emergency Diesel Generator Cooling Water Flow" subsequent to the February 9, 1990 discovery di scovery of the degraded valves revealed a complete performance of the Unit 2 DG SW flow reversal on February 25, 1990. Due to the degraded condition. condition, !( turning of the valve handwheel during the surveillance would not have resulted in actual *valve

                            ~alve pOSition position change although the plug position indicator would have shown a change. As a result, the failure to achieve actual flow reversal went unrecognized and the licensee's commitment to IE Bulletin 81-03, "Flow Blockage of Cooling                                                        My til us"U Cool ing Water to Safety System Components by Corbicula and Mytilus was not fulfilled. However, the safety significance of not performing the flow reversal in this case was minimal since the DG surveillance indicated adequate HX differential pressure and DG cooling. Since the intent was to perform the flow reversal, the licensee's surveillance program accounted for the commitment, and the safety significance in this case was minimal, this failure to achieve the actual flow reversal is not being considered a deviation from the NRC commitment. Of more concern to the NRC is the fact that these valves were known to be degraded such that the handwheel could not be used to change valve pOSition and yet the licensee did not ensure this knowledge was applied to the subsequent surveil1a~ce surveilla~ce performance. These valves were not repaired until over eight months after discovery. In addition, if only one of the two OG  DG SW valves had been degraded, the action by the operator on February 25, 1990, would have resulted in the isolation of cooling water to the DG.

However,this condition would have been identified by step 9 of DOS 6600-2, which required the operator to stand by at the DG to confirm proper SW cooling flow during the monthly DG operating surveillance test run conducted on February 25, 1990. In this case, the licensee's administrative programs were ineffective in assuring ( 13 III.15-IS III.15-18

Revision 8 April 1992 tha: tha~ the status and ramifications of degraded equipment was made ( known to appropriate personnel and reflected in decisions regarding subse,uent activities. subse~uent

e. DAP 7-14, "Control and Criteria For locked Locked Equipment and Valves,"

described the criteria for the selection of valves which were to be locked in position. Included in DAP 7-14 were manual valves which; o Maintain or could compromise the operability of an Emergency Core Cooling System (ECCS). Step 2.a (2) o Are in the flowpath of systems which are required for safe plant shutdown during post-accident situations. Step 2.a (3) Dezurik three-way valves on The inspectors observed that the DG SW Dezuri~ each of the three DGs were maintained in an unlocked condition. These valves were not listed in DOPDDP 040-M3, "loc~ed "Locked Valve List: Accessible During-During Operations," ReviSion Revision 13. The mispositioning of either one of the two DG SW valves would result in the isolation of the DG from cooling water flow. The DGs provided the emergency electrical power source for the ECCS systems. Based on the Technical Specification definition of operability, the status of the DG could compromise the functionality of the ECCS. Additionally, the DG, as defined in the UFSAR, was required for safe shutdown during design bases events, which included the simultaneous loss of offsite power. Although other manual valves were correctly loc~ed locked in the DG system, an exception had been made in this case due to the design of these particular valves which make them more difficult to ( operate. However, the intent of locking valves was to provide a positive barrier to personnel to signify the importance of that valve's position. In this case, that barrier was not particular valve1s provided and the licensee1s licensee's administrative procedure did not allow for that exception. The inspectors noted that the manual containment isolation valves on the drywell manifold sample systems were also unloc~ed unlocked on both units. These valves were also not included in DOP 040-M3. The issue of loc~ed locked manual containment isolation valves was addressed in the systematic evaluation program (SEP). As indicated in a Safety Evaluation Report dated September 24. 24, 1982, the NRC valves should be position was that manual containment isolation va1ves position such administratively controlled and locked in a closed pOSition that the valves were not fnadvertently opened during periods when containment integrity was required. This staff position on manual containment isolation valves at Dresden has been consistent with NRC 10 CFR 50. 50, Appendix A, General Design Criteria, 55, 56,56. and 57. As part of the SEP process, CECo committed, per correspondence on O*Connor. to changing the November 18, 1982, from T. J. Rausch to P. O'Connor, appropriate procedures to implement administrative controls ensuring cOntainment isolation valves would be loc~ed manual containment locked closed. The licensee's administrative procedures were consistent with this commitment. c 14 IlI.15-19

Revision 8 April 1992 Failure to maintain the DG SW three-way valves and the dryWell ( manifold sample system manual containment isolation valves in a locked condition in accordance with DAP 7-14 is considered an example of a violation (50-237/90023-02c CORP>> (DRP)) of 10 CFR 50, Appendix B, Criterion V.

f. During fuel loading on November 12, 1990, fuel loading was suspended when abnormal indications were recognized on Source Range Monitor (SRM) 23. While investigating the cause of these indications from under the reactor vessel, instrument instr"ument maintenance technicians noted that SRM 22 had dropped from its fully inserted position.

Subsequently, SRM 22 failed a response test such that it appeared SRM 22 may not have been operable and responding for a short period while loading fuel in its corresponding core quadrant. This is considered an unresolved item (50-237/90023-04 (ORP>> (DRP)) pending further review of the extent and cause of this problem.

g. The inspectors noted that the licensee's policy was not to declare Technical Specification (TS) equipment inoperable and officially enter associated TS limiting conditions for operation when the equipment was purposely rendered inoperable for the purpose of TS surveillance testing. Examples included the standby liquid control system test in which the injection path was manually isolated, the diesel generator surveillance in which manual loading of the diesel generator rendered the load shedding feature inoperable, HPCI and isolation condenser isolation instrument surveillance in which an installed jumper prevented automatic isolation and a torus to reactor building vacuum breaker instrumentation surveillance 1n in which the differential pressure transmitter was valved out-of-service. In addition, the inspectors noted that upon a control rod accumulator high water/low pressure alarm which accumulato c, the practice indicated possible inoperability of the accumulato~J was to allow up to an entire shift prior to investigating the alarm.

This permits a long delay during which the accumulator may be inoperable and action not taken to restore the accumulator to operability. These practices in regard to Technical Specification operability are considered an unresolved item (50-237/90023-05 (DRP>> (DRP)) pending further clarification of reqUirements. requirements. Three examples of a violation, one example of a non-cited violation, and no deviations were identified in this area.

5. Maintenance and Surveillances (62703, 61726, and 93702)

a. Maintenance Activities Station maintenance activities of systems and components listed below were observed or reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications.

The following items were considered during this review: 15 III.15-20 II I.1S -20

Revision 8 April 1992 ( The Limiting Conditions for Operation (LCO) were met while components or systems were removed from service; approvals approval s were we're obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicablej applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemented. Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety-related satety-re1ated equipment maintenance which may affect system performance. The inspectors witnessed or reviewed portions of the following activities: Rebuild of the 2A2 Diesel Generator Air Start Relief Valve Welding of the "e" "C" Recirculation System Riser Overlays Unit 2 Diesel Generator Service Water Three-way Valve Repair Control Rod Drive Replacement Recirculation Pump 2A Suction Valve Repair Diese1.Generator Air Start Regulator Replacement Unit 2 Diesel.Generator Various occurrences were also reviewed as follows:

 ,c (1.) On October 3, 1990, while the reactor was being defueled, defue1ed, core spray (CS) pump 28 automatically started. At the time, all low

\ pressure coolant }njection 1njection (LPCI) pumps were out of service and both CS pumps were operable. During refuel conditions, Technical Specifications only require operability of two CS pumps, two LPCI pumps, or a combination of one LPCI and one CS pump. Only one diesel generator was operable for Unit 2 in accordance with Technical Specifications for refuel conditions. This was the swing 2/3 diesel generator which supplied emergency power to CS pump 2A. Electrical maintenance personnel were performing a preventive Unit 2 diesel generator output breaker. work package on the Un;t This involved removal of the breaker from the cubicle, cleaning of the cubicle and replacement of ~a contact switch inside the switch, in series with the CS pump actuation cubicle. This switch. circuitry, was to provide information to the circuitry on circuitry. whether the diesel generator output breaker was open. The CS circuitry upstream of the switch was de-energized since an

• Signal was not present. Changing out the actual initiation signal switch did not render the pump inoperable since it was stl1lstill capable of automatic start through the load sequence portion of This,wou1d have just resulted in &

the circuitry. This*would a ten second start delay. If an actuation Signal signal occurred. this portion of the circuitry picked up in parallel to the immediate start circuitry regardless of whether an undervoltage.condition undervo1tage.condition existed. i( 16 II I .15-21 III.15-21

Revision 8 April 1992 ( The most likely cause of the automatic start was that while changing out the switch a lead may have inadvertently been grounded allowing enough voltage from the downstream circuitry to pick up the pump start relay. After ensuring that an initiation signal was not present or needed, the operators took the pump control switch to pul1-to-lock. pu11-to-10ck. No other portion of the system actuated except for the pump minimum flow valve. The CS pump was considered inoperable at that point and the appropriate action statement entered. Electrical technicians were aware that although the breaker was out-oi-service and removed from the cubicle, the Circuitry circuitry involving the switch was not out-of-service. Therefore, the instrument technicians were aware that adverse actions could occur with this activity and, therefore, took precautions in accordance with the work package including utilization of a rubber mat. The work package was discussed with Operations personnel prior to receiving permission to begin the work. This included review of associated aSSOCiated drawjngs drawings that indicated the existence of core spray ;nterloc~s. interlocks. However, it was not entirely clear from the work package and the reviewed drawings as to what the interloc~s interlocks accomplished. As such, the licensee believed that if Operations personnel were aware of the nature of these interlocks they may have halted the work activity for a few days until the CS pump was scheduled to be removed from service. As such, the licensee's corrective action was to require listing in the work package of possible specific interactions for any equipment that may have interlocks that affect other systems or contacts that may energize or de-energize equipment or related circuits. In this way, Operations reviewers would have more information on which to base decisions as to whether to let work begin. It must be noted however, that this type of decision is dependent on the individual and the circumstances such that permission to proceed may be given anyway. Therefore, this corrective action may not be sufficient to preclude repetition. However, in this case, the inspectors believed the root cause to be difficult to address since reasonable precautions were taken in changing out the switch. In addition, arriving at this root cause was by process of elimination of any other causes but was still not conclusive beyond any doubt. Further corrective action to address general concerns about events during the refueling outage is discussed in paragraph 7.b. (2. ) On October 15, (2.) IS, 1990, Unit Z 2 outboard containment isolation feedwater check valve Z20-6ZA 220-62A was mistakenly disassembled instead of the corresponding train B valve. Due to leakage problems, both the A and B B valves were to be worked on sometime during the refueling outage. The B B train had been correctly taken out-of-service in accordance with OOS DOS 1I-1279 11-1279 on October 6, 1990. The Mechanical Maintenance Foreman (MMF) responsible for the job, walked wa lked down the ODS 005 on the correc~ correct train tra i n on October 11, 1990. However, the MMF later mistakenly directed work to be c( 17 111.15-22 III. 15-22

Revision 8 April1 1992 Apri performed on the A valve. Work package 081758 clearly designated ( the B valve. In addition, sufficient identification tagging existed on the A train such that the problem would have been apparent if the tags had been checked. Quality control hold points existed in the work package but were on liter later instructions involving re-assembly of the valve. In addition. addition, Technical Staff engineers responsible for local leak rate testing examined the valve after the valve cover was removed. These individuals also failed to recognize t~at that this was no~ not the BB valve. The Technical Staff system engineer was aware of the work but did not personally view the valve since other Technical Staff personnel were performing that function. As such, the lack of attention to detail on the part of the MMF, coupled with the unquestioning reliance of other personnel that the MMF was correct. correct, caused the wrong valve to be disassembled and not discovered until October 9, 1990. OAP 15-6, "Preparation and 9,1990. Control of Work Requests," Revision 0, required work to be performed per repair manual(s), traveler/procedure, or work instructions provided in the work package. Failure to disassemble the correct valve in accordance with the work package 1s is considered to be an example of a violation (SO-237/90023-02d (50-237/90023-02d (ORP) (ORP>> of 10 CFR 50, Appendix B, Criterion V. On that date radiation protection personnel noted that doses to workers on that job were much less than expected since the 8B valve was known to be more highly contaminated than the A valve. A check as a result of this information identiffed identifIed the error. It must be noted that the disassembly. disassembly actually ( occurred prior to the generic attention to detail corrective actions discussed in paragraph 7.b. It was fortunate that safety significance in this case was minimal. The A line had been used approximately two days earlier for tilling

                                                         ~;lling the Unit 2 Therefore, if the valve had been in a reactor vessel cavity. Therefore.

disassembled state just two days earlier, the X-area (steam tunnel) would have been flooded. In addition, if the inboard containment isolation feedwater check valve hadnlt hadn't held, the reactor vessel cavity could have partially drained back through this line. The licensee was still developing event specific corrective actions at the end of the inspection per1od. period. (3.) On October 19, 1990, the inspectors identified six Appendix "R" URn emergency lights (required for safe shutdown in the event of Ia disabling fire) with the electrolyte level below the add line. The inspector observed electrolyte level varying from just below the add line to one inch below the add line. The Emergency Lighting Monthly Inspection, Dresden Electrical Surveillance (DES) 4153-02, stated that "Electrolyte level sha l1 be at the full 1line". shall ine". However, contrary to the established procedure, the licensee indicated that a practice 18 IIIl.15-23 Il.15-23

Revision 8 April 1992 had been followed such that the emergency lights need only be filled when the electrolyte level was at or below the add line. The licensee further indicated that also contrary to the established procedur.e, the determination to add distilled water was at the discretion of the maintenance personnel. Conversations with the emergency light vendor and review of the vendor technical manual indicated that allowing the electrolyte level to fall below the add line could cause damage to the battery. After the inspector identified the low electrolyte level in the emergency lighting units, the licensee initiated immediate corrective actions which consisted of: (1.) Inspected and provided maintenance on Unit 3 emergency lights requiring servicing (for example adding distilled water to a battery with low electrolyte level.) Unit 2 was defueled at the time. (2.) Review of the emergency lighting maintenance procedure. (3.) Conduct of an investigation. On November 14, 1990, the licensee indicated that an investigation report was being developed and would include an event summary, root cause(s) and corrective action(s) which would also be implemented for Unit 2. In addition, the licensee would document .( the emergency lights in the as-found condition on emergency lighting drawings. The licensee also indicated the investigation report and the marked up drawings for Unit 3 will wi 11 be tentatively completed by December 14, 1990. This is considered an unresolved item {SO-237/90023-06 (50-237/90023-06 (DRP)) pending review of the lfcenseels licensee's submittal. submitta 1.

b. Surveillance Activities The inspectors observed surveillance testing, including required Technical Specification surveillance testing, and verified for actual activities observed that testing was performed in accordance with adequate procedures. The inspectors also verified that test instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were accomplished and that test results conformed with Technical Specification and procedure requirements. Additionally, the inspectors ensured that the test results were reviewed by .*

personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel. The inspectors witnessed or reviewed portions of the following test activities: Unit 3 Rod Swapping Emergency light Light Eight Hour Discharge Test Radwaste River Discharge SPING Calibrat1on/Setpoint Calibration/Setpoint Adjustment 19 111.15-24 III.1S-24

Revision 8 April 1992 Unit 2 250 VDC Battery Discharge Test (( Source Range Monitor Checklist The following occurrences were also reviewed: (1.) On August 8.8, 1990, while calibrating the Unit 3 Torus to Transmitter, Reactor Building Vacuum Breaker A Pressure Transmitter. DPT-1622A. OPT-1622A, the instrument technician inadvertently adjusted OPT-1622B causing Vacuum Breaker B to open. DPT-1622A DPT-1622B calibration was being checked per Dresden Instrument Surveillance (DIS) 1600-20, "Torus to Reactor Building Differential Pressure Transmitter 1622A and B talibration Lalibration and Maintenance Inspection" in accordance with Work Request 094439. This and other prescribed testing was to collect data for a non-detectable failure evaluation of Rosemont (Model 1153) DPT-162ZA was valved transmitters. During the check DPT-1622A out-of-service in accordance with the procedure and was, therefore, inoperable. When the as-found readings were discovered to be outside the tolerance range described in the procedure, the instrument technician was to perform a re-calibration to correct the problem. The two transmitters were located approximately eight inches apart and access to the calibration adjustments were on the underside of the transmitters. Each of the transmitters were labelled with a small label under the transmitter. To adjust the calibration setting, the instrument technician had to turn bacKwards backwardS to where he was previously standing performing the calibration check in order to look up at the transmitter from below. (( Therefore, the transmitter that had previously been on the

  \

techniCian's left for the calibration check was then on the mistakenly right for the adjustment. As such, the technician mistaKenly adjusted t ran smit ter. DAP 15-6, "P~eparation adj us ted the wrong transmitter. aPr-epa ra t i on and Control of Wor~ Work Requests", Revision 0, required work to be performed per repair manual(s), traveler/procedure, or work instructions provided in the work package. Failing to follow the work request by adjusting the wrong transmitter is considered to be an example of a violation (SO-237/90023-02e (50-Z37/90023-02e (DRP>> of 10 CFR SO, Appendix B, Criterion V. However, safety (DRP>> significance is considered to be minimal in this case since adjustments were made in a direction that were conservative to Technical Specifications and, therefore, Vacuum Breaker B was never inoperable as to its relief function during the event. In addition, although the vacuum breaker was open for a brief time and therefore unable to perform a containment isolation function, its corresponding check valve remained closed. The vacuum breaker was immediately restored. The licensee counseled the instrument technician on the need for total job awareness especially when working in congested areas such as this. This event was also tailgated to instrument department personnel. The licensee also enhanced the labeling of both the "Z and 3 transmitters and planned to rotate the Unit *2 transmitters such that the adjustment screws could be viewed to p. from the top. ( 20 111.15-25 IIl.15-25

Revision 8 April 1992 (2.) On November 14,

14. 1990, 1990. the licensee discovered that the filter

( in the Unit 33 Reactor Building Ventilation Air media in Particulate Sampler had been misaligned in the filter holder. This allowed a portion of the sample flow to bypass the filter. This is considered to be an unresolved item (50-237/90023-07 (ORP>>) (ORP>> pending further review for the cause and significance of this event. Two examples of a violation and no qeviations deviations were.identified in this area.

6. Enaineerina Enoineerino and Technical Support (37828)

The inspectors reviewed the modification package to alter the diesel generator air start system (M-12-2-88-06). The modification was the result of a design weakness identified as a result of the Safety System Functional Inspection conducted in 1988 by the licensee. The inspectors observed the physical work of the resupport of the air receiver drain piping and verified the work was performed by qualified workers and in accordance with approved instructions and drawings contained in the work package. Additionally, Additionally. welder qualification records for those individuals welding the hanger supports were verified. No violations or deviations were identified in this area.

7. Safety Assessment/Quality Assessment/Duality Verification (35502 and 40500)

a. On October 1.

I, 1990, 1990. while Unit 2 was shutdown for a refueling outage . ( a. and fuel was being moved from the vessel to the spent fuel pool, pool. the licensee discovered that the fuel movement was out of sequence. Fuel moves were deSignated designated by the Nuclear Material Transfer Checklist (NMTC) in accordance with Dresden Technical Surveillance 8471. ~General (OTS) 8471, "General Procedure For Fuel Transfers Involving the Reactor." Step 581 of the NMTC indicated that fuel assembly XZB067 Reactor.~ X2B067 at core location 45-46 was to be transferred to Spent Fuel Storage Pool (SFSP) location F2-A7. Instead, Instead. fuel assembly X2Cl13 X2C113 at core location 43-46 was moved to that SFSP location during NMTC step 581. The error was noticed prior to movement of any other fuel assemblies and all fuel movement was halted. Safety significance was minimal since as this was offloading of fuel, fuel. a criticality concern did not exist. Further review indicated that poor communications and inattention to detail contributed to the event. The fuel assembly to be moved was the last fuel assembly in the control cell. The following step, 582. involved a transfer from a different core step. 582~ region. The Fuel Handling Supervisor went onto the fuel grapple to caution the fuel handling crew of this fact. The independent verifier and grapple operator were scheduled to swap duties starting with step 582. Therefore, Therefore. following the caution just received about that step. the independent verifier was studying a core map in regard to step 582 instead of independently verifying step 581. The fuel handling error was discussed with the current and later the fue1 oncoming crew to emphasize the importance of attention to detail an proper independent verification. The independent verifiers were instructed to communicate to the grapple operator whether or not the ( 21 111.15-26 III.lS-26

Revision 8 April Apri 1 1992 proper fuel assembly was grappled prior to moving the assembly. ( (Before the event, positive communication was necessary only if the wrong assembly was latChed.) latched.) Increased supervision to confirm the effectiveness of the independent verification was initiated. In addition, the licensee decided to expedite repairs to the core position indication system (CPIS) on the grapple which would have aided the fuel handlers to identify the correct assembly had it been entirely operable. On October 2, 1990, despite the previous corrective actions, another fuel assembly mispositioning event occurred. An Electrical Maintenance Supervisor (EMS) was on the fuel grapple to observe the operation of the CPISin preparation for repairs as discussed above. The independent verifier was discussing its operation with the EMS. Step 12 of Revision 2 of Part 7 of the NMTC prescribed movement of fuel assembly X2C160 at core location 25-28 to SFSP location F2-E1. F2-El. The grapple operator instead moved fuel assembly A20109 in core location 27-28. The independent verifier gave a cursory inspection of the core location and latched condition, while engaging in conversation with the EMS, and gave verbal permission to move the fuel assembly. The error was noted when moving the grapple to the next fuel assembly to be relocated and fuel loading was again halted. This event was again related to inattention to detail and lack of self-checking. A discussion involving management and the fuel handlers themselves was conducted to determine the best method of independent verification. It was determined that confusion still existed regarding the process the independent verifier followed during fuel moves including communications and the process was inadequately defined in appropriate procedures. In addition, external distractions were not adequately controlled on the grapple during fuel movement. A meeting was held between licensee management and all fuel handlers to stress the importance of attention to detail, independent verification and good communications. A temporary change was issued to DAP OAP 7-7, "Conduct of Refueling Operations" to restrict grapple access during fuel movement. The CPIS was also repaired prior to resuming fuel movement. The licensee also planned to revise fuel handling procedures prior to the next refueling outage on Unit 3, currently scheduled for April 1991. 1991, to clarify the duties and responsibilities of the independent verifier and to establish compensatory measures when the CPIS ;s is inoperable. Further corrective actions to address general concerns about events during the outage is discussed is paragraph 7.b. fuel~ Further review of past events, found two previous and similar fuel' loading errors on January 10 and 12, 1989 during the last Unit 2 refueling outage. The licensee had determined the root cause of these events to be fuel handler inattention to detafl. detail. As I result, a memorandum had been issued to ensure an independent verifier visually verified the correct storage and core locations in addition to verifying fuel assembly latching. It also emphasized clear and concise communication. It was evident that this corr~ctive corrective action was insufficient to prevent the later October 1, I. 1990 event. Furthermore. Furthermore, the correc~ive corrective actions from the October 1, 1990 event l 22 111.15-27 III.lS-27

Revision 8 April 1992 were also insufficient to prevent still another event on ( October 2, 1990. Inadequate corrective actions 1nin response to the January 10 and 12, 1989 and October 1, 1990 fuel assemblY assembly mispositioning events is considered to be a violation (DRP>> of 10 CFR 50, Appendix 8, (50-237/90023-08 (ORP) B, Criterion XVI. The remaining unloading of fuel and the reloading of fuel during the current refueling outage following additional corrective actions did not result in any fuel assembly mispositioning errors.

b. As described elsewhere in this report, a number of events occurred during the Unit 2 refueling outage which were indicative of personnel performance problems such as poor communications and inattention to detail. These included two fuel bundle mispositioning events, an inadvertent automatic start of a core spray pump, a reactor cavity overflow event, disassembly of the wrong feedwater isolation check valve, inadvertent draining of a diesel generator fuel 011 oil day tank.

tank, inadvertent diesel generator start and loading and several other events which are either covered in other inspection reports or were not related to reactor or radiation safety. It appears that the frequency of these types of problems increased dramatically during the Unit 2 refueling outage as compared to the last Unit 3 refueling outage. This was not a contractor control problem since the majority of events involved station personnel across several organizational boundaries. Licensee management recognized the adverse trend and instituted specific action to address personnel performance problems on a generic basis. These generic actions included speCial special meetings emphasis these events and management expectations of priorities to to emphasiS workers. Outage work activit1es activitles were temporarily reduced (substantially on Sundays) to ensure workers were well rested and to emphasize self-check program, attention to detail over schedule. In addition, a self-cheCK recently implemented for operations personnel in response to a previous violation, was expanded to the entire site. A third party review team was requested to review past events for any new insights. The inspectors observed substantial management involvement to address the problems.

c. While observing performance of a quality control (OC)

(QC) hold point in work request 95491, the inspector noted that the Q.C. inspector identified that the step was being performed incorrectly. The work request involved repairing of the air receiver tank relief valve 2A2 for the Unit 2 diesel generator. The particular QC hold point was on a step for bench setpoint adjustment of the relief valve. The mechanics had set the relief valve to "pop" fully open within the set pressure band delineated in the procedure. However. However, Aa relief valve will initially open part way in order to relieve pressure back to acceptable system pressure. If system pressure continues to rise the valve will fully open or pop. As it was set. set, the valve would have relieved below the specified tolerance band. The QC inspector explained this to the mechaniCS mechanics who then correctly Adjusted adjusted the setpoint. Followup to this problem was provided by completion of a QC Inspection Feedback Sheet by the OC QC inspector. This document is sent to the involved department to inform departmental supervision of ( 23 III.lS-28

Revision 8 April 1992 taken. the problem so that any actions they feel appropriate can be ta~en. ( However, this methodology did not provide a tracking mechanism to ensure that the root cause ;s is identified and appropriate corrective action is taken. The licensee stated that this mechanism was instituted to address lesser problems that would not be important enough to identify through other available problem reporting programs such as deviation reports. This is considered to be an unresolved item (50-237/90023-09 (DRP>> pending further review of the administrative guidance regarding these feedback sheets. sheets, type~ of problems identified in these feedback sheets, threshold criteria for other deviation reporting methods and the adequacy of actions taken by various departments in response to these feedback sheets.

d. The inspector observed the scram/engineered safety features (ESF) actuation reduction main committee meeting held on November 2, 1990.

The committee reviewed the status of corrective cOrrective actions that were being instituted in response to previous scrams SCrams and ESF actuations to prevent further occurrences. In addition, a review and discussion of recent events was performed during the meeting to ensure adequacy of planned corrective actions from a scram/ESF reduction standpoint. The status of BWR Owners Group Scram SCram Frequency Reduction Recommendation Trac~ing Tracking System items Ind and Ia recent Owners Group conference report were also discussed. This was viewed by the inspectors as a genuine effort to incorporate lessons learned from other facilities to prevent adverse occurrences. The inspectors regarded the licensee's scram/ESF reduction activities to be beneficial in light of the smaller number of scram/ESF actuations actua"tions occurring in 1990 compared to the previous year. One violation and no deviations were identified in this area.

8. Reoort Review (90713)

During the inspection period, the inspector reviewed the licensee's Monthly Operating Report for September 1990. The inspector confirmed that the information provided met the requirements of Technical Specification 6.6.A.3 and Regulatory Guide 1.16.

9. Unresolved Items Unresolved items are matters about which more information is required requ1red in order to ascertain whether it is an acceptable item, an open item, a deviation or a violation. Unresolved items disclosed during this inspection are discussed in paragraphs 4.f, 4.g, S.a.3, S.b.l 1nspection S.b.2 and 7.c.

10. Exit Interview The inspectors met with licensee representatives (denoted 1nin Paragraph 1) on November 16, 1990, and informally throughout the inspection inspect10n period, and summarized "the

                       ~he scope and findings of the inspection activities.
;(                                        24 111.15-29 III.15-29

Revision 8 April 1992 cc The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any such documents/processes as proprietary. The licensee acknowledged the findings of the inspection. . c- 25 111.15-30 III .15-30

Revision 8 1& IS.

       '~

Commonwealth Edison 1400 Oeus DDUS Place April 1992 ( 'i9 Downers Grove, illinOIS 60515 Grove. IllinOIs December 14, 1990 Mr. A. Bert Davis Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137

Subject:

Dresden Station Units 2 and 3 Report Pertaining to Unresolved Item 50-237/90023-06 NRC Docket Nos. 50-237 and 50-249

References:

(a) (a) W. Shafer (NRC) letter to C. Reed (CECo), dated December 7, 1990. (b) Conference Callan CalIon November 14, 1990 CECo (K. Peterman, E. Skowron, between CECa J. Lindahl) and NRC <D. (D. Hills, J. Holmes>. Holmes). Mr. Davis:

(

Reference (a) transmitted Inspection Report 50-237/90023 50-237/90023. and 50-249/90023 for Dresden Statton. Station. The Inspection Report contained an Unresolved Item (50-237/90023-06) regarding low electrolyte levels identtfied identified in six (6) emergency lighting units. As requested by your staff 1n 1n in the Reference (b)(b) teleconference, attached 1s is the investigatton investigation report performed Station on this unresolved 1tem. by Dresden Stat10n Item. This report, which 1s Is the Deviation Report, includes an event summary, a root cause evaluation station's Deviatton statton1s and corrective actions. Add1tionally, Additionally, the requested emergenc' lighting drawings have been directly to Mr. J. Holmes of your s:af~. provided dtrectly prov1ded Please direct any questions or comments cOlllllents on thts this matter to this off'ce~ offfce~ Respectfully, Respectfully. T.J. Kovach Nuclear Licensing Manager

Attachment:

Station Deviation Report 12-2/3-90-123 Dresden Statton cc: B. Stegel Siegel - NRR Project Manager NRR Document Control Desk Resident Inspector, Dresden D. Hills - Senior Res1dent J. Holmes - Region III Inspector MR: TK: lmw 1mw III.1S-31 III.15-31 ZNLD65217

Revision 8 April 1992 ( AITACHMENT DRESDEN STATION DEVIATION REPORT 12-2/3-90-123

(

( III.lS-32 IlI.1S-32

DEVIATION REPORT Revision 8 Apri April1 1992 I avo NO. NO. 12 -- 2/3 - 90  :~ 123 I (( PART 1 I TITLE or DEVIATION TITLE DEVIATION STA UNIT vEAR UNtT Inadequate Safe Shutdown Emergency Light V(AR NQ. I . OCCURRED 10/19/90 tonn tOni! 1000 Re: Re.: El.rtrnl El r 1 *** L . . . l, Duo tn Mon.n.ment O.fici.ncv OATE TIME ArrECTED 4100 SYSTEM AFFECTED PLANT STATUS AT TIME or EVENT TESTING NfA NIA YES Fire rir. prQitt~,eC:'t~i~orn~~-1~~~0~E=R=.=f=u:e1::!:R=un=)~~PO~W~E~R~(r.:.)~0::(9=7:)=r.===-~~WO::R:K~R~E~Q~U~E:ST~N~O~' 1--2I;-=-=-~IY_E_S PrQ~~~ ____ ~ MODE Refurl !Run) PQWER(r.) Q (97)% ____________________________________ ~ WORK REQUEST NO. ________________ __ ~ I:=I __~I~X:: ______________ I_x_ OESCRIPTION Of DESCRIPTION or EVENT At 1000 hours on October 19, 1990 the NRC Resident Inspector notified the Electrical Maintenance Haintenance Department (EMO)(EMe) Haster Master Electrician of possible deficiencies regarding siveral slvlra' Safl shutdown (SSD) {SSD} e~ergency lights, lights. The lights were installed to mtet requ;re~nts of 10CFRSO meet the requirements Appendix R, 10CFRSO App,nd;x R. Section III.J. The EHO immediately responded to these thesl concerns by re-plrforming rl-performin9 the routine maintenance check on the SSD SSO emergency lights in accordance accordanc. with Dresden Electrical Electr;cal Surveillance (DES) 4153-2, Monthly Emergency Lighting Monthly Honthly Inspection. To verify operability, eight hour battery discharge discha~ge tests were performed pe~fonmed on selected emergency lights which Wire were found to have low electrolyte levels. As the SSD SSO @mergency emergency lights are not controlled by the Technical Specifications or Dresden Administrative Technical Requirements (DATR). No Limiting Conditions for Operations (LCD) (LCO) were entered into or exceeded during this event. SIGNIrICANT EVENT PER NOD DIRECTIVE OP.TO POTENTIALLY SIGNIfICANT OP.l0 YES NO I_I L. . . .LI L.L-I NO 1 HOUR 10crR50.72 NRC RED PHONE 10CFR50.72 I_I I_I

~ ('.                                                                                I X NO     NO                                                            11/13/90
'(
  \

NOTIrICATION HADE NOTIfICATION I1 _ _ 144 HOUR - - - l_x_1

                                                                         -T-H-T M

• AT PART 2 OPERATING ENGINEER'S COHHENTS Tech Staff and EHO are working on long term corrective actions. This Deviation Report was issued as a matter of documentation and to establish a detailed review. ~eview.

l_x__ 1 NON REPORTABLE EVENT l_x_1 NOTIFICA TION,____-'N ..../....A:.-____ NOTIFICATION NlA 30 DAY REPORTABLE/10CFR 3D REPORTABLE/10CFR REG ON III DATE TIME TIHE I I_ -I I 5 DAY REPORT PER 10CFRlI 10CFR21 HI !!is j,v'IDlkh lIl!4l2a 11114/90 1547 I_I N50 N5D DATE TIME TIHE ANNUAL/SPECIAL REPORT REQUIRED I_I I_I CECO CORPORATE NOTIFICATION HADE A.. LR. /1, A.LR. /1'_ __ __ ___ ____ _ _ 1----1 1__ 1 IF If ABOVE NOTIFICATION NOTIfICATION IS PER 10CFR21 IOCfR21 L.E.R. /I TElECOPY TELECOPY NIA T PRELIHINARY REPORT COMPLETED AND REVI[W£O COI1I'LETED AI<<) REVIEWED e. A. lank

8. Zank INVESTIGATION REPORT &RESOLUTION ACCEPTED BY STATION REVIEW

(' ( RESOLUTION APPROVED AND

     '-- ..        AUTHORIZED fOR    rOR DISTRIBUTION DATE B6_5176 (form 86-5176      (rorm 15-52-1)           4-12-90 DVR/B7 eVR/87                                                                        III.lS-33 IIl.lS-33

Revision 8 prl'1 1992 April A DEVIATION OEVIATION INVESTIGATION REPORT (OIR) Fo .... Rev Fol"II R.y rae; 1 ; ty Nallllt Facility Nafrllt I PAGE ( Or.. den NNuclear Drnden Pn.er Station Ichar P"..,er Stabon Un; UnHt 213 II IOFI 0 Iorio Ti tle Ue Inadequate Safe Shutdown Emergency light Light Electrolyte Levels QUI:: tQ Hinig~m!:n~ Du!: Hanigfmfnt Q~fi~i!:n~~ Q£fi~ifn~~

         ~VENT QAT~ QATE                                          OIR !!l!~ER till!l!lER                       REPQRT QATE REPORT     DATE OPERATING                          ~~i~ii~i'
                                                                                                                                                                    ~~~~~m'
                                                               ~i
                                                               ~~

SEQUENTIAL REVISION

                                                                                     ~~ NUMBER HQNTH DAY YEAR                                                       ZZW;f/Z:

zzzn;IlI: Itt iii~zi~i; m~Zm; MODE MOOE IMl;LNTH MONTH DAY YEAR STAA !UNIT Sf UNIT YEAR II I~ NUMBER NUHeER t'I I NUHeER IHIlNTH IN POWER ~~~~~~~~I m~~m; azzzzzz~ 1I 1o 10 11 19 99 10 I1 12 22 13 1919 10 - II 12 I 33 - oI0 010 1I 12 1I 11 19 ,g 10 LEVEL 10'0 ~mm~' JW~~~ 19 17 I~~~~~W/: I:QNTAI:T FQg

                                                                           ~QNTA~T      FQ~ THIS QIR DIR NAME rri c Skowron Te~hn;CJlI
            !=:ri~                   Technical Staff Enainl!er Enoin.er ExL     Ext. 2485 AREA CODE ~

_8l11~ I TELEel::K!tf~ TELEeH2HE

                                                                                                                                      ~i4121-121912 81115914Izl_I"qt?
                                                                                                                                                           !jJalDEB tflltlltB
                                        ~QHP~tTt I:QHP~ETE gNE  QNE LINE FQR EA,H EA'H ~QHPQNENT         FAILURE QES~BleEa
                                                                                 ,QMPQNENT fAILURt       QESI:BIaEQ IN THIS Bte2RI BEeQR!

CAUSE SYSTEH COMPONENT MANUFAC-SYSTEM REPORTABLE REPORTAB HANUFAC-TURER TO NPROS

                                                                                          ~~~~~~~ CAUSE SYSTEH COMPONENT ""NUFAC-
                                                                                          ~~~~~i~

II/un HANUFAC-IO NPRO IIIIIZI MER TUI[II TO Kip IT Ill. Iv lola 13 N

                                                                                          ~~~~~~~                    I         I I JI          1I II_II E

E BIT B 1R IY T Tlot813

                                                                                          ~iiiiii
                                                                                          ~mm                                  11 1t           I I JI 11                11111 I                             j~0J~i iJ~m~i                    J          , I 11               I1 II 1 SUPPLEHENTAL REPORT EXPECTEP SUPPLEHENTAl                                                                                            I_TH It<<>NTH DAV   DAY   'Y r'

EXPECTED SUBMISSION SUBHISSION

    -I, VES_(

YES__ (;i f yes rn.,nl.t. DPITTtlLSUliHIS5_IQN

                           ,.. comDl@tl!         EXPECTED SlJBHIS510N DATEJ   OATE}       ;-IN(t x1NQ DATE I          I TEXT             En@rgy Energy Industry Identification System (EllS) codes are identified                 identi fied in U,e        t.xt as [XX]

tho tut ,( PlANT AND SYSTEM SYSTEH IOENTIFICATION: tDENTIFII:ATION: G@neral General Electric - Bo;ling BOiling Water Reactor - 2527 HWt rated core Core thermal power Nuclear Tracking System (NTS) tracking code numbers are identified in the text as (XXX-XXX-XX-XXXXX) IQENTIFI'ATIQN: EVENT IQtNTIFICATION: Inadequate Safe Shutdown Emergency Light Electrolyte Levels Due To HanageMent Management Deficiency A. CQNQITIQNS PRIOg CQNpITIONS PRIOR TO tVENT: EVENT: Unit(s): Unit (s) : Z 2 (3) Eyent Date: Event Octob.r 19, 1990 October E,~~t E, '"' Ti TI ..

                                                                                                                                             ... ::     1000 Hours R,actor Hod.!s):

Rlactor Hod,(s): N (N) Hod. Name(s): Hode N~.(s): R.fu.l (Run) Refuel Pow.r L,v,l(s): Power L,.,l(s): OX (97%) 0% R.actor Coolant S1st.- Reactor Syst.. (RCS) Pressure(s): psi9 0 (997) psig B. QESCRIPTIQN or DESCRIPTION OF EVENT: EyEN!: At 1000 hours hour. on October Octob,r 19, 1990, with Unit 2 shutdown in the R.fuel ~dt tho Refuel ~d. and Unit 3 op.ratfng op.ratlng in In the Run modemod. at 97% 97X rat,d rated core pow.r, the cor. thennal power, tho NRC Resident Inspector notified the tho El.ctrical Electrical Klint.nance Halnt,nance D'partment (tHO) (EHO) Haster Ha.ter Electrician of possible po,sibl. deficiencies deflcienci.s regarding r.garding several

                                                                                                                             .e.er.l s.f.

saf**shutdown hutdown (SSD)

                   ~rgency lights.                 The ~rgency Th. eMlrgency lights in question were among those credit.d in the Dr.sd.n                    Dresd.n SSD Analysis report for meeting the requirements of Appendix R to IQCFRSO.                        10CFRSO. section III.J. Th.* ..-ergency       . .rgency lighting units were noted to have batteries with low water (electrolyte) levels, or to have an                                              In electrolyte density that indicated low charge or end of battery life. The EHD                                  EHO i~d;.t.ly i~diat.ly rlsponded to

( these concerns by conducting the routine maintenance check of SSD Emergency Lights in.accordanc. {n.accordanci with III .15-34

Revision 8 April 1992 DEVIATION INVESTIGATION REPORT TEXT CONTINUATION Fa"" RIY FACILITY NAME OIR NUMBER PA(i~ ( . SEQUENTIAL REVISION STA 1UNIT YEAR NUleER NUll!ER Orescion N",lo'r Pnwor51atjon _Unit 213 1 12 12 13 '9 10 - 1 12 1 3 - oI0 2 OF ...0.. TEXT Identiflcatlon System Energy Industry Identification SysteM (EllS) codes are identified in the text as (XX] [XX] Dresden Electrical Surveillance (DES) 4153-2. Monthly Emergency Lighting lighting Inspection. Special Splcl.l attention was paid to recording "as found" electrolyte levels and other pertinent infol"lNtion infof"1ft&tion about the emergency Mergency ght i n9 un 1 19ht109 Ii it dur; unit ng the perfo""ance during perionnanc! of the survei 11 ance. Attachment surveillance. Attach",ent A identi fi u the "as found" identifies the fourteen worst case Unit 3 sse condition of the'fourteen e~ergency lighting units fr~ SSO emergency fr~ the October 19, 1990 inspection. The units listed are those wi th electrolyte with e1ectrolytt levels below thl lIanufacturer's "add" ,lin thllfanufactur.r's inl,

                                                                                                                                     ** or with unacceptable electrolyte specific gravity.

C. APPARENT CAUSE or EVENT: Due to the complex electro-chemical interactions of a lead-acid battery, its electrical el.ctrical charging system, maintenance, the root cause of the low battery water levels and other battery and battery operation and maintenance. deficiencies defi could ci enci es cou immediately I d not be immedi controlled ate 1y ascertained. A contro 11 ed set of "as found u battery conditi lias found" condition on data was taken immediately after the event was identified. In order to detenftine dete~ine the actual cause of the excessive water consumption found in some of the batteries iitt was concluded that a s.cond second set of "asItas found" found II battery condi t i on data woul condition wouldd be needed. The second set 0off "as lias found" found!! data woul wouldd penllH p,nftit trend; trendingng of water consumption rates under known plant ambient conditions. DES 4153-2 was repeated thirty days after the October 19, 1990 test. The following is a summary of the potential causes considered in evaluating the battlry batt.ry deficienc;.s: d.fici.ncies: (( a. electrolyte. like water, will evaporate under relatively High Ambient Temperature - The battery electrolyte, wanm ambient conditions resulting in low electrolyte levels. warm

b. Battery End of Life - Battery end of life is marked by excessiv~ water ~onsunption. consunption, and low specific gravity. These characteristics will differ among batteries du~ due to minor variations in product construction, construction. the number of discharge/charge cycles over Over the life of thl the battery. the age of the battery, battery. and any exceptional operating demands put on the battery.

c. Excessive Discharging/Charging - Excessive discharging and charging of the battery significantly decreases battery life due to cumulative damage to the ~attery plates platls during .ach lach successive discharge/charge cycle. Excessive dischargin9/chargin~

discharg;ng/charg;n~ -.y may occur through deliberate use. us., a faulty systeM. or an internal or external short resul~i~q Bharging systeM, resul;..int? in continuous si~ltan'ous si_,t.n.ous discharging/charging.

d. Cracked Battery Casing - A cracked battery case could result in leakagl leakage of the electrolyte out of the cas.. cracked battery case could be caused by ;~roper handling or accid,ntal case. A crackld accidental t.,act, i.,act, ov.~heating due to *a dirlct product defect. or intense battery overheating direct short.

I. L.ad Sulphate Excessiv. Lead Sulphlt. Plating - Sulphatl plating is an In lndication indicltion of battlry blttlry Ind of ltfl.lift. This condition will Nill increase water eonsunption, consu.ption, and dlcrease decrease the electrolyte el!ctrolyt. specific gravity. Sulphite plate shedding sh.dding will exacerbate

                                           ,xlc.rbat. battery batt.ry degradation d.gradation if suphate
                                                                                  ,uphat. flakes flakls bridge bridg. the tho b.ttery bltt.ry pl.tes pllt.s Clusing      internal short. Sulphate plating/shedding ~ay causing an int,rnal                                             May occur as Ia result of discharging the         th.

blttlry with low (below battery (b.low the tho top of battery batt.ry plates) electrolyte el.ctrolyt. llvel. llv.l.

f. Improper Maintenance - Failure to monitor and fill the battery electrolyte level lev.l is;s Ia plausable explanation of low electrolyte level.

( III.15-35 III. 15-35

Revision 8B April1 1992 Apri DEVIATION. DEVIA TIDN. INVESTIGATION INVESTlGA TION REPORT TEXT CONTlNU.TIONCONTINUATION t!2"" R'f:' F"ACILITY FACILITY NAME f,jAHE OIR NtJMI!ER PAGE ( SEQUENTIAL REVISION STA UNIT YEAR NlM!ER NUMBER Or Dresden Nucl.ar Power Station Unit 2/3 I 12 12 13 Q In - I I 2 I 3 - oI 0 3 Of C TEXT Energy Industry Identification System tEllS) I EllS) codes are identified ident;fied in the text ~s as (XX] [XX] The October 19, 1990 Inspection Revealed that ~ost most of the Unit 3 SSO SSD ~rgency lighting units had batteries ....with ith electrolyte level below the tf'le "full" "full it line. Electrolyte T!ve1s levels below the "add" line were noted in 13 of the 60 Unit 3 sse 550 emergency light batteries. and one battery was found to have dropped hydrometer discs (indicating low specific gravity) despite having an adequate .lectolyte electolyte level. 1.v.l. Interv;ews Interviews with several of the Electricians who perfOnft perform the DES 4153-Z procedure confiMRed co"fi~d that the ehctrolyte electrolyte level ....would ould not nonnally nor-mally be topped off ;iff the level lev@l was found to be above aboye the "add" lIadd ll line, Hnl, contrary to DES OES 4153-2 which specif; splci(;cally cally states that if H the thl electrolyte level llvel is below the "full"ufull ll 1lin" i ne, then distilled ....water ;s to be added to the "full" line. It should be ater is bl notld, notld. howeyer, howlvlr, that the manufacturer's maintenance instructions peMIIH pennit the electrolyt. electrolyte level levll to float between IIfull u and betweln the "full" "add tl 1lines.

            "add"         i nes.

Two T four-teen emergency lighting units identified

              .... o of the fourteen                                             identHied on October 19, 1990 (Nos.*354    (Nos.-354 and 355) had water     wate,.

levels bela below.... the "add" line and were also located in high radiatTon radiation areas. arIas. Of thesethesl two, anianI of the batte,.;es was completely dry. The batteries batteries batte,.;es for bolh both of these emergency lighting units wire replaced replacld without testing. The rooms in which these lighting units are located are also hot, and poorly lit. It is possible. therefore, therefo,.e, that due to the poor inspection conditions. conditions, and thl hurried hurr;ed c;rc~stances cirCUMstances in which the surveillance is performed, performed. that the ....water ater levels were not being properly Observed. obslrved. This is supported by the fact that identification of the water line through th, thl translucent transluclnt battlry battery casing is difficult even under unde,. ideal conditions. As for the remaining twelve battery lights identified on October 19. 1990 with electrolyte e1ectr01ytl levels below I( the "add" line. or with unacceptable specific grayity gravity readings, readings. these lighting units were wlrl located in areas of the plant which were easily accessible. accessib1@. Two units with lo~ s~wcific s~ccific gravity, Nos. 351 and 332, 332. were considered to be near end of life and were replaced without testing *. Emergency Lighting lighting Unit No. 352 was found to have an extremely 10 low.... electrolyte level. This unit was also considered to be near end of life and was replaced without testing. Finally, finally. nine lights with low electrolyteel1ctro1yte level lev,l successfully passed an eight hour discharge test. Results of the second surveillance, which was conducted thirty days after the October 19, 1990 surveillance, survlillance. showed that only one battery had a discernable disce,.nable wo,ter w;'ter level ltvel chang.. ching** This battery, No. 360. was considered to have been aging faster than the thl adjacent ligl ts, but was still not approaching Ipproaching its nonna 1I end of 1 ife. An ei ght hour di scharge test was perfol'lMc. perfonnec. ',,, light. and it.

                                                                                                           ',i. this Hght.           it was WIS confh'l!Ied confi I'IIIIed to still be acceptable. With the exception of battery No. 360 thE ~1~ctrolyte                     ~1~ctralyte levels lev.ls in th. the batt.ry cases r .... ined stable for a full thi retllained                               thirty rty days. Therefore. the proxlIllut   proxiNte cause  caust of the October Octob.r 19. 1990 "as found" water levels     ltv.1s occurring frOlll fro", 1/8" to 3/4" below the "full"       IIfulll! line lin' is                       inld.qult.

h attributed to inadequat. adherence to DES 4153-2. However, However. the underlying root cause caus. was attributed attributld to ..na,...nt

                                                                                                                                   ".,..."t deficiency d.fici.ncy in that EHD supervision provided insufficient direction to the Electr;c;ans                 El.ctric;ans p.rfon.ing the surveillance.

surv.illanc ** It shouldsh.uld also al ** be b. noted not.d that the EHD had recently r.cently been b** n assigned ass;gn.d this activity. Ictivity. andInd that thlt during durln9 the interi", the COMPleted surveillance dOCUlllentationdocu.entation had h.d not been bt.n routed to the th. 51st 5yst.. Eng;n .. r for review. Engineer It WIS was also concluded that enhanc~,nts enh.nc~ents to OES 4153-2 were needed. n**d.d. D. O. SAFETY ANALYSIS or SAfETY A.NALYSIS Qr EY£HI:. EVEHT: Altlrnative Alt'rnlt;ve or dedicatld dedicat.d shutdown capability is provided prov;dld for each specific fire fi,.e aria, IS requirld r.quirld by tOCFR50 Appendix R, IOCFR50 R. Section III.l. Oetliled Oetailed SSO procedures, which are Ire designed to account for loss of offsite power (lOOPI (LOOP) conditions, conditions. are arl provided for these activities.activitils. Emlrgeney Em.rgency lighting units of eight hour rated capacity are provided in all areas needed for local operation of SSD SSO equiPMent Iqui,..nt and in access ( and egress routes thereto, as required by 10CFRSO operat;ons under LOOP operations IOCfRSO Appendix lOOP conditions. This type of scenario ;s ApPlndix R Section III.J to support perfonaance is extr~ly plrfoMftanCI of SSD unliklly due to extlnsive fire ext,.~ly unlikely fi,.e SSO detection and suppressionSUPprlsslon equipment. III.15-36 III.lS-36

Revision 8 April 1992 DEVIATION INVESTIGATION REPORT TEXT CONTINUATION rgnn R~v ( NAHE FACILITY NAME OIR NUMBER SEQUENTIAL REVISION PAGE STA IUNIT YEAR NUItlER NUHRER Or Ore"de" ~"tl'iLr Eower_ St.Uon lloU 2/3 1 12 12 13 19 10 - 1 I 2 13 - 010 4 lor It TEXT Energy Industry Identification System (ElrS) ( EllS) codes are identified in the text as {XX} [XX) The Unit 3 emergency lighting units that were observed to have insufficient electrolyte ellctrolyt. levels, levels. i.e., i.e .* below thetne "add" line. were subjected to an eight hour discharge tut. tnt. The Th. total of nine ninl batterits batteries that were tested remained lighted at the end of the discharg! discharge test. lest. However, five battery units were replaced without being tested and conservatively can be assumed to have failed. Three of these lighting units were only trained on access acc!ss and egress pathways and were not "needed ** ded to support local operation op.ration of sse SSO equipment. Since the Operators Op~rators are ar~ prOvided provided with flaShlights the safety significance of these failures is gr.atly greatly reduced. Th. The remaining two failures were in the tht H03-1301-2 H03-130l-2 and -3 Isolation Condenser valve rooms. During perfo~ance of the SSD SSO procedures under LOOP conditions,conditions. th. thl only action required in the -2 valve room is to verify that the valve is open; in the -3 valve roOM the valve ~ust b! be throttl!d throttled to control reactor cool down rate. Operation of these Isolation Condenser valves can be coo1down easily accomplished with the illumination provided by a flashlight. In addition to r@-performing re-perfonming the routine maintenance surveillance. battery discharge testing was performed on a random sample of twenty-five percent of all acc!ssible accessible Dresden SSD ~rgency lights. This Th;s testing was perfoMfted perfoMmld in order to demonstrate the reliability of th. the SSD ~.rgency

                                                                                                                      ~ergency lighting system. A total of 43 lights were tested. with only two lights failing to r!mlin                   remain illUMinlted ill~inlted for the full eight hours. As for the two lights that failed, these were Unit Z                      2 ~ergency lights, and are not required to support SSD since Unit 2 is in cold shutdown for a refueling outage.                outag'. Hor'Dver.

Horlover~ th.s. thlSI two lights illuminated only access and egress pathways and not SSD SSO equipment or panels. The lights r~ined r ... ined lit for at least 4 hours, which would hav@ have supported initial access and egress if thlY wire Wlrl nled,d. nlldld. ( Based upon these results EHO EHD has concluded that the operability of the SSD emergency lighting is reasonably assured. Therefore. the safety significance of this event is cons;der.d considered to b, bl ~inimal. miniMal. E. CQRRECTIVE CORRECTIVE ACTIONS: Th! The EHO EMC conducted a special inspection of the Unit 3 SSD emergency energency lights under DES 4153-2. PerforMlnc. Perfonnance of this proc!dure procedure identified 14 emergency lights that had either low specific splcific gravity (dropped (droppld hydrometer discs), discs). or electrolyte level which was below th. the "add" 11nl, lin., or both (se. (sel AttachlMlnt AttachMnt A). Of the51!' fourte!n these fourteen lights. lights, five were deemed inoperable without test or e"aluation and wert wlrl replaced. The r~aining remaining nine lights were proven to b! be operable operablf by the successful c 'npletion*.."l.t;on of eight hour battery discharge tests. Trending was also performed on the thl Unit 3 SSD ~er emer !ncyIncy lights for a period p.riod of approximately one month. This study did not identify any unusual .. ~nt~r.lnce probleMS

                                                                                                    "';.ntE!lrlnCe    probl",s or electrolyte consumption characteristics.

charactlristics. Tho EHD Th. also conducted a special inspection of the Unit 2 sse EHe .lso ~rgoncy lights SSO entrg.ncy light. under undor DES 4153-2. Of the 152 lights that werew.re inspected two lights had low sp.cific gravity, two had indications of a possible internal short. and one ~rgency lighting unit battery had non-unifOnR el.ctrolyt. cons~t;o" non-unifo~ .llctrolytt conSUMption between betw.en colIs. Th. battery cells. Tho batteri.s batterios for all of th.s. ~rgency tho.o ... rgoncy lighting un;ts unit. wer.wero pra.ptly prooptly replac.d. roplacod. Th. Tho Unit 2Z SSD ~r9.ncy eMergency light inspection in.poction also identified identifi.d sev.n

                                                                                    'o.on units with .l.ctrolyt.

oloctrolytl leyels bllow the 11.11. below thl "add" lin ** Of these five wer, lin.. wire replaced without testing. The safety significance of thiS. these d.graded. d.grad.d. or potentially degrlded. ellllr,gency elller.g.ncy lights ",inlnal since Unit 2 h Tights is considered lIIinllllal is in cold shutdown for a r.fueling outage. outag ** Th. two r~ain;"g r~aining Unit Z 2 !mergency emergency lighting units with low electrolyte levels levlls w.re wire included in a group of 28 randomly selected accessible Unit 2 emergenr.y lights that were discharge tested. Thlse lighting units. together with the 15 units that were discharge tested on Unit 3, Make make up a sum total of 43 lights ((' that were discharge tested; that is, is. twenty-five p!rcent percent of all accessible emergency eMergency lights. Two of the

   \.~.         Unit 2 emergency lights failed to remain illuminated for the full eight hours. How.v.r,                 Howlv.r. the th. lights r~ai r~ained    litt for at Ileast ned Ii              east four hours. Th!se  These two emergency Ii     ghti ng unit batter; lighting                     es were batteries   wlrl replaced following the t.st.

test. III.15-37

Revision 8Il April 1992 DEVIATION INVESTIGATION REPORT TEXT CONTINUATION FgI"WI RI::Y FACILITY NAME F'ACrLITY llR NUHBER PA ;[ ( SEQUENTIAL REVISION STA IlNIT YEAR NUHBER NlN!£R Or Orosdon Nuclear Pow.r St.tion Unit 213 I 12 2 I~ ~ 10 - I I 2 I ~ - oI0 5 10F 0 TEXT Energy Industry Identification System ((EllS) EllS) codes areart identified in the text as [XX] Because (1) a significant si9n;fi~ant majority of all ;nspe~tedinspected sse SSD ~rgency lighting units ~.r, ~.r. found to be acceptable. (2) any emergency lighting units that were found to be degraded or Dr potentially degraded were promptly replaced. and (3) the relative success of the eight hour batt.ry battery discharge testing. the Station

                ~oncluded that the operability of the emergency lighting systeM ;s has concluded                                                                         is r.asonably r'lsonably assur.d.

assured. The following enhancements will be added to lhe Th. the SSD emergency e~rgency lighting ftAintenanc.

                                                                                                     ~i"t.n.nce progr.. :

1. A surveillance cover sheet requiring signature of all Mechanics perfOrMing the test. and t.chnical technical review by the EHDEHO Supervisor and the cognizant Technical Techn;cal Staff Engin Engine.r utiliz.d on all

                                                                                                        ** r will b. utiliz,d future SSO SSD e~ergency light maintenance surveillances.

2. DES 4153-2 maintenance surveillance will be revised by the EMD The OES EHD with assistance fr~ the System Syste~

Engineer Engin ** r as follows (237-200-90-12301):

a. wi 11 include acquisition of "as The procedure will lias found" lias leftll found u and "as left" electrolyte leYlls.

lev.ls.

b. The Tho procedure will wi 11 clarify the conditions when water must be added to the tho battery.

batttry.

c. The Tho procedure will wi 11 be split between SSO and balance ba1anco of plant (80P)(BOP) eMergency lights in order c( d.

to exercise better control over the SSD SSO lights. A summary of possible indications of battery degradation will be added. Th. indications to be inspected for may include excessive water consumption. differencls differences in electrolyte l.v.ls levels between battery cells, bubbling, differences in specific gravity ~.tween bubbling. d;fferences ~etw.en battery cellsc.l1s ** tc.

e. DES 4153-2 will continue to be perfonned performed monthly. However, However. EHO with the assistanCt assistanc. of the Technical Stiff T.chnieal Staff will investigate the possibility of extending the surveillanc. surveil lane. inttrval'.

int.rval. This change will be subject to upgraded ~aintenancemaintenance and adequate technical justification.

3. The Technical Staff will investigate the possibility of relocating !Mergency ~rgency light battery cabinets out of high radiation areas (237-200-90-12302).

4. An "as found u ei uas found" ght hour battery dischArge eight discharge tlSt test wi 11 b,

b. dev,lopd developd b If the EJoI)EItJ with wi th the .ssistance assist,nct of Engin.er for the Dresden SSD the Syst~ Engineer ~rgency lights. Testing for each SSD SSO eoorgoncy SSO batttr, battery light will then be p.rfo~d perfonllt!d by the EI'I)

E,.., once pel" operating cycle (237-200-90-12303). p.r op.rating F. PREVIQUS EVENTS: There were Thtre w.r. no pr.vious previous .vents of this type. G. COMpoNENT FAILURE COHPONENT FAILUBE DATA: Manyfactyrer Manyfacturer NQ!!encl,tyre NQ!!en cl a ture NU!!ber Hodel NU!I!ber Mfa. 'art Mfg. Part Nvmm.r NU!!'Iber Teledyne Inc. Emergency Lights S6L100-aO S6L100-80 *S6l100-80

                                                                                                                               *S6L100-80 This equipment ;s         NPRDS reportable. therefore.

is not NPROS therefore, an industry wide NPRDS data base search was not performed. perfo""ed. NOTE: The asterisk in the mfg. part numb~r number indicates the number of lamps"that th~ the lighting unit comes with. i.e. Le. 0,1.2.3. III.lS-38

Revision 8

         -                                                    DEVIATION'INVESTIGATION April 1992 DEVIATION "INVESTIGATION REPORT TEXT CONTINUATION r2""  Rt: v c(                     FACI LI TY NAME rACllITY   NAHE STA liN IT yEAR orR NIIMBER SEQUENTIAL NUMBER REVISION NlJIIIER PAGE-Or.sden Or      Nuclear Power Station Unit    nit 213I                1 12 12 13 19 10 - 1 I 2 I 3 -             oI0     ~  lor 0 TEXT        Energy Industry Identification System (EllS)   (EI IS) codes codls are identified in the text AS [XX]

ATTACHHENT A ATTACHMENT UNIT 3 SAFE SHUTDOWN EMERGENCY EHERGENCY LIGHT INSPECTION RESULTS RESULTS. OCTOBER 19,

19. 1990 BATTERY II

                                                 #   AS FQUNP FOUNP     AS LEtT LEFT       ILLUMINATES ILLUHINATES       COtttENTS COtt1ENTS 354        -1.5"
                                                     -l.S"            REPLACED         EQUIPMENT        NOTE 1 (BELOW) 11T~inat.s V.lv.

Ill~inAt.s VaTv. H03-1301-3 H03-130T-3 355 EMPTY EHPTY REPLACED EQUIPMENT EQUIPHENT HIGH RADIATION AREA ITT~in.t.s Valv. Ill~inAt'$ VaTv. H03-130t-2 H03-130T-2 351 +.5" REPLACED PATH HYDROMETER HYDROHETER OISCS DISCS DROPPEO DROPPED 360 -.75" tI LLED rILLED PANEL PASSED 8B HOUR DISCHARGE TEST 358 -.5" tILLED FILLED RACK PASSED 8 HOUR DISCHARGE TEST Ie{ .'"" 331 -.25" FILLED PANEL PASSED 8 HOUR DISCHARGE

               -                                                                                        TEST 353        -.-.5"5"          FILLED rILLED           PANEL PANEl           PASSED 8 HOUR DISCHARGE TEST 337        -.2S"
                                                     -.25"             fILLED FILLED           EQUIPMENT EQUIPHENT       PASSED 8 HOUR DISCHARGE TEST 342        -.5"
                                                     -.5 11            FILLED           PANEL           PASSED 8 HOUR OJ 0; lCHARGE TEST 330        -.25"
                                                     -.2S"             rIlLED FILLED           PATH            PASSED 8 HOUR DISCHARGE TEST 349         _T"
                                                     -1"               rILlED FILLED           PATH            PASSED 8 HOUR DISCHARGE TEST 3SZ 352        -3"               REPLACED          PATH            REPLACED DUE 0\lE TO EXCESSIVELY LOW ELECTROLYTE LEVEL 332         _1"
                                                     -1"                REPLACED         PATH            HYDROMETER DISCS DROPPED
 ;;C

_.5" PATH PASSED 8 HOUR DISCHARGE C~_ 335 -.5 11 FILLED rILLED TEST NOTE 1: THIS EMERGENCY LIGHT IS LOCATED IN A HIGH RADIATION AREA. THE SA TTERY WAS REPLACED RATHER THAN BATTERY TESTED IN PLACE DUE TO ALARA CONSIDERATIONS. II 1.15-39

Revision 8 April 1992 DEVIATION INVESTIGATION REPORT TEXT CONTINUATION Fg[!! !Iv : FACILITY NAHE NAME orR Nt","'. PAGE ( SEQUENTIAL REVISION STA UNIT YEAR tM4!ER NlHI£R Orfsd Drug.nn NNuel.ar 1 r P Po"'or.Station

                              ~                Unit 213            1 12 2 13 19 10 -     1 I 2 I 3 -    oI 0   7  IOF !l TEXT      Energy                                    (ErrS) codes are identtfi.d Ene~9Y Industry Identification SysteM (EllS)                idlntified in the text IS (XX)

(XX] c::; t:;?  ? . ,r'

                                                                "-T r            ....,r-
                                                                                 'Ir'
                          .75" L
                              .l                             FULL LINE ADD LINE ADO
                               ,.. ~

TYPICAL MODEL 3100S 31005 SIX VOLT BATTERY

 ;(.
 '(

III. 15-40 III.1S-40

Commonwealth Edison 1~no 1~no Opus Opus Place Place Revision 88 Downers Downers Grove. IllinoIS 60515 Grove. IllinOIS 60515 Apri April1 -1992 1992 (- ( I.

7. 1991 January 7, Mr. A. Bert Davis Regional Administrator - Rill U.S. Nuclear Regulatory Commission 799 Roosevelt Road Ellyn. II 60137 Glen Ellyn,

Subject:

Dresden Nuclear Power Station Units 2 and 3 Response To Notice of Violation Contained in Inspection Report 50-237/90023; 50-249/90023 NRC Docket Nos. 50-237 and 50-249

Reference:

W. D. Shafer letter to Cordell Reed dated November 28, 28. 1990 transmitting NRC Inspection Report 50-237/90023; 50-249/90023

Dear Mr. Davis:

This letter provides the Commonwealth Edison Company (CECo) response (attached) to the subject three Level IV violations transmitted by the referenced NRC Inspection Report for Dresden Station. The three violations identified were: 1) 1( inadequate training to assure satisfactory knowledge of plant administrative requirements, 2) the failure to follow procedures and instructions and 3) inadequate requirements,2) corrective actions with regard to fuel bundle mispositioning events. ceCo CECo has reviewed the Notice of Violations and in all but one example agrees that the violations occurred as described. The corrective actions detaUed detailed in the response will bring the Station into compliance and will prevent similar violations from occurring. CECo recognized a negative trend in human periormance, performance, personnel-related events and views this as a serious matter. Actions were taken to address each event, but we also have taken additional timely, broad, comprehensive actions to address the identified negative trend. These actions are beyond those in the detailed responses to the Notice of Violation and are listed in Attachment A. If your staff has any questions or comments concerning this letter, please refer them to Rita Radtke, Compliance Engineer at 708/515-7284.

                                             ,~
                                               ~

{c-i( cc: B.L. Siegel, Project Manager, NRR D. E. Hills. Senior Resident Inspector NRR Document Control Desk Attachment

      /scl:ID707:1                                      III .15-41

Revision 8B April 1992 ATTACHMENT A ( Additional actions taken to address human performance. performance, personnel-related events:

• A meeting was held with all supervisors (CECo & Contractor) on October 17, 1990 to raise overall onsite awareness. Each Supervisor was required to take notes during the meeting, meet with those employees he supervises and return his notes with names of those he met with by October 18, 1990.

A station "Self Check" initiative, called VerAntSO, was introduced the week of October 22, 1990. VerAntSO is an acronym used to remind everyone of the 22,1990. self-check concept and stands for Verify, Anticipate, Stop and Qbserve.

• An IN PO assist visit to review Human Performance activities was requested INPO and performed on October 25 and 26, 1990.
• The C ECo Performance Assessment Department performed an overview of CECo outage concerns during the week of October 22, 1990.
• A licensed individual from Nuclear Quality Programs Department from another CECo station performed an overall review of in-plant activities during the week of October 22, 1990.

An in-plant walkdown and independent verification of one hundred (100) ( out-of-services was performed to verify proper isolation of equipment for maintenance activities. tC .

• IILl5-42 IIL15-42

      /scl:ID707:2

Revision Revi si on 8 RESPONSE TO IO April 1992 (, NOTICE NOIICE OF VIOLATION VIOLAIION VIOLATION VIOLAIION 1 10 CFR 50, Appendix B, Criterion II, as implemented by Commonwealth Edison's "Quality Assurance Program" Program requires indoctrination and training ll of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained. Contrary to the above, indoctrination and training of personnel performing activities affecting quality waS was inadequate in assuring proficiency was achieved and maintained as to administrative requirements as indicated in the following examples:

a. Lack of operations personnel knowledge of Dresden Administrative Procedure (DAP) 7-5, "Operating Logs and Records," Revision a 8,f and Dresden Operating Abnormal (DOA) Procedure 902-5 G-2, Revision 3, requirements for maintaining the Control Rod Drive Accumulator High Water/Low Pressure Alarm Log (AHWLPAL) resulted in the AHWLPALs for both units not being maintained between April 1990 and August 3, 1990. As such the licensee's program to identify repeat failures of accumulator alarms was not effective during that time (DRP>>.

period (50-237/90023-0la (DRP>>).

b. Lack of technical staff personnel knowledge regarding recogn1z1ng recognlzlng and

( processing conditions adverse to quality resulted in a failure to properly identify a procedural nonadherence involving maintenance of the AWHLPAL when discovered in May 1990. Because of this, corrective actions to prevent recurrence were not taken at that time.

(50-237/90023-0lb (DRP>>. (50-237/90023-01b Ibis is a Severity Level IV violation (Supplement I). This DISCUSSION On December 8, 1989, DAP 7-5, Revision R~vision 8 was approved. Ibis procedure This established an Accumulator High Water/Low Pressure Alarm Log (ABWLPAL). (AHWLPAL). On the same day, a revision to the Unit 2(3) Operators Daily Surveillance

    . Log (Appendix A) was approved which removed the weekly Control Rod Drive Accumulator Log from Appendix A so that there would not be two procedures recording the same info~ation. Ibe   The intent of the AHWLPAL was to recor~ of Control Rod Drive Accumulator Alarms in maintain an ongoing recora order to identify problem accumulators. Either the AHWLPAL or the Control Rod Drive Accumulator Log (removed from Appendix A) would have been adequate for documenting Control Rod Drive Accumulator High Water/Low Pressure Accumulator Alarms. An AHWLPAL Book was established at each Unit Operator's Opera*tor's Desk.

ZNLD:708:1 ZNLD: 708: I III. 15-43

Revision 8 Apri 1 1992 April The on-shift licensed personnel did not receive formal instruction or notification as to the maintenance of the AHWLPAL during the day-to-day operation of the plant. Therefore, the implementation of the ABWLFAL ABWLI'AL was only partly accomplished by revising the procedures. The on-shift licensed personnel, responsible for maintaining the AHWPAL, did not receive training on this program. The Technical Staff System Engineer learned in early May 1990 that ~he the AHWLPAL was not being properly maintained as required by DAP 7-5 and realized that either the operators would have to be trained on the use of the log or that the program would have to be revised. During this same time period, the Operations Department was implementing a program to independently verify accumulator valving operations. This requirement was to be implemented by having the independent verifier sign the AHWLPAL AHWLI'AL log book. To resolve these two unrelated concerns in a coordinated manner, it was decided to place the AHWLPAL AHWLI'AL back into Appendix A. This revision to DAP 7-5 took longer than expected because of other changes to the procedure that were not related to the AHWLPAL. It took three months to resolve all of the issues with DAP 7-5. On August 30, 1990, the requirements for the AHWLPAL were transferred from DAP 7-5 back to the Unit Operator's Daily Surveillance Log, Appendix A, at which time CRD accumulator logging was resumed. Action to correct the problem was taken, but it was not documented through the use of DAP 9-12, "Procedural Adherence Deficiencies." The ( .. DAP 9-12 should have been used as the System Engineer was not aware that OAP mechanism to document this need for corrective action. CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED The logging of accumulator alarms was transfered back to Appendix A on August 30, 1990. Since that date all accumulator alarms have been logged. CORRECTIVE ACTIONS TO PREVENT FURTHER FURIHER NONCQMPLIANCES

1. The Operations Department, with assistance from the Training Department and the Technical Staff, will develop a method to ensure that when procedural changes are made that alter the day-to-day routine of licensed operators, a review is made to determine what training should be completed to properly implement the change. Ibis This methodology will be in-place by March 31, 1991.

2. DAP 9-12 will be tai1kated to all station personnel by February 7, 1991. Emphasis will be directed to its purpose and use.

(

(-

II 1.15-44 III.15-44 ZNLD: 708: 2 ZNLD~708:2

Revision 8 April 1992

3. A number of events, occurring during the past six months, have

( indicated that the knowledge level of station personnel with reg,rd reg~rd to the contents of various Dresden Administrative Procedures is less leas than desired. To raise the station personnel knowledge level of the contents of the Dresden Administration Procedures, the Station's ongoing training program will be reviewed to verify that all personnel involved in activities addressed in each administrative program/procedure are appropriately trained. A matrix of Dresden Administrative training requirements will be produced by January 31, 1991 and appropriate changes will be made to the ongoing programs by June 30, 1991. PATE WHEN FULL COMPLIANCE WILL BE ACHIEVED DATE Full compliance was achieved on August 30.30, 1990 at which date all accumulator alarms were properly logged. VIOLATION 2 by 'Commonwealth Edison 10 CFR 50, Appendix B, Criterion V, as implemented bY'Commonwealth Company's Quality Assurance Program, requires that activities affecting quality be accomplished in accordance with documented instructions, procedures or drawings. Contrary to the above, activities affecting quality were not accomplished in accordance with documented instructions', procedures, or drawing. drawings in the following examples: This in a Severity Level IV violation (Supplement I). EXAMPLE a Dresden Operating Procedure (DOP) 1900-3, "Reactor Cavity-Dryer Separator Storage Pit Fill and Operation of the Fuel Pool Cooling and Cleanup cODIDunication System During Refueling," Revision 8, requires constant communication between the refueling floor and the control room while filling the cODIDunication between the refueling floor and reactor vessel. Constant communication the control room was not maintained while filling the Unit 2 reactor vessel on October 14, 1990, resulting in the overfilling of the vellel vessel into the ventilation ducts and contamination of various arealareaS of the third and fourth floors of the reactor building. (50-237/90023-2& (50-237/90023-28 (DRP>> DISCUSSION On October 13, 1990 the reactor cavity flooding evolution began. Prior to beginning the evolution, a Fuel Handling Supervisor had agreed to monitor reactor cavity level from the refueling floor. As cavity flooding progressed, the Fuel Handling Supervisor reported on the cavity level. At 0330 hours on October 14, 1990, the Fuel Handling Supervisor informed Operations that he was leaving and that the level was approximately 1 lIz 1/2 feet below the bottom of the ventilation openings. From this point on, the reactor cavity water level was no longer being continuously observed. Later, the Unit 2 NSO dispatched the Equipment ( Attendant (EA) to visually observe the cavity level. The EA erroneously reported that the level was 16 inches below the bottom of the ventilation openings. ZNLD:708:3 III .15-45 III.1S-4S

Revision 8 April 1992 ( A short time later, the control room received annunciator, "Fuel Pool High Level." The NSO immediately closed the Feedwater Low Flow Valve _AI!d _lU.!d dispatched another EA to reject water from the fuel pool cooling system to the Condensate Storage Tank. A Shift Supervisor then proceeded to the fourth floor of the Reactor Building where he saw water coming from tbe the ventilation ducts. DOP 1900-3 contains a precaution: '~aintain constant communications between the refueling floor and the Control Room while filling the reactor head cavity and the dryer/separator pit to prevent overflow into the ventilation dueting." ducting." This was not followed from the time the Fuel Handling Supervisor left the refueling floor until the alarm was received in the control room. CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

1. This event was included in a series of meetings conducted on October 17, 1990 by the Station Manager and attended by all line supervisors of CECo and Contractor organizations on site. The purpose of the meetings was to increase everyone's awareness of several recent events involving personnel error and to require subsequent meetings between these supervisors and all employees to raise the overall on-site awareness level of the need for increased attention to detail.

2. A posted operator aid was revised showing the level of the bottom of the ventilation openings to be 469 inches instead of the previously

( erroneous value of 476 inches.

3. Operations personnel involved in this event were counselled on the importance of procedure adherence.

4. Procedural adherence was addressed in a Shift Engineer's meeting held on October 24, 1990. Shift Engineers were instructed that procedures must be consulted and adhered to for all complex, unique, or infrequent evolutions.

ACTIQNS TO PREVENT FURTHER NONCQHPLIANCES CORRECTIVE ACTIONS NQNCOMPLIANCES

1. DOP 1900-3 will be revised to clarify when continuous visual monitoring is required when flooding the refuel cavity. The requirement for visual monitoring will be specified so that continuous monitorinJ will not be required for slow moving evolutions when level is more tftan three feet from the final desired level. The procedure will be revised by March 31, 1991.

2. The Instrument Maintenance Department will revise DIP 0260-01, Figure 1 to correctly depict the bottom of the cavity ventilation openings at 469 inche~

inches by March 31, 1991.

3. To aid visual estimates of water level, the Technical Staff will evaluate methods of providing a graduated scale in the Reactor cavity and in the Dryer/Separator pit for Units 2 and 3. An acceptable method will then be implemented by March 31, 1991.

( ZNLD:708:4 ZNLD: 708:4 III .15-46 III.lS-46

Revision 8 April 1992 c-( 4. This event was reviewed by licensed operators during Cycle 8 of _ continuing training. This was completed on Dec~mber _continuing Dec_ember 7, 1990. It will be covered for non-licensed operators by March 1, I, 1991.

5. The EA involved in this event developed an article for the Station's monthly newsletter, discussing the importance of attention to detail, procedural adherence, and the concept of self-checking one's actions.

DATE WHEN FULL CQMPLIANCE COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on October 14, 1990 when the reactor cavity flooding evolution was completed. EXAMPLE b Specific practices required by OAPDAP 3-5, "Out of Service and Personnel Protection Cards," Cards,1t Revision 22, were not followed as to preparation, review; review,- approval, documentation and independent verification in the removal and return to service of the Unit 2 diesel fuel oil day tank drain valve on October 20, 1990. This resulted in the inadvertent draining of the day tank when the drain valve was placed in the incorrect position. (50-237/90023-02b (DRP>> DISCUSSION In preparation for the cleaning of the Unit 2 Diesel Generator Main Fuel i( IC. Oil Storage Tank, the Diesel Fuel Oil Transfer Pump Suction valve was shut and the Fuel Oil Day Tank Drain valve was checked to be shut by'aby-a member of the Operations Staff on October 8 or 9, 1990. "Do Not Operate" tags supplied by the storage tank cleaning contractor were placed on the valves. No CECo Out-of-Service was written for the tank cleaning. On October 20, 1990, between 10:30 and 11:00 am, the same member of the Operations Staff opened the Transfer Pump Suction Valve. In addition, he opened the Day Tank Drain Valve, even though he had checked that this valve was shut approximately 12 days prior. These valve manipulations were performed without the knowledge of on-shift Operations or procedural guidance. At approximately 11:20 am on Saturday, October 20, 1990, the "Unit 2 Hi/Lo Level" alarm was received in the Control Diesel Generator Day Tank Hi/La Room. At approximately the same time, two members of the Technical Staff were in the vicinity of the diesel generator room and observed a strong odor of fuel oil. Upon entering the room, they noticed fuel oil on the floor. CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

1. A member of the Technical Staff traced the source of the fuel spill to the drain on the Unit 2 Fuel Oil Day Tank, found drain valve 2-5212-500 approximately one to two turns open and closed the valve.

Approximately 600 gallons of fuel oil were spilled. ( IIL15-47 III.l5-47 ZNLD: 708: 5 ZNLD:708:5

Revision 8 Apri1 April 1992 ( 2. The involved Operations Supervisor was counseled on the need to iuteract ith Operations Department shift personnel to ensure that interact ....with all valves necessary to adequately isolate a component are included on the appropriate Out-Of-Service. The involved individual was reminded that unauthorized valve manipulation is against plant policy and could lead to personnel injury or equipment damage. QQ.RRECTIVE

   ~RRECTIYE   ACTIONS TO PREVENT FURTHER NONCOMPLIANCE

1. The Day Tank drain valves on all three emergency diesels were locked shut. The Locked Valve Checklist and System Checklist will be revised to include the Day Tank Drain on all emergency diesels in the locked closed position by March 31, 1991.

2. The details of this event were reviewed with all station personnel at the October 18, 1990 tailgate meetings, emphasizing the need to Out-Of-Service program and the hazards of properly use the Out-af-Service unauthorized equipment manipulations.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on October 20, 1990 when the drain valve was closed. EXAMPLE c i(, DAP 7-14, "Control and Criteria For Locked Equipment and Valves," the flowpath of systems required Revision 2, requires manual valves in theflowpath for plant shutdown during post-accident situations or which provide a controlled path to the environs, including primary and secondary containment isolation valves to be locked. Prior to November 1990, manual valves including the Units 2, 3 and 2/3 diesel generator service water three-way valves and the Units 2 and 3 drywell manifold sampling system containment isolation valves were not locked or designated to be (50-237/90023-02c (DRP>>. locked (SO-237/90023-02c DISCUSSION The Diesel Generator Service Water flow reversal valves were erroneously excluded from the locked valve checklist. This was due to the valves having a mechanical locking device which prevents the valves from repositioning. The Station believed that the locking device (which does not have a keyed lock) fulfilled the administrative requirements of the locked valve program. CECo does not believe that a requirement existed to "lock closed" the 2(3)-8507-500 through 521 Drywell Air Sample Valves. The USNRC had, on two separate occasions, the opportunity to review and assess the acceptability of those valves. On each occasion they found the design to be acceptable: 'e II 1.15-48 ZNLD: 708:6

Revision 8 April 1992 ( 1. In response to NUREG 0578, in a letter dated February 25, 1980, CECo supplied information on primary containment isolation valves. An excerpt from that letter reads as follows, "All non-essential systems that provide a possible open path out of the primary containment were found to be either isolated by isolation signals, by check valves that would prevent flow out of the containment, by manual valves that are normally closed during reactor operation, or as in the case of instrument lines by closed piping systems." Included were the 2(3)-8507-500 through 521 valves with their classification (non-essential) and a sketch showing their configuration. The USNRC, in a letter dated March 5, 1980, responded, '~e conclude that the licensee has completed a re-determination of which containment isolation penetrations are essential or non-essential. All non-essential lines are either automatically isolated by diverse signals or technical justification has been provided. Modifications have been made to prevent inadvertent re-opening of isolation valves. Based on the above, we find that the licensee has satisfied the requirements of this item."

2. Various correspondence exists documenting the scope and depth of the USNRC review of SEP Topic VI-4, VI-4 t "Containment Isolation Systems." An NRC letter dated December 18, 1981, transmitted a draft SER on the topic and requested that CECo provide comments and additional information. CECo's response of May 21, 1982 provided information on the 2(3)-8507-500 through 521 valves. Based upon that response, the NRC issued their final SER on September 24, 1982. Table I of that

( SER provides a list of valves which they reviewed; the 2(3)-8507-500 through 521 valves are included on that list. A section of the SER titled, "Administrative Control," identifies valves which have inadequate administrative controls and which should be listed as "locked

        !tlacked closed" instead of "normally closed." These valves are listed in Table II; the 2(3)-8507-500 through 521 valves are not included. Another section titled, Manual Isolation Valves, Valves,"II lists other valves which should be in a "locked closed" position; once again the 2(3)-8507-500 through 521 valves are not included.

Although the basis of acceptability for these valves being "normally open" or "normally closed" is not provided in the SERs, it presumably is system. In any due to the Drywell Air Sample system being a closed loop sy.tem. post-TMI and SEPt case, the NRC reviewed these valves as part of po.t-TMI SEP, and did not require them to be "locked closed." and The Inspection Report and Notice of Violation reference a November 18, 1982 cOlllllitment commitment to "review all containment penetrations in the plant and not limit the scope to Table II in the SER." nAPDAP 7-14, "Control and Criteria for Locked Equipment and Valves," is also referenced. One of the DAP 7-14 criteria for locked valves is, Manual valves which provide a controlled path to the Environs, including Primary and Secondary Containment isolation valves." Since the Drywell Air Sample system is a closed loop system, leakage past these valves would not provide an uncontrolled path to the environs. CECo does not believe that operation of a system consistent with the plant's original design basis and in (,-- accordance with NRC SERs and Station administrative programs constitutes a violation of NRC requirements. rrr II r .15-49 ZNLD:708:7

Revision 8 April 1992 ( During a recent investigation into the use of a temporary sample pump to obtain drywell air samples (in which the Drywell Air Sample system's : _ closed loop was broken), questions arose relative to the ability of the original Drywell Air Sample system to withstand seismic and accident conditions. In view of this recent information, CECo is re-assessing the acceptability the system. Past performance of the system has shown only limited usefulness in its ability to locate sources of leakage into the drywell. With the recent approval of the Station's Generic Letter 88-01 submittal (in which no credit was taken for the Drywell Air Sample system) it is believed that the system may be removed. Final resolution of the Drywell Air Sample system is expected in February, 1991. As an interim measure, valves 2(3)-8507-500 through 521 have been taken out-of-service closed. These valves will remain controlled by the out-of-service or locked closed in accordance with DAP 7-14 as long as the system remains in place. Since no method of locking these valves presently exists, work requests have been written to provide a means of locking them. The valves will be added to the locked valve checklist and locked as appropriate prior to clearing the out-of-service on these valves. CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED CQRRECTIVE Temporary Procedure Change 90-408 was written against Dresden Operating Procedure (DOP) 040-M3, "Locked Valve List: Accessible During Operations," adding the Diesel Generator Service Water flow reversal valves to the Locked Valve List. The valves were also locked at that time time.* . ( NONCOMPLIANCES CORRECTIVE ACTIONS TO PREVENT FURTHER NQNCQMPLIANCES

1. DOP 040-M3 will be revised by March 31, 1991 to include the Diesel Generator Service Water three-way valves.

2. A review of other valves with mechanical locking devices will be b~ing inappropriately excluded conducted to assure that they are not being from the locked valve program by March 31, 1991.

3. DAP 7-14 criteria will be reviewed and revised as necessary to assure the locked valve criteria are easily understood by station personnel. This will be accomplished by March 31, 1991.

PATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achi~ed on November 27, 1990 when the valves were locked into position. EXAMPLE dd PAP DAP 15-6, "Preparation and Control of Work Requests," Revision 0, requires work to be performed per repair manual(s), travelers/procedures, or work instructions provided in the work package. On October 15, 1990, work prescribed for disassembly of the Outboard Containment Isolation Feedwater Check Valve 220-62B was performed instead on Outboard

 !/

Containment Isolation Feedwater Check Valve 220-62A (SO-237/90023-o2d (50-237/90023-02d

 ~
 \.... (DRP>>.*

(DRP>> ZNLD: 708:8

                                         !ILlS-50 IIUS-SO

Revision 8 April 1992 ( DISCUSSION As a result of local leak rate tests (LLRTs) perfo~ed performed on September 24 30, 1990 on the 2-220-62A and 62B check valves, the decision was made and 30. to repair the valves. A maintenance pre-job briefing and ALARA pre-job briefing were performed on October 12, 1990 prior to proceeding to the work area for disassembly of the 62B check valve. The Maintenance Supervisor accompanied his crew to the work lqcation and directed them to valve,9 but was actually the "begin work on what he believed to be the 62B valve 62A valve. Upon removal of the valve bonnet and seal ring, the valve body was found full of water and the valve disc stuck in the open position. All work was immediately stopped, the Maintenance Supervisor and Technical Staff were notified of the as found condition. Water was pumped from the valve body and work continued on valve decontamination, inspection and repair. 17, 1990 a Radiation Protection Technician (RPT) surveyed the On October 17. valve plug and seat. The RPT was concerned that work was being performed on the wrong valve and questioned the crew several times whether they were working on the correct valve. The crew indicated that they were sure they were on the correct valve and that it had been verified with their supervisor. On October 19, 1990 the Station ALARA Coordinator questioned the reported radioactive contamination levels inside the opened valve and investigated .(( the possibility of work being performed on the wrong valve. This was confirmed to be the case. - CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

1. An Out-of-Service was hung for the valve; this included closing the downstream manual valve (2-220-57A) to provide better isolation.

Appropriate radiation protection measures were taken, including completing the proper Radiation Work Permit procedures. As the 2-220-62A valve was also scheduled for overhaul due to LLRT results, work was allowed to continue under the proper work package.

2. A tailgate discussing this event was presented on December 20, 1990, emphasizing the need for self-verification and that each nuclear worker has the responsibility to assure the equipment to be worked on is the equipment ide~tified in the work package and that actions to be taken are correct.'

CORRECTIVE ACTIONS TO PREVENT FURTHER NQNCQMPLtANCES NQNCQMPLIANCES

1. A more positive identification of the 2-220-62A, 62B, 62!, and 59 valves for the Unit 2 steam pipe tunnel has been provided. Additional valve identification has been applied to the support structure located over the 62A and 62B valves. .

2. A tailgate article will be developed by January 31, 1991 to inform

.( ( plant personnel that identification tags are expected to be attached on all plant components. If equipment tags are not found, the labeling coordinator should be notified to assure components are properly tagged and operating personnel should be contacted to assist in proper component identification before starting work. ZNLD:708:9 111.15-51

Revision 8 April 1992 ( 3. This event will be incorporated into continuing training for Maintenance, Operations, Radiation Protection, and Technical Staff personnel by December 31, 1991. Emphasis will be placed on the potential significance of opening the wrong primary system boundary, opportunities by the working group and others which were available to identify that the wrong equipment was being worked on, and methods by which the correct component could have been identified (outage walkdown, pipe/penetration labeling, RWP survey maps).

4. Radiation Protection Survey Maps will be upgraded as necessary to provide for identification of equipment specified on the maps. This will be completed prior to* the next scheduled refuel outages for each unit.

PATE DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on October 19, 1990 when the 62A valve was properly removed from service. EXAMPLE e DAP 15-6, "Preparation and Control of Work Requests," Revision 0, requirements were violated on August 8, 1990, when prescribed for calibration of Unit 3 Torus to Reactor Building Vacuum Breaker A Pressure Transmitter DTP-1622A was performed instead on Pressure Transmitter DPT-1622B. This resulted in advertent opening of the Unit 3 Reactor Building Vacuum Breaker B. (50-237/90023-02e (DRP)(DRP>> DISCUSSION On August 8, 1990 the Instrument Maintenance Department (IMO) was performing a calibration check on DPT 1622A [Torus to Reactor Building 3-1601-20A) Pressure Transmitter] Vacuum Breaker (AOV 3-160l-20A) Transmitter) using procedure DIS 1600-20, "Torus to Reactor Building Differential Pressure Transmitter l622A and B B Calibration and Maintenance Inspection." The Instrument Mechanic (IM) (1M) valved-out the DPTl622A DPT1622A transmitter and connected the calibration instruments to obtain a set of a8 as found readings. The as found readings were outside the ideal calibration tolerance range on the conservative side. IM to adjust the,calibration setting for DPT 1622A For the 1M l622A or 1622B he has to get down, turn 18~*, l8a', arch his head and back under the transmitter. While being upside down, the 1M proceeded to adjust what he 3-l601-20A to bring the calibration thought to be DPT 1622A for AOV 3-1601-20A within specified instrument tolerance range. While he was making the 3-160l-20B opened with a resultant Control Room adjustment, the AOV 3-l601-20B Annunciator. The 1M had adjusted DPT 16228l622B instead of DP! DPT 1622A. l622A. ZNLD:708:10 ZNLD: 708: 10 II 1.15-52 III.15-52

Revision 8

• April 1992 c- CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

1. The Master Instrument Mechanic counselled the 1M involved in this event. Emphasis was placed on total job awareness needing to be maintained at all times a job is in progress. Also emphasis emphssis was wss placed on heightened awareness while working in congested areal areas of the plant since the possibility of poor job performance is enhanced.

2. The Master Instrument Mechanic discussed this event at a department tailgate meeting. The discussion included a review of the situations on this job and a reminder of what is expected of Instrument Maintenance Department personnel when working in congested areas.

CORRECTIVE ACTIONS TO PREVENT FURTHER NONCQMPLIANCES

1. The labeling of the ~p (OPT l622A and liP transmitters on both units (DPT 1622B) will be improved by placing a label above the transmitter and l622B) removing any labels below the transmitters.

2. The 6P (OPT 1622A and l622B) will be rotated 180 degrees liP transmitters (DPT to relocate the adjustment screws on the top of the transmitter (Work Requests D95106, D95l07, D95108~

095106, D95107, 095108, and D95109) I, 1991 (during D95l09) by June 1, next refueling outage) for Unit 3 and by October 1, I, 1991 for Unit 2. This will greatly enhance access to the adjusting screws and minimize the possibility of adjusting the wrong transmitters. DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on August 8, 1990 when the Vacuum Breaker was reclosed. VIOLATION 3 YIOLATION 10 CFR 50, Appendix B, Criterion XVI, as implemented by Commonwealth Edison's "Quality Assurance Program," Program." requires that conditions adverse to quality be promptly identified and corrected and, in the case of significant conditions, the measures assure the cause is determined and corrective action taken to prevent repetition. Contrary to the above, following the fuel bundle mispositioning eventsevent. of January 10 and 12,12. 1989, corrective actions actiona were insufficient to prevent repetition in that similar events occurred on October 1, I, 1990 and October (50-237/90023-08 (DRP>> 2, 1990. (50-237/90023-0s (Supplement I). This in a Severity Level IV violation (Supplement. ( IILlS-S3 IILlS-53 ZNLD: 708 708:: 11

Revision 8 April 1992 c ( DISCUSSION On October 1, 1990, Unit 2 was in the Refuel Mode. Fuel Handlers were unloading fuel from the reactor * .The grapple's Core Position Indication System was improperly indicating position in the east-west direction. The current fuel move was the last fuel move from the perimeter of the core. The next fuel move was to be from the interior of the core where no fuel assemblies had yet been removed. The Fuel Handling Supervisor went onto OlltO the Refueling Grapple to caution the fuel handling crew that the next transfer ~as was from a different region of the core. After the current step, the duties of the Independent Verifier and the Grapple Operator were scheduled to be .exchanged between the two men. The Grapple Operator grappled the wrong fuel assembly. As the Independent Verifier had been cautioned about the next fuel move, he was studying a core map to determine the location of the next step's fuel transfer rather than independently verifying what the Grapple Operator was doing on the current step. The fuel assembly was erroneously transferred. transferred, the Grapple Operator and Independent Verifier exchanged duties, and the "new" Grapple Operator began to perform the next step. While examining the core, the fuel handling crew discovered that the previous fuel move had been performed in error. At this time fuel moves were suspended while discussions between the Operating Engineer, Shift Engineer and Fuel Handlers took place. Prior to resumption of unloading the core, it was decided that verbal concurrence would be required from the verifier that the proper step was ('. being initiated, prior to removing a fuel bundle from the core. A further review of the event was *conducted the next morning. was'conducted On October 2, 1990, an Electrical Maintenance Supervisor (EMS) was on the Fuel Grapple to observe the operation of the Core Position Indication System in ill preparation for repairs scheduled for later in the day. These repairs were to be completed in response to a corrective action from the first unloading error. The Independent Verifier was waa discussing it. its operation with the EMS. The Grapple Operator positioned the grapple over the wrong fuel assembly. The Independent Verifier (while engaged in a conversation with the EMS) gave a cursory inspection of the grapple location and latched condition. He then gave the Grapple Operator verbal permission to move the fuel assembly. The fuel assembly a.sembly was transferred from the core. As the Grapple Operator approached the core location of the next fuel move, he recognized that the previous step was made in error. These events were similar to the fuel handling errors which occurred during D2Rli D2Rll on January 10 and 12, 1989. Those errors were also caused by inattention to detail on the part of the Grapple Operator, lack of an effective independent verification program, and poor cOmmunications between the Grapple Operator and the Independent Verifier. A memorandum IIH!IIlOrandUII had been issued by the Assistant Superintendent of Operations on January 13~ 1989 clarifying the responsibilities of the Independent Verifier. 13, The clarification only included verifying that the correct assembly was latched. This clarification was later incorporated into applicable i( procedures. II 1.15-54 ZNLD:708:12

Revision 8

• April 1992 CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

1. A discussion was held between members of the Fuel Handling Department (management and bargaining unit), Operations Management and implement an Regulatory Assurance to determine the steps necessary to implemeQt effective independent verification program. As a result of these discussions, a Temporary Procedure Change (TPC) was made to DFP 800-1, "Unit 2 (3) Master Refueling Procedure," on October 2, 1990, delineating the steps which the Independent Verifier must follow to assure that the correct fuel assembly is being grappled.

2. A meeting was held between the Station Manager, other station management, and members of the fuel handling department on the importance of attention to detail, the importance of proper independent verification, and the importance of good communications on October 2, 1990.

3. A TPC to DAP OAP 7-7, Revision 1, I, "Conduct of Refueling Operations," was made restricting access of non-fuel handling personnel on the refuel grapple while fuel was being moved.

4. The Core Position Indication System was repaired on October 2, 1990 and the rest of the core was unloaded and later reloaded without error.

CORRECTIVE ACTIONS TO PREVENT FURTHER NONCQMPLIANCES

1. Fuel handling procedures will be revised before the next refueling outage (currently scheduled to begin on March 31, 1991) to delineate the steps which the Independent Verifier must follow to assure the correct fuel assembly is being grappled *

      . 2. Applicable procedures will be revised to establish compensatory actions to be taken during fuel moves to and from the reactor with the Core Position Indication System out-of-service before the next refueling outage.

3. Applicable procedures will be revised to restrict the movement of fuel with non-fuel handling department personnel on the grapple before the next refueling outage.

4. A requirement will b~ established for fuel handlers to demonstrate e~tabli8hed independent verification program the elements of the ektablished before (or at the beginning) of each refueling outage. Good communication techniques will also be included in the demonstration.

This program will be established before the next refueling outage. DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on October 2t2, 1990 when further effective management controls were established to control activities on the refuel c (- floor and to define responsibilities of the Independent Verifier. II 1.15-55 ZNLD:708:13 2NLD:708:13

UNITED STATES Revision 8 NUCLEAR REGULATORY COMMISSION April 1992 ADril REGION III ( 79' 799 ROOSEVELT ROAD ELLYN, ILLINOIS GLEN ELLYN. ILLINOLS 60t37 601 J7 FEB 6 :291

   ~~cket No. 50-237 Docket No. 5C-249 Commonwealth Edison Company ATTN: Mr.

Hr. Cordell Corde 11 Reed Senior Vice President Opus West III 1400 Opus Place Downers Grove, IL 6C515 Gentlemen: Thar.k you for your response dated January 7, 1991, to the December 7, 1990 Thank Notice of Violation (NOV) issued with NRC Inspection Report 50-237/90023; 50-249/90023 for the Dresden Station. We have reviewed your written comments objecting to a portion of Item 2.c of the NOV. Our response addresses each of the specific comments contained in your letter. As indicated in your response, one of the criteria requiring a valve to be locked per Dresden Administrative Procedure (DAP) 7-14, "Control and Criteria for Locked Equipment and Valves" includes "manual valves which provide a Environs, including primary and secondary containment controlled path to the EnVirons, isolation valves." Your, Your assertion that the drywell air sample system is a closed loop system is correct; however, the portion of the system downstream of the isolation valves is non-safety related and, therefore, cannot be credited as preventing a path to the environment. This portion of the system is not subjected to periodic integrated leak rate test pressure and cannot be considered a primary containment boundary. No evidence was provided to support your assumption that NRC approval of the isolation design of this system was based solely upon this being a closed loop system. We believe an additional basis was a cost/benefit decision regarding the feasibility of backfitting automatic isolation provisions. NRC approval does not alter the fact that a path to the environment still exists nor does it preclude application of other requirements. Your contention that the NRC had reviewed leaving these valves unlocked could not be verified. Although the NRC did approve the containment isolation design provisions of this system, your February 25, 1980 submittal and subsequent NRC Safety Evaluation Report (SER) dated March 5, 1980 did not include valve locking requirements. With regard to Table rI and its application to Table II in the SER dated September 24, 1982, this SER and your response letter dated November 18, 1982 indicated that during the August 1982 site visit, you had agreed to review all containment penetrations and not limit the scope to Table II. No evidence was provided to indicate that all valves listed in Table I were explicitly reviewed by the NRC for inclusion in Table II. I( !(, position delineated in Section V of the September 24, 1982 SER was The NRC's pOSition that, unless it can be demonstrated acceptable on some other defined basis, isolation valves should be either automatic or locked closed. A case in point II 1.15-56 IlLlS-S6

Revision 8 April 1992 ( Commonwealth Edison Company 2 FEB a,; 6 19,; . indicating an unacceptable basis would be your request described in this same SER for exemption from Appendix J leak detection requ;rE~ents require~ents for specific Reactor Building Closed Cooling Water System containment isolation valves. The NRC rfjected rejected your justification that the closed loop nature of the system ir.sures insures its integrity in the event of a single active failure. Our conclusion stands as documented in the above inspection report that the failure to lock these valves closed was contrary to your own procedure and the requirements of 10 CFR 50, Appendix B, Criterion V as well as the SEP require~ents commitment. As your response indicated that you plan to provide a means to lock closed these valves, we have no other concerns in this area at this.time. Sincerely, Sincere ly , H ert J. Miller, Director Division of Reactor Projects cc w/enclosure: O. D. Galle, Vice President - BWR ,( Operations T. Kovach, Nuclear Licensing Manager E. O. D. Eenigenburg, Station Manager OCD/DCB DCD/DCB (RIDS) OC/LFDCB Resident Inspectors LaSalle, Dresden, Quad Cities Richard Hubbard J. W. McCaffrey, Chief, Public Utilities Division Robert Newmann, Office of Public Counsel, State of Illinois Center B. Siegel, LPM, NRR II 1.15-57

c e Commonwealth Edison 1400 OPUS Place Downers Grove, IIlrnois illInois 60515 January 7, 1991 Revision 8 Aoril 1992 Mr. A. Bert Davis Regional Administrator - Rill U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, II 60137

Subject:

Dresden Nuclear Power Station Units 2 and 3 Response To Notice of Violation Contained in Inspection Report 50-237/90023; 50-249/90023 Docket Nos. 50~237 NRC pocket 50-237 and 50-249

Reference:

W. D. Shafer letter to Cordell Reed dated November 28. 1990 transmitting NRC Inspection 50~237/90023; 50-249/90023 Report 50-237/90023;

Dear Mr. Davis:

This letter provides the Commonwealth Edison Company (CECo) response (attached) to the subject three Level IV violations transmitted by the referenced NRC Inspection Report for Dresden Station. The three violations identified were: 1) inadequate training to assure satisfactory knowledge of plant administrative requirements. 2) the failure to follow procedures and instructions and 3) inadequate requirements, corrective actions with regard to fuel bundle mispositioning events. CECo has reviewed the Notice of Violations and in all but one example agrees that the violations occurred as described. The corrective actions detailed in the response will bring the Station into compliance and will prevent similar violations from occurring. performance, personnel-related CECo recognized a negative trend in human pertonnance, events and views this as a serious matter. Actions were taken to address each event. but we also have taken additional timely. broad. broad, comprehensive actions to address the identified negative trend. These actions are beyond those in the detailed responses to the Notice of Violation and are listed in Attachment A. If your staff has any questions or comments concerning this letter, please refer them to Rita Radtke, Compliance Engineer at 708/515-7284. cl cc: S.L B.L. Siegel, Siegel. Project Manager. NRR D. E. Hills. Senior Resident Inspector NRR Document Control Desk JAN mt 8 199t Attachment Iscl:ID707:1 III.IS-58 III .15-58

Revision 8 ATTACHMENT A A Aori Aoril1 1992 Additional actions taken to address human performance, personnel-related events:

• A meeting was held with all supervisors (CECo & Contractor) on October 17, 1990 to raise overall onsite awareness. ,Each

                                                        .Each Supervisor was required to take notes during the meeting, meet with those employees he supervises and return his notes with names of those he met with by October 18, 1990.

• A station "Self Check" initiative, called VerAntSO.

VerAntSO, was Introduced introduced the week of October 22, 1990. VerAntSO is an acronym used to remind everyone of the Anticipate, Stop and Qbserve. self-check concept and stands for Verify, AntiCipate,

• An INPO IN PO assist visit to review Human Performance activities was requested and performed on October 25 and 26,1990.

26, 1990.

• CECo Performance Assessment Department performed an overview of The CECa outage concerns during the week of October 22, 1990.
• A licensed individual from Nuclear Quality Programs Department from another CECo station performed an overall review of in-plant activities during

            .the week of October 22, 1990.
            ,the

• An in-plant walkdown and independent verification of one hundred (100) .,

out-of-services was performed to verify proper isolation of equipment for out-ot-services maintenance activities.

(

  /scl:1D707:2 Iscl:10707:2 III .15-59 III.15-59

RESPONSE TO Revision 8 8 Aoril1 1992 Apri ( NOTICE OF VIOLATION \ VIOLATION 1 10 eFR CFR 50, Appendix B, Criterion II, as implemented by Commonwealth Edison's "Quality Assurance Program" requires indoctrination and training of personnel performing activities affecting quality as neces.ary nece.sary to assure that suitable proficiency is achieved and maintained. Contrary to the above, indoctrination and training of personnel aa.uring performing activities affecting quality was inadequate in aaauring proficiency was achieved and maintained as to admini.trative administrative requirement. requirements as indicated in the following examples: . a.

8. Lack of operations personnel knowledge of Dre.den Dresden Administrative Procedure (DAP) 7-5, "Operating Logs and Records," Revision 8, and Dresden Operating Abnormal (DOA) Procedure 902-5 G-2. G-2, Revision 3, requirements for maintaining the Control Rod Drive Accumulator High Water/Low Pressure Alarm Log (AHWLPAL) resulted in the AHWLPALs for both units not being maintained between April 1990 and August 3, 1990. As such the licen.ee's licensee's program to identify repeat failures of accumulator alarms was not effective during that time period (50-237/90023-01a (50-237/90023-0la (DRP>>.

b. Lack of technical staff personnel knowledge regarding recognizing and resulted in a failure to processing conditions adverse to quality re.ulted properly identify a procedural nonadherence involving maintenance of the AWHLPAL when discovered in Hay 1990. Because of thi., this, corrective actions to prevent recurrence were not taken at that time.

(50-237/90023-0lb (DRP>>. (50-237/90023-01b This is a Severity Level IV violation (Supplement I). DISCUSSION On December 8, 1989, DAP 7-5, Revision 8 was approved. Thi. This procedure established an Accumulator HighBigh Water/Low Pressure Alarm Log (AHWLPAL). (ABWLPAL). On the same day, a revision to the Unit 2(3) Operators Daily Surveillance Log (Appendix A) va. was approved which removed the veekly weekly Control Rod Drive Accumulator Log from Appendix A &0 two procedures 80 that there would not be tva recording the sameaame information. The intent of tbethe ABWLPAL va. was to maintain an ongoing record of Control lad Rod Drive Aceumulator Accumulator Alarms in order to identify problem accumulators. Either the ABWLPAL AHWLPAL or the Control Rod Drive Accumulator Loa Log (removed fro. from Appendix A) vould would have been adequate for documenting Control Rod Drive Accumulator High Bigh Pre.**sure Water/Low Pre ure Accumulator Alarms. An ABWLPAL Book was established at each Unit Operator's Desk. c. ZNLD:708:l !IL15-60 IILIS-60

Revision 8 Aoril 1992 ADril (" ( The on-shift licensed personnel did not receive formal instruction or notification as to the maintenance of the AHWLPAL during the day-to-day operation of the plant. Therefore, the implementation of the AHWLPAL was only partly accomplished by revising the procedures. The on-shift on-ahift licensed personnel, responsible for maintaining the AHWPAL, did not receive training on this program. The Technical Staff System Engineer learned in early May 1990 that the AHWLPAL was not being properly maintained as required by DAP 7-5 and realized that either the operators would have to be trained on the use of the log or that the program would have to be revised. During this same time period, the Operations Department vas was implementing a program to independently verify accumulator valving operations. This requirement was to be implemented by having the independent verifier .ign lign the A8WLPAL AHWLPAL log book. To resolve these two unrelated concerns in a coordinated manner, it was decided to place the AHWLPAL back into Appendix A. This revision to DAP 7-5 took longer than expected because of other changes to the procedure that were not related to the AHWLPAL. It took three months to resolve all of the issues with DAP 7-5. On August 30, 1990, the requirements for the AHWLPAL were transferred from DAP 7-5 back to the Unit Operator's Daily Surveillance Log, Appendix A, at vhichwhich time CRn accumulator logging was resumed. Action to correct the problem was taken, but it was not documented through the use of DAP 9-12, "Procedural Adherence Deficiencies." The System Engineer was not aware that DAP 9-12 should have been used uaed a& aa the

( mechanism to document this need for corrective action.

CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED CORRECTIYE The logging of accumulator alarms was transfered back to Appendix A on August 30, 1990. Since that date all accumulator alarms have been logged. CORRECTIYE CORRECTIVE ACTIONS ACTIQNS TO PREVENT FURTHER NQNCQMPLI6NCES NQNCQMPLIAHCES

1. The Operations DeparbDent, Department, with aalliatance

                                              ** iatance from the Training Department and the Technical Staff, viIIwill develop a method to en.ure enaure that when procedural changes are made that alter the day-to-day routine of licensed operators, a review i8  ia made to determine vhat what training .hould Ihou1d be completed to properly imp1esent implement the change. Thi' Thil methodology will be in-place by March 31, 1991.

2. DAP 9-12 will be tailgated to all station personnel by February 7, 1991. Emphasis will be directed to it. ita purpo.e purpole and use.

i(... III.1S-61 II 1.15-61 ZNLD: 708: 2 ZNLD:708:2

Revision 8 Aoril 1992

3. A number of events, occurring during the past six months, have

( indicated that the knowledge level of station personnel with resardregard

      . to the contents of various Dresden Administrative Procedures is less than desired. To raise the station personnel knowledge level of the contents of the Dresden Administration Procedures, the Station'.

Station's ongoing training program will be reviewed to verify that all personnel involved in activities addressed in each admini.trative program/procedure are appropriately trained. A matrix of Dresden Administrative training requirements will be produced by January 31, 1991 and appropriate changes will be made to the ongoing programs by 30, 1991. June 3D, . -. DATE WHEN FULL COMPLIANCE WILL BE BE ACHIEVED Full compliance was achieved on August 30, 1990 at which date all accumulator alarms were properly logged. VIOLATION 2 10 CFR 50, Appendix B B,t Criterion V, as implemented by Commonwealth Edison Company's Quality Assurance Program, requires that activities affecting quality be accomplished in accordance with documented instructions, procedures or drawings. Contrary to the above, activities affecting quality were not accomplished documented* instructions, procedures, or drawings in in accordance with documented-the following examples: 1( This in a Severity Level IV violation (Supplement I). EXAMPLE a Dresden Operating Procedure (DOP) 1900-3, "Reactor Cavity-Dryer Separator Storage Pit Fill and Operation of the Fuel Pool Cooling and Cleanup System During Refueling," Revision 8. 8, requires constant communication cOlllllUllication between the refueling floor and the control room while filling the reactor vessel. Constant communication between the refueling floor and the control room was not maintained while filling the Unit 2 reactor vessel on October 14, 1990, resulting in the overfilling of the vessel into the ventilation duct. and contamination of o*f various area. of the third and fourth floors of the reactor building. (50-237/90023-2& (DRP>> DISCUSSION On October 13, 1990 the reactor cavity flooding evolution began. Prior to beginning the evolution, a Fuel Handling Supervisor had agreed to monitor reactor cavity level from the refueling floor. As cavity flooding progressed, the Fuel Handling Supervisor reported on the cavity level. At 0330 hours on October 14, 1990, the Fuel Bandling Supervisor informed Operations that he was leaving and that the level was approximately 1I 1/2 feet below the bottom of the ventilation openings. From this point on, the reactor cavity water level was no longer being continuously observed. Later, the Unit 2 NSO dispatched the Equipment Attendant (EA) to visually observe the cavity level. The EA erroneously reported that the level was 16 incbes inches below the bottom of the ventilation openings. ZNLD:708:3 111.15-62 II I.l5-52

Revision 8 Aoril1 1992 Ao)"i A short time later, the control room received annunciator, "Fuel Pool ( High Level." The NSO inrnediately immediately closed the Feedwater Low Flow Valve and sy~tem dispatched another EA to reject water from the fuel pool cooling syli-tem to the Condensate Storage Tank. A Shift Supervisor then proceeded to the water coming .from fourth floor of the Reactor Building where he saw vater from the ventilation ducts. DOP 1900-3 contains a precaution: '~aintain constant communications between the refueling floor and the Control Room while filling the reactor head cavity and the dryer/separator pit to prevent overflow into the ventilation ducting." This vas was not followed from the time the Fuel Handling Supervisor left the refueling floor until the alarm was received in the control room. CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

1. This event was included in a series of meetings conducted on October 17, 1990 by the Station Manager and attended by all line supervisors of CECo and Contractor organizations on site. The purpose of the meetings was to increase everyone's everyonets awareness of several recent events involving personnel error and to require subsequent meetings between these supervisors and all employees to raise the overall on-site awareness level of the need for increased attention to detail.

2. A posted operator aid was revised Ihowing showing the level of the bottom of the ventilation openings to be 469 inches instead of the previously erroneous value of 476 inches.

( 3. were couo.elled Operations personnel involved in this event vere counselled on the importance of procedure adherence.

4. Procedural adherence was addressed in a Shift Engineer's meeting held on October 24, 1990. Shift Engineers were instructed that procedures must be consulted and adhered to for all complex, unique, or infrequent evolutions.

NQNCQMPLIANCES CORRECTIVE ACTIONS TO PREVENT FURTHER NONCQMPLIAHCES

1. DOP 1900-3 will be revised to clarify when continuous visual monitoring i8 is required vhen when flooding flOOding the refuel cavity. The will be specified .0 that requirement for visual monitoring viII monitoring will not be required for .1ow continuous monitorin, slow moving evolutions when level is more than three feet from the final desired level. The procedure will be revised by March 31, 1991.

2. The Instrument Maintenance Department will revise DIP 0260-01, Figure 1 to correctly depict the bottom of the cavity ventilation openings at 469 inches by March 31, 1991.

3. To aid visual estimates of water level, the Technical Staff viII will evaluate methods of providing a graduated scale in the Reactor cavity and in the Dryer/Separator pit for Units 2 and 3. An acceptable method will then be implemented by March 31, 1991.

c( ZNLD:708:4 ZNLD: 708:4 I II1.15-63 II .15-63

                                         -s-Revision 8 Apri 1 1992 April

4. This event was reviewed by licensed operators during Cycle 8 of

( (\ continuing training. This was completed on December 7, 1990. It will be covered for non-licensed operators by March 1,I, 1991. 5. S. The EA involved in this event developed an article for the Station's monthly newsletter, discussing the importance of attention to detail, procedural adherence. adherence, and the concept of self-checking one's actions. DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on October 14, 1990 when the reactor cavity flooding evolution was completed. EXAMPLE b Specific practices required by DAP 3-5, "Out of Service and Personnel Protection Cards," Revision 22, were not followed as to preparation, review, approval, documentation and independent verification in the removal and return to service of the Unit 2 diesel fuel oil day tank drain valve on October 20, 1990. This resulted in the inadvertent draining of the day tank when the drain valve was placed in the incorrect position. (SO-237/90023-02b (DRP>> DISCUSSION In preparation for the cleaning of the Unit 2 Diesel Generator Main Fuel Oil Storage Tank, the Diesel Fuel Oil Transfer Pump Suction valve was shut and the Fuel Oil Day Tank Drain valve was checked to be shut by a member of the Operations Staff on October 8 or 9, 1990. "Do Not Operate" tags supplied by the storage tank cleaning contractor were placed on the valves. No CECo Out-of-Service was written for the tank cleaning. On October 20, 1990, between 10:30 and 11:00 am, the same member of the Operations Staff opened the Transfer Pump Suction Valve. In addition, he opened the Day Tank Drain Valve, even though he had checked that this These valve manipulation. valve was shut approximately 12 days prior. The.e manipulations were performed without the knowledge of on-.hift on-shift Operations Operation. or procedural guidance. At approximately 11:20 ... the "Unit 2 am on Saturday, October 20, 1990, tbe Diesel Generator Day Tank Bi/Lo Hi/La Level" alarm was received in tbe the Control Room. At approximately the .... samee time, two members of the Technical Staff strong were in the vicinity of the diesel generator room and observed a .trong odor of fuel oil. Upon entering the room, the,they noticed fuel oil on the floor. CORRECTIVE SIEPS TAKEN AND RESULTS ACHIEVED CQRRECTIVE

1. Ipil1 A member of the Technical Staff traced the source of the fuel .pi11 to the drain on the Unit 2 Fuel Oil Day Tank, found drain valve 2-5212-500 approximately one to two turns open and closed the valve.

spilled. Approximately 600 gallons of fuel oil were .pilled. (. ZNLP: 708:5 ZNLD: II LIS-54 I.1S-54

Revision 8 April 1992

2. The involved Operations Supervisor was counseled on the need to

( shift personnel to ensure that interact with Operations Department shIft all valves necessary to adequately isolate a component are included on the appropriate Out-Of-Service. The involved individual was reminded that unauthorized valve manipulation is against plant policy and could lead to personnel injury or equipment damage. CORRECTIVE ACTIONS TO PREVENT FURTHER NONCOMPLIANCE

1. The Day Tank drain yalves valves on all three emergency diesels were locked shut. The Locked Valve Checklist and System Checklist will be revised to include the Day Tank Drain on all emergency diesels in the locked closed position by March 31, 1991.

2. The details of this event were reviewed with all station personnel at the October 18, 1990 tailgate meetings, emphasizing the need to properly use the Out-Of-Service program and the hazards of unauthorized equipment manipulations.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on October 20, 1990 when the drain valve was closed. EXAMPLE c DAP 7-14, "Control and Criteria For Locked Equipment and Valves," Revisi9n Revis~on 2, requires manual valves in the flowpath of systems required for plant shutdown during post-accident situations or which provide a controlled path to the environs environs,t including primary and secondary containment isolation valves to be locked. Prior to November 1990 1990,t manual valves including the Units 2, 3 and 2/3 diesel generator service water three-way valves and the Units 2 and 3 drywell manifold sampling system containment isolation valves were not locked or designated to be locked (50-237/90023-02c (50-237/90023-D2c (DRP>>. DISCUSSION The Diesel Generator Service Water flow reversal valves were erroneously excluded from the locked valve checklist. This vas was due to the valves having a mechanical locking device which prevents the valves from repositioning. The Station believed that the locking device (vhich(which does not have a keyed lock) fulfilled the administrative requirements of the locked valve program. CECo does not believe that a requirement existed to "lock closed" the 2(3)-8507-500 through 521 Drywell Air Sample Valves. The USNRC had, on two separate occasions, the opportunity to review and assess the those valves. On each occasion they found the design to acceptability of .those be acceptable: 1(. ( ZNLD:708:6 II I.l15-65 III. 5-65

Revision 8 Apr; April1 1992

1. In response to NUREG 0578, in a letter dated February 25. 25, 1980, CECo supplied information on primary containment isolation valves. An

( excerpt from that letter reads as follows, "All non-essential systems that provide a possible open path out of the primary containment were found to be either isolated by isolation signals, by cbeck check valves that would prevent flow out of the containment, by manual valves that are normally closed during reactor operation, or as in the case of instrument lines by closed piping systems." Included were the 2(3)-8507-500 through 521 valves with their classification (non-essential) and a sketch showing their configuration. The USNRC, in a letter dated March 5, 1980, responded, "We '~e conclude that the licensee has completed a re-determination of which containment isolation penetrations are essential or non-essential. All non-essential lines are either automatically ilolated isolated by diverse signals or technical justification has been provided. Modifications have been made to prevent inadvertent re-opening of i.olation isolation valves. Based on the above, we find that the licensee bas has satisfied the requirements of this item."

2. Various correspondence exists documenting the scope and depth of the USNRC review of SEP Topic VI-4, "Containment Isolation Systems." An NRC letter dated December 18, 18~ 1981, transmitted a draft SER on the topic and requested that CECo provide comments and additional information. CECo's response of May 21, 1982 provided information on the 2(3)-8507-500 through 521 valves. Based upon that response, the NRC issued their final SER on September 24, 1982. Table I of that SER provides a list of valves which they reviewed; the 2(3)-8507-500 through 521 valves are included on that list. A .ection section of the SER titled, ti tled, "Administrative Control," identifies valves which have inadequate administrative controls and which should be listed as "locked closed" instead of "normally closed. closed."1t These valves are listed in Table II; the 2(3)-8507-500 through 521 valves are not included. Another section titled, i'Manual "Manual Isolation Valves, Vslves," lists It other valves which should be in a ttlocked "locked closed" position; once again the 2(3)-8507-500 through 521 valves are not included.

Although the basis of acceptability for these valves being "normally open" or "normally closed" is not provided in the SERs, iitt presumably is due to the Drywell Drywe11 Air Sample system being a closed loop system. In any case, the NRC reviewed these valves as part of po.t-7MIpost-IMI and SEPt SEP, and did not require them to be "locked closed." The Inspection Report and Notice of Violation reference a November 18, cOllllllitment to ureview 1982 cOIIIIIitlllent "review all containment penetrations in the plant and limit tbe not lilllit the scope to Table II in the SER." DAP 7-14, "Control and Criteria for Locked Equipment and Valves," is allo also referenced. One of the DAP 7-14 criteria for locked valves is, t~anual '~anual valves which provide a controlled path to the Environs, including Primary and Secondary Containment isolation valves." Since the Drywell Air SUlple Sample .y.tem system it is a closed loop system, leakage past these valves would not provide an elosed uncontrolled path to the environs. CECo does not believe that operation of a system consistent with the plant's original design basis and in accordance with NRC SERs and Station administrative progrUls programs constitutes a violation of NRC requirements. ( ZNLD:708:7 rrI.lS-66 I II .15-66

Revision 8B April 1992 During a recent investigation into the use of a temporary sample pump to ( obtain drywell air samples (in which the Drywell Air Sample system's closed loop was broken), questions arose relative to the ability of the original Drywell Air Sample system to withstand seismic and accident conditions. In view of this recent information, CECo is re-a ** e.sing re-*** easing the acceptability the system. oystem. Past performance of the system oyotem has shown only limited usefulness in its ability to locate sources of leakage into the drywell. With the recent approval of the Station's Generic Letter 88-01 submittal (in which no credit was taken for the Drywell Air Sample system) it is believed that the system may be removed. Final re.olution resolution February, 1991. As an of the Drywell Air Sample system is expected in February. interim measure, valves 2(3)-8507-500 through 521 have been taken out-of-service closed. These valves will remain controlled by the Or locked closed in accordance with DAP 7-14 as long as out-of-service or the system remains in place. Since DOno method of locking these valves presently exists, work requests have been written to provide a means of locking them. The valves will be added to the locked valve checklist and locked as appropriate prior to clearing the out-of-service on these valves. CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED Temporary Procedure Change 90-408 was written against Dresden Operating Procedure (DOP) 040-M3, "Locked Valve List: Acceuible During Operations," adding the Diesel Generator Service Water flow reversal valves to the Locked Valve List. The valves were also locked at that time. c ( CORRECTIVE ACTIONS TO PREVENT FURTHER NQNCQMPLIAHCES

1. DOP 040-M3 will be revised by March 31, 1991 to include the Diesel Generator Service Water three-way valves.

2. A review of other valves with mechanical locking devices will be conducted to assure that they are not being inappropriately excluded from the locked valve program by March 31, 1991.

3. necessary to assure DAP 7-14 criteria will be reviewed and revised as nece8sary the locked valve criteria are easily understood by station personnel. This will be accomplished by March 31, 1991.

CQMPLIANCE WILL BE ACHIEVED DATE WHEN FULL COMPLIANCE Full compliance was achieved on November 27, 1990 when the valve. valves were Were locked into position. EXAMPLE d DAP 15-6, "Preparation and Control of Work Requests," Revision 0, travelerslprocedures, requires work to be performed per repair manual(s), travelers/procedures, or work instructions provided in the work package. On October 15, IS, 1990, work prescribed for disassembly of the Outboard Containment Isolation Feedwater Check Valve 220-62B was performed instead on Outboard (50-237/90023-02d Containment Isolation Feedwater Check Valve 220-62A (50-237/90023-o2d (DRP>>

• ZNLD:708:8 ZNLD: 708: 8 III.IS-57 III.15-67

Revision 8 April 1992 DISCUSSION ( As a result of local leak rate tests (LLRTs) performed on September 24 and 30, 1990 on the 2-220-52A 2-220-62A and 62B check valves. valves, the decision was made to repair the valves. A A maintenance pre-job briefing and ALARA pre-job briefing were performed on October 12, 1990 prior to proceeding to the work area for disassembly of the 62B check valve. The Maintenance Supervisor accompanied his crew to the work location and directed them to begin work on what he believed to be the 62B valve, but was actually the 62A valve. Upon removal of the valve bonnet and seal ring, the valve body was found full of water and the valve disc stuck in the open position. All work was immediately stopped, the Maintenance Supervisor and Technical Staff were notified of the as found condition. Water was pumped from the valve body and work continued on valve decontamination, inspection and repair. On October 17, 1990 a Radiation Protection Technician (RPT) surveyed the valve plug and seat. The RPT was concerned that work was being performed on the wrong valve and questioned the crew several times whether they were working on the correct valve. The crew indicated that they were sure they were on the correct valve and that it had been verified with their supervisor. On October 19, 1990 the Station ALARA Coordinator questioned the reported radioactive contamination levels inside the opened valve and investigated the possibility of work being performed on the wrong valve. This was

  ~onfirmed to be the case.

{ CORRECT lYE STEPS CORRECTIVE TAKEN IAKEN AND RESULTS ACHIEVED

1. An Out-of-Service was hung for the valve; this included closing the downstream manual valve (2-220-57A) to provide better isolation.

Appropriate radiation protection measures were taken, including procedure.. AI completing the proper Radiation Work Permit procedures. As the 2-220-62A valve was also scheduled for overhaul due to LLRT results, work was allowed to continue under the proper work package.

2. 1990, A tailgate discussing this event was presented on December 20, 1990.

self-verification and that each nuclear emphasizing the need for lelf-verification worker has the responsibility to as lure the equipment to be worked on a.sure actions to is the equipment identified in the work package and that action. be taken are correct. CORRECTIVE ACTIQNS TO PREVENT FURTHER CORRECT lYE ACTIONS FURtHER NQNCQMPLIANCES NQNCQKPLIANCES

1. A more positive identification of the 2-220-61A, 2-220-62A, 62B, and 59 valves steam pipe tunnel has been provided. Additional valve for the Unit 2 Iteam identification has been applied to the support structure located over the 62A and 62B valves.

2. A tailgate article will be developed by January 31, 1991 to inform plant personnel that identification tags are expected to be attached Ie i(

on all plant components. If equipment tags are not found, the assure components are labeling coordinator should be notified to alsure properly tagged and operating peraonnel should be contacted to aa.si.t

                                                                          ** ist starting work.

in proper component identification before Itarting ZNLD:708:9 II 1.15-68

                                          -1.0-                               Revision 8 April 1992

3. This event will ,be be incorporated into continuing training for

(" ( Maintenance, Maintenance~ Operations, Radiation Protection, and Technical Staff Operations. personnel by December 31, 1991. Fmphasis Emphasis will be placed on the" the* ,.'" potential significance of opening the wrong primary system boundary, opportunities by the working group and others which were available to identify that the wrong equipment was being worked on, and .ethods methods by which the correct component could have been identified (outage walkdown, pipe/penetration labeling, RWP survey maps). walkdoWD,

4. Radiation Protection Survey Maps will be upgraded as necessary to provide for identification of equipment specified on the maps. This will be completed prior to the next scheduled refuel outages for each unit.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on October 19, 1990 when the 626 62A valve was properly removed from service. EXAMPLE e PAP DAP 15-6, "Preparation and Control of Work Requests," Revision 0, requirements were violated on August 8, 1990, when prescribed for calibration of Unit 3 Torus to Reactor Building Vacuum Breaker A Pressure Transmitter DTP-l6226 DTP-1622A was performed instead on Pressure Transmitter DPT-1622B. This resulted in advertent opening of the Unit 3 Reactor Building Vacuum Breaker B. (50-237/90023-02e (PRP>> (DRP>> DISCUSSION On August 8, 1990 the Instrument Maintenance Department (IMO) (IHO) was performing a calibration check on DPTOPT 16226 1622A [Torus to Reactor Building 3-l601-20A) Pressure Transmitter] Vacuum Breaker (AOV 3-1601-20A) Transmitter) using procedure DIS 1600-20 1600-20,t "Torus to Reactor Building Differential Prenure Preuure Translllitter Transmitter 1622A l622A and B Calibration and Maintenance Inspection. Inspection."It The Instrument Mechanic (1M) va1ved-out valved-out the DPT1622A tran.mitter transmitter and connected the calibration instruments to obtain a set of as found readings. The as found readings were outside the ideal calibration tolerance range on the conservative side. For the 1MIM to adjust the calibration setting for DPT 1622A or 1622B he down, turn 180*, has to get down. 180', arch his head and back under the transmitter. While being upside down, the IH 1M proceeded to adjust what he thought to be DPT 162ZA 1622A for AOV 3-160l-20A 3-l60l-20A to bring the calibration within specified instrument tolerance range. While he was making the adjustment, the AOV 3-1601-20B 3-160l-20B opened with a resultant Control Room Annunciator. The 1M had adjusted DPT 1622B instead of DPT OPT 1622A. l622A. iC" ZNLD:708:10 ZNLD: 708: 10 IIL15-69 IIl.lS-69

Revision 8 April 1992 ( CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

1. The Master Instrument Mechanic counselled the 1M involved in this event. Emphasis was placed on total job awareness needing to be main tained aatt all times 8a job is in progress. Also emphas maintained h was emphasi.

placed on heightened awareness while working in conge. ted areas of congested the plant since the possibility of poor job performance is enhanced.

2. The Master Instrument Mechanic discussed this event at a department tailgate meeting. The discussion included a review of the situations on this job and a reminder of what is expected of In.trument Instrument Maintenance Department personnel when working in congested areas.

CORRECTIVE ACTIONS TO PREVENT FURTBER FURTHER NQNCgMPLIAHCES NONCOMPLIAHCES

1. The labeling of the ~p ~P transmitters on both units (DPT 162ZA 1622A and l622B) will be improved by placing a label above the transmitter and 1622B) removing any labels below the transmitters.

2. The ~P transmitters (OPT (DPT l622A and l622B) will be rotated 180 degrees to relocate the adjustment screws on the top of the transmitter (Work Requests D95106, D95l06, D95107, D95l07, D95l08 D95l09) by June 1.

D95l08,t and 095109) I, 1991 (during next refueling outage) for Unit 3 and by October It I, 1991 for Unit 2. This will greatly enhance access to the adjusting screws and minimize the possibility of adjusting the wrong transmitters. Ie DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on August 8, 1990 when the Vacuum Breaker was reclosed. VIOLATION 3 10 CFR 50. 50, Appendix B B,t Criterion XVI, as implemented by Commonwealth Program,"u requires that conditions adver.e Edison's "Quality Assurance Program. adverle to quality be promptly identified and corrected and, in the ca.e of significant conditions, the measures a.sureassure the cause i. determined and corrective action taken to prevent repetition. Contrary to the above. above, following the fuel bundle aispositioning aispo.itioning event. eventl of January 10 and 12, 1989, corrective action. actions were in.ufficient inlufficient to prevent repetition in that similar

                         .imilar event.

events occurred on October 1,I, 1990 and October 2, 1990. (50-237/90023-08 (DRP>> 2t This in a Severity Level IV violation (Supplement I). III.15-70 IIUS-70 ZNLD: 708 : 11

Revision 8 April 1992 DISCUSSION c( On October 1, 1990, Unit 2 was in the Refuel Mode. Fuel Handlers were unloading fuel from the reactor. The grapple's Core Position Indication System was improperly indicating position in the east-west direction. The current fuel move was the last fuel move from the perimeter of the core. The next fuel move was to be from the interior of the core where no fuel assemblies had yet been removed. The Fuel Handling Supervisor went onto onlo the Refueling Grapple to caution the fuel handling crew that the next transfer was from a different region of the core. After the current step, the duties of the Independent Verifier and the Grapple Operator were scheduled to be exchanged between the two men. The Grapple Operator grappled the wrong fuel assembly. As the Independent Verifier had been cautioned about the next fuel move, he was studying a core map to determine the location of the next step's fuel transfer rather than independently verifying what the Grapple Operator was doing on the current step. The fuel assembly was erroneously transferred, the Grapple Operator and Independent Verifier exchanged transferred. duties, and the "new" Grapple Operator began to perform the next step. While examining the core, the fuel handling crew discovered that the previous fuel move had been performed in error. At this time fuel moves were suspended while discussions between the Operating Engineer, Shift Engineer and Fuel Handlers took place. Prior to resumption of unloading the core, it was decided that verbal concurrence would be required from the verifier that the proper step was being initiated, prior to removing a fuel bundle from the core. A A further review of the event was conducted the next morning. On October 2, 1990, an Electrical Maintenance Supervisor (EMS) was on the Fuel Grapple to observe the operation of the Core Position Indication System inill preparation for repairs scheduled for later in the day. These repairs were to be completed in response to a corrective action from the first unloading error. The Independent Verifier va. wal diseua.ing discUising it.its operation with the EMS. The Grapple Operator positioned the Irapple over the wrong fuel assembly. The Independent Verifier (while engaged in a conversation with the EMS) gave a cursory inspection of the grapple location and latched condition. Be then gave the Grapple Operator verbal permission to move the fuel aallembly.

                                  ** embly. The fuel assembly as.embly va.

was transferred from the core. As the Grapple Operator approached the core location of the next fuel move, he recognized that the previous step was made in error. error .. These events were similar to the fuel handling errors whieh which occurred D2Rll on January 10 and 12, 1989. Those errors were also caused during D2Rli by inattention to detail on the part of the Grapple Operator, lack of an effective independent verification program, and poor communications between the Grapple Operator and the Independent Verifier. A memorandum IIIeIBOrandum had been issued by the Assistant Superintendent of Operations on January 13, 1989 clarifying the responsibilities of the Independent Verifier. The clarification only included verifying that the CQrrect correct *** embly vas ass.embly was latched. This clarification was later incorporated into applicable procedures. ZNLD:708:12 ZNLD:708:l2 III. 15-71 III.lS-71

Revision 8 April 1992 c CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

1. A discussion was held between members of the Fuel Handling Department (management and bargaining unit), Operations Management and Regulatory Assurance to dete~ine determine the steps necessary to implement an effective independent verification program. As a result of these discussions discussions,t a Temporary Procedure Change (IPC) (TPC) was vas made to DFf DFP 800-1, "Unit 2 (3) Master Refueling Procedure,"

Procedure." on October 2,

2. 1990.

delineating the steps which the Independent Verifier must follow to assure that the correct fuel assembly is being grappled.

2. A meeting was held between the Station Manager, Manager. other station management, management. and members of the fuel handling department on the importance of attention to detail, detail. the importance of proper independent verification, verification. and the importance of good communications on October 2. 1990.

3. A TPC to DAP 7-7,7-7. Revision 1,

1. "Conduct of Refueling Operations,"

Operations." was made restricting access of non-fuel handling personnel on the refuel grapple while fuel was being moved.

4. The Core Position Indication System was repaired on October 2, 2. 1990 and the rest of the core was unloaded and later reloaded without error.

CORRECTIVE ACTIONS TO PREVENT FURTHER NQNCQMPLIAHCES NQNCQMPLIANCES

1. Fuel handling procedures will be revised before the next refueling outage (curren~ly (currently scheduled to begin on March 31,31. 1991) to delineate the steps which the Independent Verifier must follow to assure the correct fuel assembly is being grappled.

2. Applicable procedures will be revised to establish compensatory actions to be taken during fuel moves to and from the reactor with the Core Position Indication System out-of-Iervice out-of-service before the next refueling outage.

3. Applicable procedures will be revised to restrict the movement of fuel with non-fuel handling department personnel on the grapple before the next refueling outage.

4. A requirement will be established for fuel handlers to demon.trate demonstrate the elements of the established independent verification program before (or at the beginning) of each refueling outage. Good communication techniques will also be included in the demonstration.

This program will be established before the next refueling outage ** DATE WHEN FULL COMPLIANCE CQMPLIANCE WILL BE ACHIEVED

 ~ull   compliance was    achieved on October    2. 1990 when further effective 2,

management controls were established to control activities on the refuel floor and to define responsibilities of the Independent Verifier. 708 : 13 ZNLD: 708: III.15-72 III. 15-72

TAB 16 Revision 8 April 1992 ( DRESDEN 2 &3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspection Report No. 50-237/90023 and 50-249/90023 Title IIL16-1 IlL16-1 Inspection Reports No. 50-237/90027 and 50-249/90026 dated January 17, 1991. III. 16-32 February 15, 1991 CECo letter from T. J. Kovach and A. Bert Davis (NRC), Response to Notice of Violation Associated with Inspection Report No. 50-237/90027 and 50-249/90026 . 50-249/90026.

• c.(

( III.16-i

   '~,-- -

Revision 8 INSPECTIOH INSPECTION REPORT

SUMMARY

April 1992 SO-237/90027; 50-237/90027; 50-249/90026 ( Inspectors: D. Hills, Peck. J. Monnlnger, Hills. M. Peck, Monnlnger. J. Holmes Inspection Scope: Routine, Routine. unannounced safety Inspection Inspection Period: November 17 through December 29,29. 1990 Violations Non-Cited Violations Unresolved Unreso 1ved Items Open Items 2 5 6 1 VIolatIons: Violations: One level IV violation was Identified for failure to follow procedure In regard to maIntenance maintenance practices on 10 eFRCFR 50 Appendix R emergency lighting batteries. No reply to this violation 1sIs required because actions had been taken to correct the Identified violation and to prevent recurrence (p. 1,6-7>.1.6-7). The second vIolation, violation. also a level IV, IV. was for failure to report to the NRC an Engineered Safety Feature (ESF) actuation In accordance with 10 eFRCFR 50.72 (p. 1, 13-14). This Violation

1. 13-14>. violation does require a reply.

'( Nonclted Violations: The following are flve five violations for which Notices of Violation are not being Issued: issued: failure to adequately control the use of under vessel platform covers such that a source range monitor was subsequently damaged during movement (p. 2, 5-6>. 5-6). failure of a Quality Control inspector Inspector to follow radiation protection administrative procedures resulted tn In contamination of the Inspector (p. 2, 9>. 9). failure to follow a procedure during a matn main steamline steamllne plug installation Installation resulted in In a small portion of clean, clean. demineralized water drained to the drywell <p. (p. 2, 16-17>. failure to maintain records, which Include a written evaluation, for the standby gas treatment system safety evaluation. (p. ?

                       <p. 2, 20-21>.

fal1ure Imposition of a civil penalty failure to post a proposed imposition associated with the use of a temporary sample pump In the drywell mainfold sampling system (p. 2, 23-24). ( ZNLD/710:10 ZNLD171 0: 10 111.16-1 III.16-1

Revision 8 April 1992 Unresolved Items: The following six unresolved items were identified: the servIce service aIr air supply to three of the Unit 3 drywall drywe11 purge and ventilation fan dampers had been disconnected with no temporary alteration tags attached to the air lines or the dampers dampers' operators. This Is I is pending review of operator Involvement involvement and safety significance (p. 2, 2. 12). failure of an electrical wiring diagram to reflect the actual plant configuration which resulted In in an unexpected (p. 2, ESF actuation <p. 2. 12-13). use of a temporary pump and hose assembly to augment the filtering capability of .the fuel pool clean-up system during refueling is pending further review regarding safety Implications and proper use of procedures and/or temporary alterations <p.(p. 2. 15). failure of a primary containment Integrated leak rate test (ILRT) due to a leaking torus to reactor building vacuum breaker Is pending further review of the adequacy of post-maintenance testing (p. 2, 2. 17). failure to obtain a technical specification change and Initiate initiate the required surveillance calibration for a switches on the control valve fast modification to limit swItches acting solenoids (whIch (which provide the scram signal to the reactor protection system on generator load reject) Is pending further review of the modification with respect to 10 eFR CFR 50.59 requirements (p. 2,2. 21-22>. 21-22). concerns with respect to conformance to Generic Letter 82-12 guidelines on overtime as 1t It Is applied to all plant staff groups (p~ (p: 2,

2. 22-23).

Open Items: Determination of the current status of rematn1ng remaining Systematic Evaluation Program Items 1sIs considered an open ,tem Item (p. 24). ( ZNL0/710:10 ZNLD/710:10 III.16-2 1II.16-2

U:"JITED U~JITEO ST:'TE5 ST:" TES NUCLE~R NUCLE .. R REGUL .. TORY REGUL).TORY CO~.1\1ISSI0'\l CO',1\lISSI0\J Revision 8 REGIG"J REGIG:'>J III til April 1992 "39 ROOSE'/E,-T

                                                '39   ROOSE'.rE.:....T =OAD
                                                                        =OAD

( ~LEN

iLEN ELL"N.

ELLvN. ILLINOI,

                                                              !!..LINOIS   5013>

50137 JAN 1 17 ;831 Docket No. 50-237 Docket No. 50-2'9 50-2~9 Commonwealth Edison Company Corde 11 Reed ATTN: r*lr. Cordell Senior Vice President Opus Hest IIIIII P Jace 1400 Opus Place Dowl1e rs Grove, rI L 60515 Oowlle Gentlemen: This refet"s refet*s to the routine safety inspection conducted by O. D. Hills, M. Peck, J. Monninger and J. Holmes of this office on November 17 through December 29, 1990, of activities at Dresden r~uclear Power Station, Units 2 and 3 authorized by Operating License Hos. Nos. DPR-19 and DPR-25 and to the discussion of our findings with Mr. E. Eenigenburg and others at the conclusion of the inspection. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative r2cords, obs~rvations. and records, obs~rvationst interviews with personnel. During this inspection, certain of your activities appeared to be in violation URC requirements of ~RC requirements, as described in the enclosed Notice. With respect to t violation A, the inspection showed that actions had been taken to correct the identified violation and to prevent recurrence. Consequently, no reply to this item is required and we have no further questions regarding this item at this time. Regarding the remaining remainil1g item, a written response is required. In addition, addition five violations were identified for which Notices of Violation t b~ing issued in accordance with the exercise of discretion delineated are not being in either 10 CFR eFR 2, Appendix C, Section V.A or V.G.l. In accordance with 10 CFR eFR 2.790 of the Commission1s Commission's Regulations, a copy of this letter and the enclosure(s) will be placed in the NRC Public Document Room

• Room.

.( IIr.16-3

Revision 8 April 1992 ( Commonwealth Edison Company 2 JAN 1 7 1991 1S91 The responses directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of Management and Budget as required by th~ the Paperwork Reduction Act of 1980, PL 96-511. We will gladly discuss any questions you have concerning this inspection. Sincere ly, Sincerely, Lt) iJ 5-~A/\L/ L) $J~I'~ij Chi~f

w. D. Shafer: Chief Reactor Projects Branch 1

Enclosures:

1. Notice of Violation

2. Inspection Report No. SO-237/90027(DRP) 50-237/90027(DRP)

No. 50-249/90026(DRP) cc w/enclosures: O. Vic~ President - BWR D. Galle, Vice Operations T. Kovach, Nuclear Licensing Manager E. D. Eenigenburg, Station Manager DCD/DCB (R OCD/DCB (RIDS) IDS) OC/LFDCB OC/LFOCB Resident Inspectors LaSalle, Dresden, Quad Cities Richard Hubbard J. w. W. McCaffrey, Chief, Public Utilities Division Robert Newmann, Office of Public Counsel, State of Illinois Center III.16-4

Revision 8 April 1992 (('

                                                                                 . ;~

NOTICE OF VIOLATION Commonwealth Edison Company Docket Nos. 50-237; 50-249 Dresden Nuclear Power Station License Nos. DPR-19; DPR-25 During an NRC inspection conducted on November 17 through December 29, 1990, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1990), the violations are listed below: A. Section III.J. of 10 CFR Part 50, Appendix R, requires emergency lighting units with at least an 8 hour battery power supply to be installed in all areas needed for operation of safe shutdown equipment and in access and egress routes thereto. Technical Specification Section 6.2, entitled "Plant Operating Procedures," requires procedures that detail the Fire Protection Program Implementation, be prepared, approved and adhered to. Contrary to the above, the licensee did not adhere to Dresden Electrical Survei llance (DES) 4153-02 t "Emergency Lighting Month Surveillance 1y Inspect Monthly ion," Inspection," I.d.(1 j

• in that distilled water was not added to the Revision 0, Section I.d.(I),

emergency light when the electrolyte level was identified on October 29, 1990, below the fill line as required by the procedure. This is a severity level IV violation (Supplement I).

8. 10 CFR 5Q.72(b)(2)(ii) 50.72(b)(2)(ii) requires the NRC to be notified within four hours of the occurrence of any event or condition that results in manual or automatic actuation of any Engineered Safety Feature (ESF).

Contrary to the above, the unexpected closure of several Unit 2 Group II primary containment isolation valves upon lifting of an electrical lead during post-maintenance testing on December 8, 1990, constituted an automatic actuation of an ESF and the NRC was not notified of the occurrence. This is a severity level IV violation (Supplement I). 16-5 II 1.16-5 III.

Revision 8 April 1992 ( Notice of Violation 2 A, the inspection showed that actions had been taken to With respect to Item At correct the identified violation and to prevent recurrence. Consequently, no reply to the violation is required and we have no further questions regarding 8, pursuant to the provisions of this matter. With respect to Item S, 10 CFR 2.201, you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply, including for each violation: (1) the corrective steps that have been taken and the results achieved; (2) the corrective steps that will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown. (-/7-~( 1-/7- ~( bated 1 ( ( I.16-6 II 1.16-6

Revision 8 April 1992 ( U. S. NUCLEAR REGULATORY COMMISSION REGION IIII REG ION II Report Nos. 50-237/90027(DRP); 50-249/90026(DRP) 50-237/90027{DRP); Docket Nos. 50-237; 50-249 License Nos. DPR-19; DPR-25 Licensee: Commonwealth Edison Company P. O. Box 767 IL 60690 Chicago, Il Facility Name: Dresden Nuclear Power Station, Units 2 and 3 Inspection At: Dresden Site, Morris, IL Inspection Conducted: November 17 through December 29, 1990 Inspectors: D. Hills M. Peck J. Monninger Approved By: ~~B. s, Projects sect1~~'lB Sect{~~* IB . ( (\. Inspection Summary Ins eetion ection durin the eriod of November 17 throu h December 29 1990 eport os.  ; resldent lnspection of previously Areas Inspected: Routlne unannounced res,dent

     ,dentified identified inspection items, licensee event reports fol1owup, followup, plant operations, maintenance/surveillance, engineering/technical support, safety assessment/quality verification, systematic evaluation program items and report review.

Results: vlolations were identified for which Notices of Violation are being Two vloJations issued. One dealt with a failure to follow procedure in regard to maintenance practices on 10 CFR 50 Appendix R emergency lighting batteries (paragraph 2). The other involved a failure to report to the NRC an Engineered Safety Feature (ESF) actuation in accordance with 10 CFR 50.72 (paragraph 4.e). ( III.16-7

Revision 8 April 1992 ( Five violations were identified for which Notices of Violation are not being issued in accordance with the exercise of discretion delineated in 10 CFR 2. 2, Appendix C, Section V.A or V.G.l. These involved a failure to adequately control the status of under vessel platform covers such that a source range monitor (SRM) was subsequently damaged during movement (paragraph 2), a failure to follow procedure involving radiation protection practices (paragraph 4). 4), a failure to follow procedure regarding a main steamline plug installation such that a small portion of the reactor cavity drained to the drywel1 drywell (paragraph 5.a). 5.a), a failure to maintain records for a standby gas treatment system (SGTS) written safety evaluation (paragraph 5.b) and a failure to post a proposed imposition of a civil penalty (paragraph 7.c). Six unresolved items were also identified. The inspector's identification that the service air supply to three of the Unit 3 drywell purge and ventilation fan dampers had been disconnected is pending review of system design and the role of the operators in the event {paragraph (paragraph 4.c}. 4.c). The failure of electrical wiring diagram 12E2697 to reflect actual plant configuration which resulted in an unexpected ESF actuation is pending review licensee corrective actions (paragraph 4.e). The inspector's of 1icensee identification of the licensee's usage of a temporary pump and hose assembly to augment filtering of the reactor cavity water during refueling without a procedure or temporary alteration is pending further review of safety implications and 10 CFR 50.59 aspects (paragraph 4.f). Failure of a primary containment integrated leak rate test (IlRTJ (ILRTJ due to a leaking torus to reactor building vacuum breaker is pending review of the adequacy of post-maintenance ./ testing (paragraph 5.b). The failure to obtain a technical specification

\,    change and initiate corresponding surveillance calibration requirements for a modification to the generator load reject scram on turbine control fast closure is pending further review of 10 CFR 50.59 implications (paragraph 5.c). Inspector concerns regarding conformance to Generic letterLetter 82-12 guidelines on overtime ;sis pending review of further plant staff groups (paragraph 7.b).                                          .

Plant Operations A review of the Operations Department in regard to overtime policy indicated that instances of exceeding Generic letter Letter 82-12 guidelines was minimal. However, certain concerns were raised in that fuel handlers, except for the However. supervisors, were not included and the level of approval for fuel handling supervisors. exceeding the guidelines delineated in administrative procedures did not appear consistent with Generic letter Letter 82 82-12 w 12 intent. The inspectors noted that general housekeeping and contamination control had deteriorated during the Unit 2 refueling outage as compared to recent previous refueling outages. The licensee planned to implement a new material condition/housekeeping/safety inspection program in January 1991. (~ .. 2 III. 16-8 111.16-8

Revision 8 April 1992 ( Maintenar.ce/Surveillance Maintena~ce/Surveillance involving maintenance personel Two instances of failing to follow procedure involvin9 were noted. These involved maintenance practices on 10 CFR 50 Appendix R emergency lighting batteries and main steam line plug installation. However, both actually occurred prior to licensee corrective actions to address personnel performance problems delineated in inspection report 50-237/90023; 50-249/90023. A review of the Maintenance Department in regard to overtime policy indicated that instances of exceeding Generic Letter 82-12 guidelines was minimal. Engineering/Technical Support Subsequent to the Notice of Violation and Proposed Imposition of Civil Penalty dated November 28, 1990 involving a 10 CFR 50.59 violation, the inspector identified concerns which indicated additional poor past practices regarding the licensee's safety evaluation process. For example, a failure to maintain records of a written safety evaluation involving SGTS is identified as a non-cited violation. In addition, two unresolved items needed further review with regard to 10 CFR 50.59 requirements. These included a failure to obtain a technical specification change regarding a modification to the generator load reject scram function and the use of a temporary pump and hose assembly, without a safety evaluation, for the reactor cavity water filtering system. The root causes associated with the failure to post a proposed impostion of a civil penalty were repetitive to the cause of a previous violation involving a failure to ensure personnel were properly trained on specific administrative requirements. The licensee already had plans to address this concern with a new administrative requirement training program to be implemented in the spring of 1991. The inspectors noted that staffing of the plant Technical Staff had increased substantially. Staffing was regarded as a weakness in the last Systematic Assessment of Licensee Performance (SAL?) (SALP) period. The licensee did not apply the Generic letter Letter 82-12 guidelines on overtime to the plant Technical Staff. Instances were identified where these guidelines were exceeded, most notably during the refueling outage with the inservice testing/inservice inspection group. No problems were noted during non-refueling outage periods. Safety Assessment/Quality Verification The inspectors noted that the rate of events indicative of personnel performance problems decreased substantially during the second half of the Unit 2 refueling outage as a result of licensee management actions delineated in inspection report 50-237/90023; 50-249/90023. c 33 111.16-9

Revision 8 Aoril 1992 April ( DETAILS DETAr LS

1. Persons Contacted Commonwealth Edison Company

        *E. Eenigenburg, Station Manager
        *L. Gerner, Technical Superintendent E. Mantel, Services Director
        *0.
        *D. Van Pelt, Assistant Superintendent - Maintenance
        *J. Kotowski, Production Superintendent J. Achterberg, Assistant Superintendent - Work Planning
        *G. Smith, Assistant Superintendent-Operations
        *K. Peterman, Regulatory Assurance Supervisor M. Korchynsky, Operating Engineer B. Zank, Operating Engineer J. Williams, Operating Engineer R. Stobert, Operating Engineer M. Strait, Technical Staff Supervisor L. Johnson, Q.C. Supervisor J. Mayer, Station Security Administrator D. Morey, Chemistry Services Supervisor D. Saccomando, Health Physics Services Supervisor K. Kociuba, Quality Assurance Superintendent
        *0. lowenstein, Lowenstein, Regulatory Assurance Analyst

('( *J.

       .*J. Harrington, Nuclear Quality Programs Inspector
        *G. Kusnik, Quality Control Inspector
        *0. Booth, Master Electrician
        *C. Oshier, Lead Health Physicist
        *R. Whalen, Assistant Technical Staff Supervisor
        *0. Gulati, Master Instrument Mechanic The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel, and contract security personnel.
      *Denotes those attending one or more exit interviews conducted informally at various times throughout the inspection period.

!( 4 III.16-10 IIL16-10

Revision 8 April1 1992 Apri ( 2. Previously Identified Inspection Items (92701 and 92702) (Closed) Violation (50-237/89022-02): A penetration in a three hour fire I~re . rated wall w~l~ of ,the reactor building was not included in design documents documents, and modifications modlflcatlons were not controlled as required by the licensee's firefir~ protection plan which was implemented in accordance with 10 eFR 50.48{a). 50.48(a). The inspector performed visual observation and reviewed documentation to verify that appropriate corrective actions were implemented implemented.*.,The The inspector has no other concerns in this area. (Closed) Open Item (50-237/90009-02): The inspector visually verified that the problem related to legibility of the medium range drywel1 drywell pressure strip chart recorder indicator scale had been corrected. The inspector has no other concerns in ~histhis area. (Closed) Unresolved Item (50-237/90023-04): Concern regarding the discovery of damage to the Unit 2, SRM 22, following the suspension of core alterations on November 12, 1990. The damage resulted when the SRM was withdrawn during an instrument response check and the drive mechanism came in contact with an.an, under vessel platform access hole cover. The contact with the access cover resulted in the SRM becoming dislodged approximately three feet below the fully inserted position. Subsequent to the damage, fourteen fuel bundles were loaded into the SRM 22 core quadrant. Technical Specification 3.10.8. 3.10.B. required SRM 22 to be operable and fully inserted to the normal operating level in the core during core alterations in that quadrant of the reactor vessel. Review of licensee fuel handling records revealed that SRM 22 did indicate the expected neutron response during fuel movement while the instrument was in a degraded condition. On November 6, 1990, work was completed under the Unit 2 reactor to install new SRM probe connectors per Work Request (WR)*95435. (WR)'95435. During the probe replacement, platform access hole covers were utilized to minimize the potential for personnel injury during the under vessel work. The work instructions accompanying WR 95435 did not address or control the use of the platform access covers. However, a hand written memorandum was issued to all instrument maintenance (1M) supervisors requiring that the access covers be removed prior to the withdrawal of any of the SRMs for the performance of the instrument response check. Additionally. Additionally, a caution tag was placed in the control room instructing the operator not to withdraw the SRMs without first receiving permission from the 1M supervisor. This same methodology was utilized without incident during the previous Unit 3 refueling outage. On November 12. 12, 1990. 1990, operations personnel received erroneous permission from the 1M supervisor prior to the withdrawal and subsequent damage to SRM 22. To prevent recurrence of this event. event, the licensee planned to revise the applicable procedures to control the use of platform access covers under the Out-Of-Service program. '( '( 5 IIF.16-11

Revision 8 April 1992 The failure to provide adequate measures to prevent inadvertent operation of the SRM drives in relation to the status of the platform covers is considered to be a violation (50-237/90027-01(DRP)) of 10 CFR 50, Appendix 8, B, Criterion XIV. However, the criteria of 10 CFR 2, Appendix C, Section V.G.1, V.G.l, for discretionary enforcement was determined to be applicable and, therefore, no notice of violation is being issued. The inspector has no further concerns in this area. (Closed) Unresolved Item (50-237/89013-02{DRS)); 50-249/89012~02(DRS)): (50-237/89013-02(DRS)); 50-249/89012-02(DRS)): The licensee indicated that the fire fighting foam concentrate shelf life would be verified and, if testing ;s is required, it would be scheduled. According to the licensee action item report (Item Number 237-100-89-01302) the licensee had replaced the foam and had initiated Dresden Fire Protection Procedure (OFPP) (DFPP) 4114-07, "Annual Fire Fighting Foam Sampling." The inspector's review of the procedure found it to be acceptable. Based on the licensee's actions, this item is considered closed. . (Closed) Unresolved Item (50-237/88010-03(DRP); SO-249/88012-03(DRS)): 50-249/88012-03(DRS)): The simultaneous spurious opening of the Target Rock Valve and Electromatic Relief Valves has a tremendous impact in reactor coolant inventory based on the limited capacity of the Control Rod Drive (eRD) (CRD) Hydraulic System to restore or maintain reactor coolant inventory. Due to the significance of this issue and its generic implications, this issue was referred to the Office of Nuclear Reactor Regulation (NRR) for resolution. NRR Safety Evaluation forwarded by letter dated July 6, 1989, from 8.B. Siegel, NRC, to T. Kovach, CECo, accepted the licensee's modification to install two new control cables in a separate tray to rectify the potential that existed for fire induced multiconductor cable fault ;n in two control cables associated with Unit 3 Target Rock Valve and Electromatic Relief Valves. The licensee provided the inspector with modification close out form (Number M12-3-88-24) MI2-3-88-24) that indicated that the work of installing two new control cables in a separate tray was completed. This item is considered closed. (Closed) Unresolved Item (50-237/90023-06(DRP)): (50-237/90023-06(DRP}): The inspectors identified six Appendix "RuHR" emergency lights with the electrolyte level below the add line. Dresden Electrical Surveillance (DES) 4153-02,

 'tEmergency "Emergency Lighting Monthly Inspection", stated that "Electrolyte level shall sha               full 11 be at the fu     line".

11 1i However, contrary to the estab ne"

• However t established 1i shed procedure, the licensee indicated that a practice had been followed such that the emergency lights' need only be filled when the electrolyte level was at or below the add line. The licensee further indicated that also contrary to the established procedure, the determination to add distilled water was at the discretion of the maintenance

                                          ~aintenance personnel. Conversations with the emergency light vendor and review of the vendor technical manual indicated that allowing the electrolyte level to fall below the add line could cause damage to the battery.

c 6 III .16-12 III.16-12

Revision 8 April 1992 ( Licensee follow-up of the October 19, 1990, inspector observations identified extremely low or empty electrolyte levels in the following emergency tights: Emergency li2ht Light Number Electrollte Electrol~te Level (1) 271 Empty (2)) (2 274 Empty (3) 275 Empty (4) 299A Empty (5) 352 Extremely Low The licensee's response to this unresolved item dated December 14, 1990, from T. J. Kovach, (CECo), to A. B. Davis, (NRC), contained a deviation report (12-2/3-90-123) which included an event summary. summary, root causes and corrective actions (which included replacing several emergency lights) for Units 2 and 3. Based on our review of this report and supporting documentation, it was determined that due to the number of emergency lights observed with extremely low or no electrolyte, and the lack of adherence to the emergency lighting inspection procedure that this unresolved item has been upgraded to a violation (SO-237/90027-02(ORP}) (50-237/90027-02(DRP)) of Technical Specification Section 6.2. Based on prompt and thorough action to prevent recurrence and commitments to revise the emergency lighting procedure, no response to this violation is required and the NRC has no further questions regarding this matter.

• c.

( Administrative Closure of Items NRC Region III management reviewed the eXisting open items for the Dresden Station and determined that the following open items will be closed administratively due to their safety significance relative to emerging priority issues and to the age of the item. The licensee is reminded that commitments directly relating to these open items are the responsibility of the licensee and should be met as committed. NRC Region III will review licensee actions by periodically sampling administratively closed items. 50-237/84027-01 50-237/85003-BB 50-237/85003-88 50-237/87006-03 50-237/89022-01 50-237/B9022-01 50-249/85003-BB 50-249/85003-88 One cited and one non-cited violation and no deviations were identified in this area. 7 III.16-13 111.16-13

Revision 8 ADril 1992 April ( 3. licensee Event Reports (LER) Followup (90712 and 92700) Licensee Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications. lER (237/90006(DRP)): Target Rock Safety-Relief Valve Failed (Closed) LER correspondin~ corrective actions were discussed in Open. This event and correspondin9 50-237/90019(DRP); 50-249/90019(DRP). inspection report 50-237/90019(DRP)j 50-249/90019{DRP). (237/90007(DRP>>: Unplanned Primary Containment Group V (Closed) LER (237/90007(DRP)): Isolation. This event and corresponding corrective actions were discussed in inspection report 50-237/90019(DRP)j 50-237/90019(DRP)i SO-249/90019(DRP). 50-249/90019(DRP). (Closed) LER (237/90008(DRP)): Failure of HPCI Steam Line line High Flow Isolation Differential Pressure Transmitter. This event and correspondin9 correspondin~ corrective actions were discussed in inspection report SO-237/90019{DRP)i 50-249/90019(DRP). 50-237/90019(DRP)j (Closed) LER (249/90006(DRP)): (249/90006(DRP)}: Failure to Establish Appropriate Fire Protection Due to Procedure Deficiency. This event and corresponding corrective actions were discussed in inspection report 50-237/90017(DRP)j 50-237/90017(ORP); SO-249/90017(DRP). 50-249/90017(DRP). ( (Closed) LER (237/90012(DRP>>: (237/90012(DRP)): Fuel Load Core Monitoring Requirements Violated Due to Management Deficiency. This event and corresponding corrective actions are discussed in paragraph 2 of this report. No violations or deviations were identified in this area.

4. Plant Operations (61715, 71707 and 93702)

The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control rOomroom operators during this period. The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of Units 2 and 3 reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspectors noted that general housekeeping and contamination control had deteriorated during the Unit 2 refueling outag~ as compared to recent previous refueling outages. The licensee planned to implement a new material condition/housekeeping/safety inspection program in January 1991. ( 8 111.16-14 II 1.16-14

Revision 8 Aor;l ADril 1992 The inspectors identified that a licensee Quality Control inspector* inspector failed to don the required protective clothing while performing a hold-point inspection in a contaminated area contrary to the requirements of Dresden Administrative Procedure (OAP)(DAP) 12-25, "Radiation Work Permit Process," Steps E.6 and F.i.e.5, F.1.e.5, and Radiation Work Permit (RWP) OG067A, on December 3. 3, 1990. The individual subsequently became contaminated and alarmed the Personnel Contamination Monitors (PCMs) when attempting to leave the Radiological Controlled Area (RCA). Following rece~pt receipt of the PCM PC~l alarm, the individual failed to contact the Radiation Protection (RPD), per the requirements of DAP 12-13, "Personal Department (RPO). 'Personal External F.8.f, and proceeded to perform self Contamination Surveys," Step F.S.f, decontamination. Failing to follow the requirements of OAPs DAPs 12-13 and 12-25 in regard to radiation protection practices is considered to be a violation (50-237/90027-03(DRP)) of Technical Specification 6.2.B6.2.8 which required adherence to radiation control procedures. Following a discussion of the incident with the inspectors on December 4, 1990, the individual then notified the RPD. The RPD documented the incident and completed the appropriate corrective action of counselling the individual involved as to the proper health physics practices at the Dresden Station. As this was considered to be an isolated occurrence, of minimal safety significance and the appropriate corrective action had been completed, a Notice of Violation is not being issued in accordance with 10 CFR 2, Appendix C, Section V.A. The inspector has no further concerns in this area. Each week during routine activities or tours, the inspector monitored the ( licensee's security program to ensure that observed actions were being implemented according to their approved security plan. The inspector noted that persons within the protected area displayed proper photo-identification badges and those individuals requiring escorts were properly escorted. The inspector also verified that checked vital areas were locked and alarmed. Additionally, the inspector also verified that observed personnel and packages entering the protected area were searched by appropriate equipment or by hand. The inspectors performed a detailed walkdown of the accessible portions of the Unit 2 containment spray system (ess), (CSS), which is a subsystem of the (LPCI) system. The inspectors concluded low pressure coolant injection (lPCI) that the ess CSS was properly aligned and in adequate condition. The inspectors verified the proper positioning of numerous isolation valves and electrical barriers in containment penetrations. In addition, the inspectors performed a walkdown of the drywell with licensee personnel to ensure that material and equipment utilized uti.1ized during the refueling outage was secured

• properly removed or secured.

-( . ( 9 111.16-15 IlLIG-IS

Revision 8 April 1992 ( The inspectors reviewed selected new procedures and changes to procedures that were implemented during the inspection period. The review consisted of a verification for accuracy, correctness, and compliance with regulatory requirements. . The inspectors verified that the licensee had implemented controls to assure guidelines presented in Generic Letter 82-12, "Nuclear Power Plant Staff Working Hours" were followed for 1licensed icensed operators. These restrictions were delineated in DAP 7-21, "Station Policy on Reactor Operator and Senior Reactor Operator Manning Levels and Overtime," Revision 1, I, which conformed to the Generic Letter. The licensee had also extended these guidelines to Level Levell1 equipment operators in accordance with the licensee's Nuclear Operations Directive (NOD) OA.13, "Overtime Guidelines." The licensee did not apply these guidelines to Level 2 equipment operators because they were not involved in safety related work. Although the licensee indicated the guidelines were applied to the fuel handling supervisors, the fuel handlers themselves were not covered. The licensee had developed an in-house computer program which operations engineering assistants utilized to track hours worked and to assure the guidelines were met. The licensee identified four cases during the year guidelines had been exceeded. Only one of of 1990 where these overtime gUidelines these had been pre-approved by management in accordance with DAP 7-21. The others involved administrative errors, one of which occurred subsequent to utilization of the computer tracking program implemented

/   tp correct these errors and resulted from a failure of operations

(". engineering aSSistants assistants to promptly enter hours into the computer. Licensee corrective action involved counseling of the engineering assistants on the event to ensure prompt data entry. There had been no identified occurrences since that corrective action. The inspectors noted that DAP 7-21 required pre-approval to exceed the overtime guidelines by the Assistant Superintendent Operations or his designee. The licensee indicated that this designee was the Operating Engineer in charge of personnel. However, Gener;c Generic Letter 82-12 indicated that such deviation be authorized by the plant manager or his deputy or higher levels of management. The Station Manager was required to review and Sign the Overtime Deviation Authorization form only subsequent to the overtime being worked. The intent was that only Senior Management be able to authorize major deviations from the overtime guidelines. At Dresden the pre-authorization prescribed in the administrative procedures was from two supervisory levels below the p1ant plant manager with a possibility for a designee at three levels below the plant ~nager. Therefore, this practice did not appear to meet the intent of the generic letter. However, the one instance where pre-approval was given in 1990 was by the Production Superintendent, one level below the plant manager, which appeared consistent with the Generic Letter 82-12 guidelines. 10 II 1.16-16

Revision 8 April 1992 ( Inspector concerns resulting from this review and comparison to licensee commitments with respect to overtime guidelines are discussed in paragraph 7.b. l.b. Various operational occurrences were also reviewed as follows:

a. On November 23, 1990, with Unit 2 shutdown in a refueling outage, a scram was received due to noise spikes on Intermediate Range Monitors (IRM) 13 and 15. All SRMs and IRMs actually received spikes. At the time, one control rod was partially withdrawn for

       ~estin£.
       ~est;n£. T~E cause We' Wc~ 'dent'-;ed                    spike due to a
                                 ;dent~;iec t~ be a voltage spi~e faulty relay in the control logic for the Lowlow Pressure Coolant Injection (LPCI) system. The relay was replaced. A similar occurrence with IRMs spiking high occurred on December 20, 1990.

The licensee has continued to investigate the cause of the spiking problems.

b. On November 27 1990, Unit 3 entered a 24 hour limiting condition for operating (LCD)

(LCO) in accordance with Technical Specification 3.7.0.3 due to a failure of Nitrogen Makeup Valve AO-1601-59. During the nitrogen makeup surveillance, this valve failed to close completely. This was of concern since this valve also served as a containment isolation valve. Techn1cal Technlcal Specifications would have allowed closing other valves upstream of this line to provide the isolation function such that the 24 hour LCD LCO would not have been entered. However, one of these valves, AO-1601-58, was in the pumpback system which provided drywell/torus differential pressure control. Since closing this valve would have caused difficulties in maintaining the required differential pressure, it was left open and the 24 hour LCO LCD was entered. However, repairs were completed prior to actually initiating a shutdown.

c. During a plant walkdown on ,December December 7,1990, 7, 1990, the inspectors noted that an access door to the 'Unit Unit 2 drywell purge and ventilation system downstream of ventilation fan 2-5708A, 2-570BA, was open about one to two inches. As this formed a portion of the secondary containment isolation boundary, the concern was that this provided approximately a 72 square inch hole from the turbine building through this boundary. As described in paragraph 6.b., although Unit 2 was in a refueling outage, the current SGrS SGTS lineup would cause suction to be drawn from this area even if the actuation was on Unit 3. It was not clear 'whether whether this breach was large enough to prevent fulfillment of SGTS function (i.e., the ability to pull a 0.25 inch differential pressure to the atmosphere in secondary containment.)

The licensee ran the system with the access door in the as-found ,( 11 III.IS-17 III .16-17

Revision 8 April 1992 ( condition to ensure that had the SGTS been called upon, the suction created in the ductwork would have pulled the access door shut. As such, the inspectors ascertained that the operability of SGTS was not affected by the open access door. No apparent cause could be ascertained by the licensee. The inspectors also noted on Oecember December 7, 1990, that the Service Air Supply to three of the Unit 3 drywell purge and ventilation fan dampers had been disconnected. No temporary alteration tags were attached to the air lines or to the dampers' operators. Following notification to the licensee, the licensee reconnected the airlines to the dampers which changed position from open to closed. (The system was not in operation at the time.) A review of Dresden Operating Procedure (OOP) (DOP) 6600-1, "Normal Venting of Orywell Drywell and Torus", Revision 5, indicated that it required disconnection of the air operators on the drywell purge fan inlet and outlet dampers and blocking or tieing the dampers in the open position prior to the operation. This was necessitated since the dampers automatically opened and closed in conjunction with fan operation and the fan was not actually operated in this procedure. A step was also included to reconnect the air operator and unblock or untie the dampers when the operation was complete. A review of the operating history of this system indicated that it was last used for venting Unit 3 on December 5, 1990, for containment pressure control. This should have been performed in accordance with DOP 6600-1. I (

  ,    However, it was questionable whether step F.1.b(2) was followed in that the air supply was found disconnected. This is considered to be an unresolved item (50-237/90027-04(DRP)) pending further review of operator involvement and safety significance.

d. On December 8, 1990, while Unit 2 was in the refuel mode, eight Group II automatic primary containment isolation valves closed following the lifting of a field wire on a main control room terminal block. The lead was lifted to facilitate a resistance and meggering check of the Main Steam Isolation Valve (MSIV) pilot solenoid coils, per DES 200-39, "Main Steam Isolation Valve Electrical Maintenance." Further review indicated that an interruption of multiple neutral ground circuits occurred when the lead was lifted. This resulted in a loss of power to the associated seal-in relays, which maintained each of the affected Group II isolation valves in their open positions. Electrical wiring diagram 12E2697 indicated that the neutral ground circuit was designed to be wired, in daisy chain fashion.

fashion, on the cabinet side of the respective terminal block. The drawing also showed four leads terminated on the cabinet side of the effected terminal point. 12 IIL16-1B IlI.16-18

Revision 8 April 1992 c (

\  However, each terminal block point was physically accommodate a maxiw.u~

maxi~u~ of three leads. phYSically limited to In apparent compensation for this design deficiency. deficiency, one of the four wires. wires, the neutral ground circuit for the Group II valves, was placed on the field side of the terminal block, sharing the same connection point as the MSIV pilot solenoid coils. DES 200-39 matched the configuration as described on Diagram 12E2697. When DES 200-39 was performed, the electrician found two leads on the field side terminal pOint. The lifting of the second lead resulted in the valve closure. The apparent root cause of the event was a discrepancy between the plant configuration and the "as-built" drawings. A review of past revisions to Diagram 12E2697 revealed that the original plant design configuration also specified the termination of four leads on the cabinet side of the terminal block point. Based on the physical limitations of the cabinet side terminal block point, the undocumented wiring configuration was estimated to have existed since initial plant startup. This is considered an unresolved item (50-237/90027-05{DRP)) (50-237/90027-05(DRP)) pending NRC review of licensee corrective actions. The licensee failed to recognize that closure of the Unit 2 Group II containment isolation valves was an ESF actuation and also failed to make the four hour report to the NRC as required per 10 CFR 50.72. The licensee indicated that the rationale for not reporting was that the loss of power occurred in the control circuitry and not the ( logic circuitry. In other words, the logic circuitry did not de-energize to open a corresponding contact in the control circuitry to cause the closure. As such, the licensee did not classify this as an ESF actuation. The licensee based this distinction on NUREG 1022, "Licensee Event Report System,lI System," Section V, which stated the fo llowing: following: Actuation of multichannel ESF Actuation Systems is defined as actuation of enough channels to complete the minimum actuation logic (i.e. activation of sufficient channels to cause activation of the ESF Actuation System). Therefore, single channel actuations, whether caused by failures or otherwise, are not reportable if they do not complete the minimum actuation logic. ( (-- 13 III.16-19 III .16-19

Revision 8 April 1992 ( It was evident that the licensee had inferred a meaning that was not intended nor supported by the other portions of NUREG 1022 or its* supplements. The paragraph noted by the licensee was intended to specifically explain the non-reportability of single channel actuations versus multichannel actuations. No further meaning can be inferred and, in fact, this paragraph made no attempt to define circuitry as separate and distinct from the logic portion of the Circuitry other portions of the actuation circuitry (i.e. control cjrcuitry). c.ircuitry). Section IV of NUREG 1022 provided a restatement of the guidance for' 10 CFR 50.73 from the statement of consideration which indicated that the criteria were based on the nature, course and consequences of the event and not on initiating events or causes of events. In addition, Section V of NUREG 1022 indicated that the NRC was interested in both events where an ESF was needed and events where an ESF operated unnecessarily since they should not be challenged frequently or unnecessarily. 10 CFR 50.72(b)(2)(ii) specifically required the reporting of any condition that results in a manual or automatic actuation of any Engineering Safety Feature (ESF). Additionally, NUREG-I022, "Licensee Event Report System," Supplement No. I, Section 11.6, clarified that an ESF actuation includes any automatic, spurious, or manual action that r~sults results in the actuation of the device to perform its intended function. In the case of the Group II isolation valves, the intended ESF safety function was the automatic closure of the valves. All ESF actuations were required to be reported (except those expected actuations that result from and were part of preplanned sequence during testing). The failure to make the required report was considered to be a violation (50-237/90027-06(DRP)) of 10 CFR 50.72(b)(2){ii). 50.72(b)(2)(ii).

e. Following receipt of upper motor guide bearing high temperature and high vibration alarms on recirculation pump 38 on December 15,IS, 1990, both recirculation pumps were reduced to minimum speed.

Recirculation pump 38 was shutdown and it's corresponding suction valve was closed in accordance with Technical Specifications. Total power reduction during the event was from about 95 to 25 percent rated thermal power. The unit remained in single loop operation at drywe11 the end of the inspection period. On December 22, 1990 a drywell entry was made to visually examine the recirculation pump motor, take oil samples and check a pump motor vibration switch. During the subsequent local leak rate test (llRT) (LLRT) on the drywell drywe1l personnel interlock doors on December 23, 1990, the inner door seal failed. Unit 2 was shut down, the seal was repaired and Unit 2 was returned* returned' to service on December 26, 1990.

(

\ '- .. 14 III.16-20 II I.16-20

Revision Revision 88 April April 1992 1992 ( f.

f. In In the the course course of of observing observing refueling refueling operations operations on on November November 18 18 1990, the 1990, the inspector inspector noted that that aa temporary pump and hos~

temporary vacuum pump hos~ assembly assembly were were utilized utilized toto augment augment thethe filtering filtering capability capability of of the the fuel fuel pool pool clean-up clean-up system. system. This This temporary temporary vacuum vacuum pump was situated on the top guide of the reactor vessel and pumped refueling water through a hose which exited the drywell cavity, ran several feet across the refueling floor, entered the refueling pool, and led to fue 1 pool the fuel poo 1 skimmer surge tank. Various concerns were identified identified including; (1) whether the use of the pump was controlled by procedures or as part of a temporary alteration/modification program, (2) whether consideration was given to the possibility of rupturing the vacuum hose and lowering the level in the refueling pool, and (3) the availability of indicators and alarms for the refuelin9 pool level. This is considered an unresolved item (50-237/90027-07(ORP)) (50-237/90027-07(DRP)) pending regard to these concerns. further review in reg~rd One cited and one non-cited violation and no deviations were identified in this area.

5. Maintenance and Surveillances (62703, 61726. 61726, and 93702)

a. Maintenance Activities Station maintenance activities of systems and components listed below were observed or reviewed to ascertain that they were

 /

conducted in accordance with approved procedures, regulatory guides . ~.. and industry codes or standards and in conformance with Technical Specifications. The following items were considered during this review: The limiting Conditions for Operation (lCOs) (leOs) were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemented. Work requests were reviewed to determine status of outstanding jobs and to assure that priority ;s is assigned to safety-related equipment maintenance which may affect system performance. ( 15 16-21 IIR16-21 Ill'.

Revision 8 ADril 1992 Acril ( Environmental Qualification Preventive Maintenance on the Unit 2 High Radiation Sample Target Rock Solenoid Valves. Environmental Qualification Preventive Maintenance on the Unit 2 Main Steam Isolation Solenoid Valves. Unit 2 SRM Cable Routing. Plant Process Pipe Labeling. Calibration of the Unit 2 Turbine Control Valve Pressure Switches. On November 21, 1990, during performance of an LLRT, clean de-mineralized water (COW) was secured to the Unit 2 drywell. This caused a plug in Main Steamline (MSL) "A" to deflate and provide a drainpath from the reactor cavity. (The reactor cavity was flooded above the main steamlines at the time.) As the corresponding inboard MSIV had been removed for maintenance, leakage entered the drywel1. drywell. This was discovered while investigating the receipt of multiple drywell sump level alarms in a short period. Up to one inch of cavity level was lost during the event. Further review indicated the cause to be a failure to follow procedure during MSL plug installation on October 13, 1990. The work package for this activity prescribed installation in ( accordance with Dresden Maintenance Procedure (OMP) 200-31, "MSL Plug Installation and RemovaJ." The procedure prescribed inflation of the seal with service air. However, upon completion it was noted that air leakage caused bubble to form in the reactor cavity which obscured fuel handler vision. The maintenance supervisor, having COW had in the past been used to inflate the been reminded that CDW seals, had the source switched to COW, contrary to procedure requirements and without informing appropriate management. As outage planning was predicated on the assumption that the inflation source was in accordance with the procedure, the LLRT was allowed to commence on the COW line. In fact, the LLRT on the Service Air Line had been purposely postponed so as to not affect the seals. Failing to follow the procedure is considered to be a violation (50-237/90027-08(ORP)) of 10 eFR 50, Appendix a, CFR 50. B, Y. However, this event was indicative of. Criterion V. of, and in the same time frame as the types of problems encountered during the first part of the current Unit 2 refueling outage. The actual failure to follow procedure occurred prior to the corrective actions taken by plant management as described in inspection report 50-237/90023; 50-249/90023 to address these problems. AS'such, this event, had it been discovered in the previous inspection period. period, would have been ( 16 III.16-22

Revision 8 Aoril 1992 April ( included as an additional example in the corresponding notice of. already :~ violation issued with that report. As the licensee had already" taken appropriate corrective actions to address this type of concern, a Notice of Violation is not being issued in accordance with exercise of discretion delineated in 10 CFR 2, Appendix C, V.G.!. Section V.G.l.

b. Surveillance Activities The inspectors observed surveillance testing, including required Technical Specification surveillance testing, and verified for actual activities observed that testing was performed in accordance with adequate procedures. The inspectors also verified that test instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were accomplished and that test results conformed with Technical Specification and procedure requirements. Additionally, the inspectors ensured that the test results were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.

The inspectors witnessed or reviewed portions of the following test activities: Nuclear Instrumentation Surveillance ( Reactor Vessel Leakage (Hydrostatic) Test Standby Liquid Contro'l Contro-l Inservice Testing Control Rod Drive Friction Testing Single Reactor Recirculation Pump Operation Surveillance Control Rod Drive Insertion Timing On December 17,1990, 17, 1990, the Unit 2 torus to reactor building vacuum breaker (2-AO-1601-20A) was discovered to be leaking excessively into the torus basement area following the pressurization of the drywell drywel1 to 13 pounds per square inch (psig) during the performance of the ILRT. Maintenance personnel were dispatched and tightened the bolts on the inboard vacuum breaker flange. This effort resulted in the sealing of the flange mating surface and the drywell was subsequently pressurized to 48 psig. The primary containment leakage rate through the degraded flange could not be quantified. However, the leakage rate was estimated to be well in excess of the Technical Specification limit of 1.6 percent, by weight, of the air, per 24 hours, at 48 psig. A review of past containment ai.r, maintenance activities indicated the vacuum breaker was replaced during the last refueling outage. Further evaluation of the work package identified that the licensee failed to perform an ILRT or LLRT on the effected containment volume following the replacement of the vacuum breaker. This issue is considered to be an unresolved item (50-237/90027-09(DRP)). pending review of specific containment leakage testing requirements. c 17 III.16-23 III .16-23

Revision 8 ADril 1992 Aoril (

~

I

\

c. The inspectors reviewed maintenance department overtime practices in consideration of Generic Letter 82-12 and 83-14 guidelines.

Maintenance Department Memorandum (MOM) No. 65 prescribed application of the guidelines to appropriate personnel. However the MOM did not prescribe a break of at least eight hours betwee~ work periods as indicated in Generic Letter 82-12 and NOD OA.13. No actual occurrences were noted, however, where this had been exceeded. Actual implementation of MOM No. 65, began in rnid-1990. mid-1990. The methodology utilized for tracking and pre-checking was dependent on the specific maintenance master. During the inspection, the maintenance masters were provided with printouts from the payroll computer system which had been programmed to indicate personnel who had surpassed the guidelines. This was to be utilized to prepare the overtime semi-annual report to the Vice President of BWR Operations. Through this printout, the licensee identified numerous unexpected instances of electrical maintenance exceeding the 72 hours in seven day guidelines. This was because that group had been applying that guideline based on a fixed week and not on a rolling during any seven day period. The other groups had applied this correctly. As a result, the electrical maintenance group changed to a rolling seven day period and the distribution of the computer printout was changed to two week intervals. Although in some cases, tracking methodology utilized was not very formal, this did not appear to be a problem except as noted above. The total ,( number of deviations was minimal; however, additional deviations

  ~.         were expected the last week of the Unit 2 refueling outage.

Inspector concerns resulting from this review and comparison to licensee commitments with respect to overtime guidelines are discussed in paragraph 7.b. One non-cited violation and no deviations were identified in this area.

6. Engineering/Technical Support (71707 and 93702)

a. The inspectors reviewed the guidelines in Generic letter Letter 82-12 in comparison to overtime worked by the plant technical staff. The licensee indicated that the plant technical staff was not covered Letter 82-12 and, therefore, the guidelines were not under Generic letter applied to this group. The Technical Staff Supervisor had issued a memorandum on October 6, 1990, that delineated corporate direction that Technical Staff personnel working on safety related work.

work, not 18 IIL16-24

Revision 8 ADril 1992 ( work more than 18 hours continuously. This was in response to a recent event at Braidwood in which technical staff personnel worked excessive hours. The memorandum emphasized fitness for duty responsibility. The licensee had taken measures to minimize extended work hours at the start of the current Unit 2 refueling outage by developing a planned schedule for the technical staff major testing evolutions based on 12 to 13 hour work days and adding backshifts to spread out the work. The licensee personnel to bacKshifts identified two instances during the current outage in which personnel exceeded 18 continuous work hours. One was for 20 hours with prior Technical Staff Supervisor approval and in the other instance the individual was sent home just after 18 hours. The licensee did not have a formal tracking mechanism to assure the self-imposed 18 hour guideline was not exceeded, but instead informally relied upon the individual and the Technical Staff group leaders. The inspector reviewed a sample of the licensee's daily attendance report for Technical Staff personnel for the month of November 1990, in which a refueling outage was conducted. The format was primarily for pay purposes only, such that all instances of exceeding guidelines would not necessarily be identifiable. However, it appeared that several instances of exceeding Generic letter Letter 82-12 guidelines had occurred with some of this involving types of work applicable to the Generic Letter. A similar sample review for October 1990, a non-outage month did not identify any such ( instances. In addition, these instances appeared to be heavily dependent on the group within the Technical Staff, most notably with. the inservice inspection!inservice inspection/inservice testing group. A staffing increase from 67 to 85 individuals between May and November 1990, indicated a concerted effort to increase the size of the plant technical staff. (The large workload was considered a weakness in Licensee Performance (SALP) the last Systematic Assessment of licensee period.) In the same time period, the normal attrition of experienced personnel was about equal to the hiring of experienced personnel. Therefore, the net effect of the increased staffing was the addition of recent college graduates with little or no experience. However, these individuals would be expected to gain experience as time progressed and to alleviate the total workload. Inspector concerns resulting from this review and comparison to licensee commitments with respect to overtime guidelines are discussed in paragraph 7.b. ( 19 III. 16-25

Revision 8 Acril 1992 Aoril ( b. On November 30, 1990, the licensee identified a problem with the SGTS lineup which could potentially allow flow to bypass the SGTS. Among other locations, the SGTS took a suction from each unit's reactor building ventilation system (RBVS). In order to address a problem identified in the mid-19l0s, mid-1970s, the licensee had tagged the reactor building ventilation to SGTS isolation valves open with their corresponding breakers racked out such that they would not close on an automatic signal. The original design called for the isolation valve on the unaffected unit to automatically close on a SGTS automatic start Signal. signal. This was to prevent flow from the affected unit's RBVS from entering the still normally operating RBVS on the unaffected unit, which would exit to the environs without passing through the SGTS. The licensee was concerned that following an initiation signal on one unit and corresponding closure of the opposite unit's RBVS to SGTS isolation valve, an automatic start signal on the other unit would then result in total isolation of the SGTS. Therefore. Therefore, a single failure could disable the SGTS. The licensee could find no record of a 10 CFR 50.59 safety evaluation for the racked out isolation valve breakers. The licensee had planned on modifications to the logic to address the original concerns but the modifications were cancelled prior to implementation. The licensee indicated that justification was that operator actions could be taken to initiate reactor building ventilation isolation on the unaffected unit when an automatic signal was received on the opposite unit. This was reflected in procedures such that the opposite unit's RBVS would be manually lined up to the SGTS on a single unit initiation signal. While reviewing applicability of motor operated valve testing to these valves on November 30, 3D, 1990, licensee personnel reviewed these previous actions and Questioned questioned whether adequate justification had been utilized. The licensee conducted an offsite dose analysis assuming no operator actions with both isolation valves remaining open. These results indicated doses well below 10 CFR 100 guidelines. The licensee planned to develop a modification that would cause both RBVS to automatically lineup to the SGTS on a single unit initiation. The licensee also planned to perform a review of existing out-of-services to ensure that similar long standing cases of system changes through out-of-service did not exist. As indicated in enforcement conference report 50-237/90023; 50-249/90023, the licensee also was developing reforms to the 10 CFR 50.59 safety evaluation process at the facility. 20 1.16-26 III.16-26 II

Revi si on 8 Revision ADril 1992 ( Although the facility change was performed through an out~of-service, the nature and long term existence clearly indicated the necessity of performing a 10 eFR CFR 50.59 safety evaluation to ensure that an unreviewed safety question did not exist. In addition, the inspectors unrev;ewed considered the justification for cancelling modifications to rectify the various concerns to be inadequate in that an appropriate -basis was not provided as to utilization of manual operator actions in place of the automatic function. Failure to maintain records, which include a written safety evaluation of this facility change, is considered to be a violation (50-237/90027-10(DRP)} (50-237/90027-10(DRP)) of 10 CFR SO.59(b)(I). 50.59(b)(I). However, as this issue was licensee identified, an unreviewed safety exist, appropriate corrective question by definition did not actually exist. actions were initiated or planned and due to the age of the initial change, a notice of violation is not being issued in accordance with exercise of discretion delineated in 10 eFRCFR 2. 2, Appendix e, C, Section V.G.!. V.G.1.

c. On December 19 19, 1990 t 1990, the licensee informed the resident inspectors t

of a problem in regard to Technical Specifications not reflecting a previous modification. The reactor protection system (RPS) scram on generator load reject was modified in 1983 as a result of a GE Technical Information letter Letter (TIL) recommendation. The limit switches on the control valve fast acting solenoids which provided the scram signal were replaced by pressure switches which actuated on low Electrohydraulic Control (EHC) oil pressure at each control valve. Technical Specification Table 4.1.1 required a functional ( check surveillance to be performed on this scram function. However, Technical Specification Table 4.1.2 did not require a corresponding calibration check to be performed since it actually was written for the old design. Technical Specification Basis indicated that this was. was, in fact, an lIon/off" "on/off" type switch for which calibration was not applicable. However, the Il"new new design" pressure switches can be calibrated. Since it was not required, the licensee was unsure whether calibration"of these switches had ever been performed. Since the licensee had performed required functional checks but not necessarily calibration checks on this scram function, its operability was questionable. The generator load reject scram was a limiting safety system setting which anticipated the rapid increase in pressure and neutron flux from a control valve closure due to load rejection coincident with a failure of the bypass valves and corresponding Minimum Critical Power Ratio (MCPR) considerations. This was not an immediate problem since Technical Specifications allowed this scram function to be bypassed at less than 45 percent steam flow. Unit 2 was still shutdown for a refueling outage and ( 21 III. 16-27

Revision 8 Aoril 1992 Unit 3 was at less than 45 percent steam flow since it was in single ((\. loop operation due to recirculation pump motor problems. This is considered an unresolved item (50-237/90027-11(DRP) (50-237/90027-11(DRP)) pending further review of this modification with respect to 10 CFR 50.59 requirements. . No violations or deviations were identified in this area.

7. Safety Assessment/Quality Verification (40500)

a. The inspector observed the licenseets licensee's Start-Up On-Site Review Committee meeting held on December 11, II, 1990. These meetings were routinely held prior to start-up to review plant work activities accomplished during the refueling outage. The content and conduct of this meeting appeared to effectively contribute to the prevention of problems during start-up monitoring and evaluating the current plant status.

b. The inspector reviewed previous licensee/NRC correspondence to determine 1; censee cOll1!litments licensee commitments to Generi Genericc Letters 82-12, "Nuclear Power Plant Staff Working Hours", and 83-14 "Definitions of Key Maintenance Personnel." In a letter from T. J. Kovach (CECo) to A. B. Davis (NRC) dated October 4, 1989, the licensee responded to NRC concerns that CECo did not appear to have sufficient measures in place to ensure that safety-related work was not jeopardized by personnel having worked too many hours. The licensee committed to c( develop a new corporate Nuclear Operations Directive (NOD) that was to ensure uniform overtime policy governing safety-related work in accordance with the guidelines included in Generic letters Letters 82-12 and 83-14. This commitment indicated that the NOD would provide guidance applicability beyond the Technical Specification minimum shift crew composition and that included in this would be maintenance personnel and chemistry and radiation protection personnel. As such, the commitment was not clear as to what other groups beyond those specifically mentioned would also be included.

The commitment also indicated an appropriate level of management would be designated to assure that overtime was approved prior to the work occurring. The subject NOD, OA.13, was issued on March 15, 1990. ( 22 III .16-28

Revision 8 Aoril 1992 ( As indicated in paragraphs 4 and 6.a, NOD OA.13 did not extend the guidelines t~e fuel handlers or to technical staff personnel.~nd. guideline~ to the personnel,.nd, as such, inclusion lncluslon of these groups was not reflected in plant practice. As the fuel handlers performed safety related work in the movement of fuel assemblies and the technical staff performed in-plant safety-related work such as local leak rate testing and inservice inspections, the licensee's commitment would appear to apply to these groups. In addition, as described in paragraph.4, although the one pre-approved overtime deviation for operations during 1990 was from an appropriate level of management, that allowed by DAP 7-21 was not consistent with Generic letter 82-12 guidance. In addition, MOM No. 65 did not contain a requirement for a break of eight hours between shifts in accordance with Generic letter 82-12 and NOD OA.13. This is considered an unresolved item (50-237/90077-12(DRP)) pending completion of this inspection activity with respect to other plant organizations.

c. On November 30, 1990, the licensee received a Notice of Violation and Proposed Imposition of Civil Penalty associated with the use of a temporary sample pump in the drywell manifold sampling system. On December 17, 1990, the inspectors observed that the Civil Penalty had not been posted. This is considered to be a violation (50-237/90027-13(DRP)) of 10 eFR (50-237/90027-13(DRP>>) CFR 19.11 in that the Proposed Imposition of a Civil Penalty was not posted within the required two days of receipt. Discussions with station Regulatory Assurance personnel.

personnel, the group which was responsible for initiating the posting process .( 'c: 2-17, "Required Posting per Dresden Administrative Procedure (DAP) 2-17. of Documents", revealed some confusion existed over the posting requirements. The responsible supervisor believed it was not required to be posted until the station submitted their response to the Notice of Violation and Proposed Imposition of Civil Penalty. The licensee posted promptly following identification by the inspectors. The cause of the posting failure was related to the inadequate training of *Regulatory.

          'Regulatory Assurance group personnel of the requirements of DAP 2-17.

The problem of inadequate training of administrative requirements was identified as the root cause associated with a past violation of 10 eFR CFR 50, Appendix B, Criterion II, as delineated in inspection report 50-237/90023; 50-249/90023. As a result of the previous violation, the licensee was developing a program to matrix administrative training requirements with position descriptions with full implementation planned for the spring of 1991. The purposed training program was to ensure personnel were adequately trained on the administrative ( (. 23 III.16-29 rrI.16-29

Revision 8 ADril 1992 ((

  \           procedures they were required know to perform their specific duties.

As this was considered to be an isolated occurrence in regard to posting requirements and appropriate corrective actions had been formulated to address the root cause, a Notice of Violation is not being issued in accordance with 10 CFR 2, Appendix C, Section V.A. The inspectors have no further concerns in this area. One non-cited violation and no deviations were identified in this area.

8. Systematic Evaluation Program (SEP) Items NUREG 1403, "Safety Evaluation "Report related to the full-term operating Evaluation-Report Station* Table 2.1 identified SEP license for Dresden Nuclear Power Station" Integrity Plant Safety Assessment Report lIPSAR)

                                                   /IPSAR) topic resolutions to be report, confirmed by the NRC Region III office. Of the 22 items in that report.

eleven were indicated as already closed in previous inspection reports, leaving eleven remaining items to be closed. The intent is for the licensee to verify closed the remaining items with identification of the closing rationale to the NRC and a sample NRC inspection of these items to gain reasonable confidence in the licensee's information. Any items not yet closed would be identified to the NRC with anticipated closure dates. In that endeavor, the inspectors verified actual completion of the following items which the licensee indicated were closed. Item 18 - Topic VI-I0.B,2.12 VI-IO.B,2.12 (Supp.l) I Item 19 - Topic VI-IO.B,4.23.2 VI-I0.B,4.23.2 '~ Item 22 - Topic VIII-2, 4.26.2 Completion of this sample inspection as the licensee finishes the determination of the status of the remaining SEP items is considered an open item (50-237/90027-14(DRP)). No violations or deviations were identified in this area.

9. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether it is an acceptable item.

item, an open item, a deviation or a violation. Unresolved items disclosed during this inspection are discussed in paragraphs 4.c, 4.d, 4.f, S.b, 6.c and 7.b.

10. Open Items Open items are matters which have been discussed with the licensee which will be further reviewed by the inspector and which involved some actions on the part of the NRC or licensee or both. The one open item disclosed during the inspection is discussed in paragraph 8.

(( 24 111.16-30

• Revision 8B Aoril ADril 1992

( 11. Report Rev;ew Review During the inspection period, the inspector reviewed the licensee's Monthly Operating Report for October 1990. The inspector confirmed that the information provided met the requirements of Technical Specification 6.6.A.3 and Regulatory Guide 1.16.

12. Exit Interview The inspectors met with licensee representatives (denoted in Paragraph 1) 2B, 1990, and informally throughout the inspection period, on December 28, and summarized the scope and findings of the inspection activities.

The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any such documents/processes as proprietary. The licensee acknowledged the findings of the inspection. ( 25 111.16-31 11I.16-31

( f

      '\

e Commonwealth Edison 1400 Oous Place Downers Grove, IllinOIs illinoIs 60515 Revision 8 Aoril 1992 February 15. 15, 1991 Mr. A. Bert Davis Regional Administrator, Region III U.S. Nuclear Regulatory Comm\sslon Commission 799 Roosevelt Road Glen Ellyn, IL 60137

Subject:

- Dresden Nuclear Power Station Units 2 and 3 Response to Notice of Violation Associated with Inspection 50-237/90027; 50-237/90021; 50-249/90026 NRC Docket No. 50-237 and 50-249

Reference:

H.D. Shafer letter to Cordell Reed dated January 17, 1991 transmitting NRC Inspection Report 50-237/90027; 50-247/90026 Mr. Davis: Davl s: This letter provides the Commonwealth Edison Company (CECa) (CECo) response (attached) to the subject violation transmitted by the referenced NRC Inspection Report for Dresden Station. The violation involved Involved failure to report to the NRC an Engineered Safety Feature actuatton actuation tn In accordance with 10 CFR 50.72. If your staff has any Questions or comments concerning thts this letter, . please refer them to Rita Radtke, Compliance Engineer at 708/515-7284. Very truly yours. yours,

                                                                /.?/
                                                                 / P;'.:1  ovach Nuclear LI enslng Manager cc:       B.l. Siegel, Project Manager - HRR B.L. Siegel.                            NRR D.E. H111s, Hills, Senior Resident Inspector NRC Document Control Desk I mw RR: TK: 1

,( ZNLD742117 ZNlD742117

\_--

III.16-32

                                                                                    . 'r
     .                                                                              Revision 8 Acril 1992 Aoril NOTICE OF VIOLATION

( Commonwealth Edison Company Docket Nos. 50-237; 50-249 Dresden Nuclear Power Station License Nos. DPR-19; DPR-25 During an NRC Inspection inspection conducted on November 17 through December 29. 29, 1990. 1990, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions, "10 CFR Part 2. 2, Appendix C (1990), the violation is listed below: 10 CFR 50.72(b)(2)(11) 50.72(b)(2)(li) requires the NRC to be notified within four hours of the occurrence of any event or condition that results 1n In manual or automatic actuation of any Engineered Safety Feature (ESF). Contrary to the above, the unexpected closure of several Unit 2 Group II primary containment isolation valves upon lifting of an electrical lead during post-maintenance testing on December 8, 1990, constituted an automatic actuation of an ESF and the NRC was not notified of the occurrence. This is a severity level IV violation (Supplement I). DISCUSSION 'C. Dresden Station management carefully reviewed the circumstances surrounding the closure of several of the Group II isolation Isolation valves. The valve closures were not the result of a Group II signal, real or spurious. Neither channel of the Group II isolation Isolation logic circuitry actuated. The spurious closure of the valves was caused by de-energizing of the seal-in seal-In relays associated with several of the Group II valves. Page 13 of NUREG-l022 NUREG-1022,, "Ucensee "Licensee Event Report System," reads as follows:

            "Actuation" of multichannel ESF Actuation Systems 1s     Is defined as actuation of enough channels to complete the minimum actuation logiC (l.e.,  (I.e.,

activation of suff1cient sufficient channels to cause activation of the ESF Actuation System). As neither Group II Isolation channel had actuated, it It was determined that the closure of these few valves did not constitute an ESF actuation. - Discussfons Discussions were held with the Restdent Resident Inspectors concerning the reportabtlltyof reportabilityof this event. The Inspectors believed that the event was reportable, citing an Internal NRC memorandum which discussed another reportable. llcenseels licensee's proposal t9 define valid ESFs as resulting only from valid ESF to deftne Signals'. The focus of this memorandum deftnes defines an [SF ESF actuation as the system--In Dresden's case the closure of actuation of a component of an ESF system--in several Group II isolation Isolation valves. Commonwealth Edison is Is now aware of this position as to what constitutes an ESF actuation. By the time these discussions were held. held, several days had passed and the four hour reportabfllty reportability

'(     window had expired.

II I I Y.16-33 F.16-33

• Revision 8 Aoril 1992 lities to determine An informal survey was conducted of other Region III uti 11tles how the Dresden event would have been reported. Responses included both

( reportable and non-reportable. We suggest a continuing dialog with the NRC to share appropriate information (such as the internal NRC memorandum) in order to further refine our reportabl1ity reportability determinations. CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED ESF actuat'ons, Following discussions with the NRC on [SF actuations, an Emergency Notification System <ENS) (ENS) report was made in accordance with 10 CFR SO.72(b)(2)(ti) 50.72(b)(2)(ii) on February 4, 1991. CORRECTIVE ACTIONS TAKEN TO PREVENT FURTHER NONCOMPLIANCE In making 10 CFR 50.72 reportabl11ty

   . To assist in                      reportability determinations.

determinations, a memorandum will be issued by February 19, 1991 to the operating shift personnel providing this broader guidance on what constitutes an ESF. This guidance will be station procedure by May 31, 1991. incorporated into an appropriate statton DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on February 4, 1991 when all reportabtl1ty reportabillty requirements were met. reqUirements lNLD742/18 ZNLD742/18 III.16-34 111.16-34

TAB 17 Revision 8 April 1992 ( DRESDEN 2 &3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Inspection Report No. 50-237/91004 and 50-249/91004

     ~                   Title II!.17-1 II1.17-1             Inspection Reports No. 50-237/91004 and 50-249/91004 March 5, 1991.

dated Harch 111.17-11 March 27, 1991 CECo letter from T. J. Kovach to A. Bert Davis (NRC), Response to Notice of Violation Associated with Inspection Report No. 50-237/91004 and 50-249/91004. c-- ( c Co_ IIl.17-i III.17-i

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                                  ,, 'JIfl MAR 7 1991 0'Wi .!~UCl$R
                                                                  ,NUCl.~R 7

71t IGL.EN UNITED STATES UNITED STATES REGULATORY COMMISSION REGION III REGION

                                                                                   " ROOSEVEL.

ROOSEVELT GLEN EI..L.YN. III T RO,",O ELLYN, U.L.INOIS ILLINOIS ROAD S0137 10137

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L l. Docket No. 50-237 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. Cordell Reed Senior Vice President Licensing Department - Suite 300 Opus West III 1400 Opus Place Downers Grove, IL 60515 Gentlemen: This refers to the routine safety inspection conducted by Mr. J. A. Holmes of this office on January 22-29 and a walkdown on February 13, 1991, of activities at the Dresderi Nuclear Power Station, Units 2 and 3, authorized by NRC Operating Licenses No. DPR-19 and No. DPR-25, and to the discussion of our findings with Mr. E. D. Eenigenburg at the conclusion of the inspection. 'The The purpose of this inspection was to review the implementation of the routine fire protection program. The enclosed copy of our inspection inspe~tion report identifies areas examined during the course of the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. During this inspection, certain of your activities appeared to be in violation of NRC requirements, as described in the enclosed Notice. A written response is required. In accordance with 10 eFR CFR 2.790 of the Commission'S Commission's regulations, a copy of this letter, the enclosures, and your response to this letter will be placed in the NRC Public Document Room. The responses directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Action of 198_0. 1980, PL 96-511. We will gladly discuss any questions you have have'concerning "concerning this inspection. Sincerely, Sincerely. 71~ 7?/~ M. A. Ring, Chief Engineering Branch See Attached for Enclosures and Distribution ( III.17-1 IIL17-1

        ,                                                          Revision 8 Aoril 1992 Commonwealth Edison Company             2    MAR ~

M;"R!i 1991 ( Enclosures and Distribution ri..

Enclosures:

1. Notice of Violation

2. Inspection Reports No. 50*237/91004(DRS) 50-237/91004(DRS)

No. 50-249/91004(DRS) cc w/enclosures: D. Galle, Vice President - BWR Operations T. Kovach, Nuclear Licensing Manager E. D. Eenigenburg. Station Manager DCD /DCB (RIDS) DCD/DCB OC/LFDCB Resident Inspectors - Dresden, laSalle. laSalle, and Quad Cities Richard Hubbard J. Y. McCaffrey, Chief Public Utilities Division Robert Newmann. Newmann, Office of Public Counsel. Counsel, State of Illinois Center ( III. IlI.17-2 17-2

Revision 8 ADril Aoril 1992 NOTICE OF VIOLATION ( Commonwealth Edison Company Docket Nos. 50-237; 50-249 Dresden Nuclear Station Licenses No. DPR-19; No. DPR-25 Units 2 and 3 As a result of the 'inspection conducted January 22-29, and February 13, 1991, and in accordance with the -General "General Statement of Policy and Procedure for NRC Enforcement Actions", 10 CFR eFR Part 2. 2, Appendix C, (1990) (Enforcement Policy) the following violation was identified: Amendment No. 106 to Provisional Operating License No. DPR-19 (Unit 2) and Amendment No. 101 to facility Operating License No. DPR-25 (Unit 3) requires the licensee to maintain in effect all provisions of the approved fire protection program for Dresden Unit 2 and Unit 3. As part of the approved program. program, the licensee committed to install and maintain the fire detection and alarm system in accordance with the National Fire Protection Standard No 72E. 72E, which required the linear thermal detectors to be tested every six months. Contrary to the above. above, the licensee failed to conduct the six month surveillance test on the linear thermal detectors in safety-related fire zones 1.1.1.1 and 1.1.2.1 since July 31. 31, 1989. This is a Severity Level IV violation (Supplement 1). c-c Pursuant to the provisions of 10 CFR 2.201. 2.201, you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply, including' including* for the violation: (1) the corrective actions that have been taken and the results achieved; (2) the corrective actions that will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown.

                                            ~~~
                                           ?f.tfij H. A. RiTli,Chi Engineering Branch

( III. 17-3

Revision 8 April 1992 ADril U. S. NUCLEAR REGULATORY COMMISSION ( REGION III Reports No. 50-237/9l004(DRS); 50-237/91004(DRS); No. 50-249/91004(DRS) Docket Nos. 50-237; 50-249 Licenses No. DPR-19; No. DPR-25 Licensee: Commonwealth Edison Company West III Opus Vest 1400 Opus Place Downers Grove, IL 60515 Facility Name: Dresden Nuclear Power Station, Units 2 and 3 Inspection At: Morris, Illinois Inspection Conducted: January 22-29 and February 13, 1991 Inspector: ~;{6~ J~ Ho1IlleS Date Approved By: .C** F. Mai

                             . J~b~onski,     Chief enadce and Outage Section Date Inspection Summary Inspection on January 22-29 22-29, and February 13.              13, 1991 (Reports No.

50-237/91004(DRS): No. 50-249/91004lDRS>> SO-237191004{DRS); 50-249/91004(DRS>> Areas Inspected: Routine, unannounced inspection to assess the implementation of the licensee's fire protection program, which included a review of licensee action on previous inspection findings, a review of the completed fire protection surveillances, fire protection audits, fire reports and observation of a fire drill. An inspection was performed of tools required for hot shutdown and of equipment required for cold shutdown repair. The inspector utilized modules 30703, 64704, and 92701. Results: Of the areas inspect~d inspected "ne apparent violation and one unresolved item were identified. Th( T h , i e n t'ientified i f i e d the failure to develop and implement a surveillance . .~ear

                                                                   .near thermal detectors located in fire zones 1.1.1.1 and 1.1.2.1                                 ._.}.
                                                                  ..* ). The unresolved item was about the reportability requirements Wh~h         ;:, ... c~ .....

wll.... ...... *.* i.re ire pumps are inoperable (Paragraph 3.c.). l.c.)_ In general, the licensee's implementation of the fire protection program was good. good, although some areas of improvement are needed. The following follOWing strengths were identified in the licensee's fire protection program:

• The fire marshal appeared knowledgeable in fire protection protectio.n systems and initiated action regarding a trend of transformer fires.

*c
 /  ...
 \.

III.17-4

Revision 8 ADril 1992

• The fire marshal and the assistant fire marshal were dedicated and professional in addressing conce~ns.

(

• Overall implementation of the routine fire protection program appeared to be good.

to' The following weak?ess weakness was observed:

• In 1989, the licensee and the NRC identified a lack of surveillance testing for the linear thermal detection. \alen Yhen the inspector returned to evaluate licensee action in this area during this inspection, the testing was still not being conducted and the licensee's corrective actions had not been accomplished (Paragraph 2.d.)..

2.d.) { 2

• III.l7-5 III.17-S

Revision 8 April 1992 DETAILS ( 1. Persons Contacted Commonwealth Edison Company (CECa)

      *E. Eenigenburg, Station Manager
      *R.                           Marshal Black, ASsistant Fire Harshal
      *M. Churilla, Technical Staff Engineer
      *M. Dillon, Fire Marshal
      *L. Gerner, Technical Superintendent
      *R. Jackson, Technical Staff Group Leader
      *K. Kociuba, Quality Assurance. Superintendent
      *K. Peterman, Regulatory Assurance Supervisor
      *M. Strait, Technical Staff Supervisor
      *E. Skowron, Technical Staff Engineer
      *G. Smith, Operations Assistant Superintendent
      *B. Whalen, Assistant Technical Staff Supervisor The inspector also contacted other licensee personnel during the course of the inspection.
      *Denotes those attending the January 29, 1991, exit meeting.

2. Licensee Action on Previous Inspection Findings a.

8. (Open) Unresolved Item (237/88010-01 (DRS); 249/88012-0l(DRS>>);

249/88012-01(DRS>>; It was the inspector's concern that a fire in the decommissioned Unit 1I may expose operating Unit 2 safety-related areas. The licensee has requested a change to the reqUired fire protection 1 .. as indicated in a letter dated November I, program for Unit 1,' 1989, to the NRC's Office of Nuclear Reactor Regulation (NRR) regarding the Supplement to Proposed Amendment to Reflect Non-Operating Status. The inspector discussed the concern with NRR and this item will remain open pending resolution from NRR.

b. (237/88010-02(DRS); 249/88012*02(DRS>>;

(Closed) Unresolved Item (237/88010-02(ORS); 249/88012-02(DRS>>; The licensee's methodology of pulling fuses is considered a hot shutdown repair, which is not permitted by Appendix R. The hadprevlously licensee had 'previously submitted an exemption request for fuse pulling, which was addressed in the safety evaluation attached to a letter *dated July 6, 1989. The letter indicated that Region III was to verify the licensee's ability to perform the identified short-term hot shutdown repairs in a timely manner. On February 13, 1991, the inspector verified the licensee's ability to perform short term hot shutdown repairs regaring the replacement of blown control power fuses for the swing diesel generator starting controls, and removal of 20 control power fuses for the reactor relief valves. The proposed manual actions could be performed in a timely manner and no discrepancies were noted. This item is closed. 3 ( III. 17-6 III.17-6

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

Revision 8 April 1992

c. (237/88030-01(DRS); 249/88031-01(DRS>>;

(Closed) Open Item (237/88030-0l(DRS); ( Due to the unique design o!of fire wrap access covers to two pull boxes, it was requested that 3M Company review the installation of this design to ensure that the fire rating had not been invalidated. The li~ensee received a letter dated Hay Kay 3. 3, 1989. 1989, from the 3M Company, which indicated that If Company. if the installations were Were installed according to the drawings, then the installation would provide one hour of fire protection. Based on the response, this item is closed.

d. (Closed) Open Item 237/89013-01(DRS)'

237/89013-01(DRS); 249/89012-01CDRS>>; 249/89012-01(DRS>>; In the original item, the inspector had requested the six month functional test for the linear thermal detectors installed in fire zones 1.1.1.1 and 1.1.2.1. The test had not been developed; however. however, the licensee indicated that a recent audit had identified the same concern. concern, and the surveillance procedure was in the process of being developed. The licensee indicated to the inspector that the surveillance would be completed by July 21, inspector, 1989. During the current inspection, the inspector requested the completed six month functional test for the same linear thermal detectors in fire zones 1.1.1.1 and 1.1.2.1. The licensee indicated that the surveillance test had not yet been approved. The licensee's lack of surveillance testing for the thermal detectors has ( been upgraded to a violation (237/91004.01(DRS); (237/91004-01(DRS); 249/91004-0l(DRS>> of the approved fire protection program that required requir~d the detectors to be tested every six months according to the National Fire Protection Association (NFPA) Standard on Automatic Fire Detectors (NFPA 72E).

3. Routine Fire Protection Program Review (64704)

This inspection consisted of a review of completed fire protection surveillances, fire protection audits, fire reports and an observation of a fire drill. Inspections were performed of tools required for hot shutdown and of equipment required for cold shutdown.

a. Fire Protection Surveillance The inspector reviewed a sample of the l~censee's completed surveillance procedures as listed below:

DFPP 4123-6. -Unit

                          'Unit 2/3 Diesel Fire Pump Annual Capacity Check,-

Cheek,' Revision 5 DFPP 4123-7, RUnit

                          'Unit 1 Fire Pump Annual Capacity Test,R Test,'

Revision 5 4 III.17-7

Revision 8 April 1992 DFPS 4145-1, 4145-1. "Cardox System Semi-Annual Maintenance Test Data ( Sheet," Revision 3 DFPS 4183-4, "Unit

                      *Unit 2 Heat/Smoke Detector Semi-Annual Operability Test,"

Test,* Revision* Revision 0 DFPS 4~83-5, 4183-5, *Unit 3 Heat/Smoke Detector Semi-Annual Operability Test," Test,* Revision 0 DFPS 4183-6, *Unit 1,2,3, Heat/ Smoke Detector Semi-Annual Operability Test," Test,* Revision 0 No unacceptable items were identified; however. however, the following observation was noted: (I) (1) Annual Diesel Fire Pump Test The inspector observed that the fire pump surveillance test results (dated AprilS, April 5, 1990) were significantly different than the fire pump shop test curve. Procedure Number DFPP 4123-6, Revision 5, verified that the fire pump ~as was functioning properly by trending pump performance. Trends of the pump test results for at least four years did not indicate problems with the pump; however, based on the discrepancies between the fire pump surveillance test results and the fire pump shop test curve, the licensee agreed to review this concern and take appropriate actions. ( b. Fire Fjre Protection Audits (1) Technical Spe~lficatlon Spe~ification 6.0.H.l requires an independent fire protection and loss prevention program inspection and audit be performed at least once per 12 months utilizing either qualified off-site

• licensee personnel or an outside fire protection firm.

The last Annual Fire .Protection Inspection Report dated April 10-14, 1989. 1989, identified findings and observations that were either addressed or Were were scheduled to be addressed by the licensee's staff. No unacceptable resolutions were observed. (2) Technical Specification 6.0.H.2 required an inspection and audit of the fire protection program to be performed by a qualified outside fire consultant at least once every 36 months. May 8, 1990, identified items The triennial inspection of Kay that were brought to management's attention, and were resolved by the licensee. No discrepancies were observed in this area. 5 (

• III. 17-8 III .17-8

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

Revision a8 Aoril April 1992 c( c. Deviatio:: Report Review (DVR) 2/3~90-130 Deviation Report (OVR) 2/3:90-130 states, "On

                                                     *On November 20, 1990, at 0147 hours, with Unit 2 in the Refuel mode and Unit 3 in the Run mode at 95% of rated core flow, a simultaneous coolant system failure of the Unit 1 and the Unit 2/3 diesel driven fire pumps (DFPs) . occurred during weekly operability tes*ting.*

The DVR indicated that Unit 2/3 diesel fire pump failed due to the rupture of its engine coolant hose. The apparent cause of the rupture was a small tear/split from normal deterioration. This tear/split then propagated into the eventual rupture. The DVR indicated that the Unit 1 diesel fire pump had previous problems with its coolant system.and the exact cause would not be determined until the diesel was disassembled by the vendor. According to the DVR, immediate corrective actions were taken to replace the failed cooling water hose of the Unit 2/3 DFP, which was completed approximately nine hours later. The licensee indicated that this was a non-reportable event. This position does not appear consistent with Generic Letter 86-10,

        .which

_which indicated that the licensee is to report deficiencies in the Fire Protection Program which meet the criteria of 10 eFR 50.72 . and 10 eFR 50.73. This concern was discussed with the NRR project manager on February 26, 1991, and is considered an Unresolved Item (237/9l004-02(DRS); (237/91004-02(DRS); 249/9l004-02(DRS). 249/91004-02(DRS>>.

d. Fire Drill

( (

 \

On January 22, 1991, at approximately 3:30 p.m. a fire drill was initiated when a treuble and fire alarm was received in the control room from the Unit 2 diesel generator room. The fire drill postulated a fire as a result of an 011 oil spill at the Unit 2 diesel generator. The carbon dioxide system was considered out of service and not operable in the both the automatic and manual mode. The fire brigade responded fully dressed within five minutes. The brigade leader was assertive and appeared knowledgeable in directing his team in attacking the fire. The fire brigade performance was good. During the critique, the inspector indicated that all members of the fire brigade should be equipped with self contained breathing apparatus (SeRA). The licensee indicated that normally during refueling outages, due to SeRA, it was decided that time required to clean and maintain the SCRA, only two fire brigade personnel utilize SCBA. The inspector informed the licensee to consider requiring all fire brigade members to utilize seRA SCRA during the fire drill in order thac that the fire brigade members become more proficient in the use of $CRA. SCRA.

e. Redundant Safety-Related Cable The inspector verified the power cables for the control rod drive 6

{ III.17-9 III .17-9

                                                                              ... ~ .... -,.--~---.:....~"--

Revision 8 April 1992 pump 2A*302*3 2A-302-3 and power cables from the isolation condenser valve ( MQ2*130l*1 M02-1301-1 were adequately separated as required by Appendix R. No unacceptable items were"" were*- observed. f; Safe Shutdown Repair Eouipmentand Equipment "and Tools The licensee has been granted several exemptions about hot shutdown repairs. Specific pIeces pieces of equipment such as fuse pullers and fuses, are required to be readily available to accomplish hot shutdown repairs in a timely manner. Several equipment boxes and the safe shutdown equipment cart were inspected to ensure that the proper equipment was available. In addition, cold shutdown.repair equipment was also inspected. No unacceptable items were observed.

g. Fire Reports The inspector reviewed the fire reports for 1989 and 1990. The fires that occurred consisted of shorts in motor windings, electrical faults in breakers, failure of pump bearings, water leak shorting a breaker, and 50 so forth. The reported fires in many of the~*a5es the~"ases were small and insignificant and were immediately identified by plant personnel or fire detection equipment. There was, however, a trend developing regarding the fires in control transformers in nonsafety*related nonsafety-related areas where the equipment was not maintained at the same level as the equipment in safety-related areas. The fire marshal informed the appropriate personnel to address this concern. As a result, the licensee's

( proposed corrective actions included replacement of existing control transformers with new ones that have built in fuse blocks. The work will be done during routine preventive maintenance of 480 V breakers or during any corrective maintenance work. The work is tentatively expected to begin no later than April 15, 1991.

h. Plant Observations The inspector observed several areas of the reactor building and turbine building that included several hose stations, extinguishers, sprinkler valves, emergency lights and housekeeping. The inspector concluded that the eqUipment equipment was well maintained. Housekeeping In in these areas was good.

3. Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1) on January 29, 1991, and summarized the scope and findings of the inspection. The likely informational content of the inspection report was discussed with regard to documents reviewed during the inspection.

The licensee did not identify any of the documents as proprietary. The inspector also conducted a walkdown of fire areas on February 13, 1991, which resulted in no new findings. 7 (' ( III.17-10

                     ................. " .......... -.... ~ ..... ,
                   ~.100
                    ~ -100   Oeus Ocus ?'ace P'ace Downers Greve. 1:IIr.C*S      I',j,r,c*s oC51S 6C515 Revision 8 ADril 1992 March 27, 1991

• c Mr. A. Bert Davis Regional Administrator U.S. Nuclear Regulatory Commisslon Commission 799 Roosevelt Road-RIll Glen Ellyn, II 60137

Subject:

Dresden Nuclear Power Station Units 22 and 3 Response to Notice of Violation Associated with Inspection 50-237/91004: 50-237/91004; 50-249/91004 NRC Docket Nos. 50-237 and 50-249

Reference:

M.A. Ring letter to Cordell Reed dated March 5, 1991 transmitting transml ttl ng NRC Inspectton Inspection Report 50-237/91004; 50-249/91004. Mr. Davis: This letter provides the Commonwealth Edison Company (CECo) response (attached) to the subject violation transmitted by the referenced NRC Inspection Report for Dresden Station. The violation involved Involved failure to conduct a surveillance test for the linear thermal detectors located tn In specified fire zones. If your staff has any questions or comments concerning this letter, please refer them to Rita Radtke, Compliance Engineer at (708) 515-7284.

• Very truly yours, T. J. vach Nuclear licensing Licensing Manager cc: B.L. Siegel, Project Manager, NRR D.E. Hills, Senior Resident Inspector Document Control *Desk. Desk (Hash; (Washington,ngton. D.C.)

.(( RR/TK/lmw

Nl.D810/21

m.0810/21 III. 17-11

                                                           .... - ** p  --.----- .--.---..-----~~
                                                                        --.----- .... _                                     ..,.~.~                 ..

RESpoNSE TO NOTICE OF VIOLATION

RESPONSE

Revision 8 Aoril 1992 ( Company Commonwealth Edison Comoany Docket Nos. 50-237; 50-249 Dresden Nuclear Station LIcenses Licenses No. DPR-19; No. DPR-25 Un its 22 and 3 Units As a result of the inspection conducted January 22-29. and February 13. 1991, 1991. and in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions", Actions". 10 CfR CFR Part 2. Appendix C, C. (1990) (Enforcement Pol1cy) Policy) the following violation was identified: Amendment No. 106 to Provisional Operating License No. DPR-19 (Unit 2) and Amendment No. 101 to facility Operating Lfcense License No. DPR-25 (Unft (Unit 3) requires the licensee to maintain In effect all provisions of the approved fire protection program for Dresde~ Dresden Unit 2 and Unit 3. As part of the approved program, program. the licensee committed to Install and maintain the fire detection and alarm system 1n in accordance with the National Fire Protection Standard No. 72E. which required the linear thermal detectors to be tested every six months. Contrary to the above. the licensee failed to conduct the six month surveillance test on the linear thermal detectors in safety-related fire zones 1.1.1.1 and 1.1.2.1 since July 31.1989.

31. 1989.

This ;s is a Severity Level IV violation (Supplement 1) I) DISCUSSION In June 1989. Dresden Station began development of a surveillance procedure to conduct six month surveillance testing on the linear thermal detectors located tn in fire zones 1.1.1.1 and 1.1.2.1. Hark request 084775 and a special procedure were written and the surveillance was performed on July

  ~I. 1989 by the Electrical Maintenance Department.
  ~l, A permanent procedure was then drafted and routed for Station preliminary review. This permanent procedure included Included a provision to incorporate the new surveillance in    In the Dresden Statton Station General Surveillance (GSRV) computer program upon issuance Issuance of the new procedure. The GSRV computer program 1s    is used to notify cognizant personne1 personnel of pending due dates for various surveillance tasks. However. due to significant de1ays   delays 1n        In the preliminary review process. the procedure was not completed and issued          Issued within the ttme                   time frame originally anttcipated.

anticipated. As a result. the GSRV computer program was never updated and two surveillance tests were missed. During the inspection Inspection on January 22. 1991. the two mtssed missed surveillances were identified. Identified. The missed surveillances resulted from an apparent deficiency in Dresden Statton's Station's process for initiating new procedures. No* mechanism existed to guarantee that surveillances would be conducted if If necessary prior to issuance of a permanent station procedure. (

I1.DB10,22

rI.DBIO,22 III.l7-12 III.17-12

_.. . . ~. Revision 8 ADril 1992 CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED (

1. Work Request 098032 to perform the required surveillance test on the linear thermal detectors ;nin safety-related fire zones 1.1.1.1 and 1.1.2.1 was initiated and completed on January 27.

27.1991, 1991, and the results were found to be satisfactory.

2. A request to include this surveillance immediately in The Dresden Initiated on March Station General Surveillance Program (GSRV) was initiated 12, 1991 (GSRV #DOOZ418307-D12-01).

3. The permanent station procedure, Dresden Fire Protection Surveillance CDFPS)

(DFPS) 4183-7, entitled "L1near "Linear Thermal Detection Semi-Annual Surveillance" will be approved by AprilS, 1991. CORRECTIVE ACTIONS TAKEN TO PREVENT fURTHER FURIHER NONCOMPLIANCE

                                                     <DAP) 9-2, "Procedure and Revisfon Dresden Administrative Procedure (DAP)                             Revision Processing," will be revised by June 30,1991 to add a requirement for the procedure originator to contact the Department Surveillance Coordinator to initiate requests for any new surveillances or ex1st1ngexisting surveillances where the interval Interval is being changed, to be included 1n     in The Station General Surveillance Program (GSRV) at the beginning of the procedure writing process rather than wait for issuance of the procedure if a commitment is involved. This new mechanism will ensure that any "surveillance
               . surveillance procedures are betng being written or revised.

DATE WHEN FULL COMPLIANCE CQCPLIANCE HILL lULL BE ACHIEVED AQlIEVED Full compliance was achieved on January 27, 1991 when the surveillance for Fire Zones 1.1.1.1 and 1.1.2.1 was conducted and on March 12, 12. 1991 when a request was submitted to include Include this surveillance 1n In the Dresden Station General Surveillance Program . te

1:..1:810/23

..1:BlO /23 III.17-13

DRESDEN 2&3 FIRE PROTECTION PROGRAM DOCUMENTATION PACKAGE Fire Protection Technical Specifications and License Condition controlled copies of the fire protection Technical Specifications and License Condition are assigned by the Nuclear Licensing Administrator to Station and Station Nuclear Engineering personnel. (

r r

      /

~b I {lb /

        /

r I

DRESDEN IHSPECTIOH REPORT stHiARY SlHt'.RY 237/93002; 249/93002 ( Inspectors: O. D. Schrum Inspection Scope: Fire Protection FIre Inspection Period: January 11-15 and February 18-22, 1993 Violations Non-Cited Violations Unresolved Items Open Items o o o o SUlllllary: sumary:

                           +          Steady improvements    Improvements In the fire protection program. (p. 1)
                           +          Staff is            Is ~nowledgeable.

knowledgeable. (p. 1)

                           +          strengths 'ncluded            Included correcting hardware deficiencies, performing surveillances, and training fire brigagde                                                       brlgagde members. (p. 1)               I)
                           +          Fire doors and combustibles were well controlled. <p.

Ftre (p. 1) (. +

                          +          CrItfques Critiques of fire brIgade                            brigade drills were performed well. (p.                                                    1)

I) ReliabIlity Reliability and mater'al material condItion condition of the diesel driven fire pumps was poor. (p. 1, vas I, 3)

 .. --:.--:=~-~- ..., --:--:-:-;:
                      --.-.-      .---~    ... :.; ;-...
                                  .--.':....:.~-        ...:=..:.-:.:..:.:_;--

-.'-==-="':':'::'_:-. ';"; ..... ---::-:---....=-:..- --.-

                                                                                    ."--:-:--~.-                 ** -:;
                                                                                                                                                                        - .---'         -'-.::....:;..:....-~~.-..~"I.-

Concerns with the relfabilfty reliability of the Unit 1 loop fire matn main and the over use of repetitive ~heckllsts during audits. (p. 1.4-5) 1,4-5) o Housekeeping House~eeping was excellent prior prfor to outage, outage. could have been fmproved Improved during the outage. (p. dur'ng <p. 2) Poor PH PM on the dlesel diesel fire pumps. (p. 3) o Plans are underway under~ay to replace the diesel fire pumps, 1t It has an improved PH Improved PM schedule and will be placed on the Technical Issues LIst. L1st. (p. 4)

                         +         Cond1t~ons of the batteries has Improved.

Conditions improved. (p. 4) Have been numerous tamper switch sw1tch maintenance problems on fire protectIon valves. (p. 5) protection ZJa.O'7101252 111.18-1

NUCLEAR REGULATORV REGULATORY COMMISSION REGION III ROOSEVELT "OAD 71' "OO$EVELT .. OAD CL,EN ELL. COLEN ELLVN. ILLINOIS '01 VN. IL.L,NOIS &Oll7~7 ('( MAR £ 1993 Docket No. 50-237 50-2J7 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. L. O. DelGeorge ..

                                                                                                                       ~--    -

Vice President 5 ........ Nuclear oversight and Regulatory Services Executive Towers West III 1400 Opus Place - Suite 300JOO Downers Grove, IL 60515

Dear Mr. DelGeorge:

This refers to the routine safety inspection conducted by Mr. D. Schrum of this office on January 11-15 and February 18-22, 199J. The inspection included a review of authorized activities 1993. for your Dresden Nuclear Power station, Station, Units 2 and 3. J. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the enclosed report. Areas examined d~ing during the inspection are identified in the inspection report. The inspection included a review of the implementation of the fire protection program and an asse~sment and evaluation of your actions to take corrective actions for plant problems. Within these areas, we selectively examined procedures and representative records, made observations, and

            ._conducted.j.nterviews _,with_your.

_ ,_conducted.jnterviews ° personnel~. _.with _your personnel .... _'_'_ _,_0_ <,

                                                                                  , . .'__ ,... __     '---0' ** _'_,0_" ~~
                                                                                                    *0 ,---.' t. _-_ ** _ *. __

No violations of NRC requirements were identified during the course of this inspection. In accordance with 10 CFR 2.790 of the NRCls NRC's "Rules of Practice," a copy of this letter and its enclosure will be placed in the NRC Public Document Room. We will gladly discuss any questions you have concerning this inspection. Sincerely, Cj&~ C:J&~

                                                  . ) . <;:. Wr~ght,
                                                  . / { C.                          Chief Eng~neer~ng            Branch

('-._-

Enclosure:

Inspection Reports No. 50-237/93002(DRS)i 50-2J7/93002(DRS); No. 50-249/93002(DRS) 50-249/9J002(DRS) See Attached Distribution 111.18-2

Commonwealth Edison Company.Company, 2 MAR;: AlAR:: 1993 (( Distribution cc w/enclosure: M. Lyster, Site vice President C. Schroeder, Station Manager J. Shields, Regulatory Assurance supervisor .' D. Farrar, Nuclear Regulatory O. Services Manager DCD/DCB (RIDS) OC/LFDCB Resident Inspectors-Dresden, LaSalle and Quad Cities Richard Hubbard J. W. McCaffrey, Chief Public Utilities Division . Robert Newmann,' Office of Public Newmann,'office Counsel, State of Illinois counsel, center Center J. Stang, LPM, NRR State Liaison Officer Chairman, Illinois Commerce Commission ( T. o.O. Martin, DRS .' W. L.'Axe1son, L.'Axelson, ORSSDRSS " J. E. Dyer, NRR '.' E. J. Leeds, NRR .' M. JJ.'

                       *. Jordan, DRS
 =~.-=.=-~_-_    s. __Stasek r- SRI ,.-Davis-Besse...:-=....;...:.=-=~-
    ....:=-=---- S.                 ,.-Davis-Besse~-=~~=-~'                       .......--=*-=..-"-'-'*=-==--:~-
                                                                                           =.,=,.-~.    ==~~.. - " -

w. W. Hodges, RI A. Gibson, RII S. Collins, RIV '". K. Perkins, RV 111.18-3

U. S. NUCLEAR REGULATORY COMMISSION REGION III ((

  \,

Reports No: 50-237/93002 (DRS1:-No. (DRS){*No. SO-249/93002(DRS) 50-249/93002(DRS) Docket Nos: 50-237; 50-237i 50-249 Licenses No: OPR-19; DPR-19i DPR-25 Licensee: commonwealth Edison company Executive Towers West III 1400 Opus Place-suite 300 Downers Grove, IL 60515 Facility Name: Dresden Nuclear Power Station Units 2 and 3 Inspection At: Morris, Illinois Inspection Conducted: conducted: January 11-15 and February 18-22, 1993 Inspector: 3-2-9'3 DI Schrum 0/ v it {/ Date Approved By: *t-*g.a a..&.~{

                                         *}*g.O                     ~~.

a.(,..!..,...{ :;;&-. F.vJ/ Jablonski, Chief

                                                                                                        ~-2-'13 3-2-f5 Date

( Main~nance Main~Dance and outages Section Inspection summary

     ~",,;_._-
     "'---'     _..._____ Inspection on- January 11-1-5 *and

__. .:Inspection -and -February ""18-22y-1993"c (Reports :No

                                                                           . February *,18-2h-1993"            ,No .. -=--'--~-
                                                                                                                       -'--~-

50-237(93002(DRS1! 50-237(93002(DRS). No. 50-249(93002(DRSll 50-249(93002(DRS>> Areas Inspected: Routine fire protection inspection of surveillances, equipment, f1re f~re brigade training and drills, zebra mussel problems, and fire protection aUdits. The inspector utilized selected portions of NRC inspection procedures 64704 and 92702. Results: Steady improvements continued in the fire protection program. Overall, the fire protection program was considered good. The staff was knowledgeable and had taken appropriate actions to correct issues and problems. Strengths included strengths correcting hardware deficiencies, performing surveillances, and training of fire brigade members. Fire doors and transient combustibles were well controlled. Critiques critiques of fire brigade drills were performed well. Control of fire protection concerns was adequate in the area of plant modifications. Reliability and material condition of the diesel driven fire pumps was poor. Preventive maintenance was beirig increased and the pump/engines were being considered for replacement. Concerns were identified with the reliability of the Unit 1 loop fire main and the overuse C"" ( of repetitive checklists during audits. 111.18-4

DETAILS L

1. Persons contacted

( Commonwealth Common~ealth Edison Company company (CECo)

                       *R. Black, Assistant Fire Marshal E. Carroll, Regulatory Assurance
                       *L. Cartwright, Assistant Technical Staff Supervisor     supervisor
                       *A. D'Antonio, Supervisor Quality Verification
                       *M. Dillion, Fire Marshal
                       *R. Flahive, Technical superintendent
                       *B. Gurley, Regulatory Assurance
                      *K. Housh, Technical Staff Fire System Engineer
                      *J. Kotowski, Koto~ski, Operations Manager
                      *0. Mershon, Technical staff      Staff Fire Protection Engineer M. Nagle, Fire Brigade Instructor
                      *0. Roberts, Corporate Fire Protection Engineer R. Stachniak, Operating Engineer o.

D. Winchester, Win~hester, Internal Audit Group Superintendent U. S. Nuclear Regulatory Commission commission (NRC) M. Leach, Senior Resident Inspector M. Peck, Resident Inspector

                      ~Oenotes
                      ~Denotes those individuals attending the exit meeting'on        meeting on February 22, 1993.                            ..

(  ; . ."

2. Routine Fire Protection Program Review (64704) "

Revie~ (647041 This inspection consisted of observations of pl~~~ areas_and '

=-===- re.views _of...:...f. ire.:..protection.--surveillances i;:.:maintenance-;:,on~fire' ~ --=-=-

     =..:==reviews_oL.£ire-,-proteC:tion,-surveillancesi~maintenance-;:-on=fire':                     -~-~

protection equipment, fire brigade training and drills,~fire*** drills,~fire' ,. reports, deviation reports, work requests, safety evaluations, eva~~ations, ., controls to prevent bio-fouling by zebra mussels, and 'audits audits of fire protection activities. - 2.1 Observation of Plant Areas The inspector observed several areas of the 'reactor building and turbine building. The observation included hose stations, extinguishers, sprinkler valves, emergency lights, and housekeeping. The inspector determined that the equipment was being maintained in good condition. Housekeeping was excellent prior to the outage, although housekeeping could have been improved during outage activities. For example, rags were left in work areas and large quantities of anti-contamination clothing allo~ed to accumu*late. were allowed accumu'late. The majority of the wood used during outage activities was ,treated :to.*-make :to.'*make it fire resistant. Fire resistant plastic was also bein9' used. Lubricants and oils were

                          -stored' in' 'fire' resistantcabinets**-or
         , ., properly -stored*                    resistantcabinets"-or in' steel              - ...

( 2 111.18-5

containers. Equipment areas were mostly free of oil as ~he result of equipment leaks. Appropriate controls for cutting and welding operations were being enforced. No discrepancies were c noted with sprinklers or with (ire main valves or headers. bottles were at appropriate pressures and fire extinguishers had been inspected and had a current-'inspection current--inspection date. No areas were Halon noted where sprinklers should have been-installed but were not already in place. controls were being maintained for transient combustibles and fire doors. All fire doors were functional and temporary outage cables had been routed to ensure that the fire doors were operable. 2.2 surveillances The inspector reviewed completed surveillance procedures for 1992. The surveillances were performed accurately and on time. The observations and discrepancies were corrected with the exception of the Unit 2/3 diesel fire pump. Numerous engine and pump problems were noted in surveillances DFPP 4123-5, "Unit 2/3 Diesel Fire Pump Weekly Operability." The licensee was making efforts to better utilize surveillance resources based on risk and failure rate of equipment, which helped make resources available for other efforts. 2.3 Maintenance on Fire Protection EguiDment Eguioment 2.3.1 Diesel Fire Pumps Pumos ( C- The diesel fire pumps (DFP) were poorly maintained. VerI. VerI- little preventive maintenance (PH) (PM) was done. Maintenance history showed that the DFPs had a large number of failures during the 1990 to 1993 time period. The repair data indicated that the DFPs went

    =-:-:-: __
   --=-:-:-:. --.----=
                -__~ from -_fafailure._.to.     -failure.

i lure -.. to. -fai without lure - wi .any -overall-correcti ve*-actions*--=to--=-:-~-:- thout -any ve-*actions*~to-=7'-:- correct the situation. The failures were caused by years of neglect when PM efforts were not appropriate for the importance of the DFPs, that is, for fire protection and refilling the condenser following a station blackout. PM activities did not include replacing parts that deteriorated with age, such as hoses and gaskets, and checking strainers. For example, when one of the OFP DFP engine coolant hoses burst because of age and pressure, the licensee did not replace the other hoses. The hoses were also not put on a PM schedule to be replaced. Other failures included gaskets, radiator caps, packing, and seals. The engine coolant strainer was not on the PM schedule for periodic cleaning. strainers Strainers were only cleaned in the fire main system following a problem. other system strainers had been cleaned and checked for the first time since their installation more than 20 years ago. c- 3 111.18-6

DFP1 OFP1 engine failed in 1991. The licensee could not-pinpoint the exact cause, but the engine had overheated several tfmes in the six months prior to this problem. The engine was replaced but the pump is in poor condition with very little margin to meet its ( flow requirement. Maintenance history indicated that the reliability of DFPOFP 1 increased after the engine replacement. The pump and engine are scheduled for replacement in 1993. A modification package was approved and the licensee is pursuing an equipment supplier. Repair data indicated that OFP 2/3 was in poor condition. The reliability was low. The failure rate was high and occurred even though the pump was only operated ~O-50 40-50 hours per year. As a result of an engine hose failure, OFP 2/3 failed the same time OFP1 OFPl failed. The licensee was able to make repairs within 24 hours otherwise the reactors were required to be shut down. The licensee purchased a third DFP OFP that can be temporarily connected until one of the two main pumps are repaired. The problem of shutting down the plants is solved, but the reliability has not been increased much for the two main fire pumps in the event of a fire. Both DFPs will be replaced in the 1993/1994 time frame. In addition, improved PM procedures are in the concurrence cycle for the existing pumps. Also, The PM schedule now includes checking and cleaning strainers. The technical superintendent stated during the exit that the DFPs OFPs would be put on the Technical Issues List, which assures that adequate resources will be devoted for improving the material condition of the DFPs. OFPs. ( 2.3.2 Batteries _ _ . _ . . Surveil_l.~!1c~_Feports_ surveillance reports i.ndicated indicated _that that maintenance .of the-_DFP,--,_~~~_ _of the.*.DFP,-

 =---_c"-=----ba:fterles- had
 --*-**-*----ba*fterles   haer been aa- problem       i~cluding water levels, possible p'roblem including OVercharging, overcharging, and maintaining specific gravity. The licensee had taken action to turn over* the maintenance of the batteries to the electrical group during 1992.  ~992.       Following this change the surveillance reports indicated that the condition of the batteries had improved.

2.3.3 Unit 1 Yard Fire Main Loop The Unit 1 yard fire main loop appeared to be in poor condition. The 1992 fire protection insurance log indicated that the fire loop was inoperable several times in 1991 and 1992. The problems were believed to have occurred because of being disturbed during the installation of the sewage system, and not as a result of the asbestos cement piping being made brittle because of pressure 111.18-7 III. t8-7

cycling and aging. Maintaining reliability of the loop~s important because both main fire loops are required to meet the requirements of 10 CFR 50, Appendix R. CUrrent low lo~ reliability makes it ~hether t~~s_system t~~s.system will ~ill be available ( during' i t questionable whether during a fire. 2.4 Fire Brigade, Brigade. Fire Reoorts. and-Fire and* Fire Drills Fire brigade members received extensive training, which included classroom and offsite fire fighting. The onsite fire drill requirements had been met by all brigade members who ~ho were

                                                                                                ~ere listed as qualified. _ All appropriate drill and training records were                    ~ere properly maintained.

revie~ of the fire records indicated that the fire brigade was A review ~as only required to respond twice t~ice in 1992. The twot~o events were ~ere for a motor fire and a power po~er transformer fire. The small number of responses was~as indicative of good control of combustibles, cutting/~elding activities, and housekeeping. cutting/welding Recent efforts at improvements for fire fighting include purchasing more equipment to better outfit the fire brigade

                              ~ith plans to locate the equipment at strategic members, with locations in the plant. This will        ~ill al~ow al.low a faster response to fires.

2.5 Deviation Reports and Work Requests Review ( The inspector reviewed open nuclear work ~ork requests (NWRs) :~or fire protection. The backlog was low considering the high number of NWRs that had been performed during the-year. the year. The NWRs had been properly prioritized and none of the outstanding work items

   = :~':'
   ====          appear~q,._to::..:.be.Jlighly_safety_s
        *.:-C.-c appear~~                                  ig.o if iGan:t

_to:::.be.Jlighly_safety_sig.Oif ican:t .-=....The....l>acklog .been~.==

                                                                          .-=..The-.:backlog .:..had .beeno,,_==

reduced from 175 to 139 during 1992. In addition, the fire protection Nuclear Tracking System (NTS) backlog had been reduced from 65 to 32 in 1992. -~ There have been numerous tamper switch maintenance problems on fire protection valves. Many of the problems resulted from old tamper switches and the difficulty in purchasing replacement parts. A contributing factor was that the switches s~itches were ~ere an add-on feature, which was easily knocked out of calibration. These problems were being corrected by including valves on the locked valve program with ~ith valves being maintained in position by chains and locks. Specific locks and keys will ~ill be maintained for fire protection valves. The licensee reviewed the valves to assure that those important to safety were included in this effort. Some valves had been added to or deleted from the list based on the review. c( 5 111.18-8

2.6 10 CFR 50.59 Safety Evaluations The majority of the fire protection program has been removed from the Technical Specification. This allows changes to be made to ( the fire protection program by performing a 10 CFR 50.59 safety evaluation. The inspector reviewed 10 CFR 50.59 safety evaluations issued for program changes for 1992. All of the changes were appropriate and were not detrimental to fire protection safety. Some surveillance cycles had been extended based on industry data and failure rates. The safety evaluations that delayed performing full flow testing of the fire main system for six months were based on preventing zebra mussels from entering the fire protection systems, and to give the licensee adequate time to make corrective actions. The plant is currently dealing with a bie-fouling bio-fouling problem, zebra mussel infestation, in its intake water. Zebra mussels were found last summer on screens in the intake structure. Notable efforts were being made to prevent zebra mussels from entering the fire main systems and potentially making the fire protection systems inoperable. Full flow surveillances of the fire protection system were suspended for six months to permit modifications to the systems. Hypochlorite is being injected into the service water system, which connects to the keep fill line of the main system. In addition, thermal shock treatment is also being used to kill the mussels. A modification is planned for an injection system into the fire main system. Strainer checks indicate that the zebra mussels have not entered the fire main system. The licensee has increased the surveillance ( frequency for strainers. The concentration of chemicals will be he monitored in the fire main system following the full flow tests to ensure that the system is maintained zebra mussel free.

  -- -- i::7-=-=--==-=~d-i~f:-':Fi~P;~:t~cti~n 2::=-7-=-=--==-=~~~i~T':Fi~=p;~:t~cti"c;n -
                                                   . Activities The inspector reviewed the following audits of fire protection activities: Quality Assurance/Nuclear Safety Audit Report Number 12-91-I, January 17 through 30, 1991; Quality Assurance/Nuclear Safety Audit Report Number 12-92-I, January 27 through 31~      31, 1992; and Offsite Quality verification Verification Audit Report Number 12-9J-I, 12-93-I, December 14 through 18, 1992.

Preparation for the audits was good. The audit reports were brief and did not indicate the amount of reviews that had been performed in the fire protection area. The audits had adequate detail to detect most program problems. The licensee had taken timely corrective actions for those fire protection deficiencies that were identified during the audits. The audits met regulatory requirements. 6 ( Ill. 18-9 111.18-9

ratber than In general, the audits were more compliance based rather being performance based. The licensee utilized a repetitive check list approach to auditing. The check lists indicated that activities listed had been revie~ed in detail; however, this c* ;\ continued approach could contribute to missing deficiencies year after year. For example, problems with the OFFOFP and unit Unit 1 yard loop reliability, which are discussed in ~aragraph 2.3, were not discussed in the audits.

3. Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1) on February 22, 1993, and summarized the scope and findings of the inspection. The informational content of the discusse~ with, regard to documents reviewed inspection report was discusse9 during the inspection. The licensee did not identify any of the documents as proprietary.

(:( 7 111.18-10

TAB 19 c( May* 20, 1996 May

s. Perry Mr. J. S.

site Vice President Dresden station Commonwealth Edison Company 6500 North Dresden Road Morris, IL 60450

SUBJECT:

NRC INTEGRATED INSPECTION REPORT 50-010/96002, 50-237/96002, AND 50-249/96002 AND NOTICE OF VIOLATION

Dear Mr. Per.ry:

This refers to the inspection conducted on February 14, 1996, through March 29, 1996, at the Dresden Nuclear facility. The purpose of the inspection was to determine whether activities authorized by the license were conducted safely and in accordance with NRC requirements. During this period, routine resident inspections and special planned inspections of the Fire Protection Program and the Emergency Preparedness Program were performed. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the enclosed report. inspection,l several examples of ineffective During this inspection correction actions were identified. The significant weaknesses in the use of system checklists and in the locked valve program, were the result of ineffective corrective identified by the NRC, Were actions. These examples are of concern because of the large identified problems, the multiple opportunities for scope of the identified. your staff to identify these problems prior to the NRC, and because corrective actions to previous violations were unable to prevent recurrence. Both of these issues are unresolved in this report because your initial investigations and corrective actions were still in progress. Additionally, we are concerned about two violations of NRC requirements. In one violation, ineffective resulted in corrective actions from previous equipment failures reSUlted inoperable safety equipment. In the other violation, plant procedures were not followed during testing of emergency equipment. These violations are cited in the enclosed Notice of Violation, c and the circumstances surrounding the violations are described in 111'.19-1 111.19-1

detail in the enclosed report. Please note that you are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements. 111.19-2

J. S. Perry In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response will be placed in the NRC Public Document Room (POR). (PDR). Sincerely, Original Signed by B. Jorgensen for P. L. Hiland, Chief Reactor Projects Branch 1 Docket Nos. 50-10; 50-237; 50-249

Enclosures:

1. Notice of Violations' Violations

2. Inspection Report cc w/encl: J. C. Brons, Vice President, Nuclear Support H. W. Keiser, Chief Nuclear Operating Officer T. Nauman, Station Manager Unit 1I M. Heffley, Station Manager Units 2 and 3 F. Spangenberg, Regulatory Assurance Manager D. Farrar, Nuclear Regulatory Services Manager Richard Hubbard Nathan Schloss, Economist Office of the Attorney General State Liaison Officer Chairman, Illinois Commerce Commission Document Control Desk-Licensing i

\ 111.19-3

NOTICE OF VIOLATION Commonwealth Edison Company Docket Nos. 50-237; 50-249 Dresden Station License No. DPR-19; DPR-25 Units 2 and 3 During an NRC inspection conducted on February 14, 1996, through March 29, 1996, violations of NRC requirements were identified. In accordance with the "General statement of Policy and Procedure Proced'..lre for NRC Enforcement Actions," NUREG-1600, the violations are listed below:

2. Dresden Technical Specification 6.2.A required, in part, that written procedures shall be implemented covering the activities referenced in Appendix A to Regulatory Guide 1.33, "Quality IIQuality Assurance Program Requirements (Operation),"

Revision 2, February 1978. The operation of AC and DC Revisio~ emergency power systems was referenced in Appendix A to Regulatory Guide 1.33. A. Dresden Technical Surveillance (DTS) 6600-2-02, "Diesel If Diesel Generator Fuel Consumption Test," steps 1. I.22 and I. I.4,4, required the generator load to be maintained at 2600 kW during the test. Contrary to the above, on February 16, 1996, generator load varied between 2516 kW and 2600 kW during the performance of procedure DTS 6600-2-02. B. Dresden Engineering Surveillance, (DES) 4153-04, "Emergency Lighting Discharge Test, Test,"II Revision 0, dated January IS, 1993, required that battery powered emergency lighting units needed for operation of safe shutdown equipment and in access and egress routes, as required by 10 CFR Part 50, Appendix R, shall be demonstrated by an 8-hour discharge test. Contrary to the above, during 1994 and 1995, 47 emergency lighting units required by 10 CFR Part 50, Appendix R were not discharge-tested for 8 hours as required by procedure DES 4153-04. This is a Severity Level IV violation (Supplement I). (50-237; 249/96002-05) . (50-237;249/96002-05)

2. Criterion XVI of Appendix B to 10 CFR Part 50 states, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and

\. 111.19*4 111.19-4

corrected. In the case of significant conditions atlverse to quality, the measures shall assure that the cause of the ( condition is determined and corrective action taken to preclude repetition. A. contrary to the above, the licensee failed to identify and take prompt corrective actiQns for mUltiple 4 kV breaker problems which occurred since 1989. In addition, the corrective actions taken to Notice of Violation prevent recurrence for a similar violation issued in 1989 were not effective. B. Contrary contrary to the above, the licensee failed to identify and arid take prompt corrective actions for Containment Cooling Service Water (CCSW) foreign material problems which occurred since 1994. This resulted in the failure of the "2A" CCSW pump in March 1996. This is a Severity Level IV violation (Supplement I). (50-237;249/96002-06) ( Pursuant to the provisions of 10 CFR 2.201, CornEd is hereby required to submit a written statement or explanation to the u.s. u.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region III, and a copy to the NRC Resident Inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued to show cause why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time. 111.19-5

( Dated at Lisle, Illinois this 20th day of May 1996 c 111.19-6

U.S. NUCLEAR REGULATORY COMMISSION ( REGION III Docket Nos: 50-10; 50-237; 50-249 License Nos: DPR-2i DPR-2; DPR-19; DPR-25 Report No: 50-010/96002; 50-237/96002; 50-249/96002 Licensee: Commonwealth Edison Co~pany co~pany Facility: Dresden Nuclear Station Units 1, I, 2 and 3 Location: Opus West III 1400 Opus Place - Suite 300 Downers Grove, IL 60515 ( Dates: February 14 through March 29, 1996 Inspectors: C. Vanderniet, senior Senior Resident Inspector J. Hansen, Resident Inspector D. Hills, Regional Inspector R. Jickling, Regional Inspector J. Maynen, Reactor Engineer D. Roth, Resident Inspector D. Schrum, Regional Inspector C. Settles, Inspector, Illinois Department of Nuclear Safety T. Tella, Regional Inspector Original Signed by B. Jorgensen for Approved By: P. L. Hiland, Chief Reactor Projects Bra~ch 1 111.19-7

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u.s. NUCLEAR REGULA REGULATORY COMMISSION . ',.",:,:~ ~';';}7if;;~~ TORY COMMISSION"':" :X!tij$.~~l REGION III Nos: Docket Hos: 50-10; 50-237; 50-249 Li cense Nos: license OPR-2; DPR-2; DPR-19; DPR-25

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Report No: . 50-249/96002~ .,-,'(:. 50-237 /96002; 50-249/96002:-: 50-010/96002; 50-237/96002; '::<<~~. ~~~~:~ ~g~ . Licensee: licensee:

                         "_,,,,'th E",," '~~

Comonwealth Edison Company **):.~#;;~~£I

.:>:;;f:Jfl Facnity:

Facil ity: Dresden Nuclear Station Units 1. I, Z and '3 '."--::,~;.,;;;..

                                                                                      """';:i:.'~>':'~~    ..;\.:,j;:.t::g' ~

Location: location: Opus West III Opus West II I

                                                                        *;*3.;t:z~i:;~:~r~tj~:-. ""  . ~."

1400 Opus Place Suite 300 .. , DOlffiers Grove. Downers Grove, IL 60515  ::;., Dates: February 14 through Harch 29, 1996 Inspectors: C. Vanderniet, Senior Resident Inspector J. Hansen, Resident Inspector O. D. Hills, Regional Inspector R. Jlckling, Regional Inspector Jickling, J. Haynen, Reactor Engineer Haynen. D. Roth, Resident Inspector ,,~:: D. O. Schrum. Schrum, Reglonal Regl0nal Inspector ' ..... "; ~"" ...', .i..~:.!".

                                                                                   ":... -:";"              ~ ~,.r. ';",            "",.~
                                                                                                                                        ..';0:

Settles, Inspector. C. Settles. Inspector, 111 Illinois Department of" inoi5 Oepartment of ...."::'. :'; ~;~ .~. :j.

                                                                                                                               .....'.:j-'

Nuclear Safety " .', ;: -:. .. ~-::... ~ '. .'. T. Tella, Regional Inspector . ".'.. ~ '.'

                                                                                                                                   ~

- ~'. ~.: .

J "". **

ApPI-oved By: ~~ief

                        ~zm~ief Reactor Projects Branch 1
                                                                                                                *    +
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                                                                                                                  . :~~;'~.;?,.
                                                                                                                                      , r FOR INFORMATION ONLY 9bO~~90171 9605~O 9605~0 PDR  ADOCK 05000010            111.19-8

() " PDR PDR

EXECUTIVE

SUMMARY

( Dresden Nuclear Station Units 1, I, 2 and 3 NRC Inspection Report 50-10/96002; 50-237/96002; 50-249/96002 This integrated inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 7-week period of resident inspection; in addition, it includes the results of announced inspections by regional personnel in the areas of fire protection, emergency preparedness, and radiation protection. Operations In general, the conduct of operations within the control room has been professional with good demeanor and communication exhibited. Concerns were raised, however, with the use of an inadequate operator aid (Section 01.2), the potential use of pen cartridges to jam feedwater heater controllers {Section (Section 04.2} 04.2) and improper operator response to an expected half scram (Section 04.3) . Several concerns were raised with operation activities outside of the control room during this reporting period. In particular, significant problems were identified with the licensee's electrical and valve checklist and the locked (.~~ .... valve program (Sections 03.1 and 03.2). The inspectors also identified a violation of procedure during emergency diesel generator surveillance (Section 04.1). While preforming the final drywell closeout the inspector identified numerous deficiencies including an unauthorized modification, loose fibrous insulation, a chainfall and straps still in place on a feedwater line, and screws missing from junction boxes (Section 02.1). Maintenance Maintenance of 4kV circuit breakers was identified as a problem and was cited as an example of a violation for ineffective corrective actions (Section M2.1). continued problems with foreign material interfering with the operation of the Containment Cooling Service Water system were identified and were also cited as an example of a violation for ineffective corrective actions (Section M4 . 1) . ( 2 111.19-9 111.19*9

Numerous examples of problems with the skill of the craft were identified by inspectors and discussed with licensee management (Section M4 .2) . Engineering Several known Updated Final Safety Analysis Report (UFSAR) deviations and other licensing document discrepancies were allowed to exist for long periods of time due to inadequate emphasis on plant design and licensing basis. However, the licensee's additional emphasis on resolving both the old and the more recently identified discrepancies has resulted in subsequent changes to the plant design and administrative controls (Section E4.2). The licensee performed a review of the large engineering backlog.to backlog to categorize the items according to safety significance and potential plant impact. The inspectors had no immediate concerns with the licensee's prioritization and resolution of these specific engineering requests (Section E4 . 3) . E4. Several items of compliance and noncompliance with the UFSAR were identified and documented (Sections U1 and U2) . Plant Support Briefs given to personnel performing work in high radiation and high contamination areas were thorough. Radiation protection personnel continued to maintain strict control of material entering and leaving the radiological protected area (Section RI.I) R1.l) . The overall operational status of the emergency preparedness program was good. Response facilities and eqUipment were adequately maintained and in an operational state of readiness (Sections P2.1 and P2.2). Audits and surveillance of the program satisfied the requirements of 10 eFR CFR SO.54(t) SO.S4(t) (Section P7.1). Management support for the program was good and key emergency response personnel possessed a good knowledge of emergency responsibilities and procedures PS.1 and P6.1). (Sections PS.l Overall, the fire protection program was effective at meeting its safety objectives. Most fire protection problems were identified and substantial progress was being ,( I( 3 111.19-10

made to correct those problems. Fire protection program strengths included control of transient combustibles, a low number of fires during the past 3 years, and a low number of impairments requiring a fire watch (Section Fl.l). In general, emergency lights were in good condition, however, a violation was identified for not testing emergency lights F2.l). Also, poly-vinyl for the required 8 hours (Section F2.1). chloride (PVC) pipe usage was not well controlled in the plant (Section Fl.2). cc. ( 4 111.19-11

Report Details

/""  "

( ( \ Summary summary of Plant Status unit 2 remained in cold shutdown as refueling outage D2R14 continued throughout this inspection report period. Licensee efforts have focused on the restoration of components and systems in preparation for returning the unit to service. Work in progress includes system lineups, post-modification testing, and completion of a "fast cruise" program. Unit 3 continued to operate at full power throughout this reporting period, except for short periods of power reduction for planned surveillance. The licensee also began a pre-coastdown to allow for the final cycle rod pull scheduled for the end of April 1996. I. Operations 01 Conduct of Operations (71707) 01.1 General Comments (....

 '           Using Inspection Procedures 71707 and 71711, the inspectors conducted frequent reviews of ongoing plant operations. In general, the inspectors found operations inside the control room to be conducted in a professional manner with good decorum and communication practices evident. However, conduct of operations outside the control room lacked the same adherence to station standards and management expectations. Specific concerns included the poor control and execution of system checklists and the station's locked valve program.

01.2 Incorrect Operator Aid Results in Reactor Level Problem Q1.2 On March 15, IS, the cycling of the High Pressure Coolant Injection (HPCI) steam admission valves (2-2301-4 and 2-2301-5) caused an unexpected reactor vessel level decrease of about 3 inches. At the time of the unexpected level drop, operators were using an operator aid to correct the wide range reactor vessel level indication based on reactor pressure. Using the aid, operators had determined that level was about 4 inches below the HPCI steam line. Actual

c. 5 111.19-12 111.19*12

level was higher than indicated due to a ~ 11 inch uncertainty in the wide range level instrumentation. The ( licensee's corrective action was to update the operator aid by adding a caution about the instrument uncertainty. This event demonstrated a weakness in the control of operator aids. ( c 66 111.19-13

02 Operational Status of Facilities and Equipment (71711) 02.1 Inadequate Drywell Closure Inspection by the Licensee ( The inspectors identified many deficiencies during the initial Unit 2 drywell closure inspection. The most significant items identified included: an uncontrolled modification; installation of loose, fibrous insulation; an under-tension chainfall around a main feedwater line; and numerous junction box covers missing screws with several covers ajar. Prior to the inspectors' tour, several members of the facility staff including management and Site Quality Verification (SQV) personnel had been in the drywell and had identified deficiencies. However, the inspections were inadequate due to a lack of a clear understanding of licensee management's expectations for a final drywell closeout inspection. Additional efforts were expended by the licensee to correct the identified deficiencies and perform a more thorough drywell closeout. Several discussions with senior licensee management were held to communicate concerns with the implementation of management's expectations and the effectiveness of the SQV organization. The potential uncontrolled modification consisted of four support plates attached to the drywell upper level grating ( ... by wire cables. The panels were used to hold removable inspection panels in place around the reactor vessel feedwater nozzles. The "modification" had been in place for several years. Based on a 10 CFR 50.59 review, the licensee concluded there was no unreviewed safety question and initiated Document Change Request 960028 to incorporate the identified supports into station structural drawings. Regarding the fibrous insulation material, the licensee's response to NRC Bulletin 93-02, "Debris Plugging of Emergency Core Cooling Suction Strainers,lI Strainers," was to remove all loose fibrous insulation from both drywells and implement administrative controls to provide assurance that any such material used during an outage would be removed during the drywell closeout inspections. The licensee was conducting further investigation to determine how the material got into the drywell and to assess the impact of such material in the emergency core cooling system (ECCS) strainers. Numerous junction and cable pull boxes were also identified as missing several cover screws. Additionally, a junction 7 111.19-14

box labeled environmentally qualified on a reactor water clean up system isolation valve was discovered with only one screw in the cover and the cover ajar, exposing the interior of the box. The Electrical Maintenance Department (EMD) responded by putting about 10 pounds of screws into junction box covers in the drywell. The licensee was asked if any of those boxes were similarly marked environmentally qualified and what was the impact of not having adequate closure for these boxes. Further review by the licensee also has been necessary on this issue. The chainfall was removed from the drywell along with several bags of additional material. Due to the additional reviews and evaluations that were still on-going this issue will be tracked as a single Unresolved Item pending the completion of the licensee's corrective actions (50-237/96002-01) . 03 Operations Procedures and Documentation (71707) 03.1 Checklist Verification Identified Plant Configuration Problems During a table-top review, the inspectors identified several problems with completed system checklist documentation, control, and execution. A licensee SQV auditor independently identified similar problems. These issues were discussed with the licensee and two independent checklist reviews were initiated to resolve the problems. When these audits were complete, the inspectors assessed the reviews by walking down th~the outside-of-the-drywell parts of the checklist for the "Unit 2 Standby Liquid Control System, System," Dresden Operations Procedure (DOP) 1100-Ml. II II00-MI. During the walkdown, the inspectors identified a valve out of its required position, two valves missing from the checklist, and discrepancies within the checklist which had not been recorded on the checklist discrepancy-resolution sheet. Although the standby liquid control system remained operable, the errors demonstrated that the checklists execution and reviews were inadequate and not in accordance with plant procedures. The licensee responded to these new findings by staffing a 50-person, multi-discipline team to reverify 85 Unit 2 checklists. The licensee also decided that the Unit system checklists_ 2 startup would not occur until the checklist discrepancies were addressed and resolutions specified. 8 111.19-15

The results of the licensee's re-verification efforts will c be tracked as an Unresolved Item pending the checklist review completion (50-237;249/96002-02) . 03.2 Locked Valve Program Inspection Identified Plant Configuration Problems On March 8, the inspectors determined that some containment isolation valves which were identified as locked closed by the Updated Final Safety Analysis Report (UFSAR) Section 6.2.4.2.1, and Dresden Administrative Procedure (DAP) 7-14, "Control and Criteria for Locked Equipment and Valves," were not locked. More than 30 valves listed in UFSAR Table 6.2-10 were not included in the licensee's Unit 2 locked valve program implementation procedures (DOP 0040-M2 and M4) M4).. Also, there were several discrepancies in the UFSAR such as omitted,valves omitted.valves and valves with incorrect penetration listings. After discussions with the licensee, operations and engineering personnel completed a review of containment isolation valve requirements and implementation, and identified several significant discrepancies including: 9 111.19-16

• Four valves committed to being locked closed in the c

Systematic Evaluation Program that were not in the locked valve procedure.

• Several locked valves were missing placards, identifying locked open or closed, which were required by a commitment to a violation from February 1, I, 1984.
• The lock and chain were missing from the Unit 2 shutdown cooling inlet header "A" ~A" outboard drain valve

((2-10l-46A) 2 -101- 4 6A) *.

• A locked valve, operated during a surveillance, was repositioned with no independent verification of the valve's final position.
• Twelve valves listed in the UFSAR did not appear on the locked valve checklist for Unit 3.
• Two contairunent containment isolation valves installed by "exempt changes" E-12-2-95-232 and 233 were not identified as locked valves although the valves should have been added to the locked valve procedure by the modification process.

(An "exempt change" is similar to a modification.) (. As a result of the findings, the licensee developed a new technical position on locked containment isolation valves. This position required the correction of identified procedure errors, the deletion of 60 valves from the UFSAR tables, and the addition of 70 valves to the UFSAR tables. A walkdown of the valves was completed by the licensee with no containment isolation valves found out of position. The locked valve procedures were revised to incorporate the identified deficiencies and changes. The checklist team (see Section 03.1) is re-verifying locked valves during checklist execution. The inadequacy of the licenseers licensee's locked valve program has been a longstanding issue. Numerous Corrective Action Requests, Performance Improvement Forms (PIFs), and violations have been issued on this subject over the last 2 years. However, remedial actions taken have never corrected the full extent of the problem. Final resolution of the corrective actions for these findings and the results of the ( 10 111.19-17 111.19*17

licensee's locked valve verifications will be tracked as an Unresolved Item i (50-237;249/96002-03) (50-237;249/96002-03).. \. 03.3 Atmospheric Containment Atmosphere Dilution (ACAD) Operating and Surveillance Procedures' Bands Differ The inspectors noted that the ACAD system air receiver operating pressure band was being maintained at 44 to 57 psig, which was above the band in DOP 2500-01, "ACAD Dilution Subsystem Operation" (41 to 52 psig). Dresden Operating Surveillance (DOS) 2500-01, "ACAD Compressor Surveillance" and Dresden Administrative Technical Requirements (DATR) section 3/4.5 both listed a pressure band of 44 to 57 psig. In order to resolve this issue, the licensee planned to revise the ACAD procedures to reflect the installation of Nitrogen Containment Atmosphere Dilution (NCAD) system on Unit 2 and to address the UFSAR discrepancies discussed in Section U2.4 of this report. (Note that when NCAD is installed in Unit 2, the operability of Unit 2 ACAD is no longer required.) This item will be tracked as an Inspector Fo!lowup Item (50-237;249/96002-04) pending review of these changes. 04 Operator Knowledge and Performance (71707) c** 04.1 Failure to Follow Procedures During Unit 3 Emergency Diesel Generator (EDG) Surveillance On February 16, the inspectors observed field and control room performance of two Unit 3 EDG surveillance tests. At that time, "Diesel Generator Surveillance Tests" (DOS 6600-01), and Dresden Technical Surveillance (DTS) 6600-2-02, "Diesel Generator Fuel Consumption Test," were being run simultaneously. While DOS 6600-01 step I.12.c. (5) required the diesel load be maintained between 2500 and 2600 kW, DTS 6600-2 steps 1.2 and 1.4 required the generator load to be maintained at 2600 kilowatt (kW) during the test. The inspectors observed that local generator load was 2520 kW. Discussions indicated that the control room operators were not aware of the additional loading restraints of the DTS and were operating the EDG in accordance with the DOS. The inspectors informed the system engineer and Unit Supervisor of the discrepancies. A review of the test data showed that generator load varied between 2516 kW and 2600 kW during the test. This band was too low to satisfy the DTS requirements. Failure to follow 11 111.19-18

procedures was an example of a violation of Technical Specification 6.2.A (50-237;249/96002-05A). ( 04.2 Potential Improper Control of Feedwater Heater Controllers On March 15, the inspector questioned the purpose of the blue strip-chart ink cartridges that were staged by Unit 3 feedwater heater controllers. The Unit 3 Nuclear Station Operators (NSOs) and the Unit Supervisor both stated that the cartridges were to "jam the feedwater heater controllers in pull-to-stop." This practice freed the operator operator'ss hands 1 during a loss of feedwater heaters. (Note that the controllers were designed to spring return to the automatic control position when not being held in pUll-to-stop.) pull-to-stop.) The inspector immediately brought this to the attention of the Shift Operations Supervisor, who stated that use of the jams was not acceptable and the jams were immediately removed. The consequence of using jams was given by the "CAUTION "CAUTION" in I1 Dresden Operating Abnormal (DOA 3500-02), I1Loss "Loss of Feedwater Heaters," Revision 9, which stated, "By stopping the Feedwater Heater Extraction Steam Valves from closing upon a high level signal in the feedwater heater, the possibility of water induction into the turbine rises. IF the valves are stopped from closing, THEN the operator should continuously monitor the condensate level in the heaters with the high level indication. The valve should be allowed to close if the level reaches the high stop on the control room indicator." The use of jams would complicate compliance with this caution statement. The use of jams was not permitted by any station procedure and, if used, would have constituted an unanalyzed temporary alteration. Procedure DAP 07-02 E.13.e stated, "A controller may be placed in manual whenever the judgement of the operator dictates that continued automatic operation is unsafe or may cause unnecessary transients. transients." While the 11 inspectors had not observed the jams being used, the inspectors were concerned that the practice of staging control jams demonstrated a willingness to work outside of procedures to keep the plant at power. 04.3 Inappropriate Response to Expected Alarm Results in Unit 2 Scram On March 27, with the Unit 2 mode switch in refueling and all rods inserted, an expected half-scram signal was 12 111.19-19

received during surveillance testing. Rather than acknowledging the alarms, the NSO responded inappropriately ( by immediately pushing both scram pushbuttons, which initiated a manual scram. All plant equipment responded as designed and the unit was placed in a stable condition. The operator was removed from panel operation in accordance with investi-gative process and a prompt investigation the normal investigative initiated. The licensee's investigation showed that the NSO had briefed the Unit Supervisor of the anticipated half-scram. However, when the alarm actuated, the NSO reacted too quickly and scrammed the unit. The NSO was counseled concerning the event and returned to control panel operations. The licensee's immediate corrective actions appeared thorough and adequate to prevent recurrence. The inspectors planned to assess the effectiveness of the long term corrective actions when the Licensee Event Report (LERs) is issued. 08 Miscellaneous Operations Issues (92901) 08.1 (Closed) Unresolved Item 50-237/95015-08: On February 5, 1996, Unit 2 experienced an unexpected 100 psig discharge of all the Hydraulic Control Unit accumulators. The licensee determined the cause to be a sudden release of a freeze-seal. The inspector reviewed the root cause investigation

c. report and had no further questions and closed the item.

08.2 (Closed) Unresolved Item 50-237;249/95014-01 On November 27 and 29, 1995, the facility experienced trips of Unit 2 and Unit 3 fuel pool cooling water pumps respectively. The licensee has determined that the trips were the result of unexpected actuation of the fuel pool filter high pressure switch. The pumps have a discharge pressure of about 185 psig and the pressure switches have setpoints of 150 psig. Due to piping losses, however, the systems operate very close to the trip setpoint. Therefore, any fluctuation in the system could result in trips of the pumps. This Waswas particularly true when returning a filter to service. Corrective actions included recalibration of the pressure switches and revision to the system operating procedures. The inspector reviewed the root cause investigation report and had no fUrther further questions. (. 13 111.19*20 111.19-20

II. Maintenance ( M1 Conduct of Maintenance (62703) M1.1 General Comments The inspectors found that most maintenance activities observed were completed satisfactorily, however, inadequate resolution of previously identified problems continued to exist. The inspectors identified one violation concerning ineffective corrective actions regarding 4kV breaker maintenance and containment cooling service water (CCSW) foreign material exclusion (FME) controls. These problems had long histories of occurrences yet continued to remain a challenge ~oto plant operations. Also, past and present skill of the craft problems have been evident during the inspection period and have resulted in rework. M2 Maintenance Material Condition of Facility and Equipment M2.l M2.1 Inadeguate Corrective Action of 4kV Breaker Maintenance The 4kV breakers used at the station were known to have problems for many years. In 1989, the licensee was cited for failure to identify root causes for 4kV breaker failures and to take prompt corrective actions. The licensee's corrective actions included an accelerated preventive maintenance schedule for the 4kV breakers and additional direction from system engineering regarding performance of root cause analysis. The inspectors reviewed the 4kV breakers work history from 1989 through 1996. Several problems were identified during this period with 4kV breakers including four LERs and several PIFs.

• Inadvertent auto start of the Unit 2/3 EDG due to damage to a breaker linkage. (LER 2-93-06)
• Failure of an EDG output breaker to close due to a bent linkage on a 4kV main feed breaker. (LER 2-93-012)

The root cause determination and corrective actions for these failures were not effective. ( 14 111.19*21 111.19-21

• A PIF (TDF-2-94-M625) was issued in 1994 indicating breaker linkage problems on four safety-related busses.

Since this was a low priority PIF, no further investigation was performed to identify the root causes. In addition, the recommendations made to correct the linkage problems were not followed. ( 15 111.19-22

• Inadvertent auto closure of the EDG output breaker that

( caused the EDG to "motor." (LER 2-95-009) This was attributed to a failure of the Close Latch Monitoring Switch coupled with linkage binding attributed to inadequate maintenance. The corrective actions taken for this LER had not included any actions to prevent future linkage problems.

• Twelve PIFs were generated on problems with 4kV breakers during January 1996.
• Damaged linkage on the inter-tie 4kV breaker discovered during a surveillance test when the Unit 2 EDG output breaker failed to close. (LER 2-96-001)

Cl.,ER The root causes for the linkage problems were not identified in the LER. The above examples indicated that the root cause evaluations and corrective actions for the 4kV breaker linkage problems during the 1989-96 period were inadequate. As stated above, the licensee's 1989 corrective actions included system ( engineering to provide root cause analyses for the breaker problems. However, the corrective actions taken were not effective in preventing recurrence of the breaker problems. The licensee's continued failure to identify root causes for 4kV breaker problems over several years and failure to take prompt corrective actions was an example of a violation of 10 CFR Part 50, Appendix B, Criterion XVI {50-237;249/96002-(50-237;249/96002-06A) . 06A} During January 1996, the licensee upgraded the breaker investigation and selected a 10 member team to investigate the adverse trends related to 4kV breaker performance. The licensee's team concluded that the primary root cause for the breaker problems was a lack of technical documentation. This report also stated that other contributing causes were:

• Inadequate previous event root cause determination and corrective actions;
• Too many tasks assigned to the system engineer (management deficiency);

16 111.19-23

• Change-related documents not developed or not revised.

( However, the inspectors noted that even though the licensee concluded that the lack of technical documentation was the primary root cause for the breaker problems, it appeared that no efforts were initiated to evaluate whether the lack of technical documentation affected any other areas of the plant. The inspectors planned to evaluate the licensee's corrective actions when the response to the violation was issued. M4 Maintenance Staff Knowledge and Performance concerning Foreign Material M4.1 Inadeguate Corrective Actions Concerning Exclusion (FME) Results in Containment Cooling Service Water (CCSW) Inoperability On March 1,I, the 2A CCSW pump was started to support a visual inspection of the piping. Normal flow was established at about 3500 gpm when the local operator reported the pump sounded "bad," and the pwnppump was secured. The pump was restarted with maintenance and engineering personnel present but flow only carne up to about 2200 gpm at 100 psig (normal discharge pressure was about 185 psig). Initial investigation identified a small "cloth-like" rag that was lodged in the pump impeller. Corrective actions taken to ( ensure the CCSW bay was free of foreign material included a diver inspection of the suction bay and the running of all remaining Unit 2 and 3 CCSW pumps individually to ensure flow met surveillance requirements. Foreign material in the CCSW system has been a recurring problem at the facility and recent examples include:

• On November 28, 1994, two pieces of wood were discovered in 3A CCSW pump during a pump run.

Corrective actions included stating the need for better control of foreign materials.

• On January 16, 1996, a loss of flow and pressure occurred in the control room during surveillance testing being conducted on the 2D CCSW pump. The local operator reported that the 20 pump began to vibrate and sounded like it had lost suction. An investigation was performed but a root cause for this event was not identified. The pump was

( 17 111.19-24 111.19*24

retested with satisfactory results and was returned to service. ( On February 22, 1996, a small slice of wood was discovered impeding the hinge on the 3D CCSW pump discharge check valve. The valve was disassembled and the wood removed. The multiple examples of foreign material intrusion into the CCSw CCSW system demonstrated that corrective actions were ineffective in preventing repetitive occurrences. Failure to take corrective action to preclude repetition was an example of a violation of 10 CFR Part 50, Appendix B, Criterion XVI (50-237;249/96002-06B). M4.2 Skill of the Craft During tours the facility with licensee management, the inspector identified numerous examples (past and current) of poor skill of the craft work. Examples of poor mechanical skill of the craft work included the following:

         **    Inconsistent use of flat washers on flange connections.

• Inconsistent use of lock washers.
• Use of soft 50ft flat washers under torqued bolts.

(

• Hanger and small piping supports not made up tightly.

\
\.

• Misaligned flanges.
• Stacking flat washers under a bolt that was too long for application.
• Bolts too small for flange holes.

The identified problems were on the Unit 2 reactor feedwater pumps and the Unit 2/3 EDG. None of the examples posed an operability or safety concern and the licensee took immediate actions to explain skill of the craft expectations to the staff. Mechanical Maintenance Department (MMD) has initiated a display board in the work space which identified good and poor examples of skill of the craft. This was a positive action to communicate the expectation of management to the technicians regarding this issue. On March 9, an incomplete and unclear electrical work package in conjunction with weak skill of the craft resulted in del}}