ML20059C887

Intervenor Exhibit I-MFP-107,consisting of Notice of Violation & NRC Insp Rept Re Dockets 50-275 & 50-323,

ML20059C887
Person / Time
Site:	Diablo Canyon
Issue date:	08/21/1993
From:	Zimmerman R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
References
OLA-2-I-MFP-107, NUDOCS 9401060182
Download: ML20059C887 (45)

Text

. _ . _ _ _ _ _ _ _

jp - Z 15 3 2 3 - M ' ' MF 8'p95 t]P Dwis

% MReKEL l o 'l

.. p u FPHo7 ,, 4'"-

,-- / o g//< /q"

/

v.7

/

v v .1 4 1'00

'93 C 2d ?6:18 R:s e * 'm EC 75 and EC-223 rat ! fic Cn ord Elec tric Cc'pany 77 Eeale Etreet, Focr 1451 5ar t ranciscc, Califerria 9*1C6 Atterti;n: Mr J, D. Shiffer, Vice President Nuclear Pcwer Gereration Len; c en:

S W ect- N;: :ns,nec tion of ClaMe Canyon Units 1 and 2 '

% s* refers t

tc the trutine inspection ccnducted by Messrs. P. P. tiarbut, radovan, and K.

May 28, 1958. E. Johrston during the period of April 10, through This inspection examined your activities as authorized ' NRC Li enso t

tic s . CPR-E'D a nd DPR-P2. Additionally, the areas of Systen Engineering i ard Port Cause Analysis were examined by Mr. D. F. Kirsch and Mr. J. L. Crews, resce:11sely. /!

the ccnclusion of the inspection, discussions of our f incir g, were held wi th Mr. J. D. Townsend, and other merbers of your staf f.

'he terrt also includes the results of two additional inspection ef forts:

Pr. J j C. Fulsicher's examination of the Integrated Leak Rate Test of the Unit i I containment during the period of May 16 through 20,19E8; and Mr. B. Collins (of EG&G an NRC conti actor) examination of Instrumentation and Centrols during the f erieds of March 28 through April 8 and May 3 through May 19,19E8. Their exit interviews were held or, May 19 and May 20, 1988, respectively .

Aren reportexamined during this inspection are described in the enclosed inspection 1

hithin these a" as, the inspection consisted of selective t'a-inaticns of procedures and representative records, interviews with personnel, and cbservatic>ns by the inspectors.

Based cn the results of this inspection, it appears that certain of your 1 ectivities forth were not ccnducted in full compliance with NRC requirements, as set in the a tice of Violation, er. closed herewith as Aapendix A Your .

response to this rotice is to be submitted in accordance with the provisions of R c re 2.20!, as stated in Arpendix A, Notice of Violation.

FurtFer, the unresolved ite.n discussed in Section 13.c of the enclosed report. l i concernanc cperability of the Auxiliary Saltwater /Cteponent Cooling Water

! systers, ray indicate a need for increased management attention in the area of design / configuration control. As previously discussed and agreed to in our April 2?, 1988 management neeting, design / configuration control is clearly an trmortant part of your Diablo Canyon operational activities. Therefore, we reouest that you provide a discussion of your corrective action program to preclude further instances of potentially unacceptable changes to the operating riant configuration or parameters due to inadequate review or kr.owledge of the 1

i I ,

9401060182 930821 I )

PDR ADOCK 05000275 .,4 /- I O PDR L \ W, - { Lr~PCl j-h' l~ \

-. . I

L' i

e l

1 I

1 l . . . .:

h a

q 'nt sy;te~s desic,n bases.

ettesstort cf Additicr. ally, we request that ycu include your Water the crera!ility of the Auriliary Saltwater /Corpenent Cooling i

i ty;ter's in res;cese to the findirgs in the previcusly rentiCned ur=recclved ite~

.i z

l t. accerbrce wi th 10 CIR 2.790 (a), a ccpy c' this letter and the enclosures si i te placed in 1

J the MC i ublic Dccurer.: Roc .

j The res;:cnses direc ted by this letter are not tubject to the clearance

{ rec:ecres o' cf theMEO, Cf ficePLof96-511.

Managerent and Eudget as required by the Paperwork

,a 'etction Act f W.,uld you have any cuestions ccrcerning this inspection, we will be pleased n discuss t'er: with ycu.

Sincerely, hS {bh M P. P. 7irrenran, Chiet lk Reactor Projects Branch incics a es A;per.dii A, hotice of Violation Ir,sre: tion Feport Nos. 50-275/88-11 and 50-323/Ee-10 Enclosure 1 EG!.G Idaho Peport "I&C Maintenance Evaluation of th Diablo Canyon Pcwer Plant" cc w/encicsures:

% 0 5>idrcre, P3&E F r Locke, PGLE J.

D. Towrsend, PG&E (Diablo Canyon)

! A, E

Taggert, PGI,E (Diable Canyon) _ _

", , .:3 4

5. Weirlerg, IGSE (Diablo Canyon) K a

1

};

i. L. Grebel, TGLE (Ciablo Canyon) ~

btate of California [,. +i Sandra Silver (Report Only) * \ i i d '

N' cc afe^cirsures: L

>:3 (O g Project inspector 'p 0 Fesident inspector

-4< @p, Lp h docket file W A ~ h ",

G. Cock, RV 1 g6 y

, , F ;

5. Paulkenberry, RV .

J. Martin, FV '

_ Qg j A. Johnson, EV { '4 E g T. Poley, NOF i

g =

?j -

.q t_

t:c w/o repert:

J. Zellicoffer i d .

'N .S q S M. Smith i

.i I '

ya j w M ^3 t

mg i

+i

'\

I' 1

K N.

l

'  ! i

rG u

, N, c;

s

. 4 i

.1 1

i f

a.

s e

9 j REG:0 U /dat d y- g2 (e dr l LMPado an q

q,I l 3

v.[ J c h r s t o n PPNarbut MMMenconca RPZimmerman

! 6 M /90 i 6/ n/88 6/ G/88 6/.1/88 6/ii/88 f SI!, a :, i p t ) PE0uisi t Li r j niguE51 COPY ) EU ] kEQui.51 CCF Y ]

l vE5 (l0 i 1 /15'. / NO ] fEb / NC ]rTES(UEST COPY

/ NO ] YE5 / fCN ] ,

l SQD10FDR ]

t p,?)

(IE w

SI / N3 ]

l MBlume '

]

1 6//'fea

.bdg!s b. h b b k' Y )

fr E E,' /

I, i~ --

N3 ]

IU f , or.'"i f f /- 7 J g{ f* s. j

? ~ f f 5 l .' NC -l l

4 i

i l

l i

i i

l l

l l

,+ Y , p ed., > 4, ** *

, . . " ..a.

i. t- s -

$0 *f _. h ' ,, ~ * ? ,e 9

,&T. 'e , '] 'Y

- . f *. :'

_7[\,' 'l * '

', , ~ 4' l . ,

I , '

, - _ _'_ e.

e

.....~-

.. . . . . .t,

. ,

1 <

., .c . . ,c..1-  :.~;,.y

'n,s Dr, -

' ziear ics,- iler t

. M et Nes. E0-275 and 50-323 License Nos. EPR-E0 and DPP-EP brin ^

c. '. R ; Trs;ett' n corducted on April 10 thrcugh "ay ?8,1955 viola tions

+ :{ rm!rr er n wero . der ti f wn In acccrdance with the "Gereral Ste.erer: -

i t'sc, an: ;rc: e rs 'er c, r 4

3, C i;rq , tre ,r N;C Enf orce ent Actic r s,' 10 CFR Part 2, iati m are listed telow:

2 O'- Mt 9, .,rendix 3, Criter1; 1r , art, et i ;er:ns cr all 'VI, "C yrectise Action" provides, e.+rs. estatlish reasures " a assure that conditions ta:3m ;c s,u c.*e;

* , , such

',e a s f e, lures , r a l f unc t icns , def ic iencies ,

r~*l, ; a ti t it u er t c o rr+ 'tM aterial atd eau 1prert, and ncncenforrances are c.er_ _ < ;4 11*,, in the case o' sigr ificant concitions a c i t i c', the reasures shall assure thet the cause c' the r #.

or tier is @ter 1 rec anc corrective action ta' en to preclude p

'nr:rm totr e e a r, satsecuent to the identifica 1cn cf r;n:a+cr

^ '

'r-C7 at:esfor leaiackcir aoftc e stolaticn (issued in inspection report apn 1 , 10 ; , reautrec cleanliness contrcis cn March 21 and c: m1*.  : ,,

curre:tive ar1* cra! acticns taken did not preclude repetition, i

ior t t ie; en 1 c i de r. t s o f loss of cleanliness controls were  ;

.ers:nei, tr c ludir- art, 9, 12, 21, :c, and May 10,198R, by MC and licensee r tre Jni- reatter vesselthe ditctaery on April 22,19 fib of f oreign raterial  !

u;per internals.  ;

P~

n.1< 1: a Se m ity Level IV Violation (Supplement 1) applicable to Unit 1.

h .e '1+s n s' 111 te bW1o! estarliv5;e: ed, 1ficatice C.B.1 states that: " Written procedures <

. mcecu.es re:c rerled mie ented enc 1xard r;aintair,ed covering. . applicable O

in Ap; h Tetraarf 197E.. ' A of Regulatory Guide 1.33, Revision t

(,t'uar y Apperdix A of Regulatory Guide 1.33, Pevisior ,

et IFE, Section 9, ' f rocedures for Perforring Maintenance", states .e er;! rent

'4' air.tenar:e that can affect the perforrance of safety related E gitten sh:;;f te properly procedures, documented preplanned and perforced in accordance with (,

• % circum :arces' ins tructions, or drawings appropriat ; to g n

"Jintenar ce Precedure MP M-54.4, " Spi ~al Wound Gastet keplacernent uide",

G 4v e;la:ts'cr ',entdetec of Fetruary 10, '9ES, provides guid3nce on the proper s v.E4.4 spiral wcund gaskets to ensure leak free asserblies. MP E

' r. in.:iudes deta sheets recuired to be completed by the rechanics Felpro ecdttton, N-i: the Erccedure in paragraph 7.2.2.d.1 requires the use of N

C0 lubricant cn eil mating surfaces of nuts ana bolts.

C -trary a sie:s , , tc

• the a ;ove, cn April ?7,19SS, while replacing spiral wound a L nit I sefety in.)ettion relief valve header flange, y

i I

DR _

Mn. n: - i.C &nn 7'3 i rn:D

. . *f II .

s i

.g l

E4 rechanics used an ur.!uthorized lubricar.1 irstead of the prescribed felpre o p,f %EC,' anc did rat corplete ite data sheets prescribed by MP M-54.4.

. 0.0 I

d >:s is a Ee.erity Level IV violation (Supplement 1) applicable to Unit 6 1, 1 1

bb Pursuant to ite crovisiers of 10 CFR 2.201, Pacific Gas and Electric Company 4 -

i s here:iy recui red to submit a written statement or explanation to the U.S.

hu leer Ren!atory Comission, ATIN: Document Control Desk, Washington, DC

• ir s '

s MEL5 witn a ccpy to the Regional Administrator, Region V, and a copy to the N ;- Resicert :ns;ector, Diablo Canyon, within 30 days of the date of the 1 letter transmitting this Nctice. This reply should be clearly marked as a 1 'Terly to a Not'ce cf Violaticn" and should include for each violation: (1)

)pf the reascn for the violation if admitted, (2) the corrective steps that have

.g teen taken and the results achieved, (3) the corrective steps that will be pj ,i taken to avo'd further violations, and (4) the date when full c ~pliance will

,  ! te v:hieved. If an adequate reply is not received within he t.,ce specified in p l th is hetice, an order r:ay be issued to show cause why the license should not q te rcdified, suspended, cr r. voted er why such other actions as may be proper srculd not N taken. Ccnsideration may te given to extending the response 5

m ti e for r:cc1 cause shcwn.

, ,1 FOR THE NUCLEAR REGULATORY COMMISSION a

d

,i t U

)

s  ;

Dated at kalnut Creek, California 7

!\ 'lJ Mn,wrm~

R. P. 21 4 rran, Chief j

a this l1 Htay ofyy 19E8 Reactor Projects Brar.ch l \

J a

1 l i W

1 f,

A d

i Il n

I I

i it

s t

a 1

y ,.

~ U.

5. NUC EAR REGULATCRY CC""!SSICN PEGICN V 3

l 1

Repo t Nos 50-275/ M-11 and 50-323/88-10 Docket Nos: 50-275 and 50-323 I J

', cense Ncs.

Y; ETR-SC and DPR S2

+r k- Licensee 2

@~ Pacific Gas and Electric Company 77 Bea!e Street, Room 1451 vi San Francise.c, California 94106

,} Facility Name: Diablo Canyon Units 1 and 2 2

' 3 g Inspection at:

Diablo Canyon Site, San Luis Obispo County, California Inspectic, Conducted:

~

I ns pec t o rs :

n

% %sJw r n fm f kV h5 L. M.

Padovan, Resident Inspector

,  %  %> Date Signed m%a -

g% (,//g/ ?3' t.

E. Johnston, Resident Inspector r

% . %, Date signed

%n & c ~ h P. P.

f /// fff Narbut, Senior Resident Inspec{or Date Signed

% %s J. C. Pulsipher, NRR ak r ~ fr g /4 /d/

%, . Date Signed Approved by:  % _-

6 /4 /P/

M. M, Mendonca, Chief, Reactor Projects Section 1 Date Signed Summary:

i 5Insnection 0- 3 ? J / M- 10)fro Q nril 10 through May 28, 1988 (Recort Nos. 50-275/88-11 and I i

A r e a s, Inspected: The s

i operations, raintenance andinspection included routine inspections of plant i events, coen items, and surveillance activities, follow up of onsite independent inspection activities.event reports (LERs), as well as selected licensee 37700, Inspection Procedures 25026, 30702, 30703, 71881, 57050, 570B0, 60710, 61726, 62703, 70307, 70313, 71707, 71709 , 71710, this 73756, 90712,,92700, 92701, 92702, 93702, and 94703 were applie inspection.

M O ? o I O.; 14 H%i' PL6 AD:X L <M 0 0 0 7 '3 0

[1CD l _ A mw .nF-=.-k,,-"*~'~~~-' '

'~~ ~ ~ ~ ' ~

r

.] '

h

.A e i

Det.I*' U* [rs'or!ipq-U k we vio;atit rs were identified.

acticn ir- The first dealt with ineffective corrective ct eling with tr e loss of syste"' cleanliress controls as described in h p ragrart 12. d.

The seccrd violation cealt with rechanics failing to follow j

procedm s durirg r cirter..v'ce activities es described in paragraph A.a.

ln ur r esci,ec l'Er

)

c' the Aus'li ry $?'tw is described in paragrn ' 13.c. dealing with the operab..ity ater ( A%) syste"' during the pericd of tire that the Lect (>? n er J11

• erer ti8l pres sure setpoint was ra ised.

O

An the ap;.arent l% weakt. css is implied by the situaticn of uncertain cperability of

tren

',ys* r ir that it can be ccrcluded that syste"! design bases have not i successfully ray have led to, er could corrunicated to plant persennel and that the result cf this ctanges which lead to, plant personnel making system setpoint they do nc . recognize as affectirg system o, ?rability.

On additicral perceived lack inspec of tr.r concern raised during this repcrting period is the

( tirely, effective corrective acticns in dealing with s1.uaticos in in which plant perscnnel rade errors. The two exarples discussed j the reccrt treblems arc are the subject tre failure of violations; specifically repeated cleanliness cf rechanics e

fcb at to follow procedures, In both cases the I

h?rd 6as corrected but plant ranagement appeared content to allow the ret r al proc er,ses resolve the root cause of the pr oblems. The normal process involves a conccnformance ieport and a technical review group reeti ng, a process *$at can and does take months.

is an inrediate respcnse to ensure other personnel involved in similar wo are quirily alerted to the errors made.

During the reporting period there were good camples of individual plant persnnrel who cxercised an inquisitive safety nirded approach to their work Re:ific e 2rples were the identification of misaligned detectors in tne main steam inprcrer lire radiation detectors by an l&C technician, the identificaticn of sur eillance schedules for time response testing of vital instrumentation channels by an 1&C technician, and identification of the rcssibly generic prcble~ with ccntainment ventilaticn butterfly valves

'dertified by engineers involved in the integrated leak rate test.

Mditictally, .

the licensee's actions leading to the discovery of possible cereric therc;.g$ proble s with analysis.

rcot cause Westinghcuse ARD relays was noted as an example of

~- . _ . - =- - -

.y ~

b , \

.h

'9 ( i 0

CETAllS

g. F p; c. ~

I I

l i y

( . D ry rm C a r t a c t e d i 5

• J. C. Tcwnsend, Plant Manager

' *D. 8 J. " "iklusn, Acting Assistant Plant Manager, Plant Superintendent I '

'C L.

Giscion, Actirg Assistant Plant Manager for Su;. port Services l Lidridge, Quality Control Manager f t C. D ss, Cnsite Safety Peview Group

$ R. G. Iocaro, Security Supervisor

, f *I 9ennett, Acting Mair+anance Manager

' ]

• T D. A.

J.

Ta,;ert, Director Quality Support f Martin, Training Manager

.. .$ W. G.

K J V.

Croc6ett, Instrumentation and Control Maintenance Manager A L. F.

Ecots, Ctemistry and Radiation Protection Manager 3

Wc ack, Cperations Manager Q

gg

'T L. Grebel, Regulatory Compliance Supervisor

.l *5 R. Fridley, Senior Operations Supervisor R

5. Weinberg, News Service Representative W. 1 Rarp, Chairman, Onsite Safety Review Group

n ps 4 Tress!cr, Project Engineer, NECS u

The inspectors d interviewed several other licensee employees including shift fere an (Sru), reactor and auxiliary operators, maintenance Derso vel, plant 1 and technicians and engineers, quality assurance personnel

~ geretal construction /startup personnel.

l ,

e a Denotes these attending the exit interview on May 27, 1988.

?

Operational Status of Diablo Canyon Units 1 and 2 Curing the Notable reportingi period Unit I continued its second refueling outage.

occurrences

( ncluded the discovery of fatigue cracking in reactor l

i coolant suige pump lubrication system componert;, some evidence of pressurizer relays, line c.overent, possible generic problems with Westinghouse ARD a

biological growth in diesel fue' oil day tanks, combustible fire N barrier used raterial, and indications from the ILRT that 48" butterfly valves f or containment purge nj exhaust may have directionally dependent s

leak characteristics.

Unit 2 remaining at power for the reporting period.

j 3.

Op e r a t i o n a l Safety Verification (71707)

-j a. Gereral During the inspection period, the inspectors ob,erved and examined I

activities facility.

to verify the operational safety of the licensee's The cbservations and examinations of those activities l ' were conducted on a daily, weekly or monthly basis.

I

---.-,-----------------w u-,s- ,-

ME

.jy $

p 1

?' I a

I W-

. -vi 4

A se i',

3 da'l, n~

13 sis, tr e ins;ect;.s ebserved ccntrol rcom activities to

'a"ce ,,ith selected Liriting Conditicns fer Operations 1 l

(LC:s a s ; res c r ibed in the facility Technical Specifications (TS).

i,8 h Lces, 'stru entation, rece-de- traces, and other crerational

& <er_

9 fs ere ena-ired to cbtain information on plant conditions, and jg:g) 1et were re iewed fcr co pliance with regulitory requirements.

Q 'hift turr. overs were cbserved on a sa pie basis to 1,erify that all DN cer tirent in'cemation of plant status was relayed. Curing each ees d d ite irsrectors toured the accessible areas of the facility to

%6 chser,e tr e 'ollowing:

(a) Ceneral plant and equipment conditicrs.

I s' t ) cire hazards and fire fighting equipment.

O (r) 0 1diaticn prctection controls.

l (d) Ctnduct of selected activities for co pliance with the licensee's administrative controls and approved proc 'Jres. l 1

,bo 4 (e :nteriors cf electrical and control panels.

'd %]

y (f3 Inle entat,,n of selecte:* partions o' the licensee's physical se uri ty pia, k

'j ( r; ; P' ant hcusekeeping and cleanliness.

tJ h (b) Essential safety feature equipment alignment and conditions.

9 6

(i) Storago of pressurized gas bottles.

T *' e inspectors tal ked with operators in the control recm, and other p plant personnel.

The discussions centered on pertinent topics of i

5 general plant conditions, procedures, security, training, and other a s;;ec ts c f the involved work activitics, jt

• Aco'ication of the Coality Assurance Program to Diesel Generator \

ij Se l usi (ie porary Instruction 2515/93) (Closed) l g :n January 1980, the Office of Nuclear Reactor Regulatory recuested g a'i licensees to check their Quality Assurance (QA) programs with Ef respect g cil in to diesei generator (DG) fuel oil, and to include DC fuel  !

q their QA programs or provide justification for not doing so.  !

y 9 The inspector verified that the licensee has included the DG fuel a cil system in its CA program. The quality classification list specified the CC fuel oil 3 transfer filters, and transfer puTps as storage tanks, transfer strainers, "Q" This implies that the f provisions cf Appendix B to 10 CFR 50 apply.

u j

Lection 4 cf this report addresses fuel oil problems encountered I f-during this reporting period. l c

R l

}  ! I y

5 9

I I i 1

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

s 3 '.

No viciations or deviations were identified.

4. Ons4te Evont r ollow up (93702)

I

a. On April 13, 100% power, 1988, with Unit 1 in a refueling outage and Unit 2 at line post results of a routine 18 month calibration of nain steam (faur cer unit)accident steam monitoring radiation monitors required all eight line monitors to be declared inoperable.

CM tube detectors were found to be positioned in their main steamThe line shield casks such that the detectors did not fully extend into the shield radiation from aperture and were only partially sensitive to potential.

the steam lines. Incorrect detector positioning )

occurred during original installation due to pocr installation instructions and calibration procedural errors.

Accordingly, the moniters were considered to have been inoperable since August 1983 for Unit I and April 1985 for Unit 2. However, elternate proceduralized steam generatormethodologies tube ruptures. were available to identify and assess Previous identification of this problem did not occur since radiation monitor surveillance test procedure (STP) 1-18R2 specified presentation of the radiation source at the detector well top, rather than of the detector cask aperture lines. The due to the close proximity of the casks to the main steam contacted to licensee evaluate the indicated the radiation appropriateness monitor vendor would be of reportability Part 21 Further under tevlew o' licensee followup event report of this event will be done during the 50-275/88-11.

b. Cn April 16, ventilation 1998, Unit 1 experienced a spurious containment monitors RM isolation 11, 28, and due 21.to a high radiation alarm on radiation No cause was determined.

made a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> non emergency 10 CFR 50.72 report. Note: The licensee actions Licensee are addressedto reduce spu-ious actuations caused by radiation monitors in report 50-275/88-13,

c. On April 19, j ventilatten 1988, Unit 1 experienced a spurious containment isolation (ccntainment particulate). signal due to a high alarm on RM-12 No cause was determined,

d. On April 20 1988, Unit 1 fuel reload was completed.

e. On April 22, 1983, for Unit 1, the licensee determined that technical specification 3.11.1 had been violated for some period of time not exceeding one hour and 10 minutes.

The technical specification samples of the required plant vent. continuous sampling of particulate and iodine radiation ronitor RE-24. This function is usually provided by .

Due to planned work on RE-24, a temporary auxiliary sample pump had been I&C work preperly on RE-24,placed in service prior to securing RE-24 . During the thereby violating the technical specification. technicians secured the te There does not appear to be any technical consequences to this act.

The unit was in a refueling outage and local portable air monitoring

,g .yg,, my ,w---

""E "E F

- . _ - . . - - . - . - . _ . - . - _ . - - - - . = - - - -- . . _ - _ _ . .

l l

• l 1  ! s 4

i

. , L i

{ )

equipment was in place and monitoring work activities, h Additicn311y, a vent stack release, if one occurred, would contain I primarily noble gases and a smaller amount of particulate and I

[ icdine, if ary. Potential noble gas release was monitored during the entire time by RE-14A and B and showed no release.

Licensee corrective actions will be followed up through iER 88-12.

W f.

On April 23, 1998, Unit 1 erperienced a containment ventilation f isolation due to radiation monitor RM-14A spiking. No cause was determined.

! g.

On April 26, 1998, Regicnal management conducted an onsite meeting

}

with licensee management. The results are reported in Inspection Report 50-275/88-14.

y'

h. On May 5, 1988, during Unit 1 midloop operation for removal of steam

! generator nozzle dams, the licensee discovered that part of the reactor vessel vent arrangement of the temporary system, Reactor Vessel Refuelirl Level Indicatinn Systems (RVRLI"), had been removed. Specifically, RVRLIS valve 613 had been removed. The

' licensee formed an Event Investigation Team (EIT). This event had no technical consequence in that the temporary system remained vented to atmosphere as it was intended through a different path.

The error was preliminarily determined to be personnel error in that general construction personnel (GC) performed the work without a clearance. The activities discussed in this section involved apparent or potential violation of NRC requirements identified by the licensee for which appropriate licensee actions were taken or initiated. Consistent with Section IV.A of the NRC Enforcement Policy, enforcement action was not initiated by Region V.

i.

On May 5,1988, during 'he perfcrmance of a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> load test on Unit 1 diesel generator 1-1, the licensee determined that fuel oil filters were being severely clogged by biological growth. The condition identified itself as a load reduction due to fuel starvation. Operators switched to the other fuel filter and load was reestablished.

The inspector reviewed the licensee justification for continued operation of Unit 2 (Unit I was shutdown for refueling) and the inspector determined that the licensee's analysis of this condition was acceptable.

The inspector also reviewed the licensee's plan of action to correct a

the situation as well as to prevent recurrence. Initial licensee plans included tank cleaning, biocide treatment and increased testing, y

follow up of this item will be accomplished through the licensee's event report,

j. On May 9, 1988, during planned preventative maintenance of Unit I reactor coolant pump bearings and their lubrication system, the

. _ - _ . - . _ - -- - - - .. .- .. . - . . = _ - - -- . - . - . . - -

5 t

licensee noted several f ailed bolts, an extruded gasket and cracked carts in the assembly which provide rotive force and directs lubricant flow fer tFe reactor coolant purp thrust and radial bearirgs.

The ; rchlem was first identified in RCP l-2. Investigation of the r emaining puTps indicates at least one of the cracking problems may te caneric Lir ce the sa~e failure to a lesser degree was evident.

ine broken bolts and a second cracking problem may be isolated due

• o irproper asserbly or ray be a generic vibration problem.

The licensee is continuing inyt..:gative actions and has, subsequent i to the inspection period, decurented this problem in voluntary LER 50-2 75/RB-15. The licensee is planning corrective actions fer Unit 1.

The litersee has provided justification for continued operation cf Unit 2 in the LER and the inspectors review of the JC0 will be

'he subject of regional correspondence in cor. junction with NRR l review. ,

The resider.ts followed licensee actions closely. Regional and NRR project cn thismanagement ratter. Follow-up personnel were in connunication with the licensee rcrmal inspection of this item will be conducted as part of activities.

k, On May 10, 1988, testirg for reactor trip and essential safety featurethe licensee ident irstrumentation nad rot been conducted in accordance with the schedule of frequencies described in the technical specifications.

The technical specification require such instrumentation to be tested IE" nonthson where a rotational "N" basis, specifically to be tested every "N x is the number of channels of instrumentation.

Ibe licensee had not been doing this in all cases and had in effect cerfused the nurber of available channels with the number of ccmperents (e.g. steam generators) and therefore in some cases was testing at lesser frecuency than required.

The licensee deternined that Unit 1, which was shutdown, was 1;te on time respcnse testing for situations were there were two or less chanrels available.

For Unit 2, the operating plant, the licensee concluded that surveillances were not late for any time response testing but this j pcsition was predicated en an assumption that the technical specification tables (e.g. 3.3.1) did not include any requirement to test what appear to be single channel (i..e. H-1) functions every 18 ranths.

I 4

Discussions of this subject with regional and headquarter technical staf f indicated that fu ther review would be required upon licensee subnittal cf an LER dealing with the subject.

follcwed up through the LER process. The issue will be

6 1

1.

On May 11, 1988, during Unit 1 Diesel Generator 1-2's 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run, the f oamed fire barrier material around its' exhaust stack began to burn. The fire was quickly extinguished and no diesel generator inoperability occurred.

The fire barrier material apparently broke down when exposed to the heat of the exhaust pipe with time and became a flaceable material itself. The possible generic considerations of this event were related to the regional fire specialist who will perform follow up of this item.

The licensee removed the fire barrier material from the other diesel generator locations and is pursuing a design change for permanent corrective action. Fire watches have been set in the interim.

m.

On May 12, 1988, the licensee made a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> non emergency report due to the fuel Handling Building ventilation system switching to iodine removal mode due to a radiation monitor (RM-58) spike. No cause was determined for the spike.

n.

On May 13, 1988, the inspector became aware of a nonconformance report written on April 30 which dealt with a cracked stator inboard clamping ring on the motor 01 a Containment Fan Cooler Unit (CFCU) motor.

The inspector determined that the problem was not similar to that described generic cracks by NRC Information Notice 87-30 which described brackets. in large vertical electric motors in surge ring The responsible licensee engineer stated that the CFCU crack had been found as part of a visual inspection during a planned maintenance activity and was not found to be generic as determined by the inspection of the other CFCU motors (inspected to that time),

o.

On May 18, 1988, Unit I commenced pressurization for a integrated leak rate test (!LRT) of the containment. The conduct o' the test and +

its results are discussed in section 14 ur this report.

During the test it was noted the inside containment valves for containment purge supply and eyhaust (RCV 11 and FCV 660) did not appear to hold pressure.

Subscauent to the ILRT the licensee perforred tight.

a local leak rate test of the valves and found them to be The anomalous leak behavior of the valves, i.e.,

i directionally dependent leak characteristics, caused the licensee to declare the valves inoperable in Units 1 and 2 and to commente an investigation.

At the end of this reporting period, although the seal ring of one valve had been replaced, the licensee had not been able to pressurize the valve inside containment to the required level. The licensee's Event Investigation Team was continuing its efforts to repair the valves.

This matter will be followed closely as part of the routine inspection program.

s l 7 i

p.

On May 19, 1988, Unit 2 experienced a non reportable event when an ILCputer.

cc technician In attempted to remove the display screen for the plant removing the screen a short was caused which resulted in the loss of one bus of ins t r ument power (PY-2/ ). This caused a r.ur o e r manual, of bistables to trip, a number of feedwater controls to rods to step in, and letdown to isolate. Subsequent go to operator action restored 120 VAC power. Plant parameter changes during the event were minimal due to operator actions, q.

On May 20, 1989, Unit 1 exgerienced a pure water spill estimated to te 500-1000 gallons of water in the 115 foot elevation of the Auviliary Building. The spill was cat. sed by a f ailure of the f reeze seal isolating work on a CVCS valve. The licensee is investigating the cause of the free e seal failure. l r.

On April 28, 1988, the insp ct - Lecame aware of a revision to a nonconformance repo-t made on April 13, 1988. The nonconformance NCR C01-87 EM-N121 was originally written on December 2, 1987 , and dealt with malfunctions of diesel generator 1-1 during test.

%pecifically the diesel picked upload but immediately shed load.

The problem was narrowed to a binding relay (a Westinghouse ARD relay).

(40-1300 ohms) The relays binding resulted in varying contact resistance which af f ected logic circuits, Physical inspection noted concrete-like dust in the relays which was attributed (initially) to original construction cust, Subsequently the them to the manuiacturer, licensee removed some of the faulty relays and sent Westinghouse, for analysis. Westinghouse determined and stated in a reply dated March 8, 1988, that the dust like material was due to degraded solenoid potting material and that the relays had not been supplied as safety grade material.

A meeting 28, 1988. was held by the inspector with licensee personnel on April The results of the meeting indicated that 155 such relays were installed in the plant with 136 of them in the diesel generators and the remainder in non safety related uses.

Of the 136 relays, one in each diesel generator affects low voltage logic circuits their operability. in which the contact resistance problem can affect The five relays, that are affected by contact resistance, tested are in circuits used only when the diesel is being an emergency for operability, that is, in parallel with of f site power situation i.e. . De loss of of f site power (when the diesel generators actual are required to load) the 5 relays would not hamper operability.

circuits in make or The remaining 131 relays are ir 125 Vdc I contact resistance change. break situations that are not affected by the important Of these, eight are critical relays with relays. functions such as engine start and water jacket pressure As corrective actien, diesel generator with signs of degradation.the licensee has replaced all relays in The remaining critical relays will be replaced during the current Unit I refueling outage and the upcoming Unit 2 refueling outage.

~

~-

• ) '

8 I

C n a y 3 , 1988, the licensee's corrective actions, proposed cctions i

) and justification for centinued operation were discussed with regional and NRR managers, and were found acceptable.

U l Dn May 26, 1983, the licensee submitted a voluntary LER regarding '

the degradation of the relays, This item will be followed up with j

licensee's LER 50-275/88-09. i

'i

s. Fire in a Unit 2 Auxiliary Building Rad Waste Dryer Cabinet l

)

Cn May 24, 1988 at approximately 2:00 p.m. a fire was discovered inside a rad waste dryer cabinet. The dryer, located insidt

~

ventilated rad waste tent area inside the Unit 2 Auxiliary Building, 5 was being used to dry rad waste filters which had apparently collected flamable paint Chips.

/

f The fire was initially identified by a roving fire watch who N notified the Operations and Ridiation Protection departments. A health physics technician, wearing a respirator, extinguished the 3 fire by or, lugging the hea'er element, dousing the cabinet with carbon dioxide, and placing the filters and rags contained in the cabinet into a bucket of water.

The fire was out within ten minutes and an Unusual Event was r>ot declared. The licensi.e suspended all rad waste dryer operations.

j i

No violations or deviations were identified.

S.

Maintenance (627M) 7 'q The inspectors observed portions o f, and reviewed records on, selected p maintenance activities to assure compliance with approved procedures, technical specificaticar, and appr:priate industry codes and standards.

furthermore, the inspectors verified maintenance . *ivities were .

y performed by qualified personnel, in accordance witn fire protection and housekeeping controls, and replacement parts were appropriately certified, i

a. Safety injection Spectacle F1ance a.

On April 26, 1988, the inspector observed maintenance activities to reverse the spectacle flange on the safety injection relief valves return line to the pressurizer relief tank (PRT). The spectacle flange, consisting of a blind flange and an orifice, had been installed with the blind flange earlier in the outage to facilitate the local leak rate testir.g of containment penetration 71. The orifice side needed to be reinserted to return the line to service for operations and safety injection system testing.

The inspector found a number of problems with the maintenance activity:

I o

The work package in the field included only the odd numbered pages of Maintenance Procedure (MP) M-54.4, the procedure

I 9

• I i

! coverning flarge. the replacement of spiral wound gaskets use. '. this -

The mechanics were not aware of the fact that half the procedure was missing prior to identification by the inspector.

o The mechanics were not using the lubricant specified in MP M-Sa.4 f or t he lubrication of bolts. They were using Chesterton instead of Felpro N-5000.

o

The rechanics were not using the data sheets included in MP M-E4.4 for I recording flange alignment and other important data.

o The work order was poorly written, in that it did not specify

the use of the data sheets for MP M-b4.4 and in fact the cnly 1 instrc

tions given for final flange reassembly were: "At the 1 ccepletion of STP, Mech. Maint. to restore all systems to

cperating state, as required by engineer and foreman in charge."

(

The prcblems fall into two categories; an inadequate work package, i and rechanics not following the applicable procedure. The work order was written to cover both the insertion of the blind flange and the reinsertion of the orifice. When the package was reissued to thestep above fieldfor f or the reinsertion of the orifice it included only the off previcusly. the rechanics to perform which had not been signed The STP referred to was STP V-671, the local leak rate testing of the contair. ment penetration. The step does not call cut MP M-54.4 or its data sheets.

previously M-54.4 signed However, off, describe bolt torque and referred to MPOther st need to be filled out. those steps did not specify that the data sheets Although the inadequate work package contributed to the problems, the activity could have been performed correctly had the mechanics g taken the time to read the package and have it corrected or g requested guidance from their supervisor.

'"eThis was not done and as a result the wrong lubricant was used on bolts. The lubricant used had nct applications. been qualified to be used tn safety relatec solting Failure to follow MP M-54 4 is an apparent violation

, (Enforcement Item 50-275/88-11-01).

following identification by the inspector, nu-ber of immediate corrective actions: the licensee Identified a o

The bolts were cleaned and relubricated with Felpro N-5000.

O The Maintenance '.anager held a meeting with the maintenance department to discuss procedural compliance.

the need to use data sheets included in procedures, and that only the materials specified by the procedure may be used, i o The Quality Control Manager gave instructions to the QC department not to approve work orders with instructions as general as " Restore all systems worked to operating state, as recuired by engineer or foreman in charge."

s

,.m .- - -,m w -- -- v- -

'~ ~

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

Y l

10 l

l

'he licensee convened a Technica' oeview Group (TRG) to review the

' Nonconformance Report (NCR) associated with this incident. At the conclusion of this inspection period the TRG had not yet specified any further corrective actions.

i i

L. Other Maintenance Activities Observed l

The inspectors observed and found acceptable portions of the following r,aintenance activities:

i l o Auxilisry Salt Water Pump 1-1 reinstallation following l overhaul.  !

l l o Unit 1 Auxiliary Building Ventilation System damper 2A gasket i replacement.  !

[

j i

l c Control rod drive mechanism repair activities. -

l o

i Upper internals clearing operations for reassembly. t o

Repairs associated with reactor coolant pump lubricati_n system l cracking.  !

. i l

One violation and no deviations were identified.

6. Surveillance (f.1726)

).

l

! By the direct coservation and record review of selected surveillance testing, inspectors assured Compliance with I$ requirements and plant orocedures. The inspectors verified that test equipment was calibrated, i

and acceptance criteria were met or appropriately dispositioned. ,

i Su a

I -illarce activities examined during this period included:

i i o Antes 1 leak rate testing for Unit I containment described in l

sect 3 4 i I

o i

Surveillance testing of the ASW/CCW problems identified in section I 13.c. of this reLort.

i q c I

Inservice inspectic.1 testing, section 12. of this report.

j o j Surveillan.e testin!; of the main steam line radiation monitors i described in se tiel 4.a. of this report.

I 4 o Ciesel fuel oil sampling surveillance discussed in section 4.1 of j this report.

i No violations or deviations were identified.

, 7. Engineering Safety Feature Verification (71710) i i

lhe inspe- or walked down accessible portions of the Units 1 and 2 Auxiliary Saltwate; system including local and control room indication i

a i

1

.,.r,-,..- -+v.,# .- - . . . - - - , - - - - . . . . - - s.-~.-.

f,

}

u and system brep ers.

report. Findings are discussed in section 13.c. of this So violations or deviatior.s were identified.

8 Fadiological Protection (71709)

The inspectors deterrine -hether pericdically observed radiological protection practices to the licenset's program was being implemented in t

conformance with facility regulatory requirements. The policies and procedures and in compliance with supervisors and ptofessionals conductedinspectors verified that health physics activ'ttes f r eq ent plant tours to observe activiti(s, in progress and were generally aware of sigrificant plant chalterge. particularly those related to radiological conditions and/or i

PWP /.LARA consideration was found to be an integral part of each '

(Ridiiticn Work Permit).

fu violations or deviations were identified.

9 Physical Security (71881)

Security activities requirements, were observed for conformance with regulatory implementation administrative procedures of the site security plan, and screening, personnel badging, including vehicle and personnel access reasures, and protected and vital areasiteintegrity.

security fo.ce manning, compensatory checked during bac6 shift inspections. Exterior lighting was No violations or deviations were identified.

10.

Licensee Event Report Follow up (92700) a.

Statos_of LEPs Based on an in office review, the resident inspector: the fo', lowing LERs were closed out by Unit 1:

87-10, 87-16, 87-19, 87-23, 88-02, 88-03, 88-06, 88-12 Unit 2: 87-04, 87-14, 87-23 The LERs were reviewed for event description ,

root cause, corrective actions taken, generic applicability and timeliness of reporting.

h wtolations or deviations were identified.

11.

C.nen Item Fallow up (92701)

a. CC Inspector Failing To S 50-275/88-03-03; Closed) erform Inspection (Enforcement Item The inspector reviewed the licensee's response to a Notice of Violation issued on March 28, 1988 concerning a Quality Control

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

12

.~

Inspector who statped and initialed his acceptance of cleanliness on his inspection plan without visually inspecting inside the body of Valve No.'84640 for cleanliness.

The inspector reviewed the corrective actions taken and found them acceptable. Therefore, this item is closed,

b. Un3uthorized Entry to the Radiological Controls Area (Enforcement item 50-323/88-04-01; Closed)

The inspector reviewed the licensee's response to a Notice of Violation issued on March 28, 1985 concerning the unauthorized entry ,

of an individual to the Radiological Controls Area (RCA). In their l response, the licensee stated that the individual was counseled by his supervisor.

In addition, a revien determined that the existing RCA postings, procedures and training program were adequate.

However, the postings for the RCA we"e clarified to more clearly denote entry and exit points, and the barrier support was improved to reduce the amount of sag in the yellow magenta rope delineating the RCA. Based ti these actions, no further actic,.s were deemed necessary, i

Ine inspecter reviewed the actions taken including the changes to the RCA barrier and found them acceptable. Therefore, this item is closed.

c.

Revisions to Procedures Controlling Maintenance Performed on E,nergized Equipment (Follow up Item 50-275/87-04-03; Closed)

In response to findings in Inspection Report 50-275/87-04 with respect to inadvertent control rod withdrawal due to a miscom unication between 1&C and Operations, the liceasee committed to revise procedures for control of equipment required to be energ):ed during maintenance. The inspector reviewed Tagging j

Requirement Procedure AP C-751 trat information tags placed on equipment be document d which had been revised to require for installation and removal. The inspector also reviewed work orders i for systems required to be energized during maintenance and found that they required the technician to sign off that the Shift Foreman had completion.

been notified prior to performing the work and following in addition, the work orders specified where information tags were to be hung and required their removal following maintenance. Based on the above, Open Item 50-275/87-04-03 is closed.

d.

Protected Area Escort R:sponsibility (Enforcement Item 50-275/87-44-01, Closed)

NRC Inspecticn Reports 50-275/87-44 and 50-323/87-45 cor,tained a violation regarding plant security. In a March 10, 1988, letter (CCL-SS-058) PGLE addressed the identified security concerns. The inspector reviewed the licensee's corrective actions and determined them to be acceptable.

Accordingly, this item is considered closed.

Details pertaining to the corrective actions are not provided in

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

s i .

!. 13

• this report due to the security safeguards nature of the information, e.

j_goperable Unit 1 Rod Position Deviation Monitor (Open Item 50-275/P7-38-022_ Closed)

Open item 87-38-02 the "P-250 Rx Alm Axial was concernad with root cause determination of clearing during a plant evolution. Flux / Rod Pos" alarm window unexpectedly the cause of As explained in LER 87-19-01, temputer the problem was identified to be in a subroutine of the program which controlled alarm functions. The subroutine was found to were ini tialized function unpredictably if the rod bank demand values improper',

the LER. Corrective actions were described in This item is considered closed.

f.

Entry into Techni,.3' Specification (TS) 3.0.3 (Open Item 50-27L/87-38-03, _ Llosea)

This open fuse failures iteminwas concerned control with the root Cause determination of rod drives. As described in LER 50-275/87-16-01, fuse failure was attributed to poor solder connections at the fuse end caps.

in the LER. This item is consideredCorrective closed, actiores were described g.

Digital Electro-hydraulic Control (DEHC) System Halfunction (0 pen item 5'-275/87-04-01, Closed)

Ihis open control item involved inadvertent disruption of the DEHC load (software)

As corrective action, program during turbine maintenance activities .

C-3:II " Main Unit Turbine-Startupthe licensee revised Operating Procedure outage or any major to add a caution that after an should be reprogrammed. turbine maintenance, the P-2000 computer (DEHC) closed. Accordingly, this item is considered

h. Manual Valve Maintenance (0 pen Iten. 50-275/e7-01-03, Clot ; j}

A 1987 NRC team inspection identified manual valves which had not been program. greased or maintained by a preventative maintenance (PM)

In discussions with the team members licensee m;nagement

indicated the Operations Department would identify valves needed to be operated during accident and recovery periods, and these valves would be entered into a PM program.

The licensee concluded the necessary safety system valves were included in the existing sealed valve Checklist").

valve checklists (Operating Procedure K-10 " Systems Requiring Sealed to include stroking and lubrication of all sealed, once 18 months.

valvesThe every inspec This item is considered closed.

No violations or deviations were identified.

+ -- , ---

6pe--

14 '

I?. Inservice inscettien (73051)

Eeveral different rethods of nondestructive examination were observed by the inspectors. These included liquid penetrant examination (previously written up i n hRC Inspection Peport 87-42), A-scan ultrasonic examination and visual examination. TFo inspector witnessed ultrasonic examination of a Unit I reactor pressure vessel stud. The required equipment and raterials, specified in licensee procedure N-UT-3 " Ultrasonic Examination of Colting with Diameter 1 Inches or Greater," were observed to be in use, and the specific area, location and extent of the examination was clearly defined. The inspector observed personnel perform a cualification test on a calibration stardard made from a spare vessel stud, and observed ultrasonic ecuipment calibration. Transducer size, '

frequency, and type were in accordance with the procedure, and reject, de" Ting and filter settings were set at minimum values. No indications in tne stud examined were detected.

Tre inspector aise cbserved the licensee perform visual inspection of support 15-95 on the sucticn piping to RHR pump 1-2. Examinatior vf the rigid support and PSA-10 snubber was performed in accordance with ISI Procedure VT 3/4-1 " Visual Examination of Component and Piping Supports".

The as-found condition of the rigid support and snubber was acceptable, hc-ever, procecural discrepancies were found. The " Figure 1" and " Figure 2" labeling was missing from the drawings of page 12 of revision 4 of the procedure.

The " Hydraulic Snubbers" (Figure 1) diagram on page 12 centained the s ta tement ". .. subtract the 'Z' dimension. ..from the measured position setting." This statement conflicts with training provided to the ISI examiners. The drawing on page 14, above Table 1, was not clear to 151 personnel interviewed by the inspector. This drawing should be revised for clarity. Finally, on Attachment 1, page 3 l

of 3 the " post installation verification of snubber / strut washer placement" contained check-off boxes such as " thickness, 0.0. acceptable" and " remaining gap acceptable" without the procedure containing guidance on how to measure the parameters or what criteria was being used.

The ir'"ector was inforred the post installation verication was not a code recuirement. The licensee was made aware of the procedural discrepancies and plans to correct the procedures.

Code repair activities observed by the inspector, were previously datumented in NRC Inspection Report 88-07.

ho v tolations or deviations were identified.

13. Independent Inspection

a. Systen Engineering (5-37700-4) ine licensee is in the formative stages of establishing a system ,

engineering function, and has conducted information gathering

' meetings with other Region V utilities. Discussions with licensee management have not established a projected completion date for the establishment and implementation of this program.

., . ,r-- y - - - , -

l l 15 .

i

b. Post-trip s

( 5-W N -t)Posiew,_Eyents Evaluation /Roet Cause Determination

, Durirq the peiicds January 21-22 and April 20-22, 1998, the above arras were eramined by the Senior of Reactor Engineer, RV. The scope f i ndir:gs a re di scus sed below:

1) [ost-trip Review Pevision 3, cated -July Plant29, Administrative Procedure AP A-100 51, the implementation of this procedure 1985, was examined and records of were examined. for three reactor trips plant Discussions relating to AP A-100 51 and related records were held with licensee representatives and the NRC Pesident Inspectors, from which cbservations resulted: the following findings and Administrative in terms of Procedure AP A-100 51 was judged to be adequate docu~entatiorthe scope of post-trip review, evaluation and evalu' tion of plant and operatorprovides The procedure for review and response as well as the authorization of plant restart The procedure (by the Plant Superintendent).

includes the requirement that, under circumstances where the cause of a reactor t rip is not adequately erplained or where the Shift Foreman determines aaditional Staff analysis is necessary, prior to restart the Plant dat3 and Review will acprove Committee wile review the associated transient return to power operation.

Discussions with the Resident Inspection staff revealed instances where thoroughness of post-trip revi ew w as lacking in the implementation of the AP A-100 51.

documented in recent NRC Inspection Reports.These instances are inspection The Resident process staf' has also evpressed concern regarding a formal "rior to for defining plant and restart.

documenting specific a:tions rtautred In response to the Resident inspector's p ro g ram concerns, licensee management has implemented a for action plan development and implementation. (Saa section 16.b of this report for licensee management commitments in this regard).

2) f vents Evaluation and Root Cause Determination - In evaluating the licensee's programs in these areas, the following plant Quality Assurance and Administrative Procedures (APs) were e=amined and discussions relating thereto were held with responsible licensee representatives. Findings and cbservationsheld discussions resulting with from the examination of procedures and belom licensee representatives are discussed QAP 15, B , Nonconf ernances , Revision dated Ma rch 10, 1988 NPAP C-12/NPC-7.1, Identification and Resolution of Problems and Nonconformances, Revision 13, dated March 22, 1988

,_ n n ., w . - -n - .-

~

l 16

. *  ?

NPAP C-16/NPG-7.4, Human Performance Evalut ion System, Revision 1

0, dated March 3, 1986 hDAP g.j C-18/NPC-7.5, Events Investigations, Revision 0, dated July 14, 1987 M

)},

NPAP C-23/NPG-7.6, Technical Review Groups, Revision 0, dated

arch 10, 1908 r

A review of the above procedures, related plant records, and discussions with responsible plant managers and supervi w s resulted in the follow.'ng cbservations and findings:

The licensee has implemented a very effective Human Performance

$ Evaluation System (HPES) program, having been an active F pa".icipant in this INPO program from the time of its

' initiation some two years ago. This program is intended to I focus on human factor elements of plant events, and is aimed at surfacing for evaluation human factors concerns at a low 1

thres ald, e. g. "near misses" The pregram has an outreach g

a aspect, wherein employees at the plant are encouraged by direct nailings, posters (with associated forms to submit written h concerns),

corporate offices.

etc. in several locations within the plant and d During the year 1987, a total of 39 HPES 4 root cause evaluations were performed relating to various cperational/naintenance events. Approximately 25 of these were 9

) in support of the dispositioning of Nonconformance Reports (NCPs).

h f

' l The licensees procedures require formal root cause determination 135 during the for yearall1987.

NCRs, of which there were approximately  ;

f- When an additional approximately 15  !

HPES evaluations for root cause determination are added to the i

{ll number of NCRs, a total of approximately 150 events were G subjected to formal root cause determination in the year 1987.

l In discussions with the NRC inspector, the Plant Manager erpressed his view that the threshold for formal root cause determination should be lowered to include a larger population of events beyond those for which an NCR would be initiated in

, accordance with current administrative procedures. (See Exit and Management Meetings section of this report for licensee management commitments in this regard),

c.

Design Verification and Conficuration Control: The Auxiliary Saltwatec System (5-37700-1, 37700-2)

The inspector reviewed the Auxiliary 'aaltwater (ASW) system with respect to its plant.

the operating design basis and how that design is implemented in weaknesses; The inspector identified the following o

The design basis assumptions for the ASW system have not been

'ully implemented into plant procedures and alarm setpoints, l

1

17 As a result, plant operations have been conducted cuts: c des ign t.as i s assumptions requiring a review of the ASW system's past operability.

n The licensee did r.ot have an adequate program for design setpoint control.

t*e As a result, the annunciator setpoint for of differential pressure (dP) high alarm across the tube side t"e Ca*ponent Cooling Water (CCW) heat exchanger (Hx) was raised withcut the appropriate design basis review.

T*ese Cond findings a e mitigated by the licensee's current efforts in iguration Panagement.

Although at the time of this report the licensec's by described program the was in 4's development stages, the program, as licensee, would establish how design requirements and assu 7tions are to be implemented through plant operations, raintenance, and surveillance. In addition, it would establish precedaral guidance for setpoint control.

Svs'em Descriytion and Design Basis The Aid system is the ultimate heat sink, designed to cool safety related accident.loads during normal operations a.d following a design basis aischarge The system consists of two pumps headered at their through two located at the intake structure. They pump ocean water trains of 24" piping, up 85 feet over a distance of approximately the discharge of 1600 feet and through the tubes of the CCW Hxs At .

level and cascades to the ocean.the Hus the ASW is discharged at 68 feet above The tube side of the CCW Hx has a differential in the control pressure room. transmitter with a high and low annunciation The inspector design reviewed and discussed the ASW design with the system engineers at the licensee's office in San Francisco. The licensee could not provide the original design calculation Much of .

the original complete design records were took notplace kept. in the late '60s and early '70s when The syste~ was assembled around 1973 and tested in 1974 and 1975. In 1982, during the design serification program (DVP), the licensee erformed calculations L:ased on as-built conditions to verify the ASW system could meet design basis. its The limiting parameter for the ASW system was determined to be CCW (tCCA). The following a design basis Loss of Coolant Accident temperature chtrging pu p liriting component was determined to be the centrifugal F for 20 rinutes, lobe oil coolers which was rated at up to 132 degrees below It was determined that containment could be kept allowed of five containment temperature and pressure limits during a LOCA with two fan cooler units (CFCUs).

Licensee calct.lations M-3C5 Revision 3 assumes the following:

o An initial ASW temperature of 64 degrees F. Above 64 degrees F ocean of both temperature, Hx. the Technical Specifications require the use

. 18 o A pre-LOCA CCW terperature of 80 degrees F. This is based on the marimum normal CCW loads.

o The use of five CFCUs. All five CFCUs start on a Safety Injection System signal. Operator action would be required to shut down a CFCU at it's breaker, o ASW fic* cf 10,700 gpm which is based on flow taken from the manufacturers pump curve assuming "eean low-low water" level of

-2.6 feet mean sea level (MSL) and the Hx tube outlet at atmospheric pressure, o A fouling factor, used in the heat transfer coefficient of 0.001.

The results concluded that given these conditions, one train of ASW canremovethepost-LOCAheagaddedtotheCCWsystemwithouthaving the CCW outlet exceeding 132 F. The licensee did not take credit for any operator action.

Design Basis vs Plant Configuration and Procedures The inspector reviewed plant configuration and procedures against the above design basis assumptions. The following is a summary of the discrepancies found:

o The Hx dP HI alarm setpoint was 167" water whereas a clean Hx dP of 75" water was assumed in the design calculations. The following section discusses this finding in more detail.

o The Inlet bay low level alarm was set at -10' MSL whereas a level of -2.6' was assumed in the design calculations. The effect of a lower inlet bay level would be to lower suction head and consequently discharge head resulting in less flow.

o ASME Code Section XI allows pump performance to drop to 10% of its reference whereas the design ca'...lations took pump performance from the pump curve without allowing for degradation, o The CCW Hx shell side outlet temperature high alarm setpoint was set at 120 degrees whereas the highest normal operating temperature was assumed to be 80 degrees. If during normal cperations CCW temperature rose above 80 degrees, the unit would be operating outside design assumptions, o

Plant fail Procedures address actions to be taken if both ASW pumps (cross-tie with other unit) and if CCW pumps fail (reduce system heat loads such that CCW temperature is less than 95 degrees) but not actions to be taken if one ASW train does not provide sufficient cooling.

1 1

I l

i

\

l 19 I

o Plant precedures did not specifically state that cperators could re ove from service CFCUs during a LOCA to remove heat loads from the CCW system. '

o I Annunciator Resp;nse Procedure PK-0101 in sten 7a. allows  !

i operators to thrcttle the CCW Hx tube side cutlet valve if ASW j ru a dP is less that the Section XI limit. The procedure did not have operations notify engineering to evaluate the i eperability of the pump. {

• • e 'irst three l a t: ti'ty to perform findings listed raised questions of the ASW system's than assu ed in its its f unction design basisunder c onditions less conservative calculations.

It e inspector discussed these findings with the Project Engineer for Diablo Canyco who committed to provide a written analysis of ASW scstem operability to the NRC by June 7, 1988. Pending a review of the analysis this item is Unresolved (0 pen Item 50-275/88-11-02).

Ir+se findings also show that many design assueptions were not incorporated into plant operations.

ASW system, the As corrective action for the need to be licensee plans to establish what design assumptions instru entation irplemented and revise procedures, alarm setpoints, and documentation as necessary. To address these ccncerns on a larger scale, the licensee had initiated a Can'igurat ion Management program in November 1987.

the 11censee, As described by implementation in plant operations.this program would address the issue of Although the significance of these findings implementation as related to general design basis understanding and continued attention needs to be focused on this issue.is mitigated Sgtraint Control ,

The for dP inspectcr investigated the basis for the annunciator setpoint was determinedacross the that CCW Hx tubes, pressure switches PS 45 and 46. 't established the setpoint of 167" of water had been in March 1987 following a design change to install pressure change had been transmitters and switches with a higher range. The design Hr initiated in 1985 by the operations department since fouling dP across the Hxs was routinely above the existing setroint of 110" during normal operations. The engineering reviewers of the design change erroneously determined that the change did not affect equipment important to safety or equipment important to environmental quality. In the general notes contained in the design change package Project Engineering authorized C;erations to revise the setpoints for PS 45 and 46 but did not give them up by specific revising guidance drawing except to state that Operations should fol:ow a field change. 101938 (Non-Safety Instrument Setpoints) with Crerations revised the setpoint from 110" to 167" basing the revisionflow raximum on a calculation of only one limiting condition; the velocity through the tubes. The flow velocity

_ - - = . -

. - - =

20 according to the vendor should be kept below 7 feet per second; 167" correlates to 6.8 fps.

U;:o r subsequent investigation, the inspector found that safety related Drawing Nos. 060836 (for Unit 1) and 061236 (for Unit 2),

"Instru ent Setpoint Requirements" Table II lists the high alarm s e t p o i r. t for P5 45 and 46 to be 4 psid which corresponds to 110.7".

The cover note to the drawing states " Table II of this drawing lists other non-instrument Class lA setpoints which engineering has deterrined to be appropriate to meet various FSAR commitments."

This design drawing was not reviewed or changed when the setpoints of PS 45 and 46 where changed. This is a failure of Engineering ea' to reevaluate the basis f or the original setpoint and is an apparent violation of criterion III, " Design Control," of 10 CFR 50 Appendix B but will be treated as unresolved until the significance of the ASW/CCW systems operating with a 167" differential pressure setpoint is resolved. Following the meeting of the Technical Review Group for the ASW system Non Conformance Report, Operations put an administrative limit on CCW Hx tube side dP of 110" pending the resolution of the basis for the 110" setpoint. Subsequently, it was determined that the dP setpoints in Drawing Nos. 360836 and 061236 to control the low alarm setpoint satisfied the FSAR commitment for

) a control room alarm on ASW piping failure. Regardless, system perfornance is directly effected by Hx fouling and requires setpoint control.

l The licensee was in the final stages of a comprehensive revision to the setpoint control program at the time of this finding.

These revisions appear adequate to ensure that important setpoints are reviewed against the design basis.

d.

Cleanliness Control Problems (5-92700-4)

In previous resident inspector report (Inspection Report 50-275/88-07), two cleanliness problems were identified during the performance of refueling outage work. The two areas examined l

previously were the removal of thermocouple connoseals on March 21 I and spare control rod drive mechanism work on the removal of the

~21ctor vessel head on April 6, 1988.

During this reporting period the control of cleanliness problems centinued. On April 9,1988, quality control (QC) personnel i ssued a

stop work on CRDM cleanliness requirements. The stop worx was lifted later that day after corrective action was taken.

The action consisted of erecting barriers around the refueling cavity that were shown later to be ineffective. Additionally a memo was issued by engineering to the engineering task coordinators regarding cleanliness controls. Subsequent events showed that this memorandum was ineffective in precluding further occurrences.

On April 12, 1988, QC inspectors identified that cutting fluid and chips were being allowed to enter crevice areas on the reactor vessel head. Accordingly, a stop work was issued.

the licensee implemented corrective actions. Subsequently, l These corrective actions consisted of cleaning the crevices and revising the procedure for cutting to include a QC holdpoint to verify barriers

/

_ _ - - _ _ m _ ____m_ ___ _-__

21 .

were installed.

re!nstructicn evenCorrective actions did not include personnel note requir: though the procedure used had a specific caution crevices. rg steps be taken to preclude fl uids from entering the On April 22, 1958, during the attempt to reinstall the upper internals, sighting of work was sicpped by the refueling crew due to the reporte1 as debris on the upper internals which was initially tools (pliers, nuts, and washers). The debris was retrieved and determined to be a broken " tie wrap" (a plastic strap ordinarily used to secure electrical cable to cable trays) and paint chips.

The inspector attended the April 22 1988. licensee's corrective action meeting on assigned debris on the the upperresponsibility internals. to deternine the probable sou The engineers debris retrieval,inbut charge rather of had the job did not "save the evidence "upon was retrieved by the licensee and the iaspector obit placed in radioactive It w tie i, rap served that the the paint chips were yellow paint. looked old (yellowing in color as o the tie wrap prebabiy care f rom the reactorThe conclusion drawn was that cable trays. The inspector v;.sel head and its the reactor vessel then examined the work area on top of The head and noted several unsatisfactory conditions removed head was stored immediately adjacent to the refueli ng .

cavity; nost of the components on the head do not hang over the cavity, but a portion of the cable tray area does hang over the pool.

head area cable tray.The tie wrap found on the internals was directly u wraps in the cable tray area which had the potThe ential to fall.inspector found a Additionally, on the upper area of the head (where work had been stacks for CRDM weld repair access) of dirt th(underway to remove an inspector found great deal abandoned(up to 1/4" thick) including broken microphone ceramics in place since pre operational testing. The engineer in charge the of that work explained that prior to removing any DRPI coils local area dislodged wouldaround fall the DRPl coil was vacuumed , and that any dirt cavity. However, straight down and not into the refueling steel plates in thathe further explained that one of the interlocking bounced and was yetoff to be a retrieved.

strutture, and ended up in the refueling fell, y pool, cavitsame debris would only fall straight down appearedu toe.be fa lt dTherefore The inspector discussed the cleanliness situation with ethengineers in containment and with the outage manager that evening.

were recleaned and verified clean prior to recommencingAll areas asse-bly. reactor e.

we - - s + - - -e w - - - - - - - - - _ _ _ _ _

1 i .

22

(

f 4

l On May 10. 1988, licensee p; ;onnel identified cleanliness control deficiencies in the Unit 1 Spent Fuel Pool including an incomplete tool leg. Corrective actions consisted of completing the tool log. i

( Cleani f ress control problems were identified by the NRC from March 21 to April 22, 1988. Additionally, QC personnel issued two stop work 5 on the same subject and licensee identification of problems f continue.

4

) The licensee's actions up to the point of the inspectors involvement were ineffective in that they did not identify additional debris on the reactor vessel head whicn could be easily dislodged and find its i

way into the refueling cavity and possibly reactor vessel. This is

' a significant condition, because debris in the refueling cavity or reactor vessel could impact reactor operations and fuel conditions.

i This was true despite remorandums of instruction by the engineering 3

manager and increased QC surveillance. The failure to take timely ef fective corrective action to preclude recurrences of cleanliness l deficiencies is an apparent violation of 10 CFR 50 Appendix 8 1

criterion XVI (Item 50-275/88-11-03).

e. Pressurizer Surce Line Movement Trojan Nuclear Power Plant, located in Region V, has experienced movement of the pressurizer surge line possibly due to thermal stratification.

The resident inspector contacted the responsible engineer at Diablo Canyon to determine if evidence of movement or lack of it was available for Diablo Canyon. The licensee had taken measurements of the pressurizer surge line in Unit I relative to structure in 1983, 1986, and during the current refueling outage. Review of the measurements showed essentially no movement of the pressurizer surge line relative to structure. At the close of the inspection report period the licensee indicated that some evidence such as pipe burnishing indicated that in the hot condition th; pressurizer surge line may be contacting pipe whip restraints. The licensee was analyzing the findings, considering the addition of inservice instrumentation to detect thermal stratification, and planned to pursue resolution with Westinghouse. Tne licensee's resolution will  !

be followed as open item 50-275/88-11-04).

One violations and no deviations were identified 14.

Containment Integrated Leak Rate Test (ILRT) (70307 and 70313)

a. Procedure Review The inspector reviewed the Unit 1 and 2 ILRT proce9ures as described in the licensee's Surveillance Test Procedure STP M-7, Revision 7 of May 5, 1988, (and the Temporary Change Notices issued during this inspection) entitled, " Containment Integrated Leakage Rate Test ILRT), Type A." This review was to ascertain compliance with plant

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

23 -

Technical Specifications, regulatory requirerents, and applicable industrial standat ds as stated in the following documents:

o Diablo Caryon Pcwer Plant Units 1 and 2, Updated Final Safety Analysis Report (FSAR), Sections 3.8.1.7.2, 3.8.1.7,4, and 6.2.1.4 o

Diablo Canyon Power Plant, Units 1 and 2, Technical specifications, Section 3/4.6.1.2, " Containment Leakage", and 3/ 4. f>.1. 6, " Containment Structural Integri ty. "

o Appendix J to 10 CFR 50, " Prima y Reactor Containment Leakage Testing f or Water Cooled Pcwe- Heactors."

o A erican Naticnal Standard, " Leakage-Rate Testing of Contain ent Structures for Nuclear Reactors," ANSI N45.4-1972.

o inpical Report BN-TOP-1, Revision 1, " Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures dated November 1, 1972.

for Nuclear Power Plants," Bechtel Corporation, o

A~erican National Requirerents," ANSI Standard, /ANS-56.8-1981." Containment System ' eakage Testing o

IE Information Notice No. 85-71, " Containment Integrated Leak Rate Tests."

Durirg this procedure review, ebservations: the inspector made the following The proced"re requires the containment liner weld channels to be vented required. to the containment atmosphere during the tect, as is The inspector noted that, at other plants, these channels have not been vented during the test and additional safety review by the Office resolve thisof issue. Nuclear Reactor Regulation (NRR) has been required to There is a discrepancy in the procedure concerning the test acceptance criteria.

Section 5.3.3 of the procedure states that, accordance with the provisions of BN-TOP-1, Rev. 1, the end of testin 95% upper shall be lessconfidence than La. limit (UCL) for the calculated leakage rate However, in Appendix F, on the " Acceptance Criteria La. Check Form-Cata Sheet," the limit is 0.75 La, rather than 50, The NRC's position is that the regulation, Appendix J to 10 CFR recuires the acceptance criterion to be 0.75 La, as does the NRC's acceptedTopical EN-TOP-1. Report Evaluation, dated January 15, 1973 , which For the present test, the acceptance criterion of 0.75 La was in fact satisfied. Nevertheless, the inspector informed the acceptance criteria licensee requirements. that section 5.3.3 was inconsistent with  !

Section 5.4 and Appendix F of the procedure also specify that, for a 24-hour duration full pressure test according to 10 CFC 50, Appendix l

n . - - . . -

. 24 J and ANSI N45.4-1972, the calculated leakage rate shall be less than 0.75 La. However, section III.A.3.(c) of Appendix J to 10 CFR 50 requires the calculated leakage rate to be corrected for error.

Although r.o pa-ticular method is generally required, many licensees use a 95% UCL, similar to the BN-TOP-1 procedure, to account for error. For the present test, the BN-TCP-1 procedure was used. The licensee has rarked up the procedure with associated clarifications t- Le included in the next normal revision,

b. N iew of Pecords The inspector reviewed calibration records for the instrumentation used in the ILRT. That is, the twenty-four resistance tempera' 2 detectors (RTDs), six dew point temperature sensors (dew cells), and two pressure gauges used to reasure containment air mass, and the ficw ele *ent used to measure the induced leak during the verification portion of the ILRT. All instruments had been calibrated within the last six months with NES traceability. In situ checking of the instru entation had been performed within one Forth of the test.

Although the procedure did not provide instructions for containment te perature survey before the test to verify temperature sensor locations, such a survey was conducted, as observed by the inspector and discussed in the following section. The inspector requested that the survey results be included in the licensee's test report to t*se NRC, which is due within three months of ILRT completion.

Because a temperature survey will probably be performed on Unit 2 in preparation for the Unit 2 ILRT planned for Fall 1988, the licensee sFould consider developing a written procedure for this activity.

c.

Caservation of Work and Work Activities Prior to the ILRT, tha inspecto" observed a portion of the visual J inspection of the Inner surface of the containment, including the I containment liner. No evidence of structural deterioration, apparent found.

changes in appearance, or other abnormal degradation were i l

Ihe l inspector observed the containment pressurization equipment, i consisting of eight air compressors, two after-coolers, two air cryers, and connecting hoses and equipment. During the containment  !

pressurization phase, two of the air compressors were l j

out of-service, which somewhat slowed containment pressurization.

i Also, to work.

during most of the pressurization phase, one air dryer failed  ;

The tcsulting higher moisture content of the air entering containnent, nay have contributed to high relative humidity in the containment which apparently caused water condensation on dew cell No.

2 at the end of the test (during the verification phase). This l was the apparent cause of erratic readings which resulted in removal of the dew cell from service. This is discussed further below.

i l

1 The inspector witnessed a portion of the pre-test containment terperature survej. Two surveys were actually performed; one with 1 i

I 25 .

l i

1 the centainment fan eccler units running, and one without.

This information gave the licensee the option to either run or not run couldf an the be coolers during the ILRT, as the validity of RTD placement i confir.med.

the During this ILRT, the licensee chose to not run or heat fan coolers, as running then introduces additional heat sources which are difficult to control. sinks (depending on cooling water flow and t The licensee stated that the te perature weighting survey did confirm the validity of RTD positioning and factors.

included in the licensee's The inspector requested that the survey data be i

report to the NRC.

! The inspector witnessed selected portions of the following ILRT e

activities listed belc nd acted the time expended to perform each: i o

i I initial hours. pressurization to 47 psig + 2/-0 psig, approxinately 12 o ILRT I stabilization, approximately 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

o ILRT data acquisition. t 1

o Performance of test, asILRT, approximately 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />, including a failed initial discussed below. '

o

] Leak hour. rate verification test stabilization, approximately 1 c

i j

Leakage imposed rate verification test, approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> , with an 4

i which equals La, which is 0.1% per day. leak rate of 7.5 ,

Various electrical and mechanical penetrations were inspect e. d Decause crained both the typical containment isolation valve was vented and i inside and outside containment, l to fit balloonsinflation so that balloon over thewouldends of the indicate vent lines outside leakage or *c" nment, contaith containment the i isolation valve seats. The licensee .

1 t'alloons approximately every two hours during the 6sstkedfor excessive ,

these device. leaks, but did not find any through the use of this During the test stabilization period, RTD No.

21 failed high, suddenly been reading reading in the12160s.degrees F where it and other nearby RTDs has during the same period. Dew cell No. 3 exhibited erratic readings set Both sensors had their weighting factors cther nearby sensors for the duration of the test.to o zero an A few hours after starting the ILRT itself, it became apparent that the containment was leaking excessively.

the measured leakage rate (Lam) had stabilized at a value ofAfter about approximately 0.118% per day, whereas the acceptance criterion 0.75 La. equaled 0.075% per day (La 0,1% per day). Licensee personn,el searched exhaustively for leaks using soap bubble solution (Snoop)

26 l

• .I I

and other rethods, Eventuallv they found that at one of the 48-inch j

1 pu ge line penetrations, there was a preasure of 47 psig (or current containment pressure) between the two closed isolation valves. This indicated that the valve inside containment (RCV-11) was either not closed co pletely or was leaking very badly. However, during the ILRT, a few the valves had been locally (type C) leakage rate tested cnly days earlier and had passed that test easily. The valve cuiside containment (RCV-12) was found to ha.e significant packing

, leakage and some seat leakage.

A r.c t h e r purge isolation valve outside containment (FCV-661) in a different penetration was also found to have significant packing leakage, which would indicate a leaking inside containment isolation valve on this penetration.

Atnut 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> into the test, the licensee opened a vent valve (approxi-ately one inch in diameter) betweec RCV-11 and -12, in an attempt to depressurize the space between the valves. After approvirately 15 minutes, the vent valve was closed and the attempt abandoned, because the pressure between the valves had not decreased more than a few psi.

not This confirred that valve RCV-11 ,as indeed limiting leakage in any substantial way.

Subsequently, the licensee took actions to eliminate or reduce known leaks, primarily by tightening down on valve packing. When that did not reduce leakage sufficiently on valve RCV-12, the licensee took the unusual sten of adding one or more additional packing rings on the valve stem and tightened down on those. This step nearly eliminated packing leaks on valve RCV-12.

When the licensee took actions to reduce containment leakage rate by repairing, adjusting, or altering the containment pressure boundary, this caused the test to be considered a failure, in accordance with section III.A.I.(a) of Appendix J to 10 CFR 50.

In other words, the contain ent was leaking in excess of the allowable limit, and the only way to pass the test was to take steps to el4-inate leaks.

The licensee's procedure STP M-7, Rev. , also refers to this circumstance as "the initial unacceptable ILRT," which must then be followed by another, successful ILRT.

After reducing leaks, the licensee restarted the test (or started a new test) at 8:44 p.m.

on May 19, 1988, approximately 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> after the initial start of the test. Using the methodology of BN-TOP-1, I

the then test was successfully completed some 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> later. There was some delay in establishing the superimposed leakage rate flow out of the containment for the supplemental or verification test.  ;

l The licensee has run approximately 200 feet of small-diameter (0.75 inch) plastic tubing from a containment penetration to the two ,

Volumetrics thermal mass flow meters installed in instrumentation )

cabinet in the DAS (data acquisition system) shed. This long,  !

narrow tubing could only pass approximately 5 scfm, short of the needed 7.5 scfm. Therefore, the licensee resorted to a backup mechanical rotometer which was placed close to the containment penetration to allow the needed flow. With this delay and the

27

• required (by BN-TOP-1) one hour stabilization period, the verification the verificaticn test was test,started at 11:29 a.m. on May 20, 1988. During dew cell No. 2 exhibited erratic behavior "ich criteria. was appearing to cause the test to fail to meet its acceptance j When the licensee zerced its weighting factor and reassigned the original weighting factor to other dew cell s, the veri fication test passed.

was that The inspector's preliminary conclusion this action was acceptable, but, after the inspector completed his inspection andbecause it took place review cf left the site, NRC contained the licensee's justification for zeroing dew cell No. 2, final acceptabilityin the licensee's report to the NRC, will determine the of the action.

The inspector performed an independent computer calculation of leakage rates to verify that the licensee's computer program was correctly calculating leakage rates. The inspector's calculations did indeed verify this.

The liconsee's test, preliminary results for the final 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Type A which did not include Type C additions, was a total time calculated leakage rate of approximately 0.02% per day with 95%

upper confidence limit (UCL) of approximately 0.071% per day. The licensee's 0.075% maximum allowable leakage rate (0.75La) for this test was per day.

An approximately 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> verification test was performed with an imposed leak rate of approximately 7 5 SCFM or 0.1% per day of containment air mass. .

test produced a total The licensee's verification time calculated leakage rate that fell within a

the day. test acceptance criteria of approximately 0.095 to 0 145% per .

acceptance Thesecriteria.

preliminary results appear to be within the allowed

d. Conclusions At the exit meeting held on March that 20, 1988, the inspector stated section the Ill.A.failed initial test would require, in accordance with 6 of Appendix J to 10 CFR 50 and facility Techaical Specification 4.6.1.2.b.

test be reviewed and approved by the NRC.

, that the schedule for subsequent Type A test failures should occur, then a Type A test shall be performedIf each at two first,plant refueling outage or every 18 months, whichever occurs test schedule may be resumed.until two consecutive Type A tests pass that, However, the inspector emphasized few in cases specific (like this penetrations, the test) where failure NRC encourages can be the licensee to attributed to prcpose, as a formal exemption from the regulation, a corrective action increased plan which Type A test would address the problem penetrations in lieu of frequency, Such exemptions are judged on a case-by-case basis and are not automatic; it is also unlikely that the frequency licensee would schedule, andbe relieved that fromlikely a test would thehave first totest on the in be passed successfully before an exemption would be granted.

F

~

28

e. Subseauent Information After depressurizing the containment, after completing the ILRT, the licensee conducted a Type C (local leakage rate) test on RCV-11 and

-12, which passed with no repairs or adjustments to the valves. The licensee has preliminarily determined that RCV-11 (inside containment) may have been installed, maintained, and or tested improperly so that the valve leaks excessively during an ILRT (and so would during a LOCA), but not during the Type C test, which is performed by pressurizing the volume between the two isolation valves in the penetration. Thus, the Type C test measures the leakage rate through RCV-Il in a direction opposite to that w...un would occur during LOCA. It has been thought that this

" reverse-direction" testing was equivalent to testing in the "f orwa rd" d i rec tion. The licensee also found that the inside containment isolation in the second purge line, FCV-660, had the same potential problem, as did the congruent valves in Unit 2. All f our valves were declared inoperable, and Technical Specifications were satisfied. Followup will be done under routine inspection.

3 No violations or deviations were identified.

15.

Evamination of Instrumentation and Controls (I&C)

A special inspection was conducted to examine the area of instrumentation and controls. The inspection was performed by an NRC contractor from EG6G Idaho experienced in the I&C area.

The results of the examination are presented in detail as an enclosure to this report.

Areas for improvement identified by the inspector and communicated to the licensee at an exit interview conducted on May 19, 1988, included the following:

c The adequacy of procedures was found to be mixed. Repetively perf ormed surveillance test procedures (STP's) were generally found to be detailed and adequate. There was one notable exception and that was STP I-33B which is performed every refueling outage for time response testing, STP I-33B was poorly prepared despite being in preparation for several months. Loop test procedures were not addressed since they were already an issue which the licensee has laid out a plan to correct. Corrective and investigative maintenance procedures in the form of work orders were mixed in their quality from good to poor.

The licensee was cautioned to ensure that procedure review was enhanced to ensure a critical review prior to issuance and to ensure a conformance to a uniform standard of detail. The licensee stated that additional personnel (5) had been hired to achieve procedure improvements and that plans were in place for revising writers guides for future procedures to satisfy both NRC and INP0 initiatives in this area.

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

29 .

4 4

The issue of poor precedures in the I&C area and the untimely correction inspecticns. of those problems has been the subject of previous The recent e n c o u r a g i r,g.

licensee action to apply resources to the problem is The effectiveness of and timeliness of the licensee's future actions will continue to be monitored in future inspections. '

10.

.E._r i .t_

a.

Routine Exit _(30703]

1 On May 27, 1988, an exit meeting was conducted with the licensee's representatives identified in paragraph 1. The inspectors summarized this report. the sccpe and findings of the inspection as described in

b. Exit and Management Meetings (30702)

Additionally, 22, 1988, in discussions with senior plant management on April the Plant Manager committed to the folicwing:

1)

An action plan will be developed by May 6 1988, and will address the following; o

The current practice for the development of event response action plans, including schedule for implementation, will be incorporated procedure (s). in new or revised plant administrative ,

o l

Criteria will be incorporated in revised or new t administrative procedure (s) to lower the threshold for {

events which will be subjected to formal root cause de t e rm i na t i on,  !

o Specific consideration will be given to revising current Quality Control / Administrative Procedures to require root cause determination in the dispositioning of Quality Evaluation (QE) reports, of which there were a total of approximately 660 in the year 1987.

l l

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

L. sp e cA n A ff y o .,,2 7.5- / P 8 U

' 3cp - 3 z.3 / 5'8 - 10 l

J 1

I itt MAINTENANCE EVALUATION CF THE DIAEt0 CANYON POWER PtANT l

I 1 IfiiRODUCTION An evaluation of the Diablo Canyon Pcwer Plant (DCPP) Instrumentation and Control (!&C) Maintenance Department was performed during the periods '

of March 23 through April 8, and May 3 through May 19, 1989. The i guidelines used for this evaluation were the United States fluclear Regulatory Commission (NRC) Inspection Procedures 52051, 52053, 62704, and {

62705.  ;

of procedures and work orders were identified.So.te areas of concern i this Three primary inspection: areas of lit maintenance activities were the focus of 1)

Are the 11C technicians technically ccmpetent?

2)

Are the procedures used by the l&C technicians good procedures?

l

3) Do the technicians follow the procedures? I Quality Control (QC) involvemcnt with the 11C work activit i Most of the inspection effort was directed at the I&C maintenance i groups

. .portantwhich dealt with the plant protection svstems and w her systems to safety.

Therefore, the caliber of the technicians and of the llc maintenance activities. quality of the work packages were ex 2.0 CAPABILITIES OF THE TECmilCIAtis i

The maintenance and surveillance activities performed by the a

ranner with the technicians displaying an adequate lk required within the work packages. i themselves with the proper background material and systems in gain an understanding of the tasks to be performed and to obtain the necessary activities, tools and test equipment prior to beginning their work I

lhe efforts of removing the device or system from service, g pyry - '-w" '

u .

e-5 Ib tre corrective maintenance or surveillance activities, and the task of acce: table the returning c'evice er syste- back in service were all performed in an manner.

h h A few of

ossessed the technicians cr their work habits, observed were cutstanding in the skills th d sc ewhat lethargic or almost recklessly fast.A few others were observed as eith b% Overall however, the technicians see ed cenuinely interested in doing a gacd job and were conscientious and pr'ofessional with their work.

k]f n

2 .

was cbserved due to the skills of the techniciansNo unsatisfactory work f, j d  ?

3.0 ADEOUACY OF THE PROCEDURES

[j j

/ - Durinq this inspection, b ' evaluated 23 activities associated with procedures were f.p were Surveillance Tests (STPs),five of the activities we cofvewhat Maintenance, and three procedures were reviewed as examplessome 9 CCPP l&C personnel censidered good procedures.

For all but the three t

' eva ple procedures, ader.uacy of the proceduresthewere work activities of the technicians as well as evaluated. the

' observed with respect No work activities were themselves were reviewed.to the three example procedures, only the procedures

)

$ 3.1 Loop inst procedures g

jtf Q has identified then as being inadequate, and a progr update and C 1 prove these the Lcop Tests has been initiated procedures.

gI

' was the main The work activities associated with these prucedure

}k}], perform the tests in spite of the pocr proceduresinterest of th familiarity with the system. , primarily due to their

's 6

S 3.2

$ Surveillanc? Test Procedures

4 The adecuacy of the STPs was found to vary. i d those perforred on a frequent basis, contain adequate detailSome of the ST 3

instructions to efficiently co plete the task. However, even DCPP and i

.rccedures and have i ple~ented a program to update e STP w

j riant STP 1 SD awareness for Examples of the of the inadequacies of the procedures are the rewr) t Fesrcr.se Testing the Reactor Ccolant of the Reactor Ficw Trip and Transmitters Engineered Safet andmeSTP-!-

p] tcgic.

b detall and were confusing in giving direction cians.

acktoof the te 4

i the procedure more concise.STP-1-8B is being rewritten to consolid make q several ccccepts, such as human factors, and will be u r i

n...e....

4 t

precedures rewritten in the future. A review of the rough draf t of this

< procedure indicates a positive step toward standardizing and improving the precedures.

g  ; A censiderable arount of time was spent reviewing STP-1-33B and

cbserving the technicians activities while working on this task. Because

~

this procedure was a new procedure (approved 4/22/88) and time response testing of the safety systems is important, it was felt that this procedure produce in should the future. be representative of the type of procedure DCPP plans to ,

However, this new procedure (admittedly better than the old 338 procedure) had several deficiencies and the 1% personnel adm.it that the procedure is not a good one despite having spent seven months rewriting the procedure.

s~

The most significant problems with the new 338 procedure were a lack of specifics and clarity.

(

The prerequisites were vague and incomplete in describing the equipment needed to set up the test, (i.e.,"5. Toggle switch (s)." vs the actual number of switches required). The instructions for setting un the test equipment were so limited that a technician

{

performing the test for the first time, probably couldn't set up the eauioment without assistance. Even technicians that had previously performed resolve the test had difficulties and had to make several phone calls to questions.

technicians can perform The the procede-a should contain enough detail that the tas.2 without that particular procedure. requiring prior experience with p~

, An example of how a lack of adequate research and a lack of detail in the procedures creates problems was observed during the performance of Part 10 of the 338 procedure which measures the time response for the Overtemperature Delta T Reactor Trip. Initially, the technicians could not obtain repeatable results for this test. An on-the-spot tailboard between a supervising technici3 n and engineer determined that the problem was due to a module failure. A Work Order was generated to check the module and it was determined that the module had not failed. Further investigation determined the problem to be incorrect values given in the procedure for simulating the hot leg and cold leg temperature inputs. For a test as important as the safety system time response testing, adequate research and systems knowledge should ensure that the primary system temperature parameters are correctly entered in the procedure.

The research recuired to write a detailed procedure might prevent some of this type of confusion and delay. Also, dry running a new procedure can scretimes help in debugging the document so that the final result is a pre:edure that is correct and efficient to use. The I&C Manager indicated that the much as 15C policy is currently to dry run new and revised procedures as practicable.

l The DCPP 15C Department agrees that the new STP-I-338 procedure is a good model for future procedures and was not intended to be. However, to spend seven months rewriting a procedure that is known to be deficient

{

i E

'l 1  :, .

g m

appears to be self defeating.

y This effort indicates either a lack of p

comm t ent to having gocd procedures cr an only good enough to get b approach.

4 E!

encu;h test er;hasis to cc plete properly and when it cam proceed, the precedure was signed off as good enough so the task coulde h

(

3 3.3 Corrnctive Paintenance h

y Cf the five Corrective Maintenance activities Work Orders represented an excellent effort of planning observed

, cn. of the and g

preparation, procedural h considered inadecuate and unsatisfactory.and one of the Work Orde F

0 satisfactorily represent so ethirg in between these other two extre .

2 h

E Pressurizer Pressure Transmitter (PT-474) e escriptions contain

! rot fcund was in high due to the rostvisibility of theofother DCPP maintcnance the consequences of n s. Perhaps documethis t task in a rigid manner.

valves, However, the tasis performed, not performing this W;rk Order for this activity, returning to service, etc., were explained n the ve i

)

i An era cle of a Corrective Maintenance Work virtuallyOrder no with i

erformed on the pressure switches, PS-45/P5 Exchangers.

This Corrective Maintenance was generated byngEngineeri the request control room. of Operations to reduce the number of nuisa at nce alarms in the Operations because it didn't fix their problemWhen the request t Work Order.

errected and so another Work Order was written to REPLACE IPS-46A/46B" and then the directient were t AS REQUIRED TO ENSURE PROPER LOOP . OPERATION o " REPLACE ON 10 IPS-46A/468" i to complete ether systems. this task, they discovered that the powWhen the tetnnicians tr documentation to determine the effects that disturbinThe te would have on other systems in the plant .

g the power supply could not continue and was scheduled It was determined for a laterthatdate the work infor*ed the technicians that the system believed to be affOperations la in service anyway, nheduled date. ected was not so tne work could have been performed at the origina j i

{

ihis research engineer, should have been performed at the planni a supervising technician, or the planner. ng stage by an responsible for determining the ef oa

31 a n t . fects y their ef for informative procedures and work orders.They ave on the overall be given should e ailed,

\

s g, ed ,

l 7 0.

IMAGE EVALUATION  %

O 'g j/

\- TEST TARGET (MT-3) . .'- ,%

'L??

NY/ /

O'Ujk i

<\ 8l{g

g. '#

l.O 22

, i 2.0 0 L8_ __

j' l.25 !ii 1.4 HI I.6

=

nl r r#es l!!ilM =

4 150mm -

6'

~ -

~*

< , ,,kf',, p ,,s $

'.e 4

!,Ih c

,, 4 q\ ,

p 8%,4,

- . m y

m

\

4, y 1 e

'g IMAGE EVALUATION

\ / *%

TEST TARGET (MT-3)

\\/// y g3 Dg I.0 l.1 E!3 m ==

1.2 5 1.4

== fl!! =mv l]imaaI.6 b-150mm 4-

~- - - - - -

--~'*

• 6 Sh '

%p',, f ///p l , l /h 4,, n;,

e rdl

1. s .

,O V

2e IMAGE EVAL.U ATION a O '9, TEST TARGET (MT-3) b g,

////

gfp g'~

'Q,V

, j

+ ac l.0

, 2, 70 l.l L '8 s 1.4

[!l!,,ms1.25 s i.6 i wg=  ;

d 150mm 6' - - - >

d e

ye s_ , z,-/og eyY de yh *O\

ob/ ,

23 } f

3.4 Erarole Procedures

.he three procedures reviewed as examples of what responsible !!C personnel felt were good procedures were maintenance procedures.

Precedure ILC MP 4.1-1 A for checking the calibration on an audio oscillator and power amplifier is an example of an excellent procedure.

This procedure contains specific prerequisites, precautions, and i instruction , and was approved in 1982. This indicates that the ability to prepare good procedures has been available at DCPP in the past. In updating the other I&C procedures, some of the features of this procedure should be considered.

4.0 ADHERENCE TO PROCECURES for most of the Loop Tests, STPs, and Corrective Maintenance Work Orders, the technicians familiarizco themselves with the tasks to be performed and then performed the tasis per the procedures. In several cases the technicians appeared to La so familir with the proce. re that

t was difficult to determine if the technicians were actually following the precedures or sicply filling in the test data.

The only instances where the technicians obviously did not follow the procedures tne data blanksinvolved in the transferring procedure. test data from strip chart recorders to This occurred during STP-t-33B, where not only was the data not prcperly transcribed, but the strip chart recordings were not kept with the work package where the data could be reviewed.

On site follow-up showed this problem to be one of lax follow through of administrative controls of data, i.e., this problem had no technical significance however.

QUALITY CCNTROL INVOLVEMENT WITH 11C WORK ACTIVITIES

> While reviewing the I&C work packages and c' erving I&L technicians in ine field, an apparent lack of QC involvement was noted.

Several of the lit technicians stated that they didn't feel that the OC personnel were qualified to review I&C work anyway. Therefore, some time was spent how many they actually look at. reviewing the process by which QC determine When the Work Orders are generated by the ILC planners, a QC planner reviews the package using a standard checklist. If the package contains tne information required in the checklist then QC may choose to perform an inspection or surveillance on the work activity.

This checklist method seems to be a reasonable attempt at giving all work packages the same level of review and ensuring inspections are performed on a consistent basis.

T1D8 " " "

11 G

4 reviewed and how many inspections and surveillances we those Work Orders.

March 23, 1938 throughData was8,obtained for the time periods from April 3 May 13, 1988 1988. and from May 2, 1998 through L

Curing the March 28 through April B time period, 53 packages were reviewed with nine inspections and 10 surveillances perform e. d For i the time period from May 2 through May 13, 74 packages were reviewed with I 15 inspections and five surveillances performed.

li indicate that GC is involved with approxim.ately 30% of the jobs in theBoth of the field.

j inspections and surveillances were, but it appears that th involvement occurred during both time periods. of

required, few then act as many surveillances were performedIf inspections were inspections, then more surveillances were perfor

ned . When there were k

This amount of involvement (30%) appears to be a reasonable review, of amount i

t

6.0 CONCLUSION

S 1

determine if the organization was operating the best way possible.

o manner and in in an effective an average level of ability and professionalism.The s s with technician The procedures used b good and bad procedures. y the ISC Maintenance Department consitt of both represent adequate, detailed procedures.Some of the STPs and maintenance proced Maintenance, lack detail and direction. procedures r Corrective and loop Tests, planning can not only give instruct' ens and directions for pGood procedur oftask, they can also prevent the wort activities from bein erforming a control. g performed out implemented vveral programs to update However, rocedures.

and improve the ave there are no strong perceptions that these programs have commitment timely manner. by the plant staff and that the procedures pae will be u d t d a total in a The technicians adequately followed the procedures critical functions or sensitive tasks were being performed , especially when

_ _ _ _ _ - - -- -- ~

r*

APPENDIX A i

The following people were the primary contacts while performing this evaluation.

l W. G. Crockett ILC Maintenance Manager I C. A. Wetter llc Maintenance General Foreman J. J. McCann Instrument Maintenance Foreman A. G. Moore J. R. Tinlin Instrument Maintenance foreman W. L. Erown Instrument Maintenance Foreman Supervising Technician D. D. Malone Cc.mpliance Engineer L. Kase I1C Planner J. Hickman 11C Planner D. R. Ceske lead CC S;ecialist R. S. Fairchild CC Specialist

5. V. Noe CC Specialist Other persons interviewed were the 11C Maintenance Technicians and the m age ent perscnnel attending the entrance and exit meetings.

_..,n_. a wr., - - . -- -- --

4 s

i APPENDIX B Va r k Order Unil activity Systa, R0039357 2 STP-I-16A SSPS Logic Train S C0013BS5 1 Corrective Maintenance PS-46A CCW R0005365 1 STP-1-54 PT-505 Main Turbine R0;n 749 1 STP-1-SS3 FT 444 RCS RC004703 1 STP-1-SS3 FT-445 RCS R0004793 1 STP-1 883 FT-446 RCS R0010540 1 STP-1-913 Thermocouple Monitoring System '

R0022494 1 LC-21-13B LS-207 DG l-2 R0021550 I LT-21-18F TS-96 CG l-2 RC02:503 1 LT-21-lSG  ;

TS-97 DG l-2 E0020:29 1 STP-1-633 PT-474 Pressurizer RC021791 2 LC-10-4 FIC-64]B RHR 2-2 C0030212 2 \

Corrective Maintenance PT-474 Pressurizer '

RC014228 I LCV-110 (3-109)

PC-95 Aux Feed F0022327 1 LC-7-221A I

RC022323 1 TM-4110 Delta T Deviation LC-7-221B R00'2329 1 TM-4210 Delta T Deviation LC-7-221C R0022295 1 TM-431D Delta T Deviation LC-7-221D

) E0005035 1 TM 441D Delta T Deviation STP-1-72B EtlS T- 1 Seismic Trip RC005034 1 STP-I-7?B ENSI-2 Seismic Trip P0025231 1 ,

STP-1-72B ENST-3 Seismic Trip R0010932 I STP-1-33B1 Reactor Trip 5 ESF Logic l

j I l

_ , _w -

.m ..-w. _ - . - - - -

== ,

gr w-M-

_ . . . _ _ _ _ . . . _ _ _ . . _ . . _ _ _ . _ . .~. _ . _ _ _ _. . . . _ ___ . . . _ . _

t i

s- e a

Work ()rdor ,1Jnit, , Activity Syste1 CCO320$1 1 Corrective Maintenance Support for STP-I-33B1 CC032191 1 Corrective Maintenance TC-411A Module Check CCO26709 1 Corrective Maintenance Lead / Lag I,odules ft/A fl/A 15C MP 4.1-1A Test Equipment Calibration fl/A ft/A MP I-2.23-1 RVRLIS Calibration fi/A ti/A P.P l-2.14-2 Reactor Coolant RTDs 9

i I

h i

i

.. . . . - . . -- -- . . . _ - - - - .