IR 05000338/1987034
| ML20236R874 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/04/1987 |
| From: | Caldwell J, Cantrell F, King L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236R861 | List: |
| References | |
| 50-338-87-34, 50-339-87-34, NUDOCS 8711240077 | |
| Download: ML20236R874 (25) | |
Text
{{#Wiki_filter:- _ . _ _ - . - - _ v 7- . . ..
" -- e klip UNITEo STATES
' f , o * NUCLEAR REGULATORY COMMISSION a g n . REGloN H g1 U ig tol MARIETTA STREET, ATLANTA, GEORGI A 30323 "
'% /- i ***** q l
l
Report No's.: ~50-338/87-34 and 50-339/87-34 i Licensee: Virginia Electric and Power Company {
. Richmond, VA 23261 ' Docket 1Nos.: 50-338 and 50-339 License Nos.: NPF-4 and NPF-7 Facility Name: North Anca'I and 2-Inspection Ccnducted: September 18 - October 19, 1987 Inspectors: /d2- ///4 87 I wel1, Senior Resident Inspector Ddte' Signed A % /?/E fff4 b^)
L. P. King, Resident Inspector Date~ Signed Approved by: 6 3 64- // D F. Cantrell, Section Chief Dater Signed
' Division of Reactor Projects SUMMARY Scope: This routine inspection by the resident inspectors involved the following areas: plant status, unresolved items, licensee action on previous enforcement matters, licensee event report (LER followup), review of inspector follow-up _ items, monthly maintenance observation, monthly surveillance observation, ESF walkdown, verification of containment integrety, operator safety verification, operating reactor events and steam generator inspectio During the performance of this inspection, the resident inspectors conducted reviews of the licensee's backshift operations on the following days -
September 20, 30, and October 1, 2, 6, 7, 8, 9, 13, 14 and 1 Results: One Licensee . Identified Violation was identified: Violation of Technical Specification 3. (see paragraph 12) 8711240077 g71118 ,
'
PDR-G ADDCK05%lB- _ - _ - _ - _ _
- ~ -- ~
4 ,
, y ,
'
W .f ?).f ;gy , ,
,
.>_
>r
'
??j l g;
_
, . . . -.
@A T ,, ,
.
. .o,
, , .
i>, fil ' ' q :g +
,
'
y ;
' >
REPORT DETAILS
,fw e
,
'
.g .
ch 3
-
l1h 41.icenseelEmployees; Contacted-4 > *E.1W.LHarrell,? Station-Manager
; :R..C; Driscoll',LQuality' Control (QC) Manager;
- * EC kane,LAssistant Station Manager WT E *ML L. Bowling,LAssistant Station Manager
*R.- 0.l.Enfinger, Superintendent,1 Operations -
.l 'M. R;;Kansler,: Superintendent,, Maintenance if, ~ /' '
'
*A.-H.yStafford, Superintendent,. Health Physics f ,
'
.{ *J..A;uStall,1 Superintendent,lTechnicalTServices e
2 > J. . L. Downs, Superintendent,: Administrative' Services - 4 y^ .J;iR; Hayes,' Operations Coordinator-0.jA.;Heacock, Engineering Supervisor T; 0 zD. E. Thomas, Mechanical' Maintenance Supervisor i 1
;G..D! Gordon,: Electrical: Supervisor.- , ,
TR.'A.lBergquist,: Instrument Supervisor
, F.>T. Termine11a, QAiSupervisor d fJ. P.-Smith,; superintendent, engineering a, , - m s 1D.'B.lRoth,.' Nuclear Specialist TJEH.ileberstein,' Engineer e 'f"*G.LGiHarkness,LicensingCoordinator
*RD T.: Johnson,- QA Supervisor:-
,
,
, .0therilicenseef employees 7 contacted include' technicians, operators, m Lmechanics, security force members, and office personne ~
"1 >
* Attended exit' interview v, . .
+
.
. ..
NRC/ RsgionaltMa'nagement Site - Vi sit: F. S. ~ Cantrell visited North . Anna s PowerrStation t on 0ctober 13 and~14 to review plant operating performance
'- ' ~
ifollowi.ng the. restart of Unit 1 on October 12, 198 y Exit Interview (30703). .
*
, The iinspection s scope. and findings were summarized on October 19, 1987,
,( ith~ those; persons indicated in paragraph. I above. The licensee
~
w .- acknowledged _the inspectors findings. The licensee did not identify as Si a proprietary.any. of; the~ material provided to or reviewed by the inspectors [
'
during this1 inspectio Verification of the
' = -(0 pen)[ Inspector Followupu- Item ' 338, 339/87-34-01: '1 Pcompletion of the' repairs to the Unit 2 BBC Brown Boveri breakers per the l 0 10,CFR Part 21l report-(see paragraph 6).
p . >
' (0 pen)' Licensee Identified Violation 339/87-34-02: Violation of Technical il ; Specification _ (TS) 3.9.4 (see paragraph 12).
, " , 4
! '
ti <, ,
' .
,_ ___ - ._ ._ _ _. _ -_
y 8 ' !
.. .
1- .
, ..
2 l I ic .
(0 pen)~ Unresolved Item 338/87-34-02: Potential failure to follow {
e . procedure. violation which resulted in the. draining of the Refueling Water ] 3 Storage 1 Tank'(RWST)'below the TS limit (see paragraph 12). !
! Plant Status Uni l Lunit 1 began the inspection period with the unit in Mode 5, day 66 o'f the : steam generator (S/G) tube rupture outage. On September 21, the licensee
,c met with the NRC 'in Bethesda to answer questions relating to their S/G ; tube rupture report and the Westinghouse root cause determination of the ! P tube ruptur OnL 0ctober 9, the inspectors completed the review of the items relating to the restart of. Unit 1, and the licensee was issued a revised Confirmation. of Action Letter with an attached Safety Evaluation Report i
'
allowing the licensee to restart the unit and operate at pc,wer levels not
'to exceed 50%. . Unit 1 achieved criticality at approximately 2:12 p.m. on October 12 and. was on - the grid by October 13. This ended a 90-day S/G ! . tube rupture outage. The unit is presently operating at approximately 48% ' power 1with indications of approximately zero primary to secondary leakage in the' Steam Generator Uni ' Unit 2 began the inspection period in Mode 6, day 27 'of the refueling l outage. . On October 13 the licensee discovered five tubes in the "C" SG j ' which had through wall cracks even though the eddy current tests had not i indicated any problems . (see section 13 for details). The unit is 'I . presently in Mode 5, day 58 of the refueling outage, and the licensee -
intends to..have the unit restarted and back on line by October 30, 198 I Both Units The' NRC conducted an enforcement conference with the licensee in Atlanta on September 24.- The purpose of the conference was to discuss the events ; leading to an overexposure from a hot particle (see inspection report ; 338,- 339/87-30 for details) and the performance of core alterations without containment integrity on September 9 (see section 12 for details). . Unresolved Items An: Unreso'ved Item is a matter about which more information is required ; to . determine whether it is acceptable or may involve a violation or J deviation, q l One' unresolved item was identified during this inspection and is discussed )
'in paragraph 1 ]
l l l u .
,n - -
.g ;
t ,, e --
r Licensee Action on Previous Enforcement Matters (92702)
(Closed). Violation 338/87-15-02: Fuel Assembly Procedures. The licensee ,
~ has, changed procedure 1-0P-4.10 to ensure two operators are present when fuel is moved, th ' Licensee Event Report (LER) Follow-Up (90712)
'
The 'following 'LERs were reviewed and closed. The inspector verified that reporting requirements' had been met, that causes had been identified, that corrective actions appeared appropriate, that generic applicability had been considered, and that the LER forms'were complete. Additionally, the inspectors confirmed that no unreviewed safety questions were involved and that violations of regulations or Technical Specification (TS) conditions had been identifie (Closed) LER 338/86-10: Inoperable Diesel Driven Fire Protection Pum The licensee has repaired the starter motor and solenoi The pump is
'now operabl .
(Closed) LER 338/87-02, Rev. O and 1): Malfunctioning Cardreaders Resulted in Potential _ For Unauthorized / Undetected Entry Into Vital Area Thel licensee took appropriate corrective action in their preventative-maintenance-program to-reduce the probability of this recurrin (Closed) '338,339/P2185-03: K-Line Circuit Breaker BBC Brown Boveri, Inc., breakers have indicated potential insulation degradation on switchgear ' control wiring. .The manufacturer recommended protecting the insulation 1with a' field repair kit consisting of heat shrink tubing for the wiring and adhesive backed gasket for the dust cover, The work has
~
' been: completed on Unit 1, but there was not enough material to complete Unit 2. -The. licensee has the material ' on orde This item will be identified as. Inspector Followup Item 339/87-34-01 pending the completion of repair on Unit . Review of Inspector Follow-up Items (92701)
(Closed) IFI 339/86-28-08: Replacement of Air Cylinders Associated with the Auxiliary Feedwater System. The air cylinder and tubing have been l replaced with qualified cylinders during the Unit 2 outag .(Closed) IFI -338/87-24-05: Review of Differences in Emergency Plan-3, Rev. O and Rev. 1. The inspector reviewed the differences between Rev. 0
.and Rev. 1 of EP-3 " Steam Generator Tube Rupture". Although the ,
differences were signi ficant ' 'oetween the two revisions, there were no i actions taken during the tube rupture event, which could have been i incorrectly performed as a result of the change ! _
W ,- <
. ,
p y; QR, ,g ~
+> :,( n*b w ' , >
-
[ Q YiL ie _
.I, r *
- ;' Month 1'yl Maintenance l(62703)
182
' '
f,
; Station i maintenance' acti.vities ' affecting safety related systems and
, ,
' - ; components were observed / reviewed, to' ascertain- that the activities were M ' conduct'ed (in.- accordance with ' a'pproved procedures, regulatory. guides
- " ,
' 'and c industry.- codes; or. standards, ^ and in conformance with Technical Speci fic'ations. -
IThe211censee1 overhauled:the. non . return valve on "B" steam line for Unit
'
$s - . S yM 2. JIt' required weld build up' and machining. The inspector inspected the S '
'
internals beforeTand.after machining.
M, -
- '
On1 September!28, the inspectors' witnessed portions of the Motor Operated X M : Valve An'alysis' and iTest System (M0 VATS) test of 2-FW-MOV-200 This valve 4 is :the? discharge MOV for the- auxiliary feedwater steam driven
' (terryLturbine)l pump.;,The maintenance was performed under. Maintenance Procedures LEMP-P-MOV-3'(Perdictive- Analysis of Motor Operated Valves) and iMEMP.-C-MOV-1 (Inspection and = Repair of Safety Related Limitorque Valve .
ControliType"SMB-000,LSMB-00-and SB-00). 'No problems were identified by 3 the inspectors; !
, , 4 10n.; September?30,# the ' licensee identified a' problem.' associated with the ' ' :incore ; flux thimble 1 repair;per Westinghouse procedure MP - 2.3.1' VGB-2, Rev. i 0, '(Seal Table. Repair at- North . Anna Unit Two) and VEPC0 procedure 1TMMP-C-RC-8B._ This problem involved the Westinghouse technician cutting ' > -tube J-12 iinstead . of the ' desired tube J-15. The tube J-12 was not " , required lto be: repaired;, however, this inadvertent repair will not affect , ,
the use of? tube.J-12 for incore flux monitorin " I' Thei inadvertent cutting; of J-12 was performed under the ' observation of t
, ;the Westinghouse Flux Thimble Coordinator and a licensee Quality Control Inspector'. Following- the interviews conducted with- the -personnel-
-
involved with:the. event, Westinghouse determined the cause of the mistake
'to Lbe the marking' of_ the tube at the seal table instead of the tubes directly,: followed by' the crew's ' attention being diverted prior to the crepair and' finally' the technician selecting the adjacent tube to perform '
P? the repair. : Westinghouse's corrective actions were to edd a step to
- (their. procedure.to require the thimble tubes to be tagged directly. The "
repairs were1 reported. to have: been completed without further problem :This l event was an example of failure to follow a procedure, .was Lreported- by: the . licensee, and -did not result in any significant , degradation't'o any of the plant's safety related system ) m . Thef '11censee informed the -inspectors that the inspection of the 2J Emergency l Diesel Generator- (EDG) piston pin bushings has been completed.
, ' <
LThis - inspection : revealed that .the bushings were in excellent conditio L Both1 Unit 2 EDG's were inspected during the Unit 2 refueling outage (see
' *
Inspection . Report 87-29 for details on the 2H EDG) which completes the b Linspection of all four of the Diesels at North Ann The Unit 1 EDGs which had previous operating time with the inappropriate lube oil had m q! f , ,
)
w m l am _ j
g~ ' s . _ 'a
.
),b,
.
,
#
. I
,D 9jk ,.y , 4 y '.]
w ~ <- ; h ' < yd % [. c
*
.g p_Mf '
; a QMe
'
'
5-
'
,
m ,
'
, DJ , *
' , ~ ;iu . '
findicationsiof pistbn Lpin bushing extrusion Lunlike the. Unit 2 bushings w[
, - ,
TwhichLdid notihave~ operating time with the inappropriate lube oil and the-
-I jg*y ,, 9 % :; bushings;were1found to'belin. excellent: condition. - The piston pin bushings d ' ; andLlubel oi1Ewere replaced in ' Unittir and o now = all four EDGs will be'
W ' operating)withi bushings that have only been exposed to the correct lube
'
' gp o11. (ChevroniDelo: 6000). This- should solve the - piston
-
pin bushing :1 g', textrusio'n? problem.which caused EDG' failures in the;pas .] ,
~
'
f- .
. .
The Llicensee 'has completed'.th'e _ Ultrasonics (UT) inspection of ' the Unit 2' I
. . " Q. , lfeedwaters and; condensate piping-to determine wall' thickness. A total of-
'
'400 / components ~ were inspected and 149. : components were determined to-V ?] ,@W ,
.
requirereplacement. Of the .49: requiring . replacement, :10 were in . single
* * - M : phase 3 flow l systemsi and ' 39. were in the two phase ' flow systems. These
@j " icomponent: replacements wi.11.be completed ~and inspected with the exception 799 (e
, e e
of the Lflash evaporator components which 'will not be. used, prior to Unit '
"
2' restart at the end.of:0ctober. ' With.the completion of Unit 2 secondary
', ' pipingiinspections an_d. repair, theclicensee has inspected and established g , '
amase'line wallCthickness. of the suspect components in both the single
< '
cand 7 two' phase flow- secondary -systems Lfor both unit Future _ inspection T ' willu ?allowt /the - licensee-- to 'better ' determine the actual' o, . ?errosion/ corrosion rat S ' hg INo: Violat. ions' or. deviations were. identifie . . . . c .
.
r 1 97 (MonthlySurveillance'(61726)'- M The inspectors 4 observed / reviewed Technical Specification required testing
;, and. verified ~ that; testig : was . performed in caccordance with adequate '
Lprocedures, that'. test instrumentation E was calibrated, that limiting RJ y conditionsf forJ operation' (LCO): were met and ' that - any ' deficiencies
^ s Lidentified were properly-reviewed and resolved.
n
.
~
*
l0niSeptember 21, fl987; the inspector reviewed:2-PT-78.3 " Residual Heat L Removal Pump and Valve . Test" and' 2-PT-78.3.1'" Residual Heat Removal Pump LBearing : Temperature" for .2-RH P-13. A vibration analysis using the Bruhl W < and ? Karl method was attached- toL 2-PT-78.'3.1, No problems were
, ; identifie c The 1.nspectors observed a three-month periodic test of 1-SI-P-18. Heise
. . . . gages- had to be installed in. place of the normal instrumentation to give '_ * the licensee enoughEaccuracy to allow the pump to pass the differential M', 2 pressure criteria. ~ The. periodic test had failed the two previous times N using Ethe; installed _ instrumentation. There is no ' alert range for
.
-differential- pressure. in the periodic test ' because using the criteria in ASME Section XI,1 itiwould_ fall below the minimum value of 156 psig
. . , : required by;the-technical' specifications. The licensee plans to establish
,
aan. alert range. Review of the past two' periodic tests indicated that the
= pump pressureidifferential was just above 156 psig. The inspectors have & ' requested.the711censee. use heise gages for future tests to allow a trend to be' established to determine whether or not the pump is degradin ,+
l .: -
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
~
W X > >y l5 [ > % p% ; , pg , u '
"y w g;- . x
>i[ ~ #
, , , @, LThe; inspectors reviewedL1-PT-78.3 "and ~1-PT-78.3-1.for 2-RH-P-1B pump. . No ,; . ' '
problemst were -identified. The vibration data ' was: within limit The 1 , . mechanical)sealchad been replaced on the pum p
~
NoNiolations or deviation's' were identified'.
:E . . . . ' M 1 ESF System Walkdown, Verification of Containment Integrity (71710, <-
61715)) ,
' ' ThhiinspectorsH1ocallyJobservedL the . positioning of twelve penetration ' ,' < .
l valves,;bothLinside_and,outside the. containmen TheD inspectors walked.-down l the valve ; alignment for the A, B and C accumulators ( using ' valve tcheckof_f list 1-0P-7.3A on October 6,1987. No
'
problemsLwer'e-identified. Several:of'the valves had boron buildup on-the c: $ pac. king: glands which Lindicated -leakage. At the time of the' walkdown, the .
~
.g ,. g . accumulators were' not fully. pressurized. -These'i_tems_were brought to the
. attention'of' licensee management.-
4f
'
NoEviolations t or; deviations were identifie N ' L 11.l0perationa1
Safety l Verification-(71707).
,, . ~k 'By observations. during L the inspection period, .the inspectors verified ,, ' .that the'. contro1Lroom manning requirements were being met. In. addition, ~ >
lthe inspectors 1 observed shift ~ turnover to ' verify' that continuity. of . system i status .. was - maintained. The. inspectors periodically questioned shift. personnel _ relative to their awareness of plant condition j a 1
- , ~~Through log 1 review 1 and plant ' tours, the inspectors verified compliance ! - with selected ' Technical Specification i(TS) and Limiting ' Conditions for 10perations'.
In-the course.of_.the monthly activities, the resident inspectors included La Treview of the -licensee's physical security' program. The performance of various" shifts;of the security force was observed in the conduct of. daily
: activities' to include: protected' and vital areas access controls,
'
,
searching zof ~ personnel, packages and' vehicles, badge issuance and
.
,
, kretrieval,iescorting. 'of visitors, patrols and compensatory posts. In 1
- ' addition , ,' the resident inspectors observed protected area lighting,
. protected and vital areas barrier ' integrity and verified an interface Ebetween the security' organization and' operations or maintenanc ~
, s,
"4i ' - .The' inspectors kept informed, on a daily basis, of overall status of both 6 nits' and of ~ any significant safety matter related to plant operation Discussions?were held with plant management and various members of the ,; operations . staff on a regular basis. Selected portions of operating logs ;and data sheets'were reviewed dail w ,
C
,
__ _ _ . _ _ . . _ _ . _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _
(" y w, ,
,
p .
-
i
.
V 7 p i kh - The . inspectors : conducted various plant tours and made frequent visits to ,
'
H^ Lthe Control Room. Observations included: witnessing work activities in L progress; verifying the status of operating and standby safety systems .; e , and equipment; confirming valve positions, instrument and recorder readings, annuciator. alarms, and housekeepin The following. comments'were noted: On October 19,'while reviewing the SR0 log, the inspector discovered that 1 the steam generator blowdown radiation. monitors had been . inadvertently isolated.' The inspector determined . from discussion with the ' licensee that the. flow through the monitors had been checked and adjusted sometime on 0ctober 18.~ Early the morning of October 19, the blowdown monitors were discovered by an operator to be isolate The licensee has been unable to determine the cause of the isolatio Since' the authorization for the start up of Unit I was based on the
'-
licensee monitoring several radiation monitors to ensure that SG tube degradation .whf ch could result in 'a tube rupture is identified in sufficient time' to perform a controlled shutdown of the unit and one set of these radiation monitors is the steam generator _ blowdown radiation monitors, their . inadvertent isolation becomes a significant occuranc However, during the time frame the monitors were isolated, the primary to
~
secondary leak rate was approximately zero and all of the other radiation monitors were- fully operable which reduces the significance of this l particular. even The . inspectors will review the licensees investigation and corrective actions associated with this even ~ On: September. 9, - the ' inspector toured Unit 2 containment. During this
. tour the inspectors entered the secondary side of the A steam generator ; to inspect ~ the work being performed on the downcomer flow restrictor While on the platform just outside of the SG manway, the inspectors .' reviewed the entry and exit accountability log. This log is used by the Westinghouse personnel performing the SG downcomer installation to ensure ;
that all loose materials which enter the SG are accounted for or removed
'
from the SG. The log appeared to be-in order and the inspectors were l informed that the log is reviewed and all discrepancies corrected at the i end of each shift ' before the personnel on that shift are allowed to
' leave. This review was instituted because of the problems encountered j-during the Unit 1 downcomer installation (see section 13.4. A. for details of Unit 1 problems).
Also during this tour the inspectors noticed that the licensee had posted aress in the containment which were designated low dose area This informs personnel in containment, who may be waiting to commence work, of areas that will minimize the dose received. This was also done in the Unit 1 containmen _--_. gym jm , ,
.,
hQ+ f' 2 'Q .i .q; a U,N i
,
e ! w ><gg,
,<< ,
fl/ <- .n m
'
-
-
_8 y h p% g;y .
'
y ? d *Qf , N
'
6 >.
- ~-
__ >-, , < , .
""
.0' 0n t
;0ctober 8,fdurin~g _'a . tour of .the control room,. the . inspector witnessed
< q" ~
'w : portions ofyl-0P-5.5. This, procedure was: being ' used by the operator to ffill, vent andLunisolate the C RCS-loop. Both' the A= and B RCS' loopsLand
-
~c Jthe1 rest' 'of Jth'el RCS : had been filled and vented. earlier. At:the. time of
'3f ;the inspsetorsi observ'ation,; the . operators were operating ' the ' C Reactor
! Coolant ' Pump:(RCP) with the hot legL isolation valve open, the cold.;1eg
&; %
"
'
, ;isolationivalve shut; and the bypass around the _.coldileg stop Lvalve open.' I
,
:Thiscevolution _-isf required to.. continue for 90 minutes prior to' opening
' '
a5 .
. (the;coldleglstop. valve.astrequiredbyTechnicalSpecification s"$ , - L .. . _ .
M During"the refueling; outage.for Unit 2,'which commenced on August 24, the-g ?s , llicensee'. conducted . numerous 110 CFR- 50 Appendix,1 -leak rate tests. of M ; containment: penetrations _.'.. . TS 3.6.1.2 requires ' that the total of all F1 W containmentLpenetration leakage be .less :than 0.6 La. In August the W ' olicensee1 tested containment penetration 89 for the' steam jet 1 air ejector
~
'j
'
1 ,
, w < discharge .(11_ne to containment. The licensee -was- unable to measure the t', ileakagef past'_ both Lisalation ' valves in - that penetration, therefore,
. demonstrating that penetration 89 exceeded the TS limits. This penetration f's u
'
,
,
Lleak test result was-not' reviewed for compliance with TS or deportability- , . 4 ' until: September.18 because the procedure did not require a review and the
~
!
., # unit was in coldt shutdown. -On September 18, the event was determined not-a Jto beireportable;per..10 CFR:50.73, but was reviewed again on October 1 and
@p * ,
"'
;wasidetermined .to require. a,.30. day: LER. per 10 CFR 50.73. On October 9,.
ethe411censeel aiso ' rea112ed that '10 CFR150.72 > was applicable and . notified
' .
MM the;NRC'via:the ENS phone of-the' event. -The inspector has' discussed this
'
J 'situationLwith)the: licensee <and was informed _that procedures were being y a changed to'ehsure that;the Type B and C test results would be reviewed for 1 > m 2 deportability'as<soon-as they areLavailable. The inspector requested that the' licensee : address 'the delays associated with meeting the reporting
~
m1 ,
;criteriaHin' the lLER fissued describing..the event. _ The; inspector will w' f '
treview; thel LER when11t 'is issued to ' determine if the event is properly
,documentediand to' ensure that appropriate corrective actions are taken to , , ,ensureL that the reporting criteria is met;in the~ future. This situation ,
wasireported ~ priorEto the~ licensee. operating Unit 2. The unit is '
, ._ presently.in Mcde 5 Cold Shutdow .
j
,
.'No violations or deviations'were identifie .
x . . .. ,
,
'
,
: 12.) - Operating Reactor Events (93702)-
E
-
.
,The s. inspectors . reviewed activities associated with the below listed i N
'
-reactor events. ' The . review included determination of cause, safety
'
" , significance, performance.of personnel and-systems, and corrective actio The: inspectors: examined instrument recordings, computer printouts, mh .' operati on s '.--j ourna l entries, scram reports and had discussions with
'
.' operations, maintenance.and . engineeri ng support personnel as appropriat l
' 'On September;21, the licensee notified the' NRC that they had violated i Technica1' Specification. (TS) 3.9.4 (Refueling Operations-Containment
.
!
. Building i Penetrations) . This violation involved the performance of a i
'
core l alteration (the latching of approximately 10 control rods) with both , l'
, '
t ___.__-__w
w;m y
. ' "
U l
- L t 9
!
l
'the containment personnel: doors open (i.e. without containment integrity).
Since the licensee was scheduled to attend an enforcement conference on September' 24. to discuss another issue,- the NRC requested that the
-
licensee address this event as wel A description of' the event based on the inspector's review and the licensee's presentation in Region II on September-24 is as follows: i L 0n September 20, the . licensee, after verifying that. containment integrity ; had been established per 2-PT-91, placed the upper internals into the ' reactor vessel . Following the completion of this operation, the
'
Westinghouse refueling crew and the refueling SRO exited containment to ' _ allow the Westinghouse crew to be relieved. During this time. frame, both containment doors were left open. - The master procedure 2-0P-4.1 did not have provisions fo'r exiting and then re-entering the containment integrity procedure'2-PT-91. Therefore, even though no core alterations were being performed, . containment integrity should have been maintained (i.e. at least one of.the containment doors should have been closed). ;
' On_ September 21, the refuel.ing SRO and the new Westinghouse refueling i crew entered containment- to perform the control rod latching operation iper MMP-C-RC-14B, a satellite procedure of 2-0P-4.1. Both containment doors were left open and approximately 10 control rods had been latched before; the discrepancy was discovered. The discovery occurred when the shift supervisor, while performing the normal' surveillance for containment integrity, requested the status of the containment personnel doors. Upon ' discovery that both doors' were open, the control rod latching operation was secured and both containment doors were close '
The' licensee identified 'the cause of the violation to be personnel error. and Tinadequate procedures. In the presentation to the NRC on September 24, the licensee committed to perform the following corrective q actions: Revise the refueling (controlling and sub) procedures to ensure verification of containment integrity for lif ting the reactor head, I core alterations, and setting the reactor hea ; i Perform an evaluation by each department as to the need to revise i master and sub procedures to ensure initial conditions are me i Have' operations management re-emphasize the importance of the proper ) use of procedure I l Perform a human performance evaluation of the even . Discuss this event in the licensed operator requalification training and place the LER in operator required readin !
i_2r__ a _ _ _ _ _ _. j
wxy = ,. - J7 M . apf t 94 - g
, Jij >'
i.[ . &, >
'
w
,
r
-
.
i i n , a 10_ i
<
X d ;', ; j'^< ' The licensee reported; the event- as- required by 10 CFR 50.72. - A review by
:the; NRC1 determined that: the safety; significance- was minor since- the rod 1 %.m ' :latchingEoperation - only caused slight rod . motion and, ' therefore, the l . . , .
Lpotentiale for a. radiation: release was ' minor. This event will be identified:;as f a Licensee Identified Violation LIV 339/87-34-02 and a -
'
- 'w<
. V' A
,
2 notice ~' of - violation will' not be issued since it: meets the following-
'
' ? , ,
feriteria'of:10.CFR, Part 2, Appendix C: .
' ' . It was identified.by the licensee; I %. ]
[hy (It fits in Severity Level IV criteria;
'
b '. $E *
- ?It was reported; , ,
ss '
,
y-V lIt: was corrected for :the short term and if the licensee completes, f
%w : their commitmentsLas' stated above should prevent' recurrence; -l , ;
1 It :was ' nots a L violation that could reasonably be expected to have l W been prevented by the:11censee's' corrective action for a previous ! 1 violation; j e ,
I '
'&' Two other recent events at North Anna were somewhat similar to this event and~ were : discussed in ' Inspection Report 338/ 339, 87-19. On May 15,
'
,
-
f
, , 1987; Unit 1~ core alterations.were made without an operable charging pump i ' in violation Lof' Technical Specifications. . The cause was stated to be
- X < , 'personnellerror-. LThe event.was reported in LER 87-11 and identified as a
* '
' licensee : identified violation in - IR 87-19 (LIV 338, 339/87-19-17). On y- ; June ;1,1987, Unit 2 entered Mode' 3 with all three auxiliary feedwater y- _ pumps inoperable ' (no automatic '~ start capabilities)- in _ viciation of
* '
: Technical. Specifications. The cause of the problem was personnel error
.
We and procedural Inadequacie The event was reported ~ in 339 r LER 87-05 (S andiidentified as a licensee identified violation in .IR 87-19 (LIV 338,
339/87-.19-18),
.
;
. '
'E The c safety significant of. the above items does not warrent regulatory . action when L considered : individually; however, it is evident from the
..recurrance of. personnel /procedureal errors causing Technical Specification violations that increased management attention in these areas is needed torensure meticulous. compliance with requirements and prevent more serious events from occurrin '
' The licensee .. notified the inspector that at 9:17 .
on October 10,
-i Unit 1; experienced ~an : inadvertent operation of the Pressurizer Power iOperated Relief Valves (PORVs). At the time of the event, the unit was in
' Mode 5 with PCS temperature approximately 195 degrees' F. and pressure of
~
,
'
' 375'. psi The ' licensee ' identified the cause of the event to be the W., performance '.of. surveillance procedure 1-PT-212.12 which cycles the j
,
downstream-isolation valve for the Residual Heat Removal (RHR) heat
-
E i exchangers. The. RHR lineup, prior to the event, had the RHR flow through the' heat' exchangers, virtually isolated and bypassed end full component l*
-m - h
?- ^E-g "I-3:
'
. 1 f, #
, ..
11 .
.
, -
i
, cooling (CC) flow to the.other side of the heat exchangers. Consequently, I when the-RHR heat exchanger isolation valve opened, it. allowed enough cold i water:to enter the "B" Reactor Coolant System (RCS) loop to momentarily l y"- ,
reduce the "B"-loop cold ~ leg temperature (Tc) element to less than 185 .4-degrees This momentary reduction caused the PORV p_ressure setpoint to 1-automatically change from approximately . 415 psig to 350 psig. This resulted ini the cycling of one or both of the PORV The plant only i experienced a slight pressure decrease of approximately 10 psig and the l opening of the PORVs was of 'such a short duration. that they were closed q when.athe- operators observed their position following the annunciator 3
: alarm.. The other loop Tcs remained at approximately 195 degrees F and- 4 'the "B" ' loop Tc quickly returned to approximately 195 degrees F. The . < : licensee will; be modifying the . surveillance procedure, 1-PT-212.2 to prevent recurrenc ! 'The ? inspectors were notified by - the licensee that on October 14, with' ; Unit 'I operating- at 30% power, the Refueling Water Storage Tank (RWST)
was . inadvertently sluced to the fuel pool. The draining continued until
,
the 'RWST' low level alarm annunciated in the control room alerting the operators to the conditio The ' licensee entered TS action statement . 3.5.5 which required the RWST level to be returned to within TS limits
~
j
. within one hour or be in' at least hot standby within'six hours and cold '
shutdown ' within the following 30 hour The licensee was able to , terminate . the drain - down and refill the RWST to within TS limits in !
: approximately:,40 minutes. This prevented the licensee from having to shut the unit down.1The inspectors were also informed that the fuel pool ; ' level did not' exceed the~high level alarm setpoin ]
The licensee is still . evaluating the event and a human performance i
<
evaluation is being- conducted. . Based on the understanding by the inspectors, the cause of the event' appears to be an improper valve lineup ;
. caused by the failure of the licensee to follow a valve lineup procedur '
The inspectors are allowing the licensee to complete their investigation into the even This event will be identified as an unresolved item 338/87-34-01, pending 'a review by the inspector of the licensee's ;
' investigation and corrective action . Steam' Generator Inspection (52703, 61726, 71707)
Unit 1 ; The -licensee has just completed an extensive SG inspection and modifica-tion program as a result of the June 15 tube rupture event. This program resulted in a. 90 day unplanned outage for Unit The NRC has been
'
involved in =almost every aspect of -this outage relating to the SG Prior ' to - the NRC authorizing the Unit to restart on October 9, the . inspectors reviewed various aspects of the licensee's SG inspection and I
. modification progra The following is a brief description of the ' inspection that was performe ,
_ _ _ _ _ _ - . _ _ _ _ _ _
g7pTQW > ,~ ,
; ,
t
- .LV p' %k , .
w-m
,
l g lyi('d Q.:# # *
*
j) m 4
,, ,
~
. jh ,g- ,
a,m; y .,
' +
g ( .J
', ' ' 'ai Verification:that the-licensee.has. established an adequate primary to'
u m l: secondary.l.leakf rate - monitoring progra ~ The1following procedures gR Ewere reviewed: g m % 4
& , 7(1)D Standing?0rder #155 (Revision 1), dated October 9,1987,. Unit 1 i o
x* x : and Unit L 2. Primary ' to Secondary -Leakage. This . standing order _. provides : the ? operator - with' the required ' actions ~ to be taken - y,, Lba' sed onsthe~ data 1obtained from the radiation monitors and what i b[ ' ,m ,
,
i actions' to Ltake if: a' nyl of the monitors' or methods to measure l 9;' m-
'
= primary to secondary lleakage~ are inoperable.- The information-.
7 - contained in the Standing,0rder is as.follows:
-
g TITLE: ' UNIT ONE (l') UNIT -TWO (2) PRIMARY TO SECONDARY' LEAKAGE .
>
,i
<
,
"
$ [During; power. operation the following leak rate limits are g~ J, Lin1effect. -These limits :are more restrictive than Tech
~E gg Spec 3.4.6 2.C and are consistent with PT-46.2, Primary to
.
g *^ , Q* t > : Secondary Leak Rate. Determination test.
,. TR s", - , Total leakage ~of 60 < -GPD increase from- one c '
>
. surveillance interval to the next surveillance
.
' N .
. . .
interval as specified in PT-4 "
, '?
3[ ' v
,
#'
'. '
? Leakage from an ' individual steam generator *. ' of
<<160 GPD-increase from one surveillance interval to
" " "
,
Ihei ne~xt1 surveillance interval as specified in g ,7g >c
.
e
~
'PT-4 "' '
S
'
;TotalLleakage of._< 300 GP '
J c '. L Leakage from an individual steam generator * of < 100 LGP e '.- i An increasing trend based on. the latest surveillance 1
.that ' indicates > -100 GPD would be exceeded on an
'
o , ~! c
'
individual steam generator * within' ninety (90) ! minute * If the limits of (a)' or (b) are exceeded, then power is to be reduced to < 50% as soon as possible under
,
normal controlled conditions, but not to exceed ninety !$ < > . x (90) minute {
'
gM . .If the limits of (c) or (d) are exceeded, then initiate a unit shutdown to be in MODE 3, Hot Standby,
, , ; "W a;. with six (6) hours. Initiate AP-24.1, Large Steam ,
i
, ,
Generator Tube Leak Requiring Rapid Unit Shutdow l a j
[!
J'.
, ',.
n .
,
iNi t
. . Sp ,i'
M ....
g:- y ; 3 .: -_
" .st +i , - .*
4 y 13 l
:
he If the111mits of (e) are exceeded, then initiate a
-
unit shutdown to reduce power below 50*6 within ninety (90) minutes and be in MODE 3, Hot Standby, within six (6) hours of exceeding limit (e). Initiate AP-24.1, Large Steam Generator Tube Leak Requiring Rapid Unit
' ' Shutdow * If the leak rate trend is such that'the limits of (1.a) or (1.b) will be exceeded within the next 24' hour period, then ' '
notify the . Superintendent Operations or SR0 on-call . y
' If the limits of (1.a) through' (1.e) are exceeded, then carry out ' station procedure ADM-19.6, Notification. When ~
power reduction is commenced a one' (1) hour report per ; 10 CFR 50.72 (b)(1)(1)(A) is to be made even though a -
--Tech Spec' limit has not been exceede ! ; In order to perform PT-46.2, Primary to Secondary Leak Rat " -
Determination, the following equipment or systems - are ; required to be operabl ' Primary sample system to obtain .a RCS sample from any point except the pressurize Secondary sample system to obtain a steam generator
' liquid sampl Air ejector radiation monito A'ir ejector flow rate indication for each air ejecto [' ! N-16: radiation monitor on the main steam header, Multichannel analyzer - for counting various samples collected, Steam generator blowdown flow indication when blowdown is .in servic . ' Action to be taken for inoperable systems or equipment listed in item (4).
' ' If the. primary or secondary sample systems are inoperable, then the system must be returned to service within twenty-four (24) hours, provided the N-16 and air ejector radiation monitors are operable, or power reduced to 5 50% within the next ninety (90) minute If - the N-16 and air ejector radiation
. monitors are not operable, then reduce power to 5 50%
under normal controlled conditions, but not to exceed ninety (90) minutes.
! k L i b i f4 j
-- ___-_-___,
- _ __ - . _ , :.; :2 g' m ' ' , .
@ [ , , y 14 p, I R . [If the' air ejector radiation monitor is inoperable, then carry out the Tech Spec requirements every four
'
,
,
*
..
(4) hours.. If the air ejector radiation monitor is
,
'
'not returned. to service within the next : seven (7)
' days, then reduce power to'5 50% within the next four
.
,
-(t) hour If air ejector flow rate indicators are inoperable,
'
then carry out .the Tech Spec requirement If the N-16 radiation monitor is inoperable, it must be returned to service as soon as possible, and the
'
NRC ' Resident Inspector - notified within twenty-four
. (24) . hour If the N-16 radiation monitor is not returned to service within seven (7) days, then reduce power to 5 50% within the next four (4) hour .If .no Multichannel Analyzer (MCA) is operable, one '
I' must be returned to service as soon as possibl During the. time of inoperability, an RCS leak rate calculation must be performed' every eight (8) hour If a-MCA is not returned , to service within seven (7)
, '
days, then reduce power to 5 50%.within the next four ,
,
(4) hour If blowdown' flow indication is inoperable, then l estimate flow when require ! If both the N-16 and air ejector radiation monitors
[* '
are inoperable, then reduce power to > 50% under
< , normal controlled conditions, but not to exceed ninety (90) minute If other combinations of required equipment or systems are not available,- then notify Superintendent Operations (or Assistant Station Manager O&M) and the STA. A JC0 will be required as well as discussions with the NR . No deviations to this Standing Order are allowed without prior approval of the SNSOC and concurrence of the Station 7, Manager or Vice President Nuclear Operation The NRC Resident Inspector will be notified of any approved deviation * Resident Inspector Comment: Per discussions with Licensee i Management, ary increase in leak rate will be considered to come )
from one steam generator and the actions will be taken based on one steam generator until proven otherwis '
.
l
=__ _
, . _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ %' d E c ,
+ , .
,
i (2) 1-pT-46.2, Primary to Secondary Leak Rate Determinatio This
, ,
surveillance procedure provides the requirements for the periodic sampling of the primary and secondary systems, periodic
>
monitoring and logging of the radiation monitor readings and the trending of this information. The procedure contains the acceptance criteria for the data and trends and requires the operator and STA's to ensure that the requirements of Standing Order #155 are met. The following is a brief description of the m procedure: r F 1-PT-46.2 is required to be performed every 24 hours any time the Unit is operating at power levels greater than 5%. The 24 hour surveillance is broken down into six four hour interval Each of the intervals requires a certain amount of data to be collected and trended to meet the.following criteria:
(a) Monitor, log and trend the N16 radiation monitor data every four hour , '(b) Monitor.. log and trend the condenser air ejector radiation monitor count rate and leak rate in gpd every four hour (c) Monitor., log and trend .the steam generator blowdown . radiation monitor every four hour (d) Obtain and count a steam jet air ejector grab sample every eight hour , (e) Obtain a secondary side activity sample from each steam generator every 24 hours and trend the information in terms of leak' rate in gp '(f) Perform a primary RCS isotopic sample every 24 hours and log the informatio (3) 1-PT-46.2A, Condenser Air Ejector Radiation Monitor Alarm Setpoint Calculation. This procedure establishes the setpoint calibration criteria for the condenser air ejector radiation monitor and was - SNSOC (Safety Committee) approved for use on October 2, 1987. The criteria is based on establishing an early warning system of a potential tube rupture. The alarm is adjusted approximately 10 gpd above the normal operating count rate established by the most recent sample data. The procedure requires an operator aid to be placed by the monitor stating the
' a alarm setpoint in CPM and the corresponding gpd each time the setpoint is adjuste (4) 1-PT-46.2B, N-16 Radiation Monitor Alarm Setpoint Calculatio This procedure establishes the criteria for setting the alarm setpoint on the N-16 radiatiun monitor which was SNSOC approved L= :=.
g
-
,
g .
.c . g _, , . 1 s
i
'16 I
!
'
A( ' on 0ctober 2,1987. The criteria established by this procedure
.is'as follows: ! (a) First alarm set approximately 10 gpd above the normal '
reading '
. (b)_ Second' alarm set approximately 50 gpd above the normal
' ,
'
readings or 100 gpd whichever is less. An operator aid is also required.to be placed by the monitor. stating the alarm
'
setpoint (5) 1-PT-46.2C, Steam Generator Blowdown Radiation Monitor Alarm
.Setpoint Calculation. This procedure establishes the criteria e ifor setting the SG blowdown radiation monitor alarm setpoint and was approved for use by SNSOC on October 2, 1987. The criteria , < '. established .for ' the radiation monitor alarm setpoint was two times the most recent count rate. This procedure also requires an operator aid to'be placed next to the monitor to inform the + -operators of the alarm setpoin Verification of-the Stabilization of the ruptured SG tube R9-C51 per Vendor recommendation It should 'be noted that the actual .
stabil_izction was' performed by Westinghouse who is the vendor for the
SG and stabilization ' equipment. The' following procedures were
, <
reviewed and portions of, these procedures were witnessed by the inspectors:
(1). STD-FP-1987.-2976, Rev. 0,' Stabilization of Row 9, Column 51, "C" ' S/G. . Thi s ' procedure describes the equipment and procedures required .to install. the sectional tube spear and the cable
.
stabilizer in the SG "C" tube (Row 9, Column 51) cold and hot legs respectivel x ~
-.The cold leg side of R9-C51 from the seventh support plate down has. been removed and this procedure installed a sectional tube
-
spear through each of the support plates and up into the U-bend section of'R9-C51. This spear was then hydraulically expanded at each of.the support plates, the remaining section cf the U-bend.'and the tube sheet. During this operation, Westinghouse
+ ' personne1' misaligned the bladder at the seventh support plate location and the hydraulic expansion split the wall of the spear just' above the seventh support plate. Westinghouse determined
"' ' that this. situation was unacceptable and developed several methods of stabilizing the stabilize These trethods and related safety evaluations are discussed belo The hot leg section of R9-C51 was stabilized by the use of a cable with swaged sleeves, installed from the tube sheet to just above the. seventh support plat The tube sheet end of the cable was a mechanical plug which was mechanically expanded into i tr the tube sheet. The only problem Westinghouse had with the
'
[ '{
m i ._ ..
%$ UN dI ,
'
7 l >N[% 9 ';
;#H ~q i a" 3- ,; "
m ,
:..
.
r ,m', '_
,,
hpfg-[[yg
;.h [T M * ,
;_
<w x:
g 1,_ '- ..
. .s A . . .
installation ofI the. cable was associate'd .with the first. cable, ekf 5 <
'
,
$[pNNm *
' l'whichibecame; bound up in; the ~ installation equipment and was bp " . .N L idamaged. . l ALsecond cable ? exactly like the first one was LJ1
&
' , Y successfully? installe ycM :TheLeightla'djacent :tubesJ surrounding 1R9-C51 received sentinel i, '
,
,
; plugsLinJ the; tube sheetL on the icold . leg ' side and _ normal
. mechanical plugs on the. hot leg sid '
3% ; W@w@ ' ,
[(2) . STD-fps 1987-3005, iRow : 9, f Columni 51,: Tube Spear ' Modificatio . This(procedure provided:the -instructions for installing a sleeve
' '
Li,- g s.~ g4
-
into(the rupturedf areaTof; the spear. :This sleeve once installed Mb wasthydraulically_ expanded into the. spear just above the. rupture
'
Landyjust below the - seventh support . .pl ate. The sleeve , was W " _ : fsuccessfu11ylinstalled_with'theLexception.that the bladder burst g~
^' ,
, Lat fl9,000 psigiduring the _ lower expansion. The acceptance
-
; criteria was 20,500:(+500 -200) psig by. procedure. Westinghouse cf'
*
performed a'. safetyt evaluation, which will. be discussed :later,
,,
.
fE and determined the sleeve to be acceptable as'i F ,, ,, - .
.
T .
.
4 E ' g' , L . (3) . EWR _.87-558. j Safetyc Evaluation for the. stabilization of the ~ cold o: leg and Lhotc leg portio'ns of R9C. 51 with ' a jointed spear and a
"
g' .' <
,
,
,cableirespectivelyi This'EWR was approved by SNSOC on September g ,
5, L1987. LThe' safety evaluation, concluded that' the installation
*
.
'
: of; a spear?and. sup' p ort cable 'into R9-C51 of !'C" steam generator h'L cdid~'not/ represent an unreviewed-safety question.
m
~(4); -EWR 87-558A. - Safety Evaluation for. the installation of a sleeve (%
..
'intoL the c stabilizing spear in the area of the rupture, to
,
'
,
< stabilize 1.the spear. ' This evaluation concluded that the sleeve
: would prevent the . rupture in the spear from circumfirentially
.,f . expanding "and causing the spear to' result in a double ended
't ,
breakn This, EWR,, which was approved by SNSOC on October 2,
' '
..
1987,Talso Econcluded that the installation of the sleeve would e< ,
,not present an unreviewed safety questio '
Rr (5)' EWR 87-558B. _. This- EWR ' evaluated the failure of the lower
section' of the' sleeve to be expanded to a pressure of 20,500
-
~
(+500 -200) psig as ' discussed above in Section B .1 ',
g" Westinghouse determined based:on the pressure time plot recorded '
' 'during the expansion' process that the breakaway torque for a " ' '
ijoint expanded to 19,000 psi was sufficient to ensure the joint J .was tight; and within . the bounds of the engineering analysis D' s performed 'as. part of the design justification. Therefore, they
+
W , ' concluded that the condition was acceptable as i ; 'i '
:(6); Letter from- Easterling, Westinghouse SGMA Manager to l w
L. M.-Hartz, VEPCO. This letter documented the problem
'
associated with the installation of the cable into the hot leg side'of R9-C51 a's discussed in B.1. above. The inspector, aware B,- r i
'.. ,.,)
y a __ _ _ - _ _ __- '
,s - ~-- - - - - --- m g*:4['t y['.
[ W'. a
'
4 .
'
.
i
'
< ., n
&=m",~~ l ,, . c w ti
,
,g
*
._ .$'p's hk:N
'
>
,
. ' ; j. <
- { ).[ ; ~ ,: !i{
, W
. of the. problem; associated with; the installation of 'the cable,
'
:
' y$ ,
' couldL not'. find any. documentation,: evaluation or corrective Rg j (action; associatediwith the: problem in the'_ procedure used for > 'i '
the-installation. ' Ba' sed ? on 'a request from the inspector y ' .A ' < n (Westinghouse provided the (l.icensee- with. the letter to document
>c m, (the problem, on October' 9,1987, 'and the SNSOC reviewed it the
,q; same day followed by the inspectors-revie '
~
'
1 u' ? Verification that1the; .cteam L Generator. Downcomer Flow Re stri'ct'or q ' Plates have been:-installed in conformance with-vendor (Westinghouse) y4 f recommendations. sThe installation:.of these plates 1s' also discussed
< iCI_nspection' Report 87-35. The. followingE procedures were reviewed'
M
"
'
.by.the-inspectors:
:
' ' '
<
(1); : Design Change Procedure 5 (DCP) 87-02; This ' DCP., was ~ reviewed-
* _ . .* . : durings thel performance _ of' .ther insta11ation of downcomer . flow
'
a
% '
,
t restrictor plates (see Inspection ' Report 87-35 L for additional details)L . The 'DCP< wasi aiso reviewed"at its completion by the
, ,
- w inspector L,4 -(2')f : SP2.7.2 GEN 4. .nThis . isithe Westinghouse procedure used by their
: personnel mto: install the downcomer flow: restrictor plates under
' '
'
,
l , EDCP87-0 '
; s '
(3) ; EWR-87-572.1This : documents : the safety evaluation performed by
"
y' [A , , the a licensee ' to = determine if. the insta11ation of the downcomer
~
' flowirestrictor' plates into thel SG's represents an unreviewed
' s'
+ 4
,
' safety , que sti on . .The.. licensee concluded that it did represent
<
.an unreviewed ' safety ' question. and has requested the NRC review
% >
'
jand' approve the installation. The NRC approved the installation
' '
fore power levels not to exceed 50% on October 9,1987, and will m v < continue to review the installation > for 100% ' power operatio *- This approval will be irequired 'for 100% power operation
;following the steam generator tube rupture.and is need for both unit '
T Verification ' that applicable procedures were followed to ensure that N 4 . loose: parts accountability was maintained on the steam generators Th'e do11owing. 'are procedures : reviewed and observations made during i the - in'spection: - X i(1)- Quality; Assurance ~ Activity Report . AR-N-87-506. The quality , o < i , assurance inspection performed and documented by this report was l
'at tho request;of the inspector. During the performance of the
~ ~
: secondary : side steam l generator work, Westinghouse personnel-maintained a. log 'of everything that entered and exited the steam
,
p generators. The inspector reviewed and discussed the condition of;the log with Westinghouse at the completion of the steam generator' work. Numerous discrepancies were contained in the
'
,
' log - and the inspector could not determine if loose parts
.
.
5 y
' b h( f ,'I .m .
o w7% -- i z^ <v khMk p lT l h, , '
- A 7,
, , 13 s
is
- # . :
m
. ~, <
,; ; c , ,
'
M. ih m (m, ,
..
.g
. .. N y; ,?l l
, y ' '
.
,
..
~ , :acccunt' ability =hadL been properly _ maintained on the steam y*, <
4 <
~g! generators. ~ . Westinghouse personnel Lreviewed the log and 7 w. , , m ,i ' informed:1.the " inspector e that . they .had resolved all of the
) discrepancies =. However,. there.was-no documentation. indicating
" ~
-
Q; P - "7
?tha't;the discrepancies existed and how they were corrected. .
N 7~ Therefore,ithe . licensee's Qua
?toireview .the l logs and_ documeliy Assurance group .was requested
$" , nt the. discrepancies .and ensure
'
-
' '
.W that } al1 L1oose - parts 4 that entered ' the.L steam - generators were -
p? ' : accountedi for. Following ithe0 review,e the inspector ~ discussed
[(j
, -
^' ^ ..
'E 4 ;the . report with the Quality Assurance -Supervisor' and' determined P,% Jthat manyL discrepancies had been discovered and documented; by
' '
theDreport. The: inspector 1was iinformed that after the;
i f, . ' licensee's review and discussionLwith Westinghouse all'but-three lofithe discrepant' items'were' resolved. The three items were.one
~
'
'
<
"_ half' of a ' horseshoe' magnet, a six foot nylon rope. and. one weld
, 7 s >
l N ' i istub. Leach of>these items were felt not to have been left in
.3y othelsteamJgenerators, but.'since they cannot be accounted for,
.safetyLevaluationsiwere performed to determine if'the items will-UNL ff4 '. ' Jadversely; affect the steam. generators. These safety evaluations-j *
,
are: discussed below, m ' y% ,
,, , 1(2) iMSR$2.2.2fVRdW16 :Rev. 0, Installation and Removal 'of. Primary
% ,
~LManwayiIns'ertstand Nozzle Covers. This procedure provided the.
@$ 6 : instructions to remove the steam generator primary side manways qE ', y Landrinstall; aluminum nozzle ' covers in - each ~ of the RCS ~ loop openi ng s'. . Thesei nozzle covers were . installed to maintain
~ ,
j ; accountability on the. primary side of the ' steam . generators and -
, , RC ,
i Duringi the; review. of the procedure, the inspector discovered a
' requirement. for an . accountability- log to be maintained anytime
'
>
'
<
c the Lnozzlelcoversiare removed. When . .the inspector asked . to
@ .
'~
treview; the.' log, .he. was; informed that - as soon as the nozzle
; covers Jwere1 removed -the diaphragms were installed, therefore,
-
.the licensee felt there was not a need to maintain a lo The
- , . inspector requested the licensee change the procedure.to reflect
' "" '~<. ' - '
the way, the work was actually p'erformed. .The procedure change ,
'
~wasiapproved on October 9, 198 ,
Also the-inspector' discovered that the Quality Control (QC) hold j t >
.po1_nt in the procedure was after the step 'which performed the
"
% '
' final removal of' the - nozzle covers. In order for an effective )
$ "g" .
W inspection for accountability to be performed, the QC inspector
# :would have to perform the inspection just prior to the nozzio ,
4., . cover -- remova' The licensee's maintenance superintendent
., -
)
M < assured the inspector that based on discussions he had with the
.QC Jinspectors, the inspection was performed prior to nozzle
!
m
"
, cover: removal. . The Unit 2. procedures were changed to correct AB
~ this: situatio ,
,
( l
>
u e 5 ,
, l 2 ' < ! '
c r s mz= = _ _ .
m;nQ[pv v
n~'
,
, y . +
g ' '
, . .
jfy ' c
* ,
mx
, c ,
i f' ;20'
. . .
%. . . X
. "q
.
(3))'MSR : 2.2.21 VRA-6, ;Re , Post; _ Activity Signoff for Area 9h >
'
' ' Cleanliness Primary Side ~ Servicesi This procedure . documented .
-
'
, , Lthe: verification that the;1icensee; agreed with Westinghouse that-cleanliness and accountability had-been-maintaine '
,
)
l The; procedure stated that the licensee.would designate a person
# lto.. accompany Westinghouse' on _ the- inspections for cleanlines m.'m However,1 based on :the licensee personnel who ' signed off the f procedure 'and? earlier reviews of the procedu e, the inspector ' ' '
- @, .
' >
1 knew that'thecperson/who signed the procedure did not accompany % , Westinghouse..on .the . inspections. When. the . licensee .was X ,
,
0; questioned, . thel inspector was- informed that the signoffs were -l made based onLother inspections'that had_been performed by other y : personnleli :The inspector again . requested . the licensee . change l
,
:the procedure to reflect.what actually happened. This procedure 1
: change;was.~ approved by the' licensee on October 9, 198 ' ("C" Ste' am Generator only), Steam
'
..e , g ' ' (4); ? MSR f 2.4.2 L GEN-31 - Re ~ ~
. Generatory Loop Inspection and Object Retrieval, this procedure
'
'
: documented ~.the' attempt to. find .an object believed to 'have been
'
, (dropped .inEthe' "C'l RCS ' cold leg. The procedure was conducted
" t W using al Foreign 0bject Retrieval : Device (FORD) in the "C" RCS E
l , icold leg; ,however, no foreign material was found in the RCS yusingf this equipment. The licensee also radiographer the drain
'line.off' the."C" cold leg and this' did not reveal any foreign
'
material._4In atletter from Westinghouse to the licensee dated
'
.
. October 9,l.1987, the licensee was informed that it was l Westinghouse's' position that the object was not dropped into the
'. '
"C" ORCS' cold leg. This letter is discussed below.
e+
(5) 'SECL-87-245 Steam Generator Secondary Side Foreign Objects, this '
safety evaluation documented the presence.of foreign material in
,
_ the: secondary side of the "B" and "C" steam generators prior to the" tube rupture,. The objects consisted of a 2 inch piece of slag in "B"x steam generator and a .2 inch . metallic strip in the
.
>
'! C" steam : generato Attempts to remove these items were unsuccessfu The safety evaluation concluded that an unreviewed safety; question did not exist for the next fuel cycle
,
with :the foreign objects left in- the steam generator SNSOC . reviewed this safety evaluation on October 6, 198 '
,
'(6) SECL-87-502, Steam Generator Foreign ' Objects Remaining in the
,
Secondary, this was a safety evaluation of the impact of foreign objects that were discovered and/or left in the steam generator following the - recent steam generator inspections. These items are as follows: '"B" steam generator - small cylindrical object,
"C" steam generator - brass and steel nozzle and two weld stubs,
"A" steam generator - one half of a horseshoe magnet and a six foot nylon rope. Each of these items were evaluated for their
, potential adverse affects on the steam generators including
>
2:
' . . . , . ,,
- _
,_ _
-
g, 5 ,it p
.
their chemical content and. wear potential. The conclusion that Westinghouse and the licensee reached was that no unreviewed safety question existe This safety evaluation was SNSOC approved.on October 8, 198 (7) Letter f rom . R. N. Easterling, . Westinghouse SGMA Manager to
'L. N.'Hartz, VEPC0, dated October 9, 1987, Foreign Object Potential in."C" Steam Generator Cold Leg in Unit 1, this letter , ,
documented Westinghouse'siposition that a foreign object which they could not account for _ did not exist in the "C" RCS cold leg. This letter was SNSOC approved on October 9,198 .
.. (8) Letter from Nuclear Safety Engineering Department to the Safety Committee (SNSOC) dated October 9, 1987, Loose Parts and Vibration - Alarms While Shut Down. This letter documented a review of. the loose -parts and vibration alarms which existed ( during ' the Unit I heat up. The letter concluded that these ; , . alarms were normal and.did not represent any indication of loose i parts in the RCS or steam generator The SNSOC reviewed the letter on October 9, 198 Verify- the Operability of the N-16 radiation monitors. This review involved the following procedures: (1) Engineering Work Request (EWR) 87-569. This procedure installed ~ and calibrated the N-16 monitor and equipment. The actual detector was placed on the mair steam header in the turbine building with a read out and recorder in the control room. The control room instruments indica.te in gallons per day primary to secondary , leak rate. The dett:ctor was calibrated within three different sources' each having a different energy gamm This '
calibration verified that the instrument was indicating the proper energy range of gammar, so that only N-16 will be counte j (2) '1-ST-75', Correlation Between the N-16 Monitor and the Total Primary-to-Secondary Leak Rat The purpose of this procedure is to compare the N-16 readings to actual primary-to-secondary leak rates. The procedure was first performed at 30% power and the measured primary-to-secondary leak rate using wet chemistry
<
t analysis indicated zero leakage as did the N-16 monitor. The licensee will perform th$s procedure at 50% and 100% powe Also if no primary-to-secondary leakage is indicated, then the
* procedure will be reperformed if leakage develope (3) The inspector observed the operation of the N-16 monitor numerous times prior to the unit exceeding 30% powe It appeared to be fully operational, but since there was little to no primary-to-secondary leakage, the N-16 monitor almost always indicated zero. The inspector also reviewed the Alarm Response procedure for the N-16 alarm and after a few changes found it to be adequat : --:.-
, _
. _ _ _ _ _ _ _
pj ' $
* ,
L,^
,
m p h L : Miscellaneous-(1) The following ' safety evaluations were also reviewe Each of these' evaluations were reviewed by .the licensee's safety l
, committee' and. determined not to involve an unreviewed safety ,
questio l (a)? SECL-87-230 Rev.1,- Leaking Row 1 Explosive Plugs. This ' T ' evaluation documented Westinghouse's conclusion that the unit could operate for a complete fuel cycle without rick of an overpressurization failure of one of the leaking now
'
1 tubes. ' This safety evaluation was SNSOC approved on
<
October.7,:1987-. !
-(b) EWR-87-583, Safety Evaluation for the Installation of Sentinel Plugs in the Steam Generators. This evaluation 0 concluded that use of sentinel plugs (a normal . tube plug with a 13.5 mil hole in the center) to remove a tube, which did not: require plugging due to eddy current testing, from service > was acceptabl These plugs- would allow approximately 250 gpd leakage if the associated tube rupture .This would allow .the licensee the means to identify ,the failed tube without causing a severe transient on the _ plan This evaluation was SNSOC approved on September 19,.198 (c) .SECL ~ 87-437, Lodged Eddy Current Probes Within Steam Generator' Tubes. This evaluation discussed the potential problems associated with the two eddy current probes left in the "A" and. "C" steam generators. Consideration of the '
materials and chemical composition of the probes and their affect on the tubes was ma d e'. The evaluation also considered the affects'of the probes on other tubes if the tubes they were in ruptured. The evaluation concluded that based on the above considerations and since the tubes with the stuck eddy current probes would be plugged, there would be no detrimental affects' on the steam generator The Safety Committee approved the evaluation on October 7,
-
198 (d) SECL 87-453, Safety Evaluation for Bulpd Steam Generator Tubes'. This evaluation documented the effects of Westinghouse's attempt to stress relieve steam generator tubes at support plate intersections and the resulting tube deformation. All of the affected tubes were plugged and
. consequently, the evaluation concluded that there would be no detrimental effects on the associated steam generator This evaluation was approved by SNSOC on October 9, 198 _ _ _
hdkMN %ps ' N3 7" ' i., ,i x ," .
' I
'
n E yp{ , , ,' % d = t q wa - * y Mh]%%? .,# j ..
~23 75;W :. e \ & '
v.py. , 1 8' +
'
7 sg y m q ' g , L(2).!MRS/212VRA-3'Rev.4',VideoInspectio_n_andTubeIdentification f'
'
~
,
'
; ;of "Steame Generator ' Tube J Sheet. ' .This , procedure provide .
. instructions)for. the secondary- side, pressure _ test of the steam
~
' '
'
,
N gf I. . > * y generatorito. determi.ne"if ? any plugsfor tubes leak. The only Ji . l leak was discovered on the' "B" steam generator _ hot leg side from e ;upp a
,
--"a tube.that:was'aiready plugged.
+, . . . . . g#s 1 ; "' L .(3)f MRS. 2.2i2LVRA-2. Rev.' 1, . SteamL Generator Tubesheet Marking. This . M 5 " procedure" provideds instructions? for : the licensee's Quality
-
*
l&,<, x ; Control groupi to . verify that' the: right steam generator tubes
@; 9 m y . ,
'
.were plugge :
f I
/y .r' ( f
> m, * .(4)[..UnitLITSteamlGeneratorPlugList $pl * ., . . , ., ,
.
W '" ~
~
' '
'(a)';'!A" st' am' e generator had the 'following tubes Eplugged' during (theltube% rupture outage;; 80-sentinel plugs, 23-required
'
i f l , ^, > - ;
'
'
qr
,
iplugging: due' to ' eddy current . testing,11-could. not get an N, G! <
, ,' .
: eddy ; current probe' into 'the tube . and - 1-tube with a stuck ifs > ; a eddy' current probe.g d mx
.
.
.... .>, .
.
rf ut F ~ .T(. b)? '"B'! J. . team generator ha'd.the following' tub.es plugged during
'
, x
'M, * thejtube ruptu're Loutage; :12-sentinel _. plugs, 20-required JM , , ; plugging due :.to l eddy ~ cur. rent tes. ting,'.1-could not get an - f " i f' * , y^ eddy current probe Linto , the tube and 6-tubes . that were y,
'
bulged due.to support plate heat treatment.
fc ; . ~. . . .i
%' -(c) 7"C'.'i steam generator had the' following tubes plugged during
'
'
s4 ,' .
* ~
the; tube rupture 4 outage;;70-sentinel plugs, 35-requiring
'
. plugging: due. to eddy current. testing, 2-could- not get an -
.
% '
<
2 eddy current probe; into the tube,_1-failed ' tube _(R9-C51),
.and_-1-tube with a' stuck eddy current prob _
'
i ' ' >.^ 4
-
w
"
'
,I l(5) The111cen'see _' conducted training; with ; the licensed operators on
%ny s -
: October-7, 8:and;9 covering the following_ subjects:
N"\ ,
~
(a)) Oyster Creek Safety.' Limit' Violation -
g .($ Manipulating Reactor Controls Instruction
y ' 4
; l(c)! Basis.forl returning Unit:1:to service ' > -' .# : $>,< ,
1).Downcomer Modifications M ? :2)~ Steam Generator 'Insp'ection:
-
, s , 3) R9-C51sTube stabilization . _ , , ,
'
4) Primary to Secondary Leak Rate Surveillance
.
,
"z G,y
'
5); Standing Order number 155-
DS (w t The~1icensee'also conducted training with the Shift Technical Advisors on o the'new primary to secondary. leak rate surveillance procedure . -
-
Q r i F
...'.
) . -,. ('
9 y
' .
jg ' .. > _b.- _ _
h gy h b g " nh a 5 'Uwh ' ,
,
f f, ? ?? , W "
,
, , - -
p %. 1 24 %
. ., > .
qp ,m,, .,
, , e .,
s --
,
x , L> x'
; Unit' 2" o 4 <
pm
,
,
"Thehlicensee ' completed the steam generator inspection of Unit 2 generators 3 ;using th'efsame c inspection- and plugging criteria as was used on Unit 1-
'T' Jsteam ' generators.;l.The .following .is a brief of the Unit 2 plug list; "A" iSG:,h'ad 45-) Sentinel plugs 1and 9-' tubes requiring plugging due .to edd "
- ,f W ,"? '
f currenti testing,.' ."B": SG' had 129- ! sentinel plugs - and ,4- tubes. requiring
- , <
Lpluggiln'g due to-eddy current testing:and ".C" SG 44 . sentinel plugs and 17- 'j
'-
'g, l tubes requiring-'pluggi.ng due.to eddy current testin % > .
..
^ 'TheT11censeelaiso completed the installation of downcomer flow restrictors s "
'
*
?: > platesiin ;the ' Unit:2 : steam generators. ' This ' modification was performed. as
';a resultLof.the,Unitil tube; rupture even .'
, , , .+ . A"?, Duringl thelperformance" of the is'econdary side pressure test of 'the Unit ~ 2 7 w :"Cisteamfgenerator 'theilicenseeldiscovered -five row two tubes . leaking'.
. ,<g, .
JThislieakage;was: an Vindication'. of through' wall cracks in the ;1eaking
, j" , tubes. - Even lthough these; tubes had. been. throughly: inspected using the -
' - ilatestf ieddy1 current J testing l techniques these ' indications -were not
.
/ . ,
'< '
.
4 detectedt LWestinghousef re-e'xamined- the ' eddy current analysis -data for all theirow two E tubes e on?the1 Unit . 2 "C" SG. During this examination
.
, M~ , +
"
S e of,
"
c
..f 3 Westinghouse :usedt a(different set of review techniques and was. able to L
W
,
N' . detect.the' indications in the . leaking' tubes. All the' data on the row.two E 4" " itubes :forL the rest of the SGs 'in both units was 're-examined using the same yg' "
, ; techniquesLas were : used: on; the' Unit 2 "C"; SG. . Westinghouse reported.to l '
the licensee thatinciother indications _were:found.
.
'
n,. aThe-indications descovered 'on the leaking tubes in the "C" SG were located
- ' ' '
in the"lU-bend s'ection'of the tubes. These row two tubes were heat. treated t ; 'g "by; Westinghouse-in;anLattempt to relieve.the stresses in_the U-bends. The
'
'?licenseejfeels: thatLthis heat treatment' contributed to or caused these
"' ' Lthrough wall: cracks;-
- -
'
,
.i h'
gp7 ,
- i.' 4 ~
{.< -c
' -
a
.l y \ ' ? . , . ,
'I )
4 '\ , _ _ . _ . . . - _ . _ - . . . _ _ - - }}