ML20064A642
| ML20064A642 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/11/1990 |
| From: | Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NYN-90166, NUDOCS 9009170145 | |
| Download: ML20064A642 (145) | |
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s 1 l POWER ASCENSION. TEST PROGRAM a ,
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-, FINAL PHASE 2 V
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NYN-90166 September 11, 1990 t h United States Nuclear Regulatory Commission i Washington, D.C. 20555 Attention. Document Control Desk
References:
(a) Facility Operating License No. NPF-86, Docket No. 50 443 (b) NHY Letter NYN 90127 dated June 18,1990, " Power Ascension Test Program Fifty Percent Power Self Assessment Report", T. C. Feigenbaum to USNRC
Subject:
Power Ascension Test Program Final Phase 2 Self Assessment Report Gentlemce: New Hampshire Yankee (NH't , performed a self assessment of the Seabrook Statica Power Ascension Test Program (PATP). The self assessment effort was conducted in two phases, Phase 1 which assessed preparations for the PATP and Phase 2 which assessed conduct of the program. The initial Phase 2 report covering activities through the completion of testing at fifty percent power was submitted via Reference (b). This letter forwards the Gnal Phase 2 Self Assessment Report for tl.e PATP which ccvers activities conducted from the end of fifty percent power testing to the completion of the PATP. New Hampshire Yankee management personnel will discuss the self assessment as part of the presentation during the September 18, 1990 meeting regarding the PATP. Should you require additional information regarding this matter, please contact me, or Mr. Edward W. Desmarais, Independent Review Team Manager, at (603) 474-9521, Extension 3101. Very truly yours, ()vk & Ted C. Feigenbaum Enclosure TCF:JMP/ssi g\ i \ \ I New Hampshire Yankee Division of Public Service Company of New Hompshire P.O. Pox 300
- Seabrook, NH 03874
- Telephone (603) 474 9521
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United States N.uclear Rcgulatory Commission ~ September 11' 1990 ,
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- i. I P.O. Box 1149 Seabrook, NH 03874:
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.- :220 7 AREAS 0F EVALUATION v
i . . 3 12.1 Evaluation Of Power Ascension' 4.-- 14 !- 4 m 4:; m. '2.21' Evaluation Of Operations 15 -- 22 ;
, '2.3 : Evaluation Of: Maintenance and Work Control
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2.5. Evaluation'Of Radioactive Waste- 44l-- 45' h ,
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r ;" ~. . .. . 2.7- Evaluation Of Training- : 49 --~51
,, [2.8II Evaluati6n Of" Quality. Programs- 52 -- 55
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2.10 Evaluation'Of Performance: Indicators 69 -- 90 2.11' Evaluation Of' Management ~Effactiveness> 90 -- 97 j
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, Ai , N 3.0' PHASE;2 RECOMMENDATION INDEX
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{4.0 PHASE l1 RECOMMENDATION INDEX 1r 1 7 b
.' Appendix A - SAT Test. Result'Reviewa a't 75% and 100Z Power l
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,1 I h l Appendix B - Chronology of Evente j
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'The self assessment of the Power A9cension-Test Programf(PATP) was conducted: ,
} ' in ' two phases. ' Phase-1~ addressed the : readiness to begin: the - PATP 'and ; covared preparations' 'for. testing including the resolution of open f
- l. . inspection, ' audit and -'self-critique issues.- Phase 2 addressed. thel kY <
implementation of the PATP and the conduct of plant operations.. The basis. . for the assessment was the Phase.1 and 2 performance objectives and criteria contained in the Power l Ascension Self Assessment Team (SAT) Criteria' Book. M Results of the Phase 1 assessment were issued by the' Phase ' Interim Report 4
, of. December .14, 1989' and the Phase 1 Supplemental - Report of December. 22, r
1989. ~ The initial results of the Phase 2 assessment were issued by ; the -l Fifty Percent Report of: June 18,:1990. The final results.are-being issued: ! by;this report.. , This - report . covers (1) i Phase 2 performance objectives and ' criteria not . '[
;previously. evaluated bec:ause of power level prerequisites, ( 2)L the f.>' lsignificant plant problems that have-occurred since the Fifty Percent' Report
[ was f is sued , (3) areas with potential weaknesses that warranted furtheri review,Jand.(4), areas of special interest. The items covered by this report are addressed;under functional areas. These areas-are the PATP, Operations, '
- Maintenance, ' Radiation Protection. Radioactive Waste, Chemistry, Quality ;l Programs, Training, Engineering and Technical : Support,. Performance j
.I;' , ' Indicators., and Management Effectiveness.-
r The caclusions' of the assessment of activities conducted after completing-
. testing at fifty percent power follow New Hampshire Yankee (NHY) continued testing in'a deliberate, cautious, ~
.and conservative manner throughout the remainder of the PATP.
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.-- Two reactor trips occurred during the PATP. The reactor trips were I
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caused by hardware problems with the turbine generator. No operator 'i'. errors were involved. a p .
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; operators 6 and d technical ? personnelE are i proficient in noperating J and . .
gj mal'ntaining the - planti f andJ operating ' and' emergency procedures i support [y '
' safe. plant; operations.- ,l l7 m ' t; '
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pf Management was; actively involved. in the , operation and testing of thei 3
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. Technical ' Support, and . Maintenance organizations
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s y ; effectively' supported the PATP by rapidly resolving plant problems'. 1 u -
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-- - Health' Physics 1 organization ' demonstrated proficiency 'in minimizing (
c s, radiation-exposure during full power operations. .! t i MaintenanceL Program addressed progranaatic ' aspects ' of ' the ' NRC :
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' Maintenance- Inspection _ Tree. Saveral areas warrant enhancement.
These: areas include rework, maintenance' history, root:causel analysis,. plant: aging. -plant. life--extension, reliability centered preventive- N 4 4' ; maintenance, and performance measurement.. .l ( - Nuclear Quality Group demonstrated the Ollity to identify significant '{ g problems - and- to pursue ~ the ; implementation of recommended corrective , actions-from audits, surveillances, and inspections. . 1
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-- The~ number of open: Requests for Engineering' Services-(RESs) and other. .j Engineering. activities are decreasing' as expected and previously i
, planned.' The open RESs are not impacting scheduled maintenance ; and . :[
' modification activities.
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~The number of open Work Requests-(WRs) that are not non-plant related,-
J -- awaiting! plant conditions, outage work,'or future design enhancements i-m ,1 L: 7 is beginning to decrease. The open WRs-have negligible impact on' plant safety.and reliability. s Six (6) additional reconunendations are included in this report bringing the ' total number of recommendations made by the SAT during Phase 2 to. fifty-nine ' (59). None-of the recommendations affected completing the PATP or reaching /l a fully operational status. Seven (7) of the recommendations have been j , 2' e l i
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< -4; this report'provides.aLeummary of Phase!2' recommendations.: ".
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- d ' New - Hampshire Yankee has< completed' evaluating .the iniplementation and s performancel of J thel Power 1 Ascension Test' Program (PATP) : to. assure further l- 1 .that the ; required - design 1 bases were safely and conservatively validated- as I
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.the Station.was brought to full power operations. Based on the. observations
-and the evaluations of the SAT during-this process, NHY' determined that the; h
,n; e PATPfhase proven. the ability 'of ~ Seabrook ' Station , to meet < design standards 1
- g. . established: by; the Nuclear Regulatory Conunission. = NHY has also concluded L
' that the . appropriate ' interfaces required to support the. continued operation .
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.at :: f' ull' power of Seabrook Station . are in - place and properly focused lby
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Executive Management. The topics covered by this portion of the final SAT Power" Ascension Test: Program Evaluation are (1) interfaces, (2)' preparatory activities, and (3).. final conduct of the Power Ascension Test Program ~ . i o l, j~'
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. S" ry Of /Fin'dinns' 'i '
0 --iTesting' SAT Recommendations (ST)
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- O -- Conunercial Operation SAT Recommendations (ST)
- Long Term SAT Reconunindations- (LT) q
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Interfaces ' Ig . LE u
- The . Operations Group and the Power Ascension' Test Group (PATG)' were the two' 4
o . primary ' groups that ' interfaced during. and controlled the conduct of the : Power Ascension Test Program (PATP). The initial interface between these
, groups,' for: power escalation, occurred during the integrated PATP training-4 program :and continued throughout the PATP, with requalification o training, procedure revisions and-the actual conduct'of the PATP. The SAT observed a
.v. .>. . L / strong working relationship between these organizations which demonstrated a- [- 4 clear. understanding ~ of' their respective roles in the PATP. These L t observations are based on performance during trainings pre-test and pre-
. shift briefings and the final conduct of the PATP. The Shift s t -Superintendent (SS) and the Shift Test Director (STD) conducted pre-
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", ' understanding . : of . . their ' ~ responsibilities : during:Jupcoming shift- ' testing ,l F
$ activities.! The' SS and STD also verified procedure pre-requisites prior to.
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- starting al te'st, "andi monitored 'shif t y performance . for proper 1 comunand - and I
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= control j authority s between the .. Unit .' Shif t Supervisor - (USS) and ' thet Testi
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Director (TD) during the actual conduct of testing., ThelSAT' observed'.various'PATG presentations to theiStation Operation Review- 1 Connaittee (SORC)i during the conduct. of the PATP. -The~.information needed to.
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'effectivelf review: and approve the" required PATPL test procedures :;
subsequentcr N ions and test changes prior =to SORC' approval and.-continuedc 1
' testing activit les were adequately- presented. The procedures were revised L or changed, as . necessary, to support Station configuration for continued
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l operation and; testing. The PATG took deliberate measures 'to assure the ] clarity. of, completed y test packages prior to presenting to SORC . ' -These [
} -measures' were -taken to . ensure that a ' historical reviewer would have --
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,i sufficient information .to independently verify and . substantiate that: the~ j
; completed:- test results supported the appropriate procedure- acceptance ,
, ; criteria..
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'The secondary interfaces that occurred during the PATP included Maintenance.
$ Technical Support, Systems Support, Engineering, Health Physics, Chemisty , ! l Nuclear QualityJGroup '(NQG), and the. Training Department, as required. The SAT determined t. t both! Operations and the PATG took apy,ropriate action to--
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j E alleviate questions and concerns that arose during,the conduct of the PATP. s, J. - 4 [ ,0perations and the PATG intcrfaced routinely, with the- appropriate :) E i ; Maintenance Groups,-_to coordinate repairs to equipment. Some prime examples l-f
, . of this, interface during power -ascension occurred during the required re- . ) -
alignments of reactor controls, delta T and Tave instrumentation and axial
' flux --. diff erential instrumentation to more closely match actual plant.
conditions Las indicated by power . ascension test data results, using Work yg <
< Requect1. 90WR003455 to check core thermocouple H-03' which was noted to be reading. lows repairs to the heater drain discharge reducer, repair of a leak-ini the' EHC hydraulic system and repairs to the flex connector between the main generator and the generator output breaker. These interfaces minimized the - delays on the test program and ensured that Technical Specification requirements were not overlooked during the test evolutions. This interface i 5 L.m m -n .'
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coloni.includ:d' ths ; continu:d uso cf test equipment from' ths: Maintenancs; , I e 3 - Calibratien Facility and close. coordination' of; any lspecial; testi equipment' :l
\ : required;forl test. data; collection.
. 1 The interface between Operations and the PATG with the Technical Support -
[ l(Systems Support Group) was smooth and: effective. ' Technical assistance for' *
, ! the test- program was _ routinely =obtainedLfrom - either ; Technical' Support '
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-(Systeme. Support Group) or- Engineering,~ depending ~ on theicomplexity of the problem to be solved or the controlled' documents to be changed as determined
. 5, b by Engineering.. Technical Lassistance was requested to solve ' concerns -
- i identified: in such areas as stability of heater drain' system controls, EHC -
z d hydraulic lineivibration to #4 control valve, main feedwater pump vibration .; levels, stability of control interfaces between the feedwater and heater [ drain controls,:and hydrogen: consumption of the main generator. Management : L u.t ,
'showed strong commitment to L. a successful and efficient test' program by ]
[' ensuring that-. both groups provided the necessary around-the-clock coverage to'- solve technical concerns- as - they arose. The' Technical: Support 1 1
' Depa rtment.'s . required surveillances were niso effectively coordinated' .i through the SS/STD to ensure minimal impact on the power ascension testing.,
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j e Interfaces with Health Physics Department continue' in accordance with t
., Station procedures. The prime interface observed by the. SAT, during power !
ascension, was the successful implementation of the post-unit trip snubber ] outage which required,close coordination between Operations, Health Physics, sMaintenance and Technical Support. l > . O
, The weekly containment inspection entries, required by the Station Manager, l < ~
were well coordinated by the groups and provide additional detailed f
, 'information of equipment status in the containment. - over and above the Control Room. instrumentation.. These entries have demonstrated that Control f Room indication is - reliable to assess equipment status, and the Station
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h [ , Manager has relaxed ' the weekly containment entry to a longer interval so that ALARA~ considerations are optimized. iGi i l}- B, n Interfaces with the Training Department were precise and deliberate during conduct of power ascension. The Operations and PAT Groups were trained in
.an' integrated fashion. Assigned crews were subjected to both classroom and
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/during ths : PATPi Training . continuddl thecughI thO prtgram :cn a190-day cyclo 86 ;in Lwhicht refresher' training 1 was; provided . to ..the L integrated crews ' to ensure '
, . thath th'ey Lwere Ltrained ~and qualified ; on the ' most current' methodology to be - ,
Jused ' duringf the . actuali tests . :Requalification training _ classes _ emphasized' _
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[, , th'ose s sections of the L program andi procedures that had been' revisedEfor.
- additional - clarity ; and i er se, of z perforsance . from previous. training classes. :s j Simula,torL training effectively ' demonstrated proficiency _ in: the ^ areas of j
comunand _. and i control', f communications , test: termination criteria. : and u test: n acceptance criteria.- The' integrated, training also.' used a. balanced, I', ] Enerformance-oriented.: self-critique' feedback' process to strengthen 1 the' j 3 -knowledge of allitrainees.: The performance feedback : included both positive-areas and areas where improvements would make the test implementation more p, & effective. Additional emphasis was placed. on = specific crew" training to f _ assure that the most complex tests would be concentrated on for-preparation. f t The SAT. observed " crew" specific" re-training for ST-35' 'Large' Load
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(
) Reduction" .ST-38, " Unit Trip *: ST-22, ' Natural Circulation * :- and ST-39, I
' Loss _Of Off site Power Test' . ' These : observations - included' both classroom i j iand simulator - sessions'. - Comments noted during' re-training classes : were.
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} ; k --factored into the' specific test with a test change. ST-39, ' Loss of Offsite? ;[
_ Powe r' Te s t " , = te s t change.#1, is a 'specifici example of corrections -to aL
~ / d I : procedureJ resulting from .conunents generated during re-training. The STD 'i assured, that the ' specific crews" were in place prior to the' actual test performance. The - SAT ' observations indicated - that there . were no areas q x.
m , warranting: enhancements in this process. 1 observations and/or procedure reviews, showed' that the interface activities I required between the Power Ascension Test Group and' otherl' project eI ' organizations with specific.. interests in test 'results were deliberate, l 2 preci's e. and effective- in assuring the quality of the program met NHY .and I regulatory design requirements. j m
- Areas . of major ' interf ace effectiveness were found to be the command and y control, exercised between the Shift Superintendent (SS) and the Shift Test Director _(STD)Lin establishing proper initial test conditions, overall test
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4 --implementation,'the continuous communications between the Test Director (TD) '
'and the Unit Shift Supervisor (USS) during test procedure implementation, ; ~
l and the rapid support provided from the Technical Support, System Support and' Corporate - Engineering personnel in resolving test related impediments.
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' system Support engin: Ors oleo l pecvid:dt e streng point' of centcet fer system l a
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=related>information to support the yequired' testing or problem resolution.' '
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' Conduct .
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.The SNift Test' Directors maintained shift turnove!* status sheets, describing-plant status, and maintenance issues occurrir.g on their shif t _ as -well' as a .
I . chronology- of . overall . test L activities accomplishtd during the past.-shif t. These sheets and chronology'are in addition-to the chronology, maintained by ST-1, "Startup Program Administration'.. This allmred the . on-coming. PATG !l
. test crew _ to ! review previously completed activities aad to determine tbc status of'~on-going , testing and maintenance. Additionally, test . crew-turnovers involved verbal' briefings' with their shif t counterparts ; prior to '
participating in' the joint' PATG/ Operations pre-shif t/ test briefing, This' ' l methodology promoted a smooth and consistent continuation of test actisif.ies by- assuring that the on-coming' test c::ew knew the precise. status ' of - f 1
, . maintenance and testing activities that-affected continuing test activities. .;
< i Prior'to each . shift, on which testing could possibly occur, the'STD and SS '
conducted " pre-shif t/ test briefit os' ir. order to assure - that the ' integrated
.f
' shift crew clearly understood its responsibilities during the upcoming test.
i Frequently the on-shift USS would also conduct a mini-test briefing with his. '! Operations crew as . additional assurt.nce that responsibilities were clearly i understood. The-STD's and SS's closely monitored the start'of all tests to' ] assure that the appropriate- pre-requisites were completed to - support a - 1 successful test. Due to the turbine trip during implementation of ST-48, at 3 l m. 4 an ,early test plateau, which reraulted from inaccurate assumptions on the ,
- <- i status of an open EHC system Work Request, the System Readiness Program was upgraded by requiring that morte extensive independent reviews be performed prior to performing the related test. Corporate management was actively .;
involved-in this event evalut. tion and took deliberate and effective' action 39 . to develop corrective measures to minimize the potential for this type of , occurrence. This .resulted in no similar occurrences for the remainder of the PATP.
- lg) W Prior to performing a test, the Shift Test Directors validated, through ;
, . qualification record review, that the Test . personnel were qualified to
. 8 '
- l N . _ _
= ~ - -
f, , ,l- L+ y g f I . d, perfera th$ircass'ignid tsst's.i As: r:quir:d by. SM 8.1, thh STD clopossur d 'l
~3
~
/that theLJ crews ? assigned E to i specific .- tests . completed' _the a phe-requiuite: ,
%e*f (integrat'ed-training,for that: test.. ~
~
JPhior toj changingito' higher operating ~ modes. Station Managementiconducted
' ~
3 4 modei change 1 checklist meetings.- ~ The. SAT observed that f these meetings ,were' conducted Lin .a thorough, detailed'; fashion and - that . the Operations shift' ) R l : personnel clearlyLunderstood the.open issues' required'to be completed prior to~ changing modes. and followed these lists as outlined.- lThis: process was- ! , repeated? several; times.- during the - PATP and unplanned unit trip recoveries. h
, .These evaluations ' demonstrate- the- - adequacy of" this~ approach and!
" methodology. i
+' , t The SAT . observed ' proper ' command and control' in the conduct of 'the PATP.
. ' testing. ~ Respontibilities. authority and communications, internalo to both , h = . 4
. Operations 'and L the PATG, as - well as between the respective organizations. '
1:
;were' clearly understood and communicated. Test performances observed by:the 1
,[ i SAT demonstrated, that -Operations personnel - maintained ' a , strong . command t, i ( presence of the. Station during power ascension. TheyUnit Shift Supervisors- [ f (USSs) re-addressed ' topics covered ' during the pre-test briefing:to assure ! 4 that' all personnel involved with the test performance clearly understood . ' 1 _ their responsibilities and expected plant responses-prior to initiating the l test with the. Test Director. - The - primary topics re-addressed by the USSs y , ivere pre-requisites, special precautione, initial' conditions, j.
. .y _
~
responsibiliths and- expected- plant: . responses. The STDs.Iand Shift' ij i Superintend'snts -- (SSs ) continued to . maintain - close Loversight when. assured g that overalliplant operations were not compromised during test evolutions. , i , Major integrated. tests where this command presence and authority was. ' observed were S T- 3 4 ', " Load Swing *: ST-35, "Large Load Reduction": ST-38,
.f 'UnitiTrip Form 100% Power's ST-22, " Natural Circulation"; and ST-39, ' Loss j Of Offsite Power". -
Slow, cautious - and conservative conduct of the PATP was reflected- by e deliberate PATP interruptions to evaluate a particular. event, such as turbine trip, or to repair malfunctioning equipment that was detected during
- the course, of .te s ting . ' Additionally, the STD's coordinated problem s resolution with required . personnel, including contacting personnel not on i
- . shift. Some example problem areas where the STD's called appropriate l
- l. i . ,
f !: Y 1 3 F<
.;c
+
, , :g , [ support personnol ftr cosistincej weros; Ea1 pipe; crack cnithelinlet to the q
steam ' generator"Lblowdownt. flash:. tank (Sy. stem: Engineer): inadvertent p
.g.,, ,
g 1, T. Lactivation ' of the . ASDV'sL (Corporate ' Communications) T reheater drain tanki , 1
^
/
Elevel-transmitter-flange ' torquei sequence L (System Engineer): f excessive l shaf t - l f . leakage onf.' circulating : water} pump- (System Engineer Supervisor) : EHC.- n . ' hydradlic ;line' vibration '(Engineering):" and. vibration concerns ( afterl 'l relccation of EHC. low. pressure -reactor trip l: pressure' -switches-i ; (Engineering')'.- This approacha maintained a proper l balance; of effective,
, _ j
~
efficient: PATP Limplementation while concurrently maintaining c caution and-I conservatism.
' System configuration was. maintained through-the use-of Operations and Power-
, . Ascension Test-Program' configuration logs and processes. The SAT observed :i
, the. satisfactory use' of Operations logs ~ (e.g. , Danger Tag Log, Temporary K . Modification. Log,, Action Statement Status Tracking Log and the Temporary '
Setpoint Change Log) and the,PATP required processes'(e.g., Plant Material l Condition process and the. System Readiness List)r -The.'use'of.-these logs and ,i
, a j . processes,- combined with the ' utilization of . test personnel familiar with j
! -W' the specific . tests. . assured - successful completion of each test with only t 1
= minor exceptions. - Revisions and/or test changes were also generated, to -
.y ' assure- successful test completion for changing plant- operating-configurations. . The sign-off of the various configuration logs after review
{ for impact on tests is an acceptable method of completing ~ the power i
. ascension test prerequisites. These configuratloc. logs are . maintained as
},l ta permanent' plant records and can be reviewed against a given procedure based. j on dates. of test-perforance and entries to the various logs.
4
' 3 y i
During.the conduct of the PATP tests, the SAT observed timely notification-b , .~of E NQG by the . TDs to ensure - smooth implementation of each test. This process was made easier for the TDs as both 'NQG and the NRC usually had 1 personnel on shif t . around the clock to support the PATP.and perform their. own. programmatic reviews. When problems' occurred during the performance of Li 4 a test, the' Test Directors provided the STD's with the necessary information
'to revise, coordinate and schedule changes w'ith the Planning and Scheduling f Department. Accurate schedules focused the site organizations en I
g; appropriate priorities and work activities in support of testing. J: w 10
- 1. _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ - _ -
. , ,. , - L.
y C w
' :. fir.g 4~y ,t q' 'q -
, 3 ,
4 - ' i
~
. .n
= ? ths' ctnduct cffccch_PATP t3st', up,through' cad i 4
4 ; M, m ' Th% STDW t_ 1 cidssly; monittrcd ( ? ,; including the 100! power. warranty;run .:to ensure that procedures were/>einge
,g 1
gw, ~ " ) properly; followed.9 Operations and. PATP: individuals requested and' developed . r
] ,
Ask t Yn prior- to Leonducting- test evolutions.' The
~
- procedure l changes, f as nec < -
procedure 1 changes _.were app a aate'to reflect plant conditions orLimproved!
~
' test) methodology,- basedi on previous:. testing.- Significant- events ~ that, W:.. l g ' ..
occurred- during the performance- of =a test'. were documented in the' ; 78 r '( {-' . . . , _)
- chronological' log maintained by7 the TD for each
~
' t' est. - The : TD's' or L other
- 1 supporting = test; personnel? were observed? to . properly., record testEdata:
~
4 7
\ (required by their procedures.- .i L l
,' The - SAT observed the, conduct of . a portion' or the entire pc.rformance ~'of the O
,, ' , te s ts i identified z in Appendix A. . ' SAT ST Test Results Reilew AT 75! ' Power '
p. andi Appendix A, . ' SAT : ST Test - Results Review At '1002 Power'.
~
g j. The . SAT - i , observations ares recordedi on work sheets chat address the . primary topical; ~! y, areas'of " Interfaces', ' Preparatory Activities', and " Test Conduct'.. Within- 3 ie 1 feach of these-topical areas were minimum criteria that were= addressed during?
,f
.U . / .thelSAT observations ~... Key examples of:the: observations criteria follow:
' ' '[
( 2- a x InterfacesE '
) ,
a f, -- Department interfaces-to support the test. STD/SS establishment: of initial' conditions. .t Availability of-required N&TE. i I1 -Preoaratory Activities'
; ; r J --
Procedure details adequate to performLthe test.
- f. q -- _Non-standard _ plant lineups addressed.
' Acceptance criteria provided to gauge test success. ,;
Qualified OPS /PATG personnel performing test. L ( - = Plant configuration reviewed. E - Test Conduct J( i ue
- -- Pre test briefings adequate.
5
- -- Designated crews performing tests, as required.
-- Thorough communications between OPS and PATG.
11 5
- , ,i l}
\
q ng s , # < > y .. c - H e 1
<1
; y y; - r y 1
.i g YdV s-
' T:st: conduct:d.in ccccrd2nco with prce:duro.'-
F w -
.. . Test chronology thorough. '
p;wff wll _ '
~
A Upon' completion of[ a PATP > test procadure: at a given; test plateau;; the Test'-
~
x DirectorTperformed'the initial review'of-the test results package.tofassure 1 that ' allf required dataHhadL been properly recorded, that all. applicable - ! acceptance criteria were met for .the plateau - and : that all appropriate ; supporting documents ' were o enclosed (the System Engineer was consulted as ,f necessary). .This; initial review resulted in the-TestiDirector producing ai j F b _
' ' test 1 critique' that explained the results : of ' the test.- This - detailed- l v . -s y 0 review was then followed by an- additional . review: performed by. the Reactor; i
M Engineering, and ' Computer Department ~ Manager 'or the Technical, Support < a' l ' Program Manager: who' documented his analysis on the' " test results ' review' - s; , form. The! above ; detailed reviewjprocess was in accordance. with the requirements of SM 8.1.
. s Upon= completion of the. detailed review.of test results described above, the test results. package was subsequently' independently reviewed.by the.HQG with' ps 1
comments . resolved 1 prior to being- presented to SORC for final review and ' j
~
H
. approval. The presentation would be made by the : Reactor Engineering. and - d
' Computer Department Manager . or, the Technical . Support Program Manager with i the . assigned Test . Director, if available. The SORC Chairman would not ;
-approve the-test results until all SORC member' comments had been acceptably- 'I j
. resolved.-
t J L e l-[. m, The SAT E observed 1 a number of -test procedures processed through the. above~ 4 j [# ' review . cycle and concluded that the power -ascension ' test 'results were- 1
- c. ,
l 1\> thoroughly reviewed and all parties ~ involved in the-official review process J [ - . : . -
- L
, had.allL their comments thoroughly resolved prior to -final SORC approval. i pf ,
, ,This detailed and lengthy review. process took place. for all approved J v
W procedures to be-fully. implemented, partially-implemented at a test plateau !
- m o 4 / -
or repeated at:several test plateaus as required by the test program. This i
, , . process was . performed on thirty-five (35) base procedures that were L1 implemented, ninety-three (93) times during eight (8) major test plateaus.
( : , NNY has determined that the depth of the SORC review process has proven that i 1 ~it can support the plant at full power in a safe and reliable manner. Many 12 g . m/ ; ; i ~ NO) . , . ' '
-o k; , -
-l r
q' ' , g
^'
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@{'
' ..jg; th0 { programs, r: quired to c supporti th3' cpercting 'st2tiin .werol clon= ue:d:
]W yduringthe(PATPand; proven,LasLwell(e.g.Ltaggingfworkcontro3,frates'tingi, schedulingiand" surveillance performance).1 o . ,
V! ' ' @., 'y
* ~
}[
b LAs To'f . Auguetf 28,; 1990, the : SAT l had reviewed one hundred[ fortyl(140) I test , changes that. were : generated .against. the -final- power ascensioni testi procedures. 'These changes.can be broken down to'three. major categorie0, ,as' outlined below:- Vr ,
~(1) Plant ' CoEditilons .-- These_ changes were generated; due:: to_ equipment repairs, additional: vendor : checks, non-critiM instrsment failures, re-sequencing ^ steps. ori procedures. to conf um to; existing plant ~
[ conditions,'etc.
, !' s
. (2)' Imoroved - MethJw1UEY -- These changes were generated f due to test n.ethodology eing . adjusted to fit existing plant- conditions.:, improve i ,
- data- collection, . improve interfaces- with applicable- department .'
, procedures, minimias transients.,ladd vendor' plant / system iW% ' i '
' recommendations, etc.
m i I , I(3) Administrative: '-- These changes were generated; due to ' typographical-
' errors, correcting procedural references and.' equipment terminology. add. .
T data for Training Department use in qualifying the simulator to actual'
, plant conditions,' updating test procedures _to FSAR. changes, etc.
# - , 14
- j Th'e breakdown of the changes into these. categories is depictedtbelow '
e Ii Plant Conditions: -- 29Z
.3 1
f Improved Methodology -- 252
- Administrative -- 462 e
P.. i ! ,
- Based i on:anSAT member's experience with other test programs, this breakd'own g jisF consistent with . other new plants being brought to full power- through l
. initial power escalation. This assessment is based on the initial operation
'of the equipment and : systems under first. time power escalation operating L
u
' conditions, adjusting test methods to match cctual plant status and I ; conditions that were not fully known during procedu.e development and minor I';' : administrative errors that are noticed during the actual implementation of 13-i lx l "
, w r 8 a i(
wM
@ LD,' . y i xn ' W K' } m [ W W "hy ' w p , l ,
... = w ,.> .
. thi testiinstructirn.
Ng -. . . -, 4 Tht ottentien paid =to th:so^dsteils is indicative cf: ) n, a :r- ~ . . 9: n ., .c
?
..the- slow andf deliberate'ammwr' in,which the PATP was ' conducted. , ,
,. ~ 1
[ 3 1 4 1[
'4 '
InL addf. tim to. the! review of = test procedurci ch ,nges, the SAT also reviewed .
.. , . w -. .
.the stest j results i for' allitests - performed sti thel 752,- 901, andc1002 test ~
I.w. i plateaus.' This(method is' consistent with the approach the' SAT:used' prior to- ; W' 'the' approval <to exceed thef50Z power plateau. The SATLtest results reviews i _ are-J described ['in Appendix A, as - presented ' to the Management Oversight-Committee" for f thea 751~ plateau on July '14 ' .1990. ' The SAT - test - results - , reviewsf fori he t 901 and 11002 plateaus are described in Appendix A with the -
- corresponding SATf review dates identified.' The SORC reviews' of the :PATP
- final' tests e.re currently in progressfand are scheduled for completion prior' to September 18,.1990.
~
e .<
< f L-
- The ST-1 and'15; Procedures are not yet SORC approved and'are not anticipated-3- to alter the overall SAT conclusions, because they are administrative in s
' ~
nature,L' designed lto record sequences of events and final instrument settings- ; og; established in the remainder of,the actual tests. st ST-43. ' Process Computer' and - ST-46, ' Ventilation' have not yet had their
; - t D i
j
, = test data assembledtby the respective Test Directors.- Based on observations
- y. '
daring the performances and previous test'results, no design basis problems L are anticipatied that will impact the final SORC test. results approval.
~ }
Based'on'the' SAT review of theEexisting' test =results,.the SAT'~can conclude ,'9 3
?
that all tests have been properly performed and therresults demonstrate that . 1 } } -Seab ook Station-is operating in accordance with its design basis. The SAT ; ( us - ' will continue to review the remaining test results requiring SORC approval 3 to assure that no changes or exceptions are gen 9 rated to alter this analysis t
; -and conclusions.
m, -3 t {s ..' L h ' L" g Conclusion t L u , The conduct: and results-from the Power Ascension Test Program satisfy the [
~
.i
, commitments of the FSAR and Regulatory Guide 1.68, Revision 2, " Initial Test l
Programs' For Water Cooled- Nuclear Power Plants" . Seabrook Station-has been l t , proven to be able to operate at full power in a safe and reliable manner. ( , 14 i q r i-L L m
p-a; ,
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, .}
x, ' ' L2;2?ETALEkTIONOFQFEBATICIts M
',S
- y. a '
1 O W Submary
, ci y [
l p.
. ~
- j The SAT evaluated thel performance of the: Operations; group in carrying'out s :i - . thel PoweriAscension- Test Program (PATP).- Thenevaluation'was conducted by/
I( i- , , . p i 'f
; observingns hif t - activities. , reviewing' operator - log sheets and J discussing: !
Y
, , g
'I lr < -^
' items'with operators. =The chronology of events.for-the PATP is provided in Appendix- A. . The '. topics - covered in this section include reactor l trips,' an j 2
e langineered safety: feature (ESP) actuation,' control room operatore. auxiliary M operators,- at: technical.' specification violation. involving containment , Ill'ghting breaker, emergency: operating proceduresb and individual' safety.: - j 1
'The conclusion of the evaluation is that the; Operations group conducted'the. , t
-t f; ,
- PATP in a' safe...consetvative, deliberate, and almost_ flawless manner and has> '!
demonstrated . the ability to . conduct . Comercial. operations. safely 'andi creliably. $
- b 1 :
g, 4 fuggggy 'Of - Find 1 Ens f n. I': ' I l. 04--- : Testing SAT Recomendations (ST) , O -- Comercial Operation SAT Reconsnendations (ST) 0; - . : Long Term SAT Recomendations _(LT)
]
..' Reactor' Trios l
r (
+ . There : vere two automatic reactor trips at Seabrook Station during power.
~
L 1 ascension testing. On June 20, 1990 reactor power level was approximately 302 and slowly increasing to 75% when a turbine trip was activated.by ground L fault' relay 64-TG-l'. Since the reactor power level was above the f P-9
. , . reactor = protection setpoint, the turbine trip signal initiated a reactor trip.. Relay 64-TG-1 protects the last 5% ' of windings if grounds occur. ;
h
, This protection was installed as part . of the original design by 'Public lg: - . Service of New Hampshire (PSNH). The turbine manufacturer (G.E.) stated the EL protection is not part of their design and that testing could continue h"~ without the trip protection of relsy 64 TG-1. The alarm function was L 15 e,,
y .l
- i j!' r$aitad,Lbut'the trip function was disabicd to ollow further ovoluation cf '
d., _ the ground' fault relay during power ascension testing. Ground fault relay N 64-TG-1-is also discussed in Section 2.9.
}? .f i On July 5, 1990, a second reactor trip occurred with the plant at 75! power y ove to a signal generated. by low EHC pressure on Channels. I. (D5209) and III (D5213). The'setpoint of 500 psig was not reached but vibration caused c
precoure switches: I and III to. momentarily _ and .puriously actuate, c satisfying e _2/3 logic required for a reactor trip which resulted -in an immediate turbine trip. Prior to the trip thereL were no noticeable EHC [ ' pressure fluctuations. The 1000 psig low pressure alarm did not'come on nor did the standby pump start at the 1300 psig setpoint. Prior to the reactor trip, there were periodic 20 to - 40 MWe load swings and sporadic feedwater flow oscillations. The e'ent a, valuation' revealed that the switches were mounted on the actuator for , the turbine _. stop valves which is a high vibration area. Mounting the switches in a high vibration area is counter to the vendor's (ASCO) I. recommendation. The - switches are now mounted on a building' support beam with - stiff ened support to alleviate the vibration problem. A _21ex tube connection was also used instead of the rigid sensing line tubing used before.- Af ter these changes were incorporated, testing continued. The l pressure switches are also discussed in Section 2.9.
, In both trips, plant operators took appropriate and timely action to ensure the plant remained in a safe condition by entering emergency operations I procedures E-0, Reactor Trip Safety Injection, and ES-0.1, Reactor Trip Response, and implementing the requirements of OS1000.11, Post Trip to Hot Standby.
h ineered Safety Feature (ESP) Actuation One ESF actuation occurred during the PATP. Due to Operator error, the
. alternate power supply (maintenance supply) to inverter IE, which supplies Power Panel 1E (PP1E) was momentarily disconnected resulting in . the de-
- energfsation of Train 'A' radiation monitors. This includtd the containment
=
, _on-line purge (COP) radiation monitorm which f ailed the as: cisted bisi4bles 16 l
s
La the'stfo (high) stato cad fed into the ssPs as. o high rediation signal, l- s This. initiated a Train' 'A' - (CVI) containment ventilation isolation which resulted in the - closure of' COP-V1 and V4 and the shutdown of CDP Fan 73 F (Com tinment Purge Supply Fan). The Control Room received computer alarms indicating Inventer 15 power supply had . changed to the maintenance supply. An Auxiliary Operator was dispatched to investigate the. problem. Poor et usunications by the Auxiliary I Operator with?the Control ' Room was the. ma n' contributor to the ESF
' actuation.; The maintenance supply breaker was opened assuming it was tho'
' normal supply without checking with the Control Room. An RES was issued.
requesting Engineering to more accurately identify normal -'and alternate
; feeds' and develop comprehensive one line drawings which will include AC and DC feeds to vital and non-vital invertors and power panels. Training has been requested to review the incident as part of Auxiliary operator.
continuing training. Control Room Ooerators The performance of three Control Room Cperator Crews was observed and evaluated during the performance of ST-35, Large Load Reduction, including preparation and plant stabilization following the test on July 12, 19908 ST. 38 Unit Trip from 1002 Power, and ST-22 Natural Circulation, on July 29, 1990; and ST-39, Loss of Offsite Power, on August 1, 1990. The areas
- evaluated included communications, use of the (witronics system, watch
- Station' activities, managing control room. access, and control room r housekeeping.
Communications between the Unit Shift Supervisor (USB) and Board operators were clear, concise, and of suf ficient amplitude. Board operators clearly
) ; understood directions from the USS and each operator reported task
= completion to the USS.
The gaitronics system was used effectively to announce the beginning of a test, assess restrictions during tests, provide directions to Auxiliary
~
. Operators, and announce countdowns for the actual start of each test.
Gaitronics announcements were clear and professional. 17 -
w 's >
+
l' }w' ,
'The: Control Room area was i free of unauthorised ~readingsmaterial.
, . Surveillance testing and administrative activities- were postponed until' Operator ' aids such as graphs and prints
~
af ter each test was completed. were authorized = by the operations Manager. , Radio casununications were kept' j to a minimum. The Work Control Supervisor controlled personnel access to L the Control . Room to prevent any distractions of the Board opera', ors. Only ! those personnel . necessary for the . test were . permitted accost,. Only the Shif t operators and Test . personnel performing the test were 4110wed within the confines of the . control board area. The Control Rocra was .. neat and cleant only procedures and test related documents were allowsd on desks - I l Pre-shif t and pre-test briefings were held in the Technical Support Center
, prior to relieving the~ watch. These briefings were adequa.te to ensure on- !
coming L shif t - personnel understood current plant conditions and up-coming 0 testing.- JJdit).onal operators were assigned when needed for feedwater f control. l i
~ The' SAT observed breaker alignment prior to the performance .of ST-39, the l phasing of the diemel generator, and restoration from ST-22. There were no deficiencies noted with Operator performance during these observations.
o ! t' a Auxiliary Onerators j ,; i The SAT accompanied two Auxiliary Operators (AO) on their rounds to evaluate Li i A0 performance. On these rounds, one A0 was observed verifying the boron
- i thermal regeneration system (BTRS) tagout prior to adding nitrogen cover gas L to the BTRS surge tank. Both Aos were knowledgeable on the location of .
each reading on their logs, entered the current value on the log sheets and verified that the readings were within the specified tolerances. Each A0' l reviewed his logs for trends and out of specification conditions. Each A0 conducted a general inspection of plant equipment during his tour, which ( f included checking for leaks or unusual conditions in the tank farm, checking for leaks in the steam generator blowdown recirculation pump area, fan room i and heating-coils: touring the main feed water pump room, lube oil storage
! area, and turbine hall sumps and checking for deficiency tags. When a leak was found on 1-SW-V20, a Work Request was initiated by the AO. !
18 l. l' : 1: i 1, ) '
y
^l
,l ' '
)
9 One A0 was observed L entering the Radiologically Controlled Area (RCA). The , l A0 reviewed : the radiation work ; permit and. the -. radiation area survey maps.
-and noted -the radiation ! areas high : adiation areas , , and contamination j areas. The A0 properly signed in. frisked upon leaving, an! signed out. l I Each L of . the observed -AQs receivel and' provided . adequate shif t turnovers.
These turnovers were conducted on shif t work stations. - Both A0s reviewed ") the Operations Standing Orders and Night. Orders prior to the turnover and checked on " problems prior to attending the shif t briefings in the Control l Room. The shift briefings were informative and professional. Both Aos were ) attentive and involved in the exchange of information. During the shift, both Aos were observed positioning valves. These included l j closing CS-V127, stroking CS-V126, and restricting spill valve 40148. No ;
- deficiencies were noted.
,{
t lg .Overall the A0s performed a thorough plant inspection and took accurate: -
- e , readings' .
Communications with the control Room were concise and l professional.' ' Aos - investigated conditions that were unusual, such as - a lower discharge pressure in SCCW Pump A- which was resolved as the j
, characteristic of the pump. !
Containment Linhtina Breaker [ I.
~
On July 21, 1990 Operators discovered that breakers on unit substations US-
)> 11. and US-23, which feed power panel 7A and 7B in the containment, were closed and had been closed for about 78 hours as determined from the Unit Journal. ' Technical Specification 3.8.4.1 states that these breakers shall be locked open in Modes 1, 2 and 3. The exception is they may be closed for !
up to' 72 hours for work inside the containment. Personnel error was l l attributed to this incident for the shift crew's failure to enter the LCO j Li :'
' condition in the Action Statement Status Log and failure to list the ,
condition on the control Room Relief Checklist. I . Procedures for containment entry (ON1090.04), operating the Containment
., Hatches (OS1058.01) and Action Statement Tracing (OP 10.6) have bein revised 19 lt .
e o t '!s t 1 to. prev =t' a recurrcnca cf the ales:d LCO tracking. In codition, the W;rk . Control Supervisor has been assigned 'the responsibility for controlling all. I containment entries regardless'of department. I LI F
- raenev Onoratinn Procedures (EOPs) f a
l The SAT evaluated activities of the Emergency Operating Procedure (EOP) i verification task team. The team was formed to review all E0Ps and resolve ' LI1 l- all procedure change requests, ensure E0Ps are technically correct and , validated, and complete human factor enhancements. ! L ! Each procedure change request was first resolved by the BOP coordinator with the concurrence of the Operations Manager who makes the final decision. The changes were then incorporated into a rough draf t procedure by the E0P l coordinator for review by the verification team for technical accuracy, format and consistency. To accomplish this, the Verification Team used the g Plant Difference Document which compares Seabrook Station to the .
'E Westinghouse . reference plant. The reference plant document was used to !
write the ERGS (Emergency Response Guidelines). The deviation document was ' L ( also used which documents the difference between the existing plant and the l l reference plant. For example, the reference plant has- en APW system consisting of two electric driven and one steam driven emergency feedwater pumps while the Seabrook Station has one electric driven and one steam ; driven EFW pumps. The (SUFP) startup feed pump can be used as an EFW pump ' with Operator actions. The Verification Team reviews resulted in one technical discrepancy in ECA-0-2 I ensured Step 4, where adequate PCCW flow was not by valve line up which could have resulted in possible damage to
. components such as the charging and RHR pumps. There were eighteen other !
I discrepanciec where the writers guide did not match the procedure. The E0P coordinator has resolved all the discrepancies. The SAT reviewed and , concurred in the resolutions. ; Procedure validation was determined by the E0P coordinator or his designee s by filling out a Change Validation Determination Sheet OP 9.7A. If any of the first three questions were answered yes, then a Simulator validation or plant walkdoen was required. If either of the next two questions were answered yes * *.se n a table top discussion or a walkdown validation l - 20 I s, - - - , , ,
r, . (-
, y W perf rmed. If . the answers to. cll the qu3stione wero *ns', thes as:
validation was required.~ The SAT reviewed the Determination Sheet and found. , no discrepancies.-
, Simulator validation was accomplished by a team of three to four observers who observed a crew performing the E0P to be validated.. Following the simulator run, the observers and crew reviewed, discussed and comunented on the procedure. This method allowed input from the Operators which could result in further changes.
The' E0P coordinator resolved over five hundred procedure change requests, B. many of which resulted in required validations. From these validations, 101 discrepaacles were identified resulting in 31 procedure changes. The SAT reviewed a sample of twenty ' procedure changes and found no discrepancies with the procedure change review process. Human factors enhancements require specific design i:hange approval by the Station Modification Resource Conunittee ( Smit".) . Yhis- provides that IL
' instrumentation and equipment referenced in the E0Ps has been verified or repaired' to ensure accessibility, adequate lighting, and correct and clearly visim labels. The enhancements ensure the Operator's ability to easily observe instrumentation and manipulate equipment as directed by the E0P.
Training - on the revised E0Ps was also conducted. Normal classroom and simulator training was used. The SAT observed one classroom and simulator session. Good student involvement and professional presentation of training I. material was noted. Industrial Safety ) The SAT evaluated the adequacy of plant housekeeping practices during the observation of Secondary Auxiliary Operator rounds. The following problems were noted.
-- 'On the 75' level of the Turbine Building, on the left side of 'C' Low Pressure Turbine, a bearing monitor was installed which appeared to 21 I
~-- - -
, , , , 7 I4 E+ j have been . to:d fcr power ecoc:nsion" testing cad ein! cow be removed, j 7 r. Also, each MSR reheater drain tank sight glass - area has water leaking ,
on'the floor. .l l
- , on the 50 ' elevation under the. Main Generator some plastic bags were j i
inappropriately stored with what: appeared to be insulation = parts. . j valve hand - wheels were laying on the grating in the feedwater heater area.. Portable staging was unsecured, grating parts were adrift on top ;{ E ' ' " ' ' ' ' ' ' " ' ' ' ' ' ' ' ' " * " " * " ' ' ' ' " ' ' ' * * * * * " ' ' ' " ' " ' ' ' ' * " " ' '"' ' a bundle of rope was left in front of panel CP-66. l On the 40' elevation just below the generator breaker, oily rags and -
' other debris existed. !
.On the 21' level,~1 adders were not stored in a ledder storage area and- ,
tool boxes and portable work benches were not .in use and not stored. , The turbine hall sump area had buckets, ropes, wire cables and other i material adrift. I
'1 The SAT observed personnel in the Turbine Hall. All had hardhats, safety. l
- glasses, and bearing protection. Two types of hearing protection was used. 'f
~
foam plugs and covers. i Housekeeping in the. Primary Auxiliary Building (PAB) and parts of the Waste , 1 Process Building'was satisfactory. All areas'were clean and all material f was stored in : an orderly manner, with the exception of the boric acid ' j ; batching room. While the boric acid batching room . had no unauthorized, ! equipment or material, authorized material, such as funnels and an open l l barrel of boric acid, were stored in a haphazard manner instead of on l . shelves. The SAT observed ~all personnel in the PAB with proper hearing, eye 3 l- and hasd protection. The ladder entrance safety chains were in place. An > A) war, observed replacing all safety chains that he removed and the passed f throught No safety inadequacies were observed in this area. b The SAT provided a list of the discrepancies noted to Station management for correction. 22 t _ m. .
+
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- l# = ;
l The . SAT evaluated the effectiveness of the Maintenance Group in supporting . j safe .and reliable plant' operations.= The evaluation was conducted by l interviewint Maintenance. Instrumentation -and Control. Technical Support.
' Planning and Scheduling, and Engineering managers and supervisersI reviewing '
work control, design control, and planning and' acheduling documents: and ' observing field activitles. The ' maintenance performance objectives and I l criteria ' for Phase 2 ont the self assessment - of the Power Ascension Test i y Program (PATP) were addressed in the Fifty Percent Power Report., The topics discussed in this report include maintenance oriented plant problems that l
)
have-occurred since the Fifty Percent Report and areas of special interest. ] l Engineering and ' Technical Support oriented plant problems are discussed in l Section . 2. 9. The items included under plant problems in this section are L secondary system steam leaks, generator step-up unit (GSU) ' transformer-
. cooling fans, and the heater drain pipe elbow. The items included under-
'special interests are the loss of component status, snubber inspection, Work-l Request backlog, refueling outage planning, and the separate , NRC based maintenance evaluation.
- , .The conclusion'of the evaluation is that the Maintenance group continued to l
. support: safe and reliable plant operations during the demanding period of a !
test program. I ' Sc ry Of Findinas t
- 0..-- Testing SAT Recommendations (ST) 0 -- Conunercial Operation SAT Reconunendations (ST) t 1 -- Long Term SAT Reconenendations (LT) h t
Secondary System Leaks ig i us L Some secondary system steam and water leaks were observed when the moisture
; separator / reheater system was placed in service. The SAT evaluated the s 23
- . ~ . , . . . - , - - + - - . - - - -
- - ,- - - r -
p ~
, m; y .)
S ,
'i H cffictiveness L cf . the d:ficiency t g cad work contrti systems in c pt: ring
. . plant' discrepancies and the effectiveness of the Maintenance organisation in ,
l
.minimiaing: system' leaks. The evaluation was conducted by inspecting systems
'in the Turbine' Building and reviewing work control documents. }
The' SAT inspected secondary systems for leaks. The following minor valve b leaks, two to three drops per minute, were noted: 1-DM-V-0522, 1-ASC-V-' 1406, .1-ACS-V-1-373, 1-FW-V-0212, 1-HVD-V-0116, 1-HVD-V-0123, 1-HD-LG-4505 i valves 1A and 1B, 1-MD-LG-3203 valves 1A and 1B, 1-MD-LG-3213 valves 1A and ] 1B, and MD-LG-3218 valves 1A and 18. Stenst was ' noted coming from- .f extraction steam valves 1EX-V-0002 and 1EX-V-0005. Only one set of [ deficiency tage were posted. Those tage were on 1-MD-LG-3203 valves 1A and l.
'18. A set of Danger tags were posted on each of the following: .1-HD-LG- ,
4505, 1-MD-LG-3203, 1-MD-LG-3213, and 1-MD-LG-3218. Funnels had been ! installed to collect leakoff and direct the water to drains on the ground floor to avoid leakage - falling through gratings wetting equipment below. Puddling van also noted .under service water and circulating . water pipes ! ig caused by s o ting of the uninsulated pipes in hot and humid summer days.- 'f g ;
.t The SAT reviewed the package that was being prepared for a -48. hour outage
. t that was scheduled after a full power trip during the Power Ascension Test a LProgram (PATP). The package included coverage for six of the items noted
-{
l . during the SAT inspection. Those items were 1EX-V-002 und6.- Work Request ,
, (WR);i90 WOO 4043 1EX-V-0005 under WR 90W004044 1-HD-LG-45-5 under WR j
! : 90W0036633- and- 1-MD-LG-3203, 1-MD-LG-3123, and 1-MD-LG-3218' under WR 90W003835. The' remaining items noted by the SAT were in the work control *
. system under Priority 3 status with packages being prepared for j L
accomplishment during a subsequent outapo. The SAT concurred with the l . ' assigned work priorities and with the scheduling of packing adjustments needed to correct remaining SAT items for a latter and longer outage. 1
) 'Only one set of deficiency tags was noted on the twelve leaking items identified by the SAT. The SAT inspection, however, was conducted shortly
' after the deficiency tag system had been promulgated by Revision 17 of the 1
Maintenance Manual (SSMA), and backfitting the system to pre-existing , deficiencies had not been completed. Subsequent SAT inspections of the , Turbine Building noted that the deficiency tags on the 1-MD-LG-3203 valves ' had been removed following repairs during the 48-hour outage, tags had been 24 mj '
. . . . - - . . _ - _ - - - _ . _ - _ _ ~ _ _ _ _ _ - _ _ _ _ _ _ - _ _ - .-. - -
, -- - - - - .- -~
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. pined' on Cover:1: cf .the remaining SAT items, cad: new d3ficiencios wero 4
. being tagged in accordance with the deficiency tag system.-
.o Discussions with Station. and Engineering managers established that I Engineering is resolving ' action to be taken to eliminate service and
'; ,> circulating water pipe sweating during hot and humid ' susumer days. The I solution is expected to be a coating placed on the pipes during the first refueling outage. t r 5, The SAT concluded-that-leaking secondary system valves are being captured by i the work control. system are being tagged in the field to avoid duplicating i i work control paperwork, and are being repaired in a timely manner. ' i i l' Generator Sten-Un Unit (GSU) Transformer Coolina Fans I - An L inadvertent main turbine setback to 402 ' occurred when an electrical j s i- technician was performing week 1) switchyard rounds. The SAT evaluated the ) performance of the: Maintenance group in resolving the setback. This j evaluation was conducted by intarviewing Maintenance. Technical Support, j l3: Planning and Scheduling, Engineering, and Operations personnel and- by j
\
D - 4 reviewing work control documents. l
. ,. On July .2, 1990 a 402 main turbine setback occurred while an electrical ]
technician was conducting weekly Preventive Maintenance on the GSU L transformer cooling fan banks under Repetitive Task Sheet (RTS) LED-X-1A-
, E1000000. The electrician was shif ting GSU cooling fan banks to equalize I operating times, and when he turned off the operating bank before turning on Y
the standby bank, a turbine generator setback signal to 402 power was initiated. The setback was caused by both GSU cooling f an banks being de-i . energized. j,- This condition had been identified about a month earlier, and a Maintenance l: - Engineering Analyst added a handwritten note on one of the RTSs issued to the field. This note cautioned against operating the cooling fan control selector switches because a setback would occur. A RTS that was used a week after the naster RTS was changed, had been issued to the field prior to the 1 I change and had not been recalled for revision. Revision 3 to procedure a 25 I . .
i I}
' Is.
i SM7.1, Repetitive Tasks. now requirss that RTSs be r:c011cd cadFrevis:d when' ichanges'are approved. , The- previous setback also resulted in caution tage' bering placed on control ' i selector switches GSU-HBl. GSU HB2, and GSU-HB3 under Tagging Order ' 90-1450 I y to warn that' changing the switches would result- in a setback. Unfortunately,,: bank select switches 1-ED-X-1A,1-ED-X-18. . and 1-ED X-10 can l
']
also cause a setback and were not. Caution tagged (Tagging Order 90-1633)' ? i until after the second-setback. I i t Design changes are being prepared to resolve the inability to shift cooling . ,- fans without a setback. Hinor Modification (MiOD) 90-622 adds a time delay ' to the setback actuation signal so that five seconds are available to shift ; cooling fan banks before a ' setback can occur. Also, Design Coordinator i-Report (DCR) _ 90-025 adds an annunciator in the Control Room to show the cause of turbine setbacks to Operators. i I , The ' SAT concluded = that the actions proposed af ter the first setback were adequate to prevent-a' recurrence of the problem. However, implementation of l
; those actions lacked a coordinator and overseer to ensure that actions were- .,
completed before repeating the activity that caused the problem. ,The SAT. , recommends that Station Management review the Document Control Revision-
- processes for assigned responsibilities to avert a recurrent of a similar :
event.. (4054)(LT) t
, L Hester Drain Pine Elbow i lI-A heater drain elbow was replaced based on Non-Destructive Examination (NDE) 'I I information that indicated a significant reduction in wall thickness. ,
Examination of wall thickness after the elbow was removed showed no I thinning. The . SAT evaluated the approach taken to resolve the difference between the NDE and actual wall thicknees. The evaluation w&s conducted by , interviewing 'NDE, Program Support Engineers, the System Support Engineer, and Maintenance personnel, and by reviewing work control documents.
,4 :
A' leak from a crack in a reducer in heater drain system piping led to ! additional checks for pipe erosion. The wall thickness of two elbows 26
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+
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'A g downstream cf L spill valve RD-LV-4583 woro ch:ck;d und:r W rk Requ:st (WR) j 90W3624 with a UTM 110 ' Ultrasonic Digital Thickness Meter using a 5210NDT. , l i_ probe. One . of the elbowr had a 14' diameter. The wall thicknesses of ,
that elbow were found to be within specification. The other elbow had a 6' . q
', diameter. The wall ' thicknesses of the 6' elbow were measured as thin as
.003' in the _ outer bend of the elbow. The nominal thickness of the pipe is
{
.432'. The measurements were made by an individual _ in . the ' Maintenance
-l Department who was qualified to perform thickness measurements. The UTM 110 !
meter was properly calibrated. The measurements were taken in accordance j with Procedure NHY-VE-1. - Water, and possible a two phase mixture of water , ] flowed through the elbow during the measurements. i and steam, _ The 1 temperature of the elbow was 165'F using a pyrometer. n The' individual who performed the measurements rechecked the instrument using _ calibration blocks- and then repeated the measurements obtaining wall ] thicknesses of- .003' as before. The In-Service Inspection (ISI)' Lead
~
Engineer reviewed the data and decided to confirm the thinness of the elbow. with' infrared thermography using an Inframetrics 525 device. Only one UTH I
=110 ultrasonic thickness meter and SZ10NDT-1 probe was available on site. A '
DH2 ultrasonic thickness meter was available, but was not used because the ; UTM 110:was considered more accurate. The ultrasonic finw detector was not j available. The-Inframetrics 525 device indicated that the elbow could be j thinner where the .003' thickness was measured than at other locations.' n , Based en the UTH 110 and Inframetrics 525 device information, Management
- decided to replace the elbow. [
.,1 When the elbow was removed, no crosion was evident. Measurements of the l i
elbow in the shop with the UTH 110 and 5210NTD-1 probe were within 'f specification. When the elbow was heated to 165'F, the measurements ; remained within specification. Wall thickness measurements taken when a new 4
. .' elbow was installed in the heater drain system were within specification
- when no flow existed in the elbow but were as thin as .006' when flow existed. Wall thicknesses taken by the DH2 meter were consistently above ;
.5", when the meter could be read, regardless of flow conditions.
A check with the manufacturer, the specification for the UTM 110 meter, and the operator's manual established that the 5Z10NDT-1 probe should only be l used to a maximum temperature of 122*F. the minimum wall thickness that can 27 I. L___.'_._. 1 '.__ _ _ __ _ _ . . _ . _ -- -
- c. ,
x -1 g # i be me:s red with ths $110NDT-1' pecbe is ' 1208. . - cad ' t:sts have . not been . .1 conducted to determine the offoct of' vibrations:on.the meter. The meter wes ., j used in:a situation which was outside the capabilition of the meter. The Nuclear Quality ! Group conducted a detailed review of this event and j ' issued a; report, SSP #900408 on July 24, 1990. The SAT concurs with.the l conclusions of the Nuclear Quality Group (NQG)l report concerning the false . _ y , ultrasonic thickness indications. . Flow in pipes should be stopped before - -J using the UTM 110 or . similar meter for pipe - wall thickness readings, an-
!I # '
ultrasonic flaw detection scope. should be procured to . support accurately. measuring pipe wall thic hesses, a high temperature probe (5210NDT-7) should: l be procured for the UTL 110 meter, limiting . conditions for measuring and I i test equipment should be identified and included in procedures, and NDE data that is questionable should be verified by an independent method. .i Loss-of Ca=nonent status E5; Instrument process control valves downstream of root valves on instrument .; sensing lines were not all identified and included in procedures or valve' ) lineups. The SAT evaluated the plan that was developed to ensure that' all
. instrument process control valves are identified and are under configuration -;
9 j , ' control . - The evaluation was conducted by interviewing Instrumentation and control (I&C) and Configuration Management personnel and by reviewing pertinent documents.- l l The_- turbine impulse transmitter was found inoperable during the Power ,j 4 Ascension Test Program (PATP) because a proceso isolation valve, which-L < 'should ' have been open, was closed. The valve is an additional isolation , valve that_ is only installed for instruments . mounted in manufactured n : instrument racks. This valve was not normally operated by IEC personnel and' l :was not addressed by I&C programs. I&C personnel isolated these instruments by' using only the valves located next to the instruments. Double person verification of the positions of the valves adjacent to the instruments
~
ensured configuration control. To verify that the additional isolation valves on rack mounted instruments ' were still in their correct positions, a valve lineup of all safety related f 28 E .._.u__ _m m _ _ _ _ _ _[_______.m____.__m_______ _ . _ _ _ _ _ __ _ _ _ . _ _ _ _ -_____t m . _ . _-. - m ,n,e-,,..<_w.,--,-, , , . , - , .-n'- e ,- ,..e-v- m, r-r
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, .l J rock CMated ' ins trum::nt s cutside cont:inment was conducted. This ;
I '
-verification'was later expanded to include all rack mounted instruments and i then to include . the instrument ~ valves associated with safety ~ related.- and
, important-to-safety, non-rack mounted instruments. These verifications !
identify 52.mispositioned valves out of 2730 valves checked on rack mounted instruments. Thirty-five- (35) of the mispositioned. valves were drain { valves, six (6) were for instruments not. installed, five -(5) were f
. domineralized ' water valves, and six (6) were instrument isolation valves.
+
t The six (6) ' isolation valves affected four instruments.- These four. I - instruments involved two which had not been placed in service, one with an { Open Work Request, .and one for the' waste -gas particulate filter. Fortys ! seven (47)'of 933 valves associated with non-rack mounted instruments.were l mispositioned. Forty-six-(46).of the 47 valves were drain or vent valves. -
, The r6maining valve was an equalizing valve with no control function ~.
l
- The program for strengthening the configuration control of instrument ;
, process control valves includes completing Instrument Loop Diagram (ILDs)', l wiring the second isolation valves on instrument racks in the open position, i'
' inserting all -instrument. valves used to verify the valve' lineups of i protection instruments covered by Technical Specifications on Repetitive d
. Task Sheets, inserting the requirement to verify the position of second. j
- isolation valves in- calibration procedures, and requiring instrument valve verification when accomplishing Work Requests and RTSs. .
! t
. The SAT -concluded that the program for controlling l instrument process valves lis sound and comprehensive.
1, LI: 1 I ~Egfuelinn Outane Plannint Refueling. outage planning is being conducted for the 1991 outage. The SAT evaluated that- effort by interviewing Planning, Scheduling, and Outage personnel and by reviewing pertinent documents. The~ SAT has reviewed the Production Planning and Scheduling Manual (NPPS)
- with respect to the general approach to be taken in the development of a Refueling Outage Schedule.
~
s The NPPS prescribes that the Planning, ) Scheduling and Outage (PST:0) group personnel develop a base refueling outage
-[ 29 l ...o
- 1, ,
i .
plan." . The plan is to consist' cf work required to r::fu21 th3 react:r plcs any other work' that is required to restart the unit to full power.- lgy
- The PS&O has initiated work with the Operations Department that will result in the inclusion of-the major plan evolutions (MPEs) required to perform the refueling and to restore the plant systems for the return to full power. To maintain. efficiency, . corrective' maintenance work that is ' identified as requiring performance during refueling is coded for computer storage and I retrieval for inclusion in the Refueling Outage Schedule. Preventive maintenance and design changes are also identified and coded for refueling
-outages.
The refueling outage duration is defined as the time from opening to closing the main generator breaker. The first and second operating cycle durations are based on.12 month cores. Beyond the second cycle, the operation cycle is based on '18 months. Currently, the Seabrook Station first refueling
' outage is planned to start on July 27, 1991.
The generic refueling work involves the following
-- OPS Surveillances ISI Inspections
-- Type B & C Leak Rate Testing
-- Eddy Current Program
-- Mechanical PMs
-- Electrical PMs
-- Corrective Maintenance Backlog
-- Vendor Support
= ' -- HP Support /Decon
-- Sludge Lance 3 k -- Millstone RCP Mockup Due to the Power Ascension Test Program (PATP) and past forced outages, a minimal effort to date has been put toward the planning and scheduling of the first refueling outage. The PS&O management is aware that the Refueling Outage Schedule is not in the most desirable status of development and needs resources allocated to assure an effective and efficient outage. The
.I generic elements require program sequencing and fragnet development. The
< ' integration of specific corrective maintenance WRs. design changes i 30 I ..
5, Y + '
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- the first r0 fueling c ttge, 'The PS&O management is awaro that the Ref:eling.
s so e
h j; Outage Schedule is not in the most desirable status of development and'needs
~
o >1 1 ( resources allocated " tot assure an effective and efficient outage. N[ generic elements requiro program sequencing _ and fragnet development. .The' The-i
, intsgration of specific corrective maintenance WRs., design changes' (DCRs/WODs), and vendor support in the base schedule fragnet is required to
' establish the specific Refueling Outage Schedule. The first refueling t-schedule must .be ready to be implemented 90 days. before the July 27. 1991 .
start date. h.
. I'o ,
b . The refueling outage requires the involvement of Maintenance. Operations, f
-Technical- Support, Health Physics, Chemistry. Security, Engineering, and' ;
Production Services. The pre-planning meeting with representatives from i these groups'has not occurred as of' August 1990. The delay has been due to . the, heavy _ demands on all groups to assure a successful PATP. f l The planning and scheduling function has been effective: .in providing
- _ schedules' that have resulted in efficient use of outage time,.e.g. the l
forced outage for turbine repair and the 49-hour outage following ~ $7.$8, j unit trip _ from 1002 power. PS&O management, however, is concerned about. ! i u developing a Refueling Outage Schedule with only three available people to _
. schedule and coordinate the input from participating groups.
- m The SAT concluded that the planning and scheduling function throughout the q PATP ,. contributed to an effective and efficient test program and work -
+ sequencing for - forced and planned outages. An augmentation to the PS&O staff will be required to support the Refueling Outage Schedule. l
-york Reauest Backlon
-l The SAT evaluated the backlog of Work Requests (WRs) by reviewing data _from '
the NHY Work Request, RTS and DCR Matrixes Report.
~
L. i The total WR backlog tracked as tasks considered to be representative of the Station goal of 750 includes all WRs except the following: (As of August l 3 20, 1990) t
$l 1
[. >-.i,,4 j "; - -
"'i
,j' '
f
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I[.4' '~ TNon-Pla.rt'Naintena c3 -- 133 I h' , : Awaiting Plant conditions' -- '51~ j s Outage Work ~- Refsel -- 42 1 - Planned -- 117 . I - Forced -- 22 Future Design Enhancement -- 352 f (Not at the DCR/ MOD to SORC Level) , W Paperwork Closeout. -- M l TOTAL -- 1022 1 j i c The backlog as of August 20, 1990 was $19 or 69 above-the target of 750 WRs. f The VRs ; within the 819 backlog is categorised below in accordance with[ '( assigned priorities. I
'I Priority.1. E 0 1
I Priority-2 -- 131 {, Priority 3~ -- 591 {
. Priority 4
-- 97 -)
Total ~ -- 819 ! - Priority E are maintenance actions that are required to prevent or mitigate ' the, consequences of an accident, prevent the release of radioactive material ~ j ,a
=
to - the . environment, protect human . life and/or property, or restore the- ; ability of the Operator to obtain information from the plant computer. (0 ; ( WL Total) 'j t Priority 1 actions apply to repairs, replacements, or modifications required l to restore- a . system or function to operable status where the system is ~ requiredtto maintain permanent safety, reduce rask of radiation exposure to i G the public, return to full power, satisfy Technical Specification limiting i conditions of operation, return the backup plant computer to operation and f
-return accident . monitoring equipment instrumentation (AMI) to operation.
(0. total) , L
- Priority 2 actions apply to repairs, replacements or n.odifications required
; to ' correct an existing condition, which if allowed to persist, has high probability of impacting the plant's ability to maintain a system or '
32
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7, , , I ',, ~! I 'i function ' in cperablo status,' Pritrity 2 cleo lapplice . to cny personnel j , protection- and/or safetyL eoncerns not covered by Priority 1. .(131 Total. , ( 13 Safety Related). , Priority 3 actions apply to those work items which are considered to be' of medium .. priority. Priority 3 applies to those repairs,- replacements or l modifications which may be performed as manpower or other scheduled
. activities allow. . (591 Total - 72 Safety Related) l I ' ,
i Priority 4 actions apply to those work items which are considered to be of -., l low priority. Priority 4 applies to those- repairs, replacements, or ; modifications which may be performed as fill-work. (97 Total - 6 Safety ! Related) l Priority E WRs are very infrequent, but once designated Priority E, the work ' is assigned and continued to completion. Also, Priority 1 work is ' given , inunediate attention.- The backlog for Priority E and Priority 1 are j maintained low. The Priority 2 WRs contained'13 safety related tasks out of
't ~131 total. The largest backlog consists of 591 Priority 3 WRe' of which 72 are-safety related. The total safety related components of the 419 WRs as of August 20, 1990 was 91 or about 112.
Work required to be performed at the Station is constantly revjewed. -Esch' WR, as a minimum, involves the System Engineer, Operation Shif t Supe visor I or Superintendent , Department Planner, Quality Control personnel and ; Maintenance Supervision. 'The overall work effort is discussed daily at the Plan-Of-The-Day (POD) meeting. Items of importance are given' special attention at the POD including work that is delayed due to material needs. ' j Further, during startup testing. WRs were given added attention in , accordance with test procedure pre-requisites. Startup Test personnel, ;
,. along with the Shif t Superintendent, were required to verify that no open WRs related to the systems or components identified on the System Readiness
. List would affect the performance of the test.
I~ The actual number of open WRs has been manageable and controllable
; resulting in no direct delays of testing nor operation of the plant in an '
unsafe condit!on. The WR backlog has been addressed in the PATP Fif ty
- i. Percent Report resulting in a suaunary of outstanding WRs by age and in I 33 l-LI .
Wm i a o j_ 27
. o Recommendation 4013. Th3 r0 commendation is t3 ' initiate o NNY ' policy t2
- n
~
- require an ' annual review of open WRs for their impact on plant operability' ,
and justification for~ retaining long term outstanding-WRs. , q L m ..- . Snubber Insnection , An in-service inspection (ISI) of . enubbers was conducted during a short outage scheduled during the PATP. The SAT - evaluated the inspection by I>^ interviewing cognisont Program Support, Maintenance, and Planning and l 1 Scheduling personnel and by reviewing pertinent documentation. l i The ' SAT obosrved the planning to perform the .ISI snubber inspection -as ' O required by Technical Specification Sections 3/4.7.7 and 4.0.5.
' Operability of each snubber is required to be - demonstrated by a specified
,1 ISI . program. , Discussions with the Program Support ISI Engineer indicated that there was one snubber (IRH 185.RM-18) that failed operability ,
verification due to a low level of hydraulic fluid. This information ~ is { also docuarted in Repetitive Task Sheet (RTS) 90RM1867601 of July 19. 1990. .
'l Correct?"- m tenance on the snubber was performed under WR 90W004066.
f This WL . a P.'iority 1 and has been completed. The reconditioned snubber was testad and reinstalled. ; i An RES'was written requesting that an analysis be performed on the effect of I the failed snubber on the associated system. Preliminary results of the analysis indicates that the system was not detrimentally affected. , The inspection effo r t included a cooperative effort between Maintenance. Quality Control, Engineering, Utilities, Health Physics, Safety, Operations. l
. Planning and Scheduling and Technical Support. The effort was effective and ;
efficient. The visual inspection of 221 snubbers located inside the containment was completed within the 48 hour shutdown window. The snubber
-ISI was required to be performed af ter 2 months, but within 12 months, of !
commencing power operation. ,
-The SAT discussed the inspection activity with the Program Support ISI Engineer and the QC ' Department Supervisor. Also, a review of Quality I Assurance surveillance Checklist / Reports (QASR) (QASR #90-00636, 90-00633, 34 i
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90-00629 cad 90-0065) was pertermed. The rceults Cf th? QASRs indicOte that
'. tho' enubbers were satisfactory and personnel were certified to perform the ,
visual inspections . L The SAT concluded that the use- of the shutdown period to accomplish a : proactive inspection on the snubbers was successful.- Completion of that-
~
L effort 'will result in more time available during future outages by having _ fulfilled the visual requirements for snubbers in the containment and will ; l ~ avoid interrupting full power operation - for the required ini.;ial checks.- l The effort was'well coordinated and accomplished in about's ten hour period.. I 1 i Maintenance Evaluation . l I The Self Assessment . Team (SAT) has performed a Maintenance Evaluation thr,t i compares the NHY Maintenance Process with the NRC Draft Regulatory Guide, DG ] 2001, Maintenance Program For Nuclear Power Plants, August 1989. The SAT l 1 has prepared an internal report that describes the comparativ results in a ] format that parallels the NRC Maintenance Inspection Guida .ce,- June 1988 { issue. I
-r The SAT evaluation was based on f
A comparison of the NHY Maintenance Program with NRC, INPO and other -
; industry criteria:
Direct observation of the implementation of the NHY -Maintenance j -- t Program: ' Comparisons of the NHY Maintenance Program with other nuclear utility . maintenance programs:
! ,~ :
A historical review of the evolution of the NHY Maintenance Programs and ! A review of current activities that indicate the direction and continued growth of the Maintenance Program. 35 ___-_-_.__._i_____._-______.._.,__m - , _ . . -, , - , _ . , = . . , ,,,,..r.,-- ,,.%,,_v- ,,.ev- ,, ..,-g
l-e D ' i Main'tenanca et 'S brooki has steadily cy:lv:d and improved Leines- its e
. 4. . l implementation in 1986. ' The primary ; thrusts - of this development . have ; l focused on. the actual implementation of work activities and their j I
U NHYi has carefully monitored the development 'of 'more management. i c , sophisticated ~ efforts, such as Reliability Centered Maintenance' (RCH). and 3-. developed these programs on a time and resource available basis.
)
Maintenance - management control mechanisms L are generslly strong te adequate
.for the administration, tracking, planning and scheduling of work requests and repetitive maintenance activities. Management control tools exist . to ,
monitor the efficiency of maintenance efforts and administrative tracking of'
; items such as procedures, commitments and industry experience. Additional-1 management emphasis is necessary to- obtain full value - from these existing l controls. Maintenance management should also consider developing trending . programs for fundamental activities critical to their success. The trending program ~ scope should include items such as procedures expired.. RTSs deferred, average work backlog by disciplines and Station Information~ Report '
(SIR) and Operational Information Reprot (OIR) root' causes affecting { maintenance. These trending efforts should directly correlate 'with ! t maintenance specific goals and measure progress towards achieving those ! goals. NHY currently uses numerous information systems to support the ! maintenance functions. The Production Application Plan is focused on integrating many of these stand-alone systems over the next five years. Additional management attention is a pre-requisite to support improvements i in ,the : Work Control System. As an example, the SAT recommended an integrated planning, scheduling and work tracking system in the 50Z PATP ! jg us , Report. These integrated systems, when developed and implemented, should f significantly improve the efficiency and effectiveness of maintenance activities.- i The strengths of the NHY maintenance function lie in the planning, scheduling and subsequent implementation of work requests and repetitive l' tasks. The numerous groups that contr'.bute to these processes have self r '
' identified enhancements that will further strengthen these areas. These combined efforts result in high quality field maintenance efforts. The results of QC inspections and successful post-maintenance testing further support this conclusion. The SAT evaluation has indicated that improvements n to the Preventive Maintenance Programs by incorporating Predictive and 36 I.
\ .
1 , e
, j s
' Reli;bility Centered Maintenanco cinc: pts will' improve fccus, defiro sc:pe j
-and increase efficiency and effectiveness.- , .
..t t
i This' evaluation indicates that' overall. NHY has' developed or is currently . l addressing the - prograsenatic aspects of the NRC maintenance . elements. The
. progransnatic - aspects that . requite additional depth or initial development 'l include i
-- Rework i 6
-- Maintenance History Collection, Analysis and Application !
-- Root Cause Analysis !
f
-- Plant Aging )
-?
-- Plant Life Extension !
I -- A ' Preventive Maintenance Program that incorporates Predictive Maintenance ~ based upon performance monitoring and Reliability Centered - ,
, Maintenance.
ll --- Goals and Performance Measurement 1 !' ) f I e I i l 1- ,
~
t 1
.I I
F F b 37
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i
, T 2.4i ETA 12i& TIM OF EADIATION FILOTECTION l p0-' < ( .{
'l The ' SAT evaluated the effectiveness of the Radiation Protection Program in minimizing radiation exposure to personnel at the Station. The evaluation-
)
was conducted by interviewing Health _ Physics personnel, reviewing Health - Phys '.cs records , and observing the . activities of Health Physics and other ; I Station personnel in the field. The topics, covered by this section include'
. 7nield Surveys, Dose Control. Radiological Occurrence Reports (RORs). and l
l Radiological Trends.
,I The conclusion of the evaluation-is that the Radiation Protection Program is f effective in minimizing radiation exposure. ,
?
S" ry Of Findinas . ; 4 0 -- Testing SAT Recommendations (ST) 0 -- Conunercial Operation SAT Recommendation;. (ST) r
. 1 -- Long Term SAT Recommendations (LT) ;
l,
- h.
, [
t 0 Shield Surveys t . .The full power shield' survey was conducted under Startup Test. ST-41, i 1 t l Radiation Survey,- to verify that neutron and gamma dose rate levels meet ;
~
. design criteria and identify high radiation areas. The evaluation of the shield survey was conducted by observing field activities and reviewing test :
results. l- . h The full power shielding survey was performed at all of the survey points , L outside of containment. The survey results fell below the design dose rate E criteria for the survey areas. Four (4) survey points in the Containment [ Building exceeded the design crite; % of 15 mrem /hr. Three (3) of these - points were in direct streaming paths from nearby areas and were not indicative of a shielding defect. At the other survey point, the shield
. wall was not as thick as surrounding areas because a pull space had been 38 1 b .- - - - - wv-w 9w- -- --
Ev, pr , Physics staff deterinined that . the only component in the vicinity of the , i " survey point which could require maintenance while the plant is operating is a feedwater ~ 1evel ~ transmitter. Therefore, this discreFancy was considered < l [ acceptable.- In s'. follow-up survey, the general area dose' rete at - the .
-' transmitter was 15 ' mR/hr gasuna.
Because the probability of maintenance .)
- while the plant-is operating for this transmitter;is. low, this discrepancy- l was also considered acceptable. The SAT concurs with those evaluations. .l i
Three (3) - discrepancies were also noted between survey meter readings and ' o - RDMS monitors. Greater than a 201 deviation was noted. The monitors . '! s involved includes
-- RM-6540 - Volume Control Tank
-- RM-6545 - Charging Pump A l
-- RM-6534 - Seal Table Room .l In each case,- the RDMS was reading higher than the survey meter. However .
the RDMS monitors were recording exposure rates at or below . minimum
. sensitivity and outside the calibration range. The SAT recommends that once- I, radiation levels in those areas are high enough -to fall .within the 3 calibration range of the RDHS monitors, a comparison with survey meter L readings should be performed to assure that the monitors are accurately.
reflecting radiation levels. (4055)(LT) + I The SAT c 'ncluded that this full power shielding survey indicates that in ' all but tne four instances discussed above the shield walls are providing I; shielding within the design criteria. In the four areas where design criteria is not met, the discrepancies have been adequately addressed. ,I r Dose' Control The St'ation maintains personnel exposure to radiation and radioactive .;
- materials within the limits specified by 10 CFR 20 Standards For Protection Against Radiation, and makes every effort to reduce exposures to levels as n
10w as reasonably achievable (ALARA). The evaluation of dose control was . conducted by interviewing the Health Physics (HP) Department Staff regarding l . 39 i I t _._____==_________--__._-...,-.;_.-_.._._-_..,,-.--_,,-.-
. . , , ,. - - ~ - - ~~ - ~
n if 1[ , ~ l 4
.r: diction lev:Is,;cirborne activity. c ntamination conditions,' cad'cxpos:ra I ,
; considerations during the SOE to,1001. power ascension testing period.- , 1 t
Supervisors have been provided personnel exposure reports, but,negligibli [ y exposyies have occurred due to the low radiation levels-in the plant. The. ' J . HP Sr.}$ervisors have not had to plan assignments around available exposure or te equalize dose but are prepared to use these dose control and ALARA l practices. ; j jI ~ Airborne and contamination activity levels remain very low throughout the
~
plant. The low levels of radioactivity have not required HP- to consider < additional- engineering controls to reduce airborne activity below the level , that requires . respiratory protection devices. . Controls are -in place to ' . track worker exposure to airborne radioactivity using maximum permissible - concentration hours (MPC-HRS) if required. These controls were reviewed and ;
.found adequate.
I HP continued to perform routine contamination agtveys. Step-off pads were used during the L period, and precautions taken to prevent the spread of
.l contamination were found to be adequate, I, d The SAT'concludad that the Dose Control Program continued to be' implemented' in accordance with af. proved procedures by the HP Department.. The radiation, ;
contamination,- and airborne radioactivity levels continued to remain low : throughout the 50%'to 1001 test phase of the PATP. I i Radiolonical' Occurrence ReDorts fRORs) s
;; RORs are unexpected events . involving radiation and radioactive materials that - require investigation and, when possible, corrective action. .Three RORs have been issued since the Fifty Percent Power Report. The evaluation -
of the three reports was conducted by reviewing the reports cnd discussing ' pertinent events with involved personnel. I I; e .
= .-w- , - w = e -+ , w ,-v, &
1 '! j Thi threi RORs fellows
']
--~. 90-0017 - Improper Issue Of TLD
--- 90-0018'- Operator Entering High Radiation Area'Without A Survey Meter I
-- 90-0019 -' Unexpected Gas Release From Floor Drain j A description of the three RORs'follows: j
-- ROR 90-0017 '!
.1 Involved the - Issuance of a TLD to an individual who had not -
provided a current- quarter / yearly estimate of radiation exposure. f By Procedure RP 5.1, Radiation Exposure Limits and Extension ; E Requests..if an individual does not have a documented . current quarter. estimate, the individual is limited to an exposure of 312 ? mrem /qtr. However, in this instance, the 312 mrom/qtr.' limit was not applied. . l 1 The inmediate corrective action taken in response to ROR- 90-0017 l Y was to restrict the individual's access to his dosimetry devices f until the appropriate documentation was obtained. 'Other- {
<arrective actions taken by the Health Physics, Department were to L _
?.
-set up examples of all variations of initial entry paperwork that .
< s could be used and retain those examples at the control point. I Long l term corrective actions include the revision of HD0958.18 to q provide clear, step-by-step instructions for issuing dosimetry
'g. ' devices. 3-C
'ROR 90-0010 =
Involved entry into a locked high radiation area by a member of the Operations Department without the use of a survey meter. Entry without a meter violated Technical Specification- 6.11 as reported by Licensee Event Report (LER) No. 90-017-00. 1 L i After notifying the involved individual of the violation, the Health Physics staff performed a survey of the domin alley on the 41
I m - y. 7 seven foot clevation cf thi Prismry Auxiliary Building' (PAB)., In the . area traversed by : the indd .idual, radiation levels did not '
, exceed 100 mR/hr. ' Actions taken to prevent recurrence includes r ,
Taping a . sign to' the . key card reader noting the access 4 requirements ' for the area. Health . Physics : is . evaluating a l permanent sign placement on key card readers and keyed doors
'for locked high radiation areas..
Health Physics will continue to evaluate the use of alarmin's ' I dosimetry.
.I
' Health Physics will require Operations personnel to contact
$ Health Physics when access to a locked high radiation area is -;
required for routive operations. - i t
-- 'ROR 90-0019' , ?
l L r
-t Involved an ' unexpected radioactive gas release from the floor :!
L - drain in the primary sample sink room. In an attempt tn vent the
, RCS system, the pressurizer ' was lined 'up to the primary ' drain "
, tank. A relief valve weeping to the floor drain caused the gas'to l escape into the sample sink room. l l: , 1
'l As'a result of the incident, the sample sink room was posted as an
' airborne area. and a Request for Engineering Services (RES) was i i' -
1 generated to resolve the radioactive gas release'from the floor j j , drain. Long term corrective actions will be issued after the PATP. l f The Nealth Physics Department conducted routine radiation surveys during the 4' j
- PATP. Survey data and personne1' exposure records were reviewed to summarise j f ,
the radiological statu's of the plant.' 4- , Dose ~ rates outside the Containment. Building and domin alley at full, power
-remained relatively low. . Contaminated areas were also relatively 'small, under 400 square feet. The total dose received by Station personnel was 0.2
~
'I rem at the end of June 1990. Thermoluminescent dosimeter. (TLDs) tre re.ad' quarterly, but based ' on pocket dosimeter readings.. the total doseLreceived ,
by the end of the PATP was less than 0.5 rom. The SAT' concluded that- the Health . Physics Department was adequately controlling radiological hazards in the Station. il; v l ! I , l I s
~
y' 5.;., " b 'm),1 - .
- Mz ; i ,
+ a_O - . . sapE$2.5l , ETALBATICII GF RADIQACTIVE R&STE
-~
l ?' r \E Summary iU 9 -:-. < ,
! 4 3 The SAT' evaluated the effectiveness of.the Station overall and the Utilities
;LDepartmenti in L particular: !in reducing:'the- amount? of: radioactive 7 waste:
i -
- generated; by . the? Station. . the evaluation nwas : conducted - by E interviewing :
Maintenance, ' Health , Physics, :and Utilities- Department supervisors -and o reviewing radioactive, waste records. - l The ' topics covered by Ithis section-I, inclu'de th station program and the Utilities Deparunent program, m ,
~
, ' The l conclus, a of the- evaluation:.is ,that Station practices haveL been-
[* , Leffective.in minimizing the generation of radioactive waste. i A-t
,[v '
Summarv Of Findinas [ i, j -- 0 -- . Testing SAT Reconunendations (ST) (W / T2 L0.-- 'CommercialjOperation SAT Recommendations (ST) 7 n: a 1 0---- Long Term SAT Reconsnendations (LT)
, s y
4, ,.s j , StationiProgram: Ws ?,)-
, !}i ,
cThe Station- has 'a program for stressingL the- importance ,of reducing , 9 radioactive waste to people who have access to th'e Radiologically Controlled-o 7 Area ~ (RCA) . The SAT evaluated the -- effectiveness ' of the program. .The-k , ,, evaluation wes' conducted' by interviewing ' Maintenance, Health Phyiics, and w-o
. Utilities Department' supervisors.
4 The need to minimize radioactive waste is emphasized to Station personnel ~in n> l 1 f - traising courses. such as General- Employee Training- (GET), and Radiation
.~ Worker Training. -- and during daily routines by managers and supervisors.
1-Health Physics. technicians monitor compliance with minimization practices at the -control point to the RCA. Personnel are instructed to remove ( ' unnecessary wrappings from spare parts before entering the RCA and to use tools prestaged in the RCA.
. 44 4
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1 'Iv . 1 L A1 Radicoctivoi Wastoi Minisiination ' Committoo is a part. cf the; Statica I"", - organization. .- l The ' connaittee . seeks practices that- will' reduce the amount' of 7 ,1 h radioactive , waste by monitoring 1 the ; ; operating ; : experiences = : of : ' other ~-
- utilities,; -analyzing; situations that genera te . waste, ? promoting suggestions:
4
, fromLplants personnel. - monitoring 1 field f activities.. ' improving' personnel-- > '1 awareness. and' establishing training. ' I% The' SAT' concluded' that::the~ Station ~has been offactive' in~ minimizing-radioactive waste. .Only 3.7 cubic meters of waste were generated during t.he-l PATP.=
Utilities Denartment'Pronram'
~ }
The Utilities Department processes radioactive' liquids and waste.to minimize-
$' the- amount of radioactive residne. The: SAT cvaluated the effectiveness of;
'the-program ~for reducing radioactive liquid and waste. -
The evaluation was
."( ,
conductedL by . interviewing Utilities = Departmente supervisors and reviewing:
? .-
radioactive waste' records.
.; .-A trailer for surveying waste has been located in the Fuel Storage Building.
o
!Of 28,542 pounds of waste processed through the ' trailer only 3,330 pounds,
-about 122, we e found to be radioactive.- The remaining 25,212 pounds were
.classifiey e_ -contaminated trash. Approximately - 610,000 -gallons? of-
[ liquid. we' te c ve been processed through the- Chem Nuclear. domineralization F
, skids.
The SA'r . * - Utilities ' Department processes are effective in reducini . cm radioactive liquids.and waste. I I e
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L2',6i NALIETICII 0F MMISTH 'I
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-The? SAT evaluated Lthe!' effectiveness of:Ithe: Chemistry- Department inL
~
- ' 1 a ' '
l l{ w , controlling 1 primary : and secondary chemistry.- The evaluation was' conducted? j j: j bycinterviewing Chemistry Department supervisory , personne1'andj reviewing; ? c ,. . m y ,.~_ . '; ;9 chemistry; data for tho' Power , Ascension Test Program.- (PATP) - . The topics;
+
>m
.t
>a -. . y covered by this section include' primary chemistry and secondary chemistry. [ ?. , ,
<_ r .
t l6 >
;The emd.ac Les of:Lthe evaluation 'is. that. the -Chemistryi Department A 3 ,
effectively zcontroll'ed primely on! cocondary 9hemistry.
- v. , , 1 ;,
Summarv Of ! Findings g! l (l [:3;.' *
,y+
3 !
/ - 0) - , Testing SAT Recommendations (ST) p 1 '
- 0, - - - Consnercial! Operation SAT Reconunendations (ST) s
- l. .
'i, ;(
p; 1 g '
, 0'-. Long Term SAT Reconunendations (LT)' q 5
. 3 1
t
$ ! f 5. '
., y ,
.Prhaary Chemistry i <y p 4 j'
, < Chemistry Department Lpersonnels were ' involved. with obtaining..and analyzing ,4
. water samples ini support of Startup ' Test ' Procedure
- ST-42, Water . Chemistry 1 Control., "
@i' S' ST-42L wa: used to implement .the collection , of : data used . to > L mi
- . demonstrate .the chemical and radio-chemical' control analysis function as - .
- .. .a
. > - described - in the- Station FSAR. . Controls. are required -to maintain' primary - ,
- e. . .
; i chemistry within the limits - of the Station Chemistry Contral Program. The .q
!(
Chemistry Department obtained test samples for analysis at the 302, 50%, 752 ;qj I , =and~1002 power level test plateaus. g l 3%
! .y J .
1
;, , , T [ ' Limits on' water chemistry are also specified in' Technical Specification (TS) o .s. 1 (J 3.4.8( Specific Activity.- The TS requires that reactor coolant be limited
- to less than or ' equal to one microcurie per gram Dose Equivalent I-131,. and l*
o1 !; i).
. f (; > :lless than or equal to 100/E microcurie per gram of gross radioactivity. The 0 {> 'TS is applicable to all modes-of operation. The data obtained indicated that the specific activity requirements for the primary system were met. ,
'46 ;
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.Furth:r,L Chemistry.; Department. phrsonnil ~ n2tcd a v:ry low; to l impere:ptiblo:
. I
". i ' ' '
[ change lin' iodine inventory following the l1002 reactor trip con' ucted' 2 d . under
;S7038. , The; low' to imperceptible: I-131 change 1 indicatesfexcellent - reactor
' t 3 ,' fuel claddingLintegrity. 1 ig, .
>H- Primary waterLeystem. chemistry was found to be within limits of the Station-
~
Chemistry Control Program and TS 3.4.8. The . SAT' concluded' that : the j Chemistry Department.had been effective in controlling primary chemistry.' j Secondary Chemistry ,
+
p q
- The serandary side water chemistry data fell outside of limits'.during PATP.
p ; The early indication of problems with secondary water chemistry were seen at: o [sy. the E 302 e test a plateau. Sulfate was detected = as ~ being above - acceptable Westinghouse guidelines : for ~ secondary water chemistry and was noted : as. an t f exception .in ' ST-42. The decision was 'made to ' continue operatien based onj o (b
]
;J w _+ . .
consultation with the- Westinghouse'; representative chemist. The' ionic-p impurities were the L results 'of chemical constituents 1 present . within; the -'
~
o .ty l e system ' from preservatives, paint and general ~. cleanliness of : the secondary 4 A
- L sysbem. ~The' conclusion, af ter a review of 'ST 42 information; was Lthat the' 4 impuritiesL.were .not being ' introduced but were representative of what was
- . already ' hidden-out" .within the system.g The only way to clean ; out the- q systems was to continue operating.
t 1 L/ u JChemistry Department personnel employed.three methods to assist in secondary- j
, < -water chemical' cleanup.- The methods used.were: J
'tt .
5 = if *
-Dumping.The.Hotwells "
l ;
- Haximizing Steam Generator Blowdown
- E Maximizing Use Of The Supplemental .Ecolochem Deionizers- And i"
1.
; Deoxigenators.
x - The secondary water chemistry was sampled at each power level plateau, i.e., . i 302, '502, 752 and 1002. The ST-42 chronology log information indicated a
' q continuing indication of chemical impurities that resulted in exceeding j .
Westinghouse guidelines for sulfate and silica. Also, Cation conductivities of the Main-Steam, Feedwater and Steam Generators were above specifications.
- i Calculations for the Cation conductivity based on ionic content indicated 1 47 '
c. 1 e '
~
d'. { w
( i e' t f I[ I M / that5 th3' Stbam' gen:ratcr (SG)f v31u3s 'weraibelow th3 ~ 0.8"umho - specification ' limit.' l The3 SG -: water chemistry included; pH,: Na+,. chloride,c Cation 7 . conductivity, sulfate, silica, total; gamma,:and)doseequivalentI-131 data.- 1
-)
,. .Seabrook'. maintains a performance' measurement of secondary chemistry using a ,
ches# try Index. . The Chemistry Index comp'r s the concentration of selected : j impurities' to the' limiting 1 values ' for' those impurities and' normalizes the. " i results. .Thelindex requires' data -for blowdown Cation conductivity,' blowdown
- sodium and ' condensate pump.' discharge dissolved! oxygen. ' The goal' is to be below '11 ~ The iindex ranged from . a high : of some 2.67 to 1.08 at the end of f
r the PATP.: 'I [ The limiting - condition for operation on' the Leecondary . 91de is stated in TS .{ i '
, . 3.7.1.4 L i . e .", the specific activity of the Secondary Coolant System shall ,
be 'less L than or' equal to 0.it microcurie / gram Dose ' Equivalent 1-131. -No ; activity was detected in the secondary coolant. j ' }. . ' The.' secondary system water chemistry was found to have excessive impurities q and no-radioactivity. The conductivity' limits for the: steam generators were j met. . The J impurities found were primarily sulfatos, silica,- but their presence was ' not considered unusual' during the startup phase of ' a new ' y plant. The Chemistry Index shows that the impurities - are egressing as .
, expected.
l} l J t i The SAT i concluded that the ' Chemistry Department had , been effective in controlling secondary. chemistry. j [. B : ; ii >
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'y -i 2.7l ETALUATION OF TRAINING '
a s ,
, Q J
f RunstrY , j
, q.c >
'6N "
- The:- Qualification. . Tracking System and; Maintenance Requalification Program L were ; two aspects .of 2 the l trainingc- program ' that had ~ potential weaknesses'. j
~
, 4 These : areas were - evaluated to; identify - needed enhancements. . Thei topics - )
+
covered by this section include the qualification tracking system and . the- . maintenance qualification program, u
g , 'ihe conclusion of the evaluation is that .sof tware 'being developed to track 3 qualification ' status will" satisfy Station needs _and Revision 1 to = the NHY . d Qualification Manuali(NYQM)' adequately addressed'the requalification. program. , for Maintenance: personnel.- .
m Summary Of-Findinns I i 0 i-
! O - Testing SAT = Reconsnendations (ST) j 0 -- ' Commercial Operation SAT Recommendations (ST) f 3
-0 ' -1 Long Term SAT Recommendations (LT) i 4
. Ouh11fication Trackinn System L
. A 1 computer tracking system is used to record completed training, but the, j
. system is cumbersome. The SAT evaluated'the capabilities of a new tracking system -to' determine if. the system will' support Station needs. . iThe= [
.c . ,
' evaluation was cond 7cted - by interviewing Training, Operations, . Maintenance, f; ,
, Instrumentation and Control,' Chemistry, Health Physics,4 Utilities,
. Construction, Facilities. Information Resources - Security, and Emergency
. Preparedness . managers and training coordinators and reviewing scoping documents andt i he system description for the qualification tracking system that is;under development.
y.
. l The existing computer tracking system does not allow entry with a name to
- obtain information on all training received by the individual or with a job or task description to obtain the names of all individuals who have i 49
/
- q. t, n--
+ -
- ~
j yr q , , ih - 1 p c;"
- complcted rcicted3 training lrequirem:nts vithin a proscribed ^ time intcrval. :
p'"? > . , ] . The ' existing'. system is basically a: filing i system ! thatL requires L searching) ,
,k lvarious' records lto check that specific; people have! receivedL he t prescribed -
H! , training [and that I datesfof the- trainings fallL within 9 designated timei
, intervals. ;Theisystemiis time consuming to use and subject;to' error.
p The ' new' . system will be an updated : ~ version of the. Training; Information, E Managessent ; System. The' -system .w ill correlate. training completed. and! ' I ,, expiration? dates: to ; task requirements.
. names,.' time intervals,. jobs, and tasks.
The system will ~ allow entry with Information that will be' retrievable includes individual qualifications... individual training received-0 Eduring specific time intervals, and individuals qualified.for specific' tasks" and: jobs. The system is. scheduled to be operational in February 1991.': 1 Maintenance Reoualification Pronram,
'The" program for- requalifying Maintenance personnel was 'not clearly prescribed.: ,The- SAT evaluated the requalification- program for: -the
.- : Maintenance Group. The evaluation was conducted by-reviewingLthe training l' + programs for the Maintenance, Instrumentation and Control, and Utilities q
. Departments : and ~ interviewing nelected managers, supervisors, and training j coordinators.
L Revision. 0. of; the -!NYQM did not address a requalification program. for the ; Maintenance Group. Supervisors ensured that people assigned to work were 1 1
. capable of: performing the processes involved. People: acquired expertise by is assisting .those who were qualified, attending classroom training, and' l 7 : completing practical factors. Supervisors ensured that people who had been qualified. . but had not. exercised that expertise during a reasonable period' j of time, assisted others nr. worked under close supervision until skills'and' :
understanding . were redemonstrated. Documentation of the requalification 1 l steps,-except for special processes, was not.always accomplished. 1 Section 2.6, Chapter 2, Revision 1 to the NYQM now addresses the requalification program for Maintenance personnel. The section provides the i 3
. skills, tasks, and processes requiring requalification; intervals required I for requalifications and actions necessary to satisfy requalification 50 3 .
. ,. . . - .. - . .~ -
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4- 1critiria.? Training.Cotrdinatcre dscumentLth3 complotica cf r: qualification a , 3
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' requirements on; department qualification matrices".. '
p- . t.%a u ,
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a TThe ' SATJ concluded b that' LRevision '.1.' to the1 NYQH: adequately addresses thei -{
' II "i
.A
, [ ^ requalification program for Maintenance personnel. .
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, :i m l2.8" ETALU& TION OF OUALITY PROGRAMS
, 1 as Summary -
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. The SAT; evaluated j the effectiveness of .the Nuclear Quality' Group'(NQG) in_.
' identify 1.1g , saf et'y - issues and : in' improving L plant : performance , through tho'
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? corrective.' ' action 1 process. . The evaluation was conducted by!. assessing. the - . J
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! J ' involvement;L of. NQG J in the . conduct ~.of . testing L and jin the analysis of. t'est t
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results .of) the Power Ascension . Test Program (PATP)- and by -reviewing- L) 3 5, ,
' initiatives. taken by . NQGEto1 improve' the Operational Quality . Assurancej
[.
= Program'(0QAP). 'The' topics covered.by this:section include surveillance of-
,i y +
Etesting, review of ' test results,. and Operational' Quality J Assurance . Program , qH f(0QAP ) '. improvements . : r
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5 i
'The conclusions-of the evaluation are that.NQG was effectihe.in' identifying'
~
~
issues . regsrding ; the , PATP L and that NQG has improved the- OQAP. by shif ting t .l towards . p'erformance . oriented assessments ' and strengthening: the overview. of: . corrective measures. i t Sumary Of Findings $
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I P - Testing SAT Recomendations (ST) t 4 ' OP '-- Connercial Operation SAT Recomendations (ST) ; [i ' 0 ; ' -- Long Term SAT' Recommendations (LT)
'.(> <
4 Surveillance Of Testine b 3
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? k "NQG: continued to perform surveillances of testing conducted under the PATP. ['H u . The- SAT evaluatJd the effectivences of the surveillance program in-j i 1 l' : identifying: safety issues with testing. The evaluation was conducted by .;
< ; interviewing the NQG PATP Coordinator and' surveillance and, audit personnel 4 4
.' 2 and by reviewing completed surveillance reports. A Surveillar_ces were performed by Level I and II qualified pereonnel using 2 approved checklists. The checklists were documented in accordance with the OQAP Manual . and the Quality Assurance Management. NQG personnel attended 52
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W41 ;rdfesh;r: tr ining.. including crew specific ~ tr ining. . fsr Oper ters: dad Testi
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9 J personnel.l~ Surveillance and auditipersonnel attended! pre-test briefings'andL 4 j
,Q l highlighted c witness, points ' for (tests; tof be i conductedi'during .-.on-coming
;, >- , I
,W - shif ts ? ! Changes [to? test2 procedures weret reviewedi by. two.NQG personnel,;one-- i j
,. v,e , . . .w -
%'m '
ofIwhom was -Ja' _ qualified' Seniorf Reactor : Operator . (SRO) L onloachf: PATP LNQG-
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1 Examples offit' ems I'dentified by'surveillances followf j
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(( ; 1 , . 4 ,, - --- Test: ii nterruptionl.was 'not ~ declared'- when= turbine; overspeeds trip { (
- y;, w mechanism was.adjusied.' i
[4{ o f }$e,i.
, - --- The wrong revision of Procedure SM 8.1, Power -~ Ascension = Test Program,- l t
o, l l[$ W t was'being used at;the' start of one test. j' 5 3 i
' Locked-valve verification was:not; performed at end of month.
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. A' Work Request.was: not ~ identified as. affecting' an up-coming test.
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e p f ,o .' -- Procedure steps were performed out of-sequence.. ,
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. - - A. malfunctioning. flow indication instrument was not~ corrected before : i
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H . starting a test. j s y, \ !r i 1
- p. a s b t;EF 1 LThese" findings were reported to Operators in a-timely manner for resolution
+
.gg ' < [m ..and processed in accordance with surveillance procedures.-
t }[p , s [gi' ] .p : The ' SAT - concluded that ~ the surveillance program was capable of identifying ; a e %c w' ' , " probleme.with testing if problems existed. 1
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.f l Review Of Test Results N
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,Q ;NQG verified ;.that acceptance criteria for. tests conducted during the PATP. [
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' had been- met.1 The SAT evaluated the effectiveness of the verification p 4 II ! i process. -The evaluation was conducted by interviewing Technical Specialists '
= and ' audit . personnel who conducted the verifications and reviewing 7 verification, documents and problem sheets.
53 g _: jt l; 3 '; . L!" ,
i- -. f. G , y . U 1 Sixteeni(16)"of the 35!I PATP tests: werel selectoCi forf review" by Technicall
, i D1 6 l specialists.: The reviews' verified that" test results met acceptance criteria' s ore problemi sheets 1 had9 been prepared Jto':: obtaini Engineering : evaluations.. S.
-k Tests thatl wereWuoi s31ectedi for. review . by Technical SpecialistsLwere' -
'^
j] o s 9 reviewed by audit personnel. In addition to ver11'yingLthatLtest results met acceptance. criteria', i audit. ' personnel verified other attributesJinbludingy ( rtest program:renuirements. . Repetitive' tests conducted at;each plateau were [
] ; Jreviewed- by ltechnicalL specialists or audit personnel at the lowest - power 2
k $; . , . level 5 toiidentify; and resolve generic problems as- early. as possible in. thei
- testing program.
i s a
+ , j The review-of test packages by Technical: Specialists and audit personnel did
' identify deficiencies' in L two packages. Four . thermal? points in ST-52'..
Thermal Expansion Test, were outside of acceptance . values and ' wr e ( not iEN' identified on problem sheets. . The- acceptance criteria for silica' was ' noti t
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. met -and ~1dentified ' in ST-42, Water Chemistry Control'. LThese- deficiencies '
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j s were reported to PATP management and resolved before the packages'were sentL j g ,, L to the ; Station Operation Review' Conunittee (SORC) for approval'. ~ In addition. ; " y TNQG' personnel': reviewed each work package required to be completed prior to a
'[ -
l test i :and performed a walkdown of affected . systems to ? verify .. the proper L .v completion of work. -All deficiencies' were corrected prior: - to / estL t .l i L [ performance. 9 The- SAT - concluded that the NQG verificationi process for ' test results was
-t effective in identifying problem areas. 3 yb
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'T
. >i 'Doeratiional Ouality Assurance Pronram (00AP) Imorovements 1
,. .NQG.has-been shifting from a focus on programmatic and procedural overviews. 1 1
? .to performance issues and has been increasing management attention on the- i
[' corrective .. action process. The SAT evaluated' progress being made: in
' A ' refocusing thh NQG-effort. . The evaluation was conducted by interviewing NQG
$io ' personnel and reviewing recent audit and surveillance reports and closeout i ; actions. t . I i 5 ' I; a
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, NEY,f und:r1th0 Ind: pend:nt i R; view iT;am, conductcd ' cn cviluation J. cf., th%
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s , implementation and' effectiveness 'of the: OQAP.c _ Nineteen (19); recomunendations and J forty-four o (44)> " ' suggestions; were . provided - for ' . improving 7 the . program;; p #- MostL of! the: recomunendations ; and { suggestions have. been implemented with; the1 m e :l, L 6 remainder being-. track'ed , in the ? Station's comunitment ~ tracking 1 system. - Thec '" u
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B , recommendations include shifting to performance based inspections, improving; [l; :
; interfaces / through meetings, and strengthening the implementation : of .. the .
L , , , g correctivefaction process. ,
,f ;
c - . , 1 ?' All" NQG technical'. personnel ; have : been trained' on performance " based.- [ ' 1 m techniques t for audits, . surveillances, and inspections. Sixty-two - percent ~ ' (622) of the.: technical people ; have . attended: the detailed . plant systems. ' i
' course: with others ' scheduled for:1 subsequent classes. 'All technical I personnel'~ receive- quarterly' staff -technical training. . This training' i, -
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-combined:with a new: inspector certification program,'are prerequisites;for a; a, l j , performance based-QAlprogram.- The shift to performanc'e based operations}is, i
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already evident ' in. the - field > by . the type of audits and surveillances : being
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. pursued'.. 'The NQG. coverage of-the PATP is an example of a. performance based' M' '
1 a s se's ament . : .The. levels of coverage were well coordinated,! personnel
;y : involved -were qualified and : trained. and .the . findings and'< reports #
contributed positively to the-' success of the PATP. !
-Production interface meetings, involving the' Station and NQG management and supervisory -personnel,. have been instrumental in strengthening. the ;
coordination ~'and mutual understanding of the Station and NQG efforts. ! Issues > that: haveD been discussed at = those meetings ~ ' include ' implementing- IRT m recommendations - for improving the OQAP, furnishing technical specialists - . 1 ? I
~ from :. the . Station to support audits, and resolving a Management Action, .
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! Request-(MAR) concerning plant modification implementation. The MAR 11s an [
example of. managing the corrective action process.- An audit team found-a
!$ number of repetitive type problems with a design modification to the wide W range gas ' monitor. The NQG identified the generic implication' of the '
problems and. initiated-actions to correct the programmatic weaknesses.-
}
i The SAT concluded that.the NQG has improved over the past eighteen months in performing more meaningful inspections, surveillances, and audits which
. J ' identify plant significant problems. NQG has also improved in managing the ! - corrective action process.
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..> 4 li, 2597 ETAMETIGI 0F EMBIEEE .IMB TECEKICAL SUPPORT
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; Summary .
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,, .Thefresponsiveness;~and:-effe'ctiveness.'of the: Engineeringoland- Technical Il '
a Support' organizations, 'in' resolving plant problems were evaluated by . P- *
; ;considering l (1)L plant . problems that have ; occurred since . the : Phase 2' Fifty. ,g; : Percent Report was -issued.1 (2) areas f related to ' Engineering and Technical; em5. ESupport i responsibili$ies l that' ~ had ~ potential; weaknesses.f and (3) ' items of:
x 7 special interest. ; Plant problems reviewed included feedwater oscillations,.
. y i i.
'generatof neutral! ground fault, turbine generator electro-hydraulic; control 1 c(EHC) . pressure switch u. vibrations, ; steam generator high: level L signal. and-generator hydrogenileakage. The areas'withl potential weaknesses that were m
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Treviewed. includedi the1 number of . open Requests for Engineering L Services '! L ' (RESs), installation _ instructions -forL design--changes.- and' completeness { of > [ closed' Work Request'. packages.: An' item of.'special interest-.that'was reviewed' , was=the' implementation"ofLReliability Centered' Maintenance"(RCM); m u q .. The- conclusion .'of the evaluation is that Engineering and Technical, Support , fare"effeyhive'inresolvingplant.problemsandcansupportreliable'andsafei [
; plant; operations.
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, ai Summarv Of-Findinne;
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l i 0, -- Testing SAT Recommendations (ST) 3 L :
~0 -- Consercial Operation SAT Reconunendations '(ST)
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, 3 .-- Long: Term SAT Recommendations-(1.T) !
o 3 j Feedwater Oscillations { O'scillations occurred in the feedwater (FW) system due to fluctuations - in the heater drain (HD) system. ~ The SAT evaluated the effectiveness of ' i Engineering and Technical Support organizations in eliminating the
' oscillations. The evaluation was conducted by reviewing documentation and cdiscussing' steps :taken to resolve the problem with cognizant engineers.
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,. m N . Oscillations, in the-' PW ; system were observed at. the 'peginning of othe Power" ,
ih , / Ascension 1 Test Program (PATP) when the c plant;was at low power and HD pumps 'i w e ,' _ x owere not running. . The gains for the feedback loop of. the 'steamLgenerator
.-7 e ~
water? level . control andi. FW pump speed. control systems 1 were adjusted i to
' minimize the : oscillations. When powerLwasJincreased'and the HD pumps were:
- started.: large oscillations > in steam flow, feed flow.i and : steam generator m
y , levels 1 occurred." An analysis of data-taken-during-those oscillations showedL 3 W .that the;two HD pump discharge valves were oversized. The oversizing caused ~! i large fluctuations- in' FW pump suction pressures which resulted 'in
$ [ (oscillationsf inE the PW and . steam systems. - The T solution to the oversized :[
T I, 'e discharge valves was to change the cages.of.theLyalves. .The installation of; -
".a .
?
g , , l reduced 1 size = times with . reduced flow characteristics was accomplished ~ as o , yp
. part 5 of the _ _ implementation of Design , Coordination Report ' (DCR) ~ 90-0027, ;
l' ' I. Heater Drain-System Enhancements. ! L ! p.g , ... 4 Iri addition to ithe changeout _of- valve' trim for the HD tank discharge Lyalves.. + P*
.DCR : 90-0027 : provided for : relocating and replacing the ' proportional ' level . 'I l
,p controller'for1 the HD tank discharge valves., changing the level controller ' '
f, '
. for'the spill valves from integral to proportional operation, raising the lhigh.and:high-high' level settings for,-the HD tank, and adding snubbers for. .
flow transmitters.: These enhancements were: desi n d' to reduce the , sensitivity of the HD system to oscillations.by increasing the surge volume s df:the HD: tank, reducing-the sensitivity of the level control system for the
~
? HD tank,'and reducing the potential for stray control _ signals.
is .; l [ ? . 1
- DCR1 90-0027' has been effective in reducing secondary system oscillations p
- l. .r ; through the full power range. Oscillations still occur, however, especially- ,
s
.during special' evolutions such as _ cycling turbine . intercept valves. . The t
amplitude of the oscillations is usually less than five inches ( of level 1 change in'the HD tanks, but on occasion the reduction in FD flow to'the: feed-g pump. suction has resulted in a low pressure condition which has actuated the g : third condensate < pump. Three condensate pumps have been lef t running with- ! the plant at full power until secondary system information is further analyzed'and enhancements made to reduce HD system oscillations. The HD system .at Seabrook does not have a surge volume comparable to, or a control system as sophisticated as, those at similar plants. The Seabrook 57 i 1 I W ,y 7, L_0 Y :-_. ._ _ . _ . _ _ _ _ _ _ - - - _ . _ - _ _ _ _ _ _ _ _ _ , _ , - , .
( 1p 4 ( plant is mora s3nsitivs to escillotions and, tharoforo, moro cf an engineering challenge to obtain optimum performance. The combined efforts of the Operations, Technical Support. Engineering, and Maintenance organizations have been effective in improving the performance of the HD system throughout the PATP. Management direction has been timely and helpful, data collected for evaluation has been appropriate, analyses and resulting design changes have been technically sound, and teamwork and cooperation among the involved organization have been good. Consultants and representatives of the architect-engineer and equipment vendors have been I used in an advisory role with Station Management clearly directing and controlling the ef fort.
.- The focus of the on-going effort to improve the performance of the HD system is on identifying and minimizing the cause of oscillating flows to the HD tank under steady power condition and evaluating the benefits of replacing the pneumatic level controllers for the HD tank with digital controllers.
I Work Requests and design change documents exist for activities actually performed on the FW and HD systems. However, a chronology of the approach taken to resolve the oscillations, including actions taken, the bases of those actions, and results, has not been prepared. The SAT reconnends such a chronology to assist in future analyses and troubleshooting of the systems. (4056)(LT) I Generator Neutral Ground Fault The generator neutral relay de-energized causing turbine generator and reactor trips. The SAT evaluated the effectiveness of the Technical Support organization in resolving why the relay de-energized. The evaluation was conducted by field observations, document reviews, and discussions with the cognizant System Engineer, vendor representative and pertinent Operations, Maintenance, Engineering, and Technical Support personnel. A module of the generator neutral relay monitore the last 5% of the generator winding for grounds using the third harmonic of the generator I output frequency. A ground f ault draws down the third harmonic which de-energizes the relay causing actuation of the generator primary protection breaker lockout relay. The lockout relay initiates a turbine generator trip 58 I -- - - - -
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wh n power;is j ct 502; cr abova. ,l Th3 turbin3l tripic:uses o ro:cter i trip.} ,. H "u m "
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r 1 Most 2 utilities ' do.- not iuse special L ground; protection ~ for ? the ! final 52 J of-
# 1 4 / *
$ 'generatori windings.- ' The T generator manufacturer L does not7 reqdire" that L
. ' Ifo t protection. 3( Utilities.R inclhd'ing !.Seabrook, s do-~ protect !the first E 955!? ofi 0 1
4 generator, windings.againstl grounds. ' n. 9 4
'The Station' Manager held severa1' meetings with managers and. supervisors from I 4
1 lthe!-Operations," Engineering,. Technical ~~ Support, and ! Ma' int'enance- 9 i~
- hrganizations 1 toi develop a . plan. to resolve the L cause: of. the ground signal, g
Iaf .
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The meetings allo monitored progress in-carrying out the plan and modified: ,
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/ the plan as necessary. The Executive' Director. Nuclear: Production attended' .., m ! thei meetings ~ and provided d'irection~ as appropriate = The l meetings D were ' effectiveDinibuilding teamwork, incorporating the ' ideas. of - experienced - f C
- managers and. supervisors in' the approach to bettaken, and providing on~-the- ;
scene management approval of modifications to the plan.- i 4 The-System Engineer implemented-the plan.- The first steps in-the pla'niwere Qi to check that a ground 'did not exist -in generator windings and that .,the 1 , b
, relaydas functioning properly. . No- indication of a ground. existed on 'the t core monitor, and a.Doble test of the rotor indicated that the. windings were- ,,
h ' satisfactory.' Electrical'and visual checks of the generatorineutralLrelay1
*. idemonstrated ; that . the relay was functioning correctly. Checks ' for grid
, idisturban'ces, loose connections,ypersonnel actions,-transducer performance, I
u and.adeq'uate cooling. flow also showed-noicause for the' ground signal.- j
; 'l , 4 l
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.. M Thel 5% module was removed as a trip function of the generator. neutral. relay byn Temporary Modification- (TMOD)~ 9 0-0021=, : and Lmonitoring' equipmente was -
a IW 11nstalled to verify'that the module has'been' set properly. After. verifying- !: 'the' setting of the relay and discussing the situation with the relay vendor, ;j othe decis' ion will be made on whether to return the 52 ' trip function .to the 1
- J 7 *
, relay,'to use the 5% module to actuate a warning light instead of'a trip, or 7 to remove the module altogether. 3
[ Steps;taken'in troubleshooting the relay were systematic and thorough. The a i response was rapid and thorough. The exchange of information between the 4 I' .c 4 System Engineer and Operations, Engineering, and Maintenance personnel and
- , the documentation of actions taken were good. Reports prepared by the Lead Electrical System Engineer summarized actions taken, the results of the l l 59 w .
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- f (cctions',?cnd; future stcps.H ; Assist 0nceLfrom r0 pres ntativesfof'the' design"tri
~ of j the ? turbineEgeneratorf ground : protection L systmn f and L the vendor l of the ,
' relayf wa e' ' appropriately ob'tained' ~ at ( the. ~ beginning - o'f the effort.
v s l [. Industrial $ operating i experience with 52 generator winding protection- was c . i Echecked.; Only one utility was '. ider.tified- as using= tho' relay 'as n' turbine j ' '
~ generator, trip ; input signal. and ; that utility had trip' experiences similar
~
j i to' Seabrookb The utility. was: using the relays : which were manuf actured; by ; 7 different" companies. to : actuate warning. lights in two plants.: The utility. l
, 'r. did i not ' install- the relay"in the third plantL because ' of the l problems I experienced in the first'two plants.
'< 1 q
, 4 -
p ;g <'EHC-Pressure Switch Vibrations-3 m i Two. pressure = switches on the emergency trip system (ETS) fluid line of thel
,4 .
j y l , turbine Lgenerator . electro hydraulic control (EHC) system actuated . EHC ; lows '
, pressure bistables in the reactor protection system (RPS) cau' sing a reactor; d L
i trip.! The; SATi evaluated the effectiveness of lthe Technical ; Support ,
~'.
t L , organization in~ resolving the cause of the ~ low pressure signals; that .were , y
- The evaluation was conducted by field; observations '
7 s sent' to - thel RPS, + 3
;^
~ documenth reviews : and discussions with pertinent Operations. Engineering i 1
' Maintenance, and Technical: Support-personnel.
An : event Levaluation ' team, which included the System Engineer - for the' EHC Jsystem.1 was assembled under NHY -Procedureo12830 Event Evaluation : and Reduction' program, to sununarize the sequencef of' events = associated:with the
~
'l j( 1 reactor trip and to explain' how' the trip occurred. The root cause of the-
]
$ trip, which was determined under NHY' Procedure'12810 Root Cause Analysis.
. established that'.the trip was caused by vibrations closing contacts of the y a3' EHCi pressure switches.-
The three pressure sw2tches were mounted on the bottom of the: turbine stop valves. When power was' increased above 702, the s
; pressure switches. vibrated significantly, and momentary low pressure signals q
'y 7: were sent to the RPS. The RPS logic requires low pressure signals from two
' of - the ' three switches for a finite ' actuation time before initiating a reactor trip. When thoce conditions were s e. tis fied , the reactor tripped.
? '( '
The EHC hydraulic system operated properly during the vibrations with no indication of low pressure. No fluctuations in EHC pressure occurred, the EHC standby pump did not start, and a low pressure alarm did not occur. 60 e
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, 1 m c (The Station; Manager : held.- several; meetings : with' pertinent; people ~ from thei
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1I0perations.Ongineering,TechnicalSupport,andMaintenanceorganizationsto E
"n i review'. the $ symptoms of: the reactor! trip, the root cause determination,; and ; j j-correc tive l actions '.' :The: Executive' Director Nuclear Production = attended the ,!
' meetings.
i
- The System Engineer researched information ~on 'industriali ,
operating experiences and:found-that another utility had'a similar' problem.: v i .
- Also. - manufacturer's instructionsi for the . pressurelswitches were. found J that ;
cautioned against using thefswitches where vibrationsiexisted. . Engineering l l prepared Minor. Modification (MHOD) 90-626 EHC Preosure Switch Relocation,'- 'I
- to move'the. switches.to building., supports to reduce'the offect of ateam line [
{ : vibrations. - -t c; 1 2
. The SAT l concluded ' that: Engineering and Technical Support responded quicklyL .g l ,
L to the problem.1 The problem was analyzed and the design change prepared?and' ; e K , o implemented within.two days, b b . :i. ;- 1; i ( Steam-Generator Hi-Hi~ Level-
} A steam generator' Hi-H1 level signal -was.: initiated' during the reactor ' trip - 4 caused by' vibration of the EHC pressure switches. The SAT evaluated the +
effectiveness ofi Engineering 'in resolving 3 the .cause_ of the. Hi-Hi signal.:
,u m
The1 evaluation was conducted ~.by discussions with cognizant engineers and-by. review of pertinent documentation. A steam generator.Hi-Hi level signal'is initiated when steam generator level J 55 freaches 862 of the narrow range instrument span. When two of. the four 4 p , narrow range channels'of. any steam generator exceed the 86% setpoint, the ;
' turbine and FW ' pumps are tripped: FW isolation, regulating, and by-pass
,. valves are closed land automatic start of the startup feed p'.unp is- blocked.
. These actions protect the turbine from steam generator moisture carryover. ~
I
-The high - steam . generator level must be cleared 'and the reactor trip- 1
, ~switchgear closed' before the FW regulating, by-pass, and' isolation valves!
'can be opened.
1 Steam generator levels had not exceeded the 86% setpoint when the Hi-H1 level . signal ~ was initiated during the reactor trip caused by vibration of ' i 61 g . .
/ t i{ y % '
.l.
.~ .. ,_ VL 1 Enginnring li cpprcprittoly s chtck:diwith L thi m L i '
,ith n EHC ipr 0ssuron switch 3s.. n 0 p ,'F .
's
; @ .'supplieri off the NuclearJ Steam. Supply Systr.m2 (NSSS)- about .the:~ occurrence of
* ?[ 3' ' similar' synalsk ini other plants. lOttier? plants : had16xperienced aimilar' mqw
~
.. .. ,7
- signals which have : been; attributed to pressure pulses? caused : by _ the rapid i closure of ' turbine control: valves.- The.. pressure lpulsen farer transmitted - N t, i" through : water-filled impulse lines ' for steam . flow transmittere _to level
, s transmiEters; through' steam. generator taps. . Upper steam generator taps'are j
[ q sharedh by Iboth level - and flow transmitters : in ; two of the four narrow-range !
"E , ' channels.- A review'of plant data for the? reactorf trip' showed that (steam j h g)g ,
generator' levels. in! channels: .that ~s hareda upperi taps .with. steam flow, ' i transmitters did = oscillate when i the . turbine controlu valves: were1 closed'
~
' '4
, . Leve1s: int: channels- 'that' did not' -share upper. taps .with: steamt flow . i M transmittors did: not- oscillate. This confirmed the experiences of other '
j
~ plants.
~
f F The 'NSSSj supplier reconunended insta'llingL a lag time constant ' of one second ,g ; to ; allow " the plant > to: pass through load rejection 1 transients'without' a creating . false steam: generator Hi-Hi level' signals t A ' disign ' change is'- !- ~ s being prepared - to install the. lag time constant - during the next. ex' ended. t p unscheduled or. refueling outage.
~
-l' H The SAT) concluded : that' the initial- Engineering response to the problem was -
i ' rapid. The1- cause - of the Hi-Hi signal was resolved within one day of the.
, - reactor? trip. : Corrective action to 'be taken was resolved within'a week of
;, the - trip. . The proposed Jcompletion date for the design change packaget is:
W , February l 22, 1991. t + w . . < L l'~ : Turbine Generator Hydronen Leakane E:p L , Hydrogen usage in the turbine generator gas cooling system was excessive I $ during'the=PATP. The SAT evaluated the effectiveness of Technical' Support-M lin. resolving the' cause of the hydrogen leakage. The evaluation Lwas
-conducted by discussions with cognizant System Engineers and a review of p, , pertinent? documentation.
I o=- = Hydrogen usage.was. expected to be 5 to 6 bottles a day, but actual usage o averaged some.14. bottles a day during the PATP. A Hydrogen Leak Detection ; 62 is . .i___'___.__-_ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ -_ _ _ _ - -
K gi f , G '? 9 '
]
a 4_ i- , i prtgrama was;:initist d . ct!/ th3 beginning?cf th3ftest1 prtgram to: riducs: l 48
- hydrogenT consumption. ~ A Econtinuous~ effort' was made by Firefighters.:
"I'4 s n? 1 i Maintenance- Technicians.jand : System - Engineers during : th'e PATP to identify l y [ ' 7 s ;and correct 1 hydrogen leaks'. . A dall'y plot of. hydrogen usage was- used as~ a -- h
'[ (p'erforman'ce? indictor? of ' the offort.
~
,t Leaks that ' were . found and ' repaired
! included oillipipe J flanges, valve packing..'hydrogen cooler Lgaskets. - end . l shieldL seal'ing ' groove. ok #10 ' bearing, RTD ~ gronunets. .and level' detector ~ flanges.. 7 ( 1 l, . Towards the endlof the PATP and insnediately af ter a full. power reactor trip, s,. hydrogen usage : reached 24 " to 30 . bottles a day. The leak was eventually. I located in the.#10 bearing housing after.the wand =on the hydrogen detector was replaced with : tygone tubing so that the sensing point couldi be moved: J l
, closer,to the sealiassemblies. After: consulting'with the turbine' generator' <
I vendor, the _. pressure E of the s eal " oil system was raised to increase the.-
~
differential pressure ~ with the ' hydrogen system from 8 to 11 pai.; The; ; l! , increased ' pressure reduced hydrogen usage to the expected 5 to 6 bottles n' f a ,I Eday. Vibrations from the full power reactor trip are suspected to have, cocked = the - split L seals - in , the- assembly allowing more leakage. . The #10) j jy bearing 51s ; scheduled to - be opened and repaired during the next refuelingi 1 .
- l. -! , . outage, p
. The1 SAT concluded that t Management .and System Engineers demonstrated good persistency in finally rasolving the leakage problem af ter experiencing a ,
; number'of partial succ sses and setbacks.over a several month period.
, if L um Recuests For Ennir.eerinn Services (RESs) .
i
.Some 730.RESs are open. The SAT evaluated the impact of these open RESs on
' plant; safety and reliability. The evaluation was conducted by reviewing a 3
. printout of:the open RESs, checking a sample of open 1986 and 1987 RESs, and. -
discussing! RESs ,with Engineering, Technical Support, and Maintenance 7
; -managers'and supervisors.-
The SAT Fifty Percent Report discussed that RESs provide the means for departments and groups within NHY to request Engineering support. This , support included technical interpretations, engineering evaluations, special
'3) y .
63 , 1 q
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i 3-gg SR 7 .
, I* f" P studios L : cad L plint i modific:ticn i requsst's. Thi l report concludsdl?that- '
Nk
, ,;.s ..
/' . (, V L Engineering 7 wast increasing L the i- emphasis -- on T closing - RESs " and C that : this( , ~, .~n' .
, ;~
, Leffort, e combined) with an" expected decre'aseiin the number .of newl RESs being ,!
t , .
' generated > as ' the plantiEgoes into Comunercial- operation ; should-. reduce': the: i d ' number of i RESs substantially. . The SAT 1recoannended that RE8al be reviewed! .
,l annually to ensure' action was being taken.for closure; J -
ll
$, t 1
4 ,
' Since : the Fif ty ( Percent Report - was yissued, ; open RESs hve been reduced ' byL ,
p some ; 100 ? as - expected. .The' rate'of closure hast reached approximately-60 a. j i E month.- Departments. and groups have "been requested through l:their
? 'p . representatives. on the Station' Modification ' . Resource E Committee l'(SMRC) to j
( j . bring to- the attention of. that conunittee at their weekly meetings-any RESs' .
' tha t- should be rescheduled ' to support : maintenance actions. . or safe and:
1 reliable . operations . .SMRC' members also were requested to-review open;RESs. , to identify'those needing rescheduling-to support the nextLrefueling outage. No items were identified.~ Technical Support has been given; clearer. guidance. f
. oni RESs that ; System Engi' eers' . can close.
~
u. The J backlog ; of j RESs ! requires ~ m .. . .
. Engineering action.
( . Reducing the number of new' RESsD being' sent 1to ] Engineering will' assist-in clearing up the backl'g, o i Thei SAT' reviewed open RESs that were initiated in 1986' and 1987,11' and 84 ,. :respectively,;-and concurs that the items are low in priority-and are plant
' ~
enhancements. :The SAT also discussed RESs with Maintenance Supervisors and,
+ l although E the perception - is that - RESs should be answered more quickly, 1 specific RESs that are delaying ~ Maintenance actions could not be identified. j o ;
j i l The SAT concluded : that' Engineering should . continue monitoring .the . reduction '
- of ' th'eLRES backlog and request additional resources if the downward trend ,
does noti continue. .Also, the SAT concluded that departments and groups- ,
; E. should forward problem ~ areas with RESs to the SMRC for resolution through
- rescheduling or additional resources.
I , Desinn'Channe Install'ation Instructions o Some. Work Requests used to implement design changes may not have adequate installation instructions for workers. The SAT evaluated the adequacy of
? work instructions for design change packages. The evaluation was conducted I; ; 64 l
M{m
y;w , m . ;; i I *
,
- Lbh rcviewing l design' change work packag:s* cnd3- discussing tho ' cdequ2cy cf #~
' installation ~ instructions with System; Engineers and Maintenance personnel.. '
*< The.' Design : Control' Manuall(NYDC)" provides guidance on the . design process.
t
" Procedurej DC 2.1, Design Change ~ Package Processing, provides ' guidance oni
' preparing' Design Coordination Report (DCR).E packages . -- Procedure ~ DC L 2.4, Preparation 1andLProcessing of Minor Modifications (MMODs), provides guid'ance~ ;
V on preparing HMOD packages.': Tha design packages contain theresults of:the-
')' = engineering _ process for translating design t bases,c regulatory requirements,.
[ , codes -. andI standards, and constructability t into- specifications. - procurement: documents, drawings,. procedures, and instructions. Procedure MT- 3.1, 3 : DCR/HMOD : Implementation - Plan, establishes : the : proces s : for : implementing and
~
,,. . testing 'designl changes. - The DCR Implementation Plan (DIP) lists-_ the _ steps q p necessary to accomplish field' work efficiently with proper work. controls. .
- J
(_ _ 4 The Technical' Support Manager determines when ' a DIP is needed - for a DCR. i The Cognizant Implementation Engineer (CIE) determines if a DIP will be used ps for.a MMOD. The-information appropriate for DIPS is to be included in WorkL L
- Requests when: DIPS: are not 'used. DCRs and HMODs-' are implemented in the 4
. field 'in accordance' with Procedure HA 3.1, Work Request.
A . complete description _ of' work and Linstructions to workers is to be included 15 ~ the -
- ( . Description of Work / Precautions block of Work Requests. IfJ detailed- d n (instructions for complex tasks are required . procedures 1shodd be used, d
[ Work Group Supervisors implement DCRs and HMODs'using controlled. copies of
. DIPS, Work lequests, procedures, and' drawings. ;
J ' A review - of design change work - packages that contained ' DIPS showed that ; I . detailed <fic;1d proce'ss instructions were available to workers for installing the-design changes.' However, the practice of using DIPS has been reduced ins 1 ( the past couple = of years.
~
Sixteen DIPS were used in 1989, and none have l ( been used; in 1990. Design change work packages that did not'contain DI?s
;- = varied in the detail.of field instructions provided. -Approximately half of the packages sampled contained detailed instructions while the other half ' l l .
ibasically' referenced the number of the DCR/MMOD being installed. , Discussions with Maintenance personnel confirmed that the extent of
-instructions for installing design- changes has been marginal in some l
' packages.
~
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' V1' Th3 SAT'concludtd that-some d3 sign changs work packag3sLdo n2t hava cd:quato f- ,
, R m
~
. field fins tructions .- The SAT recommends that- Station Management 1 review the ' ,
j U
) progress for: creating field work packages. (4057)(LT) i
+ ,
j
~ .
Closed Work Reauest Packanes. . Administrative errors have been reported.-in closed Work RequestL packages. .- 3
; 4
;The SAT evaluated the correctness of closed ~'WorkE Eequest . packages by ; ',
i 4 , reviewing a sample of recent. packages. . ' i Section 4.2.3.20 of Procedure HA'3.1, Work Request, states that Work Group: [
; Supervisors- are -to . ensure that Work-' Requests are E properly -completed. I o Section: 4 ' 2.5.3 ; states that System Engineers--are to .ensurei that - Work
[. ' Requests are properly completed. System ' Engineers review Work JRequest { packages after, Work Group: Supervisors ' have completed their reviews , therefore, System- Engineers .have the ultimate responsibility: .for 'the' i t completeness of Work Requests. U 7" .- . A sample of 60 recently . completed Work Request packages-were selected for-review. . = Administrative errors were found _ on: the first 6 packages that were g
'I - chec'ked,and the review was terminated. Some of the errors noted included:
, i y ,
- Deficiency tag. was noted. as ~ removed in the- Consnent section but was e3; k ,, listed'as not' applicable on the Field Work Closecut Form.-
L. h j - 'RES was noted as written in the Comment section, but~the number'of the~
"k . -
RES was not recorded under Follow-Up on the Work Request Form. s
-- Painting was listed as follow-up action on the Field Work closeout I -s Form but was not listed under Follow-Up on the Work Request Form.
k n l -- Tagging number block was blank. "
- n p -- Packing rings noted in Comment section were different from those listed on the Valve Packing Record.
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% ; N?tsefon .Igniticn Scurc3 P3rmit" ccnflictsd cs; to wh:th3r ensi or two e ' fire watchesLwere used. .
I The .'SATS concluded"that L administrative : errors exist in; closed WorkaRequest' ,
>+
packages.; The SAT? recommends that (1) a training session be conducted for.; j < l System Engineers : which . involves ' reviewing ; closed Work . Packages to . identify , errors:and1(2) Supervisors:" sample . closed Work Packages monthly Tto: check ' on 4 gm - progress being made.:-(4058)(LT)
> -{
.i
,g Reliability Centered' Maintenance-(RCM) d
; ,'}
L s e New Hampshire Yankee (NHY) is developing an RCH program.- The SAT evaluated f progress Ebeing made =in Eimplementing the. program. .The evaluationo was t [, " conducted f by reviewing - documents and ' discussing, the program with cognizant - !
' Engineering ' personnel. . ,jt[
1 L T Twenty (20)' systems:have~been prioritized'for conversion to RCM over a five. It : l L 2 year period.: Two (2). system's'are scheduled for completion in 1990. "A goal. ; to Lconvert; four (4) systems annually 'in subsequent' years has' been; i ( ,
- established.' Although the' number of systems being .,nverted:each year is
-_ irelatively'small. the returns will' be significant asJonly ten (10) systems ]
[ . , l Lwill' cover 90% of: the' risk to. the. plant based - on accident . scenarios
,3
_m.m 4
' considered in the Probabilistic Safety: Assessment (PSA)'. >
7
%, j
[ 5 .. The two . systems- scheduled"for completion fin 1990 are- the Emergency Diesel i 7 Generator- System . -(EDS) and the Emergency ' Feedwater System (EFS). The'RCH . 3
- ' cu .. .e L ,i analysis of the EDS 'has - been completed by the Reliability and Safety l
/ gl , .Dspartment. The failure modes - and effects -analysis, (FMEA) and Maintenance , ll recommendations will'be forwarded to the Maintenance Group for review:by the i
- t. t m
- niiddle g of September 1990. The-analysis of the EFS has been started'and can' lW "b'e' completed' realistically in December '1990.
. A previout ly ' unscheduled o l
. ge ,
' system. the Chemical and . Volume Control' Sy' stem (CS), will be the- third system started in 1990 and will be completed-in early 1991.
. 4 Systems con 91dered for-conversion to the RCH program were those that affect
$4. ., plant ' safety or availability. Those systems were then weighed against li _
availability impacts, regulatory commitments and interest, PSA safety 43 67 -i
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r:nkings . -.the; sc:pe - cf ~ Pr vintiv' a ' Maintenance L pregrams - cnd : the extent cf a Corrective Maintenance histories, . to - arrive: at- a < ranking for; ther systems. . n The; top ' twenty J systemb ; by ranking willTbe'~ converted to RCM during the ' next .
~
- m. '
fiveLyears.- _ ' Y I
,f- , ,
~
. Program methodology was : followedo in' the / analysis - of L the EDS.- l Critical lcomponentisi were ' determinedi f rom their effect on' system functions.- .,These components:.were then considered for. Maintenance coverage depending' on their t J t
,' - 4.
specific:ifailure,-mode 'and 'importance.. ranking.' The . Maintenance [ reconsnendations T resulting - from the analysis will . now be~ reviewed by! the: :
, . Maintenance- organization.' to impart practical considerations to' the analysis s W " '
that was basically theoretical.
+
The' SAT. concluded (that NHY has a viable,RCH conversion program. Theshstems:
.being' converted.are'those that will give'the biggest return in added; safety- '
- and-reliability.
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f) 2.10.ETALBATION OF PERF0ml&NCE IMDIC& TORS I J gj '
, . r;
,NNY.uses' fifty-nine (59) indicators to assess and monitor key aspectoiof the.
@ performance ~of'Seabrook Station. .These indicators cover the, key; performance: )
~
areas of : Operations ; ; Maintenance . Technical : Support, MaterialJ Manag'ement, Hb : Personnel,. Radiation'.' Protection. -RadioactiveE Waste,: LChemistry, . Quality :! Assurance. . and Industrial: Safety.. ~ Data. for the performance- indicators is
, . compiled' monthly.
^ The Performance Indicator Program was evaluated to determine the usefulness -
r t oft the indicators .in - improving performance.
'Also, specific _ perforr.ance' .t indicators were..usedi by the SAT in assessing plant performance during"the, i Power Ascension-Test Program-(PATP).
} t p .The. evaluation has;two conclusions. 'The;first is that the: indicators will- >
- o. . . .
become c more ' helpful' in improving : plant-' performance when : data for- steady'-: p .. .
! 1! ; state Commercial ' operations . becomes 4 available. to : produce more meaningful- ,
[ y trends.. The second conclusion ,is that Operator, . Test GroupT and i- plant j iperformance7.during the: PATP far ' exceeded the norms for 1 initialL - plant- i a ..' 4 startups. 1, 1 ..'I ti DW , h f Sdamary Of'Findinas j L : ' p ' a j L O: --- Testing SAT Recommendations (ST)
;,* e.
1 0~ - - Consnercial Operation SAT Recommendations (ST) l' u .t p L: -: ~1 -- Long. Term SAT Recommendations (LT) '! l j i-EP erformance Indicator Prograp_
- ,W -
j ,' The. Performance Indicator Program was revised in March 1990 to include-INPO' ; ii ~ <
-items . and to implement a Performance -Heasurement Report form. The SAT f
e - evaluated the effectiveness of the revised program in improving Station
- fperformance. The evaluation was conducted by reviewing monthly Performance
. :). '
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f o - iI s,i, J,T '
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o
,M courementsb ,
R ports? thrcugh July.e1990: cud"_by- cerrolating.[d: grading l a ., . 1 .s
?
D fM'> . performance > trends with corrective' activities observedlin the field.: y 1 91 ' The -indicators being ; used,its supplemented by'. a . forthcoming change to " add' s surveillance and rework. indicators, are' adequate'to monitor the performance ;
'of the Station and are consistent with those recosunended-by the. Institute of' Nudlear. Power Operations'(INPO).- Afterthedefinidionsofindicatorstin:the! i 2 Performance > Measurements Report are.made consistent with INPO guidance. thel
-[
m indicators will - support _ information needed; by INPO to prepare - semi-annuni' I ,.
' reports'on performance indicators for the nuclearc utility industr/.-
The; data 'thatD was available during the .PATP .to' quantify.'performan@ indicators was i based on varying plant : conditions which complicated ~ trend _
, analysis._- Data from _ steady-state Cosumercial operations will provide trend' '
information that will be easier to interpret and, therefore, more helpful to' ! managers 11n improving plant performance.-
}
L .c A" review of- the, indicators showed that in the Performance Measurements : i Report ;the trends for.Open Work Requests, Requests For Engineering Services; 4 !: (RESs)',' Requisition ' Status. - Non-Q Receipt Backlog, and Surface Contaminated U
' Areas were degrading or ' indicated -impending problems. Management awarenesc.
L of: these_ performance trends was evident f rom the - observation of activities [ in the field:being conducted to improve those situations.. The Maintenance fi g organization has been trying to reduce the backlog of Work _ Requests through- .!
. increased productivity and . improved preparation ' for and; coordination of I
-work. The Engineering organization Lhas been reducing the backlog _ of) RESs f
'I ' through the assignment of additional resources: and increased involvement' of ~
's t .. ..
a lt .
. System' Engineers.- 'The Purchasing, Material Requirements, and Inventory ]
b organizations :have ~ increased their efforts to reduce the additional N f[ l requisitions'resulting from the full power spare parts rampup project. The l L Inventory organization implemented ' a special' backlog reduction group to , h Khandle the buildup of. material' deliveries. The Health Physics organization p .has been emphasizing the decontamination of areas cas problems occur to i
- a
$ . minimize the buildup of these areas as power history increases.
L p Goal ' and. analysis information. on Performance Measurements Report forms was t found lacking.. Some 60% of indicators do not have goals assigned, or goals that are assigned are not specific. Examples of these indicators are ( 70 p- e. --.
+-
1.
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< -07,
';Unpbanid SEf3tiy. Systemn Actustion,9 Repetitiva Tosks e Completed.3 Rat'io! cf '
. .- . 1
],
y
, Preventive i toi Total MaintenanceL Equipment Failure, Out-Of-Service' Control k ,
[R'oom 1 Instruments. Maintenance Performance -Summary Hour Distribution,c Site W '
' LServices Pdrformance Summary l Craff Hour. Distribution, ~ Temporary Modification a
t > ' Status,J Design Coordination Report (DCR),- Materiali Deliveries / Receipt, .and-K , .
,'<4 4 .
i
= . .
i Security.- . Some 65!Loff thesindicators (dolnot0 contain an(analysis ofi data 2 1 I . ,, presented. ? Examples' of these Eindicators are Unplanned' Capability? Loss [ j
,f g IFachor. LicenseN ) Event : Reports (LERs), Procedure ~_ Changes / Revisions, : Open'.
,m -
' ~
1b - Work L Requests , Repetitive" Tasks 1 Completed, p Ratio of - Preventive tof Total- 7
~ .t
, Maintenance. . Maintenance - Performance Factor, Main anance- Hour .( ,
1 , Distribution, Site # Services 1 Perf ormance Summary, ; Craf t ' Performance, Site' ; l : ~, i
. Services- 1Perf ormance ; Summary - - . Craft Bour'-Distribution, ,Temporarf-1 b , ,
- Modification _ Status,- ' Request for' Engineering" Services, Overtime .StatsP i ' Security. - ' and . Volume : of LL' Solid =. Radwaste.. The . SAT recommends that - (1);
} , goals be L finalized and assigned to indicators andi-(2, a description' of. tho' ;
- y '
- significance Lof ? the ' data collected, the cause : of trends, and - corrective ' 1
. actions be inserted in the analysis ' block of the Performance Measurements '
l3i : Report form. . (4059)(LT) JQ .m I a P The SAT . concluded J that ' indicators will - become . more he'1pful in improvihg a ,m , , plantL performance when " additional steady-sta,te . data 11s collected, goals 1' e'stablished..and' analysis-of. data completed. .[ j sb [ ; 3
~.!
d[ PATP Indicators'
- j. .
1,4 + -Performance indicators. were selected from the' NHY Performance Measurement: LRep'rt for use.by the. SAT as one element in the' assessment of the PATP. 'The
~ '
o P
.7 lSATDevaluated the L indicators by -comparing : results against t the historical -
d i , r , : data of other plants.as'obtained from INPO, NRC, and Westinghouse documents.- ,1 . The ~ indicators ' selected- for~;assessr?nt were Unplanned- Reactor Trips,' - n l' . Unplanned Safety' System.Actuations, Licensee Event Reports (LERs),.Open Work i
- , . Requests, Ratio c of Preventive to Total Maintenance, Volume of Low ' Level j Solid ' Radioactive Waste,= collective Radiation c Exposure, Total Skin and i
Clothing Contamination, Whole Body Counts, Surface Contaminated Areas, ' ec
)
- ' Radiological =0ccurrence Reports, and the Chemistry Index.
~ a o l
2 71-l _ l
' { d' ' .
J I, [:c
- u, -
The cos:sement cf c%h of the indicat:rs s:1cct:d f 11 owes'. l
- unnia===d teactor Trina (Finure 2.10 W
- Two unplanned reactor trips occurred during the = PATP. No personnel errors were involved. One of the trips occurred when the module :
monitoring the last 52 of generator windings for grounds de-energized. .' t
/ the -generator neutral relay which actuated the generator primary I protection breaker lockout' relay. No grounds were found in the generator windings, and the concept behind the operation of the 51
[ module- is being investigated. The second trip was. caused by_ vibrations closing contacts of pressure switches for the turbine
}
generator electric-hydraulic control (EHO) system when pc.rer was _, initially raised' above 702. The pressure switches monitor the EHC emergency trip eystem fluid line ad trip the reactor when low pressure is sensed. The switches were relocated to eliminate the vibrations.. > The two trips are discussed in more detail in Section 2.2 and 2.9. l I,.c The average number of trips that occurred in startup programs at other r plants is eight. The two trips that occurred at Seabrook are a d gnificant reduction indicating that personnel, equipment. .and
- ~ procadural performances were significantly above average. !
o lp
' * :Uncinnned Safety System Actuations (Finure 2.10.2)
L . INPO defines unplanned safety system -actuatione as I, , actuations of the safety injection system and actuations of the the sum of emergency AC power system resulting from loss of power to a safeguards I bus. No unplanned safety system actuations, within the INPO 1 definition, occurred at Seabrook during . the startup program. The average number of actuations that occurred in startup programs at other plants'is 3.6. Therefore, a rscord of no actuations during the PATP
-indicates that personnel equipment, and procedural performances ' at Seabrook were significantly above average, t i
Only one unplanned safety system actuation of any kind occurred at ( Seabrook during the testing program and that actuation took place while I' 72 .
+
. yg
- I 1
'[ the plcat was in a hot standby Mod 3 3 status... N:rmal power-to a panel supplying radiation monitors for ti.e containment on-line purge' suffered a capacitor failure and blown fuse. The loss of normal power initiated f an automatic shif t to alternate power. An Auxiliary Operator who was investigating the problem mistakenly opened the breaker from the alternate power source. The . loss of power caused ~ containment on-line purge radiation monitors to fail in the safe /high state which actuated' a containment initiation isolation (CVI). This ESF actuation is also.
discussed in Section 2.2.
,
- Licenses Event Renorts-( m e) (Finure 2.10.3) i Ten (10) LERs were prepared during the testing period. April 16, 1990 through August 17, 1990. The root causes of the ten LERs can be bruken down into the following categories:
, 5 -- Personnel Error S -- Equipment 1 -- Procedure 1 -- Design The SAT assessed the primary root-causes of the five LERs attributed to personnel error as (1) right action / wrong action executions design, (2)
'l action chosen improper. (3) no action when. required. (4) incomplete defective communications, and (5) right action / wrong executions equipment. The primary root causes' of the three LERs attributed .to l,' equipment are (1) defective equipment, (2) unanticipated interactions,
.and (3)' improver environment. Each of the LERs describes isolated I events with no generic implications.
The average number of LERs that occurred in startup programs at other plants is fourteen (14). The ten that occurred at Seabrook are less
, than the industry average indicating above average performance.
I I ,, I
. .. ..i i,.... .. . . . , . .., .i
3 - Ia ' 4 0 Onan Work Reauests (F1mure 2.10.A1
, INPO provides drata on the corrective maintenance backlog in the nuclear.
utility industry that is greater than three months old. The median value of the backlog greater than three months old is 512 of the total Work Requestre, excluding outage work but including work waiting ' for
- plant conditions.
The data provided on Figure 2.10.4 for Vork Requests greater than three R.7 months old shows that 46% of Work Requests during the testing program
, were grester than three months old. The data, however, does . not include Vork Requests waiting for plant conditions. When an adjustment is made for work waiting for plant conditions, the percentage.of Work Requesto greater than three months old rises to 472.
-The comparison of these percentages shows that tho'Seabrook maintenance effort is - better than the median of the nuclear utility industry = in I The limiting ' the backlog of work to less than three months old.
backlog of Work Requests is discussed in more detail in Section 2.3. I i 1 * ' Ratio Of Preventive To Total Maintenance (Finure 2.10,5)
. q INPO provides a histogram of the ratio of preventive to. total
' maintenance.- Total maintenance is defined as corrective maintenance, preventive maintenance, and technical specification surveillance d j
I . testing. The median value of the ratio is 47.51. The ' data provided in Figure 2.10.5 shows that the ratio of preventive i to total maintenance is 14.7%, but the definition of total maintenance used in' that ratio is the total manhours charged to the Mechanical, Electrical. Instrumentation and Control, and Utilities cost centers.
, Substituting the manhours for corrective maintenance, preventive maintenance, and surveillance tecting for the manhours charged to cost W centers raises the ratio to 34.52. The ratio would be expected to be lower than the median for other plants because of the need during the test program for more corrective maintenance than would be required during normal operations. Also, preventive maintenance programs were 74
=
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. i
[, ' being ' developed and finaliced during the testing program which will. l g ' increase manhours for preventive- maintenance. during subsequent months ,
.l and raise the ratio in the short tena. . In the long term, hover.or, the ratio will decrease as more systems are converted to Reliability !
Centered Maintenance (RCM). RCH will result in more components _ being ! shif ted: from the preventive maintenance program and allowed to run l until failure. . The RCH program is discussed in Section 2.9.-. , l- ( r comparing a plant's ratio of preventive maintenance to total ., maintenance with the median ratio for the -industry is not. very '[ meaningful because the ratios depend on the maintenance policies of individv.a1 plants including the extent to which RCH has been implesented. . However, use of the ratio as a relative, measure of the. j , implementation of a plant's maintenance policy might prove beneficial. j l
\
I
- Volume Of Low Level Solid Radioactive Waste (Finure 2.10,6) 1 +
I i The . volume of low ' level, solid, radioactive waste reached 3.7 cubic. meters on August 1, 1990. The SAT considered the volume extremely l small' for a test program. The small volume attests to the effectiveness of training programs and the waste reduction' program. ' The . ' 1990 ~ goal for . solid radioactive waste assigned by INPO to each pressurized water reactor plant , is 213 cubic meters. NHY has set' a _ more challenging yearly goal of 120 cubic meters. That goal will . be - met in 1990 because only five months remain in the year. If the yearly I, goal was broken down into a monthly goal of 10 cubic meters, the ; monthly goal would also be met. 7 .: goal will be a challenge in 1991, however. because a refueling outage will be accomplished after twelve ' i months of full power operation, l (c. I The processing of waste is discussed in Section 2.5. [ 4
- I. .
I , . . . ~ _ - , , _ . . ~ ,, ., , . . . . , ,, , . . - - - . . . . . _ -,-
ni > l I I 1
. )
C' Collective anMation hnosure (Fimure 2.10.7) , i i Radiation exposure . is measured by thermoluminescent dosimeters (TLDs) for doses of record, and the TLDs are read quarterly. The 1990 goal j for collective doseiassigned by INPO to each pressurised water reactor
- is 288 RDt. If the yearly goal was broken down into a monthly gosi'the' l monthly goal would be 24 RM. NHY has'. set a more challenging monthly goal of 4.1 RM. I I The collective dose exposure for the second quarter of 1990 was .2 R W i
,g or some .067 RM a month. Reactor power had only reached 651 by the g end of June 1990 so the effectiveness of the ALARA (As Low = As 1 Reasonably Achievable) program could not be fully assessed. - Dose management, however, could be assessed. The- 200 milliram was ! distributed over a number.of people so that the highest exposure was 15 ! i millirem with one exception. That exception was a 50 millirem dose. f
. received - by a person conducting sr.ubber checks in 'the containment
. building. The SAT assessed dose arnagement as effective. :
Dose control is discussed in Section 2.4. - Jotal Skin And Clothina Contamination (Finure 2.10.8) ! The median value of the annual number of skin and clothing contaminations per nuclear power plant is 163, or .13.6 a month.' j I Seabrook had three skin contaminations during the four month testing ; program.- One case was a hand contamination of 300 cpm from a leaking i ,a. > source, a second case was a hand contamination of 400 cpm from removing
- a glove af ter working in the primary sample
- sink, and the third case :
L was contaminated boots of 3000 to 4000 cpm in the charging pump room. Although the power history was minimal during the test program, the SAT
- ( assessed radiological protection practices as good.
i !I .
>e g-( -%
4-g'
I- - 0 h is Endy teamts ' f Finure 2.10. 9) INPO has established that the median value of whole body counts for a i nuclear power plant is 0. Seabrook had whole body counts of 0 for each month of the test program. Although the power history was minimal and : no particulate - or gaseous airborne activity was detected, the SAT f assessed protective measures for internal exposures as satisfactory.- i i I
- Surface Con +==Imated Areas (Firure 2.10.10) l
[ The goal for surface contamination . should be no contaminated areas outside of the containment building. However, an achievable goal as , power history increases would be 2000 to 4000 square feet, f I The contaminated areas during the test program have been minimal. The , maximum contaminated square footage that occurred was some 400. Areas. , have been decontaminated in a timely manner as'is evident from records , that show quick' reduction of contaminated ' square footage inanediately. i following increases. The SAT assessed'the decontamination efforts in the plant as good. I
- Radioloalcal Occurrence Renorts (RORs) (Firure 2.10.11)
[ I' The average' number of RORs a month during the test program was three.- y Fifty'eight percent (582) of the RORs concerned procedure violations, [ 262 involved equipment failures. 112 inattention to detail, and 52 failure to anticipate conditions. Root causes and generic implications of RORs were sought before corrective actions were finalized. The SAT assessed the ROR system as effective in enhancing radiological controls. JI
- Chemistry Inder (Pinure 2.10.12) '
The chemistry index compares the concentration of blowdown cation
; conductivity, blowdown sodium, and condensate pump discharge dissolved
; 17 I
s s -
u l
- [
ox;. gen t3 limiting values far th:s3 items cnd thin n:rmalices the-ratict. The' median value of the chemistry index ss determined by INP0 . for thi nuclear utility industry is .22. The NNY goal is to maintain - the index less than 1. The -NNY goal -is less restrictive because l Seabrook does not have a condensate polishing system. i The = chemistry index remained above 1 during the test program as ! impr.rities in the secondary systems from preservatives were flushed out j I of the ' systems. The flushing rate- depended on power level and the blowdown . rate,. wnd Ecolochem filter and domineralizer trailere
, j j
substituted for a permanently installed polishing system. l
~
The SAT assessed the control of chemistry during the test program as . f excellent. The index reached a value of 2.67 during the initial cleanup phase, but has been reduced by. chemistry controls to 1.08. Chemistry is discussed in Section 2.6. 4 I ,
~
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. $,$% OE . ,
nas as THE seutseER OF UNPUWedED AUTOt8ADC - JUNE HOURS CfWTICAL 568.5 - REACTOR TRIPS THAT OCCUR PER 7000 HOURS OF. JULY HOURS CRITICAL 671.1 - CRITICAL OPERATION. VALUE FOR UNIT: i TOTAL UNPLAf#dED AUTOteATIC RX TRIPS X 7000 HRS. I UNPLANNED RX TRIP X 7000 HOURS = le.43 TOTAL nub 8BER OFHOURS CIWTICAL 671.1 HOURS CRITICAL L m. o n mug%.wpquwm mywwg54wmxw ww~m n - is j y .i THE COGN%NYGOAL FOR UNPLANNED AUTOttATIC i i RX TIWPS WHILE CRITICAL IS <3 AFTER COGAPLETION OF FULL POWER TEST 50G. {'
. ,o - +
Unplanned Automatic RX TRIPS //000 Hours
.s j 1see
# OF RX TVWPS WALUE FOR UNIT , , _
j i ..- _ Jan '[ Fels tear i
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~~ ~
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-_ PERFORMANCE MEASUREMENTS ~ * ~
~
PERFORMANCE RAEASURE ANALYSIS JULY 16~
$lAWM.M.M MMbn-'y %Ea::2 :.= _
~
Tees esosCATOR IS DEFBIED AS THE SUM OF THE . NO UNPUWINED SAFETY SYSTEM ACTUADONS l FOttOmets SAFETY SYSTEM ACruAnONS: HAvE OCCURRED.
-THE NUMBER OF UNPUWINED 9 ACTUADONS THAT RESULT FROM REACHW8G AN ECCS . . . _ ~
ACCTUATION SETPOSIT OR FROte A SPURIOUSI
- SGADVERTENT S SIGNAI.
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4 POWER SYSTEnd ACTUATION THAT RESULT FROts
=;
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- PERFORMANCE MEASUREMENTS
' c PERFORMANCE RAEASURE - ANALYSIS JULY 19ee issdiflAh M.W6MMMir sicia sa.- LER'S AND PERSONNEL ERRORS WILL BE . l THE NUtfSER OF LICENSEE EVENT REPORTS ANALY2ED St THE AUGUST REPORT SEfrS) fuiPORTED TO THE 80UCLEAR REGULATORY l C C ^^ _ 3 EACH400 NTH. i i' s 900TE: TIE DICENTWE GOAL IS BASED ON LEft'S
- DUE TO PERSOpWEL EfWOOR. TDW SE30CATOR
! ftEPfuiSEffTS ALL LER'S. t i o i a neSc9AF - *TAisismensGAL=ewerSe roe r mnu N t'l ! , u
~ TO RIASITA50 TIE TOTAL faut 00ER OF LEft AS LOW AS POSESLE. .
THE CORPORATE GOAL IS LER'S CAUSED BY PERSOf0NEL ERROR 1 TO BE LESS THAN OR EQUAL TO 5 P-i V' i ' LICENSEE EVENT REPOHIS
=
! 1900 LERS TO DATE ' P.E. TO DATE "- , se - I { t SW - JER S - I S Feb 4 .".- S - 3Aar S S
- Apr 11 0 i temy 13 1 .";
l Jun 13 1. .- l Jd IT l 3 l~ GOAL IS < OR = 5
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- SEABROOK STATION PERFORMANCE MEASUREMENTS PEfFOfMAANCE REASURE ANALYSIS JULY 1939 d F M E E W Y sidiflij M _6 h iams & h a THE FOLLOWWIG IS A SREAMDOUUN OF Uf0RK REQUESTS ^
TfGS REPRESENTS THE TOTAL OPEN UUORK > 3 000NTHS OWERDUE SY PRIORITY: F' .7 30 THE PLANT EXCLU0eeG TfE
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' " ' AG PLANTCODENTIOpeS PfW3 314 PRI2 23 -
/. GEUUOfut c* M DE9GN ENNMSp THE FOLLOWWW8G IS A BRJAMDOUUN OF THE l i
/ERUUOfutesrusMT TOTAL OPEN 950fWC REQUESTS BY PRIORITY. '
! i PRI 1.E- 2 , PRI2 150 1 PRI 3 . 614 - o l_ e n @dwegspwp M v~>-s qrg m upv ww -we =4 set p , j- SAA'NTASI WWOfWC REQUEST m m SELOUW 883 AT TDE r8 END OF EACH GUARTER. (750 + 15% TOLERA00CE) 4 U i i *- OPEN WORK REQUEST l ! WUOleCIEQUEST "*""*'"**"""* ' i
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SEABROOK STATION - _ F PERFORMANCE MEASUREMENTS
~ ' ~
PERFORIMNCE RAEASURE - Af0ALYSISJULY 1998 " EMSM.aeudAO4SE1 . THE RATIO OF TOTAL tAANHOURS WORKED ON PREWEBMWE AAANTENANCE BY AGECHANICAUELECTRICAL ISC Ape UTILITIES TO THE TOTAL SAANHOURS FOR TDESE COST CENTERS. w
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31 CUSIC RIETERS OF WASTE HAS RECE8WEO PACMAGeeG _ THIS BWICATOR IS DEF90ED AS THE WOLUBE OF LOW-LEVEL SOLD fnananarTIVE WASTE THAT - ' HAS SEEN Pgancramsn Age IS W FWGAL FOftf . 5MEADYFOR DN
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=
pggpoggg8A00CE aurat884E Af6ALYSISJULY ISOS NWBeOLE 90DYCOINGTS . . . THERE HAVE BEEfd 900 POSITIVE UUHOLE SODY
~
THE PURPOSE OF THE POSITfvE WUHOLE 80DY COUB8TS DUR50G THE 48000TH OFJULY. COL 9fT SDICATOR IS TO 880BSTOR THE - - - EFFECigWEBWiS8 OF PftOGftAAAS FOR LANT90G THE RAD 84TIO98DOSETOWORKERSDUE TO " seTEfBGABAYDEPOSITED RADIDACTOWITY.
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. THE STATIOpd 90AS 900T ESTAsai anaED A GOAL POR -
POSITREWantsecoVCOtssTS. o-Routine Whole Body Counts e, 1989 800NTIS.Y Jan 0 Feb. 9 l: Star O I i Apr 8 f geny e i! - Jun 8
- . l sep On 90er ,o -- , - m - ,-. , - - - ., , , , - - - , +
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$165i.?dMWWWWli?idh.i * ~,s .
AseALYSISJULY esse THE AAGOINGT OF SQUARE FOOTAGE POSTED THE AtsOUNT OF CONTAMBeATED SURFACE AREA 50NASS B froes 25 SF TO 3M SF.DUE TO ADDITIONAL AS CONTAAW84TED ed THE PLANT DUFaseG ROUTOGE , OPERATIOps EMCLUDesG CONTAINhAENT. RADIOACTIVITY De THE PLANT AS WWELL AS - IdWOOR 50 CREASES se EQUIPedENT LEAfuGE ' _ r i I
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- e t SURFACE CONTAMINATION AHLA
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~
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' AgentySgS Jidy's~gge A$$N W M M id n L.a.3 :
THEIIUtWEROFmnarsa OCCURRASICE SUtsb8ARY OF EVESIT CATEGORY , . hse 50005 GENERATED EACH 800 NTH II PROCEDURE WIOLATIO90S 5 EQUIPIN!NT FAILUfMES 2 BGATTENTION TODETAIL f 1 FAILufMETO Af0TICm4TE COOGITIOpsS - 19 TOTAL OCCURENCES
~
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' I j j, llllg l l imanum 90 u __1 _ _ ._2 _ _ . _ . . . . _ . = . ...-_.__........._m.... ._ . . _ . . . . . , , .. , . . . , , - , , .
i R$11' ETAIJETIM OF MMAABIENT EFFROTITMESS Summary Management' effectiveness and influence ' on the safe and efficient' operation .
, of the Seabrook Station was evaluated by observing field activities and meetings reviewing pertinent memorandums, letters, and reports: and-interviewing selected . managers' and supervisors.- This evaluation is l discussed under the following topics: Power Ascension Testing,- Plant' 1
Performance. Management Monitoring, Management' Assessment, and Management -i , Control of Plant Configuration, I The conclusion of' the evaluation is that management is effective in j directing safe and efficient plant operations. ! q f' Summary Of Findina y W 0, -- Testing SAT Reconsnendations (ST) O. -- Commercial Operation SAT Recossnendations (ST) j 0~ -- Long Term SAT Recommendations (LT)' [ t l ,' I Power Ascension Testina And Plant Performance f [ ) 1.-- lThe SAT continued participation in the daily Station Managers meeting throughout the 75, 90 and 100! power plateaus. These meetings primarily . bWS addressed the ' current plant status, including chemistry, status of testing i and required' actions by departments and groups to support Operations and the 1 l ; Power Ascension Test Program. The SAT observed the discussions. that pertained to adjustments to the Feedwater regulator valves, Feedwater Pumps, , l
' Heater Drain Pumn rH acker;;e valves. Hoisture Separator Reheater and Reheater I' Drain Tanks. The discussions at the morning meeting provided a status on j t
the effects to fine tune the controls and dynamic performance for these i systems. -The SAT- also observed the working meetings between Technical Support, Engineering. Instrument and Controls Department and Westinghouse to ; identify. and analyze the plant performance and to develop appropriate . corrective measures. The SAT also reviewed the desj3n packages that . 91 1 p , 7 m . . _ , - . , . . - - - . , . . - ..,,, a
-. _. . - . , - - . . - . . . - - , - . . ,,m, ..m. _ _ . ,
m, ,
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enhanced ' th) contrcl mechanisms; fer improv d' pl nt perftraatc3.- On : July
;[
s T '
=14,:1990.the SAT reviewed the status of the following modifications with tho' , f Management Oversi5ht Comunittee.
4
- Adjusted the gain setting on the Main Feedwater Pump speed controller.
i
-- Adjheted the gain setting on the Feedwater Regulator Valves. 3 i
-- . Installation of a new cam on the Heater Drain Pump discharge valve.
3
-- Fine tuning. controls on the feedwater heater string. 1 LIX The SAT also observed'each SORC meeting that reviewed Power Ascension Test program test results. Although many of the SORC members had prior first-hand knowledge of the test results, SORC processed these presentations as-
' walk-ins'. This - required that at each meeting the Test Group Manager j
*o present the results for each test, review the acceptance criteria _ explain , l
. . . t test exceptions and their basia and indicate, where applicable, concurrence' :i by Westinghouse.
l SORC followed- the review of each test with specific pointed ~ questions to ensure an appropriate and detailed review. In October 1989 the SAT 4 t I . performed a 'special review of the SORC process. One of the results from j this: review' was the conclusion that SORC eafety reviews of ' walk-ins' were l l , thorough and rigorous. The SAT observation of 8020's PATP results review 1
~
t was' consistent with the October 1989 review and conclusion. ~*
; L <
- .The SAT also observed frequent plant and test program status briefings between the Station Manager, Executive Lirector Nuclear Production and the -'
;[ Senior Vice President. Throughout the entire PATP, the Executive Director Nuclear Production attended the Station Managers morning meeting. On a periodic basis the Senior Vice President also attended this meeting.
The . SAT- also observed a sample of the Plan-Of-The-Day ( PC s ,' meetings and reviewed the daily POD schedule. The schedule and working meetings focused
- 2
-on upcoming plant conditions and testing and the required supporting work 7
, efforts by Station groups and departments. The POD schedule and meeting was an effective means of focusing work efforts in support of plant operations 92 t
' it
r m 1 l,,t *V" and t: sting. The SAT cbserv::d . Station managements ' participation '13 - the s
]
a
' turbine restart af ter. ST-38,. Unit Trip from 1001 power. After completing I > Heater Drain Pump- discharge' valve cage modifications, snubber inspections and miscellaneous maintenance, the unit was synchronized to the grid on July =
l { 31, 1990'. - During ' this' restart the number 7 bearing began ~ to exhibit ] increasing vibration. and the turbine was tripped. Operations, Test . and
]
Technical Support management reviewed plant data and presented the results { and corrective actions to the Executive Director Nuclear. Production, Station i Manager, and. Assistant . Station ' Manager. Af teri rebalancing ~ the turbine ) .s , generator the unit was successfully restarted'in order to perform ST-39 Loss j i N 4 .of Offsite Power. This course of events'with analysis, corrective actions .l 5 and management approval was typical throughout the PATP. !
~0n July 14, 1990 the SAT briefed the Management Oversight Committee on the status of testing at the 752 power plateau. Memorandum (IRT #90-050)-
details the agenda, test-results review and meeting minutes. The. Management- 3 oversight ' Committee authorized continued testing at the 90 and .1002 power .g plateaus. ~ Based on these first-hand observations, reviews of.the daily POD
.' schedule and design packages the SAT concluded that management at all levels y
;in NHY was effective in directing the Power Ascension Test Program, i
k Mananement Monitorin l L
- -l The . SAT observed several ~ examples of . monitoring by management -in diverse areas affecting plant operations. The SAT observed a meeting conducted by ;
the Station Walkdown Coordinator. This individual reports directly to the [ ,i L Station Manager and is responsible for administering the walkdown program described in the Seabrook Station Management Manual. The meeting reviewed , L I:'. the program and the - status of ~ efforts for implementing the program as defined. The intent of this program is to require plant managers to ' g routinely tour the Station and provide written reports to the Station Manager describing their observations. Although the program is relatively new, additional attention and accountability are necessary to obtain the ; expected benefits of this program. r I'
=
The SAT also reviewed the monthly NHY Performance Indicators and Management . Controls Reports which track the efficiency and distribution of resources in r 93 e w , - . - - . . , -.m.y,-- .- , , , , , 8
r , e . p , p ,
' i
[ , the Nuclear Production subdivision. Th:s3'r: ports'r0531v3 wid3 distribution [* $ tthroughout NNY and provide management with the information necessary to i monitor key performance factors. Beginning in September, the Senior Vice. , { President will conduct monthly meetings with his Executive Management; Team
~
to review these performance _ indicators. i
'The SAT observed the following meetings:
/ .-- The status of developing Reliability Centered Maintenance. {
-- The status of the Materials Task Force efforts presented to executive i management.
'.-- The-results of a Station Configuration Control Task Force presented to Station Management and the Executive Director Nuclear Production r
.j.
n , The SAT also reviewed the Station Modification Review Conunittee -(SMRC) [ reports, the proposed five-year schedules and business plan, the-efforts of-L 1 the Root Cause Analysis task force' and the status of Nuclear Plant i y - Reliability Data System - (NPRDS)~ information transfer to INP0.~ -Mansgement
- l:
, ' involvement. and attention to the status- of these efforts was. clearly. a 4
evident. h Based upon these first-hand direct observations and reviews, Executive and
^ Station management monitor the progress and performance of- on-going. and L
=special activities.
1
. Mananement Assessment i
- The SAT observed that management is involved in identifying the root cause '$
4 of problems and supporting timely corrective action. The SAT has observed L . direct Executive and Station management attention on the development of a ,
,, root cause analysis program applicable to Station Information Reports (SIRS) f and Operations Information Reports (OIRs). This program, when fully
[( ,1 developed, will also encompass trending to reveal generic areas for [ l i ( improvement. Although this program is in development, the SAT observed "
. close management attention to significant events. On August 1, 1990 the 94
l, , Health Physics Department Supervistr was conducting- a r:utin3 ' management. l tour. :His tour included verification of High Radiation' Area access.' During , j the tour he observed the ' door to the Domineraliser: Room was closed but not , q Iccked.. This door was the same door involved in LER 90-020.. The operations Manager raised this topic at the Station Manager's August 1, 1990 j morning meeting. After the initial discussion by the Station Manager, j Chemistry and Health Physics Manager and the Security Department Supervisor, l tho' Executive Director Nuclear Production directed the Chemistry and Health
' Physics Manager to determine the root cause of the event and to coordinate i
I appropriate corrective actions with Security and Engineering. The Executive l g Director' Nuclear Production also directed the Chemistry and Health Physics ' ; 3 Manager' and Security Department Supervisor to determine the possibility of-generic implications resulting from this event. ' l During the review of ^ the Feedwater. Heater Drain and Moistere Sepa rator ' I Reheater. the . SAT observed Technical Support and Engineering contacting' ] l; other nuclear utilities for performance information and operational ! l ~ characteristics of similar components. These groups also directly involved the component and equipment representatives in the analysis and development :1 of Corrective Actions. I I 1 I' .The SAT reviewed the proposed revisions to the Operational Experience Review Program (OERP) and interviewed the individual responsible for this program. ) As: a . result; of attending 'a recent OERP workshop and contacting other I 1 utilities, NHY is significantly ' revising the existing program to provide 'J additional depth and specific reconsnendations to operating departments. J L l l Based on these observations, the SAT concluded that management tracks ~l l . corrective actions through completion, provides appropriate attention and - L l direction to problem areas and incorporates lessons learned from external . industry sources. J Management Control Of Plant Confinuration The SAT assessment activities for this area of management effectiveness included reviewing the draf t report of a Station Task Force assigned to 95 '
review c:nfiguratica contral cnd ottendance at the Task F:rce proscatation
'to Nuclear Production management. The Task Force analysis succinctly I identifies and defines areas for improvement. The SAT concurs with the results of this effort and the increased b6.nefits realized from line organization. recognition and identification of areas that can be etrengthened..
The SAT also reviewed the efforts by the Materials Task Force in the area of consnercial ' grade dedication. Unit 2 retagging of spare parts, initial ~ ordering andL subsequent . reordering. The SAT ' activities included a progransnatic review, trend report review, interviews, review of the Material I Task Force status -report, review of work requests awaiting material-availability and ' direct observation of material' control in the field. The-SAT concluded that the overall control of the materials process ensures that qualified materials are inst.alled in the plant. The SAT did observe that continued attention to the_ areas identified by the Materials Task Force will improve the efficiency.of work in the field. The SAT observed the controls for plant modifications and ae st cisted decision processes. The Station Modification and Review Committee (SMRC) established a five-year schedule for the engineering and installation of plant' modifications'. This process also included a review of the priorities for Requests for Engineering Services (RESs). The SAT concluded that . the
.SMRC process provides appropriate control over planned changes to plant
. configuration. ' Minor plant modifications identified in the Station Manager's daily morning meeting are controlled through the Design Control I Program and receive SORC review and approval prior to installation. The SAT review and observation of work request packages for these minor modifications indicated that additional attention to the preparation of I a Design Installation Packages is necessary.
) The SAT also reviewed the status of the vendor manual program' implementation. The program is clearly defined and will result in issuing controlled vendor manuals that reflect the latest vendor information. The procedures to solicit, route and approve vendor information has been approved and issued. NHY has begun to upca <nd reissue vendor manuals that reflect the intent of IE Bulletin 83-28.
96
F l; , - l 1:
.a ', ,
O'
' Based . on th o) che:rvatigne Jcad o specisi - rcview' cf Maintonat:0 tsing the I
i NRC Maintenance Guidance, ' the SAT concludes that NNY has developed _ basic ; n ,
' configuration controls for plant. systems, structures. and components. ;
l~ , Management is attentive to the need for additional improvements to ' ensure excellence in this area. r I f L
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' S 3 '. 0 SIRAG&RY OF PEASE 21RCQbedEllMTIONS
[
'i' , l
.g Short Comunercial Long m 5 "'tia" ^'** h " r**1a"' Issa . I Power Ascension Testing ~. 0 0 0 0
- l 1 Operations.' -
, .0 0 2 2 j
' Maintenance And Work Control. 0 0 '. 11 11 l t
I
- Testing.'and Surveillance 0 0' 0 0 :
' I; '
' Radiation Protection. 0 0 -11 11 ;
Radioactive Waste' O O 6 6 t i'
' ' Chemistry. 0 0 1 .1 I I Training 0 0 >0 0-
,1 l-Quality Programs 0 0 0 0 l
[
; Engineering and Technical Support 0 0 5 5 'i r i Performance Indicators- 0 0 1 1 i i
i c l-l l; , Management Effectiveness 0 0 9 9- - Special Event Reports 2 0 11 15
' TOTAL 2 0 57 59 NOTE: Seven (7) of Fifty-Nine (59) Rec - dations Have Been Implemented .,
i ( l; -. . ____.________i_______._________._.._..- ., . . . . . . , _ _ . , _ . _ _ _ __ . _ , ~ . . . . _ . . . . , , _ _ _ , . , . . _ , , , a
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POWER ASCENSION PHASE 2 -- RECWWElWATI(NI'IlWEX . _(ST = Short Term, L CO = Commercial Opere, tion, LT = Long Ters) FUllCTIONAL CLD5EUUT ' AREA 0F-EVALUATION ST CO - LT MANAGER . RECEIVED ~ NUMBER DESCRIPTION-4001- Prepare An Operations Depart-r nt Instruction For The Radio X J M Grillo- Completed Radio Report System 4002 Incorporate Radio Transmission Monitoring In The Operation Department Instruction Radio Report X J M Grilla Completed 4003 Install Labels On Radio G J Kline -Completed Radio Report X Communications Devices 4004 Train Operators On The Recourse For False Communica-Radio Report .X P M Richardson -ICTS tions 4005 Upgrade The. Guidance For Control Point. Operators Radio Report X J M Grf"o - ICTS 4006 Designate Control Point D G McLain Completed Operators For Site Senvices Radio Report X 4007 Issue A Site Services -X D G McLain ICTS Procedure For~The Radio System . Radio Report 4008 Centralize Radio System Radio Report X T A Maynes. ICTS Administration i 4009 Re-Evaluate The Work Control X _8 L Drawbridge ICTS Supervisor Staffing Operations' l-
~
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.FUNCT10NAL~ ,CLOSECUT:
- NUMBER' DESCRIPTION ' AREA OF EVALUATION- ST .C .LT MANAGER ? RECEIVED
.. - ..ma c.. 4010- Operations Management' Presence ..
L In-The-Plant _.- Operations' X. lJ M Grillo ~ ICTS E _ ; p
,a 4011' Reinforce' Work Control- -
i Administrative ~ Requirements Maintenance ~ X R E Cyr _
.. ICTS t
~
- j 4012 Evaluate Need For:A' Designated Chemistry And Health Physics -
Planner Position'In The - Planning And Scheduling. . Department Maintenance X- 'R J Sherwir ICTS r k l l 4013 Develop A Policy For Periodi- ' cally Reviewing Outstanding . Work Requests Mainf- ed .X R J Sherwin ICTS.: 4014 Develop Work Request Scope - Change and Rework Performance Indicators ' Maintenance- ?X R J Sherwin- ;ICTS 4015 Provide Refresher Training On .R E Cyr/ /i Work Control Procedures
- ~ Maintenance X P M Richardson
- ICTS-l 4016A Develop And Incorporate -
Failure. Codes For Work' Requests . J And Incorporate In MA 3.1 ' Maintenance _
' X .- R.E Cyr; .ICTS .
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. NUMBER ^ . DESCRIPTION:
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~
c The_ Health Physicist:-:ALARA 14024
. s n=
should Report To The~H.P.; - N{ J J Rafalowski
~
-Radiation Protection ~X- ICTS g DepartmentiSupervisor
-c ..
4025A Increase Worker Awareness Of E The ALARA Recommendation . .
~
PLM Richardson / - , J J Rafalowski: IC?S1 Process ' Radiation Protection .X_ 40258' Increase Worker Awareness Of The Location Of The ALARA .P M Richardson /
'J-J Rafalowski, 'ICTS:
Radiation _ Protection X Suggestion Box 4026 Radiation Safety Committee Chair Should-Approve Jobs In Excess Of 25 Man-Rem Radiation Protection X LJ J Rafalowski ICTS. 4027 Revise MA 3.1 to: Facilitate'H.P. Reviews Of R E Cyr/ 4027A Work Requests Radiation Protection X :J J Rafalowski ICTS-4027B Reinforce Work Group Supervisor Responsibilities'For The R E Cyr/ Radiological Aspects Of-J J Rafalowski Radiation Protection .X ICTS Pre-Job Briefings-4027C Require Work Group Supervisor- i Participation In Post-Job
- R E Cyr/-_
ALARA Reviews Radiation Protection Xi J J Rafalowski: .ICTS 4027D Reflect The'One (1) Man-Rem - ..
,R:E Cyrf . _
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Xi TJ.J Rafalowski, ;l:
-ICTS-Unit In RP 15.1 Radiation. Protection S-m3,,
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FUNCTIONAL: '
- CLOSE0UT _.
NUMBER - DESCRIPTION ' AREA 0F EVALUATION ~ - ST :CO . LT MANAGER . RECEIVED .- "D 4028 Include The Chemistry And: > Health Physics. Manager 10n.The-Station ~ Modification And Resource Committee : Radiation' Protection.. ~ X. W A-DiProfioL
. Completed- -
4029 Expedite The Procurement Of , The Steam' Generator Mockup Radiation Protection X" P M Richardson-
~
ICTS:
.=
4030 Reference Plant Photos And _- Mockup Training In RP 15.1 Radiation Protection :-X' J J Rafalowski'- ICTS 4031 Use Reusable Cloth Booties kc-aste 'X . J J Rafalowski/ Anderson, ICTS 4032 Enhance Tool Decontamination Equipment Radwaste 'X H M Anderson- ICTS. q
~'
4033 Enhance Facility Decontamina- _ _ . tion Equipment
~
i '- Radwaste- X~ H M Anderson ICTS t 4034 Review EC 5.~4 or Maintain Cask Books As Required Radwaste X .. H M Anderson ICTS. 4035 Incorporate Radwaste .. Responsibilities In The E J Sovetsky/. . J Station Management Manuals Radwaste' X- ~R E.Cyr- Completed 4sma. w .-
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NUMBER - _ ' DESCRIPTION. _ = . AREA DE EVALUATION: . ST , CO ~LT MANAGER -RECEIVED' - _ _ , L.' -
.a.
4036 Initiate: Periodic Plant.' V Visits, External ~ Training And. ^ Attendance ~At-Conferences- 'Radwaste> -
~
~
X .. ;H M Anderson ; ICTSl -
~
_.-' ._-, -7 4037 Replace Secondary In Line J M Vargash . Chemistry Instrumentation Chemistry X' J T Linville'
. ICTS-m 4038 Conduct An Annual Review ' JMVargas/-
Of Open RESs . Engineering.& Technical. Support X- - G J Kline" -'
.; ICTS-i 4039 Expedite The Installation Of Permanent Designs To Replace- J M Vargas/
Temporary Modifications Engineering & Technical Support -X G J Kline - ICTS-L I a 4040 Integrate The Multiple-Goal- ' Programs Management Effectiveness X _ T C;Feigenbaum. ICTS. , l 4041 Establish A Quarterly l Meeting For Functional'~ Area ~ . Reporting
~
j ' Management. Effectiveness T C Feigenbaum
~
'X ICTS) l - _
l 4042 Redistribute NHY Resources To-Support Operations Activities Management Effectiveness X T C Feigenbaum' 2ICTS - h
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FUNCTIONAL -CLOSEUUT. L NUMBER DESCRIPTION AREA 0F EVALUATION' . ST CO -LT ~ ~~ MANAGER ~ RECEIVED" se L 4043 Implement A Root Cause . .. . . - Analysis Trending Program
- t. Management Effectiveness X - B L Drawbridge ICTS-4044 Integrate Multiple Performance . . . . . .
Oriented Programs- Management Effectiveness. .. X 8 L' Drawbridge; -ICTS-l 4045 Review-And/0r Enforce The L Procedure Compliance Policy-. In The Administrative Area Management Effectiveness X T C Feigenbaum' - ICTSL 4046 Increase Management And B L Drawbridge / Supervisory Presence In The _ _ R J DeLoach/-- Plant Management Effectiveness X N A Pillsbury: ICTS; 4047 Minimize The Number Of Diverse Scheduling Systems Management Effectiveness X B L Drawbridge- ICTS-- , Develop A Single Or Integrated 4048 i Configuration Management And . B L Drawbridge-Controls System Management Effectiveness X R J DeLoach ICTS-4049 Line Management And Partici - pants In The Event Reduction Process For The Turbine Trip Should Conduct ~A Self Critique Turbine Report X E W Desmarais- ICTS
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- FUNCTIONAL . CLOSEGUT1. 1 NUMBER DESCRIPTION : AREA 0F EVALUATION :ST JC0 LT - MANAGER ! RECEIVED: '_
'4050 Perform Root Cause Analysis.
Per.NHY Procedure'12810'For 4
~ Radiation Data Management .~_ .
<[
All Station Information Reports : System Report .X -D.E Moody ICTS
~
~'
_4 4051 Re-Establish The RDMS User's . GJKline/RECyr/- Group RDMS Report X WBleland .ICTS-4052 Revise. Check Source Actuation From Daily To Monthly To .G J Kline/ Match Technical Specifications
-RDMS Report X J M Vargas- .ICTS-
-4053 Enter Manhole Water Intrusion Commitments:On ICTS Manhole Water Intrusion Report X J'M Peschel- ' Completed.
4054 Review Document Control -
~
Revision Processes For Assigned - I Responsibilities To Avert:A ., Recurrence-Of Similar Event. 1 Maintenance -X :D E Moody ICTS g l 4055 Complete Comparison Of Readings - - l On'RDMS Monitors (RM-6534, ~ RM-6540,andRM-6545):ToSurvey ~
~
Meters When Radiation Levels . . t
- Permit Radiation ~ Protection X :JJRafalowskii lICTS1 4056 Prepare A Chronology of Actions Taken to Resolve Feedwater-And Heater Drain System -J M.Vargas/s Oscillations' Engineering & Technical Support X. G~J Kline - -ICTSE 4
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L-e NUMBER: - DESCRIPTION - AREA:OF EVALUATION: : F- .:CO LT : MANAGER -
< RECEIVED?
.2 -
g 4057. Review Progress.For. Creating .. y ~ 5 Field Work Packages Engineering & Techynical Support LX iD ~ E: Moodyf.; - !ICTSL v^~
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4058 Conduct Training Session For = System Engineers'On Errors. '
~
4 In Closed Work _' Packages.And -
~
~
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In Progress fcgglggs, Igf,gl i A2.1)(Commitments.
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, !2.21 Physical Plant ' Readiness
- 1 J6. 4- 10.. j m
1 p'. 9 e I
- 2'34 . Power. Ascension-0 0:
'o - 0 .. 51
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,4C. - f '. . .
i ,
,I ?' 2.4J Operations: 1.
2- 3 , ?l [ .): n:, f 2.5, -Radiation Protection .2 6' 8: g
- 2. 6 ' Chemistry. '2 2- 4 9
.5
. i.
I:; 2.7- l Maintenance- . -3 1-' !4' ' o L- r x); L " u , 2.8 Emergency Preparedness 0 0: '0 9
. 1
- 2.9 .
Quality Programs 0 0' O q 4- . y g
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2.10 TrainingL l0 0 0- E n : m.
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.' 2.11 Management' L Effectiveness ~3' 7. 10 s
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-TOTAL 18 23 41 -i 1 ,
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@Nl@A[.. ; s .. l, Q , ,' > j Anoendix A
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9$m . w:d. w s j p M ~, ST TEST RESULTS REVIBf AT 751 POWER 3
>'j 1, ay JN n i
jg , } . Presented To The Mananament Oversiaht committapg , j W $, On July 14. 1990 li T
'O{ , , !
s / e. 6 .. U-4L 1'. . ST-13 ~ OPERATFRAL ALIGIttRNT OF NUCLEAR INSTRIMENTATION g m p ' 4 obiective ' M To. ensure proper, alignment of the Gammametrics and Westinghouse.. nuclear- ;i' Tg ' instrumentation systems prior to proceeding;to each of the major:
,g m plateaus'- .
Results j Li;x ' Obtained overlap data between intermediate range channels and power t L range channel detector currents, extrapolated to full power detector- ;
,3 - currents.' The extrapolated currents at 100% were noted to be 20J g microamps higher than the results from ST-36. The NI detectors were
~
re-scaled based on ST-36 (AFD Calibration) data. Results satisfactory. m .
~.No test exceptions were generated at this plateau'and only one:(1) test
.change was' generated which was cancelled.by the Reactor Engineering'and Computer Department Hanager.
ll 2.c ST-14.1- OPERATIONAL ALIGIGENT OF PROCESS TEMPERATURE INSTRIMENTATION <
- s =Obiective-
. Ensure proper alignment of the delta T and Tavg instrument channels at- 1 J '
each of the required test plateaus. j Results Sg: . 0btained data from ST-26. Data from the-0, 30, 50, and this plateau o: i , was evaluated and extrapolated'to 100% power. This. determined that. L{J- rescaling was required to allow the plant to operate more. closely:to
~';
y: 4 its design. Loop 2 narrow range cold, which was out of tolerance =at- :j
- the 30% plateau remained reading OK.at this plateau as it did at 50%.-
h The exception-.' generated at 30% will remain open so that it.will continued.to be monitored to assure its reliability at the higher L - >
-4 plateaus.
- l ;
u One test exception was generated to require rescaling of Tave and one La ; test change was issued to adjust initial conditions to more closely match those of ST-26. I e p d ( f' e A-1 4
. ., )
I, ll, '?;
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! e F "3O ST 25P STEAM GENERATM AlmRi& TIC LETIL CONTE 0L u s ,
m l q.
.Og ,
s
-r u
',,0biective> '
9 . p .. . , - . - . i
-.f #m.c' - Demonstrate the stability and proper operation of'the main feedwater:
s
,l og <,- . pumps speed ~ control .
g,x ". , j Results > !
-All data taken for the.752-plateau was found to' acceptable.
~
tThere were no test exceptions taken at.this plateau and only:one test? change'was generated.- This change provided a more accurate method ofl .! A I 7,1 - measuringimmin steam and feedwater. delta P's. , . i o , J e j- , g '
'4b ST-26 THERM &L' POWER HEASUREMENT AND STA1EPOINT DATA COLLPe?L74 51 . .. ..
I
. Obiective ' g To determine core thermal power by calorimetric determination.and' to m m i verify the performance parameters of the main steam and feedwater' _ , ; -. .m.
i Leystems. Additionally, this test will collect other process data to- ;
, , ? support the-performance of other'ST's and establish baseline operating '
- , ;datai s :p *
, 'Results '
E. , J The test'successfully collected all data required at the 75% plateau, The test' indicated a-difference between the precision and manual; heat l~ l D ' balances which-is attributsd to timingLdifferences inTdata collection i L and' differences;in FW flow values used for each analysis. RCS flow was
-also determined to be acceptable. '
.l L.1 ; I >
There were no test exceptions generated at this plateau and only one. y
.(1) test: change.'~LThis test change would' allow an extension of the 7
]
eJ g :. ' day eclibration interval for.this test for.: steps 3.11 and 3.12 for 7 [g. feedwater' flow, main steam flow, and-main steam pressure transmitters g . required for this test. ! u o
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;5i - ST-27 ' RfART-DF ADJUSTMEET OF .TER.ERACTOR CONTROL SYSTEM
$<T, m .- -
Q' , c$.'f ,P Nective'\ D
+
l Al] - ToEobtaid and evaluat' e data necessary to' determine the Tavg program" that'wllliresult'in the; highest possible steam pressure to assure O M7 ,m
~
m ! I5,f T optimum plant efficiency, while maintaining pressure for the.. turbine' ; J " and.Tavg within requiredilimitations. This' data willbe the basis for 5' , y.; , < any'adjuatments to the reactor control system (Tref or Ta7g-
' programming module).- ,
t 4gy ' , ResultsE
, (,
-This. procedure has-not bee'n fully-completed as.of this-meeting as
control: adjustments are in-process to bring the controls closer,to: j J their design' requirements as required by the. procedure. There.'are no' other open issues.with this' test'that would prevent the Station:from- i
. going to the 902 plateau after SORC approves the' final results.
t 4
.g (. . - -
yg E '
- . . J p '6'. ST-28 G&LIBRATION OF STEAM AND FEEDWATER FLOW INSTRUMENTATION a <
Obiective' This test col'lected specific main steam and feedwater flow data, under' e operating conditions, to be used for correcting.ato a greater accuracy,- t
;the calibration values for the main steam flow transmitters and'also
. verified against-feed flow data obtained from ST-26. ' 'i 1 L Results -f l ' . . .
L Steam flow and feed flow mis-matches vere within tolerance required by , the pro'cedure. . It was noted during:this review that the. steam flow - z
. mis-matches between two transmitters;in each-of two channels, were- i loutside-the. required tolerance of 50,000 pph steam flow. lThe ;
s transmitters involved were 522/523 and 542/543. - Westinghouse- i I '
~ recommended that no respanning take place until evaluation at the 902 plateau.
't y
One test exception was generated to address to address the steam flow ,
-mis-matches noted'above. Two test changes were generated-at this j A
plateau. One change'was voided before going to SORC and one test-change put off any respanning until re-evaluation at the 902 plateau.
}
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% };g, ' L7 h ". .ST-29i CORE PERFORMANCE ETALu& TION 1
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2 0b'iective '
)
IM'% i
'Thisft,est.willverifyproperreactor. core' performance)by' obtaining
]
n'w', .incore. flux..and thermocouple maps and: analyzing data lin accordance with ^l ry" *
? approved Reactor Engineering procedures to validate' nuclear physics j
.pe rf ormance . -
.i y+>
J ,Results
~I W;~ y t' *> The results of.the'75% plateauLflux' map indicated the: measured core * , g+ #
performance parameters met the acceptance criteria. . It was noted that , the;value for peak Fry,= hot channel factor,'was withintthe 752 criteria J. f
.but was above the 1002 criteria. This'will reaufre a flux an6 nrior to 4 *m.- , 962 nower nasr Tech Snec TS 4.2.2.2.d. .This situation occurred at the-4
'502 plateau as.well and additional flux map was-performed by the use of r , a Reactor: Engineering RTS. ,h-90 WOO 3455 was issued to, check a thermocouple (H-03) which is' reading-low..
t as -!
-There were no' test exceptions or test-changes generated at this plateau I U4 for this test.
b I, y'
~
- (
)
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,' . ST POWER COEFFICIENT MEASURMENT a
Obiective ,t To verify. nuclear design predictions of the Doppler-only power l coefficient through correlative measurements of the reactor coolant ' system (RCS) temperatures-(Tavg, delta T) and core thermal power.- y b Results- ~[ 4 l .
- The; test'. data and subsequent analysis attthe 75Z plateau indicated;that-4 <
4 the difference-between the. predicted and the measured doppler- .; JB; coefficient was within the acceptance criteria of this test procedure. l f{
. There were no test exceptions taken at this- plateau and two test changes were generated. .'One was issued to repeat steps that could not a
/ II be done' because' of: a circuit -interlock setpoint being off and theLother:
'was to clarify' procedure numbers and technique.
2 O
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A 2. 9b , : 87.-34' LQ&D SWING TEST *
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*f .q' t'. automatic controliThis system
' test demonstrated performance proper plant transient for a'102-load' change'response introduced and proper~
d ; D ? , .. ' '?' m; lat the turbine' generator.- , j
',c
- 7. .
- f. (Aesults. , [
[. e ct F 1[ The' intended 102: load swing at 752 test 1 plateau was uccessfullyc ! completed.eventhough the swing was' larger than anticipated. The down. l < g..,' swing and:the up swing averaged'approximately 161.< Even with the; ".l ! 'Et , larser: load ' vins the P - l ant automatic controla r*8Ponded properly-and. '!
,* .the procedure acceptance criteria was adequately. met. ..It was noted! h 1
,g that.the-feedwater heaters experienced a larger-than; expected variation: Et
- g '&
-in level' control that' contributed.to heaters becoming isolated as.well as extraction steam, a Up f
L c The PATG will be fine-tuning the:feedwater heater level controls' prior.
- to going to the 1002. test plateau. This is expected to minimize the t
-typeLof transient experienced at this plateau. . >
The PATG has also requested that the System Support Department re-ll evaluate the technique used to create this power swing prior (to . implementing it at the~1002 test plateau., .' f ; 4 R There were no test, exceptions taken by.this procedure at this plateau and two test. changes.were generated for this plateauOneLchange was f g:' to. increase the mis-match'between'-the primary and standby. turbine 1 3 35 control signals from 82.to 10% to assure that a 102 swing was generated since the previous attempts fell slightly lower than~the planned.10% '
- swing. Theilast test. change was to allow a knowledgeable Systemc ., ,
Support Engineer to assist with the test in' lieu of the availability of- ,
- a. vendor's representative' . The first change resulted in the larger
.- .: :than' anticipated load swing described above while..the second. change:. .
permitted' continuity in testing."' 1 e E 'E
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us 'Qiild10.J;87-35 LARGE 1A&D REDDCTION % F, g f If ( Obiective ") '
",(r
.I, ,,,,
., ),
N T
- "^ -To demonstrateiproper plant transient' response and proper automatic:
[ > control system performance.for>a 502 load reduction. ' I J !- # . i
- ilesult s , J- >
1; i 4 4 The plant' sustained a,520 MWe: load reduction and a reactor power drop .q I 0i > from 772 to 31% with the automatic controls controlling properly and fl within the acceptance criteria of the procedure. The procedure .
~
s ,
, d;
' originally required a-load; reduction of 540 MWe from the plant
.) *
- ' operating load but onlyLeaw a 520 MWe reduction which was deemed-ggj y acceptable based on plant' response.
Tm! ' ,
] '
Jr;g
.There were no test exceptions generated by this; test at this' plateau and only one- test change' to' clarify test' technique and projected '. load' ;;
1 g ,. reduction value.- p i r I" R .I '. " 1 h j
- Jln H j
@t, i /,. '11.-.ST-36 AKIAL FLUX DIFFERENCE INSTRUMENTATION CALIBRATION n - , Obiective, :! s' To perform the: start-up calibration of thelexcore power range y;h j , ' detectors. It also provides specific calibration data.to be,used to .; Dg set othertcore instrumentation used for this specific, reactor.: ' lt Calculational. techniques used are those specified in approved Reactor l [ p % Engineering procedures.
-t n .. d Results t
The re-alignment =of the Axial Flux Differential-Instrumentation was-3
- successfully performed.
~
D k T Lg No test' exceptions or test changes generated. hg > p LI 3 i p , il (;. : A-6 i n im . r
,_m . __. .
7, ,,;, 3. y- , Mjfj.,J p i$f .p Y g
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^
I% k__ , W ' h 12i ST-42 ELTER MEMISTRY CONTkDL
^
Wifh@(( d u [# ge J Obiective ' "" U i g yp; i > , .. . . 1 ~; 4 7 'This test demonstrated that chemical and 1adiochemical control andL o analysis' systems' function as described lin the FSAR by maintaining' L 'h [
; N. s ' primary and; secondary chemistry within requirements of.the Station gg_j' un y~y <1
. Chemistry Control' Program.
N '
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,1 Results p_b .All data taken at this plateau was accepted _by.. Westinghouse'and the" k
; g' i. -Station Chemistry Department for continuation of testing.- One; g,y %. 'feedwater silica sample was required to be re-taken by a chemistry Q, ~ Department RTS because the-original sample was off due to the
~
e
.,1 lsensitiYity of the1 sample to contaminant carryover. This: sample was I , _ ._ / successfully taken by RTS 1-CHEM-9-2-CP-W1.
gi;
*l 1 There were no tect exceptions or te'st changes generated by this test'at
[, _this plateau. ,
;A.
~, ******o **********************************************************************-
'1 l
-IE t h-
, . l .- l!S U g l13. ST-43 PROCESS COMPUTER h,
I.4s - Obiective-
, t.
- This procedure demonstrated that tha Main Process Computer System.,
5 g@l %
* (MPCS) is receiving valid inputs from process-. variables,' performing.
's related calculations correctly and that the Safety Parameter Displayn .,'
i + System (SPDS) responds properly under conditions of MPCS, stress. ' Results-
.A.m
.I ' :
The results obtained showed good agreement between'the MPCS and the MCB.' h '[ : e and with the acceptance criteria. Two program' evaluation sheets;were "
%n z g" g .
' generated.due.to AFD and PCS leakage found.out!of procedural criteria.,
Both were evaluated as acceptable due to small; calculated values
.g; V*4 , because-the percentage error was large however the absolute.value was-
; determined to be acceptable.
There were three test exceptions generated at this plateau.. Two were b,,!> the result of data not being.taken simultaneously between'the MPCS.and
'+ ' .the MCB.. When taken simultaneously, the data was acceptable.' The,last
)
5, exception was generated because the Gammametrics Train A and B monitor i f flux data did not agree with required tolerance. This was determined glH! < to be.-acceptable after further review the calibration data of the-LgF ' instruments. This resulted in the only test change for this plateau jgy , which_ changed the procedural tolerances to match those of the calibration data. M. ,~
~
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)
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k@.. s ygh'(i ii. , ; i. .
g'[ ~yv; " fj j y r - y' ,m-
%}; .
, ,i 14.R ST-487 TURRIME GENERATOR START-UP TEST j ~, jg.j7. Obiective ' l
?To' demonstrate:that;thefturbine' generator _controle functions properly
.1 through~each of thel test plateaus'and determine what_ adjustment may be:
; 1y required to support full power, operation.
Results 7 E E4 The 75% plateau required performance parameters to be.taken with the
% turbine generator in' stable condition. Additionally..the thrust:
J 'g ' bearing wear detector (TBWD) and the underexcited reactive ampere limit hg (URAL) checksirequired no adjustments to'the setpoints. The PATG , j
; adequately < addressed the five;(5) data points that were comunitted to be
'taken as the unit was coming'back-through 502-power;
. _Two (2); test exceptions 4 and 5 were generated at this plateau to- -
delete _various data points from scan-due.-to sensor failures. This _ LE! resultediin; test change i16 to remove these points from the. data-sheets
?B to be used at hig%r. plateaus. 1These. deleted points due not. impact validity of data being~taken as sufficient instrumentation exists to?
(g prove the; turbine, generator' performance. g lll 15. ST-52 POWER ASCENSION TERMAL EEPANSION TEST OMeetive s 'To' demonstrate that selected secondary systems' thermal expansion is
; consistent with design requirements for: thermal growth'in these areas.
.Results j ;
;Results'show proper l movement of pipe to support-continued' power ascension; Fifteens (15) problem sheets were issued during the test-and
.g. 1were accepted.by Yankee Engineering.- These primarily dealt with ig( temperature related differences with'the analytical model because the-MSR's are :not yet in service, to elevated the piping temperatures to there design values. 'Further monitoring will be done by'this procedure at'the_ higher power: levels.
There,were no. test exception or test changes generated for this-procedure at this power level. i,b I 4
- I ...
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bP e ST TEST RESULTS RETIEN PCR 901 AND 1002 POWER p [9z Presented To The Mannaement Oversimht Commaigtee [ w < I On Sentenber 11. 1990 t l -
, This! presentation is- being made to~ the Management" Oversight' Committee prior to ;
4 W lcllipower ascension test-procedure ' test results"-.being processed through the-final SORC review and acceptance process.' The SAT's review of each test procedure:.is' described below for each test performed'sinceitest results were reviewed'at-the 752 test plateau. =The SAT willocontinue'to monitor the SORC ' I process andLreview the final'SORC approved test results. . Any differences that I effect the following analyses will be brought-to the attention'of:the Management 1 :
' Oversight committee..
4 s
. .s,
*************************************w**************************************** -[
]y- .1.I ST-1 STARTUP PROG TAM AIMINISTRATION ,
-Obiective. -
i.. ,w ; i i .To provide guidance for the administration and-performance of <
- k, individual tests-during the Power Ascension Test' Program and providingL
... a reconanended: sequence for. major plant evolutions and testing.
Eb
- 4;et-u Results
'. L< NOTE: PATG test review still in progress. Test results not
. l, SORC approved. ; ; y s
bL < No. test changes / Reason 11 , t No. test exceptions / Reason
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' ', t ~ 2( 87-13 ; OPERATIONAL ALIGMENT OF NUCLEAR INSTEIMENTATION !
?ip a .. j
,i c '1, Obiective; c , ,
{
. n
. ,1 4 RW-u,-
M lTo ensure proper alignment of the Gammametricafand Westinghouse nuclear: instrumentation systemscprior to proceeding to each of thefmajor;
- )
1
'j l:
O s ; plateaus l ( s . 3
.w ,
.j p -
Results ' p: 4 [ rn , Obt'ained, overlap data between intermediate' range channels and power = j 4
- range channel; detector currents to confirm value for full power / 1
; ', T l detector' currents. Final 1 test results satisfactorily met, procedure -
L , 1 4- ~ acceptance criteria. j z.y '
- s,J- -). -
Two, test exceptions were generated at this plateau to address a non-
, g linear; response and adjustment problems with the IR compensation- ;;
H O. voltages-and'four (4). test changes were generated;to address procedural . , y?
~
4 problems to< fit thettesting scenarios to existing plant conditions.- q' on - :L lThese test exceptions'and test changes were' reviewed by the SAT on " u _, ' - e y '8/6/90. _g 4 - 1'
. l- .. . - . . !
a . a: 'v I r l. i ;. 1 i f:\
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,- 1 o . > j- [ . 6y3; . ST-14.1: OPERATIONAL ALIG M EN'r OF PROCESS TEMPERATURE INSTRUMENTATION 4 i: .n '
- ;Y. .
d U,Q $.7 Obiective' t in ' (O Ensure: proper alignment'of the delta T and'Tave instrument channels at! e n, coach of the required test plateaus.
, p ;s . <
n
, o , . , . y i
( g, .(Results ; r t
, 4 m. .
T' .Obtained data from ST-26. Data from the 0, 30,.50, 75,- 90 and this 2 : plateau were evaluated. This determined that rescaling was required'to 3 b;. , ,i f allow-the plant-to' operate more closely to its design. Loop 2 narrow
. range cold, which.was out of-tolerance at the 302 plateau remained l
W 33 reading OK at this plateau as-itJdid at 502. The' exception generated ~ yl( ,g aj & '
"at:302 will remain open so that it will continued'to be monitored to q i$ cir , og r. assure its~ reliability'at the higher plateaus. -
c ,. e # 1 There were no test exceptions or test changes generated at either-the: a i (!hy 3 h f ,
~
'902 or the 1002 test' plateau as of the SAT review on 8/3/90. J i 1 0
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$ 'ST-151 REACTOR FIANT SYSTEM CETPOINT TERIFICATION
(
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% Obiective
'{
E ifl; 5 E ? ' ' 4 (To' provide airecord of setpoint modifications"made during the~ Power ', j f y ,
, y : Ascension. Test Program..
i; , 1 -c
,' f{ j % 1 Res.u111 11 s "p p .
"g 4 NOTE: 'PATG test review still'in progress. Test results not- i g;p , ^ v. , ,
SORC approved.. . i
, t i
4 'h 7 aNo.; test. changes / Reason: ; ~. b- g ) i f f .;, n No. test: exceptions / Reason- '1 7 ,
, >i s -r
, o l( T ',
.. y
- 5. -
STi22- NATURAL CIRCULATION TEST !I
- s. !
o, n .. , y . 2 Obiective' j j ,[' x % lW _ '
; . To determine ~several~ natural circulation characteristics and will, ;
J . demonstrate the: ability to remove heat from the reactor coolant system 1 "M
, .using
- natural' circulation. .M}
,,..o ,
.o o.
l I Results l P This : te's t'. satisfactorily demonstrated that ~ heat can be : removed' from:the' q" reactor.' core if the natural circulation method was required to be used. l' , Test results met the acceptance. criteria ~of the' test procedure. >
'There were'no test exceptions generated andLonly one. test change. [
1 , generated to address data'aiready being' monitored by GETARS'. as of the l SAT review on 8/16/90.. I;
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m y ., 1 y <y ,. lf. ;< b 3 p fg: i ' c 2 IST-15? 6 STRAM GEEERATM AUftM& TIC LETEb CONTROL
^
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- s >
Obiective 4 7 ,
~ay' 4
Demonstrate'the stsbility and proper.opebation of the main: fee'dwater) 1 i'/ w , pumpsfspeed' control 4 , E lf Results yl*.y-N .All~d ta taken for.the 1002 plateau was found to belacc.eptable.and' meet 35 De :the; procedure acceptance' criteria. The correct differential:wes' c
;; observed between main feedwater flow and main Lteam fl:.a.
~
S d!T .
, There were no test changes taken'at this plateau andfonly one test i
4 ' . . exception was generated.lThis exception documented the error observed ' on(the; position indicator of'FW-FCV-540 in comparison'to:the actual jh" 1
. valve position. Wa 90W3920 was generated to correct the reading problem. The valve'was in the correct position but the Control Room.
. indicator did not read correctly.
r . There were no other test' exceptions or test changes generated at the-time of the SAT review on 9/4/90. 1
****************************************************************************** ;)
,. +
- 7. ST-25.1- FEEDWATER PIMP CAPACITY f ~ 0biective' ,
,, t To determine the maximum power level which can-be sustained with a 1 lt single main feedwater pump lin~ operation'and to: determine:the optimum 0
. turbine' setback following a loss-~of one main.feedwater' pump from full-power. operation.
'Results' This-t'est was deleted from the program by a test change to ST-1 with
~
L SORC concurrence. ;This test is not required by Reg. Guide 1.68 or the. o FSAR. l e h s L ,'. A l 1
- i. ;
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, ,l j f 84.0 ST-26L 7EERM&L POWER MEASUREMENT AMD STATEFOINT MTA COLLECTION
- 3, , ,
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- Obie~ctive i -
3 1 , .
~
9 4 : To determine 3 ore .tliermal~ poweE by_ ca'lorimetric .det'ermination 'and to - d t verify the performance ~~ parameters of the main steam and-'feedwater. I i ' systems. Additionally, this test.willLcollect other process data to;> u : support the performanceiofjother ST's'and: establish baseline operating; '! l1 f dsta'. J' ~Results' ' j"
'4 IThe test successfully. collected allCdata required'at the 902 and'1002 ,
1 , plateaus. All-data met-the acceptance criteria of the. test procedure ~. :(
" > e There was excellent agreement between the precision heat' balance and ., i lthe'HPCS heat' balance.e There was only a difference of less than. *]
;3
'0;22.The onlysignificanttitem that required clarification,during the. ,
'n ;
test, performance was? instantaneous readings-were observed that exceeded!
=the upper limit of'3411.MWt byD.48.MWt.. This.value was within the OS , , ,
- l, 11000.10 'eight:(8) hour a'verage limit of_3411 MWt. Upon this evaluation M.
Jm; , and. clarification.._the' interpretation concern was resolved.. I
,I '
There were no test exceptions or test changes: generated at-.these-plateaus and there are nojopen test' exceptions in the-SORC approved a
]
completed test,. dated 9/5/90'. N 1 q d
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,W[_ t h gj,(psy ST-27f START-UP AILTUS1 MENT OF TEE REACTOR CONTROL SY .j j
pAf 4 Obiective ', ,
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- To;obtain and' evaluate: data"necessary to determine? the Tave programi j
'r f j; 4
$ga/f}kp that will' result in the' highest possible1 steam'pressureito assure .
~
'i d Mk/ f ioptimum plant efficiency, while. maintaining l pressure for the turbine-
, , W 'g' and Tave within required limitations.. This data-will be the besis.for.'
any adjustments to thel reactor control system .(Tref or: Tave $ Af j(f'
; i
;y programming module).
t l 3
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, =
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< g . (l k[ ,' Results ,
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A %s This procedure obtained'the required data and.1 performed;the required 1 >
- a. 4 ii% .,. analysis,to determine the necessary adjustments lto Tref that'.would-h#$ allow the plant'to be run'at' optimum efficiency.- The' data collected.:
i j. j&Yl J
-for' final; analysis,=was obtained after the.MSR's were.placed in service.to assure the, plant-was configured'to itsinormal' full power
~
M/DM - . operating configuration. The final' adjustments that'were made were'as; '
- j
{ i'
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followsD '
) '
$L [' 4 Turbine impulse / pressure adjusted to'693.2 psia'
.m Tref' adjusted to587.5 degrees F to bring the S/G pressure to within'10_.
[g p' ' p '; ly.. 9
~
psi of 1000 psia.- . lI ,1 I ,' y L' Post-re-calibration data showed that the settings were= repeatable andl "
- .within.the.acceptar.:e criteria of the test. The final: operating data
. i
$ ', was;tak'en.as;follors - l h. o
' Average S/G pressure.was 1002.1 psia [
' l Average Tave was 587.4 degrees F
> Turbine impulse. pressure was'691.4: psia [
There were-no, test exceptions: generated-since the 75I test plateau ar'
) ; = only one test: change generated to allow the alignment. of. 0P-DELTA-T to -
? '
final full; power value of-Tave. There were notother test exceptions or y , test changes generated as of the SAT. review of the test results-on 9/5/90. I 5}- 7 t A 14 ! 1 m
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-;10F 8T-28: CALIERATICE OF STEAM AMD FEEDukTER FIAMF INSTRtMENTATICE ,
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This testicollected specific main st'am e and feedwater flowidatia', under ;! w-
, operating; conditions,cto:be used for. correcting, to a greater _ accuracy,c ,
a
.the' calibration-values for the main steam" flow transmitters and'also
~
L-( :wjg, i verified against feedLflow data obtained from ST-26.. ,-
,y
' ' Results4 , gi LSteam' flow. and: feed flow mis-matches _.were found to be J clightly out' of '
e
~
.. tolerance. required:by the procedure ati the 902 test.' plateau.;1This- ;
caused one: test exception to be~ generated.1 This'~ test exception ~will bei ,d
.D ;
' closed-after'1002lrescaling is completed.- Three; test:changesjwere. -,
i ; y i - generated to allow respanning'of the steam flow channels;afterothe 100Zn 'i c.
, ; datalcollections,: allow flexibility.in data collection, Land to change. ,
. .the 2 acceptance criteria of the feedwater/ steam-flow mismatches to .5-22=
.o j with sWestinghouse concurrence.- 4 I
+. .v ., a l
ThefSITreviewwas'performedon8/16/90. ' q m I^ , 3 i ' !!
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ggW.9j 4 ' L I ~11. ST-29 CORE PERFORMANCE EfAIB&fl05 W E; Obiective g This test will verify proper reactor core performance by obtaining
'7'
- . incore flux and thermocouple maps and analyzing data in accordance with approved Reactor Engineering procedures to validate nuclear physics
;~
performance. Results The results of the 902 plateau flux map indicated that all acceptance criteria was met. No test changes or test exceptions were generated at this plateau. M The results of the 1002 plateau flux map indicated the Fxy limit at I 1002 power was exceeded. However, the Fq value, of which Fxy is a part, was within the required limits. An analysis of the acceptability of this condition was performed by Reactor Engineering in calculation I #90-04 and is attached to the test procedure. A recommendation to revise M/D top of core settings and the grid alignment scheme used for map analysis was made and is being pursued by Reactor Engineering. An ig analysis of the thermocouple maps was performed by Yankee Atomic
-E Electric Company, which found overall response acceptable with the exception of the thermocouple at location (H-03) which continues to read low.
There were no test changes generated at this plateau and only one test exception to address the Fxy condition described above. As of the SAT review of these results on 8/7/90, the acceptance criteria of the
;I procedure was acceptably met.
1
- 12. ST-30 POIER COEFFICIENT ME&SUREMENT Obiective To verify nuclear design predictions of the Doppler-only power coefficient through correlative measurements of the reactor coolant system (RCS) temperatures (Tave, delta T) and core thermal power.
j'
.= Results ;g The test data and subsequent analysis at the 100% plateau indicated
_g that there was excellent agreement between the measured and the predicted doppler coefficient. The difference was .112 degrees F/2 which is within the acceptance criteria of + or .5 degrees F/2. There were no test exceptions or test changea taken at this plateau and w no open test exceptions in the final SORC approved completed test, dated 9/5/90. I . _ _ . . _
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j~&', '13. ST-341 LQ&D SWING TEST
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- ObiectMrei * ' * - i
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'This test demonstrated proper plant transient response and proper '
, , automatic control. system performance,for a'10Z load change' introduced
- at fthe turbine ~ generator.
1
- m. ,
L @ Resul's' t , 1 > ,- , . . i E ., '
.The intended;10Z load swing at'100Z5 test plateau was.successfully-lt completed. 1The-plant' automatic controls responded; properly..and the' *
'g <
procedure acceptance criteria was adequately met.; 7 , c y , Thef fine tuning of the feedwater heater level' controls prior to going j]
-to the 100Z' test plateau eliminated problems from previous test; ,
plateaus.-
~
J i
, , o , ,g II There was only' one test change taken by'.this procedure,'since its last' h
yh'*i 1 =
implementation,'-to' address a more accurate method of setting up the= '
}
l 7 turbine loadset just prior to implementation.' Additionally, there were ;
, conly two test exceptions at.this plateau to address that S/G #2 level' 2
[ ' W:- ^ g control <was not kept-in auto because the.USS was concerned that thei - ' i [,, .. level'. swing'may exceed procedure requirements may be: exceeded. Review i = i
, oflGETARS data indicates that the level 1 trend would not'have exceeded' ..
3' ltheltest' requirements The second exception addressed Tave^not return i p j ( ' i' .
# . + ;to within 1 1/2 degrees-of' Tref. This exception.was' accepted as.-is.
l
, )As oflthe: SAT' review on 7/26/90,'there were no-other test exceptions;or ,y test changes generated.
m < y
; ',*******************************************************w********************** ;l
'I: 1 5 l, , 4
- l
- 14. ST.35" LARGE LOAD REDUCTION 6 .
1. [ b <0 iective .n ( f- n i371 1' ' f To demonstrate proper plant transient' response and proper automatic :i 5
- 2i '. control 87ata= performance-for a 50% load. reduction'. 4 f
, sLResults
[x ') W g4 The plant sustained a 620 MWe load reduction;from 1200 MWe with the , J!
- i. 'M W automatic. controls controlling properly and within the acceptance 7 g '!g,l criteria of the procedure. The procedure originally required a load. t
- %i reduction.of 540 MWe from the plant operating load but saw a 620 MWe-
' * ' Jf g reduction which.was deemed acceptable based on plant response. a h There were no test exceptions or test changes generated by this test at df this plateau as'of the SAT review on 7/26/90. 4 E
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. h [,"15. 8T-36[ AIIAL FIM DIFFERENCE INSTREEENTATION CALIBRATION 7 '$j]
9lE 4 ' 4 w b* ,?obiective: hmsoa .l . I ', #._ f a T6 perform the start'-up-calibration-of theLexcore power range; . , hlh
'? u ' detectors. It_also,providesespecific.l calibration data to be used.toi >
%aq f/
, set (other core instrumentation used for this1 specific reactor.
1;
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'CalculationalL.techniquesLused"are;those specified in approved. Reactor. i , [}
FEngineering_ procedures;: ,. ! L D 3 ,
',E
, Results q a. , ?g E . The; analy' sis of ' data ihdicated ths axiali flux difference y-
.I ,
g
; instrumentation is properly' calibrated and no further adjustments.are:-
. required; jM'd y
e - m a q NoteN.excepE.ionsortestichangengenerated!asof'the'SATreviewof 'A,f ?, . he' test results:on 8/7/90. >yj n -
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- b y m 9
'16. ST-37 STEAM GENERATOR HDISTURE C&RRYOVER HE&SUREMENT
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;To determine. steam generator moisture carryover by' accurately measuringl 4 E g 'the moisture, leaving each steam generator'using Lithium-6,-a ';Fi
~ nonradioactive chemical-tracer.
, ? -s '
i: y Ql , - Results u . 4
~
- l. [ This test was controlled by the Test Director and performed by l
- Combustion Engineering using a non-radioactive tracer of J
Lithium-6. .The moisture carryover _from the' steam generators was. ' i demonstrated.to be well within the acceptance criteria of:less thanLor-- !IL [*
. .* ' equal to .252. The-moisture carryover from the steam generators: ranged' from.022.to a maximum of .06% from the "B" steam generator..
There were no test exceptions and only two test changes generated to
. allow the plant chemistry Department to determine the concentration off
-i y- LI-6-using CS 0921.107 and to eliminate the need to danger tag closed l
= vent valves while installing pipe caps after the test.
I n . There were no open test exceptions in the final SORC approved completed
, test, dated 9/5/90.
4 t \ ,, A-18 4 .
t I t 1 y ; f
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(17J87-38)' UNIT TRIP FRfM 100Z POWE5 ]!E ,
, n <*
+ '
Obiective A
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,g gp< W UfTodemonstrateJproper;planttransient:responsetoatripfrom1001.aIt-
'u '
" vill'also be.used tol verify that the actual overall'hotlegLresistance I
.,1
; temperature! detector (RTD);responseitime is conservative ~with(respectL -
m
'to the'value used in;the accident" analysis. . >
a 4
. l t
- The test results demonstrated;that the unit could perform is design!
~
function during's full.' unit, trip-from 100% power. lThe unit performance; . , I, met itsLacceptance criteria as defined in thei test procedure. ' n There were no test' exceptions generated during i.his tnst and onlygone :
-test change _ generated:to address administrative' corrections within the -
procedure _as of the SAT. review on 8/16/90.- ', pI,1 ******************************************************************************- l l 1! 18.-ST-39 LOSS OF OFFSITE POWER TEST
- j' i Obiective' Y
i' o-To demonstrate:that.the dynamic response of the emergency electrical. u K j power; system:will'~be in accordance with design under the condition of al . l! loss of all offsite power coincident with a loss'of the main. generator.:
~
p Resu'lts 1 7 This. test demonstrated that-tife dynamic response of the plant emergency j-L
, electrical' power system responded in accordance with design to provide sufficient power to safely shutdown the plant in the. event of a loss'of L ;
g , off site' power. The plant characteristics met-the' acceptance criteria. p, of.the' procedure. ' There was!only-one test exception to address the. fail over'of the'MPCS
~
and how data was,interpre.ted from other.. sources to' adequately address ".'I4' data collection. There were five~ test changes-generated to address s crew specific conunents, correct circulating water pump lube water line-J , L sups, add restoration-steps'for the motor drive fire, water pump to its c}
. original power source, delete VCT boron sample requirements, and'given
~
6 , the USS the flexibility to restore the. circulating water system valve ; alignment to suit plant restoration from this test. There were.no other test changes or test exceptions as of.the SAT 4
' review on 9/4/90'.
Iv ,
~
A-19 4 4 _F - hm .. ' _ _ ___ - - .' _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _ _ _ _m. _ _ ,_ , _ . . . _ + . . , _ . _ , - . _ , _. 4
. - -- - . - . . ~- . . .~
7 , ; E <
'10. 87-40' ERSS ACCEPTANCE TEST
^
{ i I/ Obiective - J w I 3 To demonstrate the reliability of the NSBS by maintaining the plant at _ I, < o rated output (+02. -52) for 250 hours w) nout a load reduction or a. plant trip resulting from an NSSS malfunctica, and to measure NS$$ outtut'at.its warranted rating.
;j q
Rtsults
; .The test.results demonstrated that the unit could run at full reactor' I<
power for its warranted duration and output without a load reduction or i a plant trip resulting from an NSSS malfunction. During this' test.'the 1 PATG properly took test interruptions.and docutented several reduced' l power periods to perform operations weekly toch spec surveillances on. I
?.he turbine control valves, as well as a power reduction resulting from an EHC leak in,the secondary plant. Neither of these conditions ;
violated the test proceduro acceptance criteria, i There were no. test exceptions generated and only three test changes j generated as of the SAT review on 8/29/90. These changes involved clarifying the MWL values to be monitored, deleting a reference that no
.I 3 longer applied and clarifying addressed in t'ae first change,.
] : J 20.'87-41' RADIATION SURVEY f L Obiective To determine the neutron and' gamma dose rate levels at selected l ! locations throughout the plant, and to identify high radiation areas.- ' Selected radiation ~~ itors will also be verified. l . Results l: . i This test st the 1002 to t plateau demonst rated that ;s u.ation levels did not exceed there design. parameters fo there respo tive sones of shielding. HN 0960.07 was used to review and resolve 7 deficiencies ! g
~ identified durire these surveys. These deficiencies addressed 3 (
. readings that appeared high but were due to radiation streaming versus. 4 a shielding defect, 1 reading that was slightly higher than expected *
-but the - ading was in the containment opposite the pressuriser heater pull space aron which was below overall containment criteria, and 3 oI radiation monitors that exceeded the limits of R0-2 but were acceptable ;
when matened against survey meters. There were ao test changes or test exceptions at this plateau and no open test exceptions in the final SORC approved completed test, dated 8/31/90. l A-20
.__.____1_._ _ . _ _ _ _ _ _ _ _ _ _ . - _ . _ . . . . . . . , _ _ ~ - . . _ . . _ . _ _ _ _ . . . _ , , , . . . - . . . , ,
- 21. 87-42 B&TB ME3ESTRY ODETRAL Obiective This test demonstrated that chemical and radiochemical control and analysis systems function as described in the FSAR by maintaining iI primary and secondary chemistry within requirements of the Station Chemistry Control Program. +
1esults All data taken at this plateau was accepted'by Westinghouse and the Station Chemistry Department and demonstrated acceptable analyser I operation. .Some cation conductivity, sulfate and silica readings were determined to be above the requirements'of SSCP 3.2 in the main steam,. feedwater and steam generators.~- These parameters are expected to be' I slightly high during the initial full power operation of units of this design until they can be purged from the plant during steady state operation.
~There were no test exceptions oritest changes' generated by this test at this plateau and no open test exceptions in the final approved completed test, dated C/31/90.
I I g: I I I I I I I . I -
.m 4
b< w' cis. m , , , R 3 i + ' iby 22.')hT143 PB00ERS OSMITBt s S Obiective ' g.r. ; * ,
, , y This procedure demonstrated that the Main Process Computer System' ' }ys *
(MPCS) is receiving valid inputs from proc e variables, performing-
/g N ;related calculations correctly and that tM fety Parameter. Display f, System.(SPDS) responds properly under concierons of MPCS stress.
I Resultel , E' ,'The results obtained showed good agreement between the MPCS and the MCB. 3 and with the acceptance criteria. Two program evaluation sheets were 7"- generated due to AFD and RCS leakage found out of procedural criteria.' Both were evaluated as acceptable due to small calculated values- , 9= because the percentage error was large however the absolute value was l, determised to be acceptable.
* 'There were three test ex:eptions generated at this plateau. Two were I i the result of data not being taken simultaneously between'the HPCS and
-the MCB. When taken simultaneously, the data was acceptable. The lastL exception was generated because the Gasunametrics Train A and'B monitor 5
flux data did not agree with required tolerance. This was determined
.I ,
to be acceptable after further review the calibration data of the-instruments. This resulted in the only test change for this plateau; - * , 4 which changed the procedural tolerances to match those of the calibration data. h-
;No. test changes since 752/ Reason Nc. test exceptions since 752/ Reason el u
l e ss m y- . j, 3 m 1
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. . . - - . a
I e j 136 ST-44 TIMATIM AMD laDRE PARTS MDEITM I l I , Obiective To obtain RCS baseline noise. level and signature data from the TEC l I Loose Parts Monitoring System (LPMS) during steady state and transient I testing activities. Results i 1 ,l . W The acceptance criteria- for steady st. ate and transient noise data i adequately. met the acceptance criteria of the test procedure, with the " I exception of higher than originally expected background noise on channel 1-VB-Y-6825-1. Procedure RN-1714 is to be revised to incorporate this data. I l lr i There were no test changes generated at this plateau and only one test ; l exception generated to address the problem described above. The SAT ; review was performed on 8/16/90. i
*,******************************s**********************s**********************
- 24. 87-45 PROCESS EFFLUENT BADIATION MONITORING SYSTEN ;
Obiective ! To ensure the proper operation of the process and effluent radiation . monitoring systems, as detailed in the FSAR, Section 11.5. I Results- l This test demonstrated that the process radiation monitors could take ; process effluent samples and analyze them for radiation. 't I I There were no test changes generated for this plateau and only 1 test exception generated to address 15 RM samples. One sample could not be taken but was addressed with WR 90WR003984, 2 samples were low because ,3 of short lived isotopes being monitored to date, and 12 sampled were gg balow the sensitivity of the monitor. [ l : There were no other test changes or test exceptions as of the SAT ~! review on 8/29/90. I , I A-23 I _u__________________ - -
h ~ vy N ,i' '
'.l .- ! I( .0 2 . 87-46h TEET11AT1tB RTETBt GPB&BILITY TEST d
h Obiective gJ.,
;g m To demonstrate that various heating, ventilation and air. conditioning systems maintain their service environment areas within the design
.g
, limits under normal plant operating conditions.
- p. % Results NOTE: PATG test review still in progress. Test results not
% SORC approved.
No. test changes since 502/ Reasons No. test exceptions since 502/ Reasons ,
- ap:
ha i
-26.'87-48 o g TURBINE GBIERATOR BTART-UP TEST -; g y obiective
- To demonstrate that the turbine generator controls functions properly through each of the test plateaus and determine what adjustment may be
.m ,
required to support full power operation. Results
.8
- (
a The 1002 plateau required performance parameters to be taken with the
'"'bi"' 8'"'ratar 2" ah * * "diti "- ^*di'i "*2 7 th';'hr"
E,; , bearing wear detector (TBWD) and the underexcited-reactive ampere limit (URAL) checks required no adjustments to the setpoints. _ : There were three test changes taken at the 902 test plateau to clarify administrative requirements and three test exceptions taken to address faulty thermocouples and some data being lost for a 45 minute period, c o. - There were no test exceptions or test changes generated at the time of J the SAT review of the 1001 plateau test data on 8/22/S0. _v L e I J ga
-;g
_ A-24 __Y c
l't _ 4 , B 27.o ST-51 PGW R ASC MSim M E MIC TIERATIQlB TEST i 4 Obiective To monitor the dynamic response of selected portions of the main steam and feedwater systems following a unit trip from 1001 power. Results I ,This test demonstrated that the plant dynamic response following a full power trip fell within plant design criteria. Test data was forwarded
.to NHY engineering and was accepted through RES 90-473. One open test exception remains to allow post calibration of installed test I equipment. ,
There were no test changes or test exceptions generated at this plateau I as of the SAT review on 8/29/90.
- 28. ST-52 POWER ABCERSION.TEERIEL. EIPARSION TEST Obiective l' To demonstrate that selected secondary systems' thermal expansion is consistent with design requirements for thermal growth in these areas.
- Results.
Results show proper movement of pipe to support continued operation at full _ power. Twenty problem sheets were issued during the test and I either were corrected by-WRs or were accepted by Yankee Engineering evaluation of predicted growth against actual growth. There were no test exceptions or test changes generated for this procedure at this power level as of the SAT review on 8/22/90. I I I I I I; .. , I
i s
,l4 ,
- 29. ST-56 FIPIHB TIB& TION TESTIES ,
h.1 Qhit.C11Yjt
- To verify that the vibration level of selected portions of the
$. condensate, feedwater and main steam system within the Containment, MS/FW chases and turbine building are acceptable under steady state conditions at full power.
Results All data' met the test procedure acceptance criteria as of the SAT I review on 8/3/90. However, since this review it has been learned that a post calibration of a vibration meter failed its post test calibration check. This will result in additional data to be taken on the condensate system for baseline purposes. Even though this data I will be re-taken, there is no indication of'any,significant problems with the condensate system based on visual observations during the testing efforts. There were no test exceptions or test changes generated at the time of the SAT review on 8/3/90.
,l
=W 30. 87-57 DRAIN OUT APPRQ&CE Obiective To determine feedwater critical water levels where flashing begins in the draine cooler region of the heater shell.-
__ - Results
!l This test was deleted from the program by a test change to ST-1 with.
;3 SORC concurrence. This test-is not required by Reg. Guide.l.68 or the FSAR.
I I iI DI 4-3 A-26 l . . . . . . . . . . _
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- 4 i
' =
#; g$ I Annendix B
( y. _GERDHQLQGY_0E.lYENIA ,
't' 7! ,
- Uli Date Event w
t:-
- w. 3/16/90 -- Comunenced PATP 1
+ ' ' 1
, -- Initial NIS Alignment (ST-13) l
.. p %
3/17/90 -- Calibration Of Steam And Feedwater Flow Instrumentation (ST-26) I,N '
+
3/20/90 c r
, m.
Reactor Critical Shutdown Monitor Verification (ST-13)
- Ih -
-- Thermal Expansion Measurements (ST-52) 't 3/21/90 --
Shutdown To Replace Noisy Power Supply '(Rod Control) And Correct ; pj f Main Steam Valve Leak j, 3/22/90 -- Steam Dump Tests Underway (St-23) n 3/23/90 -- Steam Dump Tests Completed f - . l ::" ^... .Y ! o
.3/24/90 --
Consnenced Steam Generator Level Control Tests (ST-25) ;
-- Test Delay, Main Feedwater B Valve Problem
[
; -- Completed Required SG Level Control Tests ;
'O 3/25/90, --
51 Power Level Test Plateau Completed: SORC Approves Initiation Of . c 302 Power Level Test Plateau ! Entered Mode 1: Continued SG Level Control Tests k
'3/26/90 --
. Completed SG Level Control Tests: Preparing For Turbine Teste L 3/28/90 --
Testing Interrupted: GE Representative Evaluating Turbine Rotor j c (, Position Problem . 4/1/90l -- Testing Reconsnenced: Feedwater Pump Transfer (ST-25) 4 , f 4/2/90 -- Chest Warming For Turbine Roll
~4/3/90
-- Turbine Rolled: Control System Oscillations Observed I 4/4/90 -- Loss of Speed Signal. Turbine Trip, Evaluation Commenced :
~4/18/90 --
Turbine Testing Commencea (ST-48): Bearing Vibrations Noted v 4/23/90 -- Setting Up *)r Torsional Testing (ST-48.1) 4/25/90 -- . Coasnenced Torsional Testing 4 '. I i k Ikl.nu ,
,cy n ,
f. n. i k$* '
. _ _ -- -- - - - - - - - - ~- ~ ~
L] {'l .a . .w... - 1 Completed Torsional Testing: Did Not Attempt To Synchroniae I< '4/27/90 -- ! Generator i i i 4/29/90 -PATP Interrupted: GE Turbine Personnel Er.Dute To Start Turbine I Modifications j 5/25/90 -- Prerequisite Checking Conducted Prior To Resumption Of Torsional
- Testing (87-48.1, Revision 4) 5/28/90 -- Turbine Rolled To Rated Speed: . Testing Comunenced l
5/29/90 -- Data Collection For Torsional Testing Completed: GE Declared -! Modification To Turbine Had Corrected Resonance Problem !
-- Synchronization Test Not Necessary 5/30/90 -- Testing Commenced Transferred From By-Pass To Main Feedwater ,
Valves (ST-25): Overspeed Testing (87-48) i
-- Test Interruption For Turbine Overspeed Trip Repair 6/4/90 -- Completion of 102 - 30! Load Data Collection (St-48)
-- Entered 30! Power Level Test Plateau j t
6/5/90' -- Verification of S/G Feedpump Auto Speed Control (ST-25): ': Completion of 302 Thermal Expansion Data (ST-52) , 6/6/90 -- Turbine Shutdown: Arcing In Generator Bus Duct , 6/7/90 Verification Of Main Feedwater Reg' Valve Stability (ST-25): ; !gg -- Preparation Of MIDS For Flux Mapping (ST-29) ' L . ; 6/9/90 -- Water Chemistry Sampling (St-42): Flux Mapping At 302 Completed i I, ST-29) And First Power Coefficient Determination (ST-30): : Completed Valve And Feed Pump Testing At 302 Power (ST-25): Initial MPCS Data Taken (ST-43) 6/10/90 -- Load Swings At 102 (ST-34): Automatic Reactor Control Verification (ST-24) ;
-- Entered 502 Power Level Test Plateau -f j 6/11/90 --
Completed 302 - 50% Load Data Collection (ST-48): Shield Survey Commenced (ST-41) l' -- Fifteen Hour S/G Chemistry Holdup 1 i: I 6/12/90 -. Statepoint Data Collection (ST-26): Steam And Feedwater Flow , Calibration (ST-28) 1 y B-2
)
. f nate ' Event i
I 6/13/90 -- Adjustments To Reactor Control System (ST-27): Completed Thermal ' Expansion Observations (ST-52): Core Performance Completed (ST- ! 29): Radiation Surveys (ST-41) And Chemistry Sampling'(ST-42) > Comunenced l
<I 6/13/90 --- Feedwater Level Fluctuations Prevented Operation of MFP-B Per ST. !
25 6/14/90 -- Calibration Of AFD Instrumentation (ST-36): Loose Parts Monitoring 1 (St-44) And Ventilation System Operability Test (ST-46) Comenenced ,
.- = Power Decreased 100 MW Transient Due To Turbine Control Problem f 6/15/90- -- Power Coefficient Determination (ST-30): 102 Load Swings (ST-34) 6/16/90 -- MPCS Data Acquisition Completed (ST-43): Reactor Trip And' Shutdown From Outside The Control Room (ST-33)
-6/18/90 -- Completed 501 Power Level AFD Hessurements (ST-36) 9 6/20/90 -- Turbine Trip /Roactor Trip Event Evaluation Commenced 6/25/90 -- Reactor Critical Entered 8 Hour Action Statement For SI Accumulator A (Low Boron Concentration Sample) 6/27/90 -- Repeated $7-48 302 - 502 Load Data Collection
'E. --
Severe oscillations of Feedwater Heater Level When Second MFP (A) l3 Placed In Service (Two Heater Drain Pumps Operating) 6/28/90- --- At 652 Power Level, Ran Leak Rate Calculation For ST-43 Test ? Exception, And Ran Additional Flux Hap 6/29/90- -- Flux Happing Continued
- 6/30/90 -- Heater Drain Piping Leak, Manual Turbine Trip 7/5/90 -- Entered 752 Power Level Test Plateaus Statepoint Data (St-36) And S/G Level Control (ST-25) Completed
- -- Reactor Trip (2 Of 3 EHC Low Pressure)
" Reactor Entered Mode 1
-7/7/90 --
7/8/90 -- Entered 75% Power 1,evel Test Plateau ! 7/9/90 -- Completed 502 - 702 Load Data Collection (ST-48), Thermal Expansion Evaluation (ST-52), Steam Ahd Feedwater Flow Calibration (St-28), And Water Chemistry Sampling (ST-42) 7/10/90 -- Flux Mapping (ST-29 And ST-36) I B-3 l
l,! ?i! b' I1 ,.y, j tm Event
, . (!o 1 7/11/90J ' -- During Power Coefficient Measurement (ST-30). Throttle Pressure Limiter Intsrference Caused Megawatt Reduction And A Test Interruption-
.' --- Completed Load Swing At 75% (ST-34) 7/12/90! -- Power. Coefficient Determination (ST-30): Large Lead Reduction (ST-
; 1 34) 7/13/90 -- - AFD Calibration (ST-36): Plant Computer Validation (ST-43) 7/14/90 -- TREF Program Change Completed: Reactor Control System Adjustments complet9d For 752 (ST-27) 7/1$/90' ' -- 902 Power Level Test Plateau
-- 702 - 902 Load Data Collection (ST-48): Flux Mapping (ST-29)
I 7/16/90 -- Statepoint Data Collection At 901 (ST-26)
-- Reduced Power To 752: Process Temperature Alignment (ST-14.1) And I' Steam /Feedwater Flow Calibratton (ST-28)
. 7/19/90
-- Returned Power Level To 90% And Conducted Additional Process Temperature A11gnmert (ST-14.1)
-- 1001 Power Level Test flateau 7/20/90 -- Feedwater Heater Oscillations Forced Power Reduction To 902 -
After Approximately Six Hours, Returned To 1001: High Feedwater Flow Oscillations Required Manual Control of S/Gs A, B, and D 7/21/90 --
. Returned To 1002 Power Level Tert Plateau 7/22/90 -- 1002 Load Data Collection (ST-48): Verification of Feedwater Speed Control'(ST-25): Statepoint Data Oollection (ST-26): Chemistry Sampling (ST-42): Shiald Survey (ST-41): Moisture Carryover Test I (ST-37): Baseline Loose Parts Monitoring (ST-44): Ventilation System (ST-46) 7/23/90 -- Thermal Expansion (St-42) And Piping Vibration (ST-56), Process Effluent Data Collection (ST-45, And Core Performance Flux Maps s
(ST-29) Completed I ,7/26/90 -- Load Swings (ST-34): Accumulated AFD Penalty Minutes Required Power Reduction to <50% Returned To 1002 Power Level Preparations For Unit Trip From 1002 I 7/28/90 -- (ST-38) 7/29/90 -- Unit Trip From 1002 Power (ST-38): Vibration Measurements (ST-51): Natural Circulation Test (ST-22) , B-4 I -
I nate Event l: g 7/31/90 -- During Restart, After Synchronizing Generator To Grid Esperienced * <g High Vibration on Turbine Bearings #7 and its Reduced Power To Below P-9 to Prevent. Reactor Trip; AFter Approximately 30 Minutes. Tripped Turbine When #7 Bearing Reached 13 MILS 'I 8/1/90 -- Loss of Offtite Power Test From 20! RTP (ST-33) 8/3/90 -- Loss Of Offsite Power Test From 202 RTP (ST-39) 8/3/90 -- Returned Plant To Service: Prepared For ST-25.1, Single Feedwater Pump Capacity Test 8/4/90 -- ST-25.1 C -;eled When MFP A Drop In Suction Pressure Was
-Accompanied By Feedwater Heater Instabilities 8/5/90. -- Attempted To Reach 1001 RTP On Two Condensate Pumps: Third Pump (In Automatic) Started On Low Suct!.on Pressure AT 882 RTP
- -- Cosunenced 250 Hour Warranty Run (ST-40) 8/7/90 -- Control System Adjustments (ST-27) 8/17/90 -- Completed 250 Hour Warranty Run (ST-40) 8/19/90 -- Completed PATP I -
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