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NUCLEAR REGULATORY COMMISSION -
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4 UNITEo STATES.
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REGloN H
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101 MARIETTA STREET, N.W., SUITE 2300 j
ATLANTA. GEORGIA 303234100 -
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August 28, 1995 MEMORANDUM T0:
John A. Zwolinski, Deputy Director
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4 Division of Reactor Projects I/II, NRR a
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!FROM:
Ellis W. Merschoff, Director
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f Division of Reactor Projects
SUBJECT:
REQUEST FOR ASSISTANCE IN ADDRESSING ISSUES R GARDING ST.
LUCIE EMERGENCY DIESEL GENERATOR FUEL OIL TRANSFER SYSTEM
~3 LEAK-ISOLATION AND USING OPERATOR ACTION IN PLACE OF AUTOMATIC ACTION (TIA 95-013)
.i lRecently, the 2B Emergency; Diesel Generator (EDG) fuel oil (FO) transfer system developed a leak at St. Lucie Unit 2.
The licensee's actions'in response to this event included isolating the leak to minimize environmental contamination. The licensee performed this action under the provisions of 10 CFR 50.59. The licensee's actions in this regard have given rise to l
- generic questions involving the relationship between PRA evaluations and 1
10 CFR 50.59 requirements. The background on the issue and specific questions arising from it are detailed below.
1 System Descrintion The St. Lucie EDGF0 transfer system (for a given train) consists of a FO tank, a transfer pump, a day tank mounted on each of two.EDG engines (two engines per EDG unit), and associated piping and valves. The transfer scheme involves the pumping of F0 from the storage tank, via the transfer pump, through piping
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from the F0 building to the EDG building and then to the day tanks. The day tanks' contents are then pumped directly to the EDG engines.
3 The piping from the transfer pump discharge to the day tank is normally unisolated with the exception of normally closed solenoid valves at the day tanks. When the EDG is running, F0 is drawn from each day tank until low level signals open the solenoids, allowing a gravity feed of F0 from the F0
- tank to the day tanks. Should level continue to fall, low-low level conditions in the day tanks initiate a start signal to the transfer pump to increase the makeup rate.
Condition' Description i
Over the course of several days,.the licensee noted a decrease in 2B EDGF0 i
inventory and suspected that.a leak had developed. Through increased monitoring, the licensee determined the leak to be in the-piping ~between the
'F0 transfer pump and the' day tanks. As the piping was below grade, rapid i
~ identification and correction of the leak was impossible.
1 To: terminate the release of approximately 15 gallons per day of F0 to the
~ environment through the leak, the licensee proposed: operating with an
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isolation valve at the discharge of the F0 transfer pump closed (the valve is
-. normally. locked open). As a compensatory measure, the licensee proposed
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' dedicating a non-licensed operator to the task 'of responding to open the valve.
should the EDG start. The operator would have no concurrent event-response 1
role (e.g. fire brigade); however, he would have non-response duties to perform during the course of a shift. Additionally, the licensee proposed revising a number of procedures to include the requirement of opening the valve in the event of an EDG start.-
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S'afety Evaluation The licensee elected to perform the actions described above under the: auspices of 10 CFR 50.59. The Safety Evaluation (SE) performed pursuant to the code i
noted that two new failure modes were creatcd by the proposed action. The j
first involved a failure of the operator to arrive at and open the subject valve prior to the associated EDG's day tanks emptying. The second involved a i
1-mechanical failure.that precluded the opening cf the valve.
i The licensee performed a PRA study of the proposed change which indicated that
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- an approximately 6 per cent increase existed in the estimated frequency (per year) of the. loss of the EDG and the associated safety bus. The SE went on to state that procedures would be revised, operators trained, overall awareness heightened, and, as a result, no net increase in the probability of failure of a component important to safety would result from the proposed change.
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Questions h
In light of the licensee's conclusions, we propose the following questions:
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1.
Is the attached 10 CFR 50.59 FPL Safety Evaluation (JPN-PSL-SENS-95-013)
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considered acceptable?
2.
From a PRA perspective, is it possible to completely mitigate a risk, once introduced?
3.
Is the licensee's position (that the risk of operator failure / error can j
be mitigated, probabilistically, through procedures and training) valid?
Do probabilistic estimations of operator error rates presuppose the existence of procedures and training and, if so, can one then take credit for them in a deterministic mitigation of risk?
4.
Can.10 CFR 50.59 requirements (that-the probability of failure of:
components important to safety not be increased if no unreviewed safety question is deemed to exist) be satisfied if new failure mechanisms are.
added'to-a previously reviewed system?
4 5.
PRA insighiis.are beginning to provide a more structured evaluation e
process for proposed changes to facilities and, as-a result, are showing that. changes (in a 10 CFR 50.59 context) present finite, although small,.
increases in the probabilities of failures.
Is there a threshold value of increased probability (representing " negligible" or " insignificant"
- increases).below which 10 CFR 50.59 criteria (for demonstrating that f
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unreviewed safety questions do not exist) are satisfied?
6.
.The response to a related TIA from Region III, transmitted via letter from you to Edward Greenman dated June 23, 1993, stated in part that "NRR has no particular objection to the use of PRA in 10 CFR 50.59 evaluations but recommends that it play a supportive role, in conjunction with other inputs, such as engineering judgement and operating experience."
In the given case at St. Lucie, when PRA insights provide information counter to (as opposed to supportive to).
the 10 CFR 50.59 conclusions, is it appropriate to accept deterministic conclusions over the PRA-indicated increase in probabilities of failure?
This request has been discussed with J. Norris of the NRR staff.
If you have any questions concerning this request, please contact M. Miller (407/464-7822) or K. Landis (404/331-5509).
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Docket N'o. 50-335/389 j
License No. DPR-67/NPF-16 Attachments:
1.
FPL Safety Evaluation JPN-PSL-SENS-95-013 2.
NRC IR 50-335, 389/95-14 cc w/atts:
R. Cooper, RI l
W. Axelson, RIII J. Dyer, RIV K. Perkins, WCF0 S. Vias, RII J. Norris, NRR
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JPN-PSL4EN545413 REVISION O PAGE 2 OF 10 l
q REVIEW AND APPROVALRECORD i.
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FLANT er riverm UNIT S E
TITLE 10CPR50.59 EVALUAT50N OP5tATION WITH DIESEL OIL UtANSFER PUMP :B p;
DISCHARGE ISOLADON VALVE V17216 e'r neven
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ATTACHMENT 1 is E0'd 229P 19P LOP 831430 tuapt sad al on 1 15 W9P:0T S661-TE-40
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100FR50.59 ETALUkTION POR OPERATION WITE DIESEL OIL 1
TR&BSFER PUMN 23 DISCEARGE ISOLkTION VALYE Y17216 CLOSED i ;
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1.s ARSER&dE l'
This safety evaluation is to document the acceptability of plaat operation with Diesel oil Transfer Pump (Dorp) 23 disoharge isolation 1;.
valve va721h in the CLOSED position.
Compensatory measures shall be s stablished' to open the valve upon operaties of the SE Energeasy
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l Mosel demorator (EDs).
Y17214 is normally a LOCEED OPEN valve; 1
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<3 however, due to a suspected leak in the underground piping downstream et the valve it is desired to isolate the piping until the leak is i
identified and repairs are made or the line has been replaced.
Iselaties of the line will prevent the loss of an estimated 15 gallons per day of diesel fuel oil to the environment.
Valve V17215 is leosted la line Z-2"-00-14 which commeste the DOTPs to the 23 EDs Day Tanks.
This line is classified as safety class three (2), seismis alass I; therefore, this evaluation is ' olassiilied as j
Safety Related.
i This evaluation oomsludes that operation of the plant with valve i,
Y17214 in the CLOSED position does not impact plant safety and does j
met oomstitate an unreviewed safety question mer require a change to the teshaical specifications.
This conclusies in contingent upoa
- a implementation of the following operating restriations ame l
oempensatory actions:
I operating personnel shall be instructed to open valve V17214 as seem as possible and within 20 afautes of any unp1*==aA starting j
of the 23 EDet prior to elosing V17216 and at least twice each shift, verify l
that the 25 EDs day tanks are each filled to a 320 gallons (92%
l tall per local level indication);
valW>- V17215 aust be manually opened prior to any planned
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opesE Kea of the 28 1D4 or smytime fael oil makeup is required fest y 2R EDS day taaks; lastraskiens shall be provided to all appropriate plaat personnel
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regarding the above.
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.j This evaluation does not address system operability and plant i
operation during repair or replacement of the subject pipe.
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831550 susp:say a t on 145 W3P 01 S661-12-40
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J.e nasariatisa and Purposa
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The maderground portion of the pipeline between DOTY as and the day tanks for N 28 EDG is suspected of leaking at a rate of j
approximately is gallons per day.
The exact location of the leakage 4-is currently umidentified and efforts are underway to looste the leak 3
such that the Line saa he repaired er replaced.
The purpose of this evaluation is te allow the plant to continue operaties with the DOTP l
23 disonarge isolation valve (Y17216) la the CLOSED position, thas isolating the fuel oil leak to the environment.
compensatory actions j.
are identified in section 9.
4 3.0 Liammaing aequiremmats Valve vi7214 is part of the diesel generator fuel oil system that
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transfers inel from the Diesel oil storage Tanka (DOSTs) to the day tasks.
Per Psaa section s.s.4.1 the diesel fuel oil storage and i
j transfer system is designed to perform the fellowing fumaticam i
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a) provide oil storage capacity for at least 7 days power operation j
of one emergency diesel generator set; 4
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h) maintaia fael supply to at least one diesel generater set, j
assuming a single active or passive failure of the system j
ooincident with loss of offsite power; e) meet salmaio Category I and Quality Group C requirements; and li.
d) withstand maximum flood levels or tornade wind leadings without lossist fumation.
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!t The systaa is broken down into two subsystems (1 & a).
Each subsystem i'
oommists of a Door, a DOTP, two day +anka and asseeinted valves, piping and instrumentation.
Daring moraal operation subsystem a j
serves dissel generator A and subsystem a serves diesel gemarater as j
however, the two subsyntans esa be cross-connected at the discharge of the transfer pumps.
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Technical specifloatione 3.s.1.1 and 3.2.1.2 identify the operability i
requirements of the diesel generators.
i A similar evaluation was performed for Unit 1 in 1992 (referemoe 2).
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4.0 m==1weim me affante an anfatw j
Valve V17218 is located in line I-2"-Do-14 which ooanoots the D0Tys to
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the 23 3Ds day tanks.
This line is elassified as safety class three (3), seismie eless I.
Under normal operating conditions, Y17216 is in a teren oyEN position and provides s flow path for diesel fuel oil to the 23 EDS day tanks l
from the 23 DOTP and the DOST.
system operation is described below.
t The suotion of cory sa is seannoted to the sa DosT via linee I-sa-Do-s
& 4 and ammnal valve V17212.
The discharge of the 23 DOTP is l
conaeoted to 23 EDS Day Tanks 231 and 232 via transfer lines I-1-1/2"-
DO-12, I-2"-DO-14, I-1-1/2"-DG-260, I-1-1/2"-DG-2 42 ; shook valves t
l S0*d Ez't? 19P LOP T 831 HC suaot s+ at on, 25 we4p:0T S661-TE-40
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If, JPN-PSL4 ENS 4Se12, ser. 4 i
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j T19314 4 Ys9499; manual isolation valves V17318, V173143 and seleasid valven 83-59-131 a s3-s9-133.
The manual valves in the flew path from i
the DOST to the 300 Day Tanks are moraally leaked opea.
Seleasid valves 83-89-131 a 33-s9-133 are located in the -diesel generater building upstrema of the day tanks and provide automatie isolaties of the day tanks when the prescribed fuel oil level is attained.
Level switches Laass-4195 a Es-se-0373 start n0TP sa sad open the asseeisted i
day tank solenoid walve when the day taak level deereases to 24.s" I
(223 galloas).
Level switchen Ls-59-820s & La-59-0343 stop the DOTP 4
and elese the soleasie valve when the day tank level imoreases to
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32.S" (338 gallems).
The above operation is automatie.
j The system operation proposed by this safety evaluaties would be the same as described above with the excepties of the as DOTW disobarge l
isolation valve V17216, whisk will be elesed.
closing og valve V17315
{l will isolate the 25 300 Day Tanks from the DOTP's discharge and the Dosts.
Operating persommel will be instrusted to provide for the opening of valve T17216 in the event of a 23 3Ds ante start.
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assuming an initial day taak fuel valuas of 330 galleas, the 3Ds could i
run approximately las minutes at full lead.without replenishing the i
day tank.
This is based en the is cylinder engine which is the i
limiting factors 4
320 gallons laitial day tank fuel volume As gallons unusable day tank volume (referesse 4)
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304 galloas usable day tank volume l,j 233 gallons day tank auto fill level t
l 2.4 gpa maximum fuel consumption rate (based on referense 4 values) j _1 304 gal /t.4 gpa = 12s.7 minutes of operation c
After approaimately 40 minutes of operation, the port would be automatically started as a result of a low day tank level (330 - 223 m 97 gallens fuel volume between DOTP start /stop setpoints censumed at 2.4 gpa).
Plant operations Departneat has indicated that ma operatori saa' respond to the starting of the 23 EDs and open valve V17ais withia 20 aiantes. Thus, the DOTP will not automatically start and run j against skutoff head prior to the opening of V17318. l Based om a 15' gal / day leakage rate and the system eenfiguration, the i sise of the suspeeted hele la the pipe is estimated te be less than 1/18" diameter (referesse 3). The underground porties of I-2"-DO-14 i is at plaat elevation 13'-4" (referenee 7) which is about 38 below i grade. A recent study (referesse 6) measured ground water levels in the visimity of the Wait 2 EDS building at approminately 13' to 14' belos grade, well below tha subjoet pipe. Additiemally, filtration is provided devastreen of the day tanks. Therefore, based en the i estimated sine of the hole and location of the piping, the possibility of the Latroducties of foreign material such as sand or ground water i j is considered to be very saali. I t
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i JPEPEL4EN545413, tw; 4 Pass (et14 l., operability of the subjoet pipe has been addressed is the referesse 5
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The suspected undergrenad leak has been quantifies at i appreIiaately is gal / day. yhe 23 Dorf has a design flew rate of 25 GPM (reference 1) and provides sufficient flow margia to deliver fuel to the 23 3De to asiatain the required fuel oil level in the day p. tasks. l A risk assessment was acaduated by FFL's raa group. This assessment ] used the baseline Unit 2 psA model to estimate the change in frequeasy j of loss of the ass 4.askv bus with a loss of grid initiating event and j. the addities of two new 33 EDS failure modes (i.e., failure of the EDO fuel oil manual isolation valve to open and failure of the operater te i 1 open the elosed isolation valve). 1 men-recevery probability of l 3.013-2 was used for the operator failing to open the fuel oil isolation valve. This probability was based on the ex-eentrol model of ORCA using a 120 minute available time and a 20 minute seaa lj response time. 4 Two cases were assessed: 1a l Case 1: Baseline PSA model same ~ i case 2: Baseline model with the additlenal failure modes for the as i EDO (manual fuel oil isolation valve failing to opes and j operator failing to open the valve). The estimated frequeasy for each ease is as follows: Ii {-' Case in 1.73E-3/yr case 3s 1.843-3/yr l This indicates that the additional failure modes resulting from the j elosed fuel oil isolation valve results la an approximate 64 abange la j the estimated frequemoy per year of loss of the Unit a 233 4.15kT bus. stansaar Sased on the above scenario, sufficient time suists for an operator to jl open valve V17216 prior to Dopp as automatically starting to replenish il IDG 28 day tanks to termai levels and sufficient margia esists fren the as port to deliver the required flow rate of fuel to the 35 EDG, ocasidering Taplementation of the j actions requ,the esposted ground leakage less. ired in sectica 9.4 will provide f j-equivalent to the original seafiguration. 1 l i' 1 4g*g gggp Igp 40p ; a3f HO g ap:say aton1 25 WO8p:0T 566T-TE-40
4 1 f. i l1 JPN PELSENS95013, Bev. 4 Pass 7 et to j j s.a wa41==a ~====a arrante m--1,mi. i FAILURE CAUSB secos SYMRToots & IFFacts 1 t i \\ Loss of Puel supply te EDS 237 j v17316 Fails Operator Fails EDG 23 Failure Due To Fuel j to 0 pea to open valve starvation After Appron. thes 4 EDs 21 Available as amargeasy; Ac rever supply 4 Y17318 Fails Talve Failure Same as Above l to opea l 6.6 simat maatriatiamm 1 1!I There are me operating restriations (mode restriotions) ea'the plaat i While valve V17216 is asistained in the CLOSED position. This evaluation does require sospensatory actions which are identified la meetion 9. 7.0 Effant em Tachaisal Spanificatinam i The proposed activity will have me effect en p1' ant Teobeal specifientions. operability of tae diesel fuel oil transfer system is assured by virtue of the sempensatory actions presaribed by, this evalesties. Case valve V1721e is opened the fuel oil system will effectively be retaraed to its original design configuraties and will operate is its moraal automatio mode. 4.O Durawiewed safetw onamelaa natsmimatiem I With respect to Title to of the code of Federal Regulations, Part l 59.59, a proposed change shall he deemed to involve an unreviewed safety questions (1) if the probability of occurresse or the j sensequences of am seeident er malfumoties of equipment important to safety previously evaluated in m safety Analysis soport may be inormased, er (ii) 12 a possibility for am moeident er maatuasties of !i a ditteeems type than any evaluated previessly La the safety Analysis i naport any be erested, er (111) if the margin of safety as defined la the haces for any Technical Spesification is reduced. Essed gea the above avstantions, it osa be demonstrated that plaat operation with valve va9918 in the cIasaD position to isolate the Dost from the potential met pose pipe leak and the stated osapeasatory actions in place does an. unreviewed safety question as defined by 10cyRso.se because auch of the sevea questione presented below asa be j appropriately answereds 1) Does the proposed activity. increase the emf ty of occurrence of am acatemat prWFionely evaluated in the SARF The proposed activity involves the a train of the EDS fuel oil system. FSAR section 15.10 describes the unit response to a station blackout event. The probability of a station blaskout has not been ineressed sinoa the operators are capable of assuring that valve V17316 will be opened prior to the starting of DOTP 23. Therefore, i i g.d ggcp igp top : a 1 Ho Suapi sad 81 on 1 25 weer:01 S651-T2-40 a
1 ! l'. p. 0 e n = >. 1 Pass 8etle y there is ao' increase in the probability of occurrence of aa mooident jL previously ammirsed in the saa. 3) Doee the proposed activity laarease the consequenome of am l )f emetemet previously evaluated in the BART p i The coas r :: et as moeident previously evaluated in the sak have i act beenlaoreased slase the performasse and operaties of the as see J' will met be impacted by this change. Additionally, this change will j; not sreste a new path for uscentrolled radioactive releasse and will L met adversely affect any radiation nomitoring equipment er equipment j-whist is relied upon to mitigate radielegical consequences of am aeoident. 4 3) Does the proposed eativity increase the liity of occurrenom G, of a meltumation of equipment important to safety previously 1' evaluated in the sant !!j; The proposed activity slightly alters the method for initiating fuel i: flew from the Doefs to the EDS Day Tanks. Yelve T17216 is seraally j a &OCEED OpIN valve that does not require any actuaties in order to.. ensure a flev path from the Dosts to the ;23 EDS day tanks. This evaluaties allers V17216 to be placed in the clos 3D poeikien provided the identified compensatory actions are implemented. These eenpensatery actions assure the reliability of the 3De fuel oil supply. Additiemally, omoe vi7 sis is opened, the fuel oil tran6fer j b system fumetimes as originally designed. 11 As identified is sectica 8 of this evaluation, the failare of V17316 to opea (due to either valve or operator failure) is possible. such
- 4 a failure would result in the less of the SE EDS due to fuel starvation after apprezimately two hours of operation.
A risk asessagent was conducted by ppL's PsA group to dotaraine the change i in the reliability of the a side alsotrical power system following l implemmataties of the specified campensatory actions. Since the EDG j system is caly required to perform its safety fumaties following a l loss of effsite pesar to the safety electrical buses, failures of the j' syates were takse in conjunction with a less of offsite poser. l i In the proposed configuration, the change in frequemey of a lese of L the B side electrieal power is slightly increased; however, this i small ineresse6 is met seasidered significant when eenpled with the j fast thedt plant procedures will be modified to provide for operaters who wi13:bs specially instructed to open vi7 sis as seen as possible and withiai at alastes after an unplanned start of the 23 EDO. Based on the 'above, it aan be,ooncluded that the probability of occurrence of a unifuastion of equipment important to safety previously i evaluated in the safety analysis report has not been increased. d) Dome the proposed activity lacrease the consequences of a l ma12hnactice of equipment important to safety provinn=1y evaluated in the BARy The esasequemeem of a malfanotion 'of equipeant important to safety ^ 'previously evaluated in the SAR have not been increased since the most limiting failure would result la.the 1e " et a single EDS which i, is an analysed event. No other safety systems or equipment required l-i 60*d 229P 19P LOP I 831 h0 tuse: sam *T:n1 SS RMP:0T S661-TE-40
N - e ~h-JPN-F5LES6413, her. 4 Page9atle 1 j, for aeoissat' mitigation er radiaties semitoring are impseted. ) E) Does the poned activity exwate the possibility of an emaident j of a C type than any pawviously evaluated in the AAnt I' a failure medes and effects analysis has been performed for the ] proposed estivity. This analysis (see section 5) has identified two j potential 2ailures which would result in the failure of va7215 to j opea. Failure of the valve to open would, after apprestaately two heers, result in the less of the as koe due to fuel starvaties. The j, less of a stagle EDS is an analysed event. No other failure modes have been identified for the proposed activity. Based on the above, 4 1 the possibility of an accident of a differeat type them any i previously evaluated in the safety analysis report does met esist. 6)
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Does the proposed activity armate the possibility of a different type of aalfunction of equipamat important to safety than any j previously evaluated in the BAnt 3 ; li! As stated la discussions above, a failure modes and offsets analysis has been performed for the proposed activity. This analysin has t identified two potential failures which would result la the failure of V17315 to spea. Such a failure would ultimately' result in the less of the 25 EDS, due.te feel starvation. The less of-a single SDS i is an analysed event. No other failure modes have been identified j for the proposed activity. Additionally, easept for the operater i acties to initially open V17318, the operation of the feel 011 ji transfer system is met impacted. Ne other systems are effected by { the proposed activity. Based en the above, the possibility of a salfumation of equ t important.te safety of a different type thea any previously usted in the safety analysis report has not been l, orested. s 7) Does the proposed activity moduas the maryda of safety as datised ) in the basis for any technical specificatiaat i l The proposed activity does act reduos the margia of safety as defined j in the basis for any teohaisal specifiestien siaea the sed jj activity does met impact EDS operability (Technical Specif tiens 3.8.1.1 and 3.8.1.2). The oompensatory actions required by this evaluaties ensure a reliable supply of fuel oil to the day tanks of l4 the 23 EDS. and, upon the opening of V17216, the fuel oil transfer system vilk fumatism as designed. comm.nartur i [ 10CFhse.59 allows changes to a facility as deseribed in the s&R if j they de not involve an unreviewed safety question er if a change la the Teehaisal spesifloations is met required. as sheva in the preceding sections, the proposed shaage does met involve an unreviewed i safe question because each esmeera as pened by lacrR50.S9 that l Porta to unreviewed safety questions saa be appropriately answered j and a change to a Technical specification is met required; therefore, j prior une approval is met required. I, s I t OT*d ZE9P 19e LOP - a31f50 tuapisaag ason *S WOOS _:0T.-S66*-TE-40, _
T1'd 10101 i ', ! j.
- j JPN PELSEN545413, Rev. 4 Fees 14 etle i
9.0 natians magniraq j 1. Operating persommel shall be instrusted to ensure that Y1? tis is i opened-as soon as possible and within 30 minutes of any unplammed I starting of the 23 EDS. 4
- 2. Prior to closing Y1721s and at least twice each shift, verify that i;
the sa EDS day tanks are sash filled to y,320 gallone (93% full per F i local level indication).
- 3. Talve vi7 sis must be manually opened prior to any plassed operation of the 23 Diesel Generatar er any time fuel oil makeup is required j
for the 23 Diesel Generater day tanks. l
- 4. Review a revise plant procedures and conduet operator trainiaq as appropriate.
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- s. Restore va7sts to its normal Locrap oFue position as seem ma,
4 practical after the completion of any leak repairs or line j replacement and before the ocapletion of the next refueling outage. a. j metes' This evaluation does not address system operability and plant operation during repair or replacement of the subject l1 pipe. !g 10.0 mararammam 11 l,
- 1) St. Luale Walt 2 Ys12, Amendment 9 jj'
- 2) St. IAnale Unit a Teobaioal spesifications, Amendment 75 j
- 3).7FN-PSL-SENS-92-014, Rev. 0
- 4) Calculation pSL-IFJM-90-025, Rev. 3 S) St&R 954713 l
s) contamination Assessment Report st. Lucie Power Plant Unit 2 gr eenerator Diesel Peel storage Tanks, htlanta Testing & Ragia.erlag,septeam.,.,194 'j
- 7) Drawing 3998-6-174, Rev. 14 i =
- 8) Drawing 1998-4-044, St. 1, Rev. 27 l
- 9) Drawing 2998=G-096, St. 2, Rev. 5 y
i 4 j 11.0 Attachmenta Wome i l a a i i TT*d EE9P 19P LOP T 83tff0 tuap:say aton1 ts WOOG:0T 566T-T2-40
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UNITED STATES 4 / NUCLEAR REGULATORY COMMISSION . ?E ' t. REGloN li j 3 101 MARIETTA STREET. N.W. SUITE 2900 8 ATLANTA. GEORGIA 3023-0139 j. ff ' August 22, 1995 d Florida Power and L(ght Company ATTN: Mr.-J. H. Goldberg . President - Nuclear Division P. O. Box 14000' j Juno Beach, FL 33408-0420 m l
SUBJECT:
NRC INSPECTION REPORT NOS. 50-335/95-14 AND 50'-389/95-14 Gentlemen: q This refers to the inspection conducted on July 2 through July 29, 1995, at t the St. Lucie facility. The purpose of the inspection was to determine whether activities authorized by the license were conducted safely and in accordance with NRC requirements. At the conclusion of the inspection, the ,] findings were discussed with those members of your staff identified in the .I enclosed. report. Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews.with personnel, and observation of activities in progress. .y j Within the scope of the inspection, violations or deviations were not I l identified. 1 4 In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of '3 this letter and its enclosure will be placed in the NRC Public Document Room. Should you have any questions concerning this letter, please contact us. Sincerely, O erry andi, Acting Chief Reactor rojects Branch 2 Division of Reactor Projects Docket Nos. 50-335, 50-389 License Nos. DPR-67, NPF-16 1 ~
Enclosure:
NRC Inspection Report s. 1 cc w/ encl: See page 2 4 ATTACHMENT 2 n.,, a a t ( -- i
~.., FPL 2 cc w/enci:- J: D. A. Sager j Vice President St. Lucie Nuclear Plant P. O. Box 128 Ft. Pierce, FL 34954-0128- ), H. N. Paduano, Manager L Licensing and Special Programs Florida Power and Light Company P. O. Box 14000 - Juno Beach, FL 33408-0420 4 C. L. Burton Plant General Manager St. Lucie Nuclear Plant 4 P. O. Box 128 {g Ft. Pierce, FL 34954-0128 l Robert E. Dawson P1 ant ~ Licensing Manager St. Lucie Nuclear Plant i l' P. O. Box 128 Ft. Pierce, FL 34954-0218 ed li: - J. R. Newman, Esq. M Morgan, Lewis & Bockius - 1800 M Street, NW 4 j Washington,-D. C. 20036' John T. Butler, Esq. Steel, Hector and Davis 4000 Southeast Financial Center Miami, FL_ 33131-2398 - Bill Passetti Office of Radiation Control ll; Department of Health and Rehabilitative Services 1317 Winewood Boulevard ,.o Tallahassee, FL 32399-0700 i Jack Shreve 4 Public Counsel Office of the Public Counsel.
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c/o The Florida Legislature 111 West Madison Avenue, Room 812 Tallahassee, FL-32399-1400-i cc w/enci cont'd:' See page 3 1 .wu+e r e s g as ~su-e < M ~e .44%eae. e a + 1
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- ) ;V 4
- :s. .41 FPL. 3 1 i cc w/enci cont'd: Joe Myers, Director Division of Emergency Preparedness i Department of Community Affairs 2740 Centerview Drive. Tallahassee, FL 32399-2100 Thomas R. L. Kindred County Administrator- . St. Lucie County 2300 Virginia Avenue Ft. Pierce, FL 34982 T Charles B. Brinkman Washington Nuclear Operations j: ABB Combustion Engineering,- Inc. 12300 Twinbrook Parkway, Suite 3300 Rockville, MD 20852 I 1 a f 2 l I 4 4 i2 / a 3.e.< .,a v e3, ,rw..,-**:e %. p.. er -.ga see n e.=-- ww se., r.%<+,%,6
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UNITED STATES NUCLEAR REGULATORY COMMISSION ~~ 'l - REGloN 11 .,")e' E ATL.ANTA, GEORGIA 30323-0199 101 MARIETTA STREET. N.W., SulTE 2900 1-t A N, 1 Report Nos.: 50-335/95-14 and 50-389/95-14 Licensee: Florida Power & Light Co 9250 West Flagler Street Miami, FL 33102 1 ' Docket Nos.: 50-335 and 50-389 License Nos.: DPR-67 and NPF-16 Facility Name: St. Lucie 1 and 2 l Inspection Conducted: July 2 through July 29, 1995 Lead Inspector:
- f E
Fi 4 r R. Prevatti, Senior Reyident Date Signed Inspector M. Miller, Resident Inspector Approved by: 8 ff 1 K. t' aft 61s, Chief D4te Signed i Reactor Projects Section 2B Division of Reactor Projects i-J
SUMMARY
Scope: This routine resident inspection was conducted onsite in the areas of-plant operations review, maintenance observations, surveillance i observations, engineering support, plant support, review of I nonroutine events, followup of previous inspection findings, and + other areas, i Inspections were performed during normal and backshift hours and on weekends.and holidays. Results: Plant' operations area:. Operations continued to perform well. Operator response to a reactor trip on July 8 was excellent. Operations response to deficiencies identified during plant systems walkdowns was i satisfactory, i 1 4,,,. , o P . -. l M.7 M, [
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~ .l 3 + .i-d; .2 T' MaintenanceLand Surveillance area: l Maintenance performance was:found to be good.. Critical maintenance 3 O. on the 18 Auxiliary Feedwater Pump was performed very well; in i . con ras,.allack of proper planning and preparation resulted.in t t increased out of. service time for. preventive maintenance on the 2C Auxiliary Feedwater Pump. A personnel error during main turbine trip' surveillance testing resulted in a trip on Unit 1. - An. I&C, ' procedural weakness was--identified during testing of the 2B. Diesel. Fuel Oil Day. Tanks. i
- i Engineering area
l N Performance-in this area continued to be satisfactory. 4 Plant Support area: d Performance in this area continued to be' satisfactory. In~the areas inspected, violations or deviations were not identified. h L i h e. a .{- f I 4 i D l,3 t a r i 1 (.* ar.. ,e 4 ..s
1 J u j REPORT DETAILS l. Persons Contacted Licensee Employees j R. Ball, Mechanical Maintenance Supervisor
- E. Benkin, Plant Licensing Engineer
- W. Bladow, Site Quality Manager L.'Bossinger, Electrical Maintenance Supervisor H.-Buchanan, Health' Physics Supervisor
- C. Burton, St. Lucie Plant General Manager i
R. Dawson, Licensing Manager J D. Denver, Site-Engineering Manager 'J. Dyer,-Maintenance Quality Control Supervisor d H. Fagley, Construction Services Manager l P. Fincher, Training Manager
- R. Frechette, Chemistry Supervisor K. Heffelfinger, Protection Services Supervisor
- J.- Marchese, Maintenance Manager i
W. Parks, Reactor Engineering Supervisor 1
- C, Pell, Outage Manager
- L. Rogers, Instrument and Control Maintenance Supervisor D.. Sager, St. Lucie Plant Vice President
]
- J. Scarola, Operations Manager i
L J. West, Site Services Manager l C. Wood, Operations Supervisor j W. White, Security Supervisor Other licensee employees contacted included engineers, technicians, operators, mechanics, security force members, and office personnel. NRC Personnel
- M. Miller, Resident Inspector
- R. Prevatte, Senior Resident Inspector
- S. Sandin, Senior Operations Officer, AE00
- Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragraph.
2. Plant Status and Activities a. Unit 1 Unit 1 entered the. inspection period at full power. A reactor trip was experienced on July 8 due to personnel error during a surveill ance. test. The unit achieved criticality on July 11 and was placed back on-line on July -12. The unit remained at full power for the balance of ~ the period. 1 --n ,~ r s e. e-
.~... - - .c b, ( t 2 b. Unit 2 Unit.2 operated at essentially full power throughout the period until a planned power reduction.on' July 23 for condenser waterbox - cleaning. The unit was maintained at approximately 60 to 70 per cent power during the cleaning, and was returned to full power operation on July 28. c. 'NRC Activity K. D. Landis, Acting Chief, Reactor Projects Branch 2, NRC Region II, visited the site on July 14. His activities included meetings i with licensee management and a review of resident inspection . activities. 4 R. P. Carrion of the Division of Radiological Safety and Safeguards, .i NRC Region II.-conducted an inspection of the licensee's chemistry [. program with the NRC Region II Mobile Laboratory on July 17 and 18. 1, His activities are documented in Inspection Report 95-13. 3. Plant Operations a. Plant Tours (71707) 4 The inspectors periodically conducted plant tours to verify that monitoring equipment was recording as. required, equipment was properly tagged, operations personnel were aware of plant conditions, and plant housekeeping efforts were adequate. The inspectors also determined that appropriate radiation controls were properly established, critical clean areas' were being controlled in accordance with procedures, excess equipment or material was stored properly, and combustible materials and debris were disposed of expeditiously. During tours, the inspectors looked for the existence of unusual fluid leaks, piping vibrations, pipe hanger and seismic restraint settings, various valve and breaker positions, 4 equipment caution and danger tags, component positions, adequacy of 1 fire fighting equipment, and instrument calibration dates. Some d -tours were conducted on backshifts. The frequency of plant tours and control room visits by site management was noted. The inspectors routinely conducted main flow path walkdowns of ESF, ECCS, and support systems. Valve, breaker, and switch lineups as. well:as equipment conditions were randomly verified both locally' and in the control room. The following accessible-area ESF system a.nd area walkdowns were made to verify that system lineups were in accordance with licensee requirements for operability and equipment ,j material conditions were satisfactory: i 1) Unit 1 Boric Acid Makeup The inspector found major flowpath valves properly aligned. ~ +,. + ,,,~~_n .m,., n
l~ 3 l 2) Unit 1 Auxiliary Feedwater The inspector found major flowpath valves properly aligned. i corrosion was found breaking through exterior paint on welded 1 joints on either side of V09303 and on the downstream side of V9104. These conditions were brought to the attention of the system engineer for resolution. { Additionally, the inspector examined the governor valve stems of turbine-driven auxiliary feedwater pumps 1C and'2C for { evidence of corrosion that could inhibit free movement as identified in NRC Information Notice 94-66, Supplement 1.. No significant evidence of corrosion was identified on either stem. The inspector discussed the issue of stem corrosion with the AFW system engineer and found that the issue was being considered and tracked under STAR 950496 and that the system engineer was extremely knowledgeable of the issue. 3) Unit 2 Auxiliary Feedwater The inspector performed a walkdown of the Unit 2 AFW System in the CST area, AFW Pump Rooms Steam Trestle area, and the Unit 2 Control Room. All valves in the above areas were in the proper position for current plant conditions. General and g specific comments are itemized below. I a) General Comments: 1 (1) Nameplate identification inconsistent with description in operating procedure. b) Operating Procedure No. 2-0700022, Rev 35, " Auxiliary Feedwater - Normal Operation:" (1) SE-08-1 and V08660 were listed as located in the 2C i AFW Pump Room on the alignment of Steam Supply System when, in fact, they were in the 2A/2B AFW Pump Room. (2) V09149, V09150, V09542, V09543, V09313, V09314, V09540, V09541, V09133, V09134, V09544, V09545, V09155, V09156, V09546, V09547 were LOCKED CLOSED valves. Initial lineup per the OP was CLOSED only. (3) V09540 and V09541 were LOCKED CLOSED with no valve label or position tag attached. They appeared to be replacement valves.
- These conditions were referred to the licensee for correction.
~.
4 4 4) Unit 2 Component Cooling Water The inspector verified the major CCW flow paths, reviewed d applicable procedures and walked down the system in the CCW Surge Tank area. Unit 2 Control Room HVAC area and the CCW structure. All valves in the above areas were in the proper position for current plant conditions. General and specific comments are itemized below. 4 a) General Comments: (1) Nameplate identification inconsistent with descriptions in the system operating proce' dure. 1 (2) Description of valves differ between Administrative Procedure No. 2-0010123, Rev 67, " Administrative Control of Valves, Locks and Switches," Appendix I J and Operating Procedure No. 2-0310020, Rev 32, " Component Cooling Water - Normal Operation." (3) Tag missing on SH21339 (8" Drain SS-21-18) ICW System b) Operating Procedure No. 2-0310020, Rev 32, " Component Cooling Water - Normal Operation:" (1) V14101 & V15536 were initially aligned to the CLOSED position; however, both had a handwheel locking device installed with no associated tag indicating ] LOCKED CLOSED. (2) Line 4"-FP-126 upstream of V15536 (Fire Protection System to CCW surge tank) painted blue instead of red as on Unit 1. (3) V14559 (LS-14-6B lower isol) omitted from initial i
- lineup, i
(4) V14438 (2A CCW HX outlet piping high point vent) omitted from initial lineup. i (5) S814439 was initially aligned to the closed position; i however, a handwheel locking device was installed with an associated tag indicating LOCKED CLOSED. This valve was also shown in Administrative Procedure No. 2-0010123, Rev 67, " Administrative Control of Valves, Locks and Switches," Appendix I as LOCKED CLOSED. (6) V14187 (Chemical Feed Tank outlet) tag not attached. 1 (7) V14188 did not have a LOCKED CLOSED tag as shown in j the initial alignment.
[ c j u 5-q j, c) Off-Normal Operating Procedure No. 2-0030131, Rev 49, i l 7 " Plant Annunciator Summary:" j (1). Identified sensing element for; alarm S-12 as PT " 8B, v5ce PIS-14-8B as indicated on CWD. 7 _ (2)' Identified sensing element for alarm S-42-as TIS ~ 29-281/282, vice TIS-14-29-181/lB2 as indicated on } CWD. { l. (3) Identified control room indication as " Check FIS 10A on RTGB-206" vice FIS-14-108 for alarm S-25. -f d) FSAR Table 9.2-7, " Component Cooling Water System 1 Instrumentation Application:" lj (1) Identified CCW Hx Shell Side Outlet Radiation -i Recorders as RR-2G-1,-2, vice RR-26-1,-2 as shown on 1 CWD. i .(2) Identified Fuel Pool HX Outlet Temperature Tag Number as TE-14-2, vice TE-14-20 as shown on CWD. (3) Identified RCP & Motor Cooling Water Outlet total Combined Flow tag number as FIS-14-15F, vice FIS ! 15B and the instrument range as 0-1500 gpa vice 0-i 2000 gpm. L (4) Identified RCP & Motor Cooling Water Outlet Seal i l' Cooler HX Tag Number as TDIS, vice TIS. These conditions were referred to the licensee for correction. i b. Plant Operations Review (71707) A I i The inspectors periodically reviewed shift logs and operations l' records, including data sheets, instrument traces, and records of U equipment malfunctions. This review included control room logs and auxiliary logs, operating orders, standing orders, jumper logs, and i equipment tagout records. The inspectors. routinely observed operator alertness and demeanor during plant tours. They observed. and evaluated control room staffing, control room access, and i - . operator performance during routine operations. The inspectors conducted random off-hours inspections to ensure that operations and i security performance remained at acceptable levels. Shift turnovers were observed to~ verify that they were conducted in accordance with
- r' approved licensee procedures. Control room annunciator status was verified.
Except as noted below, no deficiencies were observed.
- ..a i:
~
- u. 4. p a n.
.-.n.
I'.. 6 3 1) Vehicle Accident in Plant Discharge Canal On July 9, an automobile was inadvertently driven into the plant discharge canal. The automobile was occupied by three teenagers, who later reported that they were looking for a place to surf. The occupants escaped by crawling out of the windows just prior to the vehicle being sucked into the'12 discharge pipe which routes water from the discharge canal, under the beach, into the Atlantic Ocean. The automobile subsequently became lodged in the discharge pipe i at a "Y" which split the 12' pipe into two discharge paths. 3 The obstruction created by the vehicle did not adversely affect j safety at the facility, as a 16' pipe also existed parallel to the 12' pipe. The combined discharge capacity was more than sufficient to pass the Effluent from both units' ICW pumps i without raising discharge canal levels to a level which would [ have resulted in a spillover of water into the adjoining mangroves. The vehicle was removed by a combination of divers, who repositioned the vehicle, and a tug boat, which pulled the vehicle from the pipe. The vehicle was subsequently raised and removed from the area. y 4 2) Unit 1 Restart e The inspector observed activities associated with the approach to criticality of Unit 1 on July 11. The evolution was supported by a reactivity manager, Reactor Engineering, and l plant management. The inspector verified that ECCs were prepared correctly and were within periods of applicability, that a 1/M plot was being prepared and maintained, and that control room staffing was adequate and controlled.
- Overall,
-{ the evolution was performed in a professional manner. The unit was placed on-line at 12:35 a.m. on July 12. 4 3) CEDM Cooling Fan Failure On July 22, Unit I control room operators noted that HVE-21B, the B CEDM cooling fan, had tripped off and that HVE-21A, the standby fan, had started. Subsequent testing indicated that the motor for HVE-218 would start and run; however, amperage readings indicated the fan to be running at no-load conditions. A containment entry and inspection revealed that the fan had failed catastrophically, resulting in a low air flow trip. The fan in question was one of two designed to draw air from the reactor. cavity around the CEDMs, pass the air through . coolers, and discharge it to the containment environment. One fan was required at all times for power operation, and a loss i of both fans required the unit to be subcritical within 45 ) ~. -
_ _ _ _ _ _ _ _.. - ~ _. 4 tO, i i. i 7 i . minutes per ONOP 2-2000030.-Rev:9 " Loss of Reactor Cavity,- - Reactor Support, CEDM, or Containment Cooling Fans." .P The failure resulted in the cocking of the fan.at an angle from ~ j horizontal', cocking of the motor shaft / fan shaft at the p coupling, damage to the variable vane linkage and supports,- damage of pitot-tubes in the discharge plenum,. and damage to' i i pillow block bearings supporting the motor /pumo union..- At the j point of failure, parts were dislodged and thrown from the 3 - unit, creating holes in the fan shroud and in the screen which covered the fan discharge. The licensee found debris scattered . about the area surrounding the fan. The debris which was ejected did not damage adjacent equipment. p At the close of the inspection period, the licensee was attempting to determine root causes and corrective actions. [ Corrective action options included repair at reduced power, repair during a shutdown, and repair during the upcoming Unit 2 refueling outage. j c. Plant Housekeeping (71707) p Storage of material and components, and cleanliness conditions of iL various areas throughout the facility were observed to determine ,i. whether safety and/or fire hazards existed. No violations or i deviations were identified.. 1 d. Clearances (71707) I 'The inspector reviewed clearances 2-95-04-052, 2-95-06-106,and 2 1 F 06-095. All tags were in place and components were found to be correctly positioned, i e. Technical Specification Compliance (71707) i; \\ Licensee compliance with selected TS LCOs was verified. This i !4 included the review of selected surveillance test results. These verifications were accomplished by direct observation of monitoring instrumentation, valve positions, and switch positions, and by review of completed logs and records. Instrumentation and recorder 4' traces were observed for abnormalities. The licensee's compliance with~LCO action statements was reviewed on selected occurrences as they. happened. The inspectors verified that related plant procedures' in use were adequate, complete,. and included the most l 'recent revisions. 1) Elevated Sea Water Temperature on July 7, the licensee noted that increased sea water temperatures were approaching the operating limits for the Unit 2-ICW/CCW heat exchangers. Sea water temperature had reached f f w d \\
lc 'i i ]- [ 8 l 1 approximately 87*F. Control. room operating curves for the heat I-exchangers, which plotted maximum allowable intake temperature against existing heat exchanger differential. pressure, were clamped such that intake temperatures in excess'of. 88*F would i result in heat exchanger inoperability. Dual heat exchanger inoperability would have necessitated entry into TS 3.0.3, t requiring a unit shutdown. j i i -The licensee's immediate actions were to check the calibration of the installed temperature indicators on the B heat exchanger (the higher reading of the two) and to install a more accurate, digital, temperature indicator in its place..The inspector observed portions of the calibration and data gathering effort and noted good involvement by the NPS, who sought to ensure that limits were not being violated. The M&TE espioyed for the measurements was verified to be within its calibration interval. The inspector. spoke to control room operators about l the issue and found that they had been issued clear instructions to commence a unit shutdown should temperature exceed 88'F. The more accurate temperature instruments indicated that intake 3 temperature plateaued at approximately 87*F. Concurrently, Engineering began to develop new operating curves which E incorporated actual heat exchanger performance data (e.g. 1'. number of tubes plugged, actual pump degradation values) to' arrive at new temperature / flow relationships. As a result, l Engineering determined that the maximum allowable temperatures f for each heat exchanger exceeded 89'F at conditions of greatest fl ow. The inspector discussed the methodologies employed in 3 deriving the curves with Engineering personnel and found them i ; to be acceptable. f. Effectiveness of Licensee Controls in Identifying, Resolving, and ij Preventing Problems j i 1) QA Audit-Review (40500) 1 i a) The inspector reviewed Q.A. Audit QSL-OPS-95-14 ) ' Corrective Action" dated June 29, 1995. This audit evaluated the implementation and effectiveness of the plant's corrective action program. The report found that ~ the program was effectively implemented but identified. ,~ three areas that needed improvement. These included: I e The database did not provide accurate information regarding the responsibility for and current status of pending corrective actions. Changes that occur in status were not always communicated to the STAR 1 Coordinator. a ym Jg 6,m%.h e.v ~
p. i 1 J 9 T e Several instances were identified where STARS - i requiring work or repair on ASME Section XI ,j. components were not routed to the ANII or ISI Coordinator. j: The authentication process for STARS'that become e-i ' quality records was not clearly delineated. This i resulted in some STARS in the. quality recordt system not meeting procedural and quality records i 4' requirements. t. The audit appeared to be detailed and provided management with a clear understanding of the current STAR system. status.
- I b)
The inspector reviewed QA Audit QSL-0PS-95-13, which [I summarized performance monitoring activities in the areas of ILRT/LLRT programs, CMM, corrections of discrepant . field conditions, Maintenance Department corrective actions, M&TE programs, and protected area controls. In r f-general, the audit found the subject activities to be performed satisfactorily. The inspector noted that a number of minor changes in M&TE-control and storage methods resulted from one of the PHONS and that the nature fi of the changes appeared to offer opportunities for greater control of M&TE. The inspector concluded that the audit j. was both detailed and multidisciplinary. 2) Post-Trip Review (92901) 4 4 The inspector attended a meeting, conducted on July 21 by Operations management, which discussed the Unit 1 High Pressure - 4 trip discussed in paragraph 4 b, below. This was the second such meeting following an automatic trip, and was designed to 3 elicit comments from plant operations and support personnel on ways to avoid similar trips.in the future. Presentations covered the circumstances surrounding the event, the effect on the unit, preliminary lessons learned and an open discussion of ,4 options to prevent recurrence. The meeting was heavily attended and input and exchanges were frank. The inspector t j' concluded that this practice continues to provide plant [ management with practical. options for reducing the number of automatic-trips in the future. f g. Followup of Operations LERs (90712) (Closed) LER 50-389/94-006, Rev 1, " Trip' Circuit Breaker Failure due 4 i .to'a Broken Piece of Phenolic B1cck Lodged in the Trip Latch Mechanism" _. - _. _.,.., ' I~ Y ~ ~ ~ '
i T 10 The licensee provided the subject LER as informational following the. failure of a TCB to open during RPS logic matrix testing in July, 1994. The incident which prompted the LER is described in IR 94-15. The licensee's corrective actions ir"olved a replacement of the subject TCB, an inspection of the remaining Unit 2 TCBs and CEA MG output breakers, an inspection of Unit 1 TCBs-(discussed in IR 94-24), and an evaluation.of the use of a locking compound'on cutoff switch phenolic block screws to prevent the backing out of the screws (believed to be responsible for the subject failure). The licensee's corrective actions have been completed. No similar conditions were noted in TCB inspections and no loose screws were found. The licensee and the vendor concluded that the application i of locking compounds was not necessary. The licensee determined i that routine, periodic, inspections would suffice to detect loosening of the subject screw. The inspector. concluded that the-licensee's actions were appropriate to the circumstances. q Revision 1 to this LER also documented a failure to perform a TS required shutdown as a result of an inoperable TCB channel. This aspect of the event was documented as VIO 94-15-01. The licensee's a corrective actions were found to be satisfactory and the violation was closed in IR 94-24. This item is closed. i h. Self Contained Breathing Apparatus (SCBA) Needs and Availability i Survey (71707, 64704) The following information was provided by the licensee in response i to a questionnaire prepared by NRC Region II: 1) Facility Name. St. Lucie Nuclear Power Plant 4 2) Event (s) which require operators in the control room to wear SCBA to safely operate / shutdown the plant. FSAR states chlorine but chlorine is no longer stored or used onsite. 3). For the limiting event, does the licensee have SCBAs available for each staff member filling a required position for cperation or safe shutdown? 5 SCBAs stored in each Control Room. 4) Are all staff members filling required positions for operations or safe shutdown SCBA qualified? No, but licensee has plans that will qualify required Operations personnel by July 31, 1995. ,l A +- "?- ?'
- - =. (4 .e l ^ I. ,2 q., 11 I l- '5) Are SCBAs readily available at required use location, p B Yes
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6) Have provisions been.provided for special needs associated with SCBA use, i.e., eye glasses with face mask inserts. i;ip No, on eye wear. Licensee will correct by July 31,.1995. [ 7) What is the minimum number of spare air bottles for each user. None provided in Control Room. Stored in fire house and RCA. l 8) Has the licensee established plans to protect personnel not 4 assigned a SCBA?: lj; Yes. . If emergency responder will be SCBA qualified. ff 9) Does the licensee have SCBAs available for NRC use? None specifically assigned to NRC, but available for issue at l HP. 10) Initials of-each resident and indicate if he/she is SCBA-qualified. 1 c RLP - Yes, MSM - Yes 11) If not qualified, discuss steps necessary to have residents SCBA qualified with your Branch Chief. N/A 1
- 12) Coment field.
] N Chlorine not onsite - FSAR will be corrected next update. U 4. Maintenance and Surveillance a. Maintenance Observations (62703) l \\ Station maintenance activities involving selected safety-related systems and components were observed / reviewed to ascertain that they were conducted in accordance with requirements.. The following items were considered during this review: LCOs were met;. activities were accomplished using approved procedures; functional tests and/or o-calibrations were perfomed prior to returning components or systems to service; quality control records were maintained; activities were S' accomplished by qualified personnel; parts and materials used were properly certified; and radiological controls were implemented as l required. -Work requests were reviewed to determine the status of-outstanding jobs and to ensure that priority was assigned to safety-L n,.,,- ~,
L 4 n i Ej 12 = l - related equipment. Portions of the following maintenance activities were observed: A; 1) 2C Auxiliary feedwater Pump Preventive Maintenance 1 The inspector observed an oil change on the 2C AFP,' conducted-i1 per PWO 62/4389.. Work was performed in accordance with 2-M- !I
- 0018, Rev 42, " Mechanical Maintenance Safety-Related Preventive i
8 Maintenance Program." - The inspector verified that proper, replacement oil.was used, that the old oil was free of visible contaminants, that :the _ final oil level was ' adequate, and that I the new oil filter was a direct replacement for the old one. i... The inspector also observed the lubrication of the turbine's ' ij trip throttle linkage, performed under PWO 62/4421, and j]. verified that the proper grease and graphite spray was used. 1l i The inspector found that the quality of the. work performed was satisfactory; however, the timeliness of the work was found to suffer from inadequate prior planning. The work had been scheduled to begin at midnight on July 18. In support of the evolution, Operations declared the subject AFP OOS at 9:20 p.m. on July 17. At 1:00 a.m., an electrician arrived at the work 2. site to disconnect a lube oil _ innersion heater which required 2 removal for the oil change to take place. This task was completed in approximately five minutes. At approximately 3:10 i - a.m., mechanics arrived to perform the oil change. As a L result, the subject pump was out of service for approximately six hours before the subject task was begun in earnest. l The inspector discussed the timeliness of the maintenance with Maintenance Supervision, who stated that the personnel' involved F in the oil change had questioned a procedure revision which changed the specification of the lubricating oil from that used the last time they had performed the task. Additional 'q-complications were experienced in employing the licensee's new-PASSPORT system to obtain spare bottles and jugs to support the
- l work.
It was acknowledged in these discussions that the job
- 1 was not properly pre-planned / pre-staged, and that the confusion could have been dealt with prior to the initiation of work.
Given the licensee's development of a critical (on-line) maintenance process, the inspector reviewed AP 0010460, Rev 3, " Critical Maintenance Management." In general, the precedure required that work on TS equipment, involving a voluntary
- entrance into a TS AS, be preplanned and expedited. However, i 9; the inspector noted that section 3.1.3 of the subject procedure stated that the procedure need not apply to " Routine preventive maintenance.on equipment required more frequently than 18 months that is not risk significant..." The subject maintenance activity constituted a quarterly PM and therefore was outside the requirements.of the procedure. The inspector discussed the issue with licensee management, who. acknowledged y
Wy m: c3 *+.e f any N = 4 b ** * '~ ' ' * " * ' " ? Y'#'
i. .? IJ 4 i i; 1-13 1i 11-the apparent dichotomy between the CMM process's mandate that time in a TS AS be minimized for some maintenance evolutions
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4 but not for others. The licensee stated that they would 2 consider the issue. l ) The' inspector concluded that no regulation was violated, as the licensee was well within the A0T for the 2C AFP and the: d maintenance in question was performed satisfactorily and within i ' the bound of the licensee's programs and procedures. However,
- (
the inspector,found that preplanning for the evolution was poor
- I and unnecessarily increased the out of service time for the 2C AFP.
qh 2) Auxiliary Feedwater Pump 1B Critical Maintenance ei.' The inspector observec maintenance activities performed on the IB AFP on July 20. The work was conducted under the guidance of AP 0010460. Rev 3. " Critical. Maintenance Management." i Specific observed activities included: o PWO 61/4933 - Replacement of pump bearing Trico oilers with indicating sight glasses and installation of oil sample test fittings. The replacement was conducted per 1 !j MP-09.01, " Auxiliary Feedwater Pumps-1A and 18 Disassembly, inspection, and Reassembly Mechanical 14 Maintenance," and Procurement Engineering evaluation ? 036912. The inspector verified that the installation was L conducted satisfactorily and in accordance the governing documents. !~ e PWO 61/4974 - IB AFP coupling and thrust bearing checks. The subject activity was conducted under 1-MP-09.01, " Auxiliary Feedwater Pumps IA and'18 Disassembly, Inspection, and Reassembly Mechanical Maintenance." The 1 inspector observed coupling disassembly and cleanup, pump j; thrust bearing endplay measurement, coupling reassembly j:' and final torquing. The inspector noted that pump endplay was acceptable (.006") and that the mechanics perfoming l the work properly reassembled and torqued the pump coupling. The torque wrench was verified to be in i calibration. i 1 Overall, the inspector found that the maintenance evolution 'was performed very well. Jobs were worked concurrently, QC coverage was detailed and thorough, parts and tools were i
- I adequately prestaged, and the evolution was completed expeditiously. The inspector noted that the time the component t
was 005, including the post-maintenance' surveillance run, was only approximately eight hours, j 1 4 .e:- o-wt >.-* - + N,a . M ^w,,
- n v
- r e
v%e i 4 '*o.~w 49-M.-
g a 1j . 14 ] '3) ~PWO 64/4966 - Unit 2 Plant Vent WRGM Loss.of Counts i The inspector observed portions of th'e. troubleshooting effort l 'r in response to a failure of the Unit 2 WRGM. -!&C personnel performing the. evolution were found to be very knowledgeable of {- the equipment's construction-and operation. -Troubleshooting was methodical and thorough. M&TE used in the effort was-i verified to be within its calibration interval. The source of )] the failure was determined to be a high voltage power supply to
- 1 the unit's detector.
-T l{ b. Surveillance Observations (61726) }} Various plant operations were verified to comply with selected TS !i requirements. Typical of these were confirmation of TS compliance-f for' reactor coolant chemistry, RWT conditions, containment pressure, control room ventilation, and AC and DC electrical sources. The j .l inspectors verified that testing was performed in accordance with 4 adequate procedures, test instrumentation was calibrated, LCOs were
- j. '
met, removal and restoration of the affected components were i accomplished properly, test results met requirements and were reviewed by personnel other than the individual. directing the test, and that any deficiencies identified during the testing were !T properly reviewed and resolved by appropriate management personnel. l The following surveillance tests were observed: 1) OP 1-0030150, Rev 74, " Secondary Plant Operating Checks and Tests, Section 8.2 through 8.8 Turbine Trip Test." q. The inspector attended the prejob briefing and found that the c; 'y procedural steps, requirements and precautions were discussed j in detail with~all personnel involved in the test, d The inspector then observed the overspeed, thrust bearing and low vacuum trip tests. The low bearing oil pressure trip could
- ,I not be done since valve V22174, low bearing oil pressure trip i
drain valve could not be operated. PWO #74457 was attached to 1, the valve indicating that work was needed. The other above tests were completed satisfactorily. U The operator then proceeded to test the 20/ET, EH Fluid Trip Header Solenoid valve and the-20-1/0PC and 20-2/0PC Overspeed Protection Solenoid valves. This test consisted of opening the 4 EH test header valves to the solenoid under test; unlocking,and closing the EH inlet' isolation valve under test; inserting and turning the trip test key. C This test was completed satisfactorily on 20/ET. When the . second solenoid valve was tested, the operator opened the EH + test header valve, V22493, and unlocked, but'did not close, the . solenoid inlet isolation valve V22482 as required by the 4 . procedure. After unlocking and removing the lock he laid down e
4 1 i i 1 15 i r the lock, read the procedure, and then inserted the test key into the 20-1/0PC test switch and turned it to the test l position. A loud noise was noted as the governor valves went shut, the turbine tripped, and the main steam safety valves,
- opened, i
j The inspector and the NWE then went to Unit.I control room. In the control. room, the. operators responded to the event as j required by E0P-01, " Standard Post Trip Actions." All rods inserted and equipment responded to the event as designed. The i reactor tripped on High. Pressurizer Pressure as a result of the i Governor and Reheat valves going shut. Steam Generator "A" experienced a high level, but operator action isolated feed and ji the level was restored to normal. Overall, operator response
- i to the event was considered excellent.
j The NLO perfonning the surveillance test openly acknowledged that he inadvertently failed to close the EH inlet isolation 4 valve V22482 per procedural step 8.6.5.(B) while performing the solenoid valve tests and that this resulted in tripping the l unit. The NWE supervising the test stated that he became too j~ involved in radio communications with the cos trol room and did i not verify that each step was completed in sequence. i j ' The inspector also noted that procedural step 8.6.5.B and Lj several other steps contained two required actions in one j - procedural step and that this may have led to the error. He l also noted that the use of_ hand held radios vice sound powered heao sets for communications may have been a contributing factor. l The unit was placed in. a stable plant condition using 1-EOP-02, - " Reactor Trip Recovery." A decision was then made to accomplish several outstanding maintenance activities prior to j plant restart. This work included: e Relocate Channel "D" NIS jumper from the control room to the Reactor Building Keyway area e Rework 3 CEA reed switches e Repair IA FW Regulating valve e Inspect / repair RCP vibration probe e Repair RPS Channel "C" Wide Range NIS (failed low after j reactortrip) e Repair Main Generator excitation power supply e Repair. loose connection on IB Motor Generator set ^ e Stroke test MV-G 8 e Repair MV-09-6 e Cleaning Main condenser Water boxes Al and B2 e Other minor maintenance activities The.above work activities, except the NIS Channel "C" Wide Range, were completed by the morning of July 9. Completion of 1 ^
~ _ F n e s ,) i l P, ,p 16 i. t lLi the repair to NIS Channel "C" Wide Range, and concerns relating .to high discharge canal levels resulting from unusually high ] tides and an. automobile' lodged in a discharge canal pipe' 4 - (discussed in paragraph 3.b.1), delayed reactor restart until July 11. The inspector reviewed the above work activities and found them t satisfactory. The. reactor trip package was also reviewed and !L it was determined that all issues had been satisfactorily i. resolved to permit plant restart. Li 2) OP 1-0700050, Rev 50, " Auxiliary feedwater Periodic Test" 3% The inspector observed the surveillance test, conducted per the-i 41 above procedure, on the IB AFP. following CM work discussed in i ,31 paragraph 4.A.2, above. The test involved an ASME Section XI t-code run of the subject pump. The inspector noted that the i l; operator conducting the test locally had procedure in-hand and that M&TE employed for obtaining vibration and temperature data was within its edibration interval. The required time interval was cbse ved prior to data collection (5 minutes), discharge pressure was greater than the minimum specified for i. compliarce with TS (1342 psig), and results were satisfactory (3241.7 ft developed head).
- f 3)
OP 2-22000508, Rev 20, "2B Emergency Diesel Generator Periodic [' Test and General Operating Instructions" l l' The inspector witnessed portions of this test, conducted July i
- 26. The test involved a fast start of the 2B EDG to satisfy TS 1
surveillance requirement 4.8.1.1.2.a.4, which required that the 4 - EDG achieve rated speed and voltage within 10 seconds at least ) l' once per 184 days. l. The inspector witnessed pre-start checks perfomed by the SNP0 and found them to be perfomed satisfactorily with procedure in-hand. The inspector observed the EDG start'and examined the operating machines for signs of previously unidentified leaks. None were noted. The machines started and loaded satisfactorily, with a start time of 9.65 seconds. 4) EDG Day Tank Level Switch Surveillance The inspector observed portions of surveillance tests, perfomed in accordance with 1&C Procedure 2-1400064L, Rev 3,2, " Installed Plant Equipment Calibration (Level)," Appendix 8, Tab lo, " Diesel-011 Day Tank Lo/Lo level Verification," to verify day tank level switch setpoints on the 2B EI)G day tanks. The tests were perfomed by attaching tygon tubes to drain valves located, hydraulically, at the bottoms of the day tanks and routing the tubes vertically to the tops'of the tanks. d ~..
l. j;. i l 17 l Rulers were then located next to the tubes to provide local level indication in the tanks to assess alarm setpoints. The test methodology for testing hi/hi level' alarms was to align the temporary standpipes with their respective day tanks and manually operate the tanks' fill solenoid valves to admit fuel until the hi/hi level alarms were received. The inspector l .noted that the I&C personnel performing the tests were sensitive to the fact-that indicated level increase rates would accelerate as the levels approached the tops of the tanks,-as the tanks were horizontally oriented cylinders. Nonetheless, while filling the 2B1 day tank, the level in the tygon tube rose rapidly and resulted in a small spill (approximately two ,I cups) of FO. The spill was quickly terminated, contained to a small area around the day tank, and cleaned up by the IEC t' i personnel performing the test. Additionally, the hi/h1 level alarm did not energize. Upon inspection, it was noted that a PWO tag was hung on the level alarm, indicating inoperability of either the circuit or the sensor. The I&C personnel performing the test acknowledged not checking the PWO tag prior to beginning the test. Testing of the lo/lo level alarms resulted in satisfactory results. The inspector discussed the performance of the test with I&C i personnel, who stated that the hi/hi level alarm did not 1 U energize due to the fact that the 282 day tank hi/hi alarm was energized as a result of performing the same test on it i previously. As the hi/hi level alarms had no reflash capability, the second day tank's alarm could not annunciate. I&C personnel conceded that the governing procedure was inadequate to test the hi/hi alarms as written, and stated that i the procedure would be revised. Possible new test methoc' ologies included: I Testing the second tank's alarm after the first tank's e j~ alarm had cleared due to. engine fuel consumption, or o Performing the test by monitoring level switch output state, as opposed to the alam annunciator I&C personnel stated that the PWO which was written to document hi/h1 level switch inoperability was most probably the result of. a similar failure in a similar test. The inspector concluded that the F0 spill could have been avoided if eith'er the tygon tubing had been run further in elevation above the day tank or if the workers performing the test had recognized that the level switch they were testing would not result in annunciation due to the alam condition in the 282 day tank. In reviewing the governing procedure, the inspector noted the following weaknesses:
4 18 The title for Tab #10 of the procedure (" Diesel Day Tank e Lo/Lo Level Verification") was misleading in that hi/hi level alarm verification was also included. This point was reinforced in the body of the procedure in step B.2 when personnel were directed to place a measurement scale from 20" to 25" up the sight glass, when hi/hi level alarm verification would also require a measurement scale at approximately 34". Personnel performing the observed test showed foresight in extending the measurement scales along the full length of the sight glasses. The procedure directed that tygon tubes be taped to the e top of the day tanks. The physical arrangement of the day I tanks' overflow lines was such that the F0 level could increase approximately l' above the tops of the tanks i prior to the overflow being directed away, increasing the potential for spills. 1 The inspector concluded that the performance of the subject surveillance test suffered from procedural weakness and an inadequate pre-test observation of the component to be tested. 5) Containment Anomalies Inspection - Unit 2 The inspector accompanied Unit 2 NL0s on an inspection of 1 accessible containment areas on July 25. Damage to HVE-218, described in paragraph 3.b.3, above, was noted. The status of a packing leak from V8453, a root valve for B channel SG level and pressure instruments, was inspected and found to be unchanged. Several instances of boric acid buildup on instrument tubing was also noted. Otherwise, no adverse conditions were identified. The inspector found that the NL0s conducting the inspection proceeded swiftly but were thorough in their inspections, allowing for a comprehensive tour while-maintaining dose rates ALARA. 5. Engineering Support (37551) A. Safety Evaluation JPN-PSL-SENS-95-013 The inspector reviewed the subject SE, prepared to allow operation with a manual isolation valve closed in the 2B EDG F0 line from the DOST to the day tanks. The configuration was proposed when the a leak was determined to exist in the underground line between the two 1 tanks. The action was designed to minimize the amount of F0 released to the environment until the leak could be identified and corrected. As a compensatory measure, the licensee proposed dedicating an NLO to the task of opening the closed valve in the event of an EDG start. The licensee calculated that the EDG day tanks contained enough FO to allow 126 minutes of EDG operation at full, load before ~ c
_ _ _ _. _ _ _ _ _. _. _. _ _ _. _... _ _ _ _ _ _ _. _ _ _ _ _ _ _ 7 1 ': j. 19 j ij a transfer of FO was required. The licensee'then specified that the j .NLO would be_ required to open the valve within 20 minutes of an EDG. l 4 start. Procedures were revised to include direction to open the 7 valve on an EDG start, and administrative controls were put in place to ensure that the NLO would not be_ required to perform any other immediate response duties. Additionally, the licensee _ performed a j response time test, placing the operator at the G-2 warehouse (as j. j 1 far away from the EDG as he could credibly ~be in the PA) and jj requiring the NLO to proceed to the valve and open it. The NLO performed this task in approximately seven minutes. j In considering the issue, the licensee employed PRA techniques,to estimate the increase in the risk of the loss of the 2B3 bus due to i ?. a failure of either the operator to open the valve or a failure of Ib the valve to be able-to be opened. The licensee concluded that the increase in probability was approximately 6 percent. However, in
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considering 10 CFR 50.59 criteria, the licensee concluded that no j f increase in the probability of failure of a component important to ld safety was created by the proposed action. The inspector questioned l the licensee on this issue. The licensee explained that a l deterministic conclusion of no increased probability was reached i when the existence of procedural guidance and heightened awareness l l was balanced against the approximate 6 percent increase in failure 1 probability presented by the two new failure modes, ~ i I i l1 In the context of regulatory compliance, the inspector noted that 10 i. CFR 50.59 was written in terms of absolute increases in the i j probabilitias of failure represented by a proposed change. The {j inspector continued to question whether 10 CFR 50.59 criteria could ,' ~ ever be satisfied when new failure modes are imposed on a previously reviewed system (i.e whether added risk, once qualitatively established, could be completely mitigated). The inspector concluded that insufficient guidance existed from a regulatory perspective' to take immediate issue with the licensee's rationale. l Further, the inspector concluded that the licensee had taken prudent !j measures to ensure the continued operability of the 2B EDG while. t ji !q minimizing the F0 leak's effect on the environment. The inspector j l1 referred the question to NRR for resolution. l 6. Plant Support (71750) ~ c F a.. Fire Protection 'l p. During the course of their normal tours,' the inspectors routinely examined facets of the Fire Protection Program. The inspectors 4 'i_ reviewed transient fire loads, flassable' materials storage, !q housekeeping, control hazardous chemicals, ignition source / fire risk l D reduction' efforts, fire protection training, fire protection system surveillance program, fire barriers, fire brigade qualifications, and QA reviews of the program.' No deficiencies were identified. i J f 1 4 d lg- -.) }} j
? I 1 20 j b. Physical Protection l During this inspection, the inspector toured the protected area and noted that the perimeter fence was intact and not compromised by erosion or disrepair. The fence fabric was' secured and barbed wire was angled as required by the licensee's Physical Security Plan (PSP). Isolation zones were maintained on both sides of the barrier and were free of objects which could shield or conceal an individual. The inspector observed personnel and packages entering the protected area were searched either by special purpose detectors or by a i . physical patdown for firearms, explosives and contraband. The processing and escorting of visitors was observed. Vehicles were I searched, escorted, and secured as described in the JSP. Lighting-of the perimeter and of the protected area met the 0.2 foot-candle criteria. i j In conclusion, selected functions and equipment of the security j ~ program were inspected and found to comply with the PSP requirements. c. Radiological Protection Program .j Radiation protection control activities were observed to verify that these activities were in conformance with the facility policies and
- b procedures, and in compliance with regulatory requirements. These observations included-1 e
Entry to and exit from contaminated areas, including step-off pad conditions and disposal of contaminated clothing; e Area postings and controls; e Work activity within radiation, high radiation, and i contaminated areas; ) e Radiation Control Area (RCA) exiting practices; and, l} e Proper wearing of personnel monitoring equipment, protective clothing, and respiratory equipment. g! No violations or deviations were identified. 7. Exit Interview ^ The inspection scope and findings were summarized on July 27, 1995, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection results listed below. Proprietary material is not contained in this report. Dissenting .t comments were not recaived from the licensee. ) i i w ~ i
y :. i 21 TY2g Item Number Status Descriotion LER' 50-389/94-006, Rev 1 Closed " Trip Circuit Breaker Failure due to a Broken Piece of Phenolic Block Lodged in the Trip Latch i Mechanism", paragraph 3.g.2). 8. Abbreviations, Acroisyms, and Initialisms to AE00 Analysis and Evaluation of Operational Data, Office for (NRC) AFP Auxiliary Feedwater Pump AFW Auxiliary Feedwater (system) ALARA As Low as Reasonably Achievable (radiation exposure) ANII Authorized Nuclear Inservice Inspector i ASME Code American Society of Mechanical Engineers Boiler and Pressure Vessel Code i CCW Component Cooling Water i CEA Control Element Assembly CEDM Control Element Drive Mech'anism CIS Containment Isolation System CM Critical Maintenance Management CWD Control Wiring Diagram i DG Diesel Generator t ECC Estimated Critical Concentration I ECCS Emergency Core Cooling System l EDG Emergency Diesel Generator j i E0P Emergency Operating Procedure ,I ESF Engineered Safety Feature i ESFAS Engineered Safety Feature Actuation System F0 Fuel Oil FSAR. Final Safety Analysis Report FW Feedwater. gpe: Gallon (s) Per Minute (flow rate)
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HVAC Heating Ventilation and Air Conditioning L HVE Heating and Ventilating Exhaust (fan, system, etc.) '1 HX Heat Exenanger ,I ICW Intake Cooling l late *
- 1.
ILRT Integrated Leak Rate hat (ing) IR [NRC] Inspection Rupe t ISI' Inservice Inspection (program) 1 JPN (Juno Bat.h) Nuclear Engineering LC0 TS Limiting Condition for Operation LER Licensee Erant Report LLRT Local Leak Rate Test
- i MP Mechanical Maintenance Procedure MV Motorized Valve NIS Nuclear Instrumentation System NLO Non-Licensed Operator NPS Nuclear Plant Supervisor NRR NRC Offt
- e of Nuclear Reactor Regulation i
m. c a. -, .j
f .o 22 NWE Nuclear Watch Engineer ONOP Off Normal Operating Procedure ' 005 Out Of Service OP Operating procedure OPS Operations PMON . Performance Monitoring PORY Power Operated Relief Valve PRA Probabilistic Risk Assessment psig Pounds per square inch (gage) PSL Plant St. Lucie PWO Plant Work Order QA Quality Assurance QC Quality Control 1 QSL Quality Surveillance Letter RCP Reactor Coolant Pump RPS Reactor Protection System RTGB Reactor Turbine Generator Board 'Fuff Refueling Water Tank SCBA Self Contained Breathing Apparatus SG Steam Generator SNP0 Senior ?Jclear Plant (unlicensed] Operator TCB Trip Circuit Breaker TS Technical Specification (s) r-WRGM Wide Range Gas Monitor I j i \\ 1 l t 0 i ~~
Pp t } september 13, 1995' j ST LUCIE j. Integrated Plant Performance Review A. Current Plant Status q; The unit was shutdown on August 1 as a result of Hurricane Erin..A series of problems including RCP seal failure, both PORVs inoperable due j to incorrect assembly, SDC relief valve problems, associated problems 1 with several other relief valves, inadvertent spraydown of the. a containment, catastrophic failure of 18 emergency diesel generator, and i a leaking flange on a pressurizer safety valve have prevented the unit from restarting. With the unit down the licensee planned on performing a large number of operator-work-arounds and correcting other plant deficiencies. The' next refueling outage is scheduled for April 4,1996. Unit 2 was' shutdown on April 25 for approximately seven hours.to replace a main turbine digital electro hydraulic power supply. The unit was downpowered for several days in June and July to clean condenser water boxes. The. unit was shutdown on August 1 as a result of Hurricane Erin. .l It restarted on August 4. Power was reduced from August 17 through 29 i to clean condenser water boxes'and repair various secondary plant deficiencies. The next refueling outage is scheduled for October 9, 1995. B. Manaaement Several management changes were implemented in August and September l 1995.
- l Jim Scarola was reassigned from Manager of Operations to Plant j
l; General Manager. >! i Jeff West was reassigned from Manager of Site Services to Manager of' Operations. l1 Chris Burton was reassigned from Plant General Manager to Manager of Site Services. Lee Rogers was reassigned from Instrument and Control Maintenance Supervisor to Manager of System and Component Engineering. This area was previously technical staff and is being reorganized to focus more on systems and components. P. Fulford was assigned hs Operations and Support Testing Supervisor, a new position in operations that will be responsible for inservice, surveillance, predictive, and post maintenance -testing. R. Olson was promoted to Instrument and Control Ma.intenance Supervisor. ,a
1
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A significant number of other changes were made in saintenance at the supervisory level to place new blood and stronger individuals in supervisory positions. C. Plant Performance The units experienced an unusually high number of reactor trips (4) i between March and June of 1994. Since that time, Unit I experienced two reactor trips, October 1994 and July 1995,-and Unit 2 experienced one trip in February 1995. Unit 2 was down powered in April 1995, and Unit 1 in June 1995 for maintenance. Several other units down powers occurred to clean condenser water boxes and perform other maintenance activities. Both units were shutdown as a result of Hurricane Erin. Overall operator performance was good during the plant trips and power maneuvers. March 4, 1995 - Unit 1 experienced a loss of shutdown cooling for approximately 14 minutes when a hot leg suction valve to the operating shutdown cooling train closed. Root cause has not been established; however, operator - error in manipulating valves is the most likely cause. April 25, 1995 - Unit 2 shutdown for approximately 8 hours to replace a main turbine DEH power supply. July 8, 1995 - Unit I tripped during turbine valve surveillance 5 testing. Returned to power on July 12. l l* August 1, 1995 - Unit I shutdown as a result of Hurricane Erin. Due to equipment problems and personnel errors, the unit remained has remained shutdown. Unit 2 shutdown as a result of Hurricane Erin. Unit was restarted August 4. Power reduction August 17 through 29 to clean condenser water boxes and repair equipment problems. D. performance Indicators The previous six months has shown an increase in personnel errors involving, the failure to follow procedures, inattention to detail, and the failure to maintain awareness of equipment status. The above occurrences and events that have taken place since July 1995, indicated a decrease in over all plant performance. E. Enforcement History No escalated enforcement in 1994. Enforcement conference on inoperable PZR PORV, September 25, 1995. 1
_ _. ~.. _ _ _ 4 5 F. SALE . Period ended May 1992 Period ended January 1994 1 1 1 Operations Maintenance 1 1 Engineering 1 1 f Plant Support 1 Rad. Con. 1 Security. I t Emerg. Prep. I SA/QV 1 G.- 1ll2Q INPO assessment July 1995. Category 1 INP0/WAN0 assessment April 1994 INPO outage assessment scheduled for October 1995 4 ' H. 1994 Precursor Events None l. Alleaations and DOL Cases Nine allegations are open: one fitness.for duty concern; one unattended ARM doors; one. addressed weld repairs conducted on turbine cooling water and component cooling water; one alleged discrimination for reporting safety concerns; one concerning security contractor competency; one concerning crossing training; one concerning failure to notify public of events; one concern indication inadequate receipt inspection; one DOL case - one welding supervisor received negative performance appraisal and was then removed form supervisory position J. Attachments 1. Organization Charts 2. Powee History Profiles 3. Site Integration Matrix Jj. ) i l n T n ,}}