IR 05000336/1990018
| ML20217A112 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/05/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20217A087 | List: |
| References | |
| 50-336-90-18, NUDOCS 9011200186 | |
| Download: ML20217A112 (30) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.: 50-336/90-18 Docket No.: 50-336 License N DPR-65 Licensee: Northeast Nuclear Eneroy Company P.O. Box 270 Hartford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2 Inspection at: Waterford, Connecticut Dates: August 22 through October 1, 1990 Reporting
' Inspector: Peter J. Habighorst, Resident Inspector Inspectors: William.J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector Guy S. Vissing, Nuclear Reactor Regulation, PDI-4 Approved by: beweM b Donald R. Haverkamp, Chit /
su< w ///db Date Reactor Projects Section 4A Division of Reactor Projects Inspection' Summary: Inspection on August 223 _1990 - October 1, 1990 Inspection Report No. 50-336/90-18 Areas Inspected; Routine NRC resident and specialist inspection of plant operations, surveillance, maintenance, previously identified items, engineering / technical support, committee activities, licensee event reports, and special report Results: See Executive Summary i 9011000186 90110$ ' PDR .ADOCK 05000336 O PDC i
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EXECUTIVE SUMMARY MILL'. TONE NUCLEAR POWER STATION, UNIT NO. 2 NRC REGION I INSPECTION REPORT NO. 50-336/90-18 P_lant Operations '
' One automatic reactor trip and one shutdown to support the cycle 10 refueling i outage occurred during the inspection perio j Two unresolved items were opened regarding personnel errors that resulted in an automatic reactor trip and two engineered safety feature actuations. A generic unresolved-item was opened regarding these, and other, events which represented lack of: attention to detail and inadequate configuration control of safety related equipmen Future NRC reviews will consider licensee assessment and corrective action , . A violation was identified for not having the required reactor vessel level monitoring ' system operable for. five and one-half hours during reduced inventory operations.and for failure to implement required administrative controls for a ;
e bypass jumper used in the vessel level monitoring system isolation. T.wo un - 1
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resolved items were closed during this inspection period. These items involved incorporation of one emergency procedure guideline.into an emergency operating .
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. . procedure,. and a revision to the technical specification for core alterations a in relationship.to incore instrumentation remova Licensee strength 4 1n this area.was'. demonstrated by effective-use of the outage support: organization and' daily: outage' planning meeting .
-l Radiologidal= Protection-
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10ne unresolved -item was -opened in this; area to address. licensee- actions to '
.x improve the' effectiveness = of the' post accident sampling system (PASS).. l 7i '
Representative reactor coolant total gas samples were not' acquired from the
. PASS for:the past five years.?
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LSurveillance and Maintenance- , Control -implementation,-and evaluation of neutron loggin'g in the high density tspent fuel pool-poison boxes was adequately completed.;~
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.One unresolved' item was opened in this. area concerning actions to irrprove the . .!
- oil collection capability for. the reactor coolant pump motor, specif tcally, the - 4
. oil lif t pump package. One previously identified item was closed during:this ;
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inspection periodEconcerning correct implementation of acceptance criteria.for J
: , . Heise pressure gauges calibrated in:the meteorology laborator *
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Security Reviews in this area did not identify any noteworthy finding Engineering and Technical Support LThreeLunresolved items were closed during this inspection period. The items involved updated pressurizer cooldown rate evaluations, incorporation of a licensee. commitments to NRC Bulletin 83-03, and development and control of design documentation for the emergency diesel generator saturable transformer A deviation was identified for failure to fully implement previous committed
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actions on~ the spent fuel pool boraflex coupon surveillance program, in that no direct' dosimetry measurements of cumulative gamma exposure were mad uf Safety Assessment / Quality Verification One licensee-identified non-cited violation was documented regarding a missed-
: technical _ specification. surveillance (licensee event report' 90-07). One un-resolved item-was opened.to review root cause and timeliness determinations, lregarding inadvertent engineered safety feature actuations of the enclosure : butiding filtration system (licensee event. report 90-09).
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> TABLE OF CONTENTS P_ age i 1.0 Summary of Facility Activities.............................. 1
- 2.0 Plant Operations (IP 71707/92701/92702/60710)*.............. 2 n
2.1 Control Room Observations.............................. 2 ! ' 2.2 Plant Tours............................................ 2 ! Onsite Followup of Operational Events.................. 2 2.3.1 Automatic Reactor Trip.......................... 2
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2.3.2 - Er.gineered Safety Feature Actuation Events. . . . .. ' 2.4- Control of Outage Activities........................... 7
'2.5 Reactor Vessel Level Monitoring During Reduced . - I n v e n t o ry Op e r a t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 >
2.6 Regional Temporary Instruction 87-01, Control -
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Room. Environment....................................... 10 -
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2.7 Previously Identified Items..........................., 11 - 2.7.1 (Closed) Unresolved Item 88-22-01: Emergency Operating Procedure 2534 Did Not Includeta Step to Eliminate Voiding - in the Reactor Coolant System (RCS) Steam Generator .; Tubes as Per Combustion Engineering Procedure >
. -(CEN-152) Guidelines................................... 11 ; *
2;7.2 (Closed) Unresolved Item 89-05-07: Core Alteration Technical Specification Definition During 't m 'In-Core Instrumentation Remova1........................ 11
19 3.01 Radiological Controls (IP 71707)............................
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i 3.1 ' Posting.and Controls of Radiological Areas............. 12 , 3.21 PIR 90-85,: Post: Accident Sampling-System (PASS) : Failure......................... ...... . . 12: ,
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4.0. Maintenance / Surveillance (IP 62703/61726/92702)............... 13 4 .' 1 Observation of. Maintenance. Activities................... 13 . 4.2 Observation-of Surveillance Activities................. 13
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4.2.1L. Spent: Fuel Pool Poison Box Blackness Testing...'. 14 : 4.3' l ' A' Reactor Coolant f Pump Motor 011 - Leakage. . . . . . . . . . . . . 15: ; e :4.4; NRC:Information Notice'89-90, Pressurizer Safety Lift 1
'Setpoint Shift......................................... 16: <
4.51 Steam Generator' Level Discrepancy...................... 17 iP 4.6; PreviouslycIdentified Items............................ 17- 1
?4.6.1' (Closed) (Unresolved Item 88-24-05: Inadequate' '
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Application;of Acceptance Criteria for Heise Gauge j Ca11bration...................,........................ 17! J t
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Page ; .0 Engineering / Technical Support (IP 92701/71707).............. 18
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5.1 Plant Design Modifications............................. 18
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5.2 Spent Fuel Pool Boraflex Condition..................... 18 5.3 Previously, Identified Items..................... ...... 21 ; 5.3.1 (Closed) Unresolved Item (UNR) 90-06-0?. *
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Pressurizer Cooldown Rate Discrepancy................., 21 .; 5.3.2L (Closed) Unresolved Item (UNR) 89-17-01: ; Servic'e Water. Check Valve In-Service Testing Program.. . 21 i 5.3.3- (Closed) Unresolved Item (UNR) 89-17-05: Design ; ~ Documentation for Emergency Diesel Generator , Saturable 1 Transformers................................. 22 <
- 6.0~~ Security-(IP 717107)........................................ 22 *
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(IP 92702/92700)................-............................-
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7.1 Licensee Event Reports................................. 22 4 7.2 Committee Activit1es'....................................- 24' d j i7.3 Periodic Reports...........................'............ - 24 1
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'8'0'-Management:
. . Meetings (IP130703) .
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m The- NRC Jinspection manual . inspection procedure (!P) or temporary - .
'I instruction;(TI) that was u' sed as' inspection guida.,ce is'11sted for each
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I 4 DETAILS 1.0 Summary of Facility Activities Facility Opezattnr.4 ,. E Millstone Nuclear Power Station Unit 2 (Millstone 2 or the plant) was at full rated power at the beyinring of the inspection period. On August 27, 1990, an automatic reactor trip occurred as a result of personnel erro (See detail 2.3). The plant returned to full rated power on September 4.
' On September 14, the licensee commenced a plant shutdown for the scheduled k +~ cycle 10 refueling outage. At the end of the report period the piant was p in refueling operations (mode 6) with the moisture separator tube bundle f ' replacement as the outage critical path activity. The outage is cur-h rently scheduled to last until November 7, 199 NRC Activities i
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;The inspection activities during this report period included 211 hours of inspection- during normal working hours. In addition, the review of plant - ,
operations was routinely conducted during periods of backshifts (evening
' shifts) and deep backshifts (weekend and midnight shifts). Inspections were performed during 17 backshift hours and during 5 hours of deep w >
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p :A' Region I specialist inspection of security was: conducted between August 27-31, 1990. .Results are.provided in inspection report 50-336/90-17.. '
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A: Region I specialist inspection of_ previously identified items .and 1employeeiconcerns was conducted between-September 10-14, 199 Results
; are provided in inspectionLreport 50-336/90-19.
" A1 Region I specialist inspection'of effluent monitoring was' conducted
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between' September 10-14,<1990. ResultsLare provided in inspectioncreport 50-336/90-20; A Region I' inspection of_non-routine reporting'was conducted between
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September- 24-28,.1990. Results~are provided in inspection' report 50-336/90-8 (A Region I'sp'cialist inspection of the-implementation of 10 CFR 26-
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e requirements-and associated recent events was conducted between September i
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, < 24-28, 1990. Results'are provided .in inspection reports-50-336/90-24,and
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. 2 2.0 Plant Operations 2.1 Control Room Observations
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Control room instruments were observed for correlation between channels, proper functioning, and conformance with Technical k ' i Specifications. Alarm conditions in effect and alarms received in i the control room were discussed with operator The inspector periodically reviewed the night order log, tagout log, plant incident report (PIR) log, key log, and bypass jumper log. Each of the respective logs was discussed with operation department staff. No significant observations were mad , 2.2 Plant Tours The inspector observed plant operations during regular and backshift tours of the following areas: Control Room Containment Vital Switchgear Room Diesel Generator Room Turbine Building Intake Structure Enclosure Building Engineered Safety Feature , Cubicles
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During plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were correctly made, and tu verify correct communication and equipment status. No significant observations were mad .3 Onsite Followup of Operational Events , 2,3.1 - Automatic Reactor Trip
. Summary <
On August 27 at 12:58 a.m...the reactor automatically tripped from 100i rated power as a result of personnel error during the perform-ance of a. reactor protection system (RPS) surveillance. During implementation of procedure SP-26010, " Power Range Safety Channel and Delta T Power Channel Calibration", a reactor operator failed to by-pass channel 'B' of the RPS_as detailed in procedure' step Prior to the performance of the surveillance,-the licensee' entered technical spacification action statement 3.3.1.1 table 3.3-1 action 2b for RPS-channel 'A.' The reason for being_in the action statement-
. was an identified failure of a 120 volt fuse holder in the supply line to the RPS matrix logic power. The failure'of the fuse holder- --deenergized the channel; ' A' RPS logic for local power density (LPD),
high power,.and thermal: margin low pressure (TMLP) trip signals. I preparation to repair the holder, the licensee bypassed the channel
'A' LPD, high power, and TMLP trip signal i
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As detailed in the surveillance procedure, the control room
. operator _ correctly removed the bypass condition for channel 'A' - and placed the parameters in a " tripped" condition. The .
operator f ailed, however, to place channel 'B' RPS signals for high power, TMLP and LPD in bypas Surveillance step 6.6 requires the operator to place the delta T calculator test switch to the " cal check
#2 position." - This action removed the actual RCS cold leg temperature and installed a 10-volt direct current test signal _across a variable
> < 20: kilohm resistor. The resultant test signal output was such that the RPS logic sensed a trip condition, A 2-out-of-4 logic was satisfied and a reactor trip occurre ! Sequence of Events
)i The following lists the chronology of critical plant parameters
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; during-the reactor trip; i Time Event
'A' Local Power Density from 00:57:57.568 !
Bypass-to Trip i
'B' ; Local Power Density Trip 00:58:5.719
- Reactor Trip; . 00:58:5.929 l Reactor Trip Circuit Breakers Open:; 00:58:5.784-5.794
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Normal: Station Service Supply 1 Breaker Opens . _ 00:58:6.053 !
. Reserve Station Service Supp'y .
i Breaker Closed '00:58:6.126 :
. -Operators commenced Emergency ,1 * >0perating Procedure (EOP) 2525 , "St4ndard: Post-Trip Actions" 00:58:10.912- , - 9perators completed E0P 2525 ' ,' . r .1, commenced EOP 2526 E- '! Reactor Trip. Recovery" -01:05.00 . <
JThe; inspector reviewed +he licensee _; e-trip report, post-trip _ I report? sequence of ever t log, and po.,t-trip review summary. The ! review consisted of verification of-reactor protection system !
'i ' response time,_ turbine trip time response; and-the response of picnt ,
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w / parameters:for; steam-generator level and pressure,' pressurizer m > pressure and level'.and reactor coolant system temperatures. The
, ._ ; post-trip safety function status checks were also reviewed. 'The .
QL e plant 1 response.tothetransientwasas-expected,.andoperatorstook the proper actions to stabilize the plan .tindings and' Observations
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Procedure SP-26010 is a daily RPS surveillance conducted within the ,
. control room.- The operator who^ committed the error had successfully "
completed the survelliance the' previous day with the same initial
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conditions (i.e. RPS channel A in bypass).
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l Licensen orrective actions included operations department meetings
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to -discuw the event L specifics, and a t evision to SP-26010 to verify initial conditions of-the.RPS channe The inspecto; concluded that the event was_ caused by improper implementation of a required surveillance. The licensee event report l
- will be reviewed during a subsequent I;RC inspection, to consider the !
licensee's assessment of the event and its significance in the NRC j determination of the appropriate licensee actions being taken or ! planned to assure proper future procedural adherence. This is con-fe sidered an unresolved item (50-336/90-18-01), 2. Engineered Safety Feature Actuation Events i 4 September 17 Event
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On September 17.-1990, at approximately 2:44 a.m. a containment purge valve isolation signal actuation occurred. At the time of the event 3
the unit:was in a cold shutdown condition.' The licensee reported the
, event.as required per 10 CFR 50.72(b)(2)(ii). ]
h The cause;of the event' was personnel error during a clearance of a J - tagout associated with the containment hydrogen analyzer. Speci-
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fically, a plant equipment operator was tasked-to remove a blue maintenance tag on the vital 120 volt ac. distribution center (VA20) i c breaker #16', Breaker #16 sJpplies electrical-power to the 'B' I hydrogen analyzer control ps.nel C87. The operator opened breaker
#18, which supplies channel 'B' of the engineered' safety features ]
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(ESF) actuation cabinet', instead of closing breaker #16. This , - action resulted in deenergization of one channel-of ESF, thus sat- 1 ,
isfying the actuation logic (1-out-of-4) for the isolation signa ;
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The ESF~ actuation clos'ed the four containment purge valves as i
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designed. At approximately 2:51 a.m. the control. room operators -
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', reestablished containment purge. operation, after verifying that the :l cause of.the event was understoo I
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Licensee corrective actions at the time were to discuss'the need for j
, i attention to detail with the plant squipment operator; the issue was j reviewed further at the. operations department meetin ' ;
b h . The inspector discussed the event wit' the-operator involved. The
; l ' operator. stated that the blue tag wat cated between breakers #16 ' and #18-and he: mistakenly opened bre w b #18. Upon recognition of u n ;'
therror,.he immediately reenergizea the circuit from'breake'r #18 ; and ieported to the control roo The-inspector discussed with the } operations supervisor the' appropriateness of the operator's action to l immediately reenergize a load when an error in tag restoration 1 occurre The supervisor acknowledged the inspector's comment and j l
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.i concluded'that-it was inappropriate for the operator to perform such
, an action. The supervisor discussed this item with the operator 4
. involved,-and with the operations departmen In conclusion, the purge valve isolation was caused by personnel error. No adverse safety implications resulted from the actuatio September 20 Event i
On September 20, 1990, at 11:00 a.m. an inadvertent facility II ' engineered safety feature (ESF) actuation occurred. The licensee reported the event at 11:17 a.m., as required per 10 CFR 5D.72' (b)(2)(ii). Plant conditions at the time of the actuation were cold i shutdown,. reduced inventory conditions, and one of the steam genera- ! tor primary manways remove The actuation occurred when test acti- 1 vities resulted in a pressurizer pressure signal above the setpoint ! that caused a safety injection actuation signal (SIAS), containment- ! isolation actuation signal RIAS), and an enclosure building filtra- I tion actuation signa 1 (EBFAS), The signals were processed exclusively for facility I Initial Conditiors and Cause Prior to the actuation, an instrument and controls technician was ;
' -performing a-pressurizer pressure transmitter calibration. Channel j
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' A' of- the reactor- protection system (RPS) was in bypass and'the ES ' channel' ' A' was in . inhibit as required by thr initial conditions of - d the calibration procedure. The technician went to the transmitter 4
inside containment to begin the calibration, where he identified a i conflict'with .the work activities of another technician performing a .l calibration on-the pressurizer pressure control channel. Specifi- ; cally; the relative proximity of both transmitters caused an inter- R ference-with.the calibration equipmen l
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The' technician assigned to the. channel 'A' calibration proceeded to'_
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the channel 'D' transmitter to recommence a calibration. 'The channel
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'D' transmitter was fully. exercised without the: channel 'D' reactor
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i protection system;and ESF channel bypassed and inhibited as required- 1
.by the calibration procedure. The technician realized that_the 4-channel-?D' needed to'.be' bypassed, and-thus he called the control room to.have another technician bypess and-inhibit the 'D' RPS and i ESF channels. The facility;II.ESF actuation: occurred _when the ;
technician in the control room returned channel 'A' to operat +
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The licensee's immediate actions were to_ verify that.all: equipment ; j associated with the ESF actuation operated, or was justified not' to ' operate by the existing' tag-outs. The inspector independently g l I
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: verified the licensee's actions. Equipment operation, as a result of the. actuation, was as expecte ,
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Licensee Followup Actions' The licensee estimated approximately 50 gallons were injected into the reactor coolant system, by the 'B' and 'C' charging pumps. The
'C' high pressure safety injection pump started as designed; however,
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the outlet isolation motor-operated valve was closed for mid-loop operation controls, and for low temperature over pressure protection, i The -low pressure. safety injection pump 'B' started in addition to the operating 'A' pump to increase the shutdown cooling flow rate; however, no additional injection occurred because'the common suction j was aligned to the reactor coolant system. Shutdown cooling ' operation was not adversely impacted and remained in operation during 1
, the actuation. . Reactor coolant system level increased approximately : one inch during the transient as identified by on-shift control room operators review of mid-loop level indication.
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Atlthe close of the inspection, the licensee reviews were in progress to' determine the root cause of the actuation. The technician's fail-ure to- follow the pressurizer pressure calibration procedure was identified-as part of the cause; however, this-action by itself would not have resulted in the partial actuation. Specifically,>the exer-
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I cising:of the transmitter resulted in unblocking anditripping the SIAS channel 'D' signal -resulting in a 1-out-of-4 logic. The actua- ' tion logic is;2-out-of-4. A troubleshooting plan was being developed atsthe close of the' inspection-perio ' ,
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mm The inspector will, follow future licensee; actions to resolve the ! cause of the event. NRC review will consider the licensee's asses's-ment of the event submitted in the associated licensee event: repor This itew is unresolved pending _licensae completion of deternining i # the root'cause of the event and actions to prevent recurrence J E .(5.0-336/90-18-02). ;
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September 27 Event;
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On September 27, at 10:53 a.m. the licensee reported pursuant to 10 CFR 50.72 (b)(2)(ii) an inadvertent facility 11 main steam. isolation '
~, ( MS I ) .- The plant was in mode 6 operations with reactor coolant u - temperature at 97 degrees Fahrenheit at the time, q ",' '
At~the time of the event, two inputs to'the engineered safety feature M actuation Lsystem (ESAS) were-being calibrated by instrument and controls technicians. The calibrations were being performed on
- channel 'A' steam generator pressure, and channel-'C' of' pressurizer ' *
pressure. Both channels were verified to be~ correctly placed in the
; '; > 'inhi. bit condition. The MSI signal. occurred coincident with returning the channel 'C' pressurizer pressure input to operat <
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The-root cause of event was not determined at the end of the inspec- .
. tion period. The licensee immediately suspended all ESAS surveil-lances until the cause was' understood. The inspector reviewed l licensee actions following the event, and identified that appropriate l procedures were followe >
On October 2, at 10:52 a.m. the licensee rescinded the reportability of the facility 11 MSI ESF actuation. The basis for the cancellation was review of NUREG-1022 supplement 1 question 6.9 guidelines for reporting, which concluded that an ESF actuation is not remrtable if'
.the system is not required to be operable, and that the system had been-properly removed from service such that its intendeddunction '
was'not carried out. The inspector reviewed the applicable , requirements for MSI and tagouts on MSI components to verify the f 11tcensee decisio All aspects of the guideline where fulfilled in !
.this regar . Conclusions and Assessments
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Two of the three ESF actuations were a result of personnel erro One associated with incorrect tagging and the.second due to a failure , to adhere to a calibration procedure. Report details 2.3.1, 2.5, and 7.1 describe _ events:directly related to' personnel-performance. After
~ 'the inspection report period, the'inspe-tor noted other examples.o '
inadequate control ~of tagging, resulting in a potential loss of , containment integrity during refueling operations on October 2,-1990,
* and 'an ESF actuation on October 10, 199 The events, individually,_had minimal safety significance, except .
the potential loss of-containmen?.i.ntegrity, which is still being reviewed by the inspot. tor.: Collectively, however, the events _ repre-
'sent an ' apparent lack of attention to' detail and configuration controlLof safety-related equipment. Two of the ESF actuations at'. 4 3 the end'of,the period still requiredLa root cause determinatio Inspect _ ton of the events -will consider management attention and *
E 3 ; involvement to address the documented items. Future NRC actions will . consider' licensee-actions to prevent recurrence. In the cover lettersthat trcnsmits this report, the licensee ~has been requested to-respond:to this matter,-and specifically to address these personnel-performance problems and to describe'theTactions taken singularly and collectively to prevent recurrence. This .is considered a generic
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unresolved item,-pending? review of the requested licensee response-l :during a subsequent _NRC-. inspection (50-336/90-18-03).
2.4ucontrol of Outage Activities On September 14, at approximately 8:00 p.m. the licensee commenced a plant. shutdown for the cycle'10 refueling outage. At 8:25 p.m.,'the licensee' notified the NRC operations center as required per
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i , U'_' < 10 CFR 50.72 (b)(ii)(v) of a shutdown greater than 72 hours. The scheduled outage duration is 52 days, i The inspector ob:arved plant shutdown evolutions from approximately-50% rated power until the reactor was shut down. Control of plant conditions, procedural implementation, and coordination of the shutdown.was very goo Theilicensee maintained the operations outage organization as the
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focal point to control of outage activitie The outage organization was located within the control room such that distractions for the-operating crew was minimize The organization was controlled by-two ,
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experienced senior reactor operators. A team of support operators
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supplemented plant configuration control. Inspector observation of , the operations outage organization found good overall coordination i ( and performanc . p. ; a '
-Unit planning meetings were held twice daily on weekdays, with a "
additional meetings on weekends. The meetings-offered planning j updates. discussion of previous work activities by responsible - departments, and identification of-plant problems and resolutions, e The meetings kept unit-personnel-aware of plant status, and promoted effective communications between unit and support department The inspector verified l licensee implementation of commitments to'NRC ' Generic Letter 88-17 prior to and during reduced inventory operations between September 18-19, Overall control land implementation of ." " actions'was acceptable; however, control of reactor vessel level-monitoring was deficient, as documented'in. report detail' q~ _
-Major _ outage activities during the--inspection _ period include y 1 replacement of moisture separator; tube bundles, steam generatur eddy-
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: current' testing, service water pipe' replacement,-control' room design ' , : review modifications, containment' polar crane ' modifications, . inspection:of control element assemblies and discharged fue ,
- "s assemblies, and' low pressure turbine rotor inspection .
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2.5. Reactor Vessel Level Monitoring During . Reduced Inventory Operations 3~
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On' September 19e at approximately 4:00 p.m. the . inspector identified _
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that no' reactor vessel-level; monitoring system-(RVLMS) was operable as required during the reactor. coolant system -(RCS)Edraindown evolution. At the time of-identific.ation,:the unit was in reduced = A ,
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inventory operations with' vessel level approximately:18 inches above l
.the centerline of the RCS h'ot' leg. Upon notification, the operators '
f timmediately secured RCS draindown and requested. troubleshooting of
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M *' the RVLM The licensee identified that the #3 and #8 heated junction thermocouple (HJTC) jumpers were incorrectly installe s' The jumpers were returned to the correct configuration, and RCS ,
, =draindown recommenced at 9:50 p.m. on September 1 Licensee investigation into the cause of the HJTC jumper error concluded that authorized work order (AWO) M2 89 05450 was inade-quately controlle Specificclly, workers implementing the AWO did not. document installation of the eight jumpers on station form (SF)-235.as specified in the AWO,.and the post-installation testing e per SP-2407B was deficient in identifying the incorrect jumper ,
installatio Safety Significance The HJTC' system monitors coolant. inventor / in the region above the reactor cor Redundant strings of HJTC's are arranged in the reactor vessel head area to provide indication of conditions at eight-
: distinct levels.. The system is a two-channel system each consisting of- a string 'of eight. sensors. The HJTC system is part of the "
inadequate core cooling (ICC) monitoring system installed per the 3
- provisions of NUREG-0737, Section II . l
'f One of the two channels of RVLMS was required by the licensee.to be
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-operable'during reduced inventory operations, as' documented in
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, response to' NRC Generic Letter:88-17, ' Loss of Decay Heat Removal . ' ='
~ The control room operators had available :a continuous level y ,
indication atithe' time'of_ inoperable RVLMS'.;The indication was
. 1 W ' required by procedure, was independent of:RVLMS, and used a temporary :tygon; tube showing actual level, with-continuous display'on a CCTV /within the'controlTroo ' '
e RCS draining was immedi_ately stopped upon notification of the- .j inoperabl_e RVLMS train. Other indications of. shutdown cooling . system ^ g'
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performance were operable and in use.by'the: operators. The RVLMS was
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Linoperable for approximately five hour . In conclusion, the safety significance of the particular event was ! minimal; however, the'need-to provide operators reliable and inde-
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pendent reactorLvessel level- indication during reduced inventory
, operations was necessary based on previous NRC. concerns of shutdo'wn d f ,
Jcooling performarice assidentified in generic letter 88-1 '
, Regulatory Requirements ,
Technical Specification 6.8.1 requires that procedures shall be [ established, implemented, and maintained as recommended in Appendix A_ of NRC Regulatory Guide 1.33 (February,1978).
NRC' Regulatory Guide 1.33 Appendix A 1.J. requires > , f 1:
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;; s administrative procedures to control bypass of safety functions and -
jumper control._ Licensee procedure ACP-QA-2.06C, ' Station Bypass /
- - . ' - Jumper Control for Troubleshooting. Red. Lining, and Calibration' step . 3 6.2.3.2 requires-recording the date of each lifted lead and the g"- initials of the persons performing the task on SF-235 or equivalen <
Contrary to this, -no SF-235 was completed as described in the wor control job description under AWO M2 89 05450. The resulting omis- % sion contributed in part to not having RVLMS operable, and one of the initial conditions of OP-2301E 'RCS Draindown' was not satisfied.
3 This is a violation (50-336/90-18-04).
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-In addition to the' failure to follow procedures, the inspector.noted
' insufficient work control during HJTC jumper installation and in-adequate post-installation testing. Further, operator attention to
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a. key operating parameter during RCS draindown was deficien . The event was of minimal = safety significance; however, availability of an cindependent RVLMS to' assess'RCS conditions and shutdown cooling performance was, for a short duration, deficient.- Licensee immediate corrective actions were appropriat .6 Regional Temporary Instruction 87-01,. Control Room Environment The inspection plan for temporary instruction (TI) 87-01 was based on IE Circular 81-02,-' Performance of NRC. Licensed Individuals While On
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e '~ Duty. ' The scope of d e; inspection consisted of review and .t implementation: of licensee administrative procedure ( ACP)' 6.01 X '
' Control' Room; Procedure, observation of control room activities, and s shift turnove Inspector observation's concluded the operators appeared to be alert, '
remained within their immediate areas ofiresponsibility, .and were :
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' -attentive to the instrumentation and contcols. There were no' -distracting activities, loitering of personhl or-unnecessary-l personnel.in.the1 control roo L ;The shif t turnover consisted of each operator briei%a his relie At-the. completion of the turnover, the.on shift; supervisor reviewed , -the plantistatus;and activities of the prior shift and the current *
y, > ' shift with:al1 licensed and ron-licensed operators. The turnovers
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observed were: thoroug ? The overall' control room observations.-concluded-that there is a
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professional atmosphere in the control roo . ?b
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o 2.7; Previously Identified Items- , 2.7.1 (Closed)-Unresolved Item 88-22-01: Em'ergency
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O J erating Procedure-2534 Did Not Include a Step to Eliminate Voiding in the Reactor Coolant System (RCS) Steam Generator
. Tubes as Per Combustion Engineering Procedure CEN-152 '
Guidelines This: item concerned the licensee commitment to ' incorporate CEN-152 < guidance on steam generator U-tube void formation into E0P 2534-
" Steam Generator Tube Rupture." The commitment was based on ' documented corrective actions to a previous notice of deviation : identified by.the.NRC in-inspection report 50-336/88-10 dated-August 23,.198 The. inspectoi - reviewed CEN-152, Revision 3, " Steam Generator Tube Recovery," page 17, contingency step 33.C, action for operators to'
eliminate tube voiding, in comparison to E0P 2534, Revision 8, contingencyLaction step 3.18.b. The licensee sufficiently incor-
.porated the CEN-152 guidance into E0P 2534. The inspector also Lverified the CEN-152 to the E0P step documentation review process-and identified no further inadequacies. This item is close .7.2-] Closed)-Unresolved Item 89-05-07: Core Alteration '
Technical Specification Definition Ouring In-core Instru - 3 -mentation Removal- 1
, - Previous 'NRC review, as : documented in inspection report 50-336/89-05 'detailE11.0,-concluded that in-core: instrumentation (ICI) removal- O constituted a core alteration as defined in'the Millstone 2 Technical H Specifications (TS). The associated failure to provide senior reactor operator-(SRO) coverage and . maintenance of containment -
integrity in accordance with TS 6=.2.2.e and 3.9.4 was unresolved pending ~ resolution of a:TS change and clarification planned by the licensee, c 'Infa letter dated April 10,-1990,Lthe-licensee submitted a: proposed change to the TS to change: the definition of core alteration to-specify certain activ_1 ties that are to be considered as core altera--
* ' ' tions. ' Initial'NRC review identified no technical inadequacies in-
- tthe licensee's submittal, but concluded that.the wording in'the
;11censee's proposal was not' consistent with the Combustion Engineer- >' H ing Owners Group (CE0G)l standard definition. Afteridiscussions with '
~the NRC licensing staff, the licensee-agreed _to resubmit a proposal consistent lwith the CE0G version. The'NRC' staff did not approve the
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proposal-based on internalnNRC licensing guidance that standardized amendment' specifications should not'be submitted individuall During this time interval, the NRC staff did not have a technical . disagreement on submission of the technical specification as s
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, 12 propose The 1.icensee's alternative approach was to change the'
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bases..section for ICI within the TS to identify that removal of the
' ICI was not a core alteration. The pro ess to modify the bases was . implemented under the 10 CFR 50.59 "Unr eviewed Safety Question" determination process. The inspector reviewed the licensees un-
~c reviewed safety question determination and supporting reasons and concluded no unreviewed safety question existed. This item is close .0 Radiological Controls 3.1 Posting and Controls of Radiological Areas h During plant' tours, contaminated, high airborne radiation areas were ' reviewed .with respect to boundary identification, locking require-
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ments, and appropriate control points. No significant observations l' were made, 3.2 Plant Incident Report 90-85, Post Accident Sampling System
.(PASS) Failure --On'. September 11, the licensee documented plant incident report (PIR)
90-85 on PASS' failures. The licensee's PASS system samples the reactor coolant and containment atmosphere under postulated accident . conditions -in 'accordance with requirements of NVREG-0578 and NUREG-0737; NUREG-0737'. item II.B.3 requires the licensee-to develop a PASS such 1 that accuracy, range, and sensitivity shall be adequate to-provide Lpertinent. data to operators in order to describe' radiological!and chemical status of the' reactor coolant system. On November'14, 1984~. , Lthe NRC.provided .the safety evaluation for the: Millstone 2 post
. accident sampling system. Further, on September 28, 1987,-the NRC 0- jstaff approved 1 technical. specification' amendment 120.to include PASS requirements in TS 6.18.
' The'PIR was documented by the licensee to-identify that.the PASS * E system has provided. total: gas values that are not representative of actual reactor coolant system total. gas.since 1984. The total gas measurements are made to estimate core degradation ~and corrosion potential of the coolan Typically during operation, the total gas values rang ~e between:30-40 cubic-centimeters (cc)/ki.logram (Kg). The / PASS values acquired from six month surveillance samples over the
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past f1ve' years, ranged between -139.9 cc/Kg to'+91.7 cc/Kg'. . As' ,
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providedtin NUREG-0737,: item-II.B 3, criterion 10, the desired
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, accuracy of total. gas' samples ~ is +/-10?; between' 50-2000 cc/Kg, but'
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the NRC=also.-finds accuracies up to +/-20?4 acceptabl t- ' Inspector review noted that other parameters analyzed using the PASS ~ were acceptable and 'within the required accuracie .
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The lack of an accurate method to sample total gas from the sample locations at the P/SS does not invalidate the licensee's ability to
, assess core damage giuring a postulated accident. This conclusion is based on the abilit) to obtain adequate gross radioactive liquid samples and containn ent airborne radioactivity measurements to ,
estimate postulated core degradation and corrosion potential within ! the coolan '
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1The inspector will follow licensee cor ective actions to ' develop a reprerentative total gas sample at the PASS 1ccation. This ,
-item.is unresolved pending completion of licensee actions to improve l the effectiveness of PASS total gas sampling (50-336/90-18-05). ,
4;0' Maintenance / Surveillance > 4.1 .0bservation of Maintenance Activities
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.The'inspec. tor observed and reviewed selected portions of preventive and corrective maintenance to verify-compliance _with regulations, use
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J' ofcadministrative and maintenance procedures, compliance with' codes
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and standards,. proper QA/QC involvement, use of bypass jumpers and safety tags, personnel protection, and equipment alignment and
. retest. LThe following activities were included: j -- - Authorized Work Order (AWO) M2-90-09957, Alignment Check of ' A' Main. Feedwater Pump, September 14 --:
AW0.M2-89-05450,- Remove and Replace Heated Junction a Thermocouples for Inadequate Core-Cooling Reactor Vessel Level Indication, September 13 i
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AWO M2-90-10223, Corrective' Maintenance on the Reactor ' "' Vessel Level Indication, September 19' '
, ' -- ~AWO M2-89-09732, Preventative Maintenance on the 'B'-
Reactor; Building' Component' Cooling Water Motor, y September-5;
-- 'AWO M2-90-10572,' Reactor: Fuel Shuffle, September 28 i a
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Control ofrAWO M2-89-05450 is discussed further in report detcil-2.5.-
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The remaining work: orders were acceptably complete .. 4.2 Observation of Surveillance Activities i um [ -The? inspector observed and reviewed portions.of completed surveil-
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lance tests to assess' performance in accordance with approved pro- t .
, ' 'cedures and-Limiting _ Conditions of Operatlon, removal and restoration ! , of equipment, and deficiency review:and resolution. The following- ' tests were reviewed: , -- ,SP-2404AN, Spent Fuel Pool Radiation Monitor Calibration, p' September 10 - --
SP-90-2-3,' Auxiliary Steam In-Service Test, September 13 i
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:T90-14, Auxiliary Feedwater Pump Performance Test, September 15 No'significant observations were mad . Spent Fuel Pool Poison Box Blackness Testing During the time interval between August 20 - August 30, the licensee !
-conducted neutron logging (blackness-test'ing) of_ selected spent fuel- !
pool poison boxes. The testing was implemented by a contractor (NURSVRTEC).- The licensee opted to complete blackness testing as a result of , visual examination results of spent-fuel pool coupon ~ Specifically, on July 27 the licensee-identified missing boraflex-material from a 3/8-inch vent hole during the routine coupon-surveillance program. Details of.the examination results are r a documented in inspection report 50-336/90-14, detail (
- ' Neutron logging is intended to measure thermal neutron attenuation in I the: walls'of high density spent fuel storage racks. Each storage i cavity contains a poison insert which consists.of four boraflex i panels contained.between two stainless steel sheets. The Millstone 2' I spent' fuel pool ~ storage racks were' designed and manufactured by . l Combustion Engineering and the boraflex material was manufactured by ;
Brand Industrial: Services, Inc. The storage racks were installed in
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the Millstone 2 spent fueij pool in' Mayl198 ! The: test equipment consisted of-a' fast neutron source surrounded by 1 four radially located neutron detectors; the assembly.was suspended , i- Lin a vacant storage location. The source'and detectors. transverse
, the poison box and measure neutron counts. Variations in detector . output are evaluated to " equivalent" gap sizes based-on baseline:
odata detector speed, detector time response'to equilibHum values,
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andjgap orientatio ' The. licensee completed-measurements on .105 out of a total 384 ~! locations with' the high. density fuel' storage poison boxes,in region I l'
.of'the spent fuelJpool. The scope of the examination was ~ limited .
based on.high background gamma radiation from expended fuel' stored in j hdjacent' cells. in region. I and boundary locations .in region -II . The' " higher gamma : radiation fields. result.in lower neutron counting rates,- p'oorer statistics, and smaller differences in count rates between poison;and unpoisoned section C
' , Gaps.in the boraflex were-identified in 33 location The gaps'were estimated by the vendor to be between one inch and l'.8 inches and 1 4ere located at various elevations within the active fuel regio .'
# > > 31 (q.;. - [ , s ., .
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The inspector l observed health physics controls and neutron source ^
;t' ~" . survey results, loose parts control on the spent fuel pool bridge, and surveillance testing of four poison boxes. -The inspector reviewed the vendor's procedures and associated safety analysi Questions on the safety evaluation were presented to cognizant ' licensee engineering personnel and were appropriately dispositione Conclusion '
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, The control and implementation of neutron logging in the high density spent fuel pool poison boxes was completed efficiently. Inspection report detail 5.2' documents the blackness testing results,.the.signi-ficance of the findings and the licensee's followup action .3. 'A' Reactor Coolant' pump Motor Oil Leakage -
t On August 15 and! August 27 the licensee added oil to the upper reservoir off the 'A' reactor coolant _ pump (RCP) motor. The oil was
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added as a~ result of a downward trend in oil ' level indication since early Augus ,,
.l - A total of -approximately six (6) gallons was added on the above * dates. :Thet reservoir has a capacity of approximately -120 gallon f . '
The licensee identified the cause of the oil loss to be leakage from-
;a onejinch fitting at the rear:of the lift pump-package. Approxi- :mately 50% of the oi1 lost was collected'in the oil. collection tank,- 3 , .the remaining oil accumulated at the rear of the lift pump. package, onL the pump; insulation ~, and on the. motor fram ' '
lDuring the current refueling. outage, the -licer see replaced th " 1 suspect; fitting. :The:RCP-~ oil lift pump package will b'e inspected >
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prior to.-start-up.during' oil; system operation at normal temperature
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with the moter operatin 'I
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110 CFR 50 Appendix R Section III, Subpart 0 requires 1.an' oil ' collection-system to be installed such.that-failure of'a lube oil
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eL L system'will notL lead to a fire during normal ~and' design. basis ' event Further,:the collection system shall be capable-of'collectin.g lube 01.1 from all potential pressurized'and unpressurized leakage sites in
.the RCP oil-sys;em. Leakage points to.be protected shall. include the- 3 ' lift ! pump and piping, overflow lines and- flanged connections on the-s ,
oil reservoir ,
.In1spiteioftheoilleakage[inAugust,theupperreservoir.didnotgo' 'i
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'below-a1 point in level requiring operator action to. protect the RC > moto No other indication of inadequate performance of the RCP-motor'was'noted during this time interva .
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In summary, oil leakage out of the ' A' RCP motor upper reservoir was-verified to collect both around the pump, and in the oil collection ' tank, . The' inspector verified oil around the pump during the early i part of the cycle-10 refuel outage. The leakage in August of 1990 differed from past instances when the A pump lost oil, in that most of the oil lost was collected in the collection sumps during the-
. prior events. .The leakage point was identified, and licensee corrective actions are ongoing to correct the current probic '
The inspector questioned the acceptability of the oil collection facility relative to.the performance objectives prescribed in 10 CFR
:50 Appendix R.Section.III, subsection 0. The inspector's-concern, based-on the August 1990 leakage; is whether the oil collection ,
facility .is suf ficient to prevent a . fire hazard during RCP op'eratio ~This item is unresolved pending licensee actions to improve the oil- ! collection-for the RCF motor such'that leakage from areas around the- ' oil lift pump package is adequately contained-(50-336/90-18-06).
, 4.4 .NRC=Information Notice 89-90, " Pressurizer Safety Lift Setpoint Shift" -
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LTh'e ' inspection consisted of follow-up> of licensee actions related to )
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NRCfInformation Notice 89-90. This information notice advised licensees of potential problems resulting from operating pressurizer
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safet'y.-valves (PSV). in an en' ironment different from that L ed to
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establish the~pSV 1.ift setpoints. In October 1989 Westinghouse infor-
;med -its'plantiowners;of-aLpotential deviation of the PSV set pressure'
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from the American' Society of Mechanical Engineers (ASME)'section III- !
.and plantitechnical specification requirements for plants having loop -
seals upstream of.the PSV Millstone 2 was originally constructed withtloop seals between'the l
pressurizer:andtsafety' valve In August 1983,1the licensee removed ,l the loop seals based on- previous industry r asults of the Electric
- Power- Research Institute- (EPRI)' study ,inL198 u 3-The licensee. removes the' PSV and? sends :it'to Wyle Laboratories for testing and overhaul during each refueling outage. . The Wyle Laboratories procedure for' testing the PSV was' reviewed together with the test .results' from- the ~1ast- surveillances. done in February 1989, 1 .The valve must undergo three consecut'ive acceptable lift setpoint 1 b . tests 'at' normal. operating conditions .for acceptance. The valve is -
(; lo subjected to leak tests prior.and after~the lift setpoint test f
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, In conclusion, based on the fact that the loop seals have been removed from the PSV piping and testing is completed at conditions resembling actual use, anomalies as described in Information Notice 89-90 are not present for Millstone 2. No inadequacies were noted in the inspectio .5 Steam Generator Level Discrepancy >
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On August 11,-1990, the -licensee entered technical specification (TS) requirement 3.3.1'.1 for channel 'D' steam generator level on the N I steam' generator. Entrance into the TS action statement, which was .' L based on the channel 'D' level being slightly outside the acceptance criterion during the performance of daily channel checks as detailed in procedure OP 2619-1, required that the channel be placed in bypas "The acceptance criterion requires .that all channels agree within 4% !
. level indicatio l0n August'20,:the: licensee completed the repair of the channel and '
channel 'D' was restored-to operability after'successfully completing the daily channel. check..Licens'eeLtroubleshooting of:the leve _ me - transmitter identified a malfunctioning voltage to current converte According.to manufacturer information this converter contributes-0.25% of the instrument-loop accuracy. Troubleshooting-also
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E , = identified a shiftL(less than 4%)11n transmitter output. Items , reviewed during the inspection included -licensee troubleshooting -! m activities, problem-identification, level- channel accuracies dynamic
.leve'l: response.during transients; and operability requirements ofL the channe ' < Based on' a review of transmitter cross comparisons for dynamic ,' '
response durings the August 27 reactor; trip, successful completion'o a ithe3 channel check,: and corrective maintenance activities, .the in- 1
- spector concluded.that the-level transmitter was now operable. At1 ' >
the,end of the' inspection period, the licensee was conducting the:
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required calibrations to'further assess the transmitter and to determine what additional followup actions were necessar r The licensee's actions'to address the steam generator transmitter
. level problem.were acceptabl *
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4.6 Previously Identified items '
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4 . 6 . 11 (Closed) Unresolved Item 88-24-05: Inadequate
'.i Application of Acceptance Criteria for Heise Gauge
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Calibration
'This: item was opened based on an inspector noted= discrepancy in the licensee > application of the acceptance criteria for procedure I/C
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,, 1104A "I&C Pressure Test Gauges Calibration." Paragraph 2 of I/C
:1104A defines the acceptance criterion for Heise gauges as +/-0.1% of
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e 18 fu'll scale, or. +/- one minor scale division, whichever is greate Heise gauge No'. QA-370 is a 0-5000 psig gauge with 5-psi minor scale ! divisions. The inspector noted that the acceptance criterion for calibration'of this gauge in I&C procedure 1104A-41 was +/- 10 psi, which is inconsistent with the required acceptanca criterion of
'5 ps The licensee revised data sheet I/C 1104A-41 for the 0-5000 psig Heise pressure gauge with the . required acceptance criterion of +/- 5 psi. -The inspector verified the licensee procedure revision and-considers this item close .0 Engineering / Technical Support 5'1 Plant Design Modifications g The inspector reviewed and verified. licensee implementation of ]
electrical separation criteria as part of Plant Design. Change Record 1 (PDCR) 2-17-90, ' Correction of Human Engineering Discrepancies l' r Control Room Panel-C06F. Replacement.' Specifically, the inspector
. reviewed the class IE/non-class IE t.eparation criteria as referenced R in Northeast Utilities Service: Company (NUSCo) drawing 25203-3300 The temporary: pane'l was. installed adjacent to control room' panel C07,
:and was' utilized to facilitate continuous control.of safety and
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? non-safett components, during human engineering modifications on ,!
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. panel C07.: : Electrical cable separation criteria were met and
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. implementation of the design change was acceptabl .j 5.2; Spent Fuel'Po'ol Boraflex Condition
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,1 h :On1A'ugust :24,it he, licensee: documented plant . incident report (PI_R)' d 90-78 and: reported to the NRC per 10'CFRd50.'72(b)(1)(ii) a condition 1 that'is potentially outside the design basis of the plant. The t notification was _ based on. preliminary. blackness _ testing results of '
tregi.on?I spent-fuel pool poison boxes. Previous NRC review of this ] issue was pro'vided in inspection' report 50-336/90-14. section 5.2, ,
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rwhich documented =the results of the licensee-visual examination of the spent fuel-pool boraflex coupons. Also.see section 4.2.1 of ?
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Athis' report, j
,- . Blackness Testing Results- j
- OnLSeptember 6, the licensee's contractnr-(NUSURTEC) documented
- preliminary lresults of: the blackness' test on the-spent ~ fuel pool; ^
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storage; racks,1 The ' licensee tested.approximately 27%-(105:out of . 1" 384)'of;all poison-boxes _in the region I area of the spent fuel poo ' The largest:-(axially-oriented)_ gap was estimated at 1'.8 inch +/- 0.~2 inch w'ith the average gap size at 0.6 inch. Thirty-one cells contained one or more gaps. The gaps identified were randomly '{
' distributed axially, and located in areas previously holding spent : ~ fue l f
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Licensee Actions Prior to completing the blackness testing the licensee imposed a two-by-four loading pattern for new fuel stored in the racks and a
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soluble-boron concentration in the pool of greater than 1720 parts per million (ppm). The inspector verified the controls and identified no inadequacies. Further, the licensee verified.that l
, parametric studies developed by. Combustion Engineering (CE) confirmed ,
the ability to requalify the spent fuel. racks for a two-by-four & loading pattern without the presence of boraflex material,~ and maintaining the assumptions in the initial criticality evaluation t including no-soluble poison in the SFP wate ~ n A' subsequent CE criticality evaluation concluded that,-even with,a- (
, ' 2.7"; loss.of, boraflex 'throughout the region I spent fuel pool at 'any given axial location, the reacks maintained the' licensing basis , ' ' V values-for effective multiplication factor (Keff) of 0.95, and- the criticali.ty assumptions were preserve s *
n .. Based'on the above, the safety significance of the gaps-identified in-the highidensity- storage racks was minimal, However, the development -;
.of the gaps with continued gamma exposure:in the racks is a matter . -that warrants continued licensee monitoring and evaluatio ' . Regulatory ~ Bac kground H The inspector reviewed previous licensee- technical specification (TS)-
amendment; requests and associated NRC-safety evaluations concerning installation of storage racks with'boraflex material in' the spent fuel poo L0n . July 24;L1985, the licensee proposed a change = to the1TS to modify
.the spent fuelipool. -The.- modifications increased the fuel. storage capacity with a;high density fuel rack; design. The licensee's safety
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analysis; section 4.7,< '! Poison Material In-Service Surveillance ' Program",l documented that direct dosimetry will be utilized to' ' establish ~ an accurate record of ' cumulative gamma radiation exposure ' a to the'boraflex materialain.the spent-fuel pool racks to' gather data to conoare to. the manufacturer's qualification life--of the boraflex materia The qtalification life of boraflex.is based on accumulated gamma expostre that',results, in materialf shrinkage. !The silicone rubber t
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-binder in the'boraflexfis: subject to change from gamma exposure, that; 'E results?in crosslinking of'the molecular bonds, which increases the hardness and decreases the flexibility of=the material. The design '
of.the high' density fuel storage- racks at Millstone 2 was. to allow for relatively' free expansion and contraction of boraflex material by installation' of panel vent: holes, and a manufacturer's specified clearance between the stainless steel sheaths and the' boraflex '
. mater.i al .
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! -On January.15, 1986, the NRC issued TS amendment 109 which authorized ;
the-storage capacity of the spent fuel pool from 667 to 1112 fuel e assemblies. The staff's safety evaluation documented the licensee's proposed inservice surveillance program to meet the requirements of, 10 CFR.50, Appendix A, Criterion 6 a
~ On May 21,1986,' the licensee proposed 3 change to TS for the storage .of consolidated spent fuel. The licensee's safety analysis section i 4.7, " Poison Material-In-Service Surveillance," again referenced , - direct dosimetry: measurements to establish an accurate record of !
cumulative gamma radiation exposure of the boraflex material in the spent fuel pool 1 racks,
On June 2,.1987, the NRC issued TS amendment ' 117 regarding. consolidated spent fuel. While tne staff's-safety l evaluation approv'ed the amendment request, no explicit: mention was
.given to-the boraflex coupon surveillance program,
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- On May 6,1987, the licensee proceduralized the boraflex coupon surveillance program. . Procedure SP 21026 required periodic-veri-fication of the-presence of absorber material in the spent fuel' pool-
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;j racks. The procedure did not implement a direct dosimetry measure-ment of_~ gamma radiation exposure of the boraflex material. ' Inspector 1 review concluded thatJthis failure to-implement a-licensee commitment is a deviation,- i.e., a licensee failure to: satisfy a commitment
:where the commitment. involved has not been made a legally binding
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j requirement (50-336/90-18-07). :
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Inspector re'viewz identified no other inadequacies regarding ! implementation of commitmentscfor the boraflex coupon surveillance
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program.: Northeast Nuclear Energy Company ~ (NNEco) . engineering-L personnel recognized'on May:22',-1990 that. cumulative gamma exposure Ewas1 required as part of the coupon survell. lance program, Land upon=
identification commenced calculations to determine cumulativeL , _ exposure to the boraflex materialfin!the Tpent fuel pool, I
Conclusion-
, , = Licensee failure tofimplement commitments for the boraflex coupon ' -surveillance program is a deviation; The issue was identified by-the- , '
licensee, however the lack of gamma exposure data prevented full l
* . , assessment of the manufacturer's qualification life for the'boraflex ,
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i*- g 6 -l 5.3 Previously Identified Items I
. - .
5. : Pressurizer Cooldown Rate D(_ Closed) Unresolved Item iscrepancy ]
;
-1 This item concerned licensee actions to reconcile the technical l
- specification (TS) cooldown rate limiting condition for operation j
"
with the updated stress and fatigue analysis. The present TS ' requirements were.not consistent with the reactor coolant system i
'
pressurizer fatigue-stress analysis. Specifically, TS 3.4.0. j limits pressurizer cooldown to a maximum of 200 degrees per hou i The nuclear steam system supplier developed the fatigue analysis in 1972 and-performed a stress analysis fM a cooldown rate limited to { 100 degrees F/hr, between 635-355 degrees F and 200 degnes F/hr between 355 and ambient ~. ' On-May 8, 1990,'the. licensee evaluated a-200 degree F pressurizer
'
cooldown rate consistent with TS 3.4.9'2.b as documented in calculation MP2-L0E-1256GP. ~The objective of the alculation was to <
, determine the structural and fatigue impact of cooling the pressurizer at 200 degrees F per hour for theientire cooldown cycl 'The calculation concluded.the TS cooldown rate was acceptable based '
on the= maximum fatigue usage factor and. stress intensity being-less '
. than the allowables in the American Society of Mechanical Engineer - (ASME) Code section III, 1968 editio Inspectorireview of the NUSCo calculation.and appl'icable references-identified no inadequacies in the calculation or.the results; thus, the requirements 'for pressurizer cooldown rates were preserved. This item i~s. closed,- ,
5.3.2 '(Closed); Unresolved Item 89-17-01: -Service Water Check Valve '!
-In-Servic'el esting Program '
This' item;was opened based'on the licensee failure to implement actionsL relating.to NRC Bulletin 83-03, 3 Check Valve Failures in Raw-
'", Water-Cooling Systems'of Diesel-Generators.' The licensee commitment iwas toldisassemble:and inspect the service > water discharge check valve to each diesel generator and include the~. inspection.i.n the ten year in-service test.(IST)~ progra U 'The~ licensee identified the cause of the failure to implement the NRC Bu11etin 83-03 as._ personnel-error. This issue was discussed further , -in' inspection-report.50-336/89-17 detail 5~ . -TheLsecond ten-year in-service test program for pumps and valves dated = April 4, 1990, identified that the diesel cooling discharge check valves will be disassembled, examined and stroke tested every 40 months. The inspector verified implementation within the IST progra This bulletin is' close ?- , > . , 'i ,, - [
y 22
>
*
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.
5.3.3 (Closed) Unresolved Item 89-17-05: Design Documentation Documentation for Emergency Diesel Generator Saturable Transformers , This item concerned follow-up of the under-excitation events on the !
A' emergency diesel generator (EDG) as documented in inspection report 50-336/89-17 detail 8.0,
,
Specifically, the item concerned a ! I N k of design documentation relating to the specifications for the z,3 , replacenent saturable transformers in the EDG exciter,
,
c .The inspector reviewed and discussed with cognizant engineering
'
personnel; pmhase orGr (PO) 874222 for the three saturdle
.
transformers, the vendor's (Bassler Electric) certificate of *
' compliance, source inspection results from the-licensee quality > ' assurance (QA)-organization, licensee seismic qualification results, and the results of.the post-installation in-service tes ; ~ Based on the'above in'spection, the inspector concluded that the licensee has n_ow developed adequate design control for the saturable transformers, and.thus the. item is close !
l6.,0 Securit Selected aspects'of site security were verified to be proper during -
* inspection tours, including 1 site access controls, personnel searches, l personnel monitoring,-. placement of physical barriers, compensatory-measures, guard force-staffing .and response to alarms and degraded conditions, No significant observations were made, ;
A 7.0 Safety Assessment / Quality Verificatio ' 7.1- Licensee Event' Reports Licensee-Event Report (LER) 90-07: Missed Surveillance On July- 10, 1990, the licensee submitted LER 90-07 pursuant to 10 CFR~ s
- 50-73(a)( 2)( 1)(B) . The event involved'failureJto perform a, required technical specification surveillance-for the Enclosure-Building
~
. Filtration System (EBFS) on June 11, 1990. This failure constituted-
.a violation of technical specification surveillance 4.6.5.1(d). The pl' ant was in operational mode 4 at a reactor coolant system . '
temperature of 238 degrcas T':e licensee concluded 5the cause of the event was personnel error in-E that the: plant technical specifications were not . correctly applie Specifically,. authorized work order (AWO) M2-90-05081 removed and reinstalled-the enclosure building equipment hatch. The AWO specified surveillance procedure SP-2609E should be used for retest a requirements. The surveillance procedure tests the operability of the EBFS train by verification that a negative pressure of 0.25-inch water is achieved within one minute in the enclosure buildin . f
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- The outage control room operator deleted the retest requirement after a review of technical specification surveillance requirements 4.6.5.1(b), which discusses performing a surveillance similar to SP 2609E 'every 18-months-or after a specific structural maintenance to ;
EBFS. :The maintenance performed-under AWO M2-90-05081 in the control
, room operator's view did not result in structural maintenance to- the system. -The last successful completion of SP-2609E was on November ' 20, 198 The licensee review concluded removal and reinstallation of the enclosure building equipment hatch invalidated the previous test, and thus the retest required by the AWO was correc Upon 1dentification, the licensee completed SP-2609E at 9:18 a.m. on June 11, 199 Long-term corrective actions included operations -
supervisor discussion of the event at the department meeting, and -
: discussion with theLindividual involve .' -Documents reviewed by the inspector in followup o1 LER 90-07 included: - - -
ACP-QA-202B, Retest
---
Human Performance. Evaluation System (HPFS) report M96 18
--
SP-2609E, EBFS Testing-
-' .AWO M2-90-05081 + --
Plant Incident' Report 90-81
- .The inspector-verified licensee corrective actions, and actions to , prevent recurrence, and determined the actions were, sufficien Conclusion Failure to adhere 'to. technical specification ' limiting condition for-operation 3.6.5.'1. avis'a violation.-lIn accordance with 10 CFR 2- * ' - Appendix' C subpart G.1, a notice of violation is not being issued,
since the violation 1was: Jidentified by the licensee (PIR 2-90-41; LER -
-90-07),. considered a less'significant violation of the'TS limiting - condition:for operation', properly reported (LER 90-07), corrected by adequate and timely actions, and considered not-willful, or re-current. This non-cited violation is closed (50-336/90-18-08).
LER 90-09: . Inadvertent Actuations of Train B Enclosure Building Fi1tration: System
LER 90-09 documents three separate instances ~ involving partialuinadvertent actuations of the enclosure building filtration system:(EBFAS). The licensee reported the events
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a pursuantto10CFR50.73(a)(2)(iv),aconditionthatresultedin i automatic actuation of any engineered safety feature (ESF).
The licensee has yet to identify the= root cause of the even As
'
documented in LER 90-09 the licensee determined that the EDFAS
~
actuation module ( AM 608) momentarily. tripped; however, the signal l
' ' '
did not latch i The lack of latch was verified in that no reset of !
-the AM 608 occurred based on operator actio !
(
Two of the three events occurred coincident with replacement of ! sequence 2 light bulbs on the B emergency diesel generator sequence l
'
The remaining event occurred during a planned autostart of the B-boric acid pum ;
>
j Previous NRC~ review of the first EBFAS actuation was documented in i routine inspection report 50-336/90-11 detail 7. s Future NRC review will assess the adequacy and timeliness of licensee " actions to determine the cause of the actuations, and effects on-operability of train B of EBFA This item will be tracked as an unresolved item (50-336/90-18-09). -l-LER 90-10: High Energy Line Break Door Discrepancy
-
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,!
' Previous inspector review of the corrective actions and-generic implications concerning this event are-documented in Region I i inspection report ~50-336/90-11, detail 7.4.1. During this inspection j
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the inspector- reviewed the LER and tietermined that the licensee; had ' met the 10 CFR 50.73. reporting regt- 9ment .2 Committee Activities
, >The inspector attended meetings 2-90-.107,2-90-111, and 2-90-115 o l i
the Plant Operations Review Committee (PORC) on September 10,- r
>
September 13, and September 15. The inspector noted by observation 1 j
. :that committee' administrative requirements were met. for' the meetings, i (~ ,. andlthat the. committee discharged its function's.in accordance with- j ragulatory requirements. . The inspector-observed a thorough-discus- 'sion of. matters before.the PORC and a good regard or safety in th j W >
issues.under consideration by the committe L 7. 3 Periodic Reports- l l Upon: receipt, periodic reports submitted pursuant to technical ; g;y specifications were reviewed. This review verified that the reported M information was valid and included =the required NRC data. The
' " inspector also ascertained whether any reported information should be
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