IR 05000206/1990007
| ML20042F485 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/23/1990 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20042F482 | List: |
| References | |
| 50-206-90-07, 50-206-90-7, 50-361-90-07, 50-361-90-7, 50-362-90-07, 50-362-90-7, NUDOCS 9005080382 | |
| Download: ML20042F485 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION Y
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Rep' ort Nos.
50-206/90-07,50-361/90-07,50-362/90-07 Docket Nos.-
50-206, 50-361, 50-362 License Nos.
DPR,-13, NPF.-10; NPF-15-Licensee:
Southern California Edison Company Irvine Operations Center 23 Parker Street,;
Irvine, California:.92718 Facility Na'me: "
. San,Onofre Units 1, 2 and 3 Inspectionak:5
' San Onofre.jSari Clemehty;' California-y
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Inspect'ioncdnducted:, February 11 through March 24,.1990
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C. W.JCaldsell, S'nior;. Resident Inspector Inspectors:-
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e AN. Hon,LResidentInspector-
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C;=D. Townsend, Resident Inspector y
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Accompanying.'
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Inspector:
D. F/ Kirsch, Chief, Reactor.' Safety Branch N
Approved By:
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Rea t r
ct ction 3 l
Inspection Summary i
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Inspection on February 11 through March 24, 1990 (Report-Nos.- 50-206/90-07 -
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50-361/90-07, and 50-362/90-07)
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Areas' Inspected: Routine' resident inspection of Units 1, 2 and 3.0perations Program including the following areas:
operational fsafety. verification.
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l-radiological protection, security, evaluation of plant trips and events,
monthly. surveillance activities, monthly maintenance activities,' refueling activities, independent inspection, licensee event report review,: and-followup"of-previously identified items.- Inspection procedures 30703, 40500, 61726, 62703,'71707, 71710, 82301,'90712,.92700, 92701, 93702_were utilized.
~ Safety Issues Management bstem (SIMS) Items:~ None<
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Results:-
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General Conclusions and Specific' Findings:
A review of the February 23,1990 Unit 3 reactor trip and subsequent.-
. pressurizer safety valve opening Indicated that the-licensee took conservative actions to addresscanomalies observed during the event.
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Specifically, the licensee tested and. reset the pressurizer safety valve setpoints prior to' restarting the. unit.
In addition, the licensee-also shut down Unit 2 to' reset its safety valycs based on.setpoint changes-observed in Unit 3.
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Significant-Safety Matters:1
~None..
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Sunnary of Violations:
No ne'.'
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Open-Items Sunnary:
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During this report period. one new followup item was ' opened a' d two.
n were closed.
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DETAILS-
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1; Persons Contacted
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S6uthern Cal'1fornia Edison Company
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H; Ray,VicePfesident,NuclearEngineering, Safety,andLicensing
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- R. Bridenbecker, Vice President and Site Manager q
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- H. Morgan. Station Manager i
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'B._ Katz, Nuclear Oversight Manager, NES&L i
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- K. Slagle, Deputy Station Manager
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Ra Krieger, Operations Manager
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' *L.~ Cash,' Maintenance Manager
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J. Reilly, Technical-Manager.
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M; Merlo,' Nuclear Design Engineering Manager, NES&L s
- P. Kn_ app, Health Physics Manager D. Peacor, Emergency Preparedness Manager _.
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D. Herbst.. Quality Assurance Manager, NES&L C. Chiu, Quality Engineering Manager
- J. Schramm,-Operations Superintendent,' Unit 1 V. Fisher, Operations Superintendent, Units 2/3
- R. Rosenblum, Manager, Nuchar Regulatory Affairs
- L. Brevig, Supervisor, Onsite Nuclear Licensing
- T..Calloway, Substance Abuse Program Manager
- R. Plappert, Supervisor, Technical Support & Compliance San Diego' Gas and Electric Company-
- R. Erickson, Site Representative City of Anaheim l
- G. Edwards, Site Representative
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City of Riverside
- C, Harris, Site Representative
- Denotes trose attending the exit meeting on March 22, 1990.
The_ inspectors also contacted other licensee employees during the course of the inspection, including operations shift. superintendents, control room supervisors, control room operators, QA_ and QC engineers, compli-ance engineers, maintenance craftsmen, and health physics engineers and
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2.
Plant Status Unit 1
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During the inspection period, Unit 1 operated at full power with no significant operating problems.
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Unit 2
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The Unit operated at power until March 9,[1990 when it was shut down to
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test and reset pressurizer-safety valve (PSV) lift setpoints. This'was done in response to the anomaly observed in Unit 3 (discussed below).
The Unit was returned to service on March 111990 and operated at power
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through the remainder of the period.
Unit 3 The Unit continued its record run (229 continuous days.at power) until February 23, 1990 when it' tripped due to an inadvertent main steam isolation system (MSIS) actuation during the performance of sub-group relay testing. The inadvertent actuation was caused by a defective test switch in one trip path that did not properly reset before the operator proceeded'to test the second trip path.
Following the reactor trip, a pressurizer safety valve (PSV)-lifted at a pressure lower than the Technical Specification limit.
The licensee tested and reset both PSVs
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in Mode 3 and returned the Unit to service on March 4, 1990.
Unit 3 subsequently operated at power until the end of the inspection period.
3.
Operational Safety Verification (71707)
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The inspectors performed several plant tours and verified the operabi -
lity of selected emergency systems, reviewed the tag out log and
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verified proper return to service of affected components. Particular
attention was given to housekeeping, examination for potential fire
hazards, fluid leaks, excessive vibration, and verification that j
maintenance requests had been initiated for equipment in need of i
l maintenance. The inspectors also observed selected activities by
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licensee radiological protection and security personnel to confirm l
L proper implementation of and conformance with facility policies and
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l-procedures in these areas. No discrepancies were noted.
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No violations or deviations were identified.
4.
Evaluation of Plant Trips and Events (93702)
Inadvertent ESF Actuation and Reactor Trip (Unit 3)
ped on a'high reactor 23,1990, at _10:57 p.m., the Unit trip (MSIS) which caused the On February
pressure caused by a Main Steam Isolation Signal Main Steam Isolation Valves (MSIVs) to close. The sequence of events was as follows:
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Event 10:45 p.m..
An operator completed subgroup relay testing for l
the Train A MSIS-in accordance with procedure 5023-3-3.43.33, "ESF Subgroup Relay Semi-Annual Test."
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~10:50 p.m.
The; operator commenced Train B MSIS subgroup (Approx.)
testing. A.ta11 board meeting was completed for
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testing of Train B in accordance kj p Procedure S023-3-3.43.31.
10:57:35'p.m. Train B MSIS actuation occurred.
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10:57:40 p.m.
MSIVs' HV-8204 and HV-8205 closed due to the MSIS.
10:57:42 p.m.
Core Protection Calculator'(CPC) auxiliary trip occurred lon high pressurizer pressure (due to loss of heat sink resulting from MSIV closure).
10:57:47 p.m.
Pressurizer safety valve, PSV-0201, lifted
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(Est.)
for approximately 6' seconds, i
10:57:50 p.m.
The Assistan't Control.0perator- (AC0) applied. control
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( Approx. )_
signals from the control room to open atmospheric dump; valves (ADVs)HVv8419forsteamgenerator(S/G)
E-088 and HV-8421,for S/G E-089 (by setting each ADV
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demand signal to approximately 30%). The ACO
observed that both valves appeared to be in the mid-
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position as expected (as-shown by illumination'of both the open and closed position indicators)..
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10:58 p.m.
The expected pressure decrease was noted for (Approx.)
S/G E-089, so the ACO reset HV-8421 demand to approximately 25%.
Pressure decrease response for S/G E-088 was less than expected, so the AC0 reset HV-8419 demand to approximately 65%'.
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Unit 3 was stabilized in Mode 3 with 4 reactor 11:15 p.m.
coolant pumps (RCPs) operating. A Plant Equipment (Approx.)
Operator (PEO) was dispatched to the MSIV area.
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reported that HV-8421 was open and passing steam.
By comparison, HV-8419 was passing less steam,'and two of the main steam safety valves associated with
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S/G E-088 remained open.
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12:30 a.m.
Control signals were applied (from the control room)
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to close both ADVs after S/G blowd,own was increased and warming of the steam lines was initiated.
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2:15 a.m.
The ACO again applied a controller signal to open (Approx.)
HV-8419 to equalize pressure across MSIV-HV-8205 in
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preparation for reopening. The PE0 (dispatched to
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the area) reported that HV-8419 would not open-past approximately 5%. HV-8419 was then closed from the control room, at which time the PE0 opened its bonnet drain valve.
Subsequently, HV-8419 could be opened remotely and controlled from the control room.
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The' inspector reviewed this transient and noted that 'it was bounded by.
the loss of Condenser Vacuum Transient described in the FSAR. The
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following discussion describes the cause of the event and addresses
plant equipment which did not function normally.
MSIS Subgroup Relay Testing Subgroup relay testing of the MSIS circuitry is performed semiannually.
i with Trains A and B tested by procedures S023-3-3.43.33 and.31, l
respectively. Under normal conditions, a trip path is tested by depressing a push button switch which deenergizes the subgroup relay.
This in turn causes.an associated contact to open, actuating that trip path. The trip path is normally reset by releasing the pushbutton.. A red indicating light, in parallel with and energized by the voltage drop across four series diodes on the auxiliary cabinet, is extinguished when the subgroup relay is deenergized and current through the parallel
. circuit ceases. The light is illuminated when the subgroup relay is reenergized (reset to normal condition) and current flows again through the four series diodes. A lower than normal current, due to a high resistance contact, would cause the red light to be illuminated with
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less brilliance than normal.
The licensee investigated the reason that a trip condition was present i
in one path while another path of Train B was being tested. As.a result l
of this effort, the licensee concluded that the most-probable cause was i
an observed, intermittent, high resistance contact in the pushbutton l
switch.
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The licensee also noted that the test procedure did not provide for verification of indicating light status for one trip path before the test trip is initiated in the other trip path. This procedural over-l sight was corrected, and should aid in precluding similar. situations.
Failure of ADV HV-8419 to Open As noted in the sequence of events, atmospheric dump valve (ADV) HV-8419 did not appear to open in' response to a 65% demand signal, but was
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which had accumulated in the ADV bonnet.
The licensee's practice had been to perform the bonnet condensate removal (blowdown) operation weekly in accordance with the requirements of Attachment 6 to procedure 5023-3-2.18.1, Revision 7, " Atmospheric Dump Valve Operation," for Unit 3, end Attachment 5 to thet procedure
for Unit 2.,The inspector noted that the procedure did mention in step
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4.1.1 (Paragraph 4.0 on Precautions) the following:
" Valves should be
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drained shiftly= to remove condensate." However, this precaution was not,
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. implemented by an operator round sheet or procedure.. Thus, the bonnet
. condensate was drained weekly as described above. The inspector also
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noted that condensate'had been last drained from the bonnet of HV-8419 on February 19, 1990 (prior to the draining following the February 23,
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1990 trip). Therefore, in this case the above precaution had not been j
implenented for five days.
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The licensee also contacted the valve vendor to evaluate the effect of the condensate which collects in the valve bonnet.
In a letter dated
March 21, 1990, the vendor, Control Components Inc., informed the licen-see that "even though the bonnet may have filled with condensate as a
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result of, the weekly bonnet blowdown interval, the only effect on valve operation will be approximately_a 60 second delay in valve opening past the pilot." Furthermore, in the interim disposition of a 1989 nonconfor-
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mance report (NCR) involving slow operation of an ADV, the licensee anticipated occasional failures of the ADV and established manual actions in the above procedure should a failure ~ occur. During this event, the c
operator followed the procedure and successfully opened the failed ADV manually.
In order; to-improve the performance of the ADV, the licensee revised the procedure-to drain-the ADV bonnet shiftly. Appendix R to 10 CFR 50 requiressthat;the ADV be capable of operating within 30 minutes for fire protection considerations.
Capability to operate the valve
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within this time requirement was' demonstrated during this event.
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As further abtion,5the 1_icensee intends to improve the reliability of
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the Unit 3 ADVs by: installing a design change during the refueling outage scheduled to begin in April, 1990.
This modification was implemented on Unit > 2 during the last refueling outage. The licensee also intends to install a' " percent open" feedback indication in the control room for the ADVs, giving operators direct indication of actual 4--
. valve position. The testing program the-licensee has implemented in the past (monthly testing) appeared to have= been effective until this event.
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In addition, the licensee determined that weekly testing would be performed during the remainder of the Unit 3 fuel cycle, after which time the design improvement would be installed.
Oyeration of Pressurizer Safety Valves The licensee initially determined that primary safety valve PSV-0201 opened and reseated during the transient. The valve opened prematurely.
at about 2400 psia, below the acceptance criterion of 2500 psie + 1%
provided by Technical Specifications. The determination.that PSV-0201 opened was based upon the indication of elevated tailpipe temperature downstream of PSV-0201.
To resolve these concerns,. the licensee tested and adjusted the pressur-ff izer safety valve setpoints in Mode 3 for both Unit 2 and 3 on February
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28 and March 11, respectively. The results in psia were as follows:
f ADV As-Found As-Left 2 DSV-200 2534(high)
2518 2-PSV-201 2461 (low)
2521
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3-PSV-200 2461(low)
2489 3-PSV-201-2604(high)
2512 (The Tech. Spec. low limit is 2475 and high limit is 2525.).
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In the past, the licensee has sent PSVs to Wyle Laboratory for overhaul and lift setpoint testing on a test loop. This testing was done using a valve body temperature profile established during Unit 2 startup.
This profile was also used for the Unit 3 PSVs.
However, the licensee found that the actual temperature profiles for the valves, as measured on February 23 and March 11, were different from the profile used by Wyle.
The licensee partially attributed the lift setpoint drif ting to this difference.
For the final root cause disposition, the licensee is evaluating industry problems and experience with Dresser safety valves, with particular attention to the temperature effects. The licensee also corrected incorrect wiring associated with the tailpipe temperature sensor (caused the wrong safety valve to indicate open), discovered during the setpoint testing.
Operation o_f Steam Bypass Control System Following the opening of the MSIVs, three of the four steam bypass control valves operated properly.
However, valve HV-8424 failed to open due to a malfunction of its pneumatic positioner.
The positioner was replaced and the valve tested satisfactorily.
The licensee was in the process of evaluating the failed positioner to determine the cause of failure. The licensee was replacing older model positioners with a newer model (as failures occurred), and was evaluating another type of positioner for acceptability at SONGS.
No violations or deviations were identified.
5.
Monthly Surveillance Activities (61726)
During this report period, the inspectors observed or conducted inspection of the following surveillance activities:
a.
Observation of Routine Surveillance Activities (Unit 1)
S01-12.6-3
" Fire Pump functional Test" S01-V-2.14.1
" Monthly Test of Auxiliary Feedwater Pump, SI-AFW-G-10W" S01-12.3-2
" Hot Operational Test of the Safety Injection and Containment Spray Systems" b.
ObservationofRoutineSurveillanceActivities(Unit 21 S023-V-12.2.17 "CPC (core protection calculator), COLSS (core operating limit supervisory system), and QSPDS (qualified safety parameter display system) Daily Checks" c.
Observation of Routine Surveillance Activities (Unit 3)
5023-1-2.1
" Valve - Pressurizer Safety Valve Testing"
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S023-I-2.1.1
" Valve - Verification Testing of Installed Pressurizer Safety Valve with Hydro Assist" No violations or deviations were identified.
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MonthlyMaintenanceActivities(62703)
During this report period, the inspectors observed or conducted inspection of the following maintenance activities:
a.
ObservationofRoutineMaintenanceActivities(Unit 1)
- M0 89112972000, " Replace Refueling Water Filter Pump Gland and Packing" M0 90021085000, " Replace Refueling Water Filter Pump. Inlet-Flange Bolting" M0.90031945001, " Repair Discharge Flange Leak on the East Feedwater Pump (S1-FWS-G-3A)"
During the maintenance to correct the' feedwater pump discharge, flange leak, the-inspector observed a contractor step on the pump seal supply valve, FWS-480, and inadvertently open it approximately one quarter of a turn with his foot. The inspector notified the.
Unit Shift Superintendent who dispatched an operator and the Control Room Supervisor.
It,was determined that, although the seal water pressure had increased from the normal value of approximately 275 psi to 300. psi, the condition did not affect the: operability of the
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pump.
The operator remained on the scene to observe the remaining activities and-to.. reposition the seal supply valve. Unit management
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was also notified of-this problem by the inspector. As a result of the inspector's concern, the Maintenance department initiated a Maintenance Incident Investigation Report-(MIIR) to assess this-item. The ins (206/90-07-01)pector will' review the completed MIIR as open. item
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Observation 'of Routine' MaintNan'he Activities (Unit 2)
M090032042000, " Toxic' Gas. Isolation System Train "B" Channel
Function Test.and 31 Day Calibration"-
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Observation of Routins Maintenance' Activities (Unit 3)
M090021148800,." Loop Temperature Instrumentation
~ ^ fCalibration"~
090050012000,"QSPDS'Channell"B"31DayTest"
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Unit 3 Containment Material Condition
The inspector toured Unit 3 containment on February 28, 1990.
During.that tour, the inspector noted that tools had been left v
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lying on top of S/G E-089 level tr'ansmitter 3LT-1105 (located on the 45 foot elevation)~ and pressurizer pressure transmitter-3PT-0105-3 -
-(located on the 30 foot elevation). The inspector identified this discrepancy to the licensee, who promptly removed the tools from containment. They.had apparently been left on top of the
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transmitters after completion of surveillances performed the previous day.
flo violations or deviations were identified.
7.
Engineered Safety Feature Walkdown (71710)
Unit 1 The inspector walked down the auxiliary feedwater system. The inspector-used related drawings 5178220, 5178221, and 5178222 for the inspection.
No discrepancies were noted.
Unit 2 The inspector walked down the containment emergency cooling system and-the purge system.
The inspector used related drawings and procedures S023-1-4.1, Attachment 1, " Containment Emergency Cooling-Alignment -
Unit 2." and S023-1-4.2, " Containment Purge System Alignment - Unit 2,"
for.the verification. No discrepancies were noted.
No violations or deviations were identified.
8.
ReviewofLicenseeEventReports(90712,-92700)
Through direct observations, discussion with licensee personnel, or review of the records, the following Licensee Event Reports (LERs) were closed:
Unit 1 i
89-15, Revision 1,
" Technical Specification (TS) Violation'on Safety Injection-(SI) Alignment in Mode 5" 89-33, Revision 0,
" Failure to Perform a TS Required Chemistry Surveillance"
.c 90-01, Rev'ision 0, -TS3.0.3EntryDuringDCGroundTrouble-
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90-02,kevision0,. "TS Reactor Coolant System (RCS) Sample and Analysis Requirements Not-Met Due to Personnel Error" 90-03, Revision 0,
" Failure to Sample Dddicated Safe _ Shutdown Diesel (DSD) Fuel Day _ Tank Within TS Limit Due
to Personnel Error">,
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Unit 2 83-78, Revision 1,
" Auxiliary Feedwater (AFW) Tank Level Fell Below
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TS Limit" 87-22, Revision 1.
" Fuel Handling Isolation System (FHIS) Train A and B-
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Spurious-Actuation" 88-07, Revision 1,
" Containment Purge Isolation System (CPIS) Iodine-
Channels Inoperable-Due to Detector Nonlinearity"
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89-08, Revision 0,
"CPIS-Train A Inadvertent Actuation" 89-09, Revision 0,
"InoperableLComponentCoolingWater-(CCW)andSI
' System' Snubbers" Unit 3
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89-11,' Revision 0,-
"FHIS Actuations Due-to Component' Failure"
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90.0.1, Revision 0,
"FHIS' Spurious Actuation Due to. Rad Monitor I
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No violations or deviations were identified.
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Followup 'of Previously Identified Items (92701)
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(Clos'ed) Unresolved Item (361/86-27-02)', "Pending Disposition
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Of-Pipe Break or Component Failure of AFW System in Doghouse Area" l' t
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During a previous inspection, the inspector noted that'the S/G.
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blowdown lines share the same room (referred to as the doghouse)-
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'with the isolation valves of both AFW trains. These isolation
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ialves were not includedrin the equipment environmental
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qualification (EQ) program, so the inspector questioned the~effect'
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The. inspector consulted with the Mechanical En
.the Office of Nuclear; Reactor Regulation (NRR)gineering ' Branch of
, through the Project Manager, on this matter. The NRR staff position was.that pipe breaks are tot postulated between containment isolation valves and the welded pne restraint that defines-the terminal end of the run.
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The inspector valked down the system and noted that the welded pipe restraint was mtunted on the ceiling of the room (doghouse)Dat the
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penetration which was open to the outside environment. Thus, if.a-break occurred downstream of the restraint, most of the steam would-have gone outside and not likely have affected the AFW isolation valves below..The inspector considered this explanation to be appropriate. Therefore, this item is closed.
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b'. s. '..'(Closed) Unresolved Item (361/88-25-01); " Restricted Operation Due
> To Power CalorimetriciUncertainties" i
Aiprevi6usNRCMnspectioninObtober1988'(inspectionreport 361/88-25) discussed a problem with fouling.and defouling of the feedwAter flow venturisi finipar;ticular, the licensee noted that
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ithei condition of,thenfeedwater ventur.1 (as a result of -fouling -due
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J to material from feedwater _heaterCcopper intrusion). affected the L
< secondary calorimetric accuracy. ~This in; turn caused errors in the
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nnuclear instrumentation calibration.'
As' a result of this problem,
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the licensee developed ~a ne# methodology, for approeimating' reactor-
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power; assessed the significance of-this error, and, reviewed the-
potential for 'this phenomenon during the Unit's opereting history.
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As corrective 3 action, the licensee, developed a methodology. to use l
,other.pararaeters -independent ofifeedwater flow toimeasure the -
thermal, power of the reactor.. This methodology was vali. dated and f proceduralized in S023-V-2.h " Reactor Power Verification using
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' Secondary Indicators," and was used to verify the secondary cciorimetric 'ddring-power nnsion testing; from the <1est q
refueling: outage.
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'From the new methodology developed, the licensee ^ determined th'at in.
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December 1988, power exceeded 100% slightly due to the error
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resulting from the venturi fouling /defouling. However, reactor
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power level did not exceed 102% at that time, and the. condition was j
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therefore bounded by the safety analysis.-- In addition, tne-licensee reevaluated other occasions when the flow venturilindi-
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cated improperly. The licensee found'that between December 23, 1983 and January 4, 1984, when one of the, venturi's pressure. taps--
failed, the calibr.ation error resulted in the reactor operating at a
103% power. The licensee evaluated the operating condition-and
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concluded that the safety significance was minimal due to safety 1y margins inherent in the Core Protection ' Calculator (CPC). This occurrence wat discussed in.LER 88-35, Revision 0 for' Unit 2.
The inspector considered that the-lisensee's actions were-appropriate, Therefore, this item is closed.
No violations or deviations were identified.
10. On-Site Review Committee Meeting (40500)
y The Unit 1 On-Site Review Committee (OSRC) met on March 20, 1990. The-OSRC discussed the previous month's operations and reportable events (LERs). The LERs discussed referred primarily to three missed surveil-lances. The committee discussed the' generic. implications of the number of TS surveillances missed.
It,was concluded that a. Quality Assurance L
review of the Station's implementation of newly issued TS surveillance
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requirements was in o'rder. Quality, Assurance was' requested by the OSRC.
chairman;in writing on' April 10,>1990 to, perform this review. The.
meeting ~ was conducted in a formal manner and offered opportunities for
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people to raise any safety concerns! they might have, although'none were raised at that time.
I No violations or deviations were. identified.
11.
Followup of Licensee Actions on NRC Information Notices (92701)
The inspector reviewed the licensee's action on the followicg NRC-l Information Notices.: The purpose was to dete mine whether the licensee
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had taken appropriate actions in response to the issues discussed.
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The following NRC Information Notices were reviewed *
l 86-92
" Pressurizer Safety Valve Reliability"
l 88-60;
"Setpo' int Testing of Pressurizer Safety Valves With-
Filled Loop Seals Using Hydraulic Assist Devices"
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'89-90
" Pressurizer Safety Valve Lift Setpoint Shift" The inspector notEIt' hat the licensee's Independent Safety Engineering l
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GroupL(ISEG) had analyzed the concerns identified in the Information Notices ;for potential; impact at San Oriofre. This included an assessment
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of applicability; the needifor changes to training,;additicns to.the QA audit program, and addition of the component to the Control of Problem-
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Equipment (00PE): program. No discrepancies were noted.
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No Yiolations or' deviations were, identified.
12. Fitness For Du_ty -(TI.2515/104)'
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The inspector attended selected, licensee Fitness-for-Duty (FFD) training l
sessions to determine whether required; training was being conducted to
implement the FFD program. The session _s attended included the FFD policy awareness training _ for general employees, which included ~ escort i
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training and the FFD training'forisupervisors.- The inspector considered
.the licensee's training programs for FFD to be adequate.
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No violations or deviations were identified.'
13. Unit 1 Emergency Ixerci_se (82301)-
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The licensee conducted an emergency exercise on March 6, 1990 to assess their ability to make a containment entry on a loss of all AC power and isolate either valve S1-LDS-003 or SI-LDS-001 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> as
described in the interim disposition for nonconformance report (NCR)
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S01-P-7416. The' actions described were considered necessary because.
letdown orifice isolation valve CV-203 was leaking by as much as 15 gpm i
when fully closed. The concern was that the 15 gpm leakage would eventualb cause the letdown rrelief valve to lift, passlag water to the pressurizer relief tank.
Eventually, the tank would'become full and the rupture disk would rupture, dumping water to the containment floor.
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Valves SI-LDS-003 and S1-LDS-001 are ' manual isolation' valves in the letdown system. The exercise demonstrated-that Operations was-fully preparedLfor a containment entry within'45 minutes and it was estimated.
that less than.10 additional minutes would.have been necessary to actually isolate one of these valves.
No-violations-or deviations were' identified.'
14.
Exit' Meeting (30703)
On Msrch' 22, 1990, an exit meeting'was-conducted with the licensee.
representatives identified in Paragraph 1.
The inspectors' summarized the inspection scope and findings as described in the Results section of this report.
The licensee acknowledged the inspection' findings and noted:that appropriate corrective actions.would be implemented where warranted.
The licensee did not identify as proprietary' any of the information providedtoorreviewe{by-theinspectorsLduringthisinspection:
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