IR 05000206/1990022
| ML20055F392 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/27/1990 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20055F388 | List: |
| References | |
| 50-206-90-22, 50-361-90-22, 50-362-90-22, IEIN-90-018, IEIN-90-18, NUDOCS 9007160377 | |
| Download: ML20055F392 (33) | |
Text
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. - gg-p ,p p REGION ! Report' Nos.-. _50-206/89 4 2. 50-361/89722, 50-362/89-22 Docket Nos.
50-206, 50-361, 50-362
License Nos.- _DPR-13, NPF-10, NPF-15
b L Licensee: Southern California Edison Company a-Irvine Operations Center ' 23 Parker Street ' Irvine, California 92718 , u-
- Facility Name:
San Onofre Units 1,-2 and 3 ! - Inspection _at: San Onofre, San Clemente, California .oq l Inspection conducted: April 29 through June 2, 1990 g i E LInspectors: C. W. Caldwell, Senior Resident Inspector A. L. Hon, Resident Inspector
'C. D. Townsend, Resident Inspector h 5/47dC p Approved lBy:
' P. VI. -Johnson, CTief Date Signed ReactorProjectsSection3 i ' ' Inspection Summary V.
. Inspection on' April'29, 1990-through June 2, 1990 (Report Nos 50-206/90-22, - '50-361/90-22, and 50-362/90-22) ' l Areas Inspected: Routine resident inspection of the Units 1, 2 and-3 operational safety + operations program including the following areas: " verification, radiological protection,-security, evaluation of plant trips andL monthly surveillance activities,- monthly maintenance-activities : q , -events refue1Ing, activities independent inspection, licensee event report review, ! .followupofpreviouslyidentified-items,andfollowupofviolations.
-Inspection procedures 30703, 35502-1, 37700, 37828, 41701 60705, 60710,- 61726,62703,71707,71710,71711,72700,90712,92700,9E701,and93702were r
'L utilized.- Safety Issues Management System (SIMS) Items: None . - $h
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~ Results:- General Conclusions and Specific Findings: The licensee exhibited aggressiveness.in investi ating the source of debris found:in the Unit 3 steam generators (S/G during secondary side l .slud e lancing operations.
The results of their investigation revealed . sign ficant deficiencies with-S/G feed spargers, that wi11 ' require a r substantial amount of repair.
In addition, the licensee exhibited-conservatism by deciding that a Unit-2 shutdown will-be necessary to ' inspect the:S/G spargers based _on conditions found in Unit 3 (Paragraph 8).
A concern was identified that the licensee's staff may misunderstand . i, their abilit Leation-(TS) y.to make repeated voluntary entries into Technical S)ecifi.
. action statements.
Licensee representatives stated t1at
a action statements for components which are addressed by the TS could bei
R entered repeatedly as long as the TS action statement time limits w re not-exceeded.
This appeared.to be-a nonconservative interpretation of_ p Technical Specification requirement (Paragraph 9.b).
i 'A r'eview-of several nonconformance reports (NCRs) was performed during . this and )revious inspection periods.
There was a 3erception, on the
part of tie-inspector, that improvement could be ac11eved in the technical disposition of some NCR's by incorporating.a more questioning , attitude.(Paragraph 9.b.). l, 'Significant Safety Matters: ! t None-q y Summary of Violations and Deviations:
No violations were identified. One deviation was identified concerning' the~1ack of a filter for the supply fan to the Unit 1 new fuel storage i building (Paragraph 11.b).
U Open Items Summary: During this' report period, one new followup item was opened and 11 were' l closed.
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g , r , ' DETAILS q y \\ 1.
Persons Contacted.
Southern California Edison Company H. Ray, Senior Vice President Nuclear
- H. Morgan, Vice President and, Site Manager
- R. Krieger, Acting Station Manager
- B. Katz, Nuclear Oversight Manager, NES&L K. Slagle, Deputy Station Manager
- L.-Casa, Maintenance Manager
- M. Short, Technical Manager M. Merlo, Nuclear Design Engineering Manager, NES&L (
P.'Knapp, Health Physics Manager
D. Peacor,. Emergency Preparedness Manager ' 'D. Herbst, Quality Assurance Manager, NES&L ! C. Chiu, Quality Engineering Manaser J. ~ Schramm Operations Superintencent,, Unit 1
- V. Fisher,, Operations-Superintendent, Units 2/3 R. Rosenblum, Manager, Nuclear Regulatory Affairs
- L..Brevig, Supervisor, Onsite Nuclear Licensing T. Calloway, Substance Abuse Program Manager
.*R. Plappert, Compliance Manager San Diego Gas and Electric Company i
- R. Erickson, Site Representative City of Anaheim-
- G. Edwards, Site Representative City of Riverside
'*C. Harris, Site Representative
- Denotes;those attending the exit meeting on June 5, 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, compliance engineers, maintenance craftsmen, and health-physics engineers and technicians.
' 2.
Plant Status- ! < Unit 1 During this inspection period, Unit 1 experienced two plant trips.
At - a 10:22 p.m., on April 30, 1990, the Unit tripped due to a ground in the , low flow trip circuitry for reactor coolant loop 'B'. A reactor trip is normally expected for 1-out-of-3 low flow signals, and all circuits and s . .
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. -{ ' y . y W* systems worked as designed.
The Unit returned to service on May 3, hen a i 1990. -At 2:43 a.m.,_on May 15,C' steam 1990, Unit I was manually'The event was tripped w , low level was observed in the enerator(S/G).- n - initiated by an induced power transient n vital bus.3A, causin its-a inverter to transfer contact for the 'C' S/ power supplies.During the transfer, a hi h level G deenergized, causing the associated valve ! isolation logic to close feedwater regulating valve FCV-458.
This, in: '! turn, caused the S/G level to decrease as observed by operations personnel who correctly. initiated the reactor manual trip per procedure.
The Unit. returned to service on May 20, 1990 and operated at power the remainder of the inspection period.
Unit 2
The Unit operated at power without significant operating concerns during this period.
Unit 3 ' ! The Unit continued the Cycle 5 refueling outage during this inspection period.
- 3.
Operational Safety Verification (71707) ' The inspectors performed several plant tours and verified the operability of selected emergency systems, reviewed the tagout lo'g and
verified proper return to service of affected components.
Particular ' attention was given to housekeeping, examination for potential fire J ! ' hazards, fluid leaks,had been initiated for equipment in'need of excessive vibration, and verification that maintenance requests i maintenance.
The-inspectors also observed selected activities by _.
licensee' radiological protection and security personnel to confirm a proper implementation of and conformance with facility policies and procedures in these areas.
No violations or deviations were identified.' 4.
Evaluation of plant Trips and Events (93702) Reactor Trip Due To Reactor Coolant Low Flow Signal (Unit 1) 'm i At 10:22 p.m., on April 30, 1990, Unit 1 tripped due to a reactor . coolant low flow signal caused by a ground in the reactor coolant loop 'B' flow circuitry (above 50% power, this signal automatically.
initiates a reactor trip).
I: u As a followup to this event, the licensee held a series of meetings to address root cause considerations, initiated electrical cable analysisgeneric im corrective actions.
The licensee ' and diagnostic (ECAD) testing of the cables to the reactor coolant flow transmitters and determined that the instrument cables to both the 'A' and 'B' transmitters needed replacement.
Since a plant tri) occurred last August from a similar ground in the 'C' transmitter ca)1e, cable i- . .
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replacement fortthe 'A' and 'B' transmitters had already been scheduled - - for the outage'beginning June 30 1990.
Consequently, the necessary~
cablewasalready.availableon-site.
As corrective action the replaced and the system tested satisfactorily. plifier were, subse instrument cable and the 'B' transmitter pre-am i' The Unit returned to service on May 3, 1990.- Manual Reactor Trip Due To Low Steam Generator Level (Unit 1) i On May 15, 1990 at 2:43 a.m., Unit I was manually tripped when the i water level in $/G 'C' was observed to have decreased below the required manual trip initiation point of 5% on the narrow range indicator.
The low S/G level was caused by the closure of feedwater regulating valve FCV-458.
The transient began when e.n electrician working on the steam driven auxiliary feedwater (AFW) pump steam inlet bypass solenoid valve inadvertently touched the position indication lead to-the valve casing ., (while-the wire was energized), which included an electrical transient on vital' bus 3A.
This in turn caused the vital bus inverter to automatically switch power supplies, allowing the high S/G 1evel contact to deenergize and energizing a relay which provided a close signal to y FCV-458.
FCV-458 subsequently closed and flow was terminated to S/G 'C', ~' resulting'in the low level observed by the operators.
During the investigation of the transient, the licensee concluded that it y was unsatisfactory to operate with a system that caused a reactor trip
the. licensee decided to remove the high S/G 1evel/ plies.feedwater regulating As a result, because a vital bus transferred between power sup , valve isolation feature for the following reasons: '
The= function is not required by the operating license.
- -The configuratior. is not recognized by the NRC as a steam generator - , overfill protection.
o lit was unacce) table for this configuration to cause a plant trip- - considering that it is not required or recognized as adequate S/G overfill protection, j As corrective action, the licensee implemented a temporary facility modification to remove this signal from the circuitry, recognizing that S/G overfill protection is being reviewed as part of unreviewed safety i issue 47 (in conjunction with generic letter 89-19).
The modification was then tested and the Unit returned to service on May 20, 1990.
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No violations or deviations were identified.
5.
MonthlySurveillanceActivities(61726) During this report period, the inspectors observed or conducted inspection of the following surveillance activities: ! .. .. - -.
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0bservationof_RoutineSurveillanceActivities(Unit 1) ~
. S01-3-2, " Plant Startup From Hot Standby To Minimum Load" , 501-V-5.6, " Cycle X Reactor Vessel Thermal Shield
Monitoring" -S01-12.4-2, "Non-Routine In-Service Testing Of Valves, . SI-FWS-CV-100B" , t b.
Observation of Routine Surveillance Activities (Unit 2) M090031300, " Units 2/3 Common Early Warning Fire Detection ' Surveillance Testing (Outside Containment)" s . $023-11-1.1.5, " Surveillance Requirement, Reactor Plant Protection System Logic Matrix functional Test" ' , 0)erability Determination on Unsuccessful'Surveillances ' (Jnit 2) On May 10, 1990, while the Unit wasLoperating at. power, the
- licensee performed a routine surveillance on channel 'C' of the
~ S023-II-1.1.5 specified:reactorprotection-system [DepressandHOLDtheA RPS).
Step 6.1.25 of procedure
Hold pushbutton and verify... The RPS Actuation lights (2) are
extinguished on the appropriate channel Bistable-Control Panel."
When:theinstrumentandcontrol(I&C)techniciandepressedthe pushbutton, he observed the associated relays deenergize as-expected.
However the RPS Actuation Test Lights did not-extinguishasspecIfiedbytheprocedure.
In accordance with Attac1 ment 2 to procedure 50123-1-1,3,_" Maintenance Documentation," the technician initiated a " Notification of Unsuccessful.
' Surveillance" to document the failed surveillance.
Based upon his , . -observation that-the relays did deenergize and hence the associated , reactor trip breakers (RTBs) will open when called upon, he
determined that the channel was o)erable.
Also based upon the - symptomsobserved,hesuspectedt1atthetestIIghtanomaly(wasdu
to a failure of one of the two solid state relays (SSRs).
The I path.)g of one of the two'SSRs is sufficient to open the tripThe technician openin the problem and notified the Control Room Supervisor (CRS) who concurred with the Operability determination.
On the following day, the System Cognizant Engineer learned about
the anomaly.
Because of the lack of confidence that the remaining ! identical SSR would not fail, he initiated an NCR for the affected RPS channel.
Based on this NCR, 0)erations placed the channel in bypass and conservatively opened tie associated RTBs.
Later, the i SSR was replaced and the channel was satisfactorily tested and returned to service.
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. .e l* ' ~~ ' 5, ,, ! The inspector reviewed this event and considered that while the l channel would perform its safety function with one failed SSR, the .! existing procedure to assess operability under the circumstances l did-not provide for an independent review.- Specifically the ! operabilitydeterminationupon-afailedsurveillancerelled .! primarily on the individual performing the surveillance.
It did not i saecify an independent review by someone who was knowledgeable of tie system, but not directly involved in the activity.
The licensee ,i concurred with the inspector's finding and committed-to revise
S0123-I-1.3 to include an inde)endent operability review of any t unsuccessful surveillances.
T11s revision will be implemented by
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June 22 -1990.
The inspector considered that the licensee was .i responsIvetotheconcern.
Therefore, this item is closed.
(361/90-22-01) c.
'ObservationofRoutine'SurveillanceActivities(Unit 3) , 'S023-3-3.25, "Once a Shift Surveillance" " No violations-or deviations were identified.- g l 6.
MonthlyMaintenanceActivities(62703) During this report period, the inspectors observed or conducted ! inspection of the following maintenance activities:. a.
Observation of Routine Maintenance Activities (Unit 1)
M090051700001, " Main Feedwater Pump, SI-FWS-G-3B, Replace i
.SleeveBearing(s),CompleteInspectionsAnd n' ' -Calibrate To Disposition NCR 90050137"'
M090050890000, " Auxiliary Feedwater Pump G10 Steam Inlet Bypass Solenoid Valve Repair" M089083655003, " Steam Generator Blowdown Valve SI-FWS-CV-1008,HasSevereScorIngOfThe t1 Spindle / Lower Packing Leakage-Overhaul Valve"- ll M089052955001, " Spent Fuel Pit Bridge Crane Bridge Drive i
Brake Drum Is-Scored-Replace } L M089102983000,'" Spare Spent Fuel Pit Pump, SI-SFP-G-5A,
Place The Spare Pump In Service" " Note: Inordertoplacethesarespentfuelpitpuhsically into service, the in-line spent fue pit pump had to be p i
f disconnected from the suction header and a spool piece had to be connected to route the suction flow to the spare spent fuel pit pump.
This required a maintenance order.
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Observation of Routine Maintenance Activities (Unit 2) ' None i .
- .- - - - -. . - - . - e ., , g 4, - . - l c.. ObservationofRoutineMaintenanceActivities(Unit 3) DCP6605.51 Control Room Modification M089112186000, "ESFAS and Plant Protection System Instrumentation Calibration" M089101145000, "ESFAS Auxiliary Cabinet Spare Wiring Removal" M08808033001, "LowPressureSafetyInjectionPumpMotor ! Cooler Casting Leak Repair" -No violations or deviations were identified.
.f 7.
Engineered Safety Featu*e Walkdown Unit 1.(71710)- ' The Unit 1125 VDC on-site electrical distribution system was inspected durin$thisperiod.
The inspector utilized the System Description 50-$0 140, single line drawings 5146828,5102173,5149348,5191975and
5196035 an 501-2.6-4, " Loss of DC Bus." During this inspectlon,dprocedureno' discrepancies were noted.
' No violations or deviation's were identified.
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! '8.
plantModificationandRefuelingActivities(37700,37828,60705} 60710, 71711, 72700 j i The inspector observed the following outage activities:
1 L Fuel Reloading l - L
S023-XXVI-9.6533.0.1 Anticipated Transient Without Scram I - < ' / Diverse Scram System logic Test -' S/G Feedwater Sparger Damage and. Repairs - See discussion below No discrepancies were identified during this inspection period.
S Steam Generator Feedwater Sparger Damage (Unit 3) l l During the refueling outage, the licensee found debris inside the y secondary side'of S/G E-088 and E-089 while performing sludge lancing.
' Upon entering both S/Gs, the licensee discovered significant degradation of the feedwater spargers.
A detailed analysis of the conditions was initiated..
l \\ L Each S/G was designed and constructed such that the feedwater line '
L-penetrates the S/G through a welded nozzle coupled to the distribution l box by a thermal sleeve. The feedwater sparger is welded to both sides of the box (perpendicular to the feedwater line) and the feedwater exits l the sparger through J-tubes welded along the top of the sparger.
The ! I original sparger was made of schedule 40 carbon steel A-106 material and I attached to tie distribution box by partial penetration fillet welds.
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. , y,; e , $ . - . 'When the Unit'was undergoing'preoperational testing in 1982,lt a water hammer event occurred that collapsed the sparger. As a resu it was ' repairedusingaslightlydifferentdesigninwhichtheschedule40 ' carbon steel was replaced with schedule 120 carbon steel.
This was done sothatthespargercouldwithstanda.completevacuumwhen.subjectedto full S/G operating pressure.
The new sparger was welded to the undamaged sections (pu) piece) of the original sparger that were still attached to the distri)ution box.- To enhance the pressure equalization across the.sparger during transients,ddition, a schedule 80 vent pipe the J-tubes were enlarged from two inch to three-inch in diameter.
In a with a flow diverter (Tee piece) was added to reduce gases which could collect at the top of the distribution box.
During this refueling outage,inal schedule 40 saarger had eroded.the lice < amount of metal from the orig The a - - erosion was prominent at the weld to the distri)ution box.
Also, the " vent Tee pieces of both S/G E-088 and E-089 were found to have eroded.
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(The licensee retrieved most of the debris from both S/Gs prior to (, initiation of corrective actions and plans to continue to collect debris L after completion of repairs.) Additional ultrasonic inspecticr. of the L distribution boxes and the schedule 120 spargers did not reveal wall thinning beyond specifications.
Thus, the licensee believed that the , erosion was confined to'the pup pieces and the tee pieces.
To assess the potentia 1' significance of S/G tube wear by debris, the licensee reviewed eddy current inspection data that was performed during - this outage and noted no abnormal wear.
Furthermore the licensee performed additional eddy current inspection on the-tubes surrounding o L the debris that had been found.
Most of the tubes inspected did not . show any abnormal wear with the exception that three tubes showed fretting (apparently due to the debris).
However the indications were stillwithinthetubewallthinningacceptancecrlteriaspecifiedinthe o TS.
To aid in the root cause determination, the licensee removed the damaged < l pup pieces and Tee pieces from both S/Gs for examination.
The results of ' that examination supported a theory that thermal stress from cold AFW , flow. cracked the pup pieces at the weld points.
Once'the crack was l initiated the accelerated flow and turbulence through the crack eroded L . thematerlalandfinallytoreportionsofthepuppiece.
The licensee
' reviewed the Unit operating history and found that the Unit was in Mode 3 I - for extended periods during preoperational testing.
In that condition, AFW was used to feed the S/G in a' batch mode instead of a continuous modulated flow.
This resulted in many thermal cycles on the sparger, , p which may have initiated the crack at the pup piece weld where the , stresses were concentrated.
As for the Tee pieces, water wear marks on the remnants suggested high velocity flow erosion due to the installed configuration.
The licensee began repairs to the S/Gs using a redesigned sparger pup piece.
The schedule 40 pup aieces were removed and re) laced with schedule 120 aipe as with tie rest of the sparger.
T1ey will be attachedtotiedlstributionboxbyusingaprefabricatedweld-o-letand a full penetration fillet weld arrangement.
It is expected that this ,
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' ' ,c y - , will minimize the stress concentration at the oint The Tee-ieces will-beeliminatedbecauseitwasdeterminedthatth'eyar.eoflimit!d , ' ' ' usefulness.
In addition, the rigid sparger supports will be.reconfigured to allow for axial expansion, thereby reducing thermal stresses.
Since Unit 2 has-the same S/G design, the licensee plans to shut down the ' Unit:for inspection and modification after. Unit 3 is returned.to service about July 7, 1990.
The licensee's engineering and technical . - organizations performed a safety evaluation concerning the limited continued operation of Unit 2 with the potential for sparger problems as found in Unit 3.
This evaluation was subsequently reviewed and approved by the On-Site Safety Committee and submitted to the NRC.
It addressed several operational concerns with the most significant being the . potential for a S/G tube rupture due. to debris.
The licensee determined that'from experience at SONGS and-the industry, a complete tube failure is usually preceded by increased. leakage over a period of time.Thus,foradm This would allow time for an orderly shutdown.)
' control, the licensee lowered the threshold for the air efector radiation alarm to provide early warning of increased arimary to secondary leakage.
Furthermore since the Unit's restart from tie Cycle V refueling outage on December,8,1989, primary activity has' been less than that of the secondary activity assumed in the accident analysis.
Thus, any off-site radiological release due to a tube rupture would be within the amount accepted by previous.-analysis provided that there is no concurrent fuel failure due to the tube rupture.
To provide additional conservatism, the licensee also modified the departure from nucleate boiling (ratio - (DNBR) margin in the Core Operating Limits Supervisory System COLSS).
.This was done.to ensure that no significant fuel failure damage would occur should one of the limiting transients postulated in the Safety , Analysis take place.
The NRC' reviewed the licensee's safety evaluation for continued itit 2
operation and considered it to be acceptable.
The restart of Unit 3 is estimated-to be about July 11, 1990 and Unit 2 will be shutdown in that
timeframe for inspection and modification.
The inspector will continue
to follow the' licensee's final root cause and modification efforts.. ' (362/90-22-01) L .No violations or deviations were identified.
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Independent Inspection i a.
Observation of implementation of Administrative Controls At 12:50 a.m., on May 15, 1990, theUnit1steamgenerator(S/G) blowdown valve to the plant outfall, SI-FWS-CV-1008, failed its stroke time requirement during performance of a routine in-service test (IST).
The required IST valve stroke close time was less than 5 seconds, but it actually stroked closed in greater than 15 , l seconds. At the time, SCE was preparing a change to the Unit 1 Technical Specifications (TS) to include this valve as a o L containment isolation valve as identified in TS 3.6.2.
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n i lbeing the case, SCE conservatively elected to consider that thei actions in TS 3.6.2 should be complied with.
-TS 3.6.2 states that with one'or more of the isolation valves
specified in Table 3.6.2-1 inoperable,...one of the following
' actions shall be taken:
"1; Restore the ino twithin 4 hours,perable valve (s) to OPERABLE status ' or-- 2.
Isolate each affected penetration within 4 hours by use of at , least one deactivated power operated valve secured in the isolation position, or 3.
Isolate each affected penetration within 4 hours by use of at i least one closed manual valve or blind flange, or
I 4.
Be in at least HOT STANDBY within the next 6 hours and in COLD , SHUTDOWN within the following 30 hours."
i In this case, the licensee took actions as specified in TS 3.6.2, ! action 3,.and closed CV-100B and the associated manual isolation t valves to satisfy the TS requirement.
( . The licensee then performed the following valve manipulations: ' ., At 3:24 a.m., the licensee re-opened the manual isolation - - valves and CV-100B and reestablished this flowpath to blowdown t the S/G to the outfall.
, At 7:00 a.m., blowdown was secured and the TS action was - again complied with.
.At 11:20 a.m., CV-1008 and the manual isolation valves were I - again opened to blowdown the steam generators, j > l= At 1:20 p.m., blowdown was secured.- - At 3:29 p.m., CV-1008 and the manual isolation valves were - l again opened to reestablish blowdown.
l, At 5:36 p.m., CV-100B was restroked after some minor
- l' maintenance and declared operable.
L At 8:28'a.m., on May 22, 1990, CV-100B again failed to stroke closed within the-5 seconds time requirement.
The actual stoke time was 7.69 seconds.
Consequently, the licensee shut CV-100B and ' the manual isolation valves in accordance with TS 3.6.2.
An additional stroke test was performed at 6:30 a.m. and CV-1008 stroked in 5.7 seconds.
Four minutes later tie licensee established S/G blowdown to the outfall using CV-1008 and the manual isolation . valves.- At this point, the licensee notified
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g7 ' ~ l'0 O 4 gg < a~' e , theinspectorthatproblemsexistadwiththestroke-timeofCV-100B.
l At9:30p.m.lytwotimes.in4.36and4.53secondsandwasdeclared the biowdown was secured.
Later, CV-1008 stroked R satisfactori i > operable at 10:15 p.m.
At10:08a.m.,onMay23 1990, CV-100B j stroked closed in 39.84 seconds, once again over,the required time of 5 seconds.
the isolation valve to the blowdown. tank At 6:50 p.m.,-on May 23,d satisfactorily and declared operable as, CV-100, was stroke teste - an alternate path for steam generator _ blowdown.
Subsequently, < . blowdown was commenced through the steam generator sample lines to the'outfall (via the reheater-pit sump) on May 24 1990, at 12:30- .'(Thisalignmentallowsa5gpmcontinuousblowdown.)
a.m.
The inspector recognized that CV-100B was not officially included ' , in the.TS and that the licensee was acting conservatively by , interpreting the actions specified in TS 3.6.2 to be applicable:to a CV-100B.
The adequacy of administrative controls applied to - ' operation of CV-100B were discussed with the licensee and will be ' followed up in a subsequent report. (206/90-22-01).
b.
Implementataion of Technical Specification kauirements (35502-1) During discussions concerning the administrative controls applied to CV-1008 a question arose as to when the licensee could - deliberateIyenter.theactionstatementofaTS.
Some licensee . personnel stated that they could repeatedly enter an action statement as long as the action statement time limit to restore the component or system without further action was not exceeded during.
each entry into'the action statement.
The NRC endorses voluntary- ' entry into the action statement conditions for purposes of testing and maintenance.
Therefore, the TS are structured to permit the licenseetoexercisejudgementwithinthelatitudepermittedbythe action statement language in the TS for the period allowed by the-action statement. However, the NRC does not consider that repeated _ ! L entry into an action statement can be done for convenience in a ' L manner which effectively extends the out-of-service period allowed _.
by_the action statement.
In such cases, appropriate action should be taken on the licensee's part to comply with the Technical , S)ecification or other appropriate action.(i.e. TS relief for the u L s1 ort term and a TS amendment for the long term).
This interpretation ' L was discussed with the licensee to clarify the NRC's position.
, , A question regarding the licensee's ability to make voluntary L entries into Section 3.0.3 of the Technical Specifications was also raised durin this period.
Generic Letter 87-09 ! " 4.0ofTheSkandardTechnicalSpecifications(ST$) Sections 3.0and L on The Applicability"of Limiting Conditions For Operation And Surveillance Requirements, provides the NRC's position on this issue.
A limiting condition for operation (LCO) as s l l ' . . - - --- - - - - --. - . .1
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<> ,} definad in STS 3.0.3'is not intended to be used as an operational convenience.
STS 3.0.3 permits redundant safety systems to be'out.
of service for a limited period:of time.
Voluntary entry into LC0: 3.0.3 deliberately removes a train of-a system from service when the redundant train is also inoperable.
In the-case that voluntary entry into LCO 3.0.3 is necessary for testing or maintenance and the redundant component or system is operable (i.e., lack of a specific action for a condition when only one component-or train is inoperable). steps should be taken to amend.the Technical Specifications on a timely basis to eliminate the need for routine.
voluntary entries into LCO 3;0.3.
i c.' Technical-Content Of Nonconformance Reports (35502-1) L 'The inspector reviewed several nonconformance reports (NCR's) and considered that improvement could.be achieved in the technical' disposition of some NCR's by incorporating a more cuestioning attitude.
The review was limited to NCR's that hac received j visibility due to the perceived importance of the topics.
Three n examples;are presented below:- 1) NCR S01-P-7441, discussing the potential for water hammer due-to increased flow of the auxiliary feedwater system, technically relied on the results of testing performed at -. Indian Point-2 as documented in Westinghouse Technical Bulletin TB 75-7.: 1The licensee correctly identified this TB as - . providing the basis for water hammer in-this~ system and relied on the initial. Westinghouse opinion that aaplication of this~TB-to revise the flow requirements for water 1ammer to be less restrictive was acce) table.
However, upon further review, it o was concluded that tie comparisons between-the two plants and , - ' the complicated nature of water hammer analyses made the relaxation of the flow requirements nonconservative and further study was_needed.. ' l' -2) New fuel building NCR 90040048 addressed whether the new fuel L building was required to conform to ANSI Standard N45.2.2-p and whether the building could accommodate the requirements
L established in the standard.
The licensee determined that the ! building was not required to conform to the ANSI standard, but _l that it did comply with the requirements as specified by_the-l fuel vendor.
Inspection report 206/90-17 erroneously stated
, L that the licensee acknowledged that the new fuel building conformed to the requirements of the ANSI standard.
L The facility's capability to maintain a controlled L environment was first questioned by the discovery of an
apparently missing filter in the supply fan suction.
The NCR !
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'y determined that'the filter needed to be re laced, but did not
E 'further question the adequacy of the build ng's ventilation i ' system.- During the arocess of dispositioning the NCR, the inspector observed t1at louvers on the west wall could also-affect the ventilation system operability and questioned the licensee on this issue.
It was determined by SCE that these l , louversalsoneededadjustmentsbeforetheventilationsystem '
could perform it's intended function.
, , 3) The! relief valves on the emergency diesel generator (EDG) air l receivers came into question when a potential defect of a:
Crosby model JMBV relief valve was-identified by Energy D Services Group in a 10 CFR part 21 report.
(NRCInformation-Notice No. 90-18 " Potentia 1' Problems With' Crosby Safety Relief .' Valves On Diesel, Generator Air Start Receiver Tanks," also q-discussed the problem.) The discovery was that the relief.
t valves had relieved to a point below the EDG automatic start? ! lockout: signal.
The Part 21 report was clear as to what - actions should be taken and the licensee complied.
However, ' the NRC-Information Notice questioned the seismic y qualifications of these valves and appeared to be unclear as to the cause of the )roblem.
The licensee had considered the . seismic aspects of t11s problem to a degree, but had not documented their findings in the NCR.
Due to the importance. ';of the component in question and the consequences of the worst l' case scenario, the -inspector considered that-it would have-j been prudent to address this in.the NCR.
The licensee acknowledged that there could be improvement in the ' technical aspects of these NCR's.
The resident inspectors will-j > , continue to monitor the licensee's progress in this area..This
itemisclosed(206/90-22-02).
e d.
Unit 1 Simulator Training At The Zion Training Facility (41701)- l q As an extension of a. week of NRC training conducted in Region III, a the ins?ector traveled to the training center operated by o L Westinglouse for Zion and other Westinghouse facilities to observe ~o simulator training for Unit 1 operations' personnel.
> , The inspector attended the session on the 10:00 a.m. to 6:00 p.m.
shift on May 7, 1990.
The session began with a review of two a significant operating events in a classroom setting.
The ' instructors had the operators read the material and then the issues , were discussed to fully understand the mistakes made and the ? important lessons learned.
The o)erators then entered the I simulator and,-utilizing control aoard overlays designed to make the Zion control boards look more like SONGS 1, converted the <
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1-simulator to more closely resemb'le the SONGS 1 control room.
The computer software package was handled by the Westinghouse instructor.
The operators then went through a control board familiarization period while each person went through the process-of_ lining up and synchronizing a diesel generator to its respective bus.
The shift then performed normal startup manipulations, load increases, and responded to a spurious Safety Injection actuation.
Positions on the boards were rotated so that each operator had the . opportunity to operate-a different part of-the plant.
The , -inspector' observed that the operators were not able to
instinctively locate the annunciators and equipment control devices as they appear to do at SONGS 1.
The insaector considered that' this made the operators think through eac1 action to fully ' understand each indication and manipulation.
This was especially apparent ~when an operator inadvertently manipulated an incorrect switch and noted that it was because of the differences between the-simulator and the SONGS 1 control room.
The operator realized that ' it would require increased attention to detail to manipulate the-controls'at the simulator.
As a result'of these observations, the inspector considered that the training conducted was very educational.
It was shown that the chan e in operating environment caused the operators to carefull -think through each evolution which forced them to thoroug understand exactly what-they were doing at all times.
Adding t s increased level of difficul_ty enhanced the understanding of the simulated operations, e.
Seminar On Timely Creation Of NCRs (35502-1) The inspector attended one of a number of seminars presented by-licensee personnel on the: timely creation of NCRs.. This seminar was-given to all station technical personnel and first line maintenance and planning supervisors.
The purpose of the seminar was to describe a number of SONGS events in which NCR initiation was-delayed and to describe how the inappropriate use of the " root cause" NCR prevented.the performance of a required safety evaluation in the,recent Unit 2 taper pin event.
In addition, the seminars were designed to explain the lessons learned from the . SONGS events'and tie desired methods for preventing their recurrence.
The licensee directed that individuals should write an NCR when it is believed that a nonconforming condition exists and communicate the NCR information formally.
No violations or deviations were identified.
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10.
Review of Licensee Event Reports-(90712, 92700) Through direct observations, discussion with licensee personnel, or !
- review of the records, the following Licensee Event Reports (LERs) were l-closed:
, Unit 1 I ' 90-03, Revision ~0,- " Failure to Implement Technical S)ecification > ., 92-Day" Dedicated Shutdown Diesel
- uel Day Tank
.e , ( Sample ] ' 90-04,. Revision 0, " Potential For RCS Leak Greater Than 6 GPM Due to Procedural Inadequacy" , 90-04, Revision 1, " Potential For RCS Leak Greater Than 6 GPM-Due.to Procedural. Inadequacy" '90-06, Revision 0, "Nonconservative Failure Mode of Chemical Volume and Control Valve CV-406B'! " Reactor Trip On A S 90-07,. Revision 0,' Coolant Flow Signal"purious Low Reactor i 90-08, Revision 0, " Technical Specification 3.0.3 Entry Due To ' AnInoperableSafetyInjectionPumpBreaker" 90-10, Revision 0,." Spent Fuel Pit Cooling Maximum Heat Load Greater Than Design" Unit 2 90-01,-Revision 0, " Missed Fire Watch Due to Procedural Inadequacy" [ 90-02, Revisior. O, " Toxic Gas Isolation System Actuation Due to i Packing Leak on Condensation Ammonia Addition Pump" 90-04,. Revision _0, " Missed Hourly Fire Watch Due to Failed Door'
Lock" Unit 3 ' 89-08, Revision 2, " Unusual Event Due to LPSI Pump Leakage" + 90-03, Revision 0, " Low Pressurizer Pressure Plant Protection-System Set Points Below Technical Specification Limit" , .. No violations or deviations were identified.
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Follow-UpofPreviouslyIdentifiedItems(92701) a.
(Closed) Followup Item (206/89-27-01), " Review Of OA Activitiesconcern1noEngineeringandT(nnicalGroups" Thh item concerned the perception that the nuclear oversight staff did not appear to have a clearly defined course of action for oversight of engineering and plant technical activities.
In addition, the inspector perceived that these organizations were receiving less oversight than operations, maintenance, and others.
The inspector noted that since this concern was identified, the licensee has completed a reorganization of the Safety Engineering organization.
Of the three g oups comprising Safety Engineering; one the Quality Engineering QE) group was dedicated to performing vertical slice aud ts of engineering activities.
As of this inspection period QE has performed one vertical slice audit with a second audit in, progress.
With regard to the second audit, a , number of NCRs were issued concerning implementation of an j atmospheric dump valve (ADV) design change.
To provide , coordination of quality oversight organization functions, OE findings are reviewed by site QA for incorporation into audit and surveillance plans.
, The inspector considered that the licensee's activities to address this issue were appropriate.
The inspector will continue to monitortheQE'seffortsinconjunctionwithreviewsofthesafety
assessment / quality verification functional area.
This item is closed, b, (Closed) Followup Item (206/90-17-01), "New Fuel Storage , Facility (Unit 1)" On April 5, 1990. M M a routine walkdown of the vital areas of Unit 1,l storage building,f the insp...<. f>und three excore fission chambers in the r r new fue At the tiine, tk scaffolding was being erected and the inspector wnd the door ogn.
The fission i l chambers were in a badly worn wooden box and the room was generally very dirty.
The new fuel sStage room in the fuel handling building is part of the fuel handling and storage system as defined in Chapter 9 of the Updated Final Safety Analysis Report.
Once fuel is installed in the new fuel storage racks, it is considered to be in a 3ermanent plant-location designed specifically for the purpose of 1olding ' l nuclear fuel.
Regulatory Guide 1.38 and ANSI Standard N45.2.2, For " Packaging, Shipping 3 Receiving, Storage, And Handling Of Items Nuclear c wer Plants refer to storing items in areas, "...other than its permanent location in the plant." Therefore the licensee determinedthattheyarenotcommittedtoANSIN45.2.Efor permanent storage of special nuclear materials.
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In the last resident inspection report,ing discove, red in tie 206/90-17 the ins >ector identified the generally poor housekeep ' facility including the fact that the filter to the intake fan appeared to be missing.
Section 9.1.1 of the Updated Final Safety
Analysis Report s)ecifies that new fuel is stored in the new fuel
storage room in tie fuel handling building which provides a controlled environment for the receipt inspection, and storage of the fuel assemblies.
In addition, Pip ng and Instrument Diagram - Miscellaneous Ventilation Systems Shee 2, Drawing Number 5178621-7, indicates that there is a filter at the suction of the supply fan, MVS A-27, for the new fuel storage area.
In response to the inspector's concern, the licensee wrote NCR 90040048 to address the issue.
The new fuel storage building was i quickly cleaned up and as part of the disposition to the NCR, the filter was replaced.
The ventilation system was also added to the repetitive maintenance order (RMO) program and the building was included in operator tours to ensure continued operability.
The Topical Quality Assurance Menual (TQAM) will also be revised to clarify the issue of new fuel storace requirements.
The ins also noted an apparent security de.ficiency in report 90-17. pector Since then, the inspector learned that the new fuel storage building is in a vital area and has the added feature of an alarm in the control room when the door is opened.
Also, the inspector noted that the Shift Superintendent will not allow less than two people i to be in the facility at any given time unless the person is a licensed operator.
The lack of a filter for fan MVS-A-27 in
accordance with Piping and Instrument Diagram 5178621-7 is consideredanapparentdeviation(206/90-22-03).
The inspector considered that the licensee was aggressive to implement thorough corrective actions for this dcficiency.
As a result, this item is closed, c.
(Closed) Followup Item (206/88-12-01), " Containment Spray i i Flow Low Annunciator Circuit, Design Cnarge 79-09" Designchanke79-09was$ow"caredtomakethewiringofthe re "Containmen Spray Low F annunciator circuit match the wiring i and elementary drawings.
NCR 501 P-6052 was written to identify 'l L and disposition the correction.
This was scheduled to be completed
in the upcoming Cycle XI outage, but the inspector noted that it has been removed from the outage schedule due to higher priorities.
l Inspection report 206/90-13 had closed this item based on work to be completed during the Cycle XI outage.
' " The inspector considered that it is still appropriate to leave this item closed due to the fact that the alarm is informational only and has no safety significance.
Therefore, this item is closed.
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(Closed) Followup Item (206/90-07-01), " Inadvertent Valve Manipulation Durino Maintenance" During maintenance to correct a leak on the East feedwater (FW) ! ' pump discharge flange, a contractor inadvertently changed the position of a FW pump seal water throttling valve by accidentally stepping on it.
The inspector observed the manipulation and contacted the -introl room.
Operations personnel responded and . j' adjustedthevolvebacktoitscorrectpositionandmaintained - m; surveillance of the contractor activities until work was completed.
L For corrective action to the inspector's concern the maintenance L department initiated Maintenance Incident Investigation Report i (MIIR) 90-009 as a result.
The MIIR identified inattention to detail as the root cause of the incident and reviews were conducted i with the contractor, his supervisors, and with Unit 1 SCE mechanical craft and their supervisors.
The inspector considered these corrective actions appropriate.
This item is closed.
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(Closed? Followup Item (206/87-29-06), "Desion Review Of Safety njection Interlocks" , Theinspectorperformedadesignreviewofsafetyinjection . ' interlocks associated with the main feedwater pumps.
The review revealed that the interlock associated with the safety injection ' d'schar e valves, HV851A/B and the condensate suction valves HV854A/:, was not designed,or installed as described in the F$AR.
Specifically, the FSAR stated that the HV851 interlocks were designed to prevent inadvertent opening of the HV854 valves unless the corresponding HV851 valve is completely closed. The installed t configuration meets the interlock when the HV851 is completely o>en.
The inspector continued this review by requesting a copy of tie 10 CFR 50.59 design review in order to assess this apparent , discrepancy.
, ' The evaluation, M39419, Revision 2 " Event Specific Single failure ResponseEvaluationSanOnoireNuclearGeneratingStationUnit1," was reviewed by the inspector.
It shows that the interlock "as-built" does not affect the probability or consequences of relevant design basis events.
During normal cperations, the condensatevalvesareopen,(thesafetyinjectionvalvesareclosed, and reactor coolant system RCS)pressureisabovethefeedwater , system capability.
The multiple failures required to cause an ~ inadvertentunboratedinjectionofwaterareoutsidetheSONGS1 design basis.
Undersafetyinjectioninitiationconditions the condensate valves receive a closed signal and the safety in ection - L valves receive an open signal.
Since the sequencer for eat train j gives both the pump trip and valve actuation signals on a safety ' injectionactuation,andtha limit switches for the interlocks o>erate from valve stem position rather than from the actuator, t1ere is no single failure which can leave the condensate or heater drain pu ps running and result in an open path of unborated water to the RCS.
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< Based on the inspector's review of the single response analysis, this item is closed.
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IClosed)FollowupItem(206/89-07-08)," Overpressure ) Protection System" l The item identified that the maximum relief setpoint of 500 psig
fortheoverpressureprotectionsystem(OMS)wasnonconservative.
In particular, the setpoint was based on the reactor vessel heatu and cooldown curves as calculated by design calculation D0-1562, p dated February 10, 1984.
However, the heatup and cooldown curves u
were revised on May 21 1989,thelIcensee;but,ittedLicenseeEventReport(LER) 1986 00-1562 was not revised.
On , October 6 subm
89-022 whIch was subsequently revised on November 1,1989, to report that SCE's administrative controls were inadequate.
As corrective action, the licensee committed to impinent appropriate.
, administrative controls and to conduct in-depth Achnical reviews.
' of the OMS.
On May 25, 1990, an engineering review determined that administrative controls beyond those described in LER 89-022 were I necessary to ensure adequate relieving capacity with the current
OMS design.
Specifically, at low reactor coolant system pressures, i the charging pump flowrate could exceed the flow capacity of the OMS assumed by event analyses for the most limiting failure.
The , ' circumstances of this will be reported in a supplement to LER . 89-022.
This supplemental LER will be reviewed by the inspector as part of the routine inspection program. Therefore, this item is , closed.
g.
(Closed) Followup Item (206/89-07-05), " Steam Generator Tube Discrepancies" In February 1988, eddy current testing of S/G tubes identified that , five upper rolled expansion 'oints and one lower expansion joint i didnotexistonseparatetu6esleeves.
Foi corrective action the licensee documented this condition in NCR 501-P'6419.
Thelicenseepluggedallleakingtubesleevesand .' concluded that continued operation without the rolled expansion joints was acceptable until the cycle X outage.
This co'ndition was also ?eported in Licensee Event Report (LER) 206/88-18, " Limitations - With Steam Generator Eddy Current Test Method." NRC reviews of this issuewereperformedinconjunctionwithSupplement1tothatLER ' and this item was closed in inspection report 206/89-27.
Therefore, this item is closed, h.
(Closed) Followup Item (206/86-23-01), " Resolution Of Control Room Habitabilloy" This item identified the need for followup (of the licensee'sTSC) habitab control room and technical support center . - - - -. - - . m .-- .m m . . - . .. -.. .. - , , -.., '
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l . , l III.D.3.4) in accordance with Section 1.1.1 of Supplement I to I NUREG-0737.
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The inspector reviewed this item and noted that the issue of , control room habitability is listed as an item in the January 2, 1990 Unit 1 Order Confirming License for full-term operating license (FTOL) open items.
The Order confirms that system upgrades f for the control room withTSCincluded) Unit 1.will be implemented by the ! licensee during the ele 12 outage for This issue was ! reviewed and approve by the NRC for Units 2 and 3 during the ' licensing process.
As a result, the inspector considered that this ! open item can be closed and the issue can be tracked for NRC review ! using the FTOL list.
Therefore, this item is closed.
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(Closed) Followup Item (361/88-25 04), " Improper Reset of Diesel Generator Alarms" During a previous inspection, the inspector noted that a " Fuel ! Transfer Strainer High Pressure" alarm had annunciated on the local panel and had not been reset after the surveillance was completed.
- The inspector questioned if the condition would disable inputs from i
other alarms going to the " Fuel Oil Day Tank Trouble" common alarm in the control room.
The licensee reviewed pertinent documentation and found that the above control room alarm did not have reflash capability to
indicate other alarm conditions.
However, a se>arate alarm, " Mechanical Trouble," in the control room does 1 ave full reflash - capability for the same alarm conditions.
Thus, the o)erator would be alerted to any abnormal condition and creuld go to tie local panel to check the cause of the alarm.
This item is closed.
, f,. (Closed) Followup Item (361/89-14-03), "I1 adequate Safety . ! Evaluation in Support C Continued Piant )peration With Known Design Deficiencies" During a previous inspection, the inspector reviewed an NCR ' disposition and safety evaluation associated with the Unit 2/3 AFW check valve noise NCR GR-0071.
The inspector considered that it did not adequatel justifythecontinuedoperationofUnit3until the Cycle 5 refue ing outage when the check valves were to be inspected.
In response to the concern, the licensee revised the NCR to include
an interim root cause assessment detailing the rate of check valve ~ wear, Based on this calculation the licensee concluded the check valves could perform their function for more than a refueling cycle.
The inspector reviewed the revision and found it to be satisfactory.
Therefore, this item is closed.
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(Closed) Followup Item (206/89-18-04), " Decay Heat Removal With Reduced Reactor Coolant Inventory" The licensee's implementation of the short term decay heat removal (DHR) requirements was reviewed by the NRC and closed in previous 206/pections for all three Units, as documented in Inspection Report ins 88-28 and 361/89-14.
However, the licensee's long term enhancements remain open and will be tracked by FTOL actions for Unit I during the Cycle XI refueling outage.
For Unit 2 and 3 theserequirementsremainopenandwillbeimplementedduring{he Cycle VI refueling outage scheduled for 1991 and 1992.
The licensee's implementation of upgrades will be tracked and reviewed , by the inspector in accordance with NRC Temporary Instruction (TI)-103.
Therefore, this item is closed.
11.
Follow-up On Violations a.
(Closed) Violation (206/90-08-02), " Failure To Perform The i Annual Post Acciden". Sampling System Drill" ' This item concerned the licensee's failure to perform the annual post accident samp" ling system (PASS) drill as required by procedure 50123-VIII-0.200, Emergency Plan Drills."
For actions to correct this deficiency, the licensee conducted the PASSdrillonMay8,1990atwhichtimealldrillobjectiveswere met.
In addition, the licensee also issued a revision to procedure 50123-VII-0.200 to more clearly reflect the requirements for a biennial PASS cask drill and an annual PASS analysis and sampling of in plant liquids drill.
The inspector considered that the licensee's actions were appropriate.
Therefore, this item is closed, b.
(Closed) Violation (206/89-31-01), " Failure To Take Correct've Action To Preclude Failure Of Solenoid Valves" This item identified that the licensee failed to take the proper
corrective actions to preclude failure of the solenoid operator for l normal charging isolation valve, CV-304.
This was despite the fact that there and been failures of similar solenoid valves previously.
For corrective action on this matter, the licensee determined that lubricant (Dow Corning 550) on the valves was responsible for the valve failures.
Consequently, the solenoid for CV-304 wts replaced with a similar valve in which the lubricant had been removed and these valves were ) laced on the Control of Problem Eguipment (COPE) list to identify tie potential deficiency.
In addition, the l licensee revised procedure 50123-XV-5, " Nonconforming Material, Parts or Components," to incorporate the requirement for personnel to assess the need to input information concerning nonconformances ) into the COPE program.
Procedure 501-I-8.171, " Valves - ASCO Models 206-380, 206-381 Solenoid Valve Overhaul, " was also revised to include additional cautions to ensure that Dow Corning 550 l .- n -- --- . - - - - . - - -. - - -.
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.* l o lubricant (used on gasket parts) does not come in contact with valve body components.
The inspector reviewed the licensee's actions and considered them to be appropriate.
This item is closed, c.
(Closed) Violation (206/89-31 02), " Failure To Comply With Technical Specification 3.0.3 Upon Failure Of Valve cv-304" This item concerned the licensee's failure to comply with Technical Specification (TS)3.0.3whenthesolenoidoperatorfornormal ciarging isolation valve CV-304 failed. When this occurred, it rendered CV-304 inoperable, causing the potential for a diversion of hot leg recirculation flow. The licensee's failure to comply with the TS was due to their misconception that the hot leg recirculation flow path had not been included in the TS when that system was added.
For corrective action on this matter the licensee is preparing a proposedchange(PCN)-151totheT$torevisetheactionsand surveillancesforthesafetyinjection, recirculation,and containment spray systems. This amenciment is expected to be submitted to the NRC before restart from the midcycle outage scheduled to begin June 30, 1990.
In addition, as discussed in a letter from H. 3. Ray to Region V, dated November 16, 1990. SCE indicated that additional emphasis toward conservatism in TS interpretation and implementation would be made.
The inspector reviewed the licensee's action and considered them to be adequate.
Therefore, this item is closed.
12.
Meetings with SCE Manager of Nuclear Regulatory Affairs in the Region V Offices On June 5, 1990 members of the Region V staff met with the licensee's Manager of Nuclear Regulatory Affairs in the Region V offices to discuss current regulatory issues.
Thediscussionscenteredon(1)thedesign bases heat load on the Unit I spent fuel pit, and (2) the potential for a safety indection delay in Unit 1 due to SI valve realignment deficiencies.
The Unit I thermal shield monitors and the licensee's analysis of the Unit 3 feedwater ring degradation were also discussed briefly.
the licensee discovered that the maximum Spent Fuel On April Pit (SPF) heat loa 24,1990,ds described in the Updated Final Safety Analysis Report (UFSAR) were in error. This was documented in Licensee Event Report 90-010, dated May 24, 1990.
The licensee also submitted Proposed Change No. 224 (Amendment Application No. 182), dated May 16, 1990, which would provide for appropriate changes and corrections to the UFSAR.
The changes requested would allow the licensee to aermanently install the saare cooling pump and would document higher SFP temperatures in the U;SAR.
The licensee representative provided a " hand-out" describing the . '1 0 ' ... .. .....,........,......s ,.... .... _ _ _....... . _m__._._ ___.____ _ _. _ _ _ _ _ _ _ _ _ _. _ _ _ _., _ _. _ _ _ _ - _ _. - '
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situation and requested resolution.
This " hand out" is attached to this report.
Thelicenseediscoveredthatanunexpectedsafetyinjectiontimedelay existed in Unit 1 on March 21, 1989.
This time delay was primarily due to Si valve reclignments taking longer than design basis calculations.
The problem has been documented by the licensee In Licensee Event Report Revision 0 89-011, Subsequent 1, dated April 24 1989 and Revision 1 dated December 5, 1989.
on March 14 1990 the NRC conducted a management meetingwitntheikc,enseewhichIncludedthevalverealignmentproblems.
The general conclusion at that time was that the significance of the ' problem ap) eared less than the initial review of the Licensee Event Report.
Tie meeting is documented in inspection report 50-206/90-18, . dated April 13, 1990.
During the current follow-up meeting the licensee representative reviewed the previous analysis and described the ' assumptions and conclusions of the licensee's current analyses.
The licensee has implemented design modifications to the main feed pump mini-flow valves to reduce the time delay for safety igjection to the reactor.
The current analyses demonstrate that for a worse case" analysis,includingthesafetyinjectionsystem,calculatedcore temperatures would be less than the maximum allowed.
l The licensee representative briefly described the thermal neutron noise i versus frequency monitoring which is used to monitor the Unit 1 thermal - shield. The Region V staff confirmed that this data had been sent to headquarters and that system readin s had returned to normal following
the high reading on May 23 1990.
.egarding the Unit 3 feed-water
distribution ring, the lice,nsee re)resentative discussed the licensee's initial findings, which indicate t1at the failure in the schedule 40 section of the feed ring was due to cyclic thermal faticue followed by cracking and erosion of the metal due to high water flow velocity.
The ' "T" vent damage appears to be due to high velocity erosion alone.
' 13.
ExitMeetino(30703) On June 5, 1990 an exit meeting was conducted with the licensee representatives identified in ?aragraph 1.
The inspectors summarized the inspection scope and findings as described in the Results section of this report.
I Duringthatmeetin$.,3byJune the licensee committed to u toproviNrademaintenancefor independen l procedure 50123-I-22, 1990, ' operability reviews.
See Paragraph 5.b for further details.
The licensee acknowledged the inspection findings and noted that appropriate corrective actions wot.ie be implemented where warranted.
The licensee did not identify as proprietary any of the information provided ! to or reviewed by the inspectors during this inspection.
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.', ' ' . . SdNGS 1 SPENT PUEL POOL COOLING i DECAY HEAT LOAD PROBLEM , SPENT FUEL POOL (SFP) HEAT LOADS DURING , UPCOMING FULL CORE OFF-LOAD MAY BE HIGHER THAN SPECIFIED IN THE SONGS 1 UPDATED FINAL , SAFETY ANALYSIS REPORT.
DISCOVERY POTENTIAL FOR HIGH HEAT LOAD DISCOVERED DURING TECHNICAL REVIEW OF ADMINISTRATIVE CONTROLS NECESSARY FOR FULL CORE OFF-LOAD.
RESOLUTION MODIFY SFP COOLING SYSTEM TO SIGNIFICANTLY = REDUCE TIME REQUIRED TO PLACE SPARE COOLING PUMP IN-SERVICE.
m AUGMENTED SFP COOLING SYSTEM SURVEILLANCE AT ELEVATED TEMPERATURES.
AMENDMENT APPLICATION SUBM11TED FOR NRC m APPROVAL THAT JUSTIFIES HIGHER SFP HEAT LOA. - -..-. -. - -.- -... ~ _ - - - - - - - - - - - - A.
. . . , n' i . , i , BACKGROUND e SFP HEAT LOADS SPECIFIED IN UFSAR l UNDERESTIMATED DUE TO CALCULATIONAL l ERROR.
! e UFSAR SUPPORTING CALCULATION PERFORMED ERRONEOUSLY IN 1982 AS PART OF SEP.
, a CALCULATION USED Mw ELECTRICAL VS.
Mw THERMAL TO COMPUTE DECAY HEAT LOADS.
, I e CONCLUDED LONG TIME AVAILABLE TO CONNECT / OPERATE PARTIALLY INSTALLED SPARE , COOLING PUMP.
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- -- . .... ....; . ! - i ~ - .i l SFP COOLING SYS t:M DECAY HEAT REMOVAL CAPABILITY l ! COMPARED TO CURRENT UFSAR MAX. ABNORMAL HEAT LOAD l ! ! ! TIME TO BOIL j
MAX. SFP HEAT MAX. SFP UPON LOSS OF-j PLANT CONDITION LOAD (MBtu/h) TEMP (*F) COOUNG (HR) l i ! CURRENT UFSAR MAX.
6.8 116
! l ABNORMAL HEAT LOAD [ ! , i ! l FUEL CYCLE 10 0.5 <90 No Boiling ! i (CURRENT SFP (By Observation) i FUEL INVENTORY) ! CYCLE 11 OUTAGE 8.6 138
I l FUEL CYCLE 11 14.7 179
l (WITH POSTULATED ! EMERGENCY DEFUELING) l
CORRECTED UFSAR MAX.
17.0 193
l ABNORMAL HEAT LOAD - - - - - .. . ,. -. -... ..- -- .- --..i
... _..._.._... _ _ _ _. _ _. _ _ _. _ _.. _. _ _ _ _ _ _ _ _. _ _.. . ... . i ' MODIFICATION OF SFP COOLING SYSTEM , L CYCL 811
, a MECHANICALLY CONNECT SPARE COOLING PUMP TO " SYSTEM.
, l e SAFETY-RELATED PIPING AND VALVES.
L INTERIM ELECTRICAL POWER FROM NON-SAFETY-e RELATED SOURCE.
! J = LOCAL MANUAL CONTROL.
l .. l CYCLE 12 AND BEYOND ! UPGRADE SPARE COOLING PUMP ELECTRICAL POWER i e TO SAFETY-RELATED.
e UPGRADE SPARE PUMP CONTROLS TO SAME REQUIREMENTS AS EXISTING SYSTEM.
. ' ;. RESULT: SPARE PUMP OPERABLE WITHIN APPROXIMATELY 30 MINUTES OF PRIMARY PUMP FAILURE . - ~, .-..,....-m_.-,..... .-. --.. -- _.....,.-- _.,.- _ .~.-+--.._._.~,..m__.. .. -. -. _, _ _ _,. - _... .
a.,4,e
---enn a.am,,.m-r L.mema-a ..A-e a &-, ,e-.w>n -. -_,,e w4 -
A--s e
.,.,,*
4. * ,- .? O . -
{j '
- a }r.
O '
X l , ' l
i .
! l r at ^ J k n.
l I .' . . . g U 'a $ gd i4 _ . m . ' r ! L ): X - - . ., .. _ - .. - - -.. _,. - _ _ -, _ - _ _ _ . -. .... --
... _ .._ _ _ _ _ _ . ._ _ _.. _. _ _.___. _. _ _ _ _ _ _ _. _ l - .. .. ,
, , . l CONCLUSIONS i SHORTENED TIMES TO POOL BOILING WITH HIGH s- . HEAT LOAD AND LOSS OF PUMP.
' MODS REDUCE TIME TO PLACE SPARE PUMP s
IN-SERVICE.
l l m - ADEQUATE MARGIN TO POOL BOILING MAINTAINED.
i e NRC APPROVAL OF AMENDMENT APPLICATION NEEDED PRIOR TO OFF-LOAD.
- l . l ' o < SFPCOOLSN6 - _. - - -. _ _ _ - ...... _... - . . . - . . . .. .. . .. -.. . .. . . .. -
.. - - - _ _ - - -. - -. _ - -. - - - - - - - - - - - , , .. , ,
. . l POTENTIAL FOR SAFETY INJECTION DELAY .. PROBLEM TIME DELAYS IN THE REQUIRED REALIGNMENT OF VALVES FOR SAFET/ INJECTION OPERATION COULD
. HAVE RESULTED IN A PEAK CLAD TEMPERATURE (PCT) > 2300 F IF A DESIGN BASIS LARGE BREAK LOCA HAD
OCCURRED PRIOR TO CYCLE 10.
q ,
CAUSE i MULTIPLE PREVIOUSLY UNACCOUNTED FOR SI TIME
- '
DELAYS --> TIME FOR FULL Si FLOW OF 47 SECONDS ' VS. 21 SECONDS IN SAFETY ANALYSIS.
RESOLUTION
DESIGN MODS IMPLEMENTED TO REDUCE / ELIMINATE l TIME DELAYS.
? SAFETY SIGNIFICANCE DESIGN BASIS CASE --> PCT BELOW 2300 F.
, , l ... -.. - .. .- _.
.. _. _ ., _ . - _ _, _ -.,. _. _ _
_. . _.
_ _. _. _ _ _ _ _ _ _ _ _ _ _ _ _ _.. _._.. _. _ _ ' a :" . .' BACKGROUND ' ' . ? m IN-SERVICE TESTING-PROGRAM DETERMINED LONGER VALVE STROKE TIME THAN ASSUMED IN
LOCA SAFETY ANALYSIS WOULD DIVERT FLOW i FROM CORE.
t a WIRING ERROR.ON MAIN FEED PUMP MINIFLOW i VALVE CONTRIBUTED ADDITIONAL DELAY FOR.SI i OPERATION.
, ! LER 89-011 (APRIL,1989) INCLUDED CONSERVATIVE e - CALCULATIONS OF RESULTING PCT.
> l WORST CASE WITH LOSS OF ONE SAFETY - INJECTION TRAIN --> INTERIM ACCEPTANCE CRITERIA COULD BE EXCEEDED FOR
i 28 SECONDS WITH PCT OF 2358 F.
WITH BOTH Si TRAINS AVAILABLE --> PCT WELL -
BELOW 2300 F.
( = VALVE AND WIRING MODIFICATIONS COMPLETED PRIOR TO CYCLE 10 OPERATION TO ELIMINATE POTENTIAL FOR THESE SI TIME DELAYS.
l u ' _ . .... .. . - - - - . .. . . - - -. ... .... _
_. _. _. - _ _ _ _ - - -. _ _ _ _ _. - _ _ _. _ _ _ _ _ _ - - _ _, I s, ~o., i l o ~.
. SAFETY SIGNIFICANCE , ,.
I e PRESENT EVALUATION --> PCT < 2300 F i SISLOP 2230 F - , ' SIS 2262 F - l i e CALCULATION BASED ON: - LARGE BREAK LOCA PLUS SINGLE FAILURE - ? L PARTIAL FLOW DURING VALVE OPENING -
ESTIMATED ACTUAL SYSTEM FLOWS --> 62 F - MARGIN IN PCT CONCLUSION
MINIMAL SAFETY SIGNIFICANCE . POTNLSI.SN2 . - -... - - .. . _.. -. _ __ .... - - . ._ }}