IR 05000387/1990015
| ML17157A360 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/28/1990 |
| From: | Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17157A359 | List: |
| References | |
| 50-387-90-15, 50-388-90-15, NUDOCS 9010150262 | |
| Download: ML17157A360 (31) | |
Text
U. S.
NUCLEAR REGULATORY COMMISS'ION
REGION I
Report Nos.
50-387/90-15; 50-388/90-15 License Nos.
Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Facility Name:
Susquehanna Steam Electric Station Inspection At: Salem Township, Pennsylvania Inspection Conducted:
July 22, 1990 September 1,
1990
~ Inspectors:
Approved By:
G.
S. Barber, Senior Resident Inspector, SSES J.
R.
S acr gR ident Inspe r,
SS I
P.
wetland, Chic Reactor Projects Section No.
2A Da e
Ins ection Summar Areas Ins ected:
Routine inspections were conducted in the following areas:
operations, radiological controls, maintenance/surveillance testing, emergency preparedness, security, engineering/technical support, safety assessment/
quality verification, and Licensee Event Reports, and Significant Operating Occurrence Reports.
Results:
During this inspection period, the inspectors found that the licen-see s activities were directed toward nuclear and radiation safety.
One unre-solved item was identified concerning your failure to report a condition that was apparently outside the plant's design. basis.
A single failure of the Emergency switchgear cooling coil inlet or outlet valves could potentially cause an overheating of all 4 trains of emergency switchgear.
An Executive Summary is included and provides an overview of specific inspection findings.
9010150262 90092s-",
EXECUTIVE SUMMARY Sus uehanna Ins ection Re orts 50-387/90-15 50-388/90-15 July 22, 1990 September 1;
1990
~0 erations (30702, 30703, 71707)
The "E" DG fuel oil (FO) storage tank was sampled and insolubles were found to exceed the limit of 2 mg/100 ml.
An NRC waiver of compliance was issued to allow the licensee time to drain, clean, inspect, and refill that FO-storage tank.
Action to prevent recurrence included more representative on-site and tanker truck sampling.
~
e
- ~
During a power ascension in Unit 2 following a rod sequence exchange, the powerplex computer core monitor thermal limit checking program was out of service.
Evaluation by the licensee determined that no core thermal limits were exceeded during the power ascension.
Several procedural changes were implemented to improve controls over powerplex status changes.
An airlock damper isolation occurred i n Unit 1 due to a shorted auxiliary relay-coil.
The shorted coil also led to a blown fuse which disabled the Division I secondary containment isolation function.
This was not identified until approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> following the initiating event.
The licensee has committed to provide a supplement to LER to address the reason for not detecting the loss of Division I when the fuse blew.
A slow start of the "A" Diesel Generator (DG) occurred during a test being performed to comply with the Technical Specification (TS).
A 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> restoration and 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to hot shutdown TS action requirement was entered.
The "A" DG was repaired and successfully retested in less than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> allowing the licensee to exit the action requirement.
The inspector determined that the appropriate reportabi lity requirements were met and the response to the event was acceptable.
Radiolo ical Controls (71707)
t Individual workers and Health Physics personnel implemented radiological protection program requirements.
Periodic inspector observation noted no inadequacies in the licensee's implementation of the radiological protection program.
Maintenance/Surveillance (61726, 62703)
The licensee exercised good control of maintenance and surveillance activities.
No,scrams or safety system actuations were attributable to maintenance or surveillance activitie Executive Summary (Continued)
Emer enc Pre aredness (82301)
A Health Physics Emergency Plan Drill was held on August 21.
A portion of the drill was observed by the inspector who noted frequent thorough Emergency Director briefings and good use of the position specific procedures.
~Securit (71707)
Routine observation of protected area access and egress control showed good control by the licensee.
En ineerin /Technical Su ort (71707, 92720, 93702)
A concern was reviewed involving a single failure event that coul,d lead to a
loss of all Emergency Switchgear, room cooling. during accident conditions.
Another concern was identified which involved the qualification of 'seals used in two SGTS and two DX unit ITT NH90 damper actuators.
A special inspection was conducted to review the latter concern and to examine the licensee's process in evaluating and correcting this situation.
Safet Assessment/Assurance of ualit (40500, 60705, 90712, 92700, 92701)
The licensee made a report per
CFR 50.9 to Region I on August 29 to inform the NRC that the effect of racked out breakers on seismic qualification was not previously evaluated and that the effect on these load centers is indeterminate.
An evaluation is being performed, and in the interim, all spare circuit breakers have been racked in.
LER 50-388/90-008-00 was submitted concerning the blocking open of a secondary containment boundary door.
21 times over a period of 21 days.
The licensee is performing an analysis to determine the impact of increased SGTS drawdown times on predicted offsite radiation dose effects.
A supplement to the LER will provide the results of this analysis.
An outage preparation meeting was conducted for the Unit 1 refueling outage which commenced September 12.
The inspector noted good preparation, planning, and coordination between groups involved in outage activitie SUMMARY OF OPERATIONS 1.1 Ins ection Activities DETAILS The purpose of this inspection was to assess licensee activities at Susquehanna Steam Electric Station (SSES)
as they related to reactor safety and worker radiation protection.
Within each inspection area, the inspectors documented the specific purpose of the area under review, the scope of inspection activities and findings, along with appropriate conclusions.
This assessment is based on actual observation of licensee activities, interviews with licensee personnel, measurement of radiation levels, independent calculation, and selective review of applicable documents.
Abbreviations are used throughout the text.
Attachment
provides a listing of these abbreviations.
1.2 Sus uehanna Uni't 1 Summar Unit 1 entered the inspection period at approximately 60 percent of full power due to isolating one of the main condenser waterboxes in an attempt to locate a tube leak.
The effort was unsuccessful due to the small size of the leak, and full power was restored on July 22.
Another power reduction to 60 percent was performed on August 2 through August
to plug a leaking tube in the "D" main condenser waterbox.
Full power was restored on August 7 and maintained throughout the remainder of the period.
Scheduled power reductions were also conducted during the period for control rod pattern adjustments, surveillance testing, and maintenance.
A failed auxiliary relay resulted in an airlock damper isolation on August 6.
See Section 2.2.3 for details'.3 Sus uehanna Unit 2 Summar Unit 2 operated at or near full power for all of the inspection period.
Scheduled power reductions were conducted during the period for control rod pattern adjustments, surveillance testing, and maintenance.
No ESF actuations or scrams occurred during the period.
2.
OPERATIONS 2.1 Ins ecti on Activities
The inspectors verified that the facility was operated safely and in conformance with regulatory requirements.
Pennsylvania Power and Light (PPEL)
Company management'ontrol was 'evaluated by direct observation of activities, tours of the facility, interviews and discussions with personnel, independent verification of safety system status and Limiting Conditions for Operation, and review of facility records.
These inspection activities were conducted in accordance" with NRC inspection procedure 71707.
The inspectors performed normal and backshift inspections including deep backshift inspections on July 30 from 2:00 a.m. to 6:00 a.m.;
August
from 7:00 a.m. to 12:30 p.m.; August 6 from 3:00 a.m. to':00 a.m.;
and, August 24 from 2:00 a.m. to 6:00 a.m..
2 '
Ins ection Findin s and Review of Events 2.2. 1
"E" Diesel.Gener'ator Waiver of Com liance The licensee drew quarterly fuel oil (FO) samples for all five DGs per TS 4.8. 1. 1.2.c on July 24.,
This TS requires quarterly testing of fuel oil for viscosity water, sediment, and insolubles.
One of the requirements is that the fuel oil storage tank contain less than
mg of insolubles per 100 mls While the tanks associated with the "A" through
"D" diesels tested within the limits, the "E" DG FO storage tank was out of specification (spec).
The results were:
DIESEL A
B C
D E
INSOLUBLES (mg/100m~1 0.6 1.3 1.1 0.8
'AgL'3 7'AA'he out-of-spec condition required that the licensee declare the "E" DG inoperable and TS LCO 3.8.1. 1 was entered at 1:00 p.m., July 25.
The sample takes 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to analyze.
This LCO required returning the
"E" DG to an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or a dual unit shutdown would be require The licensee reviewed the sequence of events necessary to return the
"E" DG to an operable status on July 26 with the inspector.
They determined that the LCO time limit would most likely be exceeded because of the time necessary to drain,. clean, inspect and refill the tank with sampled, acceptable FO.
A regi'onal temporary waiver of compliance was requested and granted on July 27 which permitted a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension of the action statement for A.C.
Power Source Operability.
This extension granted relief from the requirements of Technical Specification 3.8. 1. 1 and was effective from 1:00 p.m.
on July 28 to 1:00 p.m.
on July 29.
The "E" DG day and storage tanks were emptied, cleaned, inspected, refilled, and the L.C.O. cleared on the "E" DG at 11:30 p.m. July 28.
The licensee attributed the loss of stability of the diesel fuel oil to aging and the presence of copper.
Additionally, since the "E" DG storage tank is used as the receiving tank for all new fuel oil, deposits of water and sediment are more prone to collect in it than in any of the other tanks.
Corrective actions to prevent recurrence included enhanced on-site sampling methods of storage tanks and incoming tanker trucks.
Laboratory sampling requirements will also be revised to require both a running sample and a bottom sample from tankers prior to shipment to the site.
Following discussions with the licensee concerning this event the inspector concluded that the licensee's actions in response to this event were appropriate.
The scope of the root cause and corrective actions documented in the LER will be reviewed in future inspections.
Loss of Core Thermal Limit Monitorin Ourin Power Ascension - Unit 2 On July 28, the licensee determined that the Powerplex Computer Core Monitor Thermal Limit checking program was out-of-service for approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> during a power 'ascension following a control rod sequence exchange.
The rod sequence exchange was completed at 4:22 a.m.
on July 28, and operators increased power with recirculation flow at 10 megawatts electrical per'our.
At 10:38 p.m.,
a Powerplex Core Thermal Limit Alarm annunciated indicating that one fuel bundle had an Average Planer Ratio (APRAT) greater than
.975.
The TS limit is 1.000.
When operators attempted to obtain a core monitor calculation they could not.
Investigation by the licensee determined that the computed calculation was blocked.
~ The computer system was restored and a
new calculation immediately run which determined that the maximum average planer ratio (MAPRAT) (the auctioneered high value of APRAT for all fuel bundles in the core) 'was 0.999.
Power was reduced to 97 percent from 98.8 percent per reactor engineer direction by reducing recirculation flow and eight control rods were partially inserted.
MAPRAT was recalculated and found to be less than 0.990.
Power ascension was commenced and 100 percent power was reached at 5:00 a.m.
on July 2 The licensee was not able to determine why the powerplex calculation was blocked during this period, although it is routinely blocked during a rod sequence exchange by the reactor engineer in order to prevent excessive computer demand caused by control rod movements.
The reactor engineer believes he reinstated the calculation following the sequence exchange since a calculation was obtained at 4:23 a.m.
The inspector questioned the licensee about the event and whether the licensee could verify that the (APRAT) was not exceeded during the
hours in which the calculation was blocked.
The licensee stated that their nuclear fuels group was performing calculations for the period to provide that information.
A copy was later supplied to the inspector with the results of those calculations and it was noted that the maximum MAPRAT during the period that the calculation was blocked and power ascension was in progress was 0.998.
As a result, it was verified that no core thermal limits were exceeded during the period.
Additionally, a procedure change to AD-gA-138 (Control of Core Reactivity Changes at SSES)
was issued on August 28 which strengthened controls for Powerplex status changes.
The inspector considered the licensee's corrective actions appropriate and had no further questions on this event.
Failed Auxiliar Rela Causes Airlock Dam er Isolation"- Unit
At 8:15 a.m.
on August 6, the licensee identified that an auxiliary relay (XYX607553A) failed due to a shorted coil.
This short circuit caused Unit 1 Zone III airlock isolation damper (HD17534H) to fail closed.
In addition, the resultant high current from the short led to a blown logic power supply fuse (FU-5) and a consequent loss of the ability to initiate Division I secondary containment isolation.
The lo s of Division 1 of Secondary Containment Isolation was not immediately identified by the licensee.
This condition was subsequently identified at 10:00 p.m., August 6 when an NPO discovered that two Unit 2 airlock isolation dampers indicated closed with their hand switches in the open position.
Upon identification, TS LCO 3.3.2 was entered, and its action statement was complied with, since Division I was restored to an operable status by identifying the failed relay and replacing FU-5 within one hour.
During this time, the ability to initiate the redundant Division II secondary contain-ment isolation was not affected by the relay and fuse failures.
The required ENS notification for an ESF actuation was made at 11: 16 a.m.
The failed relay was replaced on August 8 and the Zone III damper restored to its normal positions
An ERT was"established to review, investigate, and analyze the event along with providing the necessary corrective actions to prevent recurrence.
The team was also tasked with identifing improvements in design and operating practices.
The team will also address PPKL's failure to initially identify that FU-5 blew as a result of a shorted relay.
The inspector discussed the event with licensee staff and reviewed available documentation on the event.
The inspector questioned whether the licensee investigation was thorough, since it took 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to discover the inoperability of Division I secondary containment isolation.
The licensee stated that they are also concerned about this and that the ERT is addressing this concern.
The results of the team's investigation will be documented in a supplement to LER 50-387/90-.817-00.
The inspector also questioned the licensee about system design conformance to applicable IEEE standards since it appears that no indication is available to alert plant operators that the secondary containment isolation function is not available on a
loss of power to the logic circuitry.
The licensee is committed to conform to IEEE 279 and Regulatory Guide 1.47, which require that indication be available for intentionally bypassing and causing the system to be inoperable, but not for component level failures which result in system wide inoperability.
The licensee stated that they believe enhancements should be made to improve their loss of power detection capability and these woul.d be evaluated in the future.
The licensee plans to provide a supplemental LER by October 31 that addresses the ERT's findings.
This LER will be reviewed in a future inspection.
2.2.4
"A" DG Slow Start On July 26, during a manual start of the "A" DG in order to comply with TS 3.8. 1. 1 action "b.2.b" due to the inoperability of the "E" DG (see Section 2.2. 1), the "A" DG failed to achieve
Hz within 10 seconds.
Actual start time was 10.87 seconds and a retest within
hour resulted in a second slow start time of 10.26 seconds.
Investigation by the licensee determined that a one inch double threaded reducer on the right bank starting air header filter was sheared off.
The reducer was replaced and the "A" DG retested with a successful start time of 8.07 seconds.
The licensee made the required 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ENS notification to the NRC as a
result of having two out of four aligned DGs inoperable due to unrelated failures.
TS 3.8. 1. 1 action "f" was entered which has a
hour restoration to 3 operable DGs, and a
12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to hot shutdown requiremen The "A" DG was successfully retested 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 40 minutes following its initial inoperability declaration and action "f" was complied with.
The original "E" DG LCO remained in effect and action was continued to restore it to an operable status within the initial 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.
The licensee determined that the reducer failure was due to a faulty casting and considered it to be an isolated incident.
The inspector reviewed documentation related to the event and discussed it with appropriate licensee personnel.
As a result, the inspector determined that appropriate reportability requirements were met since the DG was restored before a shutdown was required.
No unacceptable conditions were identified.
3.
RADIOLOGICAL CONTROLS 3. 1 Ins ection Activities PPCL's compliance with the radiological protection program was verified on a periodic basis.
These inspection activities were conducted in accordance with NRC inspection procedure 71707.
3.2 Ins ection Findin s
Observations of radiological controls during maintenance activities and plant tours indicated that workers generally obeyed postings and Radiation Work Permit requi rements.
No inadequacies were noted.
4.
MAINTENANCE/SURVEILLANCE 4. 1 Maintenance and Surveillance Ins ection Activit On a sampling basis, the inspector observed and/or revieWed selected surveillance and maintenance activities to ensure that specific programmatic elements described below were being met.
Details of this review are documented in the following sections.
4.2 Maintenance Observations The inspector observed and/or reviewed selected maintenance activities to determine that the work was conducted in accordance with approved procedures, regulatory guides, Technical Specifications, and industry codes or standards.
The following items were considered, as applicable, during this review:
Limiting Conditions for Operation were met while components or systems were removed from service; required administrative approvals were obtained prior to initiating the work; activities were
accomplished using 'approved procedures and quality control hold points were established where required; functional testing was performed prior to declaring the involved component(s)
op'erable; activities were
'ccomplished by qualified personnel; radiological controls were implemented; fire protection controls were implemented; and the equipment was verified to be properly returned to service.
These observations and/or reviews included:
Investigation of failed auxiliary relay XYX607553A to Zone III isolation dampers per WA 501201 on August 6.
Modification of the "A" DG ESW and FO piping and supports, including installation of the ESW temperature control valves per WA C03485 on August 7.
"A" DG underside and borescope inspection of pistons and cylinders per WA S04687 on August 7.
Eighteen month Radiological Calibration of the SGTS SPING Particulate, Iodine and Low Range Noble Gas Channels per WAs A02536, A02591, and A02592 on August 20.
"D" DG underside 'and borescope inspection of pistons and cylinders per WA S04689 on August 28 and 29.
4.2. 1 Im ro er HPCI Tor ue Switch Settin Unit 2 On August 1, the licensee determined that the Unit 2 HPCI 2F003 valve (steam supply outboard isolation valve) torque switch setting for closing was set at 2.5 from work performed in April of 1988 versus a
requi'red minimum setting of 2.875.
This did not affect the operability of the HPCI system due to the fact that this valve remains open for HPCI operation, however, it is required to close upon a HPCI steam line break outside containment.
The licensee evaluated operability with respect to the ability to function as part of the primary containment isolation system.
The evaluation resulted in a determination that the valve was operable per TS 3.6.3 in that it would close fully under required accident conditions.
A WA was written and the torque switch was reset to 3.0 on August l.
The inspector reviewed the SOOR, NCR and"WA concerning the improper switch setting and operability/reportabi lity determination.
No inadequacies were note.3 Surveillance Obser vations The inspector observed and/or reviewed th'e following surveillance tests to determine that the following criteria, if applicable to the specific test, were met:
the test conformed to Technical Specification requirements; administrative approvals and tagouts were obtained before initiating the surveillance; testing was accomplished by qualified personnel in accor'dance with an approved procedure; test instrumentation was calibrated; Limiting Conditions for Operations were met; test data was accurate and complete; removal and restoration of the affected components was properly accomplished; test results met Technical Specification and procedu'ral requirements; deficiencies noted were reviewed 'and appropriately resolved; and the surveillance was completed at the required frequency.
These observations and/or reviews included SI-183-312 Quarterly Calibration of Main Steam Line Pressure Channels PSL-B21-N015C, and D, performed on August 20.
SI-214-201 Monthly Functional Test of Reactor Vessel Water Level 1 Channels LIS-24221 CBD, performed on August 29.
SI-214-202 S0-152-004 4.4 Ins ect'ion Findin s
Monthly Functional Test of Drywell Pressure Channels PSH-25120 C8D, performed on August 29.
Quar terly HPCI Valve Exercising, performed on August 31.
The inspector reviewed the listed maintenance and surveillance activities.
The review noted that work was properly released before its commencement; that systems and components were properly tested before being returned to service and that surveillance and maintenance activities were conducted properly by qualified personnel.
Where questionable issues arose, the inspector verified that the licensee took the appropriate action before system/component operability was declared.
No unacceptable conditions were identified, but two activities required followup.
The details are provided belo.4.1
"E" Diesel Generator Start Failures Not Clearl Defined er Re ulator Guide 1.108 The licensee=-reviewed two slow starts of the "E" Diesel Generator with the inspector.
The licensee provided their perspective on the slow starts after a critical review of the second slow start.
A letter (PLAS-428) specifically describing the details of the two slow starts was sent to the NRC on June 14.
On April 10, the "E" Diesel Generator (DG) was started after it had been substituted for the "A" DG.
When the "E" DG was started for its post maintenance/troubleshooting run it failed to come up to rated frequency and voltage within 10 seconds as required by TS.
The actual starting time was between 11 and 12 seconds.
Various system checks were made and no root cause was found for the slow start.
A second start on the same day was made approximately 22 minutes later and was found acceptable and the diesel was returned to an operable status.
The first start (the slow start)
was classified as a non-valid test start per RG 1. 108 since it was for troubleshooting purposes.
On May 15, a
second slow start occurred on the "E" Diesel Generator after it had been substituted for the "D" DG.
The air receivers, air start, field flash, fuel oil and governor systems were investigated as potential causes for the slow start.
The licensee identified a leak from the suction flange of the fuel oil booster pump.
A pool of fuel oil was noted below the flange and a stream was noted coming from the flange.
The DG was declared inoperable and the appropriate Limiting Condition for Operation was entered.
The flange was subsequently repaired and the diesel was retested successfully.
The "E" DG was then returned to an operable status.
The system engineer examined the May 15 slow start and determined that the leaky suction flange, in all probability, also caused the April 10 slow start.
The second acceptable start for both dates was postulated to be the result of filling the fuel oil suction line as a result of the booster pump action from the first slow start.
The failure to discover the leaky flange on April 10 indicates that the licensee'
root cause investigation was not thorough enough.
The licensee should review their investigative techniques for slow diesel generator starts.
The engineer was also concerned with the RG 1. 108 classification of the April slow start.
So, he discussed it with licensee operations and compliance personnel and confirmed that P.P.5 L. had no clear established policy to differentiate post maintenance troubleshooting runs from operability runs.
Each DG start should be properly classified in advance, to establish whether a failure to come up to rated speed and voltage constitutes a. valid test failure per RG 1. 108.
The engineer's discussion confirmed that no clear procedural guidance existe The engineer and a member of the compliance group contacted the inspector to discuss the two slow starts and the licensee's lack of procedural guidance for DG starts.
The inspector stated that the intent of RG 1. 108, with respect to DG starts, is to clearly establish the purpose of each DG start.
A start for troubleshooting purposes may not be considered a valid test start, while a start to verify the correction of maintenance problems would be considered a valid test start.
Failures for valid and non-valid tests are treated differently.
A more rigorous monitoring and corrective action program is required for too many valid test start failures.
Thus, it is important that the licensee determine the purpose for each planned DG start and also to consider the root cause of any failures in relation to the purpose of the start when defining and categorizing the start failures per RG 1. 108.
The licensee committed to provide additional clarification for the proper use of RG 1. 108.
Tem orar Waiver of Com liance Forei n Material Intrusion in Diesel Generators During a boroscopic inspection of the "D" Diesel Generator (DG) on August 29, the licensee identified vertical score marks in five out of 16 cylinders.
The score marks were somewhat uniform and 360 degrees in circumference.
The licensee considered the potential causes and reviewed recent diesel generator rebuild activities.
As a part of this review, the licensee visually examined the "D" DG's turbocharger intercooler and found sandblast grit on the air side of the intercooler.
The intercoolers for three out of the five DGs had been sandblasted as a part of a maintenance activity related to the rebuilding of all five DGs.
The licensee believes that poor work practices led to the grit's introduction in the air side of the cooler.
The inspector concurred and noted that procedural control for this activity was also weak.
The licensee originally suspected this problem when they found high levels of chromium in a lube oil sample for the "D" DG.
Oil samples were taken for all five DGs.
The oil samples for the "A", "C" and "E" DG showed normal chromium levels.
Oil samples for the "8" and "D" DGs showed 0.33 and 0.40 ppm chromium, respectively.
The motion of the piston rings inside the chrome-lined cylinder liner in the presence of the sandblast grit scraped the chromium off the liner and into the crankcase oil
~
In parallel with the samples, the licensee removed cylinder heads for the "D" DG and found six cylinders with score marks.
One additional
'cylinder that was previously found acceptable by boroscope was found to have light scoring.
Sandblast grit was found in the intake manifold of the "D" DG and in the intake portion of all of the cylinder heads.
The licensee also suspected some cylinder scoring in the "B" DG because of the chromium in the'il sample.
Thus, the licensee decided to inspect the "B" DG.
The licensee entered a
TS LCO for the "8" DG at 1200, August 30 to begin the inspection.
The licensee completed their inspection and found a number of cylinder liners, pistons, and piston rings damaged.
The licensee determined that the scope of repairs would cause exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO.
In order to prevent shutting down both units while an emergency
'TS amendment was processed, the licensee requested a temporary waiver of compliance to extend the LCO to 7 days.
The NRC reviewed and approved their request since it did not compromise reactor safety."-The licensee continued their inspection and repair effort through the end of the inspection period.
The licensee'
root cause identification and corrective actions will be reviewed in future inspections.
5.
EMERGENCY PREPAREDNESS 5. 1 Ins ection Activit The licensee conducted a Health Physics. Emergency Plan Drill on August 21.
The inspector observed a portion of this drill from the licensee's technical support center (TSC).
The inspector also r'eviewed licensee event notifications and reporting requirements for events that could have required entry into the emergency plan.
5.2 Ins ection Findin s
The inspector noted that the TSC Emergency Director held frequent
'horough briefings with all lead staff members and periodically apprised the remainder of the TSC staff of plant status and direction.
Use of the new position specific procedures appeared consistently good throughout the TSC staff.
No inadequacies were identified.
No events were identified that requi red emergency plan entry.
6.
SECURITY 6.1 Ins ection Activit PP8L's implementation of the physical security program was verified on a
periodic basis, including the adequacy of staffing, entry control, alarm stations, and physical boundaries.
These inspection activities were conducted in accordance with NRC inspection procedure 7170.2 Ins ection Findin s
The inspector reviewed access and egress 'controls throughout the period.
No unacceptable conditions were noted.
7.
ENGINEERING/TECHNICAL SUPPORT 7.1 Ins ection Activit The inspector periodically reviewed engineering and technical support activities during this in'spection period.
The on-site Technical (Tech)
section, along with Nuclear Plant Engineering (NPE) in Allentown,
. provided engineering resolution for problems during the inspection period.
The Tech section generally addressed the short term resolution of problems while NPE scheduled modifications and design chang'eE, as appropriate, to provide long lasting problem correction.
The inspector verified that problem resolutions were thorough and addressed at preventing recurrences.
In addition, the inspector reviewed short term actions to ensure that the licensee's actions provided reasonable assurance that safe operation could be maintained.
7.2 Ins ection Findin s
7.2. 1 Postulated Sin le Failure Event - Loss of ESS Switch ear Room Coolin The licensee -issued an Engineering Discrepancy Report (EDR) on July 14 that documented a single failure event that could lead to a loss of all Emergency (ESS) Switchgear Room Cooling.
There are four
KV ESS busses that can be cooled by either of two trains of Emergency Switchgear Room (ESR) Cooling.
There are two sets of cooling coils in each train of ESR cooling.
One set is supplied from the Reactor Building (RB) chiller for normal cooling and another set from the Control Structure Chiller (CSC) for accident purposes.
The CSC cooling coils are equipped with one normally closed inlet and outlet valve that are normally closed and a bypass valve that is normally open.
For a
LOCA, the ESR cooler inlet and outlet valves open and the cooler bypass valve closes causing flow through the cooling coils.
For the single failure event, either ESR cooling coil inlet or outlet valve fails to open and the bypass valve opens normally which provides flow through the system but not through the cooling coils.
Thus, the ESR cooling system continues to recirculate hot air to and from each room which results in continually increasing post-LOCA room temperature.
Operator action is necessary to start the standby train of ESR cooling for the postulated single failure even The licensee is examining the ESR temperature profiles to determine the amount of time available for compensatory action.
This event was reported to NRC Region 1 per
CFR 50.9 at 10:00 a.m., July 27 during a conference call.
The licensee's use of.10 CFR 50.9 instead of 10 CFR 50.72 and 50.73 has been previously questioned by the NRC.
The licensee is continuing to assess reportability per
CFR 50.72 and 50.73.
Plant modifications are being considered to eliminate the effects of the single fai lure.
The licensee's failure to submit 50.72 and 50.73 reports for this single failure is unresolved pending final disposition of the previous inspection finding.
(UNR 50-387/90-12-03 (Common))
7.2.2 SGTS and DN Dam er Seal Environmental uglification The licensee reported a deficiency in environmental qualification (Eg)
of polyurethane seals used for safety related damper actuators.
The specific Eg deficiency concerns the qualified life of seals used in two SGTS and two DN unit ITT NH 90 series damper actuators.
The deficiency was di'scovered when one of the Eg binders for the NH 90 actuators was returned to NPE after a consultant reviewed the seal material and changed the specified material to VITON.
A special inspection was conducted to review the qualification of these seals.
The details are provided in Inspection Report 50-387/90-17.
8.
SAFETY ASSESSMENT/EQUALITY VERIFICATION 8. 1 Licensee Event Reports ( LER), Significant Operating Occurrence Report (SOORs),
and Open Item (OI) Followup (90712, 92700)
8.1.1 Licensee Event Re orts The inspector reviewed LERs submitted to the NRC office to verify that details of the event were clearly reported, including the accuracy of description of the cause and the adequacy of corrective action.
The inspector determin'ed whether further information was required from the licensee, whether generic implications were involved, and whether the event warranted onsite followup.
The following LERs were reviewed:
Unit
88-023-01 Openings Found Through Fire Rated Barriers.
This event was reviwed in NRC Inspection Report 50-387/88-2 Postulated Single Failure Could Place the Plant 'In a Condition Outside Its Design Basis.
This event was, reviewed in NRC Inspection-Report 50-387/90-12.
90-014-00 Discrepancy Between FSAR Required and Actual Diesel Fuel Inventory.
This event was reviewed in Inspection Report 50-387/90-12.
90-017-00 Secondary Containment Isolation Division I Automatic and manual Initiation Functions Lost Due to Failed Relay.
This event is reviewed in Section 2.2.3.
Unit 2 90-007-00 ESF Actuations Due to RPS EPA Breaker Spurious Trip.
This event was revi,ewed in Inspection Report 50-388/90-12.
- 90-008-00 Secondary Containment Boundary Door Blocked Open For Mork During Period From June 1 Through June 21, 1990.
8. 1. 1. 1 Onsite Followu of Licensee Event Re orts For those LERs selected for onsite followup (denoted by asterisks in Detail 8.1.1),
the inspector verified that the reporting requirements of 10 CFR 50.73 had been met, that appropriate corrective action had been. taken, that the event was adequately reviewed by the licensee, and that continued operation of the facility was conducted in accordance with Technical Specification limits.
The following findings relate to the LERs reviewed on site:
LER 50-388/90-008-00 Blocked 0 en Personnel Access Door Violates Secondar Containment Inte rit
,On June 21, the licensee determined that plant personnel had been blocking open a personnel access door between the Unit 2 reactor building 836 foot elevation and the roof hatch which provides access to the. reactor building roof.
This door is a secondary containment boundary between the reactor building ventilation Zone II and the plenum which leads to the roof hatch and the outside atmosphere.
The door had been blocked open 21 times from the period June
through June 21.
This resulted in a plant system alignment in which secondary containment leakage rates would have been greater than previously analyzed in the event of a secondary containment isolation with a single postulated damper, failure during accident condition The licensee determined the cause of this event to be personnel error in that the Equipment Release Form initiated by the work planning group did not reference affected TS 3.6.5.
Each time the door was propped open for personnel safety reasons, a security guard was stationed at the door.
The length of time the door was open is contrary to station policy and resulted in a prolonged breach of secondary containment.
A review of each of the openings showed that the longest duration was 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 46 minut'es.
TS allows 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore secondary containment integrity.
The licensee also determined that secondary containment pressure was maintained at.25 inches of vacuum water gauge with'ormal reactor building ventilation system operation throughout the times the door was propped open.
Immediate corrective actions consisted of closing the door, and issuing a memorandum to all security shift supervi sors that requi red them to contact Operations prior to any opening of the door.
Also, the door must be immediately closed and secured following personnel access, and a security officer must remain stationed where he can observe the roof access hatch during the entire work evolution.
In addition, the work group's planning section is evaluating existing guidelines and checklists. to ensure that adequate direction is provided to require consideration of HVAC zone and fire barrier breach, including applicable.TS, during performance of similar work activities.
The licensee stated that the safety impact of the propped open door would be increased SGTS drawdown times and the corresponding effects on the predicted offsite radiation dose in the event of an accident.
The significance of these differences was not previously evaluated but wi 1 1 be, to determine their effects.
The analysis results will be documented in a supplement to this LER.
The inspector was concerned with the licensee's control of secondary containment integrity, since it was breached twenty-one times by propping open the door.
Although TS 4.6.5. 1 makes a time allowance for personnel ingress and egress through doors within the secondary containment, the TS does not intend to endorse prolonged breaches.
The licensee assured the inspector that they did not intend to violate this intent of the TS.
The licensee stated that they would address prolonged breaches in the supplement to the LER.
The inspector also noted that the SOOR on this event indicated that the event was initially reported per
CFR 50.72(b)(2)(iii)(c)
which is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report for "Any event or condition that alone could have prevented the fulfillment of, the safety function of structures or systems that are needed to control the release of
radioactive material," while'he LER was reported under
CFR 50.73(a)(2)(ii) in that it was for "Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded, or that resulted in the nuclear power plant being:
(B) In a condition that was outside the design basis of the plant."
This is an apparent discrepancy in that the corresponding
CFR 50.72 reporting requirement is 50.72(b)( 1)(a)(ii) which is a one hour report.
-The licensee stated that they would address this discrepancy in the LER supplement.
Because of the inspector's concerns, the licensee stated that the supplement to the LER would be submitted by December 1.
The licensee's analysis of the effects on drawdown times and offsite radiation doses, and the additional committments made to the inspector will be reviewed following receipt of the LER supplement.
Si nificant 0 eratin Occurrence Re orts Significant Operating Occur rence Reports (SOORs)
are provided for problem identification and tracking, short and long term corrective actions, and reportability evaluations.
The licensee uses SOORs to document and bring to closure problems ident'.fied that may not warrant an LER.
The inspectors reviewed the following SOORs during the period to ascertain whether:
additional followup inspection effort or other NRC response was warranted; corrective action discussed in the licensee's report appears appropriate; generic issues are assessed; and, prompt notification was made, if required:
Unit
41 SOORs inclusive of 1-90-192 through 1-90-236 Unit 2
SOORs inclusive of 2-90-099 through 2-90-112 The following SOORs required inspector followup:
1-90-201 documented the failure of the quarterly DG fuel oil storage tank insolubles concentration to meet the acceptable limit for the "E" DG.
This item is discussed in Section 2.2. 1.
1-90-202 documented a slow start of the "A" DG.
This item is discussed in Section 2.2.4.
1-90-216 documented a failed auxiliary relay which led to an ESF actuation.
This item is discussed in Section 2. documented the propping, open of a secondary containment boundary door.'his item is discussed in Section 8. 1. 1. 1 2-90-105 documented that the Powerplex Computer Core Thermal Limit checking program was out-of-service for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> during a
power ascension.
See Section 2.2.2 for details.
2-90-107 documented an improper torque switch setting on the HPCI F003 valve.
This item is addressed in Section 4.2. 1.
8.2
CFR 50.9 Re ort The licensee contacted Region I on August 29, to report in accordance with 10 CFR 50.9, the effect of racked out circuit breaker s on Class 1E Switchgear seismic qualification.
PP&E noted that spare circuit
, breakers on-site were'routinely maintained in the racked-out p'osition.
The licensee additionally noted that breakers are also normally racked out for blocking purposes during maintenance activities.
The effect on the seismic qualification of the affected load centers is being questioned by the licensee since they believe that the test configuration was with. the b~eakers racked in.'he licensee is presently evaluating the, effect of racked-out breakers in regard to seismic qualification and is exploring possible alternatives to racking out.
All spare circuit breakers have been conservatively racked in until the analysis is complete.
8.3 Unit 1 Outa e Pre arations On August 7, the inspector attended a licensee outage preparation meeting for the Unit 1 refueling outage which commenced September 12.
The licensee presentec.
an overview of plans for outage and preoutage work which included the specific work activities, number of documents and estimated man-hours.
Additionally, the status of readiness for the outage, coordination between various groups, contractor support and access/training and, ALARA goals were discussed.
Finally, Nuclear Safety Assessment Group (NSAG) presented their assessment of the outage schedule.
From the information presented during the meeting, the inspector noted that responsibilities were assigned to key groups, interface and coordination among the various groups was addressed and integration of contractors into outage activities appeared appropriately controlled.
The NSAG evaluation addressed how the scheduled activities complied with the applicable TS requirements, and it also ensured that adequate systems were available to remove decay heat.
No unacceptable conditions were identifie.
MANAGEMENT AND'XIT MEETINGS 9. 1 Routine Resident Exit and Periodic Meetin s
The inspector discussed the findings of this inspection with station management throughout and at the conclusion of the inspection period.
Based on NRC Region I review of this report and discussions held with licensee representatives, it was determined that this report does not contain information subject to
CFR 2.790 restrictions.
9.2 Attendance at Mana ement Meetin s Conducted B
Re ion Based Ins ector s
~Dates s
'ubject July 31 Fi tne s s for Duty August
Radwaste
Transportation
~ins ection
~Re or t Ro.
90-14; 90-14 90-16'0-16
~Re ortin
~ins ector R. Albert J. Furia
Abbreviation List ATTACHMENT 1 ADS ANSI CAC CFR CREOASS DG DX ECCS EDR EP EPA ERT'SF ESW EWR FO FSAR
~ ILRT JIO LCO LER
.
LLRT LOCA LOOP NCR NDI NPE NPO NRC OI PC PCIS PMR QA RCIC RG RHR RHRSW RPS RWCU SGTS'I SO SOOR SPING TS TSC WA
.- Administrative Procedure
- Automatic Depressurization Sy'tem
- American Nuclear Standards Institute
- Containment Atmosphere Control
- Code of Federal Regulations Control Room Emergency Outside Air Supply System Diesel Generator
- Direct Expansion Emergency Core Cooling System
- Engineering Discrepancy Report
- Electrical Protection Assembly
- Event Review Team Engineered Safety Features
- Engineering Service Water
- Engineering Work Request
- Fuel Oil
- Final Safety Analysis Report
- Justifications for Interim Operation
- Limiting Condition for Operation
- Licensee Event Report
- Local Leak'ate Test
- Loss of Coolant Accident
- Non Conformance Report
- Nuclear Department Instruction
- Nuclear Plant Engineering Nuclear Plant Operator
- Nuclear Regulatory Commission
- Open Item Protective Clothing Primary Containment Isolation System
- Plant Modification Request
- Quality Assurance
-'eactor Core Isolation Cooling
- Regulatory Guide
- Residual Heat Removal Residual Heat Removal Service Water
- Reactor Protection System Reactor Water Cleanup
- Standby Gas Treatment System
- Surveillance Procedure, Instrumentation and Control
- Surveillance Procedure, Operations
- Significant Operating Occurrence Report
- SPLIT Particulate, Iodine, and Noble Gas
- Technical Specifications
, Work Authorization
0