IR 05000285/1990035
| ML20059M678 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/25/1990 |
| From: | Constable G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20059M676 | List: |
| References | |
| 50-285-90-35, NUDOCS 9010050190 | |
| Download: ML20059M678 (10) | |
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APPENDIX.
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION:IV
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NRC Inspection Report:
50-285/90-35 License: DPR-40; j
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Docket:
50-285
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Licensee: Omaha Public ' Power District (OPPD)~
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444~ South 16th Street Mall '
Omaha,-Nebraska - 68102-2247
Facility Name:
Fort Calhoun Station (FCS)
Inspection At:
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Inspection Conducted: July 29-September. 10,'1990 Inspectors:
R. Mullikin, Se or Resident Inspector i
T. Reis H t../ ctor
' Approved-
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. L. Constable, Chief, Project;Section C
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Inspection Summary
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Inspection Conducted July 29 through September 10, 1990 (Report 50-285/90-35)-
Areas Inspected:. Routine, unannounced inspection of ~onsite followup of -
events, ope ational safety verification, monthly. maintenance, surveillance, security, and radiological protection observations, and' followup of previously identified items.
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r Results: Within the areas inspected, no violations or deviations were identified.
The licensee shut down the plant on August 24, 1990, due?to a degrading seal on Reactor Coolant Pump (RCP) RC-3A. The licensee exhibited a' positive approach to safety when management made a conservative decisionito shut down.the plant before seal failure occurred. The root cause of the seal problem was not determined at the end of the inspection period.
Details of'the event are listed in paragraph 3.c.
A problem where some lubricating oil was not purchased as safety-related or important to safety is discussed in paragraph 3.a.
This issue was identified by the licensee's quality assurance (QA) organization and demonstrates the
'l effectiveness of the licensee's self-assessment capability.
The licensee
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determined that the subject oil discovered, during QA surveillance, was the
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correct oil. At the end of the inspection period, the licensee was still l
investigating whether any nonqualified lut icants were used in safety-related J
equipment.
9010050190 90o92g
{DR ADOCK 05000285
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DETAILS 1.
Persons Contacted
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- R.1 Andrews, Division Manager, Quality and Environmental; Affairs
M. Bare, System Engineer
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- J.' Chase, Manager, Nuclear Licensing and Industry Affairs e
- W. Gates, Division Manager, Nuclear _ Operations
- R. Jaworski~,. Manager, Station Engineering.
R. Johansen, Supervisor,' Maintenance Support-B. Helmandollar Procurement Engineer
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- L. Kusek, Manager,. Nuclear Safety Review Group.
- M. Lazar, Supervisor, Operations Training.
- D. Matthews, = Supervisor,' Station Licensing.
- T. Matthews, Station Licensing Engineer
- W. Orr, Manager, Quality Assurance and Quality Control
- T. Patterson,; Manager, Fort Calhoun Station
- J. Sefick, Manager, Security Services R. Short, Supervisor,.Special Services Engineering.
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- S. Willrett, Manager, Nuclear Materials and Administration
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- Denotes attendance at the monthly exit interview.
The inspectors also interviewed additional licensee personnel during the inspection period.
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Plant Status
.a The FCS operated at 100 percent power from the beginning of this inspection a
period until August 24, 1990,.when the plant began a' controlled shutdown.
The licensee ordered the shutdown due to a steadily decreasing inlet 1.
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pressure on the RCP RC-3A middle seal.
The plant entered Modes 2 (Hot Standby) and 3 (Hot -Shutdown) on August 24, 1990. Mode 4 (Cold Shutdown)~
was reached on August 26 1990. During this outage.the licensee replaced:
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the seal cartridge on RCP RC-3A,.and completed several other outage-related.
maintenance items.
The' plant achieved criticality on. September 2, 1990, and the generator was synchronized to the grid on, September 3, 1990. The plant remained at.
30 percent power on September 4-5, 1990, to perform a boric-acid soak.of-
the steam generators.
Full power was reached on September 7; 1990,'where it remained through the end of the inspection report period.
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On August 10, 1990, the licensee declared its third auxiliary feedwater -
pump operable. The new diesel driven pump is nonsafety-related and will be used to provide the primary source of feedwater to the steam generators during plant startup, hot standby, and cooldown operations'.
It will also provide a backup source of auxiliary feedwater to the steam generators.
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Onsite Fol'lowup of Events (93702)
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Lubricants Purchased as-Noncritical Quality Element (CQE)
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On August 13, 1990,-while adding lubricating oillto. turbine-driven Auxiliary Feedwater Pump FW-10, the licensee di'scovered, during a quality assurance. audit, that theLo11.had.not been purchased as
" limited CQE" (im'portant to safety) as required by the '!CQE" (safety C
related) list. The licensee further found that'most of-the lubricants o-added to rotating. equipment:at FCS were not purchased " limited-CQE."
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However, the oil added to FW-10 was determined by:the licensee to be the correct oi1~.
A ban'was placed on the addition of lubricants until a satisfactory resolutioniof -this issue was reached.
Subsequently, procurement engineering identified all lubricants in"
the warehouse that were not purchased as " limited CQE" orL"CQE."'
Those' lubricants will undergo a: material. evaluation-to determine -
whether the lubricant can be. upgraded to'either " limited CQE"~or
"CQE,".as appropriate.
If any cannot be upgraded,.then the equipment
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that the lubricant-is' presently-in will.be evaluated.to diterminei whether the lubricant needs to be replaced with qualifier oil.
This evaluation was still in progress at the~end of the inspr.mion. period.
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This problem was discovered by the licensee's QA~ orgar ization and-has been aggressively-pursued-by the onsite engineering group, b.
Environmental Qualification of Electrical Solenoid-i On August 8, 1990, the licensee' determined-that-an environmentally.
qualified (EQ) solenoid for an air-operated valve =may not have been replaced when required. Valve HCV-1108B' regulates: the supply water -
from the auxiliary feedpumps to Steam Generator "B," and'also acts
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as a containment isolation valve.
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The FCS EQ Data Sheet showed a 4.5 year replacement interval for
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this solenoid (HCV-1108B-20A). The licensee's preventive.
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maintenance program showed that the solenoid was scheduled.for.
replacement based on the above' interval.. However, the licensee could not initially determine whether the solenoid-had ever been replaced or whether the replacement interval should -be 4.5 or-
11 years. The 11 year replacement interval came from the original Sargent and Lundy EQ study.
The solenoid was installed in 1981 as part~of the remote shutdown-modification per Appendix R' requirements. The solenoid needs.to only perform its function during an event that would require the
evacuation of the control room and plant shutdown from the remote j
shutdown panels.
Another solenoid (HOV-1108B-20C), in the air J
control system for Valvo HCV-11088, has the function of operating the valve from the control room.
Solenoid HCV-11088-20C was-i replaced before its required EQ interval expired.
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The licensee's engineering' group.made a determination.that-the;
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~4.5 year replacement-interval'was overly conservative and that-
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replacement could-wait until the'next outage. When the plant shuti
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down on August 24,'1990, due to a reactor coolant pump seal problem, '
the-solenoid was replaced.
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The issue of whether the replacement intervals: for other_. solenoids,'
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that are in'the DM program,.are correct was;being reviewed by'the-
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licensee. The 11censee's action in this matter appeared'
conservative and correct.
'l c.
Reactor Coolant Pump RC-3A Lower Seal Blockage
q On August 15, 1990, the pressure. measured between the lower seal and;
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the middle seal on RCP RC-3A began to decrease from its normal:
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pressure of approximately 1400 psia..On August ~ 17,:1990,_theLlow
. pressure alarm ~was-received at 1200 psia. The1 system engineer-
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suspected a partial blockage of the pressure breakdown ~orificei
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between the lower and middle sesi, since the' pressure measured:
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between the middle and upper seal _ was also decreasing' at approximately the same rate.
On August.20, 1990, the system engineer contacted the pump; manufacturer, Byron Jackson, for assistance. 'TheElicensee was'
informed that if the lower seal were to catastrophically fail ~, the'
mtddle and upper seals, as well as the vapor barrier, wereidesigned.
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to w!thstand full reactor coolant system pressure. :The manufacturer;
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stated that the main concern would.be:if' the breakdown orifice was
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l to be comp etely blocked.
This could ca'use damage to-the: remaining l_
seals and then, if the lower sealtwas-to fail,> primary coolant could
leak into the containment.
The licensee felt assured,-f om l
discussions with Byron Jackson, that if the lower seal was to be a
blocked completely, the plant had ample time to.beisafely: shut'down..
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The decrease in middle seal inlet pressure was-relatively slow and
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steady from the initiation of the event.
Except for momentary-spikes, there were no sustained large increments in pressure, either
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increasing or decreasing. Then on August-24, 1990, at-_10:56 a.m.=,
the pressure began to rapidly oscillate. At 11:05.a.m., the-pressure
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steadied at 825 psia and began to slowly decrease until, at 3:18 p'.m.,-
it oscillated between 650-1400 psia three times.
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to increase in frequency, and by 3:55 p.m. the middle ' seal inlet
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pressure was steadily oscillating between 500-1400 psia'over a-10 second interval.
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At approximately 4 p.m. on August 24, 1990, the plant manager made the decision to shut down the plant before the lower seal became completely blocked and damage to the remaining seals occurred. At
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4:19 p.m. the licensee commenced a controlled shutdown from
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100 percent power.
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.At -approxi.nately 6-p.m. on August 24, 1990, pressure readings-j
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' ndicated thatLthe. middle. seal had: failed. Theprobablecauselofthe i
- failur was determined to be the.pressurefoscillations that~were-
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occurring.7 The system engineer was concerned that the-pressure oscillations would!al.so damage the ! remaining ' seals'. The decision'wasc
then made to' reduce power-at a faster. rate:in order to; reach atlevel.
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.where RCP RC-3Alcould be shut'down.
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.At 7:35 p.m. and 7:40 p.m. on A'ugust 24k1990, the plant ' entered Hot -
' Standby (Mode 2),'and Hot Shutdown(Mode.3),~respectively.
RCPLRC-3A-
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was secured at 7:44 p.m., and the : system engineer; determined that the seals were. n'o. longer in jeopardy' of 'failing'
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An RCS cooldown was commenced at 12:10 a.m.fon August 25,(1990;.The plant' reached Cold Shutdown (Mode 4) at approximately'6i23;p..m.L'on-
August 26.
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On: August 29, 1990, the RC-3A seal cartridge', whi.co contains all-i four: seals, was replaced..During the replacement,Li; was noticed ~
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that lock wires were missing from two of,the. bolts that attach the-
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lower seal assembly to the middle seal-assembly._'The licensee-suspected that this could be the source oftthe. blockage.. An L initial.
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inspection on August 31 to determine the exact cause of.the blockage was not successful.
The decisionLwasimade,to postpone a detailed inspection of the seal cartridge until after the' outage was completed.
The cartridge was taken apart-on. September 5,'1990, under the direct observation of the system engineer. There was no' obstruction,found in the inlet of the pressure breakdown orifice, but.there-were-some-metal balls found at the outlet.: Thetsystem. engineer suspectedLthat
.these were the lock wires which had:b'een' forced through,the breakdown orifice, which is only about 1/10 inch in diameter.: No further
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inspection was performed at that time.
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Tentative plans were to bag the lower seal-assembly", decontaminate it, and remove it to a more controlled environment..A more detailed-
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analysis of the cause of the blockage will be performed. The licensee may request assistance -from Byron Jackson in' the analysis since this -
problem was considered very ' unusual;by the manufacturer.,The NRC t
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will continue to monitor the licensee's-corrective actions.
The performance of the system engineer during this entire event demonstrated the benefits of the system engineer 'p_rogram.
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d.
Reactor Trip Signal While Shut Down
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On August 29, 1990, while the plant was in cold shutdown, a reactor.
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trip signal was generated while swapping. control element drive
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mechanism (CEDM) clutch power supplies. The power supplies were
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being swapped from Inverter "A" to "B" and from Inverter "0" to
"C" to perform testing on the Inverter "D". bypass transformer.
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transformer had been replaced due to excessive noise.
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t L While transferring-the clutch power-supply from Inverter "A" to "B,'I d
, per Operating Instruction 01-EE-4', Reactor: Trip Breaker CB-AB
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tripped as_ expected. The operator lthen inadvertently transferred:
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the power supply from Inverter:"D"-toL"C" before'. resetting Reactor
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Trip Breaker: CB-AB, as required by -procedure. When Reactor Trip.
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Breaker CB-CD also. tripped, all CEDM clutch power' supplies were lost,
- resulting in a; reactor trip. signal.
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- The reactor trip function w'assimmediately reset. The trip signal:
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did not result in'any contro1~ rod movement since-the plant was--
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The: inspectors will review the licensee's corrective actionlon'thisi j
event during the closeout of the licensee event report. LA' review.of
the procedure and the panel where'the error occurred showed nothing-d that-would have caused this error other-than; inattentiveness to
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No violations or deviations were identified in this area.
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Operational Safety Verification -(71707)
The inspectors conducted reviews and observations of selected activities
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to verify that facility operations.were performed in compliance with the.
appropriate regulatory requirements, a.
The inspector was in ti control room fromethe time.the plant'..
commenced its controll o shutdown until the' unit was suberitical.
At 4:19 p.m. on August 24, '1990,' the. licensee commenced a controlled -
shutdown from 100 percent power due t'o a-steadily decreasing inlet pressure on the RCP RC-3A middle seal.
The details of the! plant shutdown are described in paragraph'3.c.
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'The inspector-noted that the conduct of the' operators and other.-
support personnel in the control room was very professional.
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throughout the witnessed evolution.
Good' communication flow between control room personnel was noted by the inspector.-
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On August 28, 1990, the inspector witnessed the draining down of the-
reactor coolant system to support repair of the seal-on RCP RC-3A.-
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y" The inspector observed the evolution to have been well controlled and-performed in accordance with Procedures OP"6, " Hot Shutdown to Cold.
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Shutdown," and 01-RC-5, " Reactor Coolant System Draining for. Plant Shutdown."
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During the draining evolution, operators noted that the cold shutdown reactor coolant level indicator, LI-197,.was providing
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erroneous-readings. The readings were nonconservative wnen compared.
to the sight glass readings and the redundant. level indicator,
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LI-119.
The operators promptly declared LI-197 inoperable had the
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-instrument readout.in the controlL room tagged as deficient, ani
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required the: sight' glass be nionitored every 15 ~ minutes. - ALlog was established to record the:15-minute interval. readings and verify 1 l
1their correspondence'to-the remaining-functioning _. level instrument-(LI-119).. The monitoring of-the sightglass at. thel specified-intervals'is~a commitment:made by the licensee'inEits
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- response,to. Generic' Letter 88-17,~." Loss of Decay Heat Removal." On-an-intermittent l basis, the inspector confirmed.that theirequirement to verify RCS level by sightglass -observation was beingnimplemented; throughout:the day on August'28, 1990.L During the: evening of'
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' August 28, 1990, the defective LI-19_7lwas: repaired-and recalibrated.
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.It.was declared operational'at 5 a.m..on' August 29, 1990, negating; I
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.the need to takerthe 15-minute readings directly' at; the sirbtglass.;
d The inspector _ also confirmed that during mid-' loop operation. the.
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- licensee maintained operational the necessary-normaliand emergencyL
power sources.
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The inspectors witnessed selected portions of' the plant's startup.
=from the licensee's outage to' replace the RCP RC-3A seal cartridge.
On S.eptember 1-2, 1990, the inspectors were11n the contr'ol rooma
,l observing activities' prior to achieving criticality, i
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.I The inspector. observed the transition from ModeL3 (Hot Shutdown)Lto.
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Mode 2 (Hot Standby). The transition was observed to have1been-Lperformed in accordance with Procedure '0P-7,'" Reactor: Startup." LThe reactor achieved criticality on' September 2,1990, at -approximately :
8:47 p.m.
Criticality was delayed several. hours until:the'RCS hydrogen concentration could be brought.into specification.
On. September 4, 1990, the licensee suspendedothe power ascensioniat.
j 30 percent to perform the boric-acid. soak of. the steam generators.
j This was successfully completed, and.at 'I a.m. on-September: 5,.1990,
power ascension was resumed.
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Theplantreached95:percentpoweronSeptember5,L1990,-whereit
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remained until September 6, while the required calibration of the nuclear instrumentation was performed. On. September:7, 1990,-the
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plant attained the 100 percent power.. level.
The inspectors noted good communication and attention'to detail by control room personnel during this evolution.
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No violations or deviations were identified in this area.
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Monthly Maintenance Observations (62703)
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The inspectors reviewed and/or observed selected station maintenance'
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i On August 7,1990, tne inspector witnessed a selected _ portion of the
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performance of Preventive Maintenance Procedure EM-PM-EX-0200,
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"4160 Volt Circuit Breaker Inspection." This job was the annual preventive mainter.snce on the GE 4160-volt Magna-Blast breaker for Raw Water Pump AC-10B.
EM-PM-EX-0200 had instructions from GE.to inspect for striker plate deformation. The instructfons were documented in "GE Service Advice 073 (SBD) 325.1 Magna-Blast Striker Plate", dated May 9, 1978. The plate fits flat up against the case of the breaker.and-j was welded at the bottom but not at the top. GE stated that the l
top of the ' plate could deform and move away from the breaker case.
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Maintenance personnel inspect for this deformation at every annual preventive maintenance (PM) activity, and Breaker.AC-10B had not--
previously experienced this deformation.
However,Lduring this PM activity, striker plate deformation ias discovered. This condition
was noted in the PM documentation, and Maintenance Work
Order (MWD) 903750 was initiated to correct the problem. The l
GE-recammended fix to this deformation was to cut a notch out of the top of the striker plate, clamp the plate to the side of the breaker, j
and weld the plate at the notch. 4The licensee dete'rmined that the deformation did not affect the operability of the breaker. The i
breaker was returned to service until repairs could be made.
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The 1.nsoector witnessed, on August 24, 1990, a portion'of the work involving the replacement of the Inverter "B" bypass transformer,
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i The transformer had been removed from service due to excessive noise.
The transformer was replaced under MWO 903728.
The work was completed in a professional manner.
The Inverter "D" bypass transformer, which was also experiencing excessive noise was.
replaced during the outage. The licensee plans to troubleshoot the two transformers to determine the cause of the noise.
No violations or deviations were identified in-this area.
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Monthly Surveillance Observations (61726)
The inspectors observed TS-required surveillance testing on-i safety-related systems and components.
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On August 17, 1990, the inspector witnessed the monthly performance
of Surveillance Test Procedure ST-RPS-6, " Steam Generator Level i
Channels, Revision 15. The purpose of the procedure was the:
- Comparison of the. steam generator level readings;
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- Verification of pretrips, trips, and control functions; and
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Calibration and adjustment of channels to ensure the accurate reflection of actual plant conditions.
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This~ test satisfied the requirements of Technical
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Specification (TS) 3.1.
The inspector found the test to be performed
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in a professional manner, with good attention to detail and
coordination between instrumentation and control and operations..
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On the evening of August 29, 1990, the inspector observed the'
licensee perform ASME Section XI-testing of. numerous motor-and-air-operated valves, which are required to be tested during each cold shutdown.
The inspector observed the stroke testing of chemical and volume control sys'.em Valves HCV-240, -241, -249, and -265 in accordance with Procedure OP-ST-CH-3005, " Chemical and Volume Control System (CVCS)
Category A and B Valve Exercise Test."
Also witnessed was the exercising on component cooling water' Valves HCV-4388 and -D, which supply cooling water'to the reactor coolant pumps' mechanical seals. This is also required to be performed-during each cold shutdown. The evolution was found to have been-performed in accordance with Procedure.0P-ST-CCW-3004,:" Component Cooling Category A and B Valve Exercise Test."
For all-valves tested, the inspector confirmed that the valves were timed remotely and their operation observed loca'ily as is required by ASME Section XI. All valves operated within specifications and the results were appropriately documented in the test record..
During the outage, there were numerous valves which required ASME Section XI testing, since plant conditions preclude exercising them-at power.
It appeared the licensee efficiently managed.the execution of these tests to satisfy their ASME Section XI. commitments.
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l 7.
Security and Radiological Protection (RP) Observations (71707)
The inspectors verified that the physical security plan and RP' program were being implemented.
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During the inspection period, the inspectors witnessed the' performance of:
l the security force when the access control metal detectors and the. key
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card readers were inoperable for a short period of time.
The inspectors i
noted that adequate compensatory measures were taken on all occasions.
I No violations or deviations were identified in this area.
8.
Review of Previously Identified Items (92701)
During this inspection period, the following items were reviewed:
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(Closed) Inspector Followup Item 285/9032-03:. Failed Resistance-Temperature Detector (RTD) in the Loop 1 Hot Leg.
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This item concerned a failed-high RTD (TI-111H) in the Loop I hot leg. This RTD provides input to the reactor regulating system, i
which provides control functions only.
The licensee committed to
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replacing the RTD during the next refueling outage.
However, due to the RCP seal outage, the RTD was replaced during this. inspection pertod.
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b.
(Closed) Unresolved Item 285/8950-01: Status of Hydrogen Purge System (HPS) and Hydrogen Recombiners.
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This item concerned whether. hydrogen recombiners are required at FCS
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and whether the HPS should be tested.
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The licensee's response was that hydrogen.recombiners are not i
required to be installed, since the HPS is sufficient to control combustible gas concentrations.
An emergency operating procedure l
does address.the use of recombiners but provides for their use only
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if-they are available.
The testi,19 of the HPS is not currently addressed by the TS. However, a Tt change has been submitted to include. testing' requirements for the
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HPS.
The licensee'; actions appear adequate-to resolve this item.
-l No violations or deviations were identified ln this area.
l 0.
Exit Interview The inspectors met with Mr. W. G. Gates, (Division Manager, Nuclear Operations) and other members of the licensee staff on September 11, i
1990. The meeting attendees are listed in paragraph 1 of this inspection.
report. At this meeting, the inspectors summarized the scope of the
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inspection and the findings. During the exit meeting, the licensee did
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not identify as proprietary any information provided to, or reviewed by,
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the inspector.
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