IR 05000397/1986032
| ML17278B061 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 10/21/1986 |
| From: | Rebecca Barr, Dodds R, Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17278B060 | List: |
| References | |
| 50-397-86-32, NUDOCS 8611050043 | |
| Download: ML17278B061 (16) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report No:
Docket No:
Licensee:
50-397/86-32 50-397 Washington Public Power Supply System P.
O. Box 968 Richland, WA 99352 facility Name:
Washington Nuclear Project No.
(WNP-2)
inspection at:
WNP-2 Site near Richland, Washington Inspection Conducted:
te er 1 - October 4, 1986 Inspectors:
T.
dds Senior Resident Inspector
~R.
C.
err, snidest Inspector Approved by:
P.
H.
hnson, Chief React Projects Section
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Date Signed
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Date Signed Ins ection on Se tember 1 - October
1986 (50-397/86-32)
Areas Ins ected:
Routine inspection by the resident inspectors of control room operations, engineered safety feature (ESP) status, surveillance program, maintenance program, licensee event reports, special inspection topics, and licensee action on previous inspection findings.
During this inspection, Inspection Procedures 30703, 39701, 61705, 61726, 62703, 71707, 71710, 90712, 92700, 92701, 92703, and 93702 were covered.
Results:
No violations or deviations were identified.
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DETAILS Persons Contacted
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Mazur, Managing Director Shannon, Deputy Managing Director Martin, Assistant Managing Director for Operations Powers, Plant Manager Baker, Assistant Plant Manager Corcoran, Operations Manager McKay, Assistant Operations Manager Cowan, Technical Manager Graybeal, Health Physics and Chemistry Manager Peldman, Plant (}uality Assurance Manager Peters, Administrative Manager Powell, Licensing Manager Wuesterfeld, Reactor Engineering Supervisor Landon, Plant Maintenance Manager Bouchey, Director Support Services-Personnel in attendance at exit meeting The inspectors also interviewed various control room operators, shift supervisors, shift managers, engineering, quality assurance, and management personnel relative to activities in progress and records.
Plant Status At the beginning of the period, the reactor was at 100O/ power.
On September 3,
1986, due to a loss of reactor feed pump and a malfunction of recirculation flow control valve RRC-PCV-60B, the reactor scrammed on low water level.
A reactor restart was performed on September 7 with 100$ reactor power attained on September 10.
Reactor power was reduced to about 24/ for four hours on September 13 to repair switchgear motor-operated disconnects.
On September 26, reactor power was reduced to 30/ to perform a rod sequence exchange and was returned to 100/ on September 28.
Power was again reduced on October 1 because of noise associated with the No.
4 high pressure steam turbine governor valve.
The valve was closed following licensee investigation and was expected to remain so until a maintenance outage is taken.
The reactor was then gradually returned to power on the other three valves.
0 erations Verifications The resident inspectors reviewed the control room operator and shift manager log books on a daily basis.
Reviews of the Jumper/Lifted lead Logs, Non-Conformance Report Iog and the Iimiting Condition for Operation (LCO) Log were conducted to verify there were no conflicts with Technical Specifications and that the licensee was actively pursuing corrections to off-normal occurrences listed in the logs.
.Events involving unusual conditions of equipment were discussed with available control room personnel and evaluated for safety significance.
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licensee's adherence to ICO's, particularly those dealing with engineered safety features (ESF)
and ESF electrical alignment, was observed.
The inspectors routinely took note of activated annunciators on the control panels and ascertained that the control room licensed personnel on duty at the time were familiar with the reason for each annunciator and its significance.
The inspector observed access control, control room manning, operability of nuclear instruments and availability of on-site and off-site electrical power.
The inspectors made regular tours of accessible areas of the facility to assess equipment conditions, radiological controls, security, safety and adherence to regulatory requirements.
No violations or deviations were identified.
Surveillance Pro ram Im lementation The inspectors ascertained that surveillance of safety-related systems or components was being conducted in accordance with license requirements.
In addition to witnessing and verifying daily control panel instrument checks, the inspectors observed portions of several detailed surveillance tests by operators and instrument and control technicians, including the following:
PPM 7.4.7.9.1, Weekly"By-Pass Valve Testing PPM 7.4.3.8.2.1, Weekly Turbine Valve Tests PPM 7.4.8.1.1.2.11, Monthly,Diesel Generator Test PPM 7.4.3.2.1.91, Isolation Activation Reactor level 2 PPM 7.4.8.1.1.2.12, HPCS Diesel Generator Test No violations or deviations, were identified.'
Monthl Maintenance Observation Portions of selected safety-related',systems maintenance activities were observed.
By direct observation 'and review of records the inspectors determined whether these activities were consistent with ICOs; that the proper administrative controls and tag-out procedures were followed; and that equipment was properly tested before return to service.
Specific activities observed included replacement of a condensate booster pump and the high pressure testing of Static-0-Ring differential pressure
'ransmitters required by IEB-86-02.
Testing of the 4 Static-0-Ring differential pressure (DP) transmitters used for level 2 containment isolations found 3 of 4 transmitters within specifications.
The MS-IS-61B transmitter exceeded the allowed setpoint, by.6" of water pressure.
As a corrective measure all the transmitters were recalibrated so that no transmitter will trip at a level lower than-50 inches at operating pressure.
Subsequent testing will occur in accordance 'with the recommendations of the bulletin.
Several of the tests were observed by the inspectors, including recalibration testing.
The reader-worker routine used by the instrument technicians assured that the tests were performed error fre 'IUI I
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No violations or deviations were identified.
Calibratio'n of Nuclear Instrument S stems The procedures pertaining to the cal'ibration of the LP16i and APRM systems were examined for technical content and to 'assure that appropriate precautions and prerequisites had been included.
LPRM calibration is required every 1000 effective full"power hours using the Traversing In-core Probe (TIP) system.
The, calibrations are, conducted in accordance with procedures 2.1.3, 9.3.3, 9.3.4, 7.4.3.7.7 and 7.4.3.1.1.70.
No deviations or violations were identified.
ALARA Considerations'
concern was expressed to the inspectors regarding strict adherence to the principles of ALARA (as low as reasonably achievable radiation exposures)
with respect to the storage of deck grating between the shield doors and containment building equipment hatch while operating at full power. It was believed that this job should have been done the previous day when the reactor was operating at about 30$ power.
Examination of the activity disclosed that maintenance had been trying to coordinate the storage of deck grating for 'several weeks.
The concern was that the decking would be discarded during cleanup and that it would be needed in the drywell.
Security, Operations and Health Physics personnel all have to be present to open the the equipment hatch shield doors.
On September 29, 1986, Health Physics and Maintenance groups reviewed the job and decided that to put it off until a lower power run would not be required.
Work area radiation readings were expected to be below 100 mR/hr and the stay time would be about 5 minutes.
One HP technician and "two mechanics completed the job with no measurable gamma and
mRem neutron exposure each.
Total exposure for the job was no gamma and
mRem neutron (timekeeping).
No violations oz deviations were identified.
En ineered Safet Feature Verification The inspectors verified the operability of selected Engineered Safety Feature Systems on a daily basis.
Additionally, walkdowns of accessible portions of the Service Water System and the Residual Heat Removal System were performed to confirm that the licensee's system lineup procedures matched plant drawings and the as-built configuration, and that valves were in the proper position, had power available and were locked as appropriate.
The licensee's procedures were verified to be in accordance with Technical Specifications and the FSAR.
The inspectors observed the two additional damaged hangers associated with the service water 'cooling piping of the fuel pool heat exchangers.
The inspectors'nderstanding was that the re-evaluation of the service
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water system design ini'tiated as a result of the water hammer event earlier this year will take into account these damaged hangers.
The redesign effort will occur prior to and be implemented during the second refueling outage.
No violations or deviations were identified.
8.
Licensee Event Re orts The resident inspectors reviewed the following reports and supporting information on site to verify that licensee management had reviewed the events, corrective action had been taken, no unreviewed safety questions were involved, and violations of regulations or Technical Specification conditions had been identified.
IER-86-25 Reactor Hi h Pressure Scram on Jul
1986 (Closed)
The licensee's submittal of LER-86-25, dated August 22, 1986, indicated that Plant Engineering review was needed to further characterize the transient and that this information would be provided in a supplemental report.
While not addressed in the LER, discussion with Generation Engineering provided the desired analysis related to the potential adverse effects of water in the steam lines.
The engineering analysis which was completed and approved on October 1, 1986, indicated that approximatel'y 10,400 gallons of water had flowed into each steam line, resulting in filling the lines with about 8.5 inches of water that could have been 118 degrees 7 less than steam piping temperature (conservative assumption based on recirculation/feedwater inlet temperature to the reactor vessel).
The licensee's initial assessment of the event did not consider the effect of water in the steam line prior to plant restart on the evening, of July 25, 1986.
Apparently, Generation Engineering's concern for the,.
potential effect of water in the steam lines did not surface for several weeks.
This question resulted in a walkdown of main steam lines outsirle primary containment to check for damage on August 22, 1986.
No visual damage (deformed or bent components, structural or concrete damage)
was
- observed.
'urther, a walkdown inside primary containment on September 3-4, 1986, did not disclose any anomalies.
'Generation Engineering's analysis showed that the difference between water and pipe wall temperatures was insufficient to damage piping, pipe supports or the MSIVs.
, Action being taken by the licensee -- to strengthen the post trip review process to assure that all significant facets of an event have been considered prior to restarting the reactor -- will be examined during a
future inspection.
(Pollowup item 86-32-01)
LER-86-29 Dr ell H dro en Anal zer Calibrated with Wron Gas Concentration Closed Examination of the data associated wi.th the calibration check of the drywell hydrogen gas analyzers on August 6 and ll, 1986, using 25/
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hydrogen showed the as-found condition of the analyzers to be acceptable in the area of interest, with no adjustments necessary.
The indicated leveL at 25't, was 23.54/ but no adjustments were made because the calibration was accurate between 2 and 6% hydrogen where recombiner operation would be initiated.
As noted in the KER, the analyzers were being calibrated with six rather than twenty-five percent hydrogen.
This resulted from approval of a procedure to implement a techn'ical specification change request before issuance of the change by the NRC.
The generic implication of premature procedure approval will be followed under open item 86-32-02.
LER-86-29 is considered closed.
LER-86-32 Incorrect Sizin of Under round Cable (0 en)
I A phase-to-ground short was discovered on circulating water pump CW-P-1C on September 12, 1986.
Xt was subsequently discovered that underground cabling insulation had deteriorated from overheating.
Apparently, the A-E (Burns and Roe)
had applied ampacity'actors for cable in a tray rather than that applicable.to underground (buried) cabling.
Temperature in a spare conduit'10 days after the cables had been removed was found to be at 150 degrees F (due,to'eat generated by energized cables in adjacent conduit).
This cabling was being replaced, pending final resolution of the problem.
Discussion with the cognizant, engineers disclosed that the only safety related equipment that could have a similar problem was the standby service water (SSW)
pumps and HPCS service water pump.
Calibrations show that the cable for the, HPCS service water pump, which runs in a separate duct, will be good for at least 40 years.
Manufacturer data showed the cabling for the SSW pumps, which'also runs in separate ducts, to be good for at least.
4 years of continuous operation.
Since the system has only been in operation for three years, and not run continuously, the licensee determined that cable replacement could be delayed until the next refueling outage.
An outside consultant will be brought in to evaluate and re'commend interim and long term solutions to the problem.
No violations or deviations were identified.
Fire Main Water Hammer Event, A water hammer occurred in the Reactor Building standpipe RB-2 and the Technical Support Services Building (Building 88) during the conduct of a water loop system flush on September 15, 1986 (surveillance procedure 7.4.7.6.1.1.5).
Several mechanical connections were sprung, causing severe water damage in Building 88.
The connecting rings for two gated wye valves that attach to Reactor 'Building standpipe RB-2 also cracked.
One of the valves allowed a substantial amount of water to run into the Reactor Building stairwell as its corresponding block valve was partially open.
The water in the Reactor Building did not cause any damage or operating problems.
While the water damage in Building 88 was extensive, it does not appear that any quality records were destroye ~
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The inspector observed the subsequent 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> hydrostatic test at 225 psig on October 1, 1986,'f the underground piping from the fire water main to the top of RB-2 standpipe.
No leaks were identified in the standpipe and the pressure only decayed to 224 psig during the conduct of the test, indicating the system to be 'solid.
Examination of the factors contributing to the event showed that the flush procedure did not correspond to the plant cofiguration.
Specifically, a
PMR (Plant Modification Request)
had been initiated to place the warehouses fire main on a loop that can be flushed.
However, this change had not been completed and the warehouse loop connected to the upstream main.
As a result, when the main loop was valved out and the discharge valve opened to flush the warehouse loop, it only drained standpipes that were open to the small portion of the main system being used for the flushing operation.
This apparently created a vacuum in these lines.
Mhen the operators noted that they could not establish flow through the warehouse loop, they discontinued that portion of the flush and proceeded to the next portion of the test which was to flush the "industrial" loop.
Upon starting this operation, the standpipes were refilled, causing a water hammer that resulted in the aforementioned broken pipe connections.
The inspector could not discern from review of the secretary's notes from the meeting of POC (Plant Operations Committee)
on August 1, 1986, wherein the revised flush procedure was approved, that the PMR had not been completed nor that the warehouse loop connection had not been made.
The examination of the Control Room PAID showed the fire loop to be Accurately configured in this regard.
LER 86-29, discussed in paragraph 8 of this report, also described a
situation wherein a plant procedure was approved to correspond to a
technical specification change request prior to the approval of the change by NRC.
As a result, the hydrogen gas analyzers were being calibrated with six percent rather than twenty-five percent, hydrogen (by volume).
The licensee stated that the procedure review process was being upgraded to require three tiers of review prior to being presented to POC for acceptance.
This program and its implementation will be examined further.
(Followup item 86-32-02)
No deviations or violations were identified.
Licensee Actions On Previous NRC Xns ection Findin s Followu Item 86-06-04 0 erator's and Shift Mana er's Lo s (Closed)
The examination of operating logs indicated a definite improvement in the quality of material now being included in the.Shift Manager's and Control Operator's logs.
However, in, the in'spector's opinion, these logs could be enhanced and be of greater value to subsequent operations if they contained detail on action being taken, if any, or causative factors for specific deficiencies.
Examples where additional information would be useful include:
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Cause of blown fuses on September 5, ll and 12, 1986.
High radiation level indication in the Radwaste and Reactor Building.
RIB-RV-25A relief valve leakage.
The licensee was drafting a revised procedure to provide additional guidance on log entries.
Pollowup item 86-06-04 is considered closed;
. however, the inspectors will continue to monitor the quality of operating logs as part of the routine program.
11.
Mana ement Meetin The inspectors met with the Plant Manager approximately weekly during this period, to discuss inspection finding status.
On October 7, 1986 the inspectors met with members of licensee management as indicated in paragraph l to discuss the inspection findings during this perio II 1I lt
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