IR 05000482/1990002
| ML20011F227 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 02/15/1990 |
| From: | Holler E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20011F224 | List: |
| References | |
| 50-482-90-02, 50-482-90-2, NUDOCS 9003020168 | |
| Download: ML20011F227 (13) | |
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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION
' REGION IV r
g NRC Inspection Report:
50-482/90-02 Operating License:
NPF-42 Docket: 50-482 Licensee: Wolf Creek Nuclear Operating Corporation (WCNOC)
P.O. Box-411 Burlington, Kansas 66839 Facility Name: Wolf Creek Generating Station (WCGS)
Inspection At: WCGS, Coffey County, Burlington, Kansas Inspection Conducted: January I-31, 1990 Inspectors:
B. L. Bartlett, Senior Resident Inspector Project Section D, Division of Reactor Projects M. E. Skow, Senior Resident Inspector Project Section D, Division of Reactor Projects
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Approved:
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E. Jf Holler, Chivf, Froject section D Date Dtvision of Reactor Projects
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[nspv t6:n %e]_ry Inspectwn Conductet Januaryl -3) l 9y_(_ Report 50-462/90-02)
-Areas'Ingected:
Rcutine, ura nov.ced irspection including plant status,
operational ss.fet'y verification, monthly surmillarce obsenation. monthly
maintenance disevation, rev'.tw of 11cerJea event reports, and followup on a
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crevit,usly identifiec NRC item.
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Results: Within the areas inspected, two violations were identified. The i
inspectors' review of maintenance activities identified that the licensee's understanding of the safety significance of maintenance issues was clear;
h however, one example of. not properly documenting the thought process was identified in paragraph 5 as a violation.
In general.the licensee's corrective l
actions to conditions adverse to quality are usually effective; however, one example of failure to institute effective corrective actions was identified as a violation in paragraph 3.c.
Licensee management's attention to and involvement in the determination of the operability of safety equipment is usually good; however, an isolated instance where it has been less than aggressive is discussed in paragraph 3.b.
While shutting off the boron thermal a
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L regeneration system (BTRS), the licensese experienced a pressure surge within the chemical and volume control system (CVCS) that' caused body-to-bonnet leaks-
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from three diaphragm valves. Work was performed to check and tighten body-to-bonnet' bolts on several. diaphragm valves,'but the specific valves that
= were actually tightened were not recorded.' This was another example of the
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En need to make personnel at all. levels more aware of the importance of _ recognizing i
potential safety issues (paragraph 3).
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DETAILS l
1.
Persons Contacted Principal Licensee Personnel B. Withers, President and CEO
- J. A. Bailey, Vice President, Operations
- F. T. Rhodes, Vice President, Engineering and Technical Services
- G. D. Boyer, Plant Manager
'H. K. Chernoff, Supervisor, Licensing
'R. B. Flannigan, Manager, Nuclear Safety Engineering (NSE)
- R. W. Holloway, Manager, Maintenance and Modifications C. E. Parry, Director Quality Assurance (QA), WCGS
- W. M. Lindsay, Manager QA
- J. Weeks, Manager, Operations
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- J. M. Pippin, Manager, Nuclear Plant Engineering (NPE)
"S. Wideman, Licensing Specialist III
- M. G. Williams, Manager, Plant Support
- T. Deddens, Outage Manager L
The inspectors also contacted other members of the licensee's staff during the inspection period to discuss identified issues.
- Denotes those personnel in attendance at the exit meeting held on January 31, 1990.
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2.
Plant S*atus The plant operated in Mode 1 (100 percent reactor thermal powrer) during the inspection period. There were no reactor or turbine trips.
3.
Operational Safety Verifiestion (717071
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The purpose of this inspection was to ensure that the facility wot being operated safely and in co formance with license anu reoolatory
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It also wt.s to ensure that the licensee s management
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control system wa9 effectively discharging its responsibilities for continued sate operation.
The methods used to perform this inspection included direct observation of activities and equipment, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation (LCO), corrective actions, and review of facility records.
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Areas reviewed during this inspection included, but were not limited to, control room activities, routine surveillances, engineered safety feature operability, radiation protection controls, fire protection, security, plant cleanliness, instrumentation and alarms, deficiency reports, and corrective actions.
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a.
On January 12, 1990, the operators noted a small body-to-bonnet leak on Valve BG-UV7002B, " Letdown Chiller Heat Exchanger Outlet Isolation." The valve was in the BTRS.
The licensee did not consider the leak sufficient to warrant system shutdown.
The licensee was operating the BTRS for the first time in approximately 2 years as a means of removing boron from the reactor coolant system (RCS) because boron concentration in the RCS was lov, near end-of-core life.
The licensee was also trending oil level in the BTRS chiller as it decreased steadily. At 5:35 a.m. the licensee shut off the BTRS system when the oil level had gone out the bottom of the sight glass.
Approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later, the reactor operator noted that there appeared to be an approximate 4 gpm decrease in the volume control tank (VCT) level in the RCS, At about tnat time, an increase in unit vent activity was noted. The operators found Valve BG-8516, cation bed demineralizer inlet isolation, leaking. The valve was shut and the leak stopped. VCT indication became level.
The control room operator performed STS BB-004 to quantify the RCS leakage. The operators found the leakage to be about 3.4 gpm unidentified and entered LCO 3.4.6.2.
The LCO required reducing the leak to less than 1 gpm in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or placing the plant in hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. While maintenance personnel were working to tighten the body-to-bonnet studs on BG-UV7002B and BG-8516, they discovered a leak from BG-V032, spent resin pump to cation bed demineralizer isolation. Maintenance personnel also tightened the studs on BG-V032.
The leak rate surveillance, STS BB-004, was performed after the studs were tightened with satisfactory results.
During the event, six workers were contaminated with measyred contamination levels as high
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as 1200 counts per minute to their hands and heads.
The workers were decontaminated and a subsequent whole L:ody count showed low levels of Xenon. The eSntamination hvels were below threshold reporting requirements.
At about 6:27 p.m. the same evening, tnt radiation levels in the auxiliary building vent began to increase again, and level in the VCT beghn to decreats. This twe, cueratore found and tschted a leak nn BG-HV8245, r:TRS divert valve.
This len n s a'so a boc'y-to-bonnet leak and the studs were tigH;ened to stop the leak.
Tnroe of the four leaks were in the CVCS and occurred after the BTRS f
was shut off.
RCS water is diverted from the CVCS to the BTRS, before returrir.g to the RCS, when the BTRS is being used. When operators turned off the BTRS, three valves in the BTRS automatically shut and BG-HV8245 automatically opened. Operators recalled that
CVCS pressure momentarily increased while flow decreased as the l
valves repositioned. The operators stated that pressure in the CVCS
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increased from anproximately 350 psig to about 410 psig, and that flow decreased from 75 gpm to about 40 gpm before returning to normal. The operators concluded that BG-HV8245 was slower to open l
than the other valves were to close and that this condition created a i
pressure surge that caused the leaks. A change was promptly initiated
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to the BTRS operating procedure SYS BG-205, to open BG-HV8245 before turning off the BTRS. This revised procedure was subsequently tested and the CVCS did not display the pressure and flow transients
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In addition, the licensee initiated a review of SYS BG-205 based on other operating characteristics brought to light by this event.
The licensee also verified that the body-to-bonnet studs were tight on a list of approximately 70 2-inch and above diaphragm valves prior to p
restarting the BTRS. This was perfonned under troubleshooting WR 00296-90. The work accomplished was indicated on the WR as having tightened the listed valve bonnets snug tight and adjusted all travel stops, as required. The specific valves that required adjustment and tightening were not noted. The licensee stated that six valve bonnets were tightened. After discussioni. with maintenance personnel, it appeared that four of the six were the valves that had leaked.
The other two valves could not be identified with confidence.
It also appeared that the notation on the WR met the minimum requirements of Procedure MGM M00C-01, Revision 2. " Mechanical Maintenance Troubleshooting." The work instructions did not require a list of valves actually adjusted, and the maintenance personnel did not provide a list in the description of work they accomplished. However, not recording the valves worked inhibited complete evaluation of the extent that the pressure surge affected system components.
b.
During this inspection period, the inspectors developed operability corcerns regarding equipment listed in Licensee Condition 2.C.7,
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Attachment 3. Items 3b and 3d. This license condition requires, that
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prior to restart following the first refueling outage, the operating
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corporation shall have installed and operable certain instrumentation.
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monitors for releases from steam generator atmospheric duop valves.
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Although this equipment was installed in compliance with the license
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L crrdition, the inspectors observed that two of the four radiation Nonitors and one of the four reactor vessel water level (RVLIS)
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L indicators were inoperabit Upon further inspection, the inspectors L
determined that with the licensee's currect cutbge and repeir
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schedule, the RV1.!S indinter and one of the radiation monitors would be inoperable for over 1 year before being repaired. The inspectors'
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concern wds that the licensee appeared to place a lun priority On L
required safety equipment which did not have a specific Technical Specification (TS) requirement regarding availability. The inspectors reviewed the TS, Updated Safety Analysis Report, and NUREG-0881 (WCGS
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Safety Evaluation Report) and could not identify any specific l
availability requirement for these instruments. Notwithstanding i
this, the licensee should ensure that this equipment is maintained L
operable and that a prompt assessment of safety impact is made when l
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1 the equipnent becomes inoperable. Additional details regarding the equipment involved is discussed below.
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On February 19,1989 BB LI-1312 ("A" train forced flow RVLIS) had
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failed high and WR 00920-89 had been initiated to repair the
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indicator. Troubleshooting determined that the problem was most likely in the level transmitter; however, to investigate further, a reactor outage was required in order to gain access to the reactor
cavity. The licensee decided to delay the WR until an outage but did
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not place the WR on the forced outage list. This was one of'four RYLIS indicators, two forced flow and two natural circulation flow
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channels. The other three channels remained operable.
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During a routine tour of the auxiliary building, the inspector
observed the radiation monitors located on the discharge of the steam
generator atmospheric dump valves to "B" and "C" loops (AB RE-113 and
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AB RE-112) inoperable. AB RE-112 had failed on April 4,1989, and
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AB RE-113 had failed on December 17, 1989. Becaut.e of personnel safety concerns, the monitors are not scheduled for repair until an
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outage. AB RE-112 was placed on the forced outage list and, upon
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questioning by the inspector, AB RE-113 was also placed on the forced outage list.
If there had been a steam generator tube rupture, radioactivity releases would still have been detected by the normal
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condenser vacuum discharge radiation monitor and individual steam generator monitoring could have been performed by sampling.
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c.
On January 12, 1990, during a routine tour the inspectors observed a
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chainfall tied off to the fuel oil supply line to the "B" diesel generator (DG). During a seismic event, the additional load could have '5ead to a failure of the diesel fuel oil supply. While this was i
the first documented instance since Violation 4K/8538-01 on hoist /chainfalls was issued, there have bu.n three otl,er instances pointed out to the licensee, including during a tour of thu facility by NRC officials on March 7, 1988.
This is a violation of 10 CFR c
Part 50, Appendix B, Criterion XVI., " Corrective Action" (482/9002-02).
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Upon notification by the inspector, the licensee initiated prompt and effect!ve corrective action. This included adding a note to the operations' watchstander logs to chec% the cht.irralls shiftly, e
placing a hook nearby to attach the chainfalls and posting signs stating where to tie off the chainfalls, h.
4.
L Monthly Surveillance Observation (61726)_
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The purpose of this inspection was to ascertain whether surveillance of safety-significant systems and components was being conducted in accordance with TS. Methods used to perform this inspection included direct observation of licensee activities and review of records.
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i Items in this area included, but were not limited to, verification that:
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. Testing was accomplished by qualified personnel in accordance with an
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approved test procedure, i
The surveillance procedure was in confonnance with TS requiremnts.
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The operating system and test instrunentation was within its current
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calibration cycle.
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l-Required administrative approvals and clearances were obtained prior
to initiating the test.
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~ LCOs were met and the system was properly returned to service.
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t The test data were accurate and complete and the test results met TS
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requirements.
Surve111ances witnessed and/or reviewed by the inspectors are listed t'
below:
STS IC-253A Revision 5, " Analog Channel Operational Test Fuel
Building Exhaust Radiation Monitor GG RE-27," performed January 8,1990
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STS AE-201, Revision 7. "Feedwater System inservice Valve Test "
performed January 10, 1990
.STS BG-100A, Revision 7 " Centrifugal Charging System 'A' Train
It4 service Pump Test," perfonned January 10, 1990
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STS AL-103, Revition 10. " Turbine Driven Auxiliary Feedwater Pump
Inservice Punp Test," performed January 31, 1990
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Selected inspector observations are discussed below:
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The serve 111ance procedures that unre observed, appeared to te peWormed a
by knowledgeable personnel in accordince with procadures. During l
STS AL-103, the inspector noted that an instrunent line tube from
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I Valve AL-V141 to Flow Gage Al FI-49 vibrated. This is the same tubing that was discussed in NRC Inspection Report 50-482/89-05, paragraph 9.
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That report noted that the tube geometry is such that changes to its slope
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could occur from various activities after installation. The tubes from l :.
both sides of the gage are similar and run next to each other, but the L
other tube, from AL-V140, exhibited far less vibration. Neither tube is
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supported except at their ends through their connection fittings.
Licensee personnel were informed of this observation.
No violations or deviations were identified.
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5.
Monthly Maintenance Observation (62703)
The purpose of inspections in this area was to ascertain that maintenance activities on safety-related systems and components were conducted in accordance with approved procedures and TS, Methods used in this inspection included direct observation, personnel interviews, and records review.
l-Items verified in this inspection included:
Activities did not violate LCOs and redundant components were operable, i
Required administrative approvals and clearances were obtained before initiating work.
- Radiological controls were properly implemented.
Fire prevention controls were implemented.
- Required alignments and surveillances to verify postmaintenance
operability were performed.
- Replacement parts and materials used were properly pertified.
Craftsmen were qualified to accomplish the designated task and
additional technical expertise was made available when needed.
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Quality control (OC) hold poir.ts and/or checklists were used and 00 personnel observed designated work activities.
- Procedures used were adequate, approved, and up to date.
Portions of selected maintenance activities regarding the VRs listed below
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were. observed. The WRs and related documents were reviewed by the
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inspectors:
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Activity WR 03384-87 SBG02 chiller unit low on oil WR 04682-89 Obtain air pressure readings when AE FV-40 is exercised
WR 05117-89 Remove / reinstall missile shield to essential service water valye pit WR 52739-89 Motor driven auxiliary feedwater pump "A" semiannual maintenance WR 52874-89 Room Cooler SGF02A semiannual maintenance l
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WR 05373-89 Replace tubing from air start solenoid valve to barring device interlock mechanism on "A" diesel generator p
WR 00211-90 Obtain hydraulic oil sample from AE FV-40 WR 00258-90 SGB02 CVCS chiller unit oil low WR 00332-90 Alarm 21F did not come in when safety injection pump "B" breaker was racked out Selected inspector observations are discussed below:
a.
NRC Inspection Report 50-482/89-31 discussed a task force the licensee had established to identify the root causes_of the main
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feedwater isolation vahe (MFIV) 4-way slide failures. During this inspection period, the licensee received a laboratory report that i
stated the 0-ring back-up rings for the slide valves were not the correct material.
Two replacement ring kits were also found by the
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licensee's laboratory to have incorrect back-up rings. The back-up rings were found to be Buna-N instead of Viton. The back-up rings support the 0-ring in position on thimbles sitting on both sides of the internal slide assembly, creating a pressure boundary for the Fyrquel 220 hydraulic oil.
The hydraulic oil comes in contact with the back-up rings. The Buna-N rings that were installed in the plant for about 3 months were observed to swell diametrically by approximately 7 percent.
No other effect or degradation was found on the rings.
Suspected Buna-N back-up rings remained in 4-way slide t
valves in both red and yellow actuation trains of AE-FV40 and in
yellow train of AE-FV41.
The licensee performed an evaluation and documented it in a justification for continued operation. Although
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the licensee considercd the MFIVs operable, the licensee continued weekly 10 percent stroke tet.ts on the MFIVs.
In addition,
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approximately every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the licensee monitored the frequency of hydraulic pump operation as well as a check for hydraulic oil leakage.
The if tensee stated that the rings in the 4-way slide valves would be changed duri n the refueling outage scheduled to start March 0, 1990.
The licemee also stated that, most likely, the material errer in the back-up rings was nat the major cau.e of the slide salve unreliability, and that the licensee'r task for:e was continving to pe'rform root cause analysis.
On January 31, 1990, the licensee replaced a 4-way slide valve on red train of AE-TV40.
The trend of the time it took AE-FV40 to perform the 10 percent stroke tests had been increasing, The valve subsequently passed its 10 percent stroke test with a decreased stroke time. The inspectors noted that although the slide valve was replaced on the day of the weekly stroke test, the test was delayed i
until after the slide valve was replaced.
It appeared to the inspectors that an opportunity for additional performance data was lost.
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b.
During the review of WR 05117-89, the inspector noted that there was no documentation of the perfomance of a safety evaluation (SE)
regarding removal of the essential service water valve pit missile shield. The essential service water valve pit missile shield provides the protection from natural phenomena described in the Wolf Creek Cenerating Station USAR Sections 9.2.1.2.1.1 and 9.2.1.2.3.
Because the licensee did not declare the associated train of essential
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service water out.of service, a 10 CFR 50.59 review of the change was required. Licensee personnel stated that they did not believe it was required to do SEs for these missile shields nor for other missile shields (e.g., the DG fuel oil storage tank shields).
The licensee
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was infomed that failure to perform an SE was a violation of 10 CFR l
Part50.59(482/9002-01). The inspector's evaluation of the safoty significance of the failure to perform the SE identified:
The missile shield was off for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or less which neant the
probability of a tornado generated missile causing damage was not significantly increased.
The work request limited missile shield removal to one train at a
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time.
During lifting, the missile shield was not allowed to travel
above a maximum height and was net allowed over the opposite train.
In case of inclenent weather, the missile shield would be
replaced.
The licensee performance of this work while at full power did not place the unit in an unsafe cohfiguration; however, 10 CFR Part 50.59 requires that this determination be made and documented before making changes to the facility.
This violation is a repeat of Vichtion 482/8907-01, dated March 15, 1989, in which the licensee installed a temporary jumper across one cell of safety-related Bettery NK 12 without performing a safety evaluation, 6.
Review of Licensee Event Reports (LERs)
(92700)
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During this inspection period, the inspectors performed followup on Wolf Creek LERs. The LERs were reviewed to ensure that:
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Corrective action stated in the report has been properly completed or
work is in progress.
- Response to the event was adequate.
Response to the event met license conditions, commitments, or other
applicable regulatory requirements, i
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reporting requirements.
- Generic issues were identified.
The LERs discussed below were reviewed and closed:
!86-043, " Uncertainty of Proper Valve Operation Caused by
Indeterminate Actuator Wiring." During an NRC erd roreental
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qualification inspection, certain discrepancies c identified (NRC Inspection Report 50-482/86-18).
The licensee determined that some i
of the discrepancies were reportable to the NRC and issued this LER.
The discrepancies were closed in NRC Inspection Report 50-482/87-25,
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I paragraph 3.
This LER, which should have been closed in NRC Inspection Report 50-482/87-25, is closed.
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86-056, " Improper Method For Verifying Flux Doubling Setpoint." This LER concerned a failure to properly perform a TS requi. red surveillance on the flux doubling alarm. Because of an error contained in vendor supplied information, the procedure verified that the alarm was actuated by a multiplication factor of 2.16 instead of the intended factor of 2.0.
This surveillance was performed every refueling outage and was supposed to verify that the alarm setpoint had not drifted.
The licensee corrected the procedure and confirmed that the
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setpoint had not drifted. Because the setpoint was initially set
correctly and was found to still be set correctly, the error did
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t not result in the flux doubling alarm being incapable of performing its intended function. The licensee updated the vendor manual, verified with the vendor that there were no other known similar errors, and revised the affected site procedures.
This LER is
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closed.
,86-066, " Condition Prohibited By Technical Specifications." During a
startup following a refueling <>utage, the licensee allowed two of
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four Sr.fety injection accumu17. tors to go below minimum water level.
TS 3.5.1 requires the licensee to have fcur accumulators with, among other requirements, at least 6,122 gallons (23,172 litens) of torated water and, because of leakage past the isolation check valves, two accumulators dropped below this value. The licensee declared both accumulators inoperable ana within 15 minutes placed the unit in a mode in which all accumulators did not need to be operable (i.e.,
RCS pressure less than 1000 psig (6.89X10(6) pascal)).
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accumulator check valves are tested for leakage prior to returning the
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reactor to service; however, during the time in which the RCS pressure is being increased, leakage can occur.
The licensee's corrective action was to minimize the time spent with low dif ferential pressure across the valves and to verify that the valves were leak tight during subsequent testing.
During the return to service following the three outages which have occurred since this event, the licensee has experienced minimal leakage and through procedural guidance and training, has avoided repeating the error. This LER is close !
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86-067, " Technical Specification Violation Caused By Surveillance Procedure Errors." The licensee determined that two TS surveillances were not being properly performed. One involved verifying that fire dampers properly isolated during a halon actuation and the other
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dealt with verifying that control room ventilation dampers repositioned within the required time limits. The licensee modified the deficient procedures which allowed these errors to occur and then
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t properly tested the involved dampers. All dampers actuated as required. The licensee's corrective action was to review ventilation tests to ensure that any other missed surveillances were identified.
Additional missed TS surveillences were later identified by both the licensee and the inspectors. This resulted in comment in the
systematic assessment of licensee performance (SAlp) report and the
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licensee performing a 100 percent review of all TS surveillances to l
ensure that they met requirements. This LER is closed.
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87-002 and 87-001, Revision 1. " Discovery of Breached Fire Barrier
Seal." This LER dealt with the discovery by the licensee of a
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breached piping penetration seal in the auxiliary building. The
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licensee attributed the breach to a personnel error in removing, but
not replacing, the fire barrier during maintenance.
Later LERs, including 87-10, identified other deficiencies in fire barriers.
I The licensee eventually replaced 100 percent of the wall fire barriers.
Fire penetrations were labeled and extensive training was carried out.
See NRC Inspection Reports 50-482/87-10 and -89-15 and LER 87-010 for additional details.
This LER, which should have been closed when Unresolved Item 482/8710-02 and LER 87-010 were closed, is closed.
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87-009, " Violation of Technical Specification 3.7.6 - Control Room
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Positive Pressure Test Failure." A QA auditor identified a missing shim on a control room pressure boundary door. The licensee
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performed a pressure test and verified that with the shim missing, air leakage increased such that only 0.21 inches (52.2 pascal at i
68'F (20'C)) of water positive pressure could be maintained instead of the required 6.25 inches (62.2 pascal at 68'F (20'C)).
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Inspection Report 50-482/87-03, paragraph 4, discussed thir. event and listed the reasons why a safety hazard did not exist and wny a Notice of Violation was not being issued. This LER, which should have been closed at that time, is closed.
- 87-017, "Two Reactor Trips During Power Reductions Required by
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Technical Specifications." Because of degraded control rod logic cabinet cards, two separate reactor trips occurred during troubleshooting of rod control malfunctions. All required engineered safety features operated as designed. While the licensee did not determine the root cause, the licensee perceived that electronic components failed becauw of cabinet temperatures were above expected values. The licensee replaced the failed components and temporarily
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added additional cooling to the rooms. A permanent plant modification later added additional cooling and redirected the air flow. This LER is closed.
7.
Followup on a Previously Identified NRC Item (92701)
(Closed) Open Item (482/8827-07): Licensee Guidance on the Proper Selection of Grounding Devices - During the maintenance team inspection
!B conducted in September of 1988, an inspector evaluated the electrical maintenance group.
The inspector observed that there was a lack of guidance on the selection of the proper size of certain grounding devices.
This open item was written to track the licensee's resolution of the inspector's observation.
In October of 1989, the licensee revised their safety / accident prevention manual to include a table which clearly identified the required size of ground device. The inspector which identified the observation reviewed the table and determined it adequate to resolve the concern. This open item is closed.
8.
Exit Meeting (30703)
The inspectors met with licensee personnel (denoted in paragraph 1) on January 31, 1990. The inspectors summarized the scope and findings of the inspection. The licensee did not identify as proprietary any of the information provided to, or reviewed by, the inspectors.
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