IR 05000254/1987002
| ML20210R337 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/09/1987 |
| From: | Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20210R290 | List: |
| References | |
| 50-254-87-02, 50-254-87-2, 50-265-87-02, 50-265-87-2, NUDOCS 8702170386 | |
| Download: ML20210R337 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION.
REGION III
Reports No. 50-254/87002(DRP);50-265/87002(DRP)
Docket Nos. 50-254; 50-265 Licenses No. DPR-29; DPR-30 Licensee: Commonwealth Edison Company Post Office Box 767
' Chicago, IL 60690 Facility Name: Quad Cities Nuclear Power Station, Units 1 and 2
Inspection At: Quad Cities Site, Cordova, IL
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Inspection Conducted: November 30, 1986 through January 31, 1987 Inspector:
A. D. Morrongiello Approved By:
M.
ef 7/9/h7 Projects Section IC Date
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Inspection Summary Inspection on November 30, 1986 through January 31, 1987 (Reports No.
254/87002(DRP); 50-265/87002(DRP))
Areas Inspected: Routine, unannounced inspection by the resident inspector of actions on previous inspection findings; operations; radiological controls; emergency preparedness; security; refueling / outages; quality assurance; quality control; administration; routine reports; LER review; regional requests; training; and independent inspection.
Results: One violation was identified in one area (Failure to follow procedure - Paragraph 1.f(1)(g)). No violations were identified in the remaining areas inspected.
The violation is considered significant in that it resulted in a reactor scram on low condenser vacuum.
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DETAILS 1.
Persons Contacted
- R. Bax, Plant Manager T. Tamlyn, Production Superintendent T. Lihou, Operating Engineer
- R. Robey, Technical Services Superintendent
- M.,Kooi, Compliance Coordinator
- D.'Gibson, Quality Assurance
- Denotes those present at the exit interview on January 30, 1987.
The inspectors, through direct observation, discussions with licensee personnel, and review of applicable records and logs, examined the areas stated in the inspection summary and accomplished the following inspection modules.
37700 Design Changes and Modifications 42700 Procedure Review 60710 Refueling Activities 61726 Monthly Surveillance Observations 62703 Monthly Maintenance Observations 71707 Operational Safety Verification 71710 ESF System Walkdown 81072 Access Control - Packages 90713 Review of Periodic and Special Reports 92700 Onsite Review of LERs 92701 Followup - Information Notices, Part 21 Notices 92702-Followup - Violations 92705 Followup - Regional Requests 93702 Onsite Followup of Events 92703 Followup - Confirmatory Action Letters The inspector verified that activities were accomplished in a timely manner using approved procedures and drawings and were inspected / reviewed as applicable; procedures, procedure revisions and routine reports were in accordance with Technical Specifications, regulatory guides, and industry codes or standards; approvals were obtained prior to initiating any work; activities were accomplished by qualified personnel; the limiting conditions for operation were met during normal operation and while components or systems were removed from service; functional testing and/or calibrations were performed prior to returning components or systems to service; independent verification of equipment lineup and review of test results were accomplished; quality control records and logs were properly maintained and reviewed; parts, materials anc'
equipment were properly certified, calibrated, stored, and or maintained as applicable; and adverse plant conditions including equipment malfunctions, potential fire hazards, radiological hazards, fluid leaks, excessive vibrations, and personnel errors were addressed in a timely manner with sufficient and proper corrective actions and reviewed by appropriate management personnel.
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Further, additional observations were made in the following areas:
a.
, Action on Previous Items
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(Closed) Violation 254/86002-02(DRP): This violation represented a repeat event, specifically failing to report a reactor scram that was not pre-planned. Discussions were held with the licensee to clarify the meaning of " pre-planned sequence of events."
No further actions are needed.
b.
Operations (1) Unit 1 At the beginning of the inspection period Unit I was at full power. At various times during this period the unit operated on Economic Generation Control (EGC).
On December 5, at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, a unit shutdown commenced to upgrade certain drywell splices to meet Environmental Qualification (EQ) standards. On December 9, during startup, the unit scrammed on low condensor vacuum (see LER 86038 this report).
At 1345 on December 28, 1986, a drywell high level alarm was received on Unit 1.
At 1456 hours0.0169 days <br />0.404 hours <br />0.00241 weeks <br />5.54008e-4 months <br /> 650 gallons were pumped from the drywell, corresponding to a leak rate of approximately 12~gpm, which exceeded the 5 gpm limit for unidentified leakage of reactor coolant into primary containment specified in Technical Specification 3.6.0.1.
At 1505 an Unusual Event was declared and a load reduction was initiated to comply with the Technical Specification Action statement to place the unit in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
At 1550 the ENS notification was made tn NRC headquarters.
At 1607 the Resident Inspector was notified. At 2130 the source of the drywell leakage was discovered to be Reactor Building Closed Cooling Water leaking from the IF Drywell cooler. At 2330 the closed cooling water to the IF Drywell cooler was isolated, restoring the drywell leakage to approximately 1 gpm.
The Unusual Event was terminated at this time.
For the remainder of the report period the unit was either at full power or on EGC.
(2) Unit 2 j
One scram occurred during the portion of this report period in which the unit was in a refueling outage,and is discussed in the outage section of this report.
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On January 22 at 0615 the Unit 2 mode switch was placed in Run ending the refueling outage that commenced on October 11, 1986.
On January 22, 1987, at 0905 hours0.0105 days <br />0.251 hours <br />0.0015 weeks <br />3.443525e-4 months <br /> Unit 2's Reactor Core Isolation Cooling (RCIC) was declared inoperable due to an oil distribution problem discovered during startup testing.
Unit 2's High Pressure Coolant Injection (HPCI) successfully passed its Technical Specification operability test at 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />. The Unit entered into a seven day' Limiting Condition for Operation (LCO) due to RCIC being inoperable. Additionally, the Unit began a shutdown on January 23 at 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br /> due to increasing reactor water conductivity. Leaking main condenser tubes were plugged.
RCIC was repaired by providing a larger return line to the oil pump suction and repositioning a thermocouple in the return line for one of the bearings. The unit went critical at 2052 on January 25, 1987.
RCIC was again declared inoperable at 0350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br /> on January 26, 1987, because the oil slinger ring was not moving and proper oil volume in the outboard bearing was not achieved. HPCI successfully passed its Technical Specification operability test at 0715 hours0.00828 days <br />0.199 hours <br />0.00118 weeks <br />2.720575e-4 months <br />.
The RCIC outboard bearing and slinger ring were replaced and the orifice to the bearing was cleaned.
On January 26, at 1935 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.362675e-4 months <br />, RCIC successfully passed its Technical Specification operability test. At 2030 hours0.0235 days <br />0.564 hours <br />0.00336 weeks <br />7.72415e-4 months <br />, HPCI was removed from service for an INP0 recommended fast start injection test.
Prior to this the relief valves for auto blowdown were successfully tested. On January 27, at 0135 HPCI failed this special test due to a faulty relay. The relay, which worked intermittently, would, when working correctly, allow the Motor Speed Changer (MSC) to move to the High Speed Stop(HSS).
The problem with the relay was traced to two contacts not making up. The relay pickup voltage and those contacts were readjusted.
HPCI successfully passed a post-maintenance test and an overspeed test. The special fast start test was repeated and was successfully performed and HPCI was declared cperable according to Technical Specifications at 1515 hours0.0175 days <br />0.421 hours <br />0.0025 weeks <br />5.764575e-4 months <br />. At 1926 hours0.0223 days <br />0.535 hours <br />0.00318 weeks <br />7.32843e-4 months <br /> RCIC was manually initiated for a special fast start test (also an INP0 recommendation).
RCIC achieved rated flow at design pressure in 31.9 seconds. This is 1.9 seconds longer than the manufacturer's design specifi-cation stated in the FSAR.
RCIC was declared inoperable.
RCIC's purpose is to provide cooling water to the reactor core when the reactor is isolated from the main condenser coincidently with a loss of the reactor feedwater system.
It can be started manually or automatically (along with HPCI)
on a low-low water level in the reactor. Credit for RCIC is not given in the mitigation of design basis accidents and when RCIC auto initiates HPCI also initiates and provides a few of at least 5000 gpm. At 2041 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.766005e-4 months <br />, after adjusting the governor valve, RCIC met the test criteria with a time of
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27.3 seconds. At 2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br />, RCIC Technical Specification operability' tests were completed and RCIC as declared.
operable..The modification of.RCIC will be tracked as an Open-Item (254/87002-02(DRP)and 265/87002-01(DRP)). The
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unit went to approximately 50 percent power and completed the startup test program.
ior the remainder of the report period the unit was at full power or on E.G.C..
(3) Both'
t During plant tours of Units 1 and 2, the' inspector walked down.
the accessible portions of the Standby Liquid Control System and performed the applicable portions'of the-Inspection Procedure
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71710 "ESF System Walkdown."
No violations or deviations were identified.
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-Outages
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(1) Unit 1 On December 5, 1986, the unit shutdown for a planned unit outage to upgrade drywell penetration splices to EQ standards (see LER 86037 this report).
(2) Unit 2 On December 19, 1986, while performing trip checks on the Bus 23-1 to TR28 breaker, Unit 2 scrammed. The cause of the scram was the opening of the Bus 29 to Bus 28 crosstie breaker. This breaker opened when the trip test was performed and functioned as expected, i.e. it prevents parallel feeding of Bus 28 by Bus 23-1 and 24-1.
The scram signals were Loss of RPS (Reactor Protection System) to trip Channel A and B, and D Main Steam Line Monitors going downscale to trip Channel B.
(The ESS feeding "B" and "D" Main Steam Line Monitor was also fed from Bus 28 during this event). This electrical lineup was
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established due to maintenance work on Bus 23 and is not a typical lineup. Unit 2 is in a refueling outage.
(3) Both This outage represented a marked improvement in licensee
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performance compared to previous outages.
It was conducted
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a with fewer errors and fewer Essential Safety Features (ESF)
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actuations than previous outages. Several areas, however,
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could use improvement. For example, the areas of communica-
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tions and coordination have been discussed with the licensee.
I The licensee has already initiated actions to achieve better
performance in the areas mentioned.
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No violations or deviations were identified.
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' Maintenance
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The following maintenance activities were observed / reviewed:
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(1) Observed Electrical Maintenance personnel performing EQ
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inspection on 4Ky breakers on Unit 2.
.(2) Observed Instrument Maintenance personnel calibrating Main Steam. Pressure Sensor on Unit 2.
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Noviolationsbrdeviationswereidentified.
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Surveillance The following surveillance activities were observed / reviewed:
(1). Observed portions of Unit 1 startup after EQ Outage (rod
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pulls, turbine warming, and start of Reactor Feed Pump).
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(2) Observed portions of Emergency Core Cooling System Test.
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(3) Observed portions of RCIC and HPCI tests during Unit 2's
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startup.
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i (4) Observed cynchronizing Unit 2 generator to the grid.
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(5) Observed portions of rod scram timing on Unit 2.
(6) Observed testing of Automatic Depressurization System on Unit 2.
No violations or deviations were identified.
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LER Review I
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Unit 1
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(a)
(0 pen) LER 86032, Revision 00: Underexcitation Relay
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Causes Trip of 1/2 Diesel Generator When Starting RHR Pump.
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On November 8, 1986, Units 1 and 2 were both in the
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SHUTDOWN mode of operation. Due to a modification being installed in the station electrical switchyard, Transformer
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12 (Unit-1 reserve auxiliary transformer) had to be isolated and removed from service. At 0408 hours0.00472 days <br />0.113 hours <br />6.746032e-4 weeks <br />1.55244e-4 months <br /> the 1/2
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Diesel Generator (DG) was started and loaded to Bus 13-1
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to provide power to Unit 1 during the modification work.
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Subsequently the 1A Residual Heat Removal (RHR) pump was
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started to provide shutdown cooling flow to the Unit I reactor. This resulted in the 1/2 DG feed breaker to Bus
13-1 tripping, causing a loss of power to Bus 13-1 and i
associated Reactor Protection System Bus A which resulted
in closure of a portion of the Group II isolation valves i
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due to a loss of control power.
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This event was determined to be caused by the trip of the underexcitation relay.
It was subsequently disarmed and the 1/2 DG did not trip when the sequerce of events was repeated.
This event is still under investigation by the System Planning Department because the relay was determined to be not out of calibration and its circuitry was also correct.
This LER will remain Open pending submittal of a supplementary LER.
(b)
(Closed) LER 86034, Revision 00:
Failure of I-2301-5 Valve Packing Causing HPCI to be Declared Inoperable.
On November 17, 1986, at 2045 hours0.0237 days <br />0.568 hours <br />0.00338 weeks <br />7.781225e-4 months <br />, Unit I was in the
"RUN" mode at 70 percent of rated core thermal power. An Equipment Attendant (EA) on routine rounds discovered a severe packing leak on the High Pressure Coolant Injection (HPCI) system 1-2301-5 motor operated valve. This valve is a normally open steam isolation valve designed to close under Group IV isolation conditions (High Steam Flow, High Area Temperature, and Low Reactor Pressure).
The 2301-5 valve was imediately isolated and taken out of service for repairs. This caused HPCI to be inoperable.
While HPCI was inoperable to repair the packing leak on steam supply valve 2301-5, the LPCI mode of RHR, both Core Spray subsystems, and the RCIC system were all proven operable.
Appropriate NRC notification was made at 2155 hours0.0249 days <br />0.599 hours <br />0.00356 weeks <br />8.199775e-4 months <br /> to satisfy the requirements of 10 CFR 50.72.
No further actions are necessary.
(c)
(Closed) LER 86028, Revision 00 and 01: Unit 1 Reactor Core Isolation Cooling Trip Throttle Valve Tripped Closed.
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On October 3,1986, at 1055 hours0.0122 days <br />0.293 hours <br />0.00174 weeks <br />4.014275e-4 months <br />, Unit I was in the RUN mode at 79 percent of rated power. At this time, the
Unit 1 Reactor Core Isolation Cooling (RCIC) turbine trip throttle valve tripped closed, and therefore RCIC was declared inoperable. Operating personnel sent to investigate the problem could not determine the exact cause for the valve tripping closed. The only activity in progress at the time was the performance of HPCI system operability tests.
It should be noted that when the RCIC trip throttle valve tripped the HPCI Flow Rate Testing Surveillance (Q0S
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2300-SI) was in progress.
This surveillance requires
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vibration monitoring of the HPCI system. All vibration readings were within the acceptable range.
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Although this is considered to be an isolated event, the trip throttle valve latching procedure was discussed in the weekly operator meeting on October 23, 1986, and will also be emphasized in the 1987 Equipment Operator (EO)
and Equipment Attendant (EA) requalification training.
It should be noted that HPCI was run several times prior to and subsequent to this event with no adverse affects on RCIC.
No further actions are necessary.
s (d)
(Closed) LER 86033, Revision 00: Control Room Panels -
Inadequate Mounting.
This was followed by a regional based inspector and closed in Inspection Report 254/86019 (DRS).
No further actions are necessary.
(e)
(Closed)LER86037. Revision 00: Drywell Penetration Butt Splices Failed to Remain Intact While Undergoing Qualification Testing.
On December 5, 1986 Unit 1~was in the RUN mode at 100 percent core thermal power and Unit 2 was in the REFUEL mode for a refueling outage. At 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br />, the station was notified that during qualification testing performed by Wyle Laboratories, the drywell penetration butt splices failed to remain intact. The testing was intended to verify that the butt splices were adequate on Unit 1 until its next scheduled refueling outage.
Upon failure of the qualification testing, Unit 1 began shutting down and appropriate notifications were made. At 0910 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.46255e-4 months <br /> on December 6, 1986 cold shutdown was reached on Unit 1.
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The unqualified drywell penetration butt splices of both units were taped over per QMP 100-60 (Scotch Brands Tapes E.Q. Installation Instructions). This type of tape splices has been demonstrated to be capable of withstanding a j
simulated LOCA at Wyle Laboratories.
A region based inspector reviewed the licensee's procedure for taping these splices. While this LER is closed, a Part 21 Report is being tracked by DRS as an Unresolved Item.
l No further actions are necessary regarding this LER.
(f)
(Closed) LER 86036, Revision 00: Gaseous Effluent Particulate Samples Lost by Offsite Laboratory.
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The July, 1986, monthly particulate samples for the Main Chimney and the Unit 1 and Unit 2 Reactor Vent Systems were lost. The samples were collected and sent to Teledyne Isotopes - Midwest Facility to be analyzed. When the results for the month of July were not reported in a timely manner, Teledyne Isotopes was contacted. The inquiry revealed that they could not locate the particulate samples. The missing data was projected by averaging recent sample results.
Based on operating conditions, and review of noble gas, iodine and particulate effluent releases for the previous month there was no reason to assume a large increase in particulate effluent activity.
The corporate group at Nuclear Services Technical, which sets up vendor services,has been informed of this event.
This group will review bid proposals for offsite laboratory analysis and will require that the vendor have implemented a program to track samples.
No further action is needed.
(g)
(0 pen) LER 86038, Revision 00: Low Vacuum Scram During Startup Due to Personnel Error.
On December 9, 1986, Unit One was in the process of starting up per QGP 1-1, Normal Unit Startup. At 1733 hours0.0201 days <br />0.481 hours <br />0.00287 weeks <br />6.594065e-4 months <br />, the reactor scrammed due to low condensor vacuum.
All other systems functioned as expected during this event and release rates were within Technical Specifications.
The reason for condensor low vacuum was that the primary steam jet air ejectors were not valved in.
The ejectors were not valved in due to personnel error. Specifically, QGP 1-1 refers to QOP 5400-1, "Offgas System Startup,"
which requires lineup of the primary steam jet air ejectors at 400 psig reactor pressure. This step was not performed. This LER will remain Open pending implementation of the licensee's corrective actions.
Failure to adhere to the procedures for unit startup and offgas lineup is considered to be a violation as noted in the Appendix (254/87002-01(DRP)).
(2) Unit Two (a)
(Closed) LER 86018, Revision 00:
Engineered Safety Feature Actuation Due to Radiographic Testing.
On November 13, 1986, Unit 2 was shut down for the end of cycle eight Refueling and Maintenance outage. At 1645 hours0.019 days <br />0.457 hours <br />0.00272 weeks <br />6.259225e-4 months <br />, the A & B Reactor Buildings Ventilation (RBV) system radiation monitor tripped on a high radiation level of 4 milliroentgen (mR) per hour. This closed the RBV exhaust
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dampers, initiated the B Standby Gas Treatment (SBGT)
system, and caused the control room ventilation to go on 100 percent recirculation, as designed.
The root cause of this occurrence is that it was not anticipated that the radiography on the SBLC system would trip the RBV radiation monitors located in the Turbine Building. As required by station procedure QAP 900-5 (In-Plant Radiography - Required Notifications and Actions)
the radiographer notified the appropriate station personnel of the impending X-rays. The RBV radiation monitors are located in the Turbine Building, and the Shift Engineer did not reasonably anticipate the X-ray radiation would penetrate the thick concrete common wall between the Reactor and Turbine Buildings to such a degree that it would cause the RBV radiation monitors to trip.
It should be noted that radiography of welds on the SBLC had been accomplished earlier in the week without any ESF actuations.
No further actions are required.
(b)
(Closed) LER 86017, Revision 00: Linear Indication on Reactor Recirculation System Weld.
On October 11, 1986, Quad Cities Unit 2 was shutdown for refueling. On November 5, visual inspection revealed a Recirculation weld area with water seeping from a small crack. The cause of this occurrence is postulated as being intergranular stress corrosion cracking. A supplemental report will be submitted when all inspections and repairs have been completed.
This issue was tracked by regional based inspectors and was closed in Inspection Report 254/86019(DRS) and 265/86014(DRS).
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Review of Routine and Special Reports The inspector reviewed the monthly performance report for the month of December.
No violations or deviations were identified, h.
Part 21 Followup (1) A Part 21 report was issued by Valcor Engineering regarding the failure of 17-7 stainless steel springs in Valcor valves due to hydrogen embrittlement. The licensee was informed of this notification. Through a search of purchase orders and stockroom inventory, it was cencluded that said valves
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-e are not in use at the Quad Cities site. This also closes
out IE Information Notice No.~86-72:
" Failure of 17-7 Ph Stainless Steel Springs in Valcor Valves due to Hydrogen
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Embrittlement."
No further actions are necessary.
(2) A Part 21 was issued by Commonwealth Edison regarding the qualification of AMP Butt splices.
In order to address a deficiency identified during.an EQ Audit of Dresden Station by NRC (see Inspection Reports 237/86013 and 249/86015) splices taken from Quad Citten were sent to Wyle Labs for additional EQ tests. The result.s of these tests indicated that the splices
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were not EQ.
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Corrective action at Quad Cities consisted of shutting down Unit 1 (Unit 2 was in a Refueling Outage) and repairing the splices ia both units using EQ qualified materials.
This item is being tracked by the Division of Reactor Safety as an Unresolved Item.
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Regional Request The: licensee was notified about a valve related problem that occurred at Pennsylvania Power and Light's Susquehanna Plant. They discovered that Houghto #620 lubricant attacked and degraded aluminum in valves-manufactured'by Automatic Valve Corporation. This problem was
pointed out by Quality Control and that lubricant is not used at Quad
Cities.
In response to the General Electric Service Information letter (SIL)
445 entitled " Intermediate Range Monitor (IRM) Fuse Failure, the i
licensee conducted special tests which showed that if that fuse were L
blown two annunciators functioned to alert operators of a malfunction and 265/86012(DRP))y (documented in Inspection Report 254/86013(DRP)
in the IRM circuitr At Quad Cities there does not appear to be a
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problem of minor voltage surges causing unnecessary or excessive fuse failures.
Since there have been very few cases of power supply fuse
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failures over the life of the plant, the licensee did not upgrade the
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size of the power supply fuse.
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No further actions are necessary.
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IE Information Notice Followup
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(Closed) IE Information Notice No. 86-99: Degradation of Steel Containments.
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In November, 1986, Oyster Creek discovered some erosion of their
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Mark I containment drywell.
In response to this concern Quad Cities inspected the drywells of both units. The drywell thickness was
measured in eight locations and at each location at least three areas l
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were checked by ultrasonic testing (UT) performed by certified Level II UT inspectors. No thinning degradation was detected on either unit.
These results were discussed with a Region based inspector who normally reviews NDE related work at this site. The licensee is aware that a Bulletin or Generic Letter may be issued on this subject and that further analysis may be required.
No further actions are necessary.
k.
Procedure Review The following procedures were reviewed.
QEP 310-T3, Prioritized Notification Listing.
QEP 310-1, Initial Notification.
QEP 310-T7, Simplified Emergency Notification Schedule for Site Emergency.
QEP 310-T8, Simplified Emergency Notification Schedule for General Emergency.
QEP 530-S2, Monthly Test of the NRC Health Physics Network.
QMP 100-3, Fire Prevention for Welding and Cutting.
QIP 100-12, Backfilling Reactor Instruments Sensing Lines.
No violations or deviations were identified.
1.
Independent Inspection Shortly after the incident at Surry Nuclear Power Station, Commonwealth Edison commissioned Nutech to select portions of piping between the condensate booster pump and the feedwater check valves to be examined for erosion / corrosion wall thinning.
Sample locations were selected according to piping geometry, flow, and usage factor.
All pipi g inspected satisfied code design requirements (ANSI /ASME B31.1-80.
Additionally it should be noted that Commonwealth Edison initiated an Action Item Request in August,1986 to identify water lines that are vulnerable to erosion in order to start an inspection program of these lines.
No violations or deviations were identified.
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Confirmatory Action Letter Followup A confirmatory action letter was issued to Quad Cities after results from an NRC administered requalification program for licensed operators determined that the. Quad Cities requalification program was unsatisfactory. As of February 28, 1987, a licensed SR0 advisor will no longer be required since all shift assignments will be filled by license holders who have passed one of the recent NRC administered requalification exams, or who had passed an NRC license exam since October 1, 1985, or who have passed an accelerated requalification program for people who have not passed a recent NRC administered requalification exam or passed an NRC license exam since October 1, 1985. On January 16, 1987, _the licensee met with the Regional Staff to discuss plans for implementing the long term requalification improvement program. While agreement was reached on the direction this program should take, this letter will remain Open pending implementation of the program and a satisfactory rating of the requalification program by DRS.
No violations or deviations were identified.
2.
Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspectors, and which involve some action on the part of the NRC of licensee or both. The open item disclosed during the inspection is discussed in Paragraph 1.b.
3.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance, or deviations. Two unresolved items are discussed in this report in Paragraphs 1.h and 1.f.(1)(e).
Both of these items are discussed further in reports by the Division of Reactor Safety.
4.
Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1)
throughout the inspection period and at the conclusion of the inspection l
on January 30, 1987, and summarized the scope and findings of the l
inspection activities.
The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any such
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documents / processes as proprietary.
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