IR 05000245/1988012
| ML20154J728 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/15/1988 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154J726 | List: |
| References | |
| 50-245-88-12, GL-83-08, GL-83-8, IEB-85-003, IEB-85-3, NUDOCS 8809230122 | |
| Download: ML20154J728 (13) | |
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i U.S NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-245/88-12 s
Docket No.
50-245 License No.
OPR-21 Licensee:
Northeast Nuclear Energy Company Facility:
Millstone Nuclear Power Station, Unit 1 Inspection At: Waterford, Connecticut Dates:
July 19 through August 29, 1988
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Inspectors:
William Raymond, Senior Resident Inspector Lynn Kolonauski, Resident Inspector Michael Boyle, Licensing Project Manager, Office of Nuclear Reactor
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Regulation Reporting Inspector:
Lynn Kolonauski
Approved by:
Eb O.kOA h 9/ t'//es E. C. McCabe, Chief, Reactor Projects Section 1B Date Inspection Summary: Inspection from July 19 to August 29, 1988 (Report No.
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F0 2T9/88-12)
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Areas Inspected: This inspection included routine NRC re.ident inspection of pre-
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viously identified items, plant operations, physical security, the emergency operating procedure (EOP) improvement plan, temporary instructions 2515/73, 2515/95,
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and 2515/96, the justification for continued operation (JCO) with increased service and emergency service water (SW and ESW) temperatures, inservice inspection plans i
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for the control rod drive mechanism (CRDM) nold down bolts, maintenance, surveil-
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lance, MEPL (Material Equipment Parts List) program implementation, the dual-role l
SRO/STA (Senior Reactor Operator / Shift Technical Advisor) issue, and Jafety com-
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mittee activities.
Results: The inspection identified no unsafe plant conditions.
Further licensee
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and/or inspector followup is warranted on: (1) emergency operating procedure im-provements (Section 6.0), and (ii) MEPL program implementation (Section 12.0).
J 8809230122 000915 PDR ADOCK 05000245 G
PNU
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TABLE OF CONTENTS i
t PAGE 1.0 Persons Contacted....................................................
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2.0 Summary of Facility Activities.......................................
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3.0 Status of Previous Inspection Findings...............................
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3.1 (Closed) VIO 88-11-01, "MEPL Verifications Not Performed in Accordance with ACP-QA-4.03B".................................
3.2 (Closed) UNR 88-33-03, "Post Accident Sampling System (PASS)
Surveillance".................................................
l 3.3 (Closed) IFI 50-245/87-28-01; 50-336/87-24-01 "Measurement
Control Evaluation Non-Radiological Chemistry"................
l 4.0 ;*cility Tours and Operational Status Reviews........................
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4.1 Safety System Operability.......................................
l 4.2 Review of Plant Incident Reports (PIRs).........................
5.0 P hy s i c a l S e c u r i ty....................................................
6.0 Emergency Operating Procedures improvement Plan......................
7.0 NRC Temporary Instructions 2515/73, 95, and 96.......................
8.0 JC0 for High SW and ESW Temperatures.................................
I 9.0 CRDM Hold Down Bolts......................................
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10.0 Maintenance..........................................................
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11.0 Surveillance.........................................................
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12.0 MEPL Program Implementation..........................................
13.0 Update to Dual-Role SRO/STA Issue....................................
14. 0 Pl a nt Op e ra t i on s Rev i ew Conai tt e e....................................
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15.0 Management Meetings...............................................
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DETAILS 1.0 Persons Contacted J. Stetz, Unit 1 Superintendent R. Palmiert, Operations Supervisor C. Wargo, Acting Maintenance Supervisor W. Vogel, Engineering Supervisor
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M. Bigiarelli, Assistant Engineering Supervisor
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J. Quinn, Assistant Engineering Supervisor D. Wilkens, Chemistry Supervisor l
G. Cornelius, Generation Mechanical Enginsering The inspectors 21so contacted other members of the Operations, Health Physics, Instrumentation and Control, Production Test, Maintenance, and Engineering
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departments.
2.0 Summary of Facility Activities
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j Millstone 1 operated at full power throughout the inspection period with the i
j exception of short term power reductions for routine surveillances and cor-
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rective maintenance.
I On July 19 and August 9, power was reduced to 90*4 to allow main condenser
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backwashing and tube plugging in the
"D" waterbox, l
On August 19 at 3:00 a.m., a power reduction (to 65'4) btgan for MSIV 10*.
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closure testing, main condenser tube plugging and turbine stop valve testing, j
While closing the #1 turbine stop valve, reactor pressure increased from 1010 f
to 1025 psig, power increased from 3S*4 t.
97*4 and all six average power range
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i monitors (APRMs) alarmed on high flux.
..e control valves opened to control reactor pressure.
The remaining stop valves tested normally (see report detail 11.0). All other surveillances were successfully completed, and full
j power was restored at 5:30 a.m.
Power was again reduced (to 65%) at 6:04 a.m.
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due to the development of additional tube leaks in the "O" condenser.
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turbine stop valves were retested again, with no abnormal results.
The test-
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ing and tube plugging were completed and the unit returned to full power by 5:30 p,m.
However, conductivity in the "D" waterbox again increased and power was reduced to 75% by 9:20 p.m. for additional tube plugging.
Full power was reached at 5:45 a.m. on August 20.
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On August 26, turbine stop valve testing was conducted at 60's and 80*. power i
with no abnormal results.
Power was further reduced (to approximately 45%)
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to remove the "B" recirculation motor generator set from service to allow
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replacement of a failed lube oil bearing pressure regulator. A power increase
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commenced at 3:35 a.m. on August 27.
Full power was restored at 5:30 a.m.,
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and was maintained for the remainder of the inspection period.
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3.0 Status of Previous Inspection Findings (93702)
3.1 (Closed) VIO 88-11-01, "MEPL Verifications not Perfomed in Accordance
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with ACP-QA-4.03B"
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i By letter dated July 28, 1988, the NRC requested the licensee to respond j
to a violation of ACP-QA-4.038, "MEPL for In-Service Nuclear Generating
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Facilities." By letter dated August 26, 1988, the licensee reported that the root cause for the nonadherences was the fai' lure to adequately em-phasize performance of timely MEPL verifications.
Retraining of all
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Millstone 1 engineers authorized to perform MEPL evaluations by August i
30, 1988 was proposed to prevent recurrtnie.
The inspector verified this commitment by review of the training attendance records.
The licensee 41so stated that increased management attention and increased reporting
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of MEPL evaluation status by the Millstone 1 Engineering supervisor would
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In addition, the in-l j
spector learned that the Millstone 1 engineering supervisor plans to use i
i a personal computer based tracking system weekly.
The inspector found the corrective actions to be adequate and will periodically assess their s
l effectiveness in maintaining procedural ccmpliance. This item is closed.
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3.2 (Closed) UNR 88-33-03. "post Accident Sampling Syste n (PASS) Surveillance"
i As discussed in NRC inspection report 50-245/88-11, the inspector found
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that Millstone 1 Chemistry Procedure SP 8008, "PASS Reactor Coolant
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Operability Test," did not a..plicitly direct the chemistry technician
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to request that Onerations inspect the system for leakage during the
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acquisition of li,.c PASS samples.
The inspector reviewed the revised i
procedure (Rev 1, issued 7/25/88), and found adequate direction has been
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incorporated
'nis item is closed.
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3.3 (C_losed)IFI 50-245/87-2b O1; 50-336/87-24-01, "Measurement Control
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Evaluation Non-Radiological Chemistry"
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J Upon completion of the analyses of water samples (spiked samples) by the j
y iteensee and Brookhaven National Laboratory, a statistical evaluation
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I was made. The analytical comparisons produced acceptable results, as i
listed below.
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Millstone Units 1 & 2 Split Samples with Spikes Analyte Matrix Spike Millstone 1&2 B_rookhaven Fluoride (ppb)
14.7+/-0.1 14.3+/-0.1
7.2+/-0.2
<10
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Chloride (ppb)
Steam Gene:ator
16.1+/-0.1 14.7+/-0.9
8.4+/-0.2 9.5+/-2.7
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Sulfate (ppb)
16.1+/-0.1 11.3+/-0.5
8.1+/-0.1 7.9+/-2.1 Hydrazine(ppb)
Steam Generator None 20.5+/-0.7 19.2+/-0.4
1 Iron (ppb)
Feedwater El 750+/-10 731+/-15 E2 533+/-6 465+/-16
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Copper (ppb)
Feedwater El 760+/-10 700+/-0 E2 503+/-6 450+/-0 Nickel (ppb)
Feedwater El 730+/-0 720+/-16
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E2 477+/-6 750+/-20
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Chromium (ppb)
Feedwater El 733+/-6 750+/-20
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E2 493+/-23 500+/-0 Boron (ppm)
Standby Liquid None 24,921+/-140 26,240+/-90 Control Tank
This item is closed.
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4.0 Facility Tours and Operational Status Reviews (71707)
j Control instrumentation was inspected for proper functioning, correlation be-
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tween channels, and conformance with Tecnnical Specifications (TSs). Alarm
r conditions in effect and alarms received were reviewed and discussed with the operators; the inspector fwnd the operators to be cognizant of plant condi-
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l tions and indications. The inspector observed prompt and appropriate operator i
response to changing conditions. Shift turnovers were observed and found to be thorough and in conformance with ACP 6.12. "Shift Relief Procedure "
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Operating logs and Plant Incident Reports (PIRs) were reviewed for accuracy l
and adherence to station procedures.
Posting, contros, and the use of per-l sonnel monitoring devices for radiation, contamination, and high radiation
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areas were inspected.
Plant housekeeping controls were observed, including l
i control of flammable and other hazardous materials. Backshift inspections l
of the control roea were conducted on August 13 at 4:30 p.m. and August 25
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at 7:00 p.m.
All shif t personnel were found to be alert and attentive to I
their duties.
No unacceptable conditions were identified.
The following j
activities were also addressed.
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4.1 Safety System Operability (71707)
Standby emergency systems were reviewed to determine system t.,perability and readiness for auto atic initiation.
The following systems were re-viewed: feedwater coolant injection, autoiatic pressure relief, low i
pressure coolant injection, core spray, and standby gas treatment.
The i
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status of the control rod drive hydraulic control units, emergency diesel gerarator, gas turbine, and isolation condenser was also inspected.
The
reviews considered proper positioning of major flow path valves, operable normal and emergency power sources, proper operation of indications and i
j controls, and proper cooling and lubrication.
References used for the
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review included the Updated Final Safety Analysis Report, and system
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diagrams and operating procedures.
No inadequacies were identified, j
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4.2 plant Incident Reports (71707)
Selected plant incident reports (PIRs) were reviewed te (i) determine
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of the events, (iii) verify the licensee's response and corrective ac-
tions, and (iv) verify whether the licensee reported the events in ac-cordance with applicable requirements. No inadequacies were identi.fied.
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The following PIRs were reviewed: 1-83-28, 1-88-32 through 41, 1-88-42
(see reactor pressure spike during TSV testing, Section 11.0),1-88-43 and 1-88-44.
The following items also warranted inspector followup,
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i PIRs 1-83-39 and 1-83-40 discuss two separate, unexpected initiations j
of the Standby Gas Treatment System (SBGT) due to false high radiation signals from refueling floo.' channel 34/2 on August 6 and 11.
In both cases, all other refuel floor area radiation monitors (AMs) displayed l
normal readings, and subsequent radiation area surveys indicated general l
area radiation levels of approximately 2 mrem /hr, well below the SBGT initiation setpoint of 100 mrem /hr.
The SBGT trip was reset and normal
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ventilation was restored.
Both events were reported as required by 10 l
I CFR 50.72(b)(1)(vi). After the first initiation, the Instrumentation
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I and Controls (I&C) department inspected and recalibrated the equipment
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associated with ARM channel 34/2. After the second initiation.
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reinspected and then replaced the equipment despite the absence of nega-
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tive findings.
The licensee believed the most probable cause to be a
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faulty power supply in view of the simultaneous receipt of the service
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j water effluent high radiation and the 1-FDS-1 fire alarms. All three actuations were unsubstantiated and the alarm conditions cleared immedi-i ately. ARW. channel 34/2 was returned to servi:e. The inspector had no
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further questions.
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PIRs 1-88-43 and 1-88-44 discuss problems encountered while at about 45*4 power and returning the "B" recirculation pump to service after replacing
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the motor generator (MG) set lube oil bearing pressure regulator on
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August 26. When operations personnel attempted to start the M3 set, the
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M3 set generator lock-out relay actuated after the field breaker closed.
l A second attempt was also unsuccessful.
Production Test personnel de-l termined that an improper time delay setting was causing the generator
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field breaker to close prior to excitation. When repairs were completed j
y approximately nine hours later, operations personnel started the MG set
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but manually tripped it due to a 30*. reactor power spike caused by the
introduction of cold water into the core. No trip setpoint was reached, i
The speed of the "A" M3 set was further reduced before attempting to I
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restart the "B" MG set, which was then successfully restarted.
Power was then increased. The inspector interviewed the personnel associated
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i with the event and determined that their actions were in accordance with the applicable procedure, OP-301, "Reactor Recirculation System." The
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inspector specifically verified that the operators throttled the recire
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pump discharge valve open in accordance with OP-301.
The licensee is evaluating the event and is considering additional procedural direction
when returning a recirculation pump to service during plant operation.
The inspector had no further ouestions, and will routinely inspect OP-301 adequacy in the future.
5.0 Plant Security (81064)
During station tours, the inspector! verified proper implementation of selected aspects of the station security program.
These included site access controls,
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personnel searches, compensatory measures, adequacy of physical barriers, i
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reporting of security events, compensatory measures, and guard force response j
to alarms and degraded conditions.
No inadequacies were identified.
6.0 EmergencyOperatingProceduresImprovementPlan(937@
As reported in NRC inspection report 50-245/88-11, the NRC asked the licensee
to sut:,mit a letter detailing proposed corrective actions based on the E0P
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i inspection findings as presented in the June 30, 1988 exit meeting for the
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NRC team inspection of Millstone 1 E0Ps.
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As detailed in a July 29, 1988 letter, the licensee plans short-term upgrades
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to the current, Revision 2 E0Ps by October 15, 1988.
Implementation of Re-vision 4 E0Ps after a full validation and verification process is planned three months af ter the end of the April 1989 outage or by September 1,1959, whichever is later.
The NRC provided preliminary deficiencies to Millstone
i 1 management for incorporation 13 short term E0P improvements.
NRC review
of the licensee's proposed long term action plan is continuing.
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7.0 NRC Te_mporary Instructions - 2515/73, 95, and 96 ',25573/25595/25596)
The inspectors implemented the following temporary instructions (tis), in f
j whole or in part, in reviewing the associated licensee activities, i
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Tl 2515/73 "Motor-Operated Valve Common Mode Failures During Plant Transients
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Due to Improper Switch Settings."
In November 1985, the NRC issued Bulletin j
85-03 to ensure that switch settings on certain safety-related motor-operated t
valves are set and maintained so as to accor.modate the most severe loading
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expected during design basis events.
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6 The inspectors reviewed the licensee's initial response to Bulletin 85-03 (action item "e" of the talletin) dated June 11, 1936 and concluded that the licensee's selection of applicable safety-related valves, the maximum expected differential pressures, and the valve operability assurance program were ac-i ceptable and in conformance with IEB 85-03 guidelines.
The inspectors will review ths, final licensee response (action item "f" of r
Bulletin 85-03) in a future inspection.
l TI 2515/95, "BWR Recirculation Pump Trip: Multi-Pls.nt Action Item C-02."
This TI requires the inspector to verify the installation of a recirculation pump trip under cnnettions indicative of an anticipated transient without a scram l
(ATWS) event.
Section 7.6.1.4 of the Millstone 1 Updated Final Safety An-
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alysis Report (UFSAR) indicates that the ATVS system trips both recirculation i
pump motor generator field breakers on high reactor pressure or on low-low
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reactor water level.
The inspector reviewed P&ID 25202-28055 Sheet 1 and
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found the above statement to be accurate.
i TI 2515/96, "Mark I Drywell Vacuum Breeker Modifications: Multi-Plant Action Item 0-20."
This Tl requires the insMetor to verify that, the licensee has J
modified the drywell vacuum breakers if such action was required to meet the guidelines of Generic Letter (GL) 83-08.
The NRC issued GL 84-08 to request
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licensees to evaluate the structural integrity of their vacuum breakers to assure that they could withstand chugging and condensation oscillation loads during a loss of coolant accident (LOCA). On October 16, 1986, the NRC issued
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a safety evaluation on this topic anti concluded that the analyses performed by the licensee to predict vacuum breaker impact velocities and the resulting
i stresses were performed using acceptable methodologies and that the existing vacuum breakers were adequate.
Therefore, no modifications were necessary to meet the guidelines of GL 83-08.
The inspector had ro further questions
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on this item.
8.0 JC0 for High SW and E5W Temperatures
Sustained high area temperatures caused Long Island Sound, the ultimate heat
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i sink for Millstone 1, to increase in temperature during the inspection period.
The intake temperature reached a high of 73.3 degrees F en Augusa 13.
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design limit for both service and emergency service water (SW/ESW) is 75 F l
as specified in the Millstone 1 Updated FSAR; the Millstone 1 TS do not ad-
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dress high intake temperatures.
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The licensee prepared a Justification for Continued Operction (JCO) assuming
a maximum intake temperature of 77 F.
The analysis, which was mainly quali-j tative, considered the effects of the temperature increase on the post-LOCA e
containment response and on all safety related systems and components cooled
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The inspector was in attendance when the Plant Operations Re-l view Committee (PORC) approved the JC0 and unidentified no apparent deficien-
i cies. How2ver, in reviewing the licensee's procedure for safety evaluations I
(ACP-QA-3.08), the inspector questioned whether an integrated safety evalu-J
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ation was also required because the change had direct impact on systems credited in the design basis accident analyses. The change also appeared to
meet the criteria of 10 CFR $0.59 as an unreviewed safety question.
Discussion between the Region I projects section chief and the licensee's Vice President, Nuclear and Environmental Engineering, produced information that
the ability to meet the design basis at an elevated temperature was based on the existing condition of plant equipment.
The ability to permanently change the design basis limits had not been established, and the equipment suppliers had not certified their equipments for ongoing ability to meet the design basis at a higher temperature.
Consequently, the analysis would have to be redone to certify the systems for a one-time or longer SW/E5W temperature above 75 F in the future.
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Ina;much as the design limit was not exceeded at Millstone 1 and the licenses l
must appropriately justify each deviation from the present design limits or i
justify a permanent change to those limits, this matter is not being pursued
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further at this time, r
9.0 CRDM Hold Down Bolts The licensee was informed, througi a General Electric Rapid Customer Informa-l tion Service Information Letter (L' RICSIL 019, issued 5/19/88), that inser-l vice inspections (ISI) of the contr ol rod drive mechanisms (CRDMs) at Susque-j hanna indicated stress corrosion surface cracking of the CRCM hold down bolts.
j The inspector reviewed the actions of Millstone 1 in response to the S!L.
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The licensee plans to visually inspect all bolts removed during CRCM rebuild activities conducted in the April 1939 outage.
If cracks are identified by visual inspectier, the litensee will conduct magnetic particle inspections i
of the bolts, and decide further action based on the results obtained.
The
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inspector had no further questions and will follow this activity during the
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10.0 Maintenance (62703)
The inspector observed and reviewed selected aspects of the following Jafety
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related maintenance activities, including procedural adherence, obtainment
of required administrative approvals and tagoutt prior to work initiation, t
proper QA/QC involvement and personnel protection, and verification of system j
retest prior to return to service. No inadequacies were noted.
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Torus to Drywell Vacuum Breaker Position Switch Adjustment, on July 28
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11.0 Eurveillance (61726)
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The inspector observed portions of the following surveillances for ccnduct in accordance with current approved procedures, for test result compliance
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j with technical specification and administrative requirements, and fer defi-
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ciency correction in accordance with administrative requirements. The in-
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spector noted that the surveillance teams displayed thorough coordination and rigorous procedural adherence.
One sarveillance activity which required fur-
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ther inspector 'ollowup is described below.
No inadequacies were identified.
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SP 4088, teactor Vessel High Pressure Scram Functional Test" on July
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SP 40SH, ' 1rywell High Pressure Scram Functional Test" on August 2.
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SP 40SF, tTSV Closure Functional Test" on August 26.
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On August at 3:00 a.m., a reactor power reduction to 55*4 began to allow i
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perforrant e of the 10*4 MSIV closure surveillance.
The turbine stop valves
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(TSVs) were tested at 3:17 a.m., with reactor power at SS's. While closing i
the #1 TSV, reactor pressure increased from 1010 psig to 1025 psig, reactor i
power increased from SS*4 to 97?o, all six APRMs (Average Power Range Monitors)
I alarmed on high flux, and the turbine control valves opened (from 40*4 to 50*4)
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I to cottrol pressure. When the test switch was released, the #1 TSV opened l
j in a norr.,al manner. GEK specifications for TSV closure in the test moh are t
i between 12 and 16 seconds.
TSV #1 closed slightly faster, at 11.1 seconds.
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1evels ranging from 60*o to 80*4.
No abnormal responses were observed.
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personnel installeo additional monitoring in an effort to determine the cause i
i of the pressure spike.
A multipoint chart recorder monitored flow in each i
j of the main steam lines, reacter pressure, turbine control valve cam position,
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first stage reactor pressure, and TSV position.
No conclusive information
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j has been identified. At present, the licensee plans to continue with the (
entended data collection scheme during weekly TSV testing.
The inspector will
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I continue to follow the licensee's investigation.
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12.0 MEPL Program implementation I
While investigating the licensee's corrective actions taken in response te the violation cited in NRC inspection report 50-245/83-11, t,he inspector l
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learned of a second potential concern identified by the licensee that involved i
adherence to the procedures governing maintenance of the Material Equipment
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i Parts List (MEPL).
I The Production Maintenance Maintenance Management System (PKMS) is an auto-l rated computer system which stores inforniation on specific components in-I stalled in Northeast Utilities (NU) generating plants. On 0:tober 1, 1984,
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the licensee began to use FMVS to generate authorized work orders (AW0s) for Millstone 1 components.
For the NJ nuclear generating plants, FRMS includes
J an indication of the quality assurance (QA) standards that must be applied l
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when maintenance is performed on these components. The PKMS QA indicator
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field will contain either "Y", "N", or "U",
meaning that the component is j
either under QA program controls, is not under QA program controls, or has
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an undetermined QA status, respectively.
If a "U" is listed, then a MEPL evaluation must completed prior to the initiation of non-QA work.
NUSCo (Northeast Utilities Service Company) has been given the responsibility l
l to audit the FRMS QA indicators against the MEPL at intervals equal.o the
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unit's refueling cycle length. During a meeting with Millstone 1 management, i
the inspector learned that NUSCo had conducted this audit for the first time
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in 1938 because computer printouts suitable for audit purposes were unobtain-
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i able until that time.
The inspector also learned that preliminary audit re-
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suits indicated a significant 'evel of discrepancies in the PKMS data base.
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l This was the first time the plant was informed about the status of the audit, t
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which had been in p ogress for several months.
It appeared that NUSCo did not act in a timely manner, in view of the potential safety implications of these discrepancies. Millstone 1 requested immediate follow-up, and a teaa l
of NNEco (Northeast Nuclear Energy Con.pany) engineers reviewed the audit re-
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sults to determine their significance.
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Preliminary results of changes made to PKMS 6ata entries in 1986, which en-compass approximately 13,000 entries, indicated approximately 3200 discrepan-
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cies. The results of the Millstone 1 investigation revealed that the vast l
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majority of these discrepancies were in che conservative direction; that is, j
the QA field contained a "U" when a MEPL determination had been completed,
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or the field contained a "Y" when it should have contained an "N".
Sixty-one i
d discrepancies appeared to be nonconservative; that is, the QA field centsined l
an "N" instead of a "Y".
Millstone 1 engineering determined that all but two resulted from audit errers where the associated MEPL and FRMS identification numbers were mismatched, and the associated MEPL listings were actually "N".
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Furthe Millstone 1 followup of the remaining two discrepancies identified i
that one item involved work on a reactor water cleanup system relief valve.
In this case, there were no material substitutions and no installation of
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I non-QA material in a QA apolication.
The final discrepancy involved the Vital t
AC M3 set.
Ir. 1984, the PPNS data indicated "N".
Sometime later, the MEPL i
l evaluation was reperformed to declare the Vital AC MG set as QA. A number
l of automated work orders ( AW0s) were writter for the M3 set prior to the in-
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corporation of the revised evaluation into :ie MEPL.
The MEPL was actually I
J revised in June 1937, but MG set work had been completed af ter the revision l
using one of the AW0s written prior to the revision, Millstene 1 engineering
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will write a nonconfwmance report (NCR) which the inspectors will review.
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This is an unresolved item (ES-12-02) pending NRC review.
The inspectors learned that the NUSCo audit of 1986 FRMS data cid not provide
an accurate assessment of the current PRMS system status because a great nue ber of MEPL evaluations have been performed and entered since 1986. Also, l
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one of the reasons the MEPL-PVMS audit function was so far behind was that
l, computer printouts containing only QA field changes to PPMS were unob uinable
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until recently.
The original audit of the 1936 entries was performed on
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printouts of changes made to any of the PMMS fields, not just those with QA field changes.
This caused the audit to take longer and resulted in the re-petition of many of the perceived discrepancies.
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Current licensee plans for resolution of this topic include:
Current FMMS printouts (as of June 1,1988) will be used for each nuclear l
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unit.
The licensee estimates that the audits will take approximately I
four weeks per unit, once the PMMS printout is received, which is ex-
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pected shortly.
The audit will proce&d in the following order: MP-1, i
All MEPL changes made from Janeary 1,1986 to June 1,1983 will be in-i
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cluded in the audit.
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All differences between the MEPL and PMMS will be noted, and preliminary
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results will be discussed with each unit. This notification will be im-t l
mediate if the error is nonconservative.
The units will disposition the
differences identified by the audits.
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In reviewing this issue, the inspectors concluded that the licensee's proce-dures for maintenance and use of the MEPL are generally adequate.
The failure l
to conduct audits of the PMMS QA indicators vioStes ACP-0A-4.03B (NEO 6.10),
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"Use of the PMMS Data Base to Indicate Quality Assurance Program Applicability."
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However, no enforcement action is being taken at this time because the licen-see identified the failure and took sufficient action to determine actual MF,PL
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program status and resolve the identified discrepancies.
In addition, the
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failure had no significant safety impact as evidenced by the licensee's in-vestigation results described thus far. Also, the event is not reportable per NRC requirements and could not have been prevented by the corrective actions taken for a previous violation of NRC requirements.
l 13.0 U date on the Dual-Role 0/STA Issue l
f As initially reported in NRC inspection report 50-245/88-02, the NRC had in-
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dicated that the dual-role Shift Supervisor / Shift Technical Advisor (SS/STA)
did not meet the Commission policy statement for engineering expertise on i
shift.
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In a mero dated August 9,198S, the Secretary of the Comission infor.med the
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NRC Executive Director for Operations that the Commission had voted to allow
the thirty people who had already graduated from the Memphis State University and Thames Valley State Technical College programs, as well as t.he eleven candidates enrolled prior to October 1,1987, to serve as SSs/ STAS upon suc-cessful completion of their studies. Compliance with that restraint will be routinely evaluated. Otherwise, this matter b closed as an inspection item.
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14.0 Plant Operations Review Committee (4070,0)
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- l The inspector attended Plant Operations Review Committee (PORC) meetings on l
July 20 and 28, and August 3, 12, 13, 17, 25. and 29.
Technical Specifica-i tions 6.5.1 requirements for conaittee quorum were met.
The meeting agenda included reviews of Plant Incident Reports, plant design modifications, pro-l cedure revisions, and new procedures.
The inspector noted that the committet
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discharged their functions in accordance with the requirements of TS 6.5.1.
No inadequacies were identified.
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t 15.0 Management Meetings (.30703)
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i Periodic meetings were held with station manegement to discuss inspection
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I findings during the inspection period.
A summary of findings was also dis-l cussed at the conclusien of the inspection. No proprietary information was covered within the scope of the inspection.
No written material wa-provided
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to the licensee by the inspectors.
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