IR 05000322/1986014
| ML20211B370 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 10/07/1986 |
| From: | Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20211B369 | List: |
| References | |
| 50-322-86-14, NUDOCS 8610170221 | |
| Download: ML20211B370 (26) | |
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U.S. NUCLEAR REGULATORY COMMISSION Region I REPORT NO.
50-322/86-14 DOCKET N0.
50-322
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LICENSE NO.
NPF-36 LICENSEE:
Long Island Lighting Company P. O. Box 618 Shoreham Nuclear Power Station Wading River, New York 11792 INSPECTION AT: Wading River, New York INSPECTION CONDUCTED: July 16-August 31, 1986 INSPECTORS:
John A. Berry, Senior Resident Inspector Clay C. Warren, Resident Inspector b
APPROVED:
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/0/7/#(
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. R. Strosnider, Chief, Reactor Projects Date'
Section IB, Divisien of Reactor Projects
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SUMMARY: During this inspection period, July 16, 1986 through August 31, 1986, l
297 hours0.00344 days <br />0.0825 hours <br />4.910714e-4 weeks <br />1.130085e-4 months <br /> of direct inspection effort were performed by the Senior Resident Inspector and the Resident Inspector. The facility was made critical on August 3,1986 and was shutdown on August 31, 1986. During the period of critical operation, the licensee performed training startups and completed the low power testing program. The inspectors observed the conduct of surveillance and maintenance activities, and other routine resident inspection activities.
No unacceptable conditions were identified. Ten items were closed as a result of this inspection, and four inspector follow items were
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DETAILS 1.
Status of Previous Inspection Items 1.1 (closed) 82-26-07, Single Drain Valve on H2/02 Analysers This was designated as an unresolved item in NRC Inspection Report 50-322/82-26. The item related to the fact that the four drywell analyzer lines have a single drain valve between the containment atmosphere and the reactor building. This does not provide the
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redundancy of containment isolation as required by 10CFR50, Appendix
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A, Criterion 54. NRC Inspection Report 50-322/84-18 updated this inspection item. The report noted that the Shoreham Safety Analysis Report, NUREG-0420, Supplement No. 4, determined that tne present design was acceptable until the first refueling outage. The report noted that when the Shoreham operating license was issued, it would contain a condition that two isolation barriers be installed in series in all instrument lines.
Shoreham operating license NPF-36 contains this requirement as license condition 2.C.11.
Due to the fact that this is a license condition, there is no need to track it as an open item. This item is closed.
1.2 (closed) 84-38-03, Non-timely Resolution of Audit Findings This unresolved item was opened in NRC Inspection Report
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50-322/84-38 due to a lack of timely resolution to NRB Audit H-83-1.
At the time this was opened, the licensee committed to expedite the closeout of audit findings, and to resolved future audit findings in a timely manner.
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The issue of non-timely resolution of audit findings was identified again in a special inspection of allegations relating to activities in the Radiochemistry area.
(See NRC special Inspection Report 50-322/86-03 for details.)
This issue is one part of pending NRC Enforcement Action as a result of that special inspection.
Resolution of the issue will be tracked with that Enforcement Action. This item is closed.
1.3 (closed) 84-46-06, Routing of Sprinkler System Control Cables in the RCIC & HPCI Areas of the Reactor Building During a team inspection of fire protection, documented in NRC Inspection Report 50-322/86-46, this item was opened due to the fact that the control cables for the preaction sprinkler systems on elevatien 8' of the Reactor Building were routed through the area protected by the sprinkler systems. This meant that if a fire should occur, the cables could be damaged, thereby preventing the sprinkler systems from actuating and suppressing the fire.
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In NRC Inspection Report 50-322/85-23 this item was updated in that it was verified that although cables were not routed through the sprinkler system area, the preaction systems for the HPCI & RCIC pump areas may not operate properly unless certain modifications were made. The licensee committed to implement these modifications.
The modifications include relocation of some of the heat detectors that actuate the preaction system of the HPCI pump area. These de-tectors were located above the heat collector shield of the sprinkler heads. This heat collector prevented the rising heat flux from timely reaching the detector, thus delaying the operation of the sprinkler system.
The licensee also committed to relocation of one of the manual pull stations that actuates the sprinkler system of RCIC pump area. The pull stations, as originally located, may not have been accessible in the event of a fire. Within this area the licensee agreed to relocate the heat detector conduit to a level above the heat detectors. The. conduit as originally located might have been damaged, thus preventing the operation of the RCIC preaction sprinkler system.
By memorandum dated August 8, 1986, the licensee informed the Senior Resident of the completion of these modifications. The inspector reviewed Station Modification SM85-012, and Document Change Notice 85-12-15 which implemented these modifications. The inspector veri-fied completion of the work as indicated by the signatures of the installer, cognizant site engineer, and QC witness and by independent field walkdown. The inspector verified satisfactory completion of post-work functional tests of the HPCI and RCIC Turbine Area Fire Suppression Systems.
The inspector also reviewed the Safety Evalua-tion associated with this modification and had no questions.
This item is closed.
1.4 (closed) 84-46-19, No Emergency Lighting Available in Specific Locations As a result of a special team inspection conducted on December 3-7, 1984, documented in NRC Inspection Report 50-322/84-46, a deviation was written relating to the lack of Emergency Lighting in specific locations.
Spacifically, the deviation stated:
"In the course of walking through the procedure for Shutdown From Outside the Control Room, SP 29.022.01, several areas were discovered which require local operator actions from outside of the Remote Shutdown Panel and do not contain any 8-hour
emergency battery pack lighting.
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The specific locations, all in the Reactor Building, are:
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Elevation 150' North Side - area of the Reactor Building Closed Loop Cooling Water (RBCLCW) Pumps and also the Fuel Pool Cooling Pump area.
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Elevation 78' at the PRV local instrumentation panels.
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Elevation 63' at RHR Valve Room for the vent valves 01V-3124 and 01V-3125 associated with RHR valves MOV-47 and MOV-48.
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Elevation 40' NW corner near the elevator for the condensate transfer loop fill valve (04V-0016).
In addition, the installed battery pack in the 101 Diesel-Generator Room cannot be aimed at the control panel because of its location.
This is a deviation from the licensee commitment, to install self-contained 8-hour battery pack emergency lighting in all areas of the plant which could be manned to bring the plant to a '.fe cold shutdown condition, as documented in Supplement 1 to the SER, Section 9.5.4."
NRC Inspection Report 50-322/85-23 subsequently noted that the first four locations do not require emergency lighting. This left only the lighting problem in the 101 diesel generator room.
Inspection Report 85-23 noted that the licensee committed to the resolution of this item by the initiation of Design Output Package 85-037, "D.C. Emergency Lighting Modification D.G. 101", and that this item would be closed after the relocation had been physically complete.
By memorandum dated August 5, 1986, the licensee informed the inspector of the completion of Station Modification 85-051, which implemented 00P-85-037. The inspector reviewed Station Modification 85-051, Maintenance Work Request 85-3481, and the Station
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Modification Review Form documenting modification closeout.
The inspector also verified the completion of the work in the field.
The inspector had no further questions.
This item is closed.
1.5 (closed) 85-06-02 Complete Emergency Plan Requalification Training of Two Health Physics Technicians and Two Watch Supervisors This item was opened to track the completion of requalification training in emergency planning for four individuals who had not completed all training by the end of calendar year 1984. The inspector verified that this training had been completed. This item is closed.
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1.5. (closed) 85-30-01 Valve Feedback Linkage Failure During.the 5% test program in 1985, the licensee experienced
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mechanical failure of the feedback linkage on one of the two startup
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level control valves. This failure allowed excessive feedwater flow
to the reactor vessel, causing a momentary power-spike greater than
5%. (See NRC Inspection Report 50-322/85-30 for details of this 3 -
event.) Inspector follow item 85-30-01 was opened at that time to
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track the licensee's evaluation of-the event, and subsequent corrective action.
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On August 1, 1986 the licensee submitted a report on this event to the NRC. This report was submitted as LER 86-027, a voluntary j
report submitted to meet a commitment made in the Startup Test
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Report.
The LER detailed the event description, incident analysis,
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and corrective actions (both short.and long term).
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A brief review of the event follows:
On August 15, 1985, several reactor power excursions occurred.
Reactor power was initially 1.7%, the "A" reactor feed pump was in
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I speed control with the startup level controller in automatic and set
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at +37".
The 'B' Electric Hydraulic Control pressure regulator was i
in control with the pressure setpoint at 960 psig. Additional reactor power was required to run the
'B' reactor feed pump for an operation' check, therefore at 18:15, control rod withdrawal was
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commenced to raise reactor power.
i At 19:56, with reactor power at 3.8% and the No. 1 bypass valve post-l tion at 42%, a reactor period alarm was received and reactor power j
was observed to increase to 4.5% by the APRM (Average Power Range
Monitor). Control ' rod withdrawal was stopped at the direction of the l
Watch Engineer.
Personnel were stationed to monitor the EHC bypass valve position. Approximately 5 minutes later, the No. I bypass
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valve opened from 42 to-45% and then rapidly shut to 38%. A slight i
rise in reactor water level was noted. Ten seconds after the bypass
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valve oscillation, a transient similar to that at 19:56 was noted.
The Reactor Operator inserted control rods in an attempt to limit the j
transient. Additionally, the Watch Engineer directed a reduction in i
power until the cause of the transients could be determined. Control
rods were inserted and reactor power was reduced to 2.7%, and stabilized.
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In consultation with the Watch Engineer, EHC System Engineer, and
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on-shift G.E. Representatives, the Operations Manager directed the Watch Engineer to begin a controlled power increase to the point 1.
l where the 'B' reactor feed pump could be started and the cause of
the power transients could be investigated. At this time an EHC system problem was thought to be the cause of the power transients.
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At approximately 21:30 the pressure setpoint was lowered to 900 psig to open the bypass valves and observe pressure regulator operation without having to increase power in the Reactor. While lowering the pressure setpoint it was noted that short, quick operations of the setpoint pushbutton (less than one second) caused rapid bypass valve closure as the button was released. Subsequently, a power transient similar to those observed earlier occurred. The bypass valve closing stroke appeared to be the cause of this power transient.
The Watch Engineer ordered the pressure setpoint reduced with longer, more sustained setpoint pushbutton strokes.
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With the pressure setpoint at 900 psig, a controlled power increase using control rods was initiated.
Reactor power was increased and the pressure regulator setpoint was raised to 920 psig, the normal operational setting for power operation. By 2220, reactor power was stabilized at +3.8%, with pressure setpoint at 920 psig, and the No. I bypass valve 45% open. At 2230 the No. 1 bypass valve was observed to open slowly to 50% and then quickly shut to 40%. A reactor level transient of about 3" was noted, followed shortly by a reactor power transient to 6.3% on the highest observed ApRM channel. At this point, a reactor power reduction was directed by the Watch Engineer.
Reactor power was reduced by rod insertion to 1.5% and stabilized.
A hold on reactor power was then established pending an investigation of the cause of the power transients.
i The licensee's subsequent investigation, as detailed in Inspection Report 85-30, determined that the cause of the event was the failure of a mechanical feedback linkage on the ten inch startup level control valve. The feedback arm was reinstalled with a locking washer, and an inspection of similar valves was conducted. These valves, the minimum flow control valves in the recirculation line for the condensate booster pumps and the other startup level control valves, were discovered to not have locking washers on the feedback arm connection.
Locking washers were installed.
On August 17, 1985 a power increase was performed to evaluate the performance of the startup level and the EHC pressure control system.
The results of the test demonstrated proper operation of the startup level control system and the absence of power transients resulting from reactor water level changes or pressure setpoint changes.
It was concluded that the malfunction of the startup level control valve was the sole cause of the power transients which occurred on August 15, 1985.
The licensee's detailed analysis of GETARS (GE Transient Analysis Recording System) traces and chart recorder was made by the licensee.
This analysis confirmed that the cause of the event was the introduction of excess feedwater to the reactor which decreased the temperature of the water at the recirculation pump suction,
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7 hence introducing positive reactivity to the reactor. The analysis also determined that the maximum neutron power reached during the event was approximately 5.8%.
The review and corrective actions for this event are satisfactory.
The inspector has no further questions, this item is closed.
1.7 (closed) 85-30-02 Electrical / Instrumentation Panels
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This item was opened in August 1985 as a result of an inspector's tour of the reactor building. During that tour the inspector noted that eleven (11) electrical / instrumentation panels were not properly secured in that one or more cover clips were loose or missing.
The inspector met with the licensee to discuss the situation, and they committed to steps to ensure that the situation would be corrected prior to startup.
On May 5, 1986 the Maintenance Engineer sent a memo to the Main-tenance Division Manager documenting a joint QC, Maintenance Inspec-tion of the Reactor Building, primary containment and Control Build-ing.
This tour was to assure that all junction boxes were closed and secured properly and cubicle doon.s were closed. The memo concluded that any deficiencies identified were corrected, and that the mode switch could be placed in startup.
On August 4, 1986 the Resident Inspector conducted an extensive inspection of all elevations of the Reactor Building's secondary containment to assess the status of electrical / instrumentation panel covers. The inspector noted approximately 35 examples of loose cover clips, missing fasteners, missing screws, and other deficiencies. These were noted on both safety-related and non-safety related junction boxes and equipment.
The inspectors immediately brought this situation to the attention of the Plant Manager.
The inspectors were particularly concerned with the impact of these deficiencies on the environmental qualification of the equipment because the plant was operating at the time. The licensee immediately initiated a joint walkdown of the Reactor Building with QC personnel, Maintenance Division personnel and Nuclear Engineering Department personnel to correct the noted deficiencies and evaluate the situation.
The Nuclear Engineering Department (NED) conducted an evaluation to determine what, if any, impact the physical condition might have had on environmental and seismic qualification of safety-related equipment.
For non-safety related equipment, the potential impact on the seismic mounting of this equipment was considered. The reviews also considered the potential of the loose or missing hardware not being discovered and hence not being corrected,
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The licensee's conclusions are as follows:
Non-Safety Related Equipment and Junction Boxes - Loose or Missing Screws:
No effect on equipment qualification.
Non-safety related
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equipment is not within the scope of the equipment qualification program.
Visual inspection of equipment and junction boxes concluded
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that only few fasteners, at most, were missing or loose.
In many instances, the cover was also hinged.
It was determined that a sufficient number of secured fasteners existed to prevent the covers from falling off in a seismic event.
Safety Related Junction Boxes and Equipment:
Potential loose junction box covers resulting from a loose or
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missing fastener would have no effect on environmental qualification because:
(a) Safety related cables within the junction box are qualified for post break operating conditions (PB0C) or LOCA conditions in the absence of a protective enclosure.
(b) Any cable splices within the junction box are environmentally qualified without the need for a protective enclosure.
(c) Terminal blocks located in junction boxes in the Secondary Containment and associated with safety related circuits were reviewed for operability requirements during a PBOC.
Circuits required to be operable and evaluated to be sensitive to a PBOC environment had associated terminal blocks which were located within junction boxes replaced with qualified in-line splices.
(d) Any safety related equipment required to be operable during or after a PB0C was sealed at the specific device or equipment if environmental qualification testing demonstrated a need for such a seal (i.e., the junction box was not relied upon as a moisture barrier for environmental qualification).
It was determined that a sufficient number of secured fasteners existed to prevent any covers from falling off in a seismic event.
The as-found configuration would not have affected the seismic integrity of the box or of equipment contained in the box. This determination was ba:ed on the fact that the junction box cover and
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its fasteners are non-structural, non-load bearing numbers. The presence of some missing or loose fasteners would not significantly alter the dynamic response characteristics of the box assembly.
Category I Motor Control Centers (MCCs):
The MCCs for safety related equipment required for the shutdown of the plant are located in the East and West MCC Rooms at Elevation 112'.
These MCC rooms are environmentally controlled and sealed from the effects of a P80C.
These MCCs were located in these rooms as part of the initial plant design due to a lack of environmental qualification for the MCCs at that time.
Therefore, any loose.
fastener would have no impact on the ability of the MCC to withstand a PBOC environment.
It was determined that the as-found condition would not adversely affect the seismic qualification of the MCCs.
For safety related MCCs in the Reactor Building Secondary Containment which had holes or other openings, the environmental qualification testing qualified this equipnient for a PBOC environment with this configuration.
This qualification program also concluded that the MCC gasketing material did not serve a safety function and was not;re, quired to limit the entry of moisture during a pipe break.
It was concluded that a loose fastener would not affect the MCC environmenta) qualification.
The inspectors independent evaluation of the situation agreed with that conducted by NED.
Subsequent inspection of the Reactor Building confirmed that the situation had been corrected.
The inspectors concluded that based on the NED evaluation, no violation of NRC requirement existed.
However, the inspectors expressed concern that this incident represents sloppy housekeeping practices which potentially could create an environment qualificati,on problem if continued.
Although this inspection item will be closed, the inspectors will be closely monitoring licensee housekeeping practices in the future, especially as they relate to ne proper securing of electrical / instrumentation par - :s.
This item is closed.
1.6 (closed) 85-39-02 I&E Notice 85-82, Seismic Qualification of Diesel Generator Differential Protection Relays This item was opened to track the licensee's actions related to I&E Notice 85-82.
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The Information Notice describes a potentially significant safety l
problem involving the use of General Electric model 12'CFD relays which are not seismically qualified for Class IE service when'in the de-energized state. These relays are used at Shoreham for differential protection of the TDI emergency diesel generators (EDG's). A seismic event or a jarring of the cubicle in which the relays are mounted, while the EDG is not running, could possibly close the CFD relay contacts and lock out the EDG.
The licensee had initially conducted a review of the use of the CFD relay at Shoreham for the TDI's as a result of INPO notice D & MR 117, issued December, 1982.
This review concluded that the CFD relays used in conjunction with type HFA relays would not cause an EDG lockout due to the inherent time delay to energize a HFA coil.
In the case that a lockout did occur, an alarm will appear on the process computer. To provide additional means of alerting the operator of an EDG lockout, DOP 83-119 added a lockout signal to alarms 2005, 2056 and 2106, " Diesel Generator Inoperative", for each EDG.
The use of CFD relays was also reviewed by Stone and Webster for 10CFR21 reportability, in 1985, as a result of INP0 SER 18-84.
It was concluded that a reportable condition did not exist at Shoreham based on the implementation of DOP 83-119 and established maintenance, alarm response and station procedures.
Differential protection for the Colt EDGs consists of a PVD-HFA relay pair, to prevent lockout of the Colts due to spurious activations of the PVD's. General Electric, which is supplying the relays and the cubicle, has been asked to ensure that the PVDs are qualified to the SNPS seismic levels while in the de-energized state.
In addition, a lockout of a Colt EDG will be alarmed on the plant process computer and on the new Colt control board in the Main Control Room.
The licensee concluded that the use of CFD-HFA relay pairs on the TDI and the use of PVD (which is qualified in the de-energized state)-HFA relay pairs on the Colts, combined with positive annunciation of a lockout condition adequately addresses the concerns of the I&E Notice.
The inspectors concur in this conclusion. This item is closed.
1.9 (closed) 86-06-01 Training for QA/QC Personnel in Transportation Requirements This inspection item dealt with the training of personnel conducting QC inspections in the transportation area.
The inspector determined, during NRC Inspection 50-322/86-06, that only one QC inspector had attended any training course in this area. The licensee committed to provide the training to all personnel.
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This training has been completed. This item is closed.
1.10 (closed) 86-06-02 Technical Qualifications of Audit Personnel
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During NRC Inspection 50-322/86-06, in the area of Transportation Activities, the inspector noted that at least four auditors who participated in a May 20-June 25, 1985 QC audit had not received any training in the activity that was audited. This was contrary to Item 2.2 of ANSI N45.2.23-1978. As this was a preoperational inspection, the inspector did not regard this as a violation. The licensee committed to correct this situation.
The licensee has made a programmatic change to the QA Department
Audit program to require that technical specialists be present on all audit teams. This satisfactorily addresses this concern. This
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item is closed.
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2.
Review of Facility Operations 2.1 Plant Status Summary During the period covered by Inspection Report 86-14 the facility was operated at power levels up to 4.5% as allowed by Shoreham Operating License NPF-36. The Licensee conducted retests on equipment modified during the outage period and performed training criticalities for eleven license candidates.
In addition to critical operation during the inspection period, the licensee performed required surveillances and routine maintenance activities.
2.2 Completion of 5% Test Program and Main Generator Synchronization During the period covered by Inspection Report 86-14, the facility
was brought critical for the first time since placing the plant in cold shutdown on October 8,1985 for neutron source replacement and plant modifications. Operation at power was necessary to confirm that the reactor vessel level reference leg modifications, performed during the outage period, corrected the level deviation between channels that was experienced during the low power testing phase (Inspection Report 50-322/85-36) and to perform retests on systems modified or repaired during the outage.
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retest activities the licensee also conducted training startups and synchronized the main generator with the system grid for the first time.
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The reactor was made critical on August 3,1986 at 02:45, with heatup starting on August 5, 1986 after licensee personnel completed perfor-mance of training criticals. On August 6, 1986 the High Pressure Coolant Injection System was run to meet Technical Specification and retest requirements.
The HPCI system performed as designed and i
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satisfied all requirements. The plant remained at approximately 200 psig until August 8, 1986 for HPCI and Reactor Core Isolation Cooling System tuning.
The licensee completed heatup to rated temperature and pressure on August 8, 1986. During this heatup reactor level instrumentation was closely monitored for deviations between instrument channels such as those noted in Inspection Report 85-36. No level deviations were noted.
Operation at rated temperature and pressure continued, with ROIC and HPCI testing and fine tuning constituting the majority of activity, until August 11, 1986 when the unit was shutdown to repair a faulty level transmitter (B21LIS1558).
The reactor was restarted on August 12, 1986 and heatup to rated tem-perature and pressure completed on August 13, 1986.
The licensee performed scram time testing on all control rods to satisfy Technical Specification retests required after rebuilding the hydraulic control unit ASCO solenoid valves. Scram time testing was completed satis-factorily on August 16, 1986.
Licensee personnel began main ge$erator startup procedures on August 17, 1986 when the machine was rolled to rated speed and surveillance tests performed. Upon successful completion of surveillance testing, licensee personnel began main generator exciter checks prior to ex-citation of the main generator.
The initial attempt to excite the generator was unsuccessful due to an incorrect generator relay setting.
There were no further attempts to place the generator on the grid at that time as the plant was shut down due to the approach of Hurricane Charlie.
The reactor was made critical on August 22,. 1986 and brought to rated temperature and pressure on August 23, 1986.
Further turbine testing showed that the relay problem had been corrected and the main generator was excited and placed on the grid on August 26, 1986. The main generator remained synchronized with the grid for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and supplied 10MWE net throughout that time.
The plant was placed in cold shutdown on August 31, 1986. No unacceptable conditions were noted.
2.3 Operational Safety Verification The inspector toured the control room daily to verify proper shift manning, use of and adherence to approved procedures, and compliance with Technical Specification Limiting Conditions for Operation.
Control panel instrumentation and recorder traces were observed and the status of annunciators was reviewed.
Nuclear instrumentation and reactor protection system status were examined.
Radiation monitoring instrumentation, including in plant Area Radiation monitors and effluent monitors were verified to be within allowable limits, and observed for indications of trends.
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distribution panels were examined for verification of proper lineups of backup and emergency electrical power sources as required by the Technical Specification.
The inspector reviewed Watch Engineer and Nuclear Station Operator logs for adequacy of review by oncoming watchstanders, and for proper entries. A periodic review of Night Orders, Maintenance Work Requests, Technical Specification LC0 Log, and other control room logs and records were made.
Shift turnovers were observed on a periodic' basis.
The inspector also observed and reviewed the adequacy of access controls to the Main Control Room, and verified that no loitering by unauthorized personnel in the Control Room Area was permitted. The inspector observed the conduct of Shift personnel to ensure adherence to Shoreham Procedures 21.001.01, " Shift Operations" and 21.004.01, " Main Control Room - Conduct for Personnel".
2.4 Plant and Site Tours The inspector conducted periodic tours of accessible areas of plant and site throughout the inspection period. These included:
the Turbine and Reactor Buildings, the Rad Waste Building, the Control Building, the Screenwell Structure, the Fire Pump House, the Security Building, and the Colt Diesel Generator Building.
During these tours, the following specific items were evaluated:
Fire Equipment - Operability and evidence of periodic
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inspection of fire suppression equipment;
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Housekeeping - Maintenance of required cleanliness levels;
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Equipment Preservation - Maintenance of special precautionary
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measures for installed equipment, as applicable; QA/QC Surveillance - Pertinent activities were being surveilled
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on a sampling basis by qualified QA/QC personnel; Component Tagging - Implementation of appropriate equipment
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tagging for safety, equipment protection, and jurisdiction; Personnel adherence to Radiological Controlled Area rules,
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including proper Personnel frisking upon RCA exit; Access control to the Protected Area, including search
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activities, escorting and badging, and vehicle access control;
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Integrity of the Protected Area boundary.
No unacceptable conditions were identified.
3.
Licensee Reports 3.1 In Office Review of Licensee Event Reports The inspector reviewed Licensee Event Reports (LERs) submitted to the NRC to verify that details were clearly reported, including accuracy of the cause description and adequacy of corrective action.
The inspector determined whether further information was required from the licensee, whether generic implications were involved, and whether the event warranted onsite follow-up. The following LERs were reviewed.86-011 Rev. I:
Ultimate Heat Sink, Accumulation of Sediment.
86-026:
Unplanned Automatic Initiation of RBSVS "A" Train During an Instrument and Controls Surveillance Procedure when a Technician Dropped a Screwdriver in the Panel.
86-027:
Reactor Power Excursions Above 5% due to Failure of the
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Mechanical Linkage Between the Fosition Feedback Arm and the Controller on the Startup Level Control Valve.
86-028:
Non-Environmentally Qualified Jumper Wire Installed in Motor Operated Valves.
86-029:
ESF Actuations due to EPA Breaker Trips while RPS Power was being Supplied by the Alternate Feed Transformer.
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86-030:
Unplanned Automatic Actuations of ESF Systems Caused by Power Spikes on the Grid Voltage due to Thunderstorms.
86-031:
Seismic Monitoring Recorders in Control Building out of Service for more than 30 Days.
86-032:
Unexplained RWCU Isolation While Placing the Filter Demineralizers in Operation.
The inspector will track the Licensee's Engineering Evaluation and Corrective Actions to the events described in LER 86-029 and LER 86-030 (see Sections 6 and 7 of this report for further details).
Response to the event in LER 86-029 will be Inspector Follow Item 86-14-01 and Response to LER 86-030 will be tracked as Inspector Follow Item 86-14-02.
The inspectors review of LER 86-028 will be l
covered in detail in Section 8 of this report.
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l The inspector has no further concerns.
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4.
Monthly Surveillance and Maintenance Observation
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4.1 Surveillance Activities i
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.The inspector observed the performance of various surveillance tests
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to verify that; the surveillance procedure conformed to technical specification requirements, administrative approvals and tagging
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requirements were reviewed and approved prior to test initiation,
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j testing was accomplished by qualified personnel, current approved procedures were used, test instrumentation was currently calibrated,
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limiting conditions for operation were met, test data was accurately g and completely recorded, removal and restoration of affected
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components was properly accomplished, and tests.were completed
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within the required Technical Specification frequency.
During the inspection period the following surveillance activities were observed for licensee performance:
24.202.03 High Pressure Coolant Injection Pump Operability and Flow Rate Test 24.202.05 High Pressure Coolant Injection System Actuation Response
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Time Test
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24.606.01 Average Power Range Monitor Functional Test j
24.127.01 Reactor Protection System, Turbine Overspeed Protection i
System Functional Test - Turbine Stop Valves 24.127.02 Reactor Protection System Turbine Overspeed Protection System Functional Test - Turbine Control Valves
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l 24.127.03 Turbine Bypass Valve Functional Test i
i 24.119.01 Reactor Core Isolation Cooling Pump Flow Rate Test i
l 4.2 Maintenance Activities l
l The inspector observed the conduct of various maintenance activities j
throughout the inspection period. -During this observation, the inspector verified that;. maintenance activities were conducted
within the requirements of the plant's administrative procedures and
technical specifications, proper radiological controls were implemented and observed, proper safety precautions were observed,
and that activities which have the potential to impact plant j
operations are properly coordinated with the control room.
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The inspector observed the following licensee maintenance activities
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during the inspection period:
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Replacement of BUNA-N Seals in all Hydraulic Control Unit 117 and 118 Valves.
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Inspection and Replacement of Unqualified or Unidentifiable Wire in Linitorque Operators.
Troubleshooting of Inoperable Intermediate Range Monitor E.
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No unacceptable conditions were noted.
5.
Review and Followup of I&E Notices, Bulletins and Generic Letters 5.1 I&E Notices The inspector reviewed notices issued by the Office of Inspection and Enforcement during the inspection period.
Review was to determine; if the subject of the notice was applicable to the Shoreham Nuclear Power Station, and if followup of the licensee's action was required by the inspector.
The following IE Information Notices were received during the period covered by Inspection Report 86-14:
IE Notice 86-13 Supplement 1: Standby Liquid Control Squib Valves Failure to Fire.
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IE Notice 86-55:
Delayed Access to Safety-Related Areas and
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Equipment During Plant Emergencies.
IE Notice 86-57:
Operating Problems with Solenoid Operated Valves at Nuclear Power Plants.
IE Notice 86-58:
Dropped Fuel Assembly.
IE Notice 86-60:
Unanalyzed Post-LOCA Release Paths.
IE Notice 86-61:
Failure of Auxiliary Feedwater Manual Isolation Valve.
IE Notice 86-62:
Potential Problems in Westinghouse Molded Case Circuit Breakers Equipped with a Short Trip.
IE Notice 86-64:
Deficiencies in Upgrade Programs for Plant Emergency Operating Procedures.
IE Notice 86-65:
Failure of ITT Barton Model 580 Series Switches During Requalification Testing.
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IE Notice 86-66:
Potential for Failure of Replacement AC Coils Supplied by the Westinghouse Electric Corporation for Use in Class IE Motor Starters and Contactors.
IE Notice 86-68:
Stuck Control Rod.
IE Notice 86-69:
Spurious System Isolations Caused by the
Panalarm Model 86 Thermocouple Monitor.
IE Notice 86-70:
Potential Failure of all Emergency Diesel Generators.
IE Notice 86-71:
Recent Identified Problems with Limitorque Motor Operators.
IE Notice 86-72:
Failure 17-7 PH Stainless Steel Springs in Valcor Valves to Hydrogen Enbrittlement.
IE Notice 86-73:
Recent Emergency Diesel Problems.
IE Notice 86-74:
Reduction of Reactor Coolant Inventory because of Misalignment of RHR Valves.
IE Notice 86-75:
Incorrect Maintenance Procedure on Transversing Incore Probe Lines.
IE Notice 86-76:
Problems Noted in Control Room Emergency Ventilation Systems.
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IE Notice 86-77:
Computer Program Error Report Handling.
The licensees response to IE Notice 86-13 was detailed in Inspection Report 86-05 and the inspector has no further questions.
IE Notice 86-71:
Recent Identified Problems with Limitoruge Motor Operators is covered in detail under Section 8 of this report.
Facility review of IE Notice 86-74 is of special interest to the inspector and will be followed as Open Item 86-14-03.
Licensee response to all remaining Information Notices will be reviewed as part of routine resident inspections.
5.2 Generic Letters Generic Letter 86-06:
Implementation of TMI Action Item II.K.3.5., " Automatic Trip of Reactor Coolant Pumps".
Generic Letter 86-10:
Implementation of Fire Protection Requirements.
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Generic Letter 86-11:
Distribution of Products Irradiated in Research Reactors.
Generic Letter 86-12:
Criteria for Unique Purpose Exemption From Conversion.
Generic Letter 86-13:
Potential Inconsistency Between Plant Safety Analyses and Technical Specifications.
Generic Letter 86-14:
Operator Licensing Examinations.
Generic Letters 86-06, 86-33 and 86-11 have been reviewed and are not applicable to Shoreham therefore no Licensee response is necessary.
The inspector will review the Licensee's response to the remaining items.
5.3 Bulletins IE Bulletin 86-02:
Static "0" Ring Differential Pressure Switches.
IE Bulletin 86-02 described potential problems with Static "0" Ring Incorporated Switches Model 102 or 103 and requested that licensee's report on the usage of these switches in applications important to safety, as defined in 10CFR50.49(b).
The Licensee's response to this Bulletin indicates that no switches of this type are used in applications important to safety at Shoreham. The inspector has reviewed the Licensee's documentation in this area and has no further questions.
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6.
Safety System Actuations from Electrical Protection Assembly Breaker (EPA Breaker) Spurious Trips On four occasions during the period covered by Inspection Report 86-14, Reactor Protection System (RPS) Half Scrams, Nuclear Steam Supply Shutoff System (NSSSS) Half Isolations, Reactor Water Cleanup System (RWCU)
Isolation and Reactor Building Standby Ventilation System / Control Room Air Conditioning System initiations were received when the Alternate Power Supply EPA Breakers tripped.
At the time of each of the four trips, the plant was in Mode 4 with one Division of RPS power being supplied by the Alternate Power Supply while surveillance testing was being performed on the RPS motor generator sets.
The "A" Motor Generator Set was out for three events, the "B" Motor Generator Set was out for the other.
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Instrument and Control Technicians checked the setpoints of the EPA Breakers and found them to be within tolerance.
Technicians also confirmed that the Voltage Regulating Transformers were perfnrming as
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designed; this work was performed in conjunction with a manufacturer technical representative.
At this time the cause of the trips is not known and further testing is planned prior to placing the RPS buses on the alternate supply. A test procedure that will fully instrument the system has been written but not yet approved.
The inspector has reviewed the licensees actions thus far and has found them satisfactory, however final disposition of this item has not been made therefore this item will be made Open Item 86-14-01.
7.
Unplanned Automatic Actuations of ESF Systems Due to Lightning Strikes At 02:19 on August 11, 1986 the licensee's 138 KV transmission system received a lightning strike which resulted in a 4KV dip in system voltage. The voltage transient caused the operating Reactor Building Closed Loop Cooling Water Pumps to trip and auto restart, Reactor Building Standby Ventilation System to initiate in the emergency mode, Control Room Air Conditioning System to initiate in the emergency mode, Radiation Monitoring System computer to trip, Rod Sequence Control System to lock up, and the running instrument air compressors to trip. The Control Room operators took action to return the emergency systems to standby status after determining the cause of the initiation and by 03:00 all systems were returred to normal.
Plant conditions at the time of the lightning strike were mode switch in startup, Rx critical at 2% pcwer and pressure at 545 psig.
The licensee's electrical engineering group has been assigned the task of evaluating the effect of lightning strikes on plant equipment and results of the preliminary evaluation are expected by November 15, 1986.
The inspector will closely monitor thc licensee's progress in this area and will follow this action as Open Item 86-14-02.
8.
IE Information Notice 86-03 Update In response to IE Information Notice 86-03 and commitments made during a conference call to Region I on July 2,1986, the licensee conducted an inspection of all Limitorque Motor operators located inside the Primary Containment (33 operators) and Secondary Containment (220 operators).
The licensee inspection effort was primarily intended to identify and replace unqualified or unidentifiable wire types as discussed in IE Information Notice 86-03.
In addition to the primary objective, the licensee also addressed Unresolved Item 50-322/86-13-02, Missing or
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Improper Bolting Material in Limitorque Limit Switch Covers, and inspected operators for lug deficiencies, cut insulation, and correct terminal boards on the wrap material.
The inspection / replacement evaluation was conducted in two steps; with the thirty-three (33) operators inside primary containment inspected first, followed by inspection of operators in the secondary containment. Mainten-ance Work Requests were issued to cover wire replacement and limit switch cover bolting material replacement for all valves inspected with additional Maintenance Work Requests issued to cover repair or replacement outside the original scope. All work was completed prior to plant startup.
Inspection of the 33 operators within the primary containment was completed, with unqualified wire or wire of undetermined environmental qualification found and replaced in all but two (2) operators.
In addition to wire replacement, eleven (11) operators required bolting material replacement and seven (7) operators required maintenance which included repair of cut or cracked insulation, lug replacement and gearbox torquing.
Subsequent to the inspection of operators located in the primary containment, the licensee conducted an inspection of two-hundred and sixteen (216) valves located within the secondary containment and control building.
Results of this inspection remained consistent with the inspection conducted on operators located within the primary containment. Unqualified or non-verifiable wtre was replaced in all but thirty-three (33) of the operators. Additional work in the form of terminal lug replacement, flex conduit tightening, replacement of nicked conductors and limit switch replacement was performed in part on approximately fifteen (15) percent of the operators.
The licensee's effort to address the concern of Unresolved Item 50-322/86-13-02 were also completed during the conduct of this inspection.
Approximately eighteen (18) percent of the operators addressed by the inspection had some type of inadequacy in the limit switch coverassembly.
The inadequacies included improper bolting material, missing bolts, loose bolts, loose flexible couplings and torn or missing gasket materials.
All deficiencies were corrected upon discovery.
The inspectors closely followed the conduct of this inspection including direct observation of work review of licens e documentation and performance of retests. No unacceptable conditions were noted in the conduct of this work.
9.
10CFR21 Report - Containment Building Emergency Escape Airlock On August 15, 1986 the Resident Inspector informed the licensee of an apparent design deficiency in containment equipment hatch emergency es ape airlocks supplied by Enerfab Inc. (formerly W. J. Wooley).
The deficiency was identified by Consumers Power on an escape hatch installed at their Palisades Plant and involved improperly designed cams which would allow equalizing valves to be open at the wrong time,
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potentially breaching containment. The cam design at Palisades caused the inner equalizing valve to open when the outer airlock door was open
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and the outer equalizing valve to open when the inner airlock door was open.
A representative of Enerfab Inc. stated that an airlock of similar design defect may have been provided to Shoreham. This information prompted the licensee to conduct an immediate inspection of the emergency airlock operating mechanism. A Drywell entry was scheduled and airlock door and equalizing valve operation was monitored by members of the Mechanical Maintenance Department with verification by the facility Quality Control Division.
Both the outer and inner airlock doors were cycled and mechanism operation was observed by two licensee personnel within the airlock.
The inspection of the airlock operation showed that the design deficiency found by Consumers Power at the Palisades Station are not
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apparent in the emergency airlock installed at Shoreham.
The licensee intends to conduct further testing on the airlock during the present outage period and the Resident Inspector will track the conduct and results of these tests. This will be Inspector Followup Item 86-14-04.
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10. Unusual Event on August 11, 1986
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At 11:30a.m. on August 11, 1986 the licensee initiated a shutdown to make repairs to wide range level instrument 821 LIS 155B. Action statement 20 of Technical Specification 34.3.2 requires the facility to be in Hot Shutdown with 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the inoperative instrument is not returned to service within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of malfunction discovery.
The instrument was declared inoperable at 10:40a.m. and at 12:40 p.m. an Unusual Event was declared in accordance with the Licensee's Emergency Preparedness Implementing Procedures, which require the declaration of an Unusual Event whenever a shutdown is required by Technical Specification.
The plant was placed in Cold Shutdown.at 4:55 p.m. on August 11, 1986 and B21 LIS 155B was recalibrated.
Facility status at the initiation of shutdown was reactor critical at 3% power, rated temperature and pressure, with low power testing in progress. The Resident Inspector was on site throughout the shutdown and will conduct routine followup.
11. SALP Management Meeting A meeting between NRC Region I management and licensee management was held on July 29, 1986 at NRC Region I to discuss the Systematic Assessment of Licensee Performance (SALP) for Shoreham. A list of meeting attendees is presented in Attachment I.
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The Regional Administrator and his staff presented the NRC's view of the licensee's performance for the period March 1,1985 to February 28, 1986, as detailed in the Systematic Assessment of Licensee Performance, Inspection Report 50-322/85-99.
The licensee acknowledged the need for improvement in some aspects of their operation, and detailed corrective actions which had been taken, or were being planned. These corrective actions were detailed in a letter from J. D. Leonard to T. E. Murley on August 26, 1986 (SNRC-1277, Corrective Actions Discussed During the SALP Meeting Conducted July 29, 1986; Shoreham Nuclear Power Station - Unit 1, Docket No. 50-322).
Among the corrective actions committed to by the licensee to address areas of weakness were:
Increasing the time spent by Division Managers and Section Heads in the plant observing work activities.
- Relieving some of the administrative burden from the Plant Manager and Division Managers.
- Decreasing attrition and hiring more full-time LILC0 employees to fill positions.
- Stressing the importance of procedural adherence to plant personnel through night orders, during requalification training, and during staff meetings.
- Formation of an incident review board to determine root causes of station incidents.
Formulation of plans to rearrange the control room layout and remove the security console from the control room.
- Expanding LILCO involvement in the Power Ascension Test (PAT)
Program by assigning LILCO assistant test directors and coordi-nators to actively participate in PAT programs.
- Re-evaluation of the material control program including its organization and staffing.
- Improvements in the spare parts program including a bi-weekly report on spare parts and consumables and rearrangement of the spare parts catalogue.
- Initiation of a monthly meeting between the Nuclear Licensing and Regulatory Affairs Division Manager and the NRC Project Manager.
- Establishment of a new Department of Training and increasing the authorized compliment in training from nine to forty-fiv.
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NRC will continually assess these corrective actions to monitor the licensee's performarce in the noted weak areas of the SALP.
Future Inspection Reports, and the next SALP will detail the results of this monitoring.
12. Allegations Regarding Improper Activities By Health Physics Personnel (Allegation No. RI-86-A-23)
NRC Region I received a copy of an anonymous letter sent to the Long Island Lighting Company by a " concerned employee". This letter alleged that work on Health Physics instrumentation was performed by unqualified personnel, and that records were falsified to indicate that a foreman had supervised the work, when in fact he had not. The alleger further stated that he/she believed the work was conducted properly, but that he/she was concerned with the record falsification.
The licensee and NRC, upon receipt of the allegation, conducted parallel investigations into this concern. There are basically two concerns to this allegation which were investigated; 1) did the foreman's signature next to those of an unqualified technician indicate that he was present and directly supervising the technician when in fact he was not, and 2)
did the foreman intentionally go back into the records and countersign technicians signatures falsely?
Regarding these questions, interviews by the inspectors and the licensee determined the following facts:
The foreman responsible for the Health Physics Section Instrumentation functional area has been in his position for over one year. He has responsibility for the maintenance and calibration of the fixed and portable instrumentation assigned to the Health Physics Section.
Additionally, he has been appointed to evaluate the performance of technicians during the skill demonstration and qualification verification (Phase V) step of final qualification in the Instrument Repair and
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Calibration functional area.
His signature on the technician's Phase V signoff certifies the acceptability of the individuals performance.
(See Section 8.3.2, of SP 61.040.01 - Health Physics Technician Selection, Training and Qualification Program). Therefore, his assessment of technicians' performance is the basis for final qualification. The foreman has become familiar with the level of knowledge and ability of both qualified and unqualified tecnnicians assigned to him due to his time in the position, and his responsibility for certification.
Shoreham procedure SP 61.040.01 states, in part, that individuals shall not perform, unless supervised by a supervisor, any duty which they are not specifically qualified to independently perform.
It does not require continuous, direct supervision of an unqualified individual. The level and amount of supervision required is determined by the Foreman's assess-ment of the individual's ability and past performance.
This interpretation has been previously conveyed to all the foreman in the section by the Health Physics section head.
Therefore, by procedure, the foreman did not
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have to be present during the entire performance of the task.
Following the events that occurred in the Radiochemistry Section involving training qualification, Mr. W. E. Steiger, the Plant Manger, requested that all sections perform a review of their training and qualification records. The review was to determine if problems similar to those in Radiochemistry existed. Health Physics Section personnel were directed to review all related records for any discrepancies. During this review, discussions arose about the need for countersignatures by the person supervising an unqualified technician. The Health Physics section head indicated that he preferred to see countersignatures. There was some confusion on the requirement for countersigning. Tf.e foreman proceeded to perform his review of the Instrument Repair and Calibration records. His review showed that he had failed to countersign some initials of unquali-fled technicians who had performed tasks under his supervision. He had, as required by procedure, completed and signed the " reviewed by" columns, thus indicating his review and acceptance of the work performed. Addi-tionally, his signature appears on both the History Cards, and on the
" reviewed by' signoff on the Preventative Maintenance SAWS and Calibra-tion data sheet. The foreman was performing his review of the records in the Instrumentation Shop, in the presence of technicians. He decided to place his signature next to those of the unqualified technician to simply indicate that he had supervised them, since he had.
In fact, he also told a technician what he was doing.
Subsequently, the confusion over countersigning came to the attention of the Health Physics section head.
He explained that it is not required by any procedure, but is a gooa practice. He instructed all foremen to indicate, via memo, for the instances that they had neglected to countersign, a statement that they had in fact supervised the technicians.
This was done. At this point in time, the foreman stopped placing his signature on the previous records. At no time did the foreman fail to adhere to procedural requirements. He placed his initials next to those of the technicians simply to indicate that he was the person who had supervised them.
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In conclusion, it appears that no deliberate attempt to falsify records exists. Additionally, the foreman adhered to station procedures in his supervision of technician personnel.
Therefore, there is no basis for further action regarding these allegations. This matter is closed.
13. Management Meetings At periodic intervals during the course of this inspection, meetings were held with licensee management to discuss the scope and findings of this inspection.
Based on NRC Region I review of this report, and discussions with licensee representatives, it was determined that this report does not contain information subject to 10 CFR 2.790 restriction _.
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The inspectors also attended entrance and exit interviews for inspections conducted by region-based inspectors during the period.
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ATTACHMENT I JULY 29, 1986 SALP MANAGEMENT MEETING ATTENDEES
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NRC T. E. Murley - Regional Administrator R. W. Starostecki - Director, Division of Reactor Projects T. T. Martin - Director, Division of Radiation Safeguards & Security W. Johnstone - Deputy Director, Division of Reactor Safety H. B. Kister - Chief, Projects Branch No. 1, DRP J. R. Strosnider - Chief, Reactor Projects Section IB, DRP J. A. Berry - Senior Resident Inspector, Shoreham
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R. Caruso - Licensing Project Manager, NRR R. Lo - Licensing Project Manager, NRR J. M. Gutierrez - Regional Counsel B. Bordenick - Office of the Executive Legal Director R. Nimitz - Senior Radiation Specialist LILCO W. J. Catacasinos - Chairman of the Board & President J. Dye - Executive Vice President J. D. Leonard - Vice President-Nuclear Operations W. E. Steiger - Plant Manager J. L. Smith - Director, Office of Training B. R. McCaffrey - Manager, Nuclear Operations Support Department E. J. Youngling - Manager, Nuclear Engineering Department J. A. Notaro - Manager, Quality Assurance Department
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A. F. Early - General Counsel Suffolk County D. Minor - MHB Technical Associates A. Dyner - Kirkpatrick & Lockhart J. Blough - New York State Consumer Protection Board
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