IR 05000334/1985022
| ML20136D660 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/14/1985 |
| From: | Asars A, Lester Tripp, Troskoski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20136D611 | List: |
| References | |
| 50-334-85-22, NUDOCS 8511210342 | |
| Download: ML20136D660 (14) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-334/85-22 Docket No.
50-334 Licensee:
Duquesne Light Company One Oxford Center 301 Grant Street Pittsburgh, PA 15279 Facility Name:
Beaver Valley Power Station, Unit 1 Location:
Shippingport, Pennsylvania Dates:
0 tober
- 31, 1985 Inspector:
// b.
/ 8h M. T koski, Senior Resident Inspector
' date
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)4. b. GidQS
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As'a'r s, Resident Inspector
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Approved by:
, b. AMD IfNS 1".
E.
Trilih, Chief, Reactor Project tlate
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Section No. 3A, Reactor Projects Branch 3 Inspection Summary:
Inspection No. 50-334/85-22 on October 1 - 31, 1985.
Areas Inspected:
Routine inspections by the resident inspectors (64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br />) of licensee actions for previous inspection findings, plant operations, house-keeping, fire protection, radiological controls, physical security, engineered safety features verification, vendor recommended maintenance for diesel genera-tors, inspector site access, electrical equipment environmental qualification, surveillance activities and review of licensee event reports.
Results:
Two violations were identified:
failure to issue termination expo-sure reports (detail 3.d), and unauthcrized operation of a Unit 1/2 shared system boundary isolation valve (detail 3.b.5).
8511210342 851114
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TABLE OF CONTEkTS Page 1.
Persons Contacted...........................................
2.
Followup on Outstanding Items.............................. 2 3.
Plant Operations........................................... 3 a.
General............................................... 3 b.
Operations
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Plant Security / Physical Protection.................... 8 d.
Radition Controls......................................
e.
Plant Housekeeping and Fire Protection................ 9 4.
Engineered Safety Features (ESF) Veri fication............... 9 5.
Vendor Recommended Preventive Maintenance.................. 10 6.
Inspector Site Access...................................... 10 7.
Electrical Equipment Requiring Environmental Qualification.. 11 8.
Surveillance Activities................................... 12 9.
Inoffice Review of Licensee Event Reports (LERs)...........13 10.
Exit Interview............................................. 14
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DETAILS 1.
Persons Contacted J. J. Carey, Vice President, Nuclear Group R. J. Druga, Manager, Technical Services T. D. Jones, General Manager, Nuclear Operations W. S. Lacey, Plant Manager J. D. Sieber, General Manager, Nuclear Services N. R. Tonet, General Manager, Nuclear Engr. & Constr. Unit The inspector also contacted other licensee employees and contractors during this inspection.
2.
Followup On Outstanding Items The NRC Outstanding Items (OI) List was reviewed with cognizant licensee personnel.
Items selected by the inspector were subsequently ' reviewed through discussions with licensee personnel, documentation reviews and field inspection to determine whether licensee actions specified in the OI's had been satisfactorily completed. The overall status of previously identified inspection findings were reviewed, and planned and completed licensee actions were discussed for those items reported below:
(Closed) IFI (85-13-02): Develop objective criteria for verifying agree-ment of QC intercomparison of radiological environmental samples. The ob-jective of this item was the development of objective rationale to deter-mine the adequacy of the QC interlaboratory comparisons.
The inspector reviewed procedure EM-4, Environmental Monitoring Criteria for Comparing Quality Control Samples, that incorporates acceptance criteria for split and spiked samples contained in the Environmental Radiological Laboratory Intercomparison Program - Criteria for Comparing Analytical Measurements.
A review of 1985 data, analyzed to the above criteria, indicated good agreement. When several iodine samples from one of the two laboratories were outside the possible acceptance range, the licensee instituted a three sample quality check to verify that no adverse trend was developing.
This item is closed.
(Closed) Unresolved Item (84-30-01):
Resolve non-safety loads connected to safety Class 1E battery systems to assure capacity and ensure no de-gradation.
The licensee had committed to revising their operating manual procedures and routine surveillance tests for the 125 V DC control system to ensure that non-safety related loads are maintained disconnected from the safety related batteries. This was requested because selectable loads l
on the No. 5 switchboard which are normally powered by the non-safety re-lated No. 5 battery can be powered from Vital Bus 1-1A or 1-2A when troubleshooting ground indications. The inspector reviewed various proce-dures and OM Chapter 1.39.4, Operating Procedures for 125 V DC Control System.
Appropriate changes were included to ensure that non-safety re-lated loads were restored back to the No. 5 battery. Additionally, the weekly surveillance tests of the station batteries were revised to reouire a verification of the correct alignment for the DC switchboard feeder breakers.
This item is closed.
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(0 pen) Unresolved Item (85-17-02): Verify that seismic calculations con-sider the non yielding effects of the support frame on the station battery end cells. This item was last discussed in NRC Inspection Report 50-334/
85-18, Detail 5b.
The licensee has been monitoring the end cells of batteries No. 3 and 4 for crazing. Up until October 31, 1985, there was no evidence of any crazing on any of the intermediate cells.
However, during a routine tour of Battery Room No. 3, the operator noted a crack running along the bottom of cell No. 3.
The inspector noted that the crack was approximately 4" long and was not yet leaking. At the conclu-sion of this inspection, the licensee's engineering staff was evaluating the condition.
(Closed) IFI (85-20-02): Verify that the inoperable RPI system report is issued. This item is discussed in Detail 9 of this report.
(0 pen) Unresolved Item (85-20-01): Determine whether PVC isolation from nonsafety but technical specification required equipment is necessary.
This item is discussed in Detail 9 of this inspection report.
(Closed) IFI (85-20-05):
Provide supplement to LER:
85-13.
This item is discussed in Detail 9 of this inspection report.
3.
Plant Operations a.
General Inspection tours of the plant areas listed below were conducted during both day and night shifts with respect to Technical Specification (TS) compliance, housekeeping and cleanliness, fire protection, radia-tion control, physical security and plant protection, operational and maintenance administrative controls.
-- Control Room
-- Primary Auxiliary Building Turbine Building
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-- Service Building Main Intake Structure
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-- Main Steam Valve Room
-- Purge Duct Room
-- East / West Cable Vaults
-- Emergency Diesel Generator Rooms Containment Building
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Penetration Areas
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-- Safeguards Areas Varicus Switchgear Rooms / Cable Spreading Room
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-- Protected Areas
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Acceptance criteria for the above areas included the following:
-- BVPS FSAR Technical Specifications (TS)
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BVPS Operating Manual (OM), Chapter 48, Conduct of Operations
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OM 1.48.5, Section D, Jumpers and Lifted Leads
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-- OM 1.48.6, Clearance Procedures
-- OM 1.48.8, Records OM 1.48.9, Rules of Practice
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OM Chapter SSA, Periodic Checks, Operating Surveillance Tests BVPS Maintenance Manual (MM), Chapter 1, Conduct of Maintenance
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-- 10CFR50.54(k), Control Room Manning Requirements BVPS Site / Station Administrative Procedures (SAP)
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-- BVPS Physical Security Plan (PSP)
Inspector Judgement
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b.
Operations The inspector toured the Control Room regularly to verify compliance with NRC requirements and facility technical specifications (TS).
Direct observations of instrumentation, recorder traces and control panels were made for items important to safety.
Included in the re-views are the rod position indicators, nuclear instrumentation systems, radiation monitors, containment pressure and temperature parameters, onsite/offsite emergency power sources, availability of reactor protection systems and proper alignment of engineered safety feature systems. Where an abnormal condition existed (such as out-of-service equipment), adherence to appropriate TS action statements was independently verified. Also, various operation logs and records, including completed surveillance. tests, equipment. clearance permits in progress, status board maintenance and temporary operating proce-dures were reviewed on a sampling basis for compliance with technical specifications and those administrative controls listed in paragraph 3a.
During.the course of the inspection,- discussions were conducted with operators _ concerning reasons for selected annunciators and knowledge
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of recent changes to procedures, facility configuration and plant conditions. The inspector verified adherence to approved procedures for ongoing activities observed. Shift turnovers were witnessed and staffing requirements confirmed. Except where noted below, inspector comments or questions resulting from these daily reviews were accep-tably resolved by licensee personnel.
1.
A high-high steam generator level reactor trip occurred at 2:04 p.m. October 4, 1985. The main fuse on DC inverter No. 3 failed which caused a momentary interruption of power to vital bus No.
3 until an auxiliary power source could be established. Within that 45 second interval, all three steam generator level control
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channels failed low causing the main feed reg valves to go full open. The A steam generator high-high level was reached at 75%
.which tripped the turbine.
The turbine trip - reactor trip occurred and all systems functioned as expected.
NRC Inspection Repcrt 334/85-20 discussed previous work invol-ving vital bus No.
3.
The inspector had observed the fuse changeout with a qualified spare of a 500 amp rating.
The li-censee cut the failed fuse and determined that the cause was due to a high resistance connection.
If an actual power surge had occurred, only the silver connector would have vaporized leaving no visible burn spots. However, one side of the fuse had defi-nite indications of being subjected to high temperature while
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the other side appeared normal. The I&C supervisor stated that
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the apparent cause was due to a random manufacturing defect. A qualified spare was placed in the inverter and a dummy load was hooked up for a period of several days to verify operability while the inverter was on its auxiliary power supply.
The inverter was manufactured by Cyberlex as part of a 20 KVA l
uninterruptible power supply.
Inspection of the internal fuse connector did not identify any blackened or apparent hot spots associated with the fuse holder.
The reactor was restarted at about 7:30 p.m. on October 4,1985, and returned to about 42% power.
Power level was initially limited by an inoperable condensate pump on the secondary side.
2.
During the week of October 13, 1985, reactor containment sump pumpout rate increased from about 6 to 12 gph. After the trip, the licensee conducted a containment entry and identified a body to bonnent valve leak on FW-602, steam generator 1C low point drain.
The licensee has currently scheduled valve repair for the weekend of November 2, 1985.
3.
A reactor trip from full power was initiated by a spurious low reactor coolant system flow at 11:49 a.m. on October 25, 1985.
All safety systems functioned as designed.
The cause of the spurious trip was subsequently determined to be a combination of technician error and instrument valve design used for a maintenance surveillance procedure which is normally run with the reactor shutdown.
The station had been previously experiencing a number of RCS Loop 1 Channel 2 low flow alarms and I&C technicians were sent inside containment to troubleshoot the flow transmitter.
MSP 6.29, F-415 Reactor Coolant Flow Loop 1 Channel 2 18 month Cali-bration, was used as guidance for the troubleshooting. Several defective parts (oscillator circuit and a capacitor) were changed
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out and the transmitter was calibrated.
While returning the instrument valves to their normal alignment, the I&C technicians had difficulty opening what was identified as the low pressure instrument valve in Figure 1 of the MSP. In fact, this was ac-tually the high pressure valve that isolated transmitter F-415 from an instrument line common to all three flow channels. When the technician opened the valve, it popped off its seat and a momentary pressure fluctuation occurred that was common to all three instrument loops. Only the loop 3 bistable made up for approximately 40 milliseconds.
Since the loop 2 bistable was still tripped as part of the procedure, this made up the 2 out of 3 reactor protection logic. The licensee informed the inspec-tor that a procedure revision would correct the valve identif t-cation discrepancy. No other changes are contemplated for the instrument valve sequence.
4.
The reactor was restarted at about 1:55 a.m. on October 26, 1985.
Criticality was achieved with control bank D at 30 steps as op-
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posed to the estimated critical position predicted value of 91 steps.
The plant was returned to the grid at about 5:29 a.m, without event.
No further RCS low flow alarms were received.
5.
Operations personnel informed the inspector at about 12:15 p.m.,
October 31, 1985, that radioactive liquid was leaking into the Unit 2 condensate polishing building thru a 1.5 inch uncapped line from Unit 1.
The inspector observed licensee actions to evacuate personnel from the immediate area, contain the spill and isolate the flow path. Immediate frisking of about 12 craft workers in the general area identified one person with slight contamination (20-30 cps above background) on one shoe.
Subse-quent frisking of about 100 more craft workers identified one more person with slightly contaminated shoes and pant cuffs.
Any dose was determined to be less than minimum detectable thru surveys of the general area.
The leak, which was first noticed at about 12:00 p.m., was ter-minated at 12:55 p.m. after the Unit 1 boric acid hold tank (BR-TK-7) recirculation flow path was isolated. In the interim, an Unusual Event was declared due to the radioactive liquid dis-charge from an unidentified source. It was terminated at 2:55 p.m., after the leak was identified, stopped, the area isolated and personnel checked for possible contamination.
The source was positively identified when BR-79, a 1.5 inch globe valve, servir,g as the boundary isolation point between the Unit 1/2 shared system, was found partially opened.
BR-79 is located in a U.it 1 pipe trench, under grating near the hold tank. Craft personnel had been insulating several runs of pipe
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r in the cramped quarters of this trench and the licensee theor-izes that they accidentally bumped the valve open without reali-zing it.
The valve is required to be clcsed per two separate Nuclear Shift Supervisor Clearances posted in December, 1977,
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and January,1978, until Unit 2 is ready to tie into the system.
It was not tagged in the field or locked closed though the clearances were reverified every six months.
The end of that line terminates outside of the condensat polishing building and did not have a flange or high pressure cap on it.
The BVPS QA Program, Section A.2.2.5 of FSAR Appendix A, and ANSI N18.7-1972, Administrative Requirements for Nuclear Power Plants, require implementation of equipment control procedures to avoid unauthorized operation of equipment. These procedures are to include control measures such as locking or tagging to secure and identify equipment in a controlled status. The fail-ure to effectively control the operation of BR-79 is a violation (85-22-01) of the above requirements that resulted in an un-authorized release of several hundred gallons of radioactive liquid to the environment.
Discussions with operation's personnel indicated that a review would be conducted to identify all shared radioactive system boundary isolation points to ensure that each is properly tagged, locked if necessary, and that two barriers such as lyigh pressure caps, blank flanges or valves, exist between the two units. The inspector stated that the licensee should also review the mate-rial contained in IE Bulletin 80-10, Contamination of Nonrad Systems and Resulting Potential for Unmonitored Releases, to ensure that Unit 1/2 interface controls were adequate.
About 1600 gallons of water are estimated to have been released from BR-TK-7, as indicated from level changes that occurred between about 8:00 a.m. and 1:00 p.m.
Gross activity was about 2E-3 micro Ci/ml. Most of this water ended up in the Unit 2 condensate polishing building sump, which has a volume of 1500 gallons. The licensee informed the inspector that analysis of the sump and tank water indicated that they had about the same gross activity levels. Since flushing work to the sump and sub-sequent pumpout to the storm sewer system had been terminated prior to the event, it is reasonable to conclude that much of the 1600 gallons was captured.
Followup to determine how much activity was released to the environment and comparison with 10 CFR 20.106 limits is Unresolved Item (85-22-02).
At the conclusion of this inspection p'eriod, cleanup of the con-taminated areas in Unit 2, inc.luding the sump, surrounding floors and drain lines, had just begun.
Progress in this area will be tracked as Inspector Follow Item (85-22-03).
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Plant Security / Physical Protection Implementation of the Physical Security Plan was observed in the areas listed in paragraph 3a above with regard to the following:
Protected area barriers were not degraded;
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Isolation zones were clear;
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Persons and packages were checked prior to allowing entry into
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the Protected Area;
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Vehicles were properly searched and vehicle access to the Pro-tected Area was in accordance with approved procedures;
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Security access controls to Vital Areas were being maintained and that persons in Vital Areas were properly authorized; Security posts were adequately staffed and equipped, security
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personnel were alert and knowledgeable regarding position re-quirements, and that written procedures were available; and Adequate lighting was maintained.
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No discrepancies were observed.
d.
Radiation Controls Radiation controls, including posting of radiation areas, the con-ditions of step-off pads, disposal of protective clothing, completion of Radiation Work Permits, compliance with the conditions of the Radiation Work Permits, personnel monitoring devices being worn, cleanliness of work areas, radiation control job coverage, area noni-tor operability -(portable and permanent), area monitor calibration and personnel frisking procedures were observed on a sampling basis.
The inspector was informed by the licensee that two 10 CFR 20.408 termination report deficiencies were discovered in which the 90 day report period was missed for two contractor employees.
The identi-fication of one event was the result of the individual's new employer requesting the personnel exposure file.
This has been a recurring problem as discussed in NRC Inspection Reports 334/84-15 and 84-33.
For the second instance, a Severity Level V Violation was issued.
The corrective action outlined in DLC letter of January 30, 1985, failed to preclude a reoccurrence of the termination exposure report deficiencies for contractor radiation workers.
The failure to submit the termination exposure reports within 90 days is another example of a violation (85-22-04) of 10 CFR 20.408.
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Plant Housekeeping and Fire Protection Plant housekeeping conditions including general clecnliness condi-tions and control of material to prevent fire hazards were observed in areas listed in paragraph 3a. Maintenance of fire barriers, fire barrier penetrations, and verification of posted fire watches in these areas were also observed. This included the hourly fire patrols of the switchgear and cable spreading rooms.
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The licensee conducted their Annual Fire Drill with various offsite organizations on October 1,1985.
The inspector observed the inter-action that occurred between the facility personnel and the various fire departments.
The drill was well coordinated and the several local volunteer fire companies participated enthusiastically.
No problems were identified by the inspector.
The inspector was informed by the licensee on October 10, 1985, that a QA audit of the fire protection area identified an inoperable fire-rated assembly.
Two ventilation ducts connecting the relay room (CR-3) and process instrument room, were installed without fire dampers. As a one hour fire -watch had been established in this area for the structural steel issue, the licensee was in compliance with TS 3.7.15 action requirements.
A special 30 day report was issued on October 31, 1985, per TS 6.9.2.
The licensee stated that a re-evaluation of the fire damper program (IE Information Notice 83-69)
would be completed by January 16, 1986, to ensure all such required dampers have been installed. Review of the results of this action is Inspector Follow Item (85-22-05).
4.
Engineered Safety Features (ESF) Verification The operability of the High Head Safety Injection System was verified during this inspection period, by performing a walkdown of accessible portions that included the following as appropriate:
(1) System lineup procedures matched plant drawings and the as-built configuration.
(2) Equipment conditions were observed for items which might degrade per-formance.
Hangers and supports were operable.
(3) The interior of breakers, electrical and instrumentation cabinets were inspected for debris, loose material, jumpers, etc.
(4) Instrumentation was properly valved in and functioning; and had cur-rent calibration dates.
(5) Valves were verified to be in the proper position with power avail-able.
Valve locking mechanisms were checked, where required.
No deficiencies were identified.
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5.
Vendor Recommended Preventive Maintenance Technical Specification 4.8.1.1.2.b requires that the emergency diesel generators be inspected at least once per 18 months during shutdown in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service. The diesel generators used at BV-1 are General Motors Electro-Motive Division Model 999.
The vendor recommends certain extended maintenance items to be conducted at the 3, 6, and 12 year periods.
During the initial review of this item, the inspector was informed by cog-nizant licensee representatives that no special maintenance was conducted as triggered by those yearly intervals. In reviewing this matter further, the inspector was informed that the Operations Assessment Group had recent-ly conducted a review of this item as part of the vendor manual upgrade program.
It was determined at that time that over 60 discrepancies existed between the implemented preventive maintenance program and the vendor's recommended actions. These items dealt with all aspects of the various support systems.
The licensee's representatives stated that although these -particular items were referenced in the vendor's manual, they were under the opinion that the intervals specified were for diesels subjected to continuous operation and not the emergency standby function of the BV-1 diesels. They further stated that these items were in the process of being resolved with the vendor to determine specific applicability to the standby function and additionally to determine whether or not other maintenance activities might be required for diesels used in the standby status.
Followup to determine an acceptable resolution to these items is Unresolved Item (85-22-06).
6.
Inspector Site Access Inadequate badging and dosimetry issue for visiting NRC inspectors was previously reviewed in Inspection Report 334/85-21. The inspector walked through this process and found that with the existing procedures, the li-censee would not be able to meet the goal of one hour access for visiting inspectors. Badging for access to the protected area was adequate. Dif-ficulties arose when the inspector attempted to obtain dosimetry and a respirator. The radcon technicians were not comfortable with the proce-dure for issuing dosimetry to the inspector or for performing the respira-tor fit test. The licensee had previously agreed to develop brief site specific radiation worker training for NRC inspectors consisting of a tour of the RWP and respirator issue area and a description of BVPS radiation worker practices.
No tour or description of practices was given to the inspector. When these new procedures are implemented, the licensee should be able to meet the one hour access goal. Further review of dosimetry issue and site specific radiation worker training for visiting NRC inspec-tors is IFI (85-22-07).
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Electrical Equipment Requiring Environmental Qualification 10 CFR 50.49, Environmental Qualification of Electrical Equipment Impor-tant to Safety for Nuclear Power Plants, requires the licensee to esta-blish a program for qualifying safety and non-safety related electrical equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions, and certain post-accident monitoring equipment. The licensee is required to prepare a list of this equipment with information about its performance specification, electrical characteristics, and environmental conditions.
Replacement equipment must be qualified in accordance with these requirements. The inspector reviewed the licensee's Maintenance Program to verify that cor-rective and preventive maintenance performed on this electrical equipment was conducted in such a manner as to preserve those environmental qualifications.
Station Administrative Procedure, Chapter 31, Equipment Qualification Program,_was put into effect on September 10, 1985.
This procedure co-ordinates the various aspects of the EQ Program, such as qualification assessment, new equipment procurement, spare parts assessment, degradable parts / aging assessment, etc., between the various Nuclear Group functional units. Various responsibilities appear to be well-defined.
The inspector reviewed a controlled copy of the Electrical Equipment Qualification Master List dated September 18, 1985, which was compiled to meet the requirements of 10 CFR 50.49. The inspector verified that this document has been made available to the various station maintenance groups.
It identifies equipment, sorted by mark number, and various conduit seals or splices requiring special EQ considerations.
The inspector discussed the use of this list with representatives of tha I&C and Maintenance (mechanical and electrical) Departments. The inspector was informed that after Operations personnel issue a maintenance work request, the MWR is cross-checked against the electrical equipment qualification master list to determine which special controls would be required for the repair or replacement of the item.
Long term plans are to eventually update all of the maintenance procedures to assure that vendor recommended and any special environmental qualifi-cation or test requirements are incorporated. Currently, NECU is develop-ing maintenance assessment packages (MAP) to determine what must be done to preserve the environmental qualification. These MAPS are then sent to I&C or Maintenance (they were originally sent to the Procedures Group but a reorganization has dictated a change to this flowpath) to review and determine what revisions must be made to the CMPs and PMPs for EQ main-tenance.
Revisions to these procedures are then reviewed by the I&C or Maintenance Departments.
The inspector determined that the interim program appeared to be suffi-cient to ensure that maintenance work performed did not degrade the envi-ronmental qualification of electrical equipment important to safety.
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8.
Surveillance Activities To ascertain that surveillance of safety-related systems or components is being conducted in accordance with license requirements, the inspector observed portions of selected tests to verify that:
a.
The surveillance test procedure conforms to technical specification requiremeM s.
b.
Required administrative approvals and tagouts are obtained before initiating the tast.
c.
Testing is being accomplished by qualified personnel in accordance with an approved test procedure.
d.
Required test instrumentation is calibrated.
e.
LCOs are met.
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The test data are accurate and complete.
Selected test result data was independently reviewed to verify accuracy.
g.
Independently verify the system was properly returned to service.
h.
Test results meet technical specificathm ecquirements and test dis-crepancies are rectified.
1.
The surveillance test was completed at the required frequency.
The following in progress tests were sitrassed by the inspector:
OST 1.13.2, QS-P-1B Test, conducted on Octcber 3, 1985.
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OST 1.7.1, Boron Acid Transfer Pump 2A Test, conducted on
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October 8, 1985.
OST 1.7.5, CH-P-1B Test, conducted October 9, 1985.
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No problems were identified.
9.
Inoffice Review of Licensee Event Reports (LERs)
The inspector reviewed LERs submitted to the NRC:RI office to verify that the details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action. The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event war-ranted onsite followup. The following LERs were reviewed:
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LER: 85-13-1 Reactor trip due to low - low steam generator levels LER:
85-15 Automatic actuation of ESF (SI)
LER: 85-16 OT-Delta-T/0P-Delta-T reactor trip LER: 85-17 Inoperable RPI System due to computer failure.
LER 85-13 provided information on a reactor trip due to a low - low steam generator level caused by a malfunction of a Hagen V-TO-I converter used in the steam dump system. The LER failed to specify the corrective action taken to correct the malfunction. Consequently, IFI 85-20-05 was opened pending issuance of a supplement to this LER providing the necessary in-formation.
The supplement was issued on October 3,1985, and indicated that the faulty converter was replaced. The cause of failure was deter-mined to be random end of life. The IFI is closed.
The events described in LER 85-15 and associated multiple equipment problems were previously discussed in detail 3b of NRC Inspection Report 334/85-18.
Several open items associated with this event are still being actively pursued.
The overpower /overtemperature Delta-T trip that occurred on September 16, 1985, and detailed in LER 85-16 is discussed in Inspection Report 334/85-20.
Inspection Report 334/85-20 discussed details of the inoperable RPI system.
Since this event was reported as LER 85-17, IFI (85-20-02), opened to verify that Technical Specification 3.1.3.2(3) report was made, is closed.
The LER stated that the station is investigating whether further process
variable computer isolation from non-safety but TS required equipment is required and that a supplemental report following the conclusion of this review would be made.
Unresolved Item (85-20-01) is currently tracking resolution of this concern.
10.
Exit Interview Meetings were held with senior facility management periodically during'the course of this inspection to discuss the inspection scope and findings. A summary of inspection findings was further discussed with the licensee at the conclusion of the report period.
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