IR 05000334/1987006
| ML20214Q859 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/08/1987 |
| From: | Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214Q844 | List: |
| References | |
| 50-334-87-06, 50-334-87-6, NUDOCS 8706050228 | |
| Download: ML20214Q859 (12) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-334/87-06 Docket No.
50-334 Licensee:
Duquesne Light Company One Oxford Center 301 Grant Street Pittsburgh, PA 15279
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Facility Name: Beaver Valley Power Station, Unit 1 Location:
Shippingport, Pennsylvania Dates:
March 16 - April 17, 1987 Inspectors:
W. M. Troskoski, Senior Resident Inspector 7. Pi dale, Resident Inspector, BVPS Unit 1 M
Approved by:
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4. E. Trfpp, Chief, Reactor Projects Section 3A
/Date Inspection Summary:
Inspection No. 50-334/87-06 on March 16 - April 17, 1987.
. Areas Inspected:
Routine inspections by the resident inspectors (124 hours0.00144 days <br />0.0344 hours <br />2.050265e-4 weeks <br />4.7182e-5 months <br />) of licensee actions on previous inspection findings, plant operations, housekeeping, fire protection, radiological controls, physical security, inoffice review of LERs, breaker thermal overload protection, control room emergency breathing air pressuri-zation system, and Emergency Response Facility diesel generator tests.
Results:
Two Violations were identified (failure to perform adequate post-main-tenance test - detail 4.b.2, and failure to properly prepare and execute an equip-ment clearance - detail 4.e).
In addition, a potential violation of the require-ments of IEEE Standard 279-1971 may exist in the control circuits for dampers associated with the Auxiliary Building ventilation system (detail 4.b(3)).
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TABLE OF CONTENTS P.a!Le 1.
Persons Contacted....................................................
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Plant Status.........................................................
3.
Followup on Outstanding Items........................................
4.
Plant Operations.....................................................
a.
General.........................................................
b.
Operations......................................................
c.
Plant Security / Physical Protection..............................
d.
Radiation Controls..............................................
e.
Plant Housekeeping and Fire Protection..........................
5.
Inoffice Review of LERs..............................................
6.
Breaker Thermal Overload Protection..................................
7.
Control Room Emergency Breathing Air Pressurization System...........
8.
Emergency Response Facility Diesel Generator Tests...................
9.
Exit Interview.......................................................
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DETAILS 1.
Persons Contacted During the report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support inspection activities, d
2.
Plant Status Reactor operation was maintained at full power throughout this inspection period.
Major activities included efforts toward completion of the control room emergency pressurization system modifications and expansion of the Beaver Valley Site Security Plan to include Unit 2.
A planned 20 day outage to re-move the security and pressure membrane wall separating the two control rooms is scheduled for April 24, 1987.
3.
Followup on Outstanding Items The NRC Outstanding Items (01) List was reviewed with cognizant licensee per-sonnel.
Items selected by the inspector were subsequently reviewed through discussions with licensee personnel, documentation reviews and field inspec-tion to determine whether licensee actions specified in the OIs had been satisfactorily completed.
The overall status of previously identified in-spection findings were reviewed, and planned and completed licensee actions were discussed for those items reported below:
(Closed) Violation (87-02-01): Failure to perform the Power Range Monitor low setpoint functional test per OST 1.2.1 during the reactor startup of January 11, 1987.
The intpector reviewed the corrective actions outlined in DLC let-ter dated March 20, 1987.
Special Operating Order 87-1 was implemented to place caution tags on the power range channel and an entry was made on the NSS log to require performance of OST 1.2.1 the next time Unit power was re-duced below 10%.
The inspector previously observed performance of UST 1.2.1 as discussed in NRC Inspection Report 334/87-05, detail 4.b.1.
Since then, an OMCN has been issued for Startup Procedure J and Startup Checklist D of OM Chapter 50 to require performance of the OST within seven days of startup.
Licensee actions were satisfactory and this item is closed.
(Closed) Unresolved Item (86-06-01): Licensee to evaluate preventive mainten-ance (PM) practices for rubber expansion joints (REJ) to replace before end of life.
In response to this concern, the QA Unit In-Service Inspection De-partment developed the Unit 1 REJ Visual Examination Program.
This program was initiated in April, 1987, and requires that an inspection under VT-507, Visual Examination of REJ, be performed once per 6 months to detect any de-graded condition.
The program contained requirements for recording and re-porting of examination results and increased frequency inspections for any degraded condition.
Licensee actions were determined to be satisfactory and this item is closed,
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(Closed) Unresolved Item (85-06-02): Revise TS 3.6.2.3 to require separation of two chemical addition system trains, and check auto functions of QS-P 4s during refueling outages.
Beaver Valley is currently participating in the Westinghouse Owner's Group project to delete the chemical injection system and replace it with a passive tri-sodium phosphate system installed in con-tainment.
It is the station's belief that the WOG reference plant analysis work will be done by the end of 1987.
Additionally, OST 1.13.11, Quench Spray System Operability Test, was revised to check the automatic functions of the chemical injection pumps during refueling outages.
This item is closed.
(Closed) Unresolved Item (86-17-01): Mandatory ISI test deleted but startup checklist not updated.
OST 1.55A.2, ISI - Safety and Relief Valve Tracking, was deleted when the testing requirements were superseded by a BVT.
OM 1.50.3, Startup Checklists B and D, were subsequently revised to require v2rification from the ISI supervisor that all surveillance requirements were completed prior to mode elevations.
These actions are satisfactory and the item is closed.
(Closed) Unresolved Item (85-17-05): Follow station action of Anchor - Darling check valve 10 CFR 21 Report of June 11, 1985.
The inspector reviewed Engi-l neering Memorandum No. 61604 dated October, 1985, which determined that the report was applicable to only three tilting disc check valves located at Unit 2.
This item is closed.
(Closed) Unresolved Item (85-11-01): Review completion of OST 1.33.10, C0-2 Fire Protection System Test, and licensee evaluation of current fire damper maintenance (PM) program to determine the need for a Preventive Maintenance Program.
This item was last discussed in detail 2 of NRC Inspection Report 334/86-20 and was left open pending licensee evaluation of the need for fire damper PM program.
The inspector reviewed a Plant Manager's Memo dated January 12, 1987, containing that evaluation.
The licensee intends to test and PM the remaining fire dampers that were not tested during the Fifth Re-fueling Outage to establish a baseline reference point.
Future PM work would be predicated upon a recurrent failure rate after that point in time.
These actions appear satisfactory as the dampers will be periodically retested.
The inspector had no further concerns at this time.
(Closed) IFI (85-16-03): Review IST Program Changes for MOV-RH-700 and 701, and review test results.
This item had been last reviewed in NRC Inspection Report 334/86-20 and left open pending development of a permanent OST to re-place the temporary operating procedure that gathered the initial data.
OST 1.10.5, Leak Test (M0V-RH-700 and 701), has been developed and approved by the Station.
This completes the action.
(Closed) Unresolved Item (85-20-01): Evaluate PVC - control room instrumenta-tion interface requirements.
See detail 5 of this inspection report for dis-cussio,
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(0 pen) Unresolved Iten (85-18-02): Failed 0-rings and gaskets on Low Head Safety Injection (LHSI) pump control rods.
The failure of the 0-ring and gasket assemblies on both LHSI pumps control rods on August 29, 1985 rendered both ECCS subsystems inoperable.
The licensee was to investigate the failure to assure that the 0 rings and gaskets are replaced on some frequency prior to the expiration of their useful life.
The licensee subsequently performed a test to monitor LHSI surge pressure and leakage upon starting of the pumps per OST 1.11.14 on August 6, 1986, which auto-started both pumps simultane-ously.
No anomalies or pressure spikes were observed.
Based upon the posi-tive test results, the licensee considered the item closed.
However, the evaluation did not address the unresolved issue concerning the material useful life span.
This item remains open pending further licensee evaluation.
(Closed) Unresolved Item (86-20-04): Verification of power operated relief valve (PORV) stroke times.
Plant technical specifications (TSs) and proce-dures had not implemented PORV stroke times as assumed in the NRC Safety Evaluation Report (SER), dated April 4, 1983.
The licensee performed an evaluation to determine whether the plant had been operated in an unanalyzed condition, using data from the station's ASME Valve Stroke Log, when compared to SER stroke time assumptions (2.5 seconds).
The evaluation noted that the resulting maximum RCS pressure would be below specified limits. Additionally, the licensee initiated several procedural upgrades to address PORV stroke testing.
Temporary Operating Procedure (TOP) 86-42 was developed to determine if the PORV opening stroke times are less than that assumed in the NRC SER.
The licensee committed to perform TOP 86-42 prior to reaching Mode 5 during their next shutdown (scheduled April 24, 1987).
OST 1.6.8, Placing OPS in Service, and OST 1.1.10, Cold Shutdown Valve Exercise Test, have been revised to incorporate the PORV 2.5 seconds stroke time acceptance criterion.
Per-formance of TOP 86-42 will be monitored through routine resident inspections.
This item is closed.
4.
Plant Operations a.
General Inspection tours of accessible plant areas were conducted during both day and night shifts with respect to Technical Specification (TS) com-pliance, housekeeping and cleanliness, fire protection, radiation control, physical security and plant protection, operational and maintenance administrative controls.
b.
Operations During the course of the inspection, discussions were conducted with operators concerning knowledge 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, the inspector comments or questions resulting from these daily reviews were acceptably resolved by licensee personnel.
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(1) During switchout of the operating charging pump on March 24, 1987, a suction gas binding problem occurred.
CH-P-1B had just been shutdown and the A pump started.
The 8 pump was immediately re-started to assure RCP seal injection and charging flow.
After the incident, when Operations increased the VCT pressure from 20 to 45 psig, an 8% level decrease occurred indicating the development of a gas pocket in the charging pump suction line.
The gas was vented through CH-312 to the sample sink.
This phenomena had previously occurred.
At that time, station in-vestigation found that an error in the chemistry analysis formula resulted in underestimating the RCS hydrogen concentration by a factor of two.
After correction, no further similar events occurred until now.
Discussions with the Operations Supervisor indicated that the station was in the process of evaluating the current RCS gas calculation.
Also, instructions would be issued to run with the VCT pressure at the lower end of its band, and to ensure that future pressure changes were made slowly to inhibit possible off-gassing at low pressure points in the system.
This item will re-ceive routine inspector followup.
(2) During a planned radioactive liquid release of LW-TK-78 under RWDA-2984 on March 29, 1987, a portion of the contents of the 7A tank was inadvertently released.
Post-discharge calculations performed for revised RWDA 2986 determined that a total of 11,700 gallons was released; 6,600 gallons from 78 and 4,900 gallons from 7A.
A com-posite isotope sample was calculated based on previous tank samples to determine the discharge blend.
The total isotopic discharge was 9.3 micro curies that translated into a 0.138 Maximum Permissible Concentration factor.
Therefore, no technical specification limits were approached.
Investigation found that the release path from the 7A tank was through the pump discharge valve M0V-LW-112A and into the liquid waste discharge header that had been aligned to receive the 78 tank flow.
The MOV, which indicated closed on the control board was found to actually be open in the field, and the liquid waste dis-charge header was immediately isolated.
Early discovery had been precluded by the fact that there are no control room level indica-tions for LW-TK-7A or B, and discharges typically take from 4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
On March 27, 1987, MOV-LW-112A had been worked on by maintenance personnel to repair a packing leak.
The post-maintenance test specified by MWR 852417 was a " leak check", which is appropriate for a packing adjustment on a valve with no ASME valve stroke time requirements.
However, once maintenance began troubleshooting the 2 inch ball valve, it was found to have the wrong seat.
The me-chanics also ended up replacing the valve stem, ball, seals and 0-rings.
After such work, it is typical for the mechanic to be pre-
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sent during the valve stroke.
Due to a breakdown in communication, Operations stroked the valve without the mechanic present.
Accept-ance appears to have been based upon changes in the control room's valve position lights.
The station has since determined that the valve's actuator had been assembled misorientated by 90 degrees.
Since a ball valve only turns 90 degrees from full open to full closed, the position indi-cation in the control room was reversed.
No earlier radioactive liquid release was possible while LW-P-12A was running on tank re-circulation because the liquid waste discharge header had been isolated.
The failure to perform a post-maintenance test that verified proper valve performance parameters after rebuild is a Violation (87-06-01)
of SAP 3D, Maintenance Work Request.
(3) An unplanned ESF actuation occurred on April 9, 1987, when an I&C technician caused an inadvertent short on the control room radiation monitor recorder while using a screwdriver to release a connector during replacement.
The resulting Vital Bus No. 2 voltage transient caused a spurious high-high alarm condition on the Auxiliary Build-ing ventilation exhaust Train A radiation monitor.
The high-high signal initiated an automatic flow diversion to the main filter banks and the elevated release path.
The operators responded by immediately resetting the alarm signal.
Several dampers which had not completed full travel were noted to have stopped in mid position upon this reset.
The licensee subse-quently completed a review of the electrical control prints and duplicated the event a second time.
The inspectors raised a concern regarding the failure of an ESF system to complete its actuation due to signal reset. A review of flow diagrams indicated that the rad monitor sample point is downstream of the dampers and for a true high rad condition (as opposed to a momentary spike), the actuation signal should remain as long as the actual condition existed which should result in reinitiation of the damper motion after the reset signal button was no longer depressed.
This is a potential viola-tion of IEEE Standard 279-1971, Section 4.16.
Compliance with this Standard is required by 10 CFR 50.55a(h).
This item is unresolved pending further evaluation (87-06-02).
c.
Plant Security / Physical Protection Implementation of the Physical Security Plan was observed in various plant areas 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 the
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Protected Area; Vehicles were properly searched and vehicle access to the Protected
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Area was in accordance with approved procedures;
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Security access controls to Vital Areas were being maintained and
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that persons in Vital Areas were properly authorized.
Security posts were adequately staffed and equipped, security per-
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sonnel were alert and knowledgeable regarding position requirements, and that written procedures were available; and Adequate lighting was maintained.
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Within this scope, no concerns were identified.
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At the conclusion of this inspection period, the licensee brought the Unit 2 site under the Site Physical Security Plan.
The inspectors ob-served the security steps taken in support of the Unit 1/2 control room wall removal.
Actions appeared satisfactory.
d.
Radiation Controls Radiation controls, including posting of radiation areas, the conditions of step-off pads, disposal of protective clothing, completion of Radi-ation 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 monitor operability (portable
and permanent), area monitor calibration and personnel frisking proce-dures were observed on a sampling basis.
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No discrepancies were identified.
e.
Plant Housekeeping and Fire Protection Plant housekeeping conditions including general cleanliness conditions
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and control of material to prevent fire hazards were observed in various areas during plant tours. Maintenance of fire barriers, fire barrier
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penetrations, and verification of posted fire watches in these areas were I
also observed.
The inspector was notified by the licensee that the Fire Protection Sys-tem was inoperable for about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> on March 19, 1987, due to a valving error.
TS 3.7.14.1 requires the fire suppression system to be operable at all times with two high pressure pumps and an operable flow path cap-able of transferring water through the distribution piping. With both pumps out, a backup fire suppression system is required to be established
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within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The valving error effectively isolated only the motor driven fire pump (FP-P-1).
The diesel driven fire pump (FP-P-2) auto
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started to make up system pressure losses due to leakage greater than the hydroneumatic tanks capability.
Because FP-P-1 was still running and no cause for the auto start was apparent, FP-P-L was shut down.
About 1-1/2 hours later, system pressure dropped again and FP-P-2 at-tempted to auto start but failed.
An investigation was immediately initiated to determine whether or not there was any system leakage at the Unit 2 side.
An operator at the in-take structure reported control problems at FP-P-2 and it was secured to prevent equipment damage.
About 30 minutes later, the fire and safety engineer began getting reports of low header pressure from various out buildings.
A subsequent operator walkdown identified several valves out of position that had isolated the running FP-P-1.
These valves were returned to normal alignment restoring system operability.
The cause for the total failure of the fire suppression system was due to two unrelated causes.
First, equipment clearance 524812 was incor-rectly posted.
The operators used OM Print 33.1, which shows several valves with a normal configuration of open instead of closed as they actually are.
However, the control room prints which should have been used for equipment clearance work as required by OM Chapter 1.48.68, correctly identified the valves as being normally closed (the valve num-bers are green banded).
Through the use of faulty assumptions, the equipment clearance which was originally to replace a fire hose station valve (IFP-58), rendered the flow path from FP-P-1 inoperable.
This is a Violation (87-06-03) of OM Chapter 1.48.68. Mechanical and Electrical Clearance Procedures.
The second problem was the subsequent failure of FP-P-2 to auto start and come up to speed.
This is similar to the previous problems discussed in Inspection Report 334/87-05.
Plant maintenance found that two out of four batteries on the diesel had dead cells and required replacement.
After replacement of the FP-P-2 battery banks under MWR 870400 on March 21, 1987, post-maintenance testing was conducted to assure proper cell voltage and specific gravity.
The MWR specified the performance of OST 1.33.8, Weekly Diesel Engine Driven Fire Pump Operation Test, prior to operational acceptance.
The inspector reviewed this OST and noted that it only addressed TS 4.7.14.1.la, 2a and 3a, yet TS 4.7.14.1.3b contains the surveillance requirements for the 24 volt battery bank specific gravity. MSP 33.03 and 33.04 contain the latter surveillance requirements.
It appears that these tests were conducted in a less formal manner as the inspector reviewed the necessary data that was collected attached to the still open MWR package.
The inspector expressed a concern about this apparent informality in assuring that all TS required testing is completed on a component that is totally replaced.
The licensee acknowl-edged those concern _ _ _ _ _ _ _ _ _ _
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5.
Inoffice Review of Licensee Event Reports (LERs)
The inspector reviewed LERs submitted to the NRC: Region I Office to verify that the details of the event are 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 warranted onsite followup.
The following LERs were reviewed:
LER 85-17-01:
Inoperable Rod Position Indication System due to Computer Failure.
This event had been initially reviewed in NRC Inspection Report 334/85-20 and Unresolved Item 85-20-01 was opened to follow licensee evaluation of the ac-ceptability of the Process Variable Computer (PVC) design such that an inter-nal failure could affect multiple control room indications.
This LER supple-ment provides an update of the station's investigation.
That review deter-mined that a fault on the AIM-100 Common Input Bus could affect all points inputing into the faulted module.
It was determined that such a fault could affect all the RPIs, the VCT level transmitter, and all four power range NIS racks.
The total error from such a failure was expected to be small on the order of less than 1%.
Corrective actions included removing seven RPIs from every bank and three of the four power range NIS racks from their PVC inputs.
The station indicated that long term corrective actions would rewire some of the PVC inputs such that no two NIS racks or RPIs in a given bank would input into the same module.
These plans appear acceptable and the unresolved item is closed.
LER 87-04: Inoperable Fire Suppression System This LER is discussed in detail 4.e of this inspection report.
LER 87-05: Unplanned ESF Actuation This event is discussed in detail 7 of this inspection report.
6.
Breaker Thermal Overload Protection On March 31, 1987, while performing OST 1.13.6, 28 Recirculation Pump Dry Test, control room operators attempted to remotely open pump discharge valve MOV-RS-1568 from the main control board.
After control board valve indication failed to respond to the open signal, an operator was immediately dispatched and verified that the valve had actually gone to its full open position.
The operators de-energized the MOV in the open position to assure that it remained in its safety alignment.
Maintenance investigation initially centered around a possible valve limit switch problem.
However, during post maintenance testing, the valve tripped before full travel was achieved due to thermal overload.
Maintenance person-nel then replaced the thermal overload coils with identical components, which
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also failed during post maintenance testing due to thermal overload.
Further licensee investigation noted that the installed coils were undersized when compared to the associated setting sheet.
The correct sized thermal overload device was installed and satisfactorily retested.
An inspection of all emergency motor control centers (MCCs) was conducted by the station to verify proper thermal overload settings as compared to the setting sheets.
This inspection initially identified what appeared to be about a 10% discrepancy rate between the design documents (setting sheets)
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and actual installed plant hardware.
Further review indicated that beside MOV-RS-156B, the low head safety injection pump suction valve, (MOV-SI-8608)
from the containment sump (normally closed, but required to open on an SI after a low-low level in the RWST), also had an undersized thermal-overload block installed.
The remainder of the discrepancies were in the conservative direction or accounted for in various DCPs.
The thermal overload block was subsequently replaced on MOV-RS-156B and new ASME stroke data obtained.
Discussions with plant management indicated that due to possible containment integrity concerns, M0V-SI-8608 would not have its thermal overload block replaced until the April 24, 1987 outage.
This position was justified by the fact that the valve had been successfully l
stroked during the last outage and maintenance history records do not point to any previous problems.
Additionally, the MCCs are accessible to operators
at all times.
Review of licensee actions to replace the device and test MOV-SI-8608 will be followed as Unresolved Item (87-06-04).
The inspector reviewed the administrative and engineering controls placed upon the setting sheets.
Through discussions with cognizant NED personnel, it was determined that all engineering involving relay settings and electrical pro-tection is required to be performed at the DLC corporate office under the NED engineering procedures and QA program.
These calculations are triggered by one of two mechanisms.
Either the switch is replaced in kind under an MWR, or it is replaced with one of a different value under a DCP.
After an exten-sive records review, it was determined that all the discrepancies appear to have existed since initial plant startup.
Because the current record keeping
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These actions are expected to be completed in about 30 days.
No other concerns were
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7.
Control Room Emergency Breathing Air Pressurization System a.
Train B of the control room emergency breathing air pressurization system (CREBAPS) was actuated on March 21, 1987, during performance of the monthly safeguards protection system surveillance test (OST 1.1.12).
Apparently, the operator accidentally bumped a slave relay (K6268) with his hard hat.
All components associated with K626B including the bottled air system and control room HVAC dampers responded as expected.
The root
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cause was operator inattention to possible hazards associated with the SSPS cabinets.
The inspector was informed that this item would be dis-cussed in operator training.
b.
After completion of design modification on control room air bottle VS-TK-6A, TOP 87-05, Post Maintenance Tests on CREBAPS Discharge Control Components, was conducted on March 21, 1987.
The inspector reviewed the tests and found that it adequately addressed the time requirements of the TS 3.7.7.1 action statement.
As this TOP temporarily isolated the other four control room pressurization tanks, a double verification was provided for system restoration.
During performance of this TOP, the licensee confirmed the existence of a design deficiency in the safety related HVAC control circuits on loss of power.
This original design deficiency was first identified in Unit 2 as a 10 CFR 50.55(e) report on February 13, 1987.
At the inspector's request, the licensee reviewed the Unit 1 system design and found that after actuation of the CIB 60 minute timer, any loss of power and sub-sequent restoration would result in a timer reset which delays startup of the control room emergency pressurization fans.
Followup on licensee corrective actions for this design problem and for the apparent communi-cations gap between Unit 1 and Unit 2 is Unresolved Item (87-06-04).
8.
Emergency Response Facility Diesel Generator Tests The ERF DG is required to provide non-safety grade emergency power to various components important to safety.
One example is the auxiliary feedwater pump FW-P-4, installed to maet the station's 10 CFR 50, Appendix R, Fire Protection, commitments.
The inspector reviewed the last three monthly copies of OST 1.36.15, ERFS Diesel Generator Test, and found all three signed off as un-satisfactory.
On both the February and April tests, the diesel tripped on overspeed during manual sync (these control circuits are cut out during an emergency start and do not impact that safety function).
Also, an entry in the NSS log indicated that the ERFS diesel day tank was left one-half empty after the April test because both fuel oil transfer pumps were on clearance.
Discussions were held with DLC management concerning station control of com-ponents that are important to safety, but not included in the plant technical specifications, such as the ERFS diesel (including fuel system and starting batteries). The inspector was informed that this item would be reviewed by operations.
Also, a review would be conducted to determine whether the bal-ance of plant OSTs should be revised or preventive maintenance procedures developed that address the DG batteries.
9.
Ex_it 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 conclu ion of the report period.