IR 05000334/1979022
| ML19294B319 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 01/03/1980 |
| From: | Beckman D, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19294B314 | List: |
| References | |
| 50-334-79-22, NUDOCS 8002280175 | |
| Download: ML19294B319 (22) | |
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U.S. NUCLEAR REGULATORY COMMISSI0ft 0FFICE OF IllSPECTION AtlD EllFORCEMENT Region I 50-334/79-22 Report flo.
50-334 Docket No.
C License No. D R-66 Priority Category
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Duquesne Light Company Licensee:
435 Sixth Avenue Pittsburgh, Pennsylvania 15219
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Facility Name:
Inspection at:
Shippingport, Pennsylvania Inspection conducted: October 2 - tovember 2, 1979 Inspectors:
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Or
/.1 D /79 D. A. Beckman, Resident Inspector dite signed date signed
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date signed
/[/4, M//8?
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O Approved by:
E. C. McCabel Chief, Reactor Projects date/ signed
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Section No. 2, R0&tlS canch Inspection Summary:
Inspection on October 2 - November 2,1979 (Insoection P.eport No. 50-334/79-22)
Areas Inspected:
Routine resident inspector review (55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> onsite) of:
Action on previous inspection findings, plant operations, IE Bulletin and Circular followup, plans for coping with strikes, licensee event reports, periodic reports, and followup on a reactor trip.
Results:
No items of noncompliance were identified.
Region I Form 12 (dev. April 77)
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DETAILS 1.
Persons Contacted G. Eaatty, QA Engineer R. Balcerek, Maintenance Supervisor R. Borski, Senior Engineer J. Carey, Technical Assistant - Nuclear R. Conrad, Senior Engineer C. Ewing, QA Supervisor W. Glidden, QA Engineer K. Grada, Shift Supervisor R. Hansen, NSQC Engineer L. Hutchinson, Station QA E. Kurtz, QA Engineer F. Lipchick, Station QA A. Mazukna, QC Supervisor R. Prokopovich, Reactor Engineer L. Schad, Operations Supervisor J. Starr, Station Engineer R. Washabaugh, QA Manager J. Werling, Station Superintendent D. Williams, Results Coordinator H. Williams, Chief Engineer The inspector also interviewed other licensee personnel.
2.
Licensee Action Previous Inspection Findings (0 pen) Unresolved Item (79-20-02):
Preventive maintenance program for GE Type AK-2 Circuit Breakers in accordance with IE Bulletin 79-u9.
The licensee has issued Preventive Maintenance Procedures Nos.1-39DC-BAT-1-lE through 5-lE and 1-390C-BKR-1-lE and 2-IE, each titled Inspection of GE Type AK-2A Air Circuit Breakers.
Those procedures address the breakers identified in the licensee's bulletin response as being subject to preventive maintenance (as discussed in IE Inspection Report No. 334/79-20). The Preventive Maintenance Procedures were issued consistent with the schedule provided in the licensee's bulletin response letter and appear to be consistent with the technical guidance provided by the bulletin.
These procedures are scheduled for performance during the upcoming outage (December 1979), and this item will be re-examined to verify proper completio.
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(Closed) Unresolved Item (79 17-04):
Licensee to revise BVPS Operating Manual (0M) Section 1.49.4.M, Reference Guide for Estimated Critical Position (ECP) Calculations, to make the guidance consistent with the ECP-1 Foms currently in use.
The inspector reviewed Operating Manual Change Notice (OMCN) No.79-138, dated October 9,1979 which promulgated the revised OM Section.
The procedure is now consistent with the ECP-1 Fom and appears to provide adequate guidance for calculation of the ECP and associated data.
(C1csed) Inspector Follow Item (79-17-10): Corrective action for Licensee Event Report 79-26, Loss of RCS Flow.
The licensee's review of this event concluded that operator error was the proximate cause as discussed in the 14 day followup report. The inspector confirmed that the corrective action stated in the followup report had been implemented and concurred in its adequacy.
(Closed) Inspector Follow Item (79-16-10): Review actions taken in response to 10 CFR 21 Report from vendor of contairment hydrogen recombiners.
The vendor had reported a potential for recombiner power connector failures similar to that which occurred at Three Mile Island Unit 2 during recovery from the incident of March 28, 1979.
At the time of IE Inspection No. 50-334/79-16, the licensee was formulating plans for perfoming vendor recom-mended power connector inspections.
Those inspections were subsequently completed, and reported in Licersee Event Report No. 79-22 as discussed in paragraph 8 of this report.
Bas-d upon the actions described therein, this item is closed.
(Closed) Unresolved Item (79-04-04):
Cigarette smoking in radiologically controlled areas.
As described in IE Inspection Report No. 50-334/79-04, the licensee took immediate action to curtail the use of smoking materials within the station's controlled areas.
Those areas have been pariodically observed during subsequent inspections in order to detemine the effective-ness of the licensee's actions.
No new evidence of smoking has been iden-tified by the inspector.
Review of Plant Opuations a.
Shift Logs and Operatino Records Logs and operating records were reviewed against administrative pro-cedures to verify that:
Log sheet entries were filled out and initialed;
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Entries involving abnormal conditions were sufficiently
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detailed as to equipment status, lockout status, corrective action, and restoration; Log book reviews were being conducted by the staff;
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Operating orders and temporary procedures did not conflict
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with the Technical Specifications; Incident reports showed no violation of Technical Specification
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Limiting Condition for Operation (LCO) or reporting requirements; Jumper log entries did not conflict with Technical Specifications;
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and, Logs and records were maintained in accordance with
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Technical Specifications and the procedures below.
Acceptance criteria for the above review were inspector judgement, the Technical Specifications (TS), and the following procedures:
BVPS Operating Manual (0M) Chapter 48, Conduct of Operations;
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OM 1.48.3, Section H, Temporary Procedures;
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OM l.48.5, Section D, Jumpers and lifted leads;
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OM l.48.6, Clearance Procedures;
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OM 1.48.8, Records; and,
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OM 1.43.9, Rules of Practice.
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The following logs and records were reviewed for the periods indicated:
Gl-1 (Superintendent's Daily Record), Sl-1 through SI-9
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(Shift Operating Report), L1-1 through L1-5 (Reactor Operator Log), L5-13 through L5-15 (Surveillance Verification Log) for October 2 through flovember 4,1979 were reviewed periodically during this inspection.
Temporary Operating Procedures 79-27 through 79-38 covering
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the period from July 11, 1979 through October 3,197.
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Jumper and Bypass Log entries for the period from October
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2 through November 4,1979.
Special Operating Orders through Nc. 79-7 covering the
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period through September 28, 1979.
Inconsistencies were identified in the application of TS reporting requirements for operation of plant systems and equipment in degraded modes pennitted by TS LC0 action statements.
This matter is further discussed in paragraph 8 of this report.
b.
Plant Tours Inspection tours of the following plant areas were conducted at various times during this inspection, including backshifts:
Control Room (daily)
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Auxiliary Building
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Relay and Switchgear Rooms
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Penetration Areas
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Cable Vaults
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Diesel Generator Rooms Auxiliary Feed Pump Room
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Safeguards Pump Area
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Intake Structure
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Security perimeter, fences, and isolation zones The following determinations were made:
Control Room and local monitoring instrumentation were
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frequently observed to verify that instrumentation and systems required to support Mode 1 operation were in conformance with Technical Specification LC0 requirements.
The inspector verified that selected valves were positioned
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or in the condition required by Technica' Specifications for the arai:caale plant mod.
Radiation controls established by the licensee, including
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posting of radiation areas, the conditions of step-off pads, and the disposal of protective clothing were observed.
Selected radiation work pennits used for entry into radiation areas were reviewed.
Plant housekeeping conditions including general cleanliness
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conditions and control of materials to prevent fire hazards were observed.
Equipment danger and caution tags were inspected for proper
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posting and logging.
Systems and equipment in all areas toured were observed for
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the existence of fluid leaks and abnormal piping vibrations.
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Selected pipe hangers and seismic restraints were observed for proper settings / conditions.
Selected lit control board annunciators were discussed with
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control room operators to verify that the reasons for the alarmed conditions were understood and corrective action, if required, was being taken.
Control Room manning was observed frequently during the inspection.
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Shift turnovers were observed on several occasions to confinn that continuity of system status is maintained.
Acceptance criteria for the above items included inspector judgement, the requirements of 10 CFR 50.54(k), and the following procedures:
BVPS Unit 1, Systems Valve Lists and Valve Operating Drawing
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Markups; OM 1.48.5, Safety Related Systems, Valves, and Equipment:
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MM Chapter 1, Section J, Housekeeping;
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MM Chapter 1, Section H, Cleaning and Maintenance Cleaning;
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SAD 25, Housekeeping and Cleanliness Procedure; and,
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BVPS Radcon Manual, various section.
During review of Main Control Board (MCB) instrumentation and control status, the inspector noted that the MCB power range nuclear instrument meters were reading as much as 6% full power greater than the NI instrument rack indicators.
This difference is apparently due to the MCB indicaters being driven by an isolation amplifier which is not routinely recalibrated during NI calorimetric calibrations.
Calorimetric calibrations result in adjustment of the circuit's principal suming amplifier which provides signals to the NI rack indicators and all protective functions.
The above isolation implifier receives its input from the suming amplifier and will deviate from the NI rack indication if it is not adjusted af ter a summing amplifier gain adjustment.
Interviews with on shift operators indicated that the MCB indication is not recognized as the indication of record for operation but is considered to be an approximate power level value.
The operators interviewed consistently stated that all operations and surveillance were conducted using the NI rack indications.
Discussion with licensee and licensee contractor personnel indicated that a deviation will develop between the NI rack and MCB indicators if a calorimetric calibration results in an adjustment of 2-4% full power in the summing amplifier circuit.
The licensee stated that an evaluation of the above would be performed and documented in order to determine the magnitude of adjustment which could require recalibration of the isolation amplifier and would detennine the need for such adjustments to be routinely performed.
The inspector stated that the MCB indications should be as reliable as practicable in order to provide the operators with optimum indication.
This matter is unresolved. (334/79-22-01) The licensee stated that the matter would be reviewed by the Onsite Safety Committee by November 15, 1979.
c.
Reactor Trip Review On October 14, 1979, the plant experienced a Rod Control System Urgent Failure Alarm.
As an apparent result of the alanned condition, operators were unable to move Bank D, Group I Rods.
The problem appeared to be isolated to the Bank Overlap Unit in the rod control system and troubleshooting proceeded in order to determine the specific source of the problem. The ability to move the Bank D, Group I rods was restored but rod control system alarms persisted.
At 0110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> on October 16, 1979 the plant sustained a reactor trip
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from approximately 49% power.
At the time of the trip a Meter and Control Repairman was troubleshooting the Bank Overlap Unit circuitry.
The reactor trip was caused by a high negat',ve flux rate apparently due to a dropped rod.
Review by the licent e detemined that the action of the technician in itself did not cause the trip but the logic card which the technician was installing in the system had a faulty internal pin connection which apparently initiated the event.
The faulty card was replaced and the reactor was returned to power.
The inspector reviewed the circumstances above including:
Operator response to the reactor trip in accordance with Emerg-
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ency Procedure E-5, Reactor Trip and Alarm Response Procedure -
Rod Control System Urgent Failure Alam.
The effects and results of rod control system troubleshooting
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during the period October 14-16, 1979.
Compliance with applicable Technical Specifications LC0's during
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the post trip period and recovery.
Observation of the reactor startup performed on October 16, 1979
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in accordance with the BVPS OM, Section 1.50, Station Startup and verification of conformance with selected Technical Specifications.
No items of noncompliance were identified.
d.
Licensee Response to Increasino Ohio River Level On October 9-10, 1979 the Ohio River level at the plant had increased to the action level specified in Emergency Procedure E-10, Flood, Revision 16, which requires the licensee to increase river level surveillance frequency.
During this period the inspector confirmed the licensee's conformance to the above procedure and to the admin-istrative limits imposed by the licensee in accordance with agreements made with NRC:NRR pursuant to evaluation of soil liquefaction in the plant's North yard area. The river crested at 670.5 ft. MSL during the 0000-0800 hours shift on October 11, 1979 and then receded.
The review of implementation of the above procedures and controls iden-tified no items of noncomplianc.
4.
IE Bulletin Followup The inspector reviewed licensee actions taken in response to IE
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Bulletins, incluc'ing whether the written response:
was submitted within the required time period; included the information required including adequate corrective action commitments; and licensee management had forwarded copies of the response to responsible onsite management.
Inspector review also included discussions with licensee personnel and observations of items discussed below.
IEB 71 06C, Nuclear Incident at Three Mile Island - Supplement:
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The
.censee has revised Emergency Procedure E-0, ECCS Actuation (Revision 18, August 15,1979) to require tripping of all operating Reactor Coolant Pumps upon reactor trip and initiation of high pressure injection due to low RCS pressure.
The licensee also maintains control room manning consistent with the IEB direction.
The licensee's IEB response letter, dated August 28, 1979, states that IEB Items 2-5 will be accomplished in conjunction with their NSSS Vendor Owners Grcup activities.
The Owners Group is in communication with NRC:NRR.
Acceptability of the licensee's response to and implementation of Items 2-5 is unresolved pendin completion of NRC:NRR review of the Owners Group activities 334/79-22-02).
The inspector informed licensee management.that, if the schedule for completion of IEB activities provided by the Duquesne Light Company letter of August 28, 1979 could not be attained, a supplemental bulletin response must be submitted to the addressees specified by the IEB prior to exceeding the schedule dates.
Such a supplemental response should provide a revised schedule for completion.
IEB 79-14, Seismic Analysis for As-Built Safety-Related Piping
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Systems:
During this inspection the implementation of the licensee's as-built piping inspections was periodically reviewed.
The licensee's inspection program has not been completed and is continuing.
The following aspects of the licensee's program were reviewed or observed:
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DLC Inspection Procedure MEDFP-1, Inspection Procedure for Determining Agreement Between "As-Installed" Piping and "As-Built" Drawings, Revision,1, was reviewed and appears to satisfactorily implement the inspection requirements of the IE.
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The training provided to the licensee's piping inspectors was reviewed including documentation of completed training.
Current levels of manning and training appeared to be acceptable.
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Two in-process inspection packages were reviewed for Isometric Drawings Nos. 276 and 277.
The packages and data appeared to be in conformance with Procedure MEDFP-1.
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The results of inspections performed during March, June, and July 1979 as part of a prior seismic reanalysis effort were reviewed with licensee management to assess the significance of the nonconformances identified.
Although these inspections were not all performed using formal inspection procedJres and did not meet the requirements of the IEB for detailec' inspection, the licensee and his architect engineer consider then to be valid confirmation of as-built data used as inputs for the seismic reanalysis effort.
The licensee further acknowndged that the piping systems inspected during March, June, ano July 1979 would be reinspected in accordance with Procedure MELFP-1.
The inspector noted that these prior inspections had resulted in nonconformances which required seismic reanalysis but did not subsequently require modification in order to meet design requirements.
No systems had been found to be inoperable.
At the close of this inspection, no new nonconformances had been identified as requiring reanalysis or resulting in system inoperability.
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The inspector reviewed the licensee's planning for dispositioning nonconformances and was informed by the licensee that the architect-engineer has developed procedures for dispositioning nonconformances in all piping except that small bore (< 6" diameter) piping which was designed by hand calculation methods.
The inspector reviewed correspondence between DLC and the architect engineer which indicated that such procedures were in development.
The inspector infonned the licensee that, notwithstanding the above procedure status, any nonconformances must receive expeditious evaluation pursuant to the bulleti On the basis of the above inspection findings and review of the licensee's letter dated October 30, 1979, NRC:RI informed DLC on November 1,1979 that continued plant operation until
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December 17, 1979 was acceptable.
The facility will shut down for an extended refueling / modification outage prior to that date.
The inspector further informed the licensee that the report of inspection results required by the IEB should be submitted within six weeks of completion of inspections.
The inspection schedule was submitted to NRC:RI by the licensee's letter of October 30, 1979.
The licensee acknowledged the inspector's statement.
The licensee's activities pursuant to IEB 79-14 will continue to be reviewed during future inspections.
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IEB 79-15, Deep Draft Pump Deficiencies:
The inspector reviewed the licensee's response dated September 26, 1979 and a prior submittal to NRC:NRR on the same subject dated July 26, 1978.
Review confirmed that the infs: mation submitted complied with the requirements of the IEB. The licensee is planning to implement major modifications of the facility's deep draft ECCS pumps during the upcoming outage.
Those modifications are presently being reviewed by NRC:NRR and will subsequently be inspected by NRC:RI during their implementation.
Further, the information submitted pursuant to the IEB is being reviewed by an NRC Task Force.
The bulletin remains open pending completion of NRC review and inspection.
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IEB 79-18, Audibility Problems Encountered on Evacuation of Personnel from High Noise Areas: The inspector reviewed the licensee's bulletin response dated September 26, 1979, the results of an Audibility Survey performed during August 23 -
September 8, 1979, and the results of a resurvey conducted after correction of deficiencies on September 18, 1979.
The surveys identified deficiencies in office and yard area public address systems which require addition of speaker equipment.
This addition has been initiated as a design change to be implemented during the upcoming outage.
Numerous other deficiencies were noted in the maintenance of equipment and the adequacy of speaker volume settings.
The licensee has committed to establish a program which will periodically assure the audibility /
operability of speakers, particularly following extended plant outages.
The licensee stated that such a program would be established and implemented prior to plant restart from the upcoming outage.
This matter will be followed during future inspections.
(79-22-03)
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IEB 79-21, Temperature Effects on Level Measurements:
The
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inspector reviewed the licensee's response dated September 18, 1979 and the NSSS Vendor letter (DLW-79-29, dated August 29,
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1979) on which the licensee's response was based.
The NSSS Vendor provided the licensee with generic data for correction of steam generator and pressurizer level instrumentation for the effects of containment temperature, fluid pressure, and coolant density changes.
The licensee has developed plant specific data from the generic data and has found it to be more conserntive than the generic data.
The licensee's response not s that the temporary corrective action has been taken to adjust steam generator level trip setpoints upward but that final corrective action is still in development.
The inspector confinned the implementation of the temporary corrective action.
The licensee's response also noted that the procedure revisions and training required by IES Item 4 were in progress in conjunction with the NSSS Vendor Owners Group and that an implementation schedule is forthcoming pending NRC:NRR approval of the procedure reference guidelines.
The inspector informed the licensee that a supplemental response to the IEB should be forwarded to the addressees specified in the IEB providing the schedule for procedure revisions and the final corrective action for steam generator level instruments when they are available.
The licensee stated that these activities are scheduled for completion during the upcoming outage.
The acceptability of the licensee's response and final corrective actions remain unresolved.
(79-22-04).
During the above IEB reviews, the inspector noted that the information developed by station personnel in response to the IEB's and which is necessary to substantiate the licensee's sutmittals to NRC is not being consistently entered into the QA records system.
This includes data gained through plant records research, testing, review of plant operations, and calculations.
The inspector noted that the Power Station Engineering Group is presently collating such information and forwarding it to Station QA for retention.
Other station departments appear to be retaining their documentation in departmental working files and are not forwarding it for retention by the plant records syste.
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Additionally, the station does not appear to have an established method of checking or review of such information, data, or
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calculations prior to issuance of formal correspondence to NRC.
The licensee informed the inspector that the above concerns are being addressed as part of the commitment control system development as discussed in IE Inspection Report No. 50-334/79-20.
That conTaitment control system is expected to be implemented before the end of CY 1979.
The licensee stated that, in the interim, all departments would be required to submit such information to Station QA for retention and that additional attention would be paid to the accuracy and completeness of such information.
The acceptability of the licensee's actions in this regard will be reviewed during future inspections.
This item is unresolved (79-22-05).
5.
IE Circular Followup The inspector reviewed actions taken on IE Circular (s) in order to determine that the Circular was received by licensee management, that a review for applicability to the facility was performed, and that, for those applicable to the facility, appropriate corrective actions have been taken or planned.
The following Circulars were reviewed.
IEC 79-04, Loose Locking Nut on Limitorque Valve Operators:
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The circular was reviewed as dccumented in Onsite Safety Committee (OSC) Meeting Minutes No. BV-0SC-13-79, March 27, 1979.
As a result of that review, Corrective Maintenance Procedure No. 1-75-150, Valve Stem Locking Nut Inspection for Limitorque Motors, Revision 2, was established and implemented, resulting in the inspection of 197 valves.
Twelve of the valves inspected, required corrective action to permanently stake the nuts.
All others were either already staked or did not require staking.
The inspector reviewed the maintenance data for the valves and 'he NSQC General Inspection Report for the activities and had no n ether questions.
IEC 79-09, Occurrences of Split or Punctured Regulator Diaphragms
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in Certain Self-Contained Breathing Apparatus: The licensee reviewed this circular as documented in OSC Meeting Minutes No. BV-0SC-48-79 dated August 1, 1979, inventoried all self-contained breathing apparatus in use at the facility, and concluded that no units of the type noted in the IEC are onsite.
The inspector confirmed, on a sampling basis, that no such units are available for use in the plant space.
Licensee Plans for Coping with Strikes Prior to this inspection, the licensee's contract with the labor union which represents nearly all non-salaried workers at the station, including licensed operators, had expired.
Normal manning and operations were being maintained via extensions to the expired contract.
On October 10, 1979, the labor union ratified a new contract without a strike taking place.
During October 2-10, the inspector reviewed the licensee's plans for coping with a potential strike.
The planning was reviewed using the following documents for guidance and/or acceptance criteria:
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BVPS Emergency Plan; Technical Specifications, Section 6.2;
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BVPS OM, Section 1.56.4, Emergency Squad;
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ANSI N18.1 - 1971, Selection and Training of Nuclear Power
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Plant Personnel; and, 10 CFR 55, Appendix A, Licensed Operator Requalification
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Program The following were reviewed:
BVPS Emergency Manning Procedure (Strike Contingency Plan),
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dated October 1, 1979; Station Manning List for Strike, dated October 1,1979; and,
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Station training records pursuant to ANSI N18.1-1971,10 CFR
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55, Appendix A, and the BVPS Emergency Plan for those staff members identified on the Station Manning List for Strike who would be performing specific plant duties other than those which they are normally assigned.
Additionally, discussion with station management indicated that arrangements for outside services were adequate for normal and emergency operations.
No items of noncompliance were identifie.
7.
Review of Licensee Outage Planning The licensee is planning an extended outage which will commence on
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or about December 1,1979 and extend into the sumer of 1980.
During this outage the licensee plans to implement approximately 40 major design changes as well as perform the facility's first refueling.
The outage will include major involvement of the station staff, the Engineering and Construction Division, the licensee's architect engineer, and a major construction contractor.
Dcring this reporting period the inspector began a continuing review of the licensee's plannir.g an:t staging of outage activities, to include the following:
Planning and scheduling of major outage activities including
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the avai'. ability of procedures, installation instructions, and materiais; Review of plant modifications packages for design changes
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scheduled for implementation during the outage; and,
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The status of planning of QA/QC activities to be implemented by the OLC QA Department, Nuclear Services QC, and contractor QC.
At the close of this inspectio'n, a significant fractio of the documents and material necessary to support the outag' were not yet available at the site.
The licensee appears to be ma ataining detailed status of such items and considering them..dividually in the scheduling process.
Inspection of these activities is continuing and will be documented in subsequent reports.
8.
In Office Review nf Licensee Event Reports (LER's)
a.
The inspector reviewed LER's submitted to the NRC:RI office to verify that the details of the event were clearly reported, including accuracy of the description of cause, adequacy of corrective action, whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup.
The following LER's were reviewed:
- -- 79-16/03L - Incorrect Routing of Motor Operated Valve Power Cable; 79-17/03L - Inoperable Motor Operated LHS: Pump Suction
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Valve;
- -- 79-18/03L - No Residual Heat Removal Pumps Operable;
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79-19/0lT - Use of Unqualified Cable In Containment;
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79-20/0lT - Inoperable CVCS System Piping Support Baseplate;
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79-21/01T - Degradation of Emergency Bus 4160/480VAC
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Transformer Insulation Resistance; 79-22/03L - Containment Hydrogen Recombiner Potentially
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Inoperable; Special Report - One Fire Suppression Header Inoperable
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Due to Leakage; 79-23/03L - No. 1 Diesel Generator Output Breaker Failure
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to Close; 79-24/0lT - Environmental Qualification Deficiencies for
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Solenoid Valves; 79-25/03L - Channel 2 Delta-T/Tave Instrument Loop Inoperable;
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- -- 79-26/03L - Overpower Delta-T Loop Not Calibrated with Current Data;
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79-27/03L - Control Room Air Conditioning Seismic Support
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Removed;
- -- 79-28/0lT - Tripping of All Three Reactor Coolant Pumps on Underfrequency; 79-29/03L - Personnel Error Resulting in 1C Charging Pump
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Inoperability 79-30/03L - Missed Surveillance Test for Containment
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Airlock; 79-31/03L - Emergency Diesel Generator Loading Sequencer
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Interval Out of Time Tolerance; 79-32/03L - 480V Emergency Stub Bus Undervoltage Protection
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Out of Service;
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79-33/03L - NIS Power Range Low Setpoint Trip Logic
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Failure; 79-34/03L - Loop B Delta-T/Tave Instrumentation Inoperable;
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79-35/03L - Inoperable Hydraulic Snubbers;
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- -- 79-36/03L - lA LHSI Pump Recirculation Flow Less than Design; 79-37/03: - No. 4 Vital Bus Inverter Inoperable; and,
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- -- 79-41/03L - Rod Position Indication Inoperable
- Report selected for onsite followup.
Except as noted below, the inspector had no other sprific questions on the above reports:
79-23/03L - The report states that the failures reported are
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being evaluated by DLC Engineering.
The licensee stated that this matter is being reviewed in conjunction with the overall reliability of emergency electrical system (documented in IE Inspection Report No. 50-334/79-16; Unresolved Item No. 79-16-04)
and that a supplemental report will be submitted for LER 79-23 at the completion of those evaluations.
This item will be followed during future insnections (79-22-06).
79-29/03L - This report discussed mispositioned valves
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which resulted in charging pump inoperability but failed to document the damage sustained by the pump and the action taken to correct the damage.
The inspector requested the licensee to submit a supplemental report discussing these aspects of the reported event.
The licensee stated that the supplemental report would be issued on or before December 7, 1979.
This matter remains unresolved.
(79-22-07).
Special Report (Fire Suppression Water Header Leakage) -
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This report discussed a rupture of underground piping in the plant yard an event which rendered one of the redundant headers inoperable. The #
tial report submitted was intended by the licensee w satisfy prompt and fourteen day reporting requirements but did not detail the proximate cause of the leakage, failure mode of the piping, action
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taken to prevent recurrence or evaluation of the event with respect to other similar piping in the redundant
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The inspector requested the licensee to subnit a supplenental report discussing the above.
The licensee stated that the report would be subnitted on or before December 7,1979.
This matter renains unresolved.
(79-22-08)
During review of the above reports, the inspector noted that the reporting guidance of NUREG-0161, Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER) File, was not being consistently followed.
In a series of discussions with licensee personnel associated with the preparation and review of LER's the inspector stated that the following cannents were generic to most of the reports with respect to the NUREG-0161 g uidance:
The reports do not consistently reference previous events of a
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similar nature; The reports frequently do not include a description of actions
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taken to prevent recurrence of the event; The measures taken to assure that similar canponents at the facility
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are acceptable are frequently not stated; The root cause of the pro 5len or failure is not always clearly
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identified; and, The probable consequences of the event are occasionally anitted
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fran the reports.
The inspector noted that liaison with the station in regard to specific h
reports has generally established t at the infonnation not reported has been addressed internally by the licensee in an acceptable manner and thus does not present a safety problem.
The licensee acknowledged the inspector's comments.
This will be reviewed during future inspections (79-22-09).
b.
Reportability of Degraded Mode Operations During routine review of operating logs and records and plant incident reports, the inspector noted that the licensee's evaluations of reportability for operation in degraded modes permitted by Technical Specification Action Statements was not consistent with existing NRC policy.
Technical Specification 6.9.1.9.b requires that written 30 day reports (LER's) be
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submitted to NRC for:
" Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or..."
Established NRC guidance does not require such reports to be submitted for voluntary entry into degraded mode operation such as when an action statement is entered because a piece of equipment is intentionally removed from normal service for testing or preventive maintenance.
Except for surveillance testing and preventive maintenance, whenever a parameter or system enters an action mode described in the related Technical Specification LCO, no violation of the specification has occurred, but a 30 day written report is required.
The licensee was informed of this distinction and requested to review his operations in light of the above position.
The licensee has identified a number of technical specifications which are considered to be inappropriate subjects for application of the above reporting requirements.
These include secondary chemistry transient limits, axial flux differenc-limits, rod position indication operability, various instrun,entation LCO's and others.
Some of the above LC0's are currently in the process of amendment and the licensee's specific requests for relief from reporting requirements should be submitted to NRC:NRR on a timely basis for incorporation into TS.
The inspector informed the licensee that, until the TS are changed, degraded mode operations must be reported in accordance with existing requirements and guidance.
This matter will be reviewed during future inspections.
(79-22-10).
9.
Onsite Licensee Event Followuo For those LER's selected for onsite followup (denoted by asterisks in paragraph 8), the inspector verified that the reporting requirements of the Technical Specifications and Procedures SAD 14 and SAD 23 had been met, that appropriate corrective action had been taken or planned, that the event was reviewed by the licensee as required by Technical Specifications and procedure SAD 21, and that continued operation of the facility was conducted in accordance with Technical Specifications and did not constitute an unreviewed safety question as defined in 10 CFR 50.59(a)(2).
The following findings relate to the LER's reviewed onsite.
LER 79-16/03L - This LER documented the discovery and correction
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of incorrectly routed power cables for motor operated valves MOV-SI-8608 and 8648.
The inspector reviewed Design Change
Package No. 256; Corrective Maintenance Procedure 1-75-161; Revision 4, Maintenance Work Request No. 790222 completed August 3,1979; NSQC General Inspection Report, dated August 3,1979; and NCAR Nos.172 and 177 which doctment the corrective action taken.
The action planned to prevent recurrence involves implementation of a sampling verification for other cables which is to be performed during the upcoming outage.
The inspector reviewed the " Sampling Procedure for Category 1 Cable Routing" issued by DLC Engineering on August 13, 1979.
The procedure provides for a statistical sampling plan based upon MIL-STD-105D.
The sampling plan is based on the licensee's finding that both incorrectly routed cables were installed and inspected by the same construction and QC individuals during initial plant construction.
The inspector infomed the licensee that a supple ental report to this LER must be submitted upon completion of the above inspection program and any associated followup activities.
This matter will be followed during future inspections (79-22-11).
LER 79-18/03L - Previous review of this event is doctmented
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in Inspection Report No. 50-334/79-16.
As noted in that report, procedural changes were made to prevent recurrence and additional study of the reliability of energency ;1ectrical systems is in progress.
The inspector noted, durirg this inspection, that the procedural changes appear to be effective ia eliminating inadvertant safety system actuation signals and had no further questions pending the outcome of the licensee's engineering evaluation.
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LER 79-25/03L - This event involved the attempted calibration of the Channel 3 Overpower Delta-T instrtment with incorrect da ta.
The inspector's review detemined that one of the instrtment loop's RTD's had been changed prior to the event and had resulted in new loop calibration data being generated.
This information was included in draft revision to Maintenance Surveillance Procedure 6.40 which is the loop calibration procedure.
Prior to this revision being issued, the loop mal functioned and, as reported in the LER, the old RTD data was used in an attempt to calibrate the instetment.
The error was identified by the technician and the correct calibration data sheet was provided for use.
At the close of this inspection, the inspector's review of the event was still in progress.
The time necessary to incorporate data changes into maintenance procedures and make them available to working personnel appears to be an area for additional licensee action.
This matter will be reviewed during subsequent inspections (79-22-12).
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LER 79-28/0lT - This event was previously reviewed onsite
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(Inspection Report No. 50-334/79-17).
The inspector confirmed
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that the licensee's corrective action had been implemented and appeared effective through discussions with onshift operations personnel. The inspector had no further questions on the matter.
LER 79-36/03L - This event was previously reviewed onsite
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(Inspection Report No. 50-334/79-20).
The licensee performed Operating Surveillance Tests on the pump daily for the first week following the event and then weekly for the subsequent four weeks.
To date, no additional foreign material has been found nor has pump operability been affected.
This matter will be followed up as discussed in IE Inspection Report No.
50-334/79-20.
LER 79-41/03L - The inspector observed portions of this event
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while it was in progress on October 4, 1979.
The licensee's response to the multiple Rod Position Indicator Channel inoperability was completed consistent with the requirements of Technical Specification 3.1.3.1.
The inspector, however, observed that the licensee did not have a specific procedure for operator guidance in the case of such multiple RPI anoma'ies. With greater than one RPI providing anomalous readings, Technical Specification 3.0.3 must be invoked and a plant shutdown performed within one hour if the condition is not corrected.
The licensee issued Operating Manual Change Notice No.79-149 on October 25, 1979 which promulgated OM Section 1.1.4.S, Abnormal Procedure for Rod Position Indication Malfunctions which now provides a diagnostic approach to identifying and correcting multiple RPI channel drifts or malfunctions.
Additionally, during the event the licensee did not clearly establish the time at which the Technical Specification action statement was entered, resulting in an additional possibility for violating the LC0 through continued operation past the LC0 time limit.
The inspector informed the licensee that the action statement is in effect when the anomalous RPI indications are noted, not when the cause for the anomalies is identified, i.e. at the time the operator notes that RPI and Group Demand position disagree, not at the time when the operator determines whether rods are actually misaligned or the RPI is inoperable.
The licensee acknowledged the inspector's statement.
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10.
Review of Periodic Reports The inspector reviewed the following periodic reports to verify, as
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applicable, that the information required to be reported by NRC requirements had been included, that supporting information discussed in the reports is consistent with design predictions and performance specifications, that planned corrective action is adequate for resolution of identified problems, and whether any information included in the report should be classified as an abnormal occurrence.
The reports reviewed were:
Monthly Operating Reports for July, August, and September 1979
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No items of noncompliance were identified.
11.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable, items of noncompliance or deviations.
Unresolved items addressed during this inspection are discussed in paragraphs 2, 3, 4, 8, and 9 of this report.
12.
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 also provided to the licensee at the conclusion of the report period.