IR 05000346/1994011
| ML20149H362 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/10/1994 |
| From: | Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20149H351 | List: |
| References | |
| 50-346-94-11, NUDOCS 9411220147 | |
| Download: ML20149H362 (10) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION III
Report No. 50-346/940ll(DRP)
Docket No. 50-346 Operating License No. NPF-3 Licensee: Toledo Edison Company Edison Plaza, 300 Madison Avenue Toledo, OH 43652 Facility Name:
Davis-Besse Nuclear Power Station inspection At: Oak Harbor, Ohio inspection Conducted:
September 14, 1994, through October 25, 1994 Inspectors:
S. Stasek J. Shine D. Kosloff D. Shepard Approved By:
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R. D. Lanksbury, W Date Reactor Projects Section 3B
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In_sp_ection Summary Inspection on September 14. 1994. through October 25. 1994 (Report No. 50-346/940ll(DRP1)
Areas Inspected: A routine safety inspection by resident inspectors of plant operations, surveillances, maintenance, onsite engineering, plant support, and previous inspection findings j
i Results: An executive summary follows-
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i Plant Operations: Overall, performance of the operating crews was good this
inspection period. Adherence to administrative procedures was adequate.
The plant shutdown conducted on October 1 for start of the refueling outage was conducted in a controlled, conservative manner (Section 1.2)
The October 4-6 drain-down of the reactor coolant system to support steam generator nozzle dam installation was also conducted in a like manner.
However, on October 8, maintenance personnel conducted a lift of a spent fuel pool gate without notifying operations. As a result, the lift was made during a time when the auxiliary building truck bay rollup door was open, thereby impacting the ability of the emergency ventilation system to properly respond as required by Technical Specifications (Section 1.4).
Foreign material exclusion (FME)
centrols were found to be inadequately implemented for the reactor cavity / refuel canal area (Section 1.6).
This situation resulted in issuance 9411220147 941110
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PDR ADOCK 05000346 G
l of a violation.
Refueling related problems involving mispositioning of a fuel assembly, improper control rod positioning, and a failure of a refuel grapple to properly disengage from a control rod as required were identified by plant personnel with inspector review ongoing at the conclusion of the inspection period (Section 1.7).
Maintenance: Overall, surveillance and maintenance activities reviewed this inspection period were conducted in accordance with licensee procedures and regulatory requirements.
However, the methodology used to functionally test the low voltage trip setpoint for the input breaker associated with inverter YVI appeared susceptible to erroneous results and is to be further evaluated.
An inspection followup item was initiated to track this matter (Section 3.1).
Enqineerina:
Engineering support of routine unit operation as well as during the refueling outage was good. One issue associated with compaction tests associated with the dry cask storage pad backfill having been found to be less than the established acceptance criteria was identified (Section 4.1).
The pad construction was subsequently determined to be of acceptable quality.
However, the quality of certain associated documentation and the retrievability of those documents was less than desirable.
At the end of the inspection period, the licensee was completing the remaining documentation associated with the pad construction.
Plant Support:
Overall, personnel adherence to the radiation protection and security programs was good during this inspection period.
However, two radiological events occurred this period that were subsequently reviewed by Region III radiation protection specialists (and documented in inspection report 346/94010(DRSS)).
The first involved an unplanned exposure of several contract workers to airborne radioactivity during insulation removal work in containment, and the second involved the external exposure of personnel to unexpectedly high dose rates during draining of the incore instrument tank.
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w DETAILS
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1.0 Plant Operations (71707) (92901)
The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during the inspection period.
The inspectors verified the operability of selected emergency systems, reviewed tagout records, and verified tracking of limiting conditions fur operation (LC0) associated with affected components.
Tours of the containment, auxiliary and turbine buildings were conducted to observe equipment material condition and plant housekeeping. Walkdowns of the accessible portions of the following systems were conducted to verify operability by comparing system lineups with plant drawings, as-built configuration, or present valve lineup lists; observing equipment conditions that could degrade performance; and verifying that instrumentation was properly valved, functioning, and calibrated:
Emergency Diesel Generator 1-1
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Emergency Diesel generator 1-2
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Decay Heat Removal System - Train 1
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Decay Heat Removal System - Train 2
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The inspectors reviewed potential condition adverse to quality (PCAQ)
reports during the inspection period and verified known deficiencies were identified and tracked via the PCAQ reporting system.
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One violation associated with inadequate foreign material exclusion controls was identified as discussed in Section 1.6 below.
1.1 Operations Summary During the first part of the inspection period, the unit was in final power coastdown to the ninth refueling outage. On October 1, 1994 the unit shut down for start of the outage.
At the end of the inspection period, the unit remained in cold shutdown as the outage continued to progress.
1.2 Control Room Activities Control room activities were well controlled and conducted in a conservative manner.
Procedure use was good.
Control room communications were clear and utilized repeat backs when necessary.
Operators remained cognizant of control panel indications.
Shift briefs were complete and addressed the priorities for the shift.
Operator performance during the October 1, 1994, shutdown evolution was excallent.
The portions of the shutdown evolution observed by the inspector were conducted in a controlled manner and in conformance with applicable requirements.
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1.3 Reactor Coolant System Drain-Down
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On October 4-6, 1994, operations performed a drain-down of the reactor coolant system (RCS) per operating procedure DB-0P-06904, Shutdown Operations, culminating in a reactor water level reduction to 18 inches above the reactor vessel hot leg centerline elevation. This was done to allow installation of once through steam generator (OTSG) primary side nozzle dams in both OTSG lower heads.
The following was verified available during tne evolution:
redundant electrical power supplies (offsite AC sources as well as diesel generators), containment integrity and short-term hatch closure capability, redundancy of RCS level instrumentation, redundancy and operability of core temperature monitoring instrumentation, availability and redundancy of RCS inventory makeup, and existence of an adequate RCS vent pathway.
No substantive concerns were noted. The evolution was conducted in a controlled, conservative manner. At one point during the drain-down, the safety parameter display system (SPDS) became inoperable due to someone in the plant stepping on a cable; the control room SR0 stopped the drain-down until the necessary level indication was thereafter
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restored.
A subsequent discrepancy in indicated RCS level was resolved after identification that tygon tubing, installed to allow a vent path through a poly bottle filter arrangement, had collapsed causing a slight vacuum within the reactor vessel. The discrepancy was conservative in nature in that indicated level was approximately 10 inches lower than the actual level.
Once the tygon tubing was disconnected, indicated level more accurately reflected actual level, and the drain-down was then continued.
1.4 Spent Fuel Pool Gate Lift On October 8, 1994, maintenance personnel conducted a lift of a spent fuel pool (SFP) gate over the SFP without adequately communicating their
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actions to the control room.
As a result, during the time of the lift, the auxiliary building truck bay rollup door was open.
This appeared to be in violation of Technical Specification 3.9.12., that required at least one train of the emergency ventilation system (EVS) to be operable during any heavy lifts over the SFP.
For the EVS system to be operable, it must be capable of maintaining a 1/8 inch vacuum.WG in the SFP area.
With the rollup door open, the EVS system would not have been capable of maintaining the required vacuum.
At the end of the inspection period, the licensee was preparing a licensee event report (LER) on this event.
Further inspector review of this matter will be conducted following submittal of the LER.
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1.5 Temporary Instruction 2515/125, Foreign Material Exclusion Controls
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A review of the licensee's foreign material exclusion controls was performed during the inspection period utilizing Temporary Instruction (TI) 2515/125, Foreign Material Exclusion Controls. Administrative
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Procedures DB-MN-00005, Foreign Material Exclusion; DB-PN-00007, Control of Work; DB-MN-00000, Conduct of Maintenance; and associated reference procedures were reviewed.
It was determined that the licensee's program for FME was applicable to fluid systems, as well as electrical and electronic enclosures and any other area during any maintenance, modifications, or inspection activities.
The inspector concluded that appropriate programmatic controls were in place to adequately perform activities requiring establishment and use of FME controls. However, cases of implementation inadequacies had occasionally been identified in the past (reference inspection report 346/94006 and Section 1.6 below).
TI 2515/125 is closed.
1.6 Inadequate Foreign Material Exclusion Controls On October 11, 1994 during a containment walkdown, inadequate foreign material exclusion (FME) controls were observed associated with the reactor cavity / refueling canal area. The FME boundary was not strictly adhered to in that materials (i.e., plastic bags) were seen laying under the rope barrier, halfway into the FME area. A portion of a walkway railing was also being utilized as the boundary. An individual within the FME area who was questioned as to where the boundary actually was indicated that he was unsure whether the walkway was part of the FME area. Additionally, the FME access log was not adequately maintained in that material and personnel were not properly logged.
The material not properly logged included rags, plastic bags, and some amount of electrical cabling.
Once identified, the licensee took adequate corrective actions to return the FME controlled area to an acceptable condition.
Actions were taken to better define the boundary and to en:;ure adherence to it.
The access log was updated to more accurately reflect actual materials and personnel in the FME area.
In addition, an individual was assigned fulltime to control access to the FME area, and to maintain the access log.
Subsequent evaluation of the reactor cavity / refuel canal FME area identified no further concerns.
However, administrative procedure DB-MN-00005, Foreign Material Exclusion, required the reactor cavity and refuel canal be established as a cleanliness level B area.
To maintain this, the licensee designated the area as a FME exclusion zone 3 area.
Exclusion zone 3 requirements included, as a minimum, establishment of a FME boundary around the area and access control for personnel and materials entering the area.
The controls to accomplish this included ropes and tape barriers, signs around the work area, and use of a material and personnel accountability log.
Because the established FME boundary was not adequately respected to preclude unaccounted material from entering
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the work area, and the material and personnel accountability log was not
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accurately maintained, this is a viulation for inadequate adherence to an approved administrative procedure (346/94011-01(DRP)).
1.7 Refueling Related Events During reactor refueling activities, three events occurred that involved errors in the handling of fuel or control components.
In the first case, on October 20, 1994, a periphery fuel assembly was inadvertently mispositioned in the reactor core. This apparently occurred as a result of two indeper. dent personal errors by the refueling crew. During the same timeframe, a sequencing problem occurred where a control rod was inadvertently removed from the core at the same time two other control rods were out of position.
Both of these conditions were inconsistent with the assumptions of the B&W core reload analysis governing the refueling activities utilized to maintain adequate shutdown margin.
Subsequently, B&W performed a re-analysis utilizing the actual core configuration during this time and concluded that the shutdown margin was well within previously analyzed acceptance criteria.
A third event occurred on October 21, 1994, involving a failure of the control rod mast on the auxiliary fuel handling bridge to disengage from a control rod after placement of the control rod into its core location.
This was identified visually when the mast was raised following placement of the control rod and after normal grapple disengagement actions had been completed.
The control rod was then reinserted into its core position and was disengaged from the grapple by shaking the mast cables.
Inspector review of these events was not complete at the end of the inspection period. These will be considered in the aggregate as an unresolved item (346/94011-02(DRP)) pending completion of that review.
A review of the final core verification video did not identify any mispositioned fuel or components (reference Section 1.8 below).
1.8 Cycle 10 Core Verification Video The inspector reviewed the licensee's core verification video that had been prepared as a record of the as-left core configuration for the next operating cycle. All fuel assemblies, control rods, and burnable poison rod assemblies (BPRAs) were verified to be in their proper core location by comparing the serial numbers from the video to the approved cycle 10 core map. No discrepancies were noted.
2.0 Surveillance (61726)
The inspectors observed safety-related surveillance testing and verified tfat the testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting condition for operation (LCOs) were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specification and procedure requirements and were reviewed by personnel
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other than the individual directing the test, and that any deficiencies
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identified during the testing were properly reviewed and resolved by appropriate management personnel.
The following test activities were observed and/or reviewed:
DB-MM-03001, Main Steam Safety Valve Setpoint Test.
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DB-PF-03008, Containment Local Leakrate Test - Penetration 50.
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DB-SC-03080, Emergency Diesel Generator 1 Overspeed Trip Test.
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08-5C-03076, Emergency Diesel Generator 184 Day Test.
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08-SC-03081, Emergency Diesel Generator 2 Overspeed Trip Test.
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DB-SP-03219, HPI Pump 2 Quarterly Pump and Valve Test.
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No violations or deviations were identified in this area.
3.0 Maintenance (637021 Station maintenance activities of safety-related systems and components were observed and/or reviewed during the inspection period to ensure that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with technical specifications.
The following items were considered during this review:
the limiting conditions for operation (LCO) were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and fire prevention controls were implemented.
Maintenance work orders (MW0s) were reviewed to determine status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which could have affected system performance.
The following maintenance activities were observed and/or reviewed:
MWO 3-94-0728-03, Emergency Diesel Generator 2 Preventive
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Maintenance.
MWO 3-94-0698-01, YVl/YRf1 Electrical Preventive Maintenance.
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MWO 2-91-0052-36, Install Manway Access in Service Water Piping.
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MWO 3-94-3113-01, Backfill RCS Midrange Level Transmitter
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LT10577A.
No violations or deviations were identified in this area.
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3.1 Essential Inverter YVI Testing During preventive maintenance testing associated with the Essential Inverter YVI input breaker, the breaker appeared to open at a less than allowable voltage.
The test consisted of slowly decreasing the input voltage to the subject breaker until the breaker opened. The nominal value of the expected trip voltage was about 100 Vdc with the initial data indicating a trip value of about 80 Vdc.
However, the trip setpoint was subsequently determined to incorporate a time delay setting as well.
Since this was not initially recognized, the rate at which the input voltage was reduced during the test could have resulted in the trip setpoint not being adequately verified.
This matter is considered an inspection followup item (346/940ll-03(DRP)) pending completion of inspector review.
4.0 Onsite Engineerina (37551)
Selected engineering problems r events were evaluated to determine their root cause(s). The effectiveness of the licensee's controls for the identification, resolution, and prevention of problems was also examined.
The inspection included review of areas such as corrective action systems, root cause analysis, safety committees, and self assessment in the area of engineering.
No substantive concerns were noted during this review.
No violations or deviations were identified in this area.
4.1 Dry Cask Storage Pad Compaction Testing A review by Quality Assurance personnel of the completed compaction testing associated with the backfill for the dry cask storage paa identified five instances where the data was below the est ablished acceptance criteria.
The review was done after the concrete for the pad was poured. Subsequently, the licensee determined that no technical issue existed, i.e., the backfill was sufficiently compacted and the overall testing results supported that the pad was constructed to acceptable standards.
The five tests involved widely dispersed points throughout the pad and at different elevations and were all surrounded by acceptable test results.
A total of approximately 200 compaction tests were conducted during the backfill operation.
More at issue was the quality of the associated documentation as well as the retrievability of the documents.
At the end of the inspection period, the licensee was in the process of completing the final set of documents to support the acceptability of pad construction.
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5.0 Plant Support (71750)
Selected activities associated with radiological controls, radiological effluents, waste treatment, environmental monitoring, physical security,
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emergency preparedness and fire protection were reviewed to ensure
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conformance with facility procedures and regulatory requirements.
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No violations or deviations were identified in this area.
5.1 Radiological Events During the inspection period, two radiological events occurred that were subsequently reviewed by two Region III radiation protection specialists.
The first event occurred October 7,1994, and involved an unplanned exposure of several contract workers to airborne radioactivity during insulation removal work in containment.
The second event involved the external exposure of personnel to unexpectedly high dose rates during draining of the incore instrument tank on October 23, 1994.
Further review of these events are documented in inspection report 346/94010(DRSS).
6.0 Followuj of Previous Inspection Findinas (92901) (92902) (92903) (92904)
(Closed) Inspection Followup Item (346/93006-04(DRP)):
Containment air cooler (CAC) leakage.
Inspections conducted of the CACs during the i
current refuel outage did not identify any significant degradation of the CACs or associated piping.
Since the CACs appear to be acceptably maintained, this item is closed.
(Closed) Unresolved Item (346/94005-03(DRP)): Adequacy of channel functional testing. Subsequent review determined that the licensee's method of performing functional tests of the subject equipment was acceptable.
They remained in compliance with NRC licensing rcquirements as well as standard industry practice. This item is closed.
7.0 tLaDa_gement Visit On September 26-27, 1994, an NRC management visit to the site was conducted.
Region III participants included the Regional Administrator; Director, Division of Radiation Safety and Safeguards; Acting Director, Division of Reactor Safety; and Chief, Reactor Projects Section 3B. The Licensing Project Manager, representing the Office of Nuclear Reactor
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Regulation (NRR), also participated.
The visit included a tour of the facility and meetings with various individuals from the licensee's staff.
8.0 Exit Interview The inspectors met with licensee representatives (denoted in Section 8.0)
throughout the inspection period and at the conclusion of the inspection on October 25, 1994, and summarized the scope and findings of the inspection activities. The licensee acknowledged the findings.
After discussions with the licensee, the inspectors determined there was no proprietary information contained in this inspection report.
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9.0 Persons Contacted
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Toledo Edison Company.
- J. P. Stetz, Vice President, Nuclear G. A. Gibbs, Director, Engineering
- S. C. Jain, Director, Nuclear Services
- J. K. Wood, Plant Manager T. J. Myers, Director, Nuclear Assurance
- J. W. Rogers, Manager, Maintenance J. Dillich, Manager, Radiation Protection S. Byrne, Manager, Plant Operations
- J. E. Moyers, Manager, Quality Assessment
- J. L. Michaelis, Manager, Materials Management
- R. C. Zydock, Manager, Nuclear Engineering
- G. Skeel, Manager, Security W. T. O'Connor, Manager, Regulatory Affairs
- G. R. McIntyre, Supervisor, E/C Systems
- P. W. Smith, Supervisor, Compliance
- D.
L. Miller, Sr. Engineer, Licensing
- J. Hartigan, Acting Manager, Design Engineering
- D. W. Schreiner, Supervisor, ISE
- L. A. Bonker, Supervisor, Alara Services
- L. D. Myers, Shift Supervisor
- W. J. Bentley, Assistant Shift Superviser
- D. Eshelman, Superintendent, Operations
- A. Rabe, Supervisor, Quality Engineering
- J. Wissner, Supervisor, Maintenance Planning
- G. T. Duncan, Training Manager State of Ohio
- J.
Colleli, Ohio Dept. of Health
- Denotes those licensee per<onnel attending the October 25, 1994, exit meeting.
10.0 Definitions Inspection Followup Items Inspection followup items are matters that have been discussed with the licensee, which will be reviewed further by the inspectors, and which involve some action on the part of NRC or licensee or both. An Inspection followup :'mn disclosed during the inspection is discussed in
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Section 3.1.
Unresolved Items An unresolved item is a matter requiring more information in o*rder to ascertain whether it is an acceptable item, a violation, or a deviation.
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An unresolved item was identified in Section 1.7.
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