IR 05000346/1994013

Insp Rept 50-346/94-13 on 941025-1208.Violations Noted. Major Areas Inspected:Plant Operations,Surveillances,Maint, Onsite Engineering,Plant Support & Licensee Event Repts

ML20149J201
Person / Time
Site:	Davis Besse
Issue date:	12/28/1994
From:	Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20149J155	List: IR 05000346/1994013 ML20149J151 ML20149J185
References
50-346-94-13, NUDOCS 9501050076
Download: ML20149J201 (12)

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U. S. NUCLEAR REGULATORY COMMISSION REGION III I

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Report No. 50-346/94013(DRP)

Docket No. 50-346 Operating License No. NPF-3 Licensee: Toledo Edison Company Edison Plaza, 300 Madison Avenue Toledo, OH 43652 l Facility Name: Davis-Besse Nuclear Power Station Inspection At: Oak Harbor, Ohio

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Inspection Conducted: October 25, 1994,'through December 8, 1994 Inspectors: S. Stasek  !

C. Lipa l l D. Shepard Approved By: (C 12hSje 1

R. D. Lanksbury,'Ctrfef Dat'e l Reactor Projects Section 3B l

t InsDection Summary l

Inspection on October 25. 1994. throuah December 8. 1994 (Report No. 50-346/94013(DRP))

Areas Inspected: A routine safety inspection by resident inspectors of plant operations, surveillances, maintenance, onsite engineering, plant support, previously identified inspection findings, and licensee event reports.

l Results: An executive summary follows:

Plant Operations: Overall, performance of the operating crews was very good rhls inspection period. Operator cognizance of plant conditions and panel ,

todications, especially during the refueling outage, was good. Reactor '

coolant system drain down evolutions conducted during the outage, were performed in a controlled manner. The unit restart following completion of the outage was likewise conducted in a controlled, conservative manne However, inadequate implementation of foreign material exclusion controls for the reactor cavity / refuel canal area was again evidenced (Section 1.3). The FME material accountability log indicated at the time the reactor was being reassembled that some items were unaccounted for. This was not identified until several days later, thus defeating the purpose of maintaining the lo Fortuitously, the log was able to be reconciled after the fact. Additionally, 9501050076 941228 PDR ADOCK 05000346 G PM

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an FME requirement to perform a closeout inspection of the area was incorrectly signed off as "not applicable". This matter was considered a violation of the administrative procedure governing FME control Maintenance: Overall, surveillance and maintenance activities reviewed this inspection period were conducted in accordance with licensee procedures and regulatory requirements. However, an event that occurred just prior to the inspection period was reviewed that involved a Technical Specification violation when maintenance personnel lifted a spent fuel pool gate without an operable emergency ventilation system (Section 7.0). This event, due to its minimal safety significance, was categorized as a non-cited violatio Engineering: Engineering support for the refueling outage as well as during power operations was good. Evaluation of the #1 main feedpump bearing vibration problem and determination that root cause was an " oil whirl" phenomenon was excellent (Section 4.3). However, evaluation of check valve

" chattering" in the auxiliary feedwater turbine steam supply lines appeared to be lacking (Section 4.1). This issue is further discussed in NRC inspection report 50-346/9401 Plant Support: Overall, personnel adherence to the radiation protection and security programs was good this inspection period. However, three events where radioactive material was found in inappropriate locations were identified this period (Section 5.1). Further NRC review of these events is warranted and will be tracked as an unresolved ite _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .-

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DETAILS 1.0 Plant Operations (71707) (71714) (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 tarout records, and verified tracking of limiting conditions for operation (LCO) 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:

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Emergency Diesel Generator 1-1

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Emergency Diesel Generator 1-2

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High Pressure Injection System - Train 1

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Decay Heat Removal System - Train 1

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Hydrogen Dilution System - Train 2 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 syste One violation associated with inadequate foreign material exclusion controls was identified as discussed in Section 1.3 belo .1 Operations Summary The ninth refuel outage continued during the first part of the inspection period. The outage completed on November 16 and the reactor was returned to power. On December 1, a downpower was conducted to troubleshoot and repair vibration problems on the #1 main feedpum Repairs were completed on December 5, and the unit returned to full powe .2 Control Room Activities Overall, control room activities were conducted in a controlled, conservativermanner during the inspection period. Shift briefs were complete and adequately discussed work priorities. Operators were cognizant of' plant conditions and panel indications. The reactor coolant system drain down to support steam generator nozzle dam removal and reactor coolant pump seal package maintenance was well conducte Likewise, unit startup was handled in a conservative manner, in conformance with regulatory and plant requirements. Procedural adherence by the control room staff was goo . . .

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' Inadequate Foreign Material Exclusion Controls On November 9, during a review of the material accountability log (MAL)

for the reactor cavity / refuel canal area, the inspector noted that final closeout review requirements had not been completed prior to reassembly of the reactor vessel. At the time of the review, reactor reassembly had been completed and preparations were being made to enter Mode 4, Hot Shutdow Administrative procedure, DB-MN-00005, Foreign Material Exclusion, required that a closecut inspection of the area be conducted to check for foreign material. The procedure indicated that this was necessary whenever work was performed in the vicinity of open fluid systems or in other open components when foreign material left inside could cause damage or equipment failure. This requirement appeared applicable in that the reactor cavity / refuel canal area had been directly opened to the reactor coolant system (RCS). However, the applicable step in the Foreign Material Exclusion Requirements form to perform the closeout inspection had been signed off as "not applicable".

Additionally, the MAL had not been adequately reviewed in that two log entries involving two plastic bags and a clipboard had not been signed off as having been removed from the FME area. Although the licensee was subsequently able to adequately reconcile the differences, no constraints had been in place to assure any foreign material inadvertently left in the RCS would have been identified prior to the performance of startup related activities such as starting reactor coolant pumps or plant heatu Because FME controls were not properly implemented as required by DB-MN-00005, this was considered a violation for inadequate adherence to an approved administrative procedure (346/94013-01(DRP)).

It should be noted that this constitutes the third violation issued for inadequate FME control since August 1994 and the second violation for FME problems specifically relating to reactor cavity / refuel canal work this outage (reference NRC inspection reports 346/94006 and 346/94011).

In response, the licensee better defined the onsite organizations with overall responsibility for FME control (both in containment and elsewhere), and initiated revisions to DB-MN-00005, Foreign Material Exclusion, and DB-0P-06904, Shutdown Operations, to more clearly indicate when and how FME control requirements will be implemente .4 Containment Walkd:swns Containment walkdowns were conducted on several occasions during the refueling outage. Overall, housekeeping and equipment condition were deemed adequate with no substantive concerns noted. However, on November 8, the inspector noted that a nut was missing from an incore instrument tank holddown bolt and a second nut was substantially backed off. The licensee could not determine a root cause but postulated that

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it may have been'an initial construction deficiency. A potential condition adverse to quality report (PCAQR 94-1169) was initiated.

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Initial review determined safety significance was minimal in that the 1 l holddown bolts were only required when the tank was filled with water.

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This condition would only exist during shutdown conditions. The missing nut was, thereafter, replaced, and it, along with the loose nut )

4 originally identified, and one other nut, also found to be loose,' were - !

i retorqued as required.

! Prior to final containment closeout, the licensee assured all

! extraneous material had been removed. The inspector, during a last j walkdown on November 10 identified no unacceptable conditions, i

i 1.5 Plant Winterization Checklist A review of DB-0P-06913, Plant Winterization. Checklist, was conducted to -

j assess the licensee's readiness for cold weather operation. No

substantive concerns were note i I

1.6 Main Steamline Room Doors l

! In March 1994, the Occupational Safety and Health Administration (OSHA)

{ cited the licensee for a violation relating to the licensee's position

, of maintaining the watertight access doors to the main steamline (MSL)

{ rooms closed at all times including those times when personnel were

performing work in those rooms. OSHA's position was that the doors a should remain open for quick egress in an emergency to protect oersonnel i in those room '

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However, the doors were being maintained closed to ensure compliance {

with NRC requirements for mitigation of a potential high energy line 1 4 break (HELB) as well as to meet fire protection requirements for the j subject areas.

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! In a letter dated August 26, 1994, the NRC notified OSHA of NRC's

! jurisdiction over this matte Subsequently, OSHA concurred with the l NRC on the jurisdictional issue and rescinded the violation.

l on November 14, 1994, the licensee reverted to maintaining Thereafter,losed the doors c at all times other than for normal ingress / egress.

2.0 Eurveillance (61726)

! The inspectors observed safety-related surveillance testing and verified i; that the testing was performed in accordance with adequate procedures, 1 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 conformod with Technical i Specification and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies i identified during the testing were properly reviewed and resolved by l

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The following test activities were observed and/or reviewed: ,

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DB-SC-03114, SFAS Time Response Test

- DB-SC-04274, SBODG Dead Bus Load Test

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- DB-SP-04153, AFPT #2 HSS, LSS and Overspeed Trip Test DB-PF-03008, Local Leak Rate Test - Penetration 16

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- DB-PF-04167, MOV DH2736 - Differential Pressure Test Results

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- DB-NE-03212, Zero Power Physics Testing

One concern related to DB-PF-04167 was noted during this review and is

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, discussed further in Section 4.4. No violations or deviations were identified in this area.

l 3.0 Maintenance (63702)

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Station maintenance activities of safety-related systems and components

were observed and/or reviewed during the inspection period to ensure

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that they were conducted in accordance with approved procedures, 4 regulatory guides, and industry codes or standards, and in conformance 1 with technical specification The following items were considered during this review: the limiting

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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

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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

_i qualified personnel; p: arts and materials used were properly certified;

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radiological controls were implemented; and fire prevention controls

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were implemented.

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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 may affect system performance.

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MW0 3-94-0745-01 Containment Personnel Hatch Inspectio MWO 3-93-2392-01 Auxiliary Feedwater Pump #2 Five-Year Inspectio i -

MWO l-93-1219-02 Transformer 02 Tap Chang No substantive concerns were noted during these reviews. No violations or deviations were identified in this are .0 Onsite Engineerina (37551)

Selected engineering problems or 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

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examined. 'The inspection included review of areas such as correctiv '

action systems, root cause analysis, safety committees, and self assessment in the area of engineerin .1 Auxiliary Feedwater Steam Supply Check Valves " Chattering" l As a result of a recent modification conducted during the ninth ,

refueling outage, check valves MS-734. and MS-735 located in the steam )

supply line to the auxiliary feedwater turbines were modified to remove an external packing / shaft configuration. . This change returned MS-734 and MS-735 to a near previous configuration that resulted in the check valves " chattering." During the eighth refueling outage a modification was conducted in an attempt to correct the chattering problem with the subject valves and that change had inadvertently resulted in problems 4 noted with the free movement of the valve disc ;

At the conclusion of the inspection period, no further modifications l were planned. The licensee intends-on repairing any damage incurred ;

during future operating cycles during each refueling outage. The licensee believes the amount of degradation of the seating surface will l not be significant and that periodic maintenance should be sufficient to l adequately maintain the seating surfaces. However, the licensee was not ]

able to provide adequate justification to support this assumption. This matter is further discussed in NRC inspection report 346/9401 Pending inspection of the valve internals during the next refueling outage (10RF0) currently scheduled to begin in April 1996, this matter ;

is considered an inspection followup item (346/94013-02(DRP))

4.2 Qualification of Containment Fire Extinguishers During a containment walkdown, it was noted that fire extinguishers were l permanently stored on individual brackets placed strategically around the containment. Personnel indicated the intent was to maintain the ]

extinguishers inplace during power operations. This decision was -

supported by an evaluation (reference PCAQR 88-645) that was previously done to address inadvertent dropping of the extinguisher (s) from their ;

support brackets and the conclusion that this would not potentially generate missiles. Engineering indicated that this was assured through

" drop" testing of the extinguishers per the National Fire Protection Association (NFPA) code requirements. In addition, an engineering evaluation (reference NED: 93-30008) addressed post accident conditions in containment and the extinguishers' capability to withstand the high pressure and temperature conditions they would then be subjected t Again, engineering indicated that this was assured through hydrostatic testing of the extinguishers per the NFPA standard This matter is considered an inspection followup item (346/94013-03(DRP)) pending inspector review of the drop tests, and the hydrostatic tests conducted on the subject extinguisher . .- - - - . . - - . . . . . - - - _ - ..

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4.3- Main Feedpump #1 High Vibration On November 29, operators noted higher than expected bearing vibration on the #1 main feedpump. Further analysis indicated that a phenomenon termed " oil whirl" was the root cause. Enginee' ring determined that " oil whirl" could be encountered with the right combination of pump speed, oil temperature, oil pressure, and static bearing loads.- In I conjunction, the dynamic effect resulted in the observed increased vibration levels. On December 1, a downpower to 60 percent was conducted to realign the pump and to increase oil temperature to the bearing in attempts to correct the problem. On December 5, the reactor was returned to full power. Thereafter, vibration levels returned to norma ] Notor Operated Valve DH2736 Calculation Error l An error in the calculation prepared in support of differential pressure (DP) testing of motor operated valve DH2736 was identified during l inspector review of the associated DP test results package (reference j DB-PF-04167, Motor Operated Valve Differential Pressure and Flow Test). !

The subject valve was located in the boron dilution flowpath and was included as part of the licensee's program to meet Generic Letter 89-10 requirements. Once brought to the licensee's attention, a PCAQR was initiated, revisement of the calculation made, .and retesting of the l valve conducte l This matter was referred to the Region III. Division of Reactor Safety to !

incorporate into their review of the licensee.'s program to address the Generic Letter. Further details on this matter are provided in  ;

inspection report 346/9401 j Plant Suonort (71750)

Selected activities associated with radiological controls, radiological effluents, waste treatment, environmental monitoring, physical security, emergency preparedness and fire protection were reviewed to ensure conformance with facility procedures and regulatory requirement ' Radioactive Material Found Outside the Radiological Restricted Area l Three events occurred during the inspection period that involved identification of radioactive material in areas outside of the radiological restricted area (RRA) of the plant and, in one case, being shipped to an offsite facilit PCAQR 94-1070 documented five contaminated relief valves that were found in the mechanical maintenance shop. PCAQR 94-1233' documented a situation where Bartlett Nuclear Corporation, during a receipt survey of materials just returned from ,

Davis-Besse, identified some amount of contamination. The materials had not been previously identified nor shipped as radioactive material PCAQR 94-1272 documented that a pump removed from the moisture separator

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demineralizer had been transported to the mechanical maintenance shop

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without contacting radiation protection. Upon surveying the pump in the 1 shop, fixed and smearable contamination was detected.-

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This matter warrants further NRC review, and pending completion of that

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j review, is considered an unresolved item (346/94013-04(DRP)). Followun of Previous Insnection Findinas (92901) (92902) (92903) (92904)

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] (closed) Inspection Followup Item (346/93006-01(DRP)): Decay heat pump i breaker found tripped during the eighth refueling outage. Attempts to l

determine the root cause of the trip were unsuccessful. Aftwr the outage,.the breaker was relocated from the decay heat pump cubicle to j one of the condensate pumps. Additional instrumentation was also i included during this time to ensure adequate monitoring of breaker

performance. No problems were noted during the operating cycle.

t The licensee then replaced the breaker into the No. I decay heat pump

! cubicle and the decay heat pump was utilized for shutdown cooling during the recent refueling outage. Again, no problems were noted with breaker i performance. Because the breaker has not inadvertently tripped since

} the eighth refueling outage, current performance appears adequate, and 4 no real cause appeared to be forthcoming from the licensee's evaluation, j this item is considered closed.

l (Closed) Violation 346/93011-01(DRP)): Inadvertent transfers of RCS l

inventory during the eighth refueling outage. In response to the two 1 water transfer events, the licensee conducted additional . training for j operations personnel, and revised associated procedures to ensure that i inadvertent transfers as noted during that outage would not recu Licensee corrective actions on this matter appeared adequate in that no i further water transfer events occurred during the most recent refueling

. outage (RF09). Therefore, this item is closed.

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! (Closed) Inspection Followup Item (346/93013-03(DRP)): Diesel generator

! electrical panel door found open. This matter was reinforced to appropriate maintenance personnel via both shop meetings as well as

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inclusion of appropriate requirements into maintenance administrative

{ procedures. Puring the most recent refueling outage (9RF0), no further

examples were noted. This item is close .0 Followup of Licensee Event Reports (92700)

) (Closed) LER 94-003-00: Spent fuel pool gate moved in spent fuel pool.

} while emergency ventilation inoperable. This event involved a situation i during the refueling outage where the spent fuel pool gate was lifted I with both the containment purge system in operation and the auxiliary i building roll-up door open. This situation resulted in the emergency i ventilation system (EVS) not being capable of maintaining at least 1/8"

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vacuum as required by the Technical Specifications. The root cause of j this event was determined to be a personnel error by the supervisor i involved with the lift. Overall the safety significance of this event i

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was minimal in that a spent fuel pool accident as a result 'of dropping

- the gate on spent fuel was very low. The gate was not lifted over spent fuel, and the lifting mechanism had been previously inspected per procedure to ensure that it would not fai In addition if any fuel had 4 been damaged as a result of dropping the gate, immediate' corrective !

actions to mitigate the accident would include closing of the auxiliary building roll-up door and shutdown of the containment purge syste The licensee's corrective action included counseling of the individuals !

involved in the event, the conduct of tailgate sessions with other i supervisors in the mechanical maintenance department to reinforce L management's expectations in this area, and initiation of ap)ropriate !

procedural changes to incorporate the lessons learned from me even l

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This event was a violation of Technical Specification 3.9.12 which prohibited crane operations and movement of loads over the spent fuel pool with the emergency ventilation system' inoperable. However based upon the minimal safety significance, and the licensee's corrective 1 actions taken in response to the event, it was determined that this matter met the criteria specified in Section VII.B.2 of the " General Statement of Policy and Procedures for NRC Enforcement Actions," 4

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(Enforcement Policy, 10 CFR Part 2, Appendix C) and a Notice of Violation would not be issue .0 Manaaement Meetina A management meeting was held in the Region III office on November 14, 1994, to discuss several issues that had been identified during the refueling outage. These included two radiation protection related events, a fuel mispositioning event during core reload and FME implementation problems in containment. Further details of the meeting are documented in inspection report 346/9401 . Exit Interview The inspectors met with licensee representatives (denoted in Section 10.0) throughout the inspection period and at the conclusion of the inspection on December 8, 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 repor _ _ _ _ - _ - - _

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10.0 Persons Contacted 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

• S. Byrne, Manager, Plant Operations B. Donnellon, Manager, Plant Engineering

• J. E. Moyers, Manager, Quality Assessment D. Crouch, Superintendent, Mechanical Maintenance

• D. P. Ricci, Supervisor', Operations

• P. Smith, Supervisor, Compliance

• J. Barron, Supervisor, Test / Performance

• D. Schreiner, Supervisor,.ISC

• R. C. Zyduck, Manager, Nuclear Engineering

• G. Skeel, Manager, Security- ,

• T. O'Connor, Manager, Regulatory Affairs

• T. Haberland, Manager, Planning

• R. Scott, Manager, Radiation Protection

• T. Bergner, Manager, Training

• J. Lash, Manager, Design Engineering

• J. McGee, Shift Supervisor

• Denotes those licensee personnel attending the December 8,1994, exit meetin .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. Inspection followup items disclosed during the inspection are discussed in Sections 4.1 and Unresolved Items An unresolved item is a matter requiring more information in order to ascertain whether it is an acceptable item, a violation, or a deviatio An unresolved item was identified in Section Non-cited Violations The NRC uses the Notice of Violation to formally document failure to meet a legally binding requirement. However, because the NRC wants to encourage and support licensee's initiatives for self-identification and correction of problems, the NRC will not issue a Notice of Violation if

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the requirements set forth in 10 CFR Part 2, Appendix C, Section VII. or VII.B.2 are met. A violation of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued is discussed in Section .

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