IR 05000341/1998019

Insp Rept 50-341/98-19 on 981110-990104.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maintenance & Plant Support

ML20206S135
Person / Time
Site:	Fermi
Issue date:	01/21/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206S116	List: IR 05000341/1998019 ML20206S111
References
50-341-98-19, NUDOCS 9901280129
Download: ML20206S135 (15)

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U.S. NUCLEAR REGULATORY COMMISSION REGIONlli Docket No: 50-341 License No: NPF-43 l Report No: 50-341/98019(DRP)  :

l Licensee: Detroit Edison Company Facility: Enrico Fermi, Unit 2 Location: 6400 N. Dixie Hw Newport, MI 48166 l

Dates: November 10,1998 - January 4,1999 i

Inspectors: S. Campbell, Acting Senior Resident inspector

J. Larizza, Resident inspector l

l Approved by: Anton Vegel, Chief Reactor Projects Branch 6

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PDR ADOCK 05000341 G PDR

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EXECUTIVE SUMMARY Enrico Fermi, Unit 2 Fermi Inspection Report 50-341/98019(DRP)

This inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers an 8-week period of resident inspectio Operations

• The shutdown to perform repairs on a leaking pressure seal for a feedwater check valve and subsequent startup of the unit was performed in a controlled and deliberate manner (Section 01.1).

e The actions taken by operations personnel in recovering general service water intake level due to ice formation at the station intake from the lake, were performed in a timely, effective manner in accordance with established procedures (Section 01.2).

• The inspectors identified that an operability assessment was not documented for degraded condition related to a containment isolation valve. Though the degraded condition had minor safety consequences, the failure to ensure an adequate documented operability evaluation was considered a vulnerability in ensuring that safety equipment in a degraded condition was thoroughly evaluated. The licensee's quality l assurance staff had previously identified similar problems and corrective actions were in the process of being implemented (Section O4.1).

Maintenance

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e The inspectors concluded that observed surveillance test activities were performed effectively and in accordance with established procedures (Section M1.1).

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• The inspectors concluded that poor communication between operations and maintenance personnel regarding the status of out-of-service equipment resulted in the l operations personnel being unaware that a damper actuator for the reactor building ventilation system was not installed for sixteen days (Section M1.2).

e The licensee determined that an electrician was injured while working on a motor control center due to: 1) less than adequate hazard assessment performance,2) less than adequate electrical safety culture, 3) a deficiency in the interface between the site safety program and the work control program, and 4) that the injured individual failed to adequately self-check his work. The inspectors concluded that the assessment was comprehensive and that the corrective actions to prevent recurrence were acceptable (Section M8.3).

Enaineerina e Corrective actions for safety relief valve performance deficiencies and surveillance test procedure overlap problems were reasonable and comprehensive (Sections E8.1 and E8.2).

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Plant Suocort i

e The inspectors used a radiation detector and independently verified that areas were appropriately posted (Section R1.1). l e The inspectors observed that the central and secondary alarm stations were manned .

with security officers who were knowledgeable of their assigned duties. Equipment in l- these areas appeared to be functioning properly (Section S2.1).

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Ii Report Details l

Sumrnary of Plant Status Unit 2 began this inspection period at 96 percent power. On November 12,1998, operators shut down the unit to support repairs to a leaking pressure seal on a feedwater check valv The plant was subsequently restarted and power was retumed to 96 percent on November 2 .

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1. Operations 01 Conduct of Operations "

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01.1 General Comments (71707)  !

Using inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. In general, conduct of operations was professional and safety-conscious, in particular, the shutdown to repair a leaking pressure seal on a feedwater check valve and subsequent startup of the unit was performed in a controlled and deliberate manner. Specific events and noteworthy observations are detailed in the sections belo .2 ' General Service Water (GSW) Inspection Scooe (71707)

Inspectors' review of control room logs of December 26,1998, indicated that the control room noted that the GSW intake level had experienced a rapid lowering of level. The lowest level noted on the recorder was 567 feet. Operations personnel dispatched the cutside rounds operator and the turbine building rounds operator to the GSW Pump House to check the position of the GSW de-ice valve. The inspectors reviewed the i

. action statements of Alarm Response Procedures 7D22 (7D23), " General Service H2O Screen A (B) H,0 Level High/ Low," and compared them to the actions taken by operations personne Observations and Findinas Operations personnel shutdown the East Circulating Water Make-Up to minimize level loss in the GSW intake, opened the GSW de-ice valve two additional turns, and the GSW intake level returned to normal. Following the restart of the East Circulating Water Make-Up, the GSW intake level started lowering again. The outside rounds operator operated the GSW trash rake which broke up the ice that had prevented water from entering into the intake. The GSW intake level rapidly recovered to normal without further problems. The inspectors reviewed the Updated Final Safety Analysis Report and confirmed that level did not approach the minimum Updated Final Safety Analysis Report level of 558 feet. The inspectors verified that the operators' actions were consistent with applicable procedure . . - .-

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! Conclusions  !

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The actions taken by operations personnel in recovering GSW intake level due to ice formation at the station intake from the lake were performed in a timely and effective manner in accordance with established procedures.

I O2 Operational Status of Facilities and Equipment J

O2.1' Enaineered Safety Feature System Walkdowns (71707)

The inspectors used inspection Procedure 71707 to walk down accessible portions of the following engineered safety feature systems:

e Residual Heat Removal Complex e Divisions 1 and 2 Motor Control Centers (MCC) and Battery Rooms e High Pressure Core injection System Pump Room e . Hydraulic Control Unit Areas Equipment operability, material condition, and housekeeping were acceptable in most cases. Several minor discrepancies were brought to the licensee's attention and were corrected. The inspectors identified no substantive concems as a result of these walkdown ;. 04 Operator Knowledge and Performance 04.1 Documentation of Operability Evaluations on Condition Assessment Resolution Documents (CARDS)

Inspection' Scope (71707)

On December 21,1998, during a tour of the reactor building, the inspectors noted CARD 98-00488 tag on Reactor Recirculation Sample Line Valve B3100-F020, which is an outboard containment isolation valve. The CARD indicated the valve had a less than !

one drop per minute packing leak while stroke time testing the valve. The inspectors reviewed the CARD to determine if an operability evaluation to justify the acceptability of the packing leak was performe Observations and Findinas The inspectors reviewed documentation related to CARD 98-00488 and noted that an operability evaluation was not documented on the CARD. Since the safety function of the valve was to close and prevent leakage from the containment during an accident, the inspectors questioned control room operators why an operability evaluation for the leaking valve was not documented.' The inspectors were informed that documented operability evaluations were not normally performed for packing leaks on containment isolation valves since packing leaks were considered small compared to the total allowed leakage for containmen L

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l l The inspectors questioned the leak rate test engineer regarding the packing leak on l Valve B3100-F020, and found that the engineer was unaware that the leak existed.

l Consequently, this leakage was not evaluated for incorporation into the total allowed ,

leakage for containment. However, the test engineer stated that during certain postulated accident conditions, the leakage would be smaller since accident pressure (56.5 psig) was postulated to be much lower than operating pressure (1030 psig).

Although this evaluation was not documented in the CARD, the inspectors considered this assessment reasonable to justify operability of the valve and that the safety consequences of the leak was minimal. However, the inspectors were concerned that due to a lack of a documented evaluation of the potential consequences of a degraded condition involving a safety-related component, the full ramifications or potential consequences of the deficiency may not be adequately evaluate l in addition, to the inspectors' concems involving the documentation of operability evaluations for CARDS, the inspectors were informed by licensee staff that similar concerns have been brought up recently by quality assurance personnel as documented in CARD 98-18420, dated October 12,1998. Specifically, CARD 98-18420, documented that operability evaluations were not documented on the following:

• CARD 98-18700, " Inconsistencies in Motor Operated Valve (MOV) Base Design Documents,"

• CARD 98-17404, "B Reactor Water Cleanup Tripping on Low Flow," and

l CARD 98-18067 " Wrong Grease installed in Flex Couplings on Reactor l Recirculation Motor Generator Set B."

CARD 98-18420 noted that justifications of component / system operability were unclear in these examples. As of the end of this inspection period, this CARD remained open pending development of corrective action Following discussions with the inspectors, the licensee subsequently determined that since the safety function of the valve was in question, an operability determination should have been performed. The licensee identified several other examples, in addition to the examples mentioned above, where the operability block on the CARD assessment resolution sheet was inappropriately checked "not applicable." This block was usually reserved for minor, administrative, and/or non-plant issues. The inspectors reviewed Procedures MES27, " Verification of System Operability," and MQA11, " CARD,"

and determined that the threshold for using the "not applicable" block was not define The licensee was considering a revision to the procedures clarifying this fac c. Conclusions The inspectors identified that an operability assessment was not documented for a degraded condition related to a containment isolation valve. Though the degraded condition had minor safety consequences, the failure to ensure an adequate

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documented operability evaluation was considered a vulnerability in ensuring that safety l equipment in a degraded condition was thoroughly evaluated. The licensee's quality assurance staff had previously identified similar weaknesses and corrective actions, and were in the process of being implemente l l

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II. Maintenance l l

M1 Conduct of Maintenance i M1.1 General Comments i Insoection Scope (62707: 61726)

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! The inspectors observed all or portions of the following maintenance and surveillance test activities:

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• 42.302.03 Channel Functional Test of Division 2,4160-Volt Bus 65E, l

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Undervoltage Circuits

• 24.307.17 Emergency Diesel Generator (EDG) 14 - Start and Load Test

• 24.204.06 Low Pressure Coolant Injection and Suppression Pool Cooling / Spray Pump and Valve Operability Test o 24.208.03 Division 2 Emergency Equipment Service Water Pump and Valve ,

Operability Test i Observations and Findinas j These surveillance activities were properly conducted using current revisions to procedures. At the end of the EDG 14 test, the operators involved in the test requested

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l that previously initiated Work Request (WR) 000Z984694, related to replacing an l l obsolete EDG 14 generator end bearing temperature switch, be implemented before the j next scheduled surveillance. In addition, they requested that Air Receiver Outlet l Valve R3000F043D previously initiated WR 000Z984155, related to a bad valve stem  !

bushing that caused difficulty in operating the valve, be worked before the next i scheduled surveillance. The inspectors will monitor resolution to these WRs. The '

t maintenance and operations personnel were proactive in identifying the need to

[ increase the priority of previous work request . Conclusions l The inspectors concluded that observed surveillance test activities were performed l effectively and in accordance with established procedures.

l M1.2 Poor Communication of the Status of Out-of-Service Eauipment Inspection Scope (62707.61726)

On December 2,1998, while performing surveillance test procedure 24.404.02,

" Division 1 Standby Gas Treatment System Filter and Secondary Containment Isolation Damper Operability Test," the reactor building pressure unexpectedly increase Subsequent investigation revealed that the cause of the pressure increase was attributed to the failure of center modulating inlet damper to open while restarting the Reactor Building Heating, Ventilation and Air Conditioning System (RBHVAC). The RBHVAC is a system not required by Technical Specifications (TSs). The inspectors l

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followed up on the cause of this failure and the circumstances surrounding the communication of the status of this out-of-service equipmen Observations and Findinos During the licensee's investigation of the failure of the center damper to open, they discovered that the actuator for the center exhaust fan modulating inlet damper was missing. A search for the actuator ensued and it was subsequently discovered in the Instrument and Control (l&C) shop, where it was in the process of being rebuilt in accordance with preventive maintenance (PM) procedures. Maintenance on the actuator was completed and it was successfully reinstalled on the damper and the RBHVAC system was tested satisfactoril Operations personnel performed an investigation of the issue and discovered that I &C personnel requested that PM be performed on the actuator on November 16. While planning the activity, operations personnel estimated that the work activity could be completed within one shift and, therefore, a tag-out was not required as allowed by the tag-out procedure. However, after l&C personnel removed the actuator, they realized that the work would take longer than one-shift. Operations personnel were not notified of the delay in the work activity and since no tag-out existed, a record of this work activity was not communicated to the subsequent operating shift A CARD was initiated to document the problem occurrence and to track corrective actions. Preliminary corrective actions included reemphasizing communicating the status of equipment out-of-service and enhancing equipment configuration contro Conclusions  !

The inspectors concluded that poor communication between operations and ,

maintenance personnel regarding the status of out-of-service equipment resulted in the I operations personnel being unaware that a damper actuator for the reactor building !

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ventilation system was not installed for 16 day M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Thermometers Discovered in Division 1 Battery Cells Insoection Scoce (62707)

l On December 23,1998, during a tour of the Division 1 battery room, the inspectors discovered a broken thermometer in Ceil 14 and an intact thermometer in Cell 7. The inspectors reviewed the impact the thermometers had on battery operabilit Observations and Findinos The licensee identified that the thermometers had been in the cells since the batteries were installed in 1984 and attempts to remove them were unsuccessful. An evaluation documented in Potential Design Change 1288, noted that the thermometers were

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I glass / alcohol type that was not detrimental to the function of the battery. Replacement 1

' of the Division 1 battery is scheduled for the next refueling outage (RFO).  !

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The inspectors concluded that the evaluation performed by the licensee was acceptable to justify operability of the battery cells with the thermometers presen M8 Miscellaneous Maintenance issues (92902) -

M8.1 (Closed) Licensee Event Report 50-341/96017-04: Failure of safety relief valves (SRVs)

to open within TS allowed tolerance. This involved the licensee's discovery that 11 of

! the 15 pilot valves inside the SRVs would not have lifted within the TS 1 percent

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allowable tolerance.

I l Technical Specification 3.4.2.1, requires that for the safety valve function, at least 11 of l the following SRVs shall be operable with the specified code safety valve function lift setting during Operational Conditions 1,2 and 3:

-e 5 safety / relief valves at 1135 psig i 1 percent e 5 safety / relief valves at 1145 psig i 1 percent e 5 safety / relief valves at 1155 psig ! 1 percent Contrary to the above,11 of the 15 pilot valves did not lift within the TS required 1 percent allowable tolerance.

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Corrective actions included the request for a License Amendment to modify the SRV setpoint tolerance criteria from 11 percent to 13 percent, a written justification for i operating with the installed SRV pilot assemblies for the duration of Operating Cycle 6, and the installation of platinum ion beam bombarded discs during RFO 6 to prevent l oxide bonding of the pilot valve disc and seat. The licensee also committed to follow ;

industry developments related to the improved operability of SRV setpoint performanc l The inspectors verified that these actions were complete l This licensee's identification and corrective actions of this violation is being treated as a non-cited violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement

! Policy (NCV 50-341/98019-01(DRP)).

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- l L M8.2. (Closed) Licensee Event Reoort 50-341/95007-00: Technical Specification violation due to inadequate surveillance testing due to incomplete surveillance test procedure overla l

. This involved the licensee's identification that surveillance testing for certain portions of !

> the EDG output breaker and 480-Volt MCC load sequencer logic circuitry was inadequate. This was attributed to a deficient procedure, in that not all attributes of the associated logic circuits were fully verified by the procedur !

Technical Specification 6.8.1. d., requires written procedures shall be established, j implemented and maintained for surveillance and test activities of safety-related

equipment.

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Contrary to the above:  !

e Procedures 24 302.02(04)(05)(06), " Logic System Functional Test of  !

Division 1(2)(1)(2) 4160-Volt Emergency Bus 64B(65E)(64C)(65F) and 11 AE(13EC)(12EB)(14ED) Undervoltage Circuits," used to test the 4160-Volt .

system logic, were inadequate in that the procedures did not instruct positive  !

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verification that the EDG output breaker re-close permissive logic circuitry. This is an example of a violatio l e Work Requests 000Z95609, 2, 3, and 4, which were used to test the EDG output logic, were inadequate in that the WRs did not direct positive verification that the 480-Volt MCC automatically connected loads are energized through the load sequencer after the EDG starts. This is a second example of a violatio Corrective actions included: a commitment to complete the remainder of the testing to )

satisfy surveillance requirements before startup from the subsequent plant outage; to {

revise the applicable surveillance test procedures for each 480-Volt emergency bus and undervoltage circuits; and, to include positive verification that the EDG output breaker re-close permissive logic circuitry was initiated via the load shed feature. In addition, existing procedures were to be revised, or new procedures were to be developed, to ensure positive verification that after the EDG starts, the 480-Volt MCC automatically connected loads were energized through the load sequencer. The inspectors verified that these actions had been complete )

l This non-repetitive, licensee identified and corrected violation, is being treated as an NCV, consistent with Section Vll.B.1 of the NRC Enforcement Policy (NCV 50-341/98019-02 (DRP)).

M8.3 Followuo on Electrician Severely iniured while Performina PM Inspection Report 50-341/98015, dated December 7,1998, documented that a Detroit Edison electrician journeyman was injured while performing a PM activity on MCC 72 The inspectors followed up on the licensee's findings regarding the circumstances surrounding this even The summary of the root cause of the incident indicated that the licensee's hazard assessment performance was less than adequate, the electrical safety culture was less than adequate, the program to program interface between the site safety program and the work control program was deficient and that the injured individual failed to adequately self-check his wor The corrective actions to prevent recurrence being implemented include e changing procedures to incorporate lessons learned, .

.e conducting electrical safety training for maintenance and operations personnel,  ;

e developing in-processing training for non-licensee electricians, e reinforcing the expectations for peer checking, e reviewing the provision for properly using insulated tools,

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e reviewing maintenance work instructions to identify and eliminate common l

i characteristics that contributed to the event,

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e and the forming resolution subcommittees to develop proper interfaces between work control process and safety progra In addition, the licensee plans to establish an assessment review committee as part of

the work control process to review maintenance activities that may pose serious ,

l hazards. The licensee expects to have all the above actions completed by middle '

of 1999. The inspectors concluded that the licensee's corrective actions were appropriate and reasonable to prevent recurrenc . Enaineerina E8 Miscellaneous Engineering issues (92902)

E (Closed) Violation 50-341/95005-04: Failure to follow requirements delineated in Procedure NPP-MA-01, Revision 15, " Work Control." The requirements for controlling work were not followed while implementing WR 000Z591519 to monitor reactor wide range inputs and outputs during turbine trips. Corrective action included communicating to site personnel the importance of procedure adherence via staff meetings and the plant weekly newsletter. Further, work control supervision defined these expectations to work planners. The inspectors verified these corrective actions were complete E8.2 (Closed) Follow-Up Item 50-341/95010-01: For 21 direct current MOVs, the test capability under degraded voltage conditions was in question when using the standard limitorque capability equation. Specifically, the licensee's verification methods had not been validated through testing or through a peer review process recommended by industry guidelines such as the Institute of Electrical and Electronics Engineers 129 The licensee reevaluated the direct current MOVs using standard industry methods, resulting in seven valves being identified as having an insufficient capability to trip the torque switch under degraded voltage conditions. Fermi personnel performed an operability justification on all seven valves and committed to retest the seven MOVs prior to or during the RFO 5, with action to be completed by November 30,1996. The inspectors verified that these actions were completed .

[y, Plant Support Radiological Protection and Chemistry Controls R Independent Verification of Radiation Doses rates in the Reactor and Turbine Buildinas The inspectors conducted routine tours of the reactor and the turbine buildings using a portable radiation detector to determine if radiation postings reflected the dose rates in

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. . Status of Security Facilities S2.1 Tours of the Central and Secondarv Alarm Station The inspectors toured the central and secondary alarm stations and questioned the security officers on security procedures used at the plant and security equipment used in these areas. The inspectors found that the security officers were knowledgeable of their assigned duties and that security equipment appeared to be functioning properl V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at l

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the conclusion of the inspection on January 4,1998. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was ;

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i PARTIAL LIST OF PERSONS CONTACTED

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Licensee '

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P. Fessler, Assistant Vice-President, Operations ,

W. O'Connor, Assistant Vice-President, Nuclear Assessment .

T. Bergner, Director, Organization Development i K. Howard, Director, Plant Support Engineering  :

J. Moyers; Director, Nuclear Quality Assurance .

N. Peterson, Director, Nuclear Licensing l

A. Kowalczuk, Manager, Nuclear Support l

J. Plona, Manager, Technical l R. Gaston, Supervisor, Compliance' l R. Russell, Supervisor, Simulator & Training Support  !

S. Stasek, Supervisor, independent Safety Engineering Group  ?

W. Tucker, Supervisor, Nuclear Fuels and Reactor Engineering Group P. Lynch, Nuclear Shift Supervisor, Operations  :

E. Meyer, Nuclear Shift Supervisor, Operations i

- S. Booker, Superintendent, Maintenance  !

R. Eberhardt, Superintendent, Outage Management E. Kokosky, Superintendent, Radiation Protection  ;

K. Hlavaty, Assistant Superintendent, Operations  ;

K. Harsley, Engineer, Nuclear Licensing N S. Campbell, Acting Senior Resident inspector J. Larizza, Resident inspector l

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INSPECTION PROCEDURES USED IP 61726: Surveillance Observations IP 62707: Maintenance Observation IP 71707: Plant Operations IP 92902: Followup - Maintenance ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 50-341/98019-01 NCV Failure of the SRVs to open within TS allowed tolerances 50-341/98019-02 NCV Inadequate procedure to test EDG logic Closed 50-341/98019-01 NCV Failure of the SRVs to open within TS allowed tolerances 50-341/98019-02 NCV inadequate procedure to test EDG logic 50-341/96017-04 LER Failure of the SRVs to open within TS allowed tolerances 50-341/95005-04 VIO Failure to follow requirements in Procedure NPP-MA-01, Rev.15 50-341/95007-00 LER Failure to meet TS surveillance requirements for EDG logic 50-341/95010-01 IFl Test capability under degraded voltage for 21 MOVs in question Discussed None

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LIST OF ACRONYMS USED  !

CARD Condition Assessment Resolution Document i EDG Emergency Diesel Generator l GSW General Service Water l&C Instrument and Control i LER Licensee Event Report

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MCC Motor Control Center j MOV Motor Operated Valves i NCV Non-Cited Violation

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PM Preventive Maintenance RBHVAC Reactor Building Heating Ventilation and Air Conditioning RFO Refueling Outage SRV- Safety Relief Valve TS Technical Specification

~WR Work Request l

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