ML20216D886

Insp Rept 50-346/98-02 on 980108-0218.Violations Noted.Major Areas Inspected:Licensee Operations,Maint,Engineering & Plant Support

ML20216D886
Person / Time
Site:	Davis Besse
Issue date:	03/10/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20216D875	List: ML20216D871 ML20216D881 ML20216D886
References
50-346-98-02, 50-346-98-2, NUDOCS 9803170332
Download: ML20216D886 (14)

See also: IR 05000346/1998002

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! U. S. NUCLEAR REGULATORY COMMISSION

l REGIONlli

Docket No: 50-346

License No: NPF-3 l

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Report No: 50-346/98002(DRP)

Licensee: Toledo Edison Company

Fscility: Davis-Besse Nuclear Power StatOn

Location: 5503 N. State Route 2

Oak Harbor, OH 43449

Dates: January 8 - February 18,1998

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inspectors: S. Campbell, Senior Resident inspector

K. Zellers, Resident inspector

Approved by- Thomas J. Kozak, Chief,

Reactor Projects Branch 4 ,

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9003170332 980310 E

PDR ADOCK 05000346

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EXECUTIVE SUMMARY

Davis-Besse Nuclear Power Station

NRC Inspection Report No. 50-346/98002(DRP)

This inspection included aspects of licensee operations, maintenance, engineering, and plant

support. The report covers a six-week period of resident inspection.

Ooorations

. The inspectors observed that the plant was operated in a controlled, conservative

manner. Plant issues that were identified weis appropnately handled in accordance with

the licensee's comsctive action program or material deficiency program. Operators were

knowledgeable of Technical Specification (TS) operability requirements and tagouts were

implemented in accordance with the tegout procedure. Generally, the operators exhibited

good knowledge of plant equipment status and property used plant operating procedures.

Control room ope stors and equipment operators were observed to be adequately

ful511ing their duties (Sections 01.1 and 04.1).

. Plant operations were impacted on two occasions when operators displayed a lack of

attention to detail. The first example involved the inadvertent isolation of seal injection to

all four reactor coolant pump seal packages due to an operator error. The second

example involved the lifting of a letdown system relief valve due to an operator not

performing a procedure in the correct sequence. These events are two examples of a

violation of TS 6.8.1 (Section 01.2).

• The operations manager appropriately communicated physical fitness expectations to the

operating crews after being notified by the inspectors that a shift supervisor stood watch 1

with the medical condition of laryngitis (Section 06.1). ,

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• The inspectors identified one violation of 10 CFR 50.72(b)(1)(ii)(B) where the licensee did  !

not report to the NRC, within one hour of the discovery, that speed sensing circuitry for

Emergency Diesel Generator #1 was not designed per the 10 CFR Part 50, Appendix R,

design criteria for hot short protection, a condition outside the design basis of the plant

(Section F8.1).

Maintenance

. During surveillance activities, equipment was observed to perform as described by the

Updated Safety Analysis Report. Maintenance personnel communicated adequately and

adhered to procedure requirements while performing maintenance and surveillance

activities. N inspectors observed that oversight of maintenance activities was effective.

Maintenance and surveillance testing activities were professionally conducted

(Section M1.1).

Enoineerina

• The licensee demonstrated an excellent questioning attitude regarding the configuration

of a refueling drain canal valve, a valve whose open position ensured sufficient water

supply to emergency core cooling system pump suctions during an accident. After

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determining that the valve was not in its locked valve program, the licensee !mmediately

entered containment, venfied the valve was in the open posMion, locked the valve, and

plans to enter it in the locked valve program (Section E8.2).

NRC enforcement discretion was exercised, and no violation issued, in recognihon of a

licensee identified and corrected old design issue regarding the failure to protect the safe-

shutdcwn emergency diesel generator from a hot short condition (Section F8.2).

Plant Suncort

OverM, radiation protection activities relating to a diving activity in the spent fuel pool

transfer canal and a high integrity container lift were performed in a professional, well

, planned manner (Section R1.1).

The fire brigade team effectively responded to a challenging drill scenario (Section F4.1).

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Resort Details

Summary of Plant Status

The unit operated at neady full power throughout the inspection period.

Lonerations

01 Conduct of Operations

01.1 General Comments (71707)

The inspectors observed that the plant was operated in a controlled, conservative

manner. Performance issues that were identified were appropnately handled in

acconience with the licensee's corrective action program or material deficiency program.

Operators were low 4edgeable of and property applied Technical Specification (TS)

operability and limiting conditions for operation requirements. Generally, the operators

exhibited good knowledge of plant equipment status and property used plant operating

procedures. Speci'ic events and noteworthy observations are detailed in the sections

below.

01.2 Operator Failures To Follow Procedure Durina a Test and a Plant EvW%

a. Inspection Scope (71707)

On January 6 and February 11,1998, the inspectors noted that control room log entries

indicated that operator errors had potentially occurred during a routine surveillance test

and during a plant evolution. The inspectors followed up on each event.

b. Observations and Findings  :

Failure to Follow Procedure Durina Water inventory Test

On January 6,1998, while a reactor operator performed Step 4.1.10 of

Procedure DB-SP-03357, "RCS Water inventory Balance," he inadvertently closed

Reactor Coolant Pump (RCP) Seal Retum Valve MU-38, which isolates seal irgection

return for all four RCPs, instead of opening Domineralized Water isolation

Valve DW-6831B. After closing MU-38, annunciators for low seal water retum flows

unexpectedly alarmed, and the reactor operator notifuni the assistant shift supervisor that

his error had caused the alarms. Within about three seconds, the operator reopened

MU-38 which cleared the annunciator alarms. Operators then reviewed the ROP

summaries on the plant computer and determined that all of the RCP seal parameters,

including temperature, seal water retum flow, and seal cavity pressures, were normal.

Subsequent to the event, the inspectors questioned plant engineerity personnel as to

whether any degradation to the pump seal packages was observed. The engineers

reviewed the data for seal cavity pressures and seal injection temperatures and

determined that no degradation in seal performance had occurred. According to plant

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engineering personnel, a loss of seal injection would have to occur for much longer than I

three seconds before the seals would be degraded.

The inspectors reviewed the control room panel and noted that the handswitch for

valve MU-38 was located directly below the handswitch for valve DW-68318 on Safety .

Features Actuation Panel C5717. The operstar stated that the cause of his error in

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following the procedure was not employing self checking techniques. He noted that he

was not distracted, fatigued, or rushed.

Technical Specificaten 6.8.1.c states, in part, that written procedutos shall be

implemented covering surveillance and test activities of safety-related equipment.

Technical Surveillance Requirement 4.4.6.2.1.d. states, in part, that a reactor coolant

system water inventory balance test be performed at least once por 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during

steady state operation. Procedure DB-SP-03357, "RCS Water inventory Balance,"

implements TS Surveillance Requirement 4.4.6.2.1.d. Closing MU-38 instead of opening

DW-6831B as required by DB-SP-03357, Step 4.1.10, is a TS 6.8.1 violation

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(VIO 50-346/98002-ia(DRP)).

Failure to Follow Procedure Durina Letdown Water Nnuo Activity

On February 11,1998, a control room operator began adding domineralized water to the

makeup tank (MUT) by opening Batch Isolation Valve MU-40 while diverting letdown

water through the clean waste system per Procedure DB-OP-06001, " Boron

Concentration Control." He diverted the water so that boron would be removed from the  ;

reactor coolant system. While performing the procedure to reposition three-way divert

Valve MU-11 to the " Clean Waste" position (to divert the letdown flow to the clean waste ;

system), a high letdown pressure alarm unexpectedly annunciated in the control room. '

The high letdown system pressure caused Letdown Relief Valve MU-1890, set at

150 psig, to open and relieve approximately 30 gallons of water to the reactor coolant

drain tank, in response to the alarm, the operator quickly repositioned MU-11 to the

"MUT" position and the alarm cleared.

Potential Condition Adverse to Quality Report (PCAQR) 98-0233 was initiated to

document the event. The licensee began an investigation into the cause of the

unexpected high letdown system pressure. During the investigation, the licensca found

that Booster System Bypass Valve WC-3526, a valve which isolates the clean waste

system from the makeup and purification system, was inadvertently left closed. The

closed valve stopped letdown system flow which increased letdown system pressure

above the relief Valve MU-1890 setpoint.

The inspectors interviewed the operator invosved in the event and determined that he

focused on a caution statement in the procedure stating that MU-40 be opened prior to

repositioning MU-11 to the " Clean Waste" position. The caution was located above

Procedure Steps 3.5.17 through 3.5.19. Steps 3.5.17 and 3.5.19 directed that MU-40 be

opened and MU-11 be placed to the " Clean Waste" position, respectively. Because the

operator focused on the caution, he performed the action discussed in the caution

statement, but failed to open WC-3526 in accordance with Procedure Step 3.5.18.

Consequently, he did not perform Step 3.5.18 in sequence with Steps 3.5.17 and 3.5.19.

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Technical Sr4@ cation 6.8.1 a requires, in pact, that procedures recommended in

Appendix A rd Regu; story Guide 1.33, Novemter,1972, be implemented. Appendix A of

Regulatory Guide 1.33, November 1972, Sectbn A. 4, references administrative j

procedum for procedure adherence. Procedure NG-DB 00225," Procedure Use and

Adherence," is an administrative proosdure that provides requirements for procedure

adherence. 6tep 6.8.1 of Procedure NG-DB-00225 requires, in part, that procedures be

performed as numerically sequenced in the body of the procedure. Contrary to the

above, the licornee did not perform the steps of Procedure DB-OP-06001 in sequence in

that a reactor operator failed to open WC-3526 per Step 3.5.18 while performing

Procedure Steps 3.5.17 through 3.5.1g to divert water to the clean waste system. This is

a second example of a TS 6.8.1 viciation (VIO 50f348-98001-1b (DRP)).

c. Conclusions l

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Plant operations were impacted on two occasions when operators displayed a lack of

attention to detail. The first example involved the inadvertent isolation of seal injection to

all four RCP seal packages due to the operator closing the wrong valve. The second

example involved the lifting of a letdown system relief valve due to an operator not

performing a procedure in the correct sequence. These events are two examples of a

violation of TS 6.8.1.

02 Operational Status of Facilities and Equipment

O2.1 System Walkdownw (71707)

The inspectors toured accessible postions of the following engineered-safety-ft,atures and

important-to-safety systems during the inspection period:

• Emergency Diesel Generators # 1 and 2

. Low Pressure injection Trains 1 and 2

. High Pressure injection Trains 1 and 2

. Containment Spray Trains 1 and 2

. Auxiliary Feed Water Trains 1 and 2

System lineups and major flowpaths were verified to be consistent with plant

procedures / drawings and the Updated Safety Analysis Report. Pump / motor fluid levels

were within their normal bands. No substantive concems were identified as a result of

the walkdowns.

04 Operator Knowledge and Performance

04.1 Eauipment Operator Tours (71707)

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The inspectors accompanied equipment operators during their plant tours. The operators

property recorded equipment parameters and checked the status of operating pumps,

motors and switchgear as required. The inspectors determined that the operators were

knowledgeable of equipment status and requirements for log taking. The inspectors

concluded that the equipment operators appropriately conducted their plant tours.

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06 Operations Organization and Administration

06.1 EBDgg of Control Room Personnel (71707) <

While the inspectors conducted backshift inspection activities on January 10,1998, the

inspectors noted that the shift supervisor had laryngitis which was observed to cause

some difficulty in talking. The shift supervisor then finished the remaining hour of his

watch and reported sick the followir g day. He did not report on shift until he was able to

communicate better.

The inspectors informed the operations manager of this observation. The operations

manager reviewed me existing fitness-for-duty programs and found that they did not

provide for an evaluation of the medical condition of laryngitis. Subsequently, he issued a

memorandum to all operations shift personnel that communicated additional fitness for

duty concems, such as a medical condition of laryngitis or a severe cold, that should be

considered by operations personnel before assuming the watch. The inspectors

concluded that the operations manager appropriately communicated his expectations

regarding personnel medical conditions.

II. Maintenance

M1 Conduct of Maintenance

M1.1 Maintenance and Surveillance Activities (61726)(627021 l

The following maintenance and surveillance testing activities were observed / reviewed

during the inspection period:

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DB-MI-03012 (Rev 03) Channel Functional Test of Reactor Trip Breaker A,

RPS Channel 2 Reactor Trip Module Logic, and

ARTS Channel 2 Output Logic

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DB-MI-03003 (Rev 01) Channel Functional Test of Reactor Protection

System Channel 3 RC Pressure and Temperature

Trip Functions

. MWO 3-98-4616-01 EDG Air Receiver Check Valve inspection

. MWO 2-96-0005-02 Modify Component Cooling Water pump breaker

AD108 Trip

. MWO 2-97-0030-02 Raise Setpoint: PSLLRC02A4 Safety Feature

Actuation System Channel 2

During surveillance activities, equipment was observed to perform as described by the

Updated Safety Analysis Report. Maintenance personnel communicated effectively

during the maintenance and surveillance activities. Maintenanco work order packages

included all necessary references to perform the work. Maintenance craft were observed

to be conscientiously adhering to work order instructions.

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"t +y.w,; provided good oversight at a majority of the evolutions that were observed

by the inspector. The inspectors concluded that these maintenance and surveillance

testing activities were p,C5'sf; conducted.

M8 Wiecellaneous Maintenance issues (92902)

M8.1 (Closed) Unresolved item (50-348/g701501(DRP)): Post-Maintenance Testing for

Class 1E Undervoltage Relays. Through subsequent discussion of the issue with plant

maintenance personnel, the inspectors concluded that the post-maintenance testing of

the undervoltage relays was acceptable. The undervoltage relays were deslynod to be i

removed and installed from their cubicles. Routine maintenance on the relays provided

assurance that they operated propedy before reinstallation because the maintenance

activity checked the voltage and timing requirements. Addshonally, the relay targets (a i

target provides indication that a relay changed state), were checked during post- I

maintenance testing. The target check provided further assurance that the relays were

installed property. Additionally, the inspectors concluded that an 18-month timing test,

which isolated each relay from Ks parallel relay, demonstrated proper operation of the

legic. However, as a result of tne inspectors' questions, the licensee enhanced its

undervoltage relay post-maintenance testing by including the 18-month timing test to

provide additional assurance that the relay had been propedy reinstalled following its

removal.

lit. Ennineerina

E2 Engineering Support of Facilitiec and Equipment

E2.1 Enaineenna Support of the Fuel Oil Transfer System

The inspectors reviewed procedures, regulatory documents, selected PCAQRs, industry

codes, and standards associated with the EDG fuel oil transfer system. The inspectors

also reviewed the cathodic protection system, a support system needed to limit corrosion

of buried EDG fuel oil transfer pipes, to determine if the licensee was maintaining the

system. The inspectors concluded that the licensee had adequately maintained the fuel

oil system in accordance with the documents mentioned above. Further, the inspectors i

concluded that the licensee was appropnately addressing inoperable portions of the

cathodic protection system. Minor discrepancies identified by the licensee as a result of

the inspectors' questions were docuraented in the conective action system.

E8 Miscellaneous Engineering issues (92903) j

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E8.1 (Closed) Licensee Event Report (LER) (50-346/g70016-00 (DRP)): Missed Surveillance

due to Refueling Canal Drain Valve DH-g2 Not in Locked Valve Plogram. This LER

documented the licensee's discovery, during a design basis review, that Refueling Drain

Valve DH-g2 for the reactor cavity was not in the locked valve program and did not have

its position checked every 31 days as required by TS 4.5.2.a.

The licensee determined that if the drain valve had been closed, reactor coolant system

water that spilled into containment followmg a loss-of-coolent accident may have been

retained in the deep end of the refueling canal. Retaining the inventory could have

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caused a significant volume of water to be unavailable to all emergency core cooling

system (ECCS) and containment spray pumps for the recirculation phase of an accident.

The licensee initiated PCAQR 97-it$15 to document the condition, entered containment,

verified the valve was opened and locked the valve. The inspectors verified that the

licensee made provisions through the corrective action system to include the valve in the

locked valve program and procedures. This non-repetitive, licensee-identified and

corrected violation is being treated as a non-cited violation

(NCV 50-346/98002-02(DRP)) in accordance with Vll.B.1 of the NRC Enforcement Policy.  ;

The inspectors concluded that the licensee demonstrated an exceller.t questioning

attitude regarding the valve configuration for the ECCS pump suction valves. The

licensee pro-actively addressed the issue by entering containment and performing

immediate verification of the valve position.

IV. Plant Support

R1 Radiological Protection and Chemistry (RP&C) Controls

R1.1 Spent Fuel Pool Transfer Canal Divino Evolution and a Lift of a Hiah Intearity Container

(71750.71707)

The inspectors observed portions of a diving evolution to replace a corroded stanchion

nut located on the wall of the spent fuel pool transfer canal. The diving evolution was

performed in an area of the pool that had been isolated from where the spent fuel was j

stored. Additionally, the inspectors observed RP personnel performance during a lift of a '

high integrity container which stored highly radioactive resin and spent filters, from the

auxiliary building floor to a transfer container. l

The inspectors concluded that; briefs associated with these evolutions were thorough,

management oversight was effective, health physics support was good, and that

personne! implemented good radiation protection work practices. The ALARA briefs

conservatively predicted the radiological conditions, and tele-dosimetry was effectively

used to provide real time remote dose rates at appropriately placed locations. When

required, foreign material exclusion practices were good. Overall, RP personnel

performance for these two evolutions was good.

F4 Fire Protection Staff Knowledge and Performance

F4.1 Unannounced Fire Briosde Drill (71750)

The inspectors observed the fire brigade team respond to an unannounced fire drill. All

fire brigade members responded to the drill within five minutes of notification. The drill

scenario was designed to expose the fire brigade to a situation where communication

capabilities would be challenged. This was accomplished by choosing the drill location

undemeath the main steam lines in the turbine building. This complicated fire brigade

communications to the point where oral communications could only be performed by

moving fire brigade personnel to a lower noise area. A good post fire drill brief was held

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to discuss communication issues in a noisy environment. Overall, the fire brigade team

effectively responded to a challenging drill scenario.

F8 Miscellaneous Fire Protection issues (92904)

FS.1 (Closed) Unresolved item (50-346/970015-02 (DRP)). Inadequate Hot Short Protection of

Emergency Diesel Generator Speed Circuit. See Section F8.2 for closure of this issue.

F8.2 (Closed) LER (50-346/970015-00 (DRP)): Speed Sensing Circuitry for Emergency Diesel

Generator #1 Not Designed per 10 CFR Part 50, Appendix R Design Criteria for Hot Short

Protection.

Failure to Meet 10 CFR 50.72 One-Hour Reportina Reauirements.

The licensee initiated PCAQR 97-1624 on December 12,1997, after discovering that the

speed circuit for the safe shutdown emergency diesel generator was vulnerable to a hot

short condition during a postulated fire. The PCAQR stated that the tachometer circuit

had associated circuits that would cause EDG #1 not to be able to perform its inlanded

safety function. Further, the PCAQR stated that the hot short would have to occur before

the operator isolated the circuit in order for the condition to be a concom. Subsequently,

the inspectors reviewed the PCAQR and determined that since the PCAQR described a

condition where the EDG could not perform its intended safety function, then the EDG

was not designed per the requirements of 10 CFR Part 50, Appendix R, Section lli G. 2.

The inspectors concluded that information documented in the PCAQR was sufficient for

the licenser to determiae that the EDG condition was outside the design basis of the

plant.

Technical Specification 3.3.3.5.2.b stated that, with one or more inoperable control

cdcuits or electrical disconnect switches, restore the inoperable circuits or switches to an

operable status within 30 days or report this condition to the NRC. While this TS applied

in this case, the application of the TS did not change the fact that the design of the plant

did not meet the design basis requirements. However, the engineering staff did not

initially recognize this fact and consequently the urgency to perform a prompt evaluation

of the issue was reduced and a one hour report for a design deficiency was not made to

the NRC until December 18,1997.

Title 10 CFR 50.72(b)(1)(ii)(B) states, in part, that the iicensee will notify the NRC as soon

as practical, and in all cases within one hour of the occurrence of a condition outside the

design basis of the plant. Contrary to the above, the licensee failed to notify the NRC

within one hour that the speed sensing circuit for EDG #1 was not designed per the

requirements of 10 CFR Part 50, Appendix R, for hot short protection. This is a

10 CFR 50.72 violation (VIO 50-346/98002 43 (DRP))

Failure to Protect Speed Sensino Circuit From a Hot Short Condition

As mentioned above, the lack of protection of the speed sensint circuit f'om a fire

induced hot short constituted noncompliance with the design req. Mments of

10 CFR Part 50, Appendix R, a condition outside the design bat' . d the plant. The

licensee performed an!nvestigation into the cause of the condition. During its

investigation, the licensee identified that a 1986 modification was oerformed to protect

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control room circuits from a control room fire in response to Generic Letter 81-12. The

generic letter discussed the need for licensees to verify ws-Z,es with 10 CFR Part 50,

Appendix R. In 1988, while evaluating the EDG #1 speed sensing circuit for protection

from a fire, the hoensee installed an electrical disconnect switch but did not incorporate a

design to protect the speed circuit for a hot short condstion. The hoensee failed to

incorporate this protective feature into the design because the manufacturer of the speed

circuit did not mention the circuit's vulnerability to a hot short condition. On

December 18,1997, the licensee concluded that a hot short may develop and still render

EDG #1 inoperable before opening the disconnect switch. Consequently the circuit had

not been protected for a hot short condition since implementing the 1986 modifica. tion.

Title 10 CFR Part 50, Appendix R, Section lil. G. 2, requites, in part, that dssociated

non-safety circuits that could prevent operation or cause maloperation due to hot shorts,

of redundant trains of systems necessary to achieve and maintain hot shutdown

conditions which are located within the same fire area outside of primary containment, be

protected. Contrary to the above, prior to December 18,1997, the licensee failed to

provide adequate protection of EDG #1, necessary to maintain hot shutdown conditions,

in that EDG #1 was potentially unable to perform its post-fire safe shutdown function

because the speed sensing circuit for EDG #1 was susceptible to fire-induced hot shorts.

The licensee p6tformed the following immediate and effective corrective actions for this

issue:

A temporary modification to isolate the speed sensing circuit from the control

room portion of the circuit.

Reviewed all circuits to confirm that disconnects were installed and verified that

remaining circuits were adequately protected for a hot short condition.

Proposed to modify the speed sensing circuit to prevent a hot short condition and

allow tachometer indication in the control room.

The inspectors concluded that corrective actions for the issue wwe acceptable. The

violation was identified by the licensee through a review of a condition report from another

facility, the corrective actions were prompt and thorough, the violation was not likely to be

identified by routine licensee surveillances, and the violation is not reasonably linked to

current performance. As a result, this violation will not be cited in accordance with

Section Vll.B.3 of the enforcement policy (NCV 50-346/98002-04(DRP)).

V. Manaaement Meetinas

X1 Exit Meeting Summary

The inspectors presented the inspection results to members of licensee management at

the conclusion of the inspection on February 18,1998. The licensee acknowledgset the

firidings presented, but the plant manager disagreed with the one-hour reporting

requirement violation discussed in Section F8.2. The inspectors asked the licensee

whether any materials examined during the inspection should be considered proprietary.

No proprietary information was identified.

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PARTIAL LIGT OF PERSONS CONTACTED

Ljoensee

J. K. Wood, Vice President, Nuclear

J. H. Lash, Plant Manager

T. J. Myers, Director, Nuclear Support Services

L. W. Worley, Drector, Nuclear Assurance

B. F. Gessel, Director, Human Resources

R. E. Donnellon, Director, Engineering and Services

J. L. Freels, Manager, Regulatory ANairs '

M. C. Beier, Manager, Quality Assessment l

F, L Swanger, Manager, Design Basis Engineering

D. L. Eshleman, Manager, Operations

J. L. Michaelis, M,anager, Maintenance

L. M. Dontmann, PAanager, Quality Services

G. R. McIntyre, Manager-Acting, Plant Engineering

R. J. Scott, Manager, Radiation Protection

P. R. Hess, Manager, Supply ,

H. W. Stevens, Manager, Nuclear Safety & Inspections  !

C. A. Price, Manager, Business Services I

D. H. Lockwood, Supervisor, Compliance ,

D. Ricci, Supervisor, Operations 1

A. Schumaker, Supervisor, Security Support

M. J. Roder, Supervisor, Operations Work Control

D. M. imley, Superintendent, Operations

G. W. Gillespie, Superintendent, Chemistry

S. .M. Livingston, Shift Manager

T. J. Chambers, Shift Manager

S. W. Roberts, Shift Supervisor I

M. A. Koziel, . Senior Auditor, Quality Assurance

D. L. Miller, Senior Engineering  ;

G. M. Wolf, Engineer, Licensing i

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J. Dunn, Senior Training Advisor

T. Kozlowski, Licensing Student

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INSPECTION PROCEDURES USED

IP 37551: Onsite Engineering

IP 61726: Surveillance Observations

IP 62707: Maintenance Observation

IP 71707: Plant Operations

IP 71750: Plant Support Activities

IP 92902: Followup - Maintenance

IP 92903: Followup- Engineering

IP 92904: Followup - Plant Support

ITEMS OPENED AND CLOSED

Opened

50-346/98002-01a(DRP) VIO Failure to Follow Water Balance inventory Test Procedure

50-346/98002-01b(DRP) VIO Failure to Follow Procedure Use and Adherence Procedure

50 346/98002-02(DRP) NCV Failure to Perform TS Required Locked Valve Surveillance

50-346/98002-03(DRP) VIO Failure to Meet 10 CFR 50.72 One-Hour Reporting

Requirements

50-346/98002-04(DRP) NCV Emergency Diesel Generator Design Deficiency

Closed

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50-346/98002-02(DRP) NCV Failure to Perform TS Required Locked Valve Surveillance

50-346/97015-01(DRP) URI Inadequate Testing of Undervoltage Devices

50-346/97015-02(DRP) URI Inadequate Hot Short Protection of EDG Tachometer

50-346/97015-00(DRP) LER Spsed Sensing Circuitry for Emergency Diesel Generator

not Designed Hot Short Protection

50-346/97016-00(DRP) LER Refaeling Canal Drain Valve DH-92 not in Locked Valve

Program

50-346/98002-04(DRP) NCV Emergency Diesel Generator Design Deficiency

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LIST OF ACRONYMS AND INITIAUSMS USED

CFR- Code of Federal Regulations

EDG. Emstgency Diesel Generator

ECCS Emergency Core Cooling System

IR inspection Report

LER Ucensee Event Report

i MWO Maintenance Work Order

MUT ' Makeup Tank

NCV Non-Cited Violation

! NRC Nuclear Regu! story Commission

l PCAQR Potential Condition Adverse to Quality Report

! PDR Public Document Room

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RCP Reactor Coolant Pump

RG Regulatory Guide

RP Radiation Protection

TS Technical Specification

URI Unresolved item

, USAR Updated Safety Analysis Report

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VIO Violation

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