IR 05000346/1998002

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)

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! U. S. NUCLEAR REGULATORY COMMISSION l REGIONlli Docket No: 50-346 License No: NPF-3 l

1 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 0 PDR !(

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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 inspectio 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 procedure 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 effectiv 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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f Resort Details Summary of Plant Status The unit operated at neady full power throughout the inspection perio 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 progra 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 belo .2 Operator Failures To Follow Procedure Durina a Test and a Plant EvW% 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 even 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 norma 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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t engineering personnel, a loss of seal injection would have to occur for much longer than I three seconds before the seals would be degrade 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 rushe Technical Specificaten 6.8.1.c states, in part, that written procedutos shall be implemented covering surveillance and test activities of safety-related equipmen 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 roo '

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 cleare 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 setpoin 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 3.5.1 Consequently, he did not perform Step 3.5.18 in sequence with Steps 3.5.17 and 3.5.1 __

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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)). Conclusions l

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. 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 walkdown Operator Knowledge and Performance 0 Eauipment Operator Tours (71707)

l 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 tour l i

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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 bette 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 condition 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 instruction .

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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; conducte 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 remova lit. Ennineerina E2 Engineering Support of Facilitiec and Equipment E 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 syste E8 Miscellaneous Engineering issues (92903) j

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I E (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. 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 acciden 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 Polic ;

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 positio 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 containe 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 goo 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 scenari 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 issu 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 Protectio Failure to Meet 10 CFR 50.72 One-Hour Reportina Reauirement 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. 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 plan 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 18,199 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. tio 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 short 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 circui *

Reviewed all circuits to confirm that disconnects were installed and verified that remaining circuits were adequately protected for a hot short conditio *

Proposed to modify the speed sensing circuit to prevent a hot short condition and allow tachometer indication in the control roo 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 proprietar No proprietary information was identifie r  ;

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l 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 a

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 ED 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 j

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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 i

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