IR 05000263/1999006

Insp Rept 50-263/99-06 on 990813-0923.Non-cited Violations Noted.Major Areas Inspected:Operations,Engineering, Maintenance & Plant Support

ML20217M393
Person / Time
Site:	Monticello
Issue date:	10/21/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20217M383	List: IR 05000263/1999006 ML20217M380
References
50-263-99-06, NUDOCS 9910270212
Download: ML20217M393 (16)

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U. S. NUCLEAR REGULATORY COMMISSION l

REGION lil Docket No:

50-263 License No:

DPR-22 j

Report No:

50-263/99006(DRP)

Licensee:

Northem States Power Company i

Facility:

Monticello Nuclear Generating Station i

Location:

2807 West Highway 75 Monticello, MN 55362 Dates:

August 13 through September 23,1999 Inspectors:

S. Burton, Senior Resident inspector D. Wrona, Resident inspector Approved by:

- Roger D. Lanksbury, Chief Reactor Projects Branch 5 Division of Reactor Projects i

9910270212 991021'

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PDR ADOCK 05000263 G

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EXECUTIVE SUMMARY

~ Monticello Nuclear Generating Station

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NRC Inspection Report 50-263/99006(DRP)

l LThis inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspection.

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Operations The conduct of operations was generally characterized by good procedural compliance,

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appropriate evaluations of risk on related work activities, proper three-way communications, and safety-conscious performance of activities. (Section 01.1)

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The licensee properly considered risk during maintenance work and postponed the repair of a power supply in the rod position information system. The maintenance was postponed because concurrent maintenance that was scheduled on load center 109 increased the potential for a reactor trip (scram) coupled with a loss of rod position

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indication. (Section 01.2)

r Maintenance Activities specified in surveillance test procedures were performed in a professional and i

thorough manner by qualified technicians and operators who completed the activities in l-accordance with procedural requirements, using proper radiation protection practices with radiation protection technicians present as required, calibrated test equipment; good self-checking techniques, three-way communications, and good communications with the control room. Supervisors and system engineers frequently monitored job progress. (Section M1.2)

The licensee identified that with the residual heat removal system in the suppression

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pool cooling mode an accident event sequence existed that could result in the number of available emergency core cooling system pumps being less than that assumed in the Updated Final Safety Analysis Report. Because the licensee had not previously identified this potential problem, plant operating procedures had not been written to require entry into a Limiting Condition for Operation. This procedure inadequacy was identified by the inspectors as a non-cited violation. (Section M8.1)

The licensee identified that the high pressure coolant injection test return valve was

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unable to close against maximum expected differential pressure during certain assumed accident sequences. : Because the high pressure coolant injection system would be inoperable with this valve open, a Limiting Condition for Operation should have been

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had not previously identified this potential problem, plant operating procedures had not been written to require entry into the Limiting Condition for Operation. This procedure

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!nadequacy was identifiext by the inspectors as a non-cited violation. (Section M8.4)

l Engineerina A safety review performed by the licensee to show that the 12 core spray pump

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remained operable did not use the most conservative pump conditions as calculation

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inputs. However, the margin to exceeding bearing design temperatures remained large and no operability concerns were identified. (Section E1.1)

A review by the inspectors of the corrective action program as it related to inservice

testing identified several potential weaknesses and one failure to declare the " reactor building to torus vacuum breaker" inoperable when it did not meet acceptance criteria.

A non-cited violation was issued for failure to follow procedural requirements after the breaker did not meet inservice testing acceptance criteria. (Section E1.2)

Plant Sucoort The method for tracking surveillance testing requirements afforded the opportunity to

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apply 25 percent time extensions allowed for Technical Specification sJrveillance requirements to non-Technical Specification requirements. A non-cited violation of NRC requirements was identified for a failure to inventory special nuclear material at a frequency not to exceed 12 months. (Section G1.2)

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Report Details

' Summarv of Plant Status The unit began the inspection period at approximately 100 percent power and remained at that

. power evel throughout the inspection period with 2 exceptions: short power reductions of less l

than 10 percent on September 13 and September 21 for control rod pattern changes.

l. Operations

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Conduct of Operations 01.1 General Comments a.

Insoection Scope (71707)

The inspectors observed various aspects of plant operations, including compliance with Technical Specifications (TSs); conformance with plant procedures and the Updated Safety Analysis Report; shift manning; communications; management oversight; proper system configuration and configuration control; material condition; operator performance during routine plant operations; the condect of surveillance tests; and plant power changes.

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Observations and Findinos The conduct of operations was generally characterized by good procedural compliance, evaluations of risk on related work activities, proper three-way communications, and safety-conscious performance. Operator responses to annunciators were, with a f6 y exceptions, good. Operators generally announced annunciators as " expected" or

" unexpected" and reviewed the associated annunciator response procedure when required. Evolutions such as surveillance tests and plant power changes were well controlled and deliberate, and were performed in accordance with procedures. Shiit tumover briefings were comprehensive and were typically attended by the plant manager, the operations superintendent; and representatives from the scheduling, security, instrument and control, and electrical and mechanical maintenance departments. Material condition was generally good and discrepancies, such as an extension ladder not secured with rope or wire as required by the site safety manual, and extension cords in the turbine building without fault protection, were brought to the attention of the licensee and corrected. Contair..nent isolation valves were observed to be properly aligned. Specific events and noteworthy observations are detailed below.

N*ing a routine tour of the reactor building, the inspectors observed an air leak

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in an air supply line to the scram valves and notified the licensee. The licensee i

determined that an elbow was cracked and, if it had catastrophically failed, could have resulted in that rod inserting fully; consequently, the licensee fully inserted the rod and isolated the leak. The licensee initiated Condition Report

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(CR) 9902774, "Through Wall Pipe Fitting Failure in Air Header for (Hydraulic Control Unit] HCU-06-39," to document the failure and initiated Work Order

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. (WO) 9907250, * Load Audible Air Leak 06-39 Scram Air Header," to repair the fitting.

The inspectors attended an Operations Committee (OC, the onsite review

committee) meeting on August 27,1999. Several regular OC members were unavailable to attend and attemate members were assigned. The OC quorum requirements identified in 4 awl-04.07.01, Revision 12, " Operations Committee Charter," and TS 6.2.B were met. Although the meeting was called on short notice, me%e.s were provided adequate time to review the rr aterial prior to the meeting.

The inspectors observed equipment tag-outs 99-80187, " Relief

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Valve [RV-7364B] Appears to Leak By," and 99-80247, for the control rooin emergency ventilation system air conditioning u.,it condenser. No discrepancies

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were identified.

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The inspectors walked-down the accessible portions of the residual heat removal

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(RHR) and RHR service water systems. No discrepancies were identified.

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Conclusions The conduct of operations was generally characterized by good procedural compliance, appropriate evaluations of risk on related work activities, proper three-way communications, and safety-conscious performance of activities.

01.2 Aoolication of Risk insichts for Repair of Load Center (LC)-109 Normal Power Source.

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Inspection Scone (71707)

Inspectors observed operations activities associated with the repair of the normal power supply to LC-109.

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Qbservations and Findinas

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' Due to a fault on the normal power supply to LC-109, the load center was being powered through a cross-connect from LC-102. On August 13,1999, when planning I

maintenance to restore the system _to normal, the inspectors noted that the licensee appropriately evaluated risk for associated corrective maintenance activities. During the same time period, the licensee identified that a power supply to the rod position information system (RPIS) required replacement. The licensee determinef that while in

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the current configuration a fault on LC-102 or LC-109 increased the probat sty of a

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reactor trip (scram). The licensee also determined that rod position indication would be

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lost during RPIS power supply replacement. _ A scram coupled with a loss of rod position indication would result in entry into the "Fallure to SCRAM" section of the Emergency Operating Procedures (EOPs). Due to the increased risk of a scram in this configuration, the licensee elected to delay the replacement of the RPIS power supply until LC-109 was being supplied from its normal source.

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Conclusions The licensee properly considered risk during maintenance work and postponed repair of a power supply in the rod position informaticn system. The maintenance was postponed because concurrent maintenance that was scheduled on ioad center 109 increased the potential for a reactor trip (scram) coupled with a loss of rod position indication.

i 11, Maintenance M1 Corduct of Maintenance M1.1 Goperal Comments on Maintenance Activities a.

insoection Scooe (62707)

In addition to minor maintenance activities observed during routine plant tours, the inspectors observed performance of maintenance activities associated with the following documents:

WO 9906827, "Repiace RPIS Power Supply PSX6," performed on

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August 18,1999.

WO 9905452, "V-FU-3B Inboard Fan Bearing is Going Bad," performed on

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August 31,1999.

_WO 9907162, " Replace Valve FP-5," performed on September 4,1999.

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Observations ano Findinos Work performed during maintenance activities included detailed infrequent evolution briefings, lessons-leamed discussions regarding other licensee experiences, and a review of expected responses and potential risks associated with various maintenance tasks. In general, all work was performed in accordance with approved procedures and the workers were knowledgeable of their assigned tasks. The inspectors observed that maintenance supervisors and system engineers were involved in the activities. On August 31,1999, while observing maintenance activities associated with WO 9905452,

"V-FU-3B Inboard Fan Bearing is Going Bad," the inspectors identified that the waming placard affixed to the fan unit, which noted secondary containment concerns, referenced a procedure that had been deleted. The inspectors informed the licensee, who subsequently replaced the placards.

' M1.2 General Comments on Surveillance Test Activities a.

Inspection Scooe (61726)

The' inspectors ot.A.s "r reviewed the performance of all or portions of the activities contained in the follow,,,, surveillance test procedures.

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Surveillance Test Procedure 0255-06-1 A-1, Revision 43, "HPCI [High Pressure

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Coolant injection) System Test With Reactor Pressure at Rated Conditions,"

performed on August 19,1999.

Surveillance Test Procedure 00548, Revision 2, " Main Steam Line Low Pressure

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Group 1 Isolation Instrument Test and Calibration (Rx in Run)," performed on August 26,1999.

' Surveillance Test Procedure 0270, Revision 19, " Fire Protection System Valve

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Position Verification," performed on August 30,1999.

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Observation and Findinas The inspectors observed that activities specified in the surveillance test procedures were performed using proper radiation protection practices with radiation protection technicians present when required; calibrated test equipment; good self-checking techniques; three-way communications; and good communications with the control room. The inspectors frequently observed supervisors and system engineers i

monitoring job progress. Equipment deficiencies identified by technicians during the l

performance of surveillance tests were properly noted and work orders were initiated.

On August 13,1999, during performance of activities specified in Surveillance Test Procedure 0255-06-1 A-1, Revision 43, "HPCI System Test With Reactor Pressure at Rated Conditions," three of the four RHR pumps were placed in the suppression pool cooling mode of operation. The licensee entered the appropriate TS limiting conditions for operation when the RHR system was in the suppression pool cooling mode and when the HPCI system was unavailable due to testing. While the HPCI system was in

operation for the test, suppression pool temperature exceeded 90*F (degrees Fahrenheit). The licensee appropriately entered the EOPs, placed the remaining RHR pump in the suppression pool cooling mode and re-established suppression pool temperature to less than 90*F. No concems were identified.

The inspectors identified no discrepancies when they performed a spot check of valves in the fire protection flow path that were verified during TS surveillance test Procedure 0270.

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- Conclusions Activities specified in surveillance test procedures were performed in a professional and thorough manner by qualified technicians and operators who completed the activities in j

accordance with procedural requirements, using proper radiation protection practices -

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' with radiation protection technicians present as required, calibrated test equipment,

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- good self-checking techniques; three-way communications, and good communications with the control room. Supervisors and system engineers frequently monitored job progress.

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' Miscellaneous Maintenance issues (92700)

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M8.1 : (Closed) Licensee Event Report (LER) 50-263/99-002: Procedures did not require a

. Limiting Condition for Operation (LCO) entry during suppression pool cooling.

The licensee identified an event sequence that would result in the number of available emergency core cooling system pumps being less than that assumed in the USAR. If a

division of RHR were in suppression pool cooling and a large-break loss-of-coolant accident occurred with a loss of offsite power and a failure of the same division

. emergency diesel generator, the RHR system would not re-align to the low pressure

. coolant injection mode, thus diverting some of the low pressure coolant injection flow to the suppression pool.

Appendix B, Criterion V " Instructions, Procedures, and Drawings," of 10 CFR Part 50, requires;in part, that activities affecting quality be prescribed by documented instrudons of a type appropriate to the circumstances. Administrative Work Instraction 4 AWi-02.03.03, Revision 10. " Work Procedure Preparation," required, in prst, that procedures that required entry into a TS LCO specificelly state the equipment

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licensee identified that plant operating procedures did not require entry into an LCO while RHR was in suppression pool cooling. The safety significance of this event was minimal since the time limit imposed by the TS was not exceeded, the amount of time spent in su%.sion pool cooling was minimal, and the probability of a large-break loss-of-coolant accident coincident with a loss of offsite power and an emergency diesel generator failure was low. This Severity Level IV violation is being treated as a Non-Cited Violation (NCV), consistent with Appendix C of the NRC Enforcement Policy. This issue v% in the licensee's corrective action program as CR 99000966, " Single Failure Vulnerability of the RHR System when in Suppression Pool Cooling Mode"

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M8.2 (Closed) LER 50-263/99-005:. Manual scram inserted when pressure transient closes air ejector suction isolation valves and trips off-gas system.

This issue was discussed in Section 01.7 of Inspection Report 50/263-99003(DRP). No new issues were identified following the inspectors' review ol thI LER.

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. MB.3 ' (Closed) LER 50-263/99-006-During quarterly surveillance teving, HPCI declared inoperable due to drain pot alarm.

A degraded steam trap in the HPCI steam supply line caused a drain pot high level alarm during HPCI testing. Operators followed the instructions contained in the annunciator response procedure, declared the HPCI system inoperable, and entered the appropriate TS LCO. The trap subsequently drained and HPCI system opersbility was

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restored prior to exceeding TS LCO time limits. No violations of requirements were identified.' This issue was in the licensee's corrective action program as CR 99001535,

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"HPCI Declared INOP (Inoperable] due to Inlet Drain Pot High Level During Stroking MO-2036 During Surveillance Test."

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M8.4. (Closed) LER 50-263/99-QQI: ' HPCI test retum valve unable to close against maximum expected differential pressurei During a review of Generic Letter (GL) 96-05, " Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves," the licensee identified that the thrust required to close the HPCI test return valve against the maximum expected differential pressure exceeded the maximum allowable valve thrust. If the valve were open and received a signal to close, it could fail to fully close and thus HPCI would be unable to provide the required injection flow..

Appendix B, Criterion V, " Instructions, Procedures, and Drawings," of 10 CFR Part 50, requires, in part, that activities affecting quality be prescribed by documented instructions of a type appropriate to the circumstances. Administrative Work Instruction 4 AWi-02.03.03, Revision 10, " Work Procedure Preparation," required, in part, that procedures that required entry into a TS LCO shall specifically state the

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equipment that'was considered inoperable or the LCO being entered. Contrary to the j

above, the licensee identified that plant procedures did not require entry into an LCO while the HPCI test retum valve was open. The safety significance of this event was minimal because the time limit imposed by the TS was not exceeded, the amount of j

time that the HPCI test retum valve was open was minimal, and the automatic

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depressurization system (ADS) remained available to reduce pressure for the pressure cooling. This Severity Level IV violation is being treatsd as an NCV, consistent with

' Appendix C of the NRC Enforcement Policy. This issue was in the licensee's corrective action program as CR 99001566, "HPCI to be Declared inoperable with MO-2071 [HPCI Test Retum Isolation Valve] in the Open Position" (NCV 50-263/99006-02(DRP)).

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lil. Engineering E1 Conduct of Engineering E1.1 - Safety Review item (SRI) for " Core Sorav Pumo Motor Without Water Coolina".

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Inspection Scope (37551)

The inspectors reviewed SRI 95-002, Revision 1, " Core Spray Pump Motor Without

. Water Cooling," and discussed the associated calculations with the licensee. The licensee used SRis to document engineering department evaluations of equipment operability questions.

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Observations and Findinas The licensee issued SRI 95-002 to show that the 12 core spray (CS) pump remained operable with degraded flow to the motor cooler. The SRI used calculations'and test data to verify operability. ' One of the CS pump operating conditions used as an input to the calculation was a flow of 3000 gallons per minute at 300 pounds per square inch -

gauge. The inspectors identified that this pump flow condition was not the most conservative, in terms of motor temperatures. For example, the pump technical manual showed that flow rates greater than 3000 gallons per minute resulted in higher power

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requirements. Although the most conservative value for pump conditions was not used

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in the calculations, the margin to exceeding bearing design temperatures remained j

large. The inspectors had no further concems.

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A safety review performed by the licensee to show that the 12 core spray pump remained operable did not use the most conservative pump conditions as calculation inputs. However, the margin to exceeding bearing design temperatures remained large and no operability concems were identified.

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E1.2 Inservice Testina Proaram Exemotions From the Corrective Action Proaram a.

Insoection Scooe (37551)-

The inspectors reviewed Surveillance Test Procedure 0141, " Reactor Building to Torus Vacuum Breaker Operability Test," and the associated Form 3108,

"PumpNalve/ instrument Record of Corrective Action." The licensee used Form 3108 for drumenting failures of equipment during inservice testing and for subsequent evaluations.

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Observations and Findinos During the' licensee's performance of activities specified in Survei!!ance Test Procedure 0141 for Valve AO-2379, " Reactor Building to Torus Vacuum Breaker,"

the inspectors observed that the valve did not open within the time limit identified in the acceptance criteria. The licensee initiated the appropriate forms for evaluating the condition, but a condition report was not immediately initiated. By procedure, conditions documented on Form 3108 did not need to be documented on a CR.

Because the Form 3108 process had a blanket exemption from the condition reporting process,- the inspectors reviewed the Form 3108 process to determine if it adequately paralleled all of the critical elements of the licensee's corrective action program. The inspectors found that the Form 3108 process did not clarify when a condition was significant, nor did it require the same level of management review as a significant CR required. Because Form 3108s were not a component of the maintenance rule tracking database, the process could have had the potential to circumvent tracking of non-TS, maintenance rule equipment degraded conditions. The inspectors also found that the -

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process was not clear as to what portions of the corrective action program the Form 3108 process were exempted from. This established a potential for a Form 3108

to be issued that may also require a CR under other requirements, yet result in a CR not

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iThe inspectors reviewed the performance of air-operated valve AO-2379 for the last-2 years to determine if any condition reporting criteria had been overlooked. No significant trends were identified associated with the condition reporting process and Valve AO-2379.

The inspectors did find that valve AO-2379 had exceeded acceptance criteria during

. performance of a surveillance test in June 1998 and that it was not declared inoperable in accordance with Procedure 4 AWi-09.09.01, " Inservice Testing Program

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implementation." Section 6.5 of the TS required, in part, that written procedures for surveillance testing requirements that could have an effect on nuclear safety be prepared and followed. Contrary to the above, valve AO-2379 was not declared inoperable in accordance with Procedure 4 AWi-09.09.01, " Inservice Testing Program implementation." This Severity Level IV violation is being treated an NCV, consistent with Appendix C of the NRC Enforcement Policy. This issue was entered into the licensee's corrective action program as CR 99002917, " Stroke Timing of AO-2379 in 1998 resulted in closed time exceedag LST [ Limiting Stroke Time] but valve was not declared inoperable." (NCV 50-263/99006-03(DRP))

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Conclusions A review by the inspectors of the corrective action program as it related to inservice testing identified several potential weaknesses and one failure to declare the " reactor building to torus vacuum breaker" inoperable when it did not meet inservice testing acceptance criteria. A non-cited violation was issued for failure to follow procedural requirements after the breaker did not meet inservice testing acceptance criteria.

E1.3 Travelino Screen Operability (37551)

The inspectors noted an increasing trend in the number of CRs which identified sheer pin failures for the traveling screens. Because the performance of the screens and associated screen-wash system had the potential to impact the availability of the ultimate heat sink, the inspectors reviewed the licensee's CRs and engineering department response to the degrading condition at the plant cooling water intake structure. The inspectors found that the engineering department had previously noted the negative performance trend and had initiated a review to determining the cause of the problem. No negative observations were noted by the lospectors.

JV. Plant Support j

R1 Radiological Protection and Chemistry Controls R1.1 General Comments (71750)

During routine tours of the plant and observations of plant activities, the inspectors observed that access doors to locked high radiation areas were properly secured, areas were properly posted, and personnel demonstrated proper radiological work practices.

R2 Status of Radiological Protection and Chemistry Facilities and Equipment R2.1 Radiation Protection Eauioment in the Emeraency Operatina Facility (EOF) (71750)

The inspectors toured the EOF to verify equipment was staged and available for use as needed. The inspectors obserwd that the continuous air monitor (CAM) in the EOF had a " defective equipment" tag placed on it. The inspectors questioned the licensee on the status of the CAM and asked if any compensatory measures were in place. The licensee informed the inspectors that WO 9905794 was initiated for repair of the CAM

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and Operations Manual A.2-808, Revision 1. " Radiological Monitoring and Control at the EOF," contained instructions for radiation protection personnel to perform air sampling

. and analysis if the EOF CAM was inoperable. The inspectors had no further concerns.

81-Conduct of Security and Safeguards Activities S1.1 General Comments (71750)

During routine activities or tours, the inspectors monitored the licensee's security

- program to ensure that observed actions were being implemented according to the approved security plan. The. inspectors observed that persons within the protected area displayed proper photo-identification badges and that those individuals requiring escorts

were properly escorted. Additionally, the inspectors verified that observed personnel and packages entering the protected area were searched by appropriate equipment or by hand. The inspectors toured portions of the protected area perimeter fence and observed no deficiencies.

S1.2 Saggial Nuclear Material Inventory Frecuency a.

Inspection Scoce'(71750)

Inspectors reviewed the licensee's application of the TS-allowed 25 percent extension of surveillance testing intervals to non-TS requirements for inventory of special nuclear material, b.

Observations and Findinas -

The inspectors noted that the TS and non-TS surveillances were tracked utilizing the same method. Inspectors were concerned that this afforded the licensee the opportunity to apply 25 percent extensions allowed for TS surveillance test requirements.

. to non-TS requirements. The inspectors noted that 10 CFR 70.51(d) required a

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physical inventory of all special nuclear material (SNM) be performed at a frequency not to exceed 12 months and reviewed the licensee's inventory program to determine if the potential to exceed the 12-month period existed. The licensee reviewed the

SNM surveillance (inventory) Intervals and found that 25 percent extensions had

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- been applied to this requirement. The licensee initiated CR 99002817,

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" Tech Spec 25 percent Extension Allowance Applied to Required Surveillance Frequency That Has Non-Tech Spec Source," to evaluate the extent and magnitude of l

the problem. The failure to perform a physical inventory of SNM at a frequency not to

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exceed 12 months as required by 10 CFR 70.51(d) is a Severity Level IV violation and is being treated as an NCV, consistent with Appendix C of the NRC Enforcement Policy.

This issue was entered into the licensee's corrective action program as CR 99002817 (NCV 50-263/99006-04(DRP)).

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Conclusions'

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The method for tracking surveillance testing requirements afforded the opportunity to apply 25 percent time extensions allowed for Technical Specification surveillance testing requirements to non-Technical Specification requirements. A violation of NRC requirements was identified for a failure to inventory special nuclear material at a frequency not to exceed 12 months.

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F2 Status of Fire Protection Facilities and Equipment F2.1 General Comments (71750)'

During normal resident inspection activities, routine observations were conducted in the area of fire protection. No notable degradation of equipment was observed.

V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management following the conclusion of the inspection on September 23,1999. 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 identified.

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i PARTIAL LIST OF PERSONS CONTACTED

i Licensee B. Day, Plant Manager

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J. Grubb, General Superintendent Engineering j

M. Hammer, Sits Manager K. Jepson, Superintendent, Chemistry & Environmental Protection E. Reilly, General Superintendent Maintenance i

J._Rootes, Acting Manager Quality Services i

I C. Schibonski, General Superintendent Safety Assessment E. Sopkin, General Superintendent Operations.

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L. Wilkercon, Superintendent Security J. Windschill, General Superintendent, Radiation Services -

INSPECTION PROCEDURES USED IP 37551:

Onsite Engineering IP 61726:

Surveillance Observations IP 62707:

Maintenance Observations IP 71707:

Plant Operatiores IP 71750:

Plant Support Activities IP 92700:.

Onsite Followup of Written Reports of Nonroutine Events at Power Reactor i

Facilities

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ITEMS OPENED, CLOSED AND DISCUSSED Opened 50-263/99006-01-

'NCV Single failure vulnerability of the RHR system when in suppression pool cooling mode 50-263/99006-02 NCV HPCI to be declared inoperable with MO-2071 [HPCI Test Return lsolation Valve? in the open position 50-263/99006-03 NCV ~ Stroke time of AO-2379 in 1998 exceeded acceptance criterion but valve was not declared inoperable

50-263/99006-04

- NCV. Technical Specification 25 percent extension allowance applied to non-TS required surveillance frequency Closed 50-263/99006-01 NCV Single failure vulnerability of the RHR system when in suppression pool cooling mode

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'50-263/99006-02 NCV HPCI to be declared inoperable with MO-2071 [HPCI Test Return Isolation Valve] in the open position 50-263/99006-03 NCV Stroke time of AO-2379 in 1998 exceeded acceptance criterion

- but valve was not declared inoperable

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50-263/99006-04 NCV Technical Specification 25 percent extension allowance applied to non-TS required surveillance frequency ~

i 50-263/99-002 LER Procedures did not require an LCO entry during suppression pool

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cooling 50-263/99-005 LER Manual scram inserted when pressure transient closes air ejector suction isolation valves and trips off-gas system 50-263/99-006 LER During quarterly surveillance, HPCI declared inoperable due to drain pot alarm 50-263/99-007 LER HPCI test return valve unable to close against maximum expected differential pressure Discussed None

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LIST OF ACRONYMS USED CAM Continuous Air Monitor CFR Code of Federal Regulations CR Condition Report CS Core Spray

• F Degrees Fahrenheit DRP Division of Reactor Projects EDG Emergency Diesel Generator EOF Emergency Operating Facility EOP Emergency Operating Procedure ESW Emergency Service Water GL Genenc Leuer HPCI High Pressure Coolant injection IFl Inspection Followup Item IP inspection Procedure LER Licensee Event Report LC Load Center LCO Limiting Condition for Operation NCV Non-Cited Violation NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSP Northern States Power OC Operations Committee OWI Operations Work Instruction PDR Public Document Room RCIC Reactor Core Isolation Cooling

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RHR Residual Heat Removal RP&C Radiological Protection and Chemistry

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i RPIS Rod Position Information System RWP Radiation Work Permit SNM Special Nuclear Material SRI Safety Review item SRO Senior Reactor Operator Tl Temporary Instruction TIP Traversing incore Probe TS Technicel Specification URI Unresolved item WO Work Order l

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