IR 05000263/1999004

Insp Rept 50-263/99-04 on 990521-0701.No Violations Noted. Major Areas Inspected:Licensee Operations,Engineering,Maint & Plant Support

ML20210H106
Person / Time
Site:	Monticello
Issue date:	07/28/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20210H089	List: IR 05000263/1999004 ML20210H086
References
50-263-99-04, NUDOCS 9908030278
Download: ML20210H106 (15)

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. U. S. NUCLEAR REGULATORY COMMISSION i

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Docket No: 50-263 License No: DPR-22

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Report No: 50-263/99004(DRP)

Licensee: Northern States Power Company Facility: Monticello Nuclear Generating Station

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Location: 2807 West Highway 75

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Monticello, MN 55362 I

Dates: May 21 through July 1,1999

! Inspectors: S. Burton, Senior Resident inspector I

D. Wrona, Resident inspector I

Approved by: Roger D. Lanksbury, Chief Reactor Projects Branch 5 DMsion of Reactor Projects l

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f 9908030278 990728 "

gDR .ADOCK 05000263 PDR

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EXECUTIVE SUMMARY Monticello Nuclear Generating Station NRC Inspection Report 50-263/99004(DRP)

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

A reactor startup on May 27,1999, was performed in accordance with approved procedures. Infrequent evolution briefings performed for the startup were thorough and comprehensive. Reactor thermal limits were properly monitored throughout the startu (Section O1.2)

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During a reactor startup with one average power range monitor bypassed per trip system, the licensee had procedural requirements in place to ensure that the minimum number of average power range and associated intermediate power range nuclear instruments remained operable. (Section 01.3)

Maintenance

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Operability of safety-related service water pumps was properly assessed after sandblasting material was introduced into the vicinity of the equipment through the ventilation system during the preparation of the building exterior for paintin '

(Section M1.1)

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Surveillance tests and valve lineups associated with reactor startup were performed by qualified individuals using approved procedures. Deficiencies identified during the performance of activities were properly dispositioned and corrected. (Section M1.2)

• Controls on overtime utilization were adequately implemented by the licensee for the licensee staff that the inspectors assessed. The licensee's control of overtime met the Technical Specification administrative requirements goveming overtime. (Section M8.2)

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Operability determinations for equipment susceptible to failure during a high-energy line break were incomplete and did not include an assessment of the full spectrum of potential single failures as required by the safety analysis report. Rather than perform the complex analysis required to determine equipment operability for the additional single failures not previously analyzed, the licensee elected to correct the deficiency by reinforcing the degraded structure that caused the adverse condition. (Section E1.1)

The engineering department had not fully evaluated the impact of configuring the emergency diesel generator with the droop set above zero in the standby mod Procedural enhancements and an adjustment of one breaker thermal overload setpoint were performed to further ensure operability of safety-related equipment. An engineering evaluation was initiated to reconfigure the emergency diesel generator

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controls to have droop set at zero when the emergency diesel generator was in the standby mode, precluding the continual evaluation of component operability as equipment degraded or was repaired.

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l' Plant Suonort

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Although personnel were responsible to monitor their own accumula!ed dose, the licensee did not lower electronic dosimeter alarm setpoints to reflect radiological conditions during plant shutdown, a poor practice. (Section R8.1)

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ReDort Details Summary of Plant Status The unit began the inspection period shutdown due to an unplanned scram inat occurred on May 8,1999, during the prior inspection period. The reactor was restarted on May 27, the generator was connected to the electrical distribution grid on May 28, and 100 percent power was achieved on May 29,1999. With the exception of a brief reduction to 50 percent power on June 18,1999, to replace recirculation pump motor generator brushes, the unit operated at approximately 100 percent power for the remainder of the inspection perio L ODerations 01 Conduct of Operations 01.1 General Comments (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 (USAR); shift manning; communications; management oversight; proper system configuration and configuration control; housekeeping; operator performance during routine plant operations; the conduct of surveillance tests; and plant power change The conduct of operations was professional and safety-conscious. Evolutions such as

. surveillance tests and plant power changes were well controlled and deliberate, and were performed in accordance with procedures. Shift tumover briefings were comprehensive and were typically attended by the operations superintendent and representatives from the scheduling, security, instrument and control, and electrical and mechanical maintenance departments. Housekeeping was generally good and discrepancies were promptly corrected. Containment isolation valves and portions of the reactor core isolation cooling system were found to be properly aligned. Specific events and notewoithy observations are detailed belo .2 Reactor Startuo On May 27.1999 Insoection Scope (71707)

The inspectors observed portions of activities associated with restart of the reactor following the unplanned shutdown on May 8,1999. Activities observed included pre-job briefings, drywell closure, reactor restart, reactor heat-up, synchronization of the main generator to the electrical distribution grid, and power ascensio Observations and Findinas '

The inspectors observed portions of the operations committee review of startup readiness. Committee members discussed maintenance activities and equipment availability for startup. The inspectors reviewed closure documentation associated with condition reports generated as a result of the scram and found corrective actions to be reasonable and complete. Infrequent evolution briefings performed for the startup were

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thorough and comprehensive, and the shift managers involved with the briefings were clear about their expectations with respect to procedural use, safety, and operators being slow and deliberate during all activities. Reactor startup, heat-up, mode changes, and synchronization of the main generator to the electrical distribution grid were performed in accordance with procedures. During the startup, a process computer printout showed that a reactor fuel thermal limit was exceeded during the continuous withdrawal of one controi rod. Operators, with recommendation from the reactor engineering staff, promptly inserted the rod in question. Reactor engineers determined that the thermal limit had not been exceeded and that the cause of the problem was due to the methodology used by the computer during the continuous rod withdrawal to compute the thermal limits. Subsequently, the control rod in question was withdrawn one step at a time to the same position to which it had been withdrawn earlie Calculations performed throughout the step-withdrawal of the control rod demonstrated that thermal limits were acceptable. Reactor startup continued without further complicatio Conclusions A reactor startup on May 27,1999, was performed in accordance with approved procedures. Infrequent evolution briefings performed for the startup were thorough and comprehensive. Reactor thermal limits were property monitored throughout the startu O1.3 Startuo with Averaae Power Ranae Monitors (APRM) Bvoassed Insoection Scooe (71707)

The inspectors observed that plant startups were conducted with one APRM channel in each reactor trip system continually bypassed. The inspectors reviewed TSs, standard TSs, the USAR, operating procedures, and interviewed engineers with respect to this practic Observations and Findinas

The inspectors were concemed that an interrelating function between intermediate )

Range Monitors (IRM) and APRMs, which was required to be operational when moving )

the mode switch from the startup position to the run position, could inadvertently be bypassed on more than one channel per trip system. When the mode switch was being moved from the staitup position to the run position, and prior to the withdrawal of the IRMs, a downscale condition on an APRM in conjunction with an upscale condition on i

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an associated IRM would cause a scram signal to be input from the associated IRM/APRM circuit. This relationship was commonly referred to as the IRM/APRM companion scram. Because the licensee operated with one APRM per trip system always bypassed, the inspectors were concemed that barriers might not exist to prevent an operator from bypassing an IRM associated with an APRM different from the one always bypassed in a trip system. Under these circumstances, two IRM/APRM companion scram circuits would be bypassed in a single trip system and the operable number of channels of nuclear instruments would be less than the minimum number of operable channels per trip system allowed by TSs. The inspectors interviewed operations department and engineering department personnel and found that an operator aid placard had been in place in the past to prevent this condition. Personnel

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could not readily identify what barrier currently existed to prevent this condition from ,

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inadvertently occurrin A review of the TS indicated that the IRM/APRM companion scram was not specifically identified in TS table 3.1.1, " Reactor Protection System (SCRAM) instrument Requirements," as a requirement for Monticello. Further review found that the companion scram was a requirement of the basis for TS Safety Limit 2.3.A. " Fuel Cladding Integrity Safety Limit." Additionally, a review of standard TSs indicated that the companion scram was required and listed in both the instrument table and the basi Because the operability of the companion scram directly related to the basis for the fuel cladding integrity safety limit, the inspectors communicated their concem to the licensee. The inspectors noted that specific precautions to ensure that the proper number of nuclear instrument channels remained operable under all conditions did not exist. The licensee's reactor engineering group reviewed the configuration and noted that operating procedures require APRMs to be indicating greater than 5 percent of scale prior to operators transferring the mode switch from the startup position to the run position. Although this procedural requirement was not specifically identified as being established to ensure that instrument operability requirements were met, the inspectors concurred that requiring APRMs to be on scale prior to placing the mode switch in the run position effectively ensured that the minimum number of channels per trip system remained operable during startup configuration Conclusions When performing a reactor startup with one average power range monitor bypassed per trip system, the licensee had procedural requirements in place to ensure that the minimum number of average power range and associated intermediate power range nuclear instruments remained operabl . Maintenance M1 Conduct of Maintenancs M1.1 General Comments on Maintenance Activities Insoection Scooe f62707)

in addition to minor maintenance activities observed during routine plant tours, the inspectors observed performance of maintenance activities conducted in accordance with Procedure 4151-1PM, Revision 4,"SBLC [ Standby Liquid Control] Accumulator 11," ;

. and Procedure 4151-2PM, Revision 4, "SBLC Accumulator 12." Observations and Findinas l

Work performed during maintenance activities was professional and thorough. All work j

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was performed in accordance with approved procedures and the workers were knowledgeable of their assigned tasks. Appropriate radiological work permits were followed. The inspectors observed that maintenance supervisors and engineers were )

involved in the activitie i

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Additionally, the inspectors noted that during the sandblasting of an exterior wall in preparation for painting, some of the sandblasting material had been introduced through the ventilation system into the service water bays in the area of the safety-related emergency service water pumps. The licensee had also made a similar observation and evaluated the impact of the material on the operability of the safety-related pumps. The inspectors reviewed the operability determination and discussed it with the licensee. No deficiencies were note Conclusion ,

l Operability of safety-related service water pumps was properly assessed after j sandblasting material was introduced into the vicinity of the equipment through the j ventilation system during the preparation of the building exterior for paintin l M1.2 General Comments on Surveillance Test Activities Inspection Scooe (61726)

The inspectors observed or reviewed the performance of all or portions of the activities contained in the following surveillance test procedures:

• Procedure 2154-13, Revision 23, "RCIC System Prestart Valve Checklist," l performed on May 23,199 Procedure 1371, Revision 4, "Drywell Prestart inspection," performed on l May 24,199 Observation and Findinas in general, the inspectors found that the activities specified in the surveillance test procedures were performed in a professional and thorough manner and completed in accordance with the procedures. Personnel were knowledgeable and generally demonstrated effective three-part communications, self-checking, and peer-checkin When conducted, pre-job briefings were comprehensive. The inspectors frequently observed supervisors and system engineers monitoring job progress. Wher. applicable, appropriate radiation control measures were in plac The inspectors observed operators perform portions of a system lineup in accordance with the "Reactur Core isolation Cooling System Prestart Valve Checklist." During the performance of the lineup, two operators were in the room at the same time, independently working on different portions of the checklist. The inspectors questioned operators on how they verified the position of valves with missing identification tag Operators indicated that they would use the controlled piping and instrument diagrams to locate the valve. The inspectors observed a valve with a missing identification tag and noted that the operators had requested a label. No deficiancies were identifie The inspectors noted no concems when they observed system engineers perform the

"Drywell Prestart inspection." The engineers identified minor discrepancies and initiated work orders to have the issues corrected prior to startup. A small amount of miscellaneous debris was identified by the engineers and by the inspectors and removed from the drywel *

. Conclusions Surveillance tests and valve lineups associated with reactor startup were performed by qualified individuals using approved procedures. Deficiencies identified during the performance of activities were properly dispositioned and correcte M8 Miscellaneous Maintenance issues (71707,92902)

M8.1 (Closed) Licensee Event Report 50-263/98001: Containment isolation Valve Leakage Greater than Allowed by Technical Specification During refueling outage local leak rate testing, the licensee identified that the leakage associated with three containment isolation valves, one of which was the "D" inboard main steam isolation valve, caused the combined maximum flow path leakage to exceed the TS limit. The licensee also identified that the leakage associated with main steam isolation valve leakage was in excess of the TS limi This event was determined to be of low safety significance since the redundant valve for each of the three valves had minimal leakage. All three valves were repaired prior to the licensee retuming the unit to power following the refueling outag In the licensee event report, the licensee stated that the condition could have existed during some of the previous operating period and assessed the impact of the inoperable valves on total allowable leakage. The licensee determined that no limits were exceeded due to the associated in-line valves having a lower than allowable leakag Because the valves were determined to be inoperable at the time of discovery, and the plant had actually operated with leakage less than allowed by TS, the inspectors concluded that no violation of NRC requirements had occurred. This issue was entered into the licensee's corrective action program under CR 98000737, "D MSIV [ main steam isolation valve) Exceeds Allowed Leakage in Technical Specifications"; CR 98000958,

"FW-94-2 [feedwater isolation check valve] Failed LLRT [ local leak rate test) 0137-08-2 on 3/26/98"; and CR 98001091, " Local Leak Rate Failure of AO-2378 [ torus purge inboard containment isolation valve]."

M8.2 Use and Documentation of Overtime Deviations Inspection Scooe (71707)

The inspectors reviewed the following documents:

• Security ingress and egress times for portions of April and May 1999 for four individuals;

• Forms 3361, " Authorization to Exceed Overtime Work Restrictions," completed between April and May 1999;

• 4 AWi-08.10.01, Revision 3," Overtime Restrictions and Fitness for Duty Requirements"; and

• Technical Specification 6.1.F. " Administrative Controls for the Use and Authorization of Overtime."

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. Observations and Findinas The inspectors reviewed overtime use during April and May 1999, which encompassed two forced outages. The inspectors sampled individuals who had major roles associated with equipment concems identified during the outages. The inspectors used the above records to aid in determining potential deviations from overtime guideline The inspectors utilized main gate ingress and egress times as a record of time on site and based the conclusions on this assumption. The inspectors also recognized that the ingress and egress times did not necessarily reflect actual work hour The inspectors performed a spot check of the ingress and egress times for four individuals and noted that working hours were generally limited to the hours specified in TS 6.1.F and when the hours deviated from those specified in TS 6.1.F the licensee appropriately documented and approved the daviatio Conclusions Controls on overtime used were adequately implemented by the licensee for the licensee staff that the inspectors assessed. The iicensee's control of overtime met the Technical Specification administrative requirements goveming overtim Ill. Engineering E1 Conduct of Engineering E1.1 Hich-Enerav Line Break Forces on Masonry Walls I Inspection Scope (37551)

The inspectors reviewed the licensee's actions associated with inspection Followup item (IFI) 50-263/97018-06(DRP) as documented in CR 97003052, "lEB 80-11 Program inadequately Documented the Analysis and Effects of HELB Forces on Masonry Wall Need Further Review." ObservatKms and Findings inspection Followup item 50-263/97018-06(DRP) identified that formal documentation of the effects of a high-energy line break on 78 safety-related masonry walls may not exis i The walls which were evaluated in response to NRC Bulletin 80-11, " Masonry Wall l Design," did not include an evaluation of the effects of area pressurization due to a high-energy line brea The licensee subsequently determined the effects of high-energy line breaks on masonry walls and identified that two block walls, which were part of the stairwell enclosure leading to the "B" residual heat removal room, were susceptible to failur The licensee documented the resolution of this issue in CR 97003052. The licensee !

assumed that the failure of the block walls would render the "B" residual heat removal train and the "B" core spray train inoperable and provided the analysis to show how the requirements specified in the USAR were still complied wit ;

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The set of conditions that were required to be satisfied to adequately cope with a high-energy line break were specified in the USAR. Among these conditions was the ability to cope with a high-energy line break given that a single active failure could also occu The inspectors reviewed CR 97003052 and noted that the licensee did not document all applicable single active failures. The inspectors questioned which single active failures were considered and subsequently provided examples of failures that had not been reviewed which could impact the equipment operability determination. Due to the complexity of analyzing the newly postulated single failures mechanisms, the licensee elected to correct the blockwall deficiency by reinforcing the walls susceptible to failure during a high-energy line break. The inspectors determined that the impact on safety of not completing the full analyses was minimal because HELB guidelines allow for operator actions and the licensee would have been able to detect and correct adverse conditions using operator actions. This negated the need to assess single active failures along with failure of the division "B" equipment previously mentioned abov Conclusions Operability determinations for equipment susceptible to failure during a high-energy line break were incomplete and did not include an assessment of the full spectrum of potential single failures as required by the safety analysis reporte Rather than perform the complex analysis required to determine equipment operability for the additional single failures not previously analyzed, the licensee elected to correct the deficiency by reinforcing the degraded structure that caused the adverse conditio E1.2 Diesel Generator Drooo Set Above Zero Durina Normal Ooerations Insoection Scoos (37551)

The inspectors noted that Procedure 0187-2, Revision 29,"12 Emergency Diesel Generator /12 Emergency Service Water Pump System Tests," did not leave the diesel droop set at zero upon completion of the test. The inspectors noted that operation of the emergency diesel generator (EDG) in this configuration during a loss of offsite power would result in diesel speed varying as loads on the diesel were changed. The inspectors additionally reviewed the Updated Safety Analysis Report, TSs, and the EDG technical manual to assess requirements associated with operation in this configuratio . Observations and Findings i

Droop is a method of creating stability in a diesel generator speed govemor when the generator is operating in parallel with other units. A droop setting above zero ensures that the generator does not attempt to transfer loads between other generators i operating in parallel. A droop setting of zero allows the generator to accept and reject )

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load without an associated change in speed. Droop settings above zero when the generator is operating on an isolated bus, as would be expected during a loss of offsite power, would result in the generator's speed, and associated current frequency, l increasing with a reduction in load and decreasing with an increase in loa Operating on an isolated bus with a varying attemating current frequency would result in motor speeds changing and indirectly affecting all non-resistive loads on the associated bus. The inspectors were concemed that operation in this configuration had the potential to caase pumps to operate in less than design flow or overcurrent conditions, i l

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I, depending upon operation of the EDG at a low or high operating frequency, respectively. Additionally, valve motor torque may be too low or too high to permit i proper valve operation. Changes in EDG frequency also had the potential to impact the '

i operability of solid-state equipment, such as battery chargers, that rely on a specific j frequency band to remain operable.

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The inspectors reviewed the EDG govemor technical manual and found that it l recommended that droop be set at zero when the EDG was operating on an isolated

! bus. The inspectors noted that although operators were trained to maintain EDG l frequency at 60 cycles per second when the diesel was operating on an isolated bus, l procedural precautions or guidance was insufficient to ensure that frequency changes

! ' did not affect equipment operability. Additionally, the inspectors were concemed about j frequency affecting safety-related equipment operability. The licensee agreed with the inspectors and initiated condition report CR 99001403 to evaluate operability of affected components prior to restarting the reactor on May 27,199 The licensee evaluated safety-related loads that were potentially affected by operation in this configuration and determined that all loads would function as designed; however, one load, the #11 emergency service water pump, was determined to have minimal margin to the breaker thermal trip setting. A work order was written and the overload l device was reset prior to startup. Changes were also made to operating procedures to

! ensure that EDG frequency was maintained at 60 cycles per second when the EDGs were operated with droop set above zero. An engineering evaluation was initiated to reconfigure the EDG govemor controls to have droop set at zero when the EDG was in the standby mode. Operation of the EDG with droop set at zero while in standby would preclude the licensee from having to continually evaluate component operability as l . equipment degraded or was repaire Conclusions l

l The engineering departme i had not fully evaluated the impact of configuring the *

emergency diesel generate mih the droop set above zero in the standby mod Procedural enhancements ad an adjustment to one breaker thermal overload setpoint I were performed to further ensure operability of safety-related equipment. An engineering evaluation was initiated to have droop set at zero when the EDG was in the standby mode, precluding the continual evaluation of component operability as equipment degraded or was repaire E (Closed) Insnar41an Followuo item 50-263/97018-06(DRP): Effects of high-energy line break forces on masonry walls not documented. (See Section E1.1 of this report.)

E8.2 - Summarv of Checklist Review Results for Y2K Readiness (Tl 2515/141)

The inspectors conducted an abbreviated review of Y2K activities and documentation using Temporary Instruction (TI) 2515/141, " Review of Year 2000 (Y2K) Readiness of Computer Systems at Nuclear Power Plants." The review addressed aspects of Y2K management planning, documentation, implementation planning, initial assessment, detailed assessment, remediation activities, Y2K testing and validation, notification activities, and contingency planning. The inspectors used NEl/NUSMG [ Nuclear Energy Institute / Nuclear Utilities Software Management Group) 97-07, "NLclear Utility

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Year 2000 Readiness," and NEl/NUSMG 98-07, " Nuclear Utility Year 2000 Readiness Contingency Planning," as the primary references for this revie The results of this review will be combined with the results of other reviews in a summary report to be issued by July 31,199 IV, Plant Sucoort R8- Miscellaneous Radiation Protection issues R Electronic Dosimeter Dose and Dose Rate Setooints Insoection Scone (71750)

During routine tours while the reactor was shutdown, the inspectors assessed radiological area entry requirements and radiation work permit (RWP) compliance for various plant locations. Radiation work permit 42, Revision 17. "911 Turbine -

Condenser Room General Area," and RWP 44, Revision 14, '911 Turbine - Steam Jet Air Ejector Room," were reviewed as part of this assessmen Observations and Findinas On May 23,1999, the inspectors observed that the instructions specified in RWPs associated with general entry into the condenser room and the steam Jet air ejector room while the unit was shutdown required electronic dosimeter settings that were the same as if the unit were in operation. Specifically, RWPs 42 and 44 specified electronic dosimeter dose rate and accumulated dose setpoints of 2,000 millirem per hour (mrem /hr) and 200 mrem; and 3,000 mrem /hr and 200 mrem, respectively. The highest

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q radiation dose rates in the rooms, as posted, was about 10 mrem /hr. The inspectors '

were concemed that the electronic dosimeter alarm setpoints were not sufficient to alert personnel to a change in radiological condition The radiation protection staff stated that personnel were trained to periodically monitor their accumulated dose on the electronic dosimeter and that the electronic dosimeter alarms were used as a backup method of alerting operators to unexpected condition Although personnel were responsible to monitor their own accumulated dose, the licensee did not take advantage of the capabilities of the electronic dosimeter alarm setpoints to aid workers in maintaining awareness of unanticipated changes in radiological conditions. The licensee concurred that this was a poor practice and j elected to establish alarm setpoints that better coincided with shutdown condition Conclusions I

- Although personnel were responsible to monitor their own accumulated dose, the licensee did not lower electronic dosimeter alarm setpoints to reflect radiological conditions during plant shutdown, a poor practic '. 1

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S1 Conduct of Security and Safeguards Activities S General Comments (71750)

During routine act!vities 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 those individuals requiring escorts were properly escorted. The inspector also verified that vital areas were locked and alarmed. Additionally, the inspectors verified that observed personnel and packages entering the protected area were searched by appropriate equipment or by hand The inspectors toured the protected area perimeter fence and found no deficiencie F2 Status of Fire Protection Facilities and Equipment F2.1 , General Comments (7175Q)

During normal resident inspection activities, routine observations were conducted in the area of fire protection. Fire extinguishers and fire hoses were properly stored and inspected by licensee personnel. No notable degradation of equipment was observe Y, Management Meetings X1 ' Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the ;

conclusion of the inspection on July 1,1999. 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 identifie j

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PARTIAL LIST OF PERSONS CONTACTED Licensee B. Day, Plant Manager M. Hammer, Site Manager K. Jepson, Superintendent, Chemistry & Environmental Protecuon M. Lechner, Acting General Superintendent Operations L. Nolan, General Superintendent Safety Assessment E. Reilly, General Superintendent Maintenance C. 9 nibonski, General Superintendent Engineering A. A'ard, Manager Quality Services L. Wilkerson, Superintendent Security J. Windschill, General Superintendent, Radiation Services l

lNSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 71707: Plant Operations IP 71750: Plant Support Activities IP 92700 Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92902: Followup - Maintenance IP 92903: Followup - Engineering Tl 2515/141 Y2K Readinesss Assessment ITEMS OPENED, CLOSED AND DISCUSSED Ooened None Closed 50-263/98-001 LER Containment isolation valve leakage greater than allowed by t Technical Specifications 50-263/97018-06 IFl Effects of high-energy line break forces on masonry walls not documented Discussed None l

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LIST OF ACRONYMS USED AOV Air-Operated Valve APRM Average Power Range Monitor CFR Code of Federal Regulations CS Core Spray DRP Division of Reactor Projects EDG Emergency Diesel Generator IFl Inspection Followup Item IP Inspection Procedure IRM Intermediate Range Monitor LER Licensee Event Report LLRT Local Leak Rate Test mrem /hr millirem per hour NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSP Northem States Power PDR Public Document Room RCIC Reactor Core isolation Cooling RHR Residual Heat Removal RWP Radiation Work Permit TI Temporary Instruction TS Technical Specification USAR Updated Safety Analysis Report i

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