IR 05000155/1997009

Insp Rept 50-155/97-09 on 970617-0804.No Violations Noted. Major Areas Inspected:Operations,Maint,Engineering

ML20217J348
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	10/09/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20217J332	List: IR 05000155/1997009 ML20217J329
References
50-155-97-09, 50-155-97-9, NUDOCS 9710210046
Download: ML20217J348 (15)

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U.S. NUCLEAR REGULATORY COMMISSION REGION til Docket No: 50 155 License No: DPR 06 Report No: 50155/97009(DRP)

Licensee: Consumers Energy Company Facility; -

Big Rock Point Nuclear Power Plant Locatien: 10269 U.S. 31 North Charlevoix, MI 49720 Dates: June 17 August 4,1997 Inspectors: R. J. Leemon, Senior Resident inspector C. E. Brown, Resident inspector Approved by: Bruce L. Burgess, Chief Reactor Projects Branch 6 9710210046 PDR 971009 ADOCK 05000155 0 PDR_ . -

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EXECUTIVE SUMMARY Big Rock Nuclear Power Plant l NRC inspection Report No. 50155/97009 i

This routine inspection covered aspects of licensee operations, engineering, maintenance, and plant suppor OpERll9M e

Operators failed to monitor the location of the channel handling tool when straightening the cable on the spent fuel pool bridge holst winch, resulting in the tool snagging on the i

channel rack. A lack of attention to detail was exhibited by operators during this skill-of-the craft evolution. However, the actions taken after the tool snagged the channel rack i were in accordance with management's expectations and off normal procedure (ONP)

2.105, * Fuel Core Component Damage." No equipment was damaged. (Section 01.2)

e The licensee demonstrated good work control processes during the spent fuel bundle reconstitution activities. (Section 02.1)

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Observed maintenance and surveillance activilles were appropriately performed and

, accurately documented. (Section M1,1)

e The licensee demonstrated good performance in assessing the containment domineralized water supply valve's failure to meet closing time acceptability requirements. Careful assessment by maintenance personnel resulted in the development of good root causes for the valve failure and a conservative action plan to return the valve to operability. (Oection M1.2)

The inspectors identif+d a lack of questioning attitude during the planning and physical maintenance activities associated with the maintenance efforts on the containment equipment air lock ramp hydraulic system that resulted in the unexpected release of hydraulle fluid. (Section M4.1)

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The spent fuel pool (SFP) cooling system's function of maintaining the SFP within design temperature was not affected by plugging 13 tubes. However, engineering personnel:

1) Incorrect ly transferred data, desphe three party verification, resulting in a failure to plug two degraded tubes, and 2) failed to provide engineering justification for selection of the tube plugging criterion, resulting in the need for reanalysis and selection of a new tube plugging criterion, (Section E2.1)

The inspectors identified a non cited violation (NCV) for failing to document the removal of individual filter elements in the instrument air system. Although configuration control played a part in the occurrence of this NCV, the condition report documenting the absence of the filter element did not address configuration control, (Section E3.1)

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Report Details sigmmary of PlanLSjalyg The licensee continued to allow reactor power to slowly decline (coast down) for the entire inspection period. A voltage emergency was declared at 10:20 a.m. on June 24, 1997, during a period of very hot weather. During the voltage emergency, the licensee took actions to reduce all unnecessary plant loads and to secure all nonessential site loads. Actions that could threaten the stability of the system were not allowed, The voltage emergency was downgraded to a voltage alert at 8:13 p.m. on June 24 and the voltage alert was exited at 5:39 a.m. on June 26 ;997, 1, Operations 01 Conduct of Operations 01.1 GeneralComments (71707)

Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. Specific events and findings are detailed in the sections belo .2 Spent Fuel Channels Moved from Soent Fuel Pool to Channel Storaae Rach Inspection Scope The inspector observed operations personnel moving used fuel channels from the spent fuel storage racks to the fuel channel rack Observations and Findinas On June 24,1997, at 10:00 a.m., while rewinding the fuel pool winch cable, an operator caught the channel handling tool on a channel rack. The toel caught on the channel rack subsequent to the identification of a loosely wound winch cable on the winch drum. The loose cable was caused by attaching a tool to the cable winch that was too light to maintain cable tension. After identification of the loosely wound winch, operators lowered the cable and toolinto an empty channel rack. Previously, the inspector observed the winch operator monitoring the tool's position in the rack and the cable tension on the winch. The winch operator did not become aware of the tool snagging on the channel rack until it popped loose. The inspector observed that the winch cable was under strain for about 2 inches of travel. The inspector concluded that the strain placed on the cable did not appear to deflect the equipment enough to cause damage to the tool attached to the winc The refueling senior reactor operator (RSRO)immediately stopped spent fuel pool activit% and called the shift supervisor (SS). The RSRO then appropriately used off-norm 'Adure (ONP) 2.105, Fuel and Core Component Damage" to assess fuel pool equipmt amage. The winch, cable, tool, and channel rack were visually inspected with no can, age found. After the inspection, the plant manager gave permission to resume fuel channel moves. Condition Report (CR) C BRP 97 0369, " Snagging Channel

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Grapple on Channel Rack," was written and used to brief plant operators during the pre-  !'

Job briefing for the next fuel channel move i The above aulvity was performed using * skill of the craft" and had not been proceduralized; therefore, no procedure violation occurred. The inspector observed this event from the refueling floor and concludt.d that the tool should not have been lowered into a channel rack to straighten the cable out. The tool should have been lowered into a i vacant spaco in the spent fuel pool (SFP)instead. The Inspector concluded that to i prevent snagging the tool, the second operator should have been observing the tool and ,

communicating it's location to the winch operato ! Conclusion Operators failed to monitor the location of the channel handling tool when straightening the cable on the spent fuel pool bridge hoist winch, resulting in the tool snagging on the channel rack. The operators were not attentive to the location of the refueling tool while in the channel rack during this skill-of the craft evolution. However, the actions taken after the tool snagged the channel rack were in accordance with management's expectations and off normal procedure (ONP) 2.105, * Fuel Core Component Damage." i No equipment was damage Operational Status of Facilities and Equipment 0 Soent Fuel Pool Reserve Capacity Thc inspectors observed operations personnel and contractors during reconstitution of a spent fuel bundle. The contractor's procedures had been properly reviewed and approved by the licensee before the work activity was approved. In addition, extremely thorough pre job briefings were provided to the workers before the work started including detailed radiological conditions and contingency plans for unexpected occurrences. The work then proceeded carefully in a closely controlled manner. One spent fuel bundle was successfully disassembled and the fuel pins redistributed to open locations in other bundles, regaining the space necessary for a full-core off load, The licensee demonstrated good work control processes during the spent fuel bundle reconstitution activities, ll. Malntenance M1 Conduct of Maintenance M1.1 GeneraLCommenin a, jnspg.ption Segp1162703)(61726)

The inspectors observed all or portions of the following work activities:

Mainie_rtance Activitie_n e WO (Work Order): Reactor Depressurization Systein (RDS) train D isolation RDS 12612013 valve CV-4183 adjust packing

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• WO CAS 12710454: annualinspection, No.1 air compressor

• WO RDS 12710450: Instrument air filter to SV 4981

• WO SFP 12612163: tube leak on No. 2 SFP heat exchanger

• WO BLS 12611148: equipment lock, interior hydraulic unit only

• WO RDS-12710743: re torque and square up packing follower

• WO BLS 12710745: equipment lock outer door hydraulic system siWY91tlance Actlyitigj e T3014: core spray heat exchanger test

• T30 20: Uninterruptable power supply (UPS) C monthly station battery readings

• T90 07: RDS isolation valve test at power

• T90 09: core spray instrument trip test

• T9013: ASME Boller and Pressure Vessel Code Section XI CV lsolation Valve Testing Qbjervations __and Findirigs The inspectors venfied that the licensee was performing surveillances as schedule '

Also, the licensee performed surveillances that were scheduled to be performed during the system voltage alert and voltage emergency on June 24 26,1997. The delayed surveillance activities were completed as soon as the distribution system conditions allowed. All surveillances were completed within the specified periodicity plus the allowed grace period. The inspectors observed good communication techniques during the surveillance activities. The documentation for the maintenance and surveillance activities was generally accurate and up to-date, c. Conclusion Observed maintenance and surveillance activities were appropriately performed and accurately documente M1.2 Qqmineralized Wajer Syppjyjg_qontainment gtsprestion Sqppa The inspectors evaluated the licensee's corrective actions in response to the failure of a containment demineralized water supply valve, CV 4105, to meet valve timing requirements. The valve failed to meet the requirements specified in Surveillance T9013 and in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section X b, Qhietyations and Findinap On July 24,1997, the licensee notified the inspectors that CV 4105 had failed to meet the closing time requirement (closed in 17.9 seconds vice less than 10 seconds) during the performance of surveillance T9013. This surveillance is performed to verify valve stroke timing per ASME Section XI. After cycling CV 4105 twice (less than six seconds closing time during each cycle) for data, the shift supervisor (SS) declared CV 4105 inoperable J

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and directed the operators to leave the valve shut (the safety related position)in accordance with off normal procedure (ONP) 2.4," Loss of Containment Integrity. The licensee also wrote a condition report on the occurrence. The following day, the licensee i

systematically checked the condition of the air system supplying CV 4105, including filter i cleanliness, with no negative findings. After verifying that this was the first recorded instance of CV 4105 exceeding the closing time since 1988, the licensee performed the surveillance twice; recording a closing time of less than six seconds during both test The licensee evaluated the failure to meet the closing time requirements as a one time occurrence and declared the valve operable. The licensee also increased the i

survel!!anca testing interval to once each week. If the valve closing time is determined to meet or be less than closing time requirements during the next surveillance, then the surveillance interval will be extended to two weeks, and eventually back to a monthly test interva The inspectors verified that exceeding the closing time had no immediate safety consequences through a review of the Final Hazards Summary Repor1(FHSR) and Technical Specifications (TS). The demineralized water system provides normal makeup water to the emergency condenser; however, the safety related makeup sources come from the core spray or post incident systems. The demineralized water system would not be open to containment pressure unless the system was breached. Additionally, the normal pressure in the domineralized water system is approximately twice the containment pressure during an accident, therefore, leakage during a line break wouid flow into containment. The inspectors also verified that an adverse trend did not exist during previous CV 4105 closing times, Concigl The inspectors concluded that the licensee responded appropriately to assess the causal factors associated with CV 4105's failure to meet closing time acceptability requirement Careful assessment resulted in a good root cause evaluation and a conservative action plan to return the valve to operabilit M4 Maintenance Staff Knowledge and Performance M4.1 Eauipment Air Lock Door Hvdraulic System Overhaul Inspection Scone The inspectors observed maintenance personnel activities during the repair of the containment equipment air lock door hydraulic system as specified in work orders (WOs)

BLS 12611148 and 12710745, Observations and Findinos in response to the containment equipment air lock doors becoming slower and harder to operate, the licensee decided to replace the inner door hydraulic operating cylinders and hoses and clean the associated hydraulic reservoir and filters. The work proceeded well until, on July 23, an unexpected leak of pressurized hydraulic fluid occurred. The leak developed when the mechanics were removing a control valve for the external ramp (the ramp lowers into place with the equipment air lock door open to provide a track for the

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i l equipment cart). The mechanics were able to contain the leaking fluid but the extemal ramp slowly lowered untilit came in contact with the outer equipment air lock door. The l

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licensee wrote a condition report (C BRP 97 0398) to document the occurrence. The CR was inillally assigned a low significance level (three) and assigned to a work planner to develop lessons learne After evaluating the licensee's corrective action plans, the inspectors discussed the lowering of the external ramp with the plant manager. The inspectors pointed out that the leak was an unexpected occurrence during maintenance activities, and that a release of a

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pressurized fluid occurred. This type of event is classified as a near miss occurrence by i the licensee's procedures. If the extemal ramp had fallen rapidly,it could have caused serious personnelinjury. Also, a rapidly falling ramp had the potential to breach containment if it had ciruck the outer door hard enough to cause damage (the inner door was open for maintenance activities), Additionally, all of the participants (engineering, work planning, work control, operations, and maintenance personnel) involved in the job had failed to recognize the potential hazard caused by supporting the external ramp with hydraulic fluid isolated from its source. This event revealed a lack of questioning attitude ,

on the part of all parties involved in the work control proces The plant manager committed to returning the condition report to the plant review committee for reevaluation of the significance level. On July 24, the condition report was reevaluated as a significance level two and assigned to two engineers and a work planner

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to prepare a lessons learned presentation on the near miss occurrence. The presentation was presented at an all hands safety meeting on August 4,1997, to ensure that plant staff understood the contributing factors associated with this near miss occurrence prior to the final plant shutdown on August 29,1997, Qhuil90 The inspectors identified a lack of a questioning attitude on the part of plant personnel during the work control activities performed during maintenance efforts on the equipment-air lock hydraulic system that resulted in an unexpected release of hydraulle flui Ill. Enaineerina E2 Engineering Support of Facilities and Equipment E2.1 inspection. Identification. and Pluogina of No. 2 Spent Fuel Pool Heat Exchanag.Igbj klakg Inspectivn Scope The inspectors observed work activities, held discussion with plant personnel, and reviewed work packages and requests, in accordance with Manual Chapter 6270 These inspection activities were accomplished for eddy current testing activities performed on the No. 2 spent fuel pool (SFP) heat exchanger tubes for the purpose of identification of tubes with wall thinning, and the selection of tubes to be plugged to stop or prevent tube leakag M

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I Qbservations and Findinas July 20 August 2,1997, the No. 2 SFP heat exchanger was placed *out of service * to repair tube leakage as indicated by an increasing reactor cooling water (RCW) expansion tank level. The heat exchanger tubes had neither been inspected for leakage nor the walls evaluated for thinning since September 1990 when the tube bundles in the N and No. 2 SFP heat exchangers had been replaced (WOs SFP 073040 and 078351).

The SFP heat exchangers operate at 55 pilg with low. pressure chromated RCW on the shell side and SFP water on the tube side. The tubes were originally 0.049 inch thick aluminu Pressure testing of the tubes identified water leakage through three tubes. Contrary to management's expectations, the mechanics plugged the three leaking tubes before eddy current testing of the three leaking iubes in the No. 2 heat exchanger. Discussion with plant management revealed that it was their expectation that the tubes also would have eddy current inspection completed prior to plugging the leaking tube Eddy current testing identified wasting (flow erosion) on the shell side of 23 tubes, pitting on the inside diameter (the SFP side) of 12 tubes, and thinning on 35 tubes. The degraded tubes wore randomly located throughout the heat exchangers such that a pattern of tube wastage was not discernable. The criteria for identifying a degraded (or faulted) tube by eddy current inspection was a loss of greater than 20 percent of the tube wall. However, the licensee initially selected greater than 50 percent wallloss as the criteria for tube plugging. A total of 13 tubes weto plugged; 5 with pitting,5 with wastage, and 3 as a result of the pressure test. The licensee's work package Indicated that a maximum of 17 tubes could be plugged and this number was btJ ed on flow restriction considerations for the tube side of the heat exchanger and not on heat transfer limitations. Plugging too many tubes would result in excessive flow velocity. Plugging 13 of 246 tubes resulted In reduction of approximately 5 percent in the heat transfer capability. After reviewing Section 9.1.3 of the FHSR, the inspectors concluded that (the majority of the heat load in the SFP is from full core off load and that the heat transfer capability of the SFP heat exchanger was more than adequate with respect to handling the additional heat load) this 5 percent reduction in heat removal capability was not significant. The SFP cooling system was still capable of performing its functio On July 31,1997, during review of the work package for close out, a maintenance engineer determined that two tubes had been incorrectly plugged. The data, relating to the tubes to be plugged, had been incorrectly transferred twice, despite utilizing three engineers to verify its correctness. Tubes B4 and M4 were incorrectly translated as B1 and M1, resulting in good tubes 81 and M1 being plugged. Degraded tube B4, with 53 percent loss of wall thickness, and tube M4, with 57 percent loss of wall thickness, weto not plugged. A subsequent engineering calculation determined the minimum acceptable l wall thickness at 0,005 inches (about 10 percent of the design wall thickness). The licensee decided not to reopen the heat exchanger and correct the errors. Instead, the wall thickness acceptance criterion for plugging the tubes was changed to be greater than 60 percent wall thinning. Also, the licensee wrote a condition report to evaluate the human performance errors related to the transfer of data for SFP heat exchanger tubes B4 and M .

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The inscectors independently evaluated the integrity oI the SFP heat exchangers by reviewhg the requirements of the work package and the engineering calculation for the heat euchantpr tube reinimum wall thickness. The inspectors concluded that the integrity of' I4 . ' 5eal exchanger was not affected by degraded tubes B4 and M4 not

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%hn Ine 'strh s preliminary review of this event the inspectors determined that an eCm o.3 ptw sn did not exist for selecting the 50 percent thinning plugging

&*t a e# tion engineering personnel did not perform engineering calculations to ik V a whaf & iria should apply during evaluation of spent fuel pool heat exchanger G 4.nwa wn thickness. Subsequent to questioning by the inspectors, the licensee

~ eba to pv irm calculations to support the 50 percent thinning plugging criteria. The v., :vcm cr idered this inadequate engineering involvement in the work package pry ot W in . engineers should have delstmined the minimum wall thickness criteria bewc vne .iody current testing was performed. The licensee agreed with this ametre ent at the exit meetin IN Efsical maintenance activities involved with the eddy current inspection and

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F pging of the SFP heat exchanger tubes were performed in an organized, controlled nanner. Pre job briehngs and health physics (HP) briefings and practices were performed well, a Conclusion The SFP cooling system's function of maintaining the SFP within design temperature was not affected by plugging 13 tubes. However, engineering personnel performance was lacking ',n that tube plugging data was incorrectly transferred twice, despite three party verificttlon, resulting in plugging two good tubes and not plugging two degraded tube Also, engineering personnel could not provide calculations for minimum wall thickness criteria until questioned by the inspectors, E3 Engineering Procedures and Documentation

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E3.1 818.glor Depressurization System (RDS) Isolation Valve Inspection Scope The inspectors assessed the licensee's actions in response to a reactor depressurization system valve not meeting the acceptance criteria for CLOSING time during performance of T90-07,"RDS isolation Valve Test At Power." The assessment included reviewing facility change (FC) 555, "RDS Air Supply Filter," techn! cal specification (TS) 11.3.1.5, Reactor Depressurization System," Final Hazards Summary Report (FHSR) Chapter 6.9,

" Reactor Depressurization System," Administrative Procedure 3.1.1.1, " Facility Changes,"

and all other RDS and instrument air system facility, specification, and setpoint changes since 198 ObJttyeliQamd Find!O96 On July 17,1997, RDS "B" isolation valve (CV.4181) failed the closing time requirement of less than 9.0 seconds by closing in 9.31 seconds. A second (preconditioned) attempt

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resulted in a closing time of 6.91 seconds. However, both opening times were outside the acceptance limit of less than 4.0 seconds. The shift supervisor declared CV 4181 inoperable, entered a 7 day limiting condition for operations (LCO) under TS 11.3.1. and had a condition report (C BRP 97 0394) written to document the failed surveillance test. The condition report (CR) was assigned a high significance level two and given to an engineer to resolve and submit to management for approva On July 19, the licensee wrote another CR (C BRP 97 0397) to document that CV 4181 had again exceeded the closing time requirement. The licensee's investiga. On had revealed that the slower closing times had been caused by a filter cartridge being installed in S 3003B in the instrument air supply to CV 4181 on April 10,1997. An air leak on S 3003B (instrument air filter to solenoid valve (SV) 4981) had been repaired under WO RDS-12710450. Non safety related instrument air to close CV 4181 is controlled by SV 498 Discussions with the engineer assigned to resolve several condition reports revealed that filter tubes were missing for the instrument air to the RDS train solenold valves; and this fact had been discovered and documented in March 1996 on Condition Report C BRP 96 0412 " filter tube not installed in air supply filter (S 3003D) for RDS isolation valve CV 4183." The Condition Report had been given a low significance level and required only remediation and trending. Additionally, the description in C BRP 96-0412 reported that repairmen had noted that the RDS filters had not been installed in the pas The resolution to the Condition Report noted that facility change (FC) 555, implemented in August 1982, had installed an in line filter (S 33) upstream common to all the RDS isolation valve actuators. The filtering efficiency of S 33 was the same as that specified for the individual filters for each RDS solenoid control valve. However, since some dirt was observed in filter housing S 3003D, the recommendation in C BRP 96 0412 was to roinstall filter tubes in the individual RDS train filter housings as other work allowed. The resolution to C BRP 96-0412 was the impetus for the repairmen to install a filter element when repairing the air leak on solenoid S 3003B, supplying air to CV 418 The n.+ectors reviewed the facility change procedure, the design document checklist, the FHSR, the Technical Specifications, and all other facility changes, specification changes, and setpoint changes involving either RDS or instrument air. Neither the inspectors, nor the licensee, could find any reference documenting removing the filter elements from the individual RDS train filter housings (S3003A, B, C, & D). The inspectors' review of the closing times for the other RDS trains isolation valves indicated that no individual filters were installed as the times correlated to CV 4181's c osing time before the filter element was Installed in S 30038. The individual filter elements were probably removed when FC 555 was installed in 1982; however, configuration control was lost when their removal was not documented. The inspectors did not find any other instances of falling to document configuration change The inspectors concluded that leaving the filter elements out of S 3003, A through D, had no significant safety consequences as instrument air to close the RDS isolation valves is

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a non safety-related function. Additionally, the system operated with or without the Individual filter elements installed. The safety related function to open the RDS isolation valves is accomplished by spring pressure when instrument air is vented off the isolation valves.

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Technical specification 6.8.1 requires that written procedures be established, implemented, and maintained for all structures, systems, and components defined in the Big Rock Point Quality List. Administrative procedures were listed in the quality lia Administrative procedure 3.1.1.1, " Facility Changes," Revision 25, requires perfonning a

" Design Document Checklist" to ensure that all applicable documents were changed when a facility change was implemented. The same process and requirements existed in 1982 when FC 555 was implemented, but no documentation exists to suppori the removal of the individual filter elements. The failure to document the removal of the individual filler elements in 1982, or at some time between 1982 and the present, was a l

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violation of the requirements of TS 6.8.1. This licea.:,ee-ldentified and corrected violation is being treated as a Non Cited Violation, cont l stent with Section Vll.B.1 of the NRC

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Enforcement Policy (50155/g7009 01).

l l Although this was apparently an isolated failure to document removing system components, the loss of configuration controlls of concern to the NRC. In this particular case, the event did not result in significant safety consequences. However, a simitar l

occurrence in a safety related system could have serious consequences. Proper i configuration controlis mandatory at an operating plant. Configuration control can be more critical during dismantlement activities where disassembly of radioactive systems I

can have serious safety consequences if not properly controlle c, Conclusion The inspectors identified a non-cited violation for failing to document the removal of individual filter elements downstream from a common filter element in the instrument air syste IV. Plant Suonort R1 Radiological Protection and Chemistry Controls (71750)

R1,1 General Comments Using inspection Proceduret 71707 and 71750, the inspectors made frequent tours of the radiologically protected area (RPA) and discussed specific rMiological controls with the ALARA coordinator and various radiation protection (RP) technicians. The inspectors observed plant conditions and licensee performance including radiation protection oractices. In particular, good radiological performance was observed in support of the

. pent fuel module reconstitution and in support of the No. 2 SFP heat exchanger wor No contaminations or excessive personnel exposures were recorde S1 Conduct of Security and Safeguards Activities S1,1 General Commen[g During normal resident inspection activities, routine observations were conducted in the areas of .m turity and safeguards activities using Inspection Procedure 71750. No discrepancies were note l

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V. Manaaement Meetinos X1 F,KILMeglion_symmary After the conclusion of the inspection, the inspectors presented the inspection results to members of licensee management on August 6,1997. The licensee acknowledgod the findings presented.

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The licensee did not identify any of the documents or processes reviewed by the inspectors as proprietary.

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PARTIAL LIST OF PERSONS CONTACTED kk9_01 K. Powers, General Manager R. Addy, Plant Manager S. Beachum, Engineering Manager l

G. Boss, Operations Manager D. Hice, Maintenance Manager

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J. Rang, Decommissioning and Business Manager j K. Pallagi, Chemistry / Health Physics Manager W, Trubilowicz, Outage / Work Control Manager G. Withrow, Licensing Manager- .

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INSPECTlON PROCEDURES USED IP 37551: Engineering IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing -

Problems IP 61726: Surveillance Observations IP 62703: Maintenance Observation IP 64704: Fire Protection Program IP 71707: Plant Operations IP 71750: Plant Support Activities IP 73753: Inservice Inspection IP 83729: Occupational Exposure During Extended Outages IP 83750: Occupational Exposure IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92902: Followup - Engineering IP 92903: Followup Maintenance ITEMS OPENED and CLOSED Opened 155/97009 01 NCV Failure to Document Removing a System Component i QRE9A j 155/97009-01 NCV Failure to Document Removing a System Component

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LIST OF ACRONYMS USED l

l ALARA As Low As Reasonably Achievable ASME American Society of Mechanical Engineers CV Control Valve CR Condition Report DRP Division of Reactor Projects

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FHSR Final Hazards Summary Report HP Health Physics HX Heat Exchanger IP inspection Procedure IR Inspection Report LCO Limiting Condition for Operation NCV Non-Cited Violation-NRC Nuclear Regulatory Commission RCW Reactor Cooling Water RDS Reactor Depressurization System RP Radiation Protection -

RPA- Radiologically Protected Area RSRO Refueling Senior Reactor Operator SFP Spent Fuel Pool SS Shift Supervisor SV Solenoid Valve TS Technical Specification UPS Uninterruptable Power Supply a

WO Work Order

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