ML17292B531
| ML17292B531 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/29/1998 |
| From: | Laura Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Parrish J WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 9901120038 | |
| Download: ML17292B531 (37) | |
Text
CATEGORY 2 REGULATORY INFORMATXON DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9901120038 DOC.DATE: 98/12/29 NOTARIZED: NO DOCKET I FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH.NAME AUTHOR AFFILIATION SMITH,L.J.
Region 4 (Post 820201)
RECIP.NAME RECXPIENT AFFILIATION PARRISH,J.V.
Washington Public Power Supply System
SUBJECT:
Advises of planned insp effort resulting from WNP-2 Znsp Planning Review held on 98120'2.Historical listing of plant issues, general description of PIM table labels
& Xnsp plan for next eight months encl; DISTRIBUTION CODE:
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'+o UNITED STATES NUCLEAR REGULATORY COMMISSION REGION IV 611 RYAN PLAZADRIVE. SUITE 400 ARLINGTON,TEXAS 76011.8064 DEC 29 lo98 Mr. J. V. Parrish (Mail Drop 1023)
Chief Executive Officer Washington Public PoSer Supply System P.O. Box 968 Richland, Washington 99352-0968
SUBJECT:
INSPECTION PLANNING REVIEW (IPR) - WASHINGTON NUCLEAR PROJECT, UNIT2 (WNP-2)
Dear Mr. Parrish:
On December 2, 1998, the NRC staff completed a unique Inspection Planning Review (IPR) of WNP-2. The staff normally conducts Semiannual Plant Performance Reviews for all operating nuclear power plants to develop an integrated understanding of safety performance and adjust inspection resources.
However, because of the suspension of the Systematic Assessment of Licensee Performance process, we implemented an abbreviated IPR for plant issues and to develop inspection plans. The IPR for WNP-2 involved the participation of both the Reactor Projects and the Reactor Safety divisions in evaluating inspection results and safety performance trends for the period April23 to October 28, 1998.
Based on the results of this review, inspection resources have been prioritized and scheduled as listed in the inspection plan. The inspection resources for review of your radiological controls program implementation were increased by one inspection-week of direct inspection effort.
Enclosure 1 contains an historical listing of plant issues, referred to as the Plant Issues Matrix (PIM), that was considered during this IPR process to arrive at an integrated view of licensee performance trends.
The PIM includes only items from inspection reports or other docketed correspondence between the NRC and Washington Public Power Supply System.
The IPR may also have considered some predecisional and draft material that does not appear in the attached PIM, including observations from events and inspections that had occurred since the last NRC inspection report was issued, but had not yet received full review and consideration. is a general description of the PIM table labels. This material will be placed in the NRC Public Document Room.
This letter also advises you of our planned inspection effort resulting from the WNP-2 IPR review. It is provided to minimize the resource impact on your staff and to allow for scheduling conflicts and personnel availability to be resolved in advance of inspector arrival onsite.
. Enclosure 3 details our inspection plan for the next 8 months. The rationale or basis for each inspection outside the core inspection program is provided so that you are aware of the reason for emphasis in these program areas.
Resident inspections are not listed because of their ongoing and continuous nature. We willinform you of any changes to the inspection plan.
990ii20038 98i229 PDR ADQCIl'5000397 9
Washington Public Power Supply System If you have any questions, please contact me at 817-860-8137.
Sincerely, Docket No.:
50-397 License No.:
Enclosures:
- 1. Plant Issues Matrix
- 2. General Description of PIM Table Labels
- 3. Inspection Plan cc w/enclosures:
Chairman Energy Facility Site Evaluation Council P.O. Box 43172 Olympia, Washington 98504-3172 Mr. Rodney L. Webring (Mail Drop PE08)
Vice President, Operations Support/PIO Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Mr. Greg O. Smith (Mail Drop 927M)
WNP-2 Plant General Manager Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Mr. D. W. Coleman (Mail Drop PE20)
Manager, Regulatory Affairs Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Mr. Albert E. Mouncer (Mail Drop 396)
Chief Counsel Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 L.. Smith, Ac ing hief Project Branch E Division of Reactor Projects
Washington Public Power Supply System
- Mr. Paul Inserra (Mail Drop PE20)
Manager, Licensing Washington Public Power Supply System P.O. Box 968
'Richland, Washington 99352-0968 Perry D. Robinson, Esq.
Winston &Strawn 1400 L Street, NW.
Washington, D.C. 20005-3502
Washington Public Power Supply System DEC 29 j998 E-Mail report to T. Frye (TJF)
E-Mail report to D. Lange (DJL)
E-Mail report to NRR Event Tracking System (IPAS)
E-Mail report to Document Control Desk (DOCDESK)
E-Mail report to Richard Correia (RPC)
E-Mail report to Frank Talbot (FXT) bcc to DCD (IE01) bcc distrib. by RIV:
Regional Administrator DRP Director Branch Chief (DRP/E, WCFO)
Senior Project Inspector (DRP/E, WCFO)
Branch Chief (DRP/TSS)
The Chairman (MS: 16-G-15)
Deputy Regional Administrator Commissioner Dicus Commissioner Diaz Commissioner McGaffigan Commissioner Merrifield W. D. Travers, EDO (MS: 17-G-21)
Associate Dir. for Projects, NRR Associate Dir. for Insp., and Tech. Assmt, NRR SALP Program Manager, NRR/ILPB (2 copies)
W. Bateman, NRR Project Director (MS: 13-E-17)
C. Poslusny, NRR Project Manager (MS: 13-E-16)
Resident Inspector DRS;PSB MIS System RIV File Carol Gordon Records Center, INPO C. A. Hackney B. Henderson, PAO B. Murray, DRS/PSB SRls at all RIVsites DOCUMENT NAME: G:ttDRPDIR PR'tWNP To receive co of docunten Indicate In xt "C" ~
without enctosures E" ~
with encl ures 'N ~ No co RIV:AC:DRP/E E.
GAPick;df 12/17/98
'l 2-'I AS D:DR ATHo ell 12/
8 DD:DRP KEB man D:DRP TPG nn 12/
/98
/98 OFFICIALRECORD COPY
Washington Public Power Supply System SEC 29 isa E-Mail report to T. Frye (TJF)
E-Mail report to D. Lange (DJL)
E-Mail report to NRR Event Tracking System (IPAS)
E-Mail report to Document Control Desk (DOCDESK)
E-Mail report to Richard Correia (RPC)
E-Mail report to Frank Talbot (FXT)
,-bcc~to. DCD (IE01)" --
'cc distrib. by RIV:
Regional Administrator DRP Director Branch Chief (DRP/E, WCFO)
Senior Project Inspector (DRP/E, WCFO)
Branch Chief (DRP/TSS)
The Chairman (MS: 16-G-15)
Deputy Regional Administrator Commissioner Dicus Commissioner Diaz Commissioner McGaffigan Commissioner Merrifield W. D. Travers, EDO (MS: 17-G-21)
Associate Dir. for Projects, NRR Associate Dir. for Insp., and Tech. Assmt, NRR SALP Program Manager, NRR/ILPB (2 copies)
W. Bateman, NRR Project Director (MS: 13-E-17)
C. Poslusny, NRR Project Manager (MS: 13-E-16)
Resident Inspector DRS-PSB MIS System RIV File Carol Gordon Records Center, INPO C. A. Hackney B. Henderson, PAO B. Murray, DRS/PSB SRls at all RIV sites DOCUMENT NAME: G:>DRPDIR IPRtWNP To receive co of document, Indicate In x: C" ~ Co without enclosures "E ~
with encl ures N ~ No co RIV:AC'DRP/E'.
GAPick;df GS 12/17/98 I2 2-( rt>
D:DR ATHo ell 12/
8 D:DRP TPG nn 12/
/98 DD:DRP KEB man
/98 OFFICIAL RECORD COPY
PLANTI S ES MATRIX ENCLOSURE 1 DATE TYPE SOURCE ID SFA TEMPLATE CODES 10/10/98 NEG
'IR 98-21 NRC OPS 1A 3A 09/28/98 POS IR 98-11 NRC OPS 5A 5C 09/28/98 NCV IR 98-11 NRC OPS 5A 09/17/98 NEG IR 98-20 SELF OPS 3A 09/17/98 POS IR 98-20 NRC OPS 1B 09/17/98 NCV IR 98-20 LIC OPS 1B 3A ITEM DESCRIPTION During a walkdown of the service water. supply to the System A residual heat removal pump room cooler, the inspectors determined that the locking device to an isolation valve was unattached, contrary to plant procedures.
Since the valve was found in the correct position, no safety impact would have resulted. This failure to properly lock the valve is a violation of minor significance and h not subject to formal enforcement action Operations personnel were effective in the identification and resolution of conditions adverse quality.
The gold card program was useful to identify human performance Issues.
However, it was occasionally used to improperly include procedural violations and equipment issues.
This was dur to a combination of factors, which included a lack of questioning attitude and minimal managemer Involvement. The failure to initiate performance evaluation requests (two examples) circumventec the corrective action program and was identified as a non-cited violation, pursuant to Section VII.B.1 of the NRC Enforcement Policy, of Procedure PPM'1.3.12.
As a result of human error, the watertight door between the reactor building northeast stairwell an<
reskfual heat removal pump Room C was left open prior to the flooding event. The open door resulted in substantial flooding of Room C, rendering Residual Heat Removal C inoperable and complicating operator recovery from the event.
The licensee responded well to the flooding event. The shift manager made an appropriate decision to declare an Unusual Event and activate the onsite emergency response organization to quickly bring resources to bear on an unusual and complex event.
Declaration and notification of the emergency were both timely. Actions to stop the flooding and dewater the reactor building were prompt and effective.
The Technical Support Center manager failed to confer with the emergency director prior to authorizing the discharge of the stairwell floodwater to the storm drains. The error was the result o the improper placement of an emergency response requirement into an operations procedure Instead of the emergency plan implementing procedures.
The corrective actions taken to address this deficiency and evaluate the generic implications were appropriate.
A noncited violation of Technical Specification 5.4.1.a was identified for failure to followprocedure, in accordance with Section VII.B.1 of the Enforcement Policy.
October 28, 1998 WNP-2
PLANT ISS ES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 09/17/98 NEG IR 98-20 NRC OPS 1B 3A 07/18/98 POS IR 98-13 NRC OPS 1A 2B 07/07/98 POS IR 98-13 NRC OPS 1A 2B 07/02/98 POS IR 98-09 NRC OPS 1A 1C 06/07/98 VIO IR 98-13 NRC OPS 1A 3A SL IV 04/25/98 POS IR 98-06 NRC OPS 1A 04/25/98 VIO IR 98-06 NRC OPS 1A 5A SL IV ITEM DESCRIPTION The actions of the operators to start the low pressure core spray pump during the flooding event.
while in compliance with the wording of plant procedures, did not display conservative decision making. Although the actions were an attempt to maintain the maximum number of operable/available emergency core cooling system pumps, the operators failed to recognize that other potential effects could have occurred because of the flooding.
The licensee was well prepared for plant restart from the 1998 refueling outage as evidence~~
proper closure of outage activities, completion of required Technical Specification surveillance~
and adequate configuration of plant systems to support power operation.
This was improved performance over previous refueling outages.
Control room operators took appropriate steps to limitoutside interference and maintain control of the plant during the performance of postmaintenance testing on the reactor feedwater pumps following modifications to their associated hydraulic control system.
Effective command and control and three-way communication were observed.
The routine shutdown for Refueling Outage R13 was properly executed with a detailed preevolutio brief and good command and control.
Operations performance in monitoring and controlling the cooldown was improved over that observed during the March 1998 forced outage.
Poor procedure use during the restoration from an inadvertent engineered safety feature actuatior resulted in the mtspositioning of the minimum flowbypass valve for the low pressure core spray system.
Numerous control board walkdowns performed by operators failed to identify the discrepancy.
Aviolation of Technical Specification 5.4.1.a was identified for failure to follow procedure when returning the low pressure core spray system to its standby lineup.
Good command and control of the March 18 reactor startup and April3 feedwater temperatur~
reduction was evidenced by adequate planning, proper assignment of personnel responsibllit~
clear communications, and a conservative approach to implementing the activities.
Control of plant equipment was generally effective in maintaining proper plant configuration.
However, two examples were identified whe're a lack of understanding of the impact of plant configuration changes resulted in the failure to identify discrepancies between the configuration changes and the plant's licensing bases.
Specifically, operators failed to 1) recognize that inoperable drain valves for the service water spray rings were required by the final safety analysis report for freeze protection, and 2) recognize entry into a Technical Specification action statement when the emergency cooling coils for control room air conditioning, Train A, were isolated for planned maintenance. The second example was identified as a violation of Technical Specificatior 5.4.1.aforfailureto ro e im Iementwritten roceduresforcontrolof lante ui ment October 28, 1998 WNP-2
PLANT IS ES ATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPT,ION 03/17/98 VIO IR 98-05 SELF OPS 1C 4C SL IV 03/1//98 VIO IR 98-05 NRC OPS 1B 1C SL IV 03/17/98 WK IR 98-05 NRC OPS 1C 03/17/98 NEG IR 98-05 SELF OPS 1B 3B 4B 03/17/98 NOV IR 98-05 LIC OPS 1B 1C SL IV A violation of Technical Specification 5.4.1a and Regulatory Guide 1.33, with two examples of inadequate procedures, was kfentified for a Division II logic system functional test and the DNlslon illemergency diesel generator restoration.
Temporary Change Notice TCN 98-113, made to Procedure TSP.DG2/LOCA-B501, Step 7.1.33, Substep a, to override the opening of the injection valve, was Inadequate and resulted in low pressure coolant injection to the reactor vesse during the conduct of the March 12, 1998, logic system functional test.
Procedure PPM 2.7.3, High Pressure Core Spray Diesel, Revision 29, did not provide adequate direction for the shutdown of the high pressure core spray system.
Aviolation was kfentified for the failure to provide the one hour event notification in accordance with 10 CFR 50.72, paragraph (b)(1)(iv) for the valid high pressure coolant injection into the reactc vessel.
The licensee's 10 CFR 55.59, Ucensed Operator Requalification Program, did not address the make up of crew complement used in simulator training vs the control room and was considered e
significant weakness In the licensed operator requalification training program.
Although the licensee's actions prior to the main steam line isolation valve nitrogen supply line failure and overall response to the complex transient were appropriate, weaknesses with operator:
knowledge, skills and abilities were identified Involving recognition of the plant response, verifying the appropriate engineered safety feature and emergency core cooling systems actuations.
Management oversight of the control room actions was not well focused on evolving plant conditions and assuring recovery actions were appropriately implemented.
Effective managemenl control was not implemented for the procedure temporary change process and control of infrequently performed tests and surveillance.
Operator workarounds appeared in significant area involvingvessel level and pressure control, temperature monitoring and forced circulation.
Communication within the control room and with the NRC headquarters operations officerwa and did not ensure that key control room personnel were cognizant of the overall plant and systems.
A violation of Technical Specmcatlon5.4.1a and Regulatory Guide 1.33 was identified for the failure to maintain the reactor vessel temperature and upper head pressure indications within the acceptable area of the temperature/pressure curve provkfed in Procedure OSP-RCS-C102, 'RPV Vessel Cooldown Surveillance, Revision 0, Attachment 9.1, "MinimumVessel Metal Temperature VS Reactor Vessel Pressure.
October 28, 1998 WNP-2
PLANT ISSUE lttIATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 03/17/98 NEG IR 98-05 NRC OPS 5A 5B 5C The initial event review was not fullyeffective in providing a comprehensive understanding of equipment problems, procedural weaknesses and operator performance issues.
The plant restart evaluation process was needed to fullyidentify the Issues that were missed by the post scram review. This resulted In an iterative approach to Identify, analyze and resolve each of the performance issues.
03/14/98 NEG IR 98-03 NRC OPS 1A SA One instance was identified in which an operating crew did not demonstrate a
conservative'pproach to equipment operation when a non-vital lighting panel, with an unidentified ground, reenergized without an understanding of the source of the ground or a troubleshooting plan to identify the source.
03/14/98 NCV IR 98-03 SELF OPS 1A 02/19/98 NEG IR 97.20 NRC OPS 1A 1C 02/19/98 STR IR 97-20 NRC OPS 1A 3B 02/19/98 NCV IR 97-20 SELF OPS 3A LER 96-002 5B SC Inadequate self-checking and peer checking resulted in an operator error that deenergized non-vital Bus SM-2 and started the Division IIIemergency diesel generator.
Operations personnel actions in response to the transient were appropriate and prompt. The licensee's root cause analysts and corrective actions effectively addressed the human performance concerns.
The licensee's program to assure that corrective lenses for selt contained breathing apparatus (SCBA) for operators requiring them was implemented successfully. However, procedural guidanc for maintenance of the SCBA corrective'lens program was considered weak, in that periodic inventories were not required and written expectations were not provided to operators on the neec to have SCBA qualified lenses, regardless of the type of corrective lenses normally used.
The professionalism of the control room operators and shift management ownership of crew activities supported good operational performance over the inspection period. Operators were genera'lly knowledgeable of plant and equipment status with several minor exceptions.
A personnel error on the part of an equipment operator during the performance of clearance o~
activities resulted in the momentary deenergization of the Division II4160V vital bus and the I residual heat removal assist cooling of the spent fuel pool. A noncited violation was identified associated with this 1996 licensee event re rt.
October 28, 1998 NNP-2
PLANT ISS ES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 02/09/98 VIO IR 97-13 NRC OPS 5A 5C SL IV 02/09/98 POS IR 97-13 LIC OPS 3B 3A 02/09/98 NCV IR 97-13 NRC OPS 4C 4B 02/09/98 POS IR 97-13 LIC OPS 5C 01/15/98 VIO IR 97-18 NRC OPS
- 2A 1C SL IV 11/08/97 STR IR 97-17 NRC OPS 1A 1C 3B 09/28/98 POS IR 98-11 NRC MAINT SA 5C While corrective actions to resolve the material buildup problem in Valves FDR V-3 and FOR V.4 were effective, corrective actions to resolve a required reading problem were not. Violation 50-397/9611-04 willbe dosed, however, an example of a new violation of 10 CFR Part 50, Appendix B, Criterion XVI,was identified for the failure to correct the required reading issue.
The corrective actions to resolve continuing failures of the motor-to-pump coupling on the ac standby lubricating oil pump were inadequate.
This inadequacy was considered to be an example of a violation of 10 CFR Part 50, Appendix B, Criterion XVI.
There was a failure to Issue a problem evaluation request that would have promptly identified provided corrective actions for the inadvertent start of a reactor recirculation pump. This item was considered to be an example of a violation of 10 CFR Part 50, Appendix B, Criterion XVI.
Actions to address the occurrence of shorting electrical terminals during the performance of maintenance or surveillance activities were adequate and effective toward preventing a recurrenci of the events.
The new nuclear safety assurance division procedure properly addressed the technical specification procedural requirements.
In addition, licensee conducted surveillances were effectiv in assuring that other canceled procedure activities were properly. conducted.
However, there wa.
a failure to update the Final Safety Analysis Report fire protection sections.
The corrective actions that addressed the inadvertent initiation of drywell to suppression chamber bypass flowwere appropriate for the circumstances and adequate to prevent a recurrence of the events.
A number of inspector identified deficiencies in the control of transient equipment indicated weak implementation of the licensee's program to prevent seismic interactions between the equipment and safety-related components.
Three examples of a violation of plant procedures were ident'anagement involvement ln the plant curtailment for maintenance on the reactor feedwater dr turbines (RFWDT) was notable for reemphasizing expectathns and raising personnel sensitivity tc a slgnNicant evolution. The operations staff also demonstrated conservative decision-making whe maintenance on the first drive turbine was delayed while operability concerns with the high pressure core spray (HPCS) system were addressed.
Maintenance personnel were effective In the Identification and resolution of conditions adverse to quality. The work control process was properly implemented with respect to the corrective actions program.
October 28, 1998 WNP-2
PLANT I SUES NIATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 09/17/98 VIO IR 98-20 NRC MAINT 2A 2B SL IV EA 98-452 08/29/98 NEG IR 98-19 NRC MAINT 2A 5A 08/29/98 POS IR 98-19 NRC MAINT 2B 3B 07/18/98 POS IR 98-13 NRC MAINT 2B 5A 07/02/98 NEG IR 98-09 LIC MAINT 1A 3A ITEM DESCRIPTION The licensee. failed to assign a level of importance to the emergency core cooling system pump room floor drain cross-connect valves that was commensurate with their design function. As a result, the maintenance and surveillance program for ensuring their reliability, when called upon to perform that function, was Inadequate as evidenced by the failure of Valve FDR-V-609, residual heat removal pump Room C and low pressure core spray pump room floor drain cross-connect, during the flooding event. The failure of Valve FDR-V-609 to perform its intended function resultei in the flooding of the low pressure core spray pump room and complicated recovery from the~
transient. The failure to monitor the performance of the valves against established goals ort~
demonstrate reliabilityof the valves through an effective preventive maintenance program was identified as a vhlation of 10 CFR 50.65 (EA 98-452).
Material condition deficiencies were identified in the Division I 125V DC battery (low electrolyte level) and emergency diesel generator starting air system (multiple air leaks). Although neither condithn, by itself, rendered a safety-related system or component inoperable, both conditions ha the potential to adversely affect equipment performance.
The processes for identifying these conditions adverse to quality, including operator rounds, system engineer walkdowns, and surveillances, were ineffective In these instances.
The planning and Implementation of the repair of a reactor recirculation system instrument sensin line socket weld were thorough and generally well executed.
The repair plan and mockup were notable strengths.
Some minor deficiencies were identified during execution of the repair.
The licensee's actions were comprehensive in identifying and inspecting equipment in the emergency core cooling system pump rooms that was affected by the June 17 flooding event.
Efforts to dry equipment and conduct calibrations and functional tests were sufficient to verify operability. However, walkdown inspections of the fire protection system were weak in that subsequent to the walkdowns the inspectors identified ten failed system pressure gauges an loose pipe hanger on the standby gas treatment system deluge supply piping Both the reactor disassembly and the fuel shuffle were generally well executed between the contr<
room and the refueling floor. However, two instances of weak procedure use resulted in: 1) the failure to identify an Incorrect precaution ln the maintenance procedure for the reactor building overhead crane, and 2) failure to verifythat appropriate minimum temperature requirements were bein met rior to lifti the d ell u er shield blocks.
October 28, 1998 WNP-2
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 0?/02/98 VIO IR 98-09 NRC MAINT 1A SL IV 2B 3A Personnel performance in the conduct of testing excess flowcheck valves was inadequate, as evidenced by: 1) multiple examples of poor procedural adherence and procedure adequacy, 2) personnel knowledge'deficiencies on testing requirements and plant impact, 3) weak use of procedures in the field, and 4) weak command and control. In one case, performance deficiencie:
resulted in the initiation of an engineered safety features actuation signal and plant transient.
Two vhlations of TS 5.4.1.a, each with two examples, were identified regarding adequacy and use of surveillance procedures.
The violations included inadequate procedure guidance forestabiis~
and restoring from test conditions and failure to independently verifya valve location prior to~'
manipulation.
07/02/98 POS IR 98-09 NRC MAINT 3A SC 07/02/98 NEG IR 98-09 LIC MAINT 1A 3A The licensee's actions to address previously-identified weaknesses in implementing their foreign.
material controls (FMC) program for plant systems and containment have been effective in raising the sensitivity and Improving performance of plant personnel.
Ucensee performance in implementing FMC for the spent fuel pool, reactor cavity and reactor pressure vessel (RPV) was mixed. Weaknesses were identified mainly in the administrative controls of foreign materials. These included the failure to perform inventories of the spent fuel and equipment pooh prior to removal of the RPV head.
The failure to perform the inventories eliminated an objective measure of the effectiveness of FMC and was identified as a noncited violation of Plant Procedure 6.1.1.
07/02/98 NEG IR 98-09 NRC MAINT 2A 06/15/98 NEG IR 98-13 LIC MAINT 3A LER 98-010 5B The Division I emergency diesel generator (EDG) experienced multiple material deficiencies durin Refueling Outage R13 which resulted in several failures to run and/or load. The material deficiencies included: (1) the failure of the mechanical governor's motor operated potentiometer, (2) failure of the lube oil low pressure switch to reset, and (3) failure of the diesel generator output.
breaker to close due to improper setting of the breaker's trip latch check switch. The licensee's short-term corrective actions for the failures were appropriate.
Long-term actions willbe rev'n future inspection activities.
A cognitive error on the part of maintenance personnel installing the traversing incore probe instrument tubing resulted in the separation of the undervessel connection on one of the 41 tubes.
Consequently, the drive cable for one of the probes became mechanically bound when it was Inadvertently spooled into the undervessel area during a system alignment. The failure of the driv cable precluded the abilityto close its associated containment Isolation ball valve and necessitate<
a lant shutdown in accordance with Technical S ecificatlons.
October 28, 1998 WNP-2
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 04/25/98 WK IR 98-06 NRC MAINT 2A 2B 03/14/98 VIO IR 98-03 NRC MAINT 2B 3B SL IV 03/14/98 NEG IR 98-03 LIC MAINT 2A 02/19/98 NCV IR 97-20 SELF MAINT 3A LER 96-001 02/19/98 POS IR 97-20 NRC MAINT 3A 4B 01/15/98 NCV IR 97-18 NRC MAINT 2B 01/15/98 WK IR 97-18 NRC MAINT 2B 2A 11/08/97 STR IR 97-17 NRC MAINT 4B 4C 4A Although overall plant material condition remained good, the inspectors continued to find material condition deficiencies that had not been previously identified and tracked for resolution by the licensee.
Deficiencies included leakage from components outside containment that could contain highly radioactive fluid followinga loss-of-coolant accident, and a locked spring hanger on the residual heat removal system's minimum flow bypass line.
Ucensee personnel improperly applied surveillance requirement 3.0.2 to program surveillanc~
the administrative section of Technical SpeciTications.
As a result, a 25 percent surveillance.~
interval extension was inappropriately utilized for several technical programs.
Poor material condition of the plant service water system resulted in a leak that challenged the Integrity of the control room envelope as water was able to penetrate through a concrete slab interface in the control room ceiling, a boundary credited by the licensee's flooding analysis.
The licensee is currently implementing an improvement plan that should adequately address the material condition deficiencies In the plant service water system.
The failure of maintenance personnel to read and adhere to the instructions on a caution tag prior to manipulating a breaker, resulted in the loss of the Division I 125VDC critical instrument power inverter and the Initiation of several essential safety features and isolation of several containment isolation valves. The event occurred while the plant was defueled in Mode 5. A noncited violation was identiTied associated with this 1996 licensee event report.
Observed maintenance and surveillance activities were generally well coordinated and executed with appropriate craft supervision and system engineering participation.
The methodology utilized by the licensee for testing the control room emergency charcoal filters was identified as being from a different, more recent version of the standard specified inTec~
SpecTiicatlons (TS). Based, in part, upon the staffs acceptance of the version of the standar~.
utilized by the licensee, and the more conservative results produced by its methodology, the noncompliance was viewed as a minor violation.
The licensee's material condition inspection program was not fullyimplemented to maintain and assess those areas of the reactor building not routinely accessed by plant personnel.
As a result, lower standard was established for these areas and equipment and housekeeping deficiencies were allowed to persist.
The licensee's troubleshooting and repair efforts associated with the RFWDTs were well planned and executed.
The efforts resulted in Improved drive turbine performance while Identifying tentlal desi nim rovements to the turbine overnor control oils tern October 28, 1998 WNP-2
PLANT I SUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 10/10/98 NEG IR 98-21 NRC ENG 4C IR 98-05 EA 98-203 ITEM DESCRIPTION The manual startup and shutdown of the reactor core isolation cooling system for level control, followingthe March 1998 main steam Isolation valve closure, challenged the operators.
The proceduralized method to control reactor vessel level by diverting reactor core Isolation cooling flo through the test return line coukl not be accomplished because of valve design deficiencies. The method used to maintain the reactor core Isolation cooling system test return line isolation valves decreased the reliabilityof the system and challenged the containment isolation function since the valves may not have closed against high dNerential pressure.
Unresolved Item 50-397/9800 involving exclusion of the reactor core isolation cooling test return line valves from the scope maintenance rule, was determined not to be a violation of NRC requirements (EA 98-203) 09/28/98 NCV IR 98-11 NRC ENG 4B 5A 09/17/98 EEI IR 98.20 SELF ENG 1C 4A EA 98-480 09/17/98 NCV IR 98-20 LIC ENG 4A 5C Engineering personnel were not always effective in the resolution of conditions adverse to quality.
The engineering personnel's performance was indicative of a lack of attention to detaiL This was evidenced by the nonwited violation of Criterion XVIto Appendix B of 10 CFR Part 50, pursuant tc Section VII.B.1 of the NRC Enforcement Policy, for the untimely implementation of corrective actions for a condition adverse to quality. In addition, the license actions associated with the potential bypass of primary containment and water hammer evaluations indicated a lack of a questioning attitude.
The root cause evaluation for the flooding event accurately concluded that the event resulted from design inadequacies of the fire protection water supply system.
Those inadequacies allowed for the generation of destructive forces within the system that ultimately failed Valve FP-V-29D, react<
building fire protection standpipe isolation. The design inadequacies were attributed, in part, to noncompliances ref~ted to the installation of the fire pumps compared to the requirements of the National Fire Protection Association code. The failure of the fire protection system pressure boundary upon a demand actuation would preclude the abilityof the system to provide an adequate capacity of water to suppress a postulated fire and was identified as an apparent vhiation of 10 CFR 50, Appendix A, General Design Criterion 3, 'Fire Protection" (EA 98-48 The discrepancy between the actual performance of the reactor building watertight doors and theii description in the Final Safety Analysis Report as being watertight was previously identified and analyzed by the licensee.
Although the analyses were found to be technically sound In concluding that the doors could continue to perform their function with the amount of leakage predicted, they did not result in appropriate changes to the FSAR. The licensee identified this discrepancy during its followup to the fkxxflngevent and initiated appropriate action to address it. A noncited violatior of 10 CFR 50.71(e) was kfentified for failure to update Final Safety Analysis Report, in accordance with Section VII.B.1 of the Enforcement Policy.
October 28, 1998 WNP-2
0
PLANT ISS ES ATRIX DATE TYPE 08/29/98 VIO SOURCE ID SFA TEMPLATE CODES IR 98-19 NRC ENG 2B 4B 08/29/98 NCV IR 98-19 LIC ENG 2B 4B 07/18/98 POS IR 98-13 NRC ENG 4A 4C 07/18/98 VIO IR 98-13 NRC ENG 4A 5A SL IV 07/18/98 NEG IR 98-13 NRC ENG 4A 5C ITEM DESCRIPTION The postmaintenance and operability testing of the Division II emergency diesel generator were found to be thorough in assuring that the identified deficiencies were corrected.
However, the evaluation of the operability test procedure failed to identify that Technical Specifications prohibite>>
the performance of portions of the procedure during plant operations.
The failure of licensee personnel to properly review Technical Specifications during procedure development and approva was identTiied as a violation of the requirements of 10 CFR 50.59.
The instructions established for troubleshooting the Division II emergency diesel generator fai~
identify the inherent rfsk of loading the inoperable diesel generator onto its associated vitalbus~
and, as such, failed to Include appropriate contingencies and precautions.
As a result, operators did not have sufficient guidance to protect the vital bus when the voltage regulator failed and the bus deenergized on a timed overcurrent lockout. The corrective actions in response to this event were appropriate.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified for inadequate Instructions during troubleshooting.
The licensee has maintained an appropriate program to address the requirements of 10 CFR 50.59. Program implementing procedures were generally of sufficient detail to ensure that proposed activities would precipitate safety evaluations.
However, two areas were noted where procedure guidance was either weak or inconsistent with requirements.
Although the quality of th~
12 safety evaluations reviewed was not always consistent, overall the quality was good. Strength:
were noted in the training and oversight programs with regards to maintaining a sufficient pool of
. qualified safety evaluation preparers and providing timely, critical feedback on their products.
Compensatory and corrective actions taken to address design deficiencies in the fire protection system and minimize dynamic loads were generally appropriate.
However, the licensee's evaluation of the modified system's performance failed to Identify a vulnerability to water hammer following a loss of offslte power. The vulnerability was adequately addressed when the syste configuration was modified to maintain a diesel driven fire water pump operating.
The configuration of the reactor building equipment drains did not conform to the description in thr.
Final Safety Analysh Report in that a cap was not installed on the drain line from residual heat removal pump Room B. The cap was required as part of the licensee's physical controts to proter against common mode flooding. Aviolation of 10 CFR 50.59 was Identified for failure to documen a written safety evaluation for this defacto change to the facility. The licensee's corrective actions to Install a cap on the drain line and review the generic implications for other portions of the drain tems werefoundtobea ro riate.
October 28, 1998 10 WNP-2
PLANTI S ES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 04/25/98 VIO IR 98-06 NRC ENG 4A 4B SL IV 04/25/98 NEG IR 98-06 NRC ENG 4B 04/25/98 VIO IR 98-06 NRC ENG 4A 5B SL IV 04/22/98 WK IR 98-01 NRC ENG 4B 4C 04/22/98 NEG IR 98.01 NRC ENG 4B 4C 04/22/98 NEG IR 98-01 NRC ENG 4B 4C ITEM DESCRIPTION The licensee failed to provkfe adequate controls for the position of the residual heat removal system's suppression pool return valves to ensure that during operations with the return valves open, the valves'osition would be limited to meet the injection times assumed in the loss-of-coolant-accident analyses.
Specifically, with the return valves greater than approximately 40 percent open, fulllow pressure coolant injection flowto the vessel would not be achieved within th 66 seconds assumed by the analyses.
The failure to adequately translate the design requiremenb of the RHR system to plant operating procedures and instructions was identified as a violation 10 CFR Part 50, Appendix B, Criterion III (Design Control).
Following the identification of leakage from the safety-related nitrogen supply to the automatic depressurization system, engineering personnel established an adequate technical basis for system operability. However, the technical justification was not appropriately documented in the associated problem evaluation request.
Additionally, the problem evaluation request was closed without addressing the root cause of the degraded condition.
The licensee failed to recognize that operation of the residual heat removal system, with the minimum flowbypass valves closed in standby, constituted a change to the facilityas described in the Final Safety Analysis Report (FSAR) in that the original FSAR showed the valves to be open.
The licensee missed several opportunities to identify the need for a written safety evaluation to support the change. These included the development of original system operating procedures, a 1993 revision to the FSAR that changed the valves'osition on the process data sheet, and the licensee's current FSAR upgrade project. A violation of 10 CFR 50.59(b)(1) was identified.
The licensee operated the Siemens Power Corporation's fuel in Core Cycles 7-12 in excess of a revised operating limitminimum critical power ratio based on revised and conservative ANFB-112:
'orrelation constant uncertainty.
On the basis of the November 25, 1997, licensee response, the safety limitminimum critical ratio was not exceeded during the actual events and transients experienced by the plant during Core Cycles 8-12.
The licensee's analysis to determine ifthe limitcould have been exceeded during Core Cycles 8-12 did not use licensing bases assumptions, bounds, and parameters.
Administrative controh and operating limits in place during Core Cycle 7-12 would not have ensured operation viithlnthe envelope of the licensing basis.
Therefore, had the limitingtransient occurred with design basis operating conditions, tne revised safety limitcould have been reached orexceeded.
October 28, 1998 WNP-2
PLANTISS ES MATRIX DATE TYPE SOURCE.
ID SFA TEMPLATE CODES 04/22/98 POS IR 98-01 NRC ENG 4B 4C EA 98-228 03/17/98 URI IR 98-05 NRC ENG 4C ITEM DESCRIPTION Based on a review of all available information, and in particular the information provided at the conference, the NRC has determined that there was no violation of NRC requirements in this casr Despite NRC's concern over the fuel vendor's use of a limited number of data points in deriving uncertainty values, which the NRC believes resulted in nonconservative limits, the NRC agrees fundamentally with the Supply System's position that those efforts were consistent with NRC-approved guidance documents.
In that the NRC is satisfied with the corrective actions taken by tl Supply System, the NRC does not believe that this matter warrants further evaluation.
The vendor development and Implementation of the minimum critical power ratio operating a~
safety limits for WNP-2 fuel were not adequate to assure that the limits were accurate and conservative.
Licensee oversight of the fuel vendors'esign processes and controls for the nuclear fuel supplied to WNP-2 failed to detect that an inadequate technical specification limitwas developed.
The failure to establish measures to assure that the design bases were correctly translated into technical specifications was identified as an apparent violation of Criterion III, Appendix B to 10 CFR Part 50.
An unresolved item was identified for the reactor core isolation cooling system test return line throttle and isolation valves. The item involves whether the valves'erformance should have beei effectively controlled through the performance of appropriate preventive maintenance in accordance with the requirements of 10 CFR 50.65(a)(2).
03/17/98 NEG IR 98-05 NRC ENG 4B 4C 5A The effectiveness of the system walkdowns was mixed. The licensee appropriately identified
'oncerns with the containment instrument air system; however, concerns with the reactor core isolation cooling system performance and post operation condition were not promptly identified by walkdowns or plant data review.
03/17/98 VIO IR 98-05 NRC ENG 4C 1C SL IV 03/17/98 POS IR 98-05 NRC ENG 4B 5B 4C 03/14/98 VIO IR 98-03 NRC ENG 2B 4C SL IV A violation of Technical Specification 5.4.1a and Regulatory Guide 1.33 was identified for changin<
the inten't of the logic system test to allow low pressure coolant injection into the reactor ves using the temporary change notice process.
The licensee effectively identified and corrected the cause of the main steam line isolation valve containment air supply line failure. Common cause failure of the other main steam line isolation valve instrument air lines was appropriately considered.
The licensee aggressively addressed concerns with the Dlvishn II logic system performance during the event and verified the Division II logic system functionality.
A number of deficiencies were identified ln the implementation of the licensee's leakage surveillance and prevention program.
Specifically, procedur'es for performing visual and integratei leakage Inspections on the standby gas treatment system, the containment monitoring system, an the post acckfent sampling system, were inadequate in that they failed to identify all of the a
ro riate s tern corn onents to be monitored October 28, 1998 12 WNP-2
PLANT ISSUES MAT IX DATE TYPE SOURCE ID SFA TEMPLATE CODES 03/14/98 NEG IR 98-03 NRC ENG 4B 5C 02/19/98 NCV IR 97-20 LIC ENG 4A LER '6-007 02/19/98
- NCV IR 97-20 NRC ENG 4A 4C 02/19/98 NCV IR 97-20 LIC ENG 4A LER 97-001 02/19/98 NCV IR 97-20 LIC ENG 1C LER 97-002 02/09/98 POS IR 97-13 LIC ENG 4B 4A 02/09/98 WK IR 97-13 NRC ENG 5C 4B ITEM DESCRIPTION ln reviewing the testing requirements for the standby gas treatment system, the inspector identifie the potential for the system floor drains to present a bypass pathway around the filters. In response to the inspector's concerns, the licensee took appropriate action to verify that the curren leakage is acceptable, and to develop a long-term monitoring program for this potential unfiltered leakage path.
Licensee procedures for controlling the configuration of the 4160V vital switchgear breakers~
ensure that configurations would be consistent with the seismic qualification of the switchgear~
noncited violation was identified associated with this 1996 licensee event report.
Three examples were identified in which the licensee had evaluated and implemented a change tc the facility, as described in the Final Safety Analysis Report, but failed to update the report in accordance with 10 CFR 50.71 (ENG). The licensee is Implementing a broad review of the Final Safety Analysis Report to identify and correct any additional errors. A noncited violation was Identified.
In establishing the flowswitch high flowisolation setpolnt for the reactor water cleanup system biowdown line, engineering personnel did not adequately review the instrument loop design. This resulted in the application of an improper conversion factor for the flowswitch and a nonconservative high flowisolation setpoint that exceeded the maximum allowable technical specification value. A noncited violation was identified associated with this 1997 licensee event report.
Calibration and surveillance procedures for the rod block monitor system were found to be inadequate to ensure the rod block monitors were operable prior to exceeding 30 percent rated thermal power as required by Technical Specifications.
As a result, the system did not enforce ro<
blocks until power was approximately 33 percent. A noncited violation was identified associa with this 1997 licensee event report.
An adequate evaluation of the March 3, 1996, residual heat removal system test results was performed that demonstrated that the results were within the design basis.
While Engineering Directorate Manual 2.15 was properly implemented, actions were being taken t further control the number of calculation modification records for plant calculations.
A self-assessment performed by the licensee did not identify ifthe outstanding calculation modification records potentially affected the technical content of the calculations.
The NRC plans further reviei of this area durln a future lns ection October 28, 1998 13 WNP-2
PLANT ISS ES MATRI DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 02/09/98 NCV VII.B.3 IR 97-13 NRC ENG 4A 5C EA 97-612 Multiple examples of Final Safety Analysis Report inaccuracies were identified. While no safety issues or operability issues were identified, these multiple examples were indicative of a failure to update the Final Safety Analysis Report. However, the ongoing implementation of a Final Safety Analysis Report update program permitted the exercising of enforcement discretion in accordance with the revised enforcement policy.
02/09/98 VIO IR 97-13 NRC ENG SL III EA 97-573 4A 4B 01/15/98 NCV IR 97-18 LIC ENG 4A 5A 01/15/98 WK IR 97-18 NRC ENG 4C 02/09/98 POS IR 97-13 NRC ENG 4C SC 4A 5A The lack of inclusion of the high pressure core spray service water loop in the corrosion progra was appropriate considering the type of failure that occurred.
In addition, the inclusion of the pressure core spray service water system in the wall thickness measurement program was considered to be a proactive approach toward eliminating any future problems.
Contrary to 50.59(a)(1 and (a)(2) in 1985, v ithout prior Commission approval, a change was made to the facilityas described in the safety analysis report involving an unreviewed safety question.
The reactor core isolation cooling system, a system required for safe shutdown, was downgraded from safety-related to nonsafety-related which also redesignated the system such that it was no-hnger Seismic Category I. This change constituted an unreviewed safety question in that it increased the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
Contrary to 50.55a(f) and ASME IWV-1100, as of December 1994, the inservice testing for certain reactor core Isolation cooling valves, whose function was required for safety, was not impIementet as required by Section XIof the appropriate edition and addenda of the ASME Boiler and Pressun Vessel Code. Specifically, as the result of downgrading the reactor core isolation cooling system from safety-related to nonsafety-related, Valves RCIC-V-13, the head spray isolation valve; RCIC-V-19, the minimum-ffowto suppression pool isolation valve; RCIC-V-28, the auxiliary cooling to suppression pool isolation valve; RCIC-V-31, the suppression pool to RCIC suction valve; RCIC-V<0, the turbine exhaust to suppression pool isolation valve; and RCIC-V-66, the head spray Isolation valve were not timed during stroke testing In the open direction to assure that~
met specified acceptance criteria. In addition, Valve RCIC-V<5, the turbine steam supply Iso~
valve, was no longer tested for either opening or closing stroke times.
No response required The licensee IdentTiied that plant procedures for testing the automatic isolation function of reactor core Isolation cooling were inadequate in that they did not verify the proper operation of the Division II Isolation seal-In logic contact.
Identified performance Issues in the leakage surveillance and prevention program, regarding plant staff knowledge, program implementation, and procedural inconsistencies, were indicative of weal management Involvement and poor program maintenance.
However, these Issues did not result i an si nificant safe concerns.
October 28, 1998 14 WNP-2
PLANTI S ES MATRI DATE TYPE SOURCE ID SFA TEMPLATE CODES 01/15/98 NEG IR 97-18 NRC ENG 4A 10/01/97 POS IR 97-11 LIC ENG 5A 10/01/97 NEG IR 97-11 NRC ENG 5A 5C 10/01/97 POS IR 97-11 NRC ENG 5A 5C 10/01/97 NEG IR 97-11 NRC ENG 5B 10/01/97 NEG IR 97-11 NRC ENG 5C SA 10/01/97 POS IR 97-11 SELF ENG 5A 10/01/97 POS IR 97-11 LIC ENG 5A 5C 10/01/97 NEG IR 97-11 NRC ENG 5C 5A 10/01/97 NEG IR 97-11 NRC ENG 5A 5C ITEM DESCRIPTION The licensee's use of an uncontrolled database during its power uprate implementation resulted in an affected design calculation for the ultimate heat sink being missed in the review process.
The existing revision of the calculation bounded the parameters of the power uprate.
The licensee's fuel assembly examination and review of vendor Information provided an adequate basis to conclude that significant fretting damage to fuel cladding, due to broken fuel assembly debris filtersprings, had not occurred.
The licensee operated Cycles 7-12 with Incorrect and nonconservative core operating limitrepo (COLR) values for the OLMCPR. The OLMCPR was not calculated in accordance with NRC-approved topical reports referenced in Technical Specification 5.6.5.b. The licensee's staff determined that the corrected and more conservative OLMCPR was exceeded during each of the Cycles 7-12.
The licensee's on-line monitoring of the nodal core operating limits with the Powerplex Monitoring Program was adequate.
There were fuel assembly debris filters whose springs failed in Cycle 12. The potential for the failures might have been detected by a better testing and examination program of the debris filters prior to their commercial introduction.
The licensee had not completed a planned review and, as a result, had not yet determined ifthe SLMCPR would have been exceeded for anticipated operational transients.
The corrective action by the licensee to remove the fuel assembly debris filters and modify the lower support pieces was satisfactory.-
For Cycle 13 operation, (1) the licensee applied a 0.975 conservative multiplier to the operat~
limitminimum critical power ratio (OLMCPR) calculated using the ABB/CE methodology for~
resident fuel and (2) the power level of the most reactive (twice-burned) SPC resident fuel willbe lower than in the previous cycle. These conditions provided sufficient confidence that operating SPC fuel at the OLMCPR, would not challenge the safety limitminimum critical power ratio (SLMCPR) should an anticipated operational transient occur during Cycle 13.
A proposed facilitylicense amendment did not assure conservative limits for Cycle 13 operation and, thus, was not acceptable.
The licensee's initial methodology used for confirmation of the ABB/CE correlation to predict the thermal behavhr of SPC fuel was deficient in that it could not detect absolute errors in the SPC correlation, or in the application of the SPC correlation to obtain the data matrix used for the deva men of the ABB/CE correlation.
October 28, 1998 15 WNP-2
PLANTlSSUE MATRl DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPT(ON 09/28/98 POS IR 98-18 NRC PS 3A 09/28/98 POS IR 98-18 NRC PS 3A 09/28/98 POS IR 98-18 NRC PS 3A 09/18/98 STR IR 98-18 NRC PS 1C 3A 3B The EOF staffs performance was good. Emergency classifications and protective action recommendations were correct and timely. Offsite agency notifications were timely with one licensee-Identified exception. The Department of Energy notification for the site area emergency was slightly de'layed due to the loss of the primary notification system and incorrect backup telephone numbers.
One notification form was not properly completed; the date and time were omitted from the site area emergency notification form. The error was quickly recognized and verbally corrected.
A discrepancy between the emergency plan and implementing procedure~
kfentified concerning followup notifications. Appropriate corrective actions were taken to resoIIII the discrepancy.
Dose assessment and field team control activities were proper'ly performed to support protective action recommendations and validate dose projections.
Interactions with offsitc agency representatives were candid and supportive The OSC staff's performance was generally satisfactory. Three-part communications were frequently used.
Facility briefings were frequent and contained sufficient detaiL Health physics briefings tended to delay repair team dispatch because only one person conducted the briefings.
The process used to select field team members for tasks requiring self-contained breathing apparatus did not verifycorrective lense availability. Repair team documentation was incomplete and could have affected airborne dose reconstruction.
There was no emergency lighting installed In the OSC, although emergency electrical generators were available. Appropriate corrective actions were taken to address the lack'of battery-powered air samplers.
Public address announcements and station alarms could not be heard in all areas of the plant. A health physics emergency locker contained degraded supplies and insufficient quantities of protective clothing The TSC staffs perfomiance was good. Changing plant conditions were promptly and correctly analyzed to support EOF emergency classifications.
Staff briefings and technical discussions wer effective. Some key technical issues, Including recirculation pump vibration, reactor coolant~
makeup and leak rate, and standby gas treatment performance were not aggressively pursu~
The method used to assign and track repair team priorities was unclear and hampered the operations support center's (OSC's) ability to manage repair team resources.
Habitability was challenged because: (1) the outer airlock door was not fullyclosed, (2) at least one person did not frtsk prior to reentry, and (3) emergency ventilation system operation was not verified until late in the exercise Overall, performance was good. The control room, technical support center, and emergency operations facility successfully implemented most essential emergency plan functions including classification, protective action recommendations, and dose assessment.
Critiques were thorougl and self-critical.
October 28, 1998 16 NNP-2
PLANT ISS ES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 09/18/98 WK IR 98-18 NRC PS 28
. 3A 3C 09/18/98 NEG IR 98-18 NRC PS 3C 09/18/98 NEG IR 98-18 NRC PS 3A 5A 09/18/98 NEG IR 98.18 LIC PS 2A 3A 3C 09/17/98 VIO IR 98-20 NRC PS 1C 3A SL IV 09/17/98 NEG IR 98-20 NRC PS 1C 5B 5C An exercise weakness was identified in the operations support center for failure to properly monitc habitability. Airborne, contamination, and area surveys were either never performed or were not regularly performed in all areas.
A discrepancy between the emergency plan and implementing procedures was identified concerning followup notifications.
The exercise objectives were appropriate to meet emergency plan requirements.
The Initially~
submitted scenario was not acceptable because offsite doses were not challenging and would~i demonstration of some exercise objectives.
Projected offsite doses were increased to an acceptable level in the revised scenario; however, the scenario developers incorrectly computed the offsite field team sample data. As a result, the offsite doses were not consistent with expecter projected doses and did not challenge the dose assessment staff, field team members, and decision-makers.
Scenario development has been a historical problem.
In addition, the scenario developers failed to recognize that the loss of offsite power would affect OSC operations.
Last minute controller instructions and impromptu controller actions during the exercise were thorough and conscientious The Department of Energy notification for the site area emergency was slightly delayed due to the loss of the primary notification system and incorrect backup telephone numbers.
Because of competing priorities in responding to the June 17 fire protection system rupture and flooding event, required fire watches were not established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the system impairment.
The delay of approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in implementing the compensatory measures was found to be reasonable based upon the nature of the event. A second example of a failure to implement compensatory measures for a fire protection system impaIrment was identified by the inspectors during planned corrective maintenance on June 26.
A violation of Technical Specification 5.~
was kfentified for failure to followfire protection program implementing procedures; however,~
because the corrective actions were appropriate to address the root cause, no response is required.
The fire protection corrective action program was ineffective in addressing water hammer in the fire protection water supply system.
The corrective actions taken in 1984 for known water hamme concerns were only partially effective in addressing the impact of multiple pump starts on the hydraulic transients resulting from system initiation. Subsequent indications of severe hydraulic transients in the fire protection system were not evaluated and resultant component failures were treated as broke'-fix maintenance items. These component failures and industry operating experience on water hammer both represented missed opportunities to ferret out continuing system design problems.
October 28, 1998 17 WNP-2
PLANT I ES MATRI 07/23/98 NEG IR 98-14 NRC PS 3B 3C 07/23/98 STR IR 98-14 NRC PS 3C 2B 07/23/98 VIO IR 98-14 NRC PS 3C SL IV 07/23/98 POS IR 98-14 NRC PS 1C 3A 2B 3B 07/23/98 WK IR 98-14 NRC PS 1B 3B 07/16/98 NCV IR 98-12 NRC PS 1C 3A 07/16/98 STR IR 98-12 NRC PS 1C r
DATE TYPE SOURCE ID SFA TEMPLATE CODES 09/17/98 VIO IR 98-20 NRC PS 1C 5C SL IV ITEM DESCRIPTION The corrective actions trom previous Inadvertent actuations of the fire protection system were either Ineffective in addressing personnel knowledge and procedure weaknesses in the ignition source permit process or not promptly implemented.
The inadvertent actuation of the diesel generator building corridor preaction system (System 66) on June 17, occurred over 4 months aft'n almost identical event in February 1998.
Although procedural enhancements were defined shortly atter the event, the implementation of the enhancements was not scheduled until as late a:
August 1998.
A violation ot Ucense Condition 2.C.(14) and the fire protection corrective ac~
program was identified; however, because the corrective actions for the violation were appro~
no response is required.
Department staffin was lacking In health physics expertise.
A new emergency preparedness manager strengthened department problem resolution and self
-assessments.
With upper management support, emergency response organization callout capabilities were improved by expanding the use of pagers and Initiating the use of cellular telephones.
A reduction in initial training requirements and the lack of training/retraining program descriptions i
the emergency plan were identified as a violation of 10 CFR 50.54(q).
Overall, implementation of the emergency preparedness program was good. Self critical and thorough assessments of emergency plan implementation were made for two actual events.
Emergency response tacilities were operationally maintained and appropriate equipment and supplies were readily available at the primary facilities. The alternate emergency operations facilit was upgraded to avoid the need to transfer equipment and materials from the primary facility. A recent audit led to Increased emphasis on establishing and maintaining emergency response organization personnel qualifications. There was enough depth in the emergency response organization to ensure continuous staffing.
During the simulator walkthroughs, a performance weakness was identified for failure of one of tw crews to recognize that dose projections indicated a need tor protective action recommendations beyond 10 miles.
Pursuant to Secthn VII.B.1 of the NRC Enforcement Policy, a nonclted violation was identified invoMng failure to complete employment checks on two indMduals before granting temporary unescorted access.
An excellent fitness-for<uty program was in place.
Precautions had been taken to insure detectic ifindMduals attempted to circumvent the test with false specimens.
Alltesting was properly conducted and monitored. The licensee's fitness-for-duty procedures were in-depth, comprehensive.
and of excellent quali October 28, 1998 18 WNP-2
PLANT I ES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 07/16/98 NEG IR 98-12 NRC PS 1C 2A 07/16/98 STR IR 98-12 NRC PS 1C 07/02/98 STR IR 98-17 NRC PS 1A 1C 07/02/98 NEG IR 98-17 LIC PS 28 5A 07/02/98 POS IR 98-17 NRC PS 5A 07/02/98 VIO IR 98-09 NRC PS 1C SL IV ITEM DESCRIPTION One detection zone failed to alarm during a test simulating jumping into the protected area.
This single failure was not identifiable or predictable.
In general, performance in the security and access authorization was excellent.
An effective program for searching personnel, packages, and vehicles was maintained.
Proper procedures were in place to control personnel, package, and vehicle access to the protected area.
Very good protected area barriers and detection systems were maintained.
Testing of the detection aids wa performance based and ensured that system failures were discovered and corrected.
Anelfe~
testing and maintenance program was conducted.
The timely response to repair detection ai~
access control equipment, and vital area door locks and closures was instrumental in the low number of compensatory postings..The security training program and documentation of training were excellent. Security officers displayed excellent knowledge of the procedural requirements foi the task they were performing.
An excellent security event log system was in place for reporting safeguards events.
The licensee audits and self-assessment programs were excellent.
Overall, good radiohgical and meteorological monitoring programs were implemented.
Replacement of all environmental air sampler units ln 1997 reduced the number of equipment malfunctions from 19 in 1996 to 4 in 1997. The annual land use censuses were properly conducted.
Sample collection logs, shipment and release forms, and sample analyses reports were meticulously maintained at a high level of quality. Meteorological data recovery was greater than 92 percent from 1995 through 1997.
The licensee identified that the procedures used to calibrate wind speed and delta temperature Instrument loops allowed for tolerances outside the limits specified in the Final Safety Analyses Report from 1983 through 1996. A review of calibration records indicated that the actual tolerances of meteorological Instruments from 1995 through 1997 were within the Final Safety Analyses Report limits. Inadequate procedural reviews coupled with maintenance personnel performing the calibrations not being familiar with the requirements were identified as the pri contributors for this long-term procedural error.
Comprehensive radiohgical environmental operating reports were submitted in a timely manner.
These reports discussed such anomalies as detectable levets of cesium-137 and cobalt-60 found river sediment and soil samples which were attributed to releases from the Department of Energy during the operation of the old Hanford Reservation reactors.
The licensee decreased the effectiveness of its emergency plan between February 1997 and April 1998, when it reduced on-shift health physics expertise and overburdened the chemistry technicia with health physics responsibilities during emergencies.
Aviolation of 10 CFR 50.54(q) was identiTied. The licensee returned a third health physics technician to on-shift following notTiicatlon of the noncom lienee.
October 28, 1998 19 WNP-2
PLANT ISSUES MATRI DATE TYPE SOURCE ID SFA TEMPLATE CODES 06/26/98 STR IR 98-08 NRC PS 1A 1C 06/26/98 POS IR 98-08 NRC PS 5A 5B 5C 06/26/98 POS IR 98-08 NRC PS 1A 1C 06/26/98 POS IR 98-08 NRC PS 3B 06/15/98 POS IR 98-13 NRC PS 3A 06/05/98 POS IR 98-10 NRC PS 1A 06/05/98 POS IR 98-10 NRC PS 1A lTEM DESCRIPTION Overall, good solid radioactive waste management and radioactive waste/materials transportation programs were implemented.
Documentation and packages were properly prepared for shipment Good facilities were maintained for the processing, storage, and management of solid radioactive wastes and transportation activities. An effective radioactive waste inventory/accountability syster was maintained.
The volume and radioactivity ol solid radioactive waste generated during the time period 1993 through 1997 showed a continuing decline; even though, during this same ti period the station's 3-year rolling averages of the amount of solid radioactive waste generate were greater than the Industiy median for solid radioactive waste production at boiling water reactor facilities. Challenging solid radioactive waste generation goals for fiscal years 1997 and 1998 were met indicating the effective implementation of an improved solid radioactive waste minimization program.
Good performance based biennial audits of the solid radioactive waste and transportation programs were performed.
In 1997, the chemistry department performed a comprehensive self assessment of the solid radioactive waste processing program and shipping activities. In respons to these assessments, timely corrective actions and program Improvements were implemented.
Modifications to the condensate filterdemineralizer system resulted in a significant reduction in th(
amount of spent resin generated.
The personnel dose from radwaste activities showed a decreas between 1994 and 1997.
Good training and qualification programs were implemented.
Personnel involved in the processin!
packaging, and shipping of radioactive materials and wastes were properly trained and qualified.
As-low-as-reasonably-achievable planning for the troubleshooting and repair of a traversing incore probe drive cable was effective in evaluating the potential radiological hazards and'communicatinc them to the involved personnel.
Good radiological controls practices and health physics sup also contributed to dose reduction for the work.
Overall, quality department oversight of radiation protection activities was good. The quality department included a member with a strong operational radiation protection background.
Quality department operational radiation protection surveillances performed since January 1997 were intrusive and provided management with a very good assessment of program performance. The timeliness of problem evaluation requests improved during the past 6 months.
Overall, a good training program was effectively implemented.
Lesson plans were well organized, developed, and site and industry lessons learned were incorporated. The radiation protection department was appropriately involved in developing the training topics to ensure that the practica and technical corn etence of the radiation rotection staff was maintained.
October 28, 1998 20 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 06/05/98 NEG IR 98-10 NRC PS 1A 06/05/98 VIO IR 98-10 NRC PS 1A 3A SL IV 06/05/98 POS IR 98-10 NRC PS 1A 05/19/98 NCV IR 98-07 NRC PS 1A 3A 05/19/98 STR IR 98-07 NRC PS 1A 05/19/98 VIO IR 98-07 LIC PS 1A 3A SL IV 05/19/98 STR IR 98-07 NRC PS 1A ITEM DESCRIPTION The content of ALARAwork packages needed improvement.
Site lessons learned for similar worl were properly recorded in ALARAwork history packages; however, Industry lessons learned were not included. Job improvement ideas and suggestions were normally not captured from craft Ieve licensee or contractor personnel at the completion of job activities. The senior site ALARA committee was not fullysupported by the operations department.
Between January 1, 1997, and June 3, 1998, the operations representative only attended three of the five ALARACommittee meetings.
The station had not established a hot spot reduction program. Therefore, the licen~
did not know how many hot spots were present or which ones contributed significant exposure~
station workers. The licensee did not have an ALARAsuggestion tracking system to ensure suggestions were not misplaced or forgotten.
Aviolation of Technical Specification 5.4.1 was identified involving the failure of the senior site ALARACommittee to review the 1998 refueling outage (R-13) exposure goal and ALARAreviews and exposure reduction effectiveness evaluations were not performed for shielding installations.
ln general, the external exposure control program was effectively implemented.
Radiological area were properly controlled and posted.
Radiation protection personnel stationed at the radiological controlled area egress point provided appropriate and timely guidance to workers who alarmed thi personnel contamination monitors. Housekeeping within the radiological controlled area was gex Trash and laundry containers were properly maintained.
A noncited violation of Technical Specification 5.4.1.a was identified involving the failure to barricade and conspicuously post a high-high radiation areas.
Overall, good external exposure control and dosimetry programs were implemented..
All Technical SpecNcation high, and high-high radiation areas observed were properly controlled and posted.
Dosimeter placement was proper to monitor exposure fromboth uniform and nonunif photon radiation fields. Housekeeping within the radiological controlled area was good. Mat and equipment used for outage activities were properly stored and controlled. An effective train I
program for contract radiation protection technicians had been Implemented.
Aviolation of Technical Specification 5.4.1.a, with three examples, was identified involving the failure to perform proper radiological surveys.
An effective ALARAprogram had been implemented.
The licensee has made significant improvement to reduce person-rem for the period 1994-1997 as evident by yearly person-rem totals of 867 and 248 respectively. The 1998 person-rem projected dose is 255. Outage and nonouta eALARA rson-rem oalswerechallen in andinclosea reementwithactualresults.
October 28, 1998 21 WNP-2
PLAN I S ES ATRI DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION On/09/98 STR IR 98-02 NRC PS 1A 04/09/98 NCV IR 98-02 LIC LER 97-S01 PS 1C 5A 03/19/98 NEG IR 98-04 NRC PS 5A 03/19/98 NCV IR 98-04 LIC PS 3A 03/19/98 STR IR 98-04 NRC PS 3C 03/19/98 VIO IR 98-04 NRC PS 3B SL IV 04/25/98 VIO IR 98-06 NRC PS 1A 3A SL IV 5C Inconsistent expectatlons for implementing radiological controls requirements resulted in several procedure noncompliances during an instrumentation and controls surveillance performed in a posted high radiation area.
Specifically, an improper radiation work permit was utilized for the Iob, and positive access control to the high radiation area was not maintained by a qualified health physics technician.
Although the actual and potential dose consequences of the event were considered to be low, the generic implications were considered significant in that several administrative barriers to personnel overexposure were not properly implemented.
Aviolation~
Technical Specification 5.4.1.a was identified for failure to properly implement written procedu~
for radiation protection.
Implementation of the security program continued to be highly effective. Management of the security program was excellent. An effective program for searching personnel, packages and vehicles was maintained.
Excellent assessment aids provided effective and complete assessmen of the perimeter detection zones.
Alarm stations were redundant and well protected.
Good radio and te'lephone communications systems were maintained.
The compensatory measures program was effectively implemented.
Changes to security programs and plans were reported to the NRC within the required time frame. Overall, implementing procedures met the performance requirements in the physical security plan. An excellent training program that included conducting shift contingency drills had been implemented.
The licensee's on-shift security staffing was properly maintained.
A noncited violation was klentified involving the faffure to implement compensatory measures for e inoperative microwave security zone. This licensee identified violation is being treated as a noncited vhlatlon consistent with Section VII.B.1 of the NRC Enforcement Policy.
The quality of oversight of the radioactive effluent monitoring program by the quality assurance organization declined. The 1996 quality assurance audit was good, but the 1997 audit was w because the review lacked depth and performance-based input.
A noncited violation related to the calibration of the radwaste building exhaust monitor was identified The licensee maintained a good radioactive effluent monitoring program. The licensee demonstrated animproving trend in the reduction of radioactive effiuents during 1995-1997.
Ucensee personnel performed well in identifying and correcting a problem dealing with the calibration of the radwaste building exhaust monitor.
A10 CFR Part 50, Appendix B, Criterion XVIII,violation was identified because the audit team members did not have experience or training in'the special nature of the radioactive effluent monitorin r
ram.
October 28, 1998 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 03/14/98 VIO IR 98-03 NRC PS
=
1C 3B SL IV 5B Ucensee corrective actions to address weaknesses in implementing the transient combustible control program have not been effective in addressing the root cause and precluding repeat noncompliances with procedural requirements.
The root cause of these non-compliances appeared to be a lack of understanding of fire protection requirements and inattentiveness to fire protection labeling on the part of plant personnel.
02/19/98 POS IR 97-20 NRC PS 5A 5B 02/19/98 NEG IR 97-20 SELF PS 1C 4B 01/15/98 VIO IR 97-18 NRC PS 1C 4B SL IV 01/15/98 NCV IR 97-18 LIC PS 1C 5A 11/21/97 NEG IR 97-19 NRC PS 1C 11/21/97 NEG IR 97-19 NRC PS 1C 02/19/98 VIO IR 97-20 NRC PS SC 1C SL IV 5A Corrective actions to address inadequate labeling of radioactive material containers have not b~
effective in preventing recurrence, as evidenced by several recent noncompliances identified the inspectors and the licensee, and resulted in a violation of 10 CFR 20.1904(a).
Additionally, a lack of defined ownership of areas in the radwaste building contributed to poor radiological housekeeping practices on the 507 foot elevation.
The licensee's analysis and root cause evaluation of the unexpected movement of the traversing in-core probe accurately characterized the event and identified a number of areas for improvemen including personnel level of knowledge of TIP system operation and level of involvement of radiation protection supervision in the ALARAplanning process for high radiological risk jobs.
Engineering controts placed upon the traversing in-core probe drive C were insufficient in preventing movement of the probe during troubleshooting activities. The unexpected movement c the probe required personnel action to prevent the probe from withdrawing from its shielded location and into the area where the troubleshooting was being performed.
Based upon other barriers to personnel overexposure that were in place, and the immediate actions taken in response to the event, the likelihood of a significant overexposure was low.
Implementation of the licensee's program for monitoring and control of combustibles in the plant has been inconsistent in that 1) materials have been allowed to accumulate in limited access area without being property evaluated or tracked, and 2) inconsistencies in the licensee's combusti loading calculation, coupled with a relatively large backlog of modifications to the current revi the calculation, reduced the value of the calculation as a tool ln supporting plant modifications.
The licensee's failure to test the control room facsimile machine contributed to an inoperable piecr of emergency response equipment going undetected until it was required to be used during an actual event. A noncited vhlation was identified.
The radioactive material control program needed improved procedural guidance to ensure accountability of items conditionally released from the radiological controlled area.
Sealed radioactive sources were maintained and leak tested properly Problems with high radiation area controls and radiological hazard evaluations were identified; however ex sure controls were ade uate overall October 28, 1998 23 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 11/21/97 STR IR 97-19 NRC PS 5A 5B An excellent audit of the radiation protection program was conducted by the quality department.
The audit was comprehensive and effective in identifying areas of potential improvement 11/21/97 VIO IR 97-19 NRC PS 5A 5B SL IV Failure to evaluate radiological hazards associated with potential intakes of radioactive material was identified 11/21/97 VIO IR 97-19 NRC PS 1C SL IV 11/21/97 WK IR 97-19 NRC PS 1C 11/21/97 WK IR 97-19 NRC PS 5C 11/08/97 POS IR 97-17 NRC PS 1C 11/08/97 NEG IR 97-17 NRC PS 1C SC 10/24/97 VIO SL IV 10/24/97 P OS IR 97-07 NRC PS 1C IR 97-07 NRC PS 4A 10/24/97 POS IR 97-07 NRC PS 1C 10/24/97 POS IR 97-07 NRC PS 4C 11/21/97 POS IR 97-19 NRC PS 3A 1C Significant improvement was made in reducing the number of personnel contamination events Control of access to a high high radiation area was identified. Other exposure controls were implemented appropriately.
Improved guidance was needed in implementing procedures Involving the evaluation of potential internal radiological hazards, radioactive materials control, personnel contamination events, and portable radiation instruments Corrective actions by the radiation protection organization were slow and sometimes ineffective As low as reasonably achievable (ALARA)planning for several steam leak repair activities identified effective radiological controls and work practices.
The unavailability of members of the emergency response organization, along with technical and training issues related to the use of the licensee's automatic notification system, have challenged the licensee in demonstrating its ability to staff the onsite emergency response facilities in accordance with the emergency plan. The licensee's short term corrective actions to address this concern appear appropriate.
Failure to establish required vehicle control measures The installed vehicle barrier system was consistent with the summary description previously submitted to the NRC, encompassed all vital areas, and accurately described in the security plan Procedures properly addressed security surveillance, maintenance, compensatory measures, vehicle access control, and the safe shutdown of the plant.
The bomb blast analysts was consistent with the summary description and met regulatory requirements October 28, 1998 24 WNP-2
ENCLOSURE 2 Date
+pc SFA Sources ID Issue Description Co es GENERAL DESCRIPTION OF PIM TABLELABELS Actual date of an event or signiiicant issue for those items that have a clear date ofoccurrence, the date the source of the information was issued (such as the LFR date), or, for inspection reports, the last date of the Inspection period.
The categorization of the issue - see the Type Item Code table.
SALP Functional Area Codes: OPS for Operations; MAINTfor Maintenance; ENG for Engineering; and PS for Plant Support.
The document that contains the issue information: IR for NRC Inspection Report or LER for Licensee Event Report.
Identification ofwho discovered issue: N for NRC; L for Licensee; or S for Self Identifying (events).
Details ofthe issue from the LER text or from the IR Executive Summaries.
empla Cod bl.
F.D Strength Weakness F.F.I 4 VIO NCV DF.V Positive Negative LER URI +*
Licensing MISC TYPE ITEMCODES Enforcement Action Letter with CivilPenalty Enforcement Discretion - No CivilPenalty Overall Strong Licensee Performance Overall Weak Licensee Performance Escalated Enforcement Item - Waiting Final NRC Action Violation Level I, II,III,or IV Non-Cited Violation Deviation from Licensee Commitment to NRC Individual Good Inspection Finding Individual Poor Inspection Finding Licensee Event Report to the NRC Unresolved Item from Inspection Report Licensing Issue from NRR Miscellaneous - Emergency Preparedness Finding (EP),
Dedared Emergency, Nonconformance Issue, etc.
TEMPLATECODES Operational Perfonnance: A-Normal Operations; B - Operations During Transients; and C - Programs and Processes Material Condition: A - Equipment Condition or B - Programs and Processes Human Performance: A - Work Performance; B - Knowledge, Sldlls, and AbilitiesI Training; C - Work Environment Engineering/Design: A - Design; B - Engineering Support; C - Programs and Processes Problem Identification and Resolutioni A - Identiiicatlon; B - Analysis; and C-Resolutlon NOTES:
EEIs are apparent violations of NRC requirements that are being considered for~
escalated enforcement action In accordance with the "General Statement ofPolicy~
Procedure forNRC Enforcement Action"(Enforcement Policy), NUREG-1600.
Howeva; the NRC has not reached its final enforcement decision on the issues identliii by the EEIs and the PIM entries may be modified when the final decisions are made.
Before the NRC makes its enforcement decMon, the licensee willbe provided with an opportunity to either (1) respond to the apparent vIolation or (2) request a predecisiom enforcement conference.
Ums are unresolved items about which more information is required to determine whether the issue in question is an acceptable item, a devIation, a nonconformance, or i violatiori. Hovrever, the NRC has not reached its Iinal conclusIons on the issues, and th PIM entries may be modIIIed when the final conciusfons are made.
ENCLOSURE 3 WASHINGTON NUCLEAR PROJECT, UNIT2 IP - Inspection Procedure Tl - Temporary Instruction Core Inspection - Minimum NRC Inspection Program (mandatory all plants)
INSPECTION TITLE/
PROGRAM AREA NUMBER OF INSPECTORS DATES TYPE OF INSPECTION/COMMENTS IP 71001 REQUALIFICATIONPROGRAM EVALUATION IP 83750 OCCUPATIONALRADIATIONEXPOSURE IP 81700 PHYSICALSECURITY PROGRAM IP 73753 INSERVICE INSPECTION 12/7 - 11/98'ORE INSPECTION 3/15 - 19/99 CORE INSPECTION 3/29 - 4/2/99 CORE INSPECTION 4/19 - 23/99 CORE INSPECTION IP 84750 IP 84750 RADIOACTIVEWASTE TREATMENT, AND EFFLUENT AND ENVIRONMENTAL MONITORING RADIOACTIVEWASTE TREATMENT, AND EFFLUENT AND ENVIRONMENTAL MONITORING 5/24 - 28/99 CORE INSPECTION 7/12 - 16/99 CORE INSPECTION