ML17292B607
| ML17292B607 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/19/1999 |
| From: | Laura Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Parrish J WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 9903300367 | |
| Download: ML17292B607 (48) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION IV 611 RYAN PLAZA DRIVE, SUITE 400 ARLINGTON,TEXAS 76011;8064 N8 'l 9 l999 Mr. J. V. Parrish (Mail Drop 1023)
Chief Executive Officer
. Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968
SUBJECT:
PLANT PERFORMANCE REVIEW (PPR) - WASHINGTON NUCLEAR PROJECT - 2 (WNP-2)
Dear Mr. Parrish:
On February 11, 1999, the NRC staff completed a PPR of WNP-2. The staff conducts these reviews for all operating nuclear power plants to develop an integrated understanding of safety performance.
The results are used by NRC management to facilitate planning and allocation of inspection resources.
PPRs provide NRC management with a current summary of licensee performance and serve as inputs to the NRC's senior management meeting (SMM) reviews.
PPRs examine information since the last assessment of licensee performance to evaluate long-term trends but emphasize the last 6 months to ensure that the assessments reflect current performance.
The PPR for WNP-2 involved the participation of all technical divisions in evaluating inspection results and safety performance information for the period March 2, 1997, to January 25, 1999. The NRC's most recent summary of licensee performance was provided in a letter of April 3, 1997, and was discussed in a public meeting with you on April22, 1997.
As discussed in the NRC's Administrative Letter 98-07 of October 2, 1998, the PPR provides an assessment of licensee performance during an interim in which the NRC has suspended its Systematic Assessment of Licensee Performance (SALP) program.
The NRC suspended its SALP program to complete a review of its processes for assessing performance at nuclear power plants.
At the end of the review period, the NRC willdecide whether to resume the SALP program or terminate it in favor of an improved process.
The current period of detailed focus began on April 18, 1998, when operators placed the unit in Mode 3 to begin the R13 refueling outage.
Operators initiated a power ascension on June 15; however, because of problems with a transversing incore probe drive cable and subsequent rupture of a fire main, the plant did not achieve full power until July 8. On August 5, upon energizing a 4160 Vac vital bus, a voltage transient caused Reactor Feedwater Pump B to trip and a power decrease to 70 percent.
On August 7, operators initiated a shutdown because the limiting condition for operation allowed outage time was expiring. Operators returned the plant to full'power on August 23. On November 7, operators reduced power to 60 percent to investigate oscillations on Reactor Feedwater Pump B. Operators restored power to 100 percent on November 10. The plant operated at essentially 100 percent power until the end of the assessment period except for power decreases because of economic dispatch.
Overall, performance continued to be acceptable.
While operators effectively maneuvered the plant through numerous power transients in conjunction with economic dispatch, they were challenged on several occasions.
In addition, the operations organization did not effectively 9903300367 9903i9 PDR
- DOCK OS000397 8
Washington Public Power Supply System implement the corrective actions associated with the previously identified problem of operator response to nonroutine events.
On the whole, the plant was well maintained; however, several instances were identified where material condition deficiencies affected the reliability of safety-related and risk significant components.
Engineering's support of operations and maintenance activities was essential in sustaining plant operations; however, implementation weaknesses in program areas, most notably design control, were identified.
In plant support, the licensee had well established programs; sporadic instances of implementation deficiencies were identified in emergency response, fire protection, and radiological controls.
Operators performed well during routine activities, demonstrating strong command and control and effective communications.
However, they continued to demonstrate difficultyin responding to nonroutine events.
This was evidenced by weaknes'ses in the operator actions associated with the fire main.flooding event; similarly, during a licensed operator training evaluation, operators failed to initiate drywell sprays in a timely manner.
Because of the continuing concern with corrective actions not being effective in improving the response to nonroutine events, a regional initiative inspection of the corrective action program is planned.
In the maintenance area, the program and craft skills were sufficient to ensure the overall safe operation of the facility. There were, however, some instances of poor testing, such as the failure of a technician to properly isolate an instrument, which generated an inadvertent scram signal. Additionally, a question has arisen over the attention being given to the material condition and reliabilityof safety-related and risk significant components.
This concern has grown from occurrences such as the failure to properly maintain the floor drain cross-conn'ect valves, a contributing factor in the fire main flooding event.
Only core inspection efforts are scheduled for the maintenance area over the next 8 months, but there willbe a focus in the above-mentioned areas of concern.
In the engineering area, the overall effective support provided to plant operations and maintenance activities was,counterbalanced by several instances of poor engineering products.
This was evidenced by the inadequate procedures developed for troubleshooting a diesel generator and the resultant loss of the associated vital bus. The licensee recently strengthened the programs for improving safety evaluations and controlling modifications; however, there were still problems noted with design control. Several old failures to implement the design basis were identified during this period.
In addition, issues continued to be identified concerning agreement between the Final Safety Analysis Report and the as-built facility. An example was the system out-of-service signals that did not feed the low pressure core spray out-of-service annunciator as described in the Final Safety Analysis Report.
Finally, it was noted that engineers did not always use the corrective action program effectively. Only core inspection efforts are scheduled for the engineenng area over the next 8 months; however, the corrective action inspection willsample how effective engineering support and remediation activities have been.
In the plant support area, the implementation of the radiological controls program continued to provide acceptable protection to plant workers and the public. Failures to conspicuously post a radiation area and the poor establishment of a contamination zone were noted as exceptions to expected performance.
The licensee had established good programs for solid radioactive waste, transportation, meteorological monitoring, exposures;and dosimetry, as evidenced by effective training and audits, detailed documentation, and improved performance measures.
Washington Public Power Supply System Emergency preparedness performance was mixed when implementing requirements.
The security program was being effectively implemented in all areas.
Weaknesses were identified in the implementation of the fire protection program and in the effectiveness of corrective actions for fire protection issues; however, performance in the fire protection area remained satisfactory.
Only core inspection efforts are scheduled in the plant support area over the next 8 months; again, the corrective action inspection will provide insights into the licensee's effectiveness in correcting identified deficiencies.
Enclosure 1 contains a historical listing of plant issues, referred to as the Plant Issues Matrix (PIM), that were considered during this PPR process to arrive at an integrated view of licensee performance trends.
The PIM includes items summarized from inspection reports or other docketed correspondence between the NRC and Washington Public Power Supply System.
The NRC does not attempt to document all aspects of licensee programs and performance that may be functioning appropriately.
Rather, the NRC only documents issues that the NRC believes warrant management attention or represent noteworthy aspects of performance.
In addition, the PPR 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.
This material willbe placed in the PDR as part of the normal issuance of NRC inspection reports and other correspondence. provides definitions for some of the information listed in the PIM.
This letter advises you of our planned inspection effort resulting from the WNP-2 PPR 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.
details our inspection plan for the next 8 months.
Also included in the plan are NRC noninspection activities. The rationale or basis for each inspection outside the core inspection program has been provided in this letter so that you are aware of the reason for emphasis in these program areas.
Resident inspections are not listed due to their ongoing and continuous nature.
Because of the anticipated changes to the inspection program and other initiatives, this inspection schedule is subject to revision. Any changes to the schedule listed willbe promptly discussed with your staff.
If you have any questions, please contact me at 817/860-8137.
Sincerely, Lin a J. Smith, ctiHg Chief Project Branch E Division of Reactor Projects Docket Nos. 50-397 License Nos. NPF-21
Washington Public Power Supply System
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)
Vice President, Generation 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 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, N.W.
Washington, D.C. 20005-3502
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Washington Public Power Supply System,
John L. Erickson, Director Division of Radiation Protection Department of Health Airdustrial Center Building ¹5 P.O. Box 47827 Olympia, WA 98504-7827 Max E. Benitz, Jr., Chairman Board of Benton County Commissioners P.O. Box 190 Prosser, WA 99350 Sue Miller, Chair Board of Franklin County Commissioners 1016 North 4th Street Pasco, WA 99301
Washington Public Power Supply System biAR j 9 gz bcc to DCDe<fl.
bcc distrib. by RIV:
Regional Administrator DRP Director DRS Director Branch Chief (DRP/E)
Senior Project Inspector (DRP/E)
Chief, NRR/DISP/PIPB B. Henderson, PAO T. Boyce, NRR/DISP/PIPB C. Hackney, RSLO W. D. Travers, EDO (MS: 16E15)
Associate Dir. for Projects, NRR Associate Dir. for Insp., and Tech. Assmt, NRR PPR Program Manager, NRR/ILPB (2 copies)
Chief, Inspection Program Branch, NRR Chief, Regional Operations and Program Manage W. Bateman, NRR Project Director (MS: 13E16)
C. Poslusny, NRR Project Manager (MS: 13E16)
Resident Inspector DRS Branch Chiefs (3 copies)
MIS System RIV File Branch Chief (DRP/TSS)
Chief, OEDO/ROPMS C. Gordon Records Center, INPO ment Section, OEDO DOCUMENT NAME: S:)DRP<DRPDIR<PPR<WNP To receive co of document, indicate in box: "C" = Co without enclosures "E" = Co with enclosures "N" = No co RIV:SPE:DRP/E GAPick;df D:DRS AC:DRP/E LJSmith Qb ATHowell RECONCUR D:DRP KEBrockman LJSmith L,)>
3/;/99 3/ i499 3/
/99 3/
/99 3/ 0/99 OFFICIAL RECORD COPY 300301
Washington Public Power Supply System ji4R I 9 I999 bcc to DCD (IE40) bcc distrib. by RIV:
Regional Administrator DRP Director DRS Director Branch Chief (DRP/E)
Senior Project Inspector (DRP/E)
Chief, NRR/DISP/PIPB B. Henderson, PAO T. Boyce, NRR/DISP/PIPB C. Hackney, RSLO W. D. Travers, EDO (MS: 16E15)
Associate Dir. for Projects, NRR Associate Dir. for Insp., and Tech. Assmt, NRR PPR Program Manager, NRR/ILPB (2 copies)
Chief, Inspection Program Branch, NRR Chief, Regional Operations and Program Manage W. Bateman, NRR Project Director (MS: 13E16)
C. Poslusny, NRR Project Manager (MS: 13E16)
Resident Inspector DRS Branch Chiefs (3 copies)
MIS System RIV File Branch Chief (DRP/TSS)
Chief, OEDO/ROPMS C. Gordon Records Center, INPO ment Section, OEDO DOCUMENT NAME: S~DRP'tDRPDIRtPPR'LWNP To receive co of document, indicate In box: "C = Co without enclosures "E" = Co viith enclosures N" = No co RIV:SPE:DRP/E AC:DRP/E D:DRS D:DRP RECONCUR GAPick;df 3/
/99 LJSmith Q 3/ i%99 ATHoweli 3/
99 3/.'99 3/ 8/99 OFFICIALRECORKCOPY KEBrockmarjt, >5 LJSmith L63
PLANT ISSUES MATRIX ENCLOSURE 1
DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 12/22/98 POS IR 98-23 NRC OPS 1B 12/10/98 NEG IR 98-24 NRC OPS 1C 12/10/98 WK IR 98-24 NRC OPS 1A 1B 11/11/98 POS IR 98-22 NRC OPS 1A 11/5/98 VIO IR 98-22 NRC OPS 1A SL IV 10/10/98 NEG IR 98-21 NRC OPS 1A 3A 9/28/98 POS IR 98-11 NRC OPS 5A 5C Operators demonstrated a proper safety focus when responding to smoke issuing from a constant voltage supply transformer in the reactor building, in that appropriate attention and resources were given to address and control the event without losing sight of other operational responsibilities A deficiency in the requalification examination development'process was identified in that the process does not address the verification of 10 CFR 55.43 sampling for the written requalification examination.
A generic operator performance weakness was identified in the area of control board awareness, which involved repeated failures of operators to take appropriate responses to changing plant parameters or system misalignments.
Also, inconsistent communications were observed during crew briefings given during the dynamic simulator scenarios.
The control room operators demonstrated proper safety focus in responding to a partial loss of annunciators on the reactor control board. The crew's evaluation of the significance of the event and their implementation of compensatory measures were both timely and appropriate.
A methodical troubleshooting plan was effective in isolating the root cause and returning the annunciators to service During performance of a quarterly Technical Specification surveillance on the standby liquid control system; neither the procedure nor control room togs adequately tracked the equipment configuration to verify adherence to short term outage times allowed by Technical Specifications.
As a result, the status of system operability with both trains inoperable could not be accurately reconstructed.
The failure to log and track the status of the standby liquid control system was identified as a violation of the equipment control process and Technical Specification 5.4.1.a; however, because of appropriate corrective actions, no response was required 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 is not subject to formal enforcement action Operations personnel were effective in the identification and resolution of conditions adverse to quality.
January 25, 1999 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 9/28/98 NCV IR 98-11 NRC OPS 5A 9/17/98 POS IR 98-20 NRC OPS 1B 9/17/98 NEG IR 98-20 SELF OPS 3A 9/17/98 NEG IR 98-20 NRC OPS 1B 3A ITEM DESCRIPTION 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 due to a combination of factors, which included a lack of questioning attitude and minimal management involvement. The failure to initiate performance evaluation requests (two examples) circumvented the corrective action program and was identified as a non-cited violation, pursuant to Section VII.B.1 ot the NRC Enforcement Policy, of Procedure PPM'1.3.12.
The licensee responded well to the flooding event.
The shitt 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 tlooding and dewater the reactor building were prompt and effective.
As a result ot human error, the watertight door between the reactor building northeast stairwell and residual 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 actions of the operators to start the Iow 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.
7/18/98 POS IR 98-13 NRC OPS 1A 2B 7/7/98 POS IR 98-13 NRC OPS 1A 2B 7/2/98 POS IR 98-09 NRC OPS 1A 1C 6/10/98 POS IR 98-301 NRC OPS 3B The licensee was well prepared for plant restart from the 1998 refueling outage as evidenced by proper closure of outage activities, completion of required Technical Specification surveillances, 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 ot.postmaintenance testing on the reactor feedwater pumps following modiTications 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 preevolution 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.
Allapplicants (4 reactor operator and 8 senior operators) passed their initial license examinations.
January 25, 1999 WNP-2
J
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 6/10/98 STR IR 98-301 NRC OPS 1A 3B 6/10/98 WK IR 98-301 NRC OPS 3C 6/7/98 VIO IR 98-13 NRC OPS 1A 3A SL IV 4/25/98 VIO IR 98-06 NRC OPS 1A 5A SL IV 4/25/98 POS IR 98-06 NRC OPS 1A 3/17/98 VIO IR 98-05 SELF OPS 1C 4C SL IV 3/17/98 NOV IR 98-05 LIC OPS 1B 1C SL IV ITEM DESCRIPTION Good operator performance and communication practices were observed during the initial operator licensing examination.
The licensee initiallyfailed to submit an acceptable examination for administration to operator license applicants for the operating test portion of the examinations.
The final as-given examination met the requirements of NUREG-1021 and was considered good quality.
Poor procedure use during the restoration from an inadvertent engineered safety feature actuation resulted in the mispositioning of the minimum flow bypass valve for the low pressure core spray system.
Numerous control board walkdowns performed by operators failed to identify the discrepancy.
A violation of Technical Specification 5.4.1.a was identified tor failure to tollow procedure when returning the low pressure core spray system to its standby lineup.
Control of plant equipment was generally effective in maintaining proper plant configuration.
However, two examples were identified where 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 tailed to 1) recognize that inoperable drain valves for the service water spray rings were required by the final safety analysis report tor treeze 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 Specification 5.4.1.a tor failure to properly implement written procedures for control of plant equipment.
Good command and control of the March 18 reactor startup and April3 feedwater temperature reduction was evidenced by adequate planning, proper assignment of personnel responsibilities, clear communications, and a conservative approach to implementing the activities.
A violation ot Technical Specification 5.4.1a and Regulatory Guide 1.33, with two examples of inadequate procedures, was identified for a Division II logic system functional test and the Division IIIemergency 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 vessel during the conduct of the March 12, 1998, logic system tunctional 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.
A violation of Technical Specification 5.4.1a 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 provided in Procedure OSP-RCS-C102, "RPV Vessel Cooldown Surveillance," Revision 0, Attachment 9.1, "MinimumVessel Metal Temperature VS Reactor Vessel Pressure."
January 25, 1999 WNP-2
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 3/17/98 NEG IR 98-05 SELF OPS 1B 3B 4B 3/17/98 WK IR 98-05 NRC OPS 1C 3/17/98 VIO IR 98-05
.NRC OPS 1B 1C SL IV Although the actions prior to the main steam line isolation valve nitrogen supply line failure and overall response to the complex transient were appropriate; weaknesses with operators'nowledge, 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 management control was not implemented for the procedure temporary change process and control of infrequently performed tests and surveillance.
Operator workarounds appeared in significant areas involving vessel level and pressure control, temperature monitoring and forced circulation.
Communication within the control room and with the NRC headquarters operations officer was poor and did not ensure that key control room personnel were cognizant of the overall plant and systems.
The 10 CFR 55.59, Licensed Operator Requalification Program, did not address the make up of crew complement used in simulator training vs the control room and was considered a significant weakness in the licensed operator requalification training program.
A violation was identified 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 reactor vessel.
3/17/98 NEG IR 98-05 NRC OPS 5A SB 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.
3/14/98 NCV IR 98-03 SELF OPS 1A 3/14/98 NEG IR 98-03 NRC OPS 1A 5A 5B 5C Inadequate self-checking and peer checking resulted in an operator error that deenergized non-vital Bus SM-2 and started the Division III emergency diesel generator.
Operations personnel actions in response to the transient were appropriate and prompt. The root cause analysis and corrective actions effectively addressed the human performance concerns.
One instance was identified in which an operating crew did not demonstrate a conservative approach to equipment operation when a non-vital lighting panel, with an unidentified ground, was reenergized without an understanding of the source of the ground or a troubleshooting plan to identi the source.
January 25, 1999 WNP-2
PLANT ISSUES MATRIX DATE 2/19/98 TYPE SOURCE ID SFA TEMPLATE CODES NEG IR 97-20 NRC OPS 1A 1C ITEM DESCRIPTION The program to assure that corrective lenses for self contained breathing apparatus (SCBA) for operators requiring them was implemented successfully. However, procedural guidance for maintenance ot 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 need to have SCBA qualified lenses, regardless of the type ot corrective lenses normally used.
2/19/98 NCV IR 97-20 SELF OPS 3A LER 96-002 2/9/98 NCV IR 97-13 NRC OPS 4C 4B 2/9/98 POS IR 97-13 LIC OPS 3B 3A 2/9/98 POS IR 97-13 LIC OPS 5C 2/9/98 VIO SL IV IR 97-13 NRC OPS 3B 5C 2/19/98 STR IR 97-20 NRC OPS 1A 3B The professionalism ot the control room operators and shift management ownership of crew activities supported good operational performance over the inspection period. Operators were generally 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 order activities resulted in the momentary deenergization ot the Division II 4160V vital bus and the loss of residual heat removal assist cooling of the spent fuel pool. A noncited violation was identified associated with this 1996 licensee event report.
The new nuclear satety assurance division procedure properly addressed the technical specification procedural requirements.
In addition, licensee conducted surveillances were effective in assuring that other canceled procedure activities were properly conducted.
However, there was a failure to update the Final Safety Analysis Report fire protection sections.
Actions to address the occurrence ot shorting electrical terminals during the performance of maintenance or surveillance activities were adequate and effective toward preventing a recurrence of the events.
The corrective actions that addressed the inadvertent initiation of drywell to suppression chamber bypass flowwere appropriate tor the circumstances and adequate to prevent a recurrence of the events.
While corrective actions to resolve the material buildup problem in Valves FDR V-3 and FDR V-4 were eftective, corrective actions to resolve a required reading problem were not. Violation 50-397/9611-04 willbe closed, 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 tailure to issue a problem evaluation request that would have promptly identified and provided corrective actions tor the inadvertent start of a reactor recirculation pump. This item was considered to be an exam le of a violation of 10 CFR Part 50 A endix B Criterion XVI.
-January 25, 1999 5
WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 1/15/98 VIO IR 97-18 NRC OPS 2A 1C SL IV 11/8/97 STR IR 97-17 NRC OPS 1A 1C 3B ITEM DESCRIPTION A number of inspector identified deficiencies in the control of transient equipment indicated weak implementation of the program to prevent seismic interactions between the equipment and safety-related components.
Three examples of a violation of plant procedures were identified.
Management involvement in the plant curtailment for maintenance on the reactor feedwater drive turbines (RFWDT) was notable for reemphasizing expectations and raising personnel sensitivity to a significant evolution. The operations staff also demonstrated conservative decision-making when maintenance on the first drive turbine was delayed while operability concerns with the high ressurecores ra HPCS s stemwereaddressed.
January 25, 1999 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 10/7/98 VIO IR 98-22 NRC MAINT 3A 28 SL IV 9/28/98 POS IR 98-11 NRC MAINT 5A SC 9/17/98 VIO IR 98-20 NRC
=
MAINT 2A 28 SL IV EA 98.452 ITEM DESCRIPTION Maintenance personnel performance during Technical Specification required testing of the standby gas treatment system demonstrated knowledge deticiencies in the proper use of and adherence to procedures.
The failure in two instances to properly conduct the standby gas treatment system test was identified as a violation Technical Specification 5.4.1.a.
Planning and briefing for the work failed to ensure that all prerequisites were met prior to the start ot the test, and procedure steps were not performed in sequence.
Since the licensee implemented appropriate corrective actions, no response was required 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.
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 ot 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 resulted in the flooding of the low pressure core spray pump room and complicated recovery from the plant transient. The failure to monitor the performance ot the valves against established goals or to demonstrate reliabilityof the valves through an effective preventive maintenance program was identified as a violation of 10 CFR 50.65 (EA 98-452).
8/29/98 NEG IR 98-19 NRC MAINT 2A 5A 8/29/98 POS IR 98-19 NRC MAINT 28 38 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 condition, by itselt, rendered a safety-related system or component inoperable, both conditions had the potential to adversely affect equipment performance.
The processes tor 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 sensing line socket weld were thorough and generally well executed.
The repair plan and mockup were notable strengths.
Some minor deficiencies were identitied during execution of the repair.
January 25, 1999 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 7/1 8/98 POS IR 98-13 NRC MAINT 2B 5A 7/2/98 VIO IR 98-09 LIC MAINT 1A 3A SL IV 7/2/98 NEG IR 98-09 NRC
- MAINT, 2A 5B 7/2/98 NCV IR 98.09 LIC MAINT 1A 3A 7/2/98 POS IR 98-09 NRC MAINT 3A 5C ITEM DESCRIPTION The 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.
Eftorts 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 and a 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 control room and the refueling floor. However, two instances ot weak procedure use resulted in:
- 1) the failure to identify an incorrect precaution in the maintenance procedure tor the reactor building overhead crane, and,2) failure to verifythat appropriate minimum temperature requirements were being met prior to liftingthe drywell upper shield blocks.
The Division I emergency diesel generator experienced multiple material deticiencies during 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 Iow 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 short-term corrective actions tor the failures were appropriate.
Long-term actions willbe reviewed in future inspection activities.
Licensee performance in implementing FMC for the spent fuel pool, reactor cavity and reactor pressure vessel 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 pools 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.
The actions to address previously-identified weaknesses in implementing their toreign material controls (FMC) program tor plant systems and containment have been effective in raising the sensitivit and im rovin erformance of lant ersonnel.
January 25, 1999 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 7/2/98 VIO IR 98-09 'RC SL IV MAINT 1A 2B 3A Personnel performance in the conduct of testing excess flow check 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 tield, and 4) weak command and control. In one case, performance deficiencies resulted in the initiation ot an engineered safety features actuation signal and plant transient.
Two violations ot TS 5.4.1.a, each with two examples, were identified regarding adequacy and use of surveillance procedures.
The violations included inadequate procedure guidance for establishing and restoring from test conditions and failure to independently verify a valve location prior to valve manipulation.
6/15/98 NEG IR 98-13 LIC MAINT 3A LER 98-010 4/25/98 WK IR 98-06 NRC MAINT 2A 2B 3/14/98 VIO IR 98-03 NRC MAINT 2B 3B SL IV 3/14/98 NEG IR 98-03 LIC MAINT 2A 2/19/98 NCV IR 97-20 SELF MAINT 3A LER 96-001 2/19/98 POS IR 97-20 NRC MAINT 3A 4B 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 drive cable precluded the ability to close its associated containment isolation ball valve and necessitated a plant shutdown in accordance with Technical Specifications.
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 following a loss-of-coolant accident, and a locked spring hanger on the residual heat removal system's minimum flowbypass line.
Licensee personnel improperly applied surveillance requirement 3.0.2 to program surveillances in the administrative section of Technical Specifications.
As a result, a 25 percent surveillance interval extension was inappropriately utilized for several technical programs.
Poor material condition ot 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 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 Division1125VDC 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 a ro riate craft su ervision and s stem en ineerin artici ation.
January 25, 1999 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 1/15/98 WK IR 97-18 NRC MAINT 2B 2A The 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, a lower standard was established for these areas and equipment and housekeeping deficiencies were allowed to persist.
1/15/98 NCV IR 97-18 NRC MAINT 2B 11/8/97 STR IR 97-17 NRC MAINT 4B 4C 4A 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 in Technical Specifications PS).
Based, in part, upon the staff's acceptance of the version of the standard utilized by the licensee, and the more conservative results produced by its methodology, the noncompliance was viewed as a minor violation.
The troubleshooting and repair efforts associated with the RFWDTs were well planned and executed.
The efforts resulted in improved drive turbine performance while identifying potential desi n im rovements to the turbine overnor control oil s stem.
January 25, 1999 10 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 12/7/98 VIO IR 98-23 NRC ENG 4A SL IV 10/10/98 NEG IR 98-21 NRC ENG 4C IR 98-05 EA 98-203 10/7/98 VIO IR 98-22 NRC ENG 4C 2B SL IV
. 9/28/98 NCV IR 98-11 NRC ENG 4B 5A 5C ITEM DESCRIPTION The configuration of the low pressure core spray out-of-service annunciator did not conform to Final Safety Analysis Report Figure 7.3.9, in that out-of-service signals from Battery B1-1, Diesel Generator 1, and Service Water A were not supplied to the annunciator.
This design deficiency was not identified during ongoing etforts to review the accuracy ot the FSAR. The failure to ensure that the design basis, as specified in the license application, was correctly translated into drawings was identified as a violation of 10 CAR Part 50, Appendix B, Criterion lil, "Design Control."
However, because the licensee implemented appropriate corrective actions, no response was required The manual startup and shutdown of the reactor core isolation cooling system for level control, tollowing the March 1998 main steam isolation valve closure, challenged the operators.
The proceduralized method to control reactor vessel level by diverting reactor core isolation cooling tlow through the test return line could 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 reliabilityot the system and challenged the containment isolation function since the valves may not have closed against high differential pressure.
Unresolved Item 50-397/98005-05.
involving exclusion of the reactor core isolation cooling test return line valves from the scope of the maintenance rule, was determined not to be a violation of NRC requirements (EA 98-203)
The procedure for conducting bypass leakage testing of the standby gas treatment system charcoal filters was inadequate in that the procedure failed to provide sufficient instructions for injection of the challenge gas to ensure proper mixing in the filter plenum.
A violation of Technical Specification 5.4.1.e was identified for the inadequate procedure.
As a result, dispersion of the challenge gas has not always been sufficient to challenge all portions of the charcoal filters in order to verifyTechnical Specification bypass leakage requirements.
The inadequate procedure did not result in a safety issue, as subsequent testing with proper challenge gas dispersion demonstrated compliance with Technical Specifications.
Since the licensee implemented appropriate corrective actions, no response was required 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 non-cited violation of Criterion XVIto Appendix B ot 10 CFR Part 50, pursuant to 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.
January 25, 1999 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 9/17/98 NCV IR 98-20 'IC ENG 4A The discrepancy between the actual performan'ce of the reactor building watertight doors and their 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 tollowup to the flooding event and initiated appropriate action to address it. A noncited violation of 10 CFR 50.71(e) was identified tor failure to update Final Safety Analysis Report, in accordance with Section VII.B.1 of the Enforcement Policy.
9/17/98 VIO SL III 8/29/98 NCV IR 98-20 SELF ENG 1C 4A EA 98-480 IR 98-19 LIC ENG 2B 4B On June 17, inadequate fire protection system design resulted in a rupture of the WNP-2 tire protection system significantly impairing the safety capability of components important-to-safety in Residual Heat Removal Pump Room C and the low pressure core spray room due to flooding from a fire protection system rupture. The rupture occurred when excessive hydraulic forces, generated during preaction sprinkler system actuation in response to an actual fire detection system signal, caused a fire main valve to fail. NRC concluded that identification and corrective action credit waranted.
No civilpenalty applied.
(Severity Level III, EA 98-480)
The instructions established for troubleshooting the Division II emergency diesel generator failed to identify the inherent risk of loading the inoperable diesel generator onto its associated vital bus 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 8/29/98 VIO.
IR 98-19 NRC ENG 2B 4B SL IV EA 98-462 7/31/98 VIO IR 98-15 NRC ENG 4C SL IV 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 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 prohibited the performance of portions of the procedure during plant operations.
The failure of licensee personnel to properly review Technical Specifications during procedure development and approval was identified as a violation of the requirements of 10 CFR 50.59.
Design control errors were identified that did not affect equipment operability. The Mode 4 and 5 technical specification surveillance requirement acceptance criterion for condensate storage tank level did not assure the Technical Specification Bases commitment to maintain 135,000 gallons reserve in the condensate storage tank. The technical specification allowable value tor reactor vessel water Level 1 in the emergency core coolihg system instrumentation table was not correctly derived from the analytic limitfor Level 1 in that it did not include sufficient margin for post-accident
.environmental effects. These errors were determined to be examples of a violation ot Appendix B, Criterion III~ "Design Control."
January 25, 1999 12 WNP-2
PLANT ISSUES MATRIX
'DATE TYPE SOURCE ID SFA TEMPLATE CODES 7/31/98 POS IR 98-15 'RC ENG 4B 4C 7/31/98 POS
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IR 98-15 NRC ENG 4C 7/31/98 POS IR 98-15 NRC ENG 4B 7/31/98 VIO IR 98-15 NRC ENG 4B SL IV 7/31/98 VIO IR 98-15 NRC ENG 4B SL IV 7/31/98 POS IR 98-15 NRC ENG 5A 5C ITEM DESCRIPTION The installed instrumentation and controls met high pressure core spray system logic requirements.
The majority ot the high pressure core spray controls and instrumentation were installed in conformance with good human factors practices, and the licensee routinely trained the operators regarding the availability and use ot particular instruments during accidents.
High pressure core spray instrument setpoint channel check and calibration procedures were adequate to ensure safe and reliable operation, and the procedures were well written and had an adequate level of detail. The electrical distribution system for the high pressure core spray system was generally well designed.
The high pressure core spray pump had adequate available net positive suction head.
The team found that the high pressure core spray system valves were capable ot performing their functions under accident conditions.
The licensee had developed a data base that captured the relationship between calculations, so that they could identify needed calculation revisions. When a calculation was revised, the data base was used to identify all ot the other calculations that were potentially impacted by the revision.
This data base relied on calculation cross references; however, these cross references were not-always accurate or complete for older calculations Based on a sample of six modification packages, recent design modification packages were correctly prepared in accordance with the current procedures.
The plant modification requests addressed all relevant design and safety issues arid effectively verified the design changes by post-modification testing. The current tormat for a modification package was clear and easy to understand In violation of Appendix B, Criterion V, "Instructions, Procedures, and Drawings," procedures were not always adequately tollowed or prescribed.
Category 1 and 2 post-accident monitoring instruments were not all identified on the main control panel. The temperature assumptions in site ~
short-circuit calculations were not adequately verified to ensure they were representative of actual room temperatures.
In addition, the licensee had not identified and accounted for all outstanding calculation modification records, which could affect the results/conclusions of the electrical system toad tally The thermal performance test results for the high pressure core spray standby service water system demonstrated that the system was operable.
However, the Division 3 Diesel Generator cooler thermal performance test acceptance criterion did not assure acceptable thermal performance for all allowed high pressure core spray standby service water system flows in violation of Appendix B, Criterion XI, "Test Control The licensee responded to the team's tindings with a strong safety focus and effectively identified
" additional examples of issues identified by the team related to testing the Division 2 battery and consideration of post-accident environment effects in the setpoint ot reactor vessel Level 1 January 25, 1999 13 WNP-2
PLANTISSUES MATRIX 7/31/98 URI IR 98-15 NRC ENG 4B 7/18/98 POS IR 98-13 NRC ENG 4A 4C 7/1 8/98 NEG IR 98-13 NRC ENG 4A 5C 7/18/98 VIO IR 98-13 NRC ENG 4A 5A SL IV.
EA 98-398 7/15/98 ED IR 98-15 NRC ENG 4B 6/13/98 VIO IR 98-09 LIC ENG 5A SL IV DATE TYPE SOURCE ID SFA TEMPLATE CODES 7/31/98 VIO IR 98-15
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NRC ENG 5C SL IV ITEM DESCRIPTION The licensee had not appropriately assessed the significance of some breaker coordination errors in the dc system and, as a result, did not promptly correct these design deficiencies.
This was a violation of Appendix B, Criterion XVI,'Corrective Action, Establishment of appropriate leak testing for liquid secondary containment bypass valves. This matter is unresolved pending NRC review of the calculation method for evaluating the t consequences of leakage from the liquid bypass isolation valves 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 the 12 safety evaluations reviewed was not always consistent, overall the quality was good. Strengths 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 evaluation of the modified system's performance failed to identify a vulnerability to water hammer following a loss of offsite power. The vulnerability was adequately addressed when the system 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 the Final Safety Analysis 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 physical controls to protect against common mode flooding. A violation of 10 CFR 50.59 was identified for failure to document a written safety evaluation for this defacto change to the facility. The corrective actions to install a cap on the drain line and review the generic implications for other portions of the drain systems, were found to be appropriate.
The licensee had not fullyimplemented a new Technical Specification requirement and, as a result.
inappropriately credited a battery performance test for a battery service test. The team identified
, the issue as it related to the Division 3 battery.
During followup of the team's findings, the licensee identified a similar condition for the Divisions 1 and 2 Batteries.
This was significant because the previous service test had expired for the Division 2 battery, which was in violation of Technical Specification 3.0.2. This required a notice of enforcement discretion to allow continued operation During the fuel shuffle, the licensee accepted the condition of a partially elevated fuel assembly in the spent fuel pool without fullyevaluating its impact. Weak understanding and review of the plant's design basis failed to identify that fullinsertion of the spent fuel assembly is a condition of the s ent fuel criticalit anal sisdescribedin the Final Safet Anal sis Re ort January 25, 1999 14 WNP-2
~4
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 4/25/98 NEG IR 98-06 NRC ENG 4B 4/25/98 VIO IR 98-06 NRC ENG 4A 4B SL IV 4/25/98 VIO IR 98-06 NRC ENG 4A 5B SL IV EA 98-265 4/22/98 POS IR 98-01 NRC ENG 4B 4C EA 98-228 ITEM DESCRIPTION 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 ot the degraded condition.
The licensee failed to provide 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, full low pressure coolant injection flowto the vessel would not be achieved within the 66 seconds assumed by the analyses.
The failure to adequately translate the design requirements of the RHR system to plant operating procedures and instructions was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III (Design Control).
The licensee failed to recognize that operation ot 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 identity the need for a writteri 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
'urrent FSAR upgrade project. A violation ot 10 CFR 50.59(b)(1) was identified.
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 case.
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 the Supply System, the NRC does not believe that this matter warrants further evaluation.
The vendor development and implementation ot the mit)imum critical power ratio operating and safety limits tor 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 tor 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 ot Criterion III, A
endixBto10CFRPart50.
January 25, 1999 15 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 4/22/98 WK IR 98.01 NRC ENG 4B 4C 4/22/98 NEG IR 98-01 NRC ENG 4B 4C 4/22/98 NEG IR 98-01 NRC ENG 4B 4C The licensee operated the Siemens Power Corporation's fuel in Core Cycles 7-12 in excess of a revised operating limit minimum critical power ratio based on revised and conservative ANFB-1125 correlation constant uncertainty.
On the basis of the November 25, 1997, licensee response, the safety limitminimum critical power ratio was not exceeded during the actual events and transients experienced by the plant during Core Cycles 8-12.
The analysis to determine if the limitcould have been exceeded during Core Cycles 8-12 did not use licensing bases assumptions, bounds, and parameters.
Administrative controls and operating limits in place during Core Cycle 7-12 would not have ensured operation within the envelope of the licensing basis.
Therefore, had the limiting transient occurred with design basis operating conditions, the revised safety limitcould have been reached or exceeded.
3/17/98 VIO IR 98-05 NRC ENG 4C 1C SL IV 3/17/98 POS IR 98-05 NRC ENG 4B 5B 4C 3/17/98 URI IR 98-05 NRC ENG 4C
. A violation of Technical Specification 5.4.1a and Regulatory Guide 1.33 was idenlified for changing the intent of the logic system test to allow low pressure coolant injection into the reactor vessel 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 Division II logic system performance during the event and verified the Division II logic system functionality.
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 been effectively controlled through the performance of appropriate preventive maintenance in accordance with the requirements of 10 CFR 50.65(a)(2).
3/17/98 NEG IR 98-05 NRC EN G 4B 4C SA The effectiveness of the system walkdowns was mixed. The licensee appropriately identified concerns 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 lant data review.
January 25, 1999 16'NP-2
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 3/14/98 VIO IR 98-03 NRC ENG 2B 4C SL IV 3/14/98 NEG IR 98-03 NRC ENG 4B 5C 2/19/98 NCV IR 97-20 LIC ENG 4A LER 96-007 2/1 9/98 NCV IR 97-20 LIC ENG 1C LER 97-002 2/19/98 NCV IR 97-20 NRC ENG 4A 4C 2/19/98 NCV IR 97-20 LIC ENG 4A LER 97-001 2/9/98 POS IR 97-13 LIC ENG 4B 4A ITEM DESCRIPTION A number of deficiencies were identified in the implementation of the leakage surveillance and prevention program.
Specifically, procedures for performing visual and integrated leakage inspections on the standby gas treatment system, the containment monitoring system, and the post accident sampling system, were inadequate in that they failed to identify all of the appropriate system components to be monitored In reviewing the testing requirements for the standby gas treatment system, the inspector identified~
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 verifythat the current 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 did not ensure that configurations would be consistent with the seismic qualification of the switchgear.
A noncited violation was identified associated with this 1996 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 rod blocks until power was approximately 33 percent.
A noncited violation was identified associated with this 1997 licensee event report.
Three examples were identified in which the licensee had evaluated and implemented a change to 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 flow isolation setpoint for the reactor water cleanup system blowdown line, engineering personnel did not adequately review the instrument loop design. This resulted in the application of an improper conversion factor for the flow switch and a nonconservative high flow isolation setpoint that exceeded the maximum allowable technical specification value. A noncited violation was identified associated with this 1997 licensee event report.
An adequate evaluation of the March 3, 1996, residual heat removal system test results was erformed that demonstrated that the results were within the desi n basis.
January 25, 1999 17 WNP-2
PLANT ISSUES MATRIX 2/9/98 VIO SL III IR 97-13 EA 97-573 DATE TYPE SOURCE ID SFA TEMPLATE CODES NRC ENG 4A 4B 5A ITEM DESCRIPTION Contrary to 50.59(a)(1 and (a)(2) in 1985, without 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 longer Seismic Category I. This change constituted an unreviewed safety question in that it increased the probability ot 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 implemented as required by Section XI ot the appropriate edition and addenda of the ASME Boiler and Pressure 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-flowto 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-40, 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 they met specified acceptance criteria. In addition, Valve RCIC-V-45, the turbine steam supply isolation valve, was no longer tested tor either opening or closing stroke times.
No response required 2/9/98
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NCV VII.B.3 IR 97-13, NRC ENG 4A 5C EA 97-612 2/9/98 POS IR 97-13 NRC ENG 4C 5C 4A 1/15/98 NEG IR 97-18 NRC ENG 4A 2/9/98 WK IR 97-13 NRC ENG 5C 4B 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.
The lack of inclusion of the high pressure core spray service water loop in the corrosion program was appropriate considering the type of failure that occurred.
In addition, the inclusion of the high pressure core spray service water system in the wall thickness measurement program was considered to be a proactive approach toward eliminating any future problems.
While Engineering Directorate Manual 2.15 was properly implemented, actions were being taken to further control the number of calculation modification records for plant calculations.
A selt-assessment performed by the licensee did not identify it the outstanding calculation modification records potentially affected the technical content of the'calculations.
The NRC plans further review ot this area during a future inspection The 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 arameters of the ower u rate.
January 25, 1999 18 WNP-2
PLANT ISSUES MATRIX DATE 1/15/98 TYPE SOURCE ID SFA
.TEMPLATE CODES WK IR 97-18 NRC ENG 4C ITEM DESCRIPTION Identified performance issues in the leakage surveillance and prevention program, regarding plant staff knowledge, program implementation, and procedural inconsistencies, were indicative of weak management involvement and poor program maintenance.
However, these issues did not result in any significant safety concerns.
1/15/98 NCV IR 97-18 LIC ENG 4A 5A 10/1/97 NEG IR 97-11 NRC ENG 5A 5C 10/1/97 NEG IR 97-11 NRC ENG 5A 5C 10/1/97 POS IR 97-11 LIC ENG 5A 5C 10/1/97 NEG IR 97-11 NRC ENG 5C SA 10/1/97 NEG IR 97-11 NRC ENG 5C 5A 10/1/97 NEG IR 97-11 NRC ENG SB 10/1/97 POS IR 97-11 NRC ENG 5A 5C 10/1/97 POS IR 97-11 LIC ENG 5A The licensee identified 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.
The licensee operated Cycles 7-12 with incorrect and nonconservative core operating limit report (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 staff determined that the corrected and more conservative OLMCPR was exceeded during each of the Cycles 7-12.
The initial methodology used for confirmation of the ABB/CE correlation to predict the thermal behavior 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 development of the ABB/CE correlation.
For Cycle 13 operation, (1) the licensee applied a 0.975 conservative multiplier to the operating limit minimum critical power ratio (OLMCPR) calculated using the ABB/CE methodology for SPC 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.
The licensee had not completed a planned review and, as a result, had not yet determined if the SLMCPR would have been exceeded for anticipated operational transients.
A proposed facilitylicense amendment did not assure conservative limits for Cycle 13 operation and, thus, was not acceptable.
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 on-line monitoring of the nodal core operating limits with the Powerplex Monitoring Program was adequate.
The 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 filter s rin s hadnot occurred.
January 25, 1999 19 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 10/1/97 POS IR 97-11 'ELF ENG 5A ITEM DESCRIPTION The corrective action by the licensee to remove the fuel assembly debris filters and modify the lower su ort ieces was satisfacto January 25, 1999 20 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 12/15/98 NEG IR 98-23 'RC PS 4C 2B 12/14/98 VIO IR 98-23 NRC PS 3A SL IV 12/8/98 NEG IR 98-23 NRC PS 5C ITEM DESCRIPTION The licensee did not effectively ensure that the fire seal for Containment Penetration X099 was in accordance with licensee drawings and not degraded, as evidenced by the loosely packed penetration seal that issued a warm gas.
Based on gas analysis, the licensee confirmed that gas was not from containment.
In addition, the initial slow response to the inspectors'oncern of a warm gas issuing from a penetration was not commensurate with the potential safety significance of the finding The inspectors found a security officer reading a magazine, unauthorized material, while on duty in the secondary alarm station (SAS). This was a violation of License Condition 2.E for failure to meet the Commission-approved physical security plan; however, since the licensee implemented appropriate corrective actions, no response was required The immediate corrective actions associated with water leaking from an established contamination zone were adequate.
However, the corrective actions resulting from previous incidents were weak, as evidenced by water leaking from the contamination zone 10/1 3/98 NCV IR 98-22 LIC PS 2A 2B 9/28/98 NEG IR 98-18 NRC PS 3A 5A The failure to assign the appropriate priorityfor performing maintenance on post accident sampling system limitswitches was identified as a noncited violation of Technical Specification 5.4.1.a and
. maintenance procedures, consistent with Section VII.B.1 of the Enforcement Policy. Reliability and availability of the post accident sampling system had been adversely impacted by both repetitive
=
limitswitch failures and untimely maintenance.
The low priority placed upon maintenance of the limitswitches left the system in a degraded condition for 10 months and inoperable for approximately 2 months, until the failure of a quarterly operability surveillance elevated the issue.
Subsequent actions were more timely and comprehensive to address the reliabilityconcerns 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 emer enc locker contained de raded su lies and insufficient uantities of rotective clothin January 25, 1999 21 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 9/28/98 POS IR 98-18 'RC PS 3A 9/28/98 POS IR 98-18 NRC PS
. 3A ITEM DESCRIPTION The TSC staff's performance was good. Changing plant conditions were promptly and correctly analyzed to support EOF emergency classifications.
Staft briefings and technical discussions were effective. Some key technical issues, including recirculation pump vibration, reactor coolant makeup and leak rate, and standby gas treatment performance were not aggressively pursued.
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 ~
frisk prior to reentry, and (3) emergency ventilation system operation was not verified until late in the exercise The EOF staff's performance was good.
Emergency classifications and protective action recommendations were correct and timely. Oftsite agency notifications were timely with one licensee-identified exception.
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.
Appropriate corrective actions were taken to resolve the discrepancy.
Dose assessment and field team control activities were properly performed to support protective action recommendations and validate dose projections.
Interactions with offsite agency representatives were candid and supportive 9/18/98 STR IR 98-18 NRC PS 1C 3A 3B Overall, performance was good. The control room, technical support center, and emergency operations tacility successfully implemented most essential emergency plan functions including classification, protective action recommendations, and dose assessment.
Critiques were thorough and self-critical.
9/18/98 NEG IR 98-18 NRC PS 3C 9/1 8/98 NEG IR 98-18 NRC PS 3A 5A 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 limit 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 expected 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 9/18/98 NEG IR 98-18 LIC PS 2A 3A 3C The Department of Energy notification tor the site area emergency was slightly delayed due to the loss of the primary notification system and incorrect backup telephone numbers.
January 25, 1999 22 WNP-2
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 9/18/98 MISC IR 98.18-NRC PS 2B 3A 3C An exercise weakness was identified in the operations support center for failure to properly monitor habitability. Airborne, contamination, and area surveys were either never performed or were not regularly performed in all areas.
9/17/98 NEG IR 98-20 NRC PS 1C 5B SC 9/17/98 VIO IR 98-20 NRC PS 1C 5C SL IV 9/17/98 VIO IR 98.20 NRC PS 1C 3A SL IV The fire protection corrective action program was ineftective in addressing water hammer in the fire protection water supply system.
The corrective actions taken in 1984 for known water hammer 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 terret out continuing system design problems.
The corrective actions from previous inadvertent actuations ot the tire protection system were either ineftective in addressing personnel knowledge and procedure weaknesses in the ignition source permit process or not promptly implemented.
The inadvertent actuation ot the diesel generator building corridor preaction system (System 66) on June 17, occurred over 4 months after an almost identical event in February 1998.
Although procedural enhancements were defined shortly after the event, the implementation of the enhancements was not scheduled until as late as August 1998.
A violation ot License Condition 2.C.(14) and the fire protection corrective action program was identified; however, because the corrective actions tor the violation were appropriate, no response is required.
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 ot Technical Specification 5.4.1.d was identiTied for failure to followtire protection program implementing procedures; however, because the corrective actions were appropriate to address the root cause, no response is required.
Sanuary 25, 1999 23 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 9/17/98 NCV IR 98-20
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LIC PS 1B 3A 7/31/98 POS IR 98-15 NRC PS 1C 7/31/98 POS IR 98-15 NRC PS 2A 7/23/98 STR IR 98-14 NRC PS 3C 2B ITEM DESCRIPTION The Technical Support Center manager failed to confer with the emergency director prior to authorizing the discharge ot the stairwell floodwater to the storm drains. The error was the result of the improper placement of an emergency response requirement into an operations procedure instead ot the emergency plan implementing procedures.
The corrective actions taken to address this deficiency and evaluate the generic implications were appropriate.
A noncited violation ot Technical Specification 5.4.1.a was identified for tailure to tollow procedure, in accordance with Section VII.B.1 ot the Entorcement Policy.
The fire protection program was effectively implemented.
Fire equipment was being properly maintained, upgraded, and tested at the required frequencies.
The fire brigade was properly trained and qualified to perform fire fighting, and the annual medical examination requirement was being met. The fire protection program procedures were comprehensive in detailing the requirements for control ot transient combustibles, barrier impairments, and control of ignition sources.
With respect to the fire protection program, the licensee's audit and corrective action processes were effective. Several strengths identified in the 1997 audit by the licensee were confirmed by the team during this inspection, especially fire personnel knowledge and skill, and excellent material condition of the fire fighting equipment Based on the area walkdowns, the high pressure core spray and high pressure core spray standby service water system configurations were consistent with the design basis.
The plant reflects the apparent attention to housekeeping.
The team did not observe any improperly stored material or urisecured temporary equipment 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.
- 7/23/98 WK IR 98-14 NRC PS 3C 3A IR 98-14 NRC PS 1B 3B A reduction in initial training requirements and the lack of training/retraining program descriptions in the emergency plan were identified as a violation of 10 CFR 50.54(q).
During the simulator walkthroughs, a performance weakness was identified for failure of one of two crews to recognize that dose projections indicated a need tor protective action recommendations beyond 10 miles.
7/23/98 NEG IR 98-14 NRC PS 3B 3C Department staffing was tacking in health physics expertise.
January 25, 1999 24 WNP-2
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 7/23/98 POS IR 98-14 NRC PS 1C 2B 3B 7/16/98 STR IR 98-12 NRC PS 1C 7/16/98 NEG IR 98-12 NRC PS 1C
- 2A 7/16/98 STR IR 98-12 NRC PS 1C 7/1 6/98 NCV IR 98-12 NRC PS 1C 3A 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 facilities were operationally maintained and appropriate equipment and supplies were readily available at the primary facilities. The alternate emergency operations facility 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.
An excellent fitness-tor-duty program was in place.
Precautions had been taken to insure detection if individuals attempted to circumvent the test with false specimens.
Alltesting was properly conducted and monitored. The fitness-for-duty procedures were in-depth, comprehensive, and of excellent quality.
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 g'ood protected area barriers and detection systems were maintained.
Testing of the detection aids was performance based and ensured that system failures were discovered and corrected.
An eftective testing and maintenance program was conducted.
The timely response to repair detection aids, access control equipment, and vital area door locks and closures was instrumental in the low number ot compensatory postings.
The security training program and documentation of training were excellent.
Security officers displayed excellent knowledge of the procedural requirements tor 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.
Pursuant to Section VII.B.1 ot the NRC Enforcement Policy, a noncited violation was identified involving failure to complete employment checks on two individuals before granting temporary unescorted access.
7/2/98 STR IR 98-17 NRC PS 1A 1C Overall, good radiological and meteorological monitoring programs were implemented.
Replacement of all environmental air sampler units in 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.
January 25, 1999 25 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 7/2/98 VIO IR 98-09 NRC PS 1C SL IV 7/2/98 NEG IR 98-17 LIC PS 28 5A 7/2/98 POS IR 98-17 NRC PS 5A 6/26/98 POS IR 98-08 NRC
. PS 1A 1C 6/26/98 POS
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IR 98-08 NRC PS 5A 58 5C 6/26/98 STR IR 98-08 NRC PS 1A 1C 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 technician with health physics responsibilities during emergencies.
A violation of 10 CFR 50.54(q) was identified. The licensee returned a third health physics technician to on-shift tollowing notification of the noncompliance.
The licensee identified that the procedures used to calibrate wind speed and delta temperature instrument loops allowed for tolerances outside the limits specitied in the Final Safety Analyses Report from 1983 through 1996. A review ot 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 primary contributors for this long-term procedural error.
Comprehensive radiological environmental operating reports were submitted in a timely manner.
These reports discussed such anomalies as detectable levels of cesium-137 and cobalt-60 found in river sediment and soil samples which were attributed to releases from the Department of Energy during the operation of the old Hanford Reservation reactors.
Modifications to the condensate filterdemineralizer system resulted in a significant reduction in the amount of spent resin generated.
The personnel dose from radwaste activities showed a decrease between 1994 and 1997.
Good performance based biennial audits of the solid radioactive waste and transportation programs were performed.
In 1997, the chemistry department performed a comprehensive selt assessment of the solid radioactive waste processing program and shipping activities. In response ~
to these assessments, timely corrective actions and program improvements were implemented.
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 system was maintained.
The volume and radioactivity of solid radioactive waste generated during the time period 1993 through 1997 showed a continuing decline; even though, during this same time period the station's 3-year rolling averages of the amount of solid radioactive waste generated were greater than the industry 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 ro ram.
January 25, 1999 26 WNP-2
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 6/26/98 POS IR 98-08 NRC PS 3B 6/15/98 POS IR 98-13 NRC PS 3A 6/5/98 NEG IR 98-10 NRC PS 1A 6/5/98 POS IR 98-10 NRC PS 1A 6/5/98 POS IR 98-10 NRC PS 1A 6/5/98 VIO IR 98.-10 NRC PS 1A 3A SL IV 6/5/98 POS IR 98-10 NRC PS 1A ITEM DESCRIPTION Good training and qualification programs were implemented.
Personnel involved in the processing, 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 communicating them to the involved personnel.
Good radiological controls practices and health physics support also contributed to dose reduction for the work.
The content of ALARAwork packages needed improvement.
Site lessons learned for similar work 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 level 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 licensee did not know how many hot spots were present or which ones contributed significant exposure to station workers. The licensee did not have an ALARAsuggestion tracking system to ensure suggestions were not misplaced or forgotten.
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 practical and technical competence of the radiation protection staff was maintained.
In general, the external exposure control program was effectively implemented.
Radiological areas were property controlled and posted.
Radiation protection personnel stationed at the radiological controlled area egress point provided appropriate and timely guidance to workers who alarmed the ~
personnel contamination monitors. Housekeeping within the radiological controlled area was good.
Trash and laundry containers were properly maintained.
A violation 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 ALARA'reviews and exposure reduction effectiveness evaluations were not performed for shielding installations.
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 suweillances performed since January 1997 were intrusive and provided management with a very good assessment of program performance. The timeliness of roblem evaluation re uests im roved durin the ast 6 months.
January 25, 1999 WNP-2
PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 5/19/98 STR IR 98-07 'RC PS 1A 5/19/98 STR IR 98-07 NRC PS 1A ITEM DESCRIPTION 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 nonoutage ALARAperson-rem goals were challenging and in close agreement with actual results.
Overall ~ good external exposure control and dosimetry programs were implemented..
All Technical Specification high, and high-high radiation areas observed were properly controlled and posted.
Dosimeter placement was proper to monitor exposure from both uniform and nonuniform photon radiation fields. Housekeeping within the radiological controlled area was good. Materials and equipment used for outage activities were properly stored and controlled. An effective training program for contract radiation protection technicians had been implemented.
5/19/98 VIO SL IV 5/19/98 NCV IR 98-07 LIC PS 1A 3A IR 98-07 NRC PS 1A 3A A violation of Technical Specification 5.4.1.a, with three examples, was identified involving the failure to perform proper radiological surveys.
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.
4/25/98 VIO IR 98-06 NRC PS 1A 3A SL IV Inconsistent expectations for implementing radiological controls requirements resuited 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 job, and positive access control to the high radiation area was not maintained by a qualified health physics technician.
Although the actual and potential dos'e 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.
A violation of Technical Specification 5.4.1.a was identified for failure to properly implement written procedures for radiation protection.
4/9/98 NCV IR 98-02 LIC LER 97-S01 PS 1C 5A 5C A noncited violation was identifiyd involving the failure to implement compensatory measures for an-inoperative microwave security zone. This licensee identified violation is being treated as a noncited violation consistent with Section VII.B.1 of the NRC Enforcement Polic January 25, 1999 28 WNP-2
PLANTISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 4/9/98 STR IR 98-02 NRC PS 1A
.3/19/98 NCV IR 98-04 LIC PS 3A 3/19/98 NEG IR 98-04 NRC PS 5A 3/19/98 STR IR 98-04 NRC PS 3C 3/19/98 VIO IR 98-04 NRC PS 3B SL IV ITEM DESCRIPTION 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 assessment of the perimeter detection zones.
Alarm stations were redundant and well protected.
Good radio and telephone 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 on-shift security staffing was properly maintained.
A noncited violation related to the calibration of the radwaste building exhaust monitor was identified 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 weaker because the review lacked depth and performance-based input.
The licensee maintained.a good radioactive effluent monitoring program. The licensee demonstrated an improving trend in the reduction of radioactive effluents during 1995-1997.
Licensee personnel performed well in identifying and correcting a problem dealing wilh 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 monitoring program.
3/14/98 VIO IR 98-03 NRC PS 1C SL IV 3B 5B Licensee 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.
2/19/98 NEG IR 97-20 SELF PS 1C 4B Engineering controls placed upon the traversing in-core probe drive C were insufficient in preventing movement of the probe during troubleshooting activities. The unexpected movement of 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 res onse to the event the likelihood of a si nificant overex osure was low January 25, 1999 29 WNP-2
C PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES ITEM DESCRIPTION 2/19/98 POS IR 97-20 NRC PS 5A 5B 2/19/98 VIO IR 97-20 NRC PS 5C 1C SL IV The 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 improvement
~
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.
Corrective actions to address inadequate labeling of radioactive material containers have not been effective in preventing recurrence, as evidenced by several recent noncompliances identified by 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.
1/15/98 VIO SL IV IR 97-18 NRC PS 1C 4B 5A Implementation of the 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 areas without being properly evaluated or tracked, and 2) inconsistencies in the combustible loading calculation,
. coupled with a relatively large backlog of modifications to the current revision of the calculation, reduced the value of the calculation as a tool in supporting plant modifications.
1/15/98 NCV IR 97-18 LIC PS 1C 5A The failure to test the control room facsimile machine contributed to an inoperable piece of emergency response equipment going undetected until it was required to be used during an actual event. A noncited violation was identified.
11/21/97 VIO IR 97-19 NRC PS 1C SL IV Control of access to a high high radiation area was identified. Other exposure controls were implemented appropriately.
11/21/97 VIO IR 97-19 NRC PS 5A 5B SL IV 11/21/97 NEG IR 97-19 NRC PS 1C 11/21/97 WK IR 97-19 NRC PS SC 11/21/97 NEG IR 97-19 NRC PS 1C 11/21/97 WK IR 97-19 NRC PS 1C 11/21/97 POS IR 97-19 NRC PS 3A 1C Significant improvement was made in reducing the number of personnel contamination events Failure to evaluate radiological hazards associated with potential intakes of radioactive material 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 Corrective actions by the radiation protection organization were slow and sometimes ineffective Problems with high radiation area controls and radiological hazard evaluations were identified; however, exposure controls were adequate, overall Improved guidance was needed in implementing procedures involving the evaluation of potential internal radiological hazards, radioactive materials control, personnel contamination events, and ortable radiation instruments January 25, 1999 30 WNP-2
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PLANT ISSUES MATRIX DATE TYPE SOURCE ID SFA TEMPLATE CODES 11/21/97 STR IR 97-19 NRC PS 5A 5B 11/8/97 POS IR 97-17 NRC PS 1C 11/8/97 NEG IR 97-17 NRC PS 1C 5C 10/24/97 POS IR 97-07 NRC PS 1C 10/24/97 POS IR 97-07 NRC PS 4A ITEM DESCRIPTION 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 As low as reasonably achievable (ALARA)planning for several steam teak 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 automatic notification system, have challenged the licensee ~
in demonstrating its abilityto staff the onsite emergency response facilities in accordance with the emergency plan. The short term corrective actions to address this concern appear appropriate.
Procedures properly addressed security surveillance, maintenance, compensatory measures, vehicle access control, and the safe shutdown of the plant.
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 10/24/97 VIO SL IV IR 97-07 NRC PS 1C Failure to establish required vehicle control measures 10/24/97 POS IR 97-07 NRC PS 4C The bomb blast analysis was consistent with the summary description and met regulatory requirements January 25, 1999 31 WNP-2
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ENCLOSURE 2 Date Type Sources Issue Descripfiou Codes GFNERAI. DESCRIPTION OF PIM TABI.F.LABELS Actual date of an event or signilicant issue for those items that have a clear date of occurrence, the date the source of the information was issued (such as the I.ER date), or, for inspection reports, the last date of the inspection period.
The categorization of the issue - see the Type Item Code table.
SAI,P Functional Area Codes: OPS I'or Operations; MAINTfor Maintenance; ENG for Engineering; and PS for Plant Support.
The document that contains the issue information: IR I'or NRC Inspection Report or LFR for Licensee Event Report.
Identification of who discovered issue: N for NRC; L for Licensee; or S for Self Identilying (events).
Details of the issue from the LER text or I'rom the IR Executive Summaries.
Template Codes - see table.
I.iccnsing MISC TYPE ITEM CODES Fnforcement Action Letter with CivilPenalty Fnforcement Discretion - No CivilPenalty Overall Strong I.icensee Performance Overall IVeak I.icensee Performance Facalated Fnforcement Item - IVaitingFinal NRC Action Violation Level I, II,III,or IV Non-Cited Violation Deviation from l,icensee Commitment to NRC Individual Good Inspection Finding Individual Poor Inspection Finding I.icensee Fvcnt Report to the NRC Unresolved Item from Inspection Report I.icensing Issue from NRR Miscellaneous - Emergency Preparedness Finding (EP),
Declared Emergency, Nonconformance Issue, etc.
TEMPLATECODES Operational Performance: A - Normal Operations; B - Operations During Transients; and C - Programs and Processes Material Condition: A - Fquipment Condition or B - Programs and Processes Human Performance: A - IVorkPerformance; B - Knowledge, Skills, and AbilitiesI Training; C - IVorkEnvironment Engineering/Design: A - Design; B - Engineering Support; C - Programs and Processes Problem Identification and Resolution: A - Identification; B - Analysis; and C-Resolution NOTES:
EEIs are apparent violations ofNRC requirements that are being considered for escalated enforcement action in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Action" (Enforcement Policy), NUREG-1600.
However, the NRC has not reached its final enforcement decision on the issues identilied by the EEIs and the PIM entries may be modilied when the final decisions are made.
Before the NRC makes its enforcement decision, the licensee willbe provided with an opportunity to either (I) respond to the apparent violation or (2) request a prcdecisional enforcement conference.
URIs are unresolved items about which more information is required to determine whether the issue in question is an acceptable item, a deviation, a nonconfornunce, or a violation. However, the NRC has not reached its final conclusions on the issues, and the PIM entries may be modilied when the linal conclusions are made.
ENCLOSURE 3 WASHINGTON NUCLEAR PLANT, UNIT2 INSPECTION PLAN 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 83750 OCCUPATIONALRADIATIONEXPOSURE IP 73753 INSERVICE INSPECTION RADIOACTIVEWASTE TREATMENT, AND EFFLUENT AND ENVIRONMENTALMONITORING EFFECTIVENESS OF LICENSEE PROCESS TO IDENTIFY,RESOLVE, AND PREVENT PROBLEMS IP 81700 PHYSICAL SECURITY PROGRAM IP 83750 OCCUPATIONALRADIATIONEXPOSURE 3/15 - 19/99 CORE INSPECTION 4/19 - 23/99 CORE INSPECTION 5/24 - 28/99 CORE INSPECTION REGIONAL INITIATIVETO EVALUATE IMPLEMENTATIONEFFECTIVENESS 7/19 - 23/99 CORE INSPECTION 7/1 2 - 16/99 CORE INSPECTION IP 93809 IP 37001 RADIOACTIVEWASTE TREATMENT, AND EFFLUENT AND ENVIRONMENTALMONITORING SAFETY SYSTEM ENGINEERING INSPECTION 10 CFR 50.59 SAFETY EVALUATIONPROGRAM 10/18 - 22/99 CORE INSPECTION CORE INSPECTION