IR 05000289/1997010
| ML20203L798 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 02/26/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20203L774 | List: |
| References | |
| 50-289-97-10, 50-320-97-03, 50-320-97-3, NUDOCS 9803060238 | |
| Download: ML20203L798 (103) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No and 50 320 License No DPR-50 and DPR 73
Repert No and 97 03
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Licensee: GPU Nuclear Corporation
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. Facility: Three Mile Island Station, Units 1 & 2 Location: P.O. Box 480
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Middletown, PA 17057 Dates: November 2,1997 - January 24,1 ~.98 inspectors: Wayne L. Schmidt, Senior Resident inspector Samuel L. Hansell, Resident inspector Lonny Eckert, Radiation Specialist Jason C. Jang, Sr. Radiation Specialist Edvwrd B. King, Physical Security inspector Paul H. Dissett, Sr. Operations Engineer 40500:
David M. Kern, Senior Resident inspector, Beaver Valley (lead)
Fred L. Bower, Resident inspector, Calvert Cliffs William B. Higgins, Reactor Engineer, DRS Theodore A. Easlick, Senior Resident inspector, Millstone-1 Approved by: Neil S. Perry, Acting Chief Reactor Projects Branch No. 7
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9803060238 980226 PDR ADOCK 05000289 G PDR
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EXECUTIVE SUMMARY Three Mile Island Nuclear Power Station Report Nos. 50 289/9710and 50-320/97-03 This integrated inspection included aspects of NRC inspections conducted at Three Mile Island Units 1 and 2 (TMI 1 and TMl 2) including operations, engineering, maintenance, and plant support. The report covers a 13 week period of resident inspection; in addition, it includes the results of announced inspections in the areas of: the licensed operator requalification program, corrective action prograrns and security for TMI 1, and radiological liquid and gaseous effluents control programs for TMI 1 and TMI 2 Papat_gatrJiti_9m Operators responded wellin the identification of the makeup (MUl system leak in November 1997, to the failure of the power range nuclear instrument (NI) -7 lower detector, and to a clogged strainer for the 'A' decay river (DR) cooling water pump (Section 01.1).
Operators conducted the unit shutdown, to repair the MU leak, well. The control of reactor coolant system (RCS)levelin the mid-loop range, to allow repairs to the B reactor coolant pump (RCP) seal, was appropriate, as was RCS refilling, and unit restart (Section 01.1 ).
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Tne routinely scheduled management meetings iopertnions and maintenance (6:30am),
maintenance (8:00arn), and plan of the day (3:00pm)) provided good platforms for discussion of current issues and future plans (Section 01.1).
l l The human performence review board (HPRB) meeting provided an objective assessrnent of the corrective action process (CAP). The meeting included the site Vice President, site directors, and key managers. The HPRB initiative demonstrated manegsment's support for the CAP process and expectations to improve the current program (Section 01.1).
GPUN conducted a leak test of the reactor building ernergency coolers (RBECs) without having a properly reviewed procedure in placo. The use of the inservice (IST) proceduro to close the reactor river (RR) cooling water RR-V-3 valves was an intent chango to the procedure which did not receive prior review or approval before implementation. Further, a detailed review of a procedure change to conduct this testing could have lead GPUN to question the appropriateness of closing these valves with respect to the their response to NRC Generic Letter (GL) 96-06, which assumed that all the coolers remained pressurize This is a Violation of Technical Specification (TS) 6.8.2 wnich requires that intent changes to procedures be reviewed prior to implementation (VIO 97-10-01)(Section O3.1).
The plant review group (PRG) thoroughly reviewed several plant issues including: failure of an Nilower detector, identification of a possible out of design basis issue with the RBECs, maintenance on the RR-V-4 valves in a hot shutdown condition, and the impacts of a MU leak while at power (Section 06.1).
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The inspectors found that the group offsite review board (GORB) conducted a very detailed meeting covering a broad range of topic areas, with specific issues in each ersa. Separate sub committee meetings were o good use of time and allowed more detailed questioning and probing of the TMI staff, prior to presentations to the entire GORB. Site management from engineering, operations, the PRG, and nuclear safety assessment (NSA) gave good proentations of current issues, wfilch the GORB property questioned and analyzed (Section 06.2). 4 GPUN's corrective actions to address the failure to notify the state and local counties within 15 minutes of the June 21,1997, Unusual Event were appropriate. The additional qualified communicators are expected to result in the availability of an extra control teom operator to assist the on-shift personnel after a transient or emergency condition (Section 08.1).
The GPUN response to the generic Westinghouse breaker seismic concern was comprehensive and effective. The licensee event report provided a detailed description and assessment of the event. The root causo analysis was thorough and the implementation of the associated corrective action was timely (Section 08.2).
Licensed Ooerator Reaualification GPUN implemented the licensed operator requalification training program weil for those areas reviewed by the inspector. Annuallicensed operator exarns were administered appropriatNy; operators performed well and management had extensive involvement in the exam process (Section 05.1).
. Corrective Action Proararn insoection Corrective action programs at TMI are improving. The new CAP, implemented in March 1997, has lowered problem identification threshold. However, problem resolution backlog has increased, assignment of action due dates and aggregate CAP timeliness assessment have been inconsistent, and CAP extension / escalation processes have not been consistently applied. There are multiple corrective action processes in use at.TMI, which i
at times, makes it difficult for management to track overall implimentation and to quantify needed resources. The inspectors noted the engineering task tracking system (ETTS) was not effectively used as an action tracking and resource planning tool. The corrective action processes were generally understood and properly used by station personnel (Section 07.1.1).
Performance in the area of problem resolution was mixed. The identification threshold for the corrective action processes was good. Resolution of failure to meet maintenance rule (MR) performance criteria for the high pressure injection (HPI) system was excellen Response to daily issues, including potential plant challenging problems was also goo However, the actions taken to resolve several larger issues were not properly managed due to higher station priorities and a perceived lack of resources by the plant staff. Poor initial problem scoping and unrealistic corrective action schedules made it difficult for the organization to correct problems in a timely manner. Several significant engineering problems such as previously identified service water cooling system deficiencies, iii
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calculation program controls, inservice testing program scope, and configuration control weaknesses remain only partially corrected in severalinstances, causal analyses were narrowly focused and may have fixed only the immediate problem, and not the root cause (Section 07.1.1).
Self assessment performance has been mixed. NSA audits and recommendations have been good. However,in severalinstances, the line organization f ailed to take sufficient action to preclude recurrence of problems. Recent reorganization and personnel reassignments to improve GPUN's ability to correct problems and preclude recurrence were not in place long enough to assess their effectiveness. The organization has been slow to formalize a line self assessment program (Section 07.2).
Operational experience (OE) prcgram performance was mixed with some signs of improvement. Internal and external oversight groups raised several concerns regarding ineffective use of industrv experience. Several actions to address those concerns were recently implemented. The OE and vendor documents departments typically notify the correct station personnel of potentialindustry concerns for resolution. Resolution thoroughness and response detail varied widely and in some instances were insufficient for reviewers to do an independent closeout. Inadequate detail and resource constraints resulted in poor assessment of a station blackout diesel generator issue. Senior management recognized that proper OE implementation requires more resources and intends to add one full time individual to help coordinate the OE program (Section 07.3).
Saf oty committee performance was mixed. The nuclear safety and compliance committee l (NSCC) identified important performance issues and effectively communicated them to the board of directors. The GORB also identified several underlying issues and initiated appropriate action items. However, GORB failed to recognize and act on some key issues (i.e. emergency preparedness, qualified component list reclassificaMon, and motor operated valve programs) indicating inconsistent performance by GORB subcommittees. Reasonable
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actions were proposed to improve GORB/GPUN interfaces and pcrformance. The PRG performance was mixed. An issue regarding dispositioning " failures to perform TS required
surveillances within the specified time interval" was identified as a non-cited violation.
l GPUN took actions to ensure that such TS violations would be reported via a licensee event report (LER) in the future, the inspectors considered tnis to be a violation of low safety consequence, in accordance with the NRC Enforecment Policy (NCV 97-10-02).
The independent on-site review group (IOSRG) identified and reported good issues, which were properly resolved (Section 07.5).
Maintenance The electrical maintenance technicians' performance, supervision oversight and coorFnation of the RR-V-4C/D preventive maintenance and MOVATS testing were excelk nt. The system engineer manager provided excellent support throughout the testing (Sectic a M1.1).
Sur eillance test performance was excellent. In particular, the reactor protection system (R' S) weekly tests were performed after the required documentation was revised to address the bypassed 'C' RPS logic channel (Section M1.1).
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GPUN used good tracking systems and monitoring of unavailability to identity systems, structures and components (SSCs) that needed additional attention under the NRC's M Further, the CAP process and routine meeting and planning appeared to allow identification for equipment problems and the trending and tracking of planned on-line maintenance effects on unavailability. Tracking, identification and planning for improvements on SSCs in the all) criteria appeared good (Section M2.1).
Enaineerina Overall, engineering provided good support for plant operations including: questioning the need for the emergency core cooling system (ECCS) flow instruments to meet system operability; review of the MU leak; development of a temporary modification to the RCP oil leak collection system to allow remote (outside the D ring) pumpdown; review of the NI 7 lower detector failure; and in support of the replacement of the B RCP seal package (Section E1).
Engineering preformed wellin identifying the issue of a difference between the GL 96-06 submittal and the updated final safety analysis report (UFSAR), and getting the issue to the site for PRG review (See Section O.6.1). However, this instance raised questions over the
' adequacy of the review and the understanding of possible plant impacts as a result of the GL 96-06 submittal. The inspectors considered this an Unresolved item pending review to determine if GPUN violated any NRC requirements with respect to the control of the RBEC design (Section E1.1).
The calculations and testing completed on the MU system injection piping for a small break loss of coolant accident (SBLOCA) adequatoly determined the minimum MU tank levels for NPSH. The testing proved actual system performance within the bounds of the calculations. However, the inspector found differences in the methodologies employed in system resistance modeling. The inspectors also found a weakness in the test procedure and the safety ova!uation (SE) since the as tested conditions did not match the calculation condition. The inspector determined that this did net invalidate the test results. However, GPUN committed to update the SE to include an evaluation of the test conditions and results versus the calculated condition (Section E2.1).
GPUN took comprehensive actions to resolve past qua!ity classification list (OCL) long term issues. The management involvement and oversight has resulted in an improved proces These actions included a revision to engineering procedure (EP)-01 ), " Methodology for Preparing the Quality Classification List," to formalize the process and included written detailed standards related to component and program changes. In addition, the procedure was added to the safety review program described by TS 6.5.1.12 and a safety evaluation was written to document the bases for the revision (Section E3.1).
An engineering safety evaluation written in 1993 addressed the potential safety issue associated with the high temperature effects on the concrete for the primary containmen However, the engineering processes did not result in a revision to the UFSAR to reflect the increased concrete temperature limit as noted in the safety evaluation (Section E8.1).
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Plant Sunoort The licensee maintained and implemented good routine radioactive liquid and gaseous effluent control programs. The radiation monitoring system (RMS) calibration program was good, as were the ventilation system surveillance program and quality assurance (QA) and quality control (OC) programs (Section R).
The licensee maintained an effective security program. Management support was evident based on the implementation of the security program. Audits were thorough and in-depth, alarm station operators were knowledgeable of their duties and respensibilities, communications requirements were being performed in accordance with the NRC approved physical security plan (the Plan) and assessment aids had good picture quality and excellent zone overlap. Effective contrcls were in place for identifyinJ, resolving and preventing programmatic problems and security training was being performed in accordance with the NRC approved training and qualification (T&O) plan (Section S).
Based on the inspector's observations and discussions with licensee and contractor engineering and security management, the inspector determined that the licensee's provisions for land vehicle control measures satisfy regulatory requirements and licensee commitments (Section S8),
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TABLE OF CONTENTS EX E C U TIV E S U M M A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 T A BL E O F C O N T E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii R e po rt D e t a il s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1. Operations ....................................................1 U1 Conduct of Operations (71707,92901) . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 G eneral Comm ents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 03 Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . 2 03.1 Leak Checking of Reactor Building Cooling Water Isolation Valves . 2 05 Operator Training and Qualification . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 Licensed Operator Requalification Program Review (71001) . . . . . 3 06 Operations Organization and Administration . . . . . . . . . . . . . . . . . . . , 5 l 06.1 Plant Review Group Meet!ngs . . . . . . . . . . . . . . . . . . . . . . . . . . 5
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06.2 Group Offsite Review Board December 1997 Meeting . . . . . . . . . 7 07 Quality Assurance in Operations (40500) ...................,7 07.1 Effectiveness of Corrective Action Progiams . . . . . . . . . . . . . . . . 7 07. Corrective Action Programs . . . . . . . . . . . . . . . . . . 7 07.2 Self. Assessment Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 07. Quality Assurance Audits . . . . . . . . . . . . . . . . . . 19 07.2.2 Self Assessments by Line Organitations . . . . . . . . . . . . , 21 07.2.3 Self Checking Program . . . . . . . . . . . . . . . . . . . . . . . . . 22 07.3 Operating Experience Feedback Program . . . . . . . . . . . . . . . . . 23 07.4 Saf ety Review Committees . . . . . . , . . . . . . . . . . . . . . . . . . . . 25 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 08.1 (Closed) VIO 50-289/97 06-01: Failure to Notify the State and County Offsite Agencies for the June 21 Unusual Event. . . . . . . 28-08.2 (Cbsed) LER 50-289/96-001-01: Seismic Qualifica'. ion of Class IE 4160 VAC Westinghouso Breakers .....................30 ll . M aint e n a nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 M1 Conduct of Maintenance (62707,61726) . . . . . . . . . . . . . . . . . . . . . . 30 M1.1 General Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 M2 Maintenance and Material Condition of Facilities and Equipment . . . . . 32 M2.1 Review of Maintenance Rule Implementation ..............32 111. E ng i n e e ri n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 El Conduct of Engineering (37551,92901) . . . . . . . . . . . . . . . . . . . . . . . 32 E1.1 Reactor Building Emergency Cooler - Generic Letter 96-06 . . . . . 33 E2 Engineering Support of Facilities and Equipment .................34 E Review of Makeup Pump Net Positive Suction Head Calculations . 34 E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . . . . . 37 E3.1 - (Closed)VIO 50-289/97-01-01:QCL Component Downgrade for Valves NR-V-1 A/B&C; Strainer Motors DR-S-1 A&B; Auxiliary vil
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i Ventilation Fans; and Make-up valve MU V-17; VIO 50-289/97 01-02: {
Failure to Follow the Procedure Requirements of EP-011; VIO 50-289/97-01 03: Failure to include EP-011 in the Safety Review Pr o c e s s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 l E Reactor Building Primary Shield Wall High Temperature . . . . . . . 40 1 E8.2 (Closed) LER 50 289/96-002-00/01: Potential Unreviewed Safety Question Related to the Net Positive Suction Head for the Decay Heat Removal and Building Spray Pumps. . . . . . . . . . . , . . . . . . . . . 41 IV. Plant Support ................................................42 R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 42 R1.1 Implementation of the Radioactive Liquid and Gaseous Effluent Control Progr am s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 R2.1 Calibration of Effluent / Process Radiation Monitoring Systems .. 42
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R2.2 Surveillance Tests for Air Cleaning and Ventilation Systems . . . . 43 R3 RP&C Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 44 R3.1 Radioactive Effluent Release Procedures . . . . . . . . . . . . . . . . . . 44 R3.2 Review of Annual Radioactive Effluent Reports . . . . . . . . . . . . . 45 R3.3 . Review of the ODCM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . 46 S1 Conduct of Security and Safeguards Activities (81700) . . . . . . , . . . 47 S2 Status of Security Facilities and Equipment (81700) .............. 48 S2.1 Protected Area Access Control of Vehicles . . . . . . . . . , . . . . . 48 S2.2 Alarm Stations, Communications and Assessment Ald3 ....... 48 SS Security and Safeguards Staff Training and Qualification (81700) . . . . . 49 S6 Security Organization and Administration (81700) . . . . . . . . . . . . . . . . 49 S7 Quality Assurance in Security and Safeguards Activities (81700) . . . . . 50 S7.1 Audits .........................................50 S7.2 Effectiveness of Management Controls . . . . . . . . . . . . . . . . . . . 51 S8 Miscellaneous Security and Safeguards issues ..................51 S8.1 Vehicle Barrier System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 S8.2 Bomb Blast Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2 S8.3 Procedural Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3 X1 Exit M e eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3 X2 Pre Decisional Enforcement Conference Summary . . . . . . . . . . . . . . . . 53 lNSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 lTEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 LI ST O F AC RO N YMS U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5 viii
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Report Detailt Summarv of Plant Status At the beginning of the period, TMI 1 was operating at 100% power. The Unit was shut down on November 3,1997, to repair a MU instrument leak and correct RCP oil leak and seal problems. The Unit was restarted on November 11, following repairs to the MU svstem and RCPs. At the end of the period the Unit was operating at 100% powe . Operations 01 Conduct of Operations (71707,92901)
01.1 General Comments Using inspection Procedure 71707," Plant Operations," the inspectors conducted frequent
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reviews of ongoing plant operations. In general, the conduct of operations was I professional and safety conscious; specific events and noteworthy observations are detailed in the sections below. Operators responded wellin the identification of the MU leak in November 1997 and to the failure of the power range NI -7 lower detecto The operators conducted the unit shutdown, to repair the MU leak, well. The control of RCS level in the mid-loop range, to allow repairs to the B RF seal was appropriate, as was RCS refilling and unit restar The routinely scheduled management meetings (operations and maintenance (6:30am),
maintenance (8:00am), and plan of the day (3:00pm)) provided good platforms for discussion of current issues and future plan The inspectors noted excellent plant operator response to a clogged strainer for the 'A' DR pump. Operator properly declared the pump inoperable when the strainer became clogged with leaves, small fish, and other debris. The stainer was cleaned and returned to servic Plant operators performed the DR system IST to prove the pump and system flows were adequate to meet the system design function. The increased amount of debris in the river water supply did not affect the other pumps in the intake structure. The traveling screens, spray wash pumps, and trash rakes were placed in service to remove the debris from the
' intake structur On December 3,1997, the HPRB met to assess the CAP. The meeting included the site Vice President, site directors, and key managers. Both past and future CAP program issues were discussed and included: the review of CAP data trends, improvements to the CAP program, future revisions to the CAP administrative procedure, and discussion of other facilities' CAP program strengths and weaknesses. The chairman led an objective discussion of the CAP program issues and concerns at TMI and resulted in the commitment to improve the current progra . . . . . - . . . . . . . . .
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03 Operations Procedures and Documentation O3.1 Leak Checking of Reactor Building Cooling Water Isolation Valvos: Scope The inspectors reviewed the actions taken by GPUN to address a slowly decreasing nuclear service closed cooling water (N3CCW) head tank level, since the reactor river water (RR)
cooling valves to the reactor building emergency cooler (RBEC) heat exchangers were cycled for emergency safeguards (ES) testing in November 1997. NSCCW provides positive pressure on the RBECs when they are in the standby mode, b. Observations and Findinas Plant operators torqued closed the RBEC outlet valves (RR-V-4s) to the maximum allowed values because the valves were the possible cause of the NSCCW leakage. This did not cause the leakage to decrease significantl Through discussion with engineering, the operations staff determined that they would
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conduct a pressure drop test on the coolers which are normally lined up with the inlet valves (RR-V 3s) open and the outlet valves closed. in order to conduct the test the l operators closed the inlet valves to all the coolers, and monitored the pressure drop. After l approximately seven hours the operators determined that one of tne two outlet valves, for the 'C' cooler were leaking. The operators considered closing the RBEC inlet valves appropriate action since the RR-V-3s and RR-V-4s receive an open signal following an accident, j The inspectors reviewed this test and determined that there was no specific procedure that
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allowed the valves to be closed to conduct a leak test. The operators used the normal inservice test (IST) surveillance procedure 1300 3K,"lST of Reactor River Water Pumps and Valves," to close the RBEC inlet and outlet valves for seven hours to conduct the leak test. The IST procedure was written and approved to determine the RR V-3 and RR-V-4 valve stroke open and closed times, but did not allow the valves to be closed for the seven hours. The inspector determined that use of this procedure to conduct the leak test was a substantive change to the intent of the procedure. Technical Specifications (TSs) section 6.8.2, " Procedures and Programs, " requires the review and approval of the procedure application prior to implementation. It appeared that GPUN operations did not recognize that a procedure should be used only for the approved intended purpos Subsequent to the testing the engineering department determined, based on a GPUN submittal addressing NRC GL 96-06 and on calculations, that closing of the RR-V3 valves which isolated the coolers from NSCCW pressure, placed the coolers in an unanalyzed condition. Specifically, engineering did no+ a a calculation that proved that voiding and subsequent cooler damage would not oc the event of a loss of coolant accident (LOCA), if the coolers were not pressurize.a fan the NSCCW sourc _ _ _ _ _ - _
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3 Conclusions The inspectors found that GPUN conducted a test without having a properly reviewed procedure in place. The use of the IST procedure to close the RR V-3 valves was an intent change to the procedure which did not get reviewed. Further, a detailed review of a procedure change to conduct this testing could have lead GPUN to question the appropriateness of closing these valves with respect to the GL 96-06 response, which assumed that all the coolors remained pressurize The inspectors considered this a violation of TS 6.5.1.1 and 6.8.2 which require that intent changes to procedures be reviewed and approved prior to imNementation. (Violation 97-10-01)
05 Operator Training and Qualification 0 Licensed Operator Requalification Program Review (71001) Scope A scheduled inspection of the Thrce Mile Island, Unit 1 licensed operator requalification program was conducted from January 20 23,1998.The inspector reviewed portions of the Three Mile Island licensed operator reaualification training program to ensure that a continuous two year training orogram was in place and had been implemented as required by 10 CFR 55.59. The scope of the inspection included the observation of the annual operating exams administered to one operating crew of licensed operators, and the review of previously completed biennial written exams, remedial actions taken for past exam failures, and the administration of a continuing requalification training program, Observations and Findinos The inspector reviewed and witnessed the administration of the annual operating examinations to one crew of licensed operators. These observations included the conduct of three simulator scenarios and several job performance measures (JPMs) that were administered during the week. Facility evaluations were also reviewed for scenario and JPM performance by licensed operator The inspector reviewed and witnessed the performance of several JPMs that were administered during the annual exam. This review included the performance of JPMs in both the simulator and in the plant. The JPMs were relevant to operator tasks, were consistently administered by different evaluators, were technically sufficient to discriminate operator abilities, and were appropriately evaluated to identify weaknesses in performanc Three simulator scenarios, that were given to one operating crew, were also reviewe The scenarios were challenging and met the criteria set forth in the examiner standard The scenarios were diverse and utilized various abnormal and emergency operating procedures. Also, critical tasks were correctly identified within each scenario. The inspector observed operator performance during the conduct of these three simulator scenarios. Crew and individual operator performance was good. Communications, for the
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l most part, were very good, but varied from scenario to scenario. The inspector attributed I this variation to a shifting of personnel from one position to another, or into and out of the crew composition. Facility evaluators stated that they are continuously stressing three-point communication and subsequently documented weaknesses in this area following scenario debriefs. Individual and crew performance was appropriately evaluated by operations and training evaluators. Allindividuals and the crew passed all three simulator scenarios. All critical tasks were accomplished. Individual competencies were also rated acceptable in allinstances. Minor weaknesses were noted and were adequately documented for later dissemination to the crew and individual All individuals, as observed by the inspector, passed the JPM portion of their annual operating exams. Two individuals f ailed one of their JPMs, but passed overall. The in-plant JPMs were administered by operations supervisory personnel, which indicated that management is activcly involved in the examination proces The evaluations by training and operations department evaluators were appropriate for those portions of the exam observed by the inspector. Documentation was adequate as demonstrated by their review of the previous year's exam performance prior to the start of
the simulator examination. This information was reviewed by the evaluators in an effort to alert them to previously identified individual and/or crew weaknesses or shortcoming The inspector assessed the adequacy and effectiveness of remedial training conducted during the examination cycle, including training provided to licensed operators to correct deficiencies that resulted in a failure of their annual operating or biennial written exam, in this instance, the inspector reviewed the failure of a written exam for one individual and a failure of the operating exam for one other individual. Documentation of remediation included a review of areas of weakness with the individuals, additional classroom or simulator exercises, self-study and a retake of another completely different exam. In both instances, the individuals passed their retake examinatio The inspector reviewed several LERs that occurred in 1996 and 1997 in an effort to determine if any of the events were a result of inadequate training. The LERs reviewed did not indicate any deficiencies in the knowledge level of individuals or inadequate training provided by the training department. The inspector also reviewed the process by which TMl incorporated industry events into their training program. Instances were noted where scheduled training was held to discuss various industry operating events with licensed operators during regularly scheduled requalification trainin Conclusions The TMl training department had implemented a continuing licensed operator training program that met administrative and regulatory requirements. They had developed annual licensed operator requalification exams that effectively tested the knowledge and ability level of alllicensed operators in an effort to maintain continued safe operations at the facility. The annuallicensed operator requalification exams were administered and evaluated acceptably. Operator performance was acceptable during both the simulator and JPM portions of the operating exam. Appropriate action had been taken in regard to those individuals who had failed any portion of their annuallicensed operator exam. Remediation
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appeared effective, as evidenced by retake examination results. Operations and training personnel have continued to work together in an effort to ensure that operator knowledge j and ability is maintained at a high level of performance. Upper management involvement was noted throughout the exam proces Operations Organization and Administration 1 06.1 Plant Review Group Meetings The inspector attended numerous PRG meetings over the period and assessed the adequacy of the discussions and final review. The PRG reviewed several plant issues including:
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The failure of the Nl-7 lower detector and the associated operability of the reactor protection syste *
The need to have ECCS flow instruments operable, as part of the system operability requiremen *
Reportability of isolation of the RR system from the NSCCW overpressure and the affects on system operabilit *
RR V4 maintenance while in hot shutdow *
Review of operability and reportability of the MU system following pressure boundary leakage, Observations and Findinos The PRG performed wellin review of the plant issues, as follows:
Following the NI-7 lower detector failure, PRG conducted a very good review of the TS, the UFSAR and a B&W owners group position on the ability to operate with one channel of the RPS in bypass, for an extended period of tim The inspectors independently reviewed this issue and concluded that the RPS logic design supported the GPUN position that it was acceptable to operate with one channel in bypass. Specifically the other three channels in a 2 out of four logic provided adequate redundancy to ensure a reactor trip even with one other single failur *
Based on good questioning of the system engineer the PRG changed a previous -
position that the ECCS flow instruments were not needed to be functional to support ECCS system operability. The material presented to the PRG was clear and provided a good basis for changing this position. The PRG also conducted a review of past times when the flow instruments had been inoperable for longer than the TS allowed time without the system being declared inoperable. The PRG determined that there were several cases where the LCO for the systems had been exceeded based on the instruments being inoperable and determined that an LER would be writte _____
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in review of this issue the inspectors concluded that the instruments did need to be operable to support system operability, particuledy since throttling was now being relied upon to ensure adequate net positive suction head (NPSH).
PRG completed a detailed review of RBEC operability following identification by engineering that the operations department isolation of the RBEC from NCCSW may have placed the unit in a condition where the coolers had not been reviewed in their response to a design basis accident (DBA) (See Section E.1.1.) The PRG utilized NRC GL 91 18 and GL 91-18 REV1 to review this degraded plant equipment condition. The review of this issue was methodical, the PRG determined that the RR system was placed in a condition where engineering did not have calculations to verify the lack of volding in the coolers following a LOCA withoui'..a NSCCW
pressurization. However, the engineering department stated that based on engineering judgement they believed that calculations would prove that voiding l would not occur even without the over pressure.
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l The PRG determined that based on this engineenng judgement the RR system was not known to specifically to have been outside the design basis, pending completien of the additional calculation. The PRG recommended to plant management that the additional calculations to prove operability be completed within a reasonable time and that a report to the NRC would be necessary if they did not prove operability in the isolated condition. Further, it was identified that several surveillance tests would place the coolers in a condition where either the intet valves would be closed
_ _ or the NSCCW pressure would be otherwise isolated from the RBECs. The PRG determined that until the condition could be reviewed by calculations the operations
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department should take action to ensure that toe coolers are not isolated from the NSCCW syste GPUN engineering subsequently determined by calculation that the NSCCW over pressure was not needed to prevent voiding in the RBEC cooler following a DBA LOC *
The PRG properly reviewed a possible planned maintenance activity which would have allowed the breaching of the RR system to fix the RR-V-4 C and/or D RBEC isolation valves while in a hot shutdown condition. The PRG had adequate technical data available and used good safety judgement to determine that the work should not be conducted while primary containment was still necessary. Specifically, PRG recommended that the work be conducted with the unit in a cold shutdown-condition, when primary containment is not neede *
Following the identification of the MU system leak, fro n the RCP sealinjection flow transmitter sensing line, the PRG completed a good review of the TS to determine the applicable sections and whether plant operation could continue and the need for reportabilit ~
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7 Conclusions The PRG thoroughly reviewed several plant issues including: failure of a Ni lower detector,
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identification of a possible out of design basis issue wich the RBECs, maintenance on the RR V-4 valvea in a hot shutdown condition, and the impacts of a MU leak while at powe .2 Group Offsite Review Board December 1997 Meeting Scope
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The inspector attended the December 1997 meeting of the GORB to assess the prosentations and interactions with site staff, Conclusions The inspectors found that the GORB conducted a very detailed meeting covering a broad range of topic areas, with specific issues in each area. The inspectors found that the use of sub-committees that meet separately before the entire board was a good use of time and allowed more detailed questioning and probing of the TMI staff, prior to presentations to the entire GORB. Site management from engineering, operations, the PRG, and NSA gave good presentations of current issues, which the GORB properly questioned and analyze Guality Assurance in Operations (40500)
0 Effectiveness of Corrective Action Programs The inspectors reviewed various documents, interviewed licensee personnel, toured the plant and observed licensee activities to evaluate the effectiveness of licensee controls in identifying, resolving, and preventing issues that deprade the quality of plant operations or safety. Licensee controls were assessed in five specific performance areas (1) Corrective Action Programs and Resolution of Problems (2) Self Assessment Activities, (3) Operating Experience Feedback, and (4) Safety Review Committees as described in the following sections. The period of June 1996 through November 1997 was evaluate . Corrective Action Programs Scope 4 Review of the Corrective Action Processes:
The inspectors rp ,lewed the effectiveness of GPUN's revised CAP, put in place in March 1997, for prov'. ding a method of documenting and resolving significant conditions adverse
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to quality, in a timely manner. As part of the assessment the inspectors reviewed the:
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Action Process,
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Threshold for opening CAPS and the tracking and timeliness of closing issues including establishing response dates and action due dates and the control of extensions and escalation *
.stus of open issues in other GPUN corrective action systems, interviews were conducted with selected individuals involved with the licensee's problem identification process to determine the extent of their understanding of the process and willingness to report problem Adequacy of Problem Flesolution The inspectors assessed the effectiveness of GPUN's implementation of their corrective action processes including:
initialidentification and characterization of the problems;
elevation of problems to the proper level of management;
- root-cause evaluation (RCE);
implementation of corrective actions including evaluation of repetitive conditions;
the roles played by the quality verification organizations and line management in the identification and resolution of issue The inspectors conducted a detailed review of control room (CR) logs to ensure that equipment problems were being identified and properly placed in corrective action system To assess the adequacy of causal analysis and corrective actions the inspector conducted detailed reviews of selected CAPS, quality deficiency reports (ODRs), and open self-assessment issues. The inspectors also reviewed portions of GPUN's implementation of the MR and their ability to identify needed corrective actions based on increased out of service time Observations and Findinas New Corrective Action Program Process GPUN made several key improvements to the CAP, including:
. Simplification of the priority selection process, immediate priority required the accountable manager to apply as many resources as necessary to achieve completion as soon as possible. Urgent priority is required to be completed by the date assigned at the ac~ 3 Jntability review (AR) meeting, normally within seven days. A Routine priority CAP response due date shall not exceed 30 days from CAP initiatio *
The inclusion of reportaUlity and operability significance determination *
A formal expectation on deciding which level of RCE was appropriate for the event / conditio .
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- Daily management reviews of newly generated CAPS. An early management review (EMR) was added to the 6:30 am " Operations and Maintenance Status" meeting and the AR was odded to the 3:00 pm " Plan of the Day" meeting. At the EMR, operability and reportability determinations were reviewed, priorities and significance were determined, and the appropriate resolution processes were selected. During the AR, the CAP forms were reviewed with any changes in accountability, priority, significance, or due dates, discussed and agreed to by the management tea *
CAP ;s a centralized method of controlling and receiving adverse conditions. Under the current CAP, once a condition is identified on a CAP form, the management team selects one or more of the following processes to resolve a problem: LER, QDR, material nonconformance report (MNOR), work request (WR), RCE, maintenance trend action notices (MTAN), trend, none, or other. The ODRs ard MNCRs processes also have additional requirements under the QA Plan for timeliness, concurrence, corrective action verification, and effectiveness review Adverse conditions can also be identified as output of any of the following processer, MTAN, ODR, or MNCR, which are then conver+ed to CAP forms and processed via the CAP syste The inspector determined that:
The process appeared better than the previous syste *
Based on attendance at several EMR and AR meetings, CAPS were appropriately discussed by plant manager *
GPUN ured many duplicative methods for identifying and correcting condition This duplication of effort will continue until the next revision of the CAP, which is intended to create a single low threshold /high volume corrective action reporting system. The licensee expects to implement this change in March 1 ?9 The inspector noted that the CAP had changed since its implementation as the staff became more familiar with the process. At the time of the inspection, several aspects of the CAP were not proceduralized, but were proactive and being conducted by the CAP coordinator, for examole-
Review and approval, of resolution plans and directs the actions needed if the response is incomplet *
Closure process ar.d final review to verify adequacy and consistency of actions with the original problem *
Corrective action effectiveness review, known as the "Re-visit" process. The CAP coordinator had selected closcout packages for effectiveness reviews 3-6 months late .
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Additionally, a number of required data entry points listed on the CAP form were not described in the procedure. The CAP coordinator acknowledged that the CAP process continues to evolve and plans to include all of the issues identified above in the March 1998 CAP revisio Tht9shold and Tracking of issues The inepoctors teviewed the administrative controls that were established to monitor the tieneliness of the CAP, and found that:
- For a response an accountable manager may extend the response due date for up to 30 additinnel days. Extensions past that time need to be reviewed and approved by the AA grou * Once the response is received and reviewed by the CAP coordinator, a corrective actbn due date is established commensurate with the significance of the conditio * A corrective action due date can be up to 30 days past due before it is considered late. Normally, the deficiency is escalated to tha next higher management le tel, who within 10 working days shall provide a supplemental response or request the establishment of a revised due date at one of the daily AR meetings, in review of those processes, the inspectors de ;rnined:
- The problem identification threshold has been lowered through implementation of the new CAP process. Currently,873 CAPS have been identified since March 1997, indicating a reduced problem identification threshold. Approximately 300 CAFs remained open, with an additional 60 CAPS ready for closure. The inspector noted that the backlog was increasin * Under the present CAP, it appeared that this was a large workload for the two CAP coordinators to manag *
The inspectors noted that the 30 day grace period permitted beyond the assigned i :ective action date before declaring an action late appeared excessive. Thirty-one CAP corrective actions were lats (e.g. over 30 dayn past due) dunng the inspection. The CAP coordinator acknowledged that CAP corrective action escalation warranted increased attentio *
Discussions with the CAP coordinators indicated that there were no tools available for assessing the CAP extension or escalation processes. In addition, there was no method of tracking the number of extensions for the CAPS on a programmatic leve Each individual CAP would need to be reviewed to determine the number of extensions and escalations it had accumulate = = _ = =_=
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- The escalation process, Instituted in September 1997, was not actually !
Implemented and enforced until early November. The escalation proceso has not l
been consistently used. Additionally, management ha not t 'en trained on the escalation proces * Several CAPS had no due dates assigned (three C APs had no response due dates and twelve CAPS had no corrective action duc dates). Without the required due dates assigned in the system, these CAPS would never be Identified as late. The CAP coordinator Informed tne inspectors that due dates have been requested from the respotisible managers and was awaiting responses, i
- With respect to the ODR process established under the QA Plan, the inspector found that the controls over the QDR extensions process (different from the CAP program) were good. There was visible tracking methodology, with reasonable use of waiver and escalation tool The inspector found that GPUN had not properly used the computer based ETTS, to track origineering workload and corrective action item due dates. A review of the ETTS database indicated that:
- The data was incomplete (i.e., severalitems had no due dates assigned). When due dates are not established, there is no overdue notice generated, and therefore, no ability to verify timelir.ess. Since ETTS is connscted to the CAP system, the blank due dated will also appear in the CA * Some engineering personnel did not understand how ETTS action items were communicated to the CAP system, and therefore, did not understand the impac * Several engineers indicated their computers d!d not have the ability to interface with ETT * Managcment's expectations for the use of ETTS is currently being transmitted verbally to the engineering staff, since thero is no written guidance for ETTS usag * A work duration field was recently added to the database to support resource plannin The licensee intends to use ETTS for engineering workload management, and is currently revising the system to support this purpose. Engineering management plans to ensure that all engineering work is entered into ETTS by the end of 1997. The inspectors determined that the usefulness of ETTS was limited without uniform guidance or instructions,
. personnel training, and sufficient computer equipmen Other Corrective Action Systems
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The inspectors found that GPUN had numerous tracking systems in place that co .1d ;
used to effect corrective actions, that were outside the 10 CFR Appendix B OuaL"r Control Program, appearing to be used in place of the CAP system for some issucs. Specifically:
Licensing action requests (LARs), per GPUN Regulatory Affairs Instruction 94 01,
" Licensing Information Tracking System," used to process and control regulatory correspondence and commitment information inspectors noted that 18 service water system operation performance inspection (SWSOPI) items were documented u LAns and were not in the CAP program, with no control over actual completion of those corrective actions. Discussions with the Licensing staff indicated that this issue was already identified by QA and that Licensing was tasked with reviewing the LAR database to convert corrective action items to the CAP. A QA audit report dated October 10,1997, documented that several deficiencies, from the motor operated valve (MOV) Independent Review Team Assessment at Oyster Creek, were documented by LARs, which was not an approved GPUN corrective action syste As a result of that audit, the TMI Licensing staff was tasked with reviewing their LAR proces *
The surveillance deficiency report (SDR) process used to document problems i
' encountered during surveillance testing, while adequate, was not a recognized corrective action system. All surveillance problems are evaluated by the shif t supervisor for operability /reportability and SDR sheets are submitted with the completed surveillance package. Once an item is documented as an SDR, corrective actions are determined and assigned. The text field in the SDR database is used to track corrective actions, as well as document safety evaluations for ( specific issues. SDRs were trended to improve program performanc Th3 licensee informed the inspectors that the LARs, SDR and approximately 31 sther problem reporting systems (some of which are not proceduralized) are being considered for consolidation into the new CAP progra Interviews:
The inspectors interviewed numerous sits personnel to assess the effectiveness of training and the understanding of the CAP process, and found that:
Prior to implementa.lon of the AP 1080, formal training presentations were provided j to the operations department for operability /reportability, the management teams for EMR and AR, and to all station personnelin the form of departmental briefings. A review of the attendance records for the formal training indicated that the appropriate personnel had received the requisite trainin I
The operations, maintenance, cnd engineering staff had a good working knowledge of the pro::ess and procedure, and were comfortable with the level of training they receive _ _ _ _ _ _ _ _
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Interviewees understood when to initiate a WR or a CAP. There was a clear difference between use of WRs and CAPS, with WRs used for repair of equipment, and CAP forms used for issues requiring root cause determination * The interviewees that had initiated a CAP form, received feedback on the CAP for both its initial resolution, and the corrective action that was take * The staff felt that the new CAP had good management support, and there was no indication of any reluctance to raise concern * All of the staff interviewed, concidered the CAP to be a good program, which provided a single point for problem identification that did not previously exist at TMI.
- The staff also understood why the corrective action process was changed and were clear on management's expectations for the use rf the new CAP form Operator Logkeeping and Operator Identification of Day to Day Problems The inspectors reviewed CR and primary Auxiliary Building (AB) operator logs for the period April to Oct ber 1997. Out of specification (OOS) conditions were typically identified and addressed in a timely manner using appropriate corrective action processes such as work requests or CAPS. However, tha inspectors noted early indications of lookeeping f complacency. About 30 percent of the OOS readings were longstanding conditions, and many were not circled in red to draw attention to this condition. Examples included:
- 'B' reclaimed boric acid storage tank level
- Decay heat removal system (DH) P 1B pump and motor bearing remote oiler levels (Ll 1080,1077,1078)
- Waste gas delay tank drain pot level
- Heat trace circuits (panels 3A 1/3A 2/3B 1/3B 2 circuit 12 on each panel)
Auxiliary sump room access hatch lid position The inspectors discussed this observation with operations department management who stated that a program to reevaluate the normal parameter band for several of these items had recently been initiated. They believed the existing conditions were acceptable and that the log parameter specifications should be revised. However, untillogs revisions are implemented, management expected operators to red circ!a OOS readings and initiate actions as appropriate to correct the condition. Operations management stated they would reemphasize this expectation with the operators. The inspectors determined that this action was appropriat . _ _ _ _ _ _ _ _ _ _ _ _ _
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The 1D RCP upper oil reservoir lo eel indicated leakage following startup from the recent refueling outage. The inspectors reviewed operator logs, trend charts, and discussed monitoring and contingency actions with the shift supervisor (SS). The inspectors concluded that operators demonstrated good assessment, trending, and contingency planning in response to the leaking 1D RCP upper oil reservoi Resolution of CAPS The inspectors portormed a review of approximately 80 CAP forms in detail. The quality of resolution plans were generally good. The timeliness of assigned dates and completed actions were good with the exception of several SWSOPiltems as detailed belo However, the inspectors identified several causal analyses documented on CAP forms that were narrowly focused and may have corrected only the immediate problem and not the root cause. For example:
CAP T1997 0072,The breaker for 2AH C 1100 (station blackout diesel generator)
was found closed and could not be opened with the handle. The RCE found that
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the MCC PM was terminated and resources diverted from the late 1970's to l' 1990's. The PM was re established without evaluating other electrical components that may have been affected by the reductions in PM performanc *
CAP T1997 0033,Eight RPS and turbine trip limit switches were found out of tolerance low, by about 150 psi or greater, resulting in inoperable P.PS trip functions. The RCE concluded that incorrect switches were installed. The RCE did not consider the extent of the condition or the potential for incorrect setpoint calibration practice * CAP T1997 0701, Asbestos and contaminated material was found in clean wast An RCE was assigned, but was not performed because the evaluator could not determine how the waste got put in the wrong location. "An RCE could not be performed since no trail of eve,nts/ actions could be determined."
Resolution of Quality Defiriency Reports The inspectors reviewed 13 open ODRs and determined that resolution plans submitted in response to th: OA findings were timely, and within the required 30 days, However, ott 13 ODRs missed their initial corrective action completion dates and required extensions and/or escalations from the nuclear safety assessment (NSA) department. Additionally, several ODRs for larger issues had poor initial problem scoping and u 1 realistic corrective action completion schedules. The poor initial work performed to identify problem scope and schedule wasted resources and inade it difficult for the organization to correct identified problems in a timely manner. For examp'e:
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- qDR 902009," Inadequacies identified in Chapter 8 of UFSAR indicative of total UFSAR problem." The ODR was issued on July 10,1990. The response included an engineering review of the whole UFSAR to determine other discrepancies, and a plan to complete the work by end of November 1990. This was an unrealistic schedule considering the amount of work that would be required to complete the review. OA approved extensions and walved escalations as the scope of the review increased. The current completion due date is December 31,199 * ODR 902021,"Self initled ODR to identify that MU V 17, as described in the UFSAR,is needed to hdgau wnsequences of an accidenti therefore, required to be in the IST program." p G1.9 was issued on October 29,1990, and was scheduled to be completed on Jay 7,1997. The licensee's review indicated that an additional 150 valves would be affected. Based on the required resources to l
complete the work, the current cotopletion due date is March 3,199 *
ODR 972001," The current process provided in EP-000, " Calculations" is not effective in identifying and updating the calculation of record." The ODR was issued on January 8,1997 and was related to two ODRs previously issued in 1990. The original response included a two phase closecut process with completion dates in May and September of 1997. Both of the completion dates were missed, as well as the additional extensions approved by NSA. Poor initial scoping and missed completion dates have resulted in the current completion due date of December 31, 1997. The inspectors noted that 10 of 20 action items had not been assigned due dates, even though this ODR had been escalated to the Vice President of Engineering. Discussions with NSA management indicated that it is not likely that the December 1997, date will be me *
ODR 972007," Performance history indicated that resolution of deficiencies by Engineering is not timcly and NSA's implementation of GPUN's corrective action program was not effective " The ODR was initiated on February 24,1997. The interim correction actions wsre scheduled for April 14,1997, and corrective action scheduled to be completed on July 18,1997. The interim actions were not completed on time and in May 1997, the decision was made to include the required actions resulting from the Engineering and Corrective Action Processes Assessment Team. The current completion due date is December 31,199 *
ODR 972010," Configuration Control Program weaknessec" identified problems including, procedural weaknesses, procedure conflicts, and procedure noncompliance that included f ailure to perform and inadequate design verifications and safety evaluations. Corrective actions were identified within the required 30 day period. Interim corrective actions were identified and completed. Two extensions were required and the due date was missed. All actions were completed within approximately 7 months, but missed the due dat _ . - - -..- - - - - . - . . . - . - - - - - _ - - - - .
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- Audit Finding S TMl 9511,(finding 3 of 4): Engineers f ailed to follow procedures for developing modifications and did not provide critical attributes including:
setpoint, calculations, test requirements, insulating requirements, etc. for a radwaste modification. Corrective actions were provided within 32 days; escalation was required twice; and closeout required two year Service Water System Operation Performance Inspection Corrective Actions
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The inspectors reviewed a selected sample of findings from the SWSOPl self assessment to determine the status and adequacy of corrective actions. The inspectors also reviewed the systems reports for the river water, closed cooling, and emergency core cooling
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systems and the status of the SWSOPl and microbiologically induced corrosion (MIC)
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The SWSOPl self assessment was conducted in April 1995. The inspector found that, during the self assessment, the SWSOPl findings were initially entered into the LAR system. The inspectors determined that the LAR system was a task tracking system that was not considered part of the corrective action program at TMI. Closure of the LAR ltem was allowed based on identifying corrective actions and approval of a special review tea Completion of the corrective actions was not required to gain approval.
A February 19,1997, memo from the SWSOPl program engineer provided a status of the 128 specific action items frorn the self assessment. A similar more recent status of the SWSOPlitems was not available at the time of this inspection. The inspectors determined that one or more action item could be associated with a LAR. Of the 128 action items,85 were completed and closed,13 were completed and under review, and 30 remained ope The inspectors noted that, at that time, approximately two thirds of the open action items did not have due dates establishe Although no immediate safety or operability issues were identified, the inspectors concluded that the resolution of SWSOPlissues and the development of SWSOPl and MIC programs was slow. The perceived lack of resources and other higher priority (plant shutdown) items appeared to have caused delay The inspectors found the 5 LAR items selected for detailed review were still open in the LAR system. A commitment in the TMI response to the NRC 50.54(f) letter on design review indicated that the SWSOPIitems would be entered into a corrective action tracking system. The open LAR ltems were subsequently entered into the CAP system. The ETTS was identified as the system for tracking the closure of the corrective actions identified for these LAR items. The ETTS was electronically linked to the CAP system, but was not considered a corrective action program. The inspectors determined the following status for the selected items:
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- LAR 95034.25: New DR cooling water system total hydraulic ca'culetions and decay heat closed (DC) cooling water system total hydraulics and NPSH calculations were required. Completion of calculations took approximately one year. Design verification of the new calculations appeared to remain outstanding in the LAR system with approximately 7 extensions of due dates for each. The tracking item had not been accepted by the responsible individualin ETTS and no due date was assigned. Upon further review, GPUN determined that the calculations have been completed, but have not been reviewed by the special team for closeout. Further, nothing has boon done to try to update the ETTS or close the related CAP or LA No clear corrective action due date was established in the CA * LAR 95034.20: New nuclear river (NR) cooling water system calculations were required to consider all operating modes. The calculations did not appear to be complete. Due dates for the calculations have been extended 7 times in the LAR system. The tracking item was not accepted by the responsible individualin ETTS and no due date was assigned. The ETTS item stated that there was a flaw in the calculation for not considering the screens. However, upon further review, GPUN found that the calculation was completed, but has not been reviewed by the special review team for closeout. Further, nothing had been done to try to update the ETTS or close the related CAP or LAR. No clear corrective action due date was established in the CA * LAR 95034.14: New calculations were required to determine the heat exchanger duty and maximum acceptable blockage of the nuclear service (NS), DC, & RBEC heat exchangers. The LAR indicated that the evaluations had been done but the documentation and design verification were in various stages of completion. The tracking item was accepted in ETTS and a due date was assigned. Upon further review, GPUN found that the calculation had been completed, but had not been reviewed by the special team for closeout. The status had not been updated in the ETTS, CAP and LAR systems. No clear corrective action due date was established in the CA * LAR 95034.15: Established plan for evaluating the options for conducting heat exchanger performance testing for the NS, DC, & RBEC heat exchangers. The item for the DC heat exchanger was closed, but items for NS and RBEC had been extended 7 times in the L AR tracking system. The item was accepted in ETTS but no clear due date was established in the CAPS and ETTS system * LAR 95034.23: Single f ai'ure analyses were not documented as completed for the DR/DC and RR systems. This item was originally due in December 1995 and was extended 8 times in the LAR system. No clear due date was assigned in the CAP system. Additionally, the tracking item was not accepted in ETTS, and no due date was assigned. However, comments in the ETTS system indicate that the item was planned for completion in April 1998. The updated information was not entered into the LAR or CAP system _ _ . .
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The inspectors did not identify any immediate st.fety or operability issues. The inspectors reviewed examples where interim operability determinations had been performed when determined necessary by GPUN and the SWSOPl team. However, the inspactors questioned why the follow up of SWSOPlitems appeared to be slow. GPUN personnel responded that many of the outstanding corrective actions were in thw thermal hydraulics area where there was a shortage of qualified resources. Further discussions indicated that the perceived lack of resources and higher station priorities app!Ied more generally to SWSOPl item The inspectors reviewed additional status items, the river water system engineer's reports, and discussed these status items with the recently appointed SWSOPl program coordinator. The coordinator was the third individual appointed to coordinate SWSOPl items and is also responsible for the MIC program. To date, no program had been
! developed for maintaining the SWSOPl program, but action to develop such a program had
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been assigned. A MIC program had been drafted and was under peer and management l
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review. The system engineer's reports and discussions indicated that MIC related program <
efforts were performed in 1995 and early 1996. However, the program was on hold for portions of 1996 and 1997. Inspections and sampling were restarted in mid 1997. The majority of these samples remain to be analyzed. However, the individual assigned to coordinate the MIC and SWSOPl programs was also working full time as an on shif t technical advisor (STA) until the refueling outage was completed in November 199 Maintenance Rule Equipment Assessment and Problem Resolution in February 1997, during a quarterly equipment performance assessment, the MR ccordinator identified that HPl system availability performance criteria was incorrectly j entered into the monitoring database. Upon correcting the criteria entry, he noted HPl unavailability exceeded the corrected criteria (1.5 percent) and placed HPl in category a(1).
This categorization required increased monitoring and associated corrective actions which were initiated under MTAN #96 26. The licensee identified maintenance and testing ;
scheduling inefficiencies which caused HPl unavailability to be higher than necessary, 1 Scheduling coordination was revised. The inspectors verified that this has resulted in improved HPI availability. The MR coordinator also established a two tiered escalation process to require additional approvals for maintenance which would result in a MR SSCs exceeding 50 percent, and 75 percent of the performance criteria value. The extent of condition review, performed for MTAN #96 26, identified 1 additional system (ultimate heat sink) where similar scheduling improvements could be achieved. The inspectors concluded that the licensee response to HPl exceeding MR unavailability criteria was excellen Conclusiong The corrective selon programs at TMI are improving. The new CAP, implemented in March 1997, has lowered problem identification threshold. However, CAP problem resolution backlog has increased, assignment of action due dates and aggregate timeliness assessment have been inconsistent, and the extension / escalation processes have not been consistently applied. There were multiple corrective action processes in use at TMI, which at times, makes it difficult for management to track overall implamentation and to quantify
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needed resources. The inspectors ruted the ETTS was not used effectively as an action tracking and resource planning tool. Station personnel generally understood ar.d properly i used the corrective action processes. The inspectors noted that findings from an NSt. audit of tha CAP ptcoram, as well as current recommendations and program weaknesses were scheduled to be addressed in a March 1998, program revisio The inspectors concluded that the licensee's performance in the area of problem resolution we mixed. The idemification threshold for the CAP was good. Resolution of failure to rneet MR performance criteria for the HPl system was excellent. Response to daily issues, including potential plant challenging problems was also good. However, the actions taken to resoin aaverallarger issues were not properly managed due to higher station priorities -
and a porceived lack of resources by tt,s plant staff. Poor initial problem scoping and unrealistic corrnctive action schedules ruede it ditlicult for the organization to correct prob (ems in a tintely manner. Several significant engineering problems such as SWSOPi deficient.ies, calculaticn program controls, inservice testing program scope, and configurction control weaknesses remain only partially corrected, in severalinstances, i
' causal analyses were narrowly focused and may heve fixeri only the immediate problem, and not the root caus .2 Sell-Asseament ActMtles 07. Quality Assurance Audits Scope The inspectors reviewed the scope and frequency of periodic TS required NSA audits and the offectiveness of the use of audit findings by the line organization to correct issue The inspectors also reviewed the routine NSA annual plant parformance and several ten week performance assessment NSA had responsibility for carrying out audits as required by 10 CFR 50 Appendix B and TS. The audit policies were specified in corporate policy 1000 PLN 7200.01,GPUN Operational Quality Assurance Plan, Rev.10. Aoministrative procedure 1110 ADM 7218.01, Nuclear Safety Assessment Audit Program provided imolementation details, Observations and Findinas NSA Audits The inspectors reviewed a sample of completed audits addressing the Corrective Action Program (Audits 97 01 and 96 03), Plant Support Engineering (Audit 97 04), Emergency Preparedness Program (Audits 95 07 and 96-08), Operations (Audit 9614) and Plant Maintenance (Audit 97 03). I The inspectors determined that NSA performed the completed audits at the proper
' frequency, and addressed performance as well as compliance related issues. The-inspectors also found:
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- The engineering, emergency preparedness, operations, and maintenance audits were all of good qualit * The audit policies were implemented in accordance with AP * Well scoped and detailed audit checklists assured review of pertinent safety issues and departmental performance. The checklists also included attributes for reviewing corrective actions and management requested issues in the areas assesse * Significant findings were documented both on CAP forms and in ODRs for departmental response and corrective action developmen * The audits also identified minor deficiencies that were corrected during the audit and performance improvement recommendation The inspectors noted severalinstances where lir.e management f ailed to address NSA audit findings effectively. Examples included findings associated with errors in the offsite dose calculation sof tware identified by the emergency preparedness audit and improper classification of components on the quality classification list. Failure to address these problems in a timely manner contributed to the severity of NRC enforcement art"m as described in NRC Inspection Reports 50-289/97 04 and 97-01 respectivel Audit findings generally received the appropriate level of management attentio Appropriate attention was provided to the trend of engineering related audit finding Findings were identified in the areas of design control, configuration control, mod >Ccation implementation, and operations experience information review. Additional assessments by NSA identified weaknesses in the area of calculation control. The cause of these weaknesses were attributed to the reorganization of engineering in 1996 and f ailure to follow procedures, NSA properly assessed the trend of findings in engineering and increased the frequency of audits from biennial to annua Routine NSA Assessments:
NSA provided plant and corporate management with an assessment of plant performance annual'y and each ten weeks. The ten week assessments provided station and corporate management with assessment information concerning findings from audits completed during the period and the status of ongoing audits. The assessments also provided information from monitoring assessments conducted during the period including: significant findings, minor deficiencies, recommendations, and satisfactory performanc The inspectors reviewed the assessment reports for 1995 and 1996. These reports assessed the effectiveness of implementing the QA plan and supporting requirements in the areas of operations, engineering, maintenance, and plant support. The resultant performance was assessed based on the observable effects on equipment, human, and organizational performance indicators. The assessment results were scored and categorized as green, yellow, or red to f acilitate management analysi _ . . . _ . . . .
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21 Conclusions The Inspectors concluded that NSA's assessments, findings, and recommendations have been of good quality. However,in severalinstances, including the quality classification of safety equipment and the emergency preparedness program, the line organizations f ailed to take sufficient action to preclude recurrence of known problem .2.2 Self Assessments by Line Organizations Sagne The inspectors reviewed documentation and interviewed GPU NSA personnel concerning the TMI self assessment program.
1 Qbservations and Findinat
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The inspectors determined that several departmental self assessments had been conducted, but that the overall program was still under development, finding that:
- GPUN management articulated the philosophy and expectations for line organizations to conduct self assessments as early as 199 * The TMl Site Director documented senior management expectations for performing self assessments in a memo to his staff, dated October 31,199 * GPUN incorporated expectations for self assessments into the Nuclear Corporate Policy and Procedure Manual 1000 PLN 1291.01, Nuclear and Radiation Safety Plan, Rev. 8 In July 1997
- GPUN recently appointed a program owner and program coordinato * Self assessment program implementing procedures were under ongoing developmen * GPUN planned to tralr. fif ty individuals as lead assessors by the first quarter of 199 * Departments had been performing self assessments and a database had been implemented for recording self assessment results. However to date there has been no clear demonstration that self assessments have imoroved performanc * To date, no trending had been performed with the results in the databas *
During scheduled audits, NS. - rsonnel have been reviewing to verify that self-assessmente are being performed. However, because the self assessment program has been under ongoing development, these audits have not been focused to assess the quality of completed _self assessment .
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22 Conclusions The inspectors concluded that management expectations for the celf assessment program were established 'n 1994, but to date program implementation has been inconsistent. The initial procedure for implementing the self assessment program remains under development. Some line organization self assessments were performed despite absence of a station wide program document. The program was determined to be too new to assess the quality of the self assessments performe .2.3 Self Checking Program Scope The inspectors reviewed GPUN's activities to maintain site wide consistency and accountability to the Site Vice President's goals for the "Be Sure" self checking program, @servations and Findinas During ti a NRC's last 40500 inspection (NRC Inspection Report 50 289/96 04)of the i corrective action program, the inspectors noted that the Site Vice President had established l expectatic,ns for workers to be personally responsible and accountable. The report noted
) that "Be Sure" was TMi's site specific self checking program. The inspectors observed that each department was given wide latitude and flexibility for interpreting and implementing the goals. The inspectors questioned how stStlon wide consistency and accountability to the Site VP's goals were maintaine The inspectors reviewed the actions completed since the last inspection and discussed these actions with GPUN operations and training personnel. GPUN personnel stated that self-checking training was completed for all station personnel. The inspector reviewed the I
training lesson plan and the department specific examples used as follow up training. Both the Site VP's memo on the management expectations in pursuit of excellence and the "Be Sure" training lesson plan identified that self checking expectations documented in AP 1029, Conduct of Operations, applied to all personnel working at TM The inspectors also reviewed the summary data from 58 management tours of self-checking performed in 1997. These data indicated that these tours identified no adverse findings related to self-checking. Specific attributes identified for review included: (1) did not pause before taking action; (2) did not have procedure or check as found conditions; (3) did not " Touch the Tag;" and (4) did not perform post action verification. The inspectors had no further questions, Conclusions Based on licenseo information that site wide self checking training had been completed and 58 management tours to review self checking identified no findings, the inspectors concluded that GPUN had effectively implemented self checking practices at TM i l
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07.3 Operating Experience Feedback Program ScoDe The inspectors evaluated lictnsee implementation of internal and external OE feedbac Information reviewed included TMI AP 1086, Industry OE Review Process, Rev. O, licensee assessments of the OE program, the OE and vendor information databases, and personnel interviews, Observations and Findinas During the past year the GOHB, the NSA department, and Institute of Nuclear Power Operations (INPO) representatives raised OE program concerns including the station's f allu;o to review some INPO and NRC information documents through the progra GPUN took adequate corrective actions to address ODR 972021 which identified numerous
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weaknesses, O of 25 NRC Information notices (lN) or INPO significant operating event reports (SOERai were not properly dispositioned. GPUN took several actions to improve the prooram including:
- Making the shif t engineer (STA) responsibilities for reviewing INPO and NRC Information. In July 1997, STAS began directly accessing the INPO OE database in addition to the NRC home page via the Interne *
A station-wide OE database for NRC & INPO OE information was established and updated to include all NRC ins and INPO information from 1/96 to presen STAS demonstrated a clear understanding of their responsibilities and the revised OE procedure. Based on database review and personnelinterviews the inspectors determined that the STAS were properly forwarding OE information to appropriate station personnel for resolutio The inspectors reviewed the vendor issues database and the 1997 vendor bulletin file and determined that vendor information was being properly distributed for resolution. A 10 CFR Part 21 issue regarding defective emergency diesel generator piping welds was properly reviewed and appropriate actions were assigned / schedule Resolution to 11 of 12 OE database (NRC and INPO) items reviewed were acceptabl Two examples of good issue resolution were NRC IN 97 49, B&W Once Through Steam Generator Tube inspection Findings, and SOER 961, Control Room Supervision, Operation Decision Making, and Teamwork. However, responses to severalissue reviews assigned by the STAS had minimal detail which limited the ability of the reviewere to do an independent closeout, i
The inspectors noted that the engineering response to NRC IN 97 21 demonstrated a poor understanding of the purpose of NRC ins. Acceptance of a minimal engineering response
_directly contributed to poor resolution of NRC IN 97 21, Availability of Alternate AC Power Source Designed for Station Blackout (SBO) Events. NRC IN 97 21 addressed SBO diesel e m
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generator (DG) vulnerability to f ailure to start following a delayed demand during a loss of offsite power (LOOP) event. Two reasons for failure to start were depletion of the SBO DG battery and failure of the SBO DG Keep Warm systems during the LOO .
The engineering response to this IN stated that no action for TMI was needed since the condition was outside of the plant's design basis and outside of 10 CFR 50.63 requirements. The re9ponse f ailed to state that TMI was susceptible to the same condition, as described in the IN, that caused the SBO DGs to f all at two other f acilitie The inspec'. ors identified the applicability of the IN and discussed the issue with operations and engineering personnel. The engineers who responded to the IN told the inspectors that while a proceduro change to address the issue would be nice, it was not worth the effort considering their other existing work load. Operations personnel had not been aware of the station's susceptibility to this condition and reopened NRC IN 97 21 for proper resolution. This action was appropriat Responses to several additional OE issues were even more brief, with no basis for the response documented. The minimal response negated the potential benefit of the two person reviewer closeout. The OE manager stated that reviewers often relied on the experience of the assigned responders for a given issue and didn't expect the response to fully detail the basis for the answer. The OE manager further stated he would evaluate the benefit of requiring more detailed responses as compared to the additional amount of resources which would be needed for documentation and review. Tho NSA Process Improvement Plan recommended raising resources for the OE program from part time positions to one full time OE coordinator at each site. The OE manager was reviewing this for implementation at the close of this inspection, in addition, he is reviewing the possibility of integrating OE required actions into the next revision to the CAP system for visibilit MR review of industry events for SSC scope adjustments was acceptable. The MR coordinator performed a yearly review of NPRDS (Nuclear Plant Reliability Database System). No scope additions were identified, Conclusions OE program performance was mixed with some signs of improvement. Internal and external oversight groups raised several concerns regarding use of industry experienc Several actions to address those concerns were recently implemented. The OE and Vendor Documents departments typically notify the correct station personnel of potentia! industry concerns for resolution. husolution thoroughness and response detail varied widely and in some instances were insufficient for reviewers to do an independent closecut. Inadequate detail and resource constraints resulted in poor assessment of a station blackout diesel
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' generator issue. Senior manecement recognized thst proper OE implementation requires more resources and intends to add one full time individual to help coordinate the OE progra _
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07.4 Safety Review Committees i i Scoce- :
I The inspectors reviewed safety review committee performance to determine whether the
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committees provided effective oversight of station activities and whether they satisfied j associated TS and quality assurance program requirenients.
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The TMl safety review committee function is comprised of four components, the NSCC, the GORB, the PRG, and the IOSRG. TS 6.5.2 specifies requirements for the IOSRG The GPUN Operational Quality Assurance Plan specifies requirements for the GOR Observations and Findinas NSCC The NSCC originated as a license requirement in 1983 to support independent oversight for
TMI Unit 1 restart. The license requirement was deleted in 1994 based on long term
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demonstrated safe operation though seven SALP cycles. Tho GPUN board of directors decided the benefit of independent oversight, warranted continuance of the NSCC and has maintained the committee active to date. The NSCC charter is specified within GPUN 1000-PLN 1291.01, Nuclear and Radiation Safety Plan, Rev. 8. Two NSCC staff members are assigned at TMl and report directly to three outside GPUN board of directors
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members who form the NSC The inspectors interviewed the NSCC staff and reviewed the last three semi annual NSCC reports and responses submitted from the GPUN President, Report content was meaningful and responses appropriately addre;4ed the issues raised in the reports. The NSCC has had stt.ble membership and an independent chain of reporting. The monthly
- conference call among NSCC members and the bimonthly NSCC meeting established a good frequency for reviewing station activitics. The inspectors noted that the content and detail of NSCC findings were good. Recent findings included (1) makeup pump suction line overpressurization/ configuration issues, (2) corrective action identification threshold / program consolidation / resolution timeliriess (escalation) Issues, and (3) corrective actions and root cause assessments have been slow. The NSCC staff initiated a NSCC Priority Issues List in 1997 to aid in tracking and clearly communicating important plant performance issues to the board of directors. This list was a useful tool fu .dentifying issues to the NSCC board members and directing resources for further follow up. The inspectors concluded that the NSCC identified important performance !ssues and
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effectively communicated them to the board of director GORB The inspectors reviewed GPUN 1100 ADM 1010.01, Procedure for Operation and Administration of the GORB, Rev. 7, and GPUN 1000-PLN 1291.01, Nuclear & Radiation Safety Plan, Rev. 8 which describe the GORB charter and operation. GORB is an off site review committee which is tasked to identify previously unrecognized / underlying station
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problems and review the effectiveness of the internal / external OE program. The inspectors also reviewed GORB rnecting minutes and Executive Summary Reports for the past 2 years, in general, GORD identified good issues which warranted additionallicensee attention. Examples included:
- Engineering response timelines * Engineerin0 workload and resource managemen * Consolidato corrective action programs to a single program to improve program management and effectivenes * Initiation of an air operated valve maintenance progra * Consideration to establish a program to periodically overhaul the reactor coolant pump motor However, the inspectors noted that GORB f ailed to identify several key issues ( emergency preparedness weaknesses, qualified component list reclassification deficiencies, and rnotor operated valve programs) indicating inconsistent performance by GORB subcommittees. This resulted from low qunlity issue presentations from station personnel and limited questioning from GORB members. Each of these uncorrected issues was known to station personnel, but not corrected in a timely manner as previously documented in various NRC Inspection Reports. The GORB Chairman informed the inspectors that he recognized this performance weakness and initiated several corrective actions to address it. Actions included:
- A GPU/GORB Interface improvement Plan was developed including annual member performance appraisals, specific membership responsibilities and station interf ace expectations, and GORB Executive session meeting attendance by the GPUN President beginning in 199 * Revise GORB membership to infuse new ideas and industry experienc * GPUN President issued a lottar to Vice President level personnel addressing the need to improve GPUN personnel's participation in the GORB process. This letter specifically addressed the desire to engender more open, candid, non-defensive communications with GORB members rather than keeping GORB members at arms lengt Corrective actions appeared well focused, but were not implemented sufficiently to support assessment during this report perio EllG The inspectors reviewed TMl AP 1034, Plant Review Group, Rev.14, interviewed the PRG chairman, and reviewed a sample of PRG meeting minutes. The purpose of the PRG is to provide a multidisciplined review to focus on operability and reportability determination . - - - - - .- - - -_-..--..-.---_-_----
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Membership was expanded in August 1996 to include approximately 170 members from the Operations & Maintenance, Radiological Protection, and Engineering department Chairman selects 515 members, based on subject matter specialties, from the qualified pool to attend a given meeting. The inspectors determined that:
- the PRG was properly staffed with qualified members who had been designated in writing. PRO mornber qualification and blannual requalification status were properly verified through program audit * PRG meeting minutes indicated generally reasonable discussion and issue developmen ,
The inspectors noted one issue in review of PRG action specifically identifying numerous instances where GPUN, based on PRG recommendations, had not submitted licensee event .
reports for TS required surveillances which had not been performed within the TS specified i time period (surveillance interval + 25 percent grace period). Most of these were due to ;
incomplete IST program scope. Corrective actions to strengthen IST program scope have been assigned. The inspectors questioned the PRG Chairman regarding whether these events constituted conditions prohibited by TS and as such whether they were reportable under 10 CFR 50.73(all2)(i)B. The PRG Chairman stated that failure to perform TS required surveillance within the specified time intervalis nut a violation and is not reportable if the surveillance is subsequently successfully completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery as permitted by TS 4.0.2. This position was based on Interpretation of TS 4. and an NRC letter dated May 11,1993, regarding a previous f ailure to perform a TS surveillance within the required time interva The inspectors acknowledged that TS 4.0.2 permits the licensee to delay their declaration <
that the 1ppIlcable LCO was not met, by up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of discovery. That permits the licensee to delay initiating the required LCO ACTION by up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to permit additional time to complete the surveillance. As stated in NRC GL 87 09, Sections 3.0 and 4.0 of the Standard TS on the Applicability of LCOs and Surveillance Requirements, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay was permitted to preclude unnecessary plant shutdowns wnen surveillance intervals were inadvertently exceede Plant operation in accordance with TS requirements including successful TS surveillance program implementation is a condition which forms part of the basis considered by the NRC when making risk informed decisions regarding a license, including NRC review of TS license amendment request Prior to the close of the inspection report period the inspectors discussed this issue with GPUN management informing the licensing manager and the PRG Chairman that the NRC staff currently considered that missing a TS required surveillance constituted a TS violation, which was reportable under 10 CFR 50.73 (a)(2)(i)B. Based on this discussion GPUN took action to revise their reportability guidance and report via an LER any subsequent missed TS surveillanco as a TS violation. The inspectors considered this to be a violation of minor significance in accordance with the NRC Enforcement Policy. (NCV 97 10 02)
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IOSRG/
The IOSRG was formed to independently assess plant activities from the perspective of assuring nuclear safety. Procedures GPUN 1000 PLN 1291.01, Nuclear and Radiation Safety Plan, Rev. 8, GPUN 1000 ADM 1291.01, Safety Review Process, Rev.13, and GPUN 1110 ADM 1010.01,10SRG Procedure, Rev. 3 address licensee implementation of IOSRG responsibilities. The inspectors reviewed the above procedures, the annual NSA audit of IOSRG, severallOSRG issue reports from the past 18 months, and interviewed personnel to assess IOSRG implementatlor.. The inspectors determined that IOSRG membership, training, and duty assignments satisfy TS requirements. Additionally, lOSRG identified good quality issues and findings as demonstrated by identification of:
- three inadequate GDR Safety Evaluations as docurnented in CAP 1997 047 Assessment and corrective actions were good.
l * several aspects regarding the station impact from the MIC control program that had i
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not yet been incorporated into the program. Nine clear recommendations were mad *
follow up review of issues associated with certain Millstone problems were guo *
questionable configuration of instrument air components RR V 6 and the air supply line filter, issue review and disposition were good, Conclusions Safety committee performance was mixed. The NSCC identified important performance issues and effectively communicated them to the board of directors. The GORB also identified several underlying issues and initiated appropriate action items. However, GORB f ailed to recognize and act on some key issues (i.e. emergency preparedness, qualified component list reclassification, and motor operated valve programs) indicating inconsistent performance by GORD subcommittees. Reasonable actions were proposed to improve GORB/GPU interf aces and performance. PRG membership appeared well controlled with generally good performance. However, an issue regarding dispositioning the reportability of missed TS surveillances was identified as a minor violation. The IOSRG identified and reported good issues, which were properly resolve Miscellaneous Operations issues 08.1 (Closed) VIO 50 289/97-06-01: Failure to Notify the State and County Offsite Agencies for the June 21 Unusual Even Scoce(92901)
The inspectors reviewed the corrective actions associated with the previously identified offsite notification violation. The review included the CAP Report No. T1997-0391,ODR No. 972033 and Emergency Plan Implementing Procedure EPIP TMI .03, " Emergency Notifications and Call Outs."
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29 Observations and Findinas GPUN responded to notice of violation (NOV) 50 289/97 06 01,in a letter dated September 13,1997, which provided background information regarding the failure to notify the State and local counties within 15 minutes of declaring an Unusual Event (UE)
per the emergency plan (E Plan). The root cause, as determined by the licensee, included the SS's decision to use an unqualified Instrumentation and control (l&C) technician to perform the emergency notification call outs. Short term corrective actions had been previously reviewed by the inspectors and were determined to be adequate to correct and prevent the recurrence of similar problems. The long term corrective actions, GPUN's root cause analysis and training provided to the l&C technicians were reviewe To determine the effectiveness of the E Plan training, the inspectors randomly selected l&C technicians to simulate the emergency notification call out to the state and local countie The l&C technicians were familiar with the call out procedure and the location of the communication equipment. However, the technicians' training did not include the ability to use the communication equipment to contact the offsite emergency agency during a non-emergency situation. The technicians could have the opportunity to make the offsite notification during the annual training exercise. However, based on the number of I&C technicians, some technicians could wait for greater than 10 years before they would have the opportunity to make the notifications. The inspectors discussed this issue with the E-Plan and plant management. Management decided to coordinate the monthly offsite notification call out with the l&C department to provide the technicians with the opportunity to use the actual communication equipment under contro,'e i conditions. The inspectors determined that the training provided to the I&C technicitns for the emergency notification was satisf actory, in addition to the l&C qualifications, the E Plan organization developed an electronic notification form that could be sent by computer to the state and counties. The form would provide written notification of the event to the offsite emergency centers and would be sent in parallel with the required telephone riotifications. The system was tested satisf actorily in December 1997. Some improvements were noted and will be incorporated by the E-Plan organizatio The root cause analysis and associated corrective actions were comprehensive. The violation is closed, Conclusiona GPUN's corrective actions to address the f ailure to notify the state and local counties within 15 minutes of the June 21,1997, UE were appropriate. The qualified l&C technicians should result in the availability of an additional control room operator to assist the on shift personnel af ter a transient or emergency condition.
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08.2 (Close.d) LER 50 289/96 00101: Seismic Qualification of Class IE 4160 VAC Westinghouse Breaker l Insoection Scone (927001 l The inspectors reviewed a generic seismic concern related to Westinghouse 4160 Volt !
electrical circuit breakers. TMl received the information from the nuclear network, I evaluated the data and determined that the issue applied to the TMI breakers. The review l Included the completion of the long term corrective actions, field installation of the modification that corrected the problem and a review of the LE b, Observations and Findinas On November 11,1996, the licensee was informed through the industry network that when certain circuit breakers were in the " racked out" position the switchgear was no longer seismically qualified. Following the engineering department's evaluation of the issue and completion of short term corrective actions to address the immediate concerns, a modification was installed during the 12R refuel outage to provide a permanent solution to the proble To address this concern, the plant bolted a metal bracket to the floor of each 4160 Volt breaker cubicle and attached an extension to each breaker. The modification allowed any breaker not connected to the electrical bus bars to be secured to the metai bracket. The bolted anchor was designed to prevent the Westinghouse breakers, on metal wheels, from moving during a postulated seismic event, in addition, AP 1002," Rules for Protection of Employees Working on Electrical and Mechanical Apparatus," was revised to provide written direction to the plant operators for proper breaker remova The LER provided a detailed description and assessment of the event. The root cause analysis and associated corrective actions were comprehensive. The LER is closed, Conclusions The inspectors concluded that the licensee's actions in response to the generic industry concern were comprehensive and effective. The licensee event report provided a detailed description and assessment of the event. The root cause analysis was thorough and the implementation of the associated corrective action was timel II. Maintenann M1 Conduct of Maintenance (62707,61726)
M1.1 General Comments Scope The inspectors observed all or portions of the following maintenance and surveillance work activities:
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o Job Order No. 145405," Reactor Building Emergency Cooler Outlet Valve, RR V 4D, MOVATS Testing."
e Job Order Nos. 145692 and 145693,"RR V 4C Preventive Melntenance and MOVATS Testir g."
i e Surveillance Procedure 130311.37C,"HSPS 0TSG Level and Pressure Channel lll Test."
e Surveillance Procedure 1303 4.1D,"RPS Channel 'D' Test."
e Surveillance Procedure 1303 3.1," Control Rod Movement Test."
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- Surveillance Procedure 1300 3K,IST of Reactor River Water Pumps and Valves." Observations and Findinas
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The electrical maintenance personnel followed MOV procedure E 13 "Limitorque Valve Operator inspection," throughout the preventive maintenance activity Use of the MOV electrical wiring diagram by the electricians to verify that the MOV connections were i consistent with the electrical print was excellent. The electricians performed a wire by
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wire verification of the valve connections and applied the " touch the tag" self checking
standard. The electrical supervisor provided valuable insights to the technicians 1 throughout the RR V 4C/4D maintenance activities. The maintenance activities were
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departmen Surveillance test performance for the power range NI calibration, emergency diesel I generator (EDG), ES logic, control rod movement, and RPS was excellent. The RPS weekly tests were performed af ter the required documentation was revised to address the bypassed 'C' RPS logic channel. The procedure revisions were coordinated with a detailed safety evaluation that accurately addressed the impact of the procedure change Engineering personnel evaluated the actual test performance in the control roo/.s. Each
channel test was performed satisfactorily by the experienced l&C techniciai\ Thel &C technicians applied self checking practices for each step of the procedure, Ggnqlusions The electrical maintenance technicians' performance, supervision oversight and coordina3on of the RBEC isolation valves, RR V 4C/D preventive maintenance and MOVATS testing were excellent. The system engineer manager provided excellent support throughout the testin Surveillance test performance was excellent. In particular, the RPS weekly tests were
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performed af ter the required documentation was revised to address the bypassed 'C' RPS
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I M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Review of Maintenance Rule Implementation Scope The inspectors reviewed selected aspects of the GPUN MR Implementation.
, Observations and Findinas GPUN appeared to be implementing the reviewed portions of the MR well; this included:
- Good tracking of SSC performance criteria using state of the art computer system * Proper tracking of actual and planned system out of service time. GPUN also established two lower level action points to ensure management approval before exceeding the MR unavailability point. Unavailability goals to that point are listed on the weekly plan of the day, to bring added focus to planned on line work, j * Use of the CAP system as a method of identifying and tracking oquipment problems. Morning meeting discussions of CAP provided a good first cut of possible functional f ailures and maintenance preventable functional failures (MPFF).
- There were several significant systems included as a(1) SSC including: 1) the MU and LPI systems due to unavailability above the MR allowable criteria,2) the RCS
- due to the MPFF causing the PORV to be inoperable for an operating cycle,3) the CRDMs due to slow drop times (MPFFs),4) the SBO DG due to elr start solenoid failures, it appears that GPUN has plans in place to address each of these concerns and return these system to normal monitoring under a(2).
, Conclusions GPUN properly used good tracking systems and monitoring of unavailability to identity SSCs that needed additional attention. Further, the CAP process and routine meeting and planning appeared to allow identification for equipment problems and the trending and tracking of planned on line maintenance effects on unavailability. Tracking, identification and planning for improvements on SSCs in the a(1) criteria appeared goo . Enoineerina E1 Conduct of Engineering (37551,92901)
Overall, engineering provided good support for plant operations including: questioning the need for the ECCS flow instruments to meet system operability, review of the MU leak;
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development of a temporary modification to the RCP oilleak collection system to allow
. remote (outside the D ring) pumpdown; review of the NI 7 lower detector f ailure; and in support of the replacument of the B RCP seal packag ._ _ _ _ _ _ _ __. . _ _ _ _ _ .
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E Reactor Building Emergency Cooler Generic Letter 96 00 Scope The inspectors reviewed GPUN's response to GL 96 00," Assurance of Equipment Operability and Containment integrity During Design Basis Accident Conditions," related to the reactor building emergency cooler (RBEC) isolation and a leak check of the outlet valves, Observations and Findinos GPUN's February 14,1997, submittalin response to GL 96 00, assumed at the time of the DBA LOCA, that the RBECs were in a normal standby condition with the RR V3 valves open and the RR V4 valves closed with NSCCW system providing overpressure to the coolers and piping in the containment. The submittal stated "The analysis does not reveal l any voiding of the system during this evolution [ system response to the I.OCA) dua to the surge tank (NSCCW) providing makeup and assisting to maintain system pressure as the
, RR V4 valve begins to open. The system pressure is maintained above the saturation l
pressure of the fluid. The opening of the valve into a low pressure downstream condition produces a rapid outflow of fluid. This allows the system to de pressurize, but volds do not form within the cooling coils because the surge tank is able to provide sufficient mak up flow and pressure control. The event is terminated upon the start of the RBEC pumps (RR P1 A/B). ...... Therefore, since our analysis show that voiding does not occur even in l
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the worst case design basis accident, neither two phase flow nor water hammer present a concern for TMI 1 containment air coolers".
In discussion at the PRG meeting GPUN identified that engineering had not noted that the word.ng of the submittal was different than the wording in the UFSAR, specifically that the RBECs needed to be pressurized at all times to be within the bounding analysis. Further, since it was not identified, the information was not communicated to operations prior to the performance of the normal surveillance test, the isolation of the NSCCW pressure source, or prior to the conduct of the leak test on November 20,1997. Discussions in the UFSAR relate to the NSCCW pressure (above containment design pressure) preventing a containment bypass path to the environment, allowing any leakage to pass into the RB rather than out through the coolers. The UFSAR did not discuss any need for overpressure to prevent voidin Following identification of this issuo GPUN engineering completed a more detailed calculation showing that the RBECs would not void without NSCCW pressure following a DBA LOCA. GPUN also comrnitted to update the GL 96-06 submittal to clarify that the NSCCW pressure was not needed to prevent voidin '.
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34 Conclusions Engineering preformed wellin identifying the issue of a difference between the GL 06 06 submittal and the UFSAR and getting the issue to the site for PRG teview (See Section '
O.6.1). However, this instance raised questions about the adequacy of the review and the understanding of possible plant impacts as a result of the GL 90 06 submittal. The inspectors '.4nsidered this an Unresolved item pending turther review of the RBEC ,
configuratiuh design (URI g710 03.)
E2 Engineering Support of Facilities and Equipment
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E Review of Makeup Pump Net Positivo Suction Head Calculations j Ecpani GPUN completed calculations and procedure changes to address an issue with the cc trols over the MU pump NPSH GPUN documented the calculation output in the MU system operating procedure as a graph of acceptable MU tank pressure and level, below which NPSH may not be sufficient. The basis for this calculation was a SBLOCA occurring at the injection point of normal makeup into the reactor coolant system (RCS) with the subsequent system and operator respons The overall premus of the calculations was to determine a suitable maximum flowrate from the MU tank and limit it by throttling a manual globe valve such that, with the given flowrate, if the initial level was above the curve for the given pressure, the flow rate would not cause the MU pumps to experience a low NPSH, before the BWST suction valves {
automatically open on a ES signa The inspector reviewed a special test procedure, conducted by GPUN during plant heat up following the refueling outage, designed to validate the MU system flow calculations and the positioning of MU V222, the normal MU injection line manualisolation valv The inspectors reviewed the verification backup calculation conducted to validate the system resistance calculation, Observations and Findinas The inspector validated the plant conditions that would occur following the SBLOCA including: zero back pressure, the lowering of pressurizer level and opening of MU V17 automatically, operator actions to open the MU V 17 bypass (MU V217) to increase flow and to start a second MU pump. The BWST suction valves to the MU pumps will open automatically on an ES signa ._ _ __
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The calculations were broken into two pieces: ,
- First a calculation was completed to document the system resistance in a single pump and double pump configuration, and curves for the MU pump head versus flow, so that the system flow could be calculated for different positions of the M '
V222 throttle valv GPUN used an equivalent pipe length calculation completed in 1983 as the basis for
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the system resistance and substituted new resistance values for the MU V217 and MU V 222. They then developed a mathematical model of the system resistanc (flow vs system resistance)
To develop a mathematical model of the pump curves GPUN used a spreadsheet regression analysis to produce a fourth order equation that represented the pump curve, using actual test points as the basis. (flow vs pump head)
To determine the system flow, GPUN used an iterative computer program to determine what system flow would cause these two equations to be equal (pump head thm ..as equivalent to system resistance).
- GPUN then reviewed ca NPSH requirements, the MU tank capacity, and the pump flewrates over time and the associated pressure drop vs flow between the MU tank and the pump suctions to develop a minimum MU tank level curve such that operating in the normal MU tank level band would prevent approaching a NPSH limit on two operating pumps, following a SBLOCA at the normal MU injection locatio During the review of the system resistance calculation validation the inspector found:
- The resistance values used were from a more recent form of calculation than the alternate pipe length assumptions used in the calculation and while of minimal consequence only the fittings were used in the analysis (i.e., pipe lengths were not taken into account); As such, the resistance values calculated by the alternate calculation were lower than those used in the cahulation. The calculation stated that this would result in an increase in calculated system flow of approximately 10 -
12 gallons per minut The inspector validated this increase in flow throu0h independent calculations using the flow system resistance methodology specified in Crane Technical paper 41 The inspector validated selected data and assumptions including:
- Pipe resistance calculations, regression analysis for pump head curve formula determination, system flow determinations based on selected valvo positions and system back pressure,
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The inspector found the following in reviewing the special test procedure:
- The safety evaluation was well prepared and properly stated the purpose for the test and contained a very good description of the determination that it was safe to conduc The inspector did note some weaknesses in the calculations and testing:
In comparing the system resistance determination methods used in the two calculation the inspector found a difference. The system flow calculation used an equivalent pipe length s methodology which provided system resistances that were higher than the more current methodology used to calculate the pressure drop from the makeup tank to the pump suction in the NPSH calculation. The inspector determined that use of these different methodologies produced non conservativo data ( l.e., lower system flow and lower head loss between the MU tank and the pump suction) when used to determine MU tank levels needed to meet NPSH requirements. However, the inspector found that by completing the test procedure GPUN properly validated the calculations through an actual tes *
In review of the GPUN SE as compared to the actual test, the inspector found that the SE stated that the test conditions were to meet the conditions specified by the
> MU tank calculation, except that back pressure would be rated reactor pressur However, the inspector found that the test positioned the normal pressurizer level control valve MU V17 to 40% open, which was different than the 100% open as specified in the MU system flow calculation and MU pump NPSH calculatio The inspector completed an analysis of this using independent calculations and determined that the additional flowrate with the valve fully open should not have a significant effect on the flow to the reactor coolant system. However, the inspector requested that GPUN update the SE to ensure that it matched the actual test conditions and compared the results to the calculation conditio *
The assumptions for the time between the SBLOCA to: reactor trip on low RCS pressure (20 sec) and ECCS actuation on low RCS pressure (60 sec) were not presented from any specific analys;3. The system engineer stated that these times were from an original 1980 B7W analysis on the SBLOC Conclusions The calculations completed on the MU system injection piping for a SBLOCA, to determine the minimum MU tant levels for NPSH, were adequate, because testing proved the actual system performance within the bounds of the calculations, However, the inspector found differences in the methodologies employed in system resistance modeling. The method used to calculate the system flow on the discharge of the pumps was an older method (more resistance less flow) while the method used on the pump suctior was a new method ( less resistance -low pressure drop). In the application to NPSH determination these two calculation methods could lead to non-conservative determination of minimum required MU tank levels if the calculations were all
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that was used. The lower flow from the pumps would lead to a smaller drop in MU tank level for the time period between the accident and the opening of the BWST supply, so the initiallevel could be lowered. The deceased resistance on the pump suction would lead to a determination of a lower MU tank level required to maintain NPSH at the pump suctio The inspector., also found a weakness in the test procedure and the safety evaluation since the & tested conditions did not match the calculation condition. The inspector determined that this did not invalidate the test results. However, GPUN ccamitted to update the SE to i include an evaluation of the test results versus the calculated conditio E3 Engineering Procedures and Documentation E (Closed)VIO 50 289/97 0101:OCL Component Downgrade for Valves NR V-1 A/B&C; Strainer Motors DR-S 1 A&B; Auxiliary Ventilation Fans; and Make up valve MU V 17; VIO 50 289/97 0102: Failure to Follow the Procedure Requirements of EP-011; VIO 50 289/97 0103: Failure to include EP 011 in the Safety Review Proces Scono and Backaround 192901)
NRC Inspection Report No. 97 01 dated March 20,1997, addressed the initial equipment OCL weaknesses. NRC issued the OCL confirmatory action letter (CAL) on March 4, 1997,to summarize the corrective actions planned by GPU The OCL immediate and short term corrective actions were completed by GPUN and documented in a letter to the NRC on April 30,199 The inspectors reviewed the OCL documentation related to confirmatory action letter 1-97-008, Items No.1 & 2 as noted in NRC Inspection Report No. 97 05, dated July 11, 1997. The CAL commitments included takirg immediate actions to pieclude additional inappropriate equipment downgrades and to determine the impact of the equipment downgrade program at TM On July 16,1997, a meeting was held in King of Prussia, Pennsylvania, to discuss the GPUN response to CAL 197-008,ltems 3 and 4. The presentation included the findings of the independent Engineering and Corrective Action Process Assessment Team (ECAPAT)
and thu GPUN response to problems identified in the ECAPAT report. The findings included the Team's root cause analysis, associated planned corrective actions, and assessment of the potentialimpact on other engineering processes. The ECAPAT evaluation of the engineering process problems and planned corrective actions e appropriately addressed CAL ltems 3 and On August 20,1997, the NRC closed CAL 197 008, based on the satisfactory completion of the corrective actions for the four action item . _ _- _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ . _
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In addition to the review of the immeriiate and short term corrective actions, the inspectors reviewed the implementation associated with the revised QCL orogram, EP-011 pro':edure changes and applicable ECAPAT issues. A majority of the long term QCL items were completed as scheduled by December 31,1997, Qhservations and Findinas The inspector reviewed the content and implementation of engineering procedure EP-011,
" Methodology for Preparing the Quality Classification List." The corporate Engineenng Division procedure was revised on April 29,1997, and training was provided to all j
personnelinvolved with the QCL process. The procedure was added to the safety review program described by TS 6.5.1.12 and SE 945100-099 documented '.he bases for the revision. The new procedure has formalized the OCL process and included written detailed standards related to component and program changes. For example, EP-011, section 4.5.1, " Downgrades," was revised to require a written safety determination / safety evaluation when the quality classification of a component is changed from a higher to a lower classification. The definition of regulatory requirod components was clarified in the proceduto and " Exhibit 9A" provided a list of regulatory references that would result in an regulatory required classification level. The applicability of the procedure was changed to reflect the intent of the program quality control An example of engineering personnel following the new procedure was noted in the generation of four CAP forms. The CAPS were written to docurnent and evaluate components that were upgraded to a higher classification based on the engineering revie EP-011 section 4.4.1 provides the written requirement to initiate a CAP form for upgraded components. The four CAP forms were initiated to document components that were determined by engineering to belong at the nuclear safety related classificatio The components were related to electrical separation devices between NSR and non-NSR equipment, a control room ventilation panel, and a solenoid valve that supports the operation of the spent fue' oool cooling f an. An MNCR was initiated for each component to evaluate and document the operability and potential plant impact. The engineering review determined the plant impact to be minimal based on the fact that the equipment was not replaced or changed since the time the component was downgraded. The inspectors' review of the documentation and equipment concluded that the engineering assessment was appropriat To determine the quality of the original GCL process GPUN developed a sampling methodology to review the origmal documentation. The current EP-011 guidance was
'Jsed as the standard to determine the preper classification of 431 components. The procedure checklist vua used to verify the proper quality class, seismic class, OA requirements, NSR function and applicable regulatory requirements. To date only one component's original classification was upgraded, the problem was documented on a CAP for . _ _ _ . . . . . . .
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l The inspectors reviewed a select sample of safety determinations and e aluations to support the change in the component status from safety related to deskory required ind from regulatory required to non-safety related/other. The component u..onnation required by the recently revised engineering procedures EP011 and EP 016 was documented on the proper forms. The documentation was complete, clear and provided a logical written evaluation that justified each change in component classification. The QCL documentation related to the following list of components was reviewed: ,
DH V 061 A/B, " Low Pressure injection Boundary Valve Isolation from the Caustic Supply Line" DH V 165A/B," Decay Heat RemovalTest connection Drain Valve"
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DH V-166A/B, " Decay Heat Removal Check Valve for DH V 4A/B" MU V 17, " Normal Makeup to the RCS" MU V-18, " Normal Makeup Isolation Valve" DR S 1 A/B, " Decay River Motor Strainer Motors" NR-V-1 A,B&C, " Nuclear River Water Fump Discharge Valves" RR S-1 A/B, " Reactor River Water Strainer Motors" RR V-6, " Reactor Building Cooling Coil Back pressure Valve Regulator" SF V84A/B, " Test Connections for the Fuel Transfer Tube Cover Plate "O" Rings" Auxiliary Building Ventilation System Components in addition the corresponding safety determinations / evaluations were reviewed:
SE-000211-011,"QCL Classification of the Makeup Control components for MU V-17" SE-000212-028,"QCL Classification of the Low Pressure injection Boundary Valve DH V 061 A/B" SE-000534 005,"QCL Classification cf the RR-V-6 Actuator Air Regulator"
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40 Conclusions The root cause analysis and associated corrective actions were comprehensive and timel The OCL Violations are close The quality classification list long term corrective actions were cornprehensive. The management involvement and oversight has resulted in an ;mproved procurement and QCL proces EP-011," Methodology for Preparing the Quality Classification List," was revised to formalize the OCL process and included written detailed standards related to es lponent and program changes. For example, EP-011, section 4.5.1, " Downgrades," was revised to require a written safety determination / safety evaluation when the quality classification of a component is changed from a higher to a lower classification, in addition, the procedure was added to the safety review program described by TS 6,5.1.12 and a safety evaluation was written to document the bares for the revisio E8 Miscellaneous Engineering issues E Reactor Building Primary Shield Wall High Temperatute i
. ' Scope The inspector reviewed GPUN action in dealing with the elevated temperatures at the RB primary shield wall, including the alarm response procedure, SE, UFSAR, and a PRG meeting.
I Observations and Findinas The inspectors reviewed the RB temperature values on the plant process computer. The temperature readings associated with computer point A0424 were gradually trending higher since the November 11,1997, plant startup . The RB air temperature read 199*F on December 18,1997. The inspector reviewed the UFSAR section 5.6.2 b., " Ventilation and Puge Systems." The UFSAR stated that during normal operating periods the RB concrete temperatures adjacent to the reactor vessel, seal ring and nozzles should not exceed 1500F, except local areas, such as around penetrations, which shall not exceed 200 The high temperatures were attributed to small gaps in the insulation around the RCS hot leg and cold leg pipes that penetrate the primary shield wall (pipes that run from the reactor vessel to the steam generators and from the once through steam generators (OTSG) back to the reactor vessel). A modification to address the insulation gaps was canceled in the 1993 time frame due to no noticeable concrete degradatio After discussions with the engineering department, the system engineer initiated a CAP form to evaluate the potential safety concern. CAP T1997-0941, documented the issue and stated that the American Concrete Institute (ACl) specifies that the concrete temperature around penetrations must remain below 200 F unless testing is performed to
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determine the affects of the elevated temperature on coni. rete. In addition, industry research indicated that significant strength reduction of concrete does not occur at temperatures below 300'F. TMI engineering performed SE-000153-005,to document that there have been occasions when the RB air temperatures, in close proximity of the penetrations, have exceeded 200 F. The maximum temperature recorded was 281 *F and in no caso did any air or concrete temperature exceed 300' ,
f.s recommended by the ACI guidance, TMl hes performed visualinspections of the six penetrations that have experienced air temperatures in excess of 200 F. The inspections noted that there was no deterioration of the concrete or corrosion of the reinforcing bars at any location. The inspections were performed during the 10R refueling outage in 1993, the 11R outage in 1995, and the 12R outage in September 199 The CAP form also addressed the need to review the UFSAR change process to determine why the 1993 safety evaluation did not result in an update to the design bases temperature limit of the RB concrete. TMI initiated a request, PFU No. 98 T1-131,to submit the required paper work to update and correct UFSAR section 5.6. The plant review group met on January 5,1998, to discuss the RB concrete high temperature issue and to determine if the issue was reputable. The meeting provided a good discussion about the possible reportability aspects of the issue, and more importantly, that the plant was confident that the potential safety concerns were and continue to be addressed by the engineering department. The PRG concluded that the issue was not reportabl Conclusions The engineering safety evaluation written in 1993 addressed the potential safety issue associated with the high temperature effects on the concrete for the primary containmen However, the engineering processes did not result in the revision to the UFSAR to reficct the increased concrete temperature limit as noted in the safety evaluatio E8.2 (Closed) LER 50-289/96-002-00/01: Potential Unreviewed Safety Question Related to the Net Positive Suction Head for the Decay Heat Removal and Building Spray Pumps, The unresolved item (URI), 50 289/96-201-14, associated with the decay heat removal (DHR) and building spray system (BS) pumps NPSH was reviewed in detail and closed in NRC Inspection Repoit No. 50 289/97-06/02. In conclusion, the inspectors reviewed the abnormal transient procedure and corresponding operating procedure revisions in the control room to ensure that the DHR and BS pumps will have adequate NPSH af ter a LOCA without taking credit for RB overpressure. Engineering personnel performed the detailed calculations and safety evaluations to resolve the unresolved safety question (USQ) issue prior to plant re-star The LER provided a detailed description of the event, assessment, and appropriatc corrective actions that were completed to close the URI. The LER is closei I
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IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Implementation of the Radioactive Liquid and Gaseouc Effluent Control Programs Scooe (84750)
The inspection consisted of: (1) a tour of the plant, including the control room. (2) review of liquid and gaseous effluent release permits, and (3) review of unplannedlunmonitored release pathways, if any, Observations and Findinas The inspectors toured all Units 1 & 2 effluent RMS, selected air cleaning systems, and the control rooms. All effluent RMS and air cleaning systems were operable at the time of this inspectio The inspectors reviewed selected radioactive liquid and gas release permits and associated procedures. Radioactive liquid and gas release permits contained: (1) gamma measurement results; (2) tritium measurement results; (3) projected dose calculation results; (4) cumulative dose contributions from radioactive gas and liquid releases for the current calendar quarter; (5) RMS readings (before, during, and end of release); and (6)
alert and alarm setpoints. The inspectors determined that the licensee followed associated procedures and the Offsite Dose Calet.lation Manual (ODCM) requirement Conclusions Based on the above reviews, the inspectors determined that the licensee implemented the radioactive liquid and gaseous effluent control progrcms wel R2 Status of RP&C Facilities and Equipment R2.1 Calibration of Eff!uent/ Process Radiation Monitoring Systems Scone (84750)
l The inspectors reviewed the most recent calibration results for the following list of effluent / process RMS as des %nated for each unit. The inspectors also reviewed the licensee's RMS tracking and trending result Unit 1 RM-L6, Liquid Radwaste Effluent Monitor RM-L7, Station Discharge Liquid Monitor RM-L12, IWTS/lWFS Discharge Line Monitor
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RM-A4G, Fuel Handling Building Exhaust Noble Gas Monitor
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RM-A6G, Auxiliary Building Exhaust RM-A8G, Station Vent Exhaust RM-A5G, Condenser Vent Noblo Gas Monitor RM-A15, Condenser Vent Noble Gas Monitor RM A7G, Waste Gas Holdup System Noble C 's Monitor RM A9G, Containment Purge Noble Gas Monitor Unit 2 2HP-R 219G, Station Vent Noble Gas Monitors 2HP-R 255G, Reactor Building "A" Train Purge Exhaust Monitor Observations and Findinas The I&C Department had the responsibility to perform both electronic and radiological calibrations. The Radiological Engineering staff also reviewed radiological calibration results. The system engineer had overall responsibility for operability and system evaluation, including tracking ard trending analyse The above RMS were calibrated, operable, and maintained according to the ODCM requirements for both units. The licensee continued to perform channel calibrations quarterly and every 18 months or during each refueling outage, as required by the surveillance requirements specified in the ODCM. During a review of licensee calibration data, the inspectors noted that data pertaining to licensee plateau checks en operating high voltage had not been documented. The inspectors did not note a case in which the operating high voltage had been set inappropriately. The inspectors were informed that l&C technicians had performed plateau checks to verify that operating high voltage had been set appropriately. Cognizant licensee personnel stated that calibration procedures will be reviewed and appropriated actions will be takan accordingl The system engineer performed a statistical analysis to determine the linearity as well as conversion factors. The inspectors reviewed tracking and trending analyses for radioactive liquid and gaseous effluent RMS and had no further questions in this are Conclusions Based on the above evaluation, the inspectors concluded that this program area was goo R2.2 Surveillance Tests for Air Cleaning and Ventilation Systems Scooe (84750)
The inspectors reviewed the licensee's: (1) most recent surveillance test results, and (2)
performance summaries to determine the implementation of TS and UFSAR requirements for the following systems:
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- Control Room Emergency Vent Reactor Building Purge Exhaust Auxiliary and Fuel Handling Building Fuel Handling Building Engineered Safety Features Air Treatment Unit 2 Reactor Building Purge Exhaust Auxiliary Building Cleanup Fuel Handling Building Cleanup Observations and Findinas
The inspectors noted that deficiencies identified during surveillance testing were corrected and as left conditions met the licensee's acceptance criteri The licent t's TS sps...,- riegulatory Position C.6.a of Regulatory Guide (RG) 1.52, i
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Revision 2, March 1978, a:s the requirement for the laboratory testing af the charcoal. RG l 1.52 references ANSI N509-1976," Nuclear Power Plant Air-Cleaning Units and Components." ANSI N5091976 specifies that testing is to be performed in accordance with paragraph 4.5.3 of RDT M 161T, " Gas Phase Adsorbents for Trapping Radioactive lodine and lodine Components." Charcoal efficiency testing was conducted by a vendor service. In addition to the RDT M 161T methodology, the licensee conducted for -
information only tests using more stringent charcoal challenge testing methodology. The inspectors considered this to be a good practice, Conclusions The inspectors concluded that the licensee maintained a good program for air cleaning system R3 RP&C Procedures and Documentation R3.1 Radioactive Effluent Release Procedures Scoce (84750)
The inspectors reviewed the following selected procedures to determine whether the licerisee could implement the radioactive liquid and gaseous effluent control programs effectivel ,
6610-ADM-4250.01, Releasing Radioactive Liquid Waste 6610-ADM-4250.11, Releasing Radioactive Gaseous Effluents-Waste Gas Decay Tanks A/B/C 6610-ADM-4250.12, Releasing Radioactive Gaseous Effluents-Reactor i Building Purge
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45 Observations and Findinos The inspectors noted that the effluent control procedures were sufficiently detailed and well written to control routine effluent releases effectively. Procedures were being updated to reflect the current ODCM. Release permits were properly completed, including projected dose calculation as described in Section R1.1 of this inspection repor Conclusions Based on the sbove review, the inspectors determined that the lice asce had good procedures to implement the radioactive liquid and gaseous effluen control programs effectivel R3.2 Review of Annual Radioactive Effluent Reports Scoce (84750)
The inspectors reviewed the 1995 and 1996 Annual Radioactive Effluent Reports to verify implementation of the TS/ODC Observations and Findinas These reports provided total quantities of liquid and gaseous efflet nt released from both units and included projected doses to +he public. The inspectors datermined that the licensee met the TS/ODCM reportiri -
irements and the reports contained the information specified in the ODCM. No obvious omissions, trends or anomalous measurements were identifie Projected doses to the public were well below the TS limits. For example, projected doses to the public due to radioactive liquid releases in 1995 and 1996 were 0.58 mrem and
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0.11 rnrem, respectively, well below the TS limit (3 mrem / year). l The inspectors noted some good efforts on the part of the licensee which have been beneficial toward maintaining radioactive liquid effluents as low as reasonably achievable (ALARA) such as the addition of the Auxiliary Building Sump 'ilters and oil skimmer and modifications which permitted the use of previously abandoned Unit 2 storage tank C9nclus.io.ns No discrepancies were noted pertaining to the Annual Radioactive Effluent Report Additionally, the licensee completed modifications which were beneficial toward maintaining radioactive liquid effluents ALARA.
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R3.3 Review of the ODCM Scoco(84750)
Inspection consisted of: (1) review of setpoint calculation methodologies; (2) review of selected parameters for calculating projected doses; and (3) seview of radioactive liquid and gaseous discharge pathway i Observations and Findinas in March 1995, the licensee transferred the Radiological Effluent Technical Specifications (RETS) from the TS to the ODCM in accordance with NRC Generic Letter 89-01. During the previous inspection conducted in Janua:y 1996, the inspector reviewed selected portions of the ODCM directly related to RETS inspection and had noted no significant differences from T During this inspection, the inspectors reviewed the ODCM (Revision 16, effective date:
1 June 22,1997)in the areas of: (1) setpoint calculation methodologies for effluent RMS; (2) parameters for calculating projected dose to the public; and (3) radioactive liquid and gaseous effluent discharge pathways. The ODCM containeri setpoint calculation methodologies for radioactive liquid and gaseous effluent RMS. The inspectors also noted that the ODCM contained all relevant parameters (Regulatory Guide 1.109, NUREG 0133, and site specific). Radioactive liquici and gaseous effluent pathway diagrams were also illustrated as require Conclusions
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Based on the above reviews, the inspectors determined that the licensee's ODCM contained all necessary information and guidance to perform the radioactive liquid and pnous effluent control program R7 Quality Assurance in RP&C Activities Scone (84750)
The inspection consisted of reviews of the 1997 QA audits, and chemistry measurement laboratory QA/Q Observations and Findinos The 1997 Audit covered the effluent control program including ODCM implementatio Audit scope was good. No findings of regulatory significance were identifie The inspectors toured the chemistry laboratory and discussed with chemistry staff regarding QA/QC programs to determine whether the licensee had adequate controls with respect to sampling, analyzing, and evaluating data for implementing the effluent control
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programs. The inspectors reviewed rasults pertaining to: (1) the intra laboratory and inter laboratory comparison; (2) blind duplicate samples; (3) reproduction techniques (reproducibility for sampling and analyzing); and (4) instrument control charts. OC data provided indication that the licensee implemented very good quality control, fgngiusions Based on the above observations, the inspectors determined that the licensee maintained very good QA/QC program S1 Conduct of Security and Safeguards Activities (81700) Scope The inspector reviewed the security program, as implemented, to determine whether it met the licensee's commitments in the NRC-approved security plan (the Plan) and NRC regulatory requirements. The security program was inspected during the periods of November 17-21,1997. Areas inspected included: management support and management effectiveness; audits and effectiveness of management controls; alarm stations; communications and assessment aids; protected area access control of vehicles; training and qudfication; and the vehicle banier system,
-- Obse: mons and Findinas Managemera support is ongoing as evidenced by the allocation of resources to allow supervisory and non-supervisory security personnel to participate in bench- marking initiatives, the procurement and transition of new weapons, procurement and installation of 4 a stand alone computer to enhance the control of safeguards information; manning levels have remained constant to permit effective program implementation and management is effectively administrating the security program. Audits were thorough and in-depth and effective controls were in place for identifying, resolving, and preventing programmatic publems. Alarm station operators were knowledgeable of their duties and responsibilities, communications requirements were being performed in accordance with the NRC-approved physical security plan (the Plan) and assessment aids had good picture quality and excellent zone overlap. Vehicles were being controlled in accordance with the Plan and security training was being performed in accordance with the NRC-approved training and qualification (T&O) pla Based on the inspector's observations and discusdions with security management, and licensee and contractor engineering, the inspector determined that the licensee's provisions for land vehicle control measures satisfy regulatory requirements and licensee commitments, Conclusions The inspector determined that the licensee was conducting its security and safeguards activities in a manner that protected public health and safety and that the program, as implemented, met the licensee's commitments and NRC requirement . _ _ . __ ._ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ -
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S2 Status of Security Facilities and Equipment (81700)
S2.1 Protected Area Access Control of Vehicles Scope The inspector determined whether the licensee properly controlled access of all vehicles to the protected area (PA) in conformance with the Plan and regulatory requirements, Observations, Findinas and Conclusions
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On November 1E,1997, the inspector observed s!te protection officers (SPOs) performing vehicle searches. Additionally, the inspector discussed vehicle authorization and escort requirements with security management and SPOs and determined that vehicles requiring PA access were being controlled as required in the Plan and applicable procedures, h S2.2 Alarm Stations, Communications and Assessment Aids E Scope The inspector determined whether the Central Alarm Station (CAS) s ad Secondary Alarm Station (SAS) were: (1) equipped with appropriate alerm, surveillance and communication capability, (2) continuously manned by operators, and (3) use independent and diverse systems so that no single act can remove the capability of detecting a threat oad calling for assistance, or otherwise responding to the threat, as required by NRC regulation < Observations and Findinas Observations of CAS and SAS operations verified that the alarm stations were equipped with the appropriate alarm, surveillance, and communication capabilities. Interviews with CAS and SAS operators found them knowledgeable of their duties and responsibilitie he inspector also verified through observations and interviews that the CAS and SAS operators were not required to engage in activities that would interfere with the assessment and response functions, and that the licensee had exercised communication methods with the local law enforcement agencies as committed to in the Pla Additionally, on November 18,1997, the inspector evaluated the effectiveness of the assessment aids, by observing on closed circuit television (CCTV), a walkdown of the P The inspector determined that the assessment aids in both alarm stations had good picture quality and excellent zone overla Conclusions The alarm stations, assessment aids, and communications met the licensee's Plan commitments and NRC requirements, o
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S5 Security and Safeguards Staff Training and Qualificatic : (51700) Scope The inspector determination whether members of the security organization were trained and qualified to perform each assigned security related job task or duty in accordance with the NRC-approved T&Q pla < Observations and Findinas On November 20,1997, the inspector randomly selected and reviewed T&Q records for
) nine SPOs. Physical and firearms requalification records were inspected for armed SPOs and site protection shift supervisors. The inspector found that the training had been
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conducted in accordance with the T&Q Plan and was properly documente During discussions with the security training staff and security management, the: inspector was informed that new response weapons were purchased to enhance the F ensee's tactical response capabilities and that transition training on the new weapons had been completed. Additionally, on November 19,1997, the inspector observed classroom
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requalification training on the use of oleoresin capsicum spray. The instructor was very knowledgeable on the subject matter and made an excellent presentation. The inspecor interviews i a number of SPOs to determine if they possessed the requisite knowledge and ability to c. cry out their assigned dutie Conclusions The inspector determined that training had been conducted in accordance with the T&O plan. Based on the S"Os responses to the inspector's questions and inspector's observations, the traa.ing provided by the security training staff was considered effectiv S6 Security Organization and Administration (81700) Scope The inspector conducted a review of the level of management support for the licensee's physical security program and the effectiveness of management relative to the r administration of the security program, Observations and Findinas The inspector reviewed various program enhancements made since the last program inspection, which was conducted in October 1996. These enhancements included ti e allocation of resources to allow supervisory and non-supervisory security personnel to participate in bench-marking initiatives, the procurement and transition of new weapons, procurement and installation of a stand-alone computer to enhance the control of safeguards information, and manning levels have remained constant to permit effective program implementatio _ _ _
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The inspector reviewed the Security Manager's position in the organizational structure and reporting chain. The Security Manager reports directly to the Vice President, Human and Administrative Services who reports directly to the Office of the Presiden Conclusions Management support for the physical security program was determined to be effective. No problems with the organizational structure that would be detrimental to the effective implementation of the secur'ty and safeguards programs were note S7 Quality Assurance in Security and Safeguards Activities (81700)
S7.1 Audits Scope The inspector reviewed the license 3's QA report of tha NRC required security program audit to determine if the licensee's commitments as contained in the Plan were being satisfied, Observations and Findinas The inspector reviewed the 1997 QA audit of the security program, conducted May 29 -
August 25,1997, (Audit No. S-TMI-97-05) and the 1996 QA audit of the titness-for-duty (FFD) program, conducted December 3,1996 - January 15,1997,(Audit No. O COM-96-01). The audits were found to have been conducted in accordance with the Plan and fitness for duty (FFD) rule. To enhance the effectiveness of the audits, the audit teams-included independent technical specialist The security audit report identified one deficiency associated with the course of fire used during weapons qualification. The FFD audit report identified one deficiency associated with procedural adherence during the collection of specimens. The inspector determined that the findings were not indicative of programmatic weaknesses, and the findings would enhance program effectiveness. The inspector determined, based on discussions with security management and FFD staff, and a review of the responses to the findings, that the corrective actions were effective, Conclusi9ng The audits revwed were very comprehensive in scope and depth, the findings were reported to the ppropriate levels of management, and the audit program was being effectively administere l l
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S7.2 Effectiveness of Manager 1ent Controls Sup_q The inspector conducted a review to determine if the licensee had controls for identifying, resolving and preventing programmatic problems, Observations and Findinas The inspector determined that mntrols were in place. They included the performance of a semi annual assessment by corporate security, the NRC-required annual QA audit, and a continual shift observation program by supervision. Aoditionally, the licensee has implemented a formalized self-assessment program which requires each site protection shift supervisor to conduct two assessments per year, in areas determined by the Security Manager. The licensee also utilizes industry data, such as violations of regulatory requirements identified by the NRC at other facilities, as a criterion for self assessmen Conclusions A review of documentation applicable to the licensee controls, including results, indicated that security performance errors were being minimized and were effectively implemented to identify and resolve potential weaknesse S8 Miscellaneous Security and Safeguards issues (Tl 2515/132)
S8.1 Vehicle Barrier System Backaround On August 1,1994, the Commission amended 10 CFR Part 73, " Physical Protection of Plants and Materials," to modify the design basis threat for radiological sabotage to include the use of a land vehicle by edversaries for transporting personnel and their hand-carried equipment to the proximity of vital areas and to include the use of a land vehicle bom The amendments required reactor licensees to install vehicle control measures, including vehicle barrier systems (VBSs), to protect against the malevolent use of a land vehicl Regulatory Guide 5.68 and NUREG/CR-6190 were issued in August 1994 to provide guidance acceptable to the NRC by which the licensees could meet the requirements of the amended regulation A letter dated April 23,1996 from the licensee to the NRC forwarded Revision 34 to its physical security plan that detailed the actions implemented to meet the requirements of 10 CFR 73.55 (c)(7),(8), and (9) and the design goals of the " Design Basis Land Vehicle" and " Design Basis Land Vehicle Bomb." An NRC April 22,1997, letter advised the licenset that the changes = bmitted had been reviewed and were determined to be consistent with the pro' .ans of 10 CFR 50.54(p) and were acceptable for inclusion in the NRC-approved securit,- pla ___
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This inspection, conducted in accordance with NRC Inspection Manual Temporary Instruction 2515/132, * Malevolent Use of Vehicles at Nuclear Power Plants," dated January 18,1996, assessed the implementation of the licensee's vehicle control measures, including vehicle barrier systems, to determine if they were commensurate with regulatory requirements and the licensee's physical security plan, Scope The inspectc' reviewed documentation that described the VBS and physically inspected the as-built VBS to verify it was consistent with the licensee's summary description submitted to the NRC, S
] Observations and Findinas The inspector's walkdown of the VBS and review of the VBS summary description disclosed that the as-built VBS was consistent with the summary description and met or exceeded the specifications in NUREG/CR-619 Conclusions The inspector determined that there were no discrepancies in the as built VBS or the VBS summary descriptio S8,2 Bomb Blast Analysis Scoce The inspector reviewed the licensee's documentation of the bomb blast analysis and verified actual standoff distances provided by the as-built VBS, Observations and Findinag The inspector's review of the licensee's documentation of the bomb blast analysis determined that it was consistent with the summary decription submitted to the NRC, The inspector also verified that the actual standoff dix s provided by their as-built VBS were consistent with the minimum standoff distances cak Jlated using NUREG/CR-619 The standoff distances were verified by review of scaled drawings and actual field measurements, Conclusions No discrepancies we:. noted in the documenta^ ion of bomb blast analysis or actual l
standoff distances provided by the as-built VB . _ . - - _ - -
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S8.3 Procedural Controls SenDa The inspector reviewed applicable procedures to ensure that they had been revised to include the VB Observations and Findinos The inspector reviewed the licensee's procedures for VBS access control measures, surveillance and compensatory measures. The procedures contained effective controls to provide passage through the VBS, provide adequate surveillance and inspection of the VBS, and provide adequate compensation for any degradation of the VBS, Conclusions The inspector's review of thn procedures applicable to the VBS disclosed no discrepancie X1 Exit Meeting Summary At the conclusion of the reporting period, the resident inspector staff conducted an exit meeting with TMI management on February 3,1998, summarizing Unit 1 inspection activities and findings for this report period. On November 21,1997, a regionalinspector conducted an exit meeting with licensee management summarizing Unit 2 inspection activities in the area of radiologicalliquid and gaseous effluents control progra TMl staff comments concerning the issues in this report were documented in the applicable report section. No proprietary information was identified as being included in the repor X2 Pre Decisional Enforcement Conference Summary On December 22,1997, a predecisional enforcement conference was held to discuss the events and issues involving apparent violations related to the 12R refueling activities. The details of the apparent violations are described in Inspection Report No. 50 289/97-09, dated December 2,1997. The meeting was held between the NRC and GPUN at the NRC Regien 1 Office in King of Prussia, Pennsylvania. The purpose of the meeting was to obtain information to enable the NRC to make an enforceme:.t decision, such as understanding of the facts, root cause(s), missed opportunities to identify the apparent violations sooner, corrective actions, significance of the issues and the need for lasting ai.d effective corrective actions. Handouts from the meeting are enclosed with this repor _ _ _ _ _ _ _ _ _ _ _ _J
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INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726: Surveillance Observations IP 62707: Maintenance Observation IP 71707: Plant Operations IP 71750: Plant Support Activities IP 81700: Physical Security Program For Power Reactors IP 84750: Radioactive Waste Treatment, and Effluent and Environmental Mwitoring IP 92901: Followup - Plant Operations IP 92902: Followup - Maintenance IP 92903: Followup - Engineering IP 92904: Followup - Plant Support ITEMS OPENED, CLOSED, AND DISCUSSED l Ooened VIO 9710-01 Failure to have a procedure reviewed and approved for RBEC valve leak check - SLIV NCV 9710-02 Failure to make 50.73 reports for missed TS surveillances completed satisfactorily withing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identificatio l URI 97-10-03 Review RBEC GL 96-06 submittal process to determine if any NRC requirements were violated with respect to design control Closasi VIO 97-06-01 Failure to Notify the State and County Offsite Agencies for the June 21 Unusual Even EA 97-117/
. VIO O2013 QCL Component Downgrade for Valves NR-V-1 A/8&C; VIO 02023 QCL Component Downgrade for Strainer Motors DR-S-1 A&B; {'
VIO 02033 QCL Component Downgrade for Auxiliary Ventilation Fans; VIO O2043 QCL Component Downgrade for Make-up valve MU V-17; VIO 02053 Failure to include EP-011 in the Safety Review Process; VIO 02063, 02073 Failure to Follow the Procedure Requirements of EP 01 LER 96-001-01 Seismic Qualification of Class IE 4160 VAC Westinghouse Breaker LER 96-002-00/01 Potential Unreviewed Safety Question Related to the Net Positive Suction Head for the Decay Heat Removal and Building Spray Pump _ _ _ _ _ _ _ _ _ _ _
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LIST OF ACRONYMS USED AB Auxiliary Building AEC Atomic Energy Commission ALARA As low As Reasonably Achievable AR Accountability Review (CAP) 3:00 pm AP Administrative Procedure (GPUN)
ASME American Society of Mechanical Engineers BS Building Spray System BWST Borated Water Storage Tank CAP Corrective Action Process CAS Central Alarm System CCTV Closed Circuit Television CFR Code of Federal Regulations CR Control Room l
CF,3 Control Room Operator l DBA Design Basis Accident DBD Design Basis Documents DC Decay Heat Closed Cooling Water System DCP_ Design Change Package DH Decay Heat Removal System DR Decay River Cooling Water System ECCS- Emergency Core Cooling System EDG Emergency Diesel Generator eel Escalated Enforcement issue EMR Early Management Review (CAP) 6:30am EP Engineering Procedure EPIP Emergency Plan and Implementing Procedure
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ES Emergency Safeguards-same as ECCS ESF Engineered Safety Feature ETTS Engineering Task Tracking System FFD Fitness for Duty GL Generic Letter GORB Group Offsite Review Board GPUN GPU Nuclear (Licensee)
HEPA High Efficiency Particulate HPl High Pressure Injection (MU)
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HPRB Human Performance Review Board IB Intermediate Building l&C Instrument and Control IFl Inspection Followup item IN Information Notice (NRC)
INPO Institute of Nuclear Power Operation IR Inspection Report IST Inservice Testing Program JO Job Order
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JPM Job Performance Measure LAR Licensing Action Request LCO Limiting Condition of Operation LER' Licensee Event Report *
LOCA Loss of Coolant Accident LOR Licensed Operator Requalification LPl Low Pressure injection (DH)
MIC Microbioiogically induced Corrosion MNCR Material Nonconformance Report MOV Motor Operated Valve MPFF Maintenance Preventable Functional Failure MR Maintenance Rule (NRC MTAN Maintenance Trend Action Notice MU Makeup System NSCCW Nuclear Service Closed Cooling Water System NCV Non Cited Violation NPSH Net Positive Suction Head (Pumps)
NI Nuclear Instrument NR Nuclear River Cooling Water System NRC Nuclear Regulatory Commissic $
NS Nuclear Service Cooling Water stem NSA' Nuclear Safety Assessment NOV Notice of Violation ODCM Offsite Dose Calculation Manual-OE Operational Experience OOS Out of Specification OTSG Once Through Steam Generator PA Protected Area PCR Procedure Change Request PMT Post-Maintenance / Modification test PORV- Power Operated Relief Valve (Pressurizer) '
PRA Probabilistic Risk Assessment PRG Plant Review Group OCL Quality Classification List ODR Ouality Deficiency Report QA Quality Assurance Q Quality Verification RB Reactor Building (Primary Containment)
RBEC Reactor Building Emergency Coolers RC Radiological Controls RCA Radiological Control Area RCE Root Cause Evaluation RCP Reactor Coolant Pump RCS Reactor Coolant System RG Regulatory Guide RMS Radiation Monitoring System RPS Reactor Protection System RR Reactor River Cooling Water System
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GUNUCLEAR -
i Three Mile Island Nuclear Generating Station NRC Predecisional Enforcement Conference Region I King of Prussia: December 22,1997 l
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Agenda 1.0 Introduction J. W. Langenbach 2.0 Event Descriptions & Assessments
- Missed PMT on the PORV M. J. Ross
- RCS Overfill Event
- OTSG Locked High Radiation Area D. Ethridge !
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- Hot Particle Skin Contamination 3.0 Enforcement Policy J. S. Wetmore Regulatory Assessments
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4.0 Summary & Conclusions J. W. Langenbach 5.0 IFI 50-289/97-09-01 H. Crawford Mid-loop Operations 12/22/97 2
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GPU NUCLEAR i
Introduction J. W. Langenbach Vice President and Director TMI
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Introduction
- Four apparent violations were identified in the NRC Inspection Report 97-09:
- Failure to perform the Tech. Spec. required PMT on the PORV in 11 R resulted in its inability to operate, if called upon, for the two year operating cycle from Oct.1995 to Sept.1997 for de-pressurization events. The calculated increase in CDF was 16%.
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- An SS failed to display a high procedure compliance standard during the RCS filling and venting, resulting in an excessive flow-rate and overflow of approximately 50 gals. of RCS water out the CRDM vents and an apparent violation of procedural control !
- A contract worker failed to follow a high radiation control procedure. This event appears to be a violation ofTech. Spe .8.1 in that procedures for locking high radiation areas were not followed. This event appears similar to prob! ems that occurred in 1993 and 1995 refueling outage ,
- Adequate radiological surveys were not performed during the reactor re-assembly activity. As such, adequate hot particle controls were not in place and did not prevent a personnel skin contamination potentially violating Tech. Spec. 6.11. -
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Introduction (Continued)
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- GPUN agrees that failure to perform the required Post Maintenance Test following the PORV replacement in the 11R is a violation of 1 Tech. Spec. 4.2.2 IST requirement ;
- GPUN recognizes that the SS acted in poorjudgement while carrying out the fill and vent evolution, in that sufficient caution was not adhered to per procedure guidanc GPUN agrees that a contract worker failed to follow procedures with regard to high radiation controls. However, electronic surveillance
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was maintained providing continuous access control during this event.
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GPUN believes that the new defense-in-depth method employed in 12R makes this event dissimilar to any previous problems with locked high rad area control GPUN recognizes that the hot particle controls program contains weaknesses which can be improved upon. GPUN agrees that this event can be considered a violation of Tech. Spec. 6.1 /22/97 5 l
Executive Summary
- GPUX agrees that these apparent violations were !
serious events that warrant management attention and follow-u Al; events were Licensee identifie Immediate and comprehensive corrective actions were implemented in quick response to the even Long-term action plans were developed to prevent recurrenc /22/97 6
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Missed PMT Event l
M. J. Ross Director, Operations & Maintenance
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Missed PMT Event Description
- Event Overview::
Failure to perform the Tech. Spec. required PMT on the 11R PORV resulted in a failure to discover a wiring error, where a technician incorrectly landed the lead intended for terminal Sl of the PORV actuation solenoid. The PORV was unavailable for service if called upon until its replacement during 12Ps.
- Sequence of Events:
- 11R wire landed incorrectly on the PORV solenoid; i independent verification failed to identify the problem; and, post maintenance test (PMT) was not performed after re-connectio R as part of the cooldown test sequence, while opening the PORV, the fbses blow. Valve replaced and solenoid mis-wired again, testing causes fuses to blow and engineering investigates causes. Valve rewired and tested appropriatel .
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C Missed PMT Event Assessment (Root Cause Evaluation & Corrective Actions)
- Root Causes:
- Less than adequate self-checking during 11R installatio ,
- Less than adequate independent verification followe Less than adequate procedure structure / content and l usage, specifically: no PMT checkoff provided; ,
excessive cross-referencing / branching used; and !
incomplete guidance (content) in the Job Order Package l
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to direct the performance of a PM Contributing Factors:
- The lack of detail, and possible erroneous information on certain design drawings; and, failure to use the vendor manual to augment or check the design drawings during the independent verificatio /22/97 9
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Missed PMT Event Assessment
(RCE & Corrective maons - Continued)
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- Corrective Actions:
- Interview and coach the technician who improperly '
landed the lead and failed to discover it during self-checking, and the independent verifier who did not
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detect the landing error. The purpose of the interview is to critically examine the technician's personal self-checking technique and to coach the technician and independent verifier on how to improve applicable work practices. - Actions Complete The requirement to perform PMT for this activity shall be clarified through procedure changes to 1041,1401-2.1 and to the Job Order program. Procedure ovmers i shall confer to reduce cross-referencing and branchm In progres /22/97 10
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Missed PMT Event Assessment ~
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(Corrective Actions - Continued)
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The Job Order (JO) shall clearly list all required PMTs in one location so that it can be reviewed for completion with surety. - A memo and briefing were provided to all planners on AP-1071 "PMT ,
Guidelines," as it relates to JO development. A training handout was provided to O&M penonnel to ensure time proper conduct of PMT Print SS-209-034 and the Electrical Cable Information System changed to include sufficient detail to allow verification oflanded leads. - In Progres .
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Missed PMT Event Assessment (Corrective Actions - Continued)
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- The vendor manual diagrams which clearly show the correct wiring configuration shall be included in the Job l Order Package for this task for subsequent performances of this task. - In Progres A process study including statistical analysis of how PMT renuirements for other components and systems are controlled will be performed to identify and correct programmatic weaknesses. - In Progras /22/97 12 i _
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, GU NUCLEAR RCS Overfill Event M. J. Ross Director, Operations & Maintenance
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RCS Overfill Event Description Event Overview:
- 15 Oct.1997 - day 40 of the 12R refueling outage, with Decay Heat System in service per OP 1104-4. The filling and venting of the Reactor Coolant System (RCS) was in progress using OP 1103-2. At l 0651 hours0.00753 days <br />0.181 hours <br />0.00108 weeks <br />2.477055e-4 months <br />, an SRO stationed at the Reactor Vessel head saw water issuing from several of the Control Rod Drive Mechanism (CRDM)
vents. The event was terminated when an operator closed DH-V-58, which had been opened a few minutes earlier to admit water to the RCS from the Borated Water Storage Tank (BWST).
Event Consequences;
- The reactor vessel head and CRD components were not degraded by the spill of borated wate One stator was initially affected by moisture in the connector, but eventually achieved acceptable resistance reading Some loose surface contamination was added to the reactor head assembly and associated areas, which did not require additional radiation protection measure /22/97 14 E
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RCS Overfill Event Description (Continued) .
- Event Summary:
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- A Clarification:
There were no specific violations of the Fill & Vent procedure, but the SS failed to comply with several
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normal work practices as specified in AP-1029,
" Conduct of Operations," namely:
Pre-job briefing; Use of 3-point communications; and, Use of proper communications equipmen /22/97 15
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RCS Overfill Event Assessment (Root Cause Evaluation & Corrective Acticas) 1
- Problem Statement A flow path from the BWST to the RCS was initiated at a time when it was inappropriate to do so because of the risk of overfill while the water level was within the relatively small volume of the CRD motor tube Root Causes: .
- Information misunderstood: SS did not understand Management's
, expectation that the BWST would not be used for filling the RCS when the pressurizer level was above 100."
- Incorrect Assumptions: the SS assumed he was filling to 390" in the Pressurizer and needed much more water than available in the RCB Conduct of the Evolution: the evolution, which was significant was started during a Shift Turnove .
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RCS Overfill Event Assessment (RCE & Corrective Actions - Continued)
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- Corrective Actions- !
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Modify the format of the Ops. Dept. Outage Shift Turnover Meeting to include a final summary " repeat back" by the on-coming Shift Supervisor.- Complete Revision of procedure , iO4-4 to provide a more specific warning concerning use of the BWST as a fill source. The warning shall specify a pressurizer level of 100" or above which fill I
from the BWST is prohibited. - Complete Emphasize Management's expectation that it is inappropriate to perform significant plant evolutions while the shift turnover is in progress. - Complete /22/97 17
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RCS Overfill Event Assessment
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(RCE & Corrective Actions - Continued)
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- Contributing Factors:
l Supervisory Methods:
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Tasks and individual accountabilities were not made clear to workers.
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Crew teamwork was less than adequat Verbal Communications:
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The consequences of potential errors were not discussed before starting wor Inaccurate messages were transmitted due to informal methods used for communication.
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Improper communications equipment was use /22/97 18
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RCS Overfill Event Assessment (RCE & Corrective Actions - Continued)
- Corrective Actions For Contributing Factors:
- Ops. Dept. will re-evaluate methods used to observe and evaluate the team skills ofsupervisors, foremen and team members. Ensure awareness of the special challenges in supervision and communication for temporary replacement personnel on-shift, by use of focused crew briefings providing opportunity for discussion and feedback. - In Progress
- Continue to reinforce the use of"3-Point" Communications per AP-1029 for communication involving task assignments and all work groups that interface with the control room. - Ongoing.
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- Management will review this event with all Ops and Training
personnel to stress the expectations regarding the use of the M&I phone system for major plant evolutions and that the conduct and content of pre-job briefings be pursuant to procedure AP-102 Ongoin /22/97 19
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Hot Particle Event D. Ethridge Radiological Health / Safety Dir., TMI
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OTSG Access Event i
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l D. Ethridge
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Radiological Health / Safety Dir., TMI
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OTSG Access Event Description
> - For 12R, TMI used an improved method oflocked high -
radiation control for OTSG manways to proactively provide defense-in-depth and support better task ALARA:
- Continuous Rad Con Technician (RCT) surveillance of OTSG manway via camera;
- Presence of a platform worker in the area while the
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OTSG manways were unlocked; and,
- Communication between the platform worker and the RCT observing the camer The worker's defined responsibilities, which were a part of specific task training, included:
- The expectation that the worker would remain in the area when the OTSG was unlocked; and,
- To respond as directed by the RCT to prevent inadvertent acces /22/97 21
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i OTSG Access Event Description Sequence of Events:
- At completion of his task, the platform worker left the area with the OTSG manway shield door shut, unlocked and monitored by camera.
- After about one hour, the RCT monitoring via camera ,
realized the platform worker was not present, immediately
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contacted personnel in the Reactor Building, and had the shield door locke i i
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OTSG Event Assessment (Root Cause Evaluation & Corrective Actions)
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- Root Cause: 1
- The worker's understanding of his responsibilities was l less than adequat Corrective Action:
- Prior to additional entries, each OTSG worker was required to properly state their locked high radiation area control and access responsibilitie Event Consequences:
- A procedural violation that was mitigated by continuous surveillance of the unlocked manwa /22/97 23
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OTSG Event Summary l
- A contracted craft personnel failec. to follow the GPUN locked high radiation arocedur Constant remote surveillance was maintained during the event, such that, positive control of the area was maintained to prevent unauthorized personnel entr This event is dissimilar to any prior event
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characterized by an inadvertent loss of access control to a locked high rad area, due to the defense-in-depth approach used by GPUN'.
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Hot Particle Event Description Sequence of Events:
- Fuel transfer canal drained, and cleaned to release from hot I particle controls. The reactor vessel head seal plate was to be lifted and parked. A pre-job discussion was conducted between the work crew and Rad Con Technician (RCT).
- Upon raising the seal plate, a survey was performed and numerous hot particles were discovered by the RCT. The RCT reorganized the work effort, and directly supervised the removal of hot particles from the are A hot particle area was not formally established, nor was Rad Con supervision notified of the problem.
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Hot Particle Event Description (Continued)
- Following an assessment by the RCT that the hot particle i problem was resolved, the crew performed seal plate gasket removal. Upon completion, two hot particles were identified on the face of a worke Additional hot particle cleanup was required prior to work resumptio The worker was decontaminated. The assessed dose was 14 rem skin and 0.05 rem deep-dos /22/97 27
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Hot Particle Event Assessment '
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(Root Cause Evaluation & Corrective Actions)
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- Root Causes:
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- Specific actions to address emergent hot particles were not clearly stated in the RWP procedure.
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i l - Contributing Factors:
- Rad Con supervision did not grticipate in the pre-job plannin Communications with Rad Con supervision prior to and during the task were not sufficient.
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Hot Particle Event Assessment (RCE & Corrective Actions - Continued)
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- Active ALARA Re iews and Radiation Work Permits
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were reviewed and no other ambiguous hot particle l i
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control requirements were identified. - Complete The hot particle controls program was revised to address weaknesses identified. - Complete .
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- - A 12R Lessons Learned action was issued to improve
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planning. Changes to be implemented prior to 13 !
12P_2/97 29
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Hot Particle Event Summary
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- Using the existing procedural guidance, the RCT responc ed to the uncovering of the hot particle GPUX agrees that the TMI hot particle controls arocedure needed to be strengthene Rad Con Supervisor notification.
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- More rigid requirements specified:
- Improved monitoring requirement "Shoulds" changed to "Shalls."
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GU NUCLEAR Enforcement Policy
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Regulatory Assessment J. S. Wetmore Manager, Nuclear Safety & Licensing
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Enforcement Assessment for Apparent Violations Inspection Report No. 50-289/97-09
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Atparent ApplianeSedion(s)of 'Identifcatiert Corra: tim Bforammt Violation (s) BformnentPblicy Craft Adiort Cmit IMscretiort Faihreto conduct required Supplement. N'A N'A N'A Post MaintemnceTest Section D.3-SestsityInti (PNTI)in violationof IV "Afaihretonret TechnicalSpecificatio regulatoryrequirementstlut I 4.2 -ISTrequaaiui hastnxxethanminorsafety emiimiutai significance" -------------- ---------------------- --------------------
Yes Yes- Aftmired whrremxted Yes. Isohtedoccmtnc ad tested,perIST mpiaiuis. Boadbased NSAnonitoringconductedto detcrnireotherpotertialarras ofmissed PMrs. Img-term cOITetforactionsfiumud Failure to folknv Supplemert N'A N'A N/A procedurts bytheSS GPUNhas concluded that this dtring RG fill &sent proceduresiolation does not emlutioit . meet the threshok! fora I
Inti-III violatiort
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PossibleNCVcaruhdate Yes Yes-SS cxxxiseled. and Yes. Minirmisafay p r o xlure 3o u .di o u consequence /22/97 32
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Enforcement Assessment for Apparent Violations Inspection Report No. 50-289/97-09 4ymanf AynalleSettiors(s)of Idss@Eamniorr Sgistrvrerr lCarershe Ma(aeion(s) BforoprertIhfry Grudf i AaliorrCrair IMst7tidorr VxisknofTednici Sypicnut IV N'A N'A N'A 4 Specifictkn 68.I,in tlur, GLNInsmdahldW tlis ticpunireRr pocxxkresiokakndrstu mutarmyavtrd owxa nustirtireshoid Era {
kxiallighradorinata IMIII violskn acass pnh ntt - ------------------ -------------- ---------------------- ----------- --------
strialyadunit IbmNeITVcutkine Ye Yes Imrulisecomrtiw Yes Axxssautrdwas aaxntakcngxnditunu nuirtaineti VxitinofTainical Supplaint I N'A N'A N'A Spxilictkm 6.11,in tium, GPUNinsardukxltin tlis ticpixuiral punkmalvcakrussdxsnot raymuruts Erfut nus tirtividnkiRra puticleartrdswueless IrwilII siolatin timadetpste --------------------- -------------- ---------------------- --------------------
IhmNelTVankkte Ye Yes. Innulineanutim Yes NacquaIts adianstakut ouirmlabomrryavay 111gtamu uulicaciorsin orassiwdiclimits progesstoptriudere-
OCCLITUtraITpICfCd
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12/22/97 33 l
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GU NUCLEAR l
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Summary & Conclusions
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J. Langenbach
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Vice President and Director TMI
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Summary & Conclusions
- GPUN agrees that these apparent violations were serious events that warrant management attention
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and follow-u All events were Licensee identified, immediate corrective actions were taken and planned long
- tern actions as appropriate.
t - There were no impacts upon safety; the occurrence of a personal skin contamination did not exceed regulatory or TMI's admmistrative radiation erotection limit None of the events is representative of a Severity Level III violation in NRC's Enforcement Polic /22/97 35
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GU NUCLEAR
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IFI 50-289/97-09-01 l Mid-loop Operations l
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Howard Crawford
! Manager, Equipment Reliability Programs ,
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TMI-1 Risk Management Philosophy
- Outage Fuel Protection Criteria (OFPC) Defense in depth and critical safety function (CSF)
approac i 2. Defines amount of equipment to be protected for each CS . Remain above " Required" mode (Tech Specs), try to maintain " Enhanced".
- Plan drain down work during drain down after new fuel loaded (reduced decay heat).
- Do not drain down while core is off-loaded because: Increase personnel radiation exposur . Increase risk with fuel in spent fuel poo . Increase outage length (2 Days).
12/22/97 37
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TMI-1 Pre 12R Risk Management Pre-Outage Review Identified DH-P-1B pump replacement while fuel transfer canal filled.
, "A" Decay Heat Removal System RCS check valve and discharge valve to be worked during RCS mid-loop condition after new Cycle 12 fuel is loade t 12/22/97 38
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TMI-1 Emergent 12R Work
- During cooldown, DH-P-1B developed a gasket lea Decision made to take 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of critical path time to repair pump (pump is scheduled to be replaced later in outage) while "A" Decay Heat Removal System was operating and both OTSGs available for heat removal.
- While operating, DH-P-1 A developed a seal leak. Pump could still perform its decay heat removal function.
- Revised schedule to fix seal leak after DH-P-1B replacement and prior to start of fuel transfer canal drain dow l
- Task completed in accordance with revised schedule, but new seal also leake /22/97 39
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PRG Risk Review Meeting Options Discussed Keep fuel transfer canal Sooded until seal leak is .
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fixe Leave pump as is with seal leak (able to perform c ecay heat removal function) and =cpair later in outage when OTSGs are availabl Continue to work outage schedule as planned and return the pump to full service as soon as possibl Option chose /22/97 40
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Risk Considerations l For Repair Path Selected
- BWST ("an equivalent reservoir as an available heat sink" ,
per T.S. bases) gravity drain will remain availabl Pre-outage risk evaluation of schedule included taking "A"
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Decay Heat Removal System out of service during second drain dow DH-P-1 A could be returned to a functional condition in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> at any time during the work on the seal.
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- Det erred work on "A" Decay Heat Removal System discharge valve. This would allow the DH-P-1 A (once returned within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) to be used with the "B" side pipin I2/22/97 4I l
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IFI - Summary Inspection Report 97-09 identifies two concerns: 1
- The evolution appears to reflect a poor safety interpretation of the Tech. Spec. bases
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statements; and,
- Management's decision to enter mid-loop condition with only one means of decay heat removal available appears to be a poor minimization of shutdown risk.
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