IR 05000271/1997011
| ML20198P545 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 12/23/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198P509 | List: |
| References | |
| 50-271-97-11, NUDOCS 9801220150 | |
| Download: ML20198P545 (26) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
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Docket No.-
.50-271 Licensee No.-
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Repon No.
97-11 Licensee:
Vermont Yankee Nuclear Power Corporation Facility:
Vermont Yankee Nuclear Power Stat:on--
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Location:'
Vernon, Vermont
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Dates:
October 31 - December 6,1997 Inspectors:
William A. Cook, Senior Resident inspector
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Edward C. Knutson, Resident inspector Approved by:
Curtis J. Cowgill, lil, Chief, Projects Branch 5 Division of Reactor Projects
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EXECUTIVE SUMMARY Vermont Yankee Nuclear Power Station NRC Inspection Report 50 271/97-11 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a five week period of resident inspection.
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Operations Control room operator response to the November 25 reactor scram was good and in accordance with off normal and emergency operating procedures. A significant contributing cause of the automatic reactor scram was a failure to provide an appropriate level of review and approval of the 345 KV switching orders which was contrary to Technical Specification 6.5 and a cited violation (VIO 97 11 01). Station management and PORC response to this transient and the resolution of plant technical and human performance concerns prior to unit start-up were appropriate. The temporary modification and associated actions to address the generator runback circuit problem were adequately implemented, but operator train!ag on the TM was weak and administrative oversight not comprehensive.
Violation 96-05-01, plant operation impact of high pressure coolant injection (HPCI) system design change, was reviewed. VY's corrective actions appropriately addressed the causes
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of the violation, as evidenced by no similar occurrences since the event. Accordingly, this violation was closed.
During a regularly scheduled NSAR committee meeting, the inspector observed that the presentations were informative and that many of the NSAR committce members actively engaged the presenters in additional discussion. The inspector determined that the agenda had been revised to focus routinely on the known problem areas of corrective action effectiveness and human performance. _ Based upon this review and the resident inspector stafi's routine inspection and assessment of the off site committee's plant oversight functions, inspection follow item 96 200-06 was closed.
Maintenanca VY's quality control measures for the brazing activities witnessed during inspection 50-271/95-10were determined to have been in accordance with ANSI B31.1 requirements and established VY welding and OC inspection guidelines. Accordingly, unresolved item 9510 01 was closed.
Unresolved item 95-14-01, concerning maintenance requirements for the alternate cooling system, was reviewed. Routine preventive maintenance requirements for alternate cooling
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system and RHRSW piping were found to have been properly implemented, and the unresnimd item was closed.
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Unresolved item 95 25 04,concerning an administrative error in the teat gas mixture for Advanced Offgas (AOG) system hydrogen monitor Technical Specification (TS) surveillance testing was reviewed. The licensoe's actions to address this error were appropriate and the previously observed weaknesses in VY's biennial procedural revision process have been adequately addressed, it was concluded that the error did not constitute a violation of NRC requirements. This unresolved item was closed.
The inspector identified an inconsistency in the licensee's documentation pertaining to the expected life of the AOG hydrogen recombiner catalyst. The inspector reviewed the individual action plan items and concluded that VY had implemented reasonable and timely measures to monitor hydrogen recombiner catalyst performance. The inconsistency did not constitute a violation of NRC requirements. The associated unresolved item, URI 96-03-07, was closed.
The inspector reviewed Licensee Event Report (LER)97-015, concerning an instrument check that was not completed in accordance with Technical Specifications due to personnel error. Review of the event had been previously documented in inspection report no. 50 271/97 06and dispositioned as a non-cited violation. The LER was determined to have been appropriately written and properly submitted in accordance with 10 CFR 50.73, and was closed.
Enaineerina Unresolved item 9519 01,regarding the adequacy of controls for the conduct of system manipulations for performance monitoring, was reviewed. This issue was the result of an inspector observation of system performance monitoring, during which a procedural limit was found to have been exceeded. Inspector follow up and day to-day monitoring of plant activities indicated that the type of intrusive system perfctmance monitoring conducted during inspection period 9519 was isolated and not a programmatic concern. The failure to have properly operated the RBCCW system per OP 2181 was a violation of minor safety significance and not cited. Accordingly, the unresolved item was closed.
Violation 96-05-04,concerning deficiencies in VY's addressal of a battery room wall seismic qualification issue, was reviewed. The inspector noted that a significant number of design issues have been identified by VY during 1997, and that these issues have been appropriately dealt with within the corrective action program. Accordingly, this violation was closed.
Violation EA 96 210, concerning a single failure vulnerability of the RHR minimum flow valves, was reviewed. Although VY was credited both with identifying the issue and taking prompt, comprehensiva corrective action, enforcement action was taken due to the length of time that the condition had existed and the number of orior opportunities that existed to identify and correct the condition. The significant number of design issues that have been identified through VY's design basis documentation program indicates that the licensee has taken action to address this issue. Accordingly, this violation was closed.
The inspectors reviewed LER 96-015, concerning the omission of overpressure relief protection for piping sections that are isolated by the primary containment isolation iii
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system. Based upon tese.VY staff having identified this design issue and having
implemented comprehensive and timely corrective' actions to resolve this "old design.
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issue," it was not cited, in accordance with the NRC Enforcement Policy, Section Vll.B.3.
LER 96-015 and associated unresolved item, URI 96 08 02, are closed.-
~ Plant Suooort Unresolved item 97-03 07, criticality accident requirements, was reviewed. This issue
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involved the failure to have in place either a criticality monitoring system for storage and handling of new (non-Irradicted) fuel or an NRC approved exemption to this requirement contained in 10 CFR 70.24. Although it was concluded that a violation of 10 CFR 70.24
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existed, the NR!: determined that, for this issue, it is appropriate to exercise enforcement-discretion for violations involving special circumstances in accorda.1ce with Section Vil B.6 of the Enforcement Policy. Pending the amendment to 10 CFR 70.24, further enforcement action will not be taken for failure to meet 10 CFR 70.24 provided an exemption to this regulation is obtained before the next receipt of fresh fuel or before the next planned movement of fresh fuel.
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TABLE OF CONTENTS
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EX EC UTI VE SUM M ARY.............................................. ' il TABLE O F CO NTENT S.. -...................... -........................ v
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Summ ary of Plant Status............................................ 1 1. Operations
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Conduct of Operations.................................... 1 01.1 Reactor Scram Due To Switchyard Electrical Transient......... 1 08' - Miscellaneous Operations issues. -............................. 5 -
08.1 (Closed) Violation 96-05-01: Plant operation impact of high pressure coolant injection (HPCI) system design change.............. 5
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i 08.2 -(Closed) Inspection Follow item 96-200-06:- Follow up on Nuclear Safety and Audit Review (NSAR) Committee plans to resolve known proble m a re a s..................................... 6
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I I. M a int e n a nc e................................................... 7 M1 Conduct of Maintenance................................... 7 M1.1. Maintenance Observations............................ 7 M1.2 L Surveillance Observations............................. 8 M8 Miscellaneous Maintenance issues............................ 9 M8.1 (Closed) Unresolved item 95-10-01: Quality Control inspection and verification of in process brazing........................ 9 M8.2 (Closed) Unresolved item 95-14-01: Verify maintenance requirements for alternate cooling system........................... 9 M8.3 (Closed) Unresolved item 95 25-04: Advanced Offgas (AOG) system hydrogen monitor Technical Specification (TS) surveillance requirement identifies an incorrect test gas mixture.......... 10 M8.4 (Closed) Unresolved item 96-03-07: Resolution of inconsistent recombiner catalyst lifetime in the AOG system -............ 11
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M8.5 (Closed) LER 97-015: Instrument check not completed in accordanca with Technical Specifications due to personnel error......... 11 M8.6 (Closed) Unresolved item 96 03 04: Torus water temperature administrative limit reduced........................... 11 Ill. Enginee ring.................................................. 12 E8 Miscellaneous Engineering issues..........................,. 12 E8.1- (Closed) Unresolved item 95-19-01: Adequacy of controls for the
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conduct of system manipulations for performance monitoring... 12 E8.2- (Closed) Violation 96-05-04: Battery room block wall seismic qualification methodology acceptability and operability determination i m p a c t c...................'..................... 1 2 E8.3 (Closed) LER 96-001, Supplemer't 2:- Technical Specification 4.6.E -
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i Program sc ope.................................... 13
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E8.4 - (Closed) Violation EA 96-210:= RHR Minimum Flow Valve Single-
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Failure Vulnerability...................'.............. 13 E8.5 : (Closed) NCV 9711-02 and LER 96-15: Original B31.1 ANSI Code -
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Section that required overpressure relief for isolated pip!ng was not'
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considered during original design. -...................... 14
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E8.6 (Closed) LER 50-271/96 028: Inadequate field labeling of safety class wiring and drawing updates result in the failure to maintain electrical separation requirements described in nuclear safety system design bases.
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IV. Plant Support
................................................16 P8-Miscellaneous EP issues................................... 16
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P8.1 - (Closed) NCV 97-11-03 and Unresolved Item 97-03-07: Criticality i
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Accident Requirements............................. 16
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-F8 Miscellaneous Fire Protection issues -......................... 17 i
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F8.1 (Closed) Unresolved item 95-21-02: Appendix R and Fire Protection Program implementation issues......................... = 17
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V. Management Meetings
..........................................17-X1 Exit Meeting Summary.................................... 17
.X3 Review of Updated Final Safety Analysis Report (UFSAR)............17
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INSPECTION PROCEDURES USED..........................,,......... 18
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ITEMS OPENED, CLOSED, AND DISCUSSED.............................. 19 PARTI AL LIST OF PERSONS CONTACTED,... '........................... 20 LIST OF ACRONYM S U SED.......................................... 21
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Report Details Summarv of Plant Status
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- At the beginning of the inspection period, Vermont Yankee (VY) was operatir'g et 100 percent power. During the period of November 7-12, plant power was reduced to
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approximately 60 percent for single rod scram time testing and for localized core power suppression testing to identify the location of a leaking fuel rod that developed on October -
26. This testing succeeded in identifying the single leaking fuel bundle. A control rod configuration was developed to suppress power generation in that location and allow the return to full power operation. The plant was returned to 100 percent power on November 16. On November 25, an electrical transient on the main generator due to a maintenance activity in the VY 345 KV switchyard resulted in a reactor scram. The scram was uncomplicated, and plant restart was commenced the following day. The main generator was closed on the grid on November 27 and full power operation was achieved on December 1. With the exception of a power reduction to conduct planned surveillance testing, the plant operated at 100 percent power for the remainder of the inspection period, l. Operations
Conduct of Operations'
01.1 Reactor Scram Due To Switchvard Electrical Transient a.
Insoection Scoos (93702,71707)
At 6:48 A.M. on November 25, a reactor scram occurred due to an electrical fault that occurred during manipulation of a line disconnect in the 345 KV switchyard at VY. The inspectors reviewed the circumstances that led up to the event, observed operator response to the scram, and monitored VY's actions to identify the cause and prepare for plant restart, b.
Observations and Findinas Background During power operations, VY is normally connected to the electrical grid by three outgoing power lines. These lines are normally supplied by the main generator through a ring bus in the VY 345 KV switchyard. Circuit breakers (interrupt devices that are designed to be operated under electricalload)in the ring bus allow the individual outgoing linos _to be de-energized, if required. Disconnect switches (interrupt devices that are used for maintenance isolation and not designed to be operated under load) on the outgoing lines allow the lines to remain isolated from the ring bus, so that the circuit breakers originally used to isolate a line can be re-closed. This ability is important because the configuration of the ring bus affects
' Topical headings such as 01, M8, etc.,'are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic.
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2 the amount of load that the main generator is able to transmit to the grid. The alignment of the distribution system is controlled by Vermont Electric Power Company (VELCO) through an offsite dispatcher. Operations department personnel perform switching orders, as directed by VELCO, on that portion of the distribution system that is located at VY.
The Electrical Transient j
On the morning of November 25, VELCO was preparing to conduct maintenance on one of the outgoing lines from VY (the Scobie line, circuit 379), which required this line to be de energized. The standard process for de-energizing an outgoing line was to open the associated circuit breakers in the ring bus (one on either side of the line), and then to open and lock the disconnect switch. This sequence precluded any operation of the disconnect switch when the conductors on either side of the switch were energized. On this occasion, however, VELCO was concerned for ring bus stability, due to the existing grid conditions. To expedite reestablishing the ring bus, the sequence that was used was to open the associated circuit breakers, open the disconnect, and then re close the circuit breakers prior to locking the disconnect open. As a result, the VY side of the circuit 379 disconnect was energized at the time that operators were directed to lock the disconnect roen.
Control over the position of a disconnect switch during maintenance can be maintained either by danger-tagging power to the motor operator in the off position, or by disconnecting the operating mechanism from the motor operator and installing a lock to prevent the mechanism from being rotated, in this instance, VELCO directed that the latter technique be used. Two VY auxiliary operators (AOs) were dispatched to the 345 KV switchyard to perform this operation. After the operating mechanism had been disengaged, the operators found that the holes for attaching the lock (one hole in the rotating mechanism and the other in a fixed portion of the unit) were not al;gned closely enough to allow the lock to be installed. One of the operators attempted to rotate the operating mechanism slightly to align the holes; however, a significant amount of force was required to achieve breakaway torque.
As a result, when movement did occur, the mechanism rotated further than had been desired. Rotation of the operating mechanism caused the three conducting
. arms of the disconnect switch (located about 50 feet above the operators) to rotate back towards the closed position. As a result, the conducting arms were sufficiently close to the main line contacts to draw an arc from the energized VY side through to the de-energized circuit 379. The operator noted that this condition existed for several seconds before he was able to rotate the mechanism back to the open position and install the lock.
Plant Response The resulting electrical transient on the main generator initiated a turbine runback to reduce generator load. The protective feature that actuated was the stator cooling runback circuit, which initiated upon detecting a mismatch between stator water cooling system flow and main generator output current (which was high due to the arcing across the circuit 379 disconnect). The turbine runback is designed to
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reduce main generator output to approximately 28 percent over a period of two
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minutes, and should only occur while a mismatch condition exists. in this instance, however, the turbine runback continueo to completion, even though the condition that required it existed for only a few seconds. The stator cooling runback circuit is not part of the reactor protection system; its function is equipment protection, and it is classified as a non-nuclear safety system.
The affect of a turbine runback is to cause the ten main steam bypass valves to
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automatically open, as required, to maintain 100 percent steam flow from the reactor. A secondary affect is that feedwater heating is reduced, because the steam supply for the feedwater heaters is extraction steam from the main turbine; as turbine load is reduced, less steam is available for feedwater heating. The result of lower feedwater temperature, in turn, is to add positive reactivity to the reactor (due to the negative temperature coefficient of reactivity) and therefore, to cause
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reactor power to increase. To prevent an automatic reactor protective action, operator action is required to compensate for this response by reducing reactor power. In this instance, operators responded appropriately in accordance with operating procedure OT-3119," Loss of Stator Cooling" and reduced reactor power i
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by reducing reactor recirculation flow. However, reduced recirculation flow also causes the high flux reactor scram setpoint to be reduced (due to flow biasing), and additional power reduction by control rod insertion is required to avoid an automatic scram. As the control room operators attempted to continue the manual reactor power reduction by control rod insertion, an automatic reactor scram occurred at approximately 85 percent reactor power.
l Plant response to the scram was uncomplicated and safety systems responded as designed. The inspector observed operator response and determined that it was deliberate and in accordance with appropriate procedures. In working to stabilize plant conditions, the inspector noted that the operators were concerned about lowering main condenser vacuum. Steam to the main condenser air ejectors had been secured earlier in the recovery to reduce the reactor cooldown rate; however, vacuum decreased more rapidly than anticipated and was approaching the point at which the main steam isolation valves (MSIVs) would be required to be shut.
Closure of the MSIVs was not desirable because it would eliminate the main condenser's availability as a heat sink, and require use of emergency core cooling systems for residual heat removal. Operators succeeded in arresting the decrease in vacuum, and eventually in restoring normal main condenser vacuum, without having to close the MSIVs. Securing auxiliary steam loads to control plant cooldown rate is allowed by procedure; however, the inspector considered that securing steam to the air ejectors to control a coo!down that was the result of feedwater addition may have been premature, given that a significant amount of steam was still being rejected to the main condenser. This observation was acknowledged by Operations management who stated that it would be examined as part of the post-trip. critiqu.
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Management Oversight The inspectors attended a series of department manager and Plant Operations Review Committee (PORC) meetings conducted following the November 25 plant scram and prior to unit start up on November 26. The inspectors observed that these meetings were well atter.ded, purposeful, and generally effective. The department manager meetings held shortly after the scram established clear assignments for event / problem follow-up and clear expectations for the plant staff assigned to these tasks. The post-scram / pre start up PORC meeting (meeting No.
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97-90) held on November 26 was observed by the inspector to have been especially
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useful in identifying the human performance shortcomings which directly contributed to this event. It was also usefulin developing the appropriate recommendations to the plant manager for addressing these shortcomings prior to unit restart.
The inspector noted that the PORC members' review of the preliminary root cause evaluation for this event was extensive. The PORC identified that a significant contributor to the November 25 reactor scram and the April 24,1997 reactor scram was the f ailure to provide effective oversight for an activity that had plant impact j
potential. Specifically, the licensee did not review the tagging order. The order revised the standard practice of maintaining the appropriate portion of the ring bus
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completely deenergized while disengaging the drive gear. This failure to have reviewed and approved the 345 KV switchyard switching orders was a violation of Technical Specification 6.5 and was cited (VIO 97-11-01).
The inspector's observation of the PORC's examination of the stator cooling runback sequence and generator protection circuit response, again, noted excellent.
PORC member participation and safety assessment. Troubleshooting and testing of the circuit was inconclusive in determining why the runback signal sealed-in for one minute and 18 seconds. The voltage / current transient resulting from the disconnect arcing lasted only five cycles ( 5/60th of a second ) as recorded by the generator supervisory circuit and plant process computer. Additionally, technical staff examination of the runback circuit identified that the current / flow comparator was a 24 VDC supplied component functioning in a 125 VDC protective circuit. VY engineering staff review of this anomaly concluded that this was not a desirable design, but not necessarily the cause of the circuit seal-in malfunction. Based upon the above technicalinformation and inconclusive troubleshooting, the PORC examined numerous recommendations for resolving this equipment problem. PORC concluded that modifying the generator runback circuit (bypassing the current / flow comparator) via a temporary modification was the most prudent course of action.
The inspectors viewed this interim corrective action as appropriate.
Temporary Modification VY modified the runback circuit via a temporary modification (TM No. 97-26). TM 97 26 removed the comparator input signc! and left the remaining stator cooling system high temperature and low system pressure inputs available. Design engineering staff review concluded that the current / flow comparator was ir. stalled in
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the runback circuit in 1979, as an anticipatory loss of stator cooling function which would detect an insufficient stator cooling flow condition before either the high temperature or low pressure conditions develepod. Early detection and the resultant generator runback was v;ewed to lessen the potential for generator damage and potentially lessen the severity of the reactor plant transient.
Inspector review of the implementation of this TM identified that limited communication and training was provided to the operating shift crews on the impact of TM 97 26 on plant response and/or alarm response procedures. The inspectors determined that there was an absence of a clear understanding of the modified generator runback circuit by the operators interviewed. However, inspector discussions with a number of chift operators determined that,in spite of the lack of training on TM 97-26, the control room and local stator cooling slarm response procedures provided adequate procedural guidance to address a degraded or loss of stator cooling condition. The inspector identified that control room annunciater 7 B-5, " Stator Cooling Runback" was " orange dotted" in accordance with Operations Procedure (OP) 3140, " Alarm Response," to reflect the modified runback circuit.
However, the local stator cooling panel annunciator SC A-3, "turbir.e runback," was not similarly " orange dotted." This administrative oversight was promptly corrected by the on shift supervisor when identified by the inspector. The inspector noted that training was supplemented after his observations.
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Conclusions Overall, control room operator response to the November 25 reactor scram was good and in accordance with off-normal and emergency operating procedures. A significant contributing cause of the automatic reactor scram was a failure to provide an appropriate level of review and approval of the 345 KV switching orders which was contrary to Technical Specification 6.5 and a cited violation (VIO 9711-01). Station management and PORC response to this transient and the resolution of plant technical and human performance concerns prior to unit start-up were appropriate. The temporary modification and associated actions to address the generator runback circuit problem were adequately implemented, but operator training on the TM was weak and administrative oversight not comprehensive.
Miscellaneous Operations issues 08.1 (Closed) Violation 96-05-01: Plant operation imoact of hiah oressure coolant iniection (HPCI) system desian chanae (92901)
On December 7,1995, replacement of the HPCI turbine trip pushbutton per the Installation and Test (l&T) procedure for Engineering Design Change 95 408 caused an unanticipated automatic system realignment. The cause of the realignment was that the electrical isolation that was specified in the l&T procedure for replacement of the pushbutton (CRP9-39 fuses 23A-F1 and -F2 removed) also caused the HPCI suction transfer logic to initiate, thus causing the HPCI suction to realign from the condensate storage tank to the torus. Subsequent investigation revealed that the primary containment isolation system (PCIS) logic for HPCI had also been degraded
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as a result of the fuse removals. This event was discussed in detailin inspection report 50 271/95 25 and was cited as a violation of 10 CFR 50, Appendix B, Criterion XIV, " Inspection, Tests, and Operating Status," in inspection report 50-271/96-05.
The inspector reviewed licensee event report (LER)96-005, " Failure to ic'entify and take Technical Specification required actions during design change installation activities due to human error," dated March 28,1996, which discussed this event.
The cause of the event was determined to be human errors in preparation of the installation procedure and the design change. The governing procedures were considered to have contained sufficient guidance to have avoided the problem, and corrective actions focused on promulgating lessons learned from the event. In addition, members from appropriate departments were designated to participate in advanced preparation and review for future similar on-line maintenance and design
installation activities.
In their reply to the notice of violation, dated July 19,1996, VY indicated that,
" additional review and assessment of the event determined that a contributing factor was that the expectations for Operations Department review of l&T procedures was not well understood (e.g. expectation that a thorough review of proposed tagouts be performed consistent with the level of review performed under the switching and tagging process)." The reply indicated that procedure AP 4001,
" Installation, Test and Special Test Procedures," would be revised to incorporate these expectations. The inspector reviewed AP-6001, revision 20, dated September 26,1997. This revision incorporated recommended changes that resulted from review of the l&T processes and procedures used by six other utilities.
Enhancements included increased interdepartmental participation during walkdowns -
and procedure generation, and improved coordination with operations during equipment tagout and alignment evolutions.
The inspector concluded that VY's corrective actions appropriately addressed the causes of the vlolation, as evidenced by no similar occurrences since the event.
Accordingly, violation VIO 96-05-011s closed.
08.2 (Closed) Insoection Follow item 96-200-06: Follow-uo on Nuclear Safety and Audit Review (NSAR) Committee olans to resolve known oroblem areas _ (92901)
The inspector attended portions of the regularly scheduled NSAR committee meeting of November 13 and 14,1997. The inspector observed presentations and committee member review of selected Quality Assurance (QA) audits, a Systemn Engineering Program presentation, a status report on the service water system, a Human Performance improvement Program report, and a presentation on the site Corrective Action Program which included a discussion of recent performance trends. The inspector observed that the presentations were informative and that many of the NSAR committee members engaged the presenters in further explanation of the prepared information and chsilenged the bases for the presenters'
assessments. ' A few new committee follow up items were assigned and previously opened NSAR committee items were assessed for closur _
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The inspector determined from a follow-up discussion with the NSAR committee chairman that the agenda had been revised to focus routinely on the known problem i
areas of corrective action effectiveness and human performance. In addition to the formal agenda topic presentations, the inspector learned that various site-generated monthly performance trend reports were now routinely distributed to the committee members. The inspector also noted that NSAR committee-sponsored QA audits have recently placed more focus in the areas of procedural quality and adherence, at the request of the committee.
Based upon this review and the resident inspector staff's routine inspection and assessment of the off site committee's plant oversight functions, inspection follow item IFl 96-200-06is closed, 11. Maintenance M1 Conduct of Maintenance M1.1 Maintenance Observations a.
insoection Scoom (62707)
The inspectors observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry code and technical specification requirements were satisfied, adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon comp!stion of post maintenance testing, b.
Observations. Findinas. and Conclusions The inspector observed portions of the following maintenance activities:
Reactor power suppression testing to identify the location of a leaking fuel
rod, observed November 7-9 A smallleak developed from a fuel rod on October 26. as indicated by a step increase on the steam jet air ejector radiation mon: tor. Grab sample results indicated that the activity discharge rate at the air ejector had increased from the pre-event level of 600 microcuries per second (;/Ci/sec) to 2600 Ci/sec.
A reactor coolant sample showed that iodine concentration had elso increased from 1.5E-4 Ci/ml (dose equivalent 1131) to 3.1E-4 Ci/ml. The suppression testing succeeded in identifying the fuel bundle that contains the leaking fuel rod.
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The inspector observed that this activity was well controlled. Although the activity required several days of reduced power operation, the inspector observed no evidence of time pressure on the operators. VY's decision to complete core-wide testing after the leaking fuel rod was identifieo (thereby verifying that there were no other leaking rods) also indicated management's support for obtaining as much information as possible about the condition prior to resuming full power operation.
Scram solenoid pilot valve (SSPV) replacement for hydraulic control unit 06-
19, observed November 10 The SSPV was replaced after it was determined to be buzzing; this condition has been demonstrated to be an indicator of incipient failure, as discussed in inspection report 50-271/97-04. The inspector observed no problems during the replacement and post maintenance testing.
"A" service water subsystem maintenance, observed November 10-12
This limiting condition for operation (LCO) maintenance period included preventive maintenance on the "C" SW pump and motor. No problems were noted, and activities were completed well within the 15-day allowcd outage time.
M1.2 Surveillance Observations a.
Inspection Scope (61726)
The inspector observed portions of surveillance tests to verify proper calibration of test instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to LCOs, and correct post-test system restoration.
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Observations, Findinas, and Conclusions The inspector observed portions of the following surveillance test:
High Pressure Coolant injection System Quarteriy Surveillance, observed
December 3,1997 The inspector observed that the test was well controlled. No problems were note.
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MG Miscellaneous Maintenance issues M8.1 LQlosed) Unresolved item 95-10-01: Ouality Control inspection and verification of in-orocess brazina a.
Backoround and Insocction Scope (92902)
During an earlier inspection, the inspector witnessed brazing activities associated with the replacement of the reactor recirculation units (RRUs) 7 and 8 heat exchangers. The inspector observed no independent Quality Control (OC) inspector oversight of those activities, other than the QC verifications of consumable materials issued from the warehouse. As a result, the inspector questioned the adequacy of quality assurance measures for this activity, even though no specific brazing problems were observed by the inspector. The licensee acknowledged the inspector's observation and conducted a review of their brazing control processes to ensure compliance with established practices and industry standards. The inspector reviewed the licensee's response to this item and verified the minimum regulatory requirements were being satisfied, b.
Observations and Findinas The inspector reviewed VY's internal response to this issue documented by memorandum dated April 5,1996. The memo confirmed VY's initial response to the inspector that the Quality Assurance (QA) Program had been adequately adhered to, for the brazing activities observed. The inspector independently verified, by a records review, that the brazer who mformed the work was qualified to ASME Section IX standards, as required by AN5 931.1, Section 128.
No in process QC inspections were required by ANSI e61.1, nor were in-process inspections stipulated in the work package by the responsible weld engineer.
Consequently, the inspector determined that appropriate QC measures were invoked for this brazing activity and that the qualified brazer was principally responsible per code for ensuring quality in this special process. The inspector noted that VY did clarify their contractor Quality Control Inspection Guidelines, (VY/ FPS 70010) as a result of the inspector's observations.
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Conclusions The licensee's quality control measures for the brazing activities witnessed during inspection 95-10 were in accordance with ANSI B31.1 requirements and t,stablished VY welding and QC inspection guidelines. This unresolved item (URI 9510-01)is closed.
M8.2 (Closed) Unresolved item 95-14-01: Verify maintenance reauirements for alternate coolina systqm (92902)
This unresolved item was initiated to provide a tracking item to verify the licensee's j
implementation of routine preventive maintenance (PM) requirements for the
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10 alternate cooling system. As previously documented in inspection report 50-271/9514, section 2.2.2, the NRC staff concluded that VY's 10-year test interval to demonstrate satisfactory hydraulic performance per Generic Letter (GL) 8913, Action ll, was acceptable. However, the alternate cooling system and residual heat removal service weter (RHRSW) piping PM requirements had not yet been
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formalized per GL 8913, Action Ill.
With the assistance of the service water systems engineer, the inspector reviewed the Maintenance Planning and Control (MPAC) computer data base, (VY's formal maintenance department record system), to verify that alternate cooling system and
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RHRSW piping PM requirements were entered and being implemented. The inspector examined a sampling of the most recently completed Work Orders (WOs)-
for the associated PM ttsks and verified satisfactory implementation of the selected PM activities. For example, the inspector confirmed that the safety related portion of the alternate cooling system (west cooing tower, CT-2) was receiving routine (repetitive) PM in accordance with VY's PM Basis No, M307, revision 3. WO No.
97 06891, completed May 30,1997, documented the results of the last CT-2 structural inspection per Vermont Yankee Equipment Manual (VYEM) No.146.
Deficient conditions were appropriately resolved via separate WOs, as annotated in the MPAC records. The inspector also examined a sampling of the completed PM records for the RHRSW piping and verified that the piping inspections had been
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performed in accordance with OP 5265, " Service Water component inspection and acceptance criteria." This unresolved item (URI 95-14-01)is closed.
M8.3 (Closed) Unresolved item 95 25 04: Advanced Offaas (AOG) system hydrooen monitor Technical Specification (TS) surveillance reauirement identifies an incorrect test oas mixtutg (92902)
During an earlier inspection period, the licensee icientified that AOG hydrogen monitor surveillance requirements specified in TS (TS Taole 4.9.2, note 4) were incorrect. As documented in inspection report 50-271/95 25, prompt corrective action was taken to ensure proper testing of the hydrogen monitors and an amendment to TS Table 4.9.2 was initiated. By letter dated February 5,1996, VY submitted proposed change No.180, Administrative Change to Vermont Yankee
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Technical Specification to correct typographical errors and text inconsistencies. The NRC staff's review of this February 5,1996 submittal was ongoing, as of the close of this inspection period. The VY staff has taken reasonable and timely action to resolve the TS administrative error.
Included in the inspector's original review of this issue was a conclusion that the biennial procedure review process had not been sufficiently thorough to have identified this TS error earlier (the hydrogen monitor TS was revised in the early 1980's). To address this observation, the licensee revised their administrative control procedure (AP)-0037, " Plant Procedure," to significantly enhance the biennial procedure review criteria. The revised procedure ensured a " comprehensive review" of the entire procedure, with a verification of all procedure steps against the FSAR, Technical Specification, Quality Assurance Program (YOQAP-1-A), and applicable prog;am requirements (IST, Fire Protection, EQ, E-Plan, etc.). The
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inspectors have observed, over the past several months, improved biennial review of procedures. This has ueen demonstrated by the generation of Event Reports for significant deficiencies identified via the biennial review, and the sometime extensive summary of changes memorandums addressed to the Plant Operations Review Committee, which are attached to the current revision of a plant procedure.
The licensee's actions to address the administrative error in the TS were appropriate and the previously observed weaknesses in the VY procedures biennial revision process have been adequately addressed. The inspector concluded that no violations of regulatory requirements occurred. This unresolved item (URI 95 25 04)
is closed.
M8.4 (Closed) Unresolved item 96-03-07: Resolution of inconsistent recombiner catalvst lifetime in the AOG system (92902)
During an inspection of the advanced offgas (AOG) system, the inspector identified an inconsistency in the licensee's documentation pertaining to the expected life of the AOG hydrogen recombiner catalyst. In response, the VY staff initiated an action plan via their commitment tracking system (CTS) to address this issue.
By engineering staff memorandum VYl 59/96 Revision 1, dated August 28,1996, the licensee formulated an action plan to verify the current performance characterization of the hydrogen recombiner catalysts and to routinely monitor their performance to better project their replacement, if necessary, prior to the end of the facility's lifetime (same as the estimated catalyst's lifetime). The inspector reviewed the individual action plan items (CTS items INS 960307-01 through 06)
and verified that recombiner temperature monitoring has been implemented in accordance with AP-0150," Conduct of Operations and Operator Rounds," Revision
- 31. Barsed upon the above, the inspector concluded that VY had implemented reasonable and timely measures to monitor hydrogen recombiner catalyst performance and that no violation occurred. This unresolved item (URI 96-03 07)is closed.
M8.5 (Closed) LER 97 015: Instrument check not comoleted in accordance with Technical Soecifications due to corsonnel error (92700).
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The event discussed in LER 97-015, dated September 24,1997 was previously reviewed by the inspectors, as documented in inspection report 50-271/97-06, section M1.3. As documented in report 97-06, this violation was not cited. LER 97 015 was reviewed using the guidance of Inspection Procedure 92700 and-determined to have been appropriately written and properly submitted in accordance with 10 CFR 50.73. LER 97-015 is closed.
M8.6 (Closed) Unresolveo item 96-03-04: Torus water temperature administrative limit reduced This unresolved item was examined by the NRC staff in inspection report 50-
.271/97 201,section E1.1.2.2. As documented in report 97-201, the inspect rs
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4 examined plant operation with this concern and the timeliness of the licensee's corrective actions for this design issue. Consequently, two separate unresolved items were generated (URI 97 201-01 and URI 97-201-02). Accordinely, URI 96-03-04 is administratively closed, and the two new unresolved items will track the
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individualissue's resolution and NRC staff disposition of the same.
Ill. Engineering E8 Miscellaneous Engineering issues E8.1 (Closed) NCV 97-11-02 and Unresolved item 95-19-01: Adeauacy of controls for the conduct of system manioulations for performance monitorina (92903)
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This item was initiated based upon the inspector observing an instance where the reactor building closed cooling water (RBCCW) heat exchanger outlet valve was throttled, per the operating procedure (OP-2181) for the purpose of collecting data on heat exchanger thermal performance and flow-induced vibration. During the review of this activity, the inspector observed that both heat exchanger service water inlet to-outlet differential pressure readings exceeded the operating procedure maximum limit of 10 psid. Neither of the individuals involved with the data collection were aware of this condition nor took action to address it prior to the inspector's intervention. The inspector was concerned that the RBCCW system may not have been appropriately operated per procedure and that the controls for system performance monitoring may not be adequate.
Follow up inspection of this activity identified that VY management developed similar concerns. The auxiliary operator log sheets were revised to include the acceptable RBCCW service water flow differential pressure (d/p) band and all heat exchanger d/p limits and operating parameters were verified. The procedure biennial review process was also modified to include a wrification of all system design and limiting parameters. In addition, an assessment of this event and the lessons learned was included in the Operations Department " Read and Sign" notebook (log #96-34)for mandatory Operations Department staff review. The
" Road and Sign" re-enforced VY management expectations with respect to procedural adherence and expected supervisory involvement when operating systems for data collection.
Inspector follow-up and day-to-day monitoring of plant activities indicated that the type of intrusive system performance monitoring conducted during inspection period 95-19 was isolated and not a programmatic concern. There has been no observed recurrence of this type of on-line system performance monitoring. The failure to have properly operated the RBCCW system in accordance with OP-2181 constituted a violation of minor safety significance and was treated as a Non-Cited Violation, consistent with Section IV of the NRC Enforcement Policy. NCV 97-11-02 and unresolved item URI 95-19-01 are closed.
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E8.2 (Closed) Violation 96-05 04: Batterv room block wall seismic ovalification methodoloov acceotability and operability determination imoact (92903)
On March 5,1996, the VY engineering staff identified that the block walls that separate and support the station batteries were not seismically qualified per the current design basis requirements. This condition was not entered into the
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corrective action process until March 12,1996, which constituted an inappropriate delay in the plant staff's operability assessment of the station batteries.
Additionally, VY's initial projection (ict the condition would not be corrected until the first quarter of 1997 represented acutptance of an excessive delay in implementing corrective action. This cor cition was discussed in detail in inspection report 50-271/96-03,and was cited as a violation of 10 CFR 50, Appendix B, Criterion XVI, " Corrective Action," in inspection report 50 271/96 05.
The inspector reviewed LER 96-008, " Error identified in block wall calculation due to personnel error which places plant outside of design basis." The cause of the event and corrective actions discussed in that document relate to the deficient material condition rather than the timeliness of corrective action. The inspector reviewed Vy's reply to the notice of violation, which identified the cause as being personnel error on the part of engineering management, in that reporting the condition to plant management was delayed pending completion of an engineering analysis. Corrective actions consisted of discussions and training to emphasize the need for timeliness of operability determinations with respect to issue revelation.
The reply also discussed appointment of a new engineering management team and -
reorganization of the VY engineering organization, and pointed out their progress in lowering the threshold of problem discovery.
The inspector noted that a significant number of design issues have been identified by VY engineering during the year 1997, in general, these issues have been expeditiously entered into the corrective action program and immediate operability determinations have been promptly performed. Where appropriate, continued operation with degraded or nonconforming conditions has been justified through the licensee's Basis for Maintaining Operation (BMO) process. The inspector assessed that the quality of operability determinations resulting from this process was good, and that the timeliness of proposed corrective actions was reasonable.
Accordingly, violation VIO 96 05 04 is closed.
E8.3 (Closed) LER 96-001. Suontement 2: Technical Specification 4.6.E not met due to components not included in the inservice Test Proaram scooe l92700)
The Inservice Testing Program issues documented in LER 96-001, Supplement 2, dated January 16,1997, were previously examined as documented in inspection reports 50-271/96-03,section 3.3.1, and 50-271/9611,section E8.2. Supplement 2 updated the corrective actions implemented by the VY staff. The inspector reviewed LER 96-001, Supplement 2 using the guidance of Inspection Procedure 92700 and found the licensee's updated response to this event to have been appropriate. LER 96-001, Supplement 2 is close.
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E8.4 (Closed) Violation EA 96-210: RHR Minimum Flow Valve Sinole Failure Vulnerability (92903)
This issue was examined by the NRC staff in inspection report 50-271/96-07and reported by the licensee in LER 96-010. As discussed in the notice of violation, dated August 23,1996, VY was credited both with identifying the issue and taking prompt, comprehensive corrective action, in this case, however, enforcement action was taken due to the length of time that the condition had existed and the number of prior opportunities that existed to identity and correct the condition. The sigt.lficant number of design issues that have been identified through VY's design basis documentation program indicates that the licensee has taken action to address this issue. Accordingly, violation EA 96 2101s closed.
E8.5 (Closed) NCV 97-11-03 and LER 96-15: Oriainel B31.1 ANSI Code Section that reauired overoressure relief for isolated cloina was not considered durina oriainal desian, a.
Insnection Scone (92700)
The inspectors reviewed LER 96-015, to assess the adequacy of the VY staff's sctions with respect to identification, reporting, evaluation, and resolution of that issue. This review included the examination of the associated Event Report (ER),
immediate operability determination, Basis for Maintaining Operation (BMO), and the related safety evaluation. The inspector verified proper implementation of immediate corrective actions, verified and discussed the adequacy of interim compensatory measures with responsible engineering and operations staff, and assessed the overall adequacy and timeliness of proposed actions to ensure the proper resolution and/or a comprehensive programmatic review of the design issue.
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Observations and Findinas This issue involved the licensee's determination that the original piping construction code (ANSI B31.1,1967) stipulated that portions of piping that could exceed design pressure limits, if isolated, should be protected from thermal over-pressurization by a pressure relieving device. Based upon industry operating experience review, the VY staff determined that a number of piping systems penetrating containment were susceptible. This issue was previously reviewed in inspection report 96-08 and an unresolved item (URI 96-08-02)was assigned for further NRC staff review and follow-up.
As stated in LER 96-015, based upon the event being an original design and construction issue, the root cause determination was not conclusive. However, it appeared that personnel involved with initial plant design and construction failed to apply Section 101.4.2 of ANSI B31.1-1967to those plant systems penetrating containment which could be isolatea in the event of a LOCA and resultant primary containment isolation system initiation. As previously documented in inspection report 50-271/96-08,the affected systems included portions of reactor building closed cooling water (RBCCW) piping, residual heat removal (RHR) and shutdown
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cooling piping, main steam line drains piping, liquid radwaste piping, and recirculation system sample piping.
By letter dated January 28,1997, VY responded to Generic Letter (GL) 96-06,
" Assurance of equipment operaisility and containment integrity during design basis accident conditions," dated September 30,1996. As stated in VY's September 30 letter, to address the potential thermal overpressurization of isolated piping penetrating containment, VY implemonted engineering design changes during the Fall 1996 refueling outage. Relief valves were installed in the RBCCW and radwaste system piping, and check valves were installed in the main steam line drains, RHR shutdown cooling, and recirculabon sample piping.
The inspectors found that the above licensee identified "old design issue" (reference Enforcement Guidance Memorandum 96-005, dated October 21,1996) was appropriately dispositioned in a timely fashion within the guidance of Generic Letter 91-18 and 10 CFR 50, Appendix B, Criterion XVI, " Corrective Action." The inspectors observed that the licensee's reporting of these issues was consistent with 10 CFR 50.72 and 50.73 requirements. The inspector observed that VY promptly investigated this issue and developed corrective action in advance of the NRC staff's issuance of GL 96 06.
c.
Conclusions The extended period without operability of the containment system is a violation of 10 CFR Appendix B, Criteria Ill, Design Control. The violation was identified by the licensee as a result of a voluntary initiative, corrective actions were prompt and comprehensive, the violation was not likely to be identified by routine licensee efforts such as normal surveillance or quality assurance activities and the violation is not reasonably linked to current performance. As a result, this apparent violation of NRC requirements will not be cited in accordance with Section Vll.B.3 of the NRC Enforcement Poliev. NCV 50-271/97-1103,LER 96-015 and URI 96-08 02 are Closed.
E8.6 (Closed) LER 50-271/96-028:Inadeauate field labelina of safety class wirina and prawina undates result in the failure to maintain electrical separation reauirements described in nuclear safety system desian bases. (92700)
This LER describes a f ailure to meet cable separation criteria resulting from an improper labeling of wiring during a 1976 modification. While the original modification error only involved mislabeling of the wiring, subsequent modifications to the mislabeled wiring resulted in a loss of physical separation between safety divisions. The NRC identified electrical separation concerns in NRC inspection report 50-271/97-03 and opened an unresolved item (URI 97-03-02) at that time.
Additional similar concerns were raised by the NRC AE team as described in inspection report 50-271/97-201. The NRC continues to follow the licensee's activities regarding electrical separation as part of the resolation of the previously identified unresolved item. Therefore, this LER is administratively closed and all further NRC response will be tracked via URI 97-03-0.
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IV. Plant Support P8 Miscellaneous EP issues P8.1 (Closed) NCV 97-11-04 and Unresolved item 97 03-07: Criticality Accident Reauirements (92904)
This issue involved the failure to have in place either a criticality monitoring system for storage and handling of new (non-irradiated) fuel or an NRC approved exemption to this requirement contained in 10 CFR 70.24.
10 CFR 70.24 requires that each licensee authorized to possess more than a small amount of special nuclear material (SNM) maintain in each area in which such materialis handled, used, or stored a criticality monitoring system which will energize clearly audible alarm signals if accidental criticality occurs. The purpose of 10 CFR 70.24 is to ensure that, if a criticality were to occur during the handling of SNM, personnel would be alerted to that fact and would take appropriate action.
Most nuclear power plant licensees were granted exemptions from 10 CFR 70.24 during the construction of their plants as part of the Part 70 license issued to permit the receipt of the initial core. Generally, these exemptions were not explicitly renewed when the Part 50 operating license was issued, which contained the combined Part 50 and Part 70 authority. In August 198'i, the Tennessee Valley Authority (TVA), in the course of reviewing the operating licenses for its Browns Ferry f acilities, noted that the exemption to 10 CFR 70.24 that had been granted during the construction phase had not been explicitly granted in the operating license. By letters dated August 11,1981, and August 31,1987, TVA requested an exemption from 10 CFR 70.24. On May 11,1988, NRC informed TVA that "the previously issued exemptions are still in effect even though the specific provisions of the Part 70 licenses were not incorporated into the Part 50 license."
Notwithstanding the correspondence with TVA, the NRC has determined that, in cases where a licensee received the exemption as part of the Part 70 license issued during the construction phase, both the Part 70 and Part 50 licenses should be examined to determine the status of the exemption. The NRC view now is that unless a licensee's licensing basis specifies otherwise, an exemption expires with the expiration of the Part 70 license. The NRC intends to amend 10 CFR 70.24 to provide for administrative controls in lieu of criticality monitors.
The NRC has concluded that a violation of 10 CFR 70.24 existed. The NRC has also determined that numerous other licensees have similar circumstances that were caused by confusion regarding the continuation of an exemption to 10 CFR 70.24 originally issued prior to issuance of the Part 50 license. After considering all the factors that resulted in these violations, the NRC has concluded that while a violation did exist, it is appropriate to exercise enforcement discretion for Violations involving Special Circumstances in accordance with Section Vil B.6 of the " General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600. Pending the amendment to 10 CFR 70.24, further enforcement action will not be taken for failure to meet 10 CFR 70.24, provided an l
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enforcement action will not be taken for failure to meet 10 CFR 70.24, provided an exemption to this regulation is obtained before the next receipt of fresh fuel or before the next planned movement of fresh fuel. NCV 971104 and unresolved item URI 97-03-07 are closed.
F8 Miscellaneous Fire Protection issues F8.1 (Closed) Unresolved item 95 21-02: Accendix R and Fire Protection Proaram implementation issues (92904)
This unresolved item broadly addressed the large number of related Appendix R and Fire Protection (FP) Program issues which were identified by the VY staff in late 1995. As a result of several of these identified discrepancies, compensatory fire watches were established in accordance with TS-pending resolution of the technical or programmatic issue. The inspector observed that initially, a number of minor fire watch performance lapsec occurred as a result of either poor understanding of the compensatory measures or miscommunication of the fire watch requirements.
Significant VY management attention was focused on these performance issues and the broader Appendix R and FP Program technicalissues, resultirc in sweeping changes and noteworthy improvements in the program and overall 'V staff performance.
During an earlier inspection this year (reference inspection report 50-271/97 80, dated October 16,1997) the NRC concluded that VY's overall response to the cited violations and additional licensee identified and corrected problems were appropriate. As a consequence, enforcement discretion was exercised for the related Appendix R and FP Program violations identified via the licensee's corrective action processes. Accordingly, this unresolved item is administratively closed, as it identified no additional technical or performance issues that had not already been addressed by the VY staff. Unresolved item URI 95-21-02is closed.
V. Management Meetings X1 Exit Meeting Summary
The resident inspectors met with licensee representatives periodically throughout the inspection and following the conclusion of the inspection on December _,
1997. At that time, the purpose and scope of the inspection were reviewed, and the preliminary findings were presented. The licensee acknowledged the preliminary inspection findings.
X3 Review of Updated Final Safety Analysis Report (UFSAR)
A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR description. While performing the inspections discussed in this report, the inspector = reviewed the applicable portions of the UFSAR that related to the area inspected. The inspectors verified that the UFT AR wording was consistent with the observed practices and procedures and/or parameter..
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18 INSPECTION PROCEDURES USED
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f 61726 Surveillance Observations 62707-Maintenance Observations 71707.
Plant Operations 92700 Onsite Follow-up of Written Reports of Nontoutine Events at Power Reactor Facilities 92901 Followup - Plant Operations 92902 Followup - Maintenance 92903 Followup - Engineering 92904 Followup - Plant Support 93702 Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPENED, CLOSED, AND DISCUSSED L. EN VIO 97-11-01 Failure to review and approve switchyard switching orders resulted in reactor scram CLOSED V:0 96-05 01 Plant operation impact of high pressure coolant injection (HPCI)
system design change IFl 96-200-06 Follow-up on Nuclear Safety and Audit Review (NSAR) Committee plans to resolve known problem areas URI 9510 01 Quality Control inspection and verification of in-process brazing
- URI 9514-01 Verify maintenance requirements for alternate cooling system URI 95 25 04 Advanced Offgas (AOG) system hydrogen monitor TS surveillance requirement identifies an incorrect test gas mixture URI 96-03-07 Resolution of inconsistent recombiner catalyst lifetime in the AOG system
'LER 97-015 Instrument check not completed in accordance with Technical Specifications due to personnel error URI 96-03-04 Torus water temperature administrative limit reduced
' URI 95-19-01 &.
Adequacy of controls for the conduct of system manipulations for
. NCV 97-1102 performance monitoring VIO 96-05-04 Battery room block wall seismic qualification methodology acceptability and operability determination impact
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Original B31.1 ANSI Code Section that required overpressure relief for NCV 9711-03 isolated piping sections was not considered during original des!gn URI 96 08-02-Ocsign deficiency in containment piping penetrations
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- URI 95 21-02 Appendix R and fire protection program implementation issues
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.LER96 028 Inadequate field labeling'of safety class wiring and drawing updates
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G. Meret, Plant Manager-F. Helini Tech. Services Superintendent.
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LIST OF ACRONYMS USED I
LBMO" Basis for Maintaining Operation
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