IR 05000213/1993016
| ML20059K434 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 11/04/1993 |
| From: | Doerflein L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20059K425 | List: |
| References | |
| 50-213-93-17, NUDOCS 9311160060 | |
| Download: ML20059K434 (24) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I-Docket /
50-213/93-17 Report No.
Ijcense No.
DPR-61 Licensee:
Connecticut Yankee Atomic Power Company (CYAPCo)
P. O. Box 270 Hartford, CT 06141-0270 Facility:
Haddam Neck Plant Location:
Haddam Neck, Connecticut Dates:
August 29,1993 to October 2,1993 Inspectors:
Alan B. Wang, Project Manager Howard J. Rathbun, Reactor Engineer Intern Peter J. Habighorst, Resident Inspector William J. Raymond, Senior Resident Inspector i!93 W
%.4 oPu E
I Approved by:
Lawrence T. Doerflein, ChiefI Date Reactor Projects Section No. (A Areas Instected: NRC resident inspection of plant operations, maintenance, engineering and technical support and plant support activities. As an initiative, the inspector reviewed the quality of Haddam Neck operating procedures and their conformance with industry standards.
Results: See Executive Summary i
9311160060 931105'
PDR ADOCK 05000213 G
PDR;
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t EXECUTIVE SUMMARY
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HADDAM NECK PLANT INSPECTION 50-213/93-17 i
Plant Operations
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Safe facility operation was noted throughout the period as the plant continued routine power operations. Operators performed very well to control the plant during load reductions and in response to equipment problems, including a leak on the 1A main feedwater pump, and an inoperable post accident sample system (PASS) valve. The shift supervisor made a conservative decision in response to the inoperable PASS valve and showed a high regard for containment integrity and safe plant operations. The semi-vital power supply was found to be fully operable during a walkdown of engineered safety features systems. Actions to prepare the site for the arrival of Hurricane Emily were prompt and thorough.
Operations procedures reviewed during this inspection conformed to ANSI 18.7-1976.
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Recent technical specification amendments were implemented by issuing appropriate and a
timely procedure changes. The licensee has identified some procedure deficiencies during recent Quality Services Department (QSD) surveillances.
Maintenance
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Maintenance and surveillance 'esting completed during the period was acceptably performed in accordance with the administrative requirements. Special tests of the diesel generator
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room ventilation system was acceptably performed, but the test identified the need for further
engineering reviews to resolve a longstanding issue regarding the design of the intake configuration.' The licensee executed an extensive test plan to check the suspect components in the MCC5 failure on June 27,1993; however, the root cause of the event remained unidentified.
The actions by operations, maintenance, engineering and chemistry personnel were very good to evaluate and resolve an operability issue with isolation valve SS-SOV-151D in the post
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accident sample system. There was good communication and coordination amongst all.
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parties involved to assure prompt resolution of the operability concern. The plant operators, in particular, did a good job to keep good oversight of the operability determinations and
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troubleshooting efforts, and demonstrated good " ownership" of plant status and the repair of its components.
Encineerine and Technical SuppgIl CYAPCo implemented significant changes to the leak seal program to improve vendor -
oversight, quality services department involvement, improved consistency with industry
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guidelines provided in EPRI NP-6523-D, "On Line Irak Sealing," and increased management attention and approval of leak sealing. The progum was further improved in l
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response to NRC comments provided during this inspection by requiring recommended
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training of job supervisor and vendor personnel, assuring engineering evaluations explicitly require replacement stud acceptance, and by assuring that ASME Section III requirements are met for replacement studs on Category I components.
The licensee has established formal procedural guidance and controls to evaluate each plant design change, for which 10 CFR 50.59 is applicable, to determine whether an unreviewed safety question (USQ) exists or a technical specification (TS) change is required. The need for a safety evaluation is defined in the plant design change record (PDCR) procedure and NEO 3.12 provides appropriate guidance for. preparing the safety evaluation, which.
determines if a pmposed plant change is safe and satisfies the requirements of 10 CFR-50.59. The control room operators were aware of the TS Clarifications book and the procedure to establish new clarifications. This procedure is effective in implementing a consistent interpretation of the TSs, and provides the basis for the interpretation and a method to revise those TSs that need to be changed.
Plant Sutmort Several previous inspection items were brought to resolution. Periodic reports reviewed during the period were satisfactory and met technical specification requirements.
Radiological controls were acceptably implemented during the period. A plant worker suffered a significant injury while working inside the containment on June 10,1993. The -
inspector found that the actions by licensee and emergency response personnel conformed with the requirements of EPIP 1.5-11, " Personnel Injuries." Based on a review of the licensee's investigation files and interviews of some of the personnel involved with event, the i
inspector concluded that licensee health physics and emergency response actions were acceptable to assure the radiological safety of the injured worker and offsite response personnel. No violations of 10 CFR 50 or 10 CFR 20 were identified.
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SUMMARY OF FACILITY ACTIVITIES L
The inspection period began with the unit at 30% full power (FP) following repairs to.
address a leak in extraction steam valve MS-V-560. The unit returned to full power -
operation at 4:35 a.m. on August 29. Unit load was reduced to 45% FP at.10:50 a.m. that-same day due to leakage from the inboard seal on the 1 A main _feedwater pump. The plant
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returned to full power operation at 4:20 p.m. on August 31 following seal replacement.
The unit remained at full power until September 27 when Technical Specification 3.6.1 was -
conservatively entered in response to an inoperable isolation valve in the post accident sample system. An Unusual Event was declared and a plant shutdown was initiated at 1:30 p.m. Load reduction was halted at 87% FP and the plant was returned to full power at 6:20 s
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p.m. the same day after actions were taken to assure containment integrity was' met. The unit remained at full power for the remainder of the inspection period.
2.0 PLANT OPERATIONS q
In addition to normal utility working hours, the review of plant operations was routinely -
conducted during portions _of backshifts (evening shifts) and deep backshifts (weekend and night shifts). Inspection coverage was provided for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> during backshifts and'15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> during deep backshifts.
2.1 Operational Safety Verification This inspection consisted of selective examinations of control room activities, operability reviews of engineered safety feature systems, plant tours, review of the problem identification systems, and attendance at periodic planning meetings. Control room reviews.
consisted of verification of staffing, operator procedural adherence, operator cognirance of control room alarms, control of technical specification limiting conditions of operation,.and electrical distribution verifications. Administrative control procedure (ACP) - 1.0-23,.
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" Operations Department Shift Staffing Requirements," identifies the minimum staffing requirements. During the inspection period, these requirements 'were met.
The inspectors reviewed the onsite electrical distribution system to verify proper electrical line-up of the emergency core cooling pumps and valves, the emergency diesel _ generators,.
radiation monitors, and various ' engineered safety feature equipment.The inspectors alsoL.
verified valve lineups, position of locked mannal valves, power supplies, and flow paths for.
the high pressure safety injection system, the low pressure safety injection system, the containment air recirculation system, the service water system, and the emergency diesel generators. No deficiencies were noted.
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Bypass jumpers were' reviewed against the requirements of ACP 1.2-13.1, with emphasis on proper installation and the content of the safety evaluations. The inspector reviewed all jumpers for age, and verified that Plant Operations Review Committee (PORC) evaluations were completed to disposition longstanding evaluations. The jumpers reviewed were found to be in accordance with administrative requirements.
Tapouts Equipment tagout 93-1206 was reviewed to verify conformance with the applicable sections of ACP 1.2-14.2. Tagouts were reviewed to verify that the proper equipment was tagged, equipment identified within technical specifications was appropriately controlled, and equipment isolation was proper based on work observations, controlled drawings, and procedural guidance. Tagging operations were reviewed by comparing the tags installed within the plant with the tagout sheets maintained in the control room. Equipment reviewed was appropriately isolated and the tagouts met the technical specification requirements and administrative controls.
Lop-Keeping and Turnovers The inspectors reviewed control room logs, night order logs, plant incident report logs, and crew turnover sheets. No discrepancies or unsatisfactory conditions were noted. The inspectors observed crew shift turnovers and determir.;d they were satisfactory, with the shift supervisor controlling the turnover. Plant conditions and evolutions in progress were discussed with all members of the crew, The information exchanged was accurate. Control room trouble reports were reviewed for age, planned action, and operator awareness of the reason for the trouble report. The majority of trouble reports reviewed were recent, with few longstanding items.
During attendance at daily planning meetings the inspector noted discussions were held on maintenance and surveillance activities in progress. The inspectors conducted periodic plant tours were conducted in the primary auxiliary building, turbine building, and intake struc-tures. Plant housekeeping was satisfactory.
2.2 Procedural Quality Review The inspection consisted of a review of randomly selected operations procedures to verify the -
content of the procedure conformed to technical specifications limits, adequate controls exist for safety related equipment, and that the technical content of the procedure conforms to ANSI 18.7-1976, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants." The review also verified that procedure changes were processed as a result of technical specification amendments (as required). Finally, the inspector reviewed the quality services department (QSD) surveillance results concerning procedural qualit.
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The inspector reviewed the technical content of the following operations procedures:
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k AOP 3.2-12, Loss Of Residual Heat Removal System AOP 3.2-49, less of Annunciator Indications
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EOP 3.1-21, Pressurizer Spray Valve Malfunction NOP 2.9-1, Placing the Residual Heat Removal System in Service
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NOP 2.3-4, Shutdown from Hot Standby to Cold Shutdown NOP 2.9-2, Removing the Residual Heat Removal System from Service NOP 2.1-1, Startup From Cold Shutdown to Hot Standby ANN 4.9-34B, Semi-Vital Normal Supply Low Voltage AOP 3.2-15, IAss of Vital Bus SUR 153A&B, AC and DC Distdbution Normal Configuration EOP 3.1-46, Total Loss of Semi-Vital Power The inspector concluded that all procedures conformed to the technical format requirements of ANSI 18.7-1976. The inspector noted that the integrated safety evaluation for AOP 3.2-
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12, Revision 10 was satisfactory and appropriately addressed the mitigation strategy and
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equipment usage following a loss of residual heat removal. The safety evaluation was
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thorough and complete. The inspector notd that CYAPCo implemented timely and i
appropriate procedure changes for technical specification' amendments #162, #161, and #160.
Except as discussed below, the inspector had no further comments on these procedures.
i The inspector reviewed AOP 3.2-49 instructions to the operator on actions to take in the
event control room annunciators were not operable, The procedure directed the operator to
" increase surveillance of main control board indications" upon a loss of control room
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annunciators. The inspector asked several operators how they would implement this ir struction. The operators provided acceptable but different answers, such as1" increase use
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of the plant process computer" to " increase attention to indications on the main control board." The inspector noted that possible improvements could be made in AOP 3.2-49 by'
defining better what actions management expects the operators to take to " increase l
surveillance of the control board indications."
NOP 2.3-4 states that the pressurizer spray nozzle design permits spray line operation with a differential temperature of 300* Fahrenheit (F) for 150 cycles throughout plant life (5 occurrences per year for 30 years). The procedure directs the operator to maintain differential temperature less than 200*F, and cautions to operator to record temperature and
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spray valve operation whenever temperature exceeds 200 F (consistent with technical specification 3.4.9.2.c). The inspector questioned how the licensee tracks the times when the differential temperature exceeded 200*F and how many cycles have been used. CYAPCo
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engineering personnel stated this information would be provided for inspector review. The inspector will review this information during future inspections as pan of the routine inspection program.
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The inspector reviewed the surveillances performed by QSD in 1993 to evaluate the frequency and type of procedural deficiencies, and licensee corrective actions.
Approximately 15% of the 1993 QSD surveillances identified procedural deficiencies. A majority of the deficiencies were written for engineering procedures (inservice test and reactor engineering). The procedural deficiencies involved wrong nomenclature of equipment or wrong location of plant equipment, no minimum tolerance values on instrument accuracies and torque specifications, and inappropriate guidance to install test equipment.
The inspector did not identify any safety significant procedural deficiencies. CYAPCo took appropriate corrective actions for each specific deficiency. The inspector considers eacn of the QSD surveillance issue examples of " inattention-to-detail" during the review and val;dation process for procedure revisions. The inspector noted that CYAPCo has not initiated a review ofits procedure program as a result of the QSD surveillance fmdings.
Station management is planning actions to address the improvements in the review process.
Conclusion The inspector determined that operations procedures conformed to ANSI 18.7-1976. The inspector learned that appropriate and timely procedure changes resulted from recent technical specification amendments. The inspector will review licensee actions on the following items during future inspections: (1) the program to track spray nozzle differential temperatures in excess of 200 F per NOP 2.3-4; (2) the definition of " increased surveillance of main control board indications" in AOP 3.2-49; and, (3) CYAPCo actions to address procedural deficiencies as identified in QSD surveillances.
2.3 Preparations for Hurricane Emily The inspector reviewed licensee action over the period of August 30 - September 1 to track the progress of Hurricane Emily and to prepare the Haddam Neck site for the potential land fall of the storm on the Connecticut coast.
Plant operators entered abnormal operating procedure AOP 3.2-5, " Natural Disasters," on August 30 as the hurricane came within 400 miles of the site. Preliminary action completed in accordance with the AOP included: assuring the availability of important supplies on site, such as food, fuel oil, bottled water, etc; surveying the site area and removing or securing loose material and equipment; maximizing levels in certain plant tanks; deferring scheduled preventive maintenance on the emergency diesel generators and testing the units to assure they were fully operable and in standby; and tracking the progress of the storm with input from the national weather service and the load dispatcher (CONVEX). The inspector reviewed the completion of the actions listed above and conducted a tour of the site.
Preparations for the storm were suspended on September 1 after the hurricane moved east out into the Atlantic ocean. No inadequacies were identified. Licensee actions to prepare for the storm were prompt and thoroug.
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2.4 Engineered Safety Features System Walkdown In addition to observations during routine plant tours, the inspector conducted a walkdown of the semi-vital power supply system. The references used for the review included procedures ANN 4.9-34B, " Semi-Vital Normal Supply Iow Voltage", SUR 153A&B, "AC and DC Distribution Normal Configuration", EOP 3.1-46, " Total Loss of Semi-Vital Power", and the 120 Volt System One Line Diagram 16103-300013.
The inspector verified system operability through reviews of breaker lineups, control room indications, equipment conditions, surveillance tests, and design documents and drawings.
Outstanding tags and trouble reports were reviewed; none impacted system operability. The breaker alignment listed in the procedures and drawings were compared with the physical plant and no discrepancies were noted. Breaker alignments supported system operability in accordance with the requirements of Technical Specification (TS) 3.5. and 3.8. This review confirmed that the semi-vital power system was fully operational in the standby mode. No discrepancies were noted.
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The inspector reviewed licensee engineering studies of the semi-vital power system, its
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interdependencies wi31 MCC5, and a risk assessment of the system (reference Electrical Separation Study - ISAP Topic 1.64). Significant plant vulnerabilities were identified by the
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risk assessment and addressed by modifications during the Cycle 17 and prior refueling outages. The vulnerabilities include the loss of important instrumentation, and the initiation of certain plant transients. PDCR 942 provided an altemate emergency supply to the semi-
vital bus from MCC-12-II, which eliminated MCC5 as the sole supply and prevents the loss of semi-vital power on loss of MCC5. The electrical separation modifications completed per PDCR 1317 further reduced the plant vulnerability to a loss of coolant accident by relocating
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power for redundant pressurizer valves off of the semi-vital bus. The inspector determined
the licensee took good initiatives to address plant vulnerabilities.
2.5 Inoperable Containment Isolation Valve
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The unit was operating at full power on September 27 when a routine sample of the containment was taken using the post accident sample system (PASS). The PASS system is normally isolated from the containment during plant operations using containment isolation valves SS-SOV-151A,151B,151C and 151D. All four of these valves must be operable during plant operations. The valves are opened to allow chemistry technicians to draw'an air sample from the containment. After taking a sample on September 27, plant operators noted that valve SS-SOV-151D appeared to have stuck in the "open" condition based on the presence of dual " red" and " green" indications in the main control room. The shift supervisor declared the valve inoperable at 12:30 p.m. and entered the action statement for Technical Specification 3.6.3. The specification allows for continued power operation with
the valve inoperable, as long as the redundant isolation valve in the line, SS-SOV-151C, is j
operable and secured in the isolated position. Valve 151C was verified closed. Actions j
l were taken in conjunction with chemistry personnel to investigate the status of valve 151D.
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- During his further review of the requirements for plant operations, the shift supervisor noted j
that Technical Specification 3.6.1 and the definition for " containment integrity" could be read to require that both containment isolation valves in the PASS line would have to be operable for continued plant operations. A request for a clarification of the specification from site
engineering did not indicate that less than two valves were required to be operable to meet the definition of having " containment" established. Technical specification 3.6.1 does not -
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allow plant operation unless containment integrity exists. Thus, the shift supervisor concluded that the most conservative aspect of TS 3.6.1 was not met, and he directed the
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operators to initiate a controlled shutdown to meet the action statement of the specifications.
An Unusual Event was declared and a plant shutdown was initiated at 1:30 p.m.
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The Unit Director convened a meeting with his operations and engineering staff to review the status of the plant, the PASS system and the technical specification requirements. The NRC
inspector attended the meeting to observe the deliberations. This review determined that
containment integrity requirements were satisfied with valve 151C secured in the closed
position, and that Technical Specification 3.6.3 was the only applicable specification for the
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condition of the PASS valves. Entry into Technical Specification 3.6.1 was appropriate
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under the circumstances confronting the shift supervisor, but conservative. When this position was conveyed to the control room, the load reduction was halted and the plant was stabilized at 87% FP. The shift supervisor tagged the power supply to the 151C valve in the
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"off" position. Actions were taken to notify the NRC that the unusual event had been
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terminated. The inspector identified no inadequacies in either the licensee's actions to comply with the requirements for containment integrity, or the implementation of the -
The plant was retumed to full power at 6:20 p.m. the same day after actions were taken to assme containment integrity was met. Investigations continued to correct the problem with valve 151D.' The plant operators remained actively involved in the ongoing operability determinations and troubleshooting activities for valve 151D. The operators critically i
reviewed corrective action plans to assure plant safety was not compromised. The licensee subsequently determined that the valve had in fact closed upon demand, but a problem in the position indication circuitry did not accurately show the status of the valve.
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3.0 MAINTENANCE 3.1.
Maintenance Observation
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The inspectors observed various corrective and preventive maintenance activities for compliance with procedures, plant technical specifications, and applicable codes and
standards. The inspectors also verified appropriate quality services division (QSD)
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involvement, proper use of safety tags, appropriate equipment alignment and use ofjumpers, adequate radiological and fire prevention controls, appropriate personnel qualifications, and adequate post-maintenance testing. Portions of activities that were reviewed included:
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e PMP 9.5-252,480 Volt Bus Breaker Setpoint Verification e
'AWO 93-11897, MS-SV-14, Flange leakage e.
AWO 93-12998, Replace MS-SV-14
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e AWO 93-12373, Pressurizer Heater Group A Breaker Replacement-
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AWO 93-12388, Heater Group A Breaker 52 X Relay j
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AWO 93-12678, SS-SOV-151D Dual Indication 3i e
AWO 93-12701, SS-SOV-151D Leakage Test e
AWO 93-04224, EDG "A" Intake Damper Installation j
3.1.1 AWO 93 04224, Emergency Diesel Generator ' A' Intake Duct Damper Installation r
On September 14, the inspector observed maintenance and contractor personnel install the i
intake duct damper for the ' A' emergency diesel generator. The inspector noted that
scaffolding supporting the installation did not impact diesel generator access or operability, i
Workers were aware of compensatory actions if the diesel generator automatically started.
Appropriate fire watches were stationed based on grinding / welding and removal of the fire -
j deluge system for the generator cubicle. The inspector noted no unacceptable conditions l
during the work observation.
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3.1.2 AWO 93-12373, Pressurizer IIcater Group A Breaker Replacement Following a routine test with the plant at full power on September 20, backup Group A of
l the pressurizer heaters did not reenergize at 9:14 a.m. when the operator attempted to control
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it from the main control board. Loss of the heater had no immediate impact on the control-of pressurizer level or on plant operation. ' The operators entered Technical Specification - (TS) 3.4.3.b, which allows for continued plant operation up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with _one inoperable:
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pressurizer heater group. Licensee investigation identified a problem with the operation of
the circuit breaker, located in compartment 2B of 480 volt Bus 4. A spare breaker 'was
installed in compartment 2B under AWO 93-12373 and tested satisfactorily two times. Plant
operators declared the heater group operable and' exited the TS action statement at 11:15--
a.m. on September 20.
The circuit breaker was a Westinghouse model DB-25 type breaker. Subsequent licensee
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investigation determined that the 52 X-relay failed to pick up properly and was the cause for
the breaker to close on demand. ' AWO 93-12388 was issued to replace the 52 X-relay and
document his. failure. The licensee noted this event represents another example of a-i previously identified concern regarding the operation of 52 X-relays in DB-25 circuit'
breakers (reference NRC_ Reports 93-80 and 93-12). The resolution'of the generic issue raised by these failures will be reviewed as part of the NRC followup of the previous j
concerns. The inspector had no further comment regarding the licensee actions to assure
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operability of the pressurizer heaters.
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3.1.3 AWO 93-11897, MS-SV-14 Flange Leakage
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On September 7, the licensee identified a minor body to bonnet steam leak from the flange l
area of main steam safety valve MS-SV-14. The valve was newly installed during the Cycle 17 refueling outage as part of a design change to provide for alternate methods to conduct a r
rapid plant cooldown. Licensee evaluations determined that the main gasket in the body to bonnet flange was leak tight, and the leak was most likely due to a failure of a grafoil gasket in a port on the valve head that provides steam to the " solenoid unloader", which is part of
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the control circuit in the power operated relief valve. No further action was taken to replace or repair the valve during the present inspection period. Site and Northeast Utilities Service Company (NUSCO) engineering was enlisted to study options to address the problem,
including the use of Furmanite to provide a temporary repair. The licensee approached the Furmanite repair option with full consideration of the NU corporate policy on the use of i
Furmanite, and in view of the lessons learned on the use of that process from recent events at Millstone 2. Refer also to paragraph 4.1 of this report for further discussion oflicensee actions concerning leak sealing actions.
The inspector reviewed the condition of MS-SV-14 when the leak was identified and
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periodically during the inspection period. This review determined that the leak was minor and had no impact on any other component of the main steam system or adjacent plant equipment. This review determined that continued operation with the minor leak did not impact plant or worker safety. The leakage remained minor through the end of the period.
i Licensee actions to address MS-SV-14 will be followed as part of the routine NRC review of plant operations.
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3.1.4 AWO 93-12678, SS-SOV-151D Dual Indication Licensee actions to identify and respond to this inoperable containment isolation valve are i
discussed in Section 2.5 above. The inspector reviewed activities under AWO 93-12678 to troubleshoot and repair the valve. The licensee's reviews determined that the valve was in fact closed even though the dual indication (when present) indicated that the valve was open.
The dual indication was determined to be caused by a faulty reed switch in this solenoid operated valve. The corrective actions were to replace the valve with a spare.
The inspector reviewed the actions by operations, maintenance, engineering and chemistry perscanel to evaluate and resolve this issue. There was good communication and coordination amongst all parties involved to assure prompt resolution of the operability concern. The plant operators, in particular, did a good job to keep good oversight of the
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operability determinations and troubleshooting efforts.
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3.2 Surveillance Observation The inspectors witnessed selected surveillance tests to determine whether: frequency and
action statement requirements were satisfied; necessary equipment tagging was performed;
test instrumentation was in calibration and properly used; testing was performed by qualified personnel; and, test results satisfied acceptance criteria or were properly dispositioned.
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Portions of activities associated with the following procedures were reviewed:
SUR 5.1-153A, AC and DC Distribution Normal Configuration e
ST 11.7-128, PDCR 1339 Preoperational Testing e
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e ST 11.7-129, EDG 2A Heat Generation Test Summary of Test Plan for PIR 93-139, MCC5 Auto Bus Transfer e
3.2.1 Auxiliary Feedwater Pump Surveillance On September 14,1993 during performance of surveillance procedure SUR 5.1-13B,
" Auxiliary Feed Pump P-32-1B Functional Test," the licensee noted that the acceptance criteria for speed and steam inlet pressure were not met. CYAPCo concluded that the 'B'
Auxiliary Feed Pump was operable since the technical specification surveillance requirements were met, and no in-service test (IST) acceptance criteria were violated.
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The inspector reviewed the basis of CYAPCo's conclusion with cognizant engineer personnel, reviewed technical specification operability requirements, and AFW expected system response. The inspector concluded that CYAPCo's operability determination was appropriate. The inspector will review future CYAPCo actions to understand anomalous control fur tion response in future inspections.
3.2.2 ST 11.7-128, PDCR 1339 Preoperational Testing The inspector reviewed the results of the licensee tests completed under ST 11.7-128 to measure the ventilation flow rate in the diesel generator rooms. The tests were performed to address concerns previously identified (reference NRC report 50-213/92-80) regarding the adequacy of the room ventilation flow to support both the operation of the diesel engine and to assure adequate cooling of the room. Using heat generation rate data originally supplied by the diesel manufacturer, the licensee had previously determined that the minimum flow rate needed from the room ventilation system was 36,000 cfm, which would supply 11,000 cfm needed for engine operation, and 25,000 cfm for room cooling. Previous testing had shown that room ventilation flow rate was less than 36,000 cfm for the existing room ventilation intake configuration. Thus, to maintain the diesel generators operable, the licensee instituted administrative controls to assure that the diesel room doors were kept open.
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Actions to modify the intake system configuration were completed during the Cycle 17 outage under plant design change record (PDCR) 1339. The testing completed during this inspection period with the modified damper showed the following room ventilation flow l
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EDG-2A 33,109 cfm
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EDG-2B 33,858 cfm The flow for both rooms was less than the minimum acceptable values of 36,000 cfm. This
testing was completed with an open intake duct, and prior to the installation of the back draft dampers. It was expected that the back draft dampers should further reduce the total flow into the room. Based on the above, the licensee maintained the established administrative controls in effect to assure diesel operability. The matter was referred to site and NUSCO for further review to resolve the issue. Actions were in progress at the close of the
inspection under ST 11.7-129 to measure the actual heat input into the room when the diesel l
is operating. It was expected that the heat generation rate by the engine was less than
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previously assumed and that actual values would support lower flow rates from the room s
ventilation system. CYAPCo actions to resolve this issue will be reviewed during subsequent
routine inspections of licensee engineering and surveillance activities. No inadequacies were identified regarding the testing completed under ST 11.7-128, or to maintain the operability
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of the diesels.
3.2.3 MCC 5 Transfer Failure - Test Plan Results i
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The automatic bus transfer (ABT) for motor control center (MCC) 5 failed during testing on
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June 27,1993. Plant information report (PIR)93-139 described the event and the licensee's immediate corrective actions. The long term corrective actions included a detailed test plan to investigate the causes of the ABT failure. The test plan focused on two components:
Agastat relay 62-6A, and 52X relay from breaker 11C. These components were deemed to
be the most probable cause of the ABT failure analysis of the June 27 event. The inspector reviewed the test results, as documented in memorandum MSM 93-218, with the Maintenance Engineer.
The test plan was very extensive and included several inspections, checks, tests and measure-l ments. The items tested by the plan included cyclic testing of the Agastat relay and the 52X relay individually, and combined cyclic testing of the Agastat and 52X relay in a DB-25 breaker in a special test circuit. The relays and the breaker were subjected to over.1500
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operations during the combined test. The components performed as expected during all tests.
All checks and measured parameters (e.g., contract resistances, coil drop-out voltages, etc.)
were well within acceptable ranges and compared well with values measured on new components. No physical evidence was found that would prevent the suspect components from working as designed. Other tests were conducted to measure the performance of the components under less than ideal conditions (e.g., undervoltage). The breaker performed l
satisfactorily in all tests.
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The test program failed to positively identify the root cause of the MCC5 ABT failure on June 27. Misoperation of the suspect components remains the most probable giuse for the event. Extensive inspections and testing were not successful in causing a repeat of the initial failure. One possible (although not probable) cause still under investigation by the licensee was the " sticking together" of the stationary and moveable cores in the 52X relay. ' The licensee was able to repeatedly demonstrate that certain of these components can be made to stick together on the work bench. This was last demonstrated in the presence of NRC inspectors on September 8. This issue was under review with the breaker vendor (Westinghouse). Possible causes (such as residual magnetism and adhesion of dissimilar materials used to plate the cores) and corrective actions were under evaluation.
The test plan was thorough to adequately measure pertinent parameters in the suspect components that could cause the ABT failure. The test plan was systematically and professionally executed to check the suspect components. The inspector concluded that the cause of the June 27 ABT failure was indeterminable at this time, in spite of the efforts by maintenance and engineering personnel. No inadequacies in the CYAPCo corrective actions were identified.
The licensee has several measures in place (as identified in the NRC AIT Report 93-80) to enhance the reliable operation of the MCC5 ABT. As committed to the AIT, a proposed change to the technical specifications to allow online testing of the ABT was submitted to the NRC on August 18, 1993. CYAPCo actions relative to MCC5 will be reviewed during subsequent routine inspections of plant operating activities.
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4.0 ENGINEERING AND TECHNICAL SUPPORT The inspectors reviewed selected engineering activities. Particular attention was given'to
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safety evaluations, plant operations review committee approval of modifications, procedural
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controls, post-modification testing, procedures, operator training, and UFSAR and drawing revisions.
4.1 Leak Scaling Program CYAPCo initiated a review and upgrade of its program to install leak seals as a result of an event at Millstone Unit 2 (documented in NRC report 50-336/93-18). The program was also changed to address a new NU corporate policy dated September 20,1993 on " Repairs
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During Power Operations." The inspection scope consisted of a review of the applicable procedures and licensee changes to the procedures during the inspection period, a comparison of the current procedures to the Electric Power Research Institute (EPRI) NP-6523-D Guidelines entitled "On-line I2ak Scaling," and a review of the leak repairs currently installed in the facility.
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CYAPCo has three principal procedures to perform leak sealing. The procedures are CMP
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8.5-149. " Leak Sealing of a Flange or Bonnet;" ENG 1.7-83, " Engineering Guidelines for Irak Sealing Installations"; and CMP 8.5-207, "FW-FCV-1301, 2, 3, 4 Leak Sealing of Main Feedwater Regulating Valves". CMP 8.5-207 provides instructions on leak sealing
feedwater regulating valve packing. ENG 1.7-83 provides guidelines for the site / corporate
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engineering evaluations required for the nonconformance report (NCR) written to perform leak sealing on QA category 1 components. Procedure CMP 8.5-149 provides one method for leak sealing valve packing leakage, and four methods for leak sealing valve flange or bonnet leakage.
CYAPCo used two vendors in the past to implement the procedures (Furmanite and Team, l
Inc.). Both vendors are on the approved vendors list under conditional requirements. The i
requirements are that licensee personnel to verify field activities and evaluate performance of
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the vendors on an annual basis.
i Procedures CMP 8.5-149 and ENG 1.7-83 were revised and reviewed in Plant Operations l
Review Committee (PORC) meetings on September 15 and 16. Sorae of the changes to CMP 8.5-149 were to: (1) require CYAPCo job supervisor familiarity with the leak sealing
procedure and to provide continuous oversight of the vendor; (2) provide a PORC approved safety evaluation prict to leak repairs on either safety or non-safety related components; (3)
stipulate pressure gauge accuracy for injection of sealing compounds; (4) require Unit Director approval for all leak seal repairs; (5) specify maximum allowable torque valves for valve fasteners; and, (6) add more Quality Department Inspection Hold points on valve torque verification. CYAPCo also revised ENG 1.7-23 to incorporate additional guidance from EPRI NP-6523-D. The revised procedure added injection compressibility factors, and now requires the development of safety evaluations.
The inspector reviewed the EPRI NP-6523-D guidance with the revised ENG 1.7-23. The inspector noted that the procedures did not ensure that sealing cap nuts for Class I components met the ASME Section III criteria. Further, the engineering procedure did not i
specifically require an evaluation of replacement studs for system pressures greater than 600 psig (EPRI NP-6523-D section 5.3) and CMP 8.5-149 did not include the recommended minimum training requirements for the vendor personnel and the job supervisor (reference
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EPRI NP-6523D).
The inspector discussed the above issues with the system engineer who explained that the cap nut qualification would be considered in the authorized work order
process (AWO), and that stud replacement is considered in the engineering evaluation, even though not required by the procedures. The licensee initiated procedure changes to address these issues.
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As of the inspection period, one leak seal was installed on a QA Category I component, the No. I main feedwater pump check valve bonnet. This maintenance was observed by the inspector and documented in NRC Report 50-213/93-16.
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Conclusion
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CYAPCo implemented significant changes to the leak seal program to improve vendor oversight, to provide additional quality services department involvement, to provide better consistency with industry recommendations provided in EPRI NP-6523-D, and increased management attention and approval of leak scaling. The inspector noted that the program could be enhanced by having the procedures address minimum training requirements and practices already in place. The licensee concurred with the inspector and initiated changes to
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station procedures to address the inspector's comments. Licensee implementation of the leak sealing will be reviewed during subsequent routine inspections.
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4.2 Safety Evaluations and Technical Specification Clarifications
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During the period of September 8-11,1993, the NRR Project Manager and a Reactor Engineer Intern performed a' review of the licensee 10 CFR 50.59 review process and Procedure ODI-173, Technical Specification (TS) Clarifications at Haddam Neck Plant.
Procedures and specific examples of 10 CFR 50.59 modifications and TS clarifications were reviewed.
10 CFR 50.59 Review
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CYAPCo has established formal procedural guidance for evaluating each change, test or experiment for which 10 CFR 50.59 is applicable to determine whether an unreviewed safety question (USQ) or a change to the TSs is involved. Formal procedural guidance is contained in Nuclear Engineering and Operations Procedure (NEO) 3.03, " Plant Design Change Records'(PDCR)," and NEO 3.12 or Administrative Control Procedure 1.2-6.9, " Safety Evaluations." NEO 3.03 provides detailed instructions and forms for preparing a PDCR.
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As part of this procedure, the requirement for an integrated safety evaluation is determined.
The procedure provides the methodology for determining if a detailed analysis is required or
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if the PDCR does not alter plant safety performance and can be completed with minimal -
engineering design work. In either case the closeout process assures that sound judgements
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were made and that the design and imp!ementation of the PDCR followed the appropriate procedures.- Ifit is determined that an integrated safety assessment is needed, the preparer is referred to NEO 3.12. NEO 3.12 provides the process of preparing a safety evaluation, which determines if a proposed plant change is safe and satisfies the requirements of 10 CFR 50.59. The procedure contains instructions and flow charts for evaluating the plant change and includes seven questions which help *he preparer decide if the plant change is a USQ.
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The PDCRs reviewed during this inspection included replacement of emergency diesel generator (EDG) heat exchanger temperature instrumentation, replacement of station service
transformers, modification of containment air recirculation fan safety injection actuation circuits, EDG cubicle ventilation modifications, and modernization of feedwater controls.
These PDCRs provided the inspectors with examples of changes which were determined to require minimal design review (did not alter plant safety), detailed review without a safety a
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evaluation, and detailed review with a safety evaluation. In each case, the inspectors found that the licensee's implementation of the PDCRs followed the appropriate procedures, were correctly assigned and coordinated, were reviewed by the appropriate departments and the need for a safety evaluation was conservatively decided. In cases where it was determined that a safety evaluation was required, the appropriate documentation, as required by NEO 3.12, was completed in detail and provided a sound basis for the USQ determination. The inspectors noted that the procedure should delineate the fact that any plant change requiring a TS change requires prior NRC staff review and approval. While there are steps to determine if a TS change is required, the steps do not clearly state that this would preclude the plant from making the change under 10 CFR 50.59.
In summary, the inspectors verified that the licensee has established formal procedural guidance and controls to evaluate each plant design change, for which 10 CFR 50.59 is applicable, to determine whether a USQ exists or a TS change is required. The need for a safety evaluation is defined in the PDCR procedure and NEO 3.12 provides appropriate guidance for preparing the safety evaluation, which determines if a proposed plant change is safe and satisfies the requirements of 10 CFR 50.59.
Technical Soecification Clarifications The inspectors reviewed Procedure ODI-173, Revision 3, "TS Clarifications". This procedure provides a method of logging and tracking clarifications of TS. A formal book of TS clarifications is maintained and reviewed by the Operations Manager. In addition, it is the Operations Manager's responsibility to determine if TS changes are needed. The procedure provides instructions for logging and review of TS clarifications. When a clarification of TS is made, the operator involved in the decision will complete a TS Clarification sheet. This sheet provides a discussion of the TS, the TS title, associated TSs.
and guidance on deciding if a TS change is required. The current TS Clarification book had 29 clarifications of which four were determined to require TS changes. The inspectors agreed with the licensee that the remaining 25 TS clarifications were not of sufficient safety significance to warrant a TS change. The inspectors noted that the four clarifications recommended for TS changes have remained in the log book for two years. One of the TS changes has been submitted to the NRC while the other three are still being processed. The inspectors agree, however, that the decision to process four TS changes was conservative.
The inspectors were unable to observe the operators' use of the TS Clarification book.
However, the inspectors confirmed that the control room operators were aware of the TS Clarifications book and the procedure to establish new clarifications. The inspectors concluded that this procedure is effective in implementing a consistent interpretation of the TSs, and provides the basis for the interpretation and a method to revise those TSs that need to be change,
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5.0 PLANT SUPPORT 5.1 Radiological Controls During routine inspections of the accessible plant areas, the inspectors observed the imple-mentation of selected portions of the licensee's radiological controls program. Utilization and compliance with radiation work permits (RWPs) were reviewed to ensure that detailed-descriptions of radiological conditions were provided and that personnel adhered to RWP requirements. The inspectors observed control of access to various radiologically controlled areas and the use of personnel monitors and frisking methods upon exit from those areas.
Posting and control of radiation areas, contaminated areas and hot spots, and labelling and control of containers holding radioactive materials were verified to be in accordance with licensee procedures. Health physics technician control and monitoring of these activities were determined to be good.
5.1.1 Investigation of Worker Injury Inside Containment On June 10,1993, a worker was injured while working inside the Haddam Neck containment building. The inspector reviewed the circumstances of the accident at that time to determine the nature of the injuries, the impact on plant nuclear safety and the potential for a radiological hazard offsite. There was no impact on plant nuclear safety, and there was no radiological hazard offsite. The inspector reviewed the cause of the accident and noted the licensee actions taken to preclude recurrence.
During this inspection period, the inspector reviewed the results of the licensee's final
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investigation of the incident, including the circumstances surrounding the event and its causes. This inspection was limited to a review to determine whether licensee actions prior to and after the accident provided for the radiological safety of the workers and the public.
- This review was done with representatives from the licensee's maintenance, safety and personnel departments. The inspector also met during this period with a representative of the Occupational Safety and Health Administration (OSHA). The purpose of this meeting was to assure that all pertinent information available to the NRC regarding the accident was also available to OSHA in its review of the accident and industrial safety at Haddam Neck.
Summary of Event On June 10,1993, the plant was in a shut down condition for a refueling and maintenance outage. Two contractor workers with Nuclear Support Services (NSS) performed work on the containment air recirculation (CAR) fans inside the containment. The work was performed near CAR fan #3 in the outer annulus area on the middle level of the containment building. The radiological hazards for thejob were controlled under radiation work permit 93-0056. In addition to the two NSS workers, contract workers with Babcock and Wilcox
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(B&W) Company were in the area pulling cables and performing activities to clean steam
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i generator #3. Stone and Webster (S&W) contract employees were also in the area perform-
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ing work associated with the feedwater system modifications. Finally, other CYAPCo and plant workers were either in or passed through the area for other outage related activities.
The NSS employees were working in a sitting or kneeling position on components spread out on a work area on the floor. A CAR fan damper was in storage near the NSS work area, l
propped up against a block wall. The damper is made out of sheet metal, measures about 4
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feet by 5 feet by 1 foot, and reportedly weighs about 300 pounds. The licensee's investiga-
tion determined that the damper had been stoled in the area prior to June 10 in preparation
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for subsequent installation. The damper was secured to the wall with a rope. A licensee j
maintenance supervisor had observed that the rope was in place at about 1:00 p.m., and was
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noted to be in place by NSS workers as late as 4:30 p.m. on the day of the accident.
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However, at the time of the accident, the damper was no longer secured by the rope and was resting in an upright position against the wall. At about 9:30 p.m., the CAR fan damper fell over and landed on top of one of the NSS workers. The licensee was unable to deter-mine how and when the rope became unsecured.
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Other workers immediately helped the worker and contacted the control room for assistance.
A medical emergency was declared. The individual sustained injuries that required treatment offsite. Plant and emergency medical personnel responded to assist the worker, remove his protective clothing and check him for radiological contamination. His protective clothing was removed at the scene. The worker was checked by health physics personnel found to be
free of contamination. He was transferred to the ambulance that was brought into the
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protected area. The injured worker was removed from the site by 10:30 p.m. and was transported to Middlesex Hospital for treatment. A health physics technician accompanied i
the injured person to the hospital to provide assistance as needed to hospital personnel.
Followup action by the licensee and its contractors included: securing the damper in place; reviewing the circumstances of the accident and interviewing all workers in the area at the
time of the event; a review of containment work areas to verify safe conditions; discussion of the event during the daily meeting on June 11; and, a followup discussion with plant work groups with a focus on the need to verify safe conditions in the plant work areas and the
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need to recheck conditions periodically. An assessment of the actions by emergency responders was conducted by CYAPCo employees who were not involved in the event. This assessment had specific focus on the adequacy of health physics controls. The assessment results were described in Audit No. HPA-93-18 dated June 30,1993. Although this
assessment identified no deficiencies in health physics practices, some recommendations were made to enhance health physics controls and procedures.
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NRC Findings and Conclusions The inspector reviewed this event to determine whether licensee actions were in accordance with 10 CFR 20, " Standards For Protection Against Radiation." The inspector further reviewed the medical emergency actions taken by the licensee to verify conformance with
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EPIP 1.5-11, " Personnel Injuries." Radiological conditions in the area where the accident occurred were described on CYAPCo survey sheets dated June 9,1993 for the "#3 CAR Fan", and June 10, 1993 for the " containment mid-level outer annulus." Radiation and I
contamination levels in the work area, and in the access ways from the containment access to the work area, were low and close to background levels that are normally found outside of the radiologically controlled area.
The inspector reviewed the requirements of RWP 93-0056 and found them acceptable to protect the workers from the exposure and contamination hazards documented on the survey
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forms. The actual exposures received during the work period by the injured worker and others in the work party were low and well within regulatory limits. Based on the above, the inspector concluded that the licensee controls to assure the radiological safety of the worker were proper. As noted earlier, the industrial safety aspects of this accident were discussed with OSHA for their review.
The inspector also found that the actions by licensee and emergency response personnel conformed with the requirements of EPIP 1.5-11. Since the worker was free of radioactive contamination when transported to the hospital, no 10CFR50.72 report by the licensee to the
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NRC was required. Based on a review of the licensee's investigation files and interviews of
some of the personnel involved with event, the inspector concluded that licensee health physics and emergency response actions were acceptable to assure the radiological safety of the injured worker and offsite response personne:. Based on the above and within the scope of this inspection, no violations of 10 CFR 50 or 10 CFR 20 were identified.
1, 5.2 Review of Written Reports L
Periodic reports were reviewed for clarity, validity, accuracy of the root cause and safety l
significance description, and adequacy of corrective action. The inspectors determined
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whether further information was required. The inspector also verified that the reporting requirements of Technical Specification 6.9 had been met. The following reports were reviewed:
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1993 Steam Generator Tube Inspection Report
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No inadequacies were identified.
5.3 Follow-up of Previous Inspection Findings
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Licensee actions taken in response to open items and findings from previous inspections were reviewed. The inspectors determined if corrective actions were appropriate and thorough and i
whether previous concerns were resolved. Items were closed where the inspector deterrgined that corrective actions would prevent recurrence. Those items for which additional licensee action was warranted remain open. The following items were reviewed:
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(Onen) Inspector Follow Item 93-01-01 < Emereency Diesel Generator Tachometer Not e
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Classified as Safety Related Nor Subiected to Routine Calibrations
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This item concerned an observation by the inspector, that the emergency diesel generator tachometer is used to satisfy technical specification surveillance requirements, however the instrument is not considered safety-related, nor is it subjected to a routine surveillance
program. The licensee committed to further review the calibration program for diesel instruments. CYAPCo initiated controlled routing 93-0402 to evaluate the need to perform routine calibrations of instruments used in emergency diesel generator surveillances, and to
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learn if other surveillances of safety related equipment use instruments not subject to routine calibrations. CYAPCo decided to calibrate the emergency diesel generator tachometers every
refueling outage pursuant to preventive maintenance procedure PMP 9.5-36, "EG-2A and
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EG-2B Emergency Diesel Generator Outage Inspection and Preventive Maintenance." The assignment directs completion of the procedure revision by December 1,1993. CYAPCo
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initiated additional controlled routing assignments to the remaining line departments to identify instruments that a,e not routinely calibrated and are used to satisfy technical specification requirements. The assignments are due by December 31,1993. Additionally, Connecticut Yankee Nuclear Review Board issued an item carried forward (ICF) 93-07-05 to ensure that the concern is being address by a controlled routing. This item will remain open, pending NRC review of the basis for NRB's action, and the completion of actions by CYAPCo.
(Closed) Insocctor Follow Item 92-20-02. AFW Hydraulic Hoses - PIR 92-178
The licensee's preliminary conclusions, as documented in NRC Inspection Report 92-20, were that the Goodyear auxiliary feedwater pump hose failed on October 6,1992 due to incompatibility with the operating hydraulic fluid. The fluid in use was Houghtosafe 620, an ethylene glycol and water hydraulic fluid. The Goodyear hoses are a conventional metal-braid reinforced rubber hose, constructed of a nitrile rubber tube and neoprene rubber cover.
Subsequent licensee discussions with both Goodyear and E.F. Houghton confirmed that Houghtosafe 620 is compatible with the nitril/ neoprene hoses. This confirmed the adequacy
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of the engineering design reviews in support of PDCR 91-1127.
The licensee's subsequent and final conclusions for this event were described in PIR 92-178, which provided the results of the engineering evaluation and root cause analysis for the failure of the hydraulic hose. The failure investigation was also summarized in Technical Report TR-MCC-267 dated April 23,1993, by ABB Combustion Engineering. Laboratory analysis of the failed hoses identified a foreign substance, not Houghtosafe 620, absorbed into the elastomer cover of the hose. Laboratory testing shond the substance to be an organic ester-based fluid. The substance resembled the fluid originally found in the sumps of
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the hydraulic skids, which was shipped with the skids when supplied to Haddam Neck.
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When mixed, the ester-based fluid floated on the surface of the Houghtosafe 620. The
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hydraulic pump discharge hoses are located within the pump skid sump and are partially submerged in the hydraulic fluid. The hydraulic hoses w re found to be degraded at a point corresponding to the surface of the liquid in the sum;.
The licensee's investigation found that the hydraulic skids were shipped to Haddam Neck with a fluid already m the sump reservoir. Even though CY Maintenance flushed and wiped-down the skids prior to refilling the sumps with Houghtosafe 620, enough of the ester-based fluid remained in the sump to attack the hoses. The licensee concluded that the skids were-tested with a petroleum based hydraulic oil by a Woodward sub-contractor prior to shipment of the units to Haddam Neck. The subcontractor was unaware of the incompatibility between
the fluid and the hose material.
The inspector identified no inadequacies with CYAPCo's engineering evaluation or the conclusions related to the October 1992 hose failures. The root cause investigation and
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engineering evaluation were complete and very thorough. The final results of the investiga-tion confirmed the adequacy of the corrective actions taken in 1992 to address the failed
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hoses. This item is closed.
(Closed) Unresolved Item 92-08-01. Surveillance Program Tracking Deficienci,
This item was last reviewed in Inspection 92-26, which found that the licensee had *.nproved As methods to schedule the conduct of technical specification surveillances. That inspection j
I also noted a significant improvement in surveillance tracking, as evidenced in the generally large reduction in the number of missed surveillances in the 1991 to 1992 period. The item l
was left open pending the development of additional actions to further strengthen the program
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and to address the recommendations of June 3,1992 review of the surveillance tracking program by the human performan~e evaluation system (HPES) coordinator.
This item was reviewed during this inspection to verify the completion of the corrective
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actions. CYAPCo issued supplement I to LER 92-05 dated May 12, 1993, to report that the root cause of the previous performance issues was a program failure. The LER also de-scribed action taken to assure the proper implementation of the program.
l Subsequent actions by CYAPCo on this issue included active involvement by management at
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the Unit Director level E assess the existing program and the options for improvement, as described in the following internal memoranda UD-93-032, UD-92-143, and, UD-92-110.
CYAPCo reviewed the existing tracking program and options to improve it, and considerext~
j assigning additional personnel in each department with the prime responsibiSty for surveil-i lance tracking. CYAPCo also considered the development of a new computerized tracking system to monitor administration of the program. CYAPCo elected to strengthen the existing i
manual tracking programs. Each plant department uses two independent surveillance
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tracking systems. Each department also has a primary and a backup person assigned who are j
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capable of administering the programs. Changes to enhance the program were incorporated in procedure ACP 1.0-59, " Technical Specification Surveillance Tracking," Revision 4, dated August 10, 1993.
The inspector noted that no licensee event reports (LER) were issued for the failure to complete a surveillance on time in the last 20 months. The last LER on this subject was LER 92-05 dated February 13, 1992. Based on the subsequent performance in completing surveillances, the corrective actions were effective and the program to track surveillances is achieving a high level of success. This item is closed.
6.0 EXIT MEETINGS During this inspection, periodic meetings were held with station management to discuss inspection observations and findings. At the close of the inspection period, an exit meeting was held with Mr. J. Stetz to summarize the conclusions of the inspection. No wTitten material was given to the licensee and no proprietary information related to this inspection was identified.
In addition to the exit meeting for the resident inspection held on October 4,1993, the following meetings were held for inspections conducted by Region I based inspectors.
Inspection Reporting Areas E g o rt N o.
Dates Inspector Inspected 3 -213/93-18 9/27-10/2 L. Peluso Environmental Monitoring Program
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