IR 05000213/1992026
| ML20034F597 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 02/25/1993 |
| From: | Doerflein L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20034F593 | List: |
| References | |
| 50-213-92-26, NUDOCS 9303040037 | |
| Download: ML20034F597 (24) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
Report No.
50-213/92-26 License No.
DPR-61 Licensee:
Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06141-0270
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Facility:
Haddam Neck Plant Location:
Haddam Neck, Connecticut inspection Dates:
December 13,1992 to January 9,1993 Inspectors:
William J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector Robert De La Espriella, Reactor Engineer Trainee
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Approved by:
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Lawrence T. Doerflein, Chief [/
Dat'e
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Reactor Projects Section No. #A Areas Inspected:
NRC resident inspection of phnt operations, ' radiological controls',
maintenance, surveillance, engineering and technical support, licensee self-assessment, licensee l
event reports (LERs) and previously identified items.
Inspection initiatives selected were operational experience feedback program and verification of the locked valve checklist.
Results: See Executive Summary
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9303040037 930226 PDR ADOCK 05000213 O
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EXECUTIVE SUMMARY i
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IIADDAM NECK PLANT INSPECTION 92-26
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Plant Operations f
The control room operator's response to reactor coolant system leakage from the post accident sample system module was prompt and effective to mitigate the event. The inspector noted good
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coordination between Operations, Chemistry, and Health Physics personnel to stop the leakage,
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assess the consequences, and to decontaminate the area.
- Maintenance and Surveillante i
The inspector determined that licensec actions to investigate and correct the cause of the leak from the post accident sample module were good. The procedures to collect and process a
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reactor coolant system sample in a harsh radiological environment' following a postulated i
accident are appropriate.
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CV APCo specialists displayed good attention to detail and procedural adherence'during routine.
i surveillance testing of the reactor protection system instrumentation and the charging system.
The inspector found CYAPCo action to detail and expand the lubricating oil analysis program for safety and non-safety related pieces of equipment was a good initiative.
Eneineerine and Technical Support
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The inspector noted good licensee assessment which identified a failure to update operations
critical drawings and surveillance procedures following a plant modification. The inspector also i
found corrective actions to be appropriate. Enforcement discretion was exercised based on
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prompt corrective action, and self-identification (section 4.1).
i Safety Assessment and Ouality Veri 6 cation The inspector identified a procedural adherence' issue involving the failure to document the review of past Plant Information Reports with similar root causes for recurrent events. This item is unresolved (section.5.3).
Quality assurance (QA) audits were effective in identifying CYAPCo corrective actions needed
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to assure good performance. Although the inspector noted that the bases for corrective actions could be better documented in some cases, the actions taken were appropriate. The corrective.
actions in response to QA findings contributed to the prevention of future problems.
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. SUMMARY OF FACILITY ACTIVITIES The unit operated at full power throughout the inspection period. No significant maintenance or operational problems occurred. On January 1,1993, Haddam Neck marked its 25th year of commercial operation.
On December 30,1992, Mr. T. T. Martin, Regional Administrator, NRC Region I, toured the facility and met with plant supervisors and managers. Overall, the plant physical condition was found to be satisfactory.
2.0 PLANT OPERATIONS (71707 and 93702)
The inspectors routinely reviewed plant operations during normal utility working hours and portions of backshifts (evening shifts) and deep backshifts (weekend and night shifts). During this report period, inspection coverage was provided for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during backshifts and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during deep backshifts.
2.1 Operational Safety Verification
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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 cognizance of control room alarms, control of technical specification limiting conditions of operation, and electrical distribution verifications.
Administrative control procedure (ACP) 1.0-23, " Operations Department Shift Staffing Requirements," identifies the minimuu staffing requirements. During the inspection period, the inspectors noted the control room staffing during power operations met these requirements.
The inspectors reviewed the onsite electrical distribution system to verify proper electrical lineup
of the emergency core cooling pumps and valves, the emergency diesel generators, radiation monitors, and various engineered safety features. The inspectors also verified valve lineups,
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positions of locked manual valves, power supplies, and flow paths for safety systems, including 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.
Log-Keeping and Turnovers i
The inspectors reviewed control room logs, night order logs, plant information report logs, and crew turnover sheets. No discrepancies or unsatisfactory conditions were noted. The inspectors observed crew shift turnovers and determined 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. The inspectors also reviewed control room trouble reports for age, Connecticut Yankee Atomic Power Company (CYAPCo)
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planned action, and operator awareness of the reason for the trouble report. Most trouble reports -
reviewed were recent, with few longstanding items.
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During attendance at daily planning meetings the inspector noted discussions were held on i
maintenance and surveillance activities in progress, and work control and authorizations. The inspectors conducted periodic plant tours in the primary auxiliary building, turbine building, and intake structures. The inspector determined plant housekeeping was satisfactory.
2.2 Radiological Controls During routine inspections of the accessible plant areas, the inspectors observed the implementat-ion of selected portions of the licensee's radiological controls program. The inspectors reviewed utilization and compliance with radiation work permits (RWPs) to ensure that detailed-descriptions of radiological conditions were provided and that personnel adhered to RWP.
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requirements. The inspectors observed controls of access to various radiologically controlled areas and the use of personnel monitors and frisking methods upon exit from those areas. The inspectors also noted posting and control of radiation areas, contaminated areas and hot spots, and labelling and control of containers holding radioactive materials were in accordance with licensee procedures. The inspectors determined that the health physics technician control and monitoring of these activities were good.
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2.3 Follow-up of Events Occurring During Inspection Period 2.3.1 Post-Accident Sampling System Leakage
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On December 22,1992, at approximately 9:20 a.m., the control room operators received the sample room area radiation monitor (RMS-36) and the volume control tank (VCT) rate of change alarms.
At the time, chemistry personnel were performing procedure CHM 7.6-23,
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" Performance Testing of Reactor Coolant Post Accident Sampling Module." This procedure acquires a reactor coolant system (RCS) sample through the Post Accident Sampling System (PASS). During the process, the chemistry technician observed that the actual RCS sample flowrate was approximately four times what was expected. The target flowrate in the procedure is 16 percent (%) (approximately 0.8 gallons per minute (gpm)) as read on the installed flowmeter, while the actual flowrate measured during the sample was 69%.
In response to the alarms, control room operators verified that RMS-36 radiation levels increased and VCT level decreased. Based on recognition of the RCS leakage and awareness of the chemistry sample in progress, and following consultation with chemistry management, the operators isolated the RCS from the PASS module by shutting valve (DH-MOV-544) at 9:31 a.m. This stopped the leak. Shortly thercafter operators shut valve (SS-AOV-950) for two-valve
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RCS boundary protection. The estimated leakage over the eleven minutes was approximately fifty gallons. The estimated leak rate from the PASS module was 3.5 gpm.
The leakage from the PASS module was directed to the floor drains (by design).to the aerated drain tank. A restriction in the drain system allowed RCS water to backup through the floor
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drains into an adjacent room (Drumming room). Operators directed health physics and auxiliary
operators to evaluate the extent of contamination in the affected areas.
All areas were decontaminated and restored to the pre-leakage radiological conditions. As a precautionary measure, chemistry personnel sampled the yard storm drains. No plant radioactivity was identified in the storm drains.
Chemistry determined that 0.48 curies were discharged in the gaseous release. When averaged over one hour, the release was 25 % of the allowable maximum permissible concentration for the isotopes in the mixture. The inspector verified that the release was significantly below TS
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requirements. Control room operators consulted the applicable emergency plan implementing procedures and the TS for radioactive releases. No reports were made or required.
The inspectors observed the control room operator response to the RCS leakage from the PASS module and determined it was prompt and thorough. Operators appropriately entered and exited technical specification (TS) 3.4.6.2. (unidentified leakage greater than 1.0 gpm). The inspectors noted good coordination between operations, chemistry, and health physics personnel in response
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to the event. The cause of the leakage and the licensee's corrective actions are further discussed in report detail 3.1.
2.4 Documentation of Fire Watch Rounds The licensee determined that on December 18,1993, the fire watch established for the 'A' and
'B' emergency diesel generator rooms had failed to log his readings for the 3:00 p.m. hourly round. The matter was described and investigated under plant information report (PIR)92-204.
The inspector reviewed the licensee's actions to disposition this issue.
The licensee determined by review of security records and discussions with the auxiliary operator
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that the fire barriers (the open doors to the diesel rooms) had been inspected. A security guard stationed in the area confirmed that the operator was at the doors at 3:00 p.m. However, the operator failed to enter the room to sign off on the fire watch round sheet, thus failing to document his last inspection prior to leaving for the day. The operator had properly completed the check for the seven previous rounds during his shift. The licensee instructed the operator on management's expectations regarding the proper completion of fire watch duties and.
disciplined the individual. The inspector had no further questions regarding the licensee's corrective actions.
The licensee initially classified the event as a matter reportablu to the NRC under 10 CFR 50.73 for failing to satisfy the requirements of TS 3.7.7 for degraded fire rated assemblies. The TS 3.7.7 limiting condition for operation requires, for an inoperable fire rated assembly, that a fire watch patrol inspect the areas (on both sides of the degraded assembly) at least once per hour.
The licensee questioned whether the TS had been met since the operator, although he could observe portions of the diesel generator rooms through the open doors, was physically on only one side of the barrier. The issue was referred to the site Fire Protection Engineer for resolution. In an memorandum dated December 21,1992, engineering established the position
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that "... the fire watch patrol person must inspect / observe the fire areas on both sides of the
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s inoperable barrier to satisfy himself/herself that there is no fire condition developing which could i
challenge the inoperable barrier." However, the licensee does not normally require that 100%
of the area on both sides of the barrier be observed by a fire patrol. Based on the above, the licensee determined that the TS requirements had been satisfied for the 3:00 p.m. round on December 18. Notwithstanding the above, the licensee expects the fire watch personnel to enter the diesel room to complete the inspections, and provided this instruction to fire watch personnel.
The inspector reviewed the technical specification requirements and its bases, ad reviewed the
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fire areas on both sides of the degraded barrier. The inspector noted that each diesel room has two areas; the area closest to the open door houses the switchgear for the respective diesel; and, the back area where the diesel skids are located. A significant portion of the switchgear area could be observed from one side of both open doors; a portion of the diesel skid could be
observed from the door only for the 'B' diesel generator. The inspector noted that the diesel skid areas were protected by operable fire detection and suppression systems.
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concluded that a valid assessment of a potential fire hazard could be made by observing the
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switchgear area only. The inspector did not identify inadequacies in the licensee's conclusion.
3.0 MAINTENANCE AND SURVEILLANCE (61726,62703 and 71707)
3.1 Maintenance Observation
The inspectors observed various corrective and preventive maintenance activities for compliance l
with procedures, plant technical specification 2, and applicable codes and standards. The l
inspectors also verified appropriate use of safety tags, equipment alignment and use ofjumpers, radiological and fire prevention controls, personnel qualifications, and post-maintenance testing.
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Portions of activities that were reviewed included:
i Post-Accident Sampling System Repairs a
t During a routine reactor coolant system sample using the PASS on December 22, a leak occurred i
from the sample capture chamber. The inspector reviewed the operational and radiological i
impacts of the event (as described in Section 2.3.1 above); there was no impact on worker, plant
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or public safety. The inspector also reviewed the CYAPCo activities to identify the cause for i'
the leak ar.d to restore the PASS to an operable status.
CYAPCo determined that the leak occurred as a result of a degradation in the " quick-connect"
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coupling used on the sample capture chamber normally used to collect samples from the PASS.
The capture chamber uses two couplings, one on each sample line, to connect the shielded j
chamber to the PASS module. The portion of the coupling on the chamber side of one of the j
sample lines had degraded such that, although a connection could be made, the connection could i
not withstand full pressure when the PASS was connected to the reactor coolant system. The i
primary sample capture chamber was tagged and removed from further service pending repair.
l A spare sample capture chamber was tested and found to operate satisfactorily. The inspector
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observed the hydrostatic test conducted on December 24 to verify that the system was leak tight.
The PASS system was returned to service on December 24 using the spare chamber.
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CYAPCo plans additional action as part of the follow-up of this event per PIR 92-206. This includes consideration of alternating use of the sample chambers.
Also, present sample procedures already include instruccon to assure the sample connection is leak tight prior to obtaining a sample. The licensee is considering enhancements to these instructions to better ensure the connection is leak tight. The inspector had no further questions regarding the licensee investigation and repairs for the PASS failure.
During his review of the PASS repairs, the inspector noted that the PASS sample module is located in the northeast corner of the HP Facilities building. During sample collection, the PASS is controlled from a control panel located in the chemistry laboratory. The panel is separated from the PASS module by distance and concrete block walls to shield plant personnel from the radioactive source term that will be present in the PASS sample lines during the post-accident period. The inspector noted that the PASS design did not include shielding around the sample lines in the. HP Facilities building. Since the licensee's plans assume that equipment in rooms adjacent to the PASS module area will be used for processing samples after an accident, the inspector questioned what provisions were established to assure the adjacent facilities could
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be used as intended.
The inspector reviewed EPIP 1.5-39, " Post Accident Sampling of Reactor Coolant," and plant drawings 16103-28038, " Isometric Drawing of Post Accident Sampling Piping," and 16103-
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26057, " Post Accident Sampling System." The inspector also walked down the PASS sample lines from the primary auxiliary building to the PASS sample module, and to the PASS control
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panel. The sample procedure includes provisions to flush radioactive fluid from the sample lines
to allow personnel access to the PASS module to collect the shielded sample chamber. The sample lines would be flushed from the blowdown tank room in the auxiliary building back to the PASS module, thus assuring personnel access to the adjacent rooms in the HP facilities building. Based on the above, the inspector had no further questions regarding the licensee's
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capability to collect and process a sample following an accident.
CMP 8.5-140, BT-1 A, B, C Post-Discharge, Recharge, or Equalize Charge of Station
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Batteries
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The inspector reviewed licensee testing of the 'C' station battery using procedure CMP 8.5-140.
The inspector determined that the preventive maintenance activities were completed successfully.
However, on December 21, the battery charger stopped providing an equalizing charge prior to -
the desired 208 hour0.00241 days <br />0.0578 hours <br />3.439153e-4 weeks <br />7.9144e-5 months <br /> charging time. PIR 92-205 was initiated to document and investigate the i
event.
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CYAPCo determined that the charger automatically shutdown after 199 hours0.0023 days <br />0.0553 hours <br />3.290344e-4 weeks <br />7.57195e-5 months <br /> when a timer
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internal to the charger reached its maximum set limit. The timer limit had not been recognized in the past because equalizing charge times were less than 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. The lic usee noted that the charger for the 'B' battery also had a timer. As with the 'C' battery, equalizing charge times -
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on the 'B' battery were always less than the timer limit. The 'A' battery charger does not have an internal timer. After resetting the timer, the equalizing charge for the 'C' battery was-completed without further problem. The licensee plans on revising the procedure to ensure the
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timer is reset for future battery equalizing charges. The inspector had no further questions regarding battery testing.
3.2 Surveillance Observation
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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. Portions of activities associated with the following procedures were reviewed:
SUR 5.7-144, Inservice Testing of 'B' Charging Pumps, and CII-CV-260,263,272
and 277 On January 6, the inspector observed performance of Sections 6.9 through 6.14 of SUR 5.7-144.
The surveillance objective was to acquire charging pump performance data in order to maintain operability. This quarterly surveillance test is intended to satisfy the requirements of TSs 4.1.2.3.1, 4.1.2.4.1, and 4.5.1.c.3 and section 4.0.5. The surveillance was performed by in-service test personnel, and supported by health physics and operations department personnel.
The inspector noted the results of the surveillance were acceptable. The inspector also noted good procedural adherence by personnel implementing the procedure and good health physics i
support for work within contaminated and high radiation areas.
SUR 5.2-126, Steam Generator Wide Range Level Overfill Protection Operational
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On December 29, the inspector observed performance of Sections 6.2 through 6.4 of SUR 5.?-
126. Instrument and Control specialists performed the surveillance in the control room and the
'B' switchgear room. The objective of the surveillance was to verify proper operation of the
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engineered safety feature protection circuitry. The inspector determined the surveillance data was within acceptance criteria, and the specialists displayed good attention to detail and procedural
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adherence.
SUR 5.1-126, Locked Valve Checklist
The inspector independently verified the position and condition of approximately forty percent
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of the required locked valves listed on attachment 7.4 of SUR 5.1-126. The inspector did not identify any valves to be in the incorrect position or unlocked. The inspector also evaluated the
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" Locked Valve Status Sheet," and concluded that appropriate approval, reasons for valve position change, and license conditions were adhered to. The inspector identified one set of valves routinely changed in accordance with the Locked Valve Status Sheet. Diesel fuel oil valves (FO-
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133A and FO-133B) were locked closed vs. the required position oflocked open. The basis for the change in position was to address a long-standing maintenance problem of upstream leak-by
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from air-operated valves (FO-LCV-1700A and FO-LCV-1700B). Closure of valves FO-133A i
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l and FO-133B prevents unintentional filling of the diesel underground storage tank from the above I
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i ground fuel oil storage tank (TK-33-1 A). The inspector verified TS limits for the storage tank and operability of fuel transfer to each respective diesel was maintained during repositioning of
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FO-133A and FO-133B.
During the plant walkdown, the inspector identified minor deficiencies concerning missing valve labels on valves FO-V-142B and SW-V-694. Also, SUR 5.1-126 had the wrong technical specification reference identifying the requirement locking the outlet valve for the boric acid makeup tank (BA-V-399). The inspector presented the deficiencies to CYAPCo management
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during the inspection period. Additionally, the inspector evaluated 1992 PIRs to determine if any deficiencies in control of locked valves were identified by the licensee. No PIRs were identified. The inspector determined CYAPCo maintained acceptable controls for manuallocked
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valves.
i 3.3 Lube Oil Analysis Program i
The inspector reviewed the revision to the maintenance department's lube oil analysis program.
The program change was prompted by previous licensee identified weaknesses, and the development of a predictive maintenance program and actions to address the maintenance rule
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(10 CFR 50.65 effective January 10, 1996). The previously identified program weaknesses
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inchided sampling techniques and the technical basis for sample frequencies.
l The program became effective on December 28, 1992, with the issuance of maintenance department instruction (MDI)-74, " Lube Oil Analysis Program." Instruction MDI-74 contains sample frequencies, methods to acquire samples, and evaluation criteria to include additional equipment into the program. Controls within the program include quarterly trend analysis reports, and requirements for storage of samples. The equipment monitored by the program consists of compressors, generators, power packs, and pumps with approximately thirty percent
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being safety-related.
CYAPCo also approved MDI-82, Evaluation of Lube Oil Analysis," on January 6,1993. This instruction provides a methods for evaluating the results of the lube oil analysis. A systematic
approach is used for evaluating equipment performance. This approach considers items such as contaminants expected, trends, sample verification, vendor input, and vibration data. Based on his review of the program, the inspector concluded that it included key elements for success; however, the program is still in the early stages of implementation. The inspector determined
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that the development of the lube oil analysis program was a good initiative by the licensee.
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I 4.0 ENGINEERING AND TECIINICAL SUPIORT (71707)
4.1 Plant Modification Turrmver The inspector reviewed the plant design change record (PDCR) turnover process and conditions necessary for release of the modification from the engineering department to the operations department. The inspection was a followup of CYAPCo's identification of an operations critical drawing (OCD) error. The issue was identified on January 5,1993, and documented in PIR 93-002.
The inspection consisted of interviews with operations and engineering personnel, review of PDCR 1148, " Modifications for Appendix J in Vater-to-Air Correlation Penetrations," review of applicable administrative control procedures (ACPs), and CYAPCo's past performance with drawing changes as a result of plant modifications.
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Background i
Administrative control procedure (ACP)
1.2-3.2., " Administration of PDCR Tumover, Preoperational Testing, and Release for Operation," section 6.11 states that generally at the time
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of release, the modification should be ready for unrestricted operation. Attachment 12.6 to the ACP, the " engineering release transmittal form," documents the modification status and provides a checklist of actions to facilitate engineering release. This includes the following general actions
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expected to be completed: update of technical specification databases, " redline" OCDs and submit drawing change request (DCR), perform pre-operations training, implement procedure and setpoint changes, complete authorized work orders, and, complete pre-operational testing. Upon
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completion, the engineering and operations managers review the release form, and the shift supervisor accepts the modification.
The plant engineer prepares the engineering release transmittal form. The engineer is responsible for verifying 'as-built' marked-up OCDs, and submits the DCR to the Generation Facilities records for revision to the assigned drawing. In addition to drawings, the as.Jgned engineer
determines the procedures requiring change as a result of the PDCR. This is accomplished by
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having the applicable departments identify the necessary procedure changes, and specifying when
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the change will be implemented. The departments are requested to specify when the procedure change will occur; prior to PDCR turnover to operations or within 60 days of turnover. The
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assigned engineer will prepare an internal control routing to the department to track procedural
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revision within 60 sixty days of PDCR turnover.
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Plant design change record no.1148 was implemented during the cycle 16 refueling outage. The j
PDCR modified thirteen containment penetrations by installing either manual gate or globe valves and associated drain valves in the main process lines. The modification was completed and turned over to Operations on February 4,1992. The modification required five OCDs to be updated, and twelve surveillance and normal operating procedures to be revised.
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The licensee identified and documented in PIR 93-002 that OCD 16103-26018 Sheet 8,
" Chemical Volume Control System-Operations Flow Diagram," and 161203-26008 Sheet 9,
" Component Cooling Water System-Operations Flow Diagram," did not include the manual main process valves installed per PDCR 1148. Further, the licensee determined that surveillance procedures SUR 5.1-159A, " Boron Injection Path Valve Lineup and Metering Pump Test (Modes 1 through 4)," and SUR 5.1-159B, " Boron Injection Path Valve Lineup and Metering Pump Test ( Modes 5 and 6)," did not include charging fill header valve (FW-V-351) position verification.
The above surveillances implement TS requirements 4.1.2.1.c. and 4.1.3.2. for the boration flow path.
Findines The inspector noted the licensee performed an in-depth self-assessment on the adequacy of technical specification surveillance procedures. As a result of this review, the licensee identified
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the discrepancy noted above and initiated the PIR.
The inspector interviewed the CYAPCo engineer responsible for PDCR-1148 and noted that, in retrospect, the engineer felt he had not completed a thorough review. The engineer had appropriately upgraded the OCD P& ids for the chemical volume control system and component cooling water system; however, he failed to consider and update the Operations Flow Diagrams.
Also, a contributing root cause to the problem was that the Operations department failed to
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identify the appropriate SURs requiring changes per PDCR 1148 to the CYAPCo engineer.
Although the appropriate normal operating procedures were revised; the operations department failed to identify SUR 5.1-159A and SUR 5.1-159B as requiring revision.
The inspector evaluated the operability issue associated with the failure to verify the position of all valves in the boration flow path. The technical specifications require that reactivity control
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be provided through an operable boration flow path to the reactor coolant system (RCS), and that
valves in the flow path be verified operable monthly. There are two charging discharge headers that provide a flow path to the RCS (normal charging header, and fill header). The operator has
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the option to choose either one to satisfy the TS requirement; both paths are routinely used
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depending on the plant operating mode and configuration. Valve FW-V-351 is located in the fill header and its required position is open. The fill header was credited as an operable flow path-i between Februry 6-10,1992. The charging header was credited the remainder of the operating
cycle. The licensee determined that between February 6-10 no information. exists to conclude that the normal charging was not available and operable. Further at PDCR turnover, SUR 5.7-
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54, " Inservice and Local Leak Rate Testing of Imop Fill Check Valve FH-CV-296 Penetration
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P-69" was performed on January 29,1992. The surveillance procedure restoration step requires valve FH-V-351 to be open. No other procedure existed at the time to potentially re-positioned '
valve FH-V351. Based upon the above, the inspector determined that the licensee had not j
violated the requirements for an operable boration flowpath.
The inspector evaluated CYAPCo's past performance to control OCDs associated with recent plant modifications. The scope of the review included previous NRC inspection reports and 1991 and 1992 CYAPCo PIRs. NRC inspection reports 50-213/92-13,50-213/92-15, and 50-213/92-
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24 document review of eight PDCR's. The reviews primarily consisted of evaluation of design inputs, safety evaluations, and tests conducted. The inspector did not identify any unsatisfactory conditions.
Seven PIRs written between 1991-1992 concerned licensee identified discrepancies with operations critical drawings, as follows: OCDs were red-lined to denote changes resulting from modifications, but the revised drawings were not issued within the time specified by the administrative procedures; drawings not identified as operations critical on the OCD list were found in the OCD files; and, the as-built OCDs did not accurately reflect the actual piping system - these errors were not traccable to modification activity. The PIRs were 91-015,91-123,91-302,91-304,92-093,92-117 and 92-126. Licensee corrective actions included resolving the drawing and plant configuration issues, and adding guidance in ACP 1.2-3.2, Attachment 12.4,
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to ensure timely submittal of drawings for revision. The inspector concluded that the corrective actions were appropriate to correct the deficiencies :dentified in the PIRs. The inspector identified no direct correlation between modification activities and failure to update OCDs.
At the close of the inspection period, the licensee locked-open valve FW-V-351, submitted changes to the OCD, and revised procedures SUR 5.1-159A and SUR 5.1-159B. The inspector considers the licensee identified failure to update applicable operations critical drawing and surveillances for the PDCR 1148 modification a violation of 10 CFR 50 Appendix B, Criterion
III, " Design Process." Enforcement discretion will be exercised because the licensee's efforts
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in identifying and correcting the violation meet the criteria specified in Section VII.B of the Enforcement Policy.
i Conclusion i
The inspector noted that as a result of an in-depth self-assessment, the licensee identified a failure
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to update an OCD, and a failure to complete a technical specification surveillance following a pisnt modification. The inspector did not identify any direct correlation between the modification process and failures to update OCDs. The inspector also noted that the licensee satisfactorily resolved the deficiency.
5.0 SAFETY ASSESSMENT AND QUALITY VERIFICATION (40500,71707,90712, and 92701)
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5.1 Evaluation of Licensee Self-Assessment Process The inspector evaluated a portion of the licensee's self-assessment program to determine whether the programs contribute to the prevention of problems. The programs include monitoring and evaluating plant performance, providing assessments and findings, and communicating and following up on corrective action recommendations.
The inspection included Nuclear Review Board (NRB) and quality assurance audits, past NRC evaluation of licensee self-assessment audits, and observations and effectiveness of the two oversight committees (Plant Operations Review Committee and the Nuclear Review Board). The
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inspector reviewed written material, inteniewed personnel, and verified selected corrective actions.
NRB and Quality Services Department Audits The inspector randomly selected and reviewed ten audits between 1991 and 1992. The audits reviewed are listed in Attachment 1. The inspector concluded that sixty percent of the audits resulted in CYAPCo plant staff development of corrective actions. The corrective actions
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typically included: procedural revisions, reinforcement to workers of the existing program requirements, Final Safety Analysis Report (FSAR) changes, and initiation of non-conformance reports (NCRs).
The inspector noted appropriate action by line management to initiate a programmatic review of component replacement schedules (CRS) for Limitorque motor-operated valves.
The programmatic review was in response to audit finding in report A30195, " Environmental
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Qualification Program At Connecticut Yankee." A majority of the procedural revisions were a result of audit reports A25054 and A25064, " Connecticut YankeeTechnical Specifications." The audits identified errors concerning FSAR descriptions, and incorporation of technical specification requirements. An example was a deficiency in the FSAR listing service water valves that close on either a safety injection signal or a loss of normal power signal. The technical specification surveillance procedure accurately and completely tested the isolation valves. One audit finding in A25064 resulted in a condition prohibited by technical specifica-tions and was reported pursuant to 10 CFR 50.73 (LER 92-013-00). The finding was that a technical specification surveillance procedure was performed without the correct revision. The surveillance procedure was to verify a core reactivity balance within cycle specific predicted
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values.
Approximately twenty percent of the findings resulted in no corrective actions based on re-evaluation or clarification of the issue by CYAPCo personnel. The inspector concluded the licensee actions were appropriate. The inspector identified that auditors verified corrective actions prior to audit closure, and CYAPCo proposed corrective actions were timely.
The inspector noted that the NRB reviewed all unresolved items and findings of the audits.' The
NRB generally accepted line managements corrective actions and actions to prevent recurrence.
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The NRB's audit of staff performance reviewed all of the 1991 QSD reports, and the 1991.
Combined Utility Assessment Report. The oversight committee concluded that additional line management attention was necessary in the work process, setpoint change program, and'
regulatory commitment tracking programs. The conclusions were carried forward and tracked for consideration in future NRB activities.
The inspector concluded that CYAPCo corrective actions to audits were generally responsive to
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the issues. However, documentation of the resolution and basis appeared insufficient for the -
l following issues: cleanliness controls for safety-related work orders; basis to adjust QSD
surveillances on equipment storage areas; reportability determination of Limitorque motor-operated valves; and root cause determination for a technical specification surveillance ~
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deficiency. Based on discussions with cognizant individuals, the inspector determined that appropriate actions had taken place, notwithstanding the lack of formal documentation.
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Previous NRC Inspections of Audits i
The inspector reviewed the 1991 - 1992 NRC assessments of licensee audits. Previously identified issues were the failure to perform a Process Control Program audit pursuant to technical specification 6.5.2.7.h., (report 50-213/92-16) and insufficient depth and independence of an annual emergency planning audit pursuant to 10 CFR 50.54(t) (report 50-213/92-23).
Audits reviewed in the security area, radiological environmental monitoring, fire protection, and effluent controls program were generally very good in depth and scope. Appropriate and timely responses to audit findings were apparent.
Oversight Committees Based on periodic observations of PORC meetings, both past and during this report period, the inspector concluded the committee conducted critical and thorough evaluations.' Examples of thorough evaluations included: quality assurance control questions of the reactor fuel vendor; reasons for fluctuations in failed fuel pin estimates; and, re-start from two 1992 reactor trips.
Based on attendance at two NRB meetings in 1992 and a review of meeting minutes, the inspector also concluded that the oversight committee performed its function well. Committee frequency and topics exceeded the requirements of technical specifications 6.5.2.4. and 6.5.2.6.
The meeting minutes were thorough and complete. A recent initiative added to the NRB agenda included plant issues or issues worthy of discussion as determined by the members. Effectiveness of this initiative has yet to be evaluated. The inspector concluded that the NRB continues to effectively track items carried forward (ICF) for resolution.
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Conclusion
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The inspector found that the self-assessment audits resulted in required corrective actions that contributed to the prevention of problems. The NRB individually, and programmatically
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evaluated audit results for completeness and trends in performance. CYAPCo corrective actions
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i to audits were timely. Some documentation of corrective action basis was insufficient, however, appropriate action had been taken based on follow-up interviews with cognizant personnel.
Overall, the committees were effective in fheir oversight functions.
5.2 Review of Recent Licensee Event Reports (LERs)
Periodic and Licensee Event Reports (LERs) were reviewed for clarity, validity, accuracy.of the
root cause and safety significance description, and adequacy of corrective action. The inspectors determined whether further information was required. The inspectors also verified that the reporting requirements of 10 CFR 50.73 and Technical Specification 6.9 had been met. The following reports were reviewed:
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LER 92-18-01: Invalidated Manufacturer Testing Renders Fire Wrap Inoperable The licensee issued this supplemental report dated December 31,1992, to retract LER 92-18.
The action was taken based on the additional review of Generic Letter 91-18 which showed that i
the initial operability assessment was overly conservative and in error, in that the subject fire barriers are best considered as degraded, but operable. The licensee's present engineering evaluation shows that, while the Appendix R requirements for a one hour fire barrier would not
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be satisfied, the existing fire barriers that use Thermo-bg 330 insulation are capable of performing the intended function of protecting the enclosed cables from postulated fire scenarios.
At Haddam Neck, the duration of the postulated exposure fires in areas protected by Thermo-Lag 330 is much less than one hour. Thus, the Thermo-bg installations are a qualification, rather than an operability issue. The inspector independently reviewed the CYAPCo conclusion and noted it was consistent with NRC guidance on reportability and operability matters.
CYAPCo actions on the barriers containing Thermo-Lag material have been reviewed in previous inspection reports (50-213/92-12, 92-15 and 92-18). The licensee is seeking relief from the Appendix R requirements, and evaluations continue to identify the need for additional corrective actions. Resolution of this issue is tracked by inspection item 92-15-01.
LER 92-22: Failure to Document Surveillance This reportable event was previously documented in resident inspection report 50-213/92-21.
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The issue was conservatively reported by the licensee as a failure to check the status of the low pressure safety injection core deluge valves monthly, when in fact the valve positions were routinely verified to be in the correct position daily. However, the daily actions were not documented and thus the checks could not be credited to satisfy the technical specification requirements. The LER was accurate and complete in the description of the event, and the safety evaluation.
The licensee correctly identified the cause of the event as an inadequate surveillance procedure (SUR 5.1-4) that was created in 1989 when the licensee failed to incorporate the requirements of Amendment 121 into the surveillance test procedure. The short ar.d long term corrective actions, inspected in report 50-213/92-21 and formalized in LER 92-22, were acceptable. The licensee initiated independent reviews to identify any other potential inadequacies in the operations surveillance procedures.
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The inspector evaluated whether the event in LER 92-22 should have been prevented by corrective actions for previous LERs and concluded that this was not the case. LERs 91-07,91-09 and 92-06 also involved surveillance procedure deficiencies and appear to be related events.
However, the inspector determined that LERs 92-06,91-07 and 92-22 are not causally linked.
Additionally, although LER 91-09 also resulted from the licensee's conversion to standard technical specifications, the corrective actions for that event consisted of a line item verification
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that the present technical specification testing requirements were completely incorporated in the -
operations surveillance test program. The discrepancy in LER 92-22 would not reasonably be identified in a line item review since the core deluge valves are not specifically listed in the
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specifications, but are two of many valves in the emergency core cooling system flow path. The corrective action for the present event will provide a much more thorough verification of surveillance procedure accurary down to a level detail not demanded by the event in LER 91-09.
The inspector had no further questions at this time concerning the event or the licensee's corrective actions.
5.3 Review of LERs for the SALP Cycle The purpose of this inspection was to review the Licensee Event Reports (LERs) submitted over the current Systematic Assessment of License Performance (SALP) cycle (July 14,1991 to -
January 9,1993). The inspector reviewed licensee records and NRC inspection reports, and interviewed licensee personnel. Records rev'ewed included: LERs 91-13 through 91-30 and 92-01 through 92-22; over 50 Plant Information Reports (PIRs) and Controlled Ror.iags (CRs)
related to the LERs; and, administrative and surveille.nce procedures, operator logs, and engineering reports. Specific areas inspected were: LER event descriptions; root causes; corrective actions; PIR and LER tracking and trending; and, LER trends over the current SALP
cycle.
5.3.1 LER Trends Analysis The inspector performed a trends analysis of the LERs, which is summarized below. The inspector noted the following findings:
Noteworthy Improvements The inspector noted a significant decrease in the overall number of reports under 10 CFR 50.73 since the last SALP cycle. The total number of reports decreased from fifty to thirty nine, with five planned maintenance activities conservatively reported as events.
The five planned maintenance and test activitbs resulting in temporary shutdown of spent fuel pool cooling
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typified an overly conservative interpretation of reporting criteria 10 CFR 50.73.(a)(2)(v); the licensee has stopped issuing reports involving this activity. Overall, the number of reports issued during the present SALP cycle is largely driven by a conservative approach to reporting and the licensee's identification of design deficiencies, which reflects a strong safety ethic and the
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effectiveness of the programs to identify, evaluate and report potential safety problems.
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The decline in reportable events was direct attributable to the improved performance by the plant staff, evidenced by the decrease in the number of personnel errors from 20 to 8. Significant improvements were ned in two areas: only one event involving reactor operations was due to a licensed operator error, and the number of events due to inadequate procedures dropped from eleven to three. These significant improvements were due in part to increased management ettention to these areas, enhanced training programs, and procedure upgrades.
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Missed Technical Specification Surveillances i
The inspector reviewed the reporting of missed surveillances at the Haddam Neck facility since 1989. This issue was assessed in the last SALP as an area requiring improved management attention. An improved trend was noted in the significant decline in missed surveillances in 1992 over previous years.
The inspector conducted a detailed review of the missed technical specification surveillances and various surveillance tracking systems due to their importance in the continued safe operation of the plant.
The inspector noted that a 1989 Human Performance Evaluation Periodic Report discussed missed surveillances. This report commented that "... it appears probable that with the assorted ways of tracking surveillances, the potential for future missed / late surveillances will remain..
." The licensee's HPES Coordinator issued Human Performance Evaluation (HPES) Report C92-001, dated June 3,1992. In this report, he concluded that "... Missed surveillances are not a new problem."
The inspector also noted that CYAPCo Nuclear Safety Engineering published a " Human Factors Study of Missed Surveillances At CY, Final Report" dated December 17, 1992. This report documents 5 missed surveillances in 1989, 7 in 1990, and 2 in the first half of 1991. There has been one additional missed surveillance in 1992. The report cited the revision to CY's technical specifications (effective 1988), which added new surveillance requirements, as a contributing factor. The report also states in part: "The most common corrective actions for missed surveillance events was to modify procedures, discipline personnel, improve training, or some
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combination of the above. These measures have had indifferent success in preventing additional events."
The inspector reviewed the current methodology of scheduling and tracking surveillance tests by the Maintenance and Instrumentation and Control (I&C) planning groups. The surveillance tracking process was quite different for each group. The inspector determined through interviews.
with planning personnel that there are eight groups at Haddam Neck which track and perform surveillances, each with their own unique surveillance tracking system. The inspector concluded there is a high degree of confidence that with current planning personnel and management emphasis, the two surveillance tracking programs reviewed will be reasonably successful in -
preventing future missed surveillances. However, there appears to be little assurance that a
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surveillance will not be missed with new or inexperienced planning personnel.
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surveillance tracking methods are not documented to the extent that inexperienced planners can ensure that surveillances are not missed. Current methods rely on the abilities of the individuals running the system.
The inspector also noted that the second verifications of completed surveillances appear to be.
only partially effective in preventing a missed surveillance. Weekly supervisory checks of the surveillances scheduled / completed during the previous week may not find a missed surveillance
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until after the event becomes reportable, if the periodicity of the surveillance is one month or less. Additionally, the inspector noted a weakness in tracking the end of the surveillance grace
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period (required due date) One of the two tracking methods reviewed by the inspector did not
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have these required due dates calculated, and both did not have the dates readily available on j
their respective surveillance tracking status boards for supervisory review. These status boards are used by planners and supervisors to check if a surveillance has been completed or is overdue.
The inspector discussed these findings with the licensee. Following a missed surveillance on February 13,1992 (LER 92-05), the licensee began an evaluation of current surveillance tracking methods. The evaluation is tracked by the licensee under CYAPCo Controlled Routmg 92-0545, due March 31,1993. The licensee plans to take appropriate actions based on the results of this evaluation. The inspector concluded that management attention to tnis area since 1991 has resulted in a reduction in the number of missed surveillances.
Inspection Item (UNR) 92-08-01 in May 1992, tracks the licensee's corrective action te prevent the recurrence of missed surveillances.
NRC Inspection Report 92-14 providen an update to UNR 92-08-01, documenting a review of interim measures to prevent recurrence of missed surveillances. The licensee's corrective actions appear to have a positive effect c1 preventing missed surveillances. This item will remain open pending completion oflong tern corrective actions in March 1993, and subsequent review by the NRC.
Fouling of CAR Fan Service Water Filters The inspector noted repeated fouling of the Containment Air Recirculation (CAR) fan service water " Adams" filters, due to silt in the Connecticut River. Two service water filters (1 in service,1 in standby) remove particulates greater than 0.005" from the cooling water supplied to the CAR fan cooling coils and motor coolers. A non safety-grade backwash system runs continuously to ovnd the time it takes the filters to clog. Related LERs include 90-01,90-23, 90-32,92-12 and 52-14. These events resulted in either assumed or actual inoperability of two or more of the CAR fans. CAR fans are safety re!ated fans which are the primary means of removing heat to reduce containment pressure after a design basis LOCA. The inspector conducted a detailed review of these events due to the importance of the CAR fans during a design basis accident.
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On February 2,1990, the licensee completed an engineering evaluation which determined that the ciogging or breaking of the CAR fan service water filters would render the CAR fans
inoperable following a design basis LOCA. As a corrective action, motor operated valves
(remotely operated by a manual push-button) were installed, allowing the filters to be bypassed during a DBA. The inspector reviewed PDCR 991, " Service Water Adams Filter Bypass Line Mechanical System Engineering Safety Evaluation," and determined that the engineering judgement supporting the modification was consistent with the information known at that time
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(i.e., past satisfactory performance of CY heat exchangers supplied by unfiltered service water,
and past operator experience with silt / debris levels in the Connecticut River).
On October 27,1990, the Adams filters were clogged by heavy silt and debris from cleaning activities on the intake structure trash racks. Operators opened the bypass MOV to allow continued flow to the CAR fans. Two of the four CAR fan cooling coils became fouled with
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debris, causing low service water flow through the coils. The licensee shut down the plant to
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clean the cooling coils (LER 90-23). This event was reviewed in NRC Inspection' Report 92-10.
The safety evaluation erroneously concluded that non-filtered service water to the CAR fans was acceptable (i.e., by opening the bypass MOVs) based on operating experience with non-safety grade main turbine heat exchangers.
The inspector considers this error understandable considering the information available to the licensee at the time.
The licensee further identified that the operating requirements for the bypass MOVs (specified in the PDCR 991 safety evaluation) were not incorporated into the operating procedures.Section IV.A of the safety evaluation titled " Normal Power Operation" states in part "...The bypass line will be closed during normal operation and will not be permitted to be opened except for valve testing purposes..." During this event, the bypass MOVs were opened during normal operations, allowing the CAR fan cooling coils to be fouled. Although the licensee provided instruction in the EOP on when to use the bypass valve, he failed to provide operating restrictions on use of the valves on the normal operating procedures. Following the October 1990 event, the licensee incorporated PDCR 991 design operating ~ restrictions for the filter bypass MOVs into Operations Department Instructions, and documented the event and corrective actions in LER 90-23.
On December 27,1990 and again on April 15, 1992, heavy silt from the Connecticut River caused the Adams filters to clog, causing the inoperability of the CAR fans. These events were documented in LERs 90-32 and 92-12. Corrective actions included additional guidance for operators to improve plant response te high river silt conditions, and initiating an engineering evaluation to investigate potential design changes for the Adams filters and bypass MOVs.
On June 8,1992, a monthly routine service water flow test conducted by the licensee determined
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that the maximum allowed differential pressure (dp) across the Adams filters was achieved in approximately two minutes due to river conditions at that time. On June 10, the licensee completed an evaluation of the filtet design basis, and concluded that post-DBA, a minimum of 30 minutes was required for the filters to reach the maximum allowed dp. The licensee notified the NRC of the questionable operability of the filters between June 8 and 10, documenting the event in LER 92-14. Corrective actions included stationing an operator at the filter bypass MOV
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manual switch when filter dp reached the maximum allowed value, frequent testing of the filter fouling rates, and a design change to automatically open the filter bypass MOVs on a valid safety injection or a high containment pressure signal. The design change was completed under PDCR 1294 in November 1992, eliminating all required Adams filter fouling rate testing, and post-DBA operator actions for filter / CAR fan operability.
The inspector concluded that the corrective actions to prevent the inoperability of the CAR fans following each of the events were adequate. Although they shared a similar root cause, (clogging of the Adams filters due to high silt levels in the Connecticut River), the inspector considers that the licensee has diligently worked through a complex engineering issue, with the best information available at the time, achieving adequate results.
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Documentation of Recurrent Events Administrative Control Procedure (ACP) 1.2-16.1, " Plant Information Reports," step 6.9.4.c.1 states in part:
... Review previous events; the review should be for a minimum of the
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previous two calendar years.. Additionally, all LERs shall be reviewed for prior reportable PIRs/LERs; describe the reason for the similarity and discuss the root causes and action to prevent recurrence..."
The inspector noted that PIRs90-253, 90-309,91-282, related to missed technical specification surveillances, and PIRs90-276, 90-332,92-105, 92-123, related to the inoperability of the service water Adams filters. The later PIRs did not list the required PIRs with similar root:
causes, did not describe the reasons for their similarities, and did not discuss the root causes and previous actions to prevent recurrence.
The inspector discussed this finding with the licensee. The licensee acknowledged the deficiency; however, based on a evaluation of the actions taken for the specific technical issues, the licensee concluded that the intent of the referenced ACP was met, though not documented, in that past events and corrective actions were considered by the responsible engineer as problems recurred.
The licensee agreed to address this deficiency.
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This item is unresolved pending further NRC review of licensee corrective actions to better document his reviews of recurrent PIR events in accordance with the requirements of ACP 1.2-16.1 (Unresolved item 50-213/92-26-01).
5.3.2 Conclusions The inspector concluded that the licensee's LER event descriptions, documented root causes, corrective actions, and PIR/LER tracking, trending and closeout were good, with the above noted exceptions. The inspector found that the licensee's invest:gation and resolution of reportable events is largely driven by a conservative approach to reporting and the licensee's identification of design deficiencies. Overall, the reportable issues identified and resolved during the present SALP cycle, contributed to the continued safe ope ation of the plant, as well as protecting the health and safety of the public, and reflects on the licensee's strong safety ethic and the effectiveness of the programs to identify, evaluate and report potential safety problems. Licensee documentation of actions to investigate prior events and to evaluate root causes could be improved.
5.4 Reactor Coolant Pump Seal Cooling Following the 1979 accident at the Three Mile Island (TMI) Nuclear Plant, the NRC reviewed the potential for failure of reactor coolant pump (RCP) seals following a loss of off-site power.
This led to the establishment of the TMI Task Action Plan (TMI-TAP) item II.K.3.25 (in NUREG 0737). This item required licensees to evaluate the integrity of the RCP seals for a two-hour period following a loss of off-site powe _.
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The purpose of this inspection was to evaluate the acceptability of licensee actions regarding resolution of TMI-TAP item II.K.3.25. Previous NRC inspection of this item was documented
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-in resident inspection report 50-213/90-20 dated February 7,1991.
On June 26,1991, the NRC issued a safety evaluation report for NUREG-0737 item H.K.3.25
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which concluded that the licensee proposed resolution of operator action to reinstate RCP sealing
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cooling following a loss of off-site power complied with the requirements. The licensee stated that plant procedures ensure that RCP seal cooling would be reinstated within twenty minutes after a partial or complete loss of off-site power. The inspector reviewed emergency operating
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procedures (EOPs) 3.1-10, " Partial Loss of AC," and ES-0.1, " Reactor Trip Response," to
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verify adequate operator guidance exists for reinstating seal cooling with a partial or complete loss of off-site power. The operator actions are to check that RCP bearing temperatures and to verify status of the charging and component cooling water pumps on the emergency bus. The
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instructions for exactly how to reinstate seal cooling depends on whether the RCP temperature
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is above or below 170 degrees Fahrenheit. Based on a review of the EOPs and discussions with operators, the inspector determined that the seal cooling would be re-established within the stated time frames, and thus seal integrity would be assured. The inspector considers TMI-TAP II.K.3.25 closed.
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As documented in NRC Generic Letter 91-07, the NRC staff stated that remlution of Generic Issue 23, " Reactor Coolant Pump Seal Failures and Its Possible Effect or c;tation Blackout",
would be considered in future reviews of Haddam Neck. On August 4,1992, CYAPCo submitted an update report to summarize the status of its actions under the Integrated Safety
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Assessment Program (ISAP). ISAP Topic 1.18, " Reactor Coolant Pump Seal Modifications,"
remains open pending resolution of Generic Issue 23.
5.5 Operational Experience Feedback Program l
The inspector reviewed CYAPCo's actions to address industry events as described in NRC Information Notices (ins). This inspection was a continuation of the review begun in NRC Inspection 50-213/92-20 to assess the effectiveness of the operating experience feedback l
program.
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The inspector randomly selected NRC ins issued from 1988-1992 for review, as listed m Attachment I to this report. CYAPCo's review and action for each issue was evaluated to determine whether appropriate follow-up action was taken. The effectiveness of the action was further evaluated by reviewing plant information reports written after the IN was issued to
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determine whether any plant events correlated with the IN topics.
i The inspector determined that the operational feedback program was acceptable and effective for
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the period reviewed. All ins were tracked in the licensee's system for processing industry.
i events. The inspector found engineering evaluations and reviews adequately addressed those l
issues applicable to Haddam Neck. The inspector also noted the issues reviewed received timely closure. The inspector independently verified plant design features and programs credited by
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CYAPCo to ensure they adequately addressed the IN issue. The inspector did not identify any
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5.6 Follow-up of Previous Inspection Findings The inspector reviewed licensee actions taken in response to an open item from a previous
. inspection. The inspector determined if corrective actions were appropriate, and whether previous concems were resolved. The following item was reviewed:
5.6.1 (Closed) IFI 50-213/92-20-03: Adequacy of the Operating Experience Review Program
L The' inspector previously noted the CYAPCo program for reviewing industry experience for applicability to Haddam Neck was sound. However, the inspector questioned the effectiveness of the program when three events occurred at Haddam Neck which were the subject of previous Information Notices. This matter was reviewed further in this inspection by reviewing CYAPCo actions for ins issued in the 1990-1992 time period (see Section 5.5 above). The inspector found the engineering evaluations and reviews adequately addressed those issues applicable to Haddam Neck. The inspector determined the operational feedback program was acceptable and effective for the period reviewed. The inspector concluded the three previous inadequate responses appear to be isolated events. The inspector had no further questions and considers 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 aeeting was held to summarize the conclusions of the inspection. No written material was given to the licensee and no proprietary information related to fais inspection was identified. The exit -
meeting for this inspection was held on January 22,1993.
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.,.e-P ATTACHMENT 1
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DOCUMENTS REVIEWED
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INFORMATION NOTICES IN 88-83, Inadequate Testing of Relay Contacts In Safety-Related Logic Systems, dated 10/19/88 IN 90-18, Potential Problems With Crosby Safety Relief Valves Used On Diesel Generator Air Start Receiver Tanks, dated 3/9/90 IN 90-42, Failure of Electrical Power Equipment Due to Solar Magnetic Disturbances, dated 6/19/90
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e IN 90-68, Stress Corrosion Cracking of Reactor Coolant Pump Bolts, dated 10/30/90 IN 90-69, Adequacy Of Emergency And Essential Lighting, dated 10/31/90 IN 91-36, Nuclear Plant Staff Working Hours, dated 6/10/91
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IN 91-70, Improper Installation of Instrumentation Modules, dated 11/4/91
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IN 91-78, Status Indication of Control Power For Circuit Breakers Used In Safety-Related Applications, dated 11/28/91 IN 91-85, Potential Failures of Thermostatic Control Valves for Diesel Generator Jacket Cooling
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Water, dated 12/26/91
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IN 92-18, Potential For Imss of Remote Shutdown Capability During A Control Room Fire,
dated 2/28/92 IN 92-44, Problems With Westinghouse DS-206 And DSL-206 Type Circuit Breakers, dated 6/18/9 IN 92-48, Failure Of Exide Batteries, dated 7/2/92 IN 92-49, Recent Loss or Severe Degradation of Service Water Systems, dated 7/2/92 NRB AND QUALITY ASSURANCE AUDITS
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1991 Connecticut Yankee Facility Performance Evaluation
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A25073 - Unit Staff Training A25065 - CY Personnel Qualifications - 1992 A30194 - Training Program Modifications
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A60281 - Cleaning, Storage, Shipping, - CY
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A30191 - Department Training - CY A30195 - EQ Program at CY A25054 - CY Technical Specifications
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A25064 - CY Technical Specifications '
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A25060 - 1991 Fire Protection /IAss Prevention