IR 05000213/1993006

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Insp Rept 50-213/93-06 on 930328-0508.No Violations Noted. Major Areas Inspected:Plant Operations,Radiological Controls,Maint,Surveillance,Ler & Periodic Repts
ML20044H270
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 06/03/1993
From: Doerflein L
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20044H266 List:
References
50-213-93-06, 50-213-93-6, NUDOCS 9306080127
Download: ML20044H270 (20)


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U.S. NUCLEAR REGULATORY COMMISSION

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REGION I

e Report No.

50 213/93-06 License No.

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Licensee:

Connecticut Yankee Atomic Power Company P. O. Box 270

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Hartford, CT 06141-0270 Facility:

Haddam Neck Plant location:

Haddam Neck, Connecticut Dates:

March 28,1993 to May 8,1993 i

Inspectors:

William J. Raymond, Senior Resident Inspector.

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Peter J. Habighorst, Resident Inspector Approved by:

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'6393 Lawrence T. Doerflein, Chief

' Date Reactor Projects Section No. 4A,hDRP

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Areas Inspected: NRC resident inspection of plant operations, radiological controls, maintenance, surveillance, licensee event reports, and periodic reports.

Re.sults: See Executive Summary

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EXECUTIVE SUMMARY

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HADDAM NECK PLANT INSPECTION 50-213/93-06 Plant Operations i

The inspector concluded that licensee personnel displayed good sensitivity and took l

appropriate response for abnormal Connecticut River conditions (i.e., high level and heavy debris). Licensee actions to protect plant equipment were timely and effective.

The inspector found that the licensee implemented appropriate corrective actions to preclude

the recurrence of the outboard containment hatch equalization valve failure. The valve failure resulted in a momentary loss of containment integrity during power operations.

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t Maintenance and Surveillance

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The inspector determined licensee personnel adhered to procedures and were knowledgeable of the required actions during the maintenance and surveillance activities observed. The

licensee appropriately perfonned an operability surveillance for the containment air i

recirculation coolers during periods of adverse service water debris caused by high river

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levels.

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Engineering and Technical Support The licensee did not rcquest a technical specification change on the augmented inservice

inspection program for a 1991 plant design change which added four pipe-to-pipe welds on the main steam system. The inspector found this was based on a commitment to the NRC (i.e., augmented erosion / corrosion program) which superseded the need for the license

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change. The inspector found that the licensee met the technical specification inservice inspection requirements since all the required inspection points were inspected within the required time interval with acceptible results.

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Safety Assessment and Ouality Verification The licensee identified a failure to implement a past Licensee Event Report corrective action l

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commitment. The NRC considers this issue unresolved, and will evaluate the effectiveness of the proposed licensee actions to improve the tracking and implementation of regulatory commitments.

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l SUMMARY OF FACILITY ACTIVITIES The unit operated at full rated power throughout the inspection period. The licensee continued prepamtions for the refueling outage. This included receipt inspection of the new reactor fuel.

2.0 PLANT OPERATIONS (71707)

The inspectors routinely reviewed plant operations during normal utility working hours, and portions of backshifts (evening shifts) and deep backshifts (weekend and night shifts).

Inspection coverage was provided for fourteen hours during backshifts and fifteen 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. Centrol 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

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distribution verifications. Normal Operating Procedure (NOP) 2.0-1, " Operations Department Shift Staffing Requirements," identifies the minimum staffing requirements.

During this period, the inspector noted that control room staffing during power operations

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met these requirements.

The inspectors reviewed the on-site electrical distribution system to verify proper electrical i

line-up of the emergency core cooling pumps and valves, the emergency diesel generators,

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radiation monitors, and various engineered safety feature equipment. The inspectors also verified valve lineups, position oflocked manual valves, power supplies, and flow paths for i

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

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generators. The inspector noted no deficiencies.

Log-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 determined they were satisfactory, with the shift supervisor controlling the turnover. All members of the crew discussed plant conditions and evolutions in progress. The information exchanged was accurate. The inspectors also reviewed control room trouble reports 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 item..

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At daily planning meetings, the inspector noted discussions on maintenance and surveillance activities in progress and planned work authorizations. The inspectors conducted periodic plant tours in the primary auxiliary building, turbine building, and intake structures. Plant housekeeping was satisfactory.

2.2 Radiological Controls During routine inspections of the accessible plant areas, the inspectors observed the -

implementation 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 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. The inspector 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 inspector noted that health physics technician control and monitoring of these activities were good.

2.3 Licensee Actions During Flood Conditions The inspector reviewed licensee actions during the period of April 1-5 as heavy rains and warm temperatures caused rivers to experience flood conditions in southern New England.

The Connecticut, Housatonic and Naugatuck Rivers reached flood stages in south-central Connecticut. Abnormal Operating Procedures (AOP) 3.2-24, " Flooding of the Connecticut River," and AOP 3.2-5, " Natural Disasters," describe licensee plans to protect plant equipment.

The site grade is twenty one feet above mean sea level (MSL). The entry conditions listed in AOP 3.2-24 require plant operators to initiate progressive protective actions if river level is observed at eleven feet MSL, or iflevels are forecasted to reach twelve feet above MSL.

The actions include enhanced monitoring and trending of river conditions (level and debris),

plant shutdown, and the installation of flood protection barriers at the entry points to buildings housing safety systems. If river levels reach fifteen feet above MSL, an unusual event emergency class would be declared according to the emergency plan.

The licensee's preparations for the high water conditions included enhanced monitoring of conditions at the site and closely following forecasted conditions at the upstream locations in Middletown and Hartford. Due to changes in Connecticut River width, river levels at Haddam were less than the levels e~xperienced at Middletown and Hartford. The licensee received information on projected river conditions from Connecticut Valley Exchange (CONVEX) and the national weather service. Although river conditions were projected to exceed twelve feet during the April 1-3 period, the level at Haddam Neck remained below nine fee.*

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i Operation's personnel reviewed AOP 3.2-24 to note the required actions. To supplement the i

forecasted information on river conditions, operators monitored the river using level

instrumentation and by direct observation at the intake structure. Increased amounts of debris

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and silt that accompanied the high water conditions caused operators and maintenance personnel to clean service water filters and strainers, and the intake trash racks at an l

increased frequency.

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The service water Adams filters clogged with silt and required continuous cleaning during the period of March 31 to April 3. Operators were very sensitive to the service water and filter conditions, and performed supplemental testing and evaluations to assess the operability of the i

containment fan coolers. See Section 3.2 of this report for further information on this topic.

Plant systems potentially impacted by the river conditions remained operable throughout the period.

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The inspector identified no inadequacies in the licensce's actions to protect plant equipment from flood conditions. Plant personnel demonstrated a good sensitivity and response to adverse river conditions. Licensee contingency planning and actions were timely and

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effective to assure plant safety.

2.4 Loss of Containment Integrity On April 27, at approximately 9:30 a.m., the licensee identified a momentary loss of

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containment integrity. The loss of containment integrity was through two containment isolation valves. The open containment isolation valves were the inboard and outboard containment hatch equalization valves. The valves are two inch rapid-action globe valves, which open to due a mechanical arm exerting axial force on the valve stem, and close with

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spring force. The licensee estimated that both valves were opened for approximately twenty seconds.

The licensee was performing a containment entry, when the designated hatch operator recognized a problem. The hatch operator noted that the inner personnel containment hatch would not open, and terminated the entry. CYAPCo investigation found that pressure could not be equalized across the inner hatch, since the outboard hatch equalization valve (CN-V-2)

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Corrective Action CYAPCo actions in response to this event were to terminate the containment entry, shut CN-V-2, take actions according to technical specification action statements 3.6.1.1. and 3.6.1.3.b., initiate plant information report (PIR)93-056, and evaluate the condition for

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reportability (i.e.,10 CFR 50.72). The failure of valve CN-V-2 to shut was excessive packing force. The packing force caused valve stem resistance to be greater than the t

available spring force to close the valve.

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The inspector verified that valve operation was sensitive to packing adjustments. The

maintenance department's actions to prevent recurrence were to loosen the packing gland, and

inject lock-tight on the gland nut. Additionally, the licensee replaced the packing follower.

The replaced packing follower provided equal force on all surfaces of the packing rings.

The licensee had also disassembled the valve and identified valve stem galling. At the end of-i the inspection period, the licensee did not identify the cause of the valve stem galling. To i

address this issue, CYAPCo was considering placing a caution sign near to the valve to say i

that packing adjustments are prohibited on the valve.

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Following the valve maintenance, CYAPCo successfully tested the outboard equalization valve for operation and leak-tightness using surveillances SUR 5.1-62B, " Personnel Hatch Full Pressure Leak Test," and SUR 5.1-62A, " Personnel Hatch Reduced Pressure Leak Test." The inspector reviewed the results of the smveillance, and determined that the results

were within the acceptance criteria, and adequately demonstrated successful valve performance.

CYAPCo decided that the condition was not reportable to the NRC. The bases were that the

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condition did not violate action statements within the technical specifications, and the calculated unmonitored, unplanned release of radioactivity was significantly below the

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reporting threshold. The inspector confirmed the licensee's reponability conclusion using.

10 CFR 50.72 and emergency plan implementing procedure (EPIP) 1.5-1, " Emergency

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Assessment," Attachment 12.4.

The inspector reviewed past performance of the containment hatch outboard equalization

valve. In April 1992, CYAPCo initiated PIR 92-090. PIR 92-090 concluded that valve CN-

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V-2 failed to shut due to excessive packing force. The corrective action for PIR 92-090 was to identify in the computerized component identification system, that packing adjustments affect valve operation. The statement cautions workers on the affects of packing adjustments

on valve operation. The past corrective action was ineffective in precluding valve failure

since personnel made during field adjustments and packing adjustments without a work order.

The inspector found the recent corrective actions would preclude field adjustments affecting

valve operation.

The inspector concluded that the licensee adhered to technical specificailons, implemented -

appropriate actions to preclude recurrence, and that no report of the event was required.

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Corrective actions for the past failure of CN-V-2 were ineffective to preclude recurrence of l

this event.

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3.0 MAINTENANCE AND SURVEILLANCE (61726,62703)

3.1 Maintenance Observation The inspectors observed portions of the following maintenance activities for compliance with procedures, plant technical specifications, and applicable codes and standards. The inspectors also verified if appropriate quality services division (QSD) involvement, appropriate use of

safety tags, acceptable fire prevention controls, and appropriate personnel qualifications. The inspector also verified the correct return to service of safety equipment. Portions of activities reviewed included:

CY 93-3264, EG-2A Governor Speed Changer

CY 93-3659, EDG Ventilation Modifications

CY 93-4273, New Fuel Receipt Inspection

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CY 93-4274, Unload New Fuel From Shipping Containers

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EG-2A Governor Speed Changer The inspector reviewed licensee actions under authorized work order (AWO) 93-3264 to repair the governor on the 'A' emergency diesel generator (EG-2 A). During a routine inspection of equipment conditions, an auxiliary operator noted loose cap screws on the governor speed changer motor. The cap screws are mounted to the top plate of the governor -

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housing. The licensee declared EG-2A out of service at 4:46 p.m., on April 13, pending investigation and repair.

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The licensee found that four cap screws had become partially loose due to vibration over time. Mounting bolts that hold the motor from the underside of the housing were found to be

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secure. Thus, the speed changer motor was still firmly attached such that the governor was fully functional in the "as-found" condition. The licensee tightened the four loose cap screws and secured them in place with lock wire. The licensee inspected the mounting and governor installation on EG-2B, and found it to be acceptable. A PIR was written to document the discrepancy. The PIR has been assigned to plant engineering for review to find what additional means of securing the cap screws are warranted.

EG-2-A was returned to service at 9:30 p.m., on April 13, following a successful retest. The inspector reviewed the governors on both diesels and verified both were secure. No inadequacies were noted in the licensee's corrective actions. The inspector noted the auxiliary operator demonstrated good attention to detail by identifying this discrepancy.

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EDG Ventilation Modifications i

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The inspector observed work in progress under AWO 93-3659 to modify the ventilation supply for the emergency diesel generators in accordance with plant design change record (PDCR) 1339, "EDG Ventilation Modifications." The modifications were necessary to

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address deficiencies in the ventilation flow rate into the rooms, as described in NRC Inspection Reports 50-213/92-12 and 50-213/92-14. Under PDCR 1339, the licensee will -

remove the solid steel sheeting at the ventilation intake, and replace it with fixed vaned louvers to increase the effective flow area of the opening. The steel frame for the louvers will also be reinforced.

The licensee established controls for the modifications that would allow the work to proceed ~

without impacting the operability of the emergency diesels. The work controls included consideration for personnel and equipment protection, the control of fire hazards, and the control of work materials to prevent impacting the ventilation flowrate. The inspector reviewed work in progress on April 14 for conformance with the work package controls and requirements. The inspector also reviewed the completed and planned welding against that required by the weld data sheets in the AWO package. The inspector determined workers were knowledgeable of the construction plan and details, and the controls to assure personnel and equipment protection. No discrepancies were noted.

3.2 Surveillance Observation i

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:

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SUR 5.1-17A. Diesel Generator (EG-2A) Operability Test

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The inspector reviewed the completed results for the retest of EG-2A following maintenance.

The diesel starting time and loading were acceptable to demonstrate operability as required by the technical specifications. The in pector also determined that the engine was properly restored to standby service by conducting a walkdown of the skid, and by reviewing tagging clearance 93-0283 to verify that components affected by the tags were properly aligned to support diesel operations. No discrepancies were identified.

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SUR 5.2-124.2. Reactor Trio Logic Cabinet A2 Coincidence Test

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On April 26, the inspector observed instrument and control (I&C) specialists perform

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technical specification (TS) surveillance procedure SUR 5.2-124.2. The surveillance tested the coincidence reactor protection system logic. The I&C specialists manipulated individual input parameter trip / bypass switches and observed corresponding logic lights. The inspector verified the surveillance test results met the acceptance criteria. The inspector noted that the

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I&C specialists displayed good procedural adherence during the surveillance.

  • i SUR 5.1-155B. Diesel Fire Pump (P-5-1 A) Monthly Test

On April 29, the inspector observed CYAPCo auxiliary operators perform TS surveillance

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SUR 5.1-155B. The surveillance was to prove diesel fire pump operability following maintenance repairs to the failed governor solenoid. The licensee identified the failure of the solenoid on April 26, during a failed start of the diesel. The diesel fire pump was declared inoperable, and CYAPCo complied with technical specification action statement 3.7.6.1.

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During diesel operation on April 29, the auxiliary operators noted excessive fuel oil leakage from the fitting between the fuel solenoid and the governor. On April 30,1993, following repairs to the fitting, the diesel was subsequently restored to an operable status. The inspector observed good operator recognition of a problem during the surveillance. CYAPCo actions to restore the fire protection equipment were appropriate.

SUR 5.7-159. Main Steam and Feedwater High Energy Line Break (HELB)

Inspection Points On April 6,1993, the inspector observed an auxiliary operator implement TS surveillance SUR 5.7-159. The surveillance objective was to perform a leak inspection of selected portions of the main steam and main feedwater system as required by TS 4.0.6.a.2 and 4.0.6.b.3. The inspector confirmed that the auxiliary operator evaluated each inspection area. The surveillance results were satisfactory and the operator was knowledgeable of the surveillance objectives.

i The inspector compared steam and feedwater system pipe weld inspection points in TS Table 4.0-2 with SUR 5.7-159. The inspector found that the surveillance procedure had more inspection points than the TS table. Report detail 4.3 documents the specific details of

this issue.

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SUR 5.1-13. Auxiliary Feed Pumo P-32-1 A Functional Test

i On May 4, the inspector observed technical specification surveillance SUR 5.1-13A. The surveillance objective was to verify that the 'l A' auxiliary feed pump operates acceptably on recirculation. The test also verifies correct manual valve position in the auxiliary fecdwater system.

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The inspector verified the 'A' auxiliary feed pump technical specification surveillance criteria were met. The inspector observed proper communication and coordination between auxiliary operators, control room personnel, in-service test (IST) technicians and security guards. The

operators adhered to the procedure as written. During the surveillance, the inspector noted that control room opentors record main control board (MCB) pump discharge pressure (SUR 5.1-13A, step 6.2.6.b). The recorded MCB pressure was 950 psig. The inspector noted that local pressure indicator (PI-1326A) indicated approximately 1,000 psig. PI-1326A is the indication used by IST personnel during TS 4.0.5 surveillances. SUR 5.1-13A, step 6.2.6.b requires the operator to verify discharge pressure (at the MCB) is between 800 to 1000 psig with a caution statement not to exceed 1,000 psig. The inspector questioned CYAPCo on the reasons for the discharge pressure limit and the use of the MCB gauge. CYAPCo stated that pump discharge pressure is recorded to ensure that the pump is not at shut-off head. At the end of the inspection period, the system engineer was evaluating the need to clarify the intent of the procedure caution statement.

SUR 5.7-118. Inservice Testine of CAR Fan Service Water Supoly Header Check

Valves and CAR Fan Cooling Coils

On March 31, CYAPCo authorized throttling the normally open Adams filter outlet isolation valves SW-V-838A and SW-V-838B. The activity was controlled pursuant to administrative control procedure (ACP) 1.2-5.3, " Evaluation of Activities / Evolutions Not Controlled By Procedure." The throttling of valves SW-V-838A and SW-V-838B increased the differential pressure for backwashing the Adams filters and reduced the debris accumulation in the filter.

On April 5, during a plant walkdown, the inspector questioned the operability of the

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containment air recirculation (CAR) coolers based on the reduced service water flow and the frequent cleaning of the Adams filters. The inspectors questioned CAR cooler operability in light of TS requirement 4.6.2.a.2. which requires the heat removal from each CAR cooler to be equal to or greater than 26.5 X E6 British thermal units (BTUs) per hour. SUR 5.7-118

implements Surveillance Requirement 4.6.2.a.2.

The licensee performed SUR 5.7-118 and determined that the No. 2 CAR cooler was inoperable due to insufficient heat removal capability. CYAPCo proceeded with two actions:

(1) to seek NUSCo engineering support to extrapolate the heat removal curves for temperatures less than 50 degrees Fahrenheit ( F) as the heat removal curves do not extend

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below 50 degrees; and, (2) flush the differential pressure instrumentation. NUSCo engineering concluded that the "as-found" data was acceptable. CYAPCo flushed the differential pressure instrument sensing line and performed the surveillance on the same day.

The surveillance results were acceptable, and the licensee exited the applicable forty-eight hour technical specification requirement. The inspector considers CYAPCo's action to perform the surveillance at an increased frequency was appropriate to assure CAR coolers

were operable under adverse river conditions.

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4.0 ENGINEERING AND TECIINICAL SUPPORT (71707 and 62710)

The inspectors reviewed selected engineering activities. Particular attention was given to i

safety evaluations, plant operations review committee approval of modifications, procedural

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controls, post-modification testing, procedures, operator training, and Updated Final Safety Analysis (UFSAR) and drawing revisions.

4.1 Main Steam Line Supports The inspector discussed with cognizant licensee personnel, the applicability at Connecticut Yankee of an issue at another nuclear power plant, concerning an overstress condition of the

main steam line support system. At the Beaver Valley facility, the licensee discovered that

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under existing steam flow conditions, that rapid closure of the main steam isolation valves (MSIV's) create stresses in the steam piping supports that exceed the original design allowable valves. The original design considered only the effects of a turbine trip, not that of a main steam isolation.

NUSCo engineering concluded that the main turbine stop valve closure was more limiting to main steam support stresses than closure of the MSIV's. The conclusion was supported by the designed closure time of the MSIVs in comparison to the main turbine stop valves. The MSIV closure results in a turbine trip. The MSIV's are designed for full closure between six to eight seconds, whereas the stop valves close less than a second.

The pressure stresses are minimized during a main turbine trip because of control system response of the high pressure steam dump, and the design of the main steam supports. The design of the high pressure steam dump is to open within two and one-half seconds of a turbine trip. The design of the steam dump control system would minimize the pressure

surge. The main steam line supports are equipped with slider feet allowing unrestricted

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lateral movement.

Based on the NUSCo engineering evaluation, the inspector concluded that reasonable justification exists that the main steam line supports would not be in an overstressed condition during an MSIV closure event.

4.2 New Fuel Receipt Inspection The inspector reviewed activities in progress periodically during the inspection to receive and inspect new fuel, and to store it temporarily in the new fuel storage vault. The work was completed through the cooperative efforts of reactor engineering, operations, maintenance, health physics, security and quality control personnel. The following plant procedures were used as references to review the activities in progress: SNM 1.4-3, "Dctailed Inspection of-l

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New Fuel Assembly and RCCAs," Revision 12; SNM 1.4-2, " Removing Fuel from Shipping Containers," Revision 18, and AWOs 93-4273,4274 and 4275. The inspector reviewed the licensee's inspection results for fuel bundles X28, X29, X31 and X33 as documented on

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SNM 1.4-3 Checklist 7.1, and witnessed the inspection in progress on April 26 and May 6, for fuel bundles X12, X27, and X35.

The inspector noted good coordination between the members of the work party to complete the unloading, inspection and storage activities in a controlled and efficient manner. Reactor engineeting (RE) personnel provided good oversight and control of the operations.

Applicable procedures, checklists and forms were available and in use. QC coverage of the activities on May 6, was completed in accordance with preplanned checklists and through in-plant observation of the work in progress. Licensee personnel followed the procedures and administrative controls governing the activity.

The inspector independently verified, on a sampling basis, that the prerequisite and initial conditions of SNM 1.4-2 and 1.4-3 were met. The inspector verified that QC accepted tools and gages were in use. For fuel bundles X12, X27, and X35, the inspector independently verified by direct observation that the following items were satisfactory: fuel bundle identification and storage locations; upper end fitting leaf springs, cap screws and lock cups; spacer grids and instrument sheaths; fuel rod straightness, cleanliness and appearance; bottom nozzle locknuts, end caps and support legs; and, control rod sheath clearances.

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No unacceptable conditions were identified. The inspector noted all personnel involved with the work on April 26 and May 6, were familiar with assigned tasks and equipment. In particular, the RE personnel responsible for the fuel inspections were qualified, and were very knowledgeable of the fuel bundle characteristics, the governing inspection procedures, and the detailed inspection techniques and acceptance criteria. The inspector concluded that the inspection of new fuel was completed in a safe and orderly manner.

4.3 Augmented Inservice Inspection Program The inspector found that TS surveillance procedure SUR 5.7-159, " Main Steam and Feedwater HELB Inspection Points," differed from TS Table 4.0-1, " Augmented Inservice Inspection Program," and Table 4.0-2, " Weld Locations on Steam Supply Lines to Auxiliary Feedwater Pumps." The inspector noted that SUR 5.7-159 had more weld inspection locations than the TS requirement.

The inspection objective was to understand the basis for the discrepancies. The inspection consisted of document reviews, discussions with CYAPCo and NUSCo personnel and observation of SUR 5.7-15.-

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In 1977, the NRC staff issued TS amendment.No.16 which added the augmented inservice inspection program (TS 4.0.6.) The program consisted of periodic weld inspections (volumetric and visual) and leakage inspection of portions of the main steam and feedwater systems outside containment. The portion of each system was identified based on the potential high energy line break effects to both auxiliary feedwater pumps and controls. The -

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NRC staff concluded that the augmented inservice inspection provided reasonable assurance i

of continued integrity of high energy piping systems outside containment.

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In March 1991, CYAPCo proposed, and the NRC approved, TS amendment No.139 based on plant modifications in 1987 to add one weld and two repaired welds to TS Table 4.0-1.

During the 1991-1992 refueling outage, CYAPCo performed extensive pipe weld exams, pipe weld repairs and valve replacements in portions covered by the augmented inservice i

inspection program. The activities were controlled by authorized work orders and/or plant modifications. One modification, Plant design Change Evaluation (PDCE)89-144,

" Replacement of MS-NRV-18,28,.38,48 and Addition of Test Connections," replaced four

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main steam non-return valves for the auxiliary feedwater turbines. Design change notice DCY-P-003-92 added a spool piece to the downstream side of each valve. The spool piece added four pipe-to-pipe welds in the inspection area of the augmented inspection program.

Based on the design change notice, CYAPCo engineering personnel initiated a TS change request.

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The change had been submitted to NUSCo licensing department; however, no action to process the change was initiated. Procedure NEO 4.02, " Proposed Technical Specification Change Requests and Emergency Waiver Request," defines the process for the initiation, review, approval and disposition of a proposed TS change request. At the end of the inspection period, the inspector identified that the TS change request submitted by ISI engineering was not processed according to NEO 4.02 by NUSCo licensing personnel.

Prior to the 1991-1992 refueling outage, the NRC staff reviewed the licensee's proposed modifications to the auxiliary feedwater system. The NRC:NRR staff stated to CYAPCo that augmented ISI was not an acceptable alternative to provide assurance of the high energy line pipe integrity. On December 24,1992, CYAPCo committed to the NRC staff to initiate an augmented erosion / corrosion program for piping in the main steam and feedwater systems affecting auxiliary feedwater system operation. On April 15, 1993, the Nuclear Review Board approved proposed technical specification change C-1-93 to remove the existing TS 4.0.6 (Augmented ISI) and replace it with an augmented erosion / corrosion inspection.

Based on the above, NUSCo licensing personnel decided not to process the TS change to the existing TS 4. _-

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The inspector reviewed documents related to weld examination results and repairs during the 1991/1992 refueling outage. The examinations were appropriately implemented within the

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requirements of the ASME code and CYAPCo ISI program expectations. The inspector observed appropriate implementation of the monthly leakage surveillance (SUR 5.7-159).

The inspector determined that CYAPCo appropriately implemented the requirements of TS 4.0.6.

j Conclusion CYAPCo did not request an amendment to TS 4.0.6. (based on modifications made under a design change notice DCY-P-003-92 to PDCE-89-144) since a commitment (i.e., augmented erosion / corrosion program) to the NRC superseded the need for the license change. The i

inspector discussed with CYAPCo personnel the importance for the TS requirements to accurately reflect plant conditions. The licensee acknowledged the inspector's comment. The inspector determined no safety significance exists for this issue since all the required non-destruction examinations and leakage inspections had occurred within the required time intervals.

4.4 Fuel Rod Defects During Manufacturing On April 22, during the discussions at a daily planning meeting (POD), the inspector learned that the fuel vendor (B&W Fuels) identified cladding defects in the heat-affected zone of the upper fuel rod end cap weld. The clad defects were identified using radiographic examinations during the fuel assembly process. The fuel vendor initiated a corrective action program to perform two-view radiographic examinations on all fuel rods. As of April 29, the vendor inspected approximately 6,000 fuel rods and rejected twenty-six. The fuel rods were rejected based on a clad thickness less than designed minimum wall. The accepted rods were assembled and shipped to Connecticut Yankee during the inspection period.

NUSCo reactor engineering performed onsite reviews of the fuel vendor's activities during the corrective action process. The vendor had not identified the root cause of the clad defect at the end of the inspection period. The NRC:NRR Vendor Inspection Branch was notified by the inspector of the fuel defects. The Vendor Inspection Branch will consider the generic implications, and follow-up fuel vendor activities.

5.0 SAFETY ASSESSMENT AND QUALITY VERIFICATION (40500,90713,90712, and 92701)

5.1 Plant Operations Review Corrunittee The inspectors attended several Plant Operations Review Committee (PORC) meetings. The inspector noted technical specification 6.5 requirements for required member attendance were

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met. The meeting agendas included procedural changes, proposed changes to the Technical Specifications, Plant Design Change Records, and minutes from previous meetings. The

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inspector noted PORC meetings were characterized by frank discussions and questioning of the proposed changes. In particular, good discussions and committee member interaction

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existed for the temporary modification to install the shutdown risk emergency diesel generator; PDCR 1312, " Diesel Generator Exhaust Fan Controller;" and PDCR 1361,

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" Pressurizer Surge Line Support." Items for which adequate review time was not available were postponed to allow committee members time for further review and comment. The inspector concluded that the comnittee closely monitored and evaluated plant performance and conducted a thorough self-assessment of plant activities and programs.

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5.2 Review of Written Reporc

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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

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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:

LER 93-001-01, Fire Door Opened Without Entering Limiting Condition of Operation and Establishing u Fire Watch r

CYAPCo provided this supplemental report to document the results of a programmatic review l

of past corrective actions for fire door related LERs. Past NRC inspection, as documented in

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report 50-213/93-01, closed LER 93-001. CYAPCo's programmatic review concluded that individual corrective actions for each event were appropriate. The licensee's review and conclusions reflected the NRC's independent conclusions as documented in report

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50-213/92-18. The supplemental LER closed.

I.ER 92-020-01, Steam Generator Level Malfunction with Inadequate Means of Inserting a

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Trip Signal This supplemental LER documents CYAPCo's faihire to meet a corrective action commitment in LER 92-020. The commitment was to install temporary manual trip switches for each steam generator by September 30,1992. The supplemental LER was dated April 16, 1993.

L The licensee recognized that it failed to meet the commitment to initiate and track prescribed j

actions and that resolution of the procurement controls of the temporary switches remained

unresolved. The licensee attributed this problem to a failure in the management oversight of

'l the commitment tracking process. This deficiency was also highlighted during a Quality

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Assurance (QA) audit in March 1993..At the close of the inspection period, the licensee was verifying commitments from LERs dated back to 1988 had been completed. The review was still in progress. Subsequent NRC inspections will evaluate the extent and effectiveness of licensee actions to fulfill LER corrective action commitments. This item is considered unresolved (URI 93-06-01).

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In addition to the failure to meet a corrective action commitment, CYAPCo also documented the results of a programmatic review of technical specifications to identify any other non-reactor protection system channel trip signal that could not be inserted within the required time interval. CYAPCo identified that the emergency bus undervoltage channel of the ESFAS could not be placed in a tripped condition within one hour. CYAPCo developed a procedure to accomplish this evolution. The inspector verified that procedure CMP 8.8-1,

" Installation and Removal of Bus 8 and Bus 9 Undervoltage Trip Signals," was approved to allow for placing one channel in a tripped condition as required by TS 3.3-2, Table 3.3-2.

This LER is closed.

LER 93-002, Control Rod 31 and 32 Malfunedon CYAPCo reported a manual actuation of the reactor protection system. NRC verified licensee causal analysis and short-term corrective actions in report 50-213/93-03. CYAPCo

long term actions to be completed during the Cycle 17 Refueling Outage are to replace the lockwasher for the mechanical fastener on the coil conMetors in the rod control system.

Pending future inspections to verify CYAPCo corrective actions, this LER remains open.

  • Monthly Operating Report 93-04 5.3 Follow-up of Previous Inspection Findings The inspector reviewed licensee actions taken in response to open items and fm' dings from previous inspections. The inspectors determined if corrective actions were appropriate and thorough, and whether the previous concerns were resolved. Items were closed where the

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inspector determined that corrective actions would prevent recurrence. Those items for which additional licensee action was warranted remain open. The following items were reviewed:

(Closed) IFI 50-213/92-20-01: Auxiliary Feedwater Suction Line Pine Stresses Due to Water Hammer This item questioned if the auxiliary feedwater piping and supports exceeded stress limits

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beyond what could be surmised from a visual inspection. A water hammer event on November 3,1992, precipitated the issue.

On February 25,1993, CYAPCo engineering initiated a request of NUSCo engineering (LOE-93-RA034) to determine if the allowable stress limits were exceeded in the auxiliary feedwater suction piping and supports. NUSCo engineering concluded that the pipe stresses experienced during the water hammer event were approximately one-half of the material yield

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stress. The assumed inlet pipe stresses used a system pressure of 2,000 psig, plus the normal deadweight stresses. The pressure shock was assumed to be 2,000 psig since the auxiliary feedwater discharge pressure gauge-was not damaged. The range of the gauge is from 0 to 2000 psig. The inspector considers this issue closed.

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The inspector reviewed other corrective actions by CYAPCo to prevent recurrence of a water hammer affecting AFW operation. Water hammer has been reported by CYAPCo on three

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other occasions as documented in plant information reports (PIRs)92-154, 91-130, and 90-268. The phenomenon and root cause was identified; however, corrective actions from past PIRs appeared to be ineffective to prevent the occurrence. Previous corrective actions i

were to replace the AFW discharge check valves (FW-CV-153, and FW-CV-184),

replacement of the AFW to feedwater header check valves (FW-CV-156-1, 2, 3, and 4), and the addition of temperature monitoring upstream of the AFW discharge check valves.

CYAPCo corrective actions for the November 3 event (PIR 92-187) are to replace FW-CV-156-1, 2, 3, and 4 valves during the upcoming refueling outage. The replacement valve uses a substantially different design to prevent leak-by. In addition, the licensee changed the associated procedures to isolate suction pressure gauges, and provide a operator caution step to slowly open the discharge manual valve.

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(Closed) Violation 92-01-01. Containment Valve LD-TV-230 This item was last reviewed in NRC Inspection Report 50-213/93-01. Corrective actions to address the identified deficiency were verified at that time. The matter was left open at the time pending further NRC review of the licensee's process for tracking and addressing commitments to the NRC. NRC concerns in this area are described further in Section 5.2 of this report and will be tracked under a new unresolved item number. Since the concerns

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originally associated with containment valve LD-TV-230 have been satisfactorily addressed, item 92-01-01 is closed.

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(Closed) Violation 92-04-01. Diesel Generator load Shedding

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This item was last reviewed in NRC Inspection Report 50-213/92-15. Corrective actions to address the identified deficiency were verified at that time. The matter was left open at the time pending further NRC review of the licensee's process for tracking and addressing commitments to the NRC. NRC concerns in this area are described further in Section 5.2 of this report and will be tracked under a new unresolved item number. Since the concerns originally associated with the diesel generator load shed surveillance testing have been satisfactorily addressed, item 92-04-01 is closed.

5.4 Minimum Temperature for Criticality The inspector reviewed the information contained within NRC Technical Information Summary (TIS) RIII-93-05. The information stated that Zion Nuclear Power Station was in the process of standardizing their technical specifications and queried the basis for the minimum temperature for criticality of 500*F. Westinghouse confirmed that Zion was not

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analyzed for critical temperatures below the no-load operating temperature (547'F). The

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licensee also identified that the power range nuclear instrumentation were outside of their analyzed range of temperatures. The licensee concluded that the current departure from nucleate boiling (DNB) ratio could be met during a design 5 sis accident with initial temperatures below 530 F.

The inspector verified that CYAPCo technical specification (TS) 3.1.1.6 requires the reactor coolant average temperature to be greater than or equal to the no load temperature of 525 F for criticality. The TS basis for the temperature limitation is that: (1) the moderator temperature coefficient is within its analyzed temperature range; (2) the trip system instrumentation is within its normal operating range; (3) the pressurizer is capable of being in an operable status with a steam bubble; and (4) the reactor vessel is above its minimum reference transition temperature. The inspector also verified that Updated Final Safety Analysis Report (UFSAR), Section 15.2.0., lists the minimum reactor coolant inlet temperature to be 520.8*F for accident analysis assumptions. The inspector considers this issue closed.

6.0 MEETINGS 6.1 Exit Meetings During this inspection, the inspector held periodic meetings with station management to

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discuss inspection observations and findings. At the close of the inspection period, an exit -

meeting was held on May 19,1993, to summarize the conclusions of the inspection. No written material was given to the licensee and no proprietary information related to this inspection was identified.

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In addition to the resident inspector's exit meeting, the following exit meetings were held for inspections conducted by Region I based inspectors during the period.

Inspection Reporting Areas Reoort No.

Dates Insoector Insoected 50-213/93-04 4/12-4/16/93 Stewart In-service Test /

Check Valve Program

50-213/93 4/26-4/30/93 D' Antonio Operator Examination

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6.2 Meetings with Local Officials On the dates indicated during this inspection period, the inspectors held meetings with the First Selectwomen of the following towns: East Haddam, April 13; Haddam, April 15; and

Colchester, April 23. The purpose of the meetings was to introduce the resident inspectors to l'

the local officials, and to describe generally the role of the NRC, in particular, the inspection of activities at Haddam Neck. The inspector felt that the meetings were beneficial to openiug-

a line of communication with the town officials.

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