IR 05000382/1999005

Insp Rept 50-382/99-05 on 990228-0410.Non-cited Violations Noted.Major Areas Inspected:Operations,Maintenance, Engineering & Plant Support

ML20206S451
Person / Time
Site:	Waterford
Issue date:	05/10/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20206S446	List: IR 05000382/1999005 ML20206S441
References
50-382-99-05, 50-382-99-5, NUDOCS 9905210138
Download: ML20206S451 (28)

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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION 3 REGION IV l l

l Docket No.: 50-382 l

License No.: NPF-38 j Report No.: 50-382/99-05 l

Licensee: Entergy Operations, In l Facility: Waterford Steam Electric Station, Unit 3 ,

Location: Hwy.18 Killona, Louisiana j Dates: February 28 through April 10,1999 Inspectors: T. R. Farnholtz, Senior Resident inspector J. M. Keeton, Resident inspector Approved By: P. Harrell, Chief, Project Branch D ATTACHMENT: Supplemental Information l

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9905210138 990510 l PDR ADOCK 05000382 O PDR

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_E_XECUTIVE SUMMARY Waterford Steam Electric Station, Unit 3 NRC Inspection Report 50-382/99-05 This routine, announced inspection included aspects of operations, maintenance, engineering, and plant support activities. The report covers a 6-week period of resident inspectio Operations

In general, operators performed very good throughout the refueling outage and the subsequent plant startup. The use of procedures and procedure adherence was goo Operators exhibited a cautious approach and a qu estioning attitude when involved in infrequently performed evolutions. Operations management was directly involved in daily outage activities (Section 01.2).

• The licensee appropriately addressed three instances in which new control element assemblies became disconnected from the grapple hook during movement or placement into fuel assemblies. The three control element assemblies were replaced (Section O2.1).

The cleanliness and material condition of the containment building and equipment following completion of refueling activities was considered adequate (Section O2.2).

• Several examples of human performance errors in the area of operations were identified during the refueling outage. These errors resulted in additional work, radiological ;

exposure, and potential equipment damage. The licensee's actions were appropriate in response to each event (Section 04.1).

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• A violation was identified for the failure to ensure an adequate valve lineup was performed prior to operating Low-Pressure Safety injection Pump B. This resulted in potential damage to safety-related equipment when the pump was operated for approximately 30 minutes with the suction valve closed. In addition, control room operators demonstrated lack of a questioning attitude when they did not adequately address unexpected indications when the pump was started. This Severity Level IV violatior'is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 99-0382 (Section 04.2).

Maintenance

• Overall performance of maintenance during Refueling Outage 9 was acceptabl However, many examples of human error were noted resulting in real or potential equipment damage. In some instances, equipment was returned to service following maintenance activities that were improperly performed. The licensee took appropriate actions in response to each event (Section M4.1).

• A violation was identified for the failure to properly torque the eight bolts on the inlet flange of Pressurizer Safety Valve RC-317A. This flange makes up part of the reactor coolant system pressure boundary. The cause of the event was identified as incorrect

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use of the hydraulic wrench used to torque these fasteners to their final value. This Severity Level IV violation is being treated as a noncited violation consistent with l Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective

! action program as Condition Report 99-0265 (Section M4.2).

! Quality Assurance inspectors were very active and highly visible throughout the plant during Refueling Outage 9 and contributed to the safe performance of maintenance and modification activities. The inspectors considered self-assessment efforts to be very good (Section M7).

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When a resistance temperature detector could not meet the time response requirements of the Technical Requirements Manual, licensee engineers took appropriate actions to resolve the issue (Section E1.1).

The licensee's actions to correct a packing leak from a shutdown cooling system isolation valve were appropriate. An engineering evaluation that provided the basis for not testing the motor-operated valve after retorqueing the packing gland nuts was '

determined to be sound and based on good engineering judgment (Section E1.2).

The licensee's recovery plan to ensure acceptable performance of Agastat E7000 series j time delay relays used in the emergency diesel generator sequencer system was '

considered adequate. Past errors by the licensee resulted in the relays not receiving adequato maintenance prior to the establishment of the recovery plan, in addition, i manufacturers' recommendations and industry experience and communications l regarding these relays were not applied effectively and in a timely manner (Section E1.3).

The licensee's previous actions to correct a condition involving deteriorated lube oil hoses on the nonsafety-related reactor coolant pump motors were inadequate to resolve the problem. The hoses had been replaced every refueling outage since the plant was placed in service due to deterioration of the outer coating. This deterioration had the i potential to interfere with the proper operation of the pump's lubricating oil system. The need to repeatedly replace the hoses resulted in unnecessary radiological exposure to workers (Section E1.4).

Engineering personnel provided good support of plant activities during the refueling outage. System engineers were active and involved in the maintenance activities being performed on their assigned systems and components. The availability of technical knowledge and resources facilitated the successful completion of maintenance activities (Section E4.1).

A noncited violation was identified for the failure to perform required surveillance testing of four containment vacuum relief valves. The valves were not classified as safety and relief valves and therefore were not being tested in accordance with the appropriate section of the ASME Section XI Code. This Severity Level IV violation is being treated

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3-as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 99-0344 (Section E4.2).

Plant Support

• Health Physics technician and plant worker performance during the refueling outage was good. Postings, radiation work permits, and as-low-as-is-reasonably-achievable principles were used effectively to minimize radiation worker exposure (Section R4.1).

• The performance of the security organization during the inspection period was acceptable. However, several events occurred, which demonstrated a lack of attention to detail and questioning attitude by security personnel (Section S1.1).

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Report Details Summarv of Plant Status l

At the beginning of this inspection period, the plant was shut down and in Mode 6 for Refueling '

Outage (RFO) 9. On March 8,1999, core alterations began and were completed on March 1 The reactor was started up on April 1 and reactor power reached 100 percent on April 6. At the )

end of this inspection period, the plant was operating at 100 percent powe I. Operations 01 Conduct of Operations (71707)

O General Comments (71707)

The inspectors performed continual observations of refueling operations, control panel manipulations, special tests, and infrequently performed evolutions. The inspectors observed operators utilize good self-checking and peer-checking techniques when manipulating plant equipment. Operators generally utilized good communication techniques; however, the inspectors observed that the implementation of these techniques degraded as the outage progresse .2 Refuelina Operations. Midlooo Operation. and Postrefuelina Plant Startuo Insoection Scoce (71707)

The inspectors conducted frequent observations of control room operations during refueling operations, midloop operations, and plant startup. Procedures and surveillance tests required to be completed for refueling, maintenance, and plant startup were reviewe Observations and Findinas During March 3-8,1999, the licensee encountered several delays during preparation for refueling. Following removal of the reactor head and filling the refueling cavity, operators were preparing to remove the reactor vessel upper guide structure. The operators discovered that one of the alignment pins for the upper guide structure lifting rig had been installed in the wrong stud hole in the reactor vessel flange. A diver was sent in with a special tool to move the alignment pin to the correct stud hole locatio The licensee initiated Condition Report (CR) 99-0259 to enter this error into their corrective action progra After the upper guide structure had been removed, the refueling operators began removing the incore instruments. The operators discovered that the incore instrument cutting device would not automatically cut the instruments to the required length for storing in the canisters. Thus, the time to dispose of the incore instruments and the accompanying radiation doses the operators received were substantially increase Additional problems were encountered with polar crane and jib crane operability and availabilit !

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l On March 8, core alterations were commenced. A major problem encountered during core alterations involved control element assemblies (CEA) becoming ungrappled as discussed in Section O2.1 of this report. The inspectors observed core alterations while accompanying the refueling supervisor on the refueling machine in containment. Core alterations were completed on March 19 and the reactor was reassemble On March 20, the inspectors observed operators drain water from the reactor coolant system (RCS) to lower level to midloop. The operators were very cautious during the l draindown. They ensured that all three levelindications were in agreement by periodically stopping the draindown to allow level indications to equalize prior to continuing. Applicable procedures were in continuous use throughout the draindown and midloop operations. The inspectors noted that a shutdown cooling system watch was established throughout RCS reduced level operations. On March 21, following the completion of midloop maintenance activities, operators raised RCS level to 19 feet and exited midloop conditions. The reactor head was tensioned and Mode 5 was entered i on March 2 !

Operators performed a plant heatup and entered Mode 3 on March 29. The inspectors verified that operators established the required conditions for changing plant modes and that required surveillance tests and inspections were completed. On March 31, operators performed a reactor startup in accordance witn Nuclear Engineering Procedure NE-002-030, " Initial Criticality," Revision 5. Reactor criticality was achieved on April 1. The inspectors observed control room activities during the approach to criticality and during portions of the low power physics testing. Control room access was strictly controlled during the approach to criticality. Three operators in training were performing most of the control board manipulations under close supervision of the licensed operators. Three-way communication techniques were strictly enforced by the control room supervisor and shift superintendent. Operations management was present in the control room during all phases of the startu c. Conclusions In general, operators performed very good throughout the refueling outage and the subsequent plant startup. The use of procedures and procedure adherence was goo Operators exhibited a cautious approach and a questioning attitude when involved in infrequently performed evolutions. Operations management was directly involved in daily outage activities.

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-3-02 Operational Status of Facilities and Equipment O2.1 Disconnection of CEAs From Hoist Durina Refuelina Operations Inspection Scoce (71707)

The inspectors reviewed documentation and drawings and conducted interviews concerning three events involving the unexpected disconnection of new CEAs from the hoist during reactor refueling operation Observations and Findinas During the outage, the licensee replaced all of the CEAs in the core. CEAs were inserted into a fuel assembly using the CEA hoist, which was mounted on the refueling machine in containment. The grapple hook on the hoist was a cylindrical component with four J-shaped hooks positioned 90* apart. The grapple hook was placed over the top portion of a CEA and rotated to line up the hooks with the corresponding supports on the CEA. The weight of the CEA held the assembly in place during manipulation, j The CEAs consisted of five fingers, which had to be aligned with guides in the fuel assembly. Once the CEA was properly aligned, it was lowered into the fuel assembl On March 12, while lowering a CEA into a fuel assembly, the CEA became disconnected from the grapple hook and fellinto the fuel assembly. The CEA had been .

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inserted into the fuel assembly approximately 17 inches. The licensee performed a visual inspection of the CEA and confirmed that it was not damage A total of 71 CEAs had been successfully handled prior to this event. Following the March 12 event, the licensee revised the applicable procedure. This revision primarily changed the speed at which the CEAs were lowered into the fuel assemblies to reduce the possibility that the CEA would lift up and off the grapple hook during insertio Manipulations of CEAs were then continue On March 13, a second new CEA became disconnected from the grapple hook as it was being aligned for insertion into a fuel assembly. The CEA was supported by the top of the fuel assembly and the CEA hoist assembly and remained in an upright position; The licensee's inspection of the CEA revealed that it was not damaged. The CEA was successfully grappled and lowered into the designated fuel assembl On March 14, a third new CEA was being inserted into a fuel assembly when it became disconnected from the grapple hook. The CEA dropped approximately 12 inches into the fuel assembly. The licensee determined that the CEA was not damaged. The licensee stopped loading CEAs into fuel assemblies located in the reactor vessel. The remaining CEAs were inserted into the fuel assemblies in the spent fuel pool. The three affected CEAs were replaced. No further problems were experience The licensee reviewed these events and determined that the probable cause was a difference between the metallurgical properties of the new, unirradiated CEAs and the irradiated CEAs. They believed that the flexibility of the new CEAs made them more difficult to maneuver and insert into the fuel assemblies than the irradiated CEAs and contributed to the unexpected releases. The licensee also believed that the design of the grapple hook did not provide sufficient assurance that the CEA would remain grappled during placement into the fuel assembly. The inspectors considered the licensee's actions regarding these events to be adequate. Appropriate concern for the proper handling of reactor internal components was demonstrate Conclusions The licensee appropriately addressed three instances in which new CEAs became disconnected from the grapple hook during movement or placement into fuel assemblies. The three CEAs were rep! aced.

O2.2 Containment Walkdown and Inspection Prior to Reactor Operation Inspection Scoce (71707)

The inspectors performed a thorough and complete walkdown of the containment building following completion of refueling activities and before commencement of reactor operations. The focus of this inspection was to assess the general, as-left condition of the building and to ensure that no unsecured or foreign material was in the building that could interfere with a safety system functio Observations and Findinas .

The inspectors assessed the as-left condition of the containment building during a walkdown conducted on March 28,1999. The licensee had completed all work in the containment building with the exception of removing several scaffolds that were in place to allow inspections to be performed at normal RCS operating temperature and pressur The general condition of the building was acceptable. The inspectors identified materials such as pieces of tape, tie wrap, wire, and other small articles in various areas of the building. A bottle of leak detector solution was also recovered from an area behind a cabinet. The amount of material observed was relatively small and did not cause concern for the ability of safety-related plant equiprnent to function, as designe The inspectors found that many of the horizontal surfaces in low traffic areas contained a sigreificant amount of dust and dirt. Of particular concern was the floor of the ]

pressurizer cubicle. The inspectors communicated this concern to the licensee and this area was cleane The material condition of the building and components was adequate. Some areas of rust and corrosion were observed but were not significant. Several areas of peeling paint were observed, particularly on the end caps of the steam generator snubber oil l

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-5-reservoirs. This condition was not widespread and, therefore, did not raise a safety concern. Some damage was noted on a rubber section of a ventilation duct for the reactor head area. This condition was evaluated by the licensee as not being an immediate concern and it was entered into the maintenance progra Conclusions The cleanliness and material condition of the containment building and equipment following completion of refueling activities was considered adequate.

04 Operator Knowledge and Performance 04.1 Human Performance issues in the Area of Operations Insoection Scope (71707)

The inspectors reviewed numerous events resulting from human performance errors by Operations Department personne Observations and Findinas The inspectors reviewed numerous events, which occurred during RFO 9, that involved performance errors by operators. Although the safety significance of these events was small, the inspectors were concerned with the number of errors. Several examples are described below:

• On February 23,1999, a tagout was hung for the purpose of performing inspections of Check Valves CVC-219 and -221B. An error committed during the preparation of the tagout resulted in the failure to properly drain and vent the downstream side of Valve CVC-221B (CR 99-0187).

• During preparation for reactor refueling, one of the two alignment pins used to l

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properly position the upper guide structure lifting rig was installed in the wrong reactor vessel stud hole. This resulted in a delay in refueling activities and unnecessary radiological exposure to a diver during the process of repositioning !

the pin (CR 99-0259). l

• During the process of reactor vessel disassembly, technicians failed to install a reactor vessel stud hole protective plug. This plug serves to protect the stud ,

hole threads from damage and to prevent the introduction of borated water into the stud hole during refueling operations. No damage was noted, but additional l work was required to de-water the stud hole during reassembly of the reactor 1 vessel, resulting in unnecessary radiological exposure to worker * During RCS heat up, two charging pumps were in operation with the pressurizer level control in manual. Operators failed to notice that pressurizer level had

6-increased to 68 percent, which exceeded the specified limit of 62.5 percent. This condition existed for about 6 minutes before the operators identified and corrected the high level (CR 99-0453).

• Operators failed to ensure a proper valve lineup was completed prior to running Low-Pressure Safety injection (LPSI) Pump B. This resulted in the pump being operated for approximately 30 minutes with the suction valve closed. Control room operators demonstrated lack of,a questioning attitude when they did not adequately address unexpected indications when the pump was started. This event is described in detail in Section 04.2 of this report (CR 99-0382). l

In general, operations personnel performance was acceptable throughout the outag However, several personnel errors in the area of operations resulted in additional work, additional radiological exposure, and potential equipment damage. The licensee evaluated each of these events and acted appropriately. The inspectors were I concerned that the number and significance of some of these events could have potentially resulted in serious damage to plant equipment or personnel injur Conclusions Several examples of human performance errors in the area of operations were identified during the refueling outage. These errors resulted in additional work, radiological exposure, and potential equipment damage. The licensee's actions were appropriate in response to each even .2 Miscositioned LPSI Pumo B Suction Valve Inspection Scope (71707)

The inspectors reviewed the circumstances surrounding a mispositioned LPSI pump suction valve. This valve was required to be open following a valve lineup check, but was left in the closed position. The associated LPSI pump was then operated with no suction availabl Observations and Findinas On March 17,1999, the licensee was preparing to perform an integrated safeguards test on EDG B. As part of this' preparation, LPSI Pump B, High-Pressure Safety injection Pump B, and Containment Spray Pump B were required to be started. When LPSI Pump B was started, operators identified that there was no discharge pressure and immediately stopped the pump. The control room operators determined that this was an expected condition due to the recirculation lineup (discharge valve closed) and the pump was restarted. Upon restart, LPSI Pump B motor amps oscillated and then steadied at 25 amps. Since the amps were steady, operators concluded that the pump was operating acceptably. High-Pressure Safety injection Pump B and Containment Spray Pump B were also started. Observed recirculation flow was approximately 150 gpm and represented the total recirculation flow from all three of these pumps.

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During the conduct of the test, Electrical Bus 3B was deenergized and then reenergized from EDG B. Each of the three pumps stopped and then restarted automatically as expected. An auxiliary operator observed that the associated LPSI pump suction piping was shaking and heard some unusual noises. The operator contacted the control room and LPSI Pump B was stopped immediately. A subsequent investigation revealed that LPSI Pump B Suction Valve SI-109B was closed. The suction piping and the seal cooler were vented and the suction valve, opened. Venting continued until water flowed from the vents. The licensee determined that LPSI Pump B was run for a total of 30 minutes with the suction valve close An engineering evaluation was conducted to determine operability of the pump. In addition, a detailed inspection was made of the pump and system piping. One rigid restraint on the suction piping was not properly centered within the clevis of the pipe clamp and a second rigid restraint was loose at the pipe clamp connection. These items were repaired and the pump inservice test was conducted with satisfactory results. The pump and associated system piping were determined to be operable. The inspectors considered these actions to be adequat The licensee determined that an auxiliary operator failed to open LPSI Pump Suction Valve SI-109B during a previous valve lineup. The step was signed off as being completed, but the valve was not opened. The licensee found that excessive tasking of the auxiliary operator by different individuals, and pressure to complete these tasks in a ,

timely manner, contributed to the erro l The inspectors determined that the licensee's actions were inadequate in several ,

respects. The failure to open the suction valve during the valve lineup is a violation of l

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Operations Procedure OP-903-116," Train B Integrated Emergency Diesel Generator / Engineering Safety Features Test," Revision 6. Attachment 10.5 of this procedure specified that Valve SI-109B must be open for the performance of the test. In addition, the lack of a questioning attitude was evident when LPSI Pump B was started the second time and indications were not as expected. The control room operators convinced themselves that the observed indications (pump amps) were acceptable and i pump operation was continued. As a result of these actions, the licensee operated safety-related equipment with an inappropriate valve lineup for an extended period of tim The failure to perform a proper valve lineup prior to operating LPSI Pump B is identified as a violation. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 99-0382 (50-382/9905-01).

c. Conclusions A violation was identified for the failure to ensure an adequate valve lineup was performed prior to operating LPSI Pump B. This resulted in potential damage to safety-related equipment when the pump was operated for approximately 30 minutes with the suction valve closed. In addition, control room operators demonstrated lack of

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a questioning attitude when they did not adequately address unexpected indications when the pump was started. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 99-038 Miscellaneous Operations issues (92901)

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08.1 (Closed) Violation 50-382/9717-01: Failure to Maintain Emergency Operating

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This violation is in the licensee's corrective action program as Commitment A-2474 The corrective actions were listed in the facility Notice of Violation response letter (W3F1-97-0260) from E. C. Ewing to NRC, dated November 24,1997. The corrective act:ons to date have included development and implementation of operator training on l the use of Procedure OP-902-009," Emergency Operating Procedures Standard )

Appendices," Revision 0, which provided additional guidance to operators during the use l of emergency operating procedure II. Maintenance  ;

M1 Conduct of Maintenance (61726,62707)

The inspectors observed all or portions of the following maintenance and surveillance

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activities, as specified by the referenced work authorization or surveillance procedures:

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• 01177294 EDG B Exhaust Fan Motor Replacement

• OP-903-115 Train A Integrated EDG/ Engineering Safety Features Test

• OP-903-116 Train B Integrated EDG/ Engineering Safety Features Test

• 99003546 Static Uninterruptible Power Supply (SUPS) 38-S Replacement

• STP 990003546 Acceptan.ce Test for SUPS 3B-S

• NE-002-030 Initial Criticality

• OP-903-090 Shutdown Margin Determination

• NE-002-020 CEA insertion Time Measurement in general, the observed work activities were conducted in an acceptable and effective manner. The technicians were knowledgeable and conducted the work as required by the applicable procedures. Appropriate support perscnnelincluding health physics, quality control, and supervisory personnel were at the work site when required.

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M4 Maintenance Staff Knowledge and Performance i M4.1 Human Performance issues in the Area of Maintenance Insoection Scope (61726. 62707)

! The inspectors reviewed numerous events involving human pedormance errors in the area of maintenance that occurred or were discovered during RFO l Observations and Findinas The inspectors reviewed numerous events that occurred during RFO 9 that involved human errors during maintenance activities. Several examples are describ9d below:

• On February 25,1999, during maintenance on the turbine-driven emergency feedwater (EFW) pump turbine governor valve, the licensee discovered that a nut was missing in the internals of the valve. The purpose of the nut was to ensure the valve plug was properly seated on the tapered portion of the valve stem. Based on review of the documentation from the last time this valve was worked (RFO P), the licensee determined that this nut was inadvertently not installed during RFO 8. This issue was documented in CR 99-0203. The licenses evaluated the configuration of the valve and concluded that the EFW pump remained operabl * On February 25, during maintenance on Auxiliary Component Cooling Water ;

System Train A, the licensee discovered that a spring-loaded check valve in the j recirculation line from the auxiliary component cooling water pump was missing !

the spring. This spring assisted the disk in the closed direction to ensure proper operation. Based on review of the documentation from the last time this valve was worked, the licensee determined that the spring was inadvertently not installed during RFO 8. This issue was documented in CR 99-0209. The ,

licensee evaluated the configuration of the valve and concluded that the system :

remained operabl l

. On March 3, it was discovered that the eight bolts on the inlet flange of !

l Pressurizer Safety Valve RC-317A were not torqued properly. Three of the eight were hand tight and the other five were wrench tight. This resulted in leakage from the flange. The licensee determined that the machine used to torque the bolts had not been properly positioned due to interference from surrounding components. The inspectors considered this to be a human error since this ;

condition should have been identified following maintenance on this component conducted on September 22,1998. The licensee revised the maintenance task ;

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to require an additional torque check using a hand torque wrench following the l

use of the machine. This finding is discussed in detailin Section M4.2 of this report (CR 99-0265).

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On March 4, while performing testing on a SUPS, technicians incorrectly connected test equipment to the unit, which resulted in damage to the unit. The unit required repair and testing following this event (CR 99-0273).

• On March 13, a temporary plug that had been installed in the body of containment sump pump discharge line Valve SP-1061 became dislodged while water was being pumped from the sump. This caused a significant portion of the plant wing area to become radiologically contaminated when the sump water was discharged out of the open valve body. The plug did not have the appropriate pressure rating and failed when a downstream strainer became clogged and the plug was exposed to full sump pump shutoff head pressure (CR 99-0349).

• On March 17, technicians were attempting to adjust the motor operator on turbine-driven EFW pump steam supply Valve MS-401B. While adjusting the thrust settings on this operator, the allowable torque rating of the motor operator was exceeded several times. The technicians were focused on the thrust value and were not monitoring the torque value. The licensee disassembled and inspected the motor operator for potential damage. No damage was identified !

(CR 99-0384).

• On March 19, the licensee discovered that Containment Vacuum Breaker Valve CVR-102 was missing the magnetic striker plate assembly that is used to hold the valve in the closed position and allows the actuation setpoint to be adjusted. The as-found setpoint of this valve was determined to be lower than the technical manual specified. The missing items were installed and the valve adjusted to the required setpoint. The inspectors considered this to be a human error since these parts should have been installed prior to this valve being placed in service. This event is discussed in detail in Section E4.2 of this report (CR 99-0393).

• On April 1, two hoses associated with the control oil system on Main Feedwater Pump A were identified as being switched such that the two were inappropriately connected. In addition, one hose had not been connected upon completion of work on this pump (CR 99-0462).

• On April 4, while Main Feedwater Pump A was being placed in service, the licensee identified that the upper labyrinth seal had not been properly installed in the upper section of the inboard bearing housing. The pump was secured and disassembled to repair the seal (CR 99-0472).

Many other examples of human error in the area of maintenance were identified and reviewed. The inspectors noted that several of these events were discovered during RFO 9, but that the actual error was committed prior to this outage.

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The inspectors concluded that the overall performance of the maintenance organization during RFO 9 was acceptable. Hundreds of individual maintenance tasks were successfully completed during the outage. However, the inspectors were concerned

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-11-with the number and nature of the errors. These maintenance errors resulted in equipment damage or created the potential for equipment damage. In several instances, equipment was returned to service with inadequate maintenanc Conclusions Overall performance of maintenance during Refueling Outage 9 was acceptabl However, many examples of human error were noted, which resulted in real or potential equipment damage. In some instances, equipment was returned to service following maintenance activities that were improperly performed. The licensee took appropriate actions in response to each even M4.2 Pressurizer Safety Valve Flance Bolts Insoection Scope (62707)

During RFO 9, the licensee discovered that eight bolts on the inlet flange of Pressurizer Safety Valve RC-317A were not properly torqued. The inspectors reviewed the documentation and conducted interviews of personnel involved in replacing this valv Observations and Findinos On September 22,1998, the licensee replaced Pressurizer Safety Valve RC-317A to correct a seat leakage condition. This valve was attached to the top of the pressurizer with bolted flange connections at the inlet and outlet of the valve. Following replacement, this valve was placed in service for the remainder of the operating cycl During RFO 9, which began on February 19,1999, the licensee identified evidence of leakage from the inlet flange for Valve RC-317A. The leakage was minimal and was not detected during the performance of daily RCS leak rate calculations. Further investigation revealed that the cause of the leakage was eight bolts on the inlet flange were not properly torqued. Three of the eight bolts were hand tight and the remaining five were wrench tight (greater than hand tight but not torqued to their proper value).

The most likely cause of this event was identified as incorrect use of the hydraulic wrench used to torque these fasteners to their final value The pressurizer safety valve was replaced using mechanical maintenance Procedure MM-006-019, Pressurizer Safety Valve Maintenance," Revision Section 9.6 of this procedure described the requirements for installing the safety valve Step 9.6.11 directed the technicians to torque the inlet flange nuts to 2954 to 3265 ft-lbs in the marked sequence using specified increments. This step was shown as a concurrent dual verification point, which required a concurrent dual verifier to observe this action. While performing this part of the procedure, the mechanical maintenance technicians did not ensure that the hydraulic wrench was properly installed on the nut As a result, the hydraulic wrench reached its travel limit before the nuts were properly torqued. During the process, the hydraulic pressure for the wrench increased as expected when the nut torque increased. However, the increase in hydraulic pressure was actually due to the fact that the travel limit for the hydraulic wrench had been

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i-12- l l reached. Actual torque on the nuts was not increasing during this process. The final torque on these nuts was recorded as 3263 foot-pounds. The actual as-left condition of these nuts was hand tight on three and wrench tight on the remaining five. No independent verification of the torque on these nuts was performed following the initial installatio The failure to properly torque the eight nuts on the inlet flange for the Pressurizer Safety Valve RC-317A was identified as a violation. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 99-0265 (50-382/9905-02).

c. Conclusions A violation was identified for the failure to properly torque the eight bolts on the inlet flange of Pressurizer Safety Valve RC-317A. The cause of the event was identified as incorrect use of the hydraulic wrench used to torque these fasteners. This Severity I Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 99-026 l l

M7 Quality Assurance in Maintenance Activities Throughout RFO 9, the inspectors observed that Quality Assurance inspectors were very active and highly visible throughout the plant. Many CRs were generated that .

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I contributed to the safety verification of the maintenance and modification activities. The inspectors considered these self-assessment efforts to be very goo Ill. Enaineerina E1 Conduct of Engineering (37551)

E1.1 Resistance Temoerature Detector (RTD) Response Time Greater than Allowed by Technical Reauirements Manual a. Inspection Scope (37551. 92903)

The inspectors monitored activities related to replacement of an RCS cold leg RTD and reviewed the engineering evaluation for acceptance of the final response time.

b. Observations and Findinas On March 31,1999, while performing response time testing required by the Technical Requirements Manual, the licensee identified that RTD RCITE 0112 CA failed to

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respond within the required 8 seconds. The licensee tested three different RTDs and found that none met the time response requirements. CR 99-0460 was written to t

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J l-13-address the issue. This RTD provides input to two cold leg temperature indications on Control Panel 7 and to Core Protection Calculator (CPC) Channel A. The failed RTD rendered CPC A inoperabl l Li;em ee engineers contacted the vendor and requested assistance in resolution of the issue. The vendor provided the appropriate CPC penalty factors to account for the increased RTD response times. The engineers reviewed the vendor's response, performed a 10 CFR 50.59 screening evaluation, and determined new addressable constant values for CPC A using the penalty factors provided by the vendor. The engineers determinea that use of the new addressable constant penalty factors provided assurance that the Cycle 10 accident analysis remained valid. CR 99-0484 was written to track the corrective actions take On April 2, the Plant Operations Review Committee reviewed and approved the change to the CPC constants. The inspectors reviewed the vendor and engineering packages and concluded that the actions taken were appropriate. The new addressable constants were inserted and CPC A was declared operabl Conclusions ,

l When a resistance temperature detector could not meet the time rosponse requirements l of the Technical Requirements Manual, licensee engineers took appropriate actions to l resolve the issu l E1.2 Enaineerina Evaluation of Excessive Packha Leakaae on a Shutdown Coolina System isolation Valve Inspection Scope (37551. 92903)

The inspectors reviewed the engineering evaluation associated with a packing leak on Shutdown Cooling System Isolation Valve SI-401 A, a motor-operated valv Observations and Findi7gs On March 31,1999, during r containment walkdown, opvators found that Shutdown Cooling System Iso'ation Valve SI-401 A had a packing leak. Engineering personnel determined that it would be necessary to reton., s the valve packing gland nuts to stop the packing leak. Piocedure UNT-005-024,"MOV Testing, Maintenar>ce, and Trending Program," Revision 2, required that the valve be stroke-tested following completion of this adjustment. To test the valve, operators would need to cool down and depressurize the RCS to establish Mode 5 conditions. The plant was in Mode 3 when the leak was identifie Engineers performed an evaluation (ER-W3-99-0341) to determine if the valve could be adjusted without performing the required valve stroke. The eveiuation included a review of the valve's design basis and past test data for Valve SI-401 A and similar valves. The evaluation determined that sufficient margin existed between both previous test

-14-unseating thrust values and previous open thrust values to ensure that retorqueing the j

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valve packing gland nuts to previous as-left values would not result in the valve stroke time exceeding its limit. The engineering evaluation concluded that the gland nuts should be retorqued in 5 ft-lb increments until the leak stopped or the original torque Value of 97 ft-Ibs was achieve j Additionally, a Maintenance Action item was initiated to perform a timed valve stroke test during the next cold shutdown. The retest exemption and action plan were approved by the Plant Operations Review Committee on April 1. The valve packing gland nuts were retorqued and leakage was stopped prior to reaching the approved maximum torque of 97 ft-lb ?

The inspectors reviewed the engineering evaluation,10 CFR 50.59 screening, and actions taken by the licensee. The inspectors considered the engineering evaluation to be sound and based on good engineering judgment. The licensee's actions to correct this condition were appropriate Conclusions The licensee's actions to correct a packing leak from a shut 1own cooling sy a l isolation valve were appropriate. An engineering evaluation that provided tl.c basis for I not testing the motor-operated valve after retorqueing the packing gland nuts was '

determined to be sound and based on good engineering judgmen !

E1.3 Aaastat Relav lssues and Recovery Plan inspection Scope (37551)

The inspectors reviewed the licensee's implementation of a previously developed replacement and ct libration program for safety-related time delay relay Observations and Findinas During RFO 9, the licensee replaced numerous Agastat time delay relays used in the EDG sequencer system. The relays that were removed were testec to determine if they had been operating correctly. Many of these relays were found to be outside the required tolerances for time delay and did not pass the test The relay manufacturer recommended that Agastat E7000 series time delay relays should be replaced at the end of their 10-year qualified life (10 years from the manufacturer date code) and that they should be periodically checked for substantial changes in their timing delay. Prior to January 1996, the licensee had requirements to replace all safety-related Agastat E7000 series relays 10 years from the manufacturers date code, but periodic calibration checks had not been conducted. In 1996, an engineering evaluation was performed and a recommendation was made to either extend or delete the replacement tasks provided that periodic calibration checks were

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-15- l performed. Based on this engineering evaluation, the tasks to replace the relays were .

extended or deleted. However, as a result of an error, tasks were not i.!itiated to j periodically calibrate the relay In 1998, the licensee recognized this problem and established a recovery plan to calibrate or replace existing relays and to implement a periodic maintenance and calibration schedule for the relays. This plan required that all Agastat E7000 series l time-delay relays that are used in the EDG sequencer system be replaced at a frequency of 10 years. All the relays were to be replaced within 2 years. In addition, periodic calibration of the relays was to be accomplished within 18 months,3 years, or 6 years, depending on the relay application. These actions were consistent with the manufacturer's recommendations and with programs established at other nuclear plant sites. The inspectors considered these actions adequate to give reasonable assurance i that the safety-related Agastat relays used in the EDG sequencer system would perform l as require j The licensee had replaced a total of 33 Agastat E7000 series relays since the ;

implementation of the recovery plan in 1998. Of the 33 relays that were replaced, l 16 were found to be outside of the specified tolerance range. This represented a l 48.5 percent failure rate. In addition, a total of 19 relays had been calibrated under the i recovery plan. Six of these were found to outside the specified tolerance range. This represented a failure rate of 31.6 percent. The licensee performed operability evaluations for each relay found outside its specified tolerance range. In each case, no )

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operability concerns were identified. This was primarily due to the application of these relays. Many were in annonciator or alarm circuits or were in applications that were tolerant of a timing discrepancy. The inspectors reviewed a sampling of these operability evaluations and did not identify any concern The inspectors considered the licensee's actions with regard to Agastat relays to have been inadequate in the past to correct and maintain these components. Manufacturers'

recommendations and industry experience and communications concerning these relays were sufficient to alert the licensee that a more substantial program would be required to provide assurance of acceptable performance. This information was not applied effectively and in a timely manne c. Conclusions The licens9e's recovery plan to ensure acceptable performance of Agastat E7000 series time delay relays used in the EDG sequencer system was considered adequate. Past errors by the licensee resulted in the relays not receiving adequate maintenance prior to the establishment of tls secovery plan. In addition, manufacturers' recommendations and industry experierco and communications regarding these relays were not applied effectively and in a timely manner.

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-16-E1.4 Reactor Coolant Pumo (RCP) Motor Lubricatina Oil Hose Replacement Insoection Scope (37551)

The inspectors observed the licensee's actions regarding replacement of lubricating oil hoses on each of the four RCP motors. These hoses were inspected during RFO 9 and found to be deteriorated and in need of replacemen Observations and Findinas During RFO 9, the licensee conducted inspections of the RCP motors and found deteriorated lubricating oil hoses on each of the motors. The hoses were made from a braided material covered with a rubberized insulating material. The deterioration was limited to the rubberized covering and consisted of dried and cracked material. The condition did not result in leakage from the hoses. The licensee entered this condition into their corrective action program as CR 99-0248. Every hose on all four motors was replaced (25 hoses per pump,100 hoses total). The inspectors examined these hoses and observed dried, hardened, and cracked rubberized coating, which was falling off the braided section of the hose in small piece The inspectors reviewed this condition anJ ietermined that the licensee did not take timely actions to correct a known problem \ th thew hoses. In RFOs 1-6, the licensee replaced these hoses with like for-like components when deterioration was identified following each operating cycle. No apparent attempt was made to identify and install hoses that would be less susceptible to deterioration. During RFO 7, the hoses were replaced with a different type of hose since the original type was no longer manufactured. These hoses were placed in service for Operating Cycle 8 and inspected during RFO 8. Deterioration of the hoses was more severe than had been experienced in previous operating cycles. The hoses were replaced with identical hoses even though severe deterioration had been identified. Following Operating Cycle 9, the hoses were again inspected and the same severe deterioration was noted as was seen in RFO The condition did not result in leakage from the hoses. The hoses were replaced during RFO 9 with hoses of a different design that had been used at another nuclear plan The other plant reported improved performance with these type hoses, but their experience wu timite Although these components were not safety related, the inspectors concluded that the licensee's previous actions to address this issue were inadequate. In addition to the potential to interfere with the proper operation of the motor lube oil system, the process to inspect, remove, and replace these hoses caused unnecessary radio!ogical exposure to worker Conclusions The licensee's previous actions to correct a condition involving deteriorated lube oil hoses on the nonsafety-related RCP motors were inadequate to resolve the proble The hoses had been replaced every refueling outage since the plant was placed in

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-17-service due to deterioration of the outer coating. This deterioration had the potential to interfere with the proper operation of the pump's lubricating oil system. The need to repeatedly replace the hoses resulted in unnecessary radiological exposure to worker E4 Engineering Staff Knowledge and Performance E Enaineerina Suocort Durina the Refuelina Outaae

.2. Inspection Scope (37551)

The inspectors observed the performance of engineering personnel in the plant throughout RFO Observations and Findinas The inspectors observed that engineering personnel were actively involved with activities in the plant and provided technical support throughout the refueling outag System engineering personnel provided support during maintenance activities on their assigned equipment. The inspectors determined that the engineering support facilitated the successful completion of maintenance due to the availability of technical knowledge and resource Conclusion Engineering personnel provided good support of plant activities during the refueling outage. System engineers were active and involved in the maintenance activities being performed on their assigned systems and components. The availability of technical knowledge and resources facilitated the successful completion of maintenance activitie .

E4.2 Code Issues Reaardina Containment Vacuum Valves Insoection Scoce (37551)

The inspectors reviewed the circumstances concerning the licensee's failure to include four containment vacuum relief valves in the required Americzn Society of Mechanical Engineers (ASME)Section XI Inservice Testing (IST) progra Observations and Findinas

On March 11,1999, the licensee determined that four containment vacuum relief valves, CVR-101, -201, -102, and -202, had not been included in the ASME Section XI IST program as required by Technical Specification 4.0.5. In February 1997, the licensee began a review of the IST program for the second 10-year interval to comply with the 1989 edition of the ASME Section XI Code. This edition of the Code required that safety and relief valves be tested in accordance with ASME Section OM-1, Requirements for

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inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices." These valves were originally considered power-operated valves (CVR-101 and -201) and i simpic check valves (CVR-102 and -202) and tested in accordance with ASME Section OM-10," inservice Testing of Valves in Light-Water Reactor Power Plants."

These tests consisted of stroke-time testing the power-operated valves (CVR-101 and 1-201) every quarter, along with remote position indication verification at least once every 2 years and leakage testing as containment isolation valves in accordance with the Appendix J program. The check valves (CVR-102 and -202) were full stroke tested each time the plant was in cold shutdown. However, the design of Valves CVR-102 and-202 was such that the valves were held closed with a magnetic latch. The latch was adjustable and set to open at a predetermined differential pressure. Because of this feature and the function of all four of these valves, the valves should have been classified as vacuum relief devices. This classification would require testing in accordance with Section OM-1 at a frequency of at least once every 6 months to verify open and close capability, set pressure, and performance of any pressure and position sensing accessorie As a result of the error in classifying these valves, no setpoint verifications of the magnetic latches were performed. During RFO 9, the licensee found that the magnetic striker plate assembly was missing from the magnetic latch on Valve CVR-102. Also, the as-found setpoint value of this valve was determined to be lower than the value l specified in the technical manual. The licensee determined that this condition did not j render this valve inoperable. Had the valve been properly classified and tested, these conditions would have been identified and corrected in more timely manne The Technical Specification required surveillance interval for Valves CVR-101, -201,

-102, and -202 was exceeded due to the failure to properly consider the function of these components. The licensee issued Licensee Event Report (LER) 99-003-00 to report this event. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This. violation is in the I licensee's corrective action program as CR 99-0344 (50-382/9905-03). I c. Conclusions A violuan was identified for the failure to perform required surveillance testing of four containment vacuum relief valves. The valves were not classified as safety and relief valves and, therefore, were not being tested in accordance with the appropriate section of the ASME Section XI Code. This Severity Level IV violation is being tieated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 99-034 i

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E8 Miscellaneous Engineering lasues (92903)

E8.1 (Closed) LER 50-382/99-003: Missed ASME Section XI Surveillance Due to Contractor Suoolied Misinformc'ign

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This LER was issued as a result of a failure to perform applicable Technical Specification required surveillance tests on four containment vacuum relief valves. This event is described in detail in Section E4.2 of this repor IV. Plant Suncort R4 Staff Knowledge and Performance R4.1 Radioloaical Controls Durina RFO 9

. Insoection Scooe (71750)

The inspectors observed the radiation work practices of plant personnel and the performance of health physics (HP) technicians during RFO A Observations and Findinas During routine plant tours, the inspectors observed that radiation measurements had been posted in accordance with NRC requirements and licensee procedures. During the refueling outage, the inspectors observed that special postings had been ecubkhed in areas of elevated radiation caused by operation of the shutdown cooling system. As-low-as-is-reasonably-achievable postings and principles were effectively utilized. In general, plant workers complied with postings and HP technician instruction Due to the number and scope of work activities during the outage, the control of radiological exposure and contamination in containment was a significant challenge to HP technicians and radiation workers. Radiation work permits were generated for each l individual task and specific requirements were noted. Special equipment, such as protective clothing and cool vests, were readily available and in use throughout the plant as required. Support from HP personnel was conspicuous at all times both inside and outside containment, Conclusions Health physics technician and plant worker performance during the RFO was goo Postings, radiation work permits, and ALARA principles were used effectively to ,

minimize radiation worker exposur .

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-20-S1 Conduct of Security and Safeguards Activities S1.1 Security Activities Durina the Scheduled Refuelino Outaae a. Insoection Scope (71750)

The inspectors observed plant security personnel during the outage and reviewed condition reports concerning security event b. Observations and Findinas Security was considered adequate and generally conducted in accordance with applicable procedures and expectations during the inspection period. Personnel and vehicular access to the protected area was controlled, as required. Several instances of tailgating (inadvertently accessing a security controlled door immediately following another individual without ensuring proper access authorization was obtained after carding in) were noted. In addition, several other errors were noted:

• A security cabinet housing the override controls for Gate 1 of the protected area ,

fence was left unlocked and unattended. Although not secured, the cabinet was l equipped with a supervisory alarm that would have alerted security personnel had it been opened during the time it was left unsecured. In addition,ite cabinet was located inside the protected area (CR 99-0403).

• Security personnel did not provide medical response personnel with j thermoluminescence dosimeters when they entered the protected area to i

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respond to a medical emergency (CR 99-0409).

• An individual was not logged out of an area being controlled by a security officer maintaining a personnel access log (CR 99-0211).

Although these events were not significant, they demonstrated a lack of attention to detail and questioning attitude by security personne c. Conclusions The performance of the security organization during the inspection period was acceptable. However, several events occurred, which demonstrated a lack of attention to detail and questioning attitude by security personne V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management on April 16,1999. The licensee acknowledged the findings presented.

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The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie l l

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ATTACHMENT SUPPLEMENTAL INFORMATION l

PARTIAL LIST OF PERSONS CONTACTED l

Licensee R. F. Burski, Director Site Support C. M. Dugger, Vice-President, Operations E. C. Ewing, Director, Nuclear Safety & Regulatory Affairs C. Fugate, Operations Superintendent A. Harris, Acting Superintendent, System Engineering J. G. Hoffpauir, Manager, Operations T. R. Leonard, General Manager, Plant Operations D. C. Matheny, Refuel 9 Coordinator E. Perkins, Jr., Manager, Licensing G. D. Pierce, Director of Quality B. Thigpen, Director, Planning and Scheduling A. J. Wrape, Director, Design Engineering INSPECTION PROCEDURES USED

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37551 Onsite Engineering 60710 Refueling Activities 61726 Surveillance Observations 62707 Maintenance Observations 71707 Plant Operations 71750 Plant Support Activities 92700 Onsite LER Review 92901 Followup-Plant Operations 92902 Followup-Maintenance 92903 Followup-Engineering 92904 Followup-Plant Support

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-2-ITEMS OPENED. CLOSED. AND DISCUSSED

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Opened 50-382/9905-01 NCV' Failure to ensure a proper valve lineup was performed prior to operating LPSI Pump B (Section 04.2).

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50-382/9905-02 NCV Failure to properly torque the eight nuts on the inlet flange for Pressurizer Safety Valve RC-317A (Section M4.2).

50-382/9905-03 NCV TS-required surveillance for Section XI testing exceeded (Section E4.2). ,

Closed 50-382/9717-01 VIO Failure to maintain emergency operating procedures j (Section 08.1). l 50-382/9905-01 NCV Failure to perform a proper valve lineup prior to operating LPSI Pump B (Section 04.2).

50-382/9905-02 NCV Failure to properly torque the eight nuts on the inlet flange for Pressurizer Safety Valve RC-317A (Section M4.2).

50-382/9905-03 NCV Technical Specification required surveillance for Section XI testing exceeded (Section E4.2).

50-382/99-003 LER Missed Section XI surveillance on containment vacuum relief valves (Section E8.1)

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LIST OF ACRONYMS USED ASME American Society of Mechanical Engineers CEA control element assembly CFR Code of Federal Regulations CPC core protection calculator CR condition report EDG emergency diesel generator EFW emergency feedwater gpm gallons per minute HP health physics IST inservice testing LER licensee event report LPSI low-pressure safety injection NRC Nuclear Regulatory Commission PDR Public Document Room j RCP reactor coolant pump RCS reactor coolant system RFO refueling outage RTD resistance temperature device SUPS static uninterruptible power supply

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