IR 05000382/1999020

Insp Rept 50-382/99-20 on 990815-0925.Non-cited Violations Noted.Major Areas Inspected:Operations,Maintenance, Engineering & Plant Support Activities

ML20217L050
Person / Time
Site:	Waterford
Issue date:	10/21/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20217L047	List: IR 05000382/1999020 ML20217L042
References
50-382-99-20, NUDOCS 9910270055
Download: ML20217L050 (20)

Text

ENCLOSURE i

U.S. NUCLEAP REGULATORY COMMISSION l REGION IV l Docket No.: 50-382 License No.: NPF-38 Report No.: 50-382/99-20 Licensee: Entergy Operations, In Facility: Waterford Steam Electric Station, Unit 3 j Location: Hwy.18 Killona, Louisiana

Dates: August 15 through September 25,1h99 1 Inspectors: .T. R. Farnholtz, Senior Resident inspector J. M. Keeton, Resident inspector Approved By: P. H. Harrell, Chief, Project Branch D l

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ATTACHMENT: Supplemental Information l

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9910270055 991021 PDR ADOCK 05000302 0 PDR

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I EXECUTIVE SUMMARY Waterford Steam Electric Station, Unit 3 NRC Inspection Report 50-382/99-20 This routine, announced inspection included aspects of operations, maintenance, engineering, and plant support activities. The report covers a 6-week period of resident inspecdo Operations a . Operators took the appropriate actions to initiate a manucl reactor trip when they had indications of a failed reactor coolant pump seal. Plant systems responded to the trip, as expected (Section 04.1).

- Operations personnel performed a reactor coolam Jystem draindown in a well-controlled manner. Operators demonstrated an increased awareness of plant cooldown rates and limitations and took appropriate actions when a Technical Specification out-of-tolerance condition was identified. Appropriate actions were taken when reactor coolant system level instrumentation failed to function, as required. However, operators did not maintain an appropriate awareness of the time-to-boil and time-to-core-uncovery while j the plant was shutdown &nd in a reduced inventory condition (Section 04.2).

• Operators performed in a professional manner during a plant heatup and subsequent cooldown (Section O4.3).

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• The failure to test several charging system relay contacts is a violation of Technical 1

- Specification requirements. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy (Section 08.3).  ;

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• - Failure to provide appropriate procedures for the replacement of contrui board switch j knobs is a violation of 10 CFR Part 50, Appendix B, Criterion V. Since this issue was

! being addressed in the licensee's corrective action program, 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 Condition Report 99-0722 (f. action M2.1). j

• On three occasions, plant equipment malfunctions resulted in plant transients or negatively impacted plant activities. These problems were examples of degraded plant material condition and equipment. reliability. Housekeeping and material condition inside the containment building was adequate (Section M2.2).

• Electrical maintenance technicians, performing activities to overhaul a 6.9-kV circuit breaker, demonstrated a high degree of knowledge with regard to the equipment and procedures used. The breaker condition was good (Section M4.1).

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The Significant Event Response Team, assembled to identify the causes and recommend corrective actions for the failed reactor coolant pump heat exchanger baffle, was effective. The team's efforts were comprehensive (Section E2.1).

Plant Sucoort

The emergency planning staff identified several problems during the performance of an emergency preparedness drill. The identified problems were appropriately critiquad and training was planned to correct the problems (Section P4.1).

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Report Details Summary of Plant Status At the beginni7g of this inspection period, the plant was operating at 100 percent power. On September 10,1999, operators initiated a manual reactor trip in response to indications of a Reactor Coolant Pump (RCP) 28 seal failure. Following completion of repairs to the pump, operators began to heat up the reactor coolant system (RCS). During surveillance testing with l the plant in Mode 3, operators de%rmined that High Pressure Safety injection Check Valve SI-5128 failed to reseat. The plant was cooled down to repair the valve. At the end of this inspection period, the plant was in Mode . Operations 01 Conduct of Operations (71707)

0 General Comments (71707)

1 s l The inspectors performed frequent reviews and observations of ongoing plant i

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operations, control panel walkdowns, and plant tours. Observed activities were performed in a manner consistent with safe operation of the facility. The inspectors observed operators utilize good self- and peer-checking techniques when manipulating plant equipment. Operators generally used good communication technique !

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04 Operator Knowledge and Performance {

l 04.1 Manual F.cactor Trio Due to Indications of RCP 2B Seal Failure Inspection Scope (71707. 93702)

The inspectors observed operator actions and reviewed plant response following a manual reactor tri Observations and Findinas On September 10,1999, at 7:58 p.m., control room operators initiated a manual reactor trip when indicatiorm of a seal failure on RCP 2B were noted. The indications of this event were similar to an event which took place on August 1,1999 (see NRC Inspection Report 50-382/99-16). Control room operators observed low seal pressures and low controlled bleedoff flow, along with increasing controlled bleedoff temperatures associated with RCP 28. Just prior to these indications, a valve and loose parts l monitoring system alarm was received in the control room, which indicated a potential 1 problem with RCP 28. Based on these indications, the operators initiated a manual l

reactor trip. The inspectors considered this action to be appropriat Plant systems performed as expected following the reactor trip. The operators entered Operations Procedure OP-902-001," Reactor Trip Recovery," Revision 8. The

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i-2-emergency feedwater system automatically initiated but did not feed the steam generators since the main feedwater system continued to operate. Unnecessary steam loads were secured in a timely manner to control RCS temperatur The inspectors noted that the operators twice attempted to place the auxiliary feedwater (AFW) pump, a nonsafety-related pump, in service to allow the steam-driven main feedwater pump to be secured. Procedure OP-003-003," Condensate-Feedwater,"

Revision 14, was being used to perform this task. The AFW pump failed to start on the

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first attempt and investigation revealed no indication of fault. A second attempt was made and again the pump failed to start. Operators discovered a high discharge pressure trip flag on the breaker. Operators reduced the feedwater header pressure and successfully started the AFW pump on the third attempt. The operating main feedwater pump was then secure Conclusions Operators took the appropriate actions to initiate a manual reactor trip when they had indications of a failed reactor coolant pump seal. Plant systems responded to the trip, as expecte O4.2 RCS Draindown for RCP Maintenance Inspection Scope (71707. 92901)

On September 12 and 13,1999, the inspectors observed portions of the RCS draindown, which was performed to place the plant in a condition to allow evaluation and repair of the RCP 28 seal. The draindown was conducted in accordance with Procedure OP-001-003, " Reactor Coolant System Drain Down," Revision 1 Observations and Findinos To determine the extent of necessary repairs, the plant was placed in Mode 5 and the RCS water level was drained to a level below the RCP seal. The operators maintained l RCS level between 13.38 and 14.8 feet. A dedicated reactor operator was assigned to monitor the shutdown cooling system during the draindown and while the RCS was in reduced inventory conditions.. Communications and peer-checking were excellent. The

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reactor operators performed the evolution in a well-controlled manne The inspectors observed a pre-evolution briefing conducted by the control room i supervisor. The briefing was conducted in a professional manner. The roles and I

responsibilities of individual operators were clearly designated. However, time-to-boil and tim 6-to-core-uncovery information had not been calculated and was not available during the briefing. This information was calculated and disseminated to the operations shift after the briefing was completed. Due to the brief period of time since the reactor had been shut down, the operators would have a very limited amount of time to react in the case of a loss of shutdown cooling. The inspectors determined that the  ;

pre-evolution brief should have includec' a thorough discussion of these times. The l

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failure to calculate these times prior to, and discuss them during, the briefing indicated a lack of attention to details on the part of operators for the importance of this informatio While the plant was in a reduced inventory condition, the inspectors questioned several shift superintendents and control room supervisors as to the current time-to-boil and time-to-core-uncovery information. The requested information was not immediately available and had to be locked up or calculated. The inspectors determined that plant operators demonstrated a lack of awareness of these times. The inspectors expressed this concern to plant management and subsequently observed that this information became more readily available and was routinely discussed at morning meeting j i

During the RCS draindown conducted on September 13, operators reduced shutdown cooling system flow rates from approximately 4000 to 2000 gpm in each train, as required by procedure. Reducing flow through the shutdown cooling system heat exchangers caused a 10 to 15"F reduction in the temperature of the reactor coolant at the outlet of the heat exchangers. RCS cold leg temperature instruments indicated a 9 to 15'F reduction in temperature when flow was reduced. The licensee identified that the cooldown rate exceeded the limits specified in Technical Specification (TS) 3.4.8.1,

" Pressure / Temperature Limits Reactor Coolant System." TS 3.4.8.1.d limits the maximum RCS cooldown rate te 10'F per hour when RCS cold leg temperature is less than 135'F. Operators entered the applicabb TS action statement and restored temperature to within the limit within 30 minutes, in addition, the licensee performed an engineering evaluation to determine the effects of ine cooldown rate on the structural integrity of the RC The inspectors reviewed the engineering evaluation and concluded that it was adequat The evaluation concluded that there were no adverse effects on any component. In addition, the inspectors concluded that the operators demonstrated an awareness of plant cooldown rates and limitations that was not evident in previous cooldown evolutions, as documented in NRC Inspection Report 50-382/99-16. The licensee was evaluating alternatives to prevent exceeding the TS allowed cooldown rate during future evolution During the draindown, the RCS water level indicators fell outside the required 0.25 feet channel check. T.a of the indicators in the refueling water level indicating system indicated an RCS level that differed from the level indicated by the level sightglass and the reactor coolant level monitoring system indicator by approximately 0.4 feet. The plant operators stopped the draindown to determine the cause of the disagreement. A calibration check and venting of the indicating system instrumentation was performe No problems were identified with this system. The sensing line for the monitoring system was flushed and all indications returned to within the required tolerance. The inspectors concluded that the actions by operations personnel to identify the level indication discrepancy and stop the drain down were appropriat _ _

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i Conclusions Operations personnel performed a RCS draindown in a well-controlled manne Operators demonstrated an increased awareness of plant cooldown rates and limitations and took appropriate actions when a out-of-tolerance condition was identifie Appropriate actions were taken when RCS level instrumentation failed to function, as required. However, operators did not maintain an appropriate awareness of the time-to-

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boil and time-to-core-uncovery information while the plant was shut down and in a reduced inventory conditio .3 Plant Heatuo Followino Fjeoairs to RCP 2B Inspection Scope (71707. 92901)

The inspectors observed portions of the plant heatup and mode changes in preparation for plant startup. The inspectors verified that operator actions were in accordance with l appropriate procedures and T Observations and Findinas

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The inspectors observed operators in the control room during portions of the plant heatup and mode changes. The operators were observed to use the procedures in accordance with management expectations. Operators utilized self- and peer-checking -

trchniques when manipulating control board switches. Three-part communication

tecaniques were consistently used in the contaol room and with operators in the field.

l Control room access was appropriately contrciled to reduce congestio '

While performine surveillances in Mode 3 on September 24, High Pressure Safety injection Hot Leg injection Check Valve SI-512B failed to rer. eat. Backflow through the valve was found to be approximately 120 gpm. The operators commenced cooling down and deoressurizing the RCS to Mode 5 to perform the necessary repairs. The plant was in Mode 5 at the end of this inspection perio Conclusions Operators performed in a professional manner during a plant heatup and subsequent cooldow Miscellaneous Operations issues (92901)

08.1 (Closed) Violation (VIO) 50-382/9726-01013 (EA 97-589A): Failure to meet action requirements of TS 3. (Closed) VIO 50-382/9726-01023 (EA 97-589B): Failure to comply with procedures for monitoring plant activities.

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l-5- l (Closed) VIO 50-382/9726-01033 (EA 97-589C): Failure for comply with procedures for supervisory oversight of operator activitie (Closed) VIO 50-382/9726-01043 (EA 97-589D): Failure to comply with procedure for installing danger tag l (Closed) VIO 50-382/9726-01053 (EA 97-589E): Failure to follow procedure for operation of auxiliary component cooling water syste (Closed) VIO 50-382/9726-01063 (EA 97-589F): Failure to follow procedures for control panel waNdown (Closed) Licensee Event Report (LER) 50-382/97-027 and 97-027-01: Controller for Auxiliary Component Cooling Water Valve ACC-126A Left in Manual The inspectors verified the corrective actions described in the licensee's response letter, dated March 9,1998, to be reasonable and complete. No new issues were revealed by LERs 50-382/97-027 or 50-382/97-027-01.

08.2 (Closed) LER 50-382/97-028: Emergency Diesel Generator A Autostart Due to 86A2 Relay Trip On July 20,1997, a loss of power to Safety Bus 3A caused an automatic start of Emergency Diesei Ge.urator A. The loss of power was caused by a spurious trip of Relay 86A2. The cause of the relay trip could not be determined. Relay 86A2 was replaced and the electrical distribution system was returned to normal alignment.

08.3 (Closed) LER 50-382/97-029: Inadequate Surveillance Test of Relay Contacts in Safety-Related Logic Circuits On November 13,1997, licensee engineers discovered several engineered safety features actuation c.gnal relay contacts associated with the charging pumps and boron flow paths that had not been tested, as required by TS. Previous reviews conducted by the licensee in response to Generic Letter 96-01, " Testing of Safety-Related Logic Circuits," failed to identify that these contacts were not being tested. The failure to test these contacts rendered all three charging pumps inoperable. Operators entered TS 3.0.3 and invoked TS 4.0.3 to perform the missed surveillances. The contacts were tested and found to function, as require The licensee revised the surveillance procedures to incorporate testing of the contact Additional reviews were conducted to verify that all positions of those relay circuits were appropriately tested. The inspectors verified that the licensee completed the corrective actions described in the LE The failure to test these relay contacts was a violation of TS 4.3.2.1. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy (50-382/9920-01) .

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.08.4 (Closed) VIO 50-382/9727-01: Failure to initiate a condition repor : The inspectors verified the corrective actions described in the licensee's response letter, dated February 2,1998,' to be reasonable and complete.

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M1 . Conduct of Maintenance (61T26,62707)

The inspectors observed all or portions of the following maintenance and surveillance activities,'as specified by the referenced maintenance action items (mal) and surveillance procedures:

OP-903-094 Surveillance Test ESFAS Subgroup Relay Test-Operating j

• 405347 Overhaul of 6.9-kV Magna-Blast Breaker- I l

l * _ OP-903-068 Emergency Diesel Generator B Surveillance Run

• 409506 Repair of Reactor Coolant Pump 2B  !

l In general, the work activities were performed in an acceptable and effective manner.

l The technicians were knowledgeable and conducted the work as required by applicable

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procedures., Appropriate support personnel, including health physics, quality control, supervisory, and system engineering personnel were at the work site when require ;

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I- _M2- Maintenance and Material Condition of Facilities and Equipment M2.1 Control Switch Knob Failures ] insoection Scope (62707. 92902. 37551)

The inspectors noted an increasing trend in control switch knob failures on the main control panels. The inspectors reviewed historical data, related industry experience, and the licensee's activities to resolve the knob failure b'. ' Observations and Findinas The inspectors noted an increase in the nur Der of control switch knobs that had broken

- on the main control board. This type of switch operating knob was used extensively on ;

safety-related systems on the control board, including motor controls, valve operators, and breaker controls. The most common failure of the plastic knobs appeared to be

cracking and crumbling during manipulation by licensed operators. None of the knob fai. lures had prevented the associated switch from being oper;ted to perform its intended function following knob failure; however, in most cases the broken knob was

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l more difficult to grasp while performing the manipulation. The inspectors were concerned that a knob could fail in such a way as to prevent the associated switch from being operated, particularly during an accident scenari The licensee had not established contingency plans or staged tools to operate a broken switch during an emergency situation. In addition, the licensee had not written a condition report to enter the failures into their corrective action program. The broken knobs were being casually replaced when parts and time were availabl Based on the inspector's concerns, the licensee generated Condition Report 99-0722 to

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address the switch knob failures and to place the issue in the corrective action progra A review of the failed knobs indicated that, since 1993,20 of the 642 knobs in the j control room had failed. Five of those failures occurred in 1998 and seven failures occurred in the first half of 1999. At the end of this inspection period, the licensee was reviewing corrective actions to address the failures of the switch knobs. The licensee also established contingency plans to operate a switch with a broken kno Two of the failed knobs associated with safety-related equipment that had been l replaced were found to function incorrectly after the replacement activities. The control l switch knob for RFR0002B, Train B refrigeration compressor (essential chiller), had l l been broken and replaced in early February 1999. Subsequent to the replacement, an operator inadvertently pressed the switch, resulting in tripping of the compressor. Prior to the knob replacement, this switch did not have this push-to-trip feature, nor was it l designed to have such a feature. Caution Tag 99-012-1 was placed on the switch on February 12,1999, to alert opera * ors to this problem. Corrective maintenance to return I the switch to its proper operation was performed on July 29, in accordance with )

l MAI 401460 on July 29, and the caution tag was removed. This condition had existed 167 day .

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The second safety-related control switch knob was for the Emergency Diesel l Generator A 4-kV output breaker that had been replaced on April 20,1999. If the knob !

had been pressed, it would have caused the breaker to open. This switch did not have j this push-to-open feature prior to the knob replacement and it was not designed to 1 function in this manner. Caution Tag 99-050-1 was placed on the switch to alert l operators of this condition. The knob was replaced in accordance with MAI 402053, on August 11, to return the switch to its proper operation. The caution tag was also removed on August 11 after being in effect for 83 day '

Vendor information on knob replacement was very sketchy. Because application of these switches was so versatile, the vendor appeared to have relied on the users to develop the replacement guidance based on eacn application. A review of the two MAls r previously mentioned proviaed inadequate detail in that neither had instructions on use l

of spacers and specific knob replacement descriptions. Also, the inspectors noted no evidence that the switch knobs had received a design review to ensure that the methods used during the knob replacement activities were appropriate or that the materials application was suitable for safety systems in response to the increased failure rate.

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, I l-8-The requirements of 10 CFR Part 50, Appendix B,. Criterion V, states, in part, that activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstances. The failure to provide documented instructions for the replacement of the control board knobs is a violation of 10 CFR Part 50, Appendix B, Criterion V. Because this issue was being addressed in the licencee's corrective action

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program, 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 Condition Report 99-072 ! Conclusions l

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Failure to provide appropriate procedures for the replacement of control board switch knobs is a violation of 10 CFR Part 50, Appendix B, Criterion V. Since this issue was being addressed in the licensee's corrective action program, 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 Condition Report 99-072 I M2.2 Eauipment Availability and Material Condition I Inspection Scoce (71707. 62707. 92902)

The inspectors assessed the material condition of plant equipment by conducting plant tours, including the containment building, ant reviewing the impact of equipment failures on plant operation Observations and Findinas During this inspection period, the inspectors observed three instances in which plant operations or activities were adversely impacted by equipment problems. As previously discussed in Section O4.1 this inspection report, the seal water heat exchanger baffle for RCP 2B failed after being in service for less than 5 weeks. The failures of the baffle resulted in the initiation of manual reactor trips on August 1 and September 10,199 The inspectors were concerned that the root cause of the first failure was not adequately identdied and correcte During the RCS draindown, operators were forced to stop the evolution when the RCS water level indicating systems did not agree to within the required tolerance. Details of this example are contained in Section O4.2 of this report. These instruments are required to function properly to allow operators to remain cognizant of actual RCS water level during the entire draindown procedur During repair activities on RCP 2B, it was necessary to move large pieces of equipment and materialinto and out of the containment building through the equipment hatc During the outage, the equipment hatch operating gear mechanism failed to function. A temporary rigging apparatus was used to move the hatch open and close Maintenance personnel were required to close the hatch to establish containment l

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. integrity within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of a loss of shutdown cooling event or other accident. The hatch was cycled open and closed using the temporary rigging apparatus to ensure that this time requirement could be met. A time of 20 minutes was accomplished during this activity. The inspectors considered this to be adequate; however, this represented another example of installed plant equipment that failed to properly operat The inspectors concluded that these problems were examples of degraded plant -

material condition and equipment reliability. The failure of this equipment did not pose a threat to personnel or nuclear safety, but did impact plant operations or activitie The inspectors toured the containment building during the forced outage to assess the overall condition and the progress of work on RCP 28. The inspectors noted a pool of oil on the top of the D-ring wall near RCP 18. The source of the oil appeared to be an overhead crane operating mechanism. The licensee cleaned up the oil. Also, several areas of peeling paint were noted on the containment dome. Scaffold, rigging, tools, and other equipment were present throughout the building, but appeared to be staged and stored appropriately such that it would be removed at the completion of wor Numerous lighting fixtures were noted as not functioning. This resulted in some areas of the containment building being poorly lighted. The overall condition of the containment building was considered adequate with some areas in need of cleanin, or repai Conclusions l On three occasions, plant equipment malfunctions resulted in plant transients or negatively impacted plant activities. These problems were examples of degraded plant material condition and equipment reliability. Housekeeping and material condition inside the containment building was adequat M4 Maintenance Staff Knowledge and Performance M4.1 Overhaul of Maana-Blast 6.9-kV Breaker Inspection Scope (62707)

The inspectors observed electrical maintenance technicians perform a complete overhaul of a Magna-Blast 6.9-kV circuit breaker. The work was performed using MAI 405347.

j Observations and Findinas Electrical maintenance technicians performed a complete overhaul of a 6.9-kV circuit breaker as part of an ongoing program to refurbish all such breakers in the plant. The overhaul was performed in accordance with Maintenance Procedures ME-004-115,

"4.16/6.9 kV Magna-Blast Operating Mechanism Overhaul," Revision 0, and ME-004-131,"4.16 kV G.E. Magna-Blast Breaker," Revision 1 A

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equipment and procedures used. To accomplish the task, the two applicable procedures were used concurrently. The qualified maintenance technicians appeared to be familiar with the procedural requirements, but it was not clear to the inspectors as to how the two procedures were designed to be used together. Certain procedural steps i

were contained in one procedure while others were contained in the other procedur The inspectors considered this to be a complicated and confusing method, but also observed that the process seemed to work when accomplished by trained and knowledgeable maintenance technician The breaker was disassembled and the parts were maintained in an orderly manner for future reassembly. A thorough inspection of the removed parts was performed. The inspector considered the overall condition of the breaker to be good. No concerns with the electrical maintenance technicians' ability to perform these tasks were identifie Conclusions Electrical maintenance technicians, performing activities to overhaul a 6.9-kV circuit breaker, demonstrated a high degree of knowledge with regard to the equipment and procedures used. The breaker condition was goo M8 Miscellaneous Maintenance issues (92902)

M8.1 (Closed) VIO 50-382/9722-01: Failure to properly store equipmen The inspectors verified the corrective actions described in the licensee's response letter, dated December 18,1997, to be reasonable and complete. The inspectors concluded that this violation had been appropriately addresse Ill: Enaineerina E2 Engineering Support of Facilities and Equipment E ELCP 2B Seal Water Heat Exchanaer Baffle Failure Inspection Scope (37551. 92903)

The inspectors observed engineering support for tl.e failure of the RCP 2B seal water heat exchanger baffle and associated components that resulted in the forced shutdown and cooldown of the plant to make repair Observations and Findinas Following the manual reactor trip on September 10,1999, the licensee assembled a Significant Event Response Team (SFRT) to investigate and make recommendations to

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t correct the cause of the baffle failure on RCP 2B. The SERT obtained and reviewed all

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-11-i available data to establish operating parameters and component histories. The originai I equipment manufacturer and other industry technical personnel were called in to assist I in the identif: cation of the root cause of the failure and corrective actions to be take The plant was placed in a condition to allow disassembly of tne RCP seal asserably and j heat exchanger batfle. The licensee suspected that the baffle or baffle bolts had failed, '

as had occurred on previous occasions. These previous failures are documented :n NRC Inspection Report 5438?)99-16. When these components were disassembled, a 360 degree through-wall crack was discovered around the upper portion of the baffl This caused the baffle to break into two parts. This baffle had been in service for l approximately 5 weeks. The six baffle bolts were found to be intact and the as-found preload was within tolerance. The seal water heat exchanger was found to be slightly i scored with most of this scoring caused by previous failures of the baffle. The heat

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exchanger was t5 eWmined to be acceptable for continued servic The one-piece seal water heat exchanger baffle consisted of an inner and an outer cylinder connected at the top by a wah. The failed baffle was examined and it was determined that the crack initiated at the outside diameter of the cylinder, at the bottom !

of the transition radius to the wall, which connects the inner and outer cylinders. The fracture sudace was determined to be typical of a fatigue failure. The cause of the ;

failure was identified as marginal metal thickness of the baffle at the failure locatio The inertial mass of the water in the heat exchanger baffle was identified as exerting ;

significant forces on the baffle that had not been previously considered. A finite element analysis considering these forces predicted failure of this component in the location of i the actual failures in both the August 1 and Septernber 10 events. In addition, it was identified that the natural frequency of the one-piece baffle was approximately 108 Hertz. The vane passing frequency for the pump was approximately 100 Hert The SERT considered these two frequencies to be very close and a significant contributor to baffle loading due to the resulting stress amplification facto Two other causes were identified as contributing to the failure. The RCP shaft shoulder, to which the baffle was bolted, was determined to have an incline across the shaft diameter. This inclined surface resulted in an imbalance by putting the center of gravity of the batfie off center frcm the center of rotation. Surface irregularities were identified in the shoulder, which introduced higher mean stresses in the baffle during the bolt torque process. In addition, the motor shaft to pump shaft alignment was determined to be outside of original equipment manufacturer specifications. M;salignment increases vibration levels and creates higher cyclic forces on pump and rnotcr components. The misaUgnment was considered a less significant contributor to the forces acting on the baffle and was not corrected during this maintenance outag The pump was reassembled using a new two-piece baffle in place of the f ailed one-piece unit. The geometry of the two-piece baffle at the location of the previous failures had more than twice the thickness of the one-piece baffle. Also, the natural frequency of the two-piece baffle was 119 Hertz. The difference between this and the vane passing frequency for the pump was consicered sufficient to eliminate the concern

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of the stress amplification factor. The shaft shoulder was machined so as to achieve a surface that was perpendicular to the shaft centerline and also to eliminate the surface irregularitie The inspectors reviewed the SERT report and considered it to be comprehensive. The team considered forces and stresses on the baffle that had previously not been considered. The recommendations made to correct the condition were based on analysis and reasonable engineering judgemen ; Conclusions The SERT, assembled to identify the causes and recommend corrective actions for the failed reactor coolant pump heat exchanger baffle, was effective. The team's efforts ,

were comprehensiv I E8 Miscellaneous Engineering Issues (92903)

E (Closed) VIO 50-382/9722-02: Failure to perform inservice test baselinin l The inspectors verified the corrective actions described in the licensee's response letter, dated December 18,1997, to be. reasonable and complete. All corrective actions have been complete E8.2 (Closed) LER 50-382/97-026: Single-Failure Effects on Condensate Storage Pool Inventor On November 5,1997, licensee engineers determined that the potential existed for a single failure in any one of seven component cooling water (CCW) makeup control valves and associated surge tank level switches to cause the CCW makeup system to continuously makeup to their respective tanks. This single failure could result in flooding in the affected reactor auxiliary building areas. Also, because the CCW makeup pumps take a suction from the condensate storage pool, which is the source of water for the emergency feedwater pumps, the amount of emergency feedwater available for design-basis accidents could be reduce The root cause of this condition was attributed to an original design deficiency. The original design had net considered that the CCW system was the makeup source of water for three safety systems. Therefore, the original design had not considered the effects of single failur An engineering evaluation determined that the safety impact on risk significant systems was minimal. Procedure changes were completed to ensure early operator action to monitor and manipulate the CCW makeup system upon automatic initiation or component failur IV. Plant Support R1 Radiological Protection and Chemistry Controls During routine tours, the inspectors observed posted radiation surveys, which were required by licensee procedures and NRC regulations. A sample of doors were found to be locked as required for the purpose of radiation protection. Licensee personnel

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working in radiologically controlled areas were observed following applicable procedures for radiation protection.

P4 Staff Knowledge and Performance in Emergency Preparedness (EP)

P EP Dress Rehearsal Exercise Inspection Scope (71750. 92904)

On August 18,1999, the licensee conducted a dress rehearsal EP exercise in preparation for the graded site-wide drill planned for October 1999. The inspectors observed portions of the exercise in the control room simulator, Technical Support Center, and the Emergency Operations Facilit Observations and Findinas The EP scenario had been designed to exercise each of the facility functions and allow for coaching of the participants. The scenario involved communications problems and resolutions, security accountability issues, radiological controls, medical emergencies, and augmentatio The inspectors reviewed the critique notes from each of the facilities and found that the licensee had identified problems that paralled the inspectors' observations. These observations included: (1) participants appeared to take a casual attitude during the drill, (2) classifications of events, although accurate, were not timely, (3) some communications among the facilities were incomplete, and (4) activation of the facilities was not timel The EP staff indicated that the problems identified during the exercise will be the focus of training prior to the graded exercise scheduled for October 1999, Conclusions )

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The emergency planning etaff identified several problems during the performance of an )

emergency preparedness drill. The identified problems were appropriately critiqued and !

training was planned to correct the problem I

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J l-14-V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management on September 29,1999. The licensee acknowledged the findings presente !

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

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-ATTACHMENT

. SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee

R. F. Burski, Director Site Support J. R. Douet, Manager, Plant Maintenance C. M. Dugger, Vice-President, Operations E. C. Ewing, Director, Nuclear Safety & Regulatory Affairs C. Fugate, Operations Superintendent J. G. Hoffpauir, Manager, Operations T. R. Leonard, General Manager, Plant Operations i T. P. Lett, Superintendent, Radiation Protection E. P. Perkins, Jr., Manager, Licensing

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L. N. Rushing, Manager, Mechanical and Civil Engineering B. E. Thigpen, Manager, Planning and Scheduling A. J. Wrape, Director, Design Engineering

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i INSPECTION PROCEDURES USED 37551 Onsite Engineering 61726 Surveil lance 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 93702- Prompt Ons!te Response to Events

-2-ITEMS OPENED. CLOSED. AND DISCUSSED Opened 50-382/9920-01 NCV inadequate Surveillance Test of Relay Contacts in Safety-Related Logic Circuits (Section 08.3).

50-382/9920-02 NCV Failure to Provide Appropriate Procedures for Replacement of Control Board Switch Knobs (Section M2.1).

Closed 50-382/01013 VIO Failure to meet action requirements of TS 3. (EA 97-589A) (Section 08.1).

50-382/01023 VIO Failure to comply with procedures for monitoring plant (EA 97-5898) activities (Section 08.1).

50-382/01033 VIO Failure tor comply with procedures for supervisory oversight (EA 97-589C) of operator activities (Section O8.1).

50-382/01043 VIO Failure to comply with procedure for installing danger tags (EA 97-589D) (Section O8.1).

50-382/01053 VIO Failure to follow procedure for operation of auxiliary (EA 97-589E) component cooling water system (Section 08.1). {

50-382/01063 VIO Failure to foilow procedures for control panel walkdowns i (EA 97-589F) (Section 08.1).

50-382/97-027 LER Controller for Auxiliary Component Cooling Water Valve ,

and 97-027-01 ACC-126A Left in Manual (Section 08.1). l 50-382/97-028 LER Emergency Diesel Generator A Autostart Due to 86A2 Relay Trip (Section 08.2).

50-382/97-029 LER Inadequate Surveillance Test of Relay Contacts in Safety-Related Logic Circuits (Section 08.3).

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59-382/9920-01 NCV Inadequate Surveillance Test of Relay Contacts in Safety-Related Logic Circuits (Section 08.3).

50-382/9727-01 VIO Failure to initiate a condition report (Section 08.4)

50-382/9920-02 NCV Failure to Provide Appropriate Procedures for Replacement of Control Board Switch Knobs (Section M21).

50-382/9722-01 VIO Failure to pmperly store equipment (Section M8.1).

50-382/9722-02 VIO Failure to perform inservice test baselining (Section E8.1).

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h -3-50-382/97-026 LER Single Fai!ure Effects on Condensate Storage Pool Inventory (Section E8.2).

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Discussed l

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LIST OF ACRONYMS USED AFW auxiliary feedwater CFR Code of Federal Regulations CCW component cooling water

'EA enforcement action

~EP emergency preparedness LER licensee event report mal maintenance action item NCV noncited violation NRC Nuclear Regulatory Commission PDR Public Document Room RCP reactor coolant pump  !

RCS reactor coolant system

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. SERT Significant Event Response Team TS Technical Specification l

.VIO violation I i

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