IR 05000454/1999004
| ML20207G053 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 06/04/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20207G047 | List: |
| References | |
| 50-454-99-04, 50-454-99-4, 50-455-99-04, 50-455-99-4, NUDOCS 9906110006 | |
| Download: ML20207G053 (32) | |
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l U. S. NUCLEAR REGULATORY COMMISSION REGION lll
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Docket Nos: 50-454;50-455 License Nos: NPF-37; NPF-66 Report No: 50-454/455-99004(DRP)
Licensee: Commonwealth Edison Company Facility: Byron Generating Station, Units 1 and 2 Location: 4450 N. German Church Road Byron,IL 61010 Dates: March 30 - May 10,1999 Inspectors: E. Cobey, Senior Resident inspector B. Kemker, Resident inspector l T. Tongue, Project Engineer, Rlli C. Phillips, Braidwood Senior Resident inspector J. Adams, Braidwood Resident inspector D. Pelton, Braidwood Resident inspector C. Thompson, Illinois Department of Nuclear Safety Approved by: Michael J. Jordan, Chief
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Reactor Projects Branch 3 Division of Reactor Projects
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9906110006 990604 "
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EXECUTIVE SUMMARY Byron Generating Station Units 1 and 2 NRC Inspection Report 50-454/99004(DRP): 50-455/99004(DRP)
This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a six-week period of inspection activities by the resident staff and region based inspector Operations
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The inspectors concluded that operations of the facility were conducted in a safe, professional, and controlled manner. Operators generally adhered to the station's standards for controi room conduct, procedural adherence, annunciator response, and ;
use of three-way communications. The inspectors concluded that the licensee's response to a tube leak on the 21 A drain cooler was excellent. The licensee appropriately assessed and controlled plant conditions to effect repairs and return Unit 2 l to full power. (Section 01.1)
. The inspectors concluded that the Unit i reactor startup following refueling outage B1R09 was conducted in a safe and controlled manner. Specifically, just-in-time simulator training was performed in preparation for the startup, the heightened level of awareness briefing was performed well, startup and physics testing procedures were followed, proper three-way communication techniques were used, peer and self-checks were performed, and reactivity manipulations were precisely controlled. The inspectors concluded that operators demonstrated a good safety focus while addressing emergent problems in that operators promptly identified the unexpected conditions, stabilized plant conditions, approached the problems by identifying conservative corrective actions, and implemented those actions in accordance with approved procedures. The inspectors also concluded that supervisors demonstrated good command and control, directly supervised reactivity manipulations, and provided effective oversight of the startup activities. (Section O1.2)
. The inspectors concluded that the licensee's response to the failure of the Unit i refueling machine hoist with an irradiated fuel assembly suspended from the machine's grapple was excellent. The licensee controlled and accomplished the task of lowering the fuel assembly to a designated core location in accordance with approved procedures and with a proper focus on safety. The inspectors also concluded that the 10 CFR 50.59 safety eva!uation, which was approved by the Plant Operations Review Committee to lower the suspended fuel assembly, failed to address the licensing basis for movements of heavy loads over the reactor vessel as it pertained to the evolution; failed to address the effects of bypassing refueling machine interlocks described in the Updated Final Safety Analysis Report (UFSAR); and, failed to identify temporary changes to words in the UFSAR. A Non-Cited Violation was issued. (Section 01.3)
. The inspectors concluded that observed fuel handling evolutions were performed well and in accordance with the requirements of the Technical Specifications and station fuel handling procedures. Core alterations were strictly controlled and appropriate accountability measures were followed. (Section 01.4)
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The inspectors concluded that the licensee failed to perform an adequate technical review for the out-of-service of essential service water return header cross-connect valve (1SX011) during maintenance work to rebuild the valve's actuator. Consequently, the valve was taken out-of-service in the "open" position for the maintenance work rather than in the " closed" position as specified in Design Change Procedure 980023 A Non-Cited Violation was issued. (Section 02.1)
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The inspectors concluded that the licensee failed to control the configuration of the engineered safety features (ESF) switchgear ventilation system consistent with the design basis description in the Updated Final Safety Analysis Report (UFSAR) and that the system failure analysis described in the UFSAR was not consistent with the actual failure response of the system. Furthermore, the inspectors concluded that by not monitoring the temperature of the essential service water cooling tower electric substations, the licensee did not appropriately monitor temperatures of ESF switchgear ventilation system areas consistent with Technical Requirements Manual surveillance requirement 3.7.d.1. A Non-Cited Violation was issued. (Section O2.2)
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The inspectors concluded that the licensee failed to control the configuration of the residual heat removal system during emergency core cooling system full flow testing, which resulted in operation of the Unit 1 A safety injection pump with an inadequate suction to the pump. (Section O2.3)
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The inspectors concluded that three out-of-service errors resulted in operational events during the four week B1R09 refueling outage, which was an improvement from the B2R07 outage period when nine out-of-service errors resulted in operational event The inspectors concluded that three recent out-of-service errors had similar weaknesses previously identified by the inspectors during the crevious refueling outage (B2R07) and discussed in NRC Inspection Report 50-454/455-98011(DRP). A Non-Cited Violation was issued. (Section O2.4)
Maintenance / Surveillance
. The inspectors concluded that observed maintenance work on the 1 A reactor coolant system cold leg loop stop isolation valve was generally performed well; maintenance personnel were knowledgeable of the associated activities; and, work was performed in accordance with station procedures with one notable exception. The inspectors concluded that the licensee failed to complete a final cleanliness inspection quality control hold point prior to re-assembly of the valve which resulted in foreign material left in the reactor coolant system following the maintenance. A Non-Cited Violation was issued. (Section M1.1)
. The inspectors concluded that the observed surveillance tests were performed well and satisfied the requirements of the Technical Specifications. The inspectors identified that a step in the Unit 1 motor driven auxiliary feedwater pump monthly surveillance test procedure incorrectly stated that the procedure met surveillance test requirements that it did not meet; however, the inspectors concluded that those requirements were met by other surveillance test procedures. (Section M1.2)
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The inspectors concluded that material condition in the Unit 1 containment building prior to entry into mode 3 (hot shutdown) was good and that loose items identified by the inspectors posed no significant risk to emergency core cooling system operabilit (Section M2.1)
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The inspectors concurred with the licensee's conclusion that multiple human performance deficiencies were the cause for electrical maintenance personnel installing an electrical jumper on the wrong relay while performing surveillance testing on the 18 emergency diesel generator. This caused an inadvertent trip of the common component cooling water pump. Although this event had minimal safety impact, the inspectors were concerned that the licensee's corrective actions for similar previous human performance errors have not been totally effective at preventing their recurrence.
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(Section M4.1)
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The inspectors concluded that mechanical maintenance personnel &djueted the stem packing on the 1B main steam isolation valve (MSIV) without approved woi,'t instructions and operations department authorization for the activity, which renc'ered the valve inoperable. The inspectors concluded that operators took appropri ate actions to promptly address the operability of the 1B MSIV and correctly implemented the required actions of the Technical Specifications. A Non-Cited Violation was issue (Section M4.2)
Plant Support
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The inspectors concluded that radiologically controlled areas were properly posted and radiation workers demonstrated proper work practices to control the spread of contamination. The inspectors concluded that As-Low-As-Reasonably-Achievable principles were effectively utilized to minimize dose during the Unit i refueling outag (Section R1.1)
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Report Details Summarv of Plant Status At the start of the inspection period, Unit 1 was shutdown for refueling outage B1R0 Following completion of the refueling outage, the licensee synchronized the unit to the grid on April 25,1999, and the unit was operated at or near full power until the end of the inspection perio The licensee operated Unit 2 at or near full power for the duration of the inspection perio l. Operations 01 Conduct of Operations 01.1 General Observations Insoection Scoce (71707)
During this inspection period, the inspectors routinely observed the conduct of plant operations, with particular emphasis on Unit 1 refueling outage activities. The inspectors also monitored the licensee's actions related to the reduction of Unit 2 power for repair of the 21 A heater drain cooler. The inspectors reviewed the Limiting Conditions for Operation Action Requirements (LCOARs) during the Unit i refueling outage activities since this was the first outage since ITS implementatio Observations and Findinos
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The inspectors noted that operators generally adhered to the station's standards for 1 control room conduct, procedural adherence, annunciator response, and use of three-way communication On April 2,1999, the licensee reduced Unit 2 power level to 875 megawatts electric in response to a tube leak on the 21 A heater drain cooler. The inspectors noted that the repair effort while on-line was complex and required the use of a line stop to isolate the affected low pressure heater. The heater drain cooler's leak was repaired and Unit 2 was returned to full power on April 6,1999. The licensee appropriately assessed and controlled plant conditions to effect repairs and return the unit to full powe On February 5,1999, the licensee implemented Amendment No.106 to the station's Technical Specifications (TS). The amendment reflected the full conversion of the licensee's TS to a set of improved Technical Specifications (ITS) based on NUREG-1431, " Standard Technical Specifications - Westinghouse Plants," Revision 1, dated April 1995. The licensee identified that on April 22,1999, the requirements of TS 3.3.2 were not met with Unit 1 in mode 3 (hot standby) during reactor trip breaker testing, which resulted in entry into TS 3.0.3. The inspectors noted that prior to ITS implementation, TS 3.3.2 was applicable only in modes 1 and 2. The inspectors concurred with the licensee in that this event was not safety sigrficant and that except for this single event, the requirements of the TS were satisfied during the outag ;
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. . Conclusions The inspectors concluded that operations of the facility were conducted in a safe, professional, and controlled manner. Operators generally adhered to the station's standards for control room conduct, procedural adherence, annunciator response, and use of three-way communications. The inspectors concluded that the licensee's response to a tube leak on the 21 A drain cooler was excellent. The licensee appropriately assessed and controlled plant conditions to effect repairs and return Unit 2 to full powe .2 Unit 1 Startuo Followina Refuelino Outaae B1R09 Inspection Scope (71707)
The inspectors observed the startup of Unit i following refueling outage B1R09. The inspectors interviewed operations department personnel, observed the licensee's pre-evolution briefings and startup activities, and reviewed the following procedures:
Byron General Operating Procedure (1BGP) 100-1, " Plant Heatup," Revision 31; 1BGP 100-2, " Plant Startup," Revision 23; 1BGP 100-2A1, " Reactor Startup" Revision 15; 1BGP 100-3, " Power Ascension," Revision 27; and Byron Engineering Surveillance Requirement (BVSR) Procedure XPT-23, " Unit 1 Low Power Physics Test Program with the Westinghouse ADRC [ Advanced Digital Reactivity Computer)," Revision Observations and Findinas On April 24,1999, the inspectors observed the heightened level of awareness briefing for the Unit 1 startup and noted that all necessary personnel were in attendance. The inspectors noted that the briefing contained a sufficient level of detail to properly perform the startup. The inspectors also noted that in preparation for the startup, operations and nuclear engineering department personnel had received just-in-time simulator trainin The inspectors observed the operators performing startup activities in the control room, noting that procedures were followed, proper three-way communication techniques were used, and peer and self-checks were performed. The inspectors noted that reactivity manipulations associated with reactor criticality, physics testing, and power ascension were precisely controlled, were directly supervised by a senior reactor operator, and were clearly communicated to the unit supervisor and nuclear engineer The inspectors noted that operators were required to stop the startup on several occasions when challenged by equipment problems. Operators promptly identified the equipment problems and stabilized plar,t conditions. Conservative corrective actions were identified and implemented in accordance with approved procedures. Personnel were reminded to take the time necessary to implement corrective actions safely and correctl The inspectors noted that the unit supervisor demonstrated good command and contro The unit supervisor remained in the "at-the-controls" area of the control room and appeared to maintain an overall control room perspective. Additionally, the inspectors observed the presence of senior management in the control room during performance of the startu .
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1 Conclusions The inspectors concluded that the Unit i reactor startup following refueling outage B1R09 was conducted in a safe and controlled manner. Specifically,just-in-time simulator training was performed in preparation for the startup, the heightened level of awareness briefing was performed well, startup and physics testing procedures were followed, proper three-way communication techniques were used, peer and self-checks were performed, and reactivity manipulations were precisely controlled. The inspectors concluded that operators demonstrated a good safety focus while addressing emergent problems in that operators promptly identified equipment problems and stabilized plant I conditions. Conservative corrective actions were identified and implemented in accordance with approved procedures. The inspectors also concluded that supervisors demonstrated good command and control, directly supervised reactivity manipulations, and provided effective oversight of the startup activitie .3 Failure of Unit 1 Refuelina Machine With a Suspended Fuel Assembly Inspection Scope (71707)
The inspectors observed the licensee's response to the failure of the Unit 1 refueling machine hoist with an irradiated fuel assembly suspended from the machine's grappl The inspectors also observed Plant Operations Review Committee (PORC)
Meeting 99-38. The inspectors interviewed operations, maintenance, nuclear oversight, and engineering department personnel and reviewed the following procedures: Work Request (WR) 990033818, " Unit 1 Manipulator Crane Will Not Move in Upward Direction;" Special Procedure (SPP)99-021, " Lower Stuck Fuel Assembly From Refueling Machine While Applying Secondary Restraint;" SPP 99-023," Lower Stuck Fuel Assembly From Refueling Machine With Polar Crane;" and, Nuclear Station i Procedure (NSP) CC-3005, "10 CFR 50.59 Safety Evaluation Process", Revision l Observations and Findinas On April 4,1999, while transferring an irradiated fuel assembly out of the Unit i reactor, -
the refueling machine hoist failed mechanically with the assembly suspended .
approximately 15 feet above the lower core plate. The licensee was unable to further l raise or lower the assembly using the refueling machin Since the failure mechanism was not initially understood, the licensee was concerned that the assembly could possibly free fallinto the core if the hoist drum brake and gears were to suddenly fail. Therefore, the licensee installed a steel cable restraining device that provided a positive means to prevent the fuel assembly from being lowered or dropped. The inspectors reviewed the work instructions and observed the installation of the restraining device and noted that the work was performed in accordance with established station practices for performing emergent wor After installing the restraining device, the licensee developed an action plan to lower the fuel assembly into the core. Up to this point, the licensee had determined that the refueling machine hoist could not be used to further raise the fuel assembly. The licensee had not yet attempted to lower the fuel assembly. The inspectors reviewed the procedure and observed performance of SPP 99-021, which was approved for reinserting the fuel assembly into its original core location using the refueling machine
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hoist. The inspectors noted that SPP 99-021 contained instructions to use the polar crane auxiliary hook attached to the refueling machine upper cable assembly to provide a supplemental restraining device. The licensee determined that this supplemental restraining method would be necessary because the previously installed cable restraint would have to be removed to allow the refueling machine hoist to operate. The inspectors reviewed Updated Final Safety Analysis Report (UFSAR) Section 9.1.4. and noted that the refueling machine hoist travel stops and tool length limits prevent the fuel assembly from being raised to a height where less than 9 feet of water remain above the active region of the fuel. The inspectors determined that SPP 99-021 contained appropriate controls to ensure that the maximum refueling machine hoist movement limits would be maintained using the polar crane auxiliary hook. The licensee's attempt to lower the fuel assembly failed because the refueling machine hoist gears were boun Since the refueling machine was unable to lower the fuel assembly, the licensee developed an alternative plan (SPP 99-023) to lower the fuel assembly from the refueling machine using the polar crane auxiliary hook attached to the refueling machine upper cable assembly. The inspectors reviewed SPP 99-023 and noted that, as with the prwious use of the polar crane, the procedure contained appropriate controls to ensure that upward fuel assembly movement would not challenge the UFSAR assumption that 9 feet of water is maintained above the active region of the fuel. The inspectors also noted that SPP 99-023 contained details for the connection of the polar crane auxiliary hook to the refueling machine upper cable assembly using cables, slings, and other mechanical rigging hardware capable of supporting the weight of the fuel assembly and refueling machine gripper with a margin for safety. The inspectors observed raechanical maintenance personnel install the rigging equipment and cut the refueling machine's cable. The inspectors then observed fuel handling personnel lower the fuel assembly with the polar crane auxiliary hook into a designated core location chosen to maximize its distance from other irradiated fuel assemblies in the core. The inspectors determined that the licensee controlled and accomplished the work professionally and with a proper focus on safety. The inspectors noted that senior management oversW was evident throughout the event and nuclear oversight personnel were present in the containment building and provided oversight of the activitie Additionally, the inspectors observed PORC Meeting 99-38 that was convened to approve SPP 99-023 and reviewed the safety evaluation which was completed for the SPP and approved by the PORC. The inspectors identified that the safety evaluation failed to address the licensing basis for movements of heavy loads over the reactor vessel as it pertained to the evolution, failed to address the effects of bypassing refueling machine interlocks descr. bed in the UFSAR, and failed to identify temporary changes to words in the UFSAR, all of which were required by station procedure CC-3005. The inspectors discussed these concerns with the licensee immediately following the PORC meeting. In response to the inspectors' questions the licensee revised the safety evaluation to adequately address these concern Code of Federal Regulations Title 10 Part 50, Appendix B, Criteria V, " Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be i prescribed by documented instructions, procedures, or drawings of a type appropriate to !
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the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Nuclear Station Procedure CC-3005, Section 8.10.7, specified, in part, that changes to structures, systems, and components (SSCs) not l
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axplicitly described must be evaluated to determine if they can have the potential for affecting the function of SSCs which are explicitly described in the UFSAR. In addition, NSP CC-3005, Section 8.10.8, specified, in part, that all significant permanent or temporary changes are to be described. The licensee's failure to address the licensing basis for movements of heavy loads over the reactor vessel as it pertained to performance of SPP 99-023, to address the effects of bypassing refueling machine -
interlocks described in the UFSAR, and to identify temporary changes to words in the UFSAR is a violation of 10 CFR Part 50, Appendix B, Criteria V for failure to implement {
the procedure. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99004-01(DRP)).
This violation is in the licensee's corrective action program as problem identification I form (PIF) B1999-0125 ) Conclusions
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The inspectors concluded that the licensee's response to the failure of the Unit 1 refueling machine hoist with an irradiated fuel assembly suspended from the machine's grapple was excellent. The licensee controlled and accomplished the task of lowering l the fuel assembly to a designated core location in accordance with approved procedures
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and with a proper focus on safety. The inspectors also concluded that the 10 CFR 50.59 safety evaluation, which was approved by the Plant Operations Review Committee to lower the suspended fuel assembly, failed to address the licensing basis for movements of heavy loads over the reactor vessel as it pertained to the evolution; failed to address the effects of bypassing refueling machine interlocks described in the Updated Final Safety Analysis Report (UFSAR); and, failed to identify temporary changes to words in the UFSAR. A Non-Cited Violation was issue .4 Unit 1 Refuelina Activities Durina B1R09 (71707)
The inspectors interviewed fuel handlers, observed refueling activities in the containment and fuel handling buildings, and reviewed the following procedures: Byron Administrative Procedure (BAP) 370-3, " Administrative Control During Refueling,"
Revision 22; Byron Fuel Handling Procedure (BFP) FH-4, " Fuel Movement in Spent Fuel Pool," Revision 8; and, BFP FH-5, " Fuel Movement in Containment," Revision 8. The inspectors noted that all fuel handling activities were conducted in accordance with prescribed procedures and were properly supervised by a senior reactor operator licensed for fuel handhng. The inspectors noted that core alterations were strictly controlled and that appropriate accountability measures were followed. The inspectors concluded that observed fuel handling evolutions were performed well and in accordance with the requirements of the TS and station fuel handling procedure I l
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Operational Status of Facilities and Equipment i O2.1 Out-of-Service Error Resulted in a Failure to Control the Confiauration of Essential
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Service Water (SX) System Return Header Cross-Connect Valve Consistent With
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Desian Chanae Procedure Descriotion Inspection Scope (71707)
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j The inspectors reviewed the circumstances surrounding the licensee's failure to control the configuration of SX system return header cross-connect valve (1SX011) during
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maintenance work to rebuild the valve's actuator. The inspectors interviewed operations and engineering department personnel; reviewed applicable portions of the TS and the UFSAR; and reviewed the following documents: Work Request (WR) 960002689-01,
"SX Basin 1 A/1B Return Header Cross-Tie Isolation Valve (EOP [ Emergency Operating Procedure] Valve) Various Oil Leaks on Limitorque Actuator;" Design Change Procedure (DCP) 9800239, " Essential Service Water Return Header Cross-Tie isolation Valve / Remove the Valve Operator for Maintenance;" and, Root Cause Report 454-200-99-CAQ00007, " Failure to Follow Procedures Results in a Valve Motor Operator Taken Out-of-Service in Conflict with a Temporary Alteration."
b. Observations and Findinas On March 9,1999, the licensee discovered that SX return header cross-connect valve ,
(1SX011) had been taken out-of-service in the "open" position for maintenance work to rebuild the valve's actuator rather than in the " closed" position as specified in DCP 9800239. Although the out-of-service for 1SX011 was properly performed in the field, it did not specify the correct position for the valve because instructions contained in the ,
DCP were not reviewed by operators who prepared and approved the out-of-servic !
The inspectors noted that the task title and work scope provided on the out-of service request electronically submitted to the operations department did not address the DCP; however, it did state that the valve must be taken out-of-service in the " shut" positio The licensee identified the out-of-service error upon reassembly of the valve and subsequent clearing of the out-of-servic The safety evaluation performed in accordance with 10 CFR 50.59 for the DCP assumed that 1SX011 would be locked in the " closed" position with the valve's actuator removed, thereby separating both Unit 1 SX loops. In the event of a postulated failure ,
I that would require separation of the SX trains, the valve would already be closed. For other postulated failures, safe shutdown could be accomplished using either one of the two independent SX trains. The inspectors determined that the licensee's failure to have 1SX011 in the " closed" position during the maintenance evolution invalidated the safety evaluation. The inspectors reviewed SX system operability requirements with respect to 1SX011 being failed open and concluded that the SX system remained operable because the UFSAR allowed operation with the valve open or close Code of Federal Regulations Title 10 Part 50, Appendix B, Criteria V, " Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, BAP 330-1, " Station Equipment Out-Of-Service Procedure," Revision 29, was not of a type appropriate to the circumstances, in that, it did not provide for an adequate technical review of the out-of-service prepared for WR 960002689-01 to ensure that the conditions specified in the safety evaluation completed for DCP 9800239 were incorporated into the out-of-service. Consequently, the weaknesses in BAP 330-1 resulted in the SX return l header cross-connect valve (1SX011) being taken out-of-service in the "open" position i
for maintenance work to rebuild the valve's actuator rather than in the " closed" positio This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99004-02a(DRP)). This violation is in the licensee's corrective action program as problem identification form (PIF)
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. . Conclusions The inspectors concluded that the licensee failed to perform an adequate technical l review for the out-of-service of essential service water retum header cross-connect valve (1SX011) during maintenance work to rebuild the valve's actuator. Consequently, the valve was taken out-of-service in the "open" position for the maintenance work rather than in the " closed" position as specified in Design Change Procedure 980023 A Non-Cited Violation was issue O2.2 Failure to Control the Confiauration of Enaineered Safety Features (ESF) Switchaear Ventilation System Fans Resulted in System Operation inconsistent With the Desian Dailli i Insoection Scope (71707)
The inspectors reviewed the circumstances surrounding the licensee's failure to control the configuration of the SX cooling tower electric substation ventilation fans, which resulted in ESF switchgear ventilation system operation inconsistent with the design basis of the plant. The inspectors interviewed operations and engineering department personnel; reviewed applicable portions of the Technical Requirements Manual (TRM)
and the UFSAR; and reviewed the following procedures: Byron Operating Procedure (BOP) VX-Ei, "Switchgear Heat Removal Ventilation Electrical Lineup (Unit 1),"
Revision 6 and BOP VX-E2, "Switchgear Heat Removal Ventilation Electrical Lineup (Unit 2)," Revision Observations and Findinas On February 23,1999, a non-licensed operator discovered that three of the four SX cooling tower electric substation ventilation fans were not running. Operators restarted the three fans upon discovery that the fans should normally be running; although, one of the fans was subsequently secured due to a failed recirculating air damper. The licensee initially concluded that this was not a system configuration controlissue since l the fans were secured per an approved procedure and the rooms were at normal temperature The inspectors noted that UFSAR Section 9.4.5.4.2 identified the four SX cooling tower ;
electric substation ventilation fans as part of the ESF switchgear ventilation system. As l described in UFSAR Section 9.4.5.4.1.1, these ventilation fans are designed to remove equipment heat to maintain the switchgear room temperatures in accordance with equipment requirements. The UFSAR further stated that ESF switchgear ventilation equipment and systems operation are demonstrated during normal plant operation since this system operates continuously. The inspectors identified that by not controlling the configuration of the SX cooling tower electric substation ventilation fans, the licensee did l not operate the ESF switchgear ventilation system consistent with the design basis description in the UFSA The inspectors reviewed the status of ESF switchgear ventilation equipment and systems with respect to the maintenance rule requirements contained in 10 CFR 50.6 The inspectors noted that ESF switchgear ventilation equipment met all of the 10 CFR 50.65 paragraph (a)(2) reliability performance criteria and that no functional failures had been identified. The inspectors concluded that the licensee had clearly.
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defined reliability and availability criteria as well as a reasonable definition of a functional failur The inspectors identified that the system failure analysis described in UFSAR Table 9.4-16 was not consistent with the actual failure response of the system. Updated Final Safety Analysis Report Table 9.4-16 stated that when an ESF switchgear ventilation system supply fan experiences a " fan motor trip" condition, the result will be an airflow switch sensing high or low differential pressure across the fan actuating an alarm on the local control panel. However, for an actual" fan motor trip" condition, the low differential pressure alarm will not actuate since the low differential pressure alarm is only active when the fan motor is running. Also, by system design, when an ESF switchgear ventilation system supply fan experiences a " fan motor trip" condition, an alarm will actuate due to a disagreement between the remote (main control room panel 0PM02J) control switch position and an *M" contact for the fan motor's control circui The alarm will actuate only if the associated fan's local / remote transfer switch for the fan is selected to the " remote" position. The inspectors noted that the normal ESF switchgear ventilation system lineup procedures positioned each of the fan's local / remote transfer switches to the "locar position. Therefore, a " fan motor trip" condition would not produce an alarm to alert operators of a fault contrary to the design basis description in the UFSAR. At the end of this inspection period, the licensee was evaluating their actions regarding the operation of the ESF switchgear ventilation system and the resolution of apparent discrepancies with the UFSA t The inspectors additionally noted that TRM surveillance requirement 3.7.d.1 required area temperatures to be monitored for ESF switchgear ventilation system areas. The required surveillance frequency was every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the maximum temperature limit was 108 degrees Fahrenheit (*F). The inspectors identified that the licensee was not monitoring the temperature of the SX cooling tower electric substations in accordance with TRM surveillance requirement 3.7.d.1. Technical Specification 5.4.1.a states that written procedures shall be established, implemented, and maintained for procedures recommended in Appendix A, of Regulatory Guide 1.33, Revision 2, February 197 Appendix A, of Regulatory Guide 1.33, Revision 2. February 1978, specifies procedures for startup, operation, and shutdown of safety-related PWR [ pressurized water reactor) ;
systems as an example of typical safety-related activities that should be covered by I written procedures. The Byron Station Technical Requirements Manual, Revision 1, paragraph 1.5.c states, in part, that violations of TRM surveillance requirements shall be treated the same as plant procedure violations. The licensee's failure to monitor the temperature of the SX cooling tower electric substations is a violation of TS 5.4.1.a for failure to implement the TRM surveillance requirements. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/455-99004-03(DRP)). This violation is in the licensee's corrective action program as problem identification form (PlF) B1999-088 c. Conclusions l The inspectors concluded that the licensee failed to control the configuration of the l engineered safety features (ESF) switchgear ventilation system consistent with the design basis description in the Updated Final Safety Analysis Report (UFSAR) and that the system failure analysis described in the UFSAR was not consistent with the actual failure response of the system. Furthermore, the inspectors concluded that by not monitoring the temperature of the essential service water cooling tower electric
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substations, the licensee did not appropriately monitor temperatures of ESF switchgear ventilation system areas consistent with Technical Requirements Manual surveillance roouirement 3.7.d.1. A Non-Cited Violation was issue O2.3 Failure to Follow a Procedure Resulted in Ooeration of Unit 1 A Safety Iniection Pumo with an inadeauate Suction to the Pumo Insoection Scope (71707)
The inspectors reviewed the circumstances surrounding the start of the 1A safety l injection (SI) pump with an inadequate suction to the pump during the performance of emergency core cooling system (ECCS) full flow testing. The inspectors interviewed operations and engineering department personnel and reviewed the licensee's prompt investigation report. The inspectors also reviewed Byron Engineering Surveillance
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Requirement (BVSR) Procedure 5.5.8.SI.2-1, *ECCS System Check Valve Stroke Test,"
Revision 2 and Byron Operating Procedure (BOP) RH-6, " Placing the RH [ Residual Heat Removal] System in Shutdown Cooling," Revision 16.
. Observations and Findinas On April 13,1999, during performance of ECCS full flow testing, the 1 A SI pump was started with the pump's suction from the 1B residual heat removal (RH) train isolate The inspectors noted that the ECCS full flow testing consisted of a series of engineering .
! surveillance test procedures that aligned the ECCS in different accident mode conditions.. Surveillance test procedure 1BVSR 5.5.8.SI.2-1 aligned the 1B RH train with a suction from the reactor coolant system (RCS) hot legs and discharge to the RCS cold legs by requiring BOP RH-6 to be performed as a step in the surveillance test procedure. The 1 A SI pump was to be supplied a suction from the discharge of the 1B RH pump for the test after RH system alignment to the RCS cold legs was establishe The engineering test director recognized that the 1B RH pump was already running and aligned with a suction from the RCS hot legs by a previously performed surveillance test procedure. The test director incorrectly assumed that the intent of BOP RH-6 had been met without verifying that all the appropriate steps of BOP RH-6 had been completed and signed off completion of the step in the surveillance test procedure. However, the 1B RH pump was running in the recirculation mode of operation with the mini-flow recirculation valve open and the pump's discharge valve to the cold legs closed. Shortly after the 1 A Si pump was started, the test director and operators identified that pump flow and suction pressure were lower than expected and secured the pump. Inasmuch as the 1 A Si pump was later tested and found to be undamaged, the inspectors concluded that the event had minimal safety significance. This failure constitutes a violation of minor significance and is not subject to formal enforcement actio Conclusions The inspectors concluded that the licensee failed to control the configuration of the residual heat removal system during emergency core cooling system full flow testing, which resulted in operation of the Unit 1 A safety injection pump with an inadequate suction to the pum . -
O2.4 Review of Out-of-Service Proaram implementation Durina Refuelina Outaae B1R09 I Inspection Scope (71707)
The inspectors reviewed the implementation of the licensee's out-of-service program ;
during refueling outage B1R09 to evaluate the effectiveness of corrective actions taken i by the licensee for previously identified out-of-service errors. The inspectors also reviewed the circumstances surrounding three events involving out-of-service program implementation errors during the outage. The inspectors interviewed operations ,
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department personnel and reviewed Byron Administrative Procedure (BAP) 330-1,
" Station Equipment Out-of-Service Procedure," Revision 29 and Byron Operating Procedure (BOP) RH-3, " Fill and Vent of the Residual Heat Removal [RH] System,"
Revision 1 Observations and Findinat The inspectors reviewed out-of-service errors during the 4-week B1R09 refueling outage and noted that three operational events were attributable to out-of-service errors, which was an improvement from the B2R07 outage period in which nine operational events ,
were attributed to out-of-service errors. One of the weaknesses previously identified by j the inspectors was that operators frequently received poor descriptions of the work j scope when an out-of-service request was electronically submitted to the operations )
department. The work scope was frequently vague and did not identify where a task '
was to be performed. The inspectors noted that this same weakness contributed to the 1SX011 out-of-service event discussed in section O2.1 and the 1RY-8080 out-of-service event discussed belo l The inspectors reviewed the three operational events during B1R09 which were attributable to out-of service errors. Those events are discussed belo Failure to Follow Prescribed Seauence for Out-of-Service Clearance Resulted in Water Spilled from the Residual Heat Removal System
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On April 4,1999, while restoring the 1 A residual heat removal (RH) train to service I
following maintenance, approximately 40 gallons of water was inadvertently spilled from the system via an open vent valve to the auxiliary building floor. The licensee's subsequent investigation determined that clearance of the out-of-service was not performed in the sequence prescribed in the Out-of-Service Lift Checklist. Clearance of the out-of-service was performed concurrently with venting and filling of the system piping in accordance with BOP RH-3. Although the out-of-service specified closing system vent valves prior to restoring power to system isolation valve motor operators, operations personnel did not expect any system isolation valves to reposition when power was restored. Operators failed to close the vent valves prior to restoring power to the system isolation valves and as a result, the 1 A RH pump recirculating valve (1RH610) stroked open when its power was restored. The inspectors reviewed the system design and noted that 1RH610 automatically opens when RH pump discharge flow is less than 750 gallons-per-minute to ensure that adequate cooling is provided for the pump. Therefore, the valve stroking open was ca expected system response. The inspectors also noted that BOP RH-3, Step F.1.1 required 1RH610 to be verified closed and a note in the procedure prior to the step stated that the valve may have to be de-energized to remain closed. The failure of operations personnel to follow the
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to formal enforcement action.
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Imorocer imolementation of Out-of-Service for Feedwater System isolation Valves Resulted in a Status Control Occurrence On April 6,1999, the Unit 1 steam generator feedwater system isolation valves (1FW009A/B/C/D) were re-coupled with an out-of-service in place which required them i to be uncoupled. The licensee's apparent cause report determined that the l out-of-service was inadequate, in that no out-of-service card was in place on the valves'
stems to prevent re-coupling. The licensee's electronic out-of-service program relies on the proper sequencing of out-of-services such that only one card is placed on any one component. The out-of-service card was in place on the valves' actuator because the electronic out-of-service data base did not contain a data point for the valves' stems and j special instructions to place an out-of-service card on the valves' stems were not "
followed. Weaknesses with the licensets electronic work control system (EWCS)
related to out-of-service implementation were previously identified by the inspectors and discussed in NRC Inspection Report 50-454/455-98011(DRP). In this example, the licensee identified that the lack of an EWCS data point for FWOO9 valve stems contributed to the out-of-service error. The licensee's apparent cause report additionally identified that mechanical maintenance personnel assumed that it was acceptable to re-couple the valves' actuators to the stems with out-of-service cards on the actuators because they were " pre-programmed" to look for cards on the stems. The report stated that this practice did not meet the licensee's expectations to not work on any component with an out-of-service card in place. The inspectors noted that this expectation is not j consistent with BAP 330-1 paragraph C.3.e which allows work to be performed on equipment that is tagged out-of-service provided that the work will not interfere with the
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ability of the equipment to maintain isolation. The inspectors concluded that this out-of-service error was of minor safety significance because valve interlocks prevented the valves from being opened and operators took immediate actions to place the valves out-of-service upon recognizing the deficiency. R3-coupling the Unit i steam generator feedwater system isolation valves (1FW009A/B/C/D) with an out-of-service in place which required them to be uncoupled constitutes a violation of minor significance and is
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not subject to formal enforcement actio An Out-of-Service Error Resulted in a Non-Isolable Reactor Coolant System Pressure l Boundarv Leak Durina Pressurizer Samole Valve Reolacement
On April 11,1999, while attempting to replace the Unit 1 pressurizer water space sample valve (1RY-8080), mechanical maintenance personnel ground down approximately 90 percent of two welds and observed water leakage from the partially removed welds on the line connecting the valve to the pressurizer. The work was stopped, supervisors informed, and leakage promptly controlled. The licensee quantified the leak rate at ten gallons-per-hour and determined a maximum possible leak rate with the valve removed of 8.5 gallons-per-minute. Makeup to the system in excess of the possible leakage rate was always availabl The inspectors noted that performance of the maintenance activity required the use of an " exceptional out-cf-service," which specified that the pressurizer be drained as a condition to protect plant personnel and equipment. The licensee subsequently
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determined that the plant conditions required by the exceptional out-of-service's special instruction (pressurizer drained) had not been maintained. The pressurizer was drained l when the out-of-service was hung on April 8,1999; however, on April 9,1999, the l pressurizer and reactor cavity were filled in preparation for loading fuel into the vesse l The work on 1RY-8080 which was initially scheduled for April 8,1999 was delayed until April 10,199 j l
The inspectors reviewed the licensee's prompt investigation for this event which i identified three inappropriate actions. First, the outage scheduler, outage director, and shutdown safety manager incorrectly assumed that the work request task title accurately reflected the scope of the work to be performed and did not perform a more in-depth review of the detailed work instructions. Second, out-of-service procedural requirements were not met when maintenance personnel failed to verify the special instructions of the exceptional out-of-service during the out-of-service walkdown. Third, the work execution center supervisor signed in the 1RY-8080 job on April 11,1999, and ,
failed to recognize that the required plant conditions were not me l The irispectors reviewed BAP 330-1 and noted that the procedure did not include a qualitative method to maintain required plant conditions for exceptional out-of-service For example, a conventional out-of-service uses tags to control the configuration of equipment and establish boundaries relied upon for the protection of plant personne The existence of these controls are visible to personnel, and are documented and '
tracked by the out-of-service program. In the case of the exceptional out-of-service the licensee was unable to obtain conventional isolation and included special instructions to establish and maintain specific plant conditions (pressurizer drained) for the protection of plant personnel for the duration of the out-of-service. However, BAP 330-1 did not require qualitative methods to ensure control room operators and other plant personnel were aware of the required plant conditions, nor did it require visible warnings on plant equipment necessary to alter plant conditions. The inspectors reviewed nuclear station operator, unit supervisor, and shift manager turnover sheets from April 8,1999, through April 11,1999, and noted no entries discussing the establishment of and the need to maintain plant conditions required by the exceptional out-of-service. The failure to maintain the required plant conditions was synonymous with an unauthorized removal of an out-of-service tag and the elimination of an out-of-service boundar Code of Federal Regulations Title 10 Part 50, Appendix B, Criteria V, " Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, BAP 330-1, " Station Equipment Out-Of-Service Procedure," Revision 29, was not of a type appropriate to the circumstances, in that, it did not ensure that plant conditions required as part of the exceptional out-of-service were accomplished and effectively maintained. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the N?C Enforcement Policy (50-454/99004-02b(DRP)). This violation is in the licensee's corrective action program as problem identification form (PIF) B1999-0145 c. Conclusions The inspectors concluded that three out-of-service errors resulted in operational events during the four week B1R09 refueling outage, which was an improvement from the
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82R07 outage period when nine out-of-service errors resulted in operational event The inspectors concluded that three recent out-of-service errors had similar weaknesses previously identified by the inspectors during the previous refueling outage (B2R07) and I discussed in NRC Incpection Report 50-454/455-98011(DRP). A Non-Cited Violation i was issue Miscellaneous Operations issues (92901)
08.1 (Closed) Violation 50-454/455-98005-01a(DRP): " Failure to Return Equipment to Service as Required by OOS [Out-of-Service) Program." An out-of-service error on the chemical and volume control system occurred as a result of poor communication between the involved operators, assigning multiple tasks to be performed concurrently, i and a perceived pressure to accomplish work rapidly. Although the violation was licensee identified, it was cited because this was a repetitive issue. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for j treatment of Severity Level IV violations contained in Appendix C of the Enforcement l Policy. This violation would have been treated as a Non-Cited Violation under the new '
policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PlF B1998-0109 .2 (Closed) Violation 50-454/455-98009-01(DRP): " Unit 1 Entered Mode 4 With Portions of Containment Floor Drain inoperable." Unit 1 entered into Mode 4 (hot shutdown) from Mode 5 (cold shutdown) with the seal table room floor drain plugged. The inspectors concluded that this event should have been prevented by corrective actions from previously identified issues with the floor drain system. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for treatment of Severity Level IV violations contained in Appendix C of the Enforcement Policy. This violation would have been treated as a Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PlF B1998-0183 .3 (Closed) Violation 50-454/455-98009-02a(DRP): " Inadequate Procedure (BOP (Byron Operating Procedure] VC-2) for Operation of Control Room Ventilation." The inspectors determined that BOP VC-2 was not appropriate to the circumstances due to an inadequate technical review during a procedure revision. Additionally, on at least eight occasions, control room operators performed this procedure and did not identify the procedural deficiencies and initiate corrective actions. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for treatment of Severity Level IV violations contained in Appendix C of the Enforcement Policy. This violation would have been treated as a Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PIF B1998-0183 .4 (Closed) Violation 50-454/455-98011-01a(DRP): " Failure to implement the OOS
[Out-of-Service) Program and Ensure That the OOS Boundary Was Adequate." The licensee failed to adequately provide an isolation boundary for modification work which installed additional venting capability for the 2B chemical and volume control (CV) pum As a result, when construction personnel drilled a 1/4 inch hole into the 28 CV pump suction piping, a continuous stream of water came out of the hole. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for treatment of Severity LevelIV violations contained in Appendix C of the Enforcement
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i Policy. This violation would have been treated as a Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PlF B1998-0287 .5 (Closed) Violation 50-454/455-98011-03(DRP): " Failure to Follow BOP [ Byron Operating Procedure] RH-9 Which Resulted in Level Being Inadvertently Lost From the
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pump down of the Unit 2 refueling cavity. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for treatment of Severity Level IV violations contained in Appendix C of the Enforcement Policy. This violation would have been treated as a Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PIF B1998-02881.
I G8.6 (Closed) Violation 50-454/455-98017-02(DRP): " Failure to meet TS 3.0.3 Action Requirements for Two Trains of VC [ Control Room Ventilation] Inoperable." The licensee failed to adequately evaluate the effects of multiple impairments of the control room ventilation envelope due to insufficient procedural guidance. As a result, the safety-related VC system was rendered inoperable during two periods of time when the system was not capable of maintaining a required positive pressure greater than or equal to 0.125 inch water gauge relative to ambient pressure in areas adjacent to the control room. The inspectors reviewed the licensee's corrective actions to check for any notable weaknesses. No weaknesses were identified and the corrective actions were
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found to be acceptable. This violation is closed.
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ll. Maintenance l l
M1 Conduct of Maintenance M1.1 Unit 1 Reactor Coolant Looo Stoo isolation Valve Repair Inspection Scope (627071 The inspectors observed portions of maintenance performed on the 1 A reactor coolant system cold leg loop stop isolation valve (LSIV) and interviewed maintenance and i engineering personnel. The inspectors also reviewed the following documents: Work i l
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Request (WR) 970122919-01, " Replace Valve Guides Due to Valve Binding and Travel Stop;" Westinghouse Safety Evaluation SECL-99-035, " Byron Station Unit 1 Primary Side Loose Part (Main Loop Stop Valve Internals)," Revision 2; and, Comed
[ Commonwealth Edison] Quality Assurance Manual," Quality Assurance Program Topical Report CE-1-A," Revision 66 Observations and Findinas At the start of refueling outage B1R09, the 1 A cold leg LSIV (1RC8002A) would not close. Based upon a review of design and maintenance work history, the licensee determined the most likely cause to be mis-positioning of one or both valve guides due to failure of the guide positioning pins, which were susceptible to stress corrosion cracking. The valve guides were each held in place with three rolled positioning pins, one along each side near the top of the guide, and one in the bottom. Upon
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disassembly of 1RC8002A, the licensee identified the following damage: (1) three guide pins for one of the valve guides had failed and were not found; (2) a small piece (approximately 1 inch x 3 inches x 0.5 inch) of the valve body casting used to hold one of the guides in place had broken off and was not found; (3) one valve guide had slid down below the disk and prevented the disk from closing; and, (4) minor internal valve damage to both the valve disk and seat had resulted from valve guide vibration in the reactor coolant pump (RCP) discharge flow. During the previous operating cycle, the loose parts monitoring system detected a loose part in the 1 A cold leg, which the licensee has attributed to the 1RC8002A loose guide. Over time, the vibrating valve guide caused the small piece of the upper retaining casting to break free from the valve body. The licensee concluded that the missing piece and the three positioning pins were swept down the cold leg by the RCP discharge flow, through the reactor vessel downcomer region, and are lying near the reactor lower core components. The licensee repaired the valve during the outage and performed an engineering evaluation on the missing part The inspectors reviewed the safety evaluation completed by Westinghouse in accordance with 10 CFR 50.59 and validated by the licensee for the operation of Unit 1 with the missing loose parts. In addition to the missing loose parts from 1RC8002A, the scope of the evaluation assumed that one additional LSIV would produce loose part This hypothetically would include both valve g Jides, all six positioning pins and two broken casting pieces, which could enter the reactor vessel as loose parts. The safety evaluation concluded that the presence of potential or identified loose parts would not adversely affect plant components and safety systems during normal operation and accident condition The inspectors reviewed the circumstances involving the licensee's failure to complete a quality control (OC) inspection hold point during valve re-assembly. During the final visual cleanliness inspection, a QC inspector identified metal filings and weld slag inside of the piping and valve bowl area. Those pieces of debris were removed. Upon subsequent video examination, the QC inspector identified what appeared to be a small,
% inch long, metal chip in the piping between the LSIV and the reactor vessel nozzl The mechanical maintenance crew proceeded with valve re-assembly without having the QC inspection hold point (WR 970122919-01, step 56) satisfied. The QC inspector raised this apparent procedure violation to licensee management while the valve re-assembly work continued. The licensee did not attempt to retrieve the foreign material which would have required disassembly of the valve or access via an RCS piping thermowell. Valve disassembly would have resulted in additional radiation dose for the workers and would have required another difficult rigging evolution to remove the bonnet. Access via an RCS piping thermowell would have resulted in significant testing of the pressure boundary upon completion. Instead, the licensee requested Westinghouse to revise their original loose parts safety evaluation to include this extra piec Code of Federal Regulations Title 10 Part 50, Appendix B, Criteria X, " Inspections,"
requires, in part, that a program for inspection of activities affecting quality shall be established and executed by or for the organization performing the activity to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity. The Comed [ Commonwealth Edison] Quality Assurance Manual, " Quality Assurance Program Topical Report CE-1-A," Revision 66a, established the licensee's Quality Assurance Program as described in UFSAR Chapter 1 J
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Section 3.5.3 of the Comed Quality Assurance Manual required, in part, that when inspections must be performed before work can continue, hold points are established in appropriate documents. Consent to waive hold points are recorded prior to continuation l of work. Additionally, Section 3.5.4 of the Comed Quality Assurance Manual required '
that when acceptance criteria are not met, corrected areas are re-inspected. Changes to, or rework of, an item after inspection requires re-inspection of the affected area The licensee's failure to complete a quality controlinspection hold point during valve re-assembiy is a violation of 10 CFR Part 50, Appendix B, Criteria X. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99004-04(DRP)). This violation is in the l licensee's corrective action program as problem identification form (PlF) B1999-0136 Conclusions The inspectors concluded that observed maintenance work on the 1 A reactor coolant system cold leg loop stop isolation valve was generally performed well; maintenance personnel were knowledgeable of the associated activities; and, work was performed in accordance with station procedures with one notable exception. The inspectors concluded that the licensee failed to complete a final cleanliness inspection quality control hold point prior to re-assembly of the valve which resulted in foreign material left in the reactor coolant system following the maintenance. A Non-Cited Violation was issue M1.2 Surveillance Test Observations Inspection Scope (61726)
The inspectors interviewed operations, engineering, and regulatory assurance department personnel; reviewed the completed test documentation and applicable portions of the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications (TS); and, observed the performance of selected portions of the following surveillance test procedure . 1BOSR 3.2.9-2 Unit 1 Train B Manual Safety injection Initiation and Manual Phase A initiation Surveillance l
. 1BOSR 7.5.3-1 Unit 1 Motor Driven Auxiliary Feedwater Pump Monthly Surveillance :
. 1BOSR 7.5.3-2 Unit 1 Diesel Driven Auxiliary Feedwater Pump Quarterly l Surveillance
. 1BOSR 8.1.2-1 Unit 1 1 A Diesel Generator Operability Monthly (Staggered) and Semi-Annual (Staggered) Surveillance ,
. 1BOSR 8.4-2 125V [ Volt) DC [ Direct Current] Bus 112 Load Shed When l Cross-Tied to DC Bus 212 j
. 2BOSR 3.2.7-6468 Unit 2 ESFAS [ Engineered Safety Features Actuation I
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System] instrumentation Slave Relay Surveillance (Train B Auto SI [ Safety injection]- K646) l
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. 2BOSR 8.1.2-2 2B Diesel Generator Operability Monthly (Staggered) and Semi-Annual (Staggered) Surveillance
. 1BVSR 1.4.3-1 Unit 1 Manual Rod Drop Time
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. 1BVSR 8.1.11-2 Unit 1 18 Diesel Generator Sequencer Test - 18 Month I Observations and Findinas The inspectors concluded that the observed surveillance tests were performed well and satisfied the requirements of the Technical Specifications. On May 3,1999, the inspectors observed the performance of 1BOSR 7.5.3-1 and noted that the plant responded as designed throughout the test and that the test was properly performe However, during review of the surveillance test procedure, the inspectors noted that l step A.2 stated that the surveillance test was performed to meet TS surveillance requirements SR 3.7.5.3, SR 3.3.2.3, and SR 3.3.2.8. However, the inspectors i determined that the surveillance test procedure did not satisfy the requirements of l
SR 3.3.2.3. and SR 3.3.2.8. Based upon discussions with engineering and regulatory !
assurance department personnel, the inspectors determined that surveillance requirements SR 3.3.2.3 and SR 3.3.2.8 were met by surveillance test procedures 1BVSR 3.1.9-2, " Unit 1 Reactor Coolant Pump Bus Undervoltage Quarterly Surveillance," Revision 1 and 1BVSR 3.5.1-1 " Unit 1 Bus 141 Undervoltage Protection Monthly Surveillance," Revision 1. Although the requirements of the TS were being met, the inspectors were concerned that the procedure error indicated a misunderstanding of the actual surveillance requirements. The licensee entered this issue into their corrective action program as problem identification form (PIF) B1999-0188 Conclusions i
The inspectors concluded that the observed surveillance tests were performed well and satisfied the requirements of the Technical Specifications. The inspectors identified that a step in the Unit 1 motor driven auxiliary feedwater pump monthly surveillance test procedure incorrectly stated that the procedure met surveillance test requirements that it did not meet; however, the inspectors concluded that those requirements were met by l other surveillance test procedure ;
M2 Maintenance and Material Condition of Facilitias and Equipment M2.1 Inspection of the Unit 1 Containment Buildina Prior to Startuo Followina Refuelina Outaae B1R09 (61726)
Prior to Unit 1 entry into mode 3 (hot shutdown), the inspectors performed an inspection l of the containment building to assess material condition and search for loose debris, which if present could be transported to the containment recirculation sump and cause restriction of flow to the emergency core cooling system (ECCS) pump suctions during loss-of-coolant accident conditions. The inspectors reviewed the licensee's list of outstanding items to be addressed prior to containment building closeout and noted that there were severalitems on the list. During the inspection of the containment building, the inspectors ioentified severalloose items which should have been removed from the containment building. Those items included two radiological protection warning signs, one screwdriver, several broken plastic cable ties, a four foot length of plastic hose, several felt tip markers, loose pieces of tape, foam earplugs, and wire in addition, the inspectors identified a piece of mirror insulation hanging loose on a residual heat removal system pipe located above the containment sump screen. The inspectors also i
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identified loose screws on the containment sump screens and the containment building floor drain sump cover. The inspectors discussed these discrepancies with the licensee {
upon exiting the containment building. The licensee removed the loose items and satisfactorily addressed the other inspector identified discrepancies. The inspectors concluded that material condition in the Unit 1 containment building was good and that 4 the loose items identified by the inspectors posed no significant risk to ECCS operabilit M4 Maintenance Staff Knowledge and Performance M4.1 Human Performance Error Durina 1B Emeraency Diesel Generator Surveillance Testina l Resulted in Trippina the Component Coolina Water Pumo Suoportina Decav Heat l Removal l Inspection Scooe (61726)
The inspectors reviewed the circumstance surrounding the improper installation of an l electricaljumper during performance of Byron Engineering Surveillance Requirement I I
(1BVSR) 8.1.12-2, " Unit 1 1B Diesel Generator ESF [ Engineered Safety Feature)
Actuation Test Signal Start and Non-Emergency Trip Bypass Test, and Generator Differential Trip Test," Revision 1. The inspectors interviewed maintenance department personnel; reviewed the results of the licensee's prompt investigation into the event; and, reviewed the following documents; Common Work Practice Instruction, Nuclear Station Procedure, Conduct of Operations , Instruction 11 (CWPI-NSP-OP-1-11),
" Verification Practices," Revision 0; CWPI-NSP-OP-1-09, " Heightened Level of Awareness Briefings," Revision 0; and Byron Site Policy Memo 200.06,"Use of Pre-job and Post-job Briefs." Observations and Findinas i
On April 11, during the performance of surveillance testing on the 1B emergency diesel generator (EDG), electrical maintenance personnel installed an electrical jumper on the '
wrong relay causing an inadvertent trip of the common component cooling water (CC)
pump. At the time of the trip, the common CC pump was running and removing decay heat in conjunction with the residual heat removal system. In response to this event, I operators re-established CC system flow, system engineers performed an evaluation of the trip and determined that no damage to the pump or associated circuitry had occurred, and a prompt investigation was initiated into the cause of the erro The inspectors reviewed the licensee's prompt investigation, which identified the following apparent causes and contributing factors associated with this event:
a poorlabeling of the relays;
- electrical maintenance personnel did not attend the pre-job brief; a the system engineer provided erroneous guidance;
- the verification process procedure was not followed; a the surveillance test procedure was not followed; and
. electrical prints were not used at the job site.
l The inspectors concurred with the licensee's conclusion that the event was caused by a number of inappropriate actions which included the failure of personnel involved in this event to perform step 2.5 of 1BVSR 8.1.12-2 as written and the failure of personnel to
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correctly perform concurrent dual verification. The failure to correctly install a jumper during performance of surveillance testing on the 1B EDG constitutes a violation of minor significance and is not subject to formal enforcement action. Although this event had minimal safety impact, the inspectors were concerned that the licensee's corrective actions for similar human performance errors, which were documented in NRC Inspection Reports 50-454/455-99002(DRP); and 50-454/455-98020(DRP), have not been totally effective at preventing their recurrenc Conclusions The inspectors concurred with the licensee's conclusion that multiple human performance deficiencies were the cause for electrical maintenance personnel installing an electrical jumper on the wrong relay while performing surveillance testing on the 1B emergency diesel generator. This caused an inadvertent trip of the common l
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component cooling water pump. Although this event had minimal safety impact, the inspectors were concerned that the licensee's corrective actions for similar previous human performance errors have not been totally effective at preventing their recurrenc I l
M4.2 .Qnaooroved Maintenance Activity Performed by Maintenance Personnel Rendered the1B Main Steam isolation Valve Inocerable Inspection Scope (62707)
t The inspectors reviewed the circumstances surrounding an unauthorized maintenance )
activity that was performed on the 18 main steam isolation valve (MSIV) that rendered the valve inoperable. The inspectors interviewed operations and maintenance department personnel and reviewed the results of the licensee's prompt investigation into the even Observations and Findinos On April 24,1999, during the Unit 1 startup from refueling outage B1R09, mechanical maintenance personnel adjusted the stem packing on the 1B MSIV, an activity affecting quality, without appropriate work authorization from the operations department and rendered the valve inoperable. The inspectors noted that mechanical maintenance personnel failed to have an approved set of work instructions for the packing adjustment prior to performing the activity. The inspectors reviewed the operators' immediate actions to address the operability of the 1B MSIV and determined that operators correctly declared the valve inoperable and implemented the required actions of the T Operations personnel subsequently performed 1BVSR 7.2.1, " Unit 1 Main Steam isolation Valve Operability Test," Revision 3 and declared the valve operabl l l
Code of Federal Regulations Title 10, Part 50, Appendix B, Criteria V, " Instructions, j Procedures, and Drawings," requires, that activities affecting quality shall be prescribed l by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, mechanical maintenance personnel performed a stem packing adjustment on the 18 MSIV, an activity affecting qual.ty, without an approved set of instructions containing appropriate quantitative or i i
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qualitative acceptance criteria for determining that the packing adjustment had been l satisfactorily accomplished. This is a violation of 10 CFR Part 50, Appendix B, !
Criteria V. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99004-05(DRP)).
This violation is in the licensee's corrective action program as problem identification form (PlF) B1999-0174 Conclusions The inspectors concluded that mechanical maintenance personnel adjusted the stem packing on the 18 main steam isolation valve (MSIV) without approved work instructions and operations department authorization for the activity, which rendered the valve
inoperable. The inspectors concluded that operators took appropriate actions to !
promptly address the operability of the 18 MSIV and correctly implemented the required actions of the Technical Specifications. A Non-Cited Violation was issue l M8 Miscellaneous Maintenance issues (92902)
M8.1 (Closed) Violation 50-454/455-98005-01b(DRP): " Failure to Submit ER [ Engineering Request] Prior to Installing Gasket in 1B SX [ Essential Service Water) Pump "
Appropriate design control measures had not been taken prior to adding a gasket in the l 1B SX pump. This Severity Level IV violation was issued prior to the implementation of l the NRC's new policy for treatment of Severity Level IV violations contained in l Appendix C of the Enforcement Policy. This violation would have been treated as a !
Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PIF B1998-0109 M8.2 (Closed) Violation 50-454/455-98005-02(DRP): "NSWP [ Nuclear Station Work Procedure] A-03 Inadequate to Prevent Material Intrusion in CS [ Containment Spray]
System." A condenser tube cleaning brush found in the N system renresented an additional example of continuing poor foreign material e%sion contd.s by the licensee. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for treatment of Severity Level IV violations contained in Appendix C of the Enforcement Policy. This violation would have been treated as a Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PlF B1998-0109 M8.3 (Closed) Violation 50-454/455-98009-02b(DRP): " Failure to Have Established a Written Procedure for the Control of WRs [ Work Requests] While the WRs Were on Hold." The inspectors identified that approximately 50 WRs had senior reactor operator (SRO)
authorization signatures without work activities on the current work schedule and were considered on " hold." A violation was cited for the licensee's failure to provide adequate procedural guidance on how to process a WR that had been authorized by an SRO and subsequently put on " hold" for an extended period of time. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for treatment of Severity LevelIV violations contained in Appendix C of the Enforcement Policy. This l violation would have been treated as a Non-Cited Violation under the new policy and i therefore it is closed in this report. This violation is in the licensee's corrective action program as PIF B1998-0184 l
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' M8.4 (Closed) Violation 50-454/455-98011-06(DRP): " Multiple Examples of Failure to Follow Ventilation Barrier impairment Process." The licensee had impaired several ventilation barriers without appropriately processing barrier impairment permits. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for
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treatment of Severity Level IV violations contained in Appendix C of the Enforcement Policy. - This violation would have been treated as a Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PIF B1998-0288 M8.5 (Closed) Violation 50-454/455-98011-07(DRP): " Failure to Follow Maintenance Procedure Regarding the Control of Penetration Sleeves on the RWST (Refueling Water Storage Tank) Tunnel Hatch." The licensee had improperly impaired the RWST pipe tunnel hatch cover, which is a ventilation barrier between the auxiliary building and the outside atmosphere, by having multiple cable penetrations in the hatch cover open simultaneously. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for treatment of Severity Level IV violations contained in Appendix C of the Enforcement Policy. This violation would have been treated as a Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PlF B1998-0288 M8.6 (Closed) Violation 50-454/455-98025-03(DRP): "Unacceptab'e l Preconditioning of the Auxiliary Feedwater Pump Discharge Valves,1/2AF004A/B." The licensee
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unacceptably preconditioned the stroke time testing of auxiliary feedwater discharge I valves,1/2AF004A/B, by not measuring the stroke time until the sixth time the valves were opened during the testing. The inspectors reviewed the licensee's corrective actions to check for any notable weaknesses. No weaknesses were identified and the corrective actions were found to be acceptable. This violation is close . Enaineerina E8 Miscellaneous Engineering lasues (92700,92903)
E LClosed) Violation 50-454/455-98011-11(DRP): " Failure to Update FSAR [ Final Safety analysis Report) as Required by 10 CFR 50.71(e)." The licensee fai!e d to submit a required revision to the FSAR to reflect changes in the operation of the containment purge system described in Safety Evaluation T1-93-0152, completed on October 16, 1993. This Severity Level IV violation was issued prior to the implementation of the NRC's new policy for treatment of Severity Level IV violations contained in Appendix C of the Er'orcement Policy. This violation would have been treated as a Non-Cited Violation under the new policy and therefore it is closed in this report. This violation is in the licensee's corrective action program as PIF B1998-02891.
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E8.2 (Closed) Licensee Event Reoort (LER) 50-455/98002: " Circuit Card Failure Causes l Second Inoperable Channel and Technical Specification 3.0.3 Entry." The inspectors reviewed the licensee's corrective actions for the circuit card failure which were to replace the card and monitor card failures via the maintenance rule program. Based on the information that this was the first card of this type to fail, the inspectors concluded that the licensee's corrective actions were acceptable. The inspectors also determined that the licensee's entry into Technical Specification Limiting Condition for Operations 3.0.3 for approximately 12 minutes was correct. This LER is close E
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I IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Epdiolooical Protection Practices Inspection Scope (71750)
l The inspectors observed the posting of radiation areas, the control of locked high Iadiation areas, the application of As-Low-As-Reasonably-Achievable (ALARA)
principles, and the radiation work practices of station personnel. Special attention was ,
placed on radiological practices related to the Unit i refueling outag Observations and Findinos The inspectors reviewed the postings and control of radiologically controlled areas and j high radiation areas related to refueling activities. The inspectors noted that I radiologically controlled areas were properly posted for the condition that existed in the l areas. Rope boundaries, swing gates (where applicable) and signs were properly maintained. Access and control of personnel working in the Unit 1 containment building was conducted appropriatel j
The inspectors reviewed the ALARA controls used for containment building acces The inspectors noted that prior to containment building entry, radiation protection personnel briefed personnel on the dose rates in the affected areas, identified areas to l avoid, and identified the routes that resulted in the lowest dose. In addition, radiation I protection personnel observed workers for proper use of anti-contamination clothing, {
and moving materialin and out of the controlled are {
The inspectors observed proper radiation worker practices, which included proper use of anti-contamination clothing, proper frisking upon exit from the contaminated area, and proper use of dosimetry. Personnel properly monitored equipment out of radiologically l controlled areas and made proper use of personnel radiation monitors at the exit of radiologically controlled areas. The inspectors also noted proper and efficient separation of personnel entering and exiting the containment building. Inside the .
containment building, radiation protection technicians at the missile shield entrance were 1 observing worker practices and responded promptly to digital dosimeter alarm Following the Unit i refueling outage, the inspectors identified an area of loose surface i contamination in the form of crystallized boric acid that was not within a contamination l control boundary in the Unit 1 auxiliary building containment piping penetration are i The licensee promptly roped off and decontaminated the area. The inspectors concluded that this was an isolated incident which was not indicative of a problem with the licensee's control of contamination area Conclusions i l
The inspectors concluded that radiologically controlled areas were properly posted and radiation workers demonstrated proper work practices to control the spread of
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contamination. The inspectors concluded that As-Low-As-Reasonably-Achievable principles were effectively utilized to minimize dose during the Unit 1 refueling outag . J
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V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on May 10,1999. The licensee acknowledged the findings presented. 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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PARTIAL LIST OF PERSONS CONTACTED Licensee B. Adams, Regulatory Assurance Manager M. Jurmain, Maintenance Manager B. Kouba, Engineering Manager W. Levis, Site Vice President -
R. Lopriore, Station Manager-W. McNeill, Radiation Protection Manager T. Schuster, Work Control Manager M. Snow, Operations Manager M. Snow, Operations Manager J. Stone, Nuclear Oversight Manager I
i INSPECTION PROCEDURES USED IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 71707: Plant Operations IP 71750: Plant Support Activities IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor i Facilities IP 92901: Follow-up Plant Operations <
IP 92902: Follow-up Maintenance IP 92903: Follow-up Engineering i
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ITEMS OPENED, CLOSED, AND DISCUSSED l Ooened 50-454/99004-01 NCV Failure to implement safety evaluation procedure for lowering of suspended fuel assembly 50-454/99004-02a NCV inadequate technical review of out-of-service prepared for essential service water return header cross-connect valve maintenance 50-454/99004-02b NCV Failure to ensure plant conditions required for exceptional out-of-service were accomplished and maintained 50-454/455/99004-03 NCV Failure to implement Technical Requirements Manual surveillance requirement to monitor engineered safety l
features switchgear ventilation system areas 50-454/99004-04 NCV Failure to complete a quality control inspection hold point during 1 A cold leg loop stop isolation valve repair 50-454/99004-05 NCV 1B main steam isolation valve packing adjustment without operation's authorization and appropriate work instructions Closed 50-454/99004-01 NCV Failure to implement safety evaluation procedure for lowering of suspended fuel assembly l
l 50-454/99004-02a NCV - Inadequate technical review of out-of-service prepared for essential service water return header cross-connect valve maintenance 50-454/99004-02b NCV Failure to ensure plant conditions required for exceptional out-of-service were accomplished and maintained 50-454/455/99004-03 NCV Failure to implement Technical Requirements Manual surveillance requirement to monitor engineered safety 1 features switchgear ventilation system areas j
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I l 50-454/455/98005-01a VIO Failure to return equipment to service as required by OOS i (out-of-service) program i 50-454/455/98009-01 VIO Unit i entered mode 4 with portions of containment floor l
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drain inoperable 50-454/455/98009-02a VIO Inadequate procedure (BOP (Byron Operating Procedure)
VC-2) for operation of control room ventilation l 50-454/455/98011001a VIO Failure to implement the OOS (out-of-service) program !
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and ensure that the OOS boundary was adequate
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50-454/455/98011-03 VIO Failure to follow BOP [ Byron Operating Procedure] RH-9 which resulted in level being inadvertently lost from the SFP [ spent fuel pool)
'50-454/455/98017-02 VIO Failure to meet TS 3.0.3 action requirements for two trains of VC [ control room ventilation) inoperable 50-454/99004-04 NCV Failure to complete a quality controlinspection hold point during 1 A cold leg loop stop isolation valve repair 50-454/99004-05- NCV 1B main steam isolation valve packing adjustment without operation's authorization and appropriate work instructions 50-454/455/98005-01b VIO Failure to submit ER [ engineering request] prior to installing gasket in 1B SX [ essential service water) pump ,
50-454/455/98005-02 VIO NSWP [ Nuclear Station Work Procedure) A-03 inadequate to prevent material intrusion in CS [ containment spray)
system 50-454/455/98009-02b VIO Failure to have established a written procedure for the control of WRs [ work requests) while the WRs were on {
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50-454/455/98011-06 VIO Multiple examples of failure to follow ventilation barrier {
impairment process 50-454/455/98011-07 VIO Failure to follow maintenance procedure regarding the control of penetration sleeves on the RWST [ refueling water storage tank] tunnel hatch 50-454/455/98025-03 VIO Unacceptable preconditioning of the auxiliary feedwater i pump discharge valves,1/2AF004A/B 50-454/455/98011-11 VIO Failure to update FSAR [ Final Safety Analysis Report) as required by 10 CFR 50.71(e)
50-455/98002 LER Circuit card failure causes second inoperable channel and Technical Specification 3.0.3 entry
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LIST OF ACRONYMS USED ADRC Advanced Digital Reactivity Computer ALARA As-Low-As-Reasonably-Achievable BAP- Byron Administrative Procedure BFP Byron Fuel Handling Procedure BGP Byron General Operating Procedure BOP Byron Operating Procedure BOSR Byron Operating Surveillance Requirement BVSR Byron Engineering Surveillance Requirement CC Component Cooling Water CFR Code of Federal Regulations CS Containment Spray CV Chemical and Volume Control DC Direct Current DCP Design Change Procedure DRP Division of Reactor Projects ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EOP Emergency Operating Procedure ER Engineering Request ESF Engineered Safety Feature ESFAS Engineered Safety Feature Actuation System '
EWCS Electronic Work Control System FSAR Final Safety Analysis Report ITS Improved Technical Specifications LCO Limiting Condition for Operation LER Licensee Event Report LCOAR Limiting Conditions for Operation Action Requirement LSIV Loop Stop Isolation Valve i MSIV Main Steam isolation Valve-NCV Non-cited Violation NRC . Nuclear Regulatory Commission NSP Nuclear Station Procedure NSWP Nuclear Station Work Procedure OOS Out-Of-Service PORC Plant Operational Review Committee PWR Pressurized Water Reactor-PlF Problem identification Form QC Quality Control RCS- Reactor Coolant System RCP Reactor Coolant Pump RH Residual Heat Removal RP&C Radiological Protection and Chemistry RWST Refueling Water Storage Tank SFP Spent Fuel Pool -
SI Safety injection SPP Special Procedure SR Surveillance Requirement SRO Senior Reactor Operator SSCs Structures, Systems, and Components
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!- LIST OF ACRONYMS USED (cont'd)
l SX Essential Service Water
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i .- TRM Technical Requirements Manual l T Technical Specification
! UFSAR Updated Final Safety Analysis Report l V Volts l VC Control Room Ventilation
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