IR 05000454/1999003
| ML20205Q274 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 04/14/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20205Q263 | List: |
| References | |
| 50-454-99-03, 50-454-99-3, 50-455-99-03, 50-455-99-3, NUDOCS 9904210151 | |
| Download: ML20205Q274 (24) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION lil Docket Nos: 50-454;50-455 License Nos: NPF-37; NPF-66 Report No: 50-454/455/99003(DRP)
Licensee: Commonwealth Edison Company Facility: Byron Generating Station, Units 1 and 2 Location: 4450 N. German Church Road Byron,IL 61010 Dr s: February 17 - March 29,1999 i., sectors: E. Cobey, Senior Resident inspector B. Kemker, Resident inspector T. Tongue, Project Engineer, Rlli D. Pelton, Braidwood Resident inspector C. Thompson, Illinois Department of Nuclear Safety Approved by: Michaal J. Jordan, Chief Reactor Projects Branch 3 Division of Reactor Projects
9904210151 990414 PDR ADOCK 05000454 G PDR
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EXECUTIVE SUMMARY Byron Generating Station Units 1 and 2 NRC Inspection Report 50-454/99003(DRP); 50-455/90003(DRP)
This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 6-week period of inspection activities by the resident rtaff and region based inspector Operations
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The inspectors concluded that routine operations were conducted in a safe, professional, and controlled manner. Operators adhered to the station's standards for reactivity management, professionalism, control room conduct, procedural adherence, annunciator response, and generally used three-way communications. This observation has been consistent over the past several;nspection periods. (Section 01.1)
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The inspectors concluded that the Unit 1 reactor shutdown for Refueling Outage BIR09 was conducted in a safe and controlled manner. Specifically, the heightened level of awareness briefing was thorough, simulator training was effectively utilized by esa tors to prepare for the evolution, management oversight was evident, and operations command and control of the evolution was effective. The inspectors also concluded that operators adhered to the station's standards for reactivity management, professionalism, control room conduct, procedural adherence, annunciator response, and generally used three-way communications. (Section O1.2)
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The inspectors concluded that the licensee rendered the Unit 1 A emergency diesel generator inoperable for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while the north flood door to the 1 A diesel oil storage tank room was left open and unattended. However, no violation of regulatory requirements occurred since the licensee restored the door to comply with the technical specification limiting condition for operatio . stCO) upon discovery and within the required completion time of the LCO action statement. (Section O2.1)
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The inspectors concluded that the licensee failed to implement corrective actions that were identified following an event in November 1996, in which the Unit 2 fuel cooling pump control switch had been inadvertently bumped to the "after-trip" position, As a result, the licensee failed to prevent a recurrence of the same event. This failure constitutes a violation of minor significance and is not subject to formal enforcement action. (Section 02.2) j i
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The inspectors concluded that mis-communication and ;nadequate procedures resulted l in a potential chemical transfer accident, in which the licensee nearly transferred sodium I hydroxide (a strong caustic solution) from a delivery truck into a storage tank containing I sulfuric acid. (Section O2.3)
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The inspectors concluded that the licensee failed to control the configuration of the I treated waste system during a planned liquid effluent release evolution by performing the release via an unintended and untested release path. A Non-Cited Violation was
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issued for the licensee's failure to implement Byron Chemical Control Procedure 400-TWX01, " Liquid Radwaste Release Form for Release Tank OWX01T,"
Revision 14. (Section O2.4)
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The inspectors concurred with the licensee's conclusion that prior to January 1998 both Units 1 and 2 were operated in excess of their licensed thermal power levels as a result of a discrepancy identified with the steam generator blowdown flow totalizer recorder scaling, which affected the accuracy of the thermal power calorimetric calculation in a nonconservative direction. A Non-Cited Violation was issued. (Section 08.2)
Maintenance / Surveillance
. The inspectors concluded that minimum bend radri for two flexible hoses installed on the 1 A emergency diesel generator fuel oil drain line was violated during installation. This was due to a mechanical maintenance first-line supervisor improperly deleting the appropriate requirements for the installation and inspection of the hoses from the maintenance work instructions. In addition, mechanical maintenance personnel improperly bent replacement tubing for the fuel oil pump suction line as a result of poor bending practices, which resulted in the tube's circumference being out-of-round. A j Non-Cited Violation was issued. (Section M1.1)
. The inspectors concurred with the licensee's conclusions that the 1 A chemical and volume control (CV) pump motor-to-gearbox coupling fasteners were over-torqued. This was due to a mechanical maintenance first4ine supervisor improperly revising the torque value specified in the maintenance work instructions. In addition, mechanical maintenance personnel over-torqued the 1 A CV pump gear-to-pump coupling fasteners i due to an incorrect torque value being specified in the maintenance work instructions. A l
Non-Cited Violation was issued. (Section M1.2) l
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The inspectors concluded that the observed surveillance tests were performed well and satisfied the requirements of the Technical Specifications. The inspectors identified an error in the calculation of the acceptance criteria for the auxiliary feedwater pump full-flow testing to the steam generators; however, the error did not change the overall results of the test and were considered of minor significance. (Section M1.3)
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The inspectors concluded that an invalid high energy line break isolation of the Unit 1 steam generator blowdown system occurred during temperature switch calibration as the result of multiple human errors and inadequate communications practices. A Non-Cited Violation was issued. (Section M8.2)
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The inspectors concluded that the licensee's process for implementing and administering the instrument out-of-tolerance (OOT) trending program adequately identifies, trends, and evaluates instrument OOT conditions. Additionally, the inspectors noted that assessments performed by the station's nuclear oversight department during the past 6 months provided constructive and timely recommendations which contributed to the development of the licensee's program. (Section E2.1)
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The inspectors concluded that the temporary modifications reviewed were generally well controlled and each temporary modification had an action plan for remova (Section E2.2)
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The inspectors identified that the licensee failed to include the auxiliary feedwater pump discharge valves,1/2AF004A/B, within the scope of the inservice test (IST) progra .
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Additionally, the inspectors concluded that other numcrous examples of the failure to include passive valves within the IST program identified by the licensee represented a programmatic deficiency with the scoping of the IST program. A Non-Cited Violation was issued. (Section E8.2)
i Plant Support i
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The inspectars concluded that radiologically controlled ar9as were properly pos {
locked high radiation area doors were locked and properly controlled by radiat, j protection personnel; radiation workers demonstrated proper work practices to control j the spread of radioactivity; and As-Low-As-Reasonably-Achievable (ALARA) principles, '
such as briefings to minimize exposure to personnel were effectively utilize (Section R1.1)
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Report Details Summary of Plant Status The licensee operated Unit 1 at or near full power until March 27,1999, when the licensee began refueling outage B1R09. Unit 1 remained shutdown at the end of the inspection perio The licensee operated Unit 2 at or near full power for the duration of the inspection perio . Operations 01 Conduct of Operations 0 General Observations During this inspection period the inspectors routinely observed the conduct of -lant operations. The inspectors concluded that routine operations were conducted a a safe, professional, and controlled manner. Operators adhered to the station's standards for reactivity management, professionalism, control room conduct, procedural adherence, annunciator response, and generally used three-way communications. This observation has been consistent over the past severalinspection periods. The inspectors further noted that in an effort to address configuration control and human performance issues at the station, the licensee conducted a 10-day stand down period which started on March 11,1999, to focus on performance standards. In addition, the inspectors noted that on March 12,1999, Byron Units 1 and 2 surpassed the Westinghouse domestic dual unit continuous run record of 296 days which represerted a significant accomplishment for the license .2 Unit 1 Shutdown for Refuelino Outaae B1R09 Inspection Scope (71707)
The inspectors observed the heightened level of awareness (HLA) briefing and the Unit 1 shutdown. The inspectors interviewed operations department personnel and reviewed the following procedures:
. 1 BGP 100-4, * Power Descension," Revision 17
. 1BGP 100-5, * Plant Shutdown and Cooldown," Revision 29
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1BOSR 3.g.5-1," Unit One Turbine Mechanical Overspeed Trip Surveillance,"
Revision 1 Observations and Findinas On March 27,1999, the licensee commenced a Unit 1 shutdown for Refueling Outage B1R09. The inspectors observed the HLA briefing for the shutdown and noted that the briefing was thorough, in that, the briefing included a discussion of expected plant response, contingency actions, testing to be performed, the chain of command, and the roles and responsibilities of participants in the evolution. The inspectors noted that the briefing also incorporated lessons learned from the operating shift's simulator training for the shutdown evolution as well as lessons from industry events. The
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inspectors noted that the licensee utilized an additional licensed senior reactor operator to assist the unit supervisor with administrative and other specifically assigned supervisory functions, which allowed the unit supervisor to remain properly focused on the antire evolutio The inspectors also noted that the Unit 1 start-up feedwater (FW) pump was not available for the shutdown due to a mechanical failure. Operators effectively utilized
"just-in-time training" in the simulator to evaluate FW system response at low power using the motor driven 1 A FW pump instead of the start-up FW pump. In addition, the inspectors noted that a temporary change to BGP 100-4 which allowed operators to adjust the steam dump pressure controller output to just below actual turbine first-stage pressure provided a quicker steam dump response when the turbine was trippe The inspectors observed that station management oversight was evident during the shutdown. Specifically, management oversight was particularly evident when management stopped the performance of turbine over-speed testing when repetitive fuse failures in the turbine digital electro-hydraulic control system became a distraction to operators. The incoectors noted that operators adhered to the station's standards for reactivity management, professionalism, control room conduct, procedural adherence, annunciator response, and generally used three-way communication Conclusions The inspectors concluded that the Unit 1 reactor shutdown was conducted in a safe and controlled manner. Specifically, the heightened level of awareness briefing was thorough, simulator training was effectively utilized by operators to prepare for the I evolution, management oversight was evident, and operations command and control of the evolution was effective. The inspectors also concluded that operators adhered to the station's standards for reactivity management, professionalism, control room
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conduct, procedural adherence, annunciator response, and generally used three-way
! communication O2 Operational Status of Facilities and Equipment l O Failure to Control the Confiouration of the Unit 1 A Diesel Oil Storace Tank Room Flood I
Door Rendered the 1 A Emeraency Diesel Generator Inoperable Inspection Scope (71707)
The inspectors reviewed the circumstances surrounding the licensee's failure to control the configuration of the Unit 1 A diesel oil storage tank (DOST) room flood door, which j inadvertently rendered the 1 A emergency diesel generator (EDG) inoperable for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The inspectors interviewed operations department personnel l and reviewed applicable portions of the Technical Specifications (TS) and the Updated Final Safety Analysis Report (UFSAR).
! Observations and Findinas On March 7,1999, while performing routine security rounds, a security guard found the north flood door to the 1 A DOST room open and unattended. The security guard closed the flood door immediately after confirming that no pe,sonnel were inside the room and
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notified the shift manager. The licensee completed a prompt investigation which concluded that the 1 A DOST room flood door was inadvertently left open and unattended by a non-licensed operator who performed routine operator rounds in the DOST room The inspectors reviewed UFSAR Section 10.4.5 and determined that during a design basis turbine building flooding event (rupture of a main condenser circulating water inlet line), the turbine building could be flooded to grade level. The UFSAR further states that the auxiliary building is completely watertigh 1elow grade at the turbine building / auxiliary building interface, except for the main steam tunnel. Watertight doors, as described in UFSAR Section 9.5.4.3, protect the DOST rooms, including the fuel oil transfer pumps located inside the rooms, from flooding in the turbine building. The ability of the fuel oil transfer system to transfer fuel oil from the DOST to the day tanks is required for EDG operability. The doors are assumed to withstand a flood in the turbine building up to grade level. The inspectors concluded that the 1 A EDG was r endered inoperable for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while the 1 A DOST room flood door was open and unattended. However, no violation of regulatory requirements occurred since the licensee restored the door to comply with the TS limiting condition for operation (LCO)
upon discovery and within the required completion time of the LCO action statemen Conclusions The inspectors concluded that the licensee rendered the Unit 1 A emergency diesel generator inoperable for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while the north flood door to the 1 A diesel oil storage tank room was left open and unattended. However, no violation of regulatory requirements occurred since the licensee restored the door to comply with i the technical specification limiting condition for operation (LCO) upon discovery and within the required completion time of the LCO action statemen i O2.2 F_ailure to Imolement Corrective Actions Resulted in the inadvertent Loss of the Unit 2 Fuel Coolina Pumo Inspection Scope (71707)
The inspectors reviewed the circumstances surrounding the inadvertent repositioning of the Unit 2 fuel cooling (FC) pump control switch, which resulted in an elevated spent fuel pool (SFP) temperature. The inspectors interviewed operations and engineering department personnel and reviewed Root Cause Report 455-201-99-CAO00056, " Trip I of Spent Fuel Pit Cooling Pump 2FC01P Due to inadvertent Bumping of Control Switch." Observations and Findinas On January 29,1999, while performing operator rounds, a non-licensed operator discovered that the SFP temperature had increased approximately 10*F above the previous reading. Upon investigation, the licensee determined that the Unit 2 FC pump (2FC01P) control switch had been inadvertently moved to the "after-tr;p" position. The spent fuel pool temperature reacMd 98*F before cooling was restored by the operator, which was wc" below the SFP h @ remperature alarm setting of 149"F. The inspectors therefore ce mined that this eveni had minor safety significanc O
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The inspectors reviewed the licensee's root cause evaluation which was unable to identify how or why the pump's control switch position was inadvertently changed. The root cause evaluation found no evidence to support mischief or switch failure and therefore concluded that the most probable cause of the event was an inadvertent bumping of the pump's control switch. The inspectors noted that a similar issue with the licensee's failure to control the configuration of 2FC01P was previously identified by the licensee in November 1996 and was entered into the licensee's corrective action program. The inspectors concluded that the licensee failed to implement corrective actions that were identified following the previous evarit which could have prevented a recurrence. The licensee's failure to implement corrective actions for this condition adverse to quality is considered to be a violation of 10 CFR Part 50, Appendix B, Criterion 16," Corrective Actions." This failure constitutes a violation of minor significance and is not subject to formal enforcement actio As immediate corrective actions for this event, the licensee installed protective covers over both the Unit 1 and Unit 2 FC pumps' control switches. Additionally, the licensee completed a walkdown of previously identified components whose control switches had been identified as " inadvertently bumped" to assess the completeness of corrective actions associated with those components. The licensee entered the results of that walkdown into their corrective action progra Conclusions The inspectors concluded that the licensee failed to implement corrective actions that were identified following an event in November 1996, in which the Unit 2 fuel cooling pump control switch had been inadvertently bumped to the "after-trip" position. As a result, the licensee failed to prevent a recurrence of the same event. This failure constitutes a violation of minor significance and is not subject to formal enforcement actio O2.3 Miscommunication and an inadeauate Procedure Resulted in a Potential Chemical Transfer Accident Insoection Scope (71707)
The inspectors reviewed the circumstances surrounding a potential chemical transfer accident, in which the licensee nearly transferred sodium hydroxide (a caustic solution)
from a delivery truck into a storage tank containing sulfuric acid. The inspectors interviewed operations department personnel and reviewed the licensee's prompt investigation repor Observations and Findinas On March 9,1999, a chemical truck with sodium hydroxide arrived at Byron Station for a scheduled delivery. Two non-licensed operators and a security guard were assigned to assist with transferring the chemical from the truck to a storage tank. Based upon initial communications with the truck driver and the control room, the operators were led to believe that they were to be transferring sulfuric acid. The truck driver stated that he had " caustic acid" in his truck. While the truck driver was connecting the transfer hose to the acid tank's fill connection, the security guard asked one of the operators, "what was caustic acid?" This prompted the operators to further question the truck driver and
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l to review the shipping documents to discover that the truck contained sodium hydroxid The operators directed the truck driver to disconnect the transfer hose from the acid tank's fill connection, obtained the proper procedure for filling the caustic storage tank, directed the truck driver to connect the transfer hose to the caustic tank's fill connection, and completed the transfe The inspectors noted that station procedures did not require verification of the chemical truck's contents by review of the Bill of Lading prior to connecting the truck's transfer hose to a chemical tank. Additionally, the licensee's prompt investigation identified that Byron Station did not follow industry standard configurations for chemical fill l connections. Industry standard caustic system connections utilize stainless steel l quick-disconnect fittings while acid system connections utilize four-bolt flange l
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connections. Byron Station, however, utilized four-bolt flange connections for both acid and caustic systems. The inspectors concluded an event of potentially significant industrial safety consequences was averted by the security guard's and operators'
persistent questioning of the truck driver. In addition, the inspectors concluded that mis-communication and inadequate procedures resulted in a potential chemical transfer acciden As immediate corrective actions for this event, the licensee changed the tank connection configurations to match industry standard configurations for chemical fill connection Additionally, the licensee implemented procedural changes to require operator review of appropriate shipping documents in addition to other procedure enhancements. At the end of this inspection period, the licensee was preparing an event report evaluation to further identify root causes and recommend additional corrective actions. No violation of J
regulatory requirements were identifie Conclusions The inspectors concluded that miscommunication and inadequate procedures resulted in a potential chemical transfer accident, in which the licensee nearly transferred sodium hydroxide (a strong caustic solution) from a delivery truck into a storage tank already containing sulfuric aci O2.4 Failure to Follow a Procedure Resulted in a Radioactive Liauid Effluent Release from an Unintended and Untested Path Inspection Scope (71707)
The inspectors reviewed the circumstances surrounding an inadvertent liquid effluent release via an unintended and untested release path. The inspectors interviewed operations department personnel and reviewed Byron Chemical Control Procedure (BCP) 400-TWX01," Liquid Radwaste Release Form for Release Tank OWX01T,"
Revision 1 Observations and Findinas i On February 1,1999, a non-licensed operator performed a planned liquid effluent release evolution in accordance with BCP 400-TWX01. There were two flow paths utilized for liquid effluent releases, a "high flow" path arid a " low flow" path. Both flow paths share a common effluent radiation monitor. The isolation function of the release
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path is tested prior to each release. Byron Chemical Control Procedure 400-TWX01 directed using the " low flow" path based upon a release rate calculation of 174 gallons-per-minute; however, operators failed to follow the procedure and initiated the release via the "high flow" path. Although the "high flow" path isolation function was not tested as part of the planned liquid effluent release evolution, it was previously verified to be functional. Upon discovery that the release was being performed via an untested path, the non-licensed operator immediately contacted the control room and secured the release. The flow rate limit for the planned release was maintained throughout the release and the activity remained less than the radiation monitor setpoints calculated for the release. Therefore, no activity release limits were exceede The inspectors reviewed the licensee's apparent cause evaluation which concluded that the shift manager and field supervisor failed to have a qualified radioactive waste supervisor present when conducting the liquid effluent release, contrary to specific direction from the shift operations supervisor. The evaluation identified that the error in performing the liquid effluent release evolution occurred due to knowledge deficiencies on the part of a new field supervisor and an inexperienced operator. The evaluation additionally identified deficiencies with the procedure and operator trainin Technical Specification 6.8.1.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 1978. Appendix A of Regulatory Guide 1.33, Revision 2, February 1978, specifies liquid radioactive waste system discharge to effluents as an example of a procedure for control of radioactivity. Byron Chemical Control Procedure 400-TWX01, " Liquid Radwaste Release Form for Release Tank OWX0)T," Revision 14, is one example of a liquid radioactive waste system discharge to effluents procedure. The operstor's use N the "high flow" path rather than the " low flow" path as specified in BCP 400/IV/X01 is a violation of TS 6.8.1.a for failure to implement the procedure. This Severity LeveliV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/455-99003-01(DRP)). This violation is in the licensee's corrective action program as problem identification form (PlF) B1099-0036 Conclusions The inspectors concluded that the licensee failed to control the configuration of the ,
treated waste system during a planned liquid effluent release evolution by performing the release via an unintended and untested release path. A Non-Cited Violation was issued for the licensee's failure to implement Byrcn Chemical Control Procedure 400-TWX01, " Liquid Radwaste Release Form for Release Tank OWX01T,"
Revision 1 Miscellaneous Operations issues (92700,92901)
08.1 (Closed) Violation 50-454/455/98005-04(DRP): "Failum to Parform Safety Evaluation for Venting CV [ Chemical and Vo!Lme and Control) System to the VCT [ Volume Control Tank]." The Fcensee had failed to perform a safety evaluation in accordance with 10 CFR 50.59 prior to venting the VC f into the gaseous waste processing system. This issue was previously discussed in NRC Inspection Report 50-454/98017(DRP);
50-455/98017(DRP) and the inspectors concluded that the licensee's subsequent safety evaluation which identified no unreviewed safety questions was acceptable. The
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licensee completed a root cause evaluation to address several safety evaluation process f ailures that had been identified previously by the inspectors and an off-site review committee. The inspectors reviewed the licensee's corrective actions identified in the root cause report and founo them to be acceptable. This violation is clese .2 (Closed) LER 50-454/455/98001: " Exceed Licensed Power Level Due to Calorimetric Instrument Discrepancy." The licensee identified a discrepancy with the steam generator blowdown flow totalizer recorder scaling which affected the accuracy of the thermal power calorimetric calculation in a non-conservative direction. The licensee concluded that this discrepancy resulted in Unit 2 exceeding licensed thermal power level. Although the licensee did not perform a detailed review of Unit 1 power history, it was also concluded that this scaling discrepancy resulted in Unit 1 exceeding licensed thermal power level. The Byron Unit 1 Facility Operating License (No. NPF-37) and Unit 2 Facility Operating License (No. NPF-66), License Condition 2.C.(1) authorize, in part, the licensee to operate the facility at reactor core power leveb r.M in excess of 3411 megawatts thermal (100 percent). As a result of the aNve totalizer scaling discrepancy, both Unit 1 and Unit 2 were operated in excess of their licensed thermal power levels which is a violation of the operating licenses. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the MC Enforcement Policy (50-454/455-99003-02(DRP)). The inspectors reviewed the licensee's corrective actions for this event and found them to be ermptable. This LER is close II. Maintenance
. M1 Conduct of Maintenance M A Emeraency Diesel Generator Maintenance
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. In_s;;ection Scope (62707)
The inspectors observed portions of maintenance performed on the1 A emergency die.T generator (EDG) and interviewed maintenance, work control and engineering personnel. The inspectors also reviewed Nuclear Station Work Procedure (NSWP)
WM-10. " Preparation of Maintenance Work Packages," Revision 2; NSWP M-09,
" Flexible Metal Hose Installation," Revision 2; and, Work Request (WR) 970110236-01,
" Install Flexible Hoses on Fuel Oil Drain Lines Per DCP [ Design Change 4
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Procedure] 9600257." Observations and Findinas On March 1,1999, the licensee removed the 1 A EDG from service to replace portions of the engine's fuel oil drain line with flexible hoses in accordance with WR 970110236-01 and DCP 9600257. Additionally, a section of tubing on the fuel oil pump suction was replaced to eliminate leaking fitting The inspectors identified that the circumference of replacement tubing on the fuel oil pump suction line appeared to be out-of-round and addressed this concern with engineering department personnel. In response to the inspectors' questions, the licensee measured the tubing and deterrnined that it had not been properly bent
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because the tube bending technique used dvd not control the ovality of the tube. The inspectors concurred with the licensee's conclusion that the resultant reduction in flow area was sufficiently small, su':h that it did not produce a restriction to fuel delivery to the engine that would impact its operability. The licensee concluded that the tubing should be replaced with a properly bent tube and entered the issue into their corrective action program.
, Additionally, the inspectors reviewed WR 970110236-01 for the modification to the fuel oil drain line and noted that the DCP requirements for post construction tcsting were originally incorporated by reference to NSWP M-09. Nuclear Station Work Procedure M-09 contained the licensee's requirements for installation and inspection of flexible metal braided hose assemblies and included a requirement to verify the correct bend radius of the hoses af ter installation. The inspectors identified that by making an unauthorized minor revision to the work package instructions, a mechanical maintenance first-line supervisor deleted the use of NSWP M-09. .n the work instructions. As a result, proper verification of hose cond radii was not perfcrmed upon completion of the work. Nuclear Station Work Procedure WM-10 required that for a change of technical requirements, such as acceptance criteria, the work package be returned to the work planner for processing a major revision. The inspectors determined that the licensee failed to appropriately process a major revision to the work instructions. The inspectors also noted that although a specific refe.ence to the vendor's catalog was provided in the DCP for maintenance planning personnel to obtain the correct bend radius, the proper bend radius was not translated to the work instructions and when asked, mechanical maintenance personnel did not know the correct bend radius for installation of the hose In response to the inspector's questions, the licensee inspected the hoses and determined that the minimum bend radius for two of the engine fuel c;l drain line hoses installed on the 1 A EDG had been violated during instailation, wh.'ch resulted in the hoses remaining out-of-round af ter they were installed. ine inspectors concurred with the licensee *r conclusion that there was no immediate failure concern for the distorted hoses and that the 1 A EDG remained operabl 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 circummances and shall be accomplished in accordance with these instructions, procedurcs, or drawings. Nuclear Station Work Procedure WM-10, Section 6.12, specified, in part, that the type of work package revision be determined in accordance with Exhibit A, Work Package Revision Guidelines." Exhibit A specified a change of technical requirements, such as acceptance criteria, as a major revision which required the work package be returned to the work planner for processing. The failure of the mechanical maintenance first-line supervisor to process a major revision to WR 970110236-01 is a violation of 10 CFR Part 50, Appendix B, Criteria V. This Severity Level IV violation is being treated as a Non-Citeu Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99003-03a(DRP)). This violation is in the licensee's corrective action program as PlF B1999-0103 .
4 Conclusions The inspectors concluded that minimum bend radii for two flexible hoses installed on the 1 A emergency diesel generator fuel oil drain line was violated during installation. This was due to a mechanical maintenance first-line supervisor improperly deleting the j
appropriate requirements for the installation and inspection of the hoses from the maintenance work instructions. In addition, mechanical maintenance personnel improperly bent replacement tubing for the fuel oil pump suction line as a result of poor bending practices, which resulted in the tube's circumference being out-of-round. A Non-Cited Violation was issue M1.2 Over-Torauina of 1 A Chemical and Volume Control Pumo Couplina Fasteners Inspection Scope (62707)
The inspectors reviewed the circumstances surrounding the over-torquing of fasteners on the 1 A chemical and volume control (CV) pump motor-to-gearbox coupling. The inspect. ors interviewed maintenance personnel and evaluated Root Cause Report 454-200-99-CAQ00004, " Failure to Follow Procedures Resulted in Over-Torquing of Coupling Bolting on the 1 A CV Pump." Addit!onally, the inspectors reviewed NSWP WM-10, * Preparation of Maintenance Work Packages," Revision 2; WR 980015575-01, "1 A CV Pump - Perform Various PM [ Preventive Maintenance]
Activities"; and Operability Assessment 99-00 Observations and Findings l
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On February 8,1999, durag the performance of a coupling inspection on the !
1 A CV pump in accordance witn WR 980015575-01, mechanical maintenance personnel questioned the torque value (50 foot-pounds) specified for fasteners on the motor-to-gearbox coupling. In response to the mechanics' questions, the first-line supervisor attempted to verify the torque value. The supervisor incorrectly determined that the specified torque value was wrong and directed the mechanics to torque the <
fasteners to 110 foot-pounds. Consequently, during the coupung reassembly, the l threads on three of the fasteners were stripped. The licensee's prompt investigation !
revealed, in part, that when the mechanical maintenance first-line supervisor modified '
the work instructions he failed to follow NSWP WM-10, which resulted in the 1 A CV pump motor-to-gearbox coupling fasteners being damaged during reassembly. The licensee replaced the fasteners for the motor-to-gearbox coupling and properly torqued them to 50 foot-pound Code of Federal Regulations Title 10 CFR Part 50, Appendix B, Criteria V, " Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by documentcd instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Nuclear Station Work Procedure WM-10, Section 6.12, specified, in part, that the type of work package revision be determined in accordance with Exhibit A," Work Package Revision Guidelines." Exhibit A specified a change of technical requirements, such as torque values, as a major revision which required the work package be returned to the work planner for processing. The failure of the mechanical maintenance first-line supervisor to process a major revision to WR 980015575-01 is a violation of 10 CFR Part 50, Appendix B, Criteria V. This
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i Severity Level IV violation is being treated as a Non-Cited Violation, consistent with l Appendix C of the NRC Enforcement Policy (50-454/99003-03b(DRP)). This violation is j in the licensee's corrective action program as PIF B1999-0045 J l
In addition, the work instructions specified in WR 980015575-01 were not appropriate to the circumstances, in that, the wrong torque value was specified for the 1 A CV pump pear-to-pump coupling fasteners (50 foot-pounds versus 23 foot-pounds). The
.. ensee's investigation revealed that this error existed in maintenance work instructions for all four of the CV pumps, such that the gear-to-pump coupling fasteners had l previously been over-torqued on each of the pumps. As a result, the licensee l performed an operability assessment, which concluded that the additional stresses on the fasteners were acceptable and would not result in any premature failures. The licensee inspected the fasteners for the 1 A CV pump gear-to-pump coupling and re-torqued them to 23 foot-pounds as required. Additionally, the licensee has committed to inspecting the fasteners for the gear-to-pump coupling on each of the other CV pumps at the next opportunity and verifying that correct torque values are specified in maintenance work instructions for all safety-related couplings. Since the incorrect torque value did not result in degradation of the coupling, the inspectors concluded that the failure had minimal safety significance This failum constitutes a violation of minor significance and is not subject to formal enforcement actio Conclusions The insoectors concurred with the licensee's conclusions that the 1 A chemical and volume control (CV) pump motor-to-gearbox coupling fasteners were over-torqued. This was due to a mechanical maintenance first-line supervisor improperly revising the torque value specified in the maintenance work instructions. In addition, mechanical maintenance personnel over-torqued the 1 A CV pump gear-to-pump coupling fasteners due to an incorrect torque value being specified in the maintenance work instructions. A Non-Cited Violation was issue M1.3 Surveillance Test Observations Inspection Scoce (61726)
The inspectors interviewed operations ar.u engineering 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 0.5-2.CS.1-2 Unit 1 Train B Containment Spray System Valve Stroke Test l
. 1BOSR 3.1.5-1 Unit 1 Train A Solid State Protection System Bi-Monthly Surveillance (Staggered)
. 1BOSR 3.2.7-613B Unit 1 ESFAS Instrumentation Slave Relay Surveillance (Train B Containment isolation Phase A - Relay K613)
. 1BOSR 3.2.7-614B Unit 1 ESFAS Instrumentation Slave Relay Surveillance (Train B Containment Isolation Phase A - Relay K614)
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. 1BOSR 8.1.2-2 1B Diesel Generator Operability Monthly (Staggered) and Semi Annual (Staggered)
Surveillance
. 2BOSR 8.1.2-1 2A Diesel Generator Operability Monthly (Staggered) and Semi-Annual (Staggered)
Surveillance
. 1BVSR AF-3 Unit 1 Simultaneous Start of both AF [ Auxiliary Feedwater] Pumps with Flow to the Steam Generators
. 1BVSR DG-8 1B Diesel Generator Engine Analysis Observations and Findinas On February 26,1999, the inspectors observed the performance of 1BVSR AF-3 and noted that Unit 1 responded as designed throughout the test. However, during the subsequerv review of the surveillance test data, the inspectors identified a calculational error in the acceptance criteria for the surveillance test procedure. Specifically, the system engineer made a sign eiror during the calculation which resulted in the calculated deviation being of the same magnitude but the opposite sign (e.g.,
-4.85 percent instead of 4.85 percent). While this error did not change the overall results of the test, the inspectors noted that since the acceptance criteria for the deviation from the manufacturer's pump curve was -10 percent to '7.5 percent, this type of error could havs taasked an unacceptable test result. The engineer subsequently put this issue into the licensee's corrective action program as PlF B1999-0095 Conclusion The inspectors concluded that the observed surveillance tests were performed well and satisfied the requirements of the Technical Specifications. The inspectors identified an error in the calculation of the acceptance criteria for the auxiliary feedwater pump full-flow testing to the steam generators; however, the error did not change the overall results of the tot and were considered of minor significanc M8 Miscellaneous Maintenance issues (92700,92902)
M8.1 (Closed) LER 50-454/97022: "SSPS [ Solid State Protection System] Logic Testing Not Performed Due to inadequate Testing Procedure." During a review of Generic Letter 96-01," Testing of Safety Related Logic Circuits," the licensee determined that existing SSPS surveillance testing was inadequue because it did not test all required memory logic circuit functions. The inadequate testing involved the use of universal logic boards in a memory configuration in conjunction with the SSPS internal semi-automatic tester designed by Westinghouse. There were three memory circuit functions that were inadequately tested: the Source Range Automatic P-10 Block, Feedwater Isolation on P-14 Steam Generator Hi-Hi Level, and Feedwater isolation on Safety injection. The original testing design prescribed by the vendor did not fully verify proper operation of the SSPS logic cards. The testing results prior to this event therefore could have masked failures not detectable by the surveillance testing. The problem had existed in the SSPS since initial plant startup. The licensee has subsequently added new testing criteria to the SSPS surveillance testing procedure o successfully completed testing of the functions. Additionally, the licensee has completed their review of Generic Letter 96-01 safety-related circuitry testing. The
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licensee reported this event as a condition prohibited by the plant's TS in accordance with 10 CFR 50.73(a)(2)(i)(b).
The licensee's failure to adequately test the Source Range Automatic P-10 Block, Feedwater Isolation on P-14 S!eam Generator Hi-Hi Level, and Feedwater Isolation on Safety injection functions is considered to be a violation of TS 4.3.1.1 and TS 4.3. This failure constitutes a violation of minor significance and is not subject to formal enforcement action. Although the licensee's testing of the SSPS circuitry was not in accordance with the TS, the inspectors concluded that actions taken both immediate and long term sufficiently demonstrated SSPS operability as well as ar; appropriate focus on safety. The inspectors reviewed the licensse's corrective actions for this event and found them to be acceptable. This LER is close M8.2 (Closed) Escalated Enforcement item 50-454/99002-02(DRP): " Human Performance Errors Resulted in an Invalid High Energy Line Greak (HELB) Signal and Actuation During Temperature Switch Calibration." On January 20,1999, during the performance of Byron Instrument Surveillance (BIS) 3.3.12-201," Surveillance Calibration of Steam Generator Blowdown Area Temperature Switches (SD) Environmentally Qualified,"
Revision 5, an unexpected HELB isolation of the Unit 1 steam generator blowdown system occurred as a result of multiple human errors and inadequate communications practices. The errors included installing a fuse at the wrong step in the procedure and connecting and electrical jumper on wrong terminals. These deficiencies were discussed in detail in NRC Inspection Report 50-454/99002(DRP); 50-455/99002(DRP).
Technical Specification 6.8.1.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 1978. Appendix A of Regulatory Guide 1.33, Revision 2, February 1978, specifies that procedures are required for each surveillance test listed in TS. Byron Instrument Surveillance 3.3.12-201 is the implementing procedure foi :ne HELB isolation instrumentation operability testing required by TS 4.3.3.12. The installation of fuse FU 31 at the wrong step of the procedure and connection of the jumper across terminal points 1 and 2 at junction box 1JB1743A instead of terminal points 11 and 12 are both examples of a violation of TS 6.8.1.a. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99003-04(DRP)).
This violation is in the licensee's corrective action program as PIF B1999-0020 Ill. Enaineerina E2 Engineering Support of Facilities and Equipment E Safety-Related Instrumentation Outof-Tolerance Trendina Inspection Scope (37551)
The inspectors examined the licensee's development of an instrument out-of-tolerance (OOT) trending program. The inspectors interviewed engineering department personnel, reviewed instrument OOT identification and trending data, and reviewed Byron Administrative Procedure (BAP) 500-25," Instrument Out of Tolerance (OOT) Trending Program," Revision 1. Additionally, the inspectors interviewed nuclear !
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oversight department personnel and reviewed nuclear oversight departme.-t assessments relating to the instrument OOT trending progra Observations and Findino On October 30,1998, the licensee implemented an instrument OOT trending program to l identify, trend, and evaluate instrument OOT conditions which occur during calibration and maintenance of plant instrumentation. The inspectors noted that instrument OOT conditions were identified by maintenance personnel and were documented in the station's PIF process. Additionally, the inspectors noted that the licensee had begun evaluating long term trending data to identify potential instrument failures, to increase calibration frequencies, and to identify corrective actions to preclude repetitive OOT condition The licensee identified early in their evaluation process that many expanded as-found instrument tolerances were overly restrictive which resulted in a significant number of PlFs written documenting OOT conditions that were not indichtive of actual instrument problems. The licensee additionally identified fifty instruments which required further long term evaluation and one pressure gauge which required replacemen The inspectors also reviewed assessments performed by the station's nuclear oversight department during the past 6 months relating to the instrurnent OOT trending progra The inspectort noted that nuclear oversight department assessments provided constructive and timely recommendations which contributed to the development of the licensee's program. Specific nuclear oversight department recommendations regarded .
the inclusion of historical data for trending instrument performance, communication with J maintenance personnel to evaluate appropriate calibration frequencies for instruments, and the method for tracking instruments identified by the program as requiring additional engineering evaluatio Conclusions The inspectors concluded tha' the licensee's process for implernenting and administering the instrument OOT trending program adequately identifies, trends, and evaluates instrument OOT conditions. Additionaliy, the inspectors noted that assessments performed by the station's nuclear aversight department during the past 6 months provided construct;ve and timely recommendations which contributed to the development of the licensee's progra E2.2 Temporary Modifications Insoection Scope (37551)
The inspectors interviewed operations and engineering department personnel; reviewed ;
applicable portions of the UFSAR and TS; and eva!uated the following temporary j modification '
. 93-1-003 Thermocouples on Various RH [ Residual Heat RemovalyRY
[ Pressurizer] Lines a 93-2-022 Thermocouples on Various RH [ Residual Heat RemovalyRY
[ Pressurizer) Lines
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a 98-1-019 Repair Rod Drive Position Indication Electrical Connections
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98-2-03L Install a Resistor and Disconnect Cable 2RC708 to Allow Other Sensors to Function
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98-1-060 inntall Temporary Hose and Pressure Gauge at 1DG5008A/B Test Valves
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98-2-061 Install Temporary Hose and Pressure Gauge at 2DG5008A/B Test Valves
. 99-1-002 Install Temporary Jumper to Bypass Reactor Trip Breaker 'A'
Shunt Trip Test Pashbutton I Conclusions The inspectors concluded that the temporary modifications reviewed were generally well l controlled and each temporary modification had an action plan for remova E8 Miscellaneous Engineering issues (92700,92903)
E (Closed) LER 50-454/97016-00 and 01: " Potential Failure of Westinghouse Fuel Caused by inadequate Design Code." On October 28,1997, Westinghouse notified Byron Station that based upon current fuel design modeling for a generic limiting analysis, there was a potential to not meet the 10 CFR 50.46 acceptance criteria for maximum cladding oxidation of 17 percent, for Integral Fuel Burnable Absorber (IFBA)
fuel which was in the second cycle of operation. Westinghouse detennined that during the second half of the operating cycle some fuel may potentially exceed a design criterion that the fuel rod pellet-to-gap shall not re-open. If the gap re-opening criterion were exceeded, the 17 percent local oxidation limit following a postulated loss-of-coolant accider't cot.'d be exceeded. The licensee reported this issue as a condition potentially outside the design basis of the plant in accordance with 10 CFR 50.72(a)(2)(ii).
On Decembe 15,1997, Westinghouse completed analyses for Unit 2 Cycle 7 using plant specific conditions and concluded that all fuel design limits and 10 CFR 50.46 criteria would be met for the remainder of Cycle 7. On June 5,1998, Westinghouse completed anal)ses for Unit 1 Cycle 9 using plant specific conditions and concluded that fuel rod pellet-to gap opening would occur; however,10 CFR 50.46 criteria would still be l met for Cycle 9. Tt:e inspectors reviewed the licensee's corrective actions for this event l and found them to be acceptable. This LER is close E8.2 (Closed) Escalated Enforcement Item 50-454/455/98025-04(DRP): " Failure to include Valves in the Inservice Test (IST) Program." As documented in NRC Inspection Report 50-454/98025(DRP); 50-455/98025(DRP), the inspectors identified that the licensee failed to include the auxiliary feedwater pump discharge valves,1/2AF004A/B, within the scope of the IST program as required by ASME/ ANSI OM Part 10," Inservice Testing of Valves in Light-Water Reactor Power Plants." In response to the inspect questions, the licensee determined that numerous valves in the auxiliary feedwate essential service water, and safety injection systems were also not included within the secpe of the IST program, but should have been; consequently, the licensee perfurmed the required testing and initiated PlFs B1998-05242, B1999-00864, and B1999-0086 The inspectors concluded that the numerous examples of the failure to include passive valves within the IST program represented a progrcmmatic deficiency with the scoping of the IST progra Technical Specification Surveillance Requirement 4.0.5.a states, in part, that inservice .
testing of ASME Code Class 1,2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code.Section XI !
of the ASME Boiler and Pressure Vessel Code, IWV-1100, " Valve Testing," states that valve testing shall be performed in accordance with the requirements stated in ASME/ ANSI OM Part 10. ASME/ ANSI OM Part 10," Inservice Testing of Valves in Light-Water Reactor Power Plants," Section 1, requires, in part, that active or passive valves which are required to perform a specific function in shutting down the reauor to the cold shutdown condition, in maintaining the cold shutdown condition, or in mitigating the consequences of an accident are required to be included in the IST program. The ,
auxiliary feedwater pump discharge valves,1/2AF004A/B, are ASME Code Class 3 l valves which have a passive open safety function in mitigating the consequences of accidents. The failure to include the auxiliary feedwater pump discharge valves, 1/2AF004A/B, in the IST program is a violation of TS 4.0.5.a. This Severity Level IV l violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/455/99003-05(DRP)). This violation is in the i licensee's corrective action program as PIFs B1998-05172, B1998-05242, j B1999-00864, and B1999-0086 .3 (Closed) Violation 50-454/455/98004-03bIDRS): "Faiiure to implement an Effective Program to Resolve Long Standing OOT [Out-of-Tolerance] Issues. The licensee had failed to implement an effective corrective action program to evaluate and address repetit.ve instrument OOT conditions adverse to quality. The inspectors discusseo the licensee's dew lopment of an instrument OOT trending program in Section E2.1 of this report and concluded that the licensee's program was acceptable. This violation is close IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R Radioloal. cal Protection Practices Inspection Scope (71750) ,
The inspectors & served the posting of radiation areas, the control of locked high radiation areas, the application of As-Low-As-Reasonably-Achievable (ALARA)
principles, and the radiation work practices of station personne Observations and Findinos The inspectors reviewed the postings of radiologically controlled areas, performed confirmatory surveys in selected areas, and reviewed the radiation protection foreman's log. The inspectors concluded that radiologically controlled areas were properly posted for the condition that existed in the areas. Rope boundaries, swing gates (where applicable) and signs were properly maintained. Locked high rmtion area doors were locked and were properly controlled by radiation protection persane The inspectors reviewed the ALARA controls used on several potentially high dose activities. For example, the inspectors noted that prior to entry into locked high radiation
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O areas, radiation protection persennel briefed personnel on the dose rates in the affected areas, identified areas to avoid, and identified the routes that resulted in the lowest dose, The inspectors observed proper radiation worker practices by maintenance personnel assigned to work on the Unit 1B chemical and volume control pump. The workers properly used anti-contamination clothing, properly performed frisking upon exit from the contaminated area, and properly wore and monitored dosimetry. The inspectors also noted that personnel properly monitored equipment out of radiologically controlled areas and made proper use of personnel radiation monitors at the exit of radiologica!!y controlled areas and at the gate hous Conclusions The inspectors concluded that radiologically controlled areas were properly posted; locked high radiation area doors were locked and properly controlled by adiation protection personnel; radiation workers demonstrated proper work practices to control the spread of radioactivity; and ALARA principles, such as briefings to minimize exposure to personnel ware effectively utilize 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 March 29,1999. 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 .
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PARTIAL LIST OF PERSONS CONTACTED Licensee B. Adams, Regulatory Assurance Manager i M. Jurmain, Maintenance Manager W. Kouba, Engineering Manager l
W. Levis, Site Vice Pres dent j R. Lopriore, Station Man ger W. McNeill, Radiation Protection Manager T. Schuster, Work Control Manager M. Snow, Operations Manager J. Stone, Nuclear Oversight Manager I I
I INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering I IP 6172 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 Facilities IP 92901: Follow-up Plant Operations IP 92902: Follow-up Maintenance IP 92903: Follow-up Engineering l
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r ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-454/455/99003-0; NCV failure to follow procedure for radioactive liquid effluent release from an untested path 50-454/455/99003-02 NCV exceed licensed power level due to calorimetric instrument discrepancy 50-454/99003-03a NCV failure to process a major revision to work instructions to delete requirements for installation and inspection of safety related and regulatory related flexible metal braided hose assemblies 50-454/99003-03b NCV failure to process a major revision to work instructions to change the la chemical and volume control pump motor-to-gearbox coupling torque value 50-454/99003-04 NCV failure to follow Byron instrument Surveillance 3.3.12-201 resulted in an invalid high energy line break signal and actuation during temperature switch calibration 50-454/455/99003-05 NCV failure to include valves in the Inservice Test (IST)
Program Closed 50-454/455/99003-01 NCV failure to follow procedure for radioactive liquid effluent release from an untested path 50-454/455/98005-04 VIO failure to perform safety evaluation for venting CV
[ Chemical and Volume and Control] System to the VCT
[ Volume Control Tank)
50-454/455/99003-02 NCV exceed licensed power level due to calorimetric instrument discrepancy 50-454/455/98001 LER exceed licensed power level due to calorimetric instrument discrepancy j 50-454/99003-03a NCV failure to process a major revision to work instructions to delete requirements for installation and inspection of safety-related and regulatory related flexible metal braided hose assemblies 50-454/99003-03b NCV failure to process a major revision to work instructions to change the 1 A chemical and volume control pump motor-to-gearbox coupling torque value
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ITEMS OPENED, r!.OSED, AND DISCUSSED Closed (cont'd)
I iO-454/97022 LER SSPS [ Solid State Protection System] logic testing not j performed due to inadequate testing procedure 50-454/99003-04 NCV failure to follow Byron Instrument Surveillance 3.3.12-201 resulted in an invalid high energy line break signal and actuation during temperature switch calibration 50-454/99002-02 eel auma.i performance errors resulted in an invalid High i Energy Line Brea!- (HELB) signal and actuation during l temperature switch calibration 50-454/97016-00 and 01 LER potential failure of Westinghouse fuel caused by 1 inadequate design code I 50-454/455/99003-05 NCV failure to include valves in the Inservice Test (IST)
program 50-454/455/98025-04 eel failure to include valves in the Inservice Test (IST)
program 50-454/455/98004-03b VIO failure to implement an effective program to resolve long-standing OOT [Out-of-Tolerance) issues l
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l LIST OF ACRONYMS USED AF Auxiliary Feedwater ALARA As-Low-As-Reasonably-Achievable ANSI American National Standards Institute ASME American Society of Mechanical Engineers BAP Byron Administrative Procedure BCP Byron Chemical Control Procedure BGP Byron General Operating Procedure l
l BIS Byron Instrument Surveillance
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BOSR Byron Operating Surveillance Requirement BVSR Byron Engineering Surveillance Requirement CFR Code of Federal Regulations CV Chemical and Volume Control EDG Emergency Diesel Generator DCP Design Change Procedure -
DOST Diesel Oil Storage Tank DRP Division of Reactor Projects ERS Division of Reactor Safety ESFAS Engineered Safetr Feature Actuation Signal FC Fuel Cooling FW Feedwater HELB High Energy Line Break HLA Heightened Level of Awareness IFBA Integral Fuel Burnable Absorber IST Inservu Testing LCO Limiting Condition for Operation {
LER Licensee Event Report j NCV Non-cited Violation NRC Nuclear Regulatory Commission NSWP Nuclear Station Work Procedure OOT Out-of-Toleranco PDR Public Document Room PIF Problem Identification Form RH Residual Heat Removal RP&C Radiological Protection and Chemistry SD Steam Generator Blowdown SFP Spent Fuel Pool SSPS Solid State Protection System TS Technical Specification UFSAR Updated Final Safety Analysis Report VCT Volume Control Tank VIO Violation WR Work Request 24