IR 05000454/1999008
| ML20210A338 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 07/16/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20210A319 | List: |
| References | |
| 50-454-99-08, 50-454-99-8, 50-455-99-08, 50-455-99-8, NUDOCS 9907220072 | |
| Download: ML20210A338 (22) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION lil Docket Nos:
50-454; 50-455 License Nos:
50-454/455-99008(DRP)
Licensee:
Commonwealth Edison Company Facility:
Byron Generating Station, Units 1 and 2 l
I Location:
4450 N. German Church Road Byron,IL 61010 Dates:
May 11 - June 21,1999 Inspectors:
E. Cobey, Senior Resident inspector B. Kemker, Resident inspector
T. Tongue, Project Engineer, Rlli C. Phillips, Braldwood Senior Resident inspector J. Adams, Braidwood Resident inspector C. Thompson, Illinois Department of Nuclear Safety Approved by:
Michael J. Jordan, Chief i
Reactor Projects Branch 3 Division of Reactor Projects l
9907220072 990716 PDR ADOCK 05000454 O
PDR L
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EXECUTIVE SUMMARY Byron Generating Station Units 1 and 2
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NRC Inspection Report 50-454/99008(DRP); 50-455/99008(DRP)
This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 6-week period ofinspection activities by the resident staff and region based inspectors.
Operations The inspectors concluded that operations of the facility were conducted in a safe,
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professional, and controlled manner. Operators generally adhered to the station's standards for control room conduct, prccedural adherence, annunciator response, and use of three-way communications. (Section 01.1)
The inspectors concluded that the licensee's response to the Unit 1 automatic reactor
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trip from full power was excellent. Operators effectively controlled and stabilized plant parameters and all plant safety-related systems operated as designed. The shift manager and unit supervisor demonstrated strong command and control throughout the event. (Section 01.2)
The inspectors concluded that the Unit i reactor startup was conducted in a safe and
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controlled manner. Operators precisely controlled reactivity manipulations, followed plant startup procedures, performed peer and self-checks, and generally used proper three-way communication techniques. The inspectors also concluded that senior reactor operators demonstrated strong command and control, directly supervised reactivity manipulations, and provided effective oversight of the startup activities.
(Section 01.3)
Maintenance / Surveillance The inspectors concluded that instrument maintenance technicians who performed
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setpoint scaling adjustments associated with the Unit 2 operating temperature increase were thoroughly knowledgeable of the tasks and professionally completed the work.
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The inspectors concluded that the observed surveillance tests were generally performed
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well and satisfied the requirements of the Technical Specifications. The inspectors identified oil leaking from the 28 auxiliary feedwater pump's outboard motor end bearing while the pump was running, which was caused by operators over-filling the pump's oil reservoir. The inspectors concurred with the licensee that over-filling the reservoir did not affect the operability of the pump; however, it was a poor practice. (Section M1.2)
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The inspectors concurred with the licensee's conclusion that on May 13,1999, an
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instrument maintenance technician incorrectly removed instrument power fuses from power range nuclear instrument channel N-42 while performing a calibration of channel N-43. This was due to inattention to detail and a failure to adhere to station management's expectations for self-checking and peer-checking. This procedural adherence error resulted in an Unit 1 automatic reactor trip from full power. A Non-Cited Violation was issued. (Section M4.1)
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o The inspectors concurred with the licensee's conclusion that position verification of four
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feedwater system containment penetration high point vent valves had not been performed at the frequency required by Technical Specification Surveillance,
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Requirement SR 3.6.3.3. A Non-Cited Violation was issued. (Section M8.2)
Enoineerina
' The inspectors concluded that an engineering design change to Unit 2 which increased
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the unit's operating temperature to 583 degrees Fahrenheit was well prepared and executed. Specifically, setpoint scaling changes were appropriate for the operating temperature increase, the procedure provided clear instructions to perform the work,
. and operators received an appropriate level of training prior 'o implementing the change to the facility. (Section M1.1)
The inspectors concluded that the licensee appropriately evaluated a weld failure where
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the 1B reactor coolant system (RCS) loop bypass vent valve (1RUO29B) assembly is attached to the RCS loop bypass line. In addition, the licensee implemented acceptable corrective actions to prevent recurrence of a similar vibration induced fatigue failure.
(Section E1.1)
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The inspectors concluded that the licensee failed to incorporate post-construction testing requirements from engineering design change packages into work request instructions and failed to complete the post-construction testing requirements after replacing the fuel oil filter and strainer assemblies on both Unit i diesel generators. One example of a Non-Cited Violation was issued. (Section E3.1)
The inspectors concluded that the licensee failed to incorporate testing requirements
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from engineering design change packages into work request instructions and to accomplish the post-construction and post-modification testing requirements following modifications to the Unit 1 containment recirculation sump outlet isolation valves and the Unit 1 main steam isolation valves. Two additional examples of a Non-Cited Violation were issued. (Section E3.2)
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Report Details Summarv of Plant Status Tha licensee operated Unit 1 at or near full power until May 13,1999, when the unit
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experienced an automatic reactor trip due to an inadvertent power range high flux trip signal during the performance of instrumentation and control surveillance testing. Following the reactor trip, the licensee placed Unit 1 in cold shutdown mode to repair a reactor coolant system pressure boundary leak. The licensee synchronized the unit to the grid on May 18,1999, and the unit was operated at or near full power for the remainder of the
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. Inspection period.
' Unit 2 operated at or near full power for the duration of the inspection period.
l. Operations
Conduct of Operations 01.1 General Observations (71707)
i During this inspection period, the inspectors routinely observed the conduct of plant operations and noted that operators generally adhered to the station's standards for control room conduct, procedural adherence, annunciator response, and use of three-way communications. The inspectors noted that shift tumover briefings were performed well. Plant status, limiting conditions for operation, and major work activities were discussed in appropriate detail. The inspectors concluded that operations of the facility were conducted in a safe, professional, and controlled manner.
' 01.2 Response to Unit 1 Reactor Trio Durina Performance of Nuclear Instrumentation Calibration
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Inspection Scooe (93702)
' The inspectors responded to the control room and observed the licensee's plant recovery activities following an automatic reactor trip from full power. The inspectors reviewed the circumstances surrounding the event and interviewed operations department personnel.
b.
Observations and Findinas On May 13,1999, Unit i experienced an automatic reactor trip from full power due to an inadvertent power range high flux trip signal during the performance of instrumentation and control surveillance testing. Instrument maintenance personnel were performing a calibration of power range nuclear instrument channel N-43 when an instrument maintenance technician incorrectly removed instrument power fuses from power range channel N-42 and caused the reactor trip. This human performance error is further discussed in Section M4.1 of this report.
The inspectors observed that operators effectively controlled and stabilized plant parameters and that all plant safety-related systems operated as designed. The inspectors noted that the shift manager and unit supervisor demonstrated strong
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command and control throughout the event. In addition, the inspectors noted that l
operators adhered to the station's standards for procedural adherence, annunciator j
response, and generally used three-way communications. The licensee initiated a prompt investigation and made a 4-hcur non-emergency report to the NRC in accordance with 10 CFR Part 50.72(b)(2)(ii).
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- The inspectors noted that steam dump valve 1MS004K failed to re-close in response to l
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plant cooldown, which resulted in reactor coolant system temperature decreasing slightly below the P-12 setpoint of 550 degrees Fahrenheit (*F) for a short period of
' time. When temperature reached 550*F (decreasing), the remaining steam dumps automatically isolated as designed. Operators promptly identified the malfunction and l
manually isolated the 1MS004K steam dump valve. The remaining steam dump valves
. controlled plant temperature as designed. The inspectors noted that the cooldown was not excessive since normal operating temperature was 557'F.
Other ronsafety-related equipment failures during the event included eleven failed
. feedwater heater relief valves. The feedwater heater relief valves failed as a result of
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water hammer when the feedwater header isolation valves closed in response to the reactor trip. The inspectors noted that nine feedwater heater relief valves failed i
following a Unit i reactor trip in September 1996. Although the relief valve failures were i
a distraction for plant operators, the failures did not result in a challenge to operators in
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controlling plant cooldown and the licensee replaced the failed relief valves prior to unit
. startup. The inspectors noted in NRC Inspection Report 50-454/455-96007(DRP)
following the September 1996 Unit i reactor trip that the licensee's engineering department had been pursuing a potential modification to alleviate the lifting and damaging of feedwater heater relief valves. At the end of this inspection report period, the licensee had not yet completed a review of altemative corrective actions for this problem.
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Conclusions The inspectors concluded that the licensee's response to the Unit 1 automatic reactor trip from full power was excellent. Operators effectively controlled and stabilized plant parameters and all plant safety-related systems operated as designed. The shift i
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manager and unit supervisor demonstrated strong command and control throughout the event.
01.3 Unit 1 Startuo Followina the Reactor Trio a.
Inspection Scoos (71707)
l On May 17 and 18,1999, the inspectors observed the Unit i reactor plant startup. The l
inspectors interviewed operations department personnel, observed the licensee's heightened level of awareness briefing and startup activities, and reviewed the following procedures: Byron General Operating Procedure (1 BGP) 100-2, " Plant Startup,"
Revision 23; 1BGP 100-2A1, " Reactor Startup," Revision 15; and 1BGP 100-3, " Power l-Ascension," Revision 27.-
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Observations and Findinas On May 17,1999, the inspectors observed the heightened level of awareness briefing for the Unit 1 startup ad noted that the briefing contained a sufficient level of detail to properly perform the ab tup, addressed identified hold points and startup termination criteria, and covered the chain of command and the roles and responsibilities of the
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The inspectors observed the operators performing startup activities in the control room, and noted that peer and self-checks were performed, procedures were followed, and proper three-way communication techniques were generally used. The inspectors observed effective reactivity management and noted that manipulations associated with reactor criticality and power ascension were precisely controlled, were directly supervised by a senior reactor operator, and were clearly communicated to the unit supervisor and nuclear engineers.
The inspectors noted that the unit supervisor demonstrated strong command and control. The unit supervisor remained in the "at-the-controls" area of the control room and appeared to maintain an overall perspective of control room activities. In addition, the inspectors observed the presence of senior management and nuclear oversight personnelin the control room during performance of the startup.
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Conclusions The inspectors concluded that the Unit i reactor startup was conducted in a safe and controlled manner. Operators precisely controlled reactivity manipulations, followed plant startup procedures, performed peer and self-checks, and generally used proper three-way communication techniques. The inspectors also concluded that senior reactor operators demonstrated strong command and control, directly supervised reactivity manipulations, and provided effective oversight of the startup activities.
Miscellaneous Operations issues (92700)
08.1 (Closed) Licensee Event Report (LER) 50-454/99001: " Depressing Both Feedwater Isolation Reset Pushbuttons Leads to LCO [ Limiting Conditions for Operation) 3.0.3
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Entry." The licensee identified that on April 22,1999, the requirements of Technical Specification (TS) 3.3.2 were not met while performing control rod drop testing with Unit 1 in mode 3 (hot standby), which resulted in er tiy :ato TS 3.0.3. During control rod drop testing, the reactor trip breakers were cycled numerous times. To prevent an unwanted feedwater isolation due to a reactor trip permissive interlock (P-4) signal, the licensee's p:st practice had been to temporarily depress and hold both feedwater isolation reset pushbuttons while opening the reactor trip breakers. This action blocked both trains of feedwater isolation from occurring during the P-4 interlock actuation and also rendered both trains of feedwater isolation actuation instrumentation inoperable. Prior to the licensee's implementation of improved Standard Tc hnical Specifications (ITS) on February 5,1999, this was not an issue because the feedwater isolation function was not required in mode 3. The licensee's process for implementing the ITS failed to recognize that the testing method would result in noncompliance with the TS. The inspectors concluded that this event was not safety significant. The
- licensee reported this issue as a condition prohibited by the plant's TS in accordance with 10 CFR 50.73(a)(2)(1)(B). In response to this event, the licensee committed to
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- d review procedures affected by more restrictive changes as a result of ITS
' implementation to ensure that all required changes were made. In addition, the licensee committed to make appropriate changes to the control rod drop testing procedure as well as other procedures where the reactor trip breakers are opened in mode 3 to utilize altemative methods to avoid TS 3.0.3 entry.
Code of Federal Regulations Title 10 Part 50, Appendix B, Criterion lil, " Design Control,"
requires, in part, that measures shall be established to assure that applicable regulatory requirements for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. The licensee's failure to incorporate the mode applicability change for TS 3.3.2 into appropriate operating and surveillance test procedures constitutes a violation of minor significance and is not subject to formal enforcement action. The inspectors reviewed the licensee's corrective actions for this event and found them to be acceptable. This LER is closed.
11. Maintenance M1 Conduct of Maintenance M1.1 Unit 2 Operatina Temoerature increase a.
Inspection Scope (37551. 62707)
The inspectors reviewed Special Plant Procedure (SPP)99-019, " Unit 2 T,,,[ Reference Temperature] Setpoint increase," and its associated work requests for the Unit 2 operating temperature increase from 581 *F to 583*F. The inspectors interviewed operations, engineering, and maintenance department personnel; observed the performance of setpoint scaling activities in the field; and reviewed applicable portions of the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications (TS).
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Observations and Findinas During the week of May 24,1999, the licensee performed SPP-019 to increase the Unit 2 full power operating temperature from 581 *F to 583*F. The operating
- temperature increase improved Unit 2 cycle efficiency and resulted in an increase of approximately 4 megawatts electric generator output.
~ The inspectors reviewed the procedure and the instrument loop setpoint scaling changes required to support the operating temperature increase with the cognizant engineer. The inspectors noted that the proposed temperature increase was within the plant's design operating window of 569.1 *F to 588.4*F and that the setpoint scaling changes were appropdate for the operating temperature increase. The inspectors noted that the procedure provided clear instructions to perform the work and that operators received an appropnate level of training prior to implementing the change to the facility.
The inspectors noted that SPP-019 was written to sequence the implementation of seven setpoint scaling work requests which changed the following instrumentation control and protection functions:
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Overpower Delta Temperaturc (OPAT)
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Overtemperature Delta Temperature (OTAT)
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Reactor Coolant System Reference Temperature (T,,,)
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Control Rod Speed and Direction
Pressurizer Level Program
Auctioneered High Average Temperature Alarm Setpoint
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Turbine Loading Stop Setpoint (C-16)
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Steam Dump Load Rejection Controller Setpoints
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The inspectors observed the performance of setpoint scaling adjustments in the field
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and noted that instrument maintenance technicians were thoroughly knowledgeable of the tasks and that the work was completed professionally. The inspectors noted that approximately 3 weeks prior to actually performing the work, the same instrument maintenance technicians completed a thorough review of the procedure to ensure that it would be workable and to familiarize themselves with the work. The inspectors
- considered the workability review process, as exemplified by the technicians'
preparation for SPP-019, to be a strength in the licensee's maintenance program which contributed to the successful completion of the maintenance evolution. The inspectors also noted that the instrument maintenance supervisor regularly interfaced with the technicians in the field to provide assistance and oversight of the work.
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Conclusions The inspectors concluded that an engineering design change to Unit 2 which increased the unit's operating temperature to 583 degrees Fahrenheit was well prepared and executed. Specifically, instrument maintenance technicians were thoroughly knowledgeable of the tasks and professionally completed the work; setpoint scaling changes were appropriate for the operating temperature increase; the procedure provided clear instructions to perform the work; and operators received an appropriate
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level of training prior to implementing the change to the facility.
M1.2 Surveillance Test Observations a.
Insoection Scooe (61726)
.The inspectors interviewed operations, engineering, and maintenance department personnel; reviewed the completed test documentation and applicable portions of the UFSAR and TS; and observed the performance of selected portions of the surveillance test procedures listed below. In addition, inspectors reviewed Byron Operating Department Policy 400-22," Oil Addition to Plant Equipment," Revision 4.
1BOSR 3.2.7411 A Unit 1 ESFAS [ Engineered Safety Features Actuation
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System] Instrumentation Slave Relay Surveillance (Train A Automatic Safety injection - K611)
1BOSR 5.2.2-1 Unit 1 ECCS [ Emergency Core Cooling System] Venting
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and Valve Alignment Monthly Surveillance 1BOSR 8.1.2-2 Unit 1 1B Diesel Generator Operability Monthly
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(Staggered) and Semi-Annual (Staggered) Surveillance
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1BVSR 5.2.4-4 Unit 1 ASME [American Society of Mechanical Engineers]
Surveillance Requirements for Residual Heat Removal Pump 1RH01P
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2BOSR 3.2.7-603B Unit 2 ESFAS Instrumentation Slave Relay Surveillance
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(Train B Automatic Safety injection - K603)
2BOSR 3.2.7-6438 Unit 2 ESFAS Instrumentation Slave Relay Surveillance
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(Train B Automatic Containment Spray - K643)
2BOSR 7.5.3-2 Unit 2 Diesel Driven Auxiliary Feedwater Pump Quarterly
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Surveillance 2BVSR 5.2.F.3-2 Unit 2 ASME Surveillance Requirements for Residual Heat L
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Removal Pump 2RH01PB b.
Observations and Findings While observing the Unit 2 diesel driven auxiliary feedwater pump quarterly surveillance test, the inspectors identified a lubricating oil leak from the pump's outboard motor end bearing. The oil was spraying from the pump shaft's outboard seal and dripping down
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into the pump's bedplate. The inspectors notified the non-licensed operator monitoring
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the pump who then contacted the control room to secure the pump. The apparent cause of the oil leak was overfilling of the oil reservoir. The inspectors noted that operators did not identify that the reservoir was overfilled prior to the test run. The inspectors also noted that Byron Operating Department Policy 400-22 required that whenever oil is added to a component by operators, an entry is to be made in an Oil Addition Log maintained in the shift manager's office. The inspectors reviewed the Oil Addition Log and noted that no entry had been made for the 2B AF pump since February 21,1998. The licensee noted that the log was not appropriately updated since I
an oil addition would have to have been made since the 28 AF pump was last run a month earlier. This policy is not documented in an approved station procedure'
therefore, no regulatory requirements were violated.
The inspectors discussed the potential affect on pump operability with the licensee. The licensee stated that with the oil reservoir over-filled, oil was forced back through a recirculation line to the bearing. Once the level of the reservoir went below the recirculation line connection to the reservoir, the oil leak would stop. The inspectors concurred with the licensee that the over-filled reservoir did not affect the operability of the pump; however, noted that it was a poor practice. After verifying proper oil reservoir level, the licensee re-performed the surveillance test satisfactorily. This issue is in the licensee's corrective action program as problem identification form (PlF) B1999-02315 c.
Conclusions The inspectors concluded that the observed surveillance tests were generally performed well and satisfied the requirements of the Technical Specifications. The inspectors identified oil leaking from the 28 auxiliary feedwater pump's outboard motor end bearing while the pump was running, which was caused by operators over-filling the pump's oil reservoir. The inspectors concurred with the licensee that over-filling the reservoir did not affect the operability of the pump; however, it was'a poor practice.
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Mainte' nance Staff Knowledge and Performance M4.1 Human Performance Error Durina Performance of Nuclear Instrumentation Calibration Results in Unit 1 Reactor Trio -
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. Inspection Scone (93702)
The inspectors reviewed the circumstances surrounding the Unit 1 automatic reactor trip from full power during the performance of Byron Instrument Surveillance Requirement (BISR) Procedure 3.1.6-200,"92 Day Surveillance Calibration of Power Range Nuclear Instrumentation System," Revision 1. The inspectors interviewed operations and
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maintenance department personnel and reviewed the licensee's prompt investigation into the event.
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Observations and Findings As discussed in Section 01.2 of this report, Unit i experienced an automatic reactor trip from full power on May 13,1999, due to an inadvertent power range high flux trip signal.
While performing an instrumentation and control surveillance test on channel N-43, an instrument maintenance technician incorrectly removed instrument power fuses from power range nuclear instrument channel N-42, which resulted in the reactor trip.
Instrument maintenance technicians performing the N-43 calibration removed the instrument power fuses for N 43 per step 25a of BISR 3.1.6-200. The technicians then spent approximately 30 minutes disconnecting cables in the back of the nuclear instrument cabinet ner steps 25b through 251 of the procedure. Upon retuming to the front of the cabinet, one of the technicians noted that instrument power fuses were installed on the adjacent nuclear instrument drawer and believed that he had forgotten to remove them from N-43.. However, the technician was looking at N-42 instead of N-43 and incorrectly removed the fuses from N-42. The inspectors reviewed the licensee's prompt investigation into the event and concurred with the licensee's conclusion that the instrument maintenance technician's error occurred as a result of inattention to detail and a failure to adhere to station management's expectations for selfa:hecking and peer-checking. The inspectors noted that following this event, the licensee conducted a " stand down" of maintenance work activities to discuss the event
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and to reinforce the importance of performing error reduction techniques.
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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 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 Requirement Procedure 3.1.6-200 is the implementing procedure for the 92-day surveillance calibration of the power range nuclear instrumentation system as required by TS surveillance requirement SR 3.3.1.6. The instrument maintenance technician's removal of instrument power fuses from nuclear instrument channel N-42 while calibrating channel N-43 during performance of BISR 3.1.6-200 is a violation of TS 5.4.1.a for failure to properly implement the procedure. This failure resulted in the Unit 1 automatic trip from full power. This Severity Level IV violation is being treated as a
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Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99008-01)). This violation is in the licensee's corrective action program as PlF B1999-01956.
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Conclusions The inspectors concurred with the licensee's conclusion that on May 13,1999, an instrument maintenance technician incorrectly removed instrument power fuses from power range nuclear instrument channel N-42 while performing a calibration of channel N-43. This was due to inattention to detail and a failure to adhere to station management's expectations for self-checking and peer-checking. This procedural adherence error resulted in an Unit 1 automatic reactor trip from full power. A Non-Cited Violation was issued.
M8 Miscellaneous Maintenance issues (92700)
M8.1 (Closed) LER 50-454/98010: " Failure to Perform In-Service Testing After Maintenance Due to Poor Communications." The licensee identified that on March 18,1998, while troubleshooting an increased stroke time for valve 1SD002F (steam generator blowdown containment isolation valve), maintenance alterations were performed that rendered the valve inoperable. The applicable TS action statement was not entered and post-maintenance verification stroke testing of the valve was not performed to assure the valve's operability. Valve 1SD002F is an air-operated valve which requires air pressure to overcome the force of a spring in order to open. The valve closes under spring tension; The valve was positioned to its safe (closed) position prior to the maintenance work. During troubleshooting, the valve's air supply line was disconnected and reinstalled. The inspectors concurred with the licensee's conclusion that there were no adverse affects on plant safety as a result of this event because the work involved only disconnecting and reconnecting the valve's air supply line. In addition,1SD002F was subsequently tested satisfactorily. The licensee reported this issue as a condition prohibited by the plant's TS in accordance with 10 CFR 50.73(a)(2)(1)(B).
Technical Specification Surveillance Requirement 4.6.3.1, which was applicable at the time this LER was written, required that the isolation valves specified in Table 3.6-1 shall be demonstrated operable prior to retuming the valve to service after maintenance,
. repair, or replacement tvork is performed on the valve or its associated actuator, control
. or power circuit by performance of a cycling test, and verification of isolation time. Valve 1SD002F was a valve specified in Table 3.6-1 of the TS. The licensee's failure to perform post-maintenance verification stroke testing of 1SD002F to assure its operability constitutes a violation of minor significance and is not subject to formal enforcement action. The inspectors reviewed the licensee's corrective actions for this event and found them to be acceptable. This LER is closed.
M8.2 (Closed) LER 50-454/99002: * Design Package Falls to Classify Feedwater Vent Valves as Containment isolation Valves and Results in Missed Technical Specification Surveillance." On May 9,1999, the licensee identified that four high polat vent valves (1FW118A/B/C/D), which had been added to the Unit 1 feedwater system during i
refueling outage B1R08, met the criteria for manual containment isolation valves and had not been verified closed every 31 days as required by TS Surveillance Requirement SR 3.6.3.3. The four valves were verified closed prior to Unit i entering mode 4 (hot shutdown) on February 27,1998. Unit 1 was operated from February 27,1998, to
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March 28,1999, in an operational mode that required the position of the four valves to be verified every 31 days. The licensee reported this issue as a condition prohibited by the plant's TS in accordance with 10 CFR 50.73(a)(2)(1)(B).
Technical Specification Surveillance Requirement SR 3.6.3.3 requires, in part, that each containment isolation manual valve, remote manual valve, and blind flange that is located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions be verified closed every 31 days. Valves 1FW118A/B/C/D were manual containment isolation valves meeting 10 CFR Part 50, Appendix A, General Design Criterion 57 and were required to be verified closed every 31 days. The licensee's failure to verify every 31 days that 1FW118A/B/C/D were closed while Unit 1 was operated in a mode requiring the valves to be verified closed is considered to be a violation of TS Surveillance Requirement SR 3.6.3.3. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99008-02)). The inspectors reviewed the licensee's corrective actions for this event and found them to be acceptable. This LER is closed.
M8.3 (Closed) LER 50-454/99003: " Automatic Reactor Trip Due to Human Error Dwing
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Surveiilance Procedure." This issue is discussed in Section M4.1 of this report. A Non-Cited Violation was issued. This LER is closed.
111. Engineering E1 Conduct of Engineering E1.1 Weld Failure on Unit 1 Reactor Coolant Svstem Looo Bvoass Vent Valve Assembiv Followina Modification a.
Insoection Scope (37551)
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The inspectors reviewed the circumstances surrounding the vibration induced fatigue
failure of a weld where the 1B reactor coolant system (RCS) loop bypass vent valve (1RC80298) assembly attached to the RCS loop bypass line. The inspectors interviewed engineering and operations department personnel and reviewed the following documents: Work Request (WR) 990050226," Installation of Freeze Seal on Line 1RC21BB-8" to Facilitate Maintenance Activity on Line 1RC23AB-%"";
WR 990050965, " Install Pipe Clamp Support to Restrict Vibration"; a.nd Design Change Package (DCP) 9900157, " Provide Additional Pipe Clamp for 1RC8029B to Restrict
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Vibration Movement."
b.
Observations and Findings During an inspection of the Unit 1 containment following a reactor trip on May 13,1999, the licensee identified a steam / water leak of approximately 50 drops per minute from a cracked attachment weld where the 1B RCS loop bypass vent valve assembly joined the 1B RCS loop bypass line. Since the leak could not be isolated and was part of the RCS pressure boundery, the licensee placed Unit 1 in mode 5 (cold shutdown) to repair the leak.
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During refueling outage B1R09, which concluded just 15 days prior to the Unit i reactor trip, the % inch diameter RCS loop bypass vent valve assembly on each loop was
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changed from a Kerotest valve weighing 6 pounds with a pipe cap to an Anchor Darling
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valve weighing 16 pounds with a blank flange. The new assembly installed was a
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stainless steel pipe attached to the valve and then attached to a flange and blank flange.
The purpose of the modification was to facilitate vacuum filling of the RCS loops at the end of outages. Each assembly was socket welded onto the 8 inch bypass line and the required non-destructive testing was satisfactorily completed. The same modification was installed on Unit 2 during refueling outage B2R07. Unit 2 has been operating since May 18,1998, and no indication of a similar failure has been identified.
The inspectors reviewed the licensee's preliminary root cause evaluation and discussed its findings and corrective actions with engineering department personnel. The inspectors were initially concemed that unless the nature of the failure was fully understood and appropriate corrective actions implemented, the potential existed for additional RCS pressure boundary weld failures. The apparent cause for the 1RC8029B weld failure (prior to final root cause determination via materials analysis) was determined to be vibration induced fatigue failure with propagation of the crack initiated from outside of the weld. Prior to installation of the modification, design engineers evaluated the additional weight of the new valve and flange. ' Accelerometer readings were taken and engineers performed calculations to determine the affects of vibration.
No relationship or concem was identified prior to installing the modification that would lead to fatigue failure. ^ All four of the RCS loop bypass vent valve assemblies were compared for orientation and found to be the same. After the 1RC8029B weld failure, the other three Unit 1 RCS loop bypass vent valve assemblies were re-inspected and no indication of weld failures were found. The existing weld was cut out and a section of
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the pipe with the defect was sent offsite for materials analysis. In their evaluation, the licensee also addressed other potential causes for the weld failure induding extemally applied forces, improper welding, impure weld filler material, and inadequate design.
The inspectors considered these other potential causes and concurred with the licensee's elimination of them based upon review of all available information. The inspectors noted that the licensee's approach for evaluating the weld failure was methodical, in that it considered all potential causes and reached condusions based upon all available information.
To prevent a recurrence of the weld failure, engineers designed a support that was
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installed on all four of the RCS loop bypass vent valve assemblies prior to Unit 1 ' entry Into mode 3 (hot standby). The inspectors reviewed the design with engineering personnel and noted that the support consisted of a stiffener bracket attached to the top of the assembly using existing bolts for the blank flange and clamped to the 8 inch RCS
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bypass line. The intent of the support was to add rigidity to the valve assembly to dampen its. vibration. The licensee previously demonstrated the design to effectively resolve structural vibration problems for piping system components.
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c.
Condusions The inspectors conduCd that the licensee appropriately evaluated a weld failure where the 1B reactor coolant system (RCS) loop bypass vent valve (1RC8029B) assembly is attached to the RCS loop bypass line, in addition, the licensee implemented acceptable L
corrective actions to prevent recurrence of a similar vibration induced fatigue failure.
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p E3 Engineering' Procedures and Documentation
' E3.1 ' Post-Construction Testina Not Performed on the Unit 1 Diesel Generators a.
Insoection Scooe (37551. 62707)
l The inspectors observed portions of maintenance performed on the 1B diesel generator L
(DG) and interviewed maintenance, work control, and engineering department i
personnel. The inspectors also reviewed WR 960109400-01, * Replace Fuel Oil l
Filter / Strainer Assemblies for DG 1B," and DCP 9500347," Replace DG Fuel Oil
_
l-Filters / Strainers."
I l
b.
Observations and Findinas On April 7,1999, the licensee removed the IB DG from service to replace the fuel oil
The inspectors reviewed WR 960109400-01 and noted that the DCP requirements for post-construction testing were not incorporated into the work instructions and, as a result, the post-construction testing was not performed upon completion of the work.
Design Change Package 9500347 required the completion of the following post-
'
~ construction tests: (1) a leak test of accessible piping at operating conditions per American National Standards institute B31.1-1977, Section 137.6.2; (2) a seat leakage test on the towly installed three-way transfer valves at normal operating conditions; and (3) a stroke test of the three-way transfer valves. Although some informal checks had c
been made by mechanical maintenance personnel and the system engineer, the post-construction testing was not documented and all of the requirements were not met, in response to the inspectors' questions, the licensee entered this issue into their corrective action program and performed a prompt investigation. The licensee noted that the same work was performed on each of the four DGs and reviewed all four WRs.
Post-construction testing requirements were originally incorporated into each of the four WRs by the use of generic instructions, which were put into all work instructions by the work planner. Those generic instructions were not readily identified as the post-construction testing requirements associated with a DCP The modification work was performed on the two Unit 2 DGs during refueling outage B2R07 in the Spring of 1998 and the post-construction testing, required by the DCP, was completed because mechanical maintenance personnel performed the generic instructions that were incorporated into the post-maintenance section of the work instructions. As part of the licensee's outage improvement initiative for B1R09, the generic instructions were deleted from WRs to speed up paper close-out, thus deleting the DCP required post-construction testing requirements. The inspectors reviewed the licensee's prompt f
investigation into this issue and concurred with its conclusion that the post-construction f
testing requirements were inappropriately deleted from the WRs for both Unit 1 DGs.
Code of Federal Regulations Title 10 Part 50, Appendix B, Criterion XI," Test Control,"
requires, in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in rx srdance with written test
}'
procedures which incorporate the requirements and aueptance limits contained in applicable design documents. The test program shallinclude, as appropriate, proof
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tests prior to installation, preoperational tests, and operational tests during nuclear power plant operation of structures, systems, and components. Test results shall be documented and evaluated to assure that test requirements have been satisfied. The failure to incorporate post-construction testing requirements from engineering design change packages into work request instructions and to accomplish the post-construction testing requirements after replacing the fuel oil filter and strainer assemblies on both Unit 1 diesel generators is an example of a violation of 10 CFR Part 50, Appendix B, Criterion XI. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99008-03a(DRP)).
This violation is in the licensee's corrective action program as PIF B1Pa9-02086.
The inspectors concurred with the licensee's operability assessment for the Unit 1 DGs which concluded that the DGs remained operable. Although the post-construction testing requirements were not performed, the fuel system was functionally checked for excessive leakage nd the ability to shift filters was demonstrated during post-maintenance runs of both DGs.
c.
Conclusions The inspectors concluded that the licensee failed to incorporate post-construction testing requirements from engineering design change packages into work request instructions and failed to complete the post-construction testing requirements after replacing the fuel oil filter and strainer assemblies on both Unit i diesel generators.
One example of a Non-Cited Violation was issued.
E3.2 Review of Selected Plant Modification Work a.
Insoection Scoos (37551)
The inspectors interviewed operations, engineering, and maintenance department personnel and reviewed the following safety-related plant modification WRs and associated DCPs:
WR970036163-01 Install New Relief Valve Assembly on 1B MSIV [ Main
Steam isolation Valve]
WR970114904-01 Install Time Delay on Train A Component Cooling Water
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Pump Low Pressure Auto Start WR970114907-01 Install Time Delay on Train B Component Cooling Water
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Pump Low Pressure Auto Start WR980001099-01 Install Vent Valves on the 1B EDG Turbocharger Lube Oil
.
Filter WR980004940-01 Modify Phase "A" Containrr.,it isolation Test Circuitry to l
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L Prevent Letdown Isolation During Quarterly Slave Testing
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WR980001685-01 Incorporate Time Delay Closure for 1CC685 (Flow switch
.
WR980001687-01 incorporate Time Delay Closure for 1CC685 (Flow switch
.
WR980045231-01 1 A Containment Recirculation Sump Outlet isolation Valve
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Add Stem Protector Per DCP 970174
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WR980045233-01 1B Containment Recirculation Sump Outlet isolation Valve
Add Stem Protector Per DCP 970175 WR980053090-01 Install Termination Blocks in Junction Box for MSIV
I Solenoids
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WR980077029-01 Redundant Trip Coil Circuit for Reactor Coolant
=
Pump 6.9 Kilovolt Feeder Breakers b.
Observations and Findinas Containment Recirculation Sumo Outlet l solation Valve Modificatioriin Add Stem Protectors The inspectors reviewed WRs 980045231-01 and 980045233-01 and noted that the DCP requirements for post-construction testing and post-modification testing were not incorporated into the work instructions and, as a result, the testing was not performed upon completion of the work. Design Change Packages 970174 and 970175 required the completion of the following post-construction and post-modification tests: (1) a visual verification that the stem protector when fully engaged does not extend past the bottom edge of the actuator cover; (2) a verification of valve current signature to ensure that the stem protector is not interfering with stem travel; and (3) a visual verification, with the stem protector cap removed and in conjunction with the above current signature i
test, that with the valve in the full open position adequate clearance exists between the l
cap and top of the stem. The post-construction testing and post-modification testing l-were not documented, none of the visual verifications were completed, and a valve
current signature was performed for only one of the two valves. The inspectors l
concurred with the licensee's operability assessment for the Unit 1 containment recirculation sump outlet isolation valves which concluded that the valves remained operable. Although post-construction and post-modification testing requirements were not performed, valve stroke time and position indication testing were satisfactorily
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l completed. In addition, the valves operated acceptably during emergency core cooling i
system auto actuation testing which was performed at the end of the Unit i refueling outage and following the modifications.
!
Code of Federal Regulations Title 10 Part 50, Appendix B, Criterion XI," Test Control,"
'
requires, in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and perfomied in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The test program shallinclude, as appropriate, proof tests prior to installation, preoperational tests, and operational tests during nuclear power plant operation of structures, systems, and components. Test results shall be documented and evaluated to assure that test requirements have been satisfied. The failure to incorporate testing requirements from engineering design change packages into work request instructions and to accomplish the post-construction and post-modification testing requirements after adding stem protectors on both Unit 1 j
containment recirculation sump outlet isolation valves is an additional example of a i
violation of 10 CFR Part 50 Appendix B, Criterion XI. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC
'
Enforcement Policy (50-454/99008-03b(DRP)). This violation is in the licensee's corrective action program as PlF B1999-02305.
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installation of New Relief Valve Assemblies for Unit 1 MSIV Hvdraulic Accumulators The inspectors reviewed WR 970036163-01 and r;oted that the DCP requirements for post-modification testing were not incorporated into the work instructions and, as a result, the test results were not documented. Design Change Package 9700177
. required the completion of the following post-modification tests: (1) after a leak test has been performed, the hydraulic pressure shall be raised until the relief valve lifts (5300 to 5400 pounds per square inch (psig) maximum); (2) once lifted, the hydraulic pump shall
. be tumed off; (3) the relief valve shall reset prior to 5100 psig; and (4) this test shall be conducted on both hydraulic circuits of the actuator. The inspectors noted that provisions were not provided in the work instructions to assure that all prerequisites for the test were met, that adequate test instrumentation was available and used, and that the test was performed under suitable environmental conditions. Although the testing had been performed informally and the system engineer entered his electronic signature for completion of post-maintenance testing into the electronic work control system data base, the post-modification testing was not documented and the requirements could not be verified to have been performed acceptably. The inspectors concurred with the licensee's evaluation that the valves remained operable. Although post-modification
testing requirements were not docurnented, valve stroke time testing was satisfactorily
)
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completed and system engineering personnel witnessed successful testing of the valves' relief blocks.
Code of Federal Regulations Title 10 Part 50, Appendix B, Criterion XI," Test Control,"
'
requires, in part, that a test program shall be established to assure that all testing
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required to demonstrate that structures, systems, and components will pestform satisfactorily in service is identified and performed in accordance with written test i
procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Test procedures shall include provisions for assuring that all prerequisites, for the given test have been met, that adequate test instrumentation is
.
available and used, and that the test is performed under suitable conditions. Test l
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results shall be documented and evaluated to assure that test requirements have been satisfied. The failure to incorporate post-modification testing requirements from engineering design change packages into work request instructions and to document completion of the post-modification testing requirements after installing new relief valve assemblies on the Unit 1 MSIV hydraulic accumulators is an additional example of a violation of 10 CFR Part 50, Appendix B, Criterion XI. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (50-454/99008-03c(DRP)). This violation is in the licensee's corrective action program as PIF B1999-02329.
c.
Conclusions The inspectors concluded that the licensee failed to incorporate testing requirements from engineering design change packages into work request instructions and to accomplish the post-construction and post-modification testing requirements following modifications to the Unit 1 containment recirculation sump outlet isolation valves and the Unit 1 main steam isolation valves. Two additional examples of a Non-Cited Violation were issued.
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E8 -
Miscellaneous Engineering issues (92700)
{
i E8.1 Review of Byron Station Year 2000 (Y2K) Proaram -
The inspectors conducted en abbreviated review of Y2K activities and documentation using Temporary Instruction (TI) 2515/141, " Review of Year 2000 (Y2K) Readiness of
Computer Systems at Nuclear Power Plants." The review addressed aspects of Y2K l
management planning, documentation, implementation planning, initial assessment,
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detailed assessment, mmediation activities, Y2K testing and validation, notification activities, and contingency planning. The inspectors used NEl/NUSMG 97-07," Nuclear Utility Year 2000 Readiness" and NEl/NUSMG 98-07, " Nuclear Utility Year 2000 l
_ Readiness Contingency Planning" as the basis for this review. The results of this review L
will be combined with the results of other reviews in a summary report to be issued by July 31,1999.
E8.2 (Closed) LER 50-454/98011: " Integral Attachment Welds Not inspected in Accordance With ASME [Amorican Society of Mechanical Engineers] Code Due To Deficient ISI
[ inservice Inspection) Program Plan." The license identified in March 1998 that an l
insufficient number of ASME Code Class 2 piping integral attachment welds had been i
selected for inspaction during the first inspection interval. Contrary to the code l
specification that each integral attachment weld be examined during each inspection j
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interval, the licensee had identified and examined only one of each type of integral attachment weld on each loop. The inadequate component selection was limited to the feedwater and rnain steam systems. Other Class 2 systems were govemed by ASME i
Section XI,1974, with Summer 1974 Addenda, which used a sample type selection
. criter!a. The licensee submitted and received a relief request for the second ISI Interval requiring the inspection of only a sample of the population of integral piping weld l
attachments for both Units 1 and 2. Unit 2 was still in the first ISI interval at the time this
!
problem was identified and the ISI inspection schedule for refueling outage B2R07 was l
augmented to include a 100 percent inspection of the Unit 2 integral piping weld l
attachments.
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Technical Specification 4.0.5, which was applicable at the time this LER was written, stated in part, that inservice inspection of ASME Code Class 1,2, and 3 components
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and inservice testing of ASME Code Class 1,2, and 3 class pumps and valves shall be performed in accordance with Section XI of the ASME Boller and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except were specific written relief has been granted by the Commission pursuant to 10 CFR Part 50,
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Section 50.55a(g)(6)(i). American Society of Mechanical Engineers Boiler and Pressure
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Vessel Code Section XI,1983, with Summer 1983 Addenda, Paragraph IWC 2500,
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Table IWC 2500-1, item No. C3.20, required 100 percent inspection of integrally welded L
pipe attachments during the inspection interval. Contrary to the above, only one of each type of integral attachment weld on each loop for Unit 1 was inspected during the first inspection interval. This failure constitutes a violation of minor significance and is not subject to formal enforcement action. The inspectors reviewed the licensee's corrective actions for this event and found them to be acceptable. This LER is closed.
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IV. Plant Support R1 Radiological Protection and Chemistry Controls (71750)
During routine resident inspection activities, observations were conducted in the area of radiation protection and chemistry using inspection Procedure 71750. No discrepancies were noted.
S1 Conduct of Security and Safeguards Activities (71750)
During routine resident inspection activities, observations were conducted in the area of security and safeguards using inspection Procedure 71750. No discrepancies were noted.
F1 Control of Fire Protection Activities (71750)
During routine resident inspection activities, observations were conducted in the area of fire protection using Inspection Procedure 71750. No discrepancies were noted.
V. Manacement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on June 21,1999. 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 identified.
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PARTIAL LIST OF PERSONS CONTACTED Licensee
' B. Adams, Regulatory Assurance Manager M. Jurmain, Maintenance Manager B. Kouba, Engineering Manager J. Kramer, Work Control Manager S. Kuczynski, Nuclear Oversight Manager W. Levis, Site Vice President
~ R. Lopriore, Station Manager W. McNeill, Radiation Protection Manager M. Snow, Operations Manager INSPECTION PROCEDURES USED
' IP 37551:
Onsite Engineering 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 Facilities IP D3702 Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 50-454/99008-01 NCV Human performance error during performance of nuclear instrumentation calibration results in unit i reactor trip 50-454/99008-02 NCV Failure to verify feedwater system vent valves closed every 31 days as required for containment manual isolation valves 50-454/99008-03a NCV Failure to accomplish post-construction testing requirements after replacing fuel oil filter and strainer assemblies on both Unit 1 diesel generators 50-454/99008-03b NCV Failure to accomplish post-construction and post-modification I
testing requirements after adding stem protectors on both Unit 1 containment recirculation sump outlet isolation valves 50-454/99008-03c.
NCV ' Failure to document completion of post-modification testing requirements after installing new relief valve assemblies on the Unit 1 MSIV hydraulic accumulators I
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i ITEMS OPENED, CLOSED, AND DISCUSSED (cont'd)
Closed 50-454/99001 LER Depressing both feedwater isolation reset pushbuttons leads to LCO [ Limiting Conditions for Operation] 3.0.3 entry
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50-454/99008-01 NCV Human performance error during performance of nuclear instrumentation calibration results in unit 1 reactor trip 50-454/98010 LER Failure to perform in-service testing after maintenance due to poor communications 50-454/99002 LER Design package fails to classify feedwater vent valves as containment isolation valves and results in missed Technical Specification surveillance 50-454/99008-02 NCV Failure to verify feedwater system vent valves closed every 31 days as required for containment manual isolation valves 50-454/99002 LER Design package falls to classify feedwater vent valves as containment isolation valves and results in missed Technical Specification surveillance 50-454/99003 LER Automatic reactor trip due to human error during surveillance procedure 50-454/99008-03a NCV Failure to accomplish post-construction testing requirements after replacing fuel oil filter and strainer assemblies on both Unit 1 diesel generators 50-454/99008-03b NCV Failure to accomplish post-construction and post-modification testing requirements after adding stem protectors on both Unit 1 containment recirculation sump outlet isolation valves 50-454/99008-03c NCV Failure to document completion of post-modification testing requirements after installing new relief valve assemblies on the Unit 1 MSIV hydraulic accumulators 50-454/98011 LER Integral attachment welds not inspected in accordance with ASME
[American Society of Mechanical Engineers] code due to deficient I
ISI [ Inservice inspection] program plan i
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LIST OF ACRONYMS USED
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ASME American Society of Mechanical Engineers BGP Byron General Procedure BISR Byron Instrument Surveillance Requirement BOSR Byron Operating Survelilance Requirement BVSR Byron Engineering Surve;llanco Requirement DCP Design Change Package DG Diesel Generator DRP Division of Reactor Projects ESFAS Engineered Safety Feature Actuation Signal ISI Inservice inspection ITS Improved Standard Technical Specifications LER Licensee Event Report MSIV Main Steam Isolation Valve NEl/NUSMG Nuclear Energy Institute / Nuclear Utility Software Management Group NCV Non-cited Violation i
NRC Nuclear Regulatory Commission PIF Problem Identification Form PSIG Pounds Per Square inch RCS Reactor Coolant System SPP Special Plant Procedure SR Surveillance Requirement Tl Temporary Instruction TS Technical Specification UFSAR Updated Final Safety Analysis Report WR Work Request Y2K Year 2000
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