IR 05000454/1986048

Safety Insp Repts 50-454/86-48 & 50-455/86-46 on 861202-31. Violation Observed:Failure to Maintain Drawings in Same Configuration as as-built Plant

ML20212G941
Person / Time
Site:	Byron
Issue date:	01/14/1987
From:	Forney W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20212G897	List: IR 05000454/1986048 ML20212G893 ML20212G909
References
50-454-86-48, 50-455-86-46, IEB-83-05, IEB-83-5, NUDOCS 8701210241
Download: ML20212G941 (9)

Text

U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No. 50-454/86048(DRP);50-455/86046(DRP)

Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-60 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station, Units 1 and 2 Inspection At: Byron Station, Byron, IL Inspection Conducted: December 2 - 31, 1986 Inspectors: J. M. Hinds, J P. G. Brochman J. A. Malloy W. C. Liu Approved By:

(A fd b i L h " N W. L. Forney, Chief i/N/h Reactor Projects Section 1A Date Inspection Summary Inspection on December 2 - 31, 1986 (Reports No. 50-454/86048(DRP);

50-455/86046(DRP))

Areas Inspected: Routina, unannounced safety inspection by the resident inspectors and a region based inspector of licensee action on previous inspection findings; 50,55(e) reports; IEBs; LERs; operations summary; license conditions; surveillance; maintenance; operational safety and ESF walkdown; cold weather preparation startup testing; event followup; allegations; and management meetings.

Results: Of the ten areas inspected, no violations or deviations were identified in 9 areas; one violation was identified in the remaining area:

(failure to maintain drawings in the same configuration as the as-built plant - Paragraph 8). This violation was of minor safety significanc PDA ADOCK 05000454 G PDR

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DETAILS Persons Contacted Commonwealth Edison Company

1. T. Maiman, Vice President, Projects

1. • K. Graesser, Division Vice President, Nuclear Stations Division

1. • R. Querio, Station Manager

1. V. Schlosser, Project Manager, Byren

1. R. Tutken, Assistent Project Manager, Byron

1. • R. Pleniewicz, Production Superintendent

1. • R. Ward, Services Superintendent

1. • W. Burkamper, Quality Assurance Superintendent

1. L. Sues, Assistant Superintendent, Operating

1. G. Schwartz, Assistant Superintendent, Maintenance

1. • T. Joyce, Assistant Superintendent, Technical Services

1. • D. St. Clair, Assistant Superintendent, Work Planning W. Blythe, Operating Engineer, Unit 0 J. Schrock, Operating Engineer, Unit 1

1. D. Brindle, Operating Engineer, Unit 2 A. Chernick, Operating Engineer, Rad-Waste

1. M. Snow, Regulatory Assurance Supervisor

• E. Falb, Unit 2 Startup Testing Supervisor

1. F. Hornbeak, Technical Staff Supervisor R. Flahive, Radiation / Chemistry Supervisor P. O'Neil, Quality Control Supervisor

1. P. Johnson, Master Instrument Mechanic

1. • K. Yates, Nuclear Safety Group

1. • W. Pirnat, Regulatory Assurance Staff

1. • E. Zittle, Regulatory Assurance Staff

1. D. Flowers, ISI Coordinator

1. M. Whitemore, Radiation Chemistry Staff i

The inspector also contacted and interviewed other licensee and

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contractor personnel during the course of this inspectio # Denotes those present during the management meeting on December 16, 198 * Denotes those present during the exit interview on December 31, 198 . Action on Previous Inspection Findings (92701)

(Closed) Unresolved Item (454/86040-01(DRP)): Supplemental response for !

IE Bulletin (IEB) 83005. During a review cf Unit 2 preoperational test l 2.3.60 the inspector identified that all of the testing, required by IEB 83005, had not been performed on Essential Service Water Booster Pump ISX04P. The licensee agreed with this finding and performed additional testing and submitted a supplemental response, letter from S. C. Hunsader to J. G. Keppler, dated December 12, 1986. The inspector reviewed the response and verified that the testing required by the IEB had been i

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performed, where possible, and the data was acceptable. For these tests that could not physically be performed or performed safely the licensee provided justifications for not performing the testing. Based on this review, the inspector has no further concerns and this item is considered close . 10 CFR 50.55(e) Report Followup (92700)

(Closed) 50.55(e) Report (454/84007-EE): Crush strength of Energy Absorbing Material (EAM) used in pipe whip restraints. On October 30, 1984, the licensee notified Region III of a deficiency report regarding the crush strength of EAM used in the whip restraints. The final report was submitted on March 11, 1985. This report identified 21 whip restraints to be replaced with new EAM for each unit of Byron Statio The revised final report was submitted on October 17, 1986. This report deleted four whip restraints to provide construction, operation, and maintenance cost savings. As a result, the final number of whip restraints became 17 for each unit of the facility. The NRC inspector held discussions with licensee representatives and reviewed related documentation to ensure that the corrective actions identified in the report were adequately completed. This item is considered close . IE Bulletin (IEB) Followup (92703)

(Closed) IEB (455/83005-BB): Problems with Hayward Tyler Nuclear Code Pumps. The licensee's initial response was provided in a letter from P. L. Barnes to J. G. Keppler, dated August 12, 1983. A supplemental response was provided in a letter from S. C. Hunsader to J. G. Keppler, dated December 12, 1986. The inspector reviewed the responses and verified that the testing required by the IEB had been performed or a justification for nonperformance was provided. Based on this review this item is considered close No violations or deviations were identifie . Licensee Event Report (LER) Followup (90712 & 92700)

(Closed) LER (455/86002-LL): An in-office review was conducted for the following LERs to determine that the reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification LER N Title 455/86002 Manual Reactor Trip with Control Rod Position Indicators Mismatch No violations or deviations were identifie _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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6. Review of Conditions for Unit 2 Operating License NPF-60 (Closed) License Condition C.1 of Attachment 1 to NPF-60. This item is discussed in Paragraph 4 and based on the licensee's response to IEB 83005 this license condition is considered satisfie . Summary of Operations Unit 1 operated at power levels up to 94% until 1048 on December 23, 1986, when an end-of-cycle power coastdown was begun. The unit operated at power levels above 75% for the rest of the mont Unit 2 initially entered Mode 4 on December 11, 1986, and entered Mode 3 on December 16. At 0500 on December 18, 1986, a crack was discovered in a fill line to the 2D Safety Injection Accumulator. The accumulator was declared inoperable and the unit was placed in Mode 4 by 1056 on the same day. Following repairs to the pipe the unit reentered Mode 3 on December 19, 1986. The unit remained in Mode 3, in the startup test program, for the rest of the month. This event is discussed further in Paragraph 1 . MonthlySurveillanceObservation(61726)

Station surveillance activities of the safety related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures and in conformance with Technical Specification The following items were considered during this review: the limiting conditions for operation were met while affected components or systems were removed from and restored to service; approvals were obtained prior to initiating the testing; testing was accomplished in accordance with approved procedures; test instrumentation was within its calibration interval; testing was accomplished by qualified personnel; test results conformed with Technical Specifications and procedural requirements and were reviewed by personnel other than the individual directing the test; and any deficiencies identified during the testing were properly documented, reviewed, and resolved by appropriate management personne The following surveillance testing activities were observed / reviewed:

2B Auxiliary Feedwater pump Auto-Actuations test 2A Auxiliary Feedwater pump ASME test No violations or deviations were identifie . Monthly Maintenance Observation (62703)

Station maintenance activities of the safety related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with Technical Specification _ _ ____ ___ _ _ _ _ _ - _ _ _ . - , ._ _ ,

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The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from and restored to service; approvals were obtained prior to initiating the work; activities were accomplished using approved "

procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and fire prevention controls were implemented. Work requests were reviewed to determine the status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc The following maintenance activities were observed / reviewed:

2RY8000A Pressurizer PORV Block Valve Repack Following completion of maintenance on the pressurizer, the inspectors verified that these systems had been returned to service properl No violations or deviations were identifie . Operational Safety Verification and Engineered Safety Features System Walkdown (71707 & 71710)

The inspectors observed control room operation, reviewed applicable logs and conducted discussions with control room operators during the month of December 1986.- During these discussions and observations, the inspectors ascertained that the operators were alert, cognizant of plant conditions, attentive to changes in those conditions, and took prompt action when appropriate. The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the auxiliary, turbine, rad-waste, and Unit 2 containment buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, excessive vibrations, and to verify that maintenance requests had been initiated for equipment in need of maintenanc The inspectors verified by observation and direct interviews that the physical security plan was being implemented in accordance with the station security pla The inspectors observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection control The inspectors also witnessed portions of the radioactive waste system controls associated with rad-waste shipments and barreling.

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During the month of December 1986, the inspectors walked down the accessible portions of the 2B Residual Heat Removal (RHR) and B Combustible Gas Control systems to verify operability. During the walk down of the RHR system the inspector identified several discrepancie Four valves, 2RH018C, 2RH018D, 2RH019C, and 2RH019D, were not indicated on the system drawing, Control and Instrumentation Drawing C&ID M-2137, Sheet 1,.nor on the system valve lineup, Byron Operating Procedure B0P RH-M2, "RHR Valve Lineup - Unit 2." The valves did not have station identification tags installed. These valves are used as root isolation valves for flow transmitters (FT) 2FT-RH001, Train A, and 2FT-RH002, Train B. Also, the installed piping configuration for the FT for Train A was different than the one installed for Train B and nc indication of this difference was provided on the C&ID. These flow transmitters are used to provide an input to the process computer to measure recirculation

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flow for each RHR pum CFR 50, Appendix B, Criterion V, as implemented by the Commonwealth Edison Company's Quality Assurance Manual, Quality Requirement b.0, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings. 10 CFR 50, Appendix B,

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Criterion VI, as implemented by the Commonwealth Edison Company's Quality Assurance Manual, Quality Requirement 6.0, requires that measures shall be established to control the issuance of documents, such as inst e tions, procedures, or drawings, including changes thereto, which prescribe all activities affecting quality. These measures shall assure that documents, including changes, are reviewed for adequacy. The failure to assure that: 1) drawing M-2137, Sheet I reflected that valves 2RH018C, 2RH0180, 2RH019C, and 2RH019D were installed; 2) drawing M-2137, Sheet 1, reflected the difference between the piping arrangement in Trains 2A and 2B; and 3) valves 2RH018C, 2RH018D, 2RH019C, and 2RH019D were not listed on valve lineup B0P RH-M2 is a violation of 10 CFR 50, Appendix B, Criterions V and VI (455/86046-01(DRP)).

Facility operations observed were verified to be in accordance with the requirements established under Technical Specifications,10 CFR, and administrative procedure No other violations or deviations were identifie . Cold Weather Preparations (71714)

The inspectors verified that systems susceptible to freezing had been inspected by the licensee. Thc inspectors reviewed surveillance 080S-XFT-A1, completed on November 15, 1986, to verify that equipment susceptible to freezing had been inspected to ensure that heat tracing and space heater circuits had been energized. The inspectors also verified that the inspected systems which have been subjected to maintenance and modification during the past year had the cold weather protective measures reestablished. During the review of 0BOS XFT-A1, the inspectors noticed that heat tracing for the Unit 2 Condensate Storage Tank (CST) level transmitter was not verified to be energize The Unit 1 CST heat tracing was verified to be energized in OBOS XFT-A .

The CST is required by Technical Specification 3.7.1.3 to be in operation

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in Modes 1, 2, and 3. Unit 2 initially entered Mode 3 on December 16, 1986. Prior to entering Mode 3, the licensee verified that CST heat tracing was energized by performing the electrical system lineup which included verification that the CST heat tracing motor control center was energized. The licensee stated that the freezing temperature equipment protection surveillance, OBOS XFT-A1, will be revised to include ,

verification that the heat tracing for the Unit 2 Condensate Storage Tank level transmitter is energized. Pending review of the surveillance revision, this is considered an open item (455/86046-02(DRP)).

No violations or deviations were identified 12. Startup Test Witnessing and Observation (72302)

The inspectors witnessed performance of portions of the following startup test procedures in order to verify that testing was conducted in accordance with the operating license and procedural requirements, test data was properly recorded, and performance of licensee personnel conducting the tests demonstrated an understanding of assigned duties and responsibilitie .68.80 Reactor Protection Logic 2.69.80 Pressurizer Sprays, Heaters and Bypass Flow Adjustments 2.47.82 Thermal Power Measurement and Statepoint Data Collection No violations or deviations were identifie . Onsite Followup of Events at Operating Reactors (93702) General The inspector performed onsite followup activities for events which occurred during December 1986. This followup included reviews of operating logs, procedures, Deviation Reports, Licensee Event Reports (where available) and interviews with licensee personne For each event, the inspector developed a chronology, reviewed the functioning of safety systems required by plant conditions, reviewed licensee actions to verify consistency with procedures, license conditions and the nature of the event. Additionally the inspector verified that licensee investigation had identified root causes of equipment malfunctions and/or personnel error and had taken appropriate corrective actions prior to plant restart. Details of the events and licensee corrective actions developed through inspector followup are provided in Paragraphs b and c belo . Unqualified Equipment In the Control Circuit for the Emergency Diesel Generators for Units 1 and 2 At 1530 on December 17, 1986, with Unit 1 at 94% power and Unit 2 in Mode 3, NRC headquarters notified the resident inspectors of a 10 CFR Part 21 report by the emergency diesel generator (DG) vendor,

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l Cooper-Bessemer, on their KSV series DGs. Byron's emergency diesel generators are Cooper-Bessemer's KSV-20 serie g 1- On December 12, 1986, Sargert and Lundy (S &LL), the Byron-

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Architect-Engineer, was reviewing the DG safety related component

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list and identified two non-seismically qualified components-which

were wired into the engine overspeed circuit. The engine overspeed

circuit is safety related. The two components were a limit switch, Microswitch model BZLN-LH3, and pressure switch, Square D model 9012-BC0-22. One of the four Byron diesel generators has square D Model GAW-24 pressure switch instead of the previously mentioned

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i- switch, but it is also not seismically qualified. These components-i' provide input into two logic trains, for the DG control circuit; and are redundu t electrical backups for mechanical devices. ^A failure of either component would cause the de-energization of the emergency run relays. This in turn would de-energize both fuel oil

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solenoids and shut off the DG's fuel oil' supply. S & L also identified that a component failure would prevent the DG from i

starting on a valid actuation signal, with the DG in the standby mode. And.could result in engine shutdown which is normally blocked when the DG is in the emergency run mode, f

Upon completion of a review of the S & L findings, the vendor

notified the NRC on December 17, 1986, and the NRC then notified i the licensee. By 1800 on December 17, the licensee had completed a i

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preliminary review of the information and determined' that there was insufficient information to declare the DGs inoperable and requested .

the assistance of the vendor, S & L, and their corporate engineering department to resolve the question. In parallel with that effort,

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the licensee commenced work to install jumpers in the DG control circuits to defeat these redundant _ electrical trips. NRC Region III j personnel monitored these activities.

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By 1944 on December 17, the licensee had completed testing for the jumpers which were installed on the 1A and 2A DGs. At 2000, based on the recommendation of S & L, the vendor, and the licensee's *

! corporate engineering department, the licensee determined that the

non-seismicly qualified components did render the DGs inoperabl ,

l The licensee then declared 1B and 2B DGs inoperable, as the jumpers had not been installed yet, and entered the applicable Technical

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Specification limiting condition for operation. By 2055 on December 17, the installation and testing of jumpers on the IB and

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2B DGs had been completed and the DGs were declared operabl i'

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The inspector will further review this event in a subsequent report when the LER is issue j'

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j c. Inoperable Unit 2 Safety Infection Accumulator i

At 0333 on December 18, 1986, while in Mode 3, control room

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operators received indication of decreasing pressure and level in i the 2D Safety Injection (SI) accumulator along with indications of l

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q increasing" level in the containment sump. Operators entered containment and observed water spraying out of a 1 inch fill line for the 2D SI accumulator. The accumulator was depressurized and drained. The crack was observed to be approximately 90 and  :

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circumferential. The crack was located at the edge of a fillet 4 weld where the pipe entered a boss on the accumulator. Technical

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Specification 3.5.1 requires that the accumulator be operable in ' '

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Mode 3 when pressurizer pressure is greater than 1000 psig. By 0500 ,

on December 18 pressurizer pressure had been reduced to less than 1000 psig. At 1056 on December 18, the unit was placed in Mode The licensee removed the damaged pipe for metallurgical examinatio By 1845 on December 19, the licensee had completed replacement of the damaged pipe and refilled the accumulator. The unit reentered ,.,

Mode 3 at 192 'l t

The inspectors will review the licensee's report of this event when it is issued. No violations or deviations were identifie . Management Meetings (30702) -

On December 16, 1986, Messrs. R. F. Warnick, Chief, Reactor Projects

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Branch 1, W. L. Forney, Chiefi Reactor Projects Section 1A, and the NRC '

resident inspector staff met with licensee management and supervisorys personnel' denoted in Paragraph 1_of this report. This meeting w h held to discuss the performance of the Conduct of Maintenanc.e Improvement Program and the Unit 2 Startup testing progra g 15. Open Items '

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Open items are matters which have been discussed with the licensee, which

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will be reviewed further by the inspector and which involve some action on the part of the NRC or licensee or both. An open tiem disclosed during the inspection is discussed in Paragraph 1 ~ e'

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1 Exit Interview (30703)

The inspectors met with licensee representati'ves denoted in Paragraph 1 at the conclusion of the inspection on Decembbr 31, 1986. The inspectors sumarized the purpose and scope of the inspection and the finding The insrectors also discussed the likely informational content of the

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inspection report with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any such documents / processes as proprietar t

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