IR 05000454/1987024

Safety Insp Repts 50-454/87-24 & 50-455/87-22 on 870530- 0630.No Violations or Deviations Noted.Major Areas Inspected:Licensee Action on Previous Insp Findings,Lers, Operations Summary,Training,Surveillance & Maint

ML20235S840
Person / Time
Site:	Byron
Issue date:	07/09/1987
From:	Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20235S815	List: IR 05000454/1987024 ML20235S812
References
50-454-87-24, 50-455-87-22, NUDOCS 8707210794
Download: ML20235S840 (9)

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UQ3hNUCLEARREGULATORY. COMMISSION

, , REGION III

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[ .Repcets No. 50-454/87024(DRP)';-50-455/87022(DRP)

Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66

' Licensee: Commonwealth Edison Company

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Post Office Box-767 Chicago, IL 60690

} Facility Name: . Byron Station, Units 1.and 2 Inspection At: Byron Station, Byron, Illinois Inspection Conducted: May 30 - June 30, 1987 Inspectors: J. M. Hinds, J P. G. Brochman Approved By:

kk %WAM Nd J. M. Hinds, Jr., Chief 7/f/f 7 -

Reactor Projects Section'IA Date Inspection Summary-Inspection on May 30 - June 30,," 1987 (Report Nos. 50-454/87024(DRP);

50-455/87022(DRP))

--Areas Inspected: Routine, unannounced safety inspection by the resident inspectors < of= licensee action on previous inspection findings; LERs; operations > summary; training; surveillance; maintenance; operational safety; startup from refueling; and event followu Results: Of the nine areas inspected, no violations or deviations were identified'in eight areas. One violation was identified in the remaining area; however, in accordance with 10 CFR 2. Appendix C,Section V. A, a Notice of Violation was not issued (entry into Mode 4 with an inoperable steam generator - Paragraph 3). This violation was of minor safety. significance and did not affect the public's health and safet PDR G ADOCK 05000454 PDR

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DETAILS Persons Contacted Commonwealth Edison Company

• R.' Querio, Station Manager

• R. Pleniewicz, Production Superintendent

• R. Ward, Services Superintendent

• Burkamper, Quality Assurance Superintendent L. Sues, Assistant Superintendent, Operating

• G. Schwartz, Assistant Superintendent, Maintenance T. Joyce, Assistant Superintendent, Technical Services D. St. Clair, Assistant Superintendent, Work Planning T. Higgins, Operating Engineer, Unit 0 J. Schrock, Operating Engineer, Unit 1 D. Brindle, Operating Engineer, Unit 2 T. Didier, Operating Engineer, Rad-Waste

• M. Snow, Regulatory Assurance Supervisor F. Hornbeak, Technical Staff Supervisor

• R. Flahive, Radiation / Chemistry Supervisor P. O'.Neil, Quality Control Supervisor

• Kouba, Assistant Technical Staff Supervisor

• Walter, Assistant Technical Staff Supervisor

• D. Berg, Onsite Nuclear Safety Group

• Pirnat, Regulatory Assurance Staff

• R. Curtis, Training Staff

• E. Zittle, Regulatory Assurance Staff The inspecton; also contacted and interviewed other licensee and contractor personnel during the course of this inspectio * Denotes those present during tne er't interview on June 30, 198 . Action on Previous Inspection Findings (92701)

(Closed)OpenItem(454/86010-01(DRP)): Submission of a supplemental report to LER 454/86004. This LER addressed the failure of an ESF (engineered safety feature) component to start when required. The j licensee's investigation determined that the cause of the problem was a i failed AR-3 relay in the breaker closing circuit. The inspector requested that the licensee identify any other ESF equipment affected by this relay and any generic implications. The licensee determined that if the hand switch for the breaker is not held in the close position until the breaker closes, the contacts on the AR-3 relay will interrupt the i high current which energizes the spring release coil. The spring release coil,' when energized, releases the closing spring to close the breaker. The contacts on the AR-3 relay are designed to pass the high current, but not to interrupt it. When the contacts interrupt the high current, arcing occurs. This arcing causes pitting of the contacts and can result in the welding together of the contacts. If the contacts are

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welded shut, the spring release coil will not furiction properly and the breaker will not close. The licensee has modified procedures for operating electrical equipment and established plans for periodically inspecting the 26 AR-3 relays which control ESF equipment. Based on the corrective actions taken, the inspector has no further concerns regarding this matter, and this item is considered close . Licensee Event Report (LER) Followup'(92700)

(Cloted) LERs (454/86035-01-LL; 454/87013-LL; 454/87014-LL; 454/87015-LL; 454/87016-LL;455/87004-LL;455/87007-LL): Through direct observation, discussions with licensee personnel, and review of records, the following LERs were reviewed to determine that the deportability requirements were <

fulfilled, that immediate corrective action was accomplished, and that corrective action to prevent recurrence had been accomplished in accordance with Technical Specification LER N Title Unit 1 464/36035-01 Emergency diesel generators inoperable due to seismically unqualified components in the trip circuit due to a design erro /87013 Reactor trip signal during control rod position surveillance from logic coincidence on low steam generator level due to signal spik /87014 Mode change with an inoperable steam generator due to a misplugged tube caused by personnel erro /87.015 Inadvertent start of the 1A auxiliary feedwater pump due to personnel erro /87016 Both trains of the spray additive portion of the containment spray system were inoperable due to procedural erro Unit 2 455/87004-01 Technical specification action requirement not met for containhient isolation valves due to personnel erro /87007 Reactor trip due to ine loss of an instrument inverte With regard to LER 454/87014, this LER discusses an event from May 5 -15, 1987 when unit 1 entered Mode 4 with the ID steam generator inoperabl During the cycle 1 refueling outage, eddy current inspections were performed on the u-tubes of the steam generators. Technical Specification 4.4.5.4.a(6) requires that u-tubes with imperfections of greater than 40% of nominal wall thickness be plugged. Technical Specification 4.4.5.4.b requires that for a steam generator to be operable, all tubes which exceed the plugging limit are required to be plugged. Technical Specification 3.4.5 requires that each steam i

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l' e generator shall'be operable' prior to increasing Tave above 200 degrees F-(entry into. Mode 4).

In steam generator ID, one imperfection was discovered in the u-tube at row 29/ column 63 and was estimated to be approximately 60% of the nominal

- wall thickness. ' The eddy current inspections were performed by a contractor (Babcock & Wilcox [B&W]), who then plugged the required tubes-from both the hot leg and cold. leg sides of the steam generatori All the required tubes had been plugged by B&W, and cll the steam generators were declared operable before the unit entered Mode 4 on May 5,'1987. On May'15, 1987, an engineer in .the Weasee's technical staff was reviewing the viden-

' tapes of-the plugging operations provided by B&W and determined that the plug had been. instalied at row 28/ column 64 in the hot leg side of the steam generator,;instead of row 29/ column 63. The cold leg plug was determined to be in its correct location. This review did not result from any programmatic requirements. The unit was cooled down and a new L plug was . installed ~ in the correct locatio The-licensee believes that'the cause of the error was a cognitive per:;onnel error by two B&W personnel. Both a task leader and a quality inspector verified the location of the plug per B&W procedures, although

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this ' location was in fact ' incorrect. As corrective action the licensee l (1)' inspected all other tubes which had been plugged during this outage

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and verified that the plugs were correctly installed, and (2) revised its in-process and closeout inspection procedures to require the licensee's technical staff to review all plugging locations. B&W has also revised its procedures, and they have been reviewed by.the license Had the error gone undetected, the probability of a steam generator tube leak would have increased. The licensee has emergency procedures for combating a' steam generator tube leak. During the outage the licensee performed eddy current testing on 18,290 u-tubes and plugged 1 The failure to ensure that steam generator ID was operable prior to entry into Mode 4 is a violation of Technical Specification 3. (454/87024-01(DRP)). However, this violation meets the tests of 10 CFR 2. Appendix C,Section V.A; consequently, no Notice of Violation will be issued, and this matter is considered close The events described in LER 454/87016 are discussed in Inspection Report 454/87022(DRP). The events discribed in LERs 454/87013 and 455/87007 are discussed in Inspection Report 454/87020; 455/07019. Questions related to the events described in LER 454/86035-01 are being followed as Unresolved Item 454/87002-04(DRP);455/87002-04(DRP). No other violations or deviations were identifie . Summary'of Operations Unit I was critical at 1% power during core physics testing from the beginning of the report period until 0326 on June 2, 1987. At that time it was synchronized to the grid and then operated at power levels up to 15% until 1239 on June 3, 1987, when the unit was shut down to perform turbine balancing. The unit was synchronized to the grid at 2103 on

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Lthe same day and then operated at power levels up to 100% for the rest of the report perio Unit 2 operated at power levels up to 90% until 2045 on June 5,1987,

.when the unit was shut down for a scheduled 12 day outage. Following the

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outage the unit was taken critical at 1801 on June 19, 1987, and was synchronized to.the grid at 2200 on the same day. The unit operated at to 29,.1987, when a. manual reactor power levels up(see.97%

trip occurred until10).

paragraph 0922The onunit June was taken critical at 1258 on June.30, 1987,. and was syncronir2d to the grid at 1621.on the same day. The unit operated at power levels up to 30%.for the rest of the report perio . Training (41400 & 41701)

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The effectiveness of training programs for licensed and nonlicensed

' personnel was. reviewed by the inspectors during the witnessing of the licensee's performance of routine surveillance, maintenance, and operational activities and'during the review.of the licensee's response-to events which occurred.during June 1987. Personnel appeared to be knowledgeable of the tasks ~being performed, and nothing was observed l which indicated any ineffectiveness of trainin No violations or deviations were identifie . MonthlySury.ei11anceObservation(61726)

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Station surveillance activities of the safety-related systems and components listed below were observed / reviewed to ascertain that the 'were conducted in accordance with approved procedures and in conformance with Technical Specif.ication Monthly test of the IB Auxiliary Feedwater pump Monthly test of the 1A Diesel Generator The following items were considered during this review: the limiting cond_itions 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 No vio1'ations or deviations were identifie ,

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' 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 Alignment of the indicating meter on the Unit 2 Nuclear Instrument-ation channel-2N42 Following completion of maintenance on channel 2N42, the inspectors verified that it had been returned to service properl The following items were considered during this review: the limiting

. conditions for. operation were met while components or systems were removed:from arrd restored to service; approvals were' obtained prior to-initiating the . work;- activities were accomplished using approved

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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 rdquests 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 No violations or' deviations were identifie . Operational Safety Verification (71707, 71709 & 71881)

The inspectors observed control room operation, reviewed applicable logs and coaducted discussions with control rooi operators 'during the month of June 1987. 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. fuel handling, turbine, and rad-waste buildings were corducted to observe plant i 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 ]

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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 controls. The inspectors )

also witnessed portions of the radioactive waste system controls 1 6 ,

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associated with rad-waste shipments and barreling. During the month of

' June 1987, theLinspectors walked down the accessible portions of the DC and AC power systems -to verify operabilit Facility' operations observed were verified to be.in accordance with the requirements established under Technical Specifications,10 CFR, and administrative procedure No violations or deviations were identifie , Startup from Refueling (71711, 61702 & 61706)

The startup of Unit 1 from its first refueling occurred on' May 28,'1987, '

and is discussed in Inspection Report 454/87020. The inspectors witnessed portions of the testing performed during the startup, which included incore flux mapping and core thermal power measurement The activities were witnessed in order to verify that testing was ccaducted in accordance with the operating license and procedural requirements,'that test data was properly recorded and reviewed, and that the performance of personnel conducting the tests demonstrated an understanding of assigned duties and responsibilitie No violations or deviations were identifie . - Onsite Followup of Events at Operating Reactors (93702) General i

The inspectors performed onsite followup activities for an event which occurred during June 1987. This followup included reviews of operating logs, procedures. Deviation Reports, Licensee Event Reports (where available), and interviews with licensee personne For the event, the inspector developed a chronology, reviewed the functioning of safety systems required by plant conditions,'and reviewed licensee actions to verify consistency with procedures, license conditions, and the nature of the event. Additionally the inspector verified that the licensee investi% tion had identified root causes of equipment malfunctions and/or personnel error and had taken appropriate corrective actions prior to plant restar Details of the event and licensee corrective actions developed through inspector followup are provided belo Unit 2 - Manual Reactor Trip At_0921 on June 29, 1987, with reactor power at 97%, the 2C main feedwater pump tripped. Licensed operators commenced a manual runback of the turbine to reduce feedwater flow. At 0922, with level in the 2C steam generator at 20% and decreasing, the Shift Engineer (licensed SRO) directed the reactor operator to manually trip the reacto Reactor power was at approximately 78%; the automatic reactor trip on 10-10 steam generator level is at 17%. All safety systems functioned normally folicwing the trir, and the unit was

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stabilized in Mode 3. .Before the trip the turbine control system (Digital Electro-Hydraulic Control [DEHC]) did not respond as the operators had expecte The licensee's investigation determined that the DEhC system did operate as designed during the turbine runback. The operators had preprogrammed the DEHC for a 550 MWe load reduction from 1100 MWe at 2200 MWe/ min. However, the operator had gone from MW-0VT to MW-IN on the DEHC control panel immediately before pressing the G0 butto This caused the DEHC computer to recalculate the reference load and produced the delay in response of the DEHC. Thus. the turbine load was only reduced to approximately 950 MWe. The DEHC computer was recalculating the reference value and the operator mistook this for a failure of the DEHC system, so the operator took manual control and closed the turbine governor valves to an approximately 700 MWe load. The operator expected the turbine load to-decrease to 500 MWe, with an approximate overshoot of 200 MWe on the turbine controls, from the training he had experienced on the Byron plant-specific simulator. However, the actual turbine controls did not overshoot, and consequently the turbine load was not reduced far enough to allow steam generator levels to be stabilized before a reactor trip on lo-lo steam generator level was considered unavoidabl The licensee's investigation determined tht the cause of the feedwater pump trip was the failure of a solders oint for a capacitor in the reference speed circuit. When the capacitor failed, the reference speed for the pump went to infinity, causing the pump speed to increase rapidly and actuate the overspeed trip. The failed solder joint was repaired, and the unit was taken critical at 1258 on June 30, 1987, and synchronized to the grid at 1621 on the same da The inspectors wili review this event in a subsequent report after the LER is issue No violations or deviations were identifie . Violations for which a " Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requiremen However, because the NRC wants to encourage and support licensee initiative for self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C Section V.A. These tests are: 1) the violation was identified by the licensee; 2) the violation would be categorized as Severity Level IV or V; 3) the violation was reported to 4 the NRC, if required; 4) the violation will be corrected, including measures to prevent recurrence, within a reasonable time period; and 5) '

it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violatio A violation of regulatory requirements (mode change with an inoperable steam generator) identified during the inspection for which a Notice of Violation will not be issued is discussed in Paragraph j

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l 12. Exjt interview (30703)

The inspectors met with the licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on June 30, 1987. The inspectors summarized the purpose and scope of the inspection and the findings. The inspectors also discussed the likely informational content of the l inspection report with regard to documents and processes reviewed by the inspectors during the inspection. The licensee did not identify any such documents or processes as proprietar l

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