IR 05000373/1986021

Revised Pages to Insp Repts 50-373/86-21 & 50-374/86-20. Pages Correct Tracking Numbering Sequence in Original Rept for Identified Violations & Open Items

ML20209C364
Person / Time
Site:	LaSalle
Issue date:	09/03/1986
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20209B574	List: IR 05000373/1986021 ML20209B570 ML20209B609
References
50-373-86-21, 50-374-86-20, NUDOCS 8609090011
Download: ML20209C364 (5)

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DETAILS Persons Contacted Commonwealth Edison Company (CECO)

+*G. J. Diederich, Manager, LaSalle Station

• R. D. Bishop, Services Superintendent

• C, E. Sargent, Production Superintendent D. Berkman, Assistant Superintendent, Technical Services

+W. Huntington, Assistant Superintendent, Operations J. C. Renwick, Assistant Superintendent, Work Planning R. W. Stobert, Quality Assurance Supervisor P. Manning, Tech Staff Supervisor T. Hammerich, Assistant Tech Staff Supervisor W. Sheldon, Assistant Superintendent, Maintenance

• J. Atchley, Operating Engineer

• D. Winchester, Senior Quality Assurance Inspector

+B. B. Stephenson, Manager, Department of Nuclear Safety

+C. M. Allen, Nuclear License Administrator

+J. J. Shetterly, LSCS Shift Engineer

+L. B. Wilson, Nuclear Fuel Services

+J. G. Marshall, Director of QA, Operations

+S. L. Trubutch, Staff Attorney

+L. O. DelGeorge, Assistant Vice President

+M. S. Turbak, Licensing Director, Operating Plants

+K. L. Graesser, Division Vice President

+L. F. Gerner, Regulatory Assurance Superintendent US NRC

+ James G. Keppler, Regional Administrator

+G. C. Wright, Chief, Reactor Projects Section 2C

+C. E. Norelius, Director, Division of Reactor Projects

+*M. J. Jordan, Senior Resident Inspector, LSCS

+W. G. Guldemond, Branch Chief, Reactor Projects Branch 2

+*J. C. Bjorgen, Resident Inspector, LSCS

+C. W. Hehl, Chief, Operations Branch

+J. S. Wiebe, Project Inspector, LaSalle

+L. Kanter, Resident Inspector, Zion Station

+B. A. Azab, DRS Inspector

+B. A. Berson, Regional Counsel

+W. H. Schultz, Enforcement Coordinator

+ Denotes those attending the Enforcement Conference at RIII on August 13, 198 * Denotes those attending the exit meeting held on June 11, 198 PDR ADOCK 05000373 2 O PDR

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e Operational Safety Verification (71707)

The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the inspection period. The inspector verified the operability of selected emergency systems, reviewed tagout records, and verified proper return to service of affected components. Tours of Units 1 and 2 reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspector by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security pla The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection control The inspectors, while touring the plant, noticed oil leakage from some of the post tension connections for containment on Unit 2. This leakage was brought to the attention of the station management for evaluation based on a similar problem at the Farley Unit 2 plant addressed in IE Information Notice 85-10, Supplement 1. The licensee reported back that no problem existed with the post tension connections because less than one gallon total leakage had occurred and less than one percent of that contained in any one tension was noticed. The next grease coverage check surveillance required by technical specifications is scheduled for late-1986. Since no free standing water was noticed in the tension cap, no further action was needed at this time. The inspector also reviewed the method by which IE 2a

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In addition to the mispositioned rod, the inspectors are concerned about three other aspects of this event: the immediate operator actions; the timeliness of the identification of the error by the independent verifier; and the failure of the unit operator to make a log entry describing the error. The licensee's Procedure , LOA-RD-03 provides the applicable instructions for mispositioned control rods. For a control rod withdrawn beyond its in-sequence position, Section D.2 of this procedure requires the operator to demand process computer printouts 00-3 (Core Thermal Power), OD-7 Option 2 (Control Rod Position), and 0D-8 (Local Power Range Monitor readings) and then to consult the Nuclear Engineer for the method of returning the mispositioned rod to its correct in-sequence positio Contrary to the procedure, on May 26, 1986, the operator failed to demand the required process computer printouts and failed to consult the Nuclear Engineer prior to returning the mispositioned control rod to its correct in-sequence position. This is considered to be a violation

' (374/86020-02(DRP)). In this case, fortuitously, the operator action was what the nuclear engineer would have recommended. The inspector is concerned, however, that the operator took the action prior to consulting with the nuclear engineer. Depending on circumstances, this practice could complicate a problem rather than help it. In addition, during an interview with the unit operator subsequent to the event it became apparent that the operator was not aware that he had selected and moved another rod prior to discovery of the mispositioned rod, a result of not obtaining and reviewing the required process computer printouts, 00-3, OD-7 Option 2, and OD- The timeliness of the independent verifier noting the error was the second concern. The inspector noted that the duties of the verifier, i.e., what was expected of him, were not clearly defined. However, the action statement for Technical Specification 3.1.4.1 (Rod Worth Minimizer)

states that with the RWM inoperable control rod movement and compliance with the prescribed control rod pattern shall be verified by a second licensed operator or other technically qualified member of the technical staff who is present at the reactor control console. Thereforg this individual should have been aware of his duties even without th[e b of a prescriptive procedure. (The inspectors must note again, however, that the referenced technical specification does not apply above a reactor power of 20%. Therefore, there was no technical specification requirement for the -second operator to verify rod movement.) The licensee plans to revise the appropriate procedures to provide this guidance. Com (374/86020-03(pletion DRP)). of this action will be tracked as an open item The failure of the unit operator to log the error is another item of concern. In this case, the operator was instructed not to make an entry until the shift engineer obtained clarification on the seriousness of the error. The shift engineer then neglected to provide the unit operator with the appropriate clarification. The inspector has previously expressed concern with the adequacy of log entries. Furthermore, the inspector is concerned that the operator did not make an entry because of instructions from the shift engineer. The log book in question is the unit operator

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V log and is required to contain all pertinent information related to operation of the facility. The seriousness of the error has no bearing on whether it should be recorded. The log book is used to record facts; analyses of these facts can be done in another foru Secondly, the inspectors are concerned that the shift engineer instructed the unit operator not to make the initial entry. As the senior management representative onsite he should be aware of the procedural requirement If he so desires, his log book can describe clarifications of operational events but he should not instruct operators to not record informatio Corrective actions for this example of inadequate logs will be monitored as an open item (374/86020-04(DRPP).

In summary, a control rod inadvertently was mispositioned to Position 48 instead of to Position 24. Although not required by technical specifi-cations for the operational conditions at the time (21% power) a second operator had been assigned to the control room as a verifier but did not notice the mispositioned rod until the next rod in the sequence had been move During the month of May, the inspector walked down the accessible portions of the following systems to verify operability:

Units 1 and 2 Emergency Diesel Generators Units 1 and 2 Standby Gas Treatment Systems 3. Monthly Surveillance Observation (61726)

The inspector observed technical specifications required surveillance testing and verified for actual activities observed that testing was performed in accordance with adequate procedures, that test instrumenta-tion was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with technical specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne The inspector witnessed portions of the following test activities:

LIS-NB-214 Calibration of Reactor Vessel Pressure Switch 2821N039N LST 86-096 Unit 2 Special Test to Check Trip Point of Low Level Scram Switches LIS-NR-402 Intermediate Range Monitor Rod Block and Reactor Scram Fun tional Test LIS-NB-204 Unit 2 Reactor Vessel Low-Low Water Level RCIC Initiation and Low-Low-Low LPCS/RHR Initiation Calibration

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Discussions with licensee personnel indicate that this valve has a history of hydraulic lock, a condition that allows leakage past the valve seating surface into the bonnet area. This high pressure water then provides a hydraulic lock between the valve bonnet and the wedge, preventing valve movement. To release the hydraulic lock, the licensee routinely loosens the valve stem packing to vent the bonnet area. The licensee suspects that attempts to open the valve against the hydraulic lock may have resulted in overtorquing the operator and causing the failure. The drive sleeve failure also suggests the possibility of a fatigue failure due to the presence of rust and signs of aging in portions of the break are The inspector continues to be concerned about the possible generic problem of this failure for other limitorque valve operators as to the root cause of equipment failure and corrective actions as well as how to resolve hydraulic lock problems of valves in this system and other systems. Until licensee evaluations and corrective acticn plans are completed, this will remain as an open item (374/86020-05(DRP)).

5. Training (41400)

The inspector, through discussions with personnel and a review of training records, evaluated the licensee's training program for operations and maintenance personnel to determine whether the general knowledge of the individuals was sufficient for their assigned tasks.

t Specific areas reviewed are identified in Paragraphs 2, 3, and 4. The adequacy of training to prevent personnel from over torquing limitorque valve operators was identified as a concern. Personnel have not been provided with specific instructions that would limit the amount of force applied to valve operators. This contributed to the failure of a valve operator as noted in Paragraph This concern was identified to the licensee for evaluatio . Unit Trips (93702)

On May 9, 1985 at 9:10 a.m. CDT, LaSalle Unit 2 experienced a reactor scram from 85% reactor power. The unit scransned on low reactor water level due to the loss of power to the feedwater control system. With a loss of the feedwater control system, the "B" Turbine Driven Reactor Feedwater (TDRFP) Pump " locked up" and maintained a constant speed. The

"2A" TDRFP did not lock up and coasted down. The motor driven feedwater pump was started but its feedwater control valve locked up at 20% ope With a decrease in feedwater flow, vessel level decreased to the low reactor water level scram setpoint and the unit scrammed. The loss of power to the feedwater control system was caused by a worker accidentally bumping and tripping a 120V power supply disconnect breaker to the feedwater control system. The licensee determined the lock up of the flow control valve and the coast down of the ATDFWP was due to the loss of power to the feedwater controlle ._ - . _

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