IR 05000373/1987033

Insp Repts 50-373/87-33 & 50-374/87-32 on 871103-30. Violation Noted.Major Areas Inspected:Previous Insp Findings,Operational Safety,Surveillance,Maint,Training, LERs & Regional Requests

ML20237B720
Person / Time
Site:	LaSalle
Issue date:	12/10/1987
From:	Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20237B704	List: IR 05000373/1987033 ML20237B703 ML20237B708
References
50-373-87-33, 50-374-87-32, GL-86-02, GL-86-07, GL-86-2, GL-86-7, GL-87-06, GL-87-6, IEIN-86-001, IEIN-86-1, IEIN-87-041, IEIN-87-41, NUDOCS 8712170089
Download: ML20237B720 (11)

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O. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report Nos. 50-373/87033(DRP); 50-374/87032(DRP)

Docket Nos. 50-373; 50-374 Licenses No. NPF-11; NPF-18 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: LaSalle County Station, Units 1 and 2 Inspection At: LaSalle Site, Marseilles, IL ,

Inspection Conducted: November 3 through November 30, 1987

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Inspectors: M. J. Jordan R. Kopriva e ) /

/ 1% 7 Approved By: Ms A Ring, Chief /2-/O-8 7 Reactor Projects Section 1C Date

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Inspection Summary Inspection on November 3 through November 30, 1987 (Reports N /87033(DRP); 50-374/87032(DR?))

Areas Inspected: Routine, unannounced inspection conducted by resident inspectors of licensee actions on previous inspection findings; operational safety; surveillance; maintenance; training; Licensee Event Reports; regional requests; and information meetings with local official Results: Of the eight areas inspected, no violations or deviations were identified in seven areas; one violation was identified in the remaining area (failure to adhere to procedures - Paragraph 2).

The licensed operators again showed their proficiency in recognizing the problems with the feedwater system and by taking prompt action to prevent an unnecessary scram (Paragraph 3). However, personnel errors again caused problems at the site such that the radwaste facility needed to be decontaminated due to a spill (Paragraph 2). Also, a concern was raised by the inspectors having to do with storage of equipment in the reactor building. The overall performance of the site for the month appeared to be good, such that, both units ran the entire mont PDR ADOCK 05000373 p DCD

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DETAILS

, Persons Contacted

• G. J. Diederich, Manager,'LaSalle Station

• R. D. Bishop,'. Services Superintendent J. C. Renwick, Production Superintendent

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D. Berkman, Assistant Superintendent, Work Planning

• W. Huntington,-Assistant Superintendent, Operations P. Manning, Assistant _ Superintendent, Technical Services T. 'Hammerich, Assistant Technical Staff Supervisor W. Sheldon, Assistant Superintendent, Maintenance J. Atchley, Operating Enginee *D. A. Brown, Quality Assurance Supervisor D. Winchester, Quality Assurance Engineer

• Richter, Assistant Technical Staff Supervisor

• Denotes personnel attending the exit interview on December 1, 198 LAdditional licensee technical and administrative personnel were contacted by the inspectors during the course of the inspectio . Licensee Action on Previous Inspection Findings (92701)

(Closed) Unresolved Item 373/87030-01; 374/87029-01: During this l inspection period, the inspector was able to further investigate these unresolved items pertaining to the October 6, 1987, resin spill in the .{

spent resin tank room.~ Due to inadequate communications and several errors, approximately 100 cubic feet of spent resin was pumped into the spent. resin tank room covering the tank room floor to a depth of 4 to 10 inches. The dose rate in the room went from approximately 30 mr/hr prior to the spill to 1.5 to 2.0 rem /hr. after the spill. The investigation revealed several procedural errors,. personnel errors and inadequate outages _of equipment involve Technical Specification 6.2.A states, in part, " Detailed written procedures, including applicable checkoff lists covering items listed below shall be prepared, approved and adhered to:

The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978 "

Appendix A of Regulatory Guide 1.33 includes administrative procedures I for equipment contro LAP 900-4, " Equipment Out of Service (00S) Procedure," step F.1.g states, in part, "The ' Supervisor in Charge of the Work' has the responsibility i to assure that an inspection has been made to see that out of service  !

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have been placed correctly and that the equipment is safe to work on. L

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LAP 900-4, " Equipment Out of Service Procedure," step F.2.a states, in part, "To clear an outage the ' Supervisor in Charge of the Work' for whom the out of service cards were placed, shall be responsible for having an inspection made to assure that the equipment is cleared of his personnel, obstructions, and all personnel protection cards."

LAP 900-12. " Caution Card Procedure," step F.4.a states, in part "When the person requesting the caution card determines the card is no longer required, the requestor or his/her designate shall remove the card and deliver it to the shift engineer or the appropriate shift foreman...."

Contrary to the above, on October 6, 1987, a contractor (Morrison) crew started disconnecting the suction and discharge piping to the spent resin pump in preparation for connecting the air driven motor to the suction and discharge piping. At this time, the spent resin pump was not out of service, with personnel working around and on the syste At 5:25 p.m., the station construction engineer contacted the radwaste foreman and asked him to temporarily lift the out of service cards on the air operated valves such that the valves could be opened and the spent resin pump run. The radwaste foreman was responsible for the out of service cards on the valves and agreed to the temporary lift, however, the foreman did not perform an inspection to assure that the equipment was cleared of personnel, obstructions and all personnel prctection cards. If he/she had performed the inspection they would have noticed that the piping system was open and running the pump would have caused a spil Also, the radwaste foreman and station construction engineer noticed a caution card on the radwaste control panel indicating that the manual valve upstream of the air operated valves was closed which gave the radwaste foreman and station construction engineer false assurance that no resin beads would be spilled. The caution card was an old one from August 24, 1987, and had not been properly cleared. In fact, the manual valve was ope These items of not adhering to procedures all contributed to the resin spill in the spent resin tank room and this issue is considered a vi01ation (373/87033-01; 374/87032-01).

One violation was identified in this are . 0_perational Safety Verification (71707, 71881, 71709) The inspector observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during 3 the inspection period. The inspector verified the operability of I selected emergency systems, reviewed taguut records, and verified l proper return to service of affected components. Tours of Unit 1 l and 2 reactor buildings and turbine buildings were conducted to j observe plant equipment conditions, including potential fire hazards, fluid leaks, excessive vibrations, and to verify that j l

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, A .g 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 s stetton security plan including the following:

the appropriate' number of security personnel were on site; access control barriers were operational; protected areas were well maintained, and vital area barriers were well maintained. The, ,

inspector verified the licensee's radiological protection prog hm was implemented in accordance with the facility policies and programs and in compliance with regulatory requirement During the month of November 1987, the inspector walked down the '

accessible portions of the following systems to verify: operability: ,

Unit 2 Reactor Core Isolation Cooling Unit 2 A & B Residual Heat Removal System Unit 2A & 2B Diesel Generators b. On November 5, 1987, at 1:15 a.m. (CST), while performing operating surveillance LOS-TG-W1, " Unit 2 Turbine Weekly Surveillance," Unit 2 experienced a feedwatar cransient. The unit was operating at 92%

power with both the 2A and.2B turbine driven reactor feedwater pumps (TDRFP) supplying feedwater te the reactor. The 2A TDRFP increased <

speed, supplying feedwater to the reactor at a high rate of flo ?

Reactor vessel water level had increased to approximately +50 inches '

(+38 inches is normal operating water level) when the 2A TDRFP s tripped on an overspeed signal. Withdecreasedfeedwaterflow,the reactor vessel water level started decreasing. The water level decreased low enough that the reactor recirculation (RR) flow control valves began a runback (start closing). Through the unit operator's prompt actions, the reactor vessel level was restored by ,

use of the Unit 2 motor driven reactor feed pump (MDRFP) which >

prevented a reactor scram. During the runbacks of the RR FCVs, the A RR FCV locked up at 40% (open) and the B RR FCV locked up at 20% (open). Normally during a runback, the FCVs are supposed to runback to 12% ope The problem with the 2A TDRFP increasing speed resulted from a broken air supply line to the 2A TDRFP control valve positione This caused the 2A TDRFP control valve to go wide open increasing turbine speed and subsequently tripping on turbine oversped. The air supply line to the 2A TDRFP control valve positioner was repaired and the positioner tested. The 2A TDRFP was returned to service on November 9, 198 l The problem with the RR FCVs not running back to 12% is suspected to ,

be from excessive friction of the valve stem of the FCV. During the

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' f FCV run backs, the hydraulic power unit (HPU) for positioning the l FCVs tripped from excessive difference in FCV position demand versus actual FCV position. The HPU switched over to the secondary HPU as expected and when the secondary HPU could not reduce the difference inpositionversusdemand,ittripped/.ndlockedtheFCVin

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position. Thus, from the slow response times of the FCVs, the licensee suspects that the friction of the valve stem in the FCV is increasing. The licensee is evaluating what actions to pursue on the FCVs during the next Unit 2. refueling outage, The inspector brought to the attention of the station manag'ement his ,

concerns that the utility was using the reactor buildinas to store equipment which was currently not in use. Housekeeping u s not a problem in that the equipmei.t was stored neatly, however, the use of the reactor building to store equipment until needed is not a good practice. The inspectors noted some guard rails used tb prevent personnel from falling into an open hole when ccncrete plugs are removed from the floor were stored in Unit 2 on two metal carts against a couple of instrument lines used for the high pressure core spray system. The inspector felt that although the lines were not currently damaged, they could become damaged. Also, a wooden frame desk was constructed and stored or staged in Unit 2 for the Unit i refueling outage currently scheduled for March 13, 1988. The desk is to be used during the refueling outage by health physics to issue dosimetry and review Radiological Work Permits (RWP's) prior to entering the primary containment (drywell). The licensee had also constructed several metal cages around the reactor building for storage of all types of equipment for mechanical maintenance, electrical maintenance, instrument and control maintenance and technical staff. The cages were all tidy and neat on the inside, but storage of equipment in the reactor building is not a good practice because it may lead to eventual clutter and potential fire hazard No violations or deviations were identified in this are . MonthlySurveillanceObservation(61726]

The inspector observed Technichi Specification required surveillance testing and verified for actual activities observed that testing was performed in accordance with adequate procedures, that test instruments-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 Specification 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 inspecto witneued portions of the following test activities:

LOS-AA-W1 Techcical Specifications Weekly Surveillance (Control Rod Drive Lgling)

LOS-HP-Q1 Unit 2 High Pressure Core Spray System Inservice Test for Operating, Startup, Hot Shutdown, Cold Shutdown, and Refuel Conditions When Cycled Condensate Lines are Isolated

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a. On November 5, 1987, the licensee performed LOS-HP-Q1, " Unit 2 High Pressure Core Spray (HPCS) System Operability and Inservice Test."

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l HPCS pump differential pressure (dP). The HPCS pump dP, as measured, had exceeded the required action range specified in the Ql b '"

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, procedure. The actual pump dP was measured as 380 psid and the high

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/- value for required action is anything greater than 365 psid. A surveillance evaluation was initiated due to a parameter (s) being in the required action range. The evaluation concluded that the high HPCS pump dP value was acceptable based on the comparison with the

/ Unit 2 HPCS pump curve, which had been reviewed and found acceptable A as documented in General Electric Company's Field Deviation

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Disposition Request (FDDR) No. HA2-339 dated February 1983. The percent difference between the total dynamic head (TDH) and dP as i measured on November 5, 1987, and the pump curve TDH and dP was less than 1%.. Based on this comparison, the HPCS pump performance has not changed since February 1983. The licensee expected the HPCS pump dP to be high. When the licensee ran LOS-HP-Q1 on November 5, 1987, the data obtained was to be incorporated into the procedure as new baseline dat Procedure LOS-HP-Q1 will be revised using the guidelines from the Inservice Testing (IST) Program. Relief Request (RP-11) will be used to set new acceptance criteria for the Unit 2 HPCS pump dP. Per RP-11, the new procedure will incorporate the new base line data for the HPCS pump dP showing only a required action range and eliminating the alert range readings. The required action range for the HPCS pump dP will be 10% of the as measured pump dP on November 5, 1987. Any dP measured above the HI acceptable range or below the LO acceptable range will be considered to be in the required action rang A caution card was hung on.the Unit 2 HPCS pump control switch to alert the nuclear station operator of the new acceptance criteria for the HPCS pump dP. The caution card will be removed once procedure LOS-HP-Q1 has been revise b. Four (4) Static-0-Ring (SOR) differential pressure switches on the Unit 2 reactor core isolation cooling (RCIC) system failed to pass the diaphragm integrity test during functional test LIS-RI-401 (Unit 2 Steam Line High Flow RCIC Isolation Functional Test). The test i consists of isolating the switch and pressurizing one side of the SOR switch to check and see if the diaphragm has torn or has holes in it. The switches involved provide steam line isolations upon detection of high flow cceditions. The following failures occurred:

Date Switch (2E31) Model S/N Comments l

11-19-87 N013AB B203 85-1-2424 N007AA/AB and N013AA were sa N013AB was replaced.

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11-23-87 N013BA B203 85-1-2426 The N007AA and and N007AA B203 85-1-2419 N007AB switches 11-24-87 N007AB B203 85-1-2420 had passed the q surveillance on f November 19, 1987 )

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diaphragm integrity test on November 23, 198 As a result of the failure on Unit 2, all Unit 1 RCIC steam line '

high flow isolation switches were functionally tested in accordance with LaSalle Instrument Surveillance LIS-RI-301 (Unit 1 Steam Line High Flow RCIC Isolation Functional Test) on November 23, 198 LIS-RI-301 results were satisfactory and all switches successfully passed the diaphragm integrity tes Prior to this event (on October 21, 1987), Unit 2 RCIC steam line high flow isolation. switch 2E31-N013AA (Model B203, S/N 8512423)

failed its diaphragm integrity test during the performance of LaSalle Instrument Surveillance LIS-RI-201 (Unit 2 Steam Line High Flow RCIC Isolation Calibration). The . switch was replaced at that tim The root cause of this event has not been determined. All differential pressure switches which failed the diaphragm integrity test will be sent to the manufacturer (SOR) for inspection the week of November 30, 1987. No open item will be assigned to this item at this time. The licensee will issue a Licensee Event Report (LER)

identifying the root cause for the failure. This item will be followed by the LE No deviations or violations were identified in this are . Monthly Maintenance Observation (62703)

During the inspection period, the inspector observed portions of the following maintenance activities:

Unit 18 Diesel / Generator Fuel Oil Day Tank Suction Strainer Cleaning and I Diesel Fuel Filter Change - Work Request #L72746 Unit IB Diesel / Generator Oil Leak Inspection / Repair - Work Request

1. L70156 Unit 1A Diesel / Generator Fuel Oil Transfer Pump Work - Work Request

1. L72008 Unit 1A Diesel / Generator High Temperature Switch (Element)

Repair / Replacement - Work Request #L72568 i Unit IA Diesel / Generator Fuel Oil Day Tank Suction Strainer Cleaning -

Work Request #L72743 Unit IA Diesel / Generator Motor Control Center (MCC) Cubical Work on Contactor IVY 06C

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E s On November 8, 1987, at 8:50 p.m. (CST), the control room ventilation system isolated and transferred to the emergency makeup system. The isolation was attributed to a problem with the B ammonia detector

'in the B train for. control room ventilation. The licensee investigated the cause of the failure. The cause of the failure was due to a burned out light bulb. The inspector investigated and determined that the licensee had a preventive maintenance (PM)

program to change the light bulbs annually. The bulb was changed in February 1987. This PM was verified to be part of the computerized maintenance program. Discussions with the Master Instrument Mechanic as to whether an annual replacement was adequate to prevent unnecessary isolations determined that a review of the previous isolation due to a burned out light bulb had not occurred sooner than annually. Therefore, annual replacement would be sufficien '

No further action was required on this event. The system functioned ,

as expecte ' On November 14, 1987, at 5:29 p.m. (CST), the A control room ventilation system isolated and transferred to the emergency makeup system. The isolation was due to the A ammonia detector failur The licensee investigated the cause. The cause was determined to be the tape in the detector broke which caused the detector to sense a high ammonia and caused the isolation. The tape was replaced and the system returned to servic No violations or deviations were identifie . 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. In the areas examined by the inspector, no items of concern were identifie No violations or deviations were identified in this are . Licensee Event Reports (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following Licensee Event Reports (LERs) were reviewed to determine that deportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification /87002-01 - Failure of 0 Diesel Generator (DG) output breaker to close onto Bus 241Y. The Unit 1 operator made two attempts to close the 0 DG output breaker on Bus 241Y, but the breaker would not close. The operating department with electrical maintenance in attendance racked the breaker from connect to test and found that the breaker would close while in test. After the breaker was racked to connect, it was able

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to be closed and passed its surveillance satisfactorily. Troubleshooting efforts on the breaker close circuitry revealed no discrepancies. All breaker components, including associated closure permissive contacts, were verified to operate as designed following the event. Since this event could not be duplicated under test conditions, the cause of the output breaker failing to close is unknown. This problem has not reoccurred. This item is close /87030-00 - Reactor scram while shut down. During the course of surveillance, LES-RP-107, " Control Rod Drive Charging Water Header Pressure Time Delay Relay Calibration and Functional Test", half scrams were generated as each time delay relay was tested. A full scram occurred when a half scram condition had not been reset prior to performing the test on a relay in the remaining (energized) reactor protection system (RPS) channel. The cause of this event was inadequate communication between the electrician performing the surveillance and the unit operato The safety consequences of this event were minimal since RPS responded per design to a low CRD charging water header pressure signal. Since all control rods were fully inserted at the time of this event, no rod motion occurred as a result of the scra Corrective actions were take This item is close No violations or deviations were identified in this are . Regional Requests (92701) Per a telephone conversation with a Region III Section Chief, the inspectors were requested to find out how the licensee tested their Main Steam Isolation Valves (MSIVs) in which the section chief referenced Information Notice 85-84, " Inadequate Inservice Testing of Main Steam Isolation Valves", dated October 30, 1985. The MSIVs are supposed to be able to close without steam flow assistance or non-safety related instrument air power. The licensee has indicated that when they perform MSIV testing, they do not isolate the instrument air supply. However, they do perform an Inservice Inspection (ISI)Section III Test when they do close the MSIV This test basically isolates the air accumulators and then monitors the leakage of the accumulators. ibis test provides assurance that the accumulators will hold a specified volume of air at the required pressur The licensee has performed a functional test of the MSIV's being

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closed just on the stored air supply in the associated accumulators, l

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but this test was only performed during preoperational testing. The results of the testing were acceptabl (Closed) Generic Letter (GL) 87-06: In accordance with a Regional Request from Charles E. Norelius to Region III Senior Resident Inspector, dated October 20, 1987, the inspectors reviewed the licensee's action concerning GL 87-06. This generic letter (

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addresses the testing of pressure isolation valves (PIVs). The licensee's response was that the Technical Specifications established the required leak. rate testing of the PIVs. The LaSalle Technical Specification lists the PIVs to be tested and gives the maximum leakage allowed, the testing frequency, and action to be taken when leakage exceeds the limits. The action required by this GL is considered closed (373/87033-02; 374/87032-02). (Closed)GenericLetter(GL)86-02:# In accordance with a Regional Request from E. R. Schweibinz for C. E. Norelius and his respective

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i Branch Chiefs dated March 20, 1986, the inspectors reviewed this generic letter concerning thermal hydraulic stability. This generic letter was concerned that cores having a calculated decay ratio of 0.80/0.75 may, in fact, be on the verge of limiting cycle oscillations. The licensee for both Unit 1 and 2 reload licenses submittals in 1986 and 1987, respectively, established a core stability decay ratio of 0.60 for dual recirculation loop operabilit For a single loop operation, restrictions by the limiting condition for operation have been imposed by Technical Specification 3.4.1.1.a.2 to prevent the 0.80 to 0.75 decay ratio stabilit The action required by this GL is considered closed (373/87033-03; 374/87032-03).

/ (Closed) Generic Letter (GL) 86-07 and Information Notice (IN) 86-01: In accordance with a Regional Request from Charles E. Norelius to Region III Senior Resident Inspectors dated October 20, 1987, the inspectors reviewed the licensee's actions concerning GL 86-07. The subject of this GL was the transmittal of NUREG-1190 regarding the San Onofre Unit i loss of power and water hammer event. The licensee had previously reviewed this event under IN 86-01 which addressed the same event and determined that this would not be a problem at LaSalle due to relief valves on the feedwater heaters and suction header to the feedwater pumps which would prevent this type of damage. Also, the low pressure heaters' tubing was hydrostatically tested by the manufacturer to 1275 psig and has a working pressure of 850 psig. The generic letter, information notice, including results of investigations, and NUREG 1190 were then transferred to the training department for their incorporation into the operator requalification program as recommended by the GL. This information was included into the January and February of 1987 requal program. The GL 86-007 is considered closed (373/87033-04; 374/87032-04) and IN 86-01 is also considered closed (373/87033-05; 374/87032-05).

e. (Closed) Information Notice (IN) 87-41: In accordance with regional request dated October 27, 1987 from C. E. Norelius to Senior Resident Inspectors, the inspectors reviewed the licensee's action concerning NRC IN 87-41 concerning failure of Brown Boveri electrical circuit breakers. The licensee determined that no Brown Boveri circuit breakers were utilized at LaSalle in safety or non-safety application. Therefore, this information notice does not

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I _ apply at this facility.. However, LES-GM-103 which is accomplished on circuit breakers every 5 years requires checking the mounting bolts on the closing spring charging motor and inspecting the cubicle for loose nuts, bolts and screws and tighten if necessary, L .as recommended by this information notice . The IN 87-41 is L considered. closed.(373/87033-06;374/87032-06).

No violations or deviations were -identifie . Information Meetings With Local-Officials (94600)

On the evening'of November 17, 1987, the licensee conducted a tour of the site with approximately 35 local elected officials as'part of their public relations program and to answer any questions they had about the facility. The senior resident inspector and resident inspector were

.on site, in the plant performing inspections at the time of the tou After the tour, the SRI addressed the officials as to what the Nuclear Regulatory Commission (NRC) is, the purpose and roles the resident

' inspectors have on site, and answered any questions they had' pertaining to the NR ' 10. Exit Interview (30703)

The inspectors met with . licensee representatives (denoted in Paragraph 1)

throughout the month and at the conclusion of the inspection period and summarized the scope and findings of the inspection activities. The

' licensee acknowledged these findings. The inspector also discussed the likely-informational contents of the inspection report with regard to documents or processes reviewed by the inspector during the inspectio The licensee did not identify any such documents or processes as proprietar l i

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