IR 05000373/1986033
| ML20214R001 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 09/15/1986 |
| From: | Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20214Q979 | List: |
| References | |
| TASK-2.E.4.2, TASK-2.K.3.18, TASK-TM 50-373-86-33, 50-374-86-34, IEB-86-002, IEB-86-2, NUDOCS 8609290041 | |
| Download: ML20214R001 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-373/86033(DRP); 50-374/86034(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: July 22 through September 2, 1986 Inspectors: M. J. Jordan
J. Bjorgen R. Kopriva J. Mueller Approved By: . ght, Chief Reactor Projects Section 2C 7!/6 [[7 Date '
- Inspection Summary
' Inspection on July 22, through September 2, 1986 (Reports No. 50-373/86033(DRP); 50-374/86034(DRP)) , Areas Inspected: Routine, unannounced inspection conducted by resident ! inspectors of licensee actions on previous inspection findings; operational ! safety; surveillance; maintenance; training; followup of 10 CFR 50.54(f) request for information; Licensee Event Reports; TMI action plan requirement
- followup; and licensee action on an IE bulletin.
i Results: The licensee's overall performance during the inspection period was l good. No violations were issued and followup response to events seemed to be thorough. The thoroughness to review events was also monitored by licensee corporate personnel which aided the station in its review.
l 8609290041 860916 PDR ADOCK 05000373 G PDR
. DETAILS Persons Contacted
*G. J. Diederich, Manager, LaSalle Station *R. D. Bishop, Services Superintendent J. C. Renwick, Production Superintendent 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 Engineer *R. W. Stobert, Quality Assurance Suparvisor * Denotes personnel attending the exit interview on September 2, 198 . Licensee Action on Previous Inspection Findings (92701) (Closed) Open Item (373/81-00-143(DRP)): This open item tracked a' licensee commitment to perform a shutoff head test on at least two service water pumps at each refueling outage. Vibration readings are then to be taken on the acceptable pumps at operating flow for reference and analysi The licensee has issued surveillance LTS-1000-28, " Service Water Pump Vibration and Shutoff Head Test." The inspector reviewed this surveillance and found it technically acceptable. This item is considered close (Closed) Open Item (373/81-00-41(DRP)): This open item tracked Unit 1 License Condition 2.C(22). This condition required that prior to startup after the first refueling outage for the 125 and 250 volt direct current systems of Divisions 1 and 2 and the 125 volt Division 3 direct current system, the licensee provide instrumentation in the control room for: (1) battery current (ammeter-charge discharge), (2) battery charger output voltage (voltmeter), (3) battery charger output current (ammeter), (4) battery high discharge rate alarm, and (5) battery charger trouble alarm. The inspector verified that the required instrumentation was installed. This License Condition is considered close (Closed) Open Item (373/81-00-42(DRP)): This open item tracked Unit 1 operating License Condition 2.C(23) which required the licensee to install redundant fault current protection devices (circuit breakers and/or fuses) on each primary containment penetration circuit prior to startup after the first refueling outage. Each of the devices is required to limit a fault current surge to be less than the surge for which the penetration is qualified such that the mechanical integrity of the electrical penetrations is maintained for a single random failure of a circuit overload protection device. The inspector confirmed that the redundant devices have been installed. This item is considered close . 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
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. 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 During the month of August 1986, the inspector walked down the accessible portions of the following systems to verify operability: Unit 1 and 2 Emergency Diesel Generators Unit 1 and 2 Standby Gas Treatment Systems Unit 1 and 2 Hydrogen Recombiners On June 1,1986 LaSalle Unit 2 experienced a reactor feedwater transient which caused water level to decrease below the low water level scram set point without the unit automatically shutting down. On June 2, 1986 the NRC issued a Confirmatory Action Letter (CAL) which identified a number of actions the licensee planned to accomplish and stated that Unit 2 (and Unit 1 if the problem were generic) could not be restarted without NRC concurrenc As a result of the licensee's investigation into the June 1 event, the licensee embarked on a testing program to determine the scope of the problem with the Static-0-Ring switches, and determine which switches were acceptable for continued use to support plant operation. The testing program and its results were closely monitored by the NR The results of these tests served as the cases for a Safety Evaluation Report (SER) issued on August 7,1986 by letter from Mr. J. G. Keppler to Mr. Cordell Reed. This report concluded that the licensee's proposed actions supplemented by the actions listed in the SER, and the test results of the switches provided an adequate basis for the restart and short term operation of LaSalle County Station Unit The NRC met with the licensee on July 9, 1986 to discuss actions taken by Commonwealth Edison Company in response to the June 2, 1986 CAL. All items listed in the CAL were found to be satisfactorily completed with regard to LaSalle Unit 2. This completes open item (374/86034-01(DRP)). Based on the conclusions of the SER, the licensee's actions in response to the CAL and recommendations of the NRC staff, the NRC authorized LaSalle Unit 2 to restart and operate in accordance with the provisions of the SER by letter from Mr. J. G. Keppler to Mr. Cordell Reed dated August 7, 1986. LaSalle Unit 2 commenced restart on August 8, 1986 and performed a level drop test successfully on August 10, 1986. The unit was restarted and was synchronized to the grid on August 12, 198 . . On August 18, 1986 while at approximately 70% power, Unit 2 "A" Turbine Driven Feedwater Pump (2ATDFP) tripped on high temperature on the seal injection line for the turbine. While trying to balance the injection between the inboard and outboard seals, the turbine received a high temperature trip. The 2BTDFP was operating in unison with the 2A at the time. Upon loss of the 2ATDFP, the unit operator started the Motor Driven Feed Pump (MDFP) and recovered reactor water level before it got below 31" (Tech Spec Scram Point is 11.0"). A followup investigation by the licensee identified that during construction, a wiring installation error was mad In 1984, the station had implemented a temporary system change for the Unit 2 TDFPs to eliminate the trip of the turbine on high seal water temperature. A review of the wiring revealed that it was schematically correct but was not according to the wiring diagram which was used to design the system change. The temporary change was installed using the schematic diagrams in lieu of the wiring diagrams, thus, the wiring error was not identified at that time and resulted in the trip, of the turbine, on high exhaust temperature for the inboard seal not being bypasse After identifying the trip of the 2ATDFP, the immediate action the licensee took was considered to be good. Power was reduced slowly so that if the similar trip on the 2BTDFP was also not bypassed, the MDFP could handle a feedwater transient. Once the problem was identified as a wiring error, a check was made on the temporary system change for the 2BTDFP. This change was found to have been satisfactorily implemented and power reduction was stopped. A further look at Unit 1 identified a modification completed during the current outage removed the entire trip sensors and logic chain so the problem could not occur on Unit Other corrective actions the licensee took on this item seem to be extensive to assure other problems of a similar nature do not occur in the future. The only outstanding action that the inspectors will follow up on is the change to LEP-240-6 on installation of temporary system changes and how the handling of wiring diagrams will be used and if any testing after installation will be conducted. This will remain as an open item (374/86034-02(DRP)). 4. Monthly Surveillance Observation (61726) The inspector observed Technical Specification required surveillance testing and verified for actual activities observed that testing was performed in accordance with adequate procedures, that test instrumentation 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 . - _ _
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. The inspector witnessed portions of the following test activities: LST-86-174 - Reactor Water Level Drop Test For Static-0-Ring Switch Response LST-86-183 - Reactor Water Level Drop Test For Static-0-Ring Switch Response at Pressure LIS-NB-201 - Reactor Vessel Low Water Level Scram and Primary Containment Isolation Calibration SOR Switches B21-N024 A, B, C, and D In response to IE Bulletin 81-03, " Flow Blockage of Cooling Water to Safety System Components by Corbicula sp. (Asiatic Clam) and Mytilus s (Mussel)," LaSalle Station conducted inspections of several areas of the plant for the presence of Corbicul On August 22, 1986, a diver inspected and took sediment samples from two areas of the Lake Screen House: the north end of the service water tunnel and traveling screen bay I The diver observed that the number of Corbicula had increased in the service water tunnel since he performed a similar inspection in August 1985. This was the first year that the licensee inspected traveling screen bay I The clams taken from this area by the diver were identified as Corbicul On September 3, 1986, the High Pressure Core Spray (HPCS) Diesel Cooling Water Strainer (1EZZ-D300) was inspected and no indication of Corbicula was presen As a result of these inspections, the Environmental Affairs Department determined that the quantity of Corbicula at LaSalle is " insignificant" and would not affect the operation of plant equipmen Prior to this inspection, the licensee's follow up to their response to IE Bulletin 81-03 was not well documented. While an annual inspection of-the service water tunnel apparently was performed in 1984 and 1985 as committed to in LaSalle's response, no written procedure nor any record of the results were maintaine Due to the inspector's concerns, the licensee is in the process of writing a procedure for the performance of a Corbicula monitoring program. The procedure will generate a record of each inspection's findings for future trending and analysis. This procedure will be performed annually and is expected to be in effect by March 198 . Monthly Maintenance Observation (62703) Station maintenance activities of 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 in conformance with Technical Specification . . The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from 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 implemente Work requests were reviewed to determine 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: On July 22, 1986, the Unit 2 operator was removing the power supply fuses for valve 2E51F026, the Reactor Core Isolation Cooling (RCIC) steam line drain valve,'in preparation for inspection and repair of heat shrink splices. The out-of-service required that fuses F-14 and F-15 be remove The operator inadvertently pulled fuses F-13 in lieu of F-15. The independent verifier failed to catch the error. The error went undetected until July 24, 1986 when the midshift foreman was reviewing LOP-RI-02M, the RCIC Mechanical Checklist for Unit 2 start up. The foreman noticed that the RCIC trip and throttle valve indicated as being de-energized, but that no outage could be found that de-energized the valve. Upon reviewing the appropriate drawings, the foreman determined that fuses F-12 and F-13 supplied power to the valve. Upon checking the fuses, the foreman noticed the outage for fuse F-15 was hanging in the F-13 fuse location. The error was immediately corrected. The noted error meets the criteria for a Notice of Violation. Based on the enforcement policy of 10 CFR 2, Appendix C, however, and the extensive corrective action committed to by the licensee in the management meeting on August 5, 1986 as discussed in Paragraph 7, no violation will be issue At 12:30 a.m. on August 7, 1986, the IB emergency diesel automatically started on a low bus voltage signal when the normal feed breaker trippe An equipment operator was inside the diesel generator control cabinet at the time in preparation for hanging an outage for the diesel fuel oil priming pump. The unit process computer alarm printer noted trouble with the high pressure core spray pump room cooler and sump pumps about the time of the diesel start causing a low voltage condition on the 480 volt bus 143- The resident inspector attended the initial investigation meeting and will follow the licensee's action On August 23, 1986 at 7:45 a.m. Unit 2 while at approximately 85% power, the Reactor Water Cleanup System (RWCU) isolated due to high differential flow. The isolation was due to the lifting of a relief valve on the 2B RWCU regenerative heat exchanger. The 2A regenerative heat exchange relief valve also had weepage by its seat and so both heat exchangers were isolated. The relief valve from Unit I was removed, bench tested to assure seating, and installed on Unit 2 B heat exchanger. A followup
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. investigation by the maintenance department determined from the vendor, that the valve acts as a relief valve when used for a liquid medium and as a safety valve when used for gas medium. As such, it will " pop" open at its set point (SP) for gas, but will start opening at 90% of SP and will be full open at 110% of SP for liquid. With this new information, the licensee is evaluating replacing the valve or changing the methodology of setting the SP. In the meantime, they have assured themselves the present-valve will not lift at operating pressure, which is the discharge from the RWCU pump (1200 psi). This will remain as an open item (373/86033-01(DRP)) until final resolution on the valve setting or replacemen . Training (41400) The inspectors attended several training sessions provided to various station work groups to explain the lessons learned from a recent series of personnel errors, the progress and results of the S.0.R., In differential pressure switch testing program, and the licensee's plans for restart and operation of LaSalle Unit No items of concern were identifie . Followup of 10 CFR 50.54(f) Request for Information (71707, 30702) On August 5, 1986 a management meeting between the NRC and the licensee was held at the LaSalle Station to discuss the licensee's progress in resolving the NRC's concerns related to the overall operation of the LaSalle County Station. These concerns were expressed to the licensee by letter, dated Nove..ber 22, 1985. The meeting was attended by Mr. J. G. Keppler, NRC Region III Administrator and members of his staff, Mr. A. Bournia of the NRC office of Nuclear Reactor Regulation, and Mr. J. J. O' Conner of Commonwealth Edison and members of his staf The discussion included a progress report on the licensee's commitments for improved performance and a proposed plan for error free start up and operation of LaSalle Unit It was noted and agreed that significant progress had been made toward improved performance in all areas of concer The recent trend of increased personnel errors addressed in Inspection Reports No. 373/86018 and 374/86017 dated June 11, 1986 was also discussed along with the licensee's plans for minimizing personnel errors during the startup and operation of Unit 2. This plan was forwarded to the NRC by letter dated August 4,1986 from Mr. C. Reed to Mr. J. G. Keppler. The plan included an augmented onsite review by corporate management personnel prior to Unit 2 restart, additional onshif t personnel during startup, and increased management participation and specific hold points during power ascension to review plant status. Region III management discussed the plan with the licensee's management and concluded that the plan provided adequate management involvement to support Unit 2 restar . Licensee Event Reports (92700) Through direct observations, discussions with licensee personnel, and review of records, the following Licensee Event Reports (LER's) were reviewed to determine that reportability requirements were fulfilled,
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immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification /86031-00 - A spurious trip of the 2B Hi-Radiation Monitor actuated
"B" control room HVAC system emergency make-up train. This is an Engineered Safety Feature (ESF) actuation. The cause was a bad high voltage board in the relay indicator. The board was replace /86025-00 - A spurious trip of the 2C Hi-Radiation Monitor actuated "B" control room HVAC system emergency make-up train which resulted in an Engineered Safety Feature (ESF) actuation. Cause was unknown. The licensee is working with the vendor to discover cause. A supplemental report from the licensee is expected in September 198 /86021-00 - A spurious trip of the 2D Hi-Radiation Monitor actuated "B" control room HVAC system emergency make-up train which resulted in an Engineered Safety Feature (ESF) damper actuation. Cause was unknow Suspected that the detector had an electronics problem. The detector was replace /86030-00 - A spurious trip of the 2D Hi-Radiation Monitor actuated "B" control room HVAC system emergency make-up train which resulted in an Engineered Safety Feature (ESF) damper actuation. Cause was unknow Testing to identify the cause and to initiate corrective action will continue. A supplemental report from the licensee is expected in September 1986. The licensee will address the repetitive nature of this event (see preceding three Licensee Event Reports) in the supplemental repor /86017-00 - Reactor scram on low CRD pressure. Pump tripped on low CY tank level. This was an operator personnel erro /86022-00 - Shutdown Cooling (SDC) Isolation during surveillanc Residual Heat Removal (RHR) SDC suction inboard isolation valve inadvertently closed tripping "B" RHR pump. The cause was personnel error. A Notice of Violation was issued in Inspection Report 373/8602 /86027-00 "B" control room HVAC system antnonia detector tripped spuriously due to broken chemcassette in the detector caused by warped rubber capstan roller. This is an Engineered Safety Feature (ESF)
actuation. The rubber capstan roller and chemcassette were replace The licensee concluded that the problem did not seem to be generi A modification is scheduled to upgrade detector chemcassette /86011-00 - A Unit 2 feedwater transient caused reactor vessel level to drop below level 3 trip setpoint. Scram did not occur due to failure of level 3 switches to trip at set level. Followup is documented in Inspection Report No. 374/8602 (See Bulletin 86-02) , 374/84016-01 - Division 2 Reactor Water Cleanup (CE) high ambient temperature isolation signal was received on Unit 2 about five minutes after temperature switches were reading 85 degrees F (below 116 F trip point). The cause was unknown. This LER was revised to delete temperature isolations for the pump rooms from the Technical Specification _ . _ . __ _ . _ _ . _ _ _ _ _ _ . _ _ _ _ _ .
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374/85048-01 - Unit 2 Reactor Core Isolation Cooling (RCIC) water leg pump tripped on breaker thermal overload and could not be restarte The pump was replaced. This revision presents cause as insufficient lubrication of the pump's inboard bearin /86023-00 - Loss of continuity while installing a jumper during performance of LOP-RP-04, "RPS Bus B Transfer", caused inboard secondary containment isolation dampers to isolate and thereby tripped the reactor building ventilation system. A Notice of Violation was issued in Inspection Report No. 373/8602 /86026-00 - Primary Containment Air Lock Leak Rate Test not performed within six months prior to Unit 1 startup on December 31, 1983 contrary to Technical Specifications and 10 CFR 0, Appendix J. This was due to an error in the Computer General Surveillance Program. The next~ surveillance was performed on January 4, 1984. The surveillance schedule was correcte /85032-01 - Local Leak Rate Test of Unit 1 inboard feedwater check valve exceeded Technical Specification limits due to erosion of valve disc " soft seat" material. The LER was revised to provide additional test results and action plan to use "hard seats".
373/84054-02 - Contaminants in lubrication oil caused erratic movement of the Reactor Core Isolation Cooling (RCIC) turbine governor and a RCIC turbine mechanical overspeed. These resulted in a RCIC channel "A" steam line high flow isolation and a RCIC turbine trip, respectively. This revision presented two sources of contaminants: precipitation of a rust inhibitor and an amine vapor phase inhibitor causing corrosion of copper , cooling coils in the RCIC turbines. The oil was replaced with oil which had a different rust inhibitor and a non-amine vapor phase inhibito /86028-00 - A spurious trip of the "B" control room HVAC system ammonia detector occurred while the system was in the recirculation mode and the outside dampers were closed. The event was caused by a broken lead to a fuseholder which took control power away from the detector. The broken lead and other suspect leads were repaire /86024-00 - Several potential discrepancies in the design of certain small bore piping systems on LaSalle 1 and 2 were found in hand calculations for thermal stress on two inch and smaller diameter piping. All small bore piping subsystems designed using standard criteria via hand calculations
similar to those where the errors were found were audited to ensure they were analyzed correctly. Modifications were initiated on four subsystems to restore conformance to ASME criteri . TMI Action Plan Requirement Followup (25565) Closed (0penItem 373/81000-97C): TMI Item II.K.3.18C. Automatic Depressurization System (ADS) actuation modification was closed in Inspection Report No. 373/86017. The report stated that the Action
- Item Report (AIR) No. 81-616, which covered this TMI open item, was closed but did not state specifically that the TMI item was closed.
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Open(0penItem 373/81000-93): TMI Item II.E.4.2. Containment vent and purge valve modification. The inspector reviewed this open item and found the work associated with the modification completed. The modification package is in final review and should be completed by the week of September 8, 1986. The inspector will review this open item at that time for closur . Licensee Action on an IE Bulletin (92703)
(0 pen) IE Bulletin (373/86-02-BB; 374/86-02-BB): IE Bulletin " Static 0 Ring Differential Pressure Switches." The licensee issued a response to IE Bulletin 86-02 on July 25, 1986. On August 7, 1986, in a letter from Mr. J. G. Keppler to Mr. C. Reed of Commonwealth Edison, an Office of Nuclear Reactor Regulation (NRR) Safety Evaluation Report (SER) supporting the restart of LaSalle County Station Unit 2 was transmitted to the license The SER addresses the licensee's response to IE Bulletin 86-02 and concluded that the licensee has taken appropriate measures for the short term to assure that the systems using 50R 102 and 103 switches operate reliabl Therefore, we are able to close items 1 thru 5 which are listed under the section " Actions Required of All Licensees" within Bulletin 86-02 for LaSalle Unit 2. Item No. 6 will remain open as the licensee is still evaluating long term corrective actio Pertaining to LaSalle Unit 1, the licensee's response is adequate. The implementation of the action items listed in the bulletin are presently taking place or will be accomplished in the near future prior to and during start up of Unit On August 22, 1986 was issued as a Temporary guidance for inspection follow upInstruction of license(TI))2515/81 (es ' activities taken in response to IE Bulletin 86-02. The TI states that some licensees have only reviewed safety related systems rather than all systems important to safety as defined in 10 CFR 50.49. The inspector has reviewed the licensee's response to IE Bulletin 86-02 as outlined in TI 2515/81 and had determined that TI requirements 2515/1-04-01 A thru D have been completed appropriately with respect to all systems important to safet Item No. 2515/1-04-01 E remains open while the licensee determines what the long term corrective actions for the 50R switches will b . Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. Open items evaluated and closed during the inspection are discussed in Paragraphs 3 and . 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 j summarized the scope and findings of the inspection activitie The licensee acknowledged these findings. The inspector also discussed the likely informational content 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 }}