IR 05000373/1986025
| ML20204J344 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 08/05/1986 |
| From: | Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20204J330 | List: |
| References | |
| TASK-2.K.3.25, TASK-TM 50-373-86-25, 50-374-86-26, NUDOCS 8608110068 | |
| Download: ML20204J344 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Reports No:
50-373/86025(DRP); 50-374/86026(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
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Inspection Conducted: June 11 through July 21, 1986 Inspectors:
M. J. Jordan J. Bjorgen R. Kopriva J. Mueller Approved By:
G. C. Wright, Chief 88'
Reactor Projects Date Section 2C Inspection Summary Inspection on June 11 through July 21.1986 (Reports No. 50-373/86025(DRP);
50-374/86026(DRP))
Areas Inspected:
Routine, unannounced inspection conducted by resident inspectors of operational safety; surveillance; maintenance; training; regional requests; TMI action plan requirement followup; and Licensee Event Reports.
Results: Two examples of failure to follow procedures were identified: one in the maintenance area and one in the surveillance area. Both were the result of personnel error, i
8608110068 860804 PDR ADOCK 05000373 O
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DETAILS 1.
Persons Contacted
- G. J. Diederich, Manager, LaSalle Station
- R. D. Bishop, Services Superintendent J. C. Renwick, Production Superintendent D. Berkman, Assistant Superintendent, Technical Services W. Huntington, Assistant Superintendent, Operations R. W. Stobert, Quality Assurance Supervisor P. Manning, Technical Staff Supervisor T. Hammerich, Assistant Technical Staff Supervisor W. Sheldon, Assistant Superintendent, Maintenance J. Atchley, Operating Engineer
- D. Winchester, Senior Quality Assurance Inspector The inspectors also talked with and interviewed members of the operations, maintenance, health physics, and instrument and control sections.
- Denotes personnel attending the exit interview on July 22, 1986.
2.
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 plan.
The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls.
Both units remained in Cold Shutdown during this report period. Major activities included evaluation and testing of the S.O.R. Incorporated
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differential pressure switches, inspection and repair of heat shrink splices, and evaluation of a problem with the Intermediate Range nuclear instrumentation.
The licensee reported a security problem to the resident inspectors on July 15, 1986. The event occurred on July 14, 1986.
The event was discussed with the Region III security inspector who agreed to evaluate the incident during an inspection scheduled to start on July 22, 1986.
This evaluation will be documented in a future inspection report.
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During the month of June, the inspector walked down the accessible portions of the following systems to verify operability:
Unit 2 Division 3125 Volt Battery Unit 2 250 Volt Battery Unit 2 Residual Heat Removal Service Water Unit 2 Emergency Diesel Generators Unit 2 Standby Gas Treatment 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 instrumen-tation 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 personnel.
The inspector witnessed portions of the following test activity:
LOS-VG-MI - Monthly Operation of the Unit 2 Standby Gas Treatment System.
On June 12, 1986 while performing Procedure LES-PC-107, " Unit 1 Group 7 Isolation Logic Systems Functional Test," the Unit 1 operator inadvertently placed the Group 6 Shutdown Cooling isolation test switch 1821H-579D in test instead of Group 7 Transverse Incore Probe (TIP) isolation test switch 1821H-S19D. This resulted in isolation of the inboard shutdown cooling isolation valve and loss of shutdown cooling. The keylock switches are adjacent to each other in the control room back panel and are similar in appearance.
Operation of the wrong switch is a personnel error.
Technical Specification, Section 6.2.A.7, requires detailed written procedures for surveillance and testing requirements to be prepared, approved, and adhered to.
Procedure LES-PC-107, " Unit 1 Group 7 Isolation Logic Systems Functional Test", Step F.3.d requires switch IB21H-S19D to be placed in test to test the logic of the Transverse Incore Probe (TIP)
isolation.
Contrary to the above, the operator placed switch 1821H-S79D, the shutdown cooling system isolation test switch, in test rather than switch 1821H-S19D, which caused the isolation of shutdown cooling. This is considered a violation (373/86025-01A(DRP)).
4.
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 Specifications.
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The following items were considered during this review:
the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemented.
Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance.
The following maintenance activities were observed / reviewed:
Repacking of the 2E51-F008 Reactor Core Isolation Cooling (RCIC) System valve, Work Request L 58080.
Raychem heat shrink splice repairs in the Unit 1 outboard main steam isolation valve room, Work Request L 39599, and in the Unit 2 drywell at penetration E-9, Work Request L 59581.
The inspectors noted that the repairs that were performed in the Unit 2 drywell penetration E-9 consisted of taping over the Raychem splice with two layers of Okonite tape. The inspectors discussed this method of repair with the licensee's technical staff and a Region III specialist.
It was noted that the licensee has contracted with a test laboratory to assure that the tape over heat shrink repair method satisfies the requirements of the 10 CFR 50.49 environmental qualification program. The evaluation of this testing program will be performed by a Region III specialist and documented in a future inspection report.
The scope of inspection and repair of the Raychem heat shrink splices was briefly discussed with licensee personnel.
This issue was initially identified by the NRC during a team inspection at the Dresden Station in June 1986. Apparent misunderstanding by licensee personnel as to the minimum joint overlap, i.e., the amount of sleeve material extending on either side of the connection, as well as improper applications for heat shrink splices identified at the Dresden Station, resulted in the LaSalle Station initiating a reinspection of splices at LaSalle.
The licensee has developed a program to assure that all possible splice locations are identified, inspected, and repaired as necessary.
The evaluation of this effort by a Region III specialist will be documented in a future inspection report.
I On June 13, 1986 at 1:25 a.m., the licensee was performing LOP-RP-04, "RPS Bus B Transfer," on Unit 1 in preparation for maintenance. The procedure
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l calls for an electrician to install jumpers in panel IPA 14J to prevent a Reactor Building Ventilation System (secondary containment) isolation while j
transferring the Reactor Protection System (RPS) bus power supply. While
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installing a jumper, circuit continuity was lost causing the inboard isolation dampers to close and the supply and exhaust fans to trip.
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The dampers are designed such that the operating solenoids must be energized to maintain the isolation dampers open. As part of the procedure, an alligator clip jumper was placed across the relay contacts to maintain continuity while a lug style jumper was installed between terminal points.
The personnel involved did not realize that the alligator clip jumper was within the end points of the terminal point locations of the lug style jumper. Therefore, while the lug style jumper was being installed, the alligator clip jumper was not capable of maintaining continuity in the circuit. This resulted in the Group IV isolation (an Engineered Safety Feature (ESF) actuation).
Technical Specification 6.2.A.6 requires that detailed written procedures be prepared, approved, and adhered to for preventive and corrective maintenance operations which could have an effect on the safety of the facility. Licensee procedure LOP-RP-04 requires the installation of a jumper to prevent a Reactor Building Ventilation System isolation.
Contrary to the above, on June 13, 1986, while performing procedure LOP-RP-04, a jumper installed per the procedure did not preventa Reactor Building Ventilation System isolation. The cause of the problem was personnel error in failing to properly designate the termination points of the alligator clip jumper. This is considered to be a violation (373/86025-01B(DRP)).
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5.
Training (41400)
The inspector, through discussions with personnel and a review of training records, evaluated the licensee's training program for operations personnel to determine whether the general knowledge of the individuals was sufficient for their assigned tasks.
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The inspector observed special training for site and contractor personnel pertaining to proper installation of electrical connections using Raychem or Okonite splices. The training sessions were conducted by either plant personnel or vendor representatives explaining and demonstrating the proper use of their products.
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Upon completion of the training / instruction sessions, site and contractor personnel were able to participate in hands-on-training by preparing and
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i completing qualified electrical splices.
6.
Regional Request (92703)(92705)
a.
The Region III office, via C. Norelius' memo dated May 6,1986,
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i requested the inspector to complete Temporary Instruction 2515/77, t
" Survey of Licensee's Response to Selected Safety Issues." The following questions were requested to be answered:
Item A Reliability of Reactor Core Isolation (RCIC) System:
(1)
Is RCIC system tested for operational readiness (a) by cold, quick-start testing at appropriate intervals and (b) after
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(a) Yes.
Licensee surveillance LTS 500-8, RCIC system performance test, simulates a cold, quick-start in which the turbine has been. idle for at least one day.
It is performed every eighteen (18) months.
(b) Yes. After maintenance, the RCIC system is tested by its monthly and quarterly testing as outlined in the Technical Specifications prior to the system being declared " operable".
(2) Is (a) a documented, comprehensive preventive maintenance program carried out for RCIC system, (b) including records kept of maintenance and surveillance activities, and (c) are records of these activities used for scheduling and trend analysis?
(a) Yes. A " General Surveillance Program", LAP 100-11 is utilized at LaSalle for periodic preventive maintenance of RCIC equipment and components.
(b) A structured maintenance history system, described in Procedure LAP 300-11, has been developed in sufficient detail to provide:
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A means of documenting preventative and corrective maintenance.
ii. Means of performing trend analysis through user specified retrievals which enhance system reliability.
iii. Means of performing component failure analysis.
(c) See (b) above
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(3) Is a formal program for review of vendor service information for plant-specific applicability established?
Yes.
Detailed procedures for review of vendor service information for plant-specific applicability are established in procedures LAP 850-3 through LAP 850-6 and LAP 100-15.
Specifically addressed are: Service Information Letters (SILs), Turbine Information Letters (TIls), Technical Document Updates, and Equipment Technical Information (ETIs).
(4) Are the trip and isolation signals tested and calibrated as often as initiation signals?
Functional tests of RCIC trip and isolation signals are performed as often as those for RCIC initiation signals.
All instruments that trip the RCIC turbine or cause an isolation or auto-initiation are calibrated at least every eighteen (18) months as of January 1934.
(5) Is (a) RCIC room inspected every shift, and (b) are the humidity and temperature monitored and controlled?
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(a) Yes. The shift operators performing rounds look for a variety of potential problems such as leaks, overheating bearings, low lube oil level, and fire and safety hazards.
(b) Temperature in the RCIC room is monitored from a control room panel and verified once per day to be within the Technical Specification limits of 50 degrees F and 150 degrees F.
The RCIC room is normally ventilated with the reactor building ventilation system with a minimum supply air temperature of 50 degrees F.
If the area temperature increases to 104 degrees F, the Division I emergency room cooler equipment cooling fans and coolers initiate and provide additional cooling to the RCIC room.
There is no humidity control or monitor for the RCIC room.
(6) Has monitoring of (a) RCIC system performance and maintenance, (b) vendor operating experience recommendations, and (c) problems at other plants been assigned to a qualified engineer?
Yes. A Technical Staff Engineer is assigned to monitor each system at LaSalle including (a), (b), and (c) above.
(7) Is management review of systems performance, including tracking of implementation of remedial measures and effectiveness of those measures, being performed on a routine basis?
Deviation reports are reviewed for trend analysis each time one is written on RCIC. Therefore, the licensee sees no need for a routine review of deviation reports.
(8) Are any of the following NUREG-0737 (TMI) items not closed?
No, all are closed.
Below is a list of the Inspection Report Numbers which closed each item and the date closed.
Unit 1 Inspection Report No.
Date (a) II.K.3.13 84-033 02-13-84 (b) II.K.3.15 84-033 02-13-84
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(c) II.K.3.22 81-040 11-13-81 (d) II.K.3.24 86-011 04-14-86
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Unit 2 Inspection Report No.
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(a) II.K.3.13 83-037 10-10-83
(b) II.K.3.15 83-037 10-10-83 (c) II.K.3.22 84-036 12-04-84 (d) II.K.3.24 86-011 04-14-86 Item B Biofouling of Cooling Water Heat Exchangers:
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(1) Is instrumentation available on safety-related equipment cooled by.open-cycle service water systems for monitoring changes in flow and determining degradation of heat exchanger performance?
Instrumentation for monitoring flow and temperature on both the shell and tubt side of the Residual Heat Removal (RHR) heat exchangers is available at LaSalle. These instruments are capable of indicating a potential degradation of heat exchanger performance.
(2) Are instrument readings on safety related equipment cooled by open-cycle service water systems recorded and reviewed against design parameters (e.g. flow, differential pressure) on a routine basis?
In the course of performing procedure LTS-600-16, "RHR Heat Exchanger Capacity Verification," instrument readings of flow and temperature are compared to design values.
"ECCS Equipment Cooler Fouling Trend Surveillance", LTS-600-18, is used to obtain differential pressure measurements to allow monitoring for fouling, etc.
These LTS procedures are performed upon indication from the monthly and quarterly tests of the RHR and service water system that possible degradation of the system exists.
(3) Do procedures and training address operator actions if significant heat exchanger performance degradation resulting from fouling is detected?
Yes.
LaSalle's operating procedures detail operator actions in the event of degradation of RHR heat exchanger performance ( LOA-RH-03).
(4) Are periodic inspections performed to detect fouling in service water and fire protection systems?
Procedures LTS-200-1, " Turbine Building Closed Cooling Water Heat Exchanger Capacity Verification", and LTS-200-2, " Reactor Building Closed Cooling Water Heat Exchanger Capacity Verification", are performed to detect fouling in the service water system. However, j
they are not performed on a regularly scheduled basis.
A " Fire Suppression Water Systems Flow Test" (LTS-1000-15) is performed annually. This procedure is used to detect fouling in the fire protection water loop networks.
b.
The inspector followed up on a request frv.a the region to investigate the possibility of a failure of Intermediate Range Monitor (IRM) chass-is fuses for the positive and negative 24 Volt Direct Current (VDC)
buses resulting in a portion of the IRM being inoperative or unable to initiate a half scram. The licensee performed a test during the first week in July 1986 which confirmed that LaSalle has a problem that if l
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the (-)24 VDC IRM fuses were blown, there is no indication readily available to indicate that the IRM is inoperable. The licensee has enlisted the services of General Electric to investigate this problem and potential solutions for resolving the problem.
Pertinent information relative to this investigation is as follows:
(1) LaSalle has a 15% Average Power Range Monitor trip which is totally separate from the IRM circuitry.
(2) LaSalle uses 1.5 ampere IRM chassis fuses and four 6 ampere Source Range Monitor (SRM) and IRM panel fuses. All SRM and IRM channels on the bus would be inoperative due to fuse failures, which would be evident by indication in the control room.
(3) Two bounding design basis events were analyzed, the control rod drop accident and the continuous control rod withdrawal transient in the startup power range. The results from both analyses were found to be bounded by the 15% APRM trip.
Actions taken by the licensee are:
(1) A warning label has been placed on each SRM/IRM chassis to help the operator in diagnosing a problem.
(2) A long term hardware modification is to take place. Two SRM/IRM modifications are being evaluated and will be submitted to the licensee for review after a safety analysis has been performed.
Both modifications would initiate an instrument inoperative trip in the event of a failure of the negative power supply.
c.
(Closed) IE Bulletin 86-01. Region III, via C. Norelius memo dated June 21, 1986, requested the inspectors to evaluate the licensee's response to IE Bulletin 86-01.
IE Bulletin No. 86-01 describes a recently identified problem with minimum flow logic for which a single failure under certain accident sequences could result in all Residual Heat Removal (RHR) minimum flow bypass valves being signaled to close while all other pump discharge valves are also closed resulting in no flow through the RHR pumps. This could lead to the pumps running with no flow which could cause pump damage and result in loss of RHR functions.
The Residual Heat Removal (RHR) System at LaSalle County Station is not vulnerable to the minimum flow logic problems described in IE Bulletin 86-01 since:
(1) Each of the three loops in the RHR system (A, B, and C) has its own flow bypass valve (designated F064A, F064B, and F064C) and flow indicator switch which is independent of the other loops.
As a result, any failure of a single flow indicator switch, minimum flow bypass valve, or associated relays affects only a single loop.
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(2) Each flow indicator switch generates a signal independent of the other loop's flow indicator switches. Therefore, an erroneous signal from a single flow indicator switch affects only a single loop.
(3) Operability of the RHR system in the event of a loss of power is maintained since loop A is supplied by a different bus than Loops B and C.
Within each loop, the RHR pump and minimum flow bypass valve are on the same bus. Therefore, a loss of power to a single bus would eliminate either loop A or loops B and C.
In this case, the pump (s)
in the affected loop (s) could not start and the minimum flow bypass valve (s) could not open.
The inspector reviewed the licensee's response dated May 30, 1986 and found the content to satisfy inspection procedure 92703.
d.
The inspector, per a regional request, inspected the licensee's Standby Liquid Control (SBLC) System for properly sized fuses being installed in the SBLC control circuitry.
A problem with fuses in the SBLC system control circuits was identified at another Boiling Water Reactor (BWR) within the region.
The inspector reviewed IE Circular 77-09, " Improper Fuse Coordination in BWR Standby Liquid Control System Control Circuits", SBLC system drawings, and actual installation of fuses in the control circuit to ensure that the fuses were sized properly.
LaSalle has the correct fuse coordination installed in their SBLC system control circuits.
7.
TMI Action Plan Requirement Followup (25565)
Closed (0 pen Items 373/81000-148 and 374/81000-68): TMI Item II.K.3.25A, II.K.3.25.B.
Effect of loss of alternating current power on pump seals.
Per Safety Evaluation Report (SER), Supplement 2, Item II.K.3.(25), the licensee was participating in the BWR owners group study. Data from tests on Bingham pumps with seal designs similar to those of LaSalle were provided.
The conditions were representative of BWR recirculation pump application and, therefore, the results are applicable to in plant BWR pumps. Observed-leakages were less than five gallons per minute for more than five hours.
This is acceptable.
8.
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, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications.
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373/86016-00 - A fuel bundle was loaded into the core without an operable Source Range Monitor (SRM) in the core quadrant into which the fuel bundle was loaded.
373/86015-00 - With Unit 1 in refuel (0% power) a full reactor scram was received during performance of LIS-NR-303. Cause was a mechanical failure of scram contractor 1C71-K14A (G.E. Type CR 105).
373/86018-00 - The "A" control room HVAC system ammonia detector tripped and an Engineered Safety Feature (ESF) damper actuation occurred.
The cause was the binding of the tape cassette spool mechanism in the detector unit.
373/86008-00 - During a standby liquid control system injection test, a pressure transient to a differential pressure indicator occurred causing a momentary false reactor low water level signal and subsequently an auto start of the high pressure core spray diesel generator.
373/86014-00 - The "A" control room HVAC system ammonia detector tripped causing an Engineered Safety Feature damper to actuate. The chemcassette unit in the tape carriage mechanism was found to be broken.
373/86012-00 - The "A" control room HVAC system emergency makeup train was actuated due to a spurious trip of the hi radiation monitor. The source calibration for the radiation monitoring module was not performed by the radiation chemistry department subsequent to the instrument maintenance departments electronic calibration the previous day.
The monitor was therefore more sensitive and tripped due to background radiation.
374/86010-00 - Missed Rod Worth Minimizer (RWM) and Rod Sequence Control System (RSCS) surveillance during unit shutdown. Procedure LGP 3-1 revised to correct the problem.
373/86020-00 - Primary containment isolation.
Personnel error while installing a jumper per LTS 300-4.
374/86009-00 - MSIV isolation and scram on spurious high steam flow signal at 15% power.
Flushing of sensing lines and channel monitoring failed to identify a cause.
374/86008-00 - Reactor scram on low level due to person accidentally
bumping the feedwater control system power supply breaker.
373/86019-00 - Cracked seal welds on threaded plugs in "B" and "C" Low Pressure Coolant Injection (LPCI) lines found during hydrostatic testing.
The welds were repaired during current outage on all three LPCI systems
by welding a pipe cap over the plug area.
9.
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 Paragraph 7.
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10.
ExitInterview(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 content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.
The licensee did not identify any such documents or processes as proprietary.
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