IR 05000373/1987022
| ML20235U288 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/03/1987 |
| From: | Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20235U276 | List: |
| References | |
| 50-373-87-22, 50-374-87-22, IEB-86-002, IEB-86-2, IEIN-87-008, IEIN-87-8, NUDOCS 8710140018 | |
| Download: ML20235U288 (17) | |
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i U. S. NUCLEARLREGULATORY-COMMISSION
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REGION III
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Report Nos. 50-37.3/87022(DRP); 50-374/87022(DRP)
. Docket Nos.. 50-373; 50-374 Licenses No. NPF-11; NPF-18 Licensee:
Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690
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Facility Name:
LaSalle County Station, Units 1 and 2 Inspection At:
LaSalle Site, Marseilles, IL Inspection Conducted: July 28 through September 21, 1987
i Inspectors:
M. J. Jordan R. Kopriva
R. Lanksbury i
M. Kopp
G. Pirtle I
l Approved By:
M. A. Ring, Chief
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M/5/57 l
Reactor Projects Section IC
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Inspection Summary Inspection on July 28 through September 21, 1987 (Reports No.
50-373/87022(DRP); 50-374/87022(DRP))
Areas Inspected:
Routine, unannounced inspection conducted by resident L
and regional inspectors of lice'nsee actions on previous inspection findings; l
operational safety; surveillance; maintenance; Licensee Event Reports;
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i emergency preparedness; regional requests; outages; allegations; unit trips; I
and temporary instructions.
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Results: Of the eleven areas inspected, no violations or deviations were identified in ten areas; one violation was identified in the remaining area; however, in accordance with 10 CFR 2, Appendix C, Section V. A., a Notice of Violation was not issued (failure to follow the procedure.resulting' in a scram at no power - Paragraph 11.b)._ The violation was of minor safety significance i
and did not affect the public's health and safety.
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The licensee was successful in returning. Unit 1 to power after a 107 day
maintenance outage.
Unit 2 operated throughout the assessment period
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l successfully. The licensee needs.to. continue to' strengthen the communication techniques between' maintenance personnel and operations to prevent. unnecessary actuations of safety equipment.
8710140018 8710 ( 3 PDR ADCCK 05000w Q
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DETAILS l
1.
Persons Contacted
- G. J. Diederich, Manager, LaSalle. Station
- R. D. Bishop, Services Superintendent
- J. C. Renwick, Production Superintendent-D. Berkman, Assistant Superintendent, Work Planning
- Wc Huntington, Assistant Superintendent, Operations
- P. Manning, Assistant Superintendent, Technical Services
- T. Hammerich, Assistant Technical Staff Supervisor
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W. Sheldon, Assistant Superintendent, Maintenance J. Atchley,.0perating Engineer
- D. A. Brown, Quality Assurance Supervisor D. Enright,-Quality Assurance Engineer
- M. Richter, Assistant Technical Staff Supervisor
- Denotes personnel attending the exit interview on September 22, 1987.
Additional licensee' technical and administrative personnel were contacted by the inspectors during the course of the inspection.
2.
Licensee Action on Previous Inspection Findings-(92701)-
(Closed) Violation (373/87018-01(DRP)):
In' October 1984, a 10 CFR 50.59 review was not performed or documented on a modification to the drywell ventilation (VP) system in which drywell dampers were wired open. The station agreed that a 10 CFR 50.59 safety evaluation should have been.
performed for the temporary changes made to.the primary containment ventilation system (VP).
Following identification of this concern, a 10 CFR 50.59 safety evaluation was performed. The evaluation concluded that the changes did not' involve an unreviewed safety question.
The station modified the Unit 1 and 2 VP. systems to rep 1 ace the VP fan discharge dampers and actuators with gravity actuated backdrift dampers.
The work has been completed on Unit 2 and is partially. complete on Unit 1.
The station temporary system change procedure (LAP-240-6) will be reviewed and revised to require use of the temporary system change procedure when valves or dampers'are rendered immovable such that.they cannot attain their design failure position.
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l (Closed) Violation (374/87012-01(DRP)): On March 18, 1987, procedure
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LOP-FC-03, " Fuel Pool Cooling System-Startup, Operation, and Level Changes of the Fuel Pool Skimmer Surge Tank," was not reviewed or adhered to by personnel performing.the evolution resulting in manipulation of a
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wrong valve which caused a loss of approximately 100 gallons of contaminated water into the Unit 2 reactor building and reactor building ventilation.
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The following steps were taken to enhance the performance of equipment operators:
a.
The labeling of fuel pool valves was reviewed to ensure the wording is clear as to the valve's purpose.
b.
This example of the differences between Unit I and Unit 2 has been-added to the equipment operator training program.
c.
The event was reviewed with all equipment operators, emphasizing the importance of:
(1) Ensuring only proper equipment is operated.
(2) Seeking help if any doubt exists.
d.
The event has been discussed with all nuclear station operators, emphasizing the importance of ensuring that the non-licensed operators, are knowledgeable of their assignments.
The inspectors reviewed these actions and consider this item closed.
(Closed) Open Item (373/87011-02(DRP)):
Failure to remotely operate the inboard shutdown coc, ling isolation valve (1E12-F009) from the control room.
Normal shutdown cooling mode requires the inboard and outboard isolation valves to be open establishing a path for cooling flow when the reactor is below 212 degrees F.
On numerous occasions, the licensee was unsuccessful in remotely operating the "9" valve. The licensee would then enter the drywell and manually stroke the valve.
Analysis of valve failure rates with heat-up and cool-down rates proved inconclusive as to the cause.
Stress analysis which involved the installation of strain gages, thermocouple, and linear voltage
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differential transformers (LVDis) were used to obtain precise displace-
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ment measurements and revealed no excessive stresses which would cause the disc in the valve to bind and prevent its operation.
Subsequently, the licensee contacted plant; of similar design which use the same valves. As a result, the licensee installed a larger stem with a larger operator motor.
In addition, the valve seats were lapped and the disc re-machined.
It is believed these actions will prevent future valve failure.
(Closed) Unresolved Item (373/87019-01); (0 pen) Unresolved Item
(374/87019-01):
Emergency Core Cooling System (ECCS) motor lead splices.
An inspection of ECCS motor lead splices was conducted at the LaSalle Nuclear Generating Station on July 29, 1987, as a result of Environmental
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Qualification (EQ) concerns regarding Kerite tapes, Okonite tapes and
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bolted configurations used to splice motor leads.
The inspector reviewed the following:
a.
Environment The inspector reviewed the LaSalle FSAR Chapter 3, Table 3.11-8 and EQ zone maps to identify the environmental parameters in which the
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motors were operating. The inspector confirmed by reviewing the above referenced documents that the motors'were operating in a
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s radiation harsh only environment. 'No concerns were identified.
I b,
. Maintenance / Surveillance
.The inspector reviewed maintenance / surveillance-records and interviewed operating and engineering personnel to determine j
operational. history of the ECCS motors.
No concerns were d
identified, c.
Installation The inspector reviewed installation drawings, procedures and termination cards used to make the motor lead splices.
It appears that, as originally installed, the contractor craft and quality control personnel installed and inspected the motor lead splices as required by the design.
No concerns were identified.
d.
Walkdown The inspector performed a visual inspection of the motor lead sp.1ce connection for Unit 1 RHR motor IE72-C002A'
No concerns
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were identified.
e.
Corrective Actions (1) The licensee has committed to replacing the motor lead splices with qualified materials and configurations. Unit I splices were replaced prior to startup and Unit 2 splices will be replaced during an outage of sufficient duration to perform the work.
In addition, the licensee has submitted several documents regarding qualification of splices.as installed and qualifications of splices to be' installed.
(2) The resident inspectors observed the electrical determination and determinations on the Unit 1 A, B,-and C Residual Heat Removal (RHR) Pumps, Low Pressure Core; Spray (LPCS) Pumps, and High Pressure Core Spray (HPCS). Pump with the correct Environmental Qualification (EQ) kits. The inspectors also verified that all discrepancies identified.while performing the terminations were corrected. This' item is considered closed for Unit 1 and will remain open for Unit 2 until completion of the replacement of the Kerite kits.
(Closed) Open Item (374/85009-01):
Item pertained to concern the NRC had on gap between the storage battery cells and the seismic supports. The excessive gaps between the battery cells and seismic supports have been corrected per modification M-1-2-84-174 and M-1-2-84-175 done under work request no. L43715. -The resident inspectors verified that the modifi-cations were completed satisfactorily.
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(Closed) Open Item (373/87011-03):
Item concerned failure of Reactor Core Isolation Cooling (RCiC) water leg pump. The investigation revealed that the retainer / snap ring for the thrust bearing had become displaced.
This caused or allowed the impeller to come in contact with the pump housing. The newly installed RCIC water leg pump has an outboard bearing housing that does not require a retainer / snap ring.
The resident inspector has reviewed the licensee's actions and finds them satisfactory.
(Closed)OpenItem(373/87011-05): A review of modification package 1-1-84-003 revealed that during the review of procedures affected by this modification, surveillance procedure LIS-RP-08 was not included.
The licensee has revised LIS-RP-08.
The licensee also did an additional review of affected procedures and found two additional procedures which needed revisions.
They were LIS-PC-15 and LOP-RI-01M. The revisions to these procedures are complete..The resident inspector has reviewed the affected procedures and finds the licensee's actions adequate.
(Closed) Open Item (373/87006-05):
Item concerned the task of verifying the wiring in the limitorque motors in accordance with IE Notice 87-08. -
All of Unit 2's limitorque motor lead wiring was inspected by May 9,1987, and was found acceptable. All of the limitorque motor lead wiring for Unit I was complete by July 31, 1987, and found acceptable. The resident inspectors have reviewed the licensee's actions and finds them adequate.
No violations or deviations were identified in this area.
3.
Operational Safety Verification (71707)
a.
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 Unit I and 2 reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, 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.
During the month of August 1987, the inspector walked down the i
accessible portions of the following systems to verify operability:
Unit 1 - A, 8, and C RHR System, HPCS, and LPCS Unit 2 - 2A, 28 Diesel Generator, Standby Liquid Control System
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b.
On July 23, 1987, at approximately 3:00 a.m. (CDT) with Unit 2 at
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90% power, the Unit 2 operators noticed'an increase in the off gas j
post treatment levels.
Initial efforts were started to look for evidence of off gas charcoal bed moisture or bypass line valve
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leakage..No problems were found.
By 8:00 a.m. on July 24,.1987, there had been a steady increase in the off gas rost treatment levels.
The licensee's nucir w engineering group had reviewed l
reactor power chart recorder >, fuel preconditioning printouts, and all control room typer printouts for control rod drive (CRD)
movements.
No CRD movements had taken place. The reactor power chart recorders revealed the off gas increase was approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a minor power ramp.
This observation of off gas i
increase 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a power ramp was similar to the Unit 2 l
cycle 1 observations where a small fuel leak was found. The nuclear
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engineers continued checking the computer printouts for pussible errors in the calculations of reactor power. On July 24,:1987, the reactor vendor was asked by the licensee to run a-check on the licensee's computer printouts to verify that the proper reactor power calculations were being made. The vendor reported that the licensee's programs were satisfactory. After a full' review of the data, the licensee's fuel department believed there was a small leak in one of the new fuel bundles due to low values of the initial off gases.
By July 25, 1987, the off gas levels had peaked and by Jaly 27, 1987, they had decreased and leveled off.
The licensee's chemistry department sampled the off gas for activity and found h to be well within their administrative and Technical Specificati e limits.
Presently, Unit 2 is at 100*4 power and the nuclear enguvering and chemistry departments are monitoring the off gas post treatment levels on a routine basis. The resident inspectors will continue to monitor the off gas post treatment levels throughout this fuel cycle for Unit 2.
c.
On August 13, 1987, the Unit I control room ventilation system switched to charcoal filters due to a signal from the "B": ammonia i
detector.
The cause of the signal was a broken tape in the ammonia i
detector. The licensee replaced the tape and no further action was warranted.
d.
On August 29, 1987, at approximately 6:30 a.m. (CDT), the Unit 1 outboard isolation valve for the Reactor Water Cleanup System (RT)
closed.
The isolation valve (1G33-F004) closed due to a signal of high temperature in the cleanup flow. An operator was removing a
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jumper which prevented this isolation on high temperature flow when l
the jumper was grounded, a fuse was blown, and the isolation occurred.
The jumper had been installed for the hydrostatic test of the reactor vessel following replacement of the reac'
' vessel head during the ongoing maintenance outage. This isolation signal was not identified in the Technical Specifications as an Engineered l
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Safety Feature (ESF) isolation, however, the outboard valve was identified as a primary containment isolation valve.
The system l
functioned as expected.
High temperature in the reactor water cleanup system did not exist.
The licensee identified the cause of the fuse blowing was unintentional grounding of the jumper.
Even j
though the loose end was taped and the Equipment Operator believes
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it was taped securely, it must not have been adequate. The l,umper
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was placed ia a very hard-to-get-to place in the terminal oox, above the operators head in a very congested area of pipes, instrument lines, and bracing. The operator stood on a garbage can in order, to reach the terminal bnx because it would have been difficult to get a ladder into the area.
The root cause of the AUTO valve closure was a procedural deficiency which directed a jumper be installed in a difficult location with the resultant effect being a blown fuse, even when appropriate i
actions were taken by the worker. The licensee has taken the l
following corrective action.
(1) The blown fuse was replaced and the valve reopened.
(2) A work request was written to install banana plugs in the panel in which the jumper was located l(and in the 1(2)H13-P623 panel which is a more accessible location).
This will be accomplished before the next LOP-NB-01 is performed on Unit 1 or 2.
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(3) A new type jumper with banana plugs is on order through the Instrument Maintenance Department.
This jumoor has:an-insulated boot that covers the male plug whor, disconnected.
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The current jumpers will be replaced when the new ones arrive, l
and this should lessen the chance.of accidental grounding.-
(4) Operations, Electrical and Instrument Maintenance Department personnel will be trained to not4fy their supervisors if a l
jumper or lifted lead needs to be installed in a difficult to J
reach panel.
If necessary and allowable by the schematic and l
wiring diagrams, the procedure or checklist will be changed to install the jumper / lifted lead in an easier. to reach place, In this training, the importance of insurinp that only the I
i logic that was bypassed in the original procedure or checklist I
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is bypassed will be stressed. This training will be
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documented and be completed by Ocf.ober 15, 1987.
l e.
On September 5, 1987, at approximately 9:00 a.m. (CDT) with Unit I in Cold Shutdown., the Unit 1 "B" Reactor Protection System (RPS) Motor
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Generator (MG) set tripped causing a loss of the "1B" RPS. This
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caused actuations of the Primary Containment Isolation System (PCIS)
Groups 11 through VII on Unit I and an isolation of the Unit 2 reactor building ventilation system with an auto start of the Unit 2 Standby Gas Treatment (SBGT) system. The licensee. verified tnat all the isnlations were completed.
The licensee received reports of smoke in the area of the "1B" RPS MG set. No abnormalities vere noted in the RPS distribution panel. The Unit "1A" RPS was on alternate power and was changed back to its own MG set (normal power).
In preparation for changir.g the '1A" EPS back to normal, the reactor mode switch was moved from refuel to shutdown which caused a planned reactor scram. The licensee transferred the "1A"< RPS to norinal power and the "1B" RPS to alternate power without problems or
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incidence. The reactor scram on Unit I and the PCIS logic on Units 1 and 2 were reset.. The Unit 2 reactor building ventilation was then restarted.
The' Unit 1 Residual Heat Removal (RHR) system shutdown cooling mod, which had isolated, was. restarted'. No increase in reactor water temperature was noted. Upon further investigation, the licensee discovered 1K relay for the. motor'on the
"1B" RPS motor generstor had burned up. The relay.was replaced.
satisfactorily. Noifurther action was warranted.
,o f.
The inspector reviewed the, licensee's actions when de lake blow down recorder became inoperable.
The licensee was meeting action statement 102 for Technical Specification 3.3.7.10 by 2:eading the blow down valve indicated position at the valve and by,use of a-
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curve of valve position vs blowdown flow, which'was developed during initial startup testing, to determine the blowdown flow. When the inspector re_ viewed this. action, he was unable to identify.the.
calibration ' procedure for the rembte indication for the valve which the licensee way using to meet the/actionistatement ot the Technical
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Specifications.1 The inspector identified to the licensee personnel that they werd using an instrument to meet a safety requirement which did not Save a calibration frequency on it to assure it remained acc gate.
10 CFR 50, Appendix B, Criterion'XII states,
" Measures shall be established to assure,that tools, gages,:instru-
ments, and other measurint and tut, trig devices used in activities
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affecting quality are properly coi.tyolled, calibrated, and adjusted
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at specified periods to maintain accuracy withinLnecessary limits."
The licensee was unable to provide th calibrati_on procedure for the indicated valve position, however,ta. review of the.the data sheets t
for procedure LDS-WL-Q1, " Lake Blowdown Flow ' Indicator Channel Functional Test," identified that when the cperators performed this procedure, which checked the coatrol room indicated valve position vs, flow, the operators also voluntarily recorded.the remote alve indicated positton at the s ilve. Using this data, a calibration of the valve indicated positiod*vs flow could be determined. The licensee has initiated a change to LOS-WL-Q1 to rsquire the recording of the local valve indicated position so it can be assured s
that local valve position can' be verified accurate,with some regularity.
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The inspector was satisfied with the ikensee's actio'n'.
The licensee was able to provide the inspector with sufficient data to j'
assure the indicated valve position had not change *over time. The licensee is also in the process of changing procedure LOS-WL-Q1 to provideforacheckofvalvepositionviiflov[o'assurethevalve position indicator will noj change.
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No violations or devIetions were identified in this area.
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4.
Monthly Surveillance Observation (61726)
The inspector observed Technical Specific tion required surveillance testing and verified for actual activities observed that testing was.
l 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, l
that test results conformed with Technical Specification and procedure I
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-i personnel.
The' inspector witnessed portions of the following test activities:
.LIS-NR-405 Unit 2 Rod Block Monitor. Functional Test LIS-NR-203 Unit' 2 Average Power Range Monitor Rod Block and Scram Calibration LIS-NR-209 Unit 2 Average Power Range Monitor Gain Adjustment LIS-HP-309 High Pressure Core Spray Water Leg Line Pressure Functional Test
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No violations or deviations were identified in this area.
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5.
Monthly Maintenance Observation (62703)
During the inspection period, the inspector observed portions of the following maintenance activities:
Unit 1 Work Request L71003 Determination of Lugs on A-RHR Pump Unit 1 Work Request L70616 Installation of Correct EQ Kits on Terminal Leads for A-RHR Pump
Unit Work Request L70847 Installation of Correct EQ Kits on Terminal i
Leads for the Low Pressure Core Spray
(LPCS) Motor Unit Work Request L70848 Installation of Correct EQ Kits.on Terminal Leads for the High Pressure Core Spray
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(HPCS) Motor Unit Work Request L70617 Installation of Correct EQ Kits on Terminal Leads for the C Residual Heat Removal (RHR)
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Motor Unit Work Request L70615 Installation of Correct EQ Kits on Terminal Leads for the B Residual Heat Removal
(RHR) Motor l
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No violations or deviations were identified in this area.
j 6.
Training (41400)
The inspector noted the mechanics and Quality Control inspectors who l
installed the new Environmentally Qualified kit on the RHR pumps appeared
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to be well trained on the proper termination instructions. When i
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questioned on the instructions provided by the vendor, they were knowledgeable of the proper taping method, thickness, and type of tape to be used for each layer and appeared to know the function of each layer of tape and why some layers had to be thicker than others.
They-appeared to be well briefed and trained.
No violations or deviations were identified.
7.
Licensee Event Reports (92700)
Through direct observations, discussions with licensee personnel, and review of records, the following Licensee Event Reports (LERs) were I
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 Specifications.
(Closed) 374/87016-00 - Defective Low Pressure Core Spray (LPCS) minimum -
flow switch.
Low pressure Core Spray minimum flow switch, FS-2E21-N004, was found to be leaking water through the switch internals.during the -
performance of Instrument Surveillance LIS-LP-202 on July 13, 1987.. The leakage was very small and did not affect the operation of the switch as-l demonstrated during the calibration.
A Work Request was written to replace the defective switch with another one, like-for-like. The replacement switch was installed, calibrated and placed in service on July 15,1987.
This event was reported to the NRC as a voluntary LER in accordance with the requirements of IE Bulletin 86-02, " Static-0-Ring Differential Pressure Switches."
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(Closed) 374/87015-00 - Reactor water clean-up suction isolation valve
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closure on hi filter / demineralized inlet temperature due to failed
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temperature switch. The cause of this event was a_ failed capacitor in-
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the temperature switch (TIS-2G33-N008).
The safety consequences'of this q
event were minimal since the-outboard isolation valve of the RWCU system closed as designed. Work Request L70132 was initiated to repair the failed temperature switch. The switch was repaired by July 6, 1987, and j
no further problems have been experienced.
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(Closed) 373/87026-00 - Residual heat removal pump motor termination found to be non-environmentally qualified during surveillance testing.
The identification of improper environmentally qualified terminations was documented in Inspection Report 373/87019; 374/87019 and the followup inspection was documented in this inspection report in Paragraph 2 under Unresolved Item 373/87019-01; 374/87019-01.
(Closed) 373/87027-00 - Emergency core cooling pump system terminations did not ~ meet environmental qualification requirements. The use of.
improper termination kits for ECCS pumps'was documented.in Inspection
Report 373/87019; 374/87019 and a followup inspection by a regional based j
inspector was' conducted in July and documented in this inspection report
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in Paragraph 2 under Unresolved Item 373/87019-01; 374/87019-01.
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The following steps were taken to enhance the performance of equipment operators:
l a.
The labeling of fuel pool valves was reviewed to ensure the wording is clear as to the valve's purpose.
b.
This example of the differences between Unit I and Unit 2 has been added to the equipment operator training program, c.
The event was reviewed with all equipment operators, emphasizing the importance of:
(1) Ensuring only proper equipment is operated.
(2) Seeking help if any doubt exists.
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d.
The event has been discussed with all nuclear station operators, emphasizing the importance of ensuring that the non-licensed operators are knowledgeable of their assignments.
The inspectors reviewed these actions and consider this item closed.
(Closed) Open Item (373/87011-02(DRP)):
Failure to remotely operate the inboard shutdown cooling isolation valve (1E12-F009) from the control I
room.
Normal shutdown cooling mode requires the inboard and outboard
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isolation valves to be open establishing a path for cooling flow when the reactor is below 212 degrees F.
On numerous occasions, the licensee was unsuccessful in remotely operating the "9" valve. The licensee would then enter the drywell and manually stroke the valve.
l Analysis of valve failure rates with heat-up and cool-down rates proved l
inconclusive as to the cause.
Stress analysis which involved the l
l installation of strain gages, thermocouple, and linear voltage differential transformers (LVDTs) were used to obtain precise displace-ment measurements and revealed no excessive stresses which would cause the disc in the valve to bind and prevent its operation.
Subsequently, the licensee contacted plants of similar design which use the same valves. As a result, the licensee installed a larger stem with a larger operator motor.
In addition, the valve seats were lapped and the disc l
re-machined.
It is believed these actions will prevent future valve
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failure.
(Closed) Unresolved Item (373/87019-01); (0 pen) Unresolved Item (374/87019-01):
Emergency Core Cooling System (ECCS) motor lead splices.
An inspection of ECCS motor lead splices was conducted at the LaSalle
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Nuclear Generating Station on July 29, 1987, as a result of Environmental
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Qualification (EQ) concerns regarding Kerite tapes, Okonite tapes and bolted configurations used to splice motor leads. The inspector reviewed the following:
d.
Environment The inspector reviewed the LaSalle FSAR Chapter 3, Table 3.11-8 and EQ zone maps to identify the environmental parameters in which the
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9.
Regional Requests (92701)
On June 26, 1987, a request for inspection pertaining to high ambient temperatures in areas with electrical switchgear was issued by a
memorandum from C. E. Norelius to the Region III resident inspector i
staffs.
During a routine inspection at one of the sites in Region III,
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an. inspector noted that ambient temperatures in. areas containing electrical equipment and instrumentation were excessive. This was due primarily to the prevailing hot weather conditions. The memorandum.
requested the resident inspectors to verify that the licensee has in place requirements to monitor temperatures in such areas based'on equipment performance capabilities and has established appropriate limits on these temperatures.
During the later part of July and early August of 1987_, the inspectors, through physical walkdowns, Technical Specification review,'and procedural review, inspected the areas of the plant in'which sensitive electrical equipment and instrumentation is located. The physical inspection, Technical Specification review, and procedural review proved acceptable. Also, at this same time the site Quality Assurance group was performing an annual audit on the same subject. Their' audit also contained an equipment list, serial numbers.of equipment / instruments, calibration dates and frequencies, and Technical Specification limits, if any. The inspector also reviewed the results of the QA audit. There were no anomalies found during the QA audit. This regional request is consideredclosed(373/87022-01;374/87022-01).
No violations or deviations were identified.
10. Outages (71707)
On September 13, 1987, Unit I commenced a reactor startup after a 107 day maintenance outage.
Some of the major maintenance work activities included replacement of the B Reactor Recirculation (RR) pump, repair of the A-RR pump discharge valve, replacement of the A-RR pump seals, and replacement of the motor on the inboard isolation valve for the shutdown' -
cooling system. There were several tests which needed to be accomplished, and the unit was connected to the grid on September 14, 1987.
No violations or deviations were identified.
11. Allegation (99024)
The security inspector reviewed the below allegation during an onsite security inspection between March 2-6, 1987, at the LaSalle County Nuclear Plant.
a.
Background:
NRC Region III was advised on February 26, 1987, that, among other allegations, a contractor believed he and his co-workers were " set-up" by persons unknown. One of the reasons for his belief was that at 3:00 p.m., on February 22, 1987, the door egress button on either Door 405 or Door 413 (specific door not identified) was.
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l taped closed. The egress button for-both doors inactivates the door alarm for a preset period of time to allow personnel.tc leave the area without causing an alarm.
The preset time varies from door to i
door depending on the type of personnel traffic. The contractor believed taping the door egress button would inactivate the alarm for the entire time the button was taped in the closed position.
Door 405 allows access to the Unit 1 Personnel Access Air Locks and Door 413 allows access to the Unit 2 Personnel Access Air Locks.
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NRC Review: During the onsite inspection between March 2-6,-1987,
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the inspector confirmed that Doors 405 and 413 are not considered security doors, except when they are not controlled.as high-l radiation areas.
Both doors were.under Health Physics
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administrative controls as high radiation doors on. February.22, 1987. The inspector also confirmed that the doors are monitored by
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I the security computer system, but alarms from.the doors are not responded to by members of the security force when they are high radiation doors.
If an' alarm from either door is received and the alarm does not reset within ten minutes, the security force would advise the Health Physics Department, who would have someone respond.
Interview results also disclosed that the egress buttons.for all doors monitored by the system (including Doors 405 and 413) will-alarm after the preset time even if the egress button is activated.
The inspectors tested three doors with alarms in the same configura-tion as Doors 405 and 413 and confirmed that an alarm 'will be generated after the preset time even with the egress button depressed (taped closed).
c.
Conclusion:
The egress button will function as designed even with the button activated (taped closed). Therefore, no violations or regulatory concerns were noted. The resident inspector also verified that the alarm would occur even if the button was still depressed. This allegation will remain open pending review and followup of one additional concern by the NRC. (373/87022-02; 374/87022-02)
No violations or deviations were identified in this area.
12. Unit Trips (93702)
a.
On September 2, 1987, at 8:14 a.m. (CDT), a scram occurred on Unit i from low Control Rod Drive (CRD) charging water header pressure.
The licensee was performing a time delay. functional test'of the scram signal on one' channel which caused a half scram. The Electrical Maintenance-(EM) worker failed to verify the reset of the scram signal before testing the other channel which caused a full I
scram to occur. The unit had been in a Cold Shutdown condition since June 1, 1987, so no control rod movement occurred.
The licensee followup identified that the EM worker performing the test requested the Reactor Operator (RO) to reset'the half scram
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after each channel was tested. He initialed each step appropriately identified in the LES-RP-109 requiring verification of the half scram being reset. However, his communications with the R0 were not positive, in that he did not use good repeat-back type communications.
He accepted a response from the RO that he believed was an acknow-ledgement that the half scram was reset after each channel was tested.
The R0 was not positive in his response to the EM worker.
Repeat-back type communications were not used when the B channel was tested.
The R0 did not recall the EM worker requesting a reset of the half scram when the B channel was tested. However, he thought he
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acknowledged a question from someone, since he was performing
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various activities, and the EM worker thought the acknowledgement was to his request that the half scram was reset.
l The licensee took prompt action to prevent recurrence by:
(1) Upper management discussion with the individuals involved on expectations with communications and repeat back policy.
(2) Department heads tailgated (department discussion) this incident with their personnel.
j (3) The shift engineer and supervisor on shift were directed to monitor activities for proper communications.
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The inspector noticed an increase in the licensee communication i
skills during the restart of the unit on September 18, 1987.
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Particularly, while rolling the turbine and putting the unit on line, the repeat-back communications were noted by the resident inspector to be good.
l The failure to adhere to procedure LES-RP-109 and adequately
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verify the resetting of the half scram before testing the next J
l channel is a violation of Technical Specification 6.2.A which
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requires adherence to written procedures including surveillance and testing procedures.
(373/87022-03)
Since this scram occurred at zero power, this event was not considered self identifying and it was considered licensee identified.
This violation thus meets the test of 10 CFR 2 Appendix C, Section V.A.. Consequently, no Notice of Violation
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will be issued and this matter is considered closed.
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b.
On September 17, 1987, at 3:22 a.m. (CDT), Unit 1 scrammed from approximately 5% power due to a low reactor water level signal.
The licensee was shutting down the reactor the evening of September 16, 1987, as a result of finding four of five fast acting solenoids on the bypass valves not operable. This condition was found before entering the operating condition for which the bypass valves were required to be operable. The licensee was attempting to control the water level using a manual discharge valve I
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on the Motor Driven Feed Water Pump (MDFWP) because the automatic Feed Regulating Valve (FRV) was fully closed but was leaking. The water level started to rise. The manual valve was not closed fast enough to prevent a high level and tripping of the MDFWP. The motor driven feedwater pump was restarted and the manual valve'was closed. Water level started down and the manual valve'was not opened sufficiently fast enough to prevent the scram on low level.
All systems functioned as expected.
The operator hed difficulty in opening the manual valve due to the differential'tressure across-the valve after the MDFWP was restarted.
Also, due to a conservative set point on the Static-0-Ring level sensors, the scram occurred at approximately 19" of reactor vessel level in lieu of the 12.5" Technical Specification limits.
Further investigation by.the licensee for.the cause of the four failures to the fast acting solenoids determined that the hydraulic spools in the solenoids were sticking. The licensee discussed the sticking hydraulic spools with the General Electric representative.
Testing of the electrical system to assure adequate voltage was conducted which determined that the lowest voltage the solenoids would see would be 109 volts.
GE states the solenoids will operate at any voltage above 103 volts.
The licensee thought the sticky spools could be due to contaminated.-
Electric Hydraulic Control (EHC)' oil. The oil for the fast acting solenoids does not get much flow (only once a week during testing)
and thus the oil is not cleaned regularly.
The licensee has taken action to flush and clean the EHC oil and has particularly flushed and cleaned the EHC system associated with the fast acting solenoids.
The licensee has also increased the weekly surveillance of the bypass solenoids to a daily test to assure their operation and all five solenoids were replaced.
The daily testing will be changed back to weekly once the licensee is assured the solenoids will not stick again and one week of daily testing is completed satisfactorily.
To address the low level scram, the licensee will change the procedure for shutting down the reactor to require that level control be transferred to operation with the manual. valve by two operating personnel before transferring the mode switch from "Run" to " Shutdown." This action is expected to facilitate level control at low power levels. The licensee has also scheduled a modification on Unit 1 during the upcoming refueling outage to install a motor-on the manual valve to allow for controlling of the valve from the-control room. This modification was completed on Unit 2 during the last refueling outage and the control of feed water at low power has l
been very good on Unit 2.
No violations or deviations were identified in this area except as-discussed in Paragraph 12.a.
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13. Temporary Instructions (25579)
" Inspection of Emergency Operating Procedures" a.
Operator Training During the period November 5-7, 1986, an inspection was conducted of the licensee's upgraded Emergency Operating Procedures (E0Ps) to detennine if they had been prepared and validated in accordance with the NRC approved Procedures Generation Package (PGP). This inspection was documented in Inspection Report Nos. 50-373/86042 and 50-374/86042. During that inspection, the inspectors indicated that operator training on the new E0Ps would be covered in a future inspection. The purpose of this inspection was to evaluate operator training.
On April 21-23, 1987, operating examinations were given to license candidates at LaSalle County Station, and on June 2-11, 1987, requalification operating examinations were given to selected license holders. During the course of the simulator examinations, the examiners observed the implementation of the facilities' E0Ps to l
verify that operators were familiar with their responsibilities and l
required actions during emergencies, both individually and as a i
team; operators did not physically interfere with each other while performing the E0Ps; operators did not duplicate actions unless l
required by procedures; and where transitions from one E0P to
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another E0P or procedure were necessary, precautions were taken to ensure that all necessary steps, initial conditions, and l
prerequisites were met or completed and that the operators were l
knowicdgeable about where to enter and exit the other procedures.
b.
Inspection Results Operators did not physically interfere with each other while performing the E0Ps, and generally appeared to be knowledgeable in their performance; however, during the course of the April examinations, the examiners identified a weakness with the ability of some operators to utilize the E0Ps. This difficulty was mainly l
attributable to the format of the E0Ps (all procedures in a single l
8-1/2 x 11 inch binder) and the corresponding difficulty in reading
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them.
It appeared that the candidates preferred to work from memory rather than tackle having to actually use the E0Ps.
This issue was brought to the attention of the licensee, and operator aids in the form of flow charts to help guide the operator through the E0Ps were developed.
Simulator training was then conducted l
utilizing the new flow charts in conjunction with the E0Ps.
During the course of the June examinations, the examiners noted a significant improvement in the operators' ability to utilize the E0Ps and considered that the flow charts were a major contributor to the improvement; however, some of the operators tended to use the flow charts exclusively or to too large a degree. The new flow
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charts were still in the developmental stage and were designed to j
be operator aids.
In addition, the flow charts did not contain
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all of the information that was contained in the E0Ps themselves.
The licensee needed to stress in the licensed operator requalifica-tion training program that the flow charts were to be used as an j
aid in performing the E0Ps, but were not to be considered as a replacement since they did not include all of the steps and actions contained in the actual E0Ps.
This temporary instruction is considered closed (373/87022-4; 374/87022-3).
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l No violations or deviations were identified in this area.
I 14. Open Items Open items are matters which have been discussed with the licensee, which q
will be reviewed further by the inspector, and which involve some action j
on the part of the NRC or licensee or both.
Open items disclosed during j
the inspection are discussed in Paragraph 2.
15.
Violations For Which A " Notice of Violation" Will Not Be Issued
The NRC uses the Notice of Violation as a standard method for formalizing j
the existence of a violation of a legally. binding requirement. However, j
because the NRC wants to encourage and support licensee's initiatives i
for self-identification and correction of problems, the NRC will not l
generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V. A..
These tests are:
(1) the violation was identified by the licensee; (2) the violation would be categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including
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measures to prevent recurrence, within a reasonable time period; and (5)
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it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violation.
Violations of regulatory requirements identified during the inspection j
for which a Notice of Violation will not be issued are discussed in i
Paragraph 12.a..
16. 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 j
likely informational content of the inspection report with regard to
documents or processes reviewed by the inspector during the inspection.
j The licensee did not identify any such documents or processes as
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proprietary.
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