IR 05000373/1989021
| ML19332B891 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 11/13/1989 |
| From: | Lerch R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19332B889 | List: |
| References | |
| TASK-1.C.7, TASK-2.F.2, TASK-2.K.3.16, TASK-2.K.3.21, TASK-2.K.3.28, TASK-TM 50-373-89-21, 50-374-89-21, GL-84-23, NUDOCS 8911210280 | |
| Download: ML19332B891 (19) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
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Report Nos. 50-373/89021(DRP);50-374/89021(DRP)
Docket Nos. 50-373; 50-374 License Nos. NPF-11; NPF-18
Licensee: Commonwealth Edison Company Post Office Box 767
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Chicago, IL 60690 Tacility Name:
LaSalle County Station, Units 1 and 2 i
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Inspection At: LaSalle Site, Marseilles, Illinois Inspection Conducted: September 12 through October 23, 1989
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Inspectors:
R. Lanksbury R. Kopriva D. Jones K. Zullivan (Brookhewn)
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Approved By:
R. M. Leren, /ct%g Chief
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Reactor Projects $tction J3 Dati Mpe~:tibn Sumary Inspection fron, September 12 tl rough Octooer 23, 1989 (Report Nos.
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50 373/89021(UR77; 50474/89021(0R,Py l
Areas Inspected:
Routine, unannounced safety inspection by the resident Tnspectors of licensee action on previously identified items; operational safety; surveillance; maintenance; licensee event reports; ESF system walkdowns; training; security; refueling activities; evaluation of licensee self-assessment capability; TMI action plan requirement followup; emergency preparedness; evaluation of exercises for emergency preparedness; onsite
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followup of written reports of nonroutine events at power reactor facilities; report review; and outages.
Results: Of the sixteen areas inspected, no violations were identified and the licensee's performance generally demonstrated a conservative approach to safety issues.
The inspectors had two concerns during this report period. One is the control of contractors' qualifications and scaffolding.
See paragraphs 3b and 17.
The second concern deals with control of personnel access to in the Unit '
,drywell, which is open for the refuel outage work. See paragraph 9.b.
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DETAILS 1.
Persons Contacted Commonwealth Edison Company (Ceco)
40. Reed, Senior Vice President
+D. Galle, Vice President, BWR Operations
- +G. J. Diederich, Manager, LaSalle Station
- +W. R. Huntington, Technical Superintendent
- J. C. Renwick, Production Superintendent
~D. S. Berkman, Assistant Superintendent, Work Planning
.J. V. Schmeltz, Assistant Superintendent, Operations-r
- J. Walkington, Services Director
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- T. A. Hammerich, Regulatory Assurance Supervisor
+T. Kovach,. Nuclear Licensing Manager
+N. Kalvianakis, General Manager, BWRs
+K. Brennan,.BWR Regulatory Assurance Administrator U.- S. Nuclear Regulatory Commissien (USNRC)
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+0. J. Paperiello, Deputy Regional Adainistrator
+E. G. Greencian, Director, ORP
- +H. Miller, Dircctor, DRS
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+R. Ceoper,11, Chief, Engineering Branch, DRS
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P+R. terch, Acting Section Chief IB, DRP, RIII
+P. Shemanski, NRR/ Acting Project Director PD 3-2
+H. Li, NRR/JCSB
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+E. Weiss, Acting Chief, Operations Branch
^R. Gardaer, Chief,-PSS, DRS l
+W. Shafer, Chief, Branch 1, DRP
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- +R. Kopriva, Resident-Inspector, LaSalle R. Lanksbury, Senior Resident Inspector, LaSalle-i
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+B. Grim, Manager, Special Projects Illinois Department of Nuclear Safety (IDNS)
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- J. Roman, Resident Engineer
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+ Denotes personnel attending the meeting on October 6,1989, in the Regional Office pertaining to the August 26, 1989, Unit 2 scram.
- Denotes personnel attendin5 the exit interview on October 27, 1989.
Additional licensee technical and administrative personnel were contacted by the inspectors during the course of. the inspection.
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2.
Licensee Action on Previously Identified items (92701)
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(Closed) CAL (373/87035-03): Perform increased frequency testing of the Main Steam Isolation Valves (MSIV's) on Unit I as described in a LaSalle station letter. The licensee tested the valves
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quarterly-for the first six months of unit operation following replacement, and monthly thereafter.
In a letter dated May 4, 1989, to the licensee, the NRC approved a request to return the MSIV test frequency to quarterly, b.
NRC Region III management has reviewed the existing open items for the LaSalle Station and have determined that the open items listed in Attachment I will be closed administratively due to a lack of safety significance, age of the item, and other priority work. The licensee is reminded that commitments directly relating to these open items are the sole responsibility of the licensee and must be met as specified.
NRC Region III will review licensee actions by periodically reviewing a sample of administratively closed itemt.
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3.
Operational Saftty Verification (11707)
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The inspectors observed control rorm opmtions, reviewed applicable logs, and conducted discussient witn contiol room opa<ators during
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the inspection period. The inspectors 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, r,utiliary, and turbine buildings were corducted to
observe plant eg'Jipmtat cLnditions. These tours inclu:1ed cht.cking
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for potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had beer; initiated for
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equipment in need of maintenance. The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security plan. This included verification that the appropriate number of security personnel were on site; access control barriers were operational;
protected areas were well maintained; and vital area barriers were well maintained. The inspector verified the licensee's radiological protection program was implemented in accordance with the facility policies and programs and was in compliance with regulatory requirements, b.
The insptetors performed routine inspections of the control room during off-shift and weekend periods; these included inspections between the hours of 10:00 p.m. and 5:00 a.m..
The inspections were
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conducted to assess overall crew performance and, specifically,
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control room operator atter.tiveness during night shif ts.
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inspectors also reviewed the licensee's administrative controls regarding Conduct of Operations" and interviewed the licensee's security personnel, shif t supervisors and operators to determine if shift personnel were notified of the inspectors' arrivals onsite during off-shifts.
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The inspectors identified a concern pertaining to licensee and
contractor erected scaffolding in the control room and throughout
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the plant. Several of the scaffolds that had been erected did not conform to the licensee's requirements specified in their procedures.
The inspectors determined that both licensed and non-licensed operators were attentive to their duties, and that the inspectors'
arrivals on site appeared to have been unannounced. The licensee has implemented appropriate administrative controls related to r
the conduct of operations. These include procedures which specify fitness for duty and operator attentiveness.
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c.
On the evening of September 15, 1989, the licensee started dropping power on Unit 1 in preparation for a scheduled 13 week refueling and modification outage that commenced on September 16, 1989.
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addition to the normal refueling tctivities, the licensee scheduled i
52 modificatior.s for coropletion dping the outage.
These include Ro7m Desigr Review (ywell coollry movification, Dethiled ControlDCRDR)
completion of the dr
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replerwent fcr the rentor level instrumentation that inputs to the 2eartor Prntection System (RPS).
The licensee reduced reamr poe to 111 in preparation for manually scraming the reactor. At 9:30 p.m., the Nuclear Station Operator (W50) manually scrnned the plar.t.
all systems functioned properly. During the scrafn rccovery, as the flS0s were reviewing the operating parals, they notice 6 that they had received a Groep 7 isolation tiraversing incore Probes). The actuation of the isolhtlon wn initiated by the low reactor water level switches.
During a reactor scram, a rearter water level shrink is expected, with the amount of shrink dependent on the reactor power level at
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the time of the scram. The reactor water level switches are Static-0-Ring (SOR) switches. These particular switches are calibrated to trip at a decreasing water level of 19.5 inches and
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actually tripped at 17 inches which meets the Technical Specifica-tion required value of 12.5 inches.
The NS0s verified that an actual low reactor water level did not exist and that there were no other valid isolation signals present.
The isolation was reset, the reactor was stable and continued to cool down to enter its refueling outage. The Emergency Notification System (ENS) phone call was made for the Engineering Safety) Feature (ESF) actuation of the Group 7 isolation at 12:35 a.m. (CDT on September 16, 1989. The resident inspector was in the control room during the reactor shutdown and reactor manual scram.
The licensee has reviewed the Group 7 isolation actuation and concluded that due to the fact that the reactor water level shrink is expected and that the water level had not exceeded the Technical Specification set. point for low reactor water level, a Licensee Event Report (LEP) will not be required. The inspectors are reviewing the licensee's policy on ESF actuation that are expected.
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On October 14,1989,at2:25a.m.(CDT),thelicenseestarted withdrawing control rods on Unit 2, ending their forced outage.
The unit had initially been shutdown on August 26,1989, for a six day planned outage to replace a degraded reactor recirculation pump seal. On August 28, 1989, a licensee contractor inadvertently
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caused demineralized water to be sprayed into the main generator through the generator air purging lines (See Inspection Report i
373/89019;374/89019). The Unit 2 generator was disassembled and the generator rotor sent off site for repair. On September 15, 1989, the licensee shutdown Unit I for a scheduled refueling /
maintenance outage. The Unit I generator was disassembled and the Unit I rotor placed in the Unit 2 generator.
At 5:10 a.m., the reactor was critical with a period of 222 seconds.
At 8:50 p.m., the main generator was synchronized with the grid.
The main generator was svbsequently rernoved from the grid to perform testing of the turbine stop and control valves.
During this testing, the turbine tripped due to high vibration en the nuc.;ber H bearing (the number 11 bearing it located between the geerator retor and the generator exciter). The centrol valves were all closed and the turbine was coasting down in speeo when the trip occurred. The licentee decided to roll the turbine again with personnel monitoring the bearing vibration at the Turbine Supervisory Instrumentation (TSI) panel and locally using nortable equipment. At 4:30 a.m., the turbine tripped while being rolled
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to rated speed due to high vibration on the number 11 bearing.
The TSI and lot.ai vibration readings however, did not indicate a high vibration problem. The licensee switched the number 11 and 12 vibration probes and decided to again roll the turbine with the turbine trip on high vibration function bypassed.
Individuals were again situated to monitor the bearing vibration and were in communication with the control room ta instruct them to manually trip the turbine if high vibrations were noted. The turbine was brought to rated speed and the testing of the turbine stop valves successfully completed.
High vibration was not noted locally at the' number 11 bearing but was noted at the TSI. The licensee decided to bring the turbine up to speed and monitor the vibration as turbine load was increased. At 6:50 a.m. the generator was re-synchronized with the grid.
On October 16, at approximately 6:10 a.m. as the generator load was being increased, high vibration alarms started periodically occurring on the number 8 bearing (located between the low pressure turbine and the generator). The licensee suspended power escalation and reduced generator load by inserting control rods.
Subsequently, the licensee decided to continue loading the generator while monitoring turbine generator vibration. The generator was ultimately loaded to 100% power and the vibration levels appear to have dropped. The licensee plans during the next outage to collect vibration data while taking the turbine off line, and based upon I
that data plus data they have already obtained, balance the rotor and/or generator as necessary.
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Prior to Unit 2 startup, the inspector accompanied the licensee on a closeout inspection of the drywell. A number of tools and pieces of equipment were found that needed to be removed but no other problems were noted that would prevent startup. The inspector also expressed concern over the storage of some ramps in the area just outside the personnel access hatch. The ramps were relocated by the licensee.
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On October 19, 1989, at approximately 10:30 p.m. (CDT), the licensee crosstied the Division II DC systems of Unit I and Unit 2 in preparation for an upcoming outage on the 142Y buss (which feeds the Unit 1 Division II DC system). On October 20 at approximately 1:05 a.m., buss 142Y was de-energized.
Subsequently, a large current draw along with a -125 volt DC ground came up on Unit 2.
In an effort to determine the cause of the current draw, as well es reduce the current draw. and to locate the ground, the licensee began stripping loads from the 112Y buss (125 volt DC distribation).
During the course of this trouble shooting, the Unit 2 Standby Gas
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Treatment (SBGT) system auto started and the Unit 2 reactor building
ventilation system i olation dampers closed. The lictnsee placed
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buss 142Y back in service which relieved the current draw on the Unit 2 batteries and caused the ground to disappear. The SBGT system was stopped, the reactor building ventilation system isolation dampers opened, and the reactor building ventilation
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system restarted.
Investigation by the licensee as to the cause of the event revealed that one of the loads stripped on the DC buss powered the icgic for the SBGT system and the manual pushbutton for isolating the reactor building ventilation system. The licensee made the required Emergency Notification System (ENS) cotification at 3:10 a.m.
The licensee investigated the source of the DC ground.
This was subsequently determined to be caused by one of the annunciator panel power supplies.
The power supply was replaced and buss 142Y was taken back out-of-service to allow the preplanned outage to continue.
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On October 19, 1989, at 4:23 p.m. (CDT), breakers for Unit 1 480 VAC busses 132X and 132Y tripped open. As a result, the Unit 1 B
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reactor building ventilation supply and exhaust fans tripped causing reactor building differential pressure to approach 0 psid.
The licensee declared secondary containment inoperable and started the train A reactor building ventilation supply and exhaust fans in order to recover the required differential pressure. At 4:30 p.m.,
secondary containment integrity had been restored.
When busses 132X and 132Y tripped, power was lost to a number of pieces of equipment, including the Technical Support Center (TSC)
Uninterruptable Power Supply (UPS). The TSC.'s UPS supplies power to the Prime 1 and 2 computers.
This made the Safety Parameter Display System (SPDS) in the Emergency Operations Facility unavailable. As a result, the licensee made the required Emergency Notification System (ENS) notification at 5:40 p.m. for impairment of offsite emergency assessment capability.
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Investigation-by the licensee has indicated that at the time of the event, the Operational Analysis Department (OAD) had been performing calibrations of the electrical protection devices on the 132X and 132Y busses. At 4:04 p.m., OAD had finished the calibration of the 132Y neutral overcurrent (N00) relay and had just reinstalled it.
At 4:27 p.m., 0AD returned to the 132Y buss to reinstall the undervoltage relay that had been removed for calibration and found the buss de-energized and the NOC relay tripped. The licensee removed and checked the calibration of the NOC relay, inspected the 132X and 132Y busses, and inspected the 4.16 KV feed breaker to the transformer and found no damage or other indication as to the cause of the NOC relays tripping. At 6:20 p.m., the 132X and 132Y busses were re-energized.
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On October 21, 1989, e.t approximately 1:00 a.m. (CDT), the power supply (Duit 1 buss 136X-2) for the Standby Gas Treatment (SBST)
systems Wide Range Gas Monitor (9 REM) was taken out af serv 1ce for pteplanned modification worL This placed Unit 2 (Jnit I was shutdawn arid defteled) in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limiting c9nair, ion for operaion (tCO) per Technical Specificatien 3.3.7.5 to initicle the planned alternate method of menitoring S8ST effluent and a 7 day LCO to
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return the SBGT WRGM to operability. On October 23, 1989, at i.pproximately 12:30 p.m., the licensee commenced setting up the
alternate method (the Stack Particle, Iodine, and Noble Gas (SPING)
iconitor). At 4:30 p.m., the Shift Engineer wer notified that a problem existed with the SPIN 3's detector tube and high voltage power supply. At 5:50 p.m. when it was realized that the SPING was not going to be operable, preparations were commenced to make a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Emergency Notification System (ENS) notification in accordance with the licensee's guidance for making notifications.
The ENS notification for loss of emergency assessment capability was made at 6:45 p.m.
With the alternate monitoring method not available, the licensee had until October 24 at 1:00 a.m. to restore it, or the WRGM, or to enter Technical Specification 3.0.3 and commence a shutdown of Unit 2 within I hour. On October 23 at 9:05 p.m., the licensee made a temporary power connection to the
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high range (accident monitoring) portion of the SBGT WRGM and commenced testing to verify its availability. At 10:15 p.m., the accident monitoring portion of the SBGT WRGM was again available and the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LC0 was exited. The licensee still. remained within the 7 day LC0 to restore the SBGT WRGM to operability. At 11:15 p.m., the licensee made a followup ENS call to inform the NRC that the emergency assessment capability had been restored. On October 25, at 8:30 a.m., the SPING repairs were completed and it was declared operable and was in place as the planned alternate method of monitoring SBGT effluent.
Subsequently, the temporary power connection to the high range portion of the SBGT WRGM was removed. On October 25, at 2:00 p.m., normal power was restored to the SBGT WRGM and the 7 day LC0 was exited.
No violations or deviations were identified in this area.
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4.
Monthly Surveillance Observation (61726)
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The inspectors observed surveillance testing, including required Technical Specification surveillance testing, and verified for actual tctivities observed that testing was performed in accordance with adequate procedures.
The inspectors also verified that test instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were t
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accomplished and that test results conformed with Technical Specification and procedure requirements. Additionally, the inspectors ensured that
L the test results were reviewed by personnel other than the individual
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directing the test, and that any deficiencies identified during the
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testing were properly reviewed and resolved by appropriate manageraent
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personnel.
The fMpectors witnessed portions of the following-test activities:
1IS-NB-408 Urdt 2 Reactor Yesse? Low Low Water Level !!igh Pressure Core Spray Initiation Functional Test
LES-RR-101 Unit 1 Anticipated Transient Without Scram (ATWS) Reactor Recirculation (RR) Pump A Trip System Relay Logic Functional Test and Simulated Automatic Operation
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LIS-LC-408 Unit 2 Main Steam Isolation Valve Leakage Control Steam Line Bleed-0ff Flow Functional Test
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LOS-LP-Q2 Unit 2 Low Pressure Core Spray Valve Inservice Test for Operating Conditions 1, 2, and 3 LIS-MS-407B Unit 2 Reactor Vessel Low Water Level 1 and Level 2 Isolation Channels B & D Monthly Functional Test
LOS-TG-SA2 Turbine Valve Tightness Surveillance No viclations or deviations were identified in this area.
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Monthly Maintenance Observation (62703)
Station maintenance activities of safety related systems and components
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listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications.
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
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' outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance.
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inspection period:
Disassembly / Reassembly Unit 2 Main Generator
Disassembly Unit 1 Main Turbine Disassembly Unit 1 A Reactor feedwater Turbine No violations or deviations were identified in this area.
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LicenseeEventReportsFollowup(90712,92700)
Through direct observations, discussions with licensee personnel, and
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review of records, the following event reports were reviewed to determine
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that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in.accordance with Technical Specifications.
The following reports of nonroutine events were reviewed by the inspectors. Based on this review it was deterr
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of minor safety significance, did not represent gram deficiencies,
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were properly reported, and were properly compens-.ed for. These reports are closed:
373/89014-00 Reactor Building Isolation Damper failure and Indicating Light Circuitry Design inadequacy 373/89023-00 Spurious Closure of Reactor Recirculation Process Sample Inboard Isolation Stop Valve Due to Failed Relay 374/88014-01 Type B and Type C Total Leakage Exceeded 0.6 La During Leak Rate Yesting During Refuel Outage Two 374/89010-00 High Pressure Core Spray Inoperable Due to Division 111 Battery Charger Oscillations 374/89011-00 Spurious Reactor Protection System Actuation Due to Unknoe Cause
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374/89012-00 Shutdown Cooling Isolation During Reactor Protection System i
Bus Transfer Due' to Relay Deenergizing
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i 374/89013-00 Primary Containment Isolation During Instrument j
Surveillance Testing of LIS-MS-401 i
No violations or deviations were identified in this area.
7.
ESF System Walkdown (71707)
i The operability of selected engineered safety features was confirmed by the inspectors daring walkdown of the accessible portions of the following systems. The following items were considered during the walkdowns:
verification that procedures match the plant drawings,
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equioment conditions, housekeeping, instrumentation, valve and electrical breaker lineup status (per procedure checklist), and verification that items including locks, tags, and jumpers were properly attached and identifiable..The following systems were walked down this inspection period:
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. Unit 2 Standby Gas Treatment System
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Unit 2 A Emergency Diesel Generator Unit 1 High Pressure Core Spray System No violations or deviations were identified in this area.
8.
Training (71707)
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The inspector, through discussions with personnel, evaluated the licensee's training program for operations and maintenance personnel to determine whether the general knowledge of the individuals was sufficient for their assigned tasks.
In the areas examined by the inspector, no items of concern were identified.
No violations or deviations were identified in this area.
9.
Security (71707)
a.
The licensee's security activities were observed by the inspectors during routine facility tours and during the inspectors' site arrivals and departures.
Observations included the security personnel's performance associated with access control, security checks, and surveillance activities, and focused on the adequacy of security staffing, the security response (compensatory measures),
and the security staff's attentiveness and thoroughness. The
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security force's performance-in these areas appeared satisfactory with the exception of the item noted below.
b.
On October 4, 1989, during a routine plant tour, the inspector noted that several deck plates in the floor of the Unit.1 761' elevation were marked with a paper sign indicating that security should be contacted prior to opening.
Further inspection revealed that if the drywell (Unit I was in a refueling outage) gained to the Unit I deck plates were removed, access could be without going past the guards posted to control entry. The paper signs were also deteriorating due to people standing / walking on them. The inspector brought this to the attention of the licensee. The licensee indicated that this was not a problem since both areas were all within the vital area. The licensee also noted that the paper signs were going to be replaced with plastic signs. During a deep back shift inspection on October 11, 1989, the inspector noted that the paper signs had been replaced with plastic signs as the licensee had indicated would occur.
Because the use of the signs did not seem to be consistent with the controls (i.e., locked doors, physical barricadec, etc.) in place to prevent entry to the same area from other locations, the inspector followed up on the original concern and found several other potential pathways. On October 11, the inspector contacted the regional security inspector assigned to
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LaSalle and discussed the above concerns.
Subsequently, a special security inspection was conducted. The details of this inspection L
.can be.found in Inspection Report No. 373/89022. On October 11, the licensee took compensatory measures by posting guards in the
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appropriate locations.
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No violations or deviations were iderctified in this area.
~10.
RefuelingActivities(60710)
Portions of this module instructed the inspectors to ascertain whether the defueling/ refueling activities were controlled and conducted as p
required by Technical Specifications and approved procedures.. Items reviewed were:
That equipment checkout was satisfactorily completed
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Site management involvement in refueling preparations
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That plant conditions were being maintained as required by
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Technical Specifications t
That the correct revision of the applicable procedures were in
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use and that personnel manipulating the spent fuel bridge were qualified to do so Core monitoring during refueling operations was in accordance with
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Technical Specifications Fuel accountability methods were in accordance with established
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procedures
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Vessel and spent fuel storage pool water levels were as required by
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Technical Specifications Specification (BWR) position was as required by Technical Reactor mode switch
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Control blade checks (BWR) were conducted in accordance with
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applicable procedures Operability of refuel bridge interlocks functioned properly
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Checks of decay heat removal system were being conducted as required
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by Technical Specifications Radiation controls were in place and checked
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Areas of direct observation included the refueling floor and the control room. Defueling of the Unit I reactor commenced on September 26, 1989, and was finished on October 3, 1989.
Refueling is scheduled to begin on November 27, 1989.
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No deviations or violations were identified in this area.
11. ' Evaluation cf Licensee Self-Assessment Capability (40500)
This inspection evr'9ated the effectiveness of the licensee's self-assessme-wer ams.
The intpection focused on determining whether the licensee's 4 elf-e::essment programs contribute to the prevention of.
problems by monitoring and evaluating plant performance, providing
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assessments and findings, and communicating and following up on corrective action reccmmendations.
On October 4, 1989, the licensee held their regularly scheduled Event Frequency Reduction Meeting. The inspector had reviewed the. licensee's procedure LAP-200-7, Event Frequency Reduction Post Event Review Program, prior to attending the meeting. The items discussed pertained to resolutions for events the licensee or other licensees had experienced.
The items were discussed,_ potential resolutions reviewed actions for mitigating future events were generated, and assignment to a responsible-person was made.
i The meeting was organized well, past activities reviewed which included a brief progress report, and dates assigned for either further progress update or task completion.
No violations or deviations were identified in this area.
12. TMI Action Plan Requirement Followup (25565)
a.
(Closed) TMI Item II.F.2.4 (Units I and 2): This item required the licensee to perform certain improvements to the Reactor Water Level Instrumentation System. The objective of this requirement was to provide increased assurance that the water level instrumentation would provide an unambiguous indication of inadequate Core Cooling (ICC).
By letter dated October 26, 1984, the NRC issued Generic Letter 84-23, Reactor Vessel Water Level Instrumentation in BWR's, which identified two categories of potential improvements in BWR water level instrumentation to give increased assurance that water level instrumentation would provide the core cooling instrumentation required by TMI Action Item-(NUREG-0737) II.F.2.
By-letter dated March 2, 1987, the NRC issued a Safety Evaluation Report (SER) which documents acceptance of the licensee's proposed modifications and plan for compliance with Generic Letter 84-23. A review of this SER found it to conclude that the licensee's proposed modifications will assure that the LaSalle Units 1 and 2 reactor water level measurement systems provide the adequate core cooling instrumenta-tion required by NUREG-0737, Item II.F.2, and are acceptable.
The licensee has prepared and fully implemented the previously approved modifications for each unit. Modification number M-01-1-84-089 was prepared to address this item for Unit 1.
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modification was found to be completed on December 8, 1987.
Modification number M-01-2-84-131 eddresses this item for Unit 2 and was found to be completed on July 12, 1988.
Close-out packages
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and supporting documentation, related to-each modification, were reviewed for completeness.- No deficiencies were identified during
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this review.
Based on the above review, TMI Action Item II.F.2.4 (Units 1and2)isclosed.
o b.
(Closed) TMI Item I.C.7.1 and 2 (Unit 2): This item required the licensee to obtain. Nuclear Steam System Supply (NSSS) vendor (General Electric) review of its low power test,-power-ascension y
test, and emergency operating procedures.
In addition, NSSS vendor review of power-ascension test procedures is a condition of the Operating License for Unit 1 (Unit 1 License Condition 2.C.30b).
In Supplement No. 4 to the LeSalle County Station (LSCS) Evaluation
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L Report '(SSER-4) dated July 1982, the NRC documented its acceptance of the NSSS vendor review of low power test procedures applicable to Unit I and concluded that the license condition for Unit I had been
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acceptably resolved.-
By letter dated March 18, 1983, fromC.Schroeder(CECO).to A. Schwencer (NRC), the licensee informed the NRC that, with the exception of the lessons learned section, the power-ascension test
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. procedures applicable.to Unit 2 were identical in content to the previously approved Unit 1 power-ascension test procedures. Based on this similarity, the licensee requested that THI Action-Item I.C 7 not appear as a license condition of Unit 2.
The licensee could not provide objective evidence verifying NRC acceptance of the stated position.
However, the inspector's review of license conditions applicable to Unit 2 did not identify this item as a condition of its Operating License. Additional verification of the NSSS vendor review of Unit 2 aower-ascension test procedures was provided by the licensee in t1e form of a letter, dated
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May 31, 1984, fromJ.Ellis(G.E.) tog.-Diederich(CECO). A review of this letter found it to indicate that with the' exception-of Startup Test Procedure 27-2, Turbine Trip / Load Rejection,
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(STP-27-2), all previous General Electric startup test procedure comments had been incorporated and/or resolved satisfactorily.
Additionally, a subsequent NRC Inspection Report (Inspection Report-374/84-26, dated October 1, 1984) was found to document the witnessing of certain Unit 2 startup test procedures, including
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STP-27-2, with no items of noncompliance or deviations identified in this area.
Therefore, it appears that the NSSS vendor review of power-ascension test procedures applicable to Unit 2, has been acceptably resolved.
The adequacy of the licensee's Emergency Operating Procedures (E0PS) has been reviewed and is documented in NRC Inspection Reports Nos. 50-373/86042; 50-374/86042. Based on the inspector's review of the above referenced documentation, TMI Action item I.C.7 (Unit 2) is closed.
c.
(Closed) TMI Item II.K.3.21b (Units 1 and 2): This item required the licensee to implement a design modification for the automatic restart of the High Pressure Core Spray System (reference LaSalle SER, Section 22, item II.K.3.21). A review of NRC Inspection Report
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Nos. 50-373/84-33 and 50-374/83-37 found this modification to be completed acceptably for LSCS Units I and 2.. Based on the review of i
the above referenced documents, TMI Action Item II.K.3.21b (Units 1 and 2) is closed.
d.
(0 pen) TMI Item II.K.3.16b (Units 1 and 2):
This item required the licensee to perform an evaluation of reduction of challenges and failures of relief valves. The licensee committed to supply the results of this study to NRR for review. The inspector could find no objective evidence that (reference 22, item II.K.3.16, of the LSCS SER) the licensee forwarded the CWR Owners' Group Evaluation of NUREG-0737 Item II.K.3.16, Reduction of Challenges and Failures
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of Relief Valves, to NRR for review.
The licensee was unable to provide any evidence to support their claim that it had been submitted other than a reference in NRC Generic Letter 83-36, Enclosure 2, item 4, which makes reference to the staff having reviewed information submitted by the BWR Owners Group in response to item II.K.3.16b.
In addition, the licensee was unable to provide objective evidence which verified NRR acceptance of this study.
In an effort to resolve this issue, the reviewer contacted cognizant personnel at NRR but was unable to achieve an acceptable resolution at-that time. Therefore, TMI Action Item II.K.3.16b (Units 1 and 2)
remains open pending verification of the submittal of and acceptance of the BWR Owners Group. study by NRR.
e.
(0 pen) TMI Item II.K.3.28 (Units 1 and 2): This item required the licensee to perform an evaluation of the qualification of ADS valve accumulators. The licensee committ ad to provide the results of a BWR Owners Group Study to the NRC by January 1, 1982 (reference LSCS SER, Section 22, Item II.K.3.28). By letter dated January 27, 1982, from.R. Tedesco (NRC) to L. DelGeorge (Ceco), NRC forwarded a Request for Additional Information (RAI) which cited several concerns relating to its. review of the BWR Owners Group evaluation submitted by the licensee.
By letter dated February 19,.1982, from C. Sargent (CECO) to A. Schwencer (NRC), the licensee had forwarded its response to the RAI. The licensee was unable to provide i
objective evidence necessary to verify NRR acceptance of its response to the RAI. Therefore, TMI Action Item II.K.3.28 (Units 1 and 2) remains open pending verification of NRR acceptance of the BWR Owners Group study and the licensee's response to its RAI.
No violations or deviations were identified in this area.
13. Emergency Preparedness (71707)
a.
On September 30, 1989, at approximately 9:10 p.m. (CDT), the Shift Engineer was notified that a contractor had been injured in a fall while working in the Unit I low pressure feedwater heater bay.
The contractor apparently sustained injuries to a knee, arm, shoulder, and his back. The low 3ressure heater bay is a radioactively contaminated area and t1e injured individual was dressed in a full set of anti-contamination clothing.
In preparation for transporting the individual to the hospital, the anti-contamination clothing was cut away, with the exception of a portion under his back, and he was
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- surveyed and found not to be contaminated. The portion of L
~ anti-contamination clothing under his back w0s not removed due to the potential for back injuries.
At approximately 10:00 p.m.,
the contractor. was transported to St. Mary's hospital in i
Streator, Illinois, by ambulance. A licensce Radiation Protection Technician accompanied the. ambulance since the potential for the contractor's back and tht. piece of anti-contamination clothing to be contaminated existed. The hospital was notified that the injured-individual might be contaminated. Also at 10:00 p.m., the licensee declared an Unusual Event in accordance with their General. Site
. Emergency Plan (CSEP). At 10:20 p.m., the licensee made the r
required Emergency Notification System (ENS) notification for the declaration of an Unusual Event. At approximately 11:20 p.m., the licensee was notified that the remaining portion of the injured.
t contractor h6d been surveyed and he was found not to be contaminated. At 11:35 p.m., the licensee terminated the Unusual Event and at 11:43 p.m. the licensee made a followup ENS call to notify the NRC of the termination of the Unusual Event. The injured contractor was subsequently released from the hospital.
He sustained no broken bones but did sustain contusions and strained muscles.
7he residents have evaluated the licensee's actions during the event and determined that the licensee's actions were prompt and correct
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for.the circumstances. All notifications were made within the required time-frames.
b.
On October.7, 1989, at approximately 1:50 p.m. (CDT), the Shift Engineer was notified of an injured individual in the Unit 1 drywell.
The individual had apparently tripped and fallen while working in the drywell and sustained hip injuries. The drywell is a radioactively contaminated area and the injured individual was dressed in a full set of anti-contamination clothing.
The licensee decided to transport the individual to the hospital without first removing the anti-contamination clothing due to the hip injuries. After making-this decision, at 2:30 p.m., the licensee declared an Unusual Event in accordance with their GSEP.
At approximately 2:35 p.m., the individual was-transported to St.
Mary's hospital in Streator, Illinois, by ambulance. A licensee Radiation Protection Technician and Health Physics Technician accompanied the ambulance to the hospital. At 2:45 p.m., the licensee made the required ENS notification for the declaration of u
the unusual Event. After arrival at the hospital, the injured individual's anti-contamination clothing was removed and he was surveyed.
No contamination was found on the individual and only fixed contamination was found on the anti-contamination clothing, The licensee's technicians surveyed the ambulance, the hospital, and
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the personnel involved and found no contamination and retrieved the
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anti-contamination clothing for return to the site. At 3:35 p.m.,
the licensee terminated the Unusual Event and made a followup ENS call to notify the NRC of the termination of the Unusual Event. The injured individual was subsequently released from the hospital the same day and returned to work on October 9, 1989.
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L The residents have evaluated the licensee's actions during the event and determined that the licensee's actions were prompt and correct
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for the circumstances. All notifications were made within the
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required time frames.
No deviations or violations were identified in this atea.
Evaluation of Exercises For Emergency Preparedness (82301)
On October 17, 1989, the inspectors observed the site oortion of the
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licensee's annual medical General Site Emergency Plan '(GSEP) drill.
The drill commenced at approximately 8:40 a.m. (CDT) on the 815 foot elevation of-the Unit 1 Auxiliary Building. The drill consisted of a
~ simulated accident involving an individuel, dressed in a full set of anti-contamination clothing, who fell backwards off a ladder and sustained head injuries that were bleeding, as well as potential back injuries. The drill area was simulated to be a contaminated area.
In addition to the licensee personnel observing the drill, the Federal Emergency Management Agency (FEMA) also participated as observers.
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The. drill appeared.to be well planned and with the exception of several
. minor problems appeared to go well. The problems noted by the inspectors were as follows:
The drill was being video taped and the individual performing this
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function was sticking the camera in the participants' faces and appeared to be interfering with the participants.
'At one point during.the drill, one of the participants asked several
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af the observers, who should have been " invisible" to the
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participants, to go and get some rescue equipment.
The ambulance crew did not like the type of back board the licensee
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had, so a delay in the rescue occurred while the ambulance crew's back board was brought to the scene of the accident.
The Auxiliary Building elevator broke and this method of evacuating
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-the " injured" individual could not be used. The drill was temporarily suspended and the " injured" individual walked down the stairs to the ground level where he resumed his position on the stretcher.
No violations or deviations were identified in this area.
15. Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities (92700)
The Foxboro Company (FOXBORO) informed the NRC on October 3,1989, of a potential deficiency affecting Foxboro Model N-Ell and N-E13 pressure transmitters containing 10-50 MA type amplifiers manufactured between January 1,1988 and September 1,1989.
The N prefix indicates nuclear grade or safety related.
The licensee conducted a review and determined that they do not have any N-Ell or N-E13 Foxboro pressure transmitters.
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p The licensee did determine that they did have various E-11 and E-13 pressure transmitters in the stockroom, however, these transmitters are used in' balance of plant' systems and workmanship problems did not occur in this line.
No deviations or violations were identified in this area.
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16.
Report Review (90713)
During the inspection period, the inspectors reviewed the licensee's Monthly Operating Report for September. The inspectors confirmed that
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the information provided met the requirements of Technical Specification 6.6. A.5 and Regulatory Guide 1.16, 17. Outages (71707)
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On September 22, 1989, the residents were notified by the licensee that Engineers and Constructors (UE&C) Quality Assurance (QA) audit of United on September 20, 1989, a routine
- Catalytic, the licensee's general contractor on site, had revealed that many of their Quality Control (QC)
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inspectors did not appear to have documented evidence to show that all requirements of the UE&C - Catalytic Inspector Certification procedure had been met. Background checks were not documented in the file and it appeared that not all inspectors had performed all the practical exams required for their_ certifications. UE&C - Catalytic has about 30 QC personnel on site. At 3:00 p.m. (CDT), UE&C - Catalytic placed an -
administrative hold on all activities requiring quality inspection and surveillance. The licensee also immediately reviewed all safety-related work'in progress or completed for-its impact on equipment operability.
An initial concern dealt with the C train of the Residual' Heat Removal (RHR) system. Subsequent inspection by qualified individuals indicated that the work had been accomplished satisfactorily.
Effective as of 3:48 p.m. on September 21, a Stop Work Order was issued preventing any safety-related or Code work on safety-related equipment and regulatory related work on work packages released to the field, the issuing of any new safety-related or regulatory related packages to the field, and any safety-related AWS and ASME code welding _ activities, including weld qualification.
The licensee, in conjunction with UE&C - Catalytic, initiated an overview of other activity areas to determine if the problems initially found in the QC inspector certification area were present.
While some deficiencies were found that had the potential to cause problems, none were identified that resulted in any existing problems in completed work.
By September 27, the licensee had gained sufficient confidence in UE&C
- Catalytic's ability to fully implement their quality program that a phased lifting of the Stop Work Order was begun.
Critical safety-related work has begun for which sufficient numbers of certified QC inspectors exist. The licensee established and put into effect management controls to control the scope of safety-related work underway and established an overview of that work by their Engineering and Construction group.
In addition, UE&C - Catalytic established a corporate QA overview of the ongoing work activities.
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L On October. 3,1989, the required ' actions of UE&C - Catalytic had been completed to the licensee's satisfaction and-all work restrictions were removed. 'On.0ctober 10, 1989, during a followup audit by the licensee's F
onsite QA organization, a discrepancy was noted in the educational ~
background information supplied by one of the UE&C - Catalytic QC inspectors. The high school diploma submitted appeared to have been falsified. This concern was turned over to UE&C - Catalytic for resolution.
UE&C - Catalytic contacted the high school and determined that the diploma was indeed false. The QC inspector was questioned and.
he admitted to having supplied a false high school diploma. He was subsequently terminated. A spot check of his' work on site, plus
. interviews of his supervisors and other QC inspectors, revealed no problems.
The individual was fully qualified and, based on his work experience of approximately 17 years, could have been qualified for his
position on site without a high school diploma. This individual had
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previous ex3erience at other nuclear power plants in a QC capacity, therefore, lis name and his work experience were provided to Region III for dissemination and followup at these facilities.
No violations or. deviations were identified in this area.
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18. ManagementMeeting(30703)
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i On October 6, 1989, a management meeting was held in the Region III J
office between the NRC and Commonwealth Edison. The meeting pertained j
to the August 26, 1989, Unit 2 reactor scram where two of four reactor protection system scram relays had not tripped.
This was.the second meeting on this subject. During the first meeting, which was held September 11, 1989, the licensee had not completed their
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investigation of the reactor scram, which included extensive testing of l
the Reactor Protection System scram relays.
i Based on the.results of the licensee's investigation, Region III managers concluded that there were no safety concerns to impede the restart of Unit 2.
19. Exit Interview (30703)
Theinspectorsmetwithlicenseerepresentatives(denotedinParagraph1)
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 inspectors also discussed the likely informational contents of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.
l The licensee did not identify any such documents or processes as J
proprietary.
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ATTACHMENT 1.
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IITEM: NUMBERS
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Unit 1:'
Unit 2-
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373/81137-00 374/81000-19
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- 373/81138-00.-
374/81000-36'
373/81139-00-374/81000-38-i 373/81149-00 374/81000-49.-
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,
373/81150-00.-
374/81000-57
.373/82055-05'
-374/81000-59
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J373/84033-01d 374/83028-04
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-373/85038-03 374/84002-06 373/86002-BB-c374/84013-03-373/86007-02 374/85003-HH
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- 373/86011-01 374/85007-03-
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-373/86015-01=
374/85013-HH:
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373/86018-01-374/85018-07-373/86018-03 374/85021-02 e
373/86025-02-
-374/85022-HH
'373/86046-01 ~
,374/85025-01 t
373/87006-06:
'374/85031-03-
.
373/87030-03; 374/85039-03
,0
'373/87035-04 374/85039-07 373/87035-05 374/86002-BB 373/87035-07/
374/86016-01-373/87035-11--
!374/86036-02-373/88025-01'
374/86036-03-373/89003-02-374/86046-03
,
374/87006-03
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'374/87006-07 374/89003-04
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