IR 05000373/1987016
| ML20214Q855 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/29/1987 |
| From: | Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20214Q828 | List: |
| References | |
| 50-373-87-16, 50-374-87-16, NUDOCS 8706050226 | |
| Download: ML20214Q855 (14) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report Nos. 50-373/87016(DRP);50-374/87016(DRP)
Docket Nos. 50-373; 50-374 Licenses No. NPF-11; NPF-18 Licensee: Commonwealth Edison Company
Post Office Box 767 Chicago, IL 60690 Facility Name: LaSalle County Station, Units 1 and 2 Inspection At: LaSalle Site, Marseilles, IL Inspection Conducted: April 25 through May 18, 1987 Inspectors:
M. J. Jordan R. Kopriva J.
Malloy N
7 Approved By:
M. Ring, Ch
&
y Reactor Projects Section IU D6te
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Inspection Summary Inspection on A)ril 25 through May 18, 1987 (Reports No. 50-373/87016(DRP);
50-374/87016(DR)))
Areas Inspected:
Routine, unannounced inspection conducted by resident inspectors of licensee actions on previous inspection findings; operational safety; surveillance; maintenance; training; Licensee Event Reports; refueling activities; facility modifications; local leak rate testing; and emergency exercise.
Results: Of the ten areas inspected, one violation was identified (Paragraph 4-failure to follow Technical Specifications).
The licensee's reload of the Unit 2 core went very well with no personnel errors. The communications between the control room and refueling floor were good. The licensee had several personnel problems while performing weekly surveillances for rod exercising that reflected a poor shift turnover, inadequate completed procedure review, and failure to document a problem while performing a surveillance.
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i 8706050226 870529 PDR ADOCK 05000373
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DETAILS 1.
Persons Contacted
- G. J. Diederich, Manager, LaSalle Station
- R. D. Bishop, Services Superintendent
- J. C. Renwick, Production Superintendent D. Berkman, Assistant Superintendent, Work Planning
- W. Huntington, Assistant Superintendent, Operations
- P. Manning, Assistant Superintendent, Technical Services T. Hammerich, Assistant Technical Staff Supervisor W. Sheldon, Assistant Superintendent, Maintenance J. Atchley, Operating Engineer D. A. Brown, Quality Assurance Supervisor
- D. Enright, Quality Assurance Engineer M. Richter, Assistant Technical Staff Supervisor
- Denotes personnel attending the exit interview on May 18, 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) Open Item (373/84010-06a(DPRP)): This item concerns trouble-shooting guidance for maintenance not being uniformly applied with LAP 1300-1, " Work Requests," and QP 3-52, " Design Control for Operations-Plant Maintenance." The inspector's concern was that if troubleshooting is to be performed which impacts safety or reliability, the nature and scope of such work should be clearly defined in advance. The licensee has added Attachment D to LAP 1300-1, " Maintenance Department Trouble-shooting Worksheet," for troubleshooting work requests that could impact safety or plant reliability. The work requests are then reviewed with the shift engineer or his designee to determine actions and precautions required during troubleshooting and documented on Attachment D.
This item is considered closed.
(Closed) Open Item (373/84010-03(DPRP)): This item tracked the upgrading of electrical maintenance procedures that better define craft and quality control responsibilities with respect to conductor butt splicing activities.
The inspector verified that LEP-GM-113. " Cable Terminations and Taping,"
had been revised to include better definitions of craft and quality control responsibilities. This item is considered closed.
(Closed)Openitem(373/84010-5b(DPRP)): This item concerns a discrepancy between LAP 1300-1 and QP-3-52. LAP 1300-1 did not specifically require the Operating Engineer or the Assistant Superintendent to review and concur on Shif t Engineer-specified testing. The inspector reviewed Revision 35 of LAP 1300-1.
This now includes in Step F.3h review and signature by the Operating Engineer signifying approval of the post maintenance tests specified by the Shift Engineer. This item is considered closed.
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(Closed) Violation (373/86040-01; 373/86040-02(DRP)): On October 16, 1986, during shiftly surveillance LOS-AA-S1, the lamp test on the primary containment isolation status panel for identification of off normal indication or condition was performed, but the off normal indication of the Group 6 isolation light was not identified or acknowledged. Also on October 17, 1986, the checkoff list for the " Unit 1 Residual Heat i
Removal (Shutdown Cooling Mode) High Suction Flow Isolation Functional Test," was not adhered to in that two people did not verify that the reactor high pressure isolation instrument stop valve was open.
Immediate corrective action was taken within 30 minutes of identification of these problems. Maintenance foremen / supervisory personnel verified valve lineups on both units for the instruments involved. Also, a sample verification of 100 safety related instrument valves on each unit was
performed. All valves were in their correct positions.
Isolation logic
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lights on the control room back panels have been converted to the Green l
Board / Black Panel configuration used on the control room front panels.
The resident inspector has reviewed the corrective actions and finds the licensee's response 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.
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The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection centrols.
During the month of April 1987, the inspector walked down the
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accessible portions of the following systems to verify operability:
Unit 1 Standby Gas Treatment System Unit 1 Standby Liquid Control System b.
Pursuant to a memorandum from C. E. Norelius dated April 23, 1987, pertaining to "Off-Shift Inspection of Control Room," the inspectors performed several off-shift inspections of the licensee's site, particularly the control room.
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On May 7 and 8, 1987, the inspectors made off-shift inspections, specifically between 12:01 a.m. and 6:00 a.m.
The inspectors observed control room operations and conducted discussions with control room
operators during that time. The operators, both in the plant and in
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the control room, along with other plant personnel, were alert and attentive to duty.
No violations or deviations were identified in this area.
4.
Monthly Surveillance Observation (61726)
The inspector observed Technical Specification required surveillance testing and verified for actual activities observed that testing was performed in accordance with adequate procedures, that test instrumenta-tion was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specification and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management
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personnel.
i The inspector witnessed portions of the following test activities:
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LOS-AA-W1 Technical Specifications Weekly Surveillance LIS-NB-201 Unit 2 Reactor Vessel Low Water Level Scram and Primary Containment Isolation Calibration
a.
On April 3, 1987, with Unit 1 operating in single loop operation
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at 59% reactor power, the Unit I night shift operator started LOS-AA-WI, the weekly control rod exercising surveillance at
approximately 12:10 a.m..
This surveillance requires that each of the 185 withdrawn control rods be exercised by inserting the rod one notch, then withdrawing the control rod back to its original position to prove rod operability.
The operator performed approximately 60 control rod tests and turned the rest over to the day shift operator (7:00 a.m. to 3:00 p.m.) to complete. The night shift operator noted that some intermediate rods were double-notching (i.e., some rods inserting two notches instead of
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one and some withdrawing two notches on a withdraw signal). The
night shift operator was also receiving Rod Block Monitor and Average Power Range Monitor (APRM) Hi Alarms, so he decided to discontinue testing eleven intermediate rods on the checklist and have the remainder of the intermediate rod testing be performed on the day shift when the nuclear engineer was present.
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i eleven intermediate rods were 6-39, 14-15, 14-31, 14-47, 18-19,
18-43, 22-07, 22-23, 22-31, 22-39, and 22-55.
When the night shift operator turned over to the day shift operator,
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the night shift operator mentioned that the intermediates were double-notching. The fact that eleven intermediate rods were not
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tested was not communicated. The night shift operator failed to document in the comments section of surveillance LOP-AA-W1, that eleven intermediate rods were not tested.
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The day shift operator assigned the job of completing the surveillance to an extra operator.
This operator exercised the rest of the control rods from where the night shift operator left off, but did not notice that eleven intermediate rods had not been tested. The extra operator returned the surveillance to the day shift operator and assumed it was complete.
The day shift operator reviewed the surveillance and signed completion, but did not notice that eleven intermediate rods were not signed off as complete.
The shift control room engineer reviewed the completed surveillance and signed his review, but he failed to notice that some control rods were not exercised.
l The surveillance was then signed off on the weekly surveillance
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schedule and transmitted to the Unit 1 operating engineer. On
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April 16, 1987, the Unit 1 operating engineer reviewed the l
surveillance and noticed that eleven control rods were not
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i signed off as complete.
He confirmed that the control rods i
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that were missed on the April 3,1987 surveillance were completed and operable on the next weekly surveillance.
Technical Specification Surveillance Requirement 4.1.3.1.2 requires that all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated operable by moving each control rod at least one notch at least once per seven days whenever above
the low power setpoint of the Rod Worth Minimizer and the Rod
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i Sequence Control System (20% power).
Eleven intermediate control rods. 6-39, 14-15, 14-31, 14-47, 18-19, 18-43, 22-07, 22-23, 22-31, 22-39, and 22-55 were not demonstrated operable by moving each
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control rod at least one notch at least once a week between the
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period of March 27, 1987 to April 10, 1987. This is considered a
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violation of Technical Specification Surveillance Requirement
4.1.3.1.2a.
(373/87016-01(DRP)).
i b.
On April 30, 1987, at 10:30 a.m. (CDT), the 2C Residual Heat Removal
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(RHR) System was declared inoperable. The licensee was performing surveillance procedure LIS-RH-203, " Unit 2 RHR pump Minimum Flow Bypass (LPCI Mode) Calibration", which requires calibration of Static-0-Ring (SOR) differential pressure switch 2E12-N010CA.
During the calibration of this switch, the switch repeatedly tripped
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at greater than 30 inches of Water Column (WC). The rejectable limits for this switch tripping were less than 17.7 inches WC or
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greater than 28.3 inches WC. The Technical Specification Limiting j
Condition for Operation (LCO) for this switch was 1.78 inches WC.
The licensee replaced this switch with a new SOR differential pressure switch.
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On May 7, 1987, a regional based inspector was reviewing a Licensee Event Report (373/87-13) which identified a problem with a breaker for the "B" drywell ventilation supply fan. He requested the licensee to provide documentation as to when the five year inspection and preventive maintenance Technical Specification 4.8.3.2.b surveillance was last accomplished. The licensee had difficulty in providing the documentation. All the breakers which provide power to safety related equipment had the surveillances done during the 1986 outage for Environmental Qualification (EQ)
purposes.
Documentation of these surveillances was finally produced on May 8, 1987. Documentation of the surveillance on the remaining breakers identified in the Technical Specification which are not
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required to be environmentally qualified could not be immediately provided, so the surveillance was accomplished on those breakers on May 8, 1987.
That evening, the licensee had sufficient documentation for this surveillance such that the breakers were considered operable. The licensee continued researching records to demonstrate the surveillance was current since 1982, initial license issuance, I
through 1986 when the EQ maintenance was accomplished. The results of this review will be documented in a separate inspection report (373/87017; 374/87017).
One violation was identified in the review of this functional area.
5.
Monthly Maintenance Observation (62703)
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Station maintenance activities listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, j
regulatory guides and industry codes or standards and in conformance with technical specifications.
During the inspection period, the inspector observed portions of the i
following maintenance activities:
Unit 1 Circulating Water Pump Unit 1 Condensate Pump
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i Unit 2 Control Rod Drives l
The following observations were noted:
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a.
During the Unit 2 refueling / maintenance outage the licensee removed, i
rebuilt, and replaced several control rod drives (CRD).
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The resident inspector observed portions of these tasks and found no outstanding problems. During CRD testing, a couple of problems l
were noted.
On April 17, 1987, during testing of CRD 18-15, it was noted that the CRD would not move during normal insert or withdrawal signals.
Also, there was no drive flow during the insert or withdrawal
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signals unless the vent valves on the CRD hydraulic control unit (HCU)wereopened.
Further investigation revealed that CRD 18-15
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had been installed improperly. The CRD had been misaligned
(rotated by one bolt) when bolting the CRD flange to the CRD housing flange.
One of the two alignment pins was recessed into the CRD flange. Neither alignment pin was damaged. The CRD was
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inspected, new 'O' rings installed, and the CRD reinstalled
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properly.
Further testing proved satisfactory operation of the CRD.
On April 18, 1987, CRD 30-35 would not couple with its associated control rod (CR) blade. The CR blade was removed and a camera was
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lowered into the CR blade guide tube for inspection.
The CRD uncoupling pin was found out of the CRD, laying in the bottom of the CR blade guide tube, and bent. Further inspection of the CR r
j blade revealed the locking plug was bent also.
I A new CR blade and CRD were installed at location 30-35. The CRD
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removed from location 30-35 was inspected and the inner filter was
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found cocked with a dent in the bottom of the filter. The spud
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fingers of the CRD were found in normal condition, not bent. The records for the rebuilding and installation of the CRD were reviewed.
No anomalies were found.
The root cause for the event
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J is unknown.
It is postulated that a high pressure insert signal (i.e., scram) occurred while the CRD and CR blade were uncoupled
causing damage to the uncoupling pin, locking plug and inner filter.
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It is also postulated that the uncoupling pin was pulled from the
spud by the CR blade during the removal of the CR blade.
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The uncoupling pin possibly became wedged between the locking plug i
and the bottom of the CR blade.
Further investigation has not produced any other cause(s).
On April 20, 1987, during testing at CRD 14-23, the CRD experienced i
high withdrawal drive flow and slow withdrawal movement between positions 00-04, and normal withdrawal drive flow and withdrawal j
speed between positions 04-48.
Several more tests were performed
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on the CRD with the results indicating a potential seal problem.
The CRD was rerroved and a rebuilt / spare CRD installed. Upon
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disassembly of the CRD, the " drive down" bridge seals and radial
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seals were found mispositioned. The bridge seals were located where the radial seals should have been, and vice versa, on the drive piston. The problem was corrected and the CRD was rebuilt.
Each of these problems have been addressed by the licensee and corrective actions implemented.
The licensee stressed the importance for attention to detail, acquired the services of i
GE, the manufacturer of the CRDs, for resolution of these problems,
and reviewed these events / problems with the appro
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The resident inspector has reviewed the licensee'priate departments.
s corrective actions and finds them adequate.
The resident inspector will pursue
any new or further observation / investigation of these or similar
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events through future inspection (s),
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b.
On May 13, 1987, at approximately 5:00 a.m. (CDT), the "B" control room emergency makeup ventilation system actuated due to loss of the C and D radiation monitors. The licensee opened the panel door
to the 236X-2 Motor Control Center to hang an Out Of Service (005)
tag and an 00S tag on the inside of the door caught the breaker 23
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and tripped power to the radiation monitors for the "C" train of
the control room ventilation system. This started the emergency l
makeup ventilation system for the control room.
The breaker was l
reset immediately and control room ventilation was restored. The
licensee is revising the out of service procedure to require the use of smaller 00S cards on the 120V breakers and briefing the
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operating department personnel on this event to prevent reccurrence.
The resident inspector reviewed this event and has no further
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Concerns.
No violations or deviations were identified in this area.
6.
Training (41400)
i The inspector, through discussions with personnel and a review of training records, evaluated the licensee's training program for operations and maintenance personnel to determine whether the general
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i knowledge of the individuals was sufficient for their assigned tasks.
l In the areas examined by the inspector, no significant items of concern were identified.
l The inspector did review records and interview individuals pertaining
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to the problems with the control rod drives referred to in Paragraph 5,
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Item a.
The training appeared adequate and the individuals were
knowledgeable of their responsibilities.
No violations or deviations were identified.
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7.
Licensee Event Reports (92700)
Through direct observations, discussions with licensee personnel, and I
review of records, the following Licensee Event Reports (LERs) were i
reviewed to determine that reportability requirements were fulfilled,
inmediate corrective action was accomplished, and corrective action to i
prevent recurrence had been accomplished in accordance with Technical Specifications.
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(Closed) 373/07017-00 - On April 2, 1987, with Unit 1 in Operational Condition 1 (Run) at 55% power, Primary Containment Pressure Switch
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IC71-N002A was isolated for a length of time in excess of that
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permitted by the Technical Specifications. At the time of this event.
l an instrument surveillance Was being performed on the pressure switch.
This event was documented in inspection report 373/87011; 374/87014.
The inspector had no other concerns.
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i (Closed) 374/87005-00 - At 2:13 p.m. (CST) on April 2, 1987, with Unit 2 defueled, the 2B Diesel Generator (DG) was being tested in
accordance with LTS 800-3 (DG 2B Start and Load Acceptance i
Surveillance). At this time, the 28 Diesel Generator was shutdown l
with its control switch in AUTO. The 28 DG output breaker was racked i
to REMOTE TEST and closed. The System Auxiliary Transformer (SAT)
feed breaker to BUS 243 was open and BUS 243 was deenergized.
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i Before confirming that the SAT feed breaker to BUS 243 was closed and l
BUS 243 was reenergized, the test engineer opened the 28 DG output j
breaker, in preparation for racking the breaker to CONNECT, causing the
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28 DG to auto-start. This event was documented in Inspection Report 373/87011; 374/87012. The inspector had no other concerns.
(Closed) 374/87006-00 - At 10:44 a.m. (CST) on April 2, 1987, with Unit
>I 2 defueled, Operational Analysis Department (OAD) test engineers were
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i performing functional / logic testing in the primary containment isolation circuitry that was changed as a result of a modification. During the course of the testing, an unexpected Group II isolation Engineered Safety i
Feature (ESF) actuation occurred, i
The apparent cause of this event was inattention to detail by the OAD
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test engineers. Prior to testing, the engineers did not adequately l
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review the circuitry that would be affected by the test. This event
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was documented in Inspection Report 373/87011; 374/87012. The inspector
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has no further concerns.
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(Closed) 374/87009-00 - On April 15,1987,at11:30a.m.(CDT),with Unit 2 defueled, the 2E12-F009 residual heat removal shutdown cooling
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suction inboard isolation valve received a Group VI isolation signal on high flow. The isolation signal (Engineered Safety Feature actuation)
occurred while performing a high pressure water leak rate test on the 2E12-F008 residual heat removal shutdown cooling suction outboard i
isolation valve. The isolation signal was caused by a leaky instrument root valve that is closed per the leak rate procedure to prevent a high flow isolation signal during the test. The leaky instrument root valve
allowed the high side of the 2E31-N012BA differential pressure switch
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to pressurize, which in turn actuated the switch. There were no safety
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consequences of this event because the reactor was defueled.
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I On April 16, 1987, the leak rate test procedure was revised to include steps to valve out and open pressure equalizing valves on the differential pressure switches that were previously isolated with the instrument root valves.
This is done at the instrument rack where the
instruments are located.
Steps were also included to lift leads to
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prevent an isolation signal from causing an actuation. The leak rate test was reperformed on April 17, 1987, with the revised procedure and no actuations or isolation signals occurred. The inspector finds these
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actions adequate.
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(Closed) 373/87016-00 - At 12:15 p.m. (CST), on April 1, 1987, a
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Division I Reactor Core Isolation Cooling (RCIC) high steam flow
isolation signal was received. This event occurred after the Instrument Maintenance Department had performed the Unit 1 RCIC steam line high flow isolation response time test. At the time of l
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the event, the RCIC system was inoperable because of a water leg
pump failure. Unit I was operating at 56% power. This event was
documented in Inspection Report 373/87011; 374/07012. The inspector
had no further comments.
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(Closed) 373/87019-00 - Curing Shift 1 on April 3, 1987, the Unit I Nuclear Station Operator (licensed Reactor Operator (RO)) started the
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weekly control rod exercising surveillance in accordance with LOS-AA-W1, Technical Specifications Weekly Surve111ances.
This event was documented j
in this inspection report (see Paragraph 4). The inspector has no concerns other than those addressed in Paragraph 4.
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(Closed) 373/87013-00 - On March 12,1987, at 5:30 p.m. (CST), with j
Unit 1 in RUN at 56% power, a high drywell temperature event began on a trip of the "B" primary containment fan. Attempts to restart the fan
were unsuccessful and a unit shutdown commenced at 9:45 p.m. to comply i
with Technical Specification 3.6.1.7.
Indication for the bulk average
temperature was abnormal due to irregular ventilation flow backwards through the open discharge dampers of the tripped fan. During the event, i
the solenoids for Safety Relief Valves A. C (Automatic Depressurization System valve), and li were declared inoperable per Technical Specification 3.7.7.b.
A GSEP Unusual Event was declared at 1:30 a.m. on March 13 1987, due to the inability to confirm drywell temperature less than 135
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degrees F within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. After the discharge dampers were closed, t.ormal system flow paths were re-established and the GSEp terminated.
This event was documented in Inspection Report 373/87011; 374/87012.
The inspector had no further concerns other than those identified in
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that report.
(Closed) 373/87020-00 - On April 16,1987, at 10:29 a.m. (CDT) with Unit 1 operating at 58% power, the control room received alarms and
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indications that indicated loss of 120/208 volt power at Motor Control Center (MCC)136X-1. The 120/200 volt distribution transformer on
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136X-1 feeds Division 11 IIVAC and containment monitoring system
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equipment. An oaerator was dispatched to the MCC and discovered that i
the 480 VAC brca rer that feeds the 120/2C8 volt distribution transformer on 136X-1 was tripped off.
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At the time of the event, contractors were painting near the 136X-1 MCC and had hung a drop cloth to prevert paint overspray from enturing the
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l MCC. The drop cloth inadvertently came loose and slid down the front of the MCC, bumping the breaker and tripping it off. The safety
consequences of this event were minimal. All of the Division II loads
that were lost have redundant backup systems on Division I, and Division i
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I was operable. The contractors that were painting were informed that l
extra care is required when working around MCCs and, specifically in this case, to ensure that the drop cloth is properly secured.
It was i
i also emphasized that extra care must be taken when installing and i
removing drop cloths on MCCs.
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i The inspector inspected the area pertaining to the event and reviewed
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the licensee's actions and finds them adequate.
i (Closed) 373/87015-00 - At 8:28 p.m. (CST) on March 28, 1987, with Unit 1 in Operational Condition 1 (RUN) at 57% power, Reactor Core
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Isolation Cooling (RCIC) water leg pump tripped on breaker thermal
overloads and could not be restarted.
RCIC was declared inoperable l
in accordance with Technical Specification 3.7.3.
Consequences of i
this event were minimal due to the fact that the High Pressure Core
Spray system was operable in compliance with Action Statement "b" of i
Technical Specification 3.7.3.
The root cause of the pump failure
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was due to a procedural inadequacy of an operating surveillance i
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which was being performed at the time.
The RCIC water leg pump was replaced and the RCIC system was declared operable on April 2,1987,
a t 2:10 p.m.. LOS-RI-Q3 was revised (revision 9) to ensure that the RCIC water leg pump is shut off prior to closing valve 1E51-F010
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(RCIC pump suction from CST stop valve). The new pump was tested in
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accordance with the applicable sections of LOS-RI-Q3.
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installed water leg pump has an outboard bearing housing design which l
does not require a retainer / snap ring. All RCIC procedures and i
surveillances will be reviewed to determine if any procedure revisions
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are necessary to ensure that the water leg pump will not be operated
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withthesuctionvalve(1E51-F010) closed.
The inspector finds these actions adequate.
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(Closed) 373/87012-00 - On March 11, 1987, with Unit 1 in Condition 1
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(RUN) and Unit 2 defueled, the "A" control room HVAC system (VC) ammonia
detector tripped. An Engineered Safety Feature (ESF) damper actuation
occurred and the minimum outside air dampers went to the closed position j
while the "A" VC emergency make-up unit started. No systems or components were inoperable at the beginning of the event which con-
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I tributed to this event.
The Instrument Maintenance Department i
investigated the event and found that the chemcassette tape was broken j
in the makeup spool. The chemcassette tape was readjusted and the J
detector was reset. No further actions were required.
(Closed) 373/87014-00-At1:52p.m.(CST)onMarch 19, 1987, with Unit 1 i
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in Operational Condition 1 (Run) at 54% power, a reactor scram occurred j
duetoageneratorlockoutandsubsequentturbinetrip(turbinestop
valveclosure).
The generator lockout was caused by an electrical fault
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that occurred on the non-segregated bus ducts feeding the 6.9 and 4.1 KV
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loads from the Unit Auxiliary Transformer (UAT) TR-141.
l This event was documented in Inssection Resort 373/87011. The licensee
will replace the 6.9 and 4.1 KV aus bars, aus ducts, and wall penetra-i tions damaged by the fault.
The licensco is developing a wall
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penetration for, susceptible to cracking and contamination caused
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)y moisture intrusion.
The resident inspector finds these actions q
adequate.
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l (Closed) 374/87003-00 - On March 5, 1987, at 9:00 p.m. (CST), with Unit 2 defueled, the Unit 2 control room continuous conductivity
indication was inoperable for greater than 31 days.
This exceeds the l
timeclock as stated in Technical Specification 4.4.4.c.
The monitor was inoperable because both water supplies to the monitor had been isolated for maintenance work during the refuel outage.
The Chemistry Department was taking dip samples of the reactor cavity every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (depending on conductivity analysis), from the itme of the loss of continuous monitoring until the continuous monitoring system was
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returned to service. The Technical Specification allowed dip sam) ling
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for up to 31 days and did not identify additional action to be ta(en r
after that time. The inspector has reviewed this item and finds the
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licensee's actions adequate.
i (Closed) 373/87010-00 - On March 2, 1987, at 8:45 a.m. (CST), during the
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performance of LIS-NB-106, " Reactor Vessel Low Level Confirmed Automatic
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l Depressurization System Calibration," Level Switch LS-1821-N0388 was found with a setpoint out of tolerance and in excess of the reject limit for its application. However, the Limiting Condition for Operation limit
was not exceeded. The cause of this event was setpoint drift. Level Switch LS-1821-NO388 was replaced and the new switch was recalibrated to within allowable tolerances and placed in service. The inspector
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reviewed the licensee's actions on this item and finds them adequate.
(Closed) 374/07008-00 - At 4:00 p.m. (CDT) on April 14, 1987, with Unit 2
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at 0% power in a defueled condition, a reactor scram occurred when the made selector switch was moved from the shutdown position to the refuel position. The cause for the scram was three blown fuses in the control
rod drive charging header low pressure scram circuit. The root cause of
the blown fuses is believed to be an unintentional grounding of a portion of the scram circuitry and coil surge from intermittent contact while
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i removing lugged jumpers.
This could have occurred during a routine
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surveillance performed earlier that day.
The blown fuses were replaced, the modo switch was again switched from the shutdown position to the refuel position, and no further scrams occurred. Corrective action for this event will include procedure revisions to require checking fuses
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for continuity prior to surveillance completion. The inspector has i
reviewed the corrective actions and finds them adequate.
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Refueling Activities (60710)
J The licensee comenced refueling activities on April 26, 1987, and i
activities were completed by May 3, 1987.
During this time, the resident
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i inspector witnessed those activities both in the control room and on i
the refuel floor. Communications between the control room and refuel floor were very good.
Fuel movements were completed without error or complications.
The inspector reviewed the applicable procedures in use, ensured that the radiation controls established were adhered to, and that the refueling support systems were not rendered inoperable as a result of modifications, surveillances, or maintenance activities. Also, i
the Technical Specification requirements necessary for refueling were i
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reviewed and verified.
The inspector observed that the planning,
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coordinatin l
Very well. g, and execution of the refueling activities were performed
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No violations or deviations were identified in this area.
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Facility Modifications (37701)
The inspector verified that the modifications conform with industry codes
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and standards, applicable regulatory guides and the licensee's Quality Assurance (QA) Program. The inspector verified that the modified system was installed in accordance with the approved design by observing work j
in progress, reviewing related portions of the licensee's QA program and examining records of installation, inspection, and testing.
During this inspection period, the inspectors reviewed portions of modification
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1-2-84-048, which replaced ten containment isolation valves on the I
primary containment vent and purge system.
l No violations or deviations were identified in this area.
10. Local Leak Rate Testing (61720)
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The inspector observed testing, reviewed records and independently
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i verified calculations concerning the local leak rate program.
The i
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inspector independently verified the acceptability of the test results.
The inspector verified that penetration boundaries and isolation valves had been local leak rate tested at the required frequency since the i
l previous integrated leak rate test.
The sum of the local leak rates for all boundaries and valves subjected to local Icak rate tests met the acceptance criteria.
The local leak rate procedures were reviewed and found to utilize approved methods for testing penetration boundaries and isolation valves. The test equipment utilized during the local leak
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i rate tests observed was within calibration.
During this inspection period, the inspector observed the following valves being local leak
rate tested:
2VQO26 2VQO30 i
2VQO27 2VQ042
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2VQ029 2VQ043
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No violations or deviations were identified in this area.
11. Emergency Exercise (82301)
On April 29, 1987, the station conducted an annual exercise of the Generating Stations Emergency Plan (GSEP).
The exercise was observed
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by NRC Region !!! personnel and a resident inspector. The licensee manned the control room, the Technical Support Center (TSC), the Emergency Of fsite Facility (E0F), and the corporate headquarters center.
l The results of this exercise are documented in inspection Report 373/87014; 374/87014.
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No violations or deviations were identified in this area.
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12. Exit Interview (30703)
i 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 inspectors also discussed
j the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the
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inspection. The licensee did not identify any such documents or processes as proprietary, i
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