IR 05000416/1990011
| ML20044B118 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/02/1990 |
| From: | Cantrell F, Christensen H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20044B117 | List: |
| References | |
| 50-416-90-11, NUDOCS 9007170334 | |
| Download: ML20044B118 (10) | |
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NUCLEAR CE:ULATCRY COMM18410N -
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101 MARIETTA STREET, N.W.
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ATLANTA, OEORGI A 30323 l,
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9 Report No.:- 50-416/90-11 F,
Licensee:
Entergy Operations Inc.-
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Jackson, MS 39205 r-
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50-416 License No.: NPF-29 M
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Facility.Name: Grand Gulf Nuclear Station
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Inspection Conducted: May 19, 1990 to June 15, 1990
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Inspectors:
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. O. Christensen, Senior Resident Inspector Date Signed j
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p. L. Mathis, Resident Inspector
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1, Approved by:
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F. 5. Cantrell, SectioYi CM4f Date' Signed
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Reactor Projects Branch V
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t-Division of Reactor Projects
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SUMMARY i
Scope:r
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The[residentt inspectors conducted a routine inspection in the areas of
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o)erational safety-verification; maintenance observation; surveillance
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s o)servation; : installation and testing of modifications; action on previous inspection findings; and reportable occurrences.
The inspectors-conducted
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backshift inspections on May-23, 28 and June 12, 1990.
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.Results:
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n One.NRCidentifiednon-citedviolation-(NCV)wasidentifiedduringthisinspec-
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tion ' period.. An operator failure to follow procedures while racking-out,the
.LPCS > ump breaker which inadvertently started the pump, paragraph 3.
This NCY
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and tie recent clearance tagging problems.. paragraph 3, may be indications of
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complacency-'in routine plant operations. 'However, the licensee can and does perform complicated tasks in a controlled manner, this was demonstrated by the
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backfilling of the reactor vessel reference leg.. paragraph 4.
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.In the inspection areas of maintenance observation, surveillance observation, L
'and installation and testing of modifications,'the' licensee met the safety
objectives of these areas.
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l 9007170334 900702 PDR ADOCK 05000416 O
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REPORT DETAILS 1.
Persons Contacted Licensee Employees J. G. Cesare, Director, Nuclear Licensing W. T. Cottle, Vice President, Nuclear Operations D. G. Cupstid, Manager, Plant Projects
- L. F. Daughtery, Compliance Supervisor M. A. Dietrich, Director. Quality Programs
- J. P. Dimette, Manager, Plant Maintenance
- C W. E11saesser, Operations Superintendent C. R. Hutchinson, G M General Manager F. K. Mangan, Director, i : cit Projects and Support L. B. Moulder, Acting Manager, Plant Support
- J. V.:Parrish, Manager, Plant.0perations J. C. Roberts, Manager, Plant & System Engineering S. F. Tanner, Manager. Quality Services F. W. Titus Director, Nuclear Plant Engineering M. J. Wright, Manager, Nuclear Training G. W. Vining, Manager, Plant Modification and Construction
- G. Zinke, Superintendent, Plant Licensing Other licensee employees contacted included superintendents, supervisors, technicians, operators, security force members, and office personnel.
- Attended exit interview
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F. Cantrell, Section Chief Division of Reactor Projector, Region II, was on site June 11 and 12, 1990, to tour the site and conduct discussions P
with the resident inspectors and plant management.
2.
Plant Status The plant began and ended the inspection period in mode one, power opera-tions. On June 6,1990, the control and performance of licensed activities for Grand Gulf Nuclear Station was transferred to Entergy Operations, Inc.
3.
Operational Safety (71707, 93702)
The inspectors were aware of the overall plant status, and of any significant safety matters related to plant operations. Daily discussions
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were held with plant management and various members of the plant operating
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staff.
The inspectors made frequent visits to the control room.
The
observations. included:
the verification of instrument readings, setpoints j
and. recordings; the review of operating system status and equipment
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tagging controls; the verification of annunciator alarms, limiting i
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conditions for operation, and temporary alterations; and the review of daily journals, data sheet entries, control room manning, and access controls.
Weekly, selected engineered safety feature (ESF) systems were confirmed operable.
The inspectors verified that accessible valve flow path alignment was correct, power supply breaker and fuse status was correct and instrumentation was operational.
The Grand Gulf probabilistic risk assessment-based inspection plan was used to verify the following systems operable:
RCIC, LPCI A, B and C, and SSW A, B and C.
The inspectors conducted plant tours weekly. Portions of the control building, turbine building, auxiliary building and outside areas were visited.
The observations included safety related tagout verifications,
,f shift turnovers, sampling programs, housekeeping and general plant
conditions.
Additionally, the inspectors observed the status of fire protection equipment, the control of activities in progress, the problem identification systems, and the readiness of the onsite emergency response facilities.
The inspectors observed health physics managements involvement and awarness of significant plant activities, and observed plant radiation controls.
Periodically the inspectors verified the adequacy of physical security control.
Additionally, senior plant management was observed making routine tours of the plant.
The inspectors reviewed safety related tagouts, 892561(ADHRSystem)
and 900033 (IRM A and C), to ensure that the tegouts were properly.
prepared, and performed.
Additionally, the inspectors verified that the tagged components were in the required position.
During this inspection period, three separate events dealing with protective tagging occurred.
The first concerned the incorrect tagging of an unlabeled air valve;. the second was the failure to provide an adequate system isolation; and the third was the failure of operations to verify equipment posi.tions prior to issuing the clearance for work.
The licensee issued QDRs - to track-and implement corrective actions.
These events are continuing indications of a lack of attention by the operator or a lack of understanding on the primary function of a clearance tag.
Operations management issued night orders on the importance of the protective tagging system and conducted discussion with the shift superintendent.
The inspectors reviewed the activities associated with the events listed below.
On May 23,1990, the shift supervisor noticed a utility knife on the frisker cart at the entrance of the control room.
The knife was frisked and determined to be contaminated.
Health Physics (HP) was notified and
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the area surveyed.
No other contamination was detected.
A radiological
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deficiency report documented the occurrence and recommended the following l
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corrective actions, conduct training for the operation's department on the importance of ensuring all tools exit the RCA through the 93 foot turbine control point and that HP technicians inform the HP supervisor of radiological problems.
Routine surveys of the control room and other
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Non-RCA areas over the past year have found no other cases of contaminated tools in Non-RCA areas.
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On May 26, 1990, the low pressure core spray (LPCS) pump inadvertently
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started when an operator incorrectly racked-out the LPCS pump breaker.
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The pump was being taken out of service to allow post modification
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testing on the LPCS suppression pool. return valve (E21F012).
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operator pulled the control power fuses, de-energized the breakers charging motor, but then, discharged the closing springs before racking out the breaker.
The LPCS pump and division one SSW started. -The operator tripped the breaker and secured ths SSW system.
No LPCS system r
injection occurred.
After conducting discussions with the resident inspector, the licensee made a 10 CFR 50.72 (b)(2) (ii) report.
General Operating Instruction 04-S-04-2, Operation of Electrical Circuit Breakers, step 4.3, contains the breaker rack-out instructions.
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failure of the operator to use procedure 04-S-04-2 is a violation of T.S.6.8.1, which states, written procedures shall be established, implemented and maintained.
This NRC identified violation is not being cited because criteria specified in section V. A of NRC Enforcement Policy-were satisfied (NCV 90-11-02).
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On June 11, 1990, during the restoration of suppression pool cooling mode of RHR, one of three thermal overloads for valve E12F048B completely melted l.
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at Motor Control Center (MCC) 16B31 when the valve was stroked open.
Preliminary indications is that the opening contacts, 42F device, failed i
to open resulting in the actuation of the thermal overload.
Limit condition for operation 90-893 was entered and the problem was investigated under MNCR 061-90.
The thermal overloads and contact assembly were replaced and the valve stroked tested satisfactorily.
Nuclear plant engineering evaluated possible damage to the valve bonnet
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key or other weak link components and actuator assembly, NPE calculated the yield strength of the backseat limiting component.
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which is the lower bonnet / lower bonnet bushing interface.
The calcula-
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tions indicate that minor localized yielding was possible.
However, NPE l
believes that the bushing backing partially or fully out of the lower bonnet would be unlikely due to the contact surface.
NPE concluded in
their evaluation that the function of the valve had not been impaired, however, extended operation with flow (especially throttled flow), could
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create vibrations which might tend to loosen the lower bonnet bushing if i
the tack weld has been broken.
Plant personnel were made aware that u
persistent packing leakage may be the result of bushing seal' problems.
Due to this condition and the magnitude of the load experienced by the actuator, a thorough inspection of the valve and actuator is required in RF04.
This will be followup by the inspector and identified as inspector followup item 90-11-03.
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4.
Maintanance Observation (62703)
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During the report period, the inspectors observed portions of the maintenance activities listed below.
The observations included a review of the MW0s and other related documents for adequacy; adherence to procedure, proper tagouts, technical specifications, quality controls, and radiological controls; and the observation of work and/or retesting.
MWO DESCRIPTION E11499 Calibrate Division 2 D/G DC voltmeter.
E11502 Calibrate Division 2 D/G tachometer.
109441 Calibrate Division 2 D/G synchronous speed pressure switch.
111630 Calibrate Division 2 D/G bearing temperature trip pressure switch.
111872 Backfill reactor vessel level transmitter reference leg, channel A.
M08159 Rework Division 2 D/G jacket water leaks.
M11704 Change oil and filters in CRD pump and speed increaser gear box.
During the calibration of.the division 2 diesel generator pressure switches, on May 23, 1990, the ~ inspector noted the poor labeling of components inside panel P401.
This concern was addressed to the 1 & C-Superintendent.
Additionally, during the performance of MWO 111630, calibrate division 2 D/G bearing temperature pressure switch (P75N0328),
the technician valved. out the wrong pressure switch.
Pressure switch P75N093B was valved out, it was labeled PS32B, Pressure switch P75N032B was labeled PS150.
The MWO contained information that identified PS150 as pressure switch P75N032B, The technician realized the error and performed the calibration on the correct pressure switch.
A quality deficiency report was issued to document and correct the problem.
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licensee committed to relabelling the components inside D/G panel P400 l
and P401.
This is inspector followup item 90-11-01.
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On May 25, 1990, the inspectors observed the ) reparation and performance of MW0 11872, backfill reactor vessel condens ng pot reference leg D004A.
The backfill was recuired to correct the channel A reactor vessel water level instrument dr' ft.
The channel had failed its channel check, however, the instrument was in tolerance.
The licensee stated that during long periods of plant operations, non condensable gases accumulate
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in the reference leg condensing pots and if there is a small leak in any instrument lines the water level in the condensing pot will decrease..
Back filling the reference leg will restore the water level in the condensing pot.
The evolution was well planned. Operations reviewed and issued 20 LCOs on the effected reactor instrumentations. The coordination between 1&C and operation ensured the work was performed in a controlled manner.
During the backfill, I&C identified a leaking instrument root
valve that was repaired.
The reactor vessel level instrumentation was
returned within the channel check band.
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No violations or deviations were identified.
The results of the inspec-tion in this area indicate that the maintenance program was effective with respect to meeting safety objectives.
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5.
Surveillance Observation (61726)
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The inspectors observed the performance of portions of the surveillances listed below.
The observation included a review of the procedures for technical adequacy. conformance to technical specifications and LCOs; verification of test instrument calibration; observation of the actual surveillances; removal and return to service of the system or component; and review of the data for acceptability based upon the acceptance
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criteria.
06-IC-1821-M-1001 Safety / Relief Valve High Pressure Trip / Low Low Set Relief /ECCS Vessel Pressure Injection
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Permissive Functional Test, Channel B.
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06-IC-1821-M-1002 ReactorVesselHigh/LowPressure(RPS/RHRISOL)
Functional Test.
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06-IC-1821-M-2005 Main Steam Low Pressure Functional Test, I
Channel D.
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06-IC-1C71-M-2002 Turbine Stop Valve Trip Fluid Low Pressure t
Functional Test, Channel H.
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06-IC-1E32-M-1001 MSIV Leakage Control System Pressure Functional Test.
06-IC-1E61-M-1004 Containment and Drywell Hydrogen Analyzer A Calibration.
On May 26, 1990, during the performance of surveillance 06-IC-1E61-M-1004, Containment and Drywell Hydrogen Analyzer Calibration, the I&C technician noted that the hydrogen analyzers voltage to current (E/I) converter required a calibration accuracy greater than the accuracy of the digital
volt meter, Fluke 8600A, specified for use in the procedure.
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divisions of the containment and drywell hydtogen analyzers were declared inoperable.
Another digital volt meter, Fluke 852A, was used to perform o
the surveillance and the analyzers were returned to service later that day. The licensee is investigating the requirements for the E/I converter
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accuracy and will revise the surveillance procedure depending on the
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results. This issue was documented in QDR 90-473.
T No violations or deviations were identified.
The results of the s
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inspection in this area indicate that the surveillance program was l
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(ffective with respect to meeting the safety objective.
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6.
Installation and Testing of Modifications (37828)
The inspectors examined selected modifications to verify that work was
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performed by qualified workers and according to approved instructions, procedures and drawings.
Additionally, the inspectors reviewed the
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post-modification testing to ensure that the test adequately addressed s
the modification and established appropriate acceptance criteria.
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On April 2, 1990, the licensee identified a motor operated valve, E21F012, LPCS test return valve, that may have delf vered a thrust force exceeding the calculated thrust value for the valves limiting component,
the key bushing set screws.
This condition was documented in NRC inspection report 90-06.
The licensee implemented the repair under MWO-11533 and documented the modification in accordance with the MNCR process.
To increase the effective strength of the valve's limiting component, the two existing 5/8 inch set screws were replaced with two j
3/4 inch high strength set screws.
The two 3/4 inch set screws were
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machined from one inch diameter, ASME SA-193, grade B7, studs.
The modification was performed in accordance with the work order and the valve was operationally tested using surveillance 06-0P-1E21-Q-0002, LPCS
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M0V functional test.
The valve successfully passed the stroke time test
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and was returned to operation.
l The installation and testing of the new key brushing set screws was
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adequately performed and documented.
A few minor administrative
_ deficiencies were noted with the work package.
A quality deficiency L
report was issued by the licensee to correct these items.
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No violations or deviations were identified.
7.
~ Reportable Occurrences (90712 & 92700)
o The event reports listed below were reviewed to determine if the information provided met the NRC reporting requirements.
The determination included adequacy of event description, the corrective i
action taken or. planned, the existence of potential generic problems and the relative safety significance of each event. The inspectors used the NRC enforcement guidance to determine if the event met the criterion for licensee identified violation [
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[(Closed) LER 89-16, Lightning Induced Spikes Causes APRM Scram.
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licensee installed a lightning dissipation system.
This system reduces the potential for lightning strikes at the plant site.
Since-the i
lightning _ dissipation system has been installed several. severe thunder storms have past through the-area without causing a lightning induced scram. This item is closed.
(Closed) LER 89-15, Failure to Retest Isolation Dampers Following Maintenance..This event was documented in NRC inspection report 89-28'
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and-a licensee identified violation was cited.
The licensee has
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'imN etaented a computer maintenance work order system that address retest
control. This item is closed.
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(Closed)_ LER 89-12, Reactor Scram Due to Condenser Expansion Joint
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Failure.. This event was documented in NRC inspection report 89-19. The corrective actions listed in the LER were completed.
This item is y'
closed.
(Closed) LER 90-06, Effluent Sample Analysis Exceeds Time Limit Due to
Personnel Error.
This event was discussed in NRC inspection. report
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90-08, paragraph 6.
Non-cited violation 90-08-02'was documented.
This'
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- On May121, 1990,- the ' licensee' identified six fire doors that were TS required-but wre not being surveyed as TS fire. doors.
They were inspected as insuranco required fire door, however, a LCO.would not have been entered if the doors were inoperable. An additional.four doors'were
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discovared that were not inspected:by any procedure.
The licensee is
con. ducting a review'to determine'if additior. fire rated assemblies are
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not being surveyed.
The results of this -red 4 and' the corrective actions will be-tracked under the issued LER.-
No violations-or deviations were identified.
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Action on Previous Inspection Findings (92701,92702)
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(Closed) Unresolved Item 90-02-03, Evaluate the adequacy of the long term cooling capabilities for HPCS SSW system. A NRC memorandum dated May-17,
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1990, from the Reactor Systems Branch, NRR, to Project Directorate 11-1 states that the licensee revised -LOCA analysis appears reasonable.and-indicates tnat cladding temperature.and oxidation limits would not be.
li axceeded.
The licensee's analysis supports justification 'for continued
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0; : rations in the interim until appropriate design modifications are
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. implemented. ;The licensee is preparing a design modification, which will l
be installed during refueling outage 4.
This item is closed.
(Closed) Inspector Followup Item 90-03-04, Form 16.10(c) for CAR 2214 was
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missing a " reviewed by" signature.
The licensee identified four additional examples.
The _ licensee conducted a review of all CAR
closeouts over the past 24 months and corrected the four identified
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Additionally, QAP 16.10 was revised to provide clearer
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instruction on the CAR review process. This item is closed.
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(Closed) Inspector Followup Item 90-06-03, Division 1 D/G ESF. Walkdown items.- GGNS has an established labeling program to label systems / components
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by safety system priority. AECM 89/0100 outlines the program and determined this priority.
It -is estimated GGNS will complete the safety system
'other' labeling-by December 1991. This item is closed.
9.
Exit Interview (30703)
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-The inspection scope and findings were sunnarized,on June 14, 1990, with those persons indicated in paragraph 1 above.- The licensee did:not j
identify as proprietary any of the materials provided to or reviewed by
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the inspectors during this inspection.
The' licensee had no connent on-
the following inspection findings:
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Item Number Description and Reference
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'IFI 90-11-01 Inspector Followup Item - Relabel the components-
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in D/G panel P400 and P401, paragraph 4.
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NCV 90-11-02 Noncited Violation - Failure to follow procedure,
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results in LPCS pump start. paragraph 3.
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IFI 90-11-03 Inspector Followup Item - Review inspection on valve E12F048B.-
10. Acronyms and Initialisms q
Boiling Water Reactor
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DCP:
Design Change Package
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Diesel Generator DG
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ECCS -
Emergency Core. Cooling System
.j ESF Engineering Safety Feature
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'I&C Instruruentation and Control
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. Inspector Followup Item-
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Limiting Condition for Operation
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LER Licensee' Event Report j
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LPCI -
Low Pressure Core Injection
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LPCS -
Low Pressure Core Spray MNCR -
Material Nonconformance Report i
MWO Maintenance Work 10rder ti
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Nuclear Plant Engineering NPE
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Nuclear Regulatory Commission
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NRC
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Pressure Differential Switch PDS
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PSWi -
Plant Service Water
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JQuality Deficiency Report QDR
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' Reactor Core Isolation Cooling i
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t Reactor Protection System EC RPS
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Radiation Work Permit
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SSW. -
Standby Service Water Technical: Specification TS
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