IR 05000416/1990015
| ML20059L402 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 09/04/1990 |
| From: | Cantrell F, Christensen H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20059L394 | List: |
| References | |
| 50-416-90-15, NUDOCS 9009260282 | |
| Download: ML20059L402 (11) | |
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UNITED STATES --
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- Report.No.:
50-416/90-15-m
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tLicensee': 2 Entergy Operations,.Inc.
Jackson, MS 39205 s
Docket-No.:
50-416 License No.:
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, Facility Name:. Grand Gulf Nuclear Station
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Inspection Conducted: July,21th,rfughAugust 17, 1990
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h Irispectors: h kt 9d H. O. Christensen, Se;n orfesident inspector
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b 9)L/ !h b J.'L.'Mithi s, ResidenY In ~ esttir~
D6td Signed
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. Approved by:
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Fti Cintrell,m Sectidn 9Mef
'DiteM igned Reactor Projects Branch'1
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Division of Reactor Projects
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SUMMARY s
- Scope:
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The resident inspectors conducted a routine inspection in the following: areas:
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operational, safety. verification; maintenance' observation; surveillance observa-tion;--action on previous inspection -findings; and! reportable' occurrences..
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!The inspectors conducted backshift inspections on* July-26, 29,.31 and-August 16, 1990.
- Results:
' During '. thi s inspection two violations and one non-cited violation;wasi identifieo.
The first ' violation ' with four examples was for inadequate >
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procedure or - f ailure to follow procedures.
The second violation. was ' for.-..
inadequate corrective action on the main steam isolation' valve leakage control'
system. Both violations do not appear to be programmatic problems; however,;
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the. plant staff needs to' increase their attention to detail. This may indicate
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^a-negative; trend in plant performance. The licenseeiidentified violation is for failure.to perform required' post-maintenance testing.
During;this period the plant scrammed due to a feedwater transient,+ paragraph 3.
During.a main steam isolation valve test, inboard isolation valve F0220 failed.
to close, paragraph 6.
Additionally, the licensee reported that the division
. three DC load profile was inaccurate and that certain DC circuits may not
. receive the required operatir.g voltage, paragraph 6.
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9009260282 900906 AEOCKOSOOg6 PDR o
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l REPORT DETAILS
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. Persons Contacted
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Licensee Employees j
- J. G. Cesare,- Director, Nuclear Licensing
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W. T.~ Cottle,' Vice President, Nuclear Operations D. G. Cupstid. Manager, Plant Projects l
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L. F. Daughtery, Compliance Supervisor M. A. Dietrich, Director, Quality Programs l
- J. P Dininette, Manager, Plant Maintenance
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C. W. Elisaesser, Operations Superintendent
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C. R. Hutchinson,-GGNS. General Manager F. K,Mangan, Director, Plant Projects and Support
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- L. B. Moulder, Acting Manager, Plant Support
- J. V.J Parrish.. Manager, Plant Operations
- J. C. Roberts, Manager, Plant & System Engineering
- J. E. Reaves, Manager, Quality Services
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F. W. Titus, Director, Nuclear Plant Engineering
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, G. W. Vining', Manager,_ Plant Modification and Construction.
- G. Zinke, Superintendent,-Plant Licensing
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Other licensee employees contacted included superintendents, supervisors,-
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technicians, operators. security force members, and office personnel ~.
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l Attended' exit intervi.ew
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- P1 ant Status cThe; plant operated in mode one, power operations during this inspection "
period.
0n July 24,1990, the -unit scrammed due to a failure-of the
reactor feedwater ~ control system and returned to power _ operation-on
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July 26, 1990.
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Operational Safety (71707, 93702)
The? inspectors - were aware of the overall plant status, and of any
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l significant-safety matters related to plant operations.
Daily discussions were held with plant management and various members. of:the plant ' operating, staff.
The inspectors made frequent visits to the control roo'm.'
Observations included:
the verification of instrument y
readings',- setpoints and recordings; the review of operating system
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status and the tagging of equipment; the verification of ' annunciator
- alarms, the limiting conditions for operation, and the ~ temporary alterations; and the review of daily journals, data sheet entries, control room manning, and access controls.
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Weekly-selected engineered-safety feature '(ESF) systems were confirmed operable.
The inspectors verified that accessible valve flow = path j
alignment was correct, power supply breaker and' fuse status was correct-J and instrumentation was operational.
The inspectors verified the following systems operable:. ADS, SLCS'and LPCS.
The inspectors - conducted plant tours-weekly. Portions of the control i
building, turbine building, auxiliary building and outside areas were j
visited. ~ The observations included safety related tagout verifications, l
shift turnovers, sampling programs, housekeeping and general plant.
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conditions.
Additionally, the inspectors observed the status of. fire j
protection equipment -the. control of activities in progress, the problem
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identification systems, and: the readiness of the :onsite emerpancy response facilities.
The inspectors observed health-physics managements involvement and awareness of significant plant activities, and observed plant radiation controls.
Periodically the inspectors verified the adequacy of physical
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security-control.
Additionally,. senior plant management was observed making routine tours of the plant.
The inspectors reviewed safety related tagout, 901220(Mechanical vacuum pump, N6200010) to-ensure.that the tagout was properly prepared, and performed.
Additionally, the inspectors verified that the tagged components were in the-required position.
The inspectors reviewed the activities associated with the events listed below.
On July 20, 1990,'during the performance of surveillance 06-IC-C34-M-001,
Reactor Vessel Water Level (Level 8) Main Turbine / Reactor Feed Pump.
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Turbine Trip -Functional. Test, an electrical arc occurred in the panel where I&C technicians: were working.
Both reactor feed pump turbinas a
(RFPT) ran-back, reducing feed water.. flow. A RFPT signal. failure occurred-i and RFPT 'B' was shifted to manual control.
RFPT 'A'
continued to backdown; Reactor water level decreased to approximately 17 inches.
The reactor operator placed the master feedwater control system to manual and restored J
reactor level.
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L The licensee determined that the I&C technician inadvertently drew. an i
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arc from terminal 2.
He was attempting to connect a transmation
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calibration unit to terminal 4.
Procedure 06-IC-C34-M-001, step 5.20, L'
requires the. technician to place non-conducting tape over terminals 1-4 and 7-14.
In procedure step 5.22, the technician is: required to connect
the transmation units 'to terminals 5 and 4 and terminals 6 and 4.
During
this process the tape prevented direct access to terminal 4 ll I
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TS-6.8.1, requires written. procedure-shall be~ established, implemented -
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. placing tape ; over terminals 1-4,: the required connectionc point was
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and maintained.
Procedure 06-IC-C34-M-001 was inadequate in 7that by
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obstructed from easy access.
This is one example of a violation of TS l
6.8.1.(90-15-01).
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N On July 24, 1990, et approximately 1:35 p.m..-the plant scrammed on high reactor vessel water level, 53.4 inches, due to a malfunction in the.'B'
reactor feed pump controller.
The reactor was operating. at'100% power.
when the B0P' computer alarmed due to high RFPT ' speed rate of change for t
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both pumps.
The operator noticed that the FW controller was calling for j
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= maximum feed flow demand however, the feed pumps were not-responding to i"
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The control room operator then placed the master feedwater
controller in manual in an attempt to stabilize reactor water level. The m
water level decreased-and a half scram on low reactor water level
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occurred.
RCIC was manually initiated to increase water. level.
Reactor.
vessel water level then stabilized at the normal level of 36 inches and
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the half scram was reset.
The 'B' feed pump continued to oscillate so a power. reduction was made using recirculation FCV.--As power was decreasing a reactor scram occurred on high water level.
Level was restored using r
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' A' feed pump on the startup level controller.- Suppression pool cooling l
.'A' was then placed in service and the scram was reset.
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I RFPT Electronic Automatic Positioner (EAP) controller being 'out of H
calibration.
The controller was observed to have a deadband of
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approximately 6% which was outside of the vendor recommended setting of 1 to 2 percent.
I&C. technicians ' adjusted the deadband and made other y-tuning adjustments on the 'EAP controller under. work order 17059.
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L EAP controller was tested at 25 percent power.
The results of the test L
' indicated.that the feed pump response was satisfactory.
After reviewing a
b the data the PSRC approved two RFPT operations.
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- At approximately 55% power the B RFPT was placed in service.
I&C along.
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with. operation noticed abnormal controller characteristics.
Further
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o investigation determined that the Linear Variable Differential 1:
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Trensformers ~(LVDT) control circuit was malfunctioning.. The-LVDT and
'the associated circuits were repaired under work < order 93033.
After repair the. system. functioned satisfactorily.
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On July 25, 1990, during the shutdown the plant scrammed when I&C technicians were working on the A and C IRMs.
The reactor was at zero
power -with reactor pressure and temperature at 532 lbs and 476 F
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respectively.
Following the first scram which occurred on July 24, 1990, the subsequent operator actions were being performed according to ONEP 05-1-02-I-1, Reactor Scram.
These actions included driving in all IRMs.
IRM A and C were inoperable and had been red tagged in the control room i
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in, ths full-out position.
The. operator ' thought he had selected IRM B -
and D which =left IRM A and C in the full out position.- However, the-operator inserted IRM C and D.
When maintenance personnel entered the
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area under-the reactor vessel to performed work on the: A and.C detectors
they assumed that the two detectors withdrawn were A and C.
However, in actuality A and B' detectors were withdrawn.
When'the B detector was disconnected (which maintenance personnel had assumed to be the C detector) a tripi signal occurred on RPS division B. -- A half scram was
0 already inse ed on RPS ' A' as required by TS for having less than -3 operable IRM; a channel.
The operator contends that the red tag used to tag the IRM switches were j
too large and obstructed the operator view.
Administrative Procedure P
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01-S-06-1, Protective Tagging, was inadequate in that no provisions were J
provided for using protective -covers on push button type switches
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similiar to the IRMs switches or similar devices that would prevent -
obscuring adjacent component.
This is another example of violation of--
TS 6.8.1 for.having inadequate procedure for tagging out switches similiar
.2 to the IRMs switches (90-15-01).
Startup from the forced outage occurred on July 26, 1990.
The plant achieved criticality at 8:02 p.m.
Prior to the startup the on-shift scram analysis was reviewed by the inspectors to assure._ that the plant functioned as intended during the scram.
The analysis documented the, licensee' investigation of the-scram.
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On July 27,1990, while performing the fast closure MSIV operability E
test on inboard MSIV, B21F022A, the control power fuse (E32-F02) to
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subsystem A of the inboard MSIV leakage control system blew.
The inboard MSIV leakage control system was declared inoperable.. MSIV F022A-was reopened and fuse F02 was replaced under an existing work order.
The leakage control system was returned to operable status.
l The inspectors questioned the operability of the inboard MSIV leakage
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control system, in that the system is required to operate with the MSIVs
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The shift superintendent retested MSIV F022A and fuse F02 L
failed.
The fuse'was replaced with the MSIV closed and it failed again.
The inboard MSIV leakage control system was declared inoperable and ' a L
10CFR 50.72 report'was made.
This condition was-originally identified on April 29, 1989, when work L
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l order 93033 was written to investigate. the F02 fuse failure during a
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L MSIV operability test.
The failed fuse was replaced with MSIV. F022A opened and the work order was lef t open for further investigation.
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-August 14, 1989, the plant scramed due to a condenser boot failure.
l The MSIVs were closed during the scram and a MSly operability test was l-performed, no documented fuse failures were recorded.
On December 30, 1989, the plant scrammed due to loss of plant service water.
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failure occurred -during?the scram MSIV closure, but the fuse faQd
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during the MSIV' operability test.. 'The fuse. was replaced under existing work order. 93033, with MSIV F022A open..No further investigation was-conducted.
The fuse failures were'not documented in the SR0 or R0-logs.
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l After the fuse failure on July_ 27, 1990, trouble shooting on July 29,.
1990, found a. nick in a wire to: limit switch B21N101-A6.-located in the drywell.
The limit switch gives the-MSIV closure indication for the MSIV. leakage control subsystem ' A' initiation logic.
The nicked-wire
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resulted in a short circuit when MSIV F022A was fast closed, this caused the F02 fuse failure.
10CFR50, Appendix B,' Criteria XVI, requires that measures be established to - assure that ' conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material: and equipment are promptly identified and corrected.
The failure to correct the MSIV leakage control system fuse problem is a.
violation of 10CFR50, Appendix B, criteria XVI (90-15-02).
On July 30, 1990, while performing work under MWO 17341, MSR ' A' inlet isolation valve, N11F028A, "watar/ steam leak" from packing gland, the contract maintenance crew leak seal injected valve = N11F0288..
The licensee documented the-event in QDR 90-176., Administrative. Procedure 01-S-07-1, Control of Work on Plant Equipment and Facilities, requires that all plant work be properly authorized.
The leak seal injection on valve N11F028B was not authorized and is another example of violation of TS6.8.1.(90-15-01).
On August 10, 1990,-during a post work review of work request (WR) 7628, under which plant chilled water isolation valve, IP71F307, filter / regulator;was replaced on May 9,1990, the licensee noticed that no post maintenance test was performed to stroke time valve.F307 prior to returning the valve to service. QDR-190-90 was written to evaluate this-deficiency.
The planner 'for the WR failed.to specify stroke. timing -
retest requirements on the retest controlL form.
Valve 1P71F307 was successfully tested in accordance with Surveillance Procedure 06-0P-171-C-0003 on August 10, 1990, however, the failure to perform the required' post-maintencnce test is a licensee identified violation and is not cited because criteria specified in Section V.G.1 of NRC Enforcement Policy were satisfied.
(NCV 90-15-03)
On August 16, 1990, the control room recieved -an annunciator for RHR system ' A'. out ' of service.
An operator was dispatched to the upper control room to determine the cause of the alarm.
The RHR ' A' jockey pump was found not in service.
A control room operator went' to rack out the A RHR pump breaker. The operator mistakenly racked out the B RHR pump breaker.
Upon racking out the B pump breaker, the control room recieved
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numerous alarms indicating the B ' pump was out of-servic'e; i AlshiftJSR0
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was dispatched to take irmediate corrective actions ~ to rack in the RHR B i
pump. breaker. _ The pum s was then started L to ver.ify proper breaker.
operation.',
Incident Report 90-08-02 was written-- to document: this
--incident.; _ This is another example of failure to: follow 4 procedures
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-(90-15-01)..
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MaintenanceObservation(62703)
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-
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- procedure, proper ' tagouts, technical specifications, quality control's,-
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and radiological controls; observation of work and/or retesting;J and L
specified retest requirements.
MW0, DESCRIPTION 1-16176 Inspect / replace = desiccant for standby diesel'
starting air; dryer.
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-17059.
Monitor & troubleshoot the reactor feed pump B control loop.
17191 Rebuild dual ' solenoid, removed from plant as 1821F022B SVF201BA.& BB per - disposition of -
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MNCR 0093-90.
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18357 Inspect / bearing & bushing for Division I D/G-starting air compressor, j
No violations or deviations were identified,
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Surveillance Observation (61726)
e The inspectors observed the performance of portions of the surv'eillances l
listed, below'.
The observation included 'a. review of the: procedures' for-
-i technical adequacy, conformance to. technical specifications and LCOs;
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verification ofL test instrument calibration; observation of all 'or 'part'-
l of the actual-surveillances; removal and ' return to service of the' system i
or component; and review of the-data for acceptability based upon the
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acceptance criteria.
06-IC-1821-M-2004'
Reactor Vessel-Water Level, Level 2-and 1,
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Functional Test, Channel lC.
06-IC-1821-R-2004, Main Steam Line Isolation Valve Closure
.l Calibration, 1821-F022A.
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1 06-IC-1E21-M-1015,-
RWCU Delta Flow High Functional Test.
MSIV Leakage Control System Pressure Functional Test.
06-IC-1L11-R-0003,
- ESF 125 Volts Battery Pank Service Discharge Test.
06-EL-1L11-Q-0001, 125 Volts Battery Bank All Cell Check.
MSly Operability Test.
06-0P-1C51-V-0002, IRM Functional Test, Channel C.
00-0P-1T48-M-0002, SGTS System 'B' Operability.
During the performance of surveillance 06-0P-1B21-V-0001, MSly Operability Test, inboard MSIV F022D failed to f ast close.
The dual ASCO solenoid valve, SVF5010, on the, MSIV was disassembled for inspection.
The disc seat holder seating surface revealed a significant-compression; set.
The existing ASCO solenoid valves used an EPDM seating.
material and ' gasket.
The seating surface, as well as, the sentar protrusion of thei seating -surface were of a darker color than ',he remainder - of the disc.-
Engineering: Report 90-0023 concluded that the
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ASCO solenoid EPDM valve seats were deteriorating due to exposure-to high temperature -over a period of time in the presence of a lubricant.
All solenoid valves were replaced or refut blish for inboard MSIVs.
During refueling outage 4 the licensee plans to replace the dual solenoid valves with two single coil valves.
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No violations or deviations were identified.
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Reportable ~0ccurrences (90712 & 92700)
'The event reports listed below were reviewed to determine if the informa-
-tion provided met ' the NRC = reporting requirements.
The determination included adequacy of event description, the corrective 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' violations.
On July 24, 1990, the licensee reported that certain division three 125 VDC circuits may not receive-.the manufacturers minimum voltage values during a DBA. ' This issue was discovered during a NPE electrical system design basis review.and is based on electrical calculations.. The calcula-tions could not demonstrate that all devices receive sufficient energy to
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start and-operate und'.r DBA conditions.
The bounding device with the
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4 greatest potential voltage-deviation was the HPCS diesel generator breaker
closing coil. -The deficient: voltages have been attributed.to voltage
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- drops within long power and: control circuit runs.
Past system testing has oemonstrated the capability of ~ the as-build-125. VDC system to provide -
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sufficient energy to start and ' operate required loads.
To' reduce the-
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voltage drops, spare conductors of.the existing division three cables were used'to parallel the-cables to the HPCS diesel generator breaker (152-1701)
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auto close circuit. This work'was completed on July 25, 1990.
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A second issue identified by1 the. licensee concerns TS 4.8.2.1.d.2 '
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surveillance-for division three battery capacity test.
The TS requires a battery capacity test with a specified dummy load profile, which is
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verified to be greater than the actual emergency load. ' The licensee's j
calculations did not demonstrate the TS load' profile was greater than
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the actual load.
The calculations did show that the battery is sized.to
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support division three DC -system -loads.
The licensee performed a
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battery capacity test with the new' load profile, the test was successful.
Additionally, the licensee plans to submit a TS change to address the new
load rofile.
The review of the TS change will be an inspector followup (
item 90-15-04)
g No violations or deviations were identified.
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1 Action on Previous Inspection Findings (92701,.92702)
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(Closed) Violation 90-08-03, Failure to. maintain an operable isolation
.l group' III tripped condition.
.The _~ inspectors reviewed the licensee corrective action associated with this violation and concur that counselling the individual was adequate to p_revent recurrence of this
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incident. This item is closed.
Exit.Intervief (30703)
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The inspection scope and findings were summarized on August 17, 1990, with those persons indicated..in paragraph 1 above.
The licensee did not identify as proprietary any of the materials provided to or reviewed by a
,p the inspectors during this inspection.
The-licensee had no comment on
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A the'. following inspection findings:
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Description and Reference
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Item Number'
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-VIO 90-15-01 Severa' examples of failure to follow or inadequate procedures, paragraph 3.
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VIO 90-15-02 Failure to take adequate corrective action on the MSIV leakage control system,
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paragraph 3.
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NCV 90-15-03
' Non-cited violation ~ for failure to = perform
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thel required' post maintenance retest, V
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paragraph 3.
IFI 90-15-04 Review TS submitted on DC load profile,
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paragraph 6.
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Acronyms and.Initialisms-ADHRS-Alternate Decay Heat' Removal System Automatic Depressurization System ADS
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. Average < Power Range Monitor ASCO -
Automatic. Switch Company
.ATWS--
Ar ticipated Transient Without Scram B0P -
Balance of' Plant '
Boiling Water Reactor BWR
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CRD -
Control Rod Drive Design Basis Accidents DBA
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Direct Current DC
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-Design Change _ Package DCP
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Diesel. Generator DG
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Electronic Automitic Positioner EAP
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Emergency-Core Cooling System
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Engineering Safety FeatJre'
FCV.
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Flow Control Valve Feedwater
- FW
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HPCS -
High. Pressure Core-Spray
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Hydraulic Power-Unit HPV
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Instrumentation and Control 1&C
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Inspector Followup Item IFI
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Intermediate Range Monitor JIRM
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LCO = -
Limiting Condi_ tion for Operation
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LER -
Licensee' Event Report LLRT -
Local Leak Rate Test LPCI
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. Low Pressure' Core Injection LPCS -
Low. Pressure Core Spray-
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MNCR -
Material Nonconformance Report MSIV
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Main Steam Isolation Valve Maintenance Work Order
. MWO
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Nuclear Plant Engineering
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NPE
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Nuclear Regulatory Commission NRC-4 PDS -
Pressure. Differential Switch P&lD,-
Piping and Instrument Diagram
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PSRC -
Plant Safety Review Committee Plant. Service Water PSW
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Quality Deficiency Report QDR
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RCIC -
Reactor Core Isolation Cooling RFPT -
-Reactor Feed Pump Turbine
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Reactor Operator'
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RWCV.-
_ Reactor Water Cleanup RWP, -
Radiation Work Permit SLCS --
Standby Liquid Control System-System Operating Instruction
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Senior Reactor Operator SR0
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Standby Service Water SSW
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Temporary Change Notice TCN
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Technical Specification TS.
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Work-Order WO
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Work Request WR
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