IR 05000416/1989012
| ML20246J303 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 04/28/1989 |
| From: | Cantrell F, Christensen H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20246J288 | List: |
| References | |
| 50-416-89-12, NUDOCS 8905170044 | |
| Download: ML20246J303 (11) | |
Text
e--
.
.
.,
.
.
,
,;
AM2g-UNITED STATES.
'
-
. NUCLEAR REGULATORY COMMISslON
[.
o'
REGION 11
'
' gt --
y:
.101 MARIETTA STREET, N.W.
ATLANTA, GEORGI A 30323 i
% -
O
'
4*****.
,I
~ Report No.:
50-416/89-12
. Licensee:-
System Energy Resources, Inc.
Jackson, MS 39205 i
Docket No.:
50-416 License No.:
!
Facility Name:
Grand Gulf Nuclear Station Inspection Conducted: March 18 April.14, 1989
. Inspectors:
/
///28/87
,
H. 0..Christensengeryfor Resident Inspector Ddte Signed EAV
$
Y2$ $9 J.~L.Mathis,ResippffiInsector D&te Signed Approved by:
dMM 9/2-f/P9 F. S. Cantrell, Section ChijfV Dete S'igne1 Division of Reactor Projects SUMMARY Scope:
The resident-inspectors conducted a routine inspection in - the following areas:
operational safety verification; maintenance observation; surveillance observation; installation and testing-of modifications; outage organization; action on previous ~ inspection findings; ' and reportable occurrences.
The inspectors conducted backshift inspections on April 4,
5, 6, '11 and ' 12, 1989.
F. S. Cantrell, Section Chief, Division of Reactor Projects, was on site April 4 r J 5,1989, to conduct a plant tour and hold discussions with the resi,ent inspectors and plant management.
Results: Within the areas inspected one licensee identified violation was noted for failure to resolve and clear nonconforming material items prior to placing the alternate decay heat removal system in service.
During this inspection period the plant was in a refueling outage, which has progressed on schedule, with minimum difficulty.
The inspectors have observed a variety of activities relating to the refueling outage. The majority of the activities were carried out in accordance with the procedures and in a professional manner.
However, the inspectors have noted some minor indications of lack of attention to detail by the licensee (paragraphs 3, 5, and 6). When these items were brought to the licensee attention they were immediately addressed, l
!
8905170044 890503 PDR ADOCK 05000416 O
'
,
_ _ _ - _ _ - _
.
-
.
,
REPORT DETAILS 1.
Persons Contacted Licensee Employees J.G. Cesare, Director, Nuclear Licensing W.T. Cottle, Vice President of Nuclear Operations
- D.G. Cupstid, Superintendent, Technical Support
- L.F. Daughtery, Compliance Supervisor l
J.P. Dimmette, Manager, Plant Maintenance
- S.M. Feith, Director Quality Programs
- C.R. Hutchinson, GGNS General Manager R.H. McAnulty, Electrical Superintendent A.S. McCurdy, Technical Asst., Plant Operations Manager
- L.B. Moulder, Operations Superintendent
,
J.H. Mueller, Mechanical Superintendent J.V. Parrish, Chemistry / Radiation Control Superintendent J.L. Robertson, Superintendent, Plant Licensing
- J.C. Roberts, Manager, Performance and System Engineering S.F. Tanner, Manager, Quality Services
!
L.G. Temple, I & C Superintendent F.W. Titus, Director, Nuclear Plant Engineering M.J. Wright, Manager, Plant Support
- J.W. Yelverton, Manager, Plant Operations Other licensee employees contacted included technicians, operators, security force members, and office personnel.
NRC Personnel L. Modenos, Project Engineer
- Attended exit interview l
2.
Plant Status Unit 1 began and ended the inspection period conducting refueling outage
,
number 3.
3.
Operational Safety, (71707)
The inspectors were cognizant of the overall plant status, and of any 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 room.
Observations included the verification of instrument readings, setpoints and recordings, status of operating systems, tags and clearances on equipment controls and switches, annunciator alarms, adherence to limiting conditions for operation, temporary alterations in effect, daily journals and data sheet entries, control room manning, and access controls.
This inspection activity included numerous informal discussions with operators and their supervisors.
- _ ______
.
.
,
During tours of the control room on April 4, 5 and 6,1989, the inspectors noted that the RHR A MOV test switch was in test.
RHR A and LPCI A were required to be operable.
When the operators were questioned about the switch setting, they immediately placed the switch in the normal position and stated that they earlier had operated the valves in the system, which required them to set the switch in the test position.
The function of the test switch is to install thermal overload protection in the ECCS MOVs during testing or normal system operations and to by pass this protection during accident conditions.
T.S. 3.8.4.2, motor operated valves thermal overload protection action statement, allows the thermal overload protection to be installed in the MOVs for up to eight hours, but requires the affected valve to be declared inoperable if this time limit is exceeded.
The insoector expressed a concern that the operators were not adequately controlling the test switches to ensure that the T.S. LCO was not being exceeded. The inspectors determined through discussion that the eight hour LC0 had not been exceeded. The licensee issued a night order and placed information tags on the test switches to increase operator awareness.. Additionally, the reactor operator commenced logging the switch position, when the switch was in test.
On a weekly basis selected engineered safety feature (ESF) systems were confirmed operable. The confirmation was made by verifying the following:
that accessible valve flow path alignment was correct; power supply breaker, and fuse status were correct; and instrumentation was operational.
The following systems were verified operable:
General plant tours were conducted on a weekly basis. Portions of the control building, turbine building, auxiliary building and outside areas were visited.
The observations included safety related tagout verifications, shift turnovers, sampling programs, housekeeping and general plant conditions, the status of fire protection equipment, control of activities in progress, problem identification systems, and containment isolation.
Additionally, the licensee's onsite emergency response facilities were toured to determine facility readiness.
The inspectors noted health physics management's involvement and aware-ness of significant plant activities including plant radiation controls.
The inspectors verified licensee compliance with physical security manning and access control requirements.
Periodically the inspectors verified the adequacy of physical security detection and assessment aids.
The inspectors reviewed the following safety related tagout, 890243, Division II Diesel.
The review ensured that the tagout was properly prepared, and performed.
Additionally, the inspectors verified that the tagged components were in the required position.
-
_ _ - _ _ _ _ _ - _ _ _ _ _
f.
\\
l
,
.
.
The inspectors have noted specifically that the general manager, operations manager, operations superintendent and maintenance manager,
,
make routine tours to the plant and the control room.
I I
i The inspectors reviewed the activities associated with the below listed events:
On March 23, 1989, at approximately 4:15 a.m.
the licensee commenced lifting the moisture separator. They encountered difficulty with the lift.
Investigation by the licensee determined that two shroud head bolts (#33 and #36) had not been loosened. The licensee loosened the bolts and successfully completed the separator lift on March 26, 1989.
A visual inspection was conducted of the moisture separator and core should head bolts.
The two hold downs that were engaged during the initial lift were removed from service.
The separator has 36 hold down bolts. GE analysis indicted that only 4 bolts are required to adequately secure the separator in the reactor vessel.
On March 28, 1989, during the. performance of a LLRT, the inspector noted the entry of a contract personnel into a high radiation area without the proper dosimetry.
Followup on this item was assigned to a radiation protection inspector and is documented in NRC 1,spection report 416/89-08.
On April 12, 1989, while cycling RHR B valve, E12F003B, it was noticed that the valve did not cycle to the full open or close position.
A maintenance work order (MWO) was written to investigate and inspect the valve. It' was discovered that the locking tab for the stem to disk coupling nut had broken.
The coupling nut backed off the disk allowing the disk to separate from the stem.
When the valve was cycled the disk was driven into the valve body seat causing cracking of the body seat, disk, and stem guide. The damaged valve is being replaced with a Unit 2 valve.
Additionally the plant plans to radiograph the RHR A valve to determined if it has similiar problems.
On April 12, 1989, HPCS recevied an automatic initiation signal due to a I
spurious reactor water loPI level, level 2, signal. The Division 3 diesel auto started; however, reactor water level was at +195 inches so the injection valve remained closed.
The HPCS pump breaker was racked out for pump maintenance, therefore, no ECCS injection occurred. The licensee stated the most probable cause was a radio transmission in the vicinity of the HPCS level transmitters.
The licensee determined the event was not reportable per NUREG-1022, Supplement 1.
No violations or deviations were identified.
-
4.
Maintenance Observation (62703)
During the report pariod, 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
__ _ _
_-_
-
,
.
.
,
procedures, to " proper tagouts, technical specifications, quality controls, and radiological controls; and observation of work, and/or retesting.
MWO DESCRIPTION M91931 Division 1 DG electric driven air compressor repair.
191964 Division 1 DG pneumatic computer.
190073.
Division 1 DG voltage regulator not in auto annunciator.
M91393 Division 3 DG thermostatic valve.
M91395 Division 3 DG - perform frequent inspection.
M91379 Division 2 DG - perform turbo charger maintenance and exhaust manifold inspection.
M91400 Division 3 DG - perform maintenance run.
M92013 Feedwater check valve.
M91377 Division 2 DG - perform cold crankshaft deflections, crankcase relief valve maintenance and crankcase interior inspection.
191996 RPS 8 level annunciator.
EL2581 Calibrate 1B21-11-R660C, MSIV pilot solenoid ammeter 1H13P622.
M90569 Control room A/C compressor train A rebuild.
MWP 89/1082 Perform wiring change per DCP 86/0126.
On April 10, 1989, while performing a calibration of the MSIV Pilot solenoid ammeters, the electricians determined that the H-R659 and H-R660, A, B, C and D meters were labeled backwards, that is H-R659 meters were labeled H-R660 and the H-R660 meters were labeled H-R659. The correction of the meter labels will be an inspector followup item (416/89-12-01).
No violations or deviations were identified.
L___
.. _ _ _ _ _ _ _ _
_ _ _. _ _ _ _ _ _ _ _ _
__-
___
--. _ - - _
o
.
.
-
,
.
.
5.
Surveillance Observation (61726)
The inspectors observed the performance of portions of the surveillance listed below.
The observation included a review of the procedure for technical adequacy, conformance to technical specifications and limiting conditions for operations (LCOs), verification of test instrument calibration, observation of all or part of the actual surveillance',
removal and return to service of the system or component, and review of the data for acceptability based upon the acceptance criteria.
06-EL-1R65-R-0001, Revision 27, Motor Operated Valve Thermal Overload Protection Device Functional Test, Attachment I, M0V 52-151124.
06-IC-1821-R-2009, Revision 20, Drywell High Pressure (ECCS), Attachment III, Channel F.
06-IC-1C51-V-0001, Revision 25, Intermediate Range Monitor Calibration, Attachment III, Channel C.
06-IC-1E21-R-0001, Revision 23, Low Pressure Core Spray Discharge Line Hi/Lo Pressure Calibration.
06-IC-1E31-R-0038, Revision 23, Drywell Floor and Equipment Drain Sump Level and Flow Monitoring System Calibration.
06-ME-1M61-V-0001, Revision 32, Local Leak Rate Test, for RHR 0-ring E120003A and E12D003B.
06-0P-1821-R-0006, Revision 30, Containment Drywell and Auxiliary Building Isolation Valve Functional Test.
06-0P-1C41-M-0001, Revision 30, Standby Liquid Control Operability, Attachment II.
06-0P-1P75-R-0003, Revision 27, Standby Diesel Generator 11:
18 Month Functional Test, Attachment II and IV.
06-0P-1P75-R-0004, Revision 28, Standby Diesel Generator 12, 18 Month Functional Test, Test #4.
Simulated Loss of Offsite Power.
06-0P-1P81-R-0001, Revision 26, HPCS Diesel Generator 13, Functional Test, Attachment III.
On April 6,1989, the inspector observed the performance of a local leak rate test, using a bubble test rig for RHR 0-ring, E1200038.
The acceptance criteria for the test was no flow as indicated by no bubbles
_
___ _ __- ___ _____-____-
______ _ _ --_ __-_- _ __
7_ - - _-_ __
i
'
,
.
'
-
.
,
through the test rig water column for a period of one hour and that water level must not rise by more than one-quarter inch during the test.
The.
contract mechanic performing the test did not have a means to measure the level increase. Another mechanic used a ruler to perform the measure-ments. The 0-ring passed the bubble test after test reinitiation.
The licensee took immediate action to attach rulers to the test rigs and will
. review the procedure to determine if the acceptance criteria is adequate.
On April 7, 1989, during the performance of the Division III diesel generator 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run, 06-0P-1P81-R-001, Attachment III, the local operator inadvertently tripped the output breaker at approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> into the run.
The operator was in the process of adjusting generator reactive load when he mistakenly manipulated the output breaker control switch rather than the voltage regulator control switch.
Incident Report 89-4-5, states the following factors contributed to the event: The close proximity of the voltage regulator control switch and the output breakers, output breaker is approximately five inch below the control switch; both switch handles are identical in size, shape and color; and the voltage control switch is out of the normal field of view of the operator when viewing the VARs meter.
After the generator was tripped, the diesel was cooled down and secured.
The licensee restarted the surveillance, and the diesel passed the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run successfully.
No violations or deviations were identified.
6.
Installation and Testing of Modifications (37828)
The inspector conducted an inspection of the alternate decay heat removal system modification.
The inspection included direct observation of the installation process, installation testing, and preoperational and startup testing.
Additionally, a system walkdown was conducted, using System Operating Instruction 04-1-01-E12-1, Revision 40, RHR System, Temporary Change Notice 74, and P&ID's M-A10850, Residual Heat Removal System, and M-A072H, Plant Service Water System.
The following items were noted during the system walkdown.
-
Butterfly valves; E12-F411B, P44-F481A, P44-F481B, P44-F482A, and P44-F482B handles are installed opposite normal convention. When the valves are open the handles are perpendicular to the pipe, norm. ally the handles align with the pipe.
The system was placed into operation prior to performing system
-
preservation (e.g. painting). This is an ALARA concern, in that this area has the potential to become a high radiation area.
t j
I
_ _ - _ _. _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -
__-_
__ -_ _ _ _
_
. _ _ _
_ _ - _
._
- - _ - _ _ - _ _ _ - _ _.
.
.
=
l *
Hose connections were attached to the following drain lines:
-
P44-F4878, P44-F487A, E12-F421B, E12-F421A, and E12-F427, the drawings show the drains being capped.
The S01 04-1-01-E12-1, Attachment III D, Electrical Lineup Checksheet
-
Compon'ent Description differed from the actual equipment label, for all six breakers.
The correction of the above items will be an inspector followup item (416/89-12-02).
The inspectors observed portions of the preoperational and startup testing and reviewed the completed sections of modification special test instruc-tion (MSTI) 1E12-88-001-0-S, Alternate Decay Heat Removal System Test.
The testing was performed in accordance with the MSTI and when test changes were required, they were performed in accordance with procedure 15-S-03-2, Revision 2, Modification Special Test Instruction.
The plant service water liquid process monitor was calibrated and the alarm setpoint was set within the T.S. limit.
On April 3,1989, the plant placed ADHRS in operation as an alternate decay heat removal method. On April 4, 1989, the licensee determined that the ADHRS was placed in operation with three outstanding Discrepant Material Reports (DMRs).
The DMRs were subsequently resolved and closed the same day.
Procedure 01-5-03-7, Discrepant Material Reports, requires the preparation of MNCRs to replace DMRs before nonconforming items are installed in the plant.
Procedure 01-S-03-3, Material Nonconformance Reports (MNCRs), states that under no circumstances shall nonconforming hardware be returned to service and relied upon to fulfill its intended safety function.
The failure to transfer the DMRs to an MNCRs and then not resolve them before placing the ADHRS in operation is a violation of T.S. 6.8.1, which states that procedures shall be established, implemented and maintained.
The violation meets the criteria for licensee identified as discussed in 10 CFR 2, Appendix C, and will not be cited.
The appropriate corrective action was initiated in a timely manner to resolve the DMRs.
QDR 117-89 was written to document that an MNCR was not written to replace DMRs and that the MNCR was not resolved prior to the system placed in operation.
7.
Refueling Activities (60710)
The inspectors reviewed the Grand Gulf Refueling Outage 3 organization.
Administrative Procedure 01-S-06-42, Revision 0, described the refueling outage organization and its functions and responsibilities for the scheduling and management of refueling outages.
The Plant General Manager, appointed two Refueling Outage Directors, who coordinate and direct all pre-outage and outage activity with the authority of the General Manager.
The Outage Director are on 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> coverage.
In addition the outage organization consisted of the following:
_________-__-____ - _-__ _ _ ___ _ _ -
,
__-
__._. _ _ _ _ _ _ _ - _ _
_ - _ _ _ _ _
~
,.
-
.
,
l four outage managers, four operations outage coordinators, five local leak rate testing coordinators, three tagging preparation task force members, five refueling floor SR0s and 10 surveillance coordinators.
Refueling outage 3 was projected to take approximately 46 days, as of day l
26 the plant was on schedule.
There has been some minor changes to the schedule but none that have impacted the overall projected outage time.
8.
Reportable Occurrences (90712 & 92700)
,
The below listed event reports were reviewed to determine if the information provided met the NRC reporting requirements. The determination included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance of each event. Additional inplant reviews and discussions with plant personnel as appropriate were conducted for the reports indicated by an asterisk. The event reports were reviewed using the guidance of the general policy and procedure for NRC enforcement actions, regarding licensee identified violations.
On March 27, 1989, at 6:05 a.m.
the auxiliary building, secondary containment, received a division I isolation. The isolation was caused by a loss of power to the 4160 volt A.C. bus 15AA. The power loss was caused by a dispatcher switching operation on the Baxter to Port Gibson 115KV feeder.
This caused a momentary loss of power to the Port Gibson substation. The load shedding and sequencer sensed a 70 percent bus under voltage for 0.5 seconds and tripped the incoming feeder breaker.
The division I diesel was in a maintenance outage at the time.
The Port Gibson substation was returned to power in less than a minute and the plant recovered from the isolation with no reported problems.
On March 29, 1989, at 9:28 p.m. the licensee reported the LLRT failure of RHR A valve E12F290A.
The valve would not perform its intended function as a containment isolation valve.
The valve was repaired and and subsequently passed retesting.
On March 31, 1989, the licensee reported the failure of the type B and C LLRT for main steam line C and feed water line A valves.
The overall penetration leakage rates calculated by maximum pathway leakage exceed 0.60 LA.
The leakage calculated by minimum pathway leakage was less than 0.60 LA.
The feedwater line A check valve was repaired, resilient seats replaced, and subsequently past its LLRT.
The main steam line C, leak path was repaired and subsequently past its LLRT.
On April 3,1989, the licensee reported the failure of feedwater line B type C LLRT.
The overall maximum pathway leakage exceeded 0.60 LA, the minimum pathway leakage was zero.
The licensee repaired the feedwater check valve and retested the feedwater line, which passed.
No violations or deviations were identifie _ _ _ - _
-__
_ _ _
-
,
.
_
,
' -
9.
Action on Previous Inspection Findings (92701,92702)
(Closed) Violation 88-28-01, Failure to follow procedures, two examples.
Failure to follow annuciator system procedure and failure to follow a maintenance work procedure.
On February 28, 1989, the licensee responded to the violation.
The following corrective action have been completed; a QDR was initiated on the annunciator status logbook; operations performed an extensive review of the inoperable annunciators and issued a new deficiency documentation requirements; retraining was conducted on how to conduct log audits and the annuciator status is tracked on the daily plant status report.
For the maintenance procedure violation, the contract crew recieved remedial training. This item is closed.
(Closed) Violation 88-28-03, Failure to perform a written safety evaluation for bypassing of containment and reactor vessel isolation control system isolation signals for the RWCU system.
The licensee responded to the violation on March 16, 1989.
The following corrective actions have been completed: An additional 10 CFR 50.59 reviewer has been added to screening process; 50I 04-1-01-G33-1 was revised to delete the use of the manual bypass switch; and integrated operating instructions and S0Is will be reviewed to determine if 10 CFR 50.59 analysis are warranted.
This item is closed.
(Closed) Violation 88-28-02, Failure to have M&TE control of hydrometer used for safety-related technical specification measurements.
The licensee responded to the violation on February 28, 1989.
The licensee has procurred chemist grade hydrometers, that are certified to the National Institute of Standards and Technology. This item is closed.
(Closed) IFI 89-04-01, Review RCIC system periodic oil change procedure 07-S-14-81, to reflect the vendor recommendations.
A temporary change notice was issued on February 23, 1989. This item is closed.
10. Exit Interview (30703)
The inspection scope and findings were summarized on April 14, 1989, with those persons indicated in paragraph 1 above.
One licensee identified violation was noted for failure to resolve and clear nonconforming
material items prior to placing the alternate decay heat removal system in service.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
The licensee had no comment on the following inspection findings:
Item Number Descr_iption and Reference 89-12-01 (IFI)
Correct MSIV pilot solenoid ammeter labels 89-12-02 (IFI)
Correct ADHRS walkdown items i
i t
- - _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ - _
_ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
[
..
.
.
,
11. Acronyms and Initialisms l
'
Alternate Decay Heat Removal System ADHRS
-
Average Power Range Monitor APRM
-
CRD Control Rod Drive
-
Design Change Package l
-
L DG
-
Diesel Generator
!
-
Discrepant Material Report Emergency Core r mling System ECCS
-
'eature Engineering Sat ESF
-
-
FCCU Fuel Pool Coolis d Cleanup
-
FCV Flow Control Valt
-
-
Hydraulic Power Unit HPU
-
I&C Instrumentation and Control
-
Inspector Followup Item y
IFI
-
Limiting Condition for Operation l
LC0
-
LER
-
Licensee Event Report Local Leak Rate Test LLRT
-
-
Low Pressure Core Injection Low Pressure Core Spray LPCS
-
Material Nonconformance Report MNCR
-
Modification Special Test Instruction MSTI
-
MW0 Maintenance Work Order
-
Nuclear Regulatory Commission NRC
-
PDS Pressure Differential Switch
-
Piping and Instrument Diagram P&ID
-
Plant Service Water PSW
-
-
Quality Deficiency Report RCIC
-
Reactor Core Isolation Cooling RHR Residual Heat Removal
-
-
Reactor Protection System RWCU
-
-
Radiation Work Permit SBLC
-
Standby Liquid Control System Energy Resource Incorporation SERI
-
System Operating Instruction 501
-
Standby Service Water SSW
-
Temporary Change Notice TCN
-
Technical Specifications TS
-