IR 05000416/1990002
| ML20012C351 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/02/1990 |
| From: | Cantrell F, Christensen H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20012C350 | List: |
| References | |
| 50-416-90-02, 50-416-90-2, NUDOCS 9003210094 | |
| Download: ML20012C351 (11) | |
Text
.,
UNIT E D STDTis
,[p H0g'o NUCLE AR REGULATORY COMMISslON
['
Rt Ol0N 11 n
g g
101 MARitTT A ST Rt E T. N.W.
- ATLANTA,O!ORGI A 30323
\\e....}
Report No.: 50-416/90-02 Licensee:
System Energy Retources. Inc.
Jackson MS 39205 Docket No.:
50-416 License No.:
NPF-29 Facility Name:
Grand Gulf Inspection Conduct >d: January 20 - February 16, 1990 Inspectors:
d 3 * I ~ TO di
~0. Christens ~n. Senior Resident Inspector Date Signed e
f& A h a0M s -I ~@
. L. Mathis Resident inspector Date Signed Approved by:
$MMk,
d r-
.
Floyd 5. Cantrell. Thiel/g Date Signed Projects Section IB Division of Reactor Projects Scope:
The resident inspectors conducted a routine inspection in the following areas:
operational safety verification; maintenance observation; surveillance observation; Independent Safety Engheering Group; action on previous
!
inspection findings; and reportable occurrences.
The inspectors conducted backshift inspections on January 30. and February 8, 1990.
Results:
During this inspection no violations or deviations were identified.
One unresolved item was identified concerning the adequacy of long-term high pressure core spray service water system cooling following a design basis accident, paragraph 7.
The licensee identified this during an Improved i
Technical Specification Program review.
In the inspection areas of safety verification, maintenance observation and surveillance observation, paragraphs 3, 4. and 5. the licensee met the safety objectives of these areas.
However, weaknesses were noted with the material l
nonconformance process, paragraph 4.
This process is being evaluated by the-i licensee for improvements.
l
l The Independent Safety Engineering Group (ISEG) provides detailed and informative reports to senior plant management.
However, the timeliness for implementing ISEG recommendations needs improving, paragraph 6.
9003210094 90o3o%26
,
$DR ADOCK 0 000 PDC i
.d
-
.,
-
..
,
,
!
REPORT DETAILS j
1.
Persons Contacted Licensee Employees
'C. W. Angle, Manager, Operational Analysis Section
J. G. Cesare, Director, Nuclear Licensing i
W. T. Cottle, Vice President Nuclear Operations
!
- D. G. Cupstid. Manager, Plant Modifications and Construction
.
!
- L. F. Daughtery, Compliance Supervisor
- J. P. Dimmette, Manager, Plant Maintenance i
S. M. Feith, Director, Quality Programs i
C. R. Hutchinson, GGNS General Manager
!
F. K. Mangan, Director, Plant Projects and Support L. B. Moulder, Operations Superintendent
- J. C. Roberts, Manager, Plant & System Engineering
'
S. F. Tahner, Manager Quality Services F.W. Titus, Director, Nuclear i
Plant Engineering
,
M. J. Wright, Manager, Plant Support J. W. Yelverton, Manager, Plant Operations G. Zinke, Superintendent Plant Licensing
[
,
Other licensee employees contacted included superintendents, supervisors, technicians, operators, security force members, and office personnel.
- Attended exit interview l
F. Cantrell, Section Chief, Division of Reactor Projects, was on site i
February 6 and 7, 1990.
Tours were conducted with the resident inspectors of the Claiborne County emergency operations center, the backup emergency operations facility and the local public document room.
Additionally, j
discussions were held with plant management.
.
.
2.
Plant Status
-'
The plant began and ended the inspection period in mode 1, power operations.
3.
Operational Safety, (71707 and 93702)-
l The inspectors were cognizant of the overall plant status, and of any
.4 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.
Observa-tions 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 l
and data sheet entries, control-room manning, and access controls.
This
r i
?
- _ _ _ _ _ _ _ - _
. _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
_ _ _ _
"
',
..
.
inspection activity included numerous informal discussions with operators
,
and their supervisors.
On a weekly basis selected engineered safety feature (ESF) systems were
!
confirmed operable.
The confirmation was made by verifying that
,
accessible valve flow path alignment was correct, power supply breaker and fuse status was correct and instrumentation was. operational.
The following systems were verified operable: SLCS, CRD HPCS, ADS, and emergency power.
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 verifica-tions, 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 and
.'
the readiness of the onsite emergency response facilities.
The inspectors observed health physics management involvement and awareness of significant plant activities, as well. as, plant radiation controls.
Periodically the inspectors verified the adequacy of physical
,
security control.
,
900033 (IRM A and C), and The insp(ectors reviewed safety related tagouts, Recirculation Pump A HPU)
to ensure that the tagouts were 900201 properly prepared, and performed.
Additionally, the inspectors verified that the tagged components were in the required position.
The inspectors verified that the following containment isolation' valves were in there correct lineup: M41-F017, drywell air - purge exhaust; M41-F013, drywell air purge supply; E51-F019A, RCIC pump discharge minimum
,
flow; and E30-F594B. suppression pool level instruments.
The inspectors reviewed the activities associated with the events listed -
below.
On February 6, 1990, at ap)roximately 2:18 pm the licensee received notification of an unusual occurrence by Chem-Nuclear, at the Barnwell disposal site.
During loosening of the ratchet binders on the NUPAC 14-210 cask, gas was heard escaping from the cask.
Safety personnel determined the gas to be 100 percent flammable.
Based on the presence of explosive gases the South Carolina Department of Environment Control was notified of an unusual occurrence.
A followup conversation at 4:20 pm on February 6,1990 informed the licensee that the cask was vented and no violation had occurre *
.
,
.
...
,
i
[
During a tour of containment the inspector observed welding activities for DCP 88/0005, RWCU resin metering pump upgrade.
The work was being performed in accordance with the work package.
The inspector expressed concern with welding in containment while at power. A review of the DCPs
operational impact checklist indicated several. areas may be impacted by welding.
However, no justification was documented on the checklist to
)
allow this activity. Discussions with the managers, of plant modification
and construction, and plant licensing indicated that an adequate review of the activities was conducted; however, this review was not adequately
,
documented.
The operational impact checklist was reperformed and adequate justifications were documented.
The manager of plant modifications and
'
constructions, stated that the proc; dure 01-S-16-2, Modification Work Permit, will be revised to providt better guidance for documenting
operational impact reviews.
The review of the completed procedure revision will be an inspector followup item (90-02-01).
During the weeks of January 22 and 29, 1990, the licensee conducted I
a self assessment of the maintenance program. The liscensee's assessment team consisted of 10 individuals, one from INP0, and two from other nuclear plants.
The assessment appeared indepth and detailed.
Several recommendations were made to improve perfonnance.
These were included in-
the licensees maintenance improvement program for implementation.
'
No violations or deviations were identified. The results of the inspection in this area indicate that the program was effective with respect to i
meeting safety objectives.
- 4.
Maintenance Observation (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 procedure, proper tagouts, technical specifications, quality controls, and
,
radiological controls; observation of work and/or retesting; and specified
retest requirements.
MWO DESCRIPTION W0 3578 Monitor / troubleshoot ADS /SRV Lo-Lo set and ECCS ' injection permissive logic.
WO 3726 Change lube oil on pedestal bearing
!
for HPCS D/G.
'
W0 3728 Inspect HPCS D/G slip ring and meggar motor.
WO 3731 Calibrate and check 480 volts diesel generator voltmeter.
WO 3789 Inspect / rework flange holes per
!
07-5-14-357 and Patel electrical connections
,
for SRV 1821F0510.
L
'
.
-
.
,
MWO DESCRIPTION (cont'd)
WO 4152 Lubricate SSW to RHR isolation valve (F094)
WO 4211 Check cou) ling wear and alignment of RHR C joccey pump.
On January 30, 1990, the RCIC system was tagged out of service for routine preventive maintenance.
The inspector reviewed the tagout of RCIC to ensure that the tagout was properly prepared and performed. The inspector also witnessed the following maintenance activities associated with the RCIC outage:
WO 3145 Replace RCIC oil filters.
W0 3421 Clean and inspect RCIC trip throttle valve.
WO 3847 Periodic oil change of bearing sump for RCIC pump.
W0 3875 Clean / replace RCIC seal gland compressor air inlet filter.
WO 3876 Lubricate seal gland compressor.
W0 3898 Lubricate RCIC pump coupling.
WO 4480 Water in RCIC turbine crankcase, inspect.
During the lube oil sampling under work order 3877 a very large amount of water was found in the lube oil sump of the RCIC turbine. MNCR 011-90 was written to obtain a failure analysis concerning water in the oil.
The lube oil cooler was then removed and hydrostatically tested for leakage,.
the result was negative. The rest of the preventive maintenance scheduled was completed and the lube oil system refilled.- After the cold start retest, water leakage was observed at the turbine seals, another-oil-sample was taken and more water was found in the oil sump.
MNCR 011-90 instructed plant staff to flush the oil sump per WO 4480 and change the oil sample point to a low-point drain plug.
During the removal of the drain plug, the oil well cover plate was broken. MNCR 012-90 was written on February 2,1990 documenting this problem.
NPE dispositioned MNCR 012-90 to allow fabrication and installation of a. new RCIC turbine oil well cover plate.
The new cover plate will' be a temporary replacement until a vendor supplied component can be obtained.
The duration of the interim disposition is for 180 days.
Engineering analysis of the new cover plate showed that it can withstand stresses greater than the original plate due to the higher tensile strength of the replacement material.
During the review of MNCR 012-90, the resident inspector identified problems with the operability determination statement.
The licensee reviewed the MNCR and identified several additional concerns with the MNCR process.
These concerns were documented in QDR 43-90.
The inspector.
questioned the licensee as to the usage of an MNCR process
,
-
.,
-
..
,
.
8i i
versus a temporary alteration in correcting the broken cover plate.
Procedure 01-S-03-3, Material Non-Conformance Report, Section 6.12 allows i
implementation of design change MNCR without a DCP/MCP providing that a
Review and Approval Implementation Record (RAIR), determination form is initiated.
The procedure does not provide clear provisions for a RAIR
!
review for interim disposition to MNCRs.
The licensee is aware that problems exist with the MNCR process as
!
outlined in procedure 01-S-03-3.
Several of these problems are documented
,
in QDRs 0027-90 and 0043-90. The majority of the problems identified with
the MNCR process relates to operability.
The following problems are
examples documented in the QDRs:
.
1.
There are currently no provisions in the MNCR process for the PSRC to make operability determinations on behalf of operations.
2.
Procedure 01-S-03-3 does not provide for extending the operability
review date once it has been established.
,
t 3.
Procedure 01-S-03-3 does not provide a method for obtaining a failure analysis only.
The inspector will review the licensees corrective action for impruving the MNCR process. This is inspector followup item 90-02-02.
No violations or deviations were identified. The results of the inspection in this area indicate that the maintenance program was effective with respect to meeting safety objectives.
5.
SurveillanceObservation(61726)
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-all or part 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 criteria.
06-lC-1821-M-0014. Safety Relief Valve Tail Pipe Pressure Switch Functional Test.
06=lC-1E31-M-0023, RCIC/RHR and RCIC Steam Line Flow High (RCIC isolation) Functional _ Test,attachmentII.
06-IC-1E32-R-1001, Main Steam Isolation Valve Leakage Control System.
06-lC-1C51-SA-0001, Average Power Range Motoring Calibration.
Channel D.
06-0P-1E51-Q-0003, RCIC System Quarterly Pump Operability Verificatio.,
.
,
No violations or deviations were identified.
The results of the inspection in this area indicated that the program was effective with respect to meeting safety objectives.
6.
Independent Safety Engineering Group (40500)
The inspectors reviewed selected ISEG reports to determine if the reports were indepth, detailed, and addressed root cause problems. Additionally, selected ISEG recommendations were reviewed to determine if the recommendations were implemented by the plant. The various ISEG functions were reviewed to determine if the requirements of the TSs, FSAR and NUREG-0737 were being performed.
The following ISEG reports were reviewed:
ISEG-87-007 Electrical system walkdown.
ISEG-88-007 Evaluation of the GGNS site 10 CFR 21 screening process.
ISEG-89-004 Effectiveness review of the operating experience program.
ISEG-89-015 Investigate incident reports 89-6-6.-7.-8.
ISEG-89-021 ISEG survey report on SIL 497.
The ISEG reports were detailed, addressed root causes and made appropriate recommendations.
ISEG recommendations are tracked and the reports are not closed until the recommendations are implemented.
A review of the open recommendations indicated a number of open items that were several years ol d.
Discussion with plant staff and NPE indicate that high priority reconnendations are acted upon quickly, however routine items may be delayed and or an adequate status of progress is not provided to ISEG.
Significant recommendations are reviewed by the Vice President, Nuclear Operations, while routine ISEG recommendations are forwarded directly to the General Manager or NPE for review.
ISEG provides senior management a quarterly assessment report on incident report trends and root causes; and a monthly sunnary on plant operating experiences.
These reports, along with special reports requested by the safety review committee provide senior management with independent assessment information.
A review of ISEG records indicate that not all ISEG project reports are transmitted to records management for lifetime storage as required by NPE administrative procedure 710. This deficiency was immediately corrected.
ISEG members have the expertise and experience level to conduct meaningful, independent assessments and provide valid recommendations to senior plant managemen..
..
,
.
7 Reportable Occurrences (90712 & 92700)
f
'
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 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.
(0 pen) LER 89-19, Malfunctioning telemetry causes loss of plant service water and manual reactor scram.
This event was discussed in NRC inspection report 89-30, paragraph 3.
(Closed) LER 89-17, RWCU line break analysis does not reflect changes in isolation provisions.
This event was discussed in NRC inspection report 89-29, paragraph 6.
The licensee performed a new line break analysis, which still bounded the existing design limits and completed a 10 CFR 50.59 safety evaluation. This item is closed.
(Closed) LER 90-001, Technical Specification action delinquent due to programmatic deficiencies and personnel error.
This event was discussed in NRC inspection report 89-30, paragraph 3. This LER is administratively closed and the corrective actions will be tracked under violation 89-30-01.
'
,
On February 7, 1990, the licensee made a 10 CFR 50.72 (b) (1)(v) report on the out of service condition of the operational hot lines to the state and local agencies.
The ring down circuit failed to work.
A grounding problem was found and repaired. The hot line was satisfactorily retested.
On February 15, 1990, the licensee made a 10 CFR 50.72 (b)(1)(ii)(A)
report on the HPCS service water system.
During a design bases LOCA in which the single active failure is the loss of ESF electrical division 1, the division 1 ECCS and support systems (e.g. division 1 SSW pump and cooling tower fans, the RHR A pump and the LPCS pump) would not be available for long-term core cooling.
Therefore, the HPCS system and its support system along with division 2 ECCS would be relied upon for
,
long-term core cooling.
The HPCS SSW system, by design, pumps cooling water from the division 1 cooling tower basin to the division 3 components through the HPCS service water header. After removing heat from the HPCS
,
components, the HPCS SSW is discharged to the main SSW header which discharges to the division 1 cooling tower.
The return flow of HPCS SSW without the return flow of division 1 SSW would be insufficient to provide effective spray cooling for HPCS SSW.
It is estimated that the division 1 basin temperature would increase from its initial 75 degrees F and exceed
,
.
,
.
,
'
the 90 degrees F design temperature limit within 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br />.
Long-term cooling of the HPCS SSW may not be assured.
Nuclear Plant Engineering held discussions with GE and they indicated that long-term core cooling could be accomplished after 3 days without a core spray system. GE will perform an assessment to validate this information.
The review of the assessment, that HPCS is not required after three days, will be an
unresolved item 90-02-03.
On February 15, 1990, the licensee made a 10 CFR 50.72(b)(2)(iii) report on the diesel generator sprinkler systems.
The licensee determined that
pre-action valves for the diesel generator sprinkler system have not been qualified to demonstrate that the valves would not actuate during a seismic event, thereby charging the system.
The sprinkler piping or sprinkler heads have not been evaluated to show adequate design margin to
.
prevent leakage during a seismic event.
i The diesel generators are not qualified to operate while being subject to
>
water spray.
The licensee has isolated the fire sprinkler header and
,
stationed a permanent fire watch until a design change can be implemented.
No violations or deviations were identified.
8.
Action on Previous Inspection Findings (92701,92702)
?
(Closed) Violation 88-07-02, Failure to perform a written safety
,
evaluation regarding the tagging shut of RCIC drain valves.
The licensee denied the violation in a response dated June 2,1988. After NRC review, the violation was withdrawn in NRC letter dated January 30, 1990.
This item is closed.
(Closed) Inspector Followup Item 88-19-03, Install a keylock. bypass switch for delta flow.
The licensee conducted additional evaluations of the RWCU flow perturbations and determined that a design change was not necessary.
l The system operating procedure was revised to modify the lineup and reduce erroneous flow indications.
These corrective actions are documented in LER 89-07-01. This item is closed.
(Closed) Inspector Followup Item 89-16-01, Resolve 10 CFR 50.59 recommen-dations.
The recommendations made in NRC letter dated May 22, 1989, have
'
been or will be completed by the end of refueling outage four. This item is closed.
(Closed) Violation 89-19-01, Failure to follow chemistry 3rocedures, four examples.
The licensee admitted the violation in letter cated October 12, 1989. The following corrective actions have been completed. Surveillance Procedure 06-CH-1D17-W-0017, Gaseous Release Points, lodines, Tritium, and Particulates; was revised to visually inspect the filter cartridge.
Administrative procedure 01-S-03-3, Material Non Conformance reports, was revised to include a note on instrument air sample failures.
Chemistry
]
- .
-
..
,
i procedure 08-S-03-21, was revised to require a MNCR for failed instrument air sampics. Chemistry documentation was transmitted to document control.
A quality programs audit was performed on the chemistry department. This
item is closed.
9.
Exit interview (30703)
i The inspection scope and. findings were summarized on February 16, 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 the inspectors during this inspection. The licensee had no comment on the
following inspection findings:
Item Number Description and Reference 90-02-01 Inspector Followup Item l
Review revision to Modification Work Permit Procedure, paragraph 3
90-02-02 Inspector Followup Item Review the licensee corrective action for improving the MNCR process, paragraph 4
,
UNR 90-02-03 Unresolved Item Evaluate the adequacy of the long-term cooling capabilities for HPCS SSW system, paragraph 7
10. Acronyms and Initialisms Automatic Depressurization System ADS
-
APRM -
Average Power Range Monitor i
Control Rod Drive CRD
-
DCP Design Change Package
-
Diesel Generator D/G
-
ECCS -
Emergency Core Cooling System Engineering Safety Feature ESF
-
FSAR -
Final Safety Analysis Report HPCS -
_High Pressure Core Spray General Electric Company GE
-
I&C Instrumentation and Control
-
Inspector Followup Item IFI
-
IR Incident Report
-
ISEG -
Independent Safety Engineering Group
Limiting Condition for Operation LCO
-
LER -
Licensee Event Report Minor Charge Package MCP
-
...
..
,
.
...
-.
10-
!
MNCR -
Material Nonconformance Report'
Maintenance Work Order MWO
-
Nuclear. Plant Engineering i-NPE
-
Nuclear Regulatorv. Commission
!
NRC
-
-Quality Deficiency Report
.!
-
RAIR -
Review and Approval Implementation Record
RCIC -
Reactor Core Isolation Cooling i
-
RWCU -
SLCS --
Standby Liquid Control: System
!
Safety Relief-Valve i
-
SSW -
Standby Service Water-
!'
Temporary Change Notice TCN
-
Technical Specification TS
-
WO Work Order i
-
Quality Deficiency Report l
-
i; I
!
.'
i
.
!
!'I l,
~!'I f
I-l i
,
-
l l
'i i
d
-l
,
'
.
.