ML20246B280
| ML20246B280 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 06/27/1989 |
| From: | Cantrell F, Christensen H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20246B259 | List: |
| References | |
| 50-416-89-16, NUDOCS 8907070264 | |
| Download: ML20246B280 (11) | |
See also: IR 05000416/1989016
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UNITER) STATES
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NUCLEAR REGULATORY COMMISSION
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REGION 11
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101 MARIETTA ST, N.W.
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ATLANTA, GEORGIA 30323
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. Report No.: 50-416/89-16
Licensee:
' System Energy Resources, Inc.
Jackson, MS 39205
Docket No.: 50-416
License No.: NPF-29
Facility Name:
Grand Gulf Nuclear Station
Inspection' Conducted: May 20 - June 16, 1989
Inspectors:
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K 0. Christensen, Ffe f6r#esid' nt inspector
Da'te Signed
e
Yb45
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d/D/f9
YL Mathis, Residenyryfpsct'Or
D6te Siigned
Approved.by:
dk/d
F. 3. 'Cahlrell, Sectidrif#jtbf
Dste 5'igned
Division of of Reactor +rojects
SUMMARY
Scopp:
The resident inspectors conducted a routine inspection in the following areas:
operational- safety verification, maintenance observation, surveillance
. observation, engineering safety features (ESF) system walkdown, 10 CFR Part 21
. procedures, action on previous-inspection findings, and reportable occurrences.
The inspectors conducted backshift inspections on May 20, 21, 23, 24, 29 and
June 13, 1939.
Results:
Within the areas inspected two violations were identified:
Failure to take
adequate corrective action to prevent RCIC system isolations, paragraph 9,
and
failure to maintain post-accident sample system design control, adequate
training, and perform required samples, paragraph 9.
The post accident sample system problems appear to be related to the licensee
treating the system as a non-safety, non-technical specification system (no
= limit conditions for. operations).
Controls placed on safety-related systems
are more inclusive and the plant staff appears to focus more attention to the
safety related system status as compared to non-safety related system.
The recirculation pump shaft replacement outage was well managed.
8907070264 890627'
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ADOCK 05000416
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REPORT DETAILS
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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
- J.'P.
Dimmette, Manager, Plant Maintenance
S. M. Feith, Director,' Quality Programs
- C. R. Hutchinson, GGNS General Manager
F. K. Mangan, Director, Plant Projects and Support
R. H. McAnuity, Electrical Superintendent
A. S. McCurdy, Technical Asst., Plant Operations Manager
- L. B. Moulder, Operations Superintendent
J. H. Mueller, Mechanical Superintendent
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
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J. W. Yelverton, Manager, Plant Operations
G. Zinke, Superintendent, Plant Licensing
Other licensee employees contacted included technicians, operators,
security force members, and office personnel.
- Attended exit interview
F. S. Cantrell, Section Chief, Division of Reactor Projects, was on site
May 30 and 31, 1989, to conduct a plant tour and hold discussions with the
resident inspectors.
2.
Plant Status
Unit 1 began the inspection period in a recirculation pump maintenance
outage.
On May 31, 1989, Unit i restarted and returned to power
operations.
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
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inspection activity included numerous informal discussions with operators
and their supervisors.
On a weekly bases selected engineered safety feature 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:
Suppression pool makeup,
control rod drive system, HPCS and LPCS.
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 the
readiness of the onsite emergency response facilities.
The inspectors observed health physics management involvement and
awareness of significant plant activities, and observed plant radiation
controls.
Additionally the inspectors verified 'the adequacy of physical
security controls.
The inspector reviewed safety related tagout 892996 (Equipment drain sump
pump).
The review ensured that the tagout was 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; E22-F035, E22-F022 and E12-F339.
The ' inspectors noted that senior plant management makes routine tours to
the plant and the control room.
The inspectors reviewed activities associated with the failure of the B
recirculation pump shaft.
On May 15, 1989, the plant was taken to cold
shutdown due to high vibration on the B Recirculation Pump.
Upon
disassembly, a 300 degree crack was discovered on the lower pump shaft.
Both A and B recirculating pump rotating elements were replaced.
The
details of the shaft failure are documented in NRC inspection report
50-416/89-15. The recirculation pump maintenance outage was well planned,
scheduled and managed.
The plant conducted an orderly restart on May 31,
1989.
No v1alations or deviaticns were identified.
3.
Maintenance Observation (62703)
During the report period, the inspectors observed portions of the
maintenance ' Stivities listed oelow.
The observations included a review
of the i'WE .nd other related documents for adequacy; adherence to
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. procedure, proper 'tagouts, technical specifications, quality controls, and
radiological controls; observation of work and/or retesting; and specified
retest requirements.
MWO
DESCRIPTION
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DCP 86/0085
Extension.of the upper containment pool weir wall
'M 93729
SSW B basin fan D repair
M 93647
Recirculation pump internal inspection
M 93480
Recirculation pump internal removal and replacement
M 934E2
Recirculation pump assemble of spare rotating assembly
EL 11r3
Megger Motor (SBLC) from the breaker
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EL .'1.0
Inspect MOV C41F001B
ME 34m
Unit 1 Instrument Air Dryer Desiccant Change Out
No violations or deviations were identified.
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 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.
06-0P 1P75-M-0001, Standby Diesel Generator (SDG) II Functional Test,
Attachment 11
06-EL-1E31-M-0001, RCIC Main Steam Tunnel Isolation Delay Timer Channel
A Functional Test and Calibration, Attachment 1
06-IC-1E31-M-0022, Drywell Air Cooler Condensate Flow Rate Monitoring
Functional Test
06-IC-1E32-M-1002, MSIV Leakage Control System Functional Flow Test,
Attachment 1
06-IC-1C34-M-0001, Reactor Vessel Water Level (Level B) MT/RFPT Trip
Function Test
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06-IC-1D17-A-0012, Fuel Handling Area Ventilation Radiation Monitar
Calibration
06-IC-1E31-M-2003, Main Steam Line "C" High Flow Functional Test
No violations or deviations were identified.
6..
Engineered Safety Features System Walkdown (71710)
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The inspectors conducted a complete walkdown on the accessible portions of
the standby gas treatment system.
The walkdown consisted of the
following:
confirm that the system lineup procedure matches the plant
drawing and. the as-built configuration; identify equipment condition and
items that might degrade plant performance; verify that valves in the flow
path are in correct positions as required by procedure and that local and
remote position indications are functional; verify the proper breaker
position-at local electrical boards and indications on control boards; and
verify.that instrument calibration dates are current.
The inspectors walked down the system using system operating instruction
04-1-01-T48-1, Revision 19, SBGT and P&ID M-1102 A and B, SBGT system.
The monthly operability test for SBGT systems A and B were performed
satisfactory and the 18 month system logic and vacuum test was
successfully performed.
Additionally, the 18 month calibrations for
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drywell high pressure and reactor vessel water level were performed. The
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annual fuel handling area ventilation exhaust radiation monitor
calibrations were also performed successfully.
The SBGT electrical lineup was verified by using attachment III to the
system operating instruction.
All electrical breakers were in the
required position.
The instructions component description differed from
the breakers label name for all breakers.
A labeling program has been
implemented to correct all labeling deficiencies.
All annunciators and valve positions were in accordance with the system
operating instruction.
No violations or deviations were identified.
7.
10 CFR Part 21 Inspection (36100)
The inspectors verified that procedures and controls were established and
-implemented for 10 CFR Part 21 requirements. The initiating document for
Part 21 is through a Material Nonconformance Report (MNCR).
MNCRs are
used to document discrepancies concerning material-related documentation,
i.e., test results, certification, and etc. Administrative Procedure
01-5-03-3, Material Nonconformance Reports designates Nuclear Plant
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Engineering (NPE) as the organization responsible for evaluating whether
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deficiency or nonconformances constitute deportability pursuant to l') CFR
Part 21.
Additionally, Quality Programs screens all nonconformance
reports for potential Part 21.
Quality Assurance Procedure (QAP) 6.40,
Potential Reportable Deficiency Screening, provides guidance to be used by
the screening teviewer for Part 21.
The inspectors reviewed procedure 01-S-09-1, Revision 24, Procurement of
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Materials and Services, which requires 10 CFR 21 be required on all
Quality Level 1 and 2 material item procurement and on all Quality Level
1, 2 and 3 service contracts. The inspector performed a random sample of
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purchase orders written after January 6,1987.
All four purchase order
. contained the 10 CFR 21 applicability statement.
In addition the inspectors
selected two evaluated deviations or nonconformances not resulting in a
report to the commission to verify the following:
The item was identified for evaluation consistent with established
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procedures.
The information and data used in the evaluation appear to be factual
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and complete.
The nonconformance was evaluated, or forwarded to the purchaser for
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evaluation consistent with established procedures.
Overall it appears that the licensee has an effective program in place for
evaluating 10 CFR 21 requirements.
8.
10 CFR 50.59 Safety Evaluation
In a May 22, 1989, NRC letter to SERI, NRR documented the NRC staff's
safety evaluation results for five SERI 10 CFR 50.59 safety evaluations.
Corrective actions were recommended in three areas:
Revision of the SERI safety evaluations NPE-86-279 and PLS-86-123 to
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include adequate bases to support a determination that the changes do
not involve unreviewed safety questions.
Revision of the UFSAR to include information to show how safety
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significant cranes meet NUREG-0612, as discussed in NRC evaluation of
NPE-86-279.
Revision of the surveillance procedure for TS 4.6.6.1.b to require
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that drawdown test of secondary containment be run with the primary
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containment hatch open as discussed in the NRC evaluation of
PLS-86-136.
Tne resolution of the above recommendation will be an inspector followup
item 89-16-01.
9.
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.
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.a.
(Closed) LER 89-005, Reactor Core Isolation Cooling System Isolations
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on Indicated High Steam Line Flow. The RCIC isolations of April 29,
and May 8,1989, were documented in NRC inspection report 89-14.
Upon completion of the investigation, it was determined that the
isolations were caused by a spurious high flow signal produced by
pressure oscillations in a sensing. line for a RCIC steam line
differential pressure transmitter. The sensing lines were backfilled
with demineralized water to reduce the amount and amplitude of the
oscillation and the damping pots were increased to give a stable
signal to the transmitters.
Similar isolations occurred in December
1984, as reported under LERs 84-56 and 84-57.
The 1984 correction
was to adjust the damping of the transmitters to depress the level of
oscillations.
The adjustments were completed by early 1985. During
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refueling outage three, I&C technicians replaced the electronic
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circuit boards for the transmitters.
The work package required the
damping pots to be set to the as found setting.
The old pots were
found to be set at minimum damping, therefore the technicians set the
new circuit board pots to minimum.
A records search of work
performed on the transmitters did not reveal when or how the old pots
were readjusted.
The failure to adequately control the damping
values for the RCIC transmitters is a violation of 10 CFR 50
Appendix B, Criterion XVI Corrective Actions. The LER 89-005 will be
administratively closed and corrective actions will be tracked under
violation 89-16-02.
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b.
.0n June 7, 1989, the shift superintendent made a one hour report per
10 CFR 50.72.(b).(v) on the loss of post-accident sampling assessment
capability.
While performing a post modification retest, for DCP
87-4018, PAS system atmospheric and liquid panel modification, it was
determined that two valves, P33F719 and P33F720, which were required
to be deleted by t.he design change were installed and isolated,
preventing the capability to take a post-accident liquid radionuclides
sample.
Further investigation determined other design package
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deficiencies, which include:
Inadequate post modification system
walkdowns, several valves and piping not in accordance with as-built
drawing, inadequate post modification test'ing, performed system
retest approximately one month after completion of modification and
at 100% reactor power; and inadequate design package review and
implementation.
The package
contained vendor equipment numbers
which were not converted to SERI system numbers and the package was
not translated into adequate work instructions, that is valves
P33F719 and P33F720 were left installed contrary to design intent.
It was also determined that the system had not previously been
operated in accordance with the plants license condition, TS, or
Licensing Condition 2.C.(33).(c), requires SERI to
incorporate the requirements of Safety Evaluation Report, Supplement
4 (SSER 4), into procedures.
SSER 4. Section 22.2,II.B.3, requires a
post-accident sampling program be performed on a semiannual basis and
consists of obtaining and analyzing reactor coolant, suppression
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pool, and RHR samples chemically and radiochemically by persons
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responsible: for post-accident procedures.
SERIs Emergency Plan,
' Section 7.6.4, states that systems are installed to obtain samples
. from the following . locations:
...RHR A and B; drywell atmosphere;-
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and suppression pool." TS 6.8.3.c . requires a post-accident sampling
program be established, implemented and maintained which will ensure'
the capability to obtain and analyze a sample under accident-
conditions,- training of personnel, procedures for sampling and -
analysis',and. provisions for maintenance of the equipnent.
A review of past sampling data, January 1986 to June 1989, indicated
that semiannual samples were not conducted on RHR or suppression pool
and
the RHR L B and drywell sample paths _ were never ' tested.
Additionally, discussions with the systeri engineer and plant-
chemistry indicated that they believed a sample could only be taken
during full' system pressure.
Chemistry personnel attempted to take
samples' on April 24 and 27,and May 17,18, and 19,1989, when the
plant wasl sh'ut down and depressurized. Only small amounts of water
were obtained,- this l'ed them to believe that a sample path was
available.
The system engineer and plant chemist decided to sample
at pressurized conditions due to the sample difficulties at
depressurized conditions.
The PAS system was declared operable on
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April' 27, 1989.
The . plant operated at various power levels for
approximately 20 days prior to conducting a post modification sample
test on June.6, 1989. Chemistry was unable to obtain a sample and on
June 7,1989, the valves P33F719 and P33F720 were discovered in the
syste'n. The licensee performed a dose rate assessment calculation to
determine _'the radiation ~ 1evel that would be present during accident
conditions at the area of the two isolated valves.
The calculation
determined an exposure field of 800 rem per hour on contact.
(Other
locations _are available that could be used to obtain a sample
although the exposure maybe higher than would be expected from an
operable PAS system.)-
The. failure to maintain the PAS system ooerable, to' conduct
semiannual samples of the RHR and suppre
a pool, to demonstrate
the drywell and RHR B sample oaths, and to adequately train the
system engineer and plant chemistry in PAS system operations is a
violation of TS 6.8.3.c, and 10 CFR 50, Appendix B, Criterion Ill.
This will be documented as violation 89-16-03.
c.
On June 7, and 8, ~1989 during a hotline phone check, none of the
offsite- emergency agencies could be contacted by the TSC or the
control room emergency notification system. The licensee immediately
notified the telephone company to correct the problem. This incident
was also called into the NRC as a one hour reportable event per
d.
On June 8, 1989 a power outage in Claiborne County led to 22 a rens
being without power, the licensee reported the incident to NRC per
50.72(b)(1)(v).
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10. Action on Previous Inspection Findings
(92701,92702)
(Closed) IFI 86-02-03, Inspector Followup Item, NRR conducted review of
BWR technical specifications (TS) governing requirements'for safety relief
valve -(SRV) operability and if changes to TS should be made.
concluded that they do not have a safety concern for those BWR reactors
which have TS allowing unlimited operation with more than one SRV out of
service because:
(1) the number of SRVs required to be operabl.e by TS are
based on the overpressure protection analysis as described in the Standard
Review Plan 5.2.2, (2) the number of SRVs out of service up to the maximum
allowed by TS is unlikely to affect fuel cladding integrity during
transients and accidents described in the FSAR Chapter 15, and (3) there
are no regulatory requirements for an ATWS analysis.
However, the
discrepancy between the TS limiting conditions for operation with SRVs out
of service - and the GE recommended administrative procedure shou'ld be
resolved.
As a followup to this evaluation, NRR will address the inconsistency
between the TS and the GE recommended procedure with the BWR Owners Group
as a part of the TS Improvement Program.
The Generic Communi~ ations Branch, NRR has determined that no action is
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warranted to inform affected licensees because operation with the present
TS is not a safety significant concern.
This item is closed.
(Closed) IFI 89-04-04, Incorporate GE SIL 319 recommendations into PN
program.
Maintenance Planning & Scheduling System initiated task card
ME6044 to inspect RCIC Turbine Drive Gear Assembly every 18 months.
Procedure 07-2-14-301, RCIC Turbine Drbre Gear Assembly Inspection, is the
procedure used by mechanical maintenance personnel to inspect the RCIC
gear assembly. This item is closed.
(Closed) IFI 8b-09 n6, Design change on precoat filter isolation causing
The inspectors reviewed DCP 84/3000 and MWO 52146 which provided
an automatic start of the instrument air compressor on low air receiver
pressure and allowed operators to restart the air compressor remotely from
the control room whether it is being operated in the automatic or manual
mode. This item is closed.
(Closed) URI 86-01-02, Review test exceptions to level 1 startup test
acceptance for 10 CFR 50.59 consideration.
The inspectors reviewed IPC
86/0296 and IPC 86/363, which reviewed all startup exceptions written to
document level I criteria failures.
The findings of t,he review verified
that no further FSAR changes were required. This item is closed.
(Closed) Violation 88-07-03, Violation, Failure to follow procedure for
procurement of materials.
The licensee revised the warehouse inventory
computer system to reflect the voltage level and type as part of the model
number for all ASCO solenoid valves in stock.
Future ASCO solenoid valves
will be stored according to voltage level and type.
This item is closed.
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(Closed) IFI 88-19-02, Inspector Followup Item, Implementation and repair
of 3/4" HCD-73 line.
MWO M91252 implemented MCP 89/1011 during RF03 to
repair the vent line for valve Q1833F023A and repair the 4" drain hub.
This MCP modified the vent line and replaced valves B33F025A and B33F026A
with identical valves. This item is closed.
(Closed)IFI 87-16-05, Installation and testing of warning system for high
noise areas of the plant.
Visual alarms were installed and tested under
DCP 84/0231. This item is closed.
11.
ExitInterview(30703)
'The inspection scope and findings were summarized on June 16, 1989, 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 General flanager stated that
they view the PAS system problem very serious and that it is a result of
multipule breakdowns. .The licensee had no additional comments on the
following inspection findings:
Item Number
Description and Reference
89-16-01,
IFI
Resolution of 10 CFR 50.59 reviews,
paragraph 8.
'89-16-02,
V10
Failure to adequately control the damping
values for RCIC transmitters, paragraph 9.
89-16-03
V10
Failure to maintain the PAS system design
control, to conduct adequate training and
perform required samples, paragraph 9.
12. Acronyms and Initialisms
ADHRS-
Alternate Decay Heat Removal System
Automatic Depressurization System
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APRM -
Average Power Range Monitor
CRD -
Control Rod Drive
Design Change Package
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Diesel Generator
ECCS -
ESF -
Engineering Safety Feature
FCV -
Flow Control Valve
FSAR -
Final Safety Analysis Report
HPCS -
Hydraulic Power Unit
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1&C -
Instrumentation and Control
Inspector Followup Item
IFl
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LCO
Limiting Condition for Operation
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LER -
Licensee Event Report
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LPCI -
Low Pressure Core Injection
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LPCS -
Low Pressure Core Spray
MNCR -
Material Nonconformance Report
MWO
Maintenance Work Order
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NPE
Nuclear Plant Er.gineering
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Nuclear Regulatory Commission-
NRC
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PASS -
Post Accident Sample System
Pressure Differential Switch
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P&lD -
Piping and Instrument Diagram
PSW
Plant Service Water
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Quality Assurance Procedure
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-Quality Deficiency Report
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RCIC -
Reactor Core Isolation Cooling
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RWCU -
Radiation Work Permit
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SBLC -
Standby Diesel Generator
SDG
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SERI -
System Energy Resource Incorporation
501
System Operating Instruction
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Standby Service Water
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Temporary Change Notice
TCH
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Technical Specification
TS
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UFSAR-
Updated Final Safety Analysis Report
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