ML20234E994
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| ML20234E994 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 06/29/1987 |
| From: | Bennett W, Chamberlain D, Jaudon J, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20234E977 | List: |
| References | |
| 50-458-87-14, NUDOCS 8707080064 | |
| Download: ML20234E994 (13) | |
See also: IR 05000458/1987014
Text
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APPENDIX B
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-458/87-14
Dceket: 50-458
Licensee: Gulf States Utilities Company (CSU)
P.O. Box 220
St. Francisville, Louisiana 70775
Facility Name: River Bend Station (RBS)
Inspection.At: River Bend Station, St. Francisville, Louisiana
Inspection Conducted: May 1 through June 15, 1987
Inspectors: 1 /[- [[ ;
D. D. \ chamberlain, Senio'r Resident Inspector Date
Project Section A. Reactor Projects Branch
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W. B. Jones, R n Inspector Date
Project Sectibn A. Reactor Projects Branch
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W. R. Bennett, Project Engineer, Projects Date
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Section A Readtor Pr ects Branch
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Approved: d- k f l/ / [ //1 0 .
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( Rdactor rojects Branch
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Inspection Summary
Inspection Conducted May 1 through June 15, 1987 (Report 50-458/87-14)
Areas Inspected: Routine unannounced inspection of licensee action on previous
inspection findings, 10 CFR Part 21 Reports, maintenance witnessing, safety'.
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system walkdown, surveillance test witnessing, operational safety verification,
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inspection of stainless steel piping and in-office revieu of written reports
of non-routine events at power reactor facilities.
Results: Within the areas inspected, one violation was identified (failure
to follow surveillance test procedure, paragraph 6).
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DETAILS
1. Persons Contacted
W. J. Beck, Supervisor, Reactor Engineering
*J. E. Booker, Manager, Oversight
E. M. Cargill, Supervisor, Radiation Programs
J. W. Cook, Lead Environmental Analyst, Nuclear
Licensing
T. C. Crouse, Manager, Quality Assurance (QA)
*J. C. Deddens, Senior Vice President, River Bend
Nuclear Group
*C. L. Fantacci, Supervisor, Radiological Engineering
P. E. Freehill, Outage Manager
A. O. Fredieu, Assistant Supervisor, Operations
*D. R. Gipson, Director, Quality Services l
P. D. Graham, Assistant Plant Manager, Operations
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*E. R. Grant, Director, Nuclear Licensing
J. R. Hamilton, Director, Design Engineering
R. W. Helmick, Director, Projects
K. C. Hodges, Supervisor, Chemistry
*L. G. Johnson, Site Representative, Cajun
G. R. Kimmell, Supervisor, Operations QA
*R. J. King, Supervisor, Nuclear Licensing
A. D. Kowalczuk, Assistant Plant Manager, Maintenance
I. M. Malik, Supervisor, Quality Systems
*V. J. Normand, Supervisor, Administrative Services
*W. H. Odell, Manager, Administration
*T. F. Plunkett, Plant Manager
*M. F. Sankovich, Manager, Engineering
*R. R. Smith, Engineer, Nuclear Licensing
R. B. Stafford, Director, Operations QA ,
*K. E. Suhrke, Manager, Project Management
R. G. West, Supervisor, Instrumentation and Controls
D. W. Williamson, Supervisor, Operations
The NRC also interviewed additional licensee personnel during the
inspection period.
* Denotes those persons that attended the exit interview
conducted on June 16, 1987,
2. Licensee Action on Previous Inspection Findings
a. (Closed) Open Item (458/8615-01): Process computer assignment of the
Transverse In-Core Probe (TIP) strings were not in agreement with the
as-built TIP configuration.
The licensee initiated a process computer software change to reassign
the TIP strings to correspond to the as-built configuration. The TIP
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machines, machine indexes, Local Power Range Monitor (LPRM) pseudo ;
strings, and core locations specified on the 0D-1 output from the '
process computer were compared with the as-built drawings and found
to be in agreement. Core thermal limits were then calculated at 20
and 40 percent power levels using the corrected TIP locations for the
00-1 program and the results compared to previous core thermal limit l
calculations. The results verified that no thermal power limits were
violated.
This item is closed.
b. (Closed) License Condition 2.a: Complete and have operational the
fuel building sampling system.
The NRC inspector reviewed the preoperational test for the system and
determined that it was satisfactory. The NRC inspector also verified !
that flow could be obtained from all sample points required in the
FSAR. The inspector ensured that procedures are adequate to obtain
samples and that the system has been utilized.
This item is closed. !
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c. (Closed) Violation (458/8639-01): Failure to follow surveillance I
testprocedures(STPs). I
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Corrective action included revising STP-051-4234 and issuing a !
memorandum to all I&C personnel to reiterate the administrative >
requirements of the surveillance test program. The NRC inspector
reviewed Revision 7 to STP-051-4234 and Memoranda PIC-M-87-007 and
PIC-M-87-033 to ensure that corrective actions have been properly
performed.
This item is closed.
d. (Closed) Violation (458/8639-02): Failure to utilize proper revision
of surveillance test procedure.
Corrective action included revising Administrative Procedure ADM-015
to clarify the requirement for ensuring that procedures utilized in
the plant are current. The NRC inspector reviewed Temporary Change
Notice (TCN) 87-0268 to ADM-015, Revision 9, to ensure corrective
actions were properly performed.
This item is closed.
e. (Closed) Deviation (458/8624-01): Failure to verify and validate
certain abnormal operating procedures.
The licensee incorporated the abnormal operating procedures mentioned
in the deviation into emergency operating procedures (E0P's).The NRC
inspector reviewed E0P's 0101, 0104 and 0107 to verify proper
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incorporation. The NRC inspector interviewed licensee personnel to I
ensure that the procedures had been properly verified and validated, j
and that operations department personnel had been trained on the !
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procedures prior to implementation.
This item is closed,
f. (Closed) Unresolved Item (458/8702-02): Adequacy of the disposition J
of a nonconformance regarding temperature effects on rubber boot
material and concrete structural integrity for shield building
penetrations. (Reference Allegation Number 4-86-A-092)
The SRI was provided with copies of data taken during the startup
test program which indicated maximum temperature readings in the
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115 F range at locations in the main steam tunnel around the main
l steam line penetrations. The licensee also took additional
temperiam: >>adings and air flow measurements in the steam tunnel
during this inspection period which compared favorably with the
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startup test data. This data was used to refine assumptions for
l modeling the expected temperature gradients for the concrete l structure end penetration boot seals. The SRI reviewed l CalculaMs 12210-ES-250, " Steady State Temperature Profile in the l Shield Niilding Wall at the Main Steam Line Penetrations Z1A, Z1B, '
Z1C and Z10," and Calculation 16138.11-STR-201.120.194, " Evaluate the
Effect of Increase in Temp in Shield Bld M.S. Tunnel Area." The
results from these calculations indicate a maximum temperature for
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the rubber boot magerial of 340 F, which is below the design
temperature of 400 F, and g maximum concrete temperature of 247 F. A )
maximum temperature of 250 F was used to evaluate the reduction in ;
allowable stresses in the concrete because of the increased 1
tempe ratu res. The reduction in the strength of the concrete was
determined to be 10 percent. The calculated strength margin was
* 12.8 percent which leaves a total remaining margin of 2.8 percent.
Therefore, the concrete structure would perform as designed with the
increased temperatures. i
This unresolved item and Allegation No. 4-86-A-092 are closed. ,
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g. (Closed) Violation (458/8627-01): Failure to meet electrical
specification requirements for enclosure of an electrical cable
splice on the Division III batteries. !
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The licensec initiated a modification request (MR 86-1752) to enclose i
the cable splices in accordance with the intent of the electrical i
specification. MaintenanceWorkOrderRequest(MWOR) 101810 was l
completed to implement the modification. A review of the other
safety-related batteries and of other plant areas by the licensee
revealed no other problems with splice enclosure.
This violation is closed.
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3. 10 CFR Part 21 Reports
The SRI was provided copies of selected 10 CFR Part 21 Reports by NRC
Region IV, which may be applicable to equipment or services supplied to l
River Bend. These reports were provided to the licensee, who verified
that the reports either had been or were being evaluated for applicability !
at River Bend. Any reports that were not already entered into the !
licensee tracking system were immediately entered. A listing of reports
by date, manufacturer, and subject is provided below:
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a. 7/21/86 - General Electric Company - HFA auxiliary relay armature
binding.
b. 12/19/86 - Limitorque Corporation - Nuclear grade D.C. motors lead l
wire failure,
c. 12/9/86 - General Electric Company - AMP butt splices found not
qualified during simulated LOCA-test because of excessive leakage
current.
d. 11/10/86 - Limitorque Corporation - Motor burnout occurring because ,
of new less viscous grease leaking from the operator housing into the (
spring back area creating a hydraulic lock.
e. 8/20/86 - So11 tech, A Division of Industrial Engineering and
Equipment Company - Failure of terminal blocks on Solitech power
controllers in standby gas treatment system.
f. 1/27/87 - Boston Edison Company - Standby gas treatment system
humidity control sub system could fail because of a single failure to
a temperature switch heater or motor starter relay.
g. 2/17/87 - Foxboro Company - Capacitors may fail on Spec 200 cards
stored in a high moisture environment without power supplied.
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h, 2/10/87 - Limitorque Corporation - Geared limit switch rotors warped
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1. 1/13/87 - Morrison-Knudsen Company, Inc. - Square D Company general
purpose relay failed to drop out when deenergized.
j. 3/20/87 - Bingham /Willamette - Design causes RHR pump minimum flow to
be higher then specification value.
k. 3/30/87 - Isomedix, Inc. - Minimum radiation doses stated for
equipment qualification cannot be assured because of allowed
tolerances.
The resident inspectors will continue to provide copies of potentially
applicable 10 CFR Part 21 Reports for licensee evaluation, and a followup
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of licensee action on selected 10 CFR Part 21 Reports will be conducted
during future NRC inspections.
No violations or deviations were identified in this area of inspection.
4. Maintenance Witness
During this inspection period, the RI observed portions of selected
corrective maintenance activities to verify that maintenance activities
are being conducted in accordance with approved procedures, Technical
Specifications (TS) and appropriate industrial standards and codes. The
following activities were observed:
. Division I Diesel Generator Jacket Water Leak Repair - On May 18,
1987, the RI observed maintenance activities performed under
Maintenance Work Order (MW0) 102857 which was initiated to replace
the gaskets on the jacket water header to each of the eight
cylinders. Prior to the licensee beginning work, the RI verified
that the required administrative approvals and tagouts were obtained.
. Division 1 Diesel Generator Lube Oil Pump - On May 20, 1987, the
licensee initiated MWO 55274 to repair the electrically driven lube
oil pump after excessive oil leakage developed around the shaft seal.
The quality control (QC) inspector was present during the maintenance
activity. QC Inspection Report 87-21330 was initiated as required to
document the inspectors observations.
No violations or deviations were observed in this area of the inspection.
5. Safety System Walkdown
On June 3, 1987, the SRI and RI performed a walkdown of the low pressure
coolant injection (LPCI) mode of the "B" residual heat removal (RHR)
system. This system is required to be operational during operational
conditions 1, 2, and 3 with an established flow path capable of taking
suction from the suppress' ion pool and transferring the water to the
reactor vessel.
The system walkdown revealed the following:
. system valves located on the major flow paths were properly aligned;
. associated instrumentation was properly aligned;
. no abnormal control room instrumentation readings or alarms were
present which would prevent LPCI "B" from responding if required;
. the RHR pump bearing oil reservoirs were properly filled;
. accessible hangers and snubbers were intact; and
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. the electrical switchgear was properly aligned.
t , no problems were noted which would have affected LPCI "B"
operability.
No violations or deviations were identified in this area of the
inspection.
6. Surveillance Test Witness
the RI observed the performance of
During this inspection
surveillance period [STP) STP-511-4501, "RPS/ ISOLATION ACTUATION
test procedures
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- MSLI - MAIN STEAM LINE RADIATION - HIGH MONTHLY CHFUNCT (DI7-K610A),"
STP-051-4267, "ECCS DRYWELL PRESSURE - HIGH MONTHLY CHFUNCT; 18 MONTH
CHCAL; 18 MONTH LSFT (B21-N0948B; B21-N694B)," STP-209-3302, "RCIC PUMP 1
OPERABILITY and FLOW TEST," and Reactor Engineering Procedure REP-0014,
"USE OF TIP MACHINE."
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. STP-511-4501 - This STP was performed on May 16, 1987, to meet the
channel functional test for the RPS/ ACTUATION MSLI - MAIN STEAM LINE
RADIATION HIGH instrumentation (D17-K610A) as required by Technical
Specification (TS) Section 4.3.1.1 and 4.3.2.1 Tables 4.3.1.1-1.7 and
4.3.2.1-1.2.b. During the performance of the STP, the RI noted that
each of the technicians performing the STP had not signed
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prerequisite Step 6.1, indicating that he had read and understood the
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procedure prior to proceeding with Section 7 of the surveillance
test. This failure to sign and initial the prerequisite step was
identified by the RI as an apparent violation (458/8714-01). Review
of the final STP package indicated the procedure had received the
required operations review and that the surveillance acceptance
criteria had been meet.
. STP-051-4267 - This STP was performed on June 4,1987, to meet the l
31-day channel functional test of the ECCS Drywell Pressure High
(B21-N094B; B21-N694B) instrumentation as required by TS Section
4.3.3.1 Tables 4.3.3.1-1.B.1 and 4.3.3.1-1.B.2.b. During the
performance of this test, the RI the also noted that each of the ;
technicians performing this STP had not signed prerequisite Step 6.1, l
indicating that he had read and understond the procedure prior to
proceeding with the surveillance test. This failure to sign and l
Initial the prerequisite step was identified by the RI as an 1
additional example to apparent violation (458/8714-01). No
deficiencies with the actual performance of the STP were identified.
. STP-209-3302 - This STP was performed on June 5, 1987, to verify the
RCIC pump develops an initial flow greater than or equal to 600 gpm
in the test mode as required by TS 4.7.3.b. This STP 1
also satisfies the inservice inspection (ISI) requirements of
TS 4.0.5. The RI observed the performance of this test from both the ,
RCIC pump room and the control room. Although the RCIC turbine was i
able to attain the required RPMs, the pump flow controller was not
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able to obtain a 600 gpm output through the test return line. The
RCIC system was subsequently declared inoperable and Maintenance Work
Order (MWO) 55284 initiated. On June 7, 1987, the RCIC STP wcs again
performed and the acceptance criteria met..
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. REP-0014 - This REP, performed on May 8, 1987, utilized the
transverse incore probe (TIP) machine to provide the process
computer's 0D-1 Program with the reactor core's neutron flux profile.
The 0D-1 Program then calculated the local power range monitor (LPRM)
calibration constants and determined the BASE and BASELP
distributions used in the axial power distribution calculations. The
core thermal limits calculated from the 00-1 input and displayed on
the P-1 printout were then reviewed and the therr..al limits verified .
to be within the TS allowable limits. l
7. ,0perational Safety Verification
l The resident inspectors observed operational activities throughout the
inspection period and closely monitored operational events. Control room
activities and conduct were generally observed to be well controlled and
efficient. However, some periods of congestion were noted by the SRI.
This was attributed in part to operator license candidates on shift for
reqstred training. This was discussed with licensee management and a memo
was issued by the licensee which reiterated the limitations on trainees
and stressed the shift supervisor responsibility for assuring proper
control room conduct. The SRI noted immediate improvement in control room
congestion after this memo was issued. Proper control room staffing was
maintained, and access to the control room operational areas was
controlled. Operators were questioned regarding lit annunciators, and
they understood why the annunciators were lit in all cases. Selected
shift turnover meetings were observed, and information concerning plant
status was being covered in these meetings. A walkdown of the "B" low
pressure coolant injection (LPCI) system was conducted and the results are
documented in paragraph 5 of this report. !
Plant tours were conducted, and while general plant cleanliness was good,
certain specific plant areas were identified as needing a general cleanup.
These areas included the fuel building and the service building machine I
shop. Plant cleanliness was discussed with licensee management and action
was taken to clean specific areas. The resident inspectors will continue I
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to monitor licensee actions to maintain a high level of plant cleanliness.
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General radiation protection practices were observed and no problems were ,
noted. The SRI verified that the minimum number of armed security ;
officers were on shift during one 24-hour period. Security access control '
was observed to be efficient and orderly.
The resident inspectors also reviewed licensee actions on operational I
events and potential problems. The results of reviews of selected items i
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are described belou:
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a. Corbicula Control Program Problems: The licensee independent safety
engineering group (ISEG) conducted an evaluation of the corbicula
(asiatic clam) control program implementation and several problem i
areas were identified. These problem areas included failure to I
maintain required chlorine levels in service water systems, failure )'
l to conduct visual inspections of certain coolers on required l intervals, failure to perform visual inspection of certain coolers
because of an inability to disassemble, etc. The plant staff has
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addressed all of the ISEG concerns and actions are being implemented
to upgrade the corbicula control program. No corbicula problems have i
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been identified during any inspections performed of the nonnal
I service water system to date. The SRI will monitor licensee actions
in response to ISEG Special Analysis 87-005 as an open item ,
(458/8714-02), to assure that an effective corbicula control program i
is implemented. j
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b. Emergency Siren Problems: The licensee experienced several problems
with emergency sirens during this inspection period. The problems
included a loss of activation capability because of a loose
electrical lead, inadvertent activation of sirens in East Baton Rouge
Parish from an unknown cause and three cases of individual siren ,
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activations from lightning strikes. The licensee plans to implement
a design modification for the emergency radio system power supplies
so that a single loose lead will not cause a total loss of activation
capability. The licensee has fully investigated the false siren
activations and they implemented the new procedure for public
notification on the emergency broadcast network each time. It i
appears that there is an unavoidable time delay of approximately 45 I
minutes before notification is made. This time dela/ is caused by
the time required to verify that the activation is inadvertent and
not an actual civil defense siren activation for reasons unknown to
the staff at River Bend. The resident inspectors will continue to
monitor licensee actions on identified emergency f.iren problems.
c. Leak Detection Resistance Temperature Detector (RTD) Response Times:
The licensee was notified by Stone and Webster that incorrect RTD
response times were used in high energy line break (HELB) and
moderate energy line break (MELB) analysis for the reactor core i
isolation cooling (RCIC) and the residual heat removal (RHR) systems.
The RCIC system HELB analysis was reevaluated with credit taken for
the high steam flow switch isolation capability for large line breaks
and the RTD response times were determined to be adequate for smaller
line breaks or cracks. The RHR system MELB analysis was reperformed
and the existing equipment qualification envelope was not exceeded
using the correct RTD response times. Licensee actions in response
to this identified problem were considered prompt and effective. No
additional NRC followup of this issue is planned.
d. Reactor Pressure Transmitter Calibration Deficiencies: The licensee
discovered that the calibration of four reactor pressure transmitters
and trip units had not been factored in a required water column head
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correction. An approximate 18 pound correction should have been
applied. The identified transmitters affect control logic for the
automatic depressurization system (ADS) and the low pressure core
i spray (LPCS) and low pressure coolant injection (LPCI) systems !
injection logic. The 18 pound calibration error was more
conservative for the ADS system, but slightly less conservative for
the LPCS and LPCI injection logic. The licensee took immediate
action to recalibrates the affected electronic trip units being
supplied by the affected transmitters with a temporary shifting of
the milliamp setpoint. The licensee commited to recalibrates the
affected transmitters during the first planned or unplanned plant
outage. This commitment has been added to the forced outage schedule
for tracking purposes. The licensee also conducted a review of other
instruments for head correction errors and no other problems were
identified. The resident inspectors will verify that the affected
pressure transmitters are properly calibrated during the first plant
outage,
e. Condition Report Backlog: The licensee's condition report (CR)
program has been effective for identifying and resolving problems at
River Bend. The plant manager assigns responsibility for responses
to CRs and a choice of three time frames is identified for response.
These time frames are 24 hours, 10 days, or 30 days. A large
percentage of CRs have been assigned to engineering, and engineering
has experienced difficulty with providing responses within these time
frames. Overdue CRs within the engineering department peaked at
approximately 135 in January 1987. Since that time additional
resources have been applied to reduce that excessive backlog. The
present backlog stands at 37 CRs which exceed the 30-day time frame.
The licensee expects to have this backlog reduced to at or near zero
by the end of June 1987. The resident inspectors will monitor the
licensee's progress toward reducing and controlling overdue CR !
responses.
f. Staff Turnover: The SRI continues to monitor personnel turnover at
River Bend. A review of turnover for the period of May 1986 through
May 1987 revealed the following:
. Engineering - 20 staff members lost from an authorized staff of
92. There ar.16 open positions presently.
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. Quality Assurance - 11 staff members lost from an authorized
staff of 60. There are 4 open positions presently.
. Training - 6 Staff members lost from an authorized staff of 36.
There are 7 open positions presently. j
. Licensed Operators - 2 operators lost fram an authorized staff
of 29. There is 1 open position presently.
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The greatest area of concern continues to be the engineering
turnover. The licensee is presently supplementing their permanent
staff with contract personnel as needed and they are actively
recruiting to fill the open positions. The use of contract personnel
appears to be an effective short term method of addressing the high j
turnover in the engineering department. The SRI will
continue to monitor staff turnover during future inspections.
g. Maintenance Backlog: The SRI noted a sharp increase in overall
maintenance backlog during this inspection period. This was
discussed with licensee management and they stated that the rate at
which maintenance work orders were being issued had increased. The
senior vice president stated that he intended to bring in additional
craft personnel to reduce the backlog and to make preparations for
i the planned refueling outage in September 1987. The resident '
inspectors will monitor licensee progress in reducing and controlling
the maintenance backlog.
No violations or deviations were identified in this area of inspection.
8. Inspections of Boiling Water Reactor Stainless Steel Piping
Temporary Instruction (TI) 2515/89 was issued to verify that the boiling
water reactor (BWR) licensees have performed inspections of stainless
steel piping welds susceptible to intergranular stress corrosion cracking
(IGSCC) in accordance with Generic Letter (GL) 84-11. The requirements of
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this TI do not apply to licensees who did not install IGSCC susceptible
material . IGSCC resistant material was installed at River Bend. NRC
Inspection Reports 50-458/85-66 and 50-458/85-30 closed IE Bulletins 82-03
and 83-02, respectively, which relate to IGSCC problems and these reports
also document that IGSCC resistant material is installed at River Bend.
Although IGSCC resistant material is installed at River Bend, the licensee
will perform inservice inspections of the recirculation piping in
accordance with the ASME code. The licensee plans to use a contractor
for these inspections and they will require the contractor to use Level I
certified UT examiners with demonstrated field performance capability. If
IGSCC-type flaws are detected during these inspections, appropriate action
will be implemented.
The resident inspectors will monitor licensee actions in this area during
routine inspection activities. TI 2515/89 is closed at River Bend.
9. In-Office Review of Written Reports of Non-Routine Events at Power
gactorFacilities
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The purpose of this portion of the inspection was to oscertain whether
corrective actions discussed in the licensee event reports appeared to be
appropriate, and whether information reported satisfied reporting
requirements.
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The NRC inspector reviewed LERs 87-04 through 87-07. All reporting
requirements were found to have been met, and all reports were adequate to
assess the events reported. Corrective actions specified in the reports
appeared to be adequate to identify the root causes and to correct these
Causes. l
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No violations or deviations were identified in this area of inspection.
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10. Exit and Inspection Interview
An exit interview was conducted on June 16, 1987, with licensee 1
representatives (identified in paragraph 1). During this interview, the l
SRI and RI reviewed the scope and findings of the inspection. l
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