ML20133B863
| ML20133B863 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 08/01/1985 |
| From: | Jaudon J, Tapia J, Whittlesey K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20133B828 | List: |
| References | |
| 50-458-85-31, NUDOCS 8508060277 | |
| Download: ML20133B863 (6) | |
See also: IR 05000458/1985031
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APPENDIX B
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-458/85-31
License /CP:
CPPR-145
' Dock'et:
50-458
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Licensee: Gulf States Utilities
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P. O. Box 2951
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Beaumont, Texas 77704
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Facility Name:
River Bend Station
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Inspection At:
River Bend Station, St Francisv111e, Louisiana
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/ Inspection Conducted: April 3-8, 1985
Inspebtors:
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- K. A. W itt16sef,'Re ' tor Inspdgtor, Project
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- Sect on A, Reactor P o ect Branch 2
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apid, Rdact@ Inspector, Project Branch B
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- J./P. paudg6, Chief, Project Section A, Reactor
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Prohet Branch 1
Inspection Summary
Inspection Conducted April 3-8, 1985 (Report 50-458/85-31)
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8508060277 350002
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Areas Inspected:
Routine, announced inspection of containment structural
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integrity test and integrated leak rate test.
The inspectiontinvolved 51
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inspector-hours ~onsite by two NRC inspectors.
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Results: Within the two areas inspected, one violation was identified (failure.
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to provide adequate procedure, paragraph 3).
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DETAILS
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Persons Contacted
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Gulf States Utilities (GSU)
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~*C.-M. Coones, Civil Engineer
- T. L.L Crouse, Manager Quality Assurance
- P.~J.
Dautel, Licensing Staff Assistant
- J. C. Deddens, Vice President
- P. E. Freehill, Superintendent Startup and Test
- E. R. Grant, Supervisor Licensing
J. E. Lo'zes, Senior' QC Inspector
- G. R. Kimmell, Supervisor Operations Quality Assurance
- G.
V.' King, Plant Services Supervisor
- E. R. Oswood, Quality Assurance Engineer
C. D. Payton, Field Quality Control Level II
- T.
F. Plunkett, Plant Manager
J. E. Redmond, Senior QC Inspector
- T. E. Suhrke, Manager Project Planning and Coordination
Stone and Webster
- R. H. Bernier, Senior Advisory Engineer
- J. L. Busa, Assistant to the Chief Engineer
- F. W. Finger III,' Project Manager, Preliminary Test Organization
- R. I. Parry, Supervisor, Mechanical Test Engineering
- Indicates presence at exit interview conducted April 8, 1985.
2.
Structural Integrity Test (SIT)
The purpose of the SIT is to demonstrate the ability of the containment
vessel to withstand internal loads imposed by pressurizing to 1.15 times
the design pressure of 15 psig.
Preoperational Test Procedure No. 1
PASIT.001, Revision 1, " Pressure Test of the Steel Containment," was
reviewed and determined to meet NRC requirements and licensee connitments
listed in the' Final Safety Analysis Report.
The test, already in progress
at the time of the NRC inspector's arrival on site, was being conducted in
accordance with the reviewed and approved test procedure.
The inspectors reviewed the procedure for ultrasonic monitoring of
electrical penetrations (attachment 10.2 to 1 PASIT.001) to be performed
under step 7.19.
The ultrasonic leak detection method was implemented in
lieu of local pneumatic tests of the circumferential butt welds in'
electrical penetration nozzles.
Calibration of acoustic monitoring
equipment on a mockup in the annulus area was observed prior to
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, commencement o'f the: ultrasonic inspection by field quality control
personnel. The NRC inspectors observed and independently monitored
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inspection of several penetrations and noted that no leaks were detected.
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The. method for discerning the equipment response to leakage from response
to background noise was described to the inspectors as well asr
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demonstrated.
At the comple' ion of the SIT, the containment was depressurized.
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-Integrated Leak Rcte Test
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The preoperational containment integrated leak' rate test conducted in
accordance with Preoperational/ Acceptance Test Procedure 1-PT-57-1
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~" Integrated Leak Rate Test," was addressed during this portion of the.
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. inspection. The inspection included procedure and records review,=' test
witnessing, and independent calculations by the NRC inspectors. .The
inspection was performed in order to ascertain whether testing was
fconducted in accordance with approved procedures and, satisfied the
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specified acceptance criteria of 10 CFR 50, Appendix J and,the Final
Safety Analysis Report.
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After-a period of time at atmospheric pressure to allow for degassing of
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-structures and components inside containment subsequent to the SIT,
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pressurization of the containment vessel for the ILRT commenced.
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Stabilization commenced.after internal pressure reached 8.6 psig (23.3 -
psia), the compressors were shut down and isolated. . The atmosphere is
considered stabilized when the rate'of change of containment temperature-
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averaged over the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> minus the rate of. change in containment-
temperature averaged over the last hour is less than 0.5*F/
About
11:00a.m.onApril5,1985,theNRCinspectorswereadviseb#.that
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containment atmosphere stabilization had been achieved and the official
24-hour test had begun. . Iaitial calculated leakage was excessive and
attempts to identify the leakage source resulted in the following sequence
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ILRT configuration includes a pneumatic block of main steam isolation
valves-(MSIVs).
An increasing pressure trend was noticed on the pressure
gauge on the outboard MSIVs indicating leakage across the inboard MSIVs.
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After the: piping between inboard and outboard MSIVs equalized with
containment pressure, the downstream piping was sealed at a pressure of
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8.5 psig to minimize the effect of this leakage.
It should be noted that
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the MSIVs are supplied with-a positive leakage control system, which would
be pressurized above peak postulated accident pressure in the event of'a
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' design basis accident. Additionally, Valve 1 RHS-V15, an instrument root
. valve on Residual Heat Removal'B pump discharge piping was found out of
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position and leaking a steady stream of water.
Although the valve lineup
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sheet and control room tagout log both indicated the valve to be in the
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closed position, it was found open by the licensee. The importance of
compliance with tagging procedures was discussed at the exit meeting,
although this was considered an isolated case.
The valve was closed, but
calculated leakage remained excessive, indicating a remaining unidentified
leakage source.
Further investigation by the licensee identified a direct
leakage path from the containment vessel to the annulus via three
instrument lines associated with the containment to annulus differential
pressure monitoring portion of the annulus pressure control system.
Instrument root, isolation, and equalization valves were open for 1
HVR*PDT G0A,1 HVR*PDT 60C, and 1 HVR*PDT 60E. This allowed a direct
leakage path from containment to the annulus because six differential
pressure instruments were omitted from the ILRT valve lineup. The subject
instrument and associated root valves were added to the ILRT lineup via
Minor Change Request (MCR) Number 4; a correct lineup was achieved, and
the ILRT was restarted.
After the restart of the test, motor operated valve 1 DFR*MOV 146, a
designated containment isolation valve, was determined to be in the closed
condition.
10 CFR 50, Appendix J requires that closure of containment
isolation valves for the ILRT be accomplished by normal operation.
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this case, normal operation would indicate response of the valve to a
containment isolation signal. However, the valve, having been recently
installed under authorization of Engineering and Design Coordination
Report Number P13043B, was not yet connected to electrical supply so it
could not be closed by normal means. The valve had been hand closed prior
to initiation of the test, and it was not included in the ILRT valve
lineup, although the expressed intent had been to call 1 DFR*MOV 146 open
in the lineup and rely on check valves 1 DFR*V131 and 1 DFR*V132. The
omission of differential pressure instruments 1 HVR*P0T 60A, 1HVR*PDT 60B,
1 HVR*PDT 60C, 1 HVR*PDT 60D, 1 HVR*PDT 60E, and 1 HVR*PDT 60F and the
omission of 1 DFR*MOV 146 from the valve lineup for 1-PT-57-1 constitute a
violation for failure to have adequate procedures (458/8531-01).
1 DFR*MOV 146 was opened, and the test continued. There was no noticeable
perturbation in the test data which could be attributed to the valve
manipulation, and open drain valves outboard showed no sign of water
leakage.
It should be emphasized that as a portion of the reactor
building floor drain system for pump back suppression pool water
inventory, the referenced valve would be exposed to a water seal rather
than directly exposed to containment atmosphere.
Pending final acceptance
and test demonstrating a fluid leakage rate within technical specification
limits for 1 DFR*M0V 146, this item remains open.
(485/8531-02).
Continuation of the test indicated convergence of the calculated leak rate
and the upper confidence limit below the allowable leakage. At completion
of the 24-hour test, the superimposed leak verification portion of the
test was performed with results between the calculated and imposed
leakages within the 25% La limit.
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Subsequent to the performance of the test, the NRC inspectors obtained the
. raw data and computed the leakage rate in accordance with the Mass Point
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Data Analysis' technique. The computations performed by the NRC inspector
were, compared with the licensee's results for!the purpose of. verifying-the
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calculational procedure and confirming the results.
This' analytical
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technique confirmed _the acceptability of thefresults obtained by the-
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Exit Interview
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The NRC-inspector met with the licensee representatives ~ denoted in
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paragraph 1 at the conclusion of the inspection.- The NRC inspector
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summarized the scope and findings of the inspection.
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