ML20133B909
| ML20133B909 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 08/01/1985 |
| From: | Boardman J, Chamberlain D, Jaudon J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20133B883 | List: |
| References | |
| 50-458-85-43, IEB-77-06, IEB-77-6, IEB-79-15, IEB-79-23, IEB-79-27, IEB-80-08, IEB-80-16, IEB-80-8, IEB-84-03, IEB-84-3, NUDOCS 8508060287 | |
| Download: ML20133B909 (13) | |
See also: IR 05000458/1985043
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APPENDIX B
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U. -S.
NUCLEAR REGULATORY COMISSION
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REGION IV
NRC Inspection.Reporti 50-458/85-43
CP:
CPPR-145
Docket:
50-458
Licensee:
Gulf States Utilities Company-(GSU)
P. O. Box 2951
' Beaumont, Texas
77704
Facility Name:
River Bend Station (RBS)
Inspection At:
River Bend Station, St. Francisville, Louisiana
Inspection Conducted:
May 1 through June 15, 1985
Inspect :
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D. 7. C6am@ rTain, Senior Resident Inspector
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(pars. 1, 2, 3, 4, 5, 6, 7, 11)
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hJ yB rt9 nan, Reactor Inspector, Operations
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Secti n, Reactor Project Branch (par. 8)
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J./P. yaudo , Chief, Project Section A, Reactor
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Approved:
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J./ P. paup6n, Chief, ProIect' Section A,
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(Reac'to(Projects Branch
Inspection Summary
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Inspection Conducted May 1 through June 15, 1985 (Report 50-458/85-43)
Areas Inspected:
Routine, unannounced inspection of licensee action on
previous inspection findings, site tours, reactor protection system
preoperational test witness, reactor pressure vessel leakage test witness,
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control rod drive system full core scram test witness, reactor coolant system
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8508060287 850002
ADOCK 05000458
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hydrostatic ~ test results evaluation, construction quality assurance (QA)
~ program review,-Inspection and Enforcement (IE) Bulletin followup, and
allegation followup.' The inspection involved 147 inspector-hours onsite by
three NRC' inspectors.
Results: .Within the areas' inspected, onb violation was issued in the area of
construction QA program review (failure to obtain.NRC approval of a QA
program' change, paragraph 8).
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DETAILS
1.
Persons Contacted
Principal Licensee Employees
K. Arnstedt, Quality Assurance (QA) Engineer
- W. H. Cahill, Jr. , Senior Vice President, River Bend Nuclear Group
- T. F. Crouse, QA Manager
J. Davis, QA Engineer
- P. J. Dautel, Licensing Staff Assistant
- J. C. Deddens, Vice President, River Bend Nuclear Group
D. R. Derbonne, Supervisor, Startup and Test
S. Finnegan, Control Operating Foreman
- P. E. Freehill, Superintendent, Startup and Test
- D. R. Gipson, Assistant Plant Manager, Operations
P. D. Graham, Assistant Plant Manager, Services
R. W. Helmick, Director, Projects
K. C. Hodges, Supervisor, Quality Systems
D. Jernigan, Engineer, Startup and Test
- G. R. Kimmell, Supervisor, Operations QA
- G. V. King, Supervisor, Plant Services
J. L. Pawlik, Engineer, Startup and Test
- T. L. Plunkett, Plant Manager
- S. R. Radebaugh, Assistant Superintendent, Startup & Test
- S. F. Sawa, Control Superintendent, Startup & Test
- J. E. Spivey, QA Engineer
R. B. Stafford, Director, Quality Services
K. E. Suhrke, Manager Project Planning & Coordination
L. Sutton, QA Engineer
- P. F. Tomlinson, Director, Operations QA
- A. Valenzuela, Startup and Test
- J. Venable, Mechanical Maintenance Supervisor
D. White, Engineer, Startup and Test
Stone and Webster
D. P. Barry, Superintendent of Engineering
W. I. Clifford, Senior Construction Manager
F. W. Finger, III, Project Manager, Preliminary Test Organization (PTO)
M. Fischete, Engineer, Startup and Test
- P. H. Griffin, Site Advisory Manager
B. R. Hall, Assistant Superintendent, Field Quality Control (FQC)
Q. E. Harper, Hydro Test Engineer
D. Hill, Maintenance Engineer
R. L. Spence, Superintendent, FQC
The NRC senior resident inspector (SRI) also interviewed additional
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licensee, Stone and Webster (S&W), and other contractor personnel ~during
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lthe inspection period.
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- Denotes .those persons that attended the exit' interview conducted on
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jJune21,1985.
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[2.
Licensee Action on Previous Inspection Findings
0'en Item (458/8408-01): ' Review to determine if and how the
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- diesel; generator loading restrictions of calculation 12210-E-122 are
implemented in plant operating procedures.
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The. SRI obtained a copy.of calculation 12210-E-122, Revision 4,
" Standby Diesel generator Loading Calc. ," dated March 1,1984.
This
revision.of the calculation is based on a 3500-KW loading limit on
the diesel and does not' reflect the latest load restriction of 3130
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However, a GSU letter RBG - 20,086 dated February 6, 1985,
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.contains revisions to the FSAR "to establish a qualified load for
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each of the diesel generators." These revisions include, for diesel
1EGS*EG1A loading, a requirement.that "LPCS or RHR A pump shall be
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. manually tripped after 2.0 hr of LOCA, depending upon the available
' diesel generator sets" and for diesel 1EGS*EG1B loading, a requirement
that "RHR C is stripped manually.by the operator after 2.0 hr of
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operation after LOCA, depending upon the combination of the available
diesel generator. sets."
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Abnormal Operating Procedure A0P . .0004, Revision 1, " Loss of Offsite
Power," dated April 10, 1985, appears to address the latest loading
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1 restrictions (3130 KW) for the diesels,' but the stated action for
' manual tripping of an RHR or LPCS pump in Section 5.8 needs some
clarification.
For example, it is not clear under what conditions
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that RHR A pump is tripped,instead of the'LPCS pump.
This item will
remain open pending issue'of an approved calculation reflecting
. qualified diesel loading and pending the required clarifications in
procedure Section 5.8.
b.
(0 pen) Unresolved Item (458-8408-04):
Review of licensee program,'
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for tracking of commitments to the NRC.
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GSU has developed and implemented a commitment tracking program at
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- River Bend.
Project Procedure' No. 8.2,'" Identifying and Tracking
Project Commitments," was issued on Augu~st 9,.1984,Dto " provide
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guidelines for River Bend Nuclear Group <(R8NG) organizations for
identifying, documenting, tracking,,and closing commitments made .to
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regulatory agencies." Nuclear licensing is responsible for the
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' tracking system' with QA responsible for verification of completion .
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of commitments on a sampling basis.' ' GSU' uses the "19AC" computer
program and they have identified approximately ^2,152,t.cenitments to
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date.
The present status of the 1,191'open commitments is 717.high'
priority commitments required for fuel, load and 156 required after
fuel load, and there are 280 low priority commitments required for
fuel load and 38 required after fuel load.
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This item will remain open for the SRI to evaluate the GSU method _of
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identifying commitments that require closure prior to fuel load and
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for identifying those that can be completed after fuel' load.
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c.
(Closed) Open Item (458/8434-01):
Implementation of preoperational
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test commitments ~by the control rod drive'(CRD) hydraulic
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preoperational test procedure (1-PT-052).
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The' specific items of concern were impienentation of'the following
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Final safety Analysis Report (FSAR), Chapter 14, test commitments:
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To verify the failure mode of the CRD
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system on loss of power."
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The CRD pumps are tripped and the time for
accumulator' inoperable alarms to occur is
,. recorded as baseline data."
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"4.f. . All scram valves open on a loss of
instrument air to the CRD, system."
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- The above items were addressed in the. following manner:
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Th'e failure mode of the CRD system was
tested by verifying the scram function on a
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> loss of power to the scram' pilot solenoids.
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3.j. ~A minor change request (MCR 09) was issued
to require recording of the times for all
hydraulic control units which exhibited low
accumulator pressure alarms within 10
- minutes after tripping oflthe CRD pump.
4.f.
An acceptance criteria step 10.9 was added
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by a major change request (MRC 06) to
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reference the backup scram valve test which
demonstrates that the scram valves open on a
loss of instrument air.
This item is closed.
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d.
.(Closed) Violation (458/8507-01):- Procedures were not implemented-
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to maintain Class 8 cleanliness requirements in the spent fuel storage.
area'where the new fuel was to be stored'in accordance with the
Special Nuclear Material License issued on January 15, 1985.
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GSU took immediate action to issue an unsatisfactory inspection
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report and fuel' receipt was delayed 1 day to allow removal and
inspection of the spent fuel racks and clean up of the spent fuel
pool floor.
Following'the cleaning, the spent fuel racks were
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reassembled in the pool and fuel receipt progressed as scheduled.
Also, to prevent recurrence of this type of item, housekeeping and
cleanliness procedures shall be implemented as a prerequisite to the
governing procedures.
This item is closed.
e.
(Closed) Deviation (458/8507-02):
Preoperational test procedures
are not being provided for NRC review 60 days prior to the scheduled
test performance in all cases.
Only four preoperational test procedures remained to be submitted for
NRC review at the time of this deviation.
The.four remaining
procedures were expedited and all have now been submitted.
This item is closed.
f.
(Closed) Open Item (458/8522-07):
GSU has installed motor operated
valve'(M0V) circuit breaker trips that can be reset either manually
or automatically.
GSU has not established control to verify that all
such resets are in the manual mode.
Motor control center starters for MOVs at River Bend have both
thermal overload trip and magnetic trip devices. The magnetic trip
devices trip the manual circuit breaker to remove the overload
condition.' The magnetic, trip device then resets automatically, but
the~ circuit breaker must be manually reclosed to provide power to the
' starter.
The thermal overload trip device opens the circuit to the
motor starter to: remove the overload condition.
The thermal overload
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device'has a hand / automatic option on reset.
GSU has chosen to place
all of the thermal overload ' devices in the hand reset position.
Temporary Change Notice.No.85-131 has been issued to revise Procedure
No. CMP-1026, " Corrective Maintenance of MCC Starters," to include a
step for verifying that hand automatic reset selectors are in the
- hand position for thermal overload trip devices. GSU operations was
notified of this' condition per memorandum APM-M-85-94 dated June 13,
1985.
It was also noted during the review of this item that certain
loss of coolant accident initiated MOVs would have not thermal
overload trip devices installed.
This item is closed.
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(Closed)) Open Item (458/8522-12): A system had not been provided
for assuring that each piece of measuring and test equipment (M&TE)
is calibrated and adjusted on or before the date required.
Procedure ADM-0029, Revision 4, " Control of Measuring and Test
Equipment (M&TE)," has been revised via temporary Change Notice
85-733 to clarify the system used and the responsibilities for the
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recall of M&TE for calibration.
Also, in addition to the recall
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. requirements, the M&TE issue facility must verify that the M&TE
calibration due date is current prior to issue of the M&TE and users
of M&TE are required to verify that the calibration of M&TE is
current prior to use.
All of these requirements are intended to
preclude the use of any M&TE for which the calibration has expired.
This item is closed.
3.
Site Tours
The SRI toured areas of the site during the inspection period to gain
knowledge of the plant and to observe general job practices.
The site
tours conducted included special tours on separate occasions with
Commissioner Bernthal and with a group fron NRC Nuclear Reactor Regulation
headed by Harold Denton.
Both of these tours included the conduct of mock
scenarios on the River Bend plant simulator.
No violations or deviations were identified in this area of inspection.
' 4.
Reactor Protection System Preoperational Test Witness
The SRI witnessed portions of the reactor protection system response time
measurements testing conducted during this inspection period.
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specific testing witnessed included reactor pressure sensor response
timing, reactor vessel level sensor response timing and drywell pressure
sensor response timing.
Testing personnel experienced several problems
with the set up of the response time test equipment which caused testing
delays.
They also experienced problems with interpreting the response
time curves such that the proper ramp was generated for acceptance.
criteria purposes.
The vendor for the test equipment was brought to the
site and the test equipment problems were corrected.
Also','a uniform
method for interpreting the response time curves was formulated.
The SRI
conducted a preliminary review of several response time curves and it
appeared that the response times were within acceptance criteria limits.
The major testing remaining for the reactor protection system
preoperational test at the end of this inspection period was the
intermediate range monitor (IRM) and average power range monitor (APRM)
response time measurement testing.
No violations or deviations were identified in this area of inspection.
5.
Reactor Pressure Vessel Leakage Test Witness
This special reactor pressure vessel (RPV) leakage test (1.MPRV.002) was
performed in order to disposition a Nonconformance and Disposition (N&D)
Report No. 11275.
This N&D resulted when the review of the N-5 data
reports on the reactor pressure vessel indicated that no hydrostatic test
was performed subsequent to the installation of reactor-internals or
rework on nozzle safe ends.
This included installation of items such as
control rod drive (CRD) housings, incore housings, recirculation and
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feedwater safe end rework, jet pump.pene'tration seals, etc.
The reactor ,
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pressure-: vessel system hydro procedure (1-G-ME-15) and associated
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- documentation did not identify the welds for these items,as being'withinJ
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.the scope of the RPV hydro-test inspections.' _Therefore, during the RPV
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hydro' conducted in May 1984, documentary evidence of the inspection off
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these welds was not obtained.
'TheSRIwitnessedthe-RPVleakagetestingandweldinspectionspe[ form
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Jon May 16, 1985.
The leakage testing was performed at a design pressure
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of 1250 psig.
Initially, trouble was. experienced with obtaining the test-
_ pressure _due to excessive leakage through the gaged safety relief valves.
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Test personnel obtained apnroval from the relief valve vendor which
allowed the relief valve gags to be torqued to 30 foot pounds psig at a
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.RPV pressure of 1000 psig.
This was accomplished and they were then able
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The test pressure was held for a
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minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the performance'of the official inspections.
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The inspections were' performed and no problems were identified.
'No violations or deviations were identified in this area of inspection.
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6.
-Control Rod Drive System Full Core Scram Test Witness
During this: inspection period, two back up scram valve full core reactor
scram tests were performed to complete the control rod drive system
preoperational testing.
Also, a special full core reactor scram test was
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performed to evaluate the scram discharge volume level instrument response.
The. SRI witnessed the perfo'rmance of the special scram test on May 29, 1985.
'This special scram test was performed ~in conjunction with reactor
' protection system response time testing and the scram was initiated by a
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reactor-vessel low water signal.
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No violations or deviations were identified in this area of inspection.
7.
Reactor Coolant System Hydrostatic Test'Results Evaluation
The SRI conducted a review of the completed test results_for the RPV
system hydrostatic test (Procedure 1-G-ME-15) and for the subsequent
reactor pressure vessel' leakage test.(Procedure 1.MPRPV.002).
The
specific areas reviewed and findings noted included the following:
a.
Changes to the test procedures were documented and implemented in
'accordance with the licensee's administrative controls.
.b.
The system boundary either included all piping and equipment
protected by the safety' relief valves or documentation was provided
to show that; separate hydros were performed on equipment.or piping
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that could be isolated from the RPV.
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The water quality met all requirements.-
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The licensee held the maximum test pressure (1.25 times the design
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pressure) for at least 10 minutes during the RPV system hydro test.
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The hydrostatic test pressure did not exceed the maximum pressure
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-.The reactor coolant temperature ~was maintained above the~ nil
ductility transition temperature throughout the hydro and leak
. testing.
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All' identified test exceptions 'have been resolved, but a concern was
identified with Test Exception TE-13 for Procedure 1-G-ME-15.
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test exception addressed certain flexible hoses that were not
installed'at the time of the original hydro.
There were 40 hoses
identified on the test exception and they were apparently identified
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on system punch lists for~ installation and hydro at a later date.
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-The SRI selected 4 flexible hoses (Nos. 114, 121, 122, and 140) for a
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review of the rework documentation to determine if the required
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hydros were performed.
Of the four selected, documentation was
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test.
However, it was noted that the flange connections on these
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flex hoses had been blanked during the hydro'and only two of the.four
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- rework control forms required a subsequent operational . leak test
(OLT). This was discussed with test personnel and it was determined
that the performance of OLTs, on a flange connection that.is completed
after a hydro, has been normal practice at River Bend.
Further
review revealed that three of the four hoses received an OLT during
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the RPV leakage test per a startup trouble ticket (STT).
The fourth
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hose was not specifically. mentioned on the STT, but it was identified
on. working drawings as being inspected.
The SRI believes that no
problem was created by failure to note an OLT requirement on the
rework document and test personnel stated that the normal practice
will continue for performance of OLTs.
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The test results have been reviewed and approved by those personnel
charged with the responsibility.
The SRI also. reviewed selected vendor supplied pump and valve hydro
records and no problems were noted.
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No violations or deviations were identified in this area of inspection.
8.
Construction QA Program Review
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'A review of the licensee construction QA program revealed that on June 10,
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1983, GSU forwarded fnr approval a revision of their construction QA
program as required by 10 CFR 50.55(f)(2).
Discussions with licensee
personnel revealed that revised Section 17.1.2.4.A reducing the periodicity
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of GSU . review of all controlled documents from 1 year to 2 years was made
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after March 11 1983.
Licensee personnel further stated that this change
.had been implemented, although it had never been approved by NRC.
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Implementation of this change is in violation of 10 CFR Part 50.55(f)(3),
which requires prior NRC approval of licensee QA program changes made
after March 11, 1983, when such changes reduce commitments.
(8543-01)
9.
Inspection and Enforcement (IE) Bulletin Follow Up
The purpose of this inspection was to followup on licensee action taken in
response to Inspection and Enforcement Bulletins (IEBs),
a.
This bulletin concerned the consequences of a failure of the
refueling cavity seal. The NRC inspector reviewed the following
licensee correspondence to the NRC (Region IV).
Letter Serial
Date
Items Addressed
RBG-19487
11-29-83
. Cross Seal Failure
. Maximum Leak Rate
Because of Seal
Failure
.Make Up Water
Capacity
. Potential Effect on
Stored Fuel and
Fuel in Transfer
.0ther Consequences
RBG- 20042
02-01-84
. Emergency Operating
Procedures
RBG- 20635
04-05-85
. Time to Cladding
Damage Without
Operator Action
RBG-21023
05-15-85
. Time to Cladding
Damage Without
Operator Action
These four letters address all of the points required by IEB 84-03.
The design of the seal used at River Bend is a stainless steel
bellows assembly welded to its support structure. The maximum
credible leakage rate is within make up capacity. Total failure of
the seal without operator action could result in a problem for fuel
in transit between the reactor vessel and the containment fuel
storage pool. This is addressed in licensee Procedure A0P-0032,
which requires the fuel to be placed in either the vessel or storage
racks. All other fuel would remain covered with water, and no vital
equipment would be flooded by a complete draining caused by a bellows
failure. The bellows is also protected from direct impact by a
radiation shield and a guard ring.
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IEB 84-03 is considered closed.
b.
This bulletin concerned General Electric Series 100 containment
electrical penetrations.
River Bend does not use this type of
electrical penetration.
Therefore, there is no action required at
River Bend for IEB 77-06.
IEB 77-06 is considered closed.
c.
This IEB addressed deep draft pump deficiencies and the long term
operability of these pumps.
The NRC inspector found that the licensee had addressed operability
of pumps in the FSAR, and this was recognized in NUREG-0989, the
safety evaluation report for River Bend.
Additionally, the final
draft technical specifications contained surveillance requirements
for monthly demonstration of pump operability in accordance with the
ASME code,Section XI and an 18-month system operability test.
Since
the question of deep draft pump operability is being addressed by the
normal review process and implementation of the requirements to test
is under the routine inspection program, no additional tracking of
this IEB is warranted.
IEB 79-15 is closed for record purposes.
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d.
This IEB concerned the potential failure of emergency diesel
generators.
The failure could result if there was a large
circulating current between the exciter transformer and the
generator.
Such a circuit could be set up by connection through a
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common ground.
It was found that the design of all three emergency
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diesel generators at RBS was such that exciter transformers had a
floating primary neutral.
Subsequent testing of the emergency diesel
generators did not disclose any problems of the type discussed in
this IEB.
IEB 79-23 is considered closed.
e.
This IEB concerned the loss of non-Class 1-E instrumentation and
control power.
This IEB was not specifically directed to River Bend,
but it was included in the FSAR review, becoming Question 421.003 and
as Confirmatory Item 31 of NUREG-0989, the Safety Evaluation Report.
Since the action required by this IEB is being tracked as a confirma-
tory item, IEB 79-27 is closed for record purposes.
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f.
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This IEB addressed radiography of flued head design penetrations of
the containment.
The licensee was found to use flued head design in
both the containment and the drywell.
The licensee committed to use
radiography on all flued head design penetrations of the containment
and other nondestructive tests on other penetrations.
This IEB is closed.
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This IEB concerned Rosemont pressure transmitters, Models 1151 and
1152. When these transmitters were fitted with either "A" or "0"
output code cards, it was possible for the transmitters to have an
ambiguous output and the input signal was either an over pressure or
a reverse pressure signal.
The licensee found that there were four
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Rosemont 1152 transmitters with an "A" output code.
These four
transmitters were modified to have an "N" output board.
IEB 80-16 is considered closed.
10.
Allegation Follow Up
The NRC inspecto.a did a followup inspection of an allegation.
Backaround.
An anonymous letter was sent to both Gulf States
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Utilities (GSU) and the NRC.
This letter forwarded an internal piece
of Stone and Webster (S&W) correspondence.
This S&W correspondence
was a letter signed by engineers in the design group for small bore
pipe.
The letter complained that the group was being required to
account for on-the-job time in a log and alleged that such a time
accounting procedure was inimical to quality assurance.
Licensee Action.
GSU conducted a quality assurance review of the
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small bore piping group March 4-22, 1985.
NRC Review.
The NRC inspector reviewed the report of the licensees'
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review.
It was noted that the licensee had concluded that the
allegation was not substantiated in that the use of a time log to
account for time spent on various charge items does not have a direct
relationship to quality assurance.
It was further noted that there
were four additional concerns noted by the quality assurance review.
The NRC inspector noted that the licensee had followed up on these
four concerns and closed them.
Conclusion.
The NRC inspector concluded that the allegation was
.
substantiated in that the time log was kept but that it was invalid
as a safety concern.
This item is closed.
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11.
Exit-Interview
.
An exit interview was conducted on June 21, 1985, with licensee
representatives (identified in paragraph 1). .During this interview, the
SRI reviewed the scope and findings of the inspection.
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