ML19322A737
| ML19322A737 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/12/1972 |
| From: | Moseley N, Murphy C, Warnick R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19322A733 | List: |
| References | |
| 50-269-72-03, 50-269-72-3, 50-270-72-02, 50-270-72-2, NUDOCS 7911210769 | |
| Download: ML19322A737 (22) | |
See also: IR 05000269/1972003
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UNITED STATES
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ATOMIC ENERGY COMMISSION
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DIVISION OF COMPLIANCE
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REGloN ll - SulT E 818
230 P E Ac HT R E E ST R E ET, NoRT HWEST
T a La p=ce.a r (404)526 4s03
AT L ANT A. GEoRGt A 30303
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TESTING AND STARTUP INSPECTION REPORT
C0 Report Nos. 50-269/72-3 and 50-270/72-2
Duke Power Company
Oconee 1 and 2
Docket Nos. 50-269 and.50-270, License Nos.
CPPR-33 and 34
Category: A3/B1
Oconee County, South Carolina
Type of Licensee: PWR-2452 MW(t), B&W
Type of Inspection: Routine, Unannounced
Dates of Inspection: March 21-24, 1972
Dates of Previous Inspections: February 22-25, 1972
February 29, 1972
Principal Inspector-
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C.E.Mufty{ReftorInspector
Da'te'
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(Testing and Startup Unit)
Accompanying Inspectors:
[
7M
R. F. Nar61ck," Reactor Inspector
,,
Dite'
(Te ting and Star p Unit)
b
5 P-7A
F. Jape, Reactor faspector
Date
(Testing and Startup Unit)
Other Accompanying Personne : None
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Reviewed By:
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N. C. Mc5eley, Sfnior' Reactor
spector
D6tE/ ~
(Testing and Startup Unit)
Proprietary Information: None
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CO Report Nos. 50-269/72-3
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SECTION I
Enforcement Action
None
Licensee Action on Previously Identified Enforcement Matters.
1.
Action on the items of noncompliance relating to the low tensile
strength steel identified in CO Report No. 999-42/72-1 has been
completed.
2.
A satisfactory response was received on the-item of. noncompliance
relating to constructions audits identified.in inspection. reports
50-269/71-8 and 50-269/72-1. Action on the item of noncompliance
relating to test control identified in C0 Report No. 50-269/72-1
has not been completed as.yet.
3.
The deficiencies identfied in CO Report Nos. 50-269/71-9, 71-10
and 71-11, have not been corrected as yet.
Unresolved Items
1.
Determination of cause of failures to reactor coolant system components
and repairs to system.
2.
Revision of OP 1502/04, " Initial Fuel Loading Procedure," to
incorporate Compliance comments including the requirement for an
audible indication from in-core instrumentation in the reactor
building and/or control room, and establishment of criteria for
emergency boron injection.
(See Section IV, Paragraphs 1.g. and
1.s.)
Status of Previously Reported Unresolved Items
The inspector's questions and comments concerning hot functional
tests identified in CO Report No. 50-269/72-2 have been resolved.
(See Section III, Paragraph 3.b and 3.c)
Unusual Occurrences
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CO Report Nos. 50-269/72-3
50-270/72-2
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Persons Contacted
- D. L. Freeze - Principal Field Engineer
- J. E. Smith - Plant Superintendent
J. W. Hampton - Assistant Plant Superintendent
M. D. McIntosh - Operating Engineer
- R. M. Koehler - Technical Support Engineer
- L.
E. Summerlin - Staff Engineer
E. P. Stergakos - Engineer
- A. C. Thies - Senior Vice President, Production and Transmission
R. L. Wilson - Performance Engineer
- E. D. Brown - Maintenance Supervisor
O. S. Bradham - Instrument and Control Engineer
R. A. Morgan - Assistant Field Engineer
- By Telephone
- At Management Interview
Management Interview
The management interview was held on March 24, 1972, and the following
items were discussed.
1.
Murphy stated that Compliance would follow the repairs to the
reactor coolant system. He cautioned Smith that all repairs should
be done using approved procedures and qualified workers.
Details
of the work performed should be documented. Nondestructive testing
should also be performed by qualified inspectors and the results
documented.
Smith confirmed the inspector's understanding that
DPC would issue a formal report to DRL cn the incident and that
a preliminary report describing the incident would be issued
within ten days.
(See Section II, Paragraph 4)
2.
Murphy stated that Compliance would begin inspecting Unit 2 on a
more frequent basis and would shift the emphasis of the inspections
from construction to the testing aspects. He reviewed briefly
the discussions with the operating staff and particulary emphasised
that DPC would be expected to conform to the quality assurance
requirements of Appendix B to 10 CFR 50 during testing and sub-
sequent operation of all units.
(See Section II, Paragraph 3)
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3.
Smith confirmed the inspector's understanding that a review would
be made to determine if gate valves were being used in regulating
service and to take corrective action if any were found.
(See
Section II, Paragraph 2)
4.
Jape presented a summary of discussions held on March 21 and 22,
1972, with Hampton, McIntosh and Stargakas on the initial fuel
loading procedures.
Comments on the procedure were discussed and
received DPC concurrence.
It was stated that Compliance would
like to review the procedure again when the comments have been
incorporated by DPC.
(See Section IV, Paragraph 1)
5.
The inspector stated that he had reviewed the results of two
preoperational tests which had been performed and approved as
completed by Smith.
It was stated that these two tests appeared
to fully satisfy the acceptance criteria stated in the tests.
(See Section IV, Paragraph 2)
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CO Report Nos. 269/72-3
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SECTION II
Prepared By:
C. E. Murphy, Reactor
Inspector (Testing and
Startup)
ADDITIONAL SUBJECTS INSPECTED, NOT IDENTIFIED IN SECTION I, WHERE
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NO DEFICIENCIES OR UNRESOLVED ITEMS WERE FOUND
1.
Work Order System
The plant operating organization had previously instituted a
Work Order Systemb to assure that plant deficiencies would be
documented and corrected. The inspector reviewed the records
relating to the failure of the low pressure injection system
regulating valve and did not note any deficiencies in the records.
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(See Section II, Paragraph 2)
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DETAILS OF SUBJECTS DISCUSSED IN SECTION I
2.
Low Pressure Injection System Regulating Valve
DPC experienced a recurring failure of the ten-inch gate valve
used for flow regulation in the low pressure injection system.
Brown advised the inspector that after consultation with the valve
manufacturer, DPC has decided to either replace the valves with a
type more suitable for regulating service or to install a regulating
valve in parallel with the gate valve.2/
In response to the inspector's
questions, Brown stated that he did not know if gate valves had
been used elsewhere in the plant for regulating service.
Smith
agreed in the management interview to have a check made to determine
if there were other examples of gate valves used for regulation
and to tske corrective action if any were found.
3.
Change of Unit 2 Inspection Status
Murphy, in a meeting with the principal members of the operation staff
discussed the quality assurance requirements for plants in the testing
and operating status.
He advised them that the requirements of
Appendix B to 10 CFR 50 applied to a plant for its lifetime.
He
then discussed briefly the implications of each of the 18 criteria
1/ CO Report No. 50-269/72-1
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2/ Inquiry Report No. 50-269/72-5
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in Appendix B.
He also stated that since Unit 2 was approaching
the point that preoperational testing would be started,.the
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inspections would be performed on a more frequent basis and would
be oriented more to the testing and operating aspects.
4.
Damage to the Reactor Coolant System
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The inspector discussed the extent of the damage.to the reactor coolant
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syctem with Smith and questioned him about the events leading to
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the failures, the DPC and B&W programs for determining the cause
of the failures, the corrective actions planned, methods of repair
and the effect on the plant schedule.
The inspector also discussed
these items with McIntosh, Brown and Morgan.
In addition, the
inspector extmined the equipment and observed some of the nondestructive
testing which was in progress.
According to McIntosh, the first indication of potential trouble
was noted during heatup of the reactor coolant system on March 3, 1972.
Noises were heard in the compartment housing the once Through Steam
Generator "A" (OTSG A) when the coolant temperature reached
approximately 3500F.
(See Exhibit A)
The signal from one accelerometer
indicated that it had failed on that day.
On March 4, 1972,
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it was decided that the noise was from OTSG A.
The noise stopped
when the reactor coolant pumps were stopped.
B&W engineering
personnel were brought to the site and the decision was made
to continue the hot functional testing while noise measurements
were made. On March 7, 1972, investigations showed that 23
of the 60 temporary thermocouples in
"B" steam generator had
failed.
On March 10, 1972, at the conclusion of the first
phase of hot functional testing, cooldown of the system was begun.
The system was depressurized and drained on March 11 and 12,
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1972.
On March 13, 1972, the manway on OTSG A was removed and
metallic material was observed on the tube sheet.
The manway
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on OTSG H was also opened and inspection revealed that the thermocouple
harness conduit had been destroyed and four pieces of metallic
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material were seen on top of the tube sheet.
Inspection of the
remainder of the coolant system was initiated.
At the time of
the Compliance inspection, the following damage had been observed:
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CO Report Nos. 50-269/72-3
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All tubes ends in OTSG A were damaged extensively.
The
a.
tube sheet cladding and head cladding were also damaged.
b.
Approximately 12,000 tube ends in OTSG B were moderately
or lightly damaged. About 7,000 will require machining
and 700 rewelding.
The temporary thermocouple installation was partially destroyed.
c.
Twenty-one in-core instrument stub tubes were broken off
at the heat affected zone of the weld or slightly above.
Fourteen additional tubes were cracked.
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d.
Four in-core instrument tubes were broken off below the flow
distributor and one also broken off above.
Four additional
tubes were cracked.
Nicks and scratches were observed on the flow distributor,
e.
the flow guide vanes and the reactor vessel outlet nozzle.
f.
Two of six thermocouple guiis tubes in the upper plenum were
broken off.
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One surveillance specimen holder support bracket was cracked.
h.
Metal upset occurred at the mating edges of the thermal shield
and the lower grid assembly.
The retention welds on the eight
dowels at the lower edge of the thermal shield were broken
and one of the dowels backed out approximately three-fourth
inch.
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Nicks and scratches were found on the reactor coolant pump
impe11ers.
j. Metal particles were scattered throughout the system.
Repair of the tubes in OTSG B had been initiated at the time of the
inspection, but was limited to the removal of damaged metal at
the tube ends.
Dye penetrant testing of all welds and possible
points of damage to the reactor vessel and internals was in progress.
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A meeting was planned with DRL and Compliance for.the first week
in April to outline the repairs to OSTG A.
Thies advised the
- .nspector that B&W was undertaking a rehnalysis of the coolant
system design in an attempt to determine the cause of the failures.
A duplicate set of internals were being tested for vibration response
characteristics.
Thies, in response to the inspector's questions, stated _that the
extent of the delay to the Oconee units had not been determined
and in fact could not be determined with any accuracy until the
cause of the failures had been determined.
He said that he had
advised the Federal Power Commission that it was possible that
Unit 1 would not be available to meet the 1972 winter system
peak. He caid that formal report on the incident would be sent
to DRL.
The inspector advised. Smith that he had observed dye penetrant
testing being performed on the vessel internals and that the dye
was not being removed to the extent required by code.
He cautioned
Smith that all work on repairing the damage be done by qualified
personnel using approved. procedures and that the non-destructive
testing also be done by qualified personnel using approved procedures.
He advised Smith that the repairs and testing operations should be
well documented since Compliance would be following these operations
during each inspection.
Smith stated that the work would he done
according to the applicable codes and regulation.
He ales immediately
notified B&W to correct their PT methods.
These items were discussed
in the management interview.
The inspector does not plan any
enforcement action on the PT deficiencies since corrective action
was taken immediately and because repaired areas will be retested.
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C0 Report Nos. 50-269/72-3
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SECTION III
Prepared by R. F. Warnick, Reactor
Inspector
ADDITIONAL SUBJECTS INSPECTED, NOT IDENTIFIED IN SECTION I, WHERE NO
DEFICIENCIES OR UNRESOLVED ITEMS WERE FOUND
1.
General
Summerlin, a DPC staff engineer, has prepared a summary of the
preoperational and startup test programs.
The status as of March 23,
1972, is as follows:
Total
Approved
Tests
Results Approved or
Number
Procedures
Completed
Tests Signed Off
Tests before
hot functionals
380
380 (100%)
380 (100%)
380 (100%)
Tests during hot
functionals
160
158 ( 99%)
89 ( 56%)
60 ( 38%)
Tests prior to
core loading
143
97 ( 68%)
59 ( 41%)
51 ( 36%)
Total Preoperational
tests
683
635 ( 93%)
528 ( 77%)
491 ( 72%)
2.
Preoperational Test Program - Procedure Review
TP-1A-600-14, Revision 5, Pipe and Component Hanger Hot Deflection
and Inspection Test
TP-1B-600-27, Center CRD Venting for Startup Testing
DETAILS OF SUBJECTS DISCUSSED IN SECTION I
3.
Preoperational Test Program - Procedure Review
a.
TP-1A-330-3A, Control Rod Drive Rod Drop Time Test
Bradham reported this procedure had been rev. sed to include
dropping the slowest and fastest rods 25 times.1/
b.
TP-1B-600-22, Revision 3, Degassification Test
The procedure has been clarified to specify that the pressurizer
will be degassed.2/
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1/ C0 Report No. 50-269/72-1,Section III, Paragraph 5.a.
2/ C0 Report No. 50-269/72-2,Section III, Paragraph 4.c.
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TP-1A-600-26, Revision 2, 1500 psig High Pressure Injection
Engineered Safeguard Test
The questions raised by the inspector have been satisfactorily
resolved. The ES signal to valves HP-20 and HP-21 is no longer
being blocked (jumpered) .
The procedure checks that the isolation
valves in the purification letdown line close on high pressure
injection signal. TP-202/5 tests that the three high pressure
injection pumps start automatically and reach full speed within
six seconds after the actuation signal.
TP-202/5 also tests
that a high pressure injection string can deliver 450 gpm
at 585 psig.1/
4.
Preoperational Test Program - Evaluation of Test Results
TP-1A-201-5, Revision 2, Core Flood System Flow Test
a.
As reported earlier,2/ the inspector reviewed the completed
test and concluded that the test results were questionable
since the B tank level instrument indicated below the level
of the lower instrument tap.
The test was also evaluated by
B&W and they indicated by a letter dated July 27, 1971, that
the results were satisfactory even though the level instrument
on B tank may have indicated improperly.
B&W reco= mended the
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tank level and pressure instrumentation be checked prior to
operation.
Since then, the instrumentation has been chacked and the results
were reported by the plant superintendent to be satisfactory.
In addition, the inspector reexamined the B tank level instrument
trace from the test master file and determined that the B tank
level instrument did not indicate below the low level tap.
Both A and B tanks did meet the test acceptance criteria.
This item is considered to be resolved.
b.
TP-1A-204-4, Revision I, Reactor Building Spray System Functional
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The acceptance criteria have been changed to agree with the
way the test was conducted.
This is now considered to be
satisfactorily resolved.
1/ C0 Report No. 50-269/72-2,Section III, Paragraph 4.e.
2/ CO Report No. 50-269/72-1,Section III, Paragraph 4.b.
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SECTION IV
Prepared by: F. Jape, Reactor Inspector
ADDITIONAL SUBJECTS INSPECTED, NOT IDENTIFIED IN SECTION I, WHERE NO
DEFICIENCIES OR UNRESOLVED ITEMS WERE FOUND
None
DETAILS OF SUBJECTS DISCUSSED IN SECTION I
1.
Review of Initial Fuel Loading Procedure
An approved copy of OP 1502/04, " Initial Fuel Loading," was reviewed
by Murphy, Warnick and Jape, prior to the on-site inspection.
Each
inspector submitted comments which were collectively presented
by Jape at a meeting with DPC personnel for their consideration.
The review was done to ensure that the requirements specified
in the Oconee Technical Specifications, FSAR manuals and the
USAEC's " Guide For the Planning of Initial Startup Programs"
would be fulfilled.
The comments and the licensee's response are listed below.
In
every case, a suitable response or co=mitment was obtained from
the licensee.
a.
Comment
The relationship between DPC and B&W during initial core
loading is not covered in detail. The question regarding
how differences between B&W and DPC are to be resolved is
not covered in the procedure.
Licensee's Response
B&W personnel will be present during initial fuel loading
in an advisory capacity. DPC will have final word on all
decisions.
This is essentially the same as all other testing
work is setup.
b.
Comment
The responsibility as to who interrupts loading data and allows or
aurhorizes the next fuel assembly to be placed into the reactor is
not clear in the procedure. An organization chart showing minimum #
staffing, plans for relief men for lunch periods and shift
turnover and the chain of command would reatly improve the
procedure.
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Licensee's Response
An organization chart with duties spelled out will be. added
to the procedure.
Basically, the senior reactor operator
will be in charge of. fuel loading and will.be expected.co
utilize the technical personnel, which will be available to
him, for making the decisions necessary as the loading progresses.
c.
Comment
The procedure does not require a response test of the flux
monitors with a neutron source prior to beginning fuel loading.
Licensee's Response
The procedure will be revised to require that all four flux
monitors, two in-core and two out-of-core, be source tested
within eight hours of beginning fuel. loading.
In addition,
the first fuel assembly to be installed has a neutron source.
d.
Comment
The procedure implies that the plant superintendent can approve
changes in the requirements for nuclear flux monitors.
Licensee's Response
This is not intended..nor-planned. The plant. superintendent.
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recognizes he.cannot authorize changes _to the Technical Speci-
fications or FSAR. 'A minimum of two flux monitors should always
be in service during any change in core geomentry.
e.
Comment
The procedure does not cover the problem of equipment check-
out before resuming fuel loading operations following a signi-
ficant delay.
Who has the authority to restart loading operations
after a delay?
Licensee's Response
A test will be added to the procedure to response check at
least two neturon detectors before restarting after any eight-
hour delay or interruption in fuel loading before.
The SRO
has the authority to resume operations.
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f.
Comment
At some point in the procedure, before the two special in-core
flux monitors are to be removed from the core to load fuel
in their platg the two out-of-core flux monitors need to be
verified to be.on-scale.
Licensee's Response
This question will be resolved by adding a statement to the
" Limits and Precautions" section of the procedure to identify
the point where the out-of-core flux monitors are on-scale
and may be used in place of the two in-core monitors.
g.
Comment
One of the in-service flux monitors should have an audible
count rate signal in the control room.
Licensee's Response
The installed equipment does not have any provision for an
audible signal, and the special. in-cora monitor will. indicate
only in the. reactor building.
In a telecon, on April 27, 1972,
Hampton stated that the licensee would reconsider their position
on audible signals and/or indication of the in-core monitors
in the control. room.
h.
Comment
A communications network is required between all. personnel
involved in the fuel loading operation not, as stated in the
procedure, between the cont-C room and reactor building
personnel only.
Licensee's Response
The plan is to have a communications network between all
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personnel involved in fuel loading.
The procedure will be
corrected to show this as a requirement at all times when
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fuel is being handled.
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Comment
The procedure states that a mini =um of four bolts be used to
secure the equipment hatch during fuel loading.
The question
is raised regarding the geometry of these four bolts.
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Licensee's Response
The procedure vill state that the four bolts are to be
approximately 900 apart.
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Comment
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The procedure states that the reactor building purge system
will be tested; but the test details are not given.
Licensee's Response
The test objective will be added to the procedure and the
test will be expanded.to include a check of the reactor
building evacuation alarm. This test is to be done within
one week prior to fuel loading.
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Comment
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Items T and U in the " Limitations and Precautions" section
of the procedure appear to conflict.
Some clarification is
needed on these two items.
Licensee's Response
In itec T, reference to nuclear fuel will be deleted and item
U will be reworded. When completed, the two items will be
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three separate accident types.
1.
Comment
The provisions stated in the procedure for sampling boron appear
to be lacking sufficient detail regarding the frequency for
sampling prior to establishing equilibrium concentration and
t,he location of sampling points.
Licensee's Response
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The wording in the procedure will be clarified to resolve
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these questions of frequency and location of sampling.
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Comment
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The wording in the procedure should be direct regarding the
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analysis and reporting of results of samples soon after the
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sample is taken.
In other words, samples should not be taken
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cn swing and graveyard shift and held for the day shift to run
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the analysis.
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Licensee's Response
It was intended that samples be analyzed on a continuing
basis. The wording in the procedure will be changed to reflect
these plans.
The organization chart will show a chemist to
be available around the clock.
n.
Comment
Besides sampling anu analyzing for boron, what about other
water quality requirements?
Licensee's Response
Water chemistry specifications will be included in the procedure.
o.
Comment
The pool water level is given, but a phrase which says that
the superintendent may approve a change in this specification.
What is intended here?
Licensee's Response
This was intended to permit flexibility in case it's decided
to load fuel with the spent fuel pit dry and only the reactor
filled with water instead of having the fuel pit, canal and
reactor inundated.
This phrase will be removed and if plans
are changed to having only the reactor full of borated water,
a revised procedure will be issued.
p.
Comment
The procedure does not specifically state whether or not
a map of the spent fuel pit showing the location of each assembly
will be available in the control room and spent fuel pit.
Will such a map be used?
Licensee's Response
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No, the computer and existing accountability procedures make
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such a map unnecessary. A status map of fuel loading progress
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will be available in the control room.
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q.
Comment
Nowhere in the procedure is there a description of the 1/M
plot or how to normalize the data.
Shouldn't these be a part
of the procedure?
Licensee's Response
A description of the 1/M plot will be added and an example
of how to normalize the data will also be added to the procedure.
r.
Comment
Foreseeable emergencies need to be covered in more detail.
For example each of the following should be covered:
(1) Dropped fuel assembly.
(2) Abnormal count rate.
(3)
Fuel handling crane hoist over travel.
(4) Unresponsive neturen flux monitor.
(5) Loss of communications.
Licensee's Response
Each of these items are covered in the procedure, in some
cases indirectly.
For example, the procedure requires
communications and upon loss of communications, all fuel
movement must stop.
The only item that requires additional coverage is the fuel
handling crane hoist limit switches.
The plan is to functionally
test each limit switch within two weeks of starting fuel loading
operations.
This will be added to the procedure.
s.
Comment
The criteria for emergency boron injection-is not stated in
the procedure and the reference to OP 1104/6 for emergency
boron addition appears unwieldy.
The referenced procedure
is not designed for emergency action. _It should be possible
to inject boron by opening a single valve and starting a single
pump from the control room upon receipt of some emergency
signal.
The valve lineup should provide this capability.
Licencee's Response
This comment will be studied and an answer will be given at
a later date.
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t.
Comment
The reactor building evacuation horn is not discussed in the
procedure.
Licensee's Response
A test of the reactor building evacuation horn will be added
to the procedure.
The test will be done within one week of
initiation of fuel loading,
u.
Comment
What will be the status of the polar crane during fuel loading?
Licensee's Response
This item is covered in the Technical Specifications and there
is no need to repeat it in this procedure.
v.
Comment
Will the in-core power range flux monitors be inserted during
or after fuel loading? If it's to be done daring fuel loading,
the procedure should cover the details.
Licensee's Response
This has not been decided at this time; but, if they are to
be inserted during fuel loading, the procedure will be revised
to include the necessary details.
w.
Comment
The procedure does not specifically state if health physics
monitoring will be provided continuously during fuel loading.
What are the plans?
Licensee's Respense
Health physics coverage will be provided continously.
The
organization chart, which is to be added, will indicate this.
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C0 Report Nos. 50-269/72-3
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50-270/72-2
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x.
Comment
The status of the reactor coolant system is not stated. What
will be the status of all pumps and equipment needed for fuel
loading?
Licensee's Response
This item will be studied in more detail but, generally speaking,
the status of all equipment will be verified before and during
fuel loading operations.
y.
Comment
When were the fuel assemblies last inspected? Will there be
a final inspection prior to loading into the reactor? And,
were any fuels assembled or disassembled at Oconee?
Licensee's Response
-
Fuel inspection was completed in December 1971, when the
assemblies were placed into the spent fuel pit. There are
no plans for another inspection at this time.
And no assemblies
were assembled or disassembled at~the Oconee station.
z.
Comment
It has been found necessary at other PWR's to stop all flow
when charging the fuel assembly nearest the reactor coolant
inlet. This procedure and the Technical Specifications do
not allow this.
Licensee Response
The procedure will be revised to permit P.he option of stopping
flow-to permit insertion of an assembly near the inlet.
aa.
Comment
.
Will every operations man be given an opportunity to go through
a dry run of the fuel loading procedure.
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C0 Report No. 50-269/72-3
50-270/72-2
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Licensee's Response
No.
All personnel who will run the cranes and bridges will
be trained and will have made dry-runs.
Not every operations
man will make dry-runs.
bb.
Comment
i
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Does the fuel handling hoists have " dead-man" switch controls?
Licensee's Response
t
!
Yes.
"
cc.
Comments
1
1
In step Q, page 3, it is stated that proper seating will be
'
visually verified. How is this done?
Licensee's Response
i
Verification of proper seating of fuel assemblies will be done
,
by the E axis tape readings.
The visual requirement was
intended to simply mean to read the tape measurement. This
wording will be clarified.
dd.
Comment
.
Is there a proper orientation of fuel assemblies when placad
'
in the reactor?
Licensee's Response
No.
The fuel will be oriented with all numbers facing in the
same direction for easy reading. This is not a requirement
of the procedure.
ee.
Comment
The procedure requires all personnel entering and leaving
the containment building to log in and out.
Who will be allowed
to sign in, and where will the logbook be stationed?
T
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CO Report Nos. 50-269/72-3
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50-270/72-2
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Licensee's Response
This will be clarified.
The logbook will be at the airlock
with someone in attendance at all times.
The procedure will
specify who can and how many people will be allowed to be in
the containment vessel.
ff.
Comment
Item CC under " Limitations and Precaut' ion" states that " Class
i
A cleanliness standards will be maintained . . . ." What is
class A cleanliness?
'
Licensee's Response
The reference to class A cleanliness is incorrect.and will
be deleted.
In its place will be a statement similiar to
those contained in OP/1503/1 which prescribes rules for
control of all tools and personnel gear when working over the
reactor.
gg.
Comments
The radiation monitors, located on the bridges over the
reactor vessel, do not have a local, audible alarm. The
only light is a red light.
Is this adequate communication
for personnel protection?
Licensee's Response
This will be looked into and most likely an audible alarm
will be added to these two radiation monitors.
hh.
Ccmment
Several typographical errors were found. These are:
4
Procedure No.
Page No.
Item
As Found
Should Be
OP/506/1
6
Bl(h)
attached
lit
l
OP/1502/4
2
1st line
0115
015
OP/1502/4
2
5th line
008
C08
OP/1502/4
3
13th line
020
C20
OP/1502/4
5
23rd line
040
C40
%
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CO Reports No. 50-269/72-3
50-270/72-2
-21-
2.
Evaluatica of Test Results
The results of two preoperational tests were reviewed.
The results
of both tests appear to fully satisfy the acceptance criteria of
the test. A brief discussion of each test is presented below:
'
a.
TP 1B 270/33, Main Steam Safety Valve Lift Test
This test was conducted on March 7,and 8, 1972.
The Oconee
audit by R. A. Benfield is dated March 20, 1972, and the
" test completed approval" by Smith is dated March 21, 1972.
The test was also audited by S. E. Nabors, GORC, on March 9, 1972.
All of the audits indicated that the test was satisfactory
During performance of the test, when about four of the
sixteen valves had been tested, it was noted by McIntosh
that each relief valve had to be reset several pounds. He
then stopped the test and rechecked the specifications and
found them to be in error.
The valves which had been completed
so far were then redone and the remainder of the testing was
.
completed.
(
The test results and the licensee's evaluation are satisfactory.
b.
TP/IC/600/24, RV and SG Support Flange Temperature
This test was performed on February 23 and 24, 1972. An
audit check by Benfield is dated March 23, 1972, and Smith
has not signed the test as completed yet.
However, the data are sufficient to conclude that the test
4
has met the acceptance criterion.
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Ltr to Duke Power Company
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dtd 5/11/72
.
DISTRIEl' TION :
J. G. Keppler, RO
J. B. Henderson, R0
R0 Office of Operations Evsluation
RO AD for Indpection & Enforcement
L, DD for Reactor Projects
RO Files
DR Central Files
FDR
Local PDR
'iSIC
DTIE
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