ML19322A650
| ML19322A650 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/27/1972 |
| From: | Jape F, Murphy C, Warnick R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19322A648 | List: |
| References | |
| 50-269-72-09, 50-269-72-9, NUDOCS 7911210713 | |
| Download: ML19322A650 (22) | |
See also: IR 05000269/1972009
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UNITED STATES
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DIRECTORATE OF REGULATORY OPERATIONS
RO Inspection Report No. 50-269/72-9
Licensee: Duke Power Company
Powar Building
422 South Church Street
Charlotte, North Carolina 28201
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Facility Name: Oconee 1
Docket No.: 50-269
License No.: CPPR-33
Category: B1
Location: Oconee County, South Carolina
Type of Licensee: B&W, PWR, 2452 Mwt
Type of Inspection: Routine, Unannounced
Dates of Inspection: October 3-6, 1972
Dates of Previous Inspection: August 22-25, 1972
Principal Inspector: MM.M [d, etwdle
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R. F. Warnick, Reactor Inspector *
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Facilities Test and Startup Branch
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Accompanying Inspectors: Aan U
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'F. Jape, Reactor / nspector**
Date
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Facilities Test and Scartup Branch
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W.'D.~Kelley,iteactorfnspector*w*
Facilities Construction Branch
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A. K. Hardin, Reactor Inspector **
Date
Facilities Operations Branch
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N. Tconomos, Reactor Inspector **
Date-
Facilities Construction Branch
Other Accompanying Personnel: C. E. Murphy, Acting Chief, Facilities
Test and Startup Branch ****
Reviewed by: Y
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M/2 72-
C. E. Murphy,4ctin'g Chief
'Date
Facilities Test and Startup Branch
- At the site october 3-6, 1972.
- At the site October 3-4, 1972
- At the site October 4-6, 1972
- At the site October 6, 1972
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RO Report No. 50-269/72-9
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SU!EARY OF FINDINGS
A.
Enforcement Action
None
B.
Licensee Action on Previously Identified Enforcement Item
Use of memoranda in lieu of approved procedures has been clarified
by DPC. Thf; item is closed.
(See Details I, paragraph 2.)
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C.
Unusual C 2currences
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None
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D.
New Unresolved Items
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Criticality control during fuel handling.
(See Details III, paragraph 2.)
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E.
Status of Previously Reported Unresolvrd Items
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The following items have been satisfactorily resolved:
1.
Single-loop, two-pump operation is covered by Technical Specification
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3.1.8.6.
(See Details I, paragraph 3)
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2.
Heater and heat tracing tests, as described in test procedure TP
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210/8, will be completed prior to loading fuel into the reactor
vessel.
(See Details I, paragraph 4.)
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3.
The control rod drive cooling system will be tested during the
hot functional test program as described in test procedure TP 600/3.
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(See Details I, paragraph 5.)
4.
The loss of instrument air test has been completed and DPC has
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agreed to conduct power ascension program tests on (a) shutdown
from outside the control room at ten percent of full power, and
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(b) psuedo rod ejection at forty percent of full power but with
the rod configuration for one hundred percent power.
(See Details
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I, paragraph 6.)
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RO Report No. 50-269/72-9
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5.
An audible neutron count-rate signal will be provided in both the
control room and the reactor building during initial fuel loading.
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The emergency boron injection requirements have been included in
the initial fuel loading procedure.
(See Details I, paragraph 7.)
6.
The plant work order form has been revised to include a requirement
for testing where appropriate.
(See Details I, paragraph 8t)
7.
A licensed senior reactor operator will be directly in charge
of fuel loading operations and he will have no other concurrent
responsibilities .
(See Details I, paragraph 9.)
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The following item remain', unresolved:
The number of welding filler metal heat numbers that are recorded
on the DPC QC 36 IBM cards, but do not appear on the welding
filler metal certification file printout, has increased from 15
to 25.
(See Details II, paragraph 4.)
F.
Other Significant Finding
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The tube sheet and tube ends of OTSG A sustained minor damage
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from the whiplashing action of an air hose, in the upper dome
of the stems generator.
(See Inquiry Report No. 50-269/72-8
and Details V, paragraph 2.a.)
2.
Ninety-seven Unit 1 fuel assemblies are to be returned to B&W
for pressurization. This will result in having all fuel as-
semblies for Unit 1 pressurized.
(See Inquiry Report No.
50-269/72-9 and Detills III, paragraph 2.)
Management Interview
A management interview was held on October 6, ?972, to discuss the findings
of the inspection. The following people were in attendance:
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Duke Power Company (DPC)
1. L. Dick - Vice President, Construction
D. G. Beam - Construction Manager
C. B. Aycock - Senior Field Engineer
J. R. Wells - Manager Construction Services
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RO Report No. 50-269/72-9
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K. S. Canady - Nuclear Engineer
J. E. Smith - Plant Superintendent
J. W. Hampton - Assistant Plant Superintendent
M. D. McIntosh - Operating Engineer
R. M. Koehler - Technical Support Engineer
The following items were discussed:
1.
Insocction Reoorts To Be Placed In The Public Document Room
The licensee was inforced of the new procedure to release inspection
reports to the Public Document Room. The licensee will have an
opportunity to review the reports for proprietary information prior
to their release.
2.
Fuel Handling Procedures
Fuel handling procedures, accountability logs, and storage facilities
were inspected and the findings were discussed with DPC staff members.
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The procedure, to be followed when 97 fuel assenblies are returned to
B&W, was being prepared but had not been approved. The licensee's
representative stated that an approved procedure would be on hand
before the work commencad.
The inspector stated that criticality control requirements should be
available and should be common knowledge to personnel handling the
fuel assemblies. The inspector also stated that this needed added
emphcsis and attention. DPC indicated that they felt their men were
knowledgeable; however, they did agree to pursue the matter and take
appropriate action.
(See Details III, paragraph 2.)
3.
Initial Fuel Loading Procedure
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Three of RO:II's four comments on the initial fuel loading procedure
were resolved during the inspection as summarized below. The fourth
comment was resolved at a meeting with RO:HQ and L on October 17, 1972.
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R0 Report No. 50-269/72-9
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a.
Audible count rate signals will be provided in the control room
and the reactor building, and at least one will be in service at
all times whenever fuel is being moved within the reactor building.
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b.
A requirement will be added to the procedure that all men who will
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be handling fuel will be trained.
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A health physics technician will be added to the organization chart
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c.
enclosed with the initial fuel loading procedure.
The forth comment involved the location of the licensed senior reactor
operator (SRO) during fuel loading operations. The licensee's position
is that the SRO should be free to be in the reactor building, control
room or spent . fuel pit area. DPC stated that the SRO will not have
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any responsibilities other than fuel loading.
(See Details I, paragraph
7 and 9 and Details IV, paragraph B.)
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4.
B&W Repairs to Reactor Vessel, Steam Generators, and Reactor Internals
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Repairs and modifications to the steam generators, reactor vessel, and
the vessel internals were inspected. DPC's representative agreed to
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inform RO:II of the resolution of the fitup problem between,the reactor
vessel internals - flow distributor and the lower grid.
(See Details V,
paragraph 2)
5.
Outstanding Items List
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The status of items on the outstanding items list was discussed.
(See
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Details I, paragraphs 2 through 9.)
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DETAILS I
Prepared by: hMf[/&
Reviewed by: Mds
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Persons Contacted
Duke Power Company (DPC)
R. L. Dick - Vice President, Construction
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D. G. Beam - Construction Manager
D. L. Freeze - Principal Field Engineer
C. B. Aycock - Senior Field Engineer
K. S. Canady - Nuclear Engineer
J. E. Smith - Plant Superintendent
J. W. Hampton - Assistant Plant Superintendent
M. D. McIntosh - Operating Engineer
L. E. Schmid - Assistant Operating Engineer
S. W. Dressler - Associate Field Engineer, Piping
C. L. Thames - Health Physics Supervisor
M. Ray - Associate Engineer, Welding
A. R. Hollins - Associate Field Engineer, Welding
2.
Use of Properly Aoproved Revisions To Procedures 1/
As stated in DPC's letter from A. C. Thies to J. G. Davis, dated
September 28, 1972, it is DPC's intention to obtain correct ap-
proval for all procedures and revisions to procedures prior to
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issuing them to the field forces for use.
DPC will use memoranda
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only to clarify instructions, but not to change procedures.
The two memoranda previously used in lieu of procedures (Memorandum
from R. E. Blaisdell dated January 4,1972, entitled " Procedure for
Identification and Control of Field Fabricated Pipe [ Attachment Welds]"
and Mechanical Memorandum 8-72 from L. R. Barnes entitled " Control of
Piping Isometrics, QR-27") were observed to have been incorporated into
Oconee Procedure E.1, " Procedure for the Identification and Control of
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Field Fabricated Pipe and Welds."
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This previously identified unresolved item is considered closed.
1/ See RO Inspection Report No. 50-269/72-7,Section II, paragraph 4.
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RO Report No. 50-269/72-9
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3.
Data During Startuo for Single Loop, Two Pump Operations
Requirement 3.1.8.6 in the Technical Specifications requires DPC
to notify the AEC prior to single loop testing, and to report the
results of the single loop testing to the AEC. The AEC's written
approval is required by the same specification before subsequent
single loop operation.
This previously identified unresolved item is considered to be
satisfactorily resolved.
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4.
Heater and Heat Tracing Tests ,f
DPC has provided test procedure TP-1B-210/8, " Trace Heating System
Functional Test," in the preoperational test program to test the
adequacy of the heat tracing on the piping containing boron solu-
DPC indicated the tests will be completed prior to loading
tion.
fuel into the reactor vessel.
There are
The test procedure was reviewed by the RO:II inspector.
no further questions, and this previously identified unresolved item
is considered resolved.
Control Rod Drive Cooling System Tests 1/
5.
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DPC has provided test procedure TP-1B-600/8, " Component Cooling Sys :en
Operational Test," in their preoperational testing program.
ew was documented in
ThetestprocedurehasbeenreviewedandtherepI
the inspection report of February 22-25, 1972. _
This previously
identified unresolved item is considered resolved.
Power Ascension Test Program E/
6.
During the inspection of April 6-9, 1971, RO inspectors observed
that certain tests had been omitted from the power ascension test
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Among the tests identified were (1) loss of instrument
program.
air, (2) shutdown from outside the control room at 100% power, and
(3) rod ejection test at power.
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1/ See RO Inspection Report No. 50-269/70-12.
2/ See RO Inspection Report No. 50-269/72-2,Section III, paragraph 2.
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3/ See RO Inspecti'n Report No. 50-269/71-4, paragraph I.6.
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RO Report No. 50-269/72-9
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A
DPC informed the inspectors that the loss of instrument air test
has been completed.
In addition, DPC has agreed to conduct the
" Shutdown from Outside the Control Room" test at ten percent of full
power, and the " Pseudo Rod Ejection" test at forty percent of full
power but with the rod configuration for 100 percent power.
The completion of the loss of air test and the cotmitments to perform
the other test satisfy RO's concerns and this previously identified
unresolved item is considered resolved.
7.
Fuel Loading 1/
During the inspection of March 21-24, 1972, RO discussed coments
on the initial core loading procedure with DPC, RO's comments were
incorporated in the revised procedure.
Source flux monitor count rate signals will be audible in the control
room and in the containment building at the start of fuel loading.
DPC has agreed that one of these two signals will be in service at
all times during fuel movement in the reactor building.
Emergency boron injection requirements have been added to the procedure.
These previously identified unresolved items are considered closed.
Revision to Work Order Procedure to Include Retest Requirements E!
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During the inspection of May 16-19, 1972, the RO inspector identified
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a weakness in DPC's Station Work Orders.
(The possible need for testing
af ter completion of the work was not identified.)
Until the current supply of work order forms on hand (approximately
5000) is used, DPC indicated they are and will continue to write in
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the lower right hand corner either that "no test is required" or
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else " review for possible retest is required." The licensee indicated
that when new forms are ordered, this information will be preprinted
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on the forms so that only the correct box need be checked.
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This previously identified unresolved item is considered resolved,
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1/ See RO Inspection Report No. 50-269/72-3,Section IV, paragraphs 1.g
and 1.s.
2_/ ~ See RO Inspection Report No. 50-269/72-5,Section II, paragraph 5.
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Duties of the Senior Reactor Operator During Fuel Loading $/
9.
The licensee has stated that a licensed senior reactor operator (SRO)
will be directly in charge of fuel loading and will not be assigned
This was discussed
any other concurrent duties or responsibilities.
with RO:HQ and with Licensing on October 17, 1972, and determined to
be consistent with AEC's position.
This previously identified unresolved item is considered resolved.
10. Fuel Insoection 2/
The need for inspecting fuel assemblies prior to initial core loading
was discussed with DPC during the inspection of July 18-21, 1972. DPC
has agreed to inspect a representative number of assemblies to determine
if inspection of all assemblies is warranted.
The 97 fuel assemblies that are scheduled to be shipped to B&W for
pressurization will be inspected before they are shipped to B&W and
again when r. hey are returned to' the reactor site.
The depth and content of the inspection of fuel essemblies will be
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reviewed during a subsequent site visit.
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J/ See RO Inspection Report No. 50-269/72-7,Section II, paragraph 8.
2/ See RO Inspection Report No. 50-269/72-7, Management Interview,
paragraph 6.
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DETAILS II
Prepared By:
Reviewed By:
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1.
Persons Contacted
A.
Duke Power Company (DPC)
D. G. Beam - Construction Manager
J. M. Curtis - QA Supervisor (Charlotte-Design Engineering)
M. Ray - Associate Engineer - Welding (Charlotte-Construction Dept.)
A. R. Hollins - Associate Field Engineer - Welding
L. R. Davison - Associate Field Engineer - NDT
B.
DPC Consultant
- H. Thielsch
- By telephone conference.
2.
Unit 1 Welding Program Organization
Changes were made on October 1, 1972, to personnel assignments as
reflected in the Functional Chart outlined in the Quality Assurance
and Functions manual prepared by DPC's consultant.
In a discussion with the DPC personnel involved in the program, the
inspector found that sone uncertainty existed among some of the members
as to who had the technical responsibility for the program. This item
was discussed with the licensee's representatives who assured the inspec-
tor that this item had been previously discussed with the individuals
involved. Prior to the conclusion of the inspection, the inspector was
advised that the matter had again been discussed with each individual.
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The inspector plans no further action on this item at this time.
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3.
Status of Corrective Work
The " List of 141" is the list of Class I piping buttwelds above 8
inches in size that had radiographs with artifacts (pencil marks) and
were to be radiographed. Two of the welds have not been radiographed;
Weld No. 53B,1-B8 is encased in concrete and No. 53 E,10-B74 is in-
accessible in a pipe chase. These welds have been placed on the
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R0 Report No. 50-269/72-9
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" Variation List for Unit One Re-Radiography," and an engineering
evaluation will be made to resolve the discrepancy. Most of the
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raradiographs of the buttwelds have been reviewed and approved by
the DPC consultant and the Hartford Insurance Inspector.
The " List of 52" is the 52 Class I piping buttwelds, 8 inches and less
in size, that had radiographs with artifacts (pencil marks) that were
selected for reradiography and to be used as a basis for the engineering
evaluation and acceptance of the balance of radiographs with pencil
marks. All reradiographs of these 52 buttwelds have been reviewed and
accepted by the DPC consultant and the Hartford inspector. None were
inaccessible.
The " List of 164" is the list of Class I piping buttwe.ds, (of all sizes)
whose radiographs were disapproved, because of radiographic technique,
by Industrial Inspection Industries Inc. , (IIII) . A Level 11 radiog-
rapher has been employed by DPC to perform a 100% reevaluation of all
radiographs. All buttwelds on the list have been reradiographed with
the exception of 51A,1-114a which was inaccessible in a sleeve, and
51A, 1-128AA that has a drain line located in the center of the weld
(half coupling-saddled on). The majority of the reradiographs have
been evaluated and approved by the DPC consultant and the Hartford
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inspector.
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The " List of 44-47" is a list of buttwelds that required reradiography
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by IIII using finer grain film due to questionable indications in the
weld area. All of these buttwelds have been reradiographed and approx-
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imately thirteen required repair after they were evaluated by IIII and
DPC consultant.
A new list has been made of welds that had not been reradiographed
because they were inaccessible or where radiography w:as obstructed.
There are 12 welds listed on this " Variation List for Unit One
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Raradiography."
4.
Identification of Welding Filler Metal
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The verification of the filler metal heat numbers by DPC continues.
The number of filler metal heat numbers that appear on the QC-36
IBM cards, but do not appm.r in the filler metal certification file,
has increased from 15 to 25. DPC has listed 12 of these heat numbers
where they feel the discrepancies are due to the transposition of
numbers, such as, leaving off the last numbers, or using the wrong
prefix letter. Examples are as follows:
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Heat No. on QC-36
Possible Sol , .: ion Heat
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IBM Card
No. in Certi ' cation File
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4598
45983
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04897
04797
45982
45983
123626
12362C
05 MD 45
05 MO 43
C 129 1T
E1291T
The mill certifications for two previously unidentified heats have
been received from Acros Corporation for heat Nos. D 9291N and
E 1075L but an explanation of how these two heats of filler wire
could have been received and issued to the weldors without being in
the filler metal certification file has not been investigated.
The QC-36 card for the buttweld in system 50 (1-inch instrument line),
Iso 4, weld 4, has listed mill certification heat No. 12547, which is
a heat number format used by the McKay Company. The buttwald in the
instrument line is a 304 stainless steel valve welded to an inconel
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safe end. The QC-36 card states that the EB insert was inconel; however,
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the associate field engineer-welding stated he was informed that McKay
does not Janufacture the EB type insert nor does it produce inconel
weld filler material.
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Heat Nos. D9281 and C1292T are valid ra :1 :ertification heat numbers
for " Carpenter 21" type weld filler macerial manufactured by Arcos
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Corporation; however, the associate field engineer-welding states that
to the best of his knowledge, no " Carpenter 21" weld material has been
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used at Oconee.
Drill shavings from weld deposits made using six inconel mill certification
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heat numbers have been sent to Pittsburgh Testing Laboratory for a
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chemical analysis.
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Post-Weld Heat Treatment
DPC'r, consultant has issued report No.1027, dated September 9,
1972, entitled " Confirmation of Postwela Heat Treatment, Carbon Steel
Piping Welds, Systems Ola and 03. Unit No.1. . . ." The report con-
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tains statements from DPC personnel that the six piping buttwelds in
question, where heat treatment charts were lost, had been post-weld
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RO Report No. 50-269/72-9
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heat treated. The report states that confirmation of the post-weld
heat treatment was obtained by performing Brinell hardness measure-
ments on the weld deposit and the heat af f ected zone on each side of
the weld desposit using a portable Telebrineller hardness instrument.
The report states that the hardness reading for the "as-welded" con-
dition for these welds would be approximately 200 Brinell. The results
from the hardness measurements were in the range from 132 and 165, thus
confirming that the welds in question did receive post-weld heat
treatment.
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RO Report No. 50-269/72-9
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Prepared By:
DETAILS III
Reviewed By:
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Persons Contacted
.DPC
J. E. Smith - Station Fuperintendent
M. D. McIntosh - Operating Engineer
T. McConnell - Engineer
B. Moore - Engineer
J. W. Hampton - Assistant Station Superintendent
S. Holland - Relief Shift Supervisor
R. Koehler - Technical Support Engineer
R. Wilson - Performance Engineer, Unit 1
L. Schmidt - Assistant Operating Engineer
2.
Fuel Handling Procedure Discussion
According to the licensee, 97 Unit 1 fuel assemblies are to be
returned to B&W for pressurization.
(See RO Inquiry Report No.
dated September 22, 1972.) The procedure for removing
50-269/72-9,
the control assemblies from the fuel bundles was reviewed by Warni".k
on October 6,1972. The procedure for repackaging the fuel had no-
The licensee rep-
been completed at the time of the inspector's visit.
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resentative indicated the procedure would be prepared before any fuel
was moved and chat it would essentially be a reverse of Oconee's
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Operating Procedure, OP-1503-04, "New Fuel Assembly Inspection and
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14, 1971. OP-1503-04 contains the instruc-
Storage," dated September
tions for receiving, unloading, inspecting, and storing of new fuel.
In discussion of procedure OP-1503-04 with the licensee, questions
were asked by the inspector regarding criticality control in fuel
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Specific answere relative to the number of fuel assemblies
handling.
which could be critical in a moderated system and whether containers
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could be stacked in a close packed, unlimited, geometrical array were
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The inspector das informed by a DPC representative
not obtained.
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that criticality control measures stated in the FSAR were used.
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measures were reviewed in the Regional office following the inspection
and the information in the FSAR is not sufficient to permit a deter-
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aination of the requirements to preclude criticality during handling
Followup action will be taken during a subsequent inspection.
operations.
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In later discussions with licensee management, the inspector pointed
out that personnel handling fuel should be fully cognizant of criti-
cality control requirements. The licensee stated they would review
and augment their training program in this area.
3.
Review of Procedures
Oconee Procedure OP-1503-04, "New Fuel Assembly, Inspection and Storage,"
and OP-1103-18, " Control and Accountability Procedure for Nuclear Fuel
Material," were reviewed. The inspector raised questione en additional
instructions which had been handwritten into OP-1503-04 and asked the
licensee representative how such changes were made and approved. The
questions were resolved prior to leaving the site.
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4.
Inspection of New Fuel Storage
All of Unit 1 fuel and a portion of Unit 2 fuel is stored in the Unit 1
and 2 spent fuel storage pit. The remainder of Unit 2 fuel is stored
in the new fuel secrage building. These storage facilities and the
equipment for handling fuel were inspected. The inspector had no com-
ment regarding the status of the storage facilities or handling equipment.
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RO Report No. 50-269/72-9
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Part IV
Prepared By:
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Reviewed By: E 1. J [ h / n d h
1.
Persons Contacted
J. E. Smith - Plant Superintendent
J. W. Hampton - Assistant Plant Superintendent
R. Wilson - Performance Engineer, Unit 1
F. Hood - Junior Engineer, Unit 1
2.
Initial Fuel Loadine Procedure
The "Oconee Nuclear Station Initial Fuel Loading Procedure," dated
September 11, 1972, was reviewed and discussed with the licensee. RO's
comments and the licensee's response are listed below,
a.
Health Physics Monitoring
RO's Comment
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The procedure does not specifically state that health physics me,n-
itoring will be provided continously during fuel loading. This
same comment was discussed for the procedure dated January 8, 1971,
and at that time it was RO's understanding 1/ that HP coverage would
be included in the organization chart showing the fuel handling team.
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What are Duke Power Company's plans for covering this item?
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Licensee's Response
The organization chart, v.C.ch is shown in Enclosure 6 to the pro ,
cedure, will be revised to include health physics as part of the
team.
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If See RO Inspection Report No. 50-269/72-3,Section IV, paragraph 1.
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R0 Report No. 50-269/72-9
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b.
Fuel Handlers
RO's Comment
In the previous discussions on this procedure, the question was
asked about the training and qualifications for the fuel handlers,
and if these men would be given a practice run before actual load-
ing. The response at that time was that all personnel who operate
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the cranes and bridges will be trained and will have made several
practice runs. There is nothing in the current issue of the proce-
dure that refers to training or practice runs. How will the train-
ing of fuel handlers be documented?
Licensee's Resnonse
A statement will be added to the procedure stating that the fuel
handlers have been trained and are qualified to operate the fuel
handling equipment.
c.
Duties of SRO
RO's Comment
Item 10 on page 3 cnd item B-1 on page 8 of the procedure, states
that the SRO shall be in charge of fuel loading operations and
that he will be at the site while work is in progress. The prece-
dure should state that the SRO in charge be directly supervising
the fuel loading operations and that he have no other responsi-
bility during fuel loading.
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Licensee's Resoonse
The SRO assigned to the fuel loading operation will be in direct
communication with the control room, spent fuel pit, and the reactor
building. The SRO in charge will have no other responsibilities ,
during fuel loading.
However, it is Duke Power Company's position that his movements
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not be restricted any further by requiring him to remain in the
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reactor building.
(At a meeting with RO:HQ and L on October 17,
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1972, the licensee's position on this issue was determined to be
consistent with AEC's position and the comment is considered resolved.)
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d.
Audible Signals
RO's Comment
Item 14 on page 8 stated that the audible count rate signal may
As a minimum,
be either in the control room or reactor building.
the signal should be in the control room, and preferably it should
be at both locations.
Licensee's Resconse
Two audible signals will be provided; one in the control room and
one in the reactor building. One of these will always be maintained
If one of
in service during fuel movements within the reactor.
the two audible signals fails, work will not be stopped, instead
repair will be attempted while fuel loading continues.
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R0 Report No. 50-269/72-9
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Prepared By:~/. 7< -4/vu.*- , ~ ;_ -
DETAILS V
Reviewed By:
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1.
Persons Contacted
Babcock & Wilcox Construction Company (B&W)
a.
W. Faasse - Field Project Manager
C. D. Thompson - Field QC Supervisor
b.
J. W. Hampton - Assistant Plant Superintendent
2.
Repair Status Unit 1
a.
Stean Cenerator A
1/
In discussing the latest reported incident - concerning damage to
the tube sheet and tube ends by an air hose, B&W personnel made the
following comment: The air hose responsible for the damage had been
connected to the clean-air system and had been used to provide suit-
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able atmospheric conditions for the workers. The hose, prior to
the incident, had been doubled-over and secured with several turns
of wire and the air supply had been turned off. It is believen that
someone attempting to use the hose opened the valve and, from an
undisclosed area pulled hard enough, to cause the wire to come off
the hose. This started the hose to whiplash as it was not tied or
fastened to anything inside the dome. Reportedly, the metal coupling
at the end of the hose caused extensive superficial damage to the
tubesheet, tube ends and cladding overlay on the wall surfaces.
This incident occured prior to the lic,uid penetrant test step of '
the general repair procedure, which had been generated by B&W
Construction Company, to be used on this generator. Documentation
refers to the incident in question as a " Deviation to Field Construc-
tion Procedures." Reportedly, tube ends with nicks and dents were
1/ Inquiry Report No. 50-269/72-8
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RO Report No. 50-269/72-9
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buf fed and blended. All tubes with irregularities were inspected
visually and with appropriate gages. Welds requiring veld repair
were charted and subsequently repaired according to proegpures
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used on the generator as a result of the prior incident.--
sentially, the procedure enploys the manual GTAW processes with con-
sumable electrode and a minimum preheat temperature of 250*F.
Results of procedure and performance qualifications were found to
be in accordance with the ASME Boiler and Pressure Vessel Code,
Sections III and IX.
Inspection of repair work was done according
to B&W specification 12-2-WQ1-3, Rev. 1, which is essentially a
visual inspection proced re for welds on commercial nuclear components.
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Results of the nitrogen gas leak test completed on September 30,
1972, showed 23 leak indications. These were investigated and-
repaired.
Preparation for hydrostatic testing was underway with the actus1
testing scheduled for the weekend of the October 7.
Testing will
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be followed by final cleaning. The procedure for cleaning had not
been released at the time of the inspection.
b.
Hydrostatic testing of B Generator completed on September 9 showed.
two leaks, one of which was associated with the first of a kind
(FOAK) instrumentation while the other was located on the bottom
tube sheet. In the case of the first leak, B&W quality control
personnel decided to plug the particular tube exhibiting the leak,
while in the latter case the tube was repair welded.
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This generator is being prepared for final cleaning, to be followed
by installation of F0AK instrumentation.
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c.
Reactor Vessel
Welding of in-core instrumentation nozzles has been completed.
At the time of this inspection, workmen were reaming the I.D.
of these nozzles to drawing specifications.
Final inspection will
be performed after the aforementioned machining operation is
completed.
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RO Report No. 50-269/72-9
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Reportedly, the flev vanes in the lower head will be removed by
arc-air gouging according to a procedure developed by B&W Construction
Company for thermal cutting and gouging of components fabricated
The procedure stipulates that the stub ends
for ASME contracts.
This
will be ground and inspected af ter the cutting operation.
procedure appears to be satisfactory for the intended use.
Present plans call for reducing the number of surveillance specimen
Procedures for this pro-
holder tube assemblies from four to three.
gram are still in the planning stage, with the final revision expected
later in the month of October.
Drilling of bolt-down holes in the thermal shield is now complete.
Reportedly, these holes will be tapped af ter the thermal shield is
separated from the lower grid assembly, thus permitting the use of
The
a cutting lubricant necessitated by the tapping operation.
the threads dry while the shield was in
original intent was to cut
position with the lower grid. However this procedure was revised
as a result of tap breakage which was attributed to the nature of
the material.
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