ML19322A621
| ML19322A621 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/31/1971 |
| From: | Murphy C, Seidle W, Upright C US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19322A619 | List: |
| References | |
| 50-269-71-03, 50-269-71-3, NUDOCS 7911210693 | |
| Download: ML19322A621 (26) | |
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U. S. ATOMIC ENERGY C03CIISSION REGION II DIVISION OF COMPLIANCE Report of Inspection C0 Report No. 50-269/71-3 Licensee:
Duke Power Company Oconee 1 License No. CPPR-33 Category B Dates of Inspection:
February 24-26, 1971 Date of Previous Inspection:
February 10, 1971 Inspected By:
.f-2 4~ 7)
C. E. M0rphl, R(6ctor Inspector (Operations)
Date (In Charge)
S-2$-7/
o C. M.
'p righ epor Inspector (Operations)
Date
- 2/
Reviewed Sy:
W. C. SeidleV Senior Reactor Inspector
'Date Proprietary Information:
None SCOPE A routine, announced inspection was made of the 2,452 Mw(t) pressurized water reactor under construction near Seneca, South Carolina, known as Oconee Station No. 1.
Purposes of the insoection were:
1.
To determine the construction status and significant changes to the schedule dates.
2.
To review the outstanding construction items remaining to be com-pleted at the facility.
3.
To review the progress of the test pro!On, 4
.A, 00 Rp t. No. 50-269/71-3,
SUMMARY
Safety Items - None Nonconformance Items - Contrary to the requirements of the FSAR and established cleaning procedures, the licensee failed to properly pro-tect the reactor vessel and internals from recontamination.
(See Management Interview and Section F.)
Unusual Occurrences - None Status of Previously Reported Items - The licensee's response to the CDN relating to the relocation of the station batteries, the electrical and instrumentation documentation and the procedure for c1 aning the reactor coolant system was considered to be satisfactory.17 Other Significant Items -
1.
The damaged expansion bellows on the fuel transfer tube has been replaced with a new bellows. (See Exhibit A, Item 32.)
2.
Wells advised the inspector that to his knowledge, there had been no overflights of the facility by military or private aircraft.3/
Commercial flights over the facility are at several thousand feet in altitude and are not considered a hazard. Photographs of the f acility are made by the licensee on a periodic basis from a helicopter, but the licensee does not permit this craf t to fly above the plant nor is it permitted to fly so close that it could glide into the plant if it experienced an engine or other malfunction.
Outstanding Items - See Exhibit A for the current status of outstanding items.
Managemen t Interview - The management interview was held on February 26 and was attended by Rogers, Beam, Hunnicutt, Smith and Hamoton.
1.
The inspector expressed concern about the wood splinters in the reactor vessel internals and the polyethylene trapped under the core shield support flange. He stated that it appeared that the vessel and internals had not been properly protected af ter they had been cleaned.
Rogers stated that the ladder had been removed from the vessel immediately after the Compliance inspector had advised them 1/CDN issued January 29, 1971.
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3/ Memo from CO:HQ (O'Reilly) dated February 1,1971.
(75 CO Rpt. No. 50-269/71-3 of the problem. He further stated that they had removed all the splinters that they had located and were ~considering methods of locating and removing any that might retain. He also stated that a study was being made by the licensee and Babcock and Wilcox Company (B&W) of the possible adverse effects of the polyethylene. The inspector stated that he would want to review any in-house report that was issued relating to these items. He also advised Rogers that it appeared that the licensee was in nonconformance with his FSAR and his cleaning procedures and that a CDN would probably be issued relative to these incidents.
(See Section F.)
2.
The inspector advised Hunnicutt that he had observed unidentified stainless steel welding rods on a workbench in the auxiliary building.
He had also observed stainless steel transducer tubing without end caps or other means of protection in the same area. Hunnicutt stated that he would take corrective action to prevent recurrence of this type incident.
(See Section C.)
3.
The inspectors advised Rogers that they had observed a welding electrode holder clamped to an installed section of instrumentation tubing. Rogers stated that this was contrary to their accepted practices and would be discussed at the next foremen's meeting.
(Sae Section C.)
4.
In response to the inspector's question, Hunnicutt stated that the in-house report on the ITE relay failures had not been issued but that he would follow up on this item.
(See Exhibit A, Item 50.)~
5.
The inspectors questioned the adequacy of the drip pans on the fuel handling cranes. Hunnicutt stated that he would review the design of the cranes to assure that pans were of adequate size and installed at all the required points. Smith stated that he would verify that procedures were developed that would require the pans to be inspected and maintained.
(See Section K.)
6.
In response to the inspector's question, Smith stated that the initial fuel loading procedure would be rewritten and that the inspector's comments would be included.
(See Section H.)
7.
The inspector stated that the. revised adminis trative procedure,
" Guide for Conducting the Oconee Initial Test Program," appeared to give the test coordinator more authority to make revisions to procedures than might be permitted by 10 CFR 50, Appendix B.
Smith stated that he did not agree with the inspector's interpreta-tion, but would review the requirements of 10 CFR 50 and would take corrective action if required.
(The inspector received a telephone call from Smith on March 2,1971, and was advised that the procedure
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(75 C0 Rpt. No. 50-269/ 71-3.
would be corrected to limit the coordinator's authority to correcting typographical errors and minor mistakes.) The inspector will review this item m the next inspection.
(See Section G.)
8.
In respenme to the inspector's question, Smith stated that he had checked the operation of the elevator in the auxiliary building and has found that the doors did not close automatically at _ the penetra-tion room level.
(See Exhibit A, Item 59.) He advised the inspectors that the licensee had not decided as yet what corrective action would be taken. The inspectors stated that this would be considered an outstanding item and they would review this item on a future inspection.
9.
The inspector asked Smith if he planned to perform the reactor building controlled leak rate test at both the accident pressure and at 50% pressure.
Smith stated that he did not think that the test would be performed at both pressures but that he would advise the inspector as socn as a definite decision was made. The inspector was advised by telephone on March 2,1971, that the tests would be made at both pressures.
10.
In response to the inspector's question, Smith stated that administra-tive steps were being taken to insure that all comments were incor-porated into the plant procedures prior to their approval and use.
(See Section G.)
DETAILS A.
Persons Contacted Duke Power Company (Duke)
R. L. Dick - Manager of Construction J. C. Rogers - Construction Manager, Oconee Station and McGuire Station D. G. Beam - Project Engineer, Oconee J. R. Wells - Project Engineer, McGuire G. L. Hunnicutt - Principal Field Engineer, Oconee J. E. Smith - Plant Superintendent J. W. Hampton - Assistant Plant Superintendent M. D. McIntosh - Operating Engineer R. M. Koehler - Technical Support Engineer C. A. Price - Electrical Design Engineer J
r7s CO Rpt. No. 50-269/71-3 B.
Administration and Organization The following changes have been made to the licensee's organization since the last inspection:
1.
Rogers has been promoted to the position of Construction Manager with overall responsibility for both Oconee and McGuire.
2.
Beam has been promoted to the position of Project Engineer, Oconee.
3.
Wells has been promoted to the position of Project Engineer, McGuire, but at the time of the inspection, he had not transferred from Oconee.
4.
Hunnicutt has been pronoted to the position of Principal Field Engineer, Oconee.
5.
L. E. Summerlin, formerly Technical Sunport Engineer, Operations, has been given a staff nosition and Koehler has been nromoted to the position of Technical Support Engineer.
C.
Quality Assurance Welding During a tour of the plant, the inspectors observed a welding electrode holder clamped to an installed safety-related instrument sensing tubing. They also noticed several unidentified stainless steel welding rods on a workbench in the auxiliary building and a number of lengths of stainless steel tubing without end caps or other means of protection in the area of the workbench. The tubing was the type normally used for safety-related instrumentation transducer sensing lines. These items were brought to Hunnicutt's attention and he was reminded that welding rod which could not be identified to source had been observed during the previous inspection. Hunnicutt stated that he recognized the seriousness of these occurrences and that he would discuss the items with his inspectors and would take whatever corrective action necessary to prevent a recurrence. These items were discussed in the management interview and the inspectors plan a followup inspection during the next site visit.1/
1/C0 Report No. 50-269/71-2.
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i CO Rp t. No. 50-269/71-3 D.
Construction Progress 1.
The reactor vessel internals, except for the plenum, have been installed.
2.
Installation of the control rod drive mechanisms is in progress.
3.
The erection of the fuel handling cranes is underway. The manu-facturer's technicians are currently making modifications and adjustments to correct deficiencies noted by the licensee.
4.
Mirror insulation is presently being installed on the reactor vessel.
5.
Approximately 10% of the nuclear steam supply system flushing and hydrostatic testing has been completed.
E.
Construction Schedule The following dates were given the inspectors as the best informa-tion available to the licensee and is based on a computer printout on February 25, 1971:
1.
Reactor Coolant System Hydrostatic Test April 21,1971 2.
Reactor Building Leak Rate Test May 1, 1971 4
3.
Hot Functional Test - Start May-8, 1971
- Complete July 11, 19 71 -
4.
Keowee Functional Test July 11, 1971 5.
Fuel Loading July 1971 6.
Start Power Ascension August 31, 1971 7.
Achieve 100% Power October 25, 1971 F.
Reactor Vessel and Internals Installation of Vessel Internals - Attachment K The licensee has installed the lower grid, thermal shield, core barrel and core support shield in the reactor vessel. The inspectors reviewed the records relative to this installation. The Vessel Internals Installatien Procedure, FIP-22, included information concerning the handling, assembly and installation of the reactor internals. A B&W supplied procedure gave detailed steps for the installation of instru--
ments to measure flow-induced vibration of the reactor internals.
Tool control procedures and cleanliness requirements were included in Appendix A of this procedure. No deficiencies esce noted in these
- records, a
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O 00 Rpt. No. 50-269/71-3 -
The licensee maintained a daily summary log of significant events during the installation of the internals. The inspectors reviewed the logs for the period from January 9,1971, through February 10, 1971. The log contained information concerning guide block settings, fitup of internals into vessel,results of laboratory tests for com-ponent cleanliness, and plenum drilling logs. The welding procedures used during the assembly, welding inspection records, security watch logs and tool checklists were also included. No deficiencies were noted in these records.
During discussions of the installation with Hunnicutt, the inspectors were advised that on January 27, 1971, a strip of 4-mil-thick polyethylene had been found trap /
ed between the mating flanges of the core shield and core barrel._
The daily summary log of signif-icant events documented the actions taken by the licensee after the polyethylene had been found. Details of this incident are as follows:
The core support shield lower flange mates to the upper flange of the core barrel <
The two assemblies are bolted together with 120 bolts spaced on approximatsly 3-11/16-inch centers.
During assembly of these componenta in the vessel, the licensee had draped the 4 mil polyethylene sheeting over the walls to afford protection and to maintain the established clean conditions of the vessel and the internals.
Af ter all bolts had been installed and their retaining clips had been tack welded to them, it was found that one section of polyethylene about six feet long had been trapped between the mating surfaces.
(See Exhibit B.)
The licensee attempted to pull the trapped material out and succeeded in removing a strip approximately three feet long.
The width of this strip was appror.imately 1/8 inch. Attempts to remove the remaining piece were unsuccessful. The licensee estimates that the remaining piece is 1/8 to 3/16 inch wide and 3 feet long. Since the distance from the bolt holes to the edge of the flange is 5/8 inch, the licensee is of the opinion that this would represent the maximum possible width of the strip.
Two plies of the material are trapped for a distance of about six inches at one end. The licensee has made tests of the polyethylene sheet and found that when compressed to the degree calculated to be approximately represented by the torqued bolts, that the thickness is reduced from 4 mils to 1 mil. The j
1/ Inquiry Memorandum to Compliance Headquarters (Henderson) from Region II (Seidle) dated March 1,1971.
n CO Rpt. No. 50-269/71-3.
licensee has postulated that the polyethylene will melt during operation and does not consider that the relaxation of the bolts will be a problem.
If the material does melt and is released into the system, the licensee feels that it will be trapped in the purification system filters or will plate out on the steam generator tube sheets. The studies were not complete as yet and Hunnicutt was not able to state what would be the effects of the material in other portions of the coolant system.
Both the licensee and B&W are having chemical analyses made of the polyethylene to determine if any halogens are present.
When the inspector attempted to examine the installation, he found that the portable steel ladders had been removed and a rope ladder with wood rungs had been installed.
In climbing on this ladder, the inspector observed wood splinters f alling from the rungs. Other splinters were observed on the core barrel flange. Examination of the rungs indicated that numerous splinters possibly had been dislodged f rom the rungs and fallen into the internals. When the inspector pointed out that the use of this type ladder compromised the vessel cleanliness, the ladder was removed immediately. Because of the installation of the vessel internals, however, Hunnicutt tould not advise the inspector of a method that could be used to assure that the splinters could be located and removed from the reactor vessel. He stated that Region II would be kept appraised of the results of any studies or analyses that were made regarding the polyethylene or the splinters. The inspector stated that it appeared that the licensee may be in nonconformance with the FSAR in that the vessel had not been properly protected af ter cleaning and a CDN would possibly be issued relating to these items. The inspector will follow' up on these i. ems during future inspections.
G.
Preonerational Testing (Attachment M, PI 5800/2) 1.
General Review of Testing Program (PI 5805.01)
The preoperational testing program will be performed as described by a general guidance procedure, " Guide for Conducting the Oconee Initial Test Program," to assure that commitments in Section 13 of the FSAR will be met.
Recent revisions to the procedure were-reviewed and discussed with Smith and Hampton during the inspec-tion. The guide states that procedures will be reviewed prior to -
approval by various technical support groups such as the Duke Engineering Department and B&W..The master file contains the latest revision of each procedure with comments that have been received frop the various groups, but there was no method of
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CO.Rpt.'No. 50-269/71-3 :
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f insuring. that comments not yet received would be considered in i
the final approved procedure. Smith stated that he would consider l
withholding his signature until the Test Work Group has ' completed its final. review approximately one week prior to -conducting the L
test. The final review.will include checking that comments have been received and included.
I i.
The definition of changes that will be permitted without approval I
of.the Station. Review Committee and station superintendent had been revised and was considered unsatisfactory by the inspectors. The revision. states that modifications that do not change the intent j
of the original procedure would be made by the station test coordinators (shif t supervisors). The inspectors pointed out that such a definition i
could mean correcting a valve number or completely rewriting the procedure, depending upon the interpratstion of the individual making 1
j the change. After a lengthy discussion, Smith agreed to reword the-definition to more clearly state the type of changes to be allowed af ter a procedure has been approved for use.
Smith notified Region l
II by telephone on March 3, 1971, that the tsst coordinators would only have the authority to correct typographical errors or make such changes as correcting valve lineups.
l Documentation of the test program is contained in a master file maintained by the station superintendent's staff. A similar file 4
will be maintained in the Steam Production Department general office at Charlotte, North Carolina. The master file contains a i
correspondence folder, test specifications, test procedures, system
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description, and operating procedure pertaining to a particular test.
A working file is maintained for -use by the supervisors conducting the - test.
Any discrepancies observed during the performance of a ' test are recorded on a deficiency sheet attached to each : test procedure.
Correction of deficiencies is the responsibility of the station test coordinator assigneA L:: the test and the corrective action is recorded on the deficiency sheet when complete. Retests will
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be performed as necessary to verify the adequacy of the corrective action. Hampton stated that a system will be devised to give the status of outstanding ' discrepancies as well as the status'of i
l the overall testing program.
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~2.
Licensee's Test Results Evaluation Method (PIL 5805.02)
Evaluation of test results falls into two categories, (a) onsite
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evaluation of results against essentially go-no-go acceptance criteria-
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and ;(b) offsite evaluation of results.against more general criteria '
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C0 Rpt. No. 50-269/71-3 i i
not amenable to a complete go-no-go acceptance criteria. Examples of ' tests in the latter category are functional tests, operational tests, physics tests, power escalation tests and reactor building tests. Review by offsite groups such as Duke Engineering Depart-ment, B&W, or Bechtel, as required, will include analysis, con -
. clusions, and endorsement that the test is satisfactory or recom-mendations if the test is considered unsatisfactory.
Copies of j
.,such reports will be sent to the station superintendent for his consideration prior to approving test results. Test results are i
not considered final until approved by the station superintendent.
Safety-related tests and their results will be audited by the General Office Review Committee.
After a test is completed and the results approved, all material pertaining to that test will be filed for permanent documentation in the master file and the working file for the completed test will be discontinued. Each completed and approved test folder
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in the master file will contain the approved procedure, data collected, calculations, conclusions, final disposition of dis-crepancies, and approval of the completed test.
3.
Review, Witnessing, and Evaluation of Tests (PI 5805.03) a.
Procedure Review Comments on the following procedures were discussed with Smith, Hampton and McIntosh:
TP 1A 150 3
- Reactor Building Integrated Leak Rate Test TP 1A 150 6 2
- Reactor Building Isolation Penumatic Leak Test i
TP 1B.150 9 1
- Reference Vessel System Leak Test TP 1A 204 5
- Reactor Building Spray Pump Engineered Safeguards Test-TP 1B 210 5
- Chemical Addition and Sampling Functional Test TP 1A 230 t
- Soluble Poison Control Operational Test TP 1B 250 3
- Low Pressure Service Water Functional Test TP.lA 600 15
- Control Rod Drive System The inspector's comments on specific procedures are on file in Region II and will be included in the tests unless the licensee disagrees with the comment, in which case the differences will be discussed with the inspector. Due to the repetitive nature of the weaknesses noticed in many of lthe procedures, the
'S following general' items were discussed and will be considered in
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preparing and revising all of the Oconee test procedures:
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O-CO Rpt. No. 50-269/71-3.
(1) Provide for signof fs of items of significance and approvals.
(2). Notify required groups prior to start and af ter completion of tes ts.
(3) Give specific acceptance criteria where applicable.
(4) Reference procedures by which prerequisite conditions are es tablished.
(5) Provide concise, detailed and applicable limitations and precautions. Do not include prerequisites and procedural steps.
1 (6)
Include critical path chart prerequisites.
During the discussion of the containment integrated leak rate test, the inspectors pointed out that the test must be conducted to assure that Technical Specification requirements will be met even though the specifications are still in prep-aration and have not been accepted by DRL. The procedure did not include a controlled leak rate test to verify instru-ment sensitivity following the full pressure test, but only af ter the reduced pressure test. The inspectors considered that the control)ed leak rate test following each 24-hour integrated leak rate test was required by Section 15.4.3.1 of the proposed Technical Specifications. Smith did not believe that a controlled leak rate test was necessary at both pressures, but agreed to review the Technical Specifica-tion requirements before making his decision. Smith notified Region 11 by telephone on March 3,1971, that the procedure would be revised to include a controlled leak rate test following the full pressure test as well as the reduced pres-sure test.
b.
Flushing McIntosh stated that approximately 10% cf all flushing is comolete. Marked up flow diagrams are maintained to indicate the status of the flushing program and to assure that clean systems are not contaminatad by subsequent flushing operations.
i Test proceduree used during the reactor building spray system flush anc the low pressure injection system flush a
were reviewed and found acceptable. Both procedures were properly approved for use and discrepancies observed during the tests were recorded on the deficiency sheet attached to.
Q C0 Rpt. No. 50-269/71-3.
1 each procedure. Recorded test data met the specified acceptance criteria. All deficiencies had been cleared on the reactor building spray system flush and the test was approved by the station superintendent. All deficiencies had not been cleared on the low pressure injection system j
flush and the test had not been approved.
i
.Each procedure folder contained a test log which had.been 1
maintained by the test coordinator during the performance of the test. Test logs were reviewed by the inspector for the reactor building spray and low pressure injection i
systems and all significant events in these logs were 1
recorded on the deficiency sheets, j
Space is -provided on test cover sheets for the signatures of the test coordinator and vendor representatives witnessing,
a test, but all procedures were not signed by these persons.
Hampton stated that the signatures were not required "ntil a test is complete and the cover sheets in question involved tests that were not yet complete. Since the shift supervisor's
,r signature is specified on the cover sheet and some tests will span several different shif ts, Hampton was asked to consider having tect witnesses sign the cover sheet as the test is i
perfo rmed. Hampton stated that this was probably closer to the intent of the test program guide and the present system j
would be reviewed to assure that all responsible test witnesses are identified.
c.
Hydrostatic Testing Hydrostatic testing usually follows system flushing and is approximately 10% complete. Marked up flow diagrams are maintained to indicate the overall status of the hydrostatic
-test program.
Procedures were originally written to test an' entire sy: tem but completing the test was found to be impractical since piping installations were completed by area and not by system.
l A general procedure was devised to test portions of systems.
This procedure was reviewed and discussed with Smith, Hampton.
j and McIntosh.
The general procedure had been reviewed and approved by the Station Review Committee amd station superintendent, but the-actual procedure to be used in the field was not to be lnaviewed by the same group. Before 'the test is to -Ina conducted,,
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the test coordinator identifies the pressure boundary, makes t
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O C0 Rpt. No. 50-269/71-3 -
up valve checklists, and specifies test pressures and relief valve settings. The licensee was informed that this method of procedure preparation did not provide appropriate review and was contrary to Duke's procedure preparation guide which requires that procedures and major procedure modifications be resubmitted to the Station Review Committee and station superintendent for approval.
Smith and Hampton stated that the hydrostatic test procedure for a portion of a system would be reviewed and approved by the Station Review Committee and superintendent prior to the test.
In addition to the proposed procedure, the committee will be supplied a flow diagram showing portions of the system already tested and the portion to be tested to insure that at completion the entire system has been satisfactorily tested.
H.
Initial Core Loading Procedure Review (Attachment N-12, PI 5900)
The Initial Core Loading Procedure, OP 1502 4, was reviewed in accordance with PI 5900 and discussed with Smith, Hampton, and McIntosh. The licensee will rewrite this procedure and give con-sideration to the following items:
1.
Method of determining that specified limitations are being met on a continuing basis.
2.
Definition of " unexpected" as used in Step I.M.
3.
Longer counting times than those specified in Step II.B.4.
4.
Elimination of references to irradiated fuel.
5.
Access control to reactor building.
6.
Continuous recirculation of Lorated water.
7.
Periodic checklists for maintaining satisfactory status of all equipment and events important to safety.
8.
Final QC checks on core components.
9.
Use of plant nuclear instrumentation.
10.
Audible annunciation of source range monitors.
11.
Recording of flux monitor signals.
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12.
Checkout of fuel handling equipment.
13.
Specification of miniatua pool level.
14.
Status of mamjal containment isolation valves.
15.
Operability of emergency boron addition system and conditions for using.
16.
Water quality.
17.
Limit on number of fuel assemblies that may be in route between the fuel storage area and the reactor vessel.
18.
Critical path chart prerequisites.
19.
Haalth Physics Group participation and personnel monitoring requirements.
20.
Use of status boards and other appropriate records such as verifi-cation of proper enrichment, location, orientation and seating of components.
21.
Method used to normalize count rate af ter source or detector relocation.
22.
Job assignments giving consideration to the minimum permissible crew size and maximum allowable working hours.
23.
Criteria and method of initiating containment evacuation as indicated in FSAR, Section 7.4.3.
24.
Criteria for stopping fuel loading and authorization to continue.
25.
Involvement of groups discussed in FSAR, Section 13.1.1.
26.
Additional information on core makeup and detector locations during the loading process.
l Generation of adequate fuel handling procedures is on the outstanding items list.
1.
Control Rod Drive Mechanisms -' Attachment L
'l.
Records Review (4905.05)
The inspector reviewed the records relating to the installation of the control rod drives mechanisms to the vessel head. A M
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- 13) Rpt. No. 50-269/71-3,
detailed procedure had been prepared for each major step in the assembly of the mechanisms. The data sheets included spaces for signoffs for each assembly step and for recording the serial numbers associated with each mechanism. Revisions on the pro-cedures were documented and approved prior to use.
No deficiencies were noted in these procedures by the inspector.
The engineer responsible for the installation of the mechanisms maintains a daily log of events. This log indicated that the mechanisms which had been previously identified as having incor-rectly installed torque tubesl/ had been repaired. Other problems which had been encountered during the installation of the mechanisms were described and the resolution of the problems was documented.
Since the licensee did not maintain a separate list of the prob-lems and their resolution, it was necessary to review the log in detail to determine that a particular deficiency had been cor-rected. The inspector asked Hunnicutt if the licensee had con-sidered keeping a list of the deficiencies in order to minimize the possibility of f ailing to correct them. Hunnicutt stated that he recognized the advantage of maintaining a separate list and that he would follow up on this comment.
2.
Observation of Work (4905.06)
The licensee has established a clean area around the reactor vessel head and the assembly of the control rod drive mechanisms is being done in this area. When receiven the site, the com-ponents had been sealed in plastic bags packed in wooden crates.
They had been inspected for shipping damage during the inspection of the guide bearings and torque tubes.2/ During installation, the plastic bags were not removed until the components were placed in the clean area. At the time of this inspection, the pressure thimbles had been installed, the installation of the stator and position indication assemblies was essentially com-plete and the installation of the seismic restraints was in pro-gress. The inspector did not note any deficiencies in the assembly process and plans no further action on this item at this time.
J.
Electrical and Instrumentation 1.
Control Rod Drive Controls - Attachment H 1/C0 Report Nos. 50-269/70-8 and 50-269/70-11.
2/C0 Report Nos..50-269/70-3'and 50-269/70-11.
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00 Rpt. No. 50-269/71-3,
a.
Review Cable of QC System (5205.04)
The inspector reviewed the requirements for F.e installation of the cables and wireways for the contro' rod drive system.
The requirements for these items 'are l'.ntical to those for other electrical and instrumentation systems previously inspected.1/ Only the cable for the scram magnets are con-sidered by the licensee to be safety-related and requiring special routing and separation. The licensee's use of color coded cable simplifies the verification of separation of redundant circuits. Power cables are installed in a single layer in the tray with maintained separation and loading of these cables must conform to IPCEA requirements. Wireways for instrumentation cables are not permitted to be loaded above the side rails. No deficiencies were noted and the inspector plans no further action on this item at this time, b.
Followup Observation of Work (5205.06)
Essentially all of the cables for the control rod drive system have been installed in the auxiliary building and in the reactor building. The cables and trays for the control rod drives were installed in accordance with the licensee's procedures and the proper materials were used. A QC inspector is assigned to stay with each cable pulling crew to insure that the installa-tion is made in accordance with approved drawings and procedures.
During a tour of the installation, the inspector observed that a QC inspector was with each crew.
2.
Nuclear Instrumentation - Attachment H a.
Records Review and Observations of Work (5105.05 and.06)
The inspector reviewed the nuclear instrumentation records with Price. These records had been inspected previously and found to be deficient.2/ The licensee has now completed a review of the documentation and corrected the deficiencies which had been found. The nuclear instrumentation chassis are mounted on the main control console. A review of this installation indicated that the equipment had been installed in accordance with approved drawings. The inspector plano no further action on these items at this time.
jl/C0 Report No.- 50-269/70-10, 2/C0 Report No. 50-269/70-12.
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b.
Cables (5205.04 and.06)
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1 The QC and installation requirements for the nuclear instru-mentation cable and wireway installation are the same as for j
the control rod drives as discussed in Section J.1.
The i
inspector did not observe any deficiencies in the QC require-ments or the installation and plans no further action on these items at this time.
j 3.
Pressurizer Level Instrumentation - Attachment H Review of Cable QC System (5205.05) l The requirements for the cable and wireway installation are the same as for the control rod drive controls, Section J.l.
The inspector did not observe any deficiencies in these requirements i
and plans no further action on this item at this time.
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Uninterrupted QC Pouer System - Attachment I I
i a.
Review of Cable QC System (5205.04)
The requirements for the cable and wireway inatallations are the same as for the control rod drive controls, Section J.1.
i The inspector did not observe any deficiencies in these require-ments and plans no further action on this item at this time.
1 b.
Followup Observation of Work (5205.06)
The installation of the cables and wireways associated with j
the uninterrunted a.c. power system was reviewed by the inspec :
tor and appe. red to be in accordance with the QC requirements.
The anspector plans no further acticn on this item at this-time.
I -
5.
Battery System - Attachment I
- a. - Review of Cable QC System (5205.04)
. The requirements for the cable and wireway installation for-the station control batteries are the same as for the control rod drive control, Section J.l.
_The inspector did not' observe-any deficiencies in these requirements and plans no further action on these items 'at this time..
4 I.
i i*
. a.m
, ~
b'
C 'T CO-Rpt. No. 50-269/71-3.
b.
Followup Record Review (5205.05.a.1 and a.2)
The records relating to the NDT requirements for cables had been previously reviewed and no deficiencies had been noted.1/
The licensee has provisions for the quarantine of nonconforming cable but, to date, no cable has been received at the site, i
c.
Followup Observations of Work (5205.06)
The inspector reviewed the battery cable and wireway installa-tion. No deficiencies were noted and the inspector plans no further action on this item at this time.
K.
Miscellaneous 1.
Valve Numbering Sys tem The inspector advised Smith and Hampton that during a review of the test procedures, he had noticed that the valve numbers in the procedures did not conform to those in the FSAR.
Smith stated that the numbers in the test procedures and the operating manual were assigned by the Operating Department and the numbers on the valve identification tags would conform to these. Numbers in the FSAR had been assigned by either B&W (primarily nuclear steam supply system) or the Duke Engineering Department (primarily secondary systems). The valves on the system design drawings had originally been identified by B&W or by Duke Engineering or by both. Subsequently, the drawings had been revised to also include-the Operating Department numbers. The numbering system used by the Operating Department is similar to that used by B&W in that both use a system designation followed by a component number.
The high pressure injection system valves in both cases are designated 'MP."
The inspector pointed out that in having multiple designations on the drawings, the possibilities of operator errors were increased. In addition, in communicating with other organizations, the multiple numbers could lead to confusion and errors.
Smith stated that he would determine what could be done to minimize or eliminate the type occurrences pointed out by the inspector.
2.
Erl Handling Ecuinment In reviewing the installation of the fuel handling craros, the inspectors observed that the drip pans under some of the bearing 1/C0 Report No. 50-269/70-10.
f
,m CO Rpt. No. 50-269/71-3.
boxes appeared to be too small to hold the lubricant that might leak into them. One of the drip pans appeared to be tilted down-ward at.one' corner, further reducing its' capacity. Because of the positions of the cranes over the fuel canals, it was not possible to make a complete inspection nor an accurate evaluation of the adequacy of-the pans. This item was discussed in the management interview. Hunnicutt stated that he would review the designs to determine if the pans 'were adequately sized and at the necessary locations. Smith stated that he would verify that procedures were developed that would require periodic' inspection and maintenance of the pans.
3.
Fuel Transfer Tubes a
Hunnicutt advised the inspector that the expansion bellows on the l
fuel transfer tube which had been previously reported as damaged had been replaced with a new bellows. The inspector plans no further action on this item at this time.
Attachments:
Exhibits A and B 3
i i
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LICE!SEE Duke ?cver Cc=pany n
I FACILITY Ccenee Statien no. 1 DOCKET 2: LICFl!SE NCS, 50-269, C??R-33 REACTOR OUTSTANDING ITD'S IDE!TIFIED 1 ITEM CLOSED I ;.. sc-2, 3/5/68, Cenerete test cylinder breaks belov specs
.68-3, D.5.,
E 6/19/68 2.
68-3, 6/19/68, Unauthorized revision to Cadveld specifications 68-k, Su==ary, E
l 9/25/69
,3 68-3, 6/19/68,
! Failure to provide concrete inspector 68 h, Su =ary, E
I 9/25/69 L.
68-L, 9/25/68 Failure to properly test Cadweld splices 69-1, S"- vy, E
1/6/69 t
' 5.
69-8, 9/9/69, l Failure to properly qualify veld precedures 69-9, G, 11/3/69 KC
- 6.
69-8, 9/9/69, Failure to properly qualify velders 69-9, G, 11/3/69 v..
- 2-
- 7.
IZ3, h/11/69 Procedure for repair of arc strikes not available 70-5, Su==ary, t /1-'/7;
' 3.
CC:I, 1/8/70
!DT of core flooding valves Meno, WCS to HQ, 2/2/70 I
9 70-1, 1/6/70, Welding and NDT deficiencies, CD:t issued Memo, WCS to HQ, E
3/26/70^
p.0.
Sinchc= 69-1, Main coolant punp discrepancies Meno, WCS to HQ, 12/9/69, NC h/21/70 2
.11.
TC L, h/27/70, IIcv strength concrete Memo, WCS to HQ, E
8/7/70 2.
II3, 5/1/70 Pressure vessel safe ends Meno, WCS to HQ, 8/5/70 (13 70-6, 5/25/70, i Tendon stressing discrepancies Meno, WCS to HQ, i
8/7/70
!S..
70-8, 8/3/70, j Tendons and stress gages Mero, WCS to HQ,
E I
10/8/70
-15 70-8, 9/1/70,-
Fissures in pri: vy coolant pipe cladding FSAR, Amend.24, C':
.t 12/17/70
~ 6.
IZ3, 9/11/70,
, c.
Determination of safety system response to axial 1
I E
power imbalances
',c.
Availability of in-ccre detectors 7ar Ili..T; /IE.7 Cr, amn : 3 carety item; NC - nonccarliance or nonconformance tem: UU - unrocolvec item: IN - inqu Hy item; IEE - Reactor Inspection and ErGrecconc 'Eranch reque3T; ~O - other source o.'
ide:;&Mication A (bri c01;/ specb'y')
1 of 61
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..>o LICE:SEE Duke Power Cc=;any FACILITY Cconee Station I!o.1 j
i DOCKET & LICF"ISE NOS. 50-269, CPPE-33 i
REACTOR OUTSTANDING ITE'S,
IDE:TIFIED !
ITEM I
CLOSED l c.
Mer.sure=ents of flow and temperature during j
initial operation d.
Verification of 'oypass flov I
j e.
Verificatien of axial peak effects on DE R v
t i
f f.
Data during startup for single loop, two pump operations 3 c.
Inspection of reactor internals after completien l
of preoperational tests f h.
Field test of steam generator i
t 5
{ :..
Low strength concrete and omitted tendons Memo, WCS to HQ, i
10/8/70 l J.
?enetration room valves 70-12, Sim ary 1
12/1/70 I
i k.
Strain gauge failures Meno, WCS to HQ, i
10/8/70 r
H? and L? injection systen startup times 1.
i n.
Core flooding tank XO valve
{ n.
Reactor building sprsy pu=p performance i
t c.
Condenser cooling water crossover header valve 3
t p.
Spent fuel accident filters 1
1 j
q.
Administrative control of MCP startup I
i
- r.
Flov testa per 200/12 and 200/13 I
t.
i c.
Flow distribution chart
.i For IDEl.TI!'IED Oo?.u=n: S - sarety item; NC - noncompliance.or nonconformance ite=;-UI! - unresolveif item; IN - inquTry item; IEB - Reactor Inspection and Fr.lTrcement Branch request; O
. other source of identification (briefly specify)
^
2 or_ 6 eact A
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1.
LICENSEE tuke Pcwer Company j
g FACILITY Ocenee Station No. 1 4
i DOCKET & LICENSE NOS. 50-269, CP??.-33 5
REACTOR OUTSTANDIMG ITE':3 i
ITry.
l CLOSED i
J.Dr.e. z _.r Trn m
~
, 17 70-2, 2/19/70, j '/endor NDT records for safeguards syste=s cables 70-11, F, 10/26/70 l
C:
i 13.
70 L, 3/2?/70, i Varification of separation of transducer tubing EI l
i 19 70-8, 8/3/70, Control rod drive gdide bushings and torque tubes 71-3I, 2/24/71 a
ti 20.
70-8, 8/3/70, Cc=pletion of EP facilities m:
21.
70-8, S/3/70,.gcnpletion of HP procedures 2
3.:i j 22.
70-8, S/3/70, Cc:pletion of EP personnel training 70-12, Summary E
12/1/70 l23 70-5, 8/3/70, Crahe load test 71-1, 1/4/71 C;
, 2h.
70-6, 6/3/70, VariP/ that test procedures are properly revised and
{
E i cpproved when -changes are required
'25.
70-3, 8/3/70, l' 2ri?/ that analysis of centainment is made FSAR, Arend. 24 C:
I i 26.
70-8, S/3/70, f Adequ' ate fuel handling procedures i,
CI
- P-dn steam pipe hangers i 27 70-3,8/3/70,
+
-C!
l 2S.
70-9,9/1/70,
, steam generator skirt e.dapter indicatiens 1.::
j 29 70-9,9/:./7C,
'2? injection pu=p QC records 70-11, C,
}
g 10/26/70
- 3 40-9,9/1/70,
,2aais for particle size in flusiing procedures 70-11, G, C:
(
10/26/70 t
31.
70-9, 9/1/70,
!?rotection of instrumentation during hydro test CI
, 32.
70-10,9/26/70, Fuel transfer tube expansion joint replacement 71-3L, 2/24/71 E!
1
~~~
i 33 TC-10, 9/28/70, p.:uting of cables extehior to cable trays Memo, WCS to HQ E
1/1S/71 For IDEUTIJIID Colucn: S - safety ite:n; NC - noncompliance or nonconformance item; UU - unresolveE itc=; IN - inquify iter.; IEE - Reactor Insnection and En?crecment Branch requeliT; O - other source of identificati^on g
.('criefly specify) 3 of 6 ].
=
LICEIISEE Du.e P0ver Company
/3 FACILITY Ocenee Station No. 1 DOCKET E LICEU3E UOS.
50-269,.CPPR-33 REACTOR OUTSTA:TDING ITEMS IDE:""!FIED 1 ITEM l
CLOSED
- 3L.
ERL Ept. Iso.1, i:nstallation of c.dditional environ = ental monitoring 7/2h/70,L77 is cinzent DEL R-t. No.1, !' lent valve replacement test 35 i
7/2L/70, E
- 36. ERL Ept. No.'*l, Ctrong motion acceleremeter installation 7/2L/70, E 37 ERL Rpt. Ito 1, 2enetration rocm flow indication and adjustment 7/2h/70, d
- 33. DRL Ept. I:o.1, Instrumentation bypass keys Tech Specs Change
.7/2h/70, E l
12/70 39 ERL Ept :;o. 3,1* Internals vibration test I
9/15/70, E LO.
DRL Rpt. No. 3, Core flooding tank
- valves 9/15/70, E kl'.
70-10, 9/28/70, F:ydrostatic test pressures 71_1, 1/4/71 UN I
l L2.
70-11, 10/26/70 71 caning reactor coolant system piping and equipment 71-2, 1/25/71 CN i
L3 70-11, 10/26/70,I3ensitized stainless steel in reactor coolant pu=p E
{dischargepiping 71-1, 1/4/71 44.
IE3, 12/22/70
.eactor coolant pump tests
' afety inj ection syste= testin;;
45.
IE3, 10/30/70 3
i t
46.
70-12,12/1/70 ' ; Vibration testing - equipnent and piping 1
47.
70-12, 12/1/70
,.ocation of station batteries NC t,
48.
70-12,12/1/70 f Nucicar instru=entation vendor tests
-NC' 49.
70-12, 12/1/70
,21ectrical QC data packages NC
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go r.,:,._:,...-.<_c.a u,
-umn:
3, - carety itom; iC - noncompliance nr nonconforman::e item: ?!" - unreco_vec iter: IN - inquTry item: IF3 Reactor Ins.n.cetion and._.crec cr.
Lranch request; O - other source oJ. identil.ication g
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m LICE!SF.E Duke Power company FACILITY Oconce Station No. 1
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DOCE E: LICFl!3E : 0S. 50-26', C??R-33 9
REACTOR OUTSTANDING ITEMS
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IDENTIFIED I ITEM CLOSED 50.
70-12, 12/1/70 ITE relays UN 51.
70-12, 12/1/70 IIcater and heat tracing tests UN 52.
70-12, 12/1/70 control rod drive cooling system tests UN 53, 70-12, 12/1/70, Containment am. auxiliary building vent system filters I
- 54..FSAR, Acend 25 Installation of strain gages M 12/30/70 55.
71-2, 1/25/71 Keowee battery room ventilation UN I
56.
71-2, 1/25/71 I Switchyard battery blocking diode tests
}
UN 57.
71-2, 1/25/71 ! Re=ove temporary steam line at. 4 kv switchgear
_UN SS.
71-2, 1/25/71 ' Controlled leak rate tests E
i 59, 71-2, 125/71 l ?enetration room elevator opening UN l
t 60, 71-2, 1/25/Y1
.! Verificatien of separation of redundant circuits
.s.
3 i
- Cleanup of cabh i:renches 61.
71-2, 1/25/71 l
UN l
62.
71-2, 1/25/71 ! Adequacy of leak rate tests
_UN 63.
71-2, 1/25/71 Replacement of feedwater pipe i
64.
72.-3, 2/24/71 l Cleanliness of reactor vessal and internals NC 1
For IDE.;TI/IED Cr.;umn: 3 - care;y item; NC - noncompliance or nonconformance tem: UU - unrenoIvoif item. IN - inquiry item; Is3 - Reactor Insnection anu -.b?Ertement I, ranch requeTC; ~O - other source of identificat,i'on
(~cricf1:/ spec !.fy)
Exhibit A Page 5 of 5 m.
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.,cc Power Co=vany
- 37v,
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- w; FACILITY Oconee Station :0. 1 n
i DOCKET i': LICIC3E :iOS.
50-269. CPPR-33 RvlCTOR CUTSTA?iDING ITE:S t
4 t
xe. r e.
l CI,0a""zI>'
- p..m...
T :..+a l
l Drip pans on fuel handling cranes l 65.
71-3, 2/24/71 U::
66.
71-3, 2/24/71 l Containment leak rate tests m:
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For ;DE::TicIED Cc.;umn: S - carety 1,em; :iC - nonconcliance or nonconformance iter 4 :., '. '
unreco'.vec iter: IN - inqu Hy itom; IIB - Reactor Insr.cction and.T.n F rccrenc Eranen request.; O - other coureli of 1.dentificati~n c
(brl er1;r cpec 1.'y)
I:xhibit A j
Page 6 of 6 n - - -
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O EOLT CIRCLE 3ET!!EE MATING FLANGES OF THE CO*lE SHIELD &
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CORE BARREL SH0i!ING LOCATION OF TRAPPED POLYETHYLENE I
30I.7 NIEGERING SYSTDI a
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Exhibit 3
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