IR 05000327/1978042
| ML19259B384 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/29/1978 |
| From: | Bryant J, Vallish E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19259B383 | List: |
| References | |
| 50-327-78-42, 50-328-78-28, NUDOCS 7902090116 | |
| Download: ML19259B384 (5) | |
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Report No.:
50-327/78-42 and 50-328/78-28 Docket No..
50-327 and 50-328 License No.: CPPR-72 and CPPR-73 Category:
A3 and A2 Licensee: Tennessee Valley Authority 830 Power Building Chattanooga, Tennessee 37401 Facility Name:
Sequoyah Nuclear Plant, Units 1 and 2 Inspection at:
Daisy, Tennessee Inspection conducted:
December 11-14, 1978 Inspector:
E. J. Vallish Reviewed by:
.h.
g, u m 12 /gh J. C. Bryant\\ Chief d Date Engineering Support Section No. 1 Reactor Construction and Engineering Support Branch Inspection Summary Inspection on December 11-14, 1978 (Report Nos. 50-327/78-42 and 50-328/78-28)
Areas Inspected:
Units 1 and 2, progress of installation of safety-related mechanical components, reactor vessel internals' quality records, safety-related pipe support and restraint systecs,' licensee identified item concerning reactor vessel support and nozzle loads. This inspec-tion involved 24 inspector-hours onsite by one NRC inspector.
Results: Of the four arers inspected, no apparent items of noncompli-ante or deviations were identified.
790209006
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RII Rpt. Nos. 50-327/78-42 and 50-328/78-28 I-1 DETAILS I Prepared by:
M (n b
(2[79f7g E.J.Vallish),MechangcalEngineer Date Engineering Support Section No. 1 Reactor Construction and Engineering Support Branch Dates of Inspection: December 11-14, 1978 Reviewed by h
\\2 29 7[
J. C. Bryant, Chkef j
ate Engineering Support Section No. 1 Reactor Construction and Engineering Support Branch 1.
Persons Contacted a.
Tennessee Valley Authority (TVA)
- G.
G. Stack, Project Manager, Construction
- J. E. Wilkins, Construction Engineer H. D. Lollar, Mechanical Inspection Unit Supervisor J. A. Thompson, Mechanical Engineer D. Williams, Mechanical QC and Inspection H. Fisher, Mechanical Engineering Unit Supervisor E. C. Pendergrass, QC Record Unit R. D. Vickery, Mechanical QC Inspector D. Hafley, Power Production Engineer
- Denotes attendance at the exit interview.
2.
Licensee Actions on Previous Inspection Findings Licensee actions on previous inspection findings were not reviewed during this inspection.
3.
Unresolved Items No unresolved items were identified during this inspection.
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4.
Independent Inspection Effort - Units 1 and 2 Inspection was made of the status of installation of the safety-related mechanical components in Units I and 2, including the reactor control rod drive mechanisms, steam generators, pressurizers, reactor coolant pumps, control rod drive coolers, RHR pumps and the RHR and containment spray heat exchanger.
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RII Rpt. Nos. 50-327/78-42 and 50-328/78-28 I-2 No items of noncompliance were identified.
5.
Reactor Vessel Internals - Quality Record Review - Units 1 and 2 These internals were inspected in their stored position in the refueling canal. They are supported by the peraanently installed supports for these units and are covered with a heavy plastic sheeting. The licensee stated that field assembly of the internals are complete except for QC inspection of the Unit 2 guide tube welds recently completed. The licensee also stated that Westinghouse will certify compliance with the procurement requirements for the internals and deliver a Quality Control Release form to that effect for each unit's intervals after hot functional testing of that unit. Handling during assembly and initial installations were performed in accordance with Westinghouse Procedure No. 2463A64 entitled, " Reactor Internals Assembly". Handling was performed by Westinghouse with the licensee's engineers performing surveillance of these operations.
Monthly storage inspections are performed by the licensee and the records of these inspections from June 1977 through December 12, 1978 were reviewed for both Units 1 and 2.
These records confirmed that adequate control of cleanliness was maintained and the inspec-tion of the internals reported earlier in this paragraph verified the recorded conditions.
The licensee's representative stated that the next handling of the Unit 1 internals will be done by Power Production personnel using their permanent operating procedures.
No items of noncompliance were identified.
6.
Safety-Related Pipe Support and Restraint Systems - Units 1 and 2 The following approved work procedures were reviewed to determine licensee commitments and installation requirements for these support systems and to determine if these installation procedures will assure the uncompromised functioning of the dynamic snubbers after installation:
a.
Installation Instruction (II)-36 RS, " Orientation and Alignment" b.
II-38 R6, " Inspection of Site Fabricated Assemblies" II-66 R6, " Inspection of Supports" c.
d.
II-Al R1, " Steam Generator Hydraulic Snubber Inspection"
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RII Rpt. Nos. 50-327/78-42 and 50-328/78-28 I-3 e.
II-A3, " Inspection and Cycling of Shock Suppressors" The snubber overhaul and calibration shop located near the Unit I hot machine shop area was inspected.
employees supervise the shop activities, with craft support from construction forces.
This function has just started removing, re-building and re-calibrating the Unit I hydraulic snubbers and reinstalling them. The licensee's representative stated that the following DPP procedures prescribe the snubber activity the shop performs, and that these procedures are being dev~ eloped and processed for TVA approval.
a.
MI 6.13 A, " Removal and Reinstallation of Snubbers" b.
MI 6.13 B, " Functional Testing" c.
MI 6.13 C, " Repair and Rebuilding of Hydraulics" d.
MI 6.13 D, " Rebuilding Mechanical Snubbers" The mechanical snubbers will also be renovated, the licensee stated, based upon the manufacturer's recommended instructions, which are on order. Overhauled and reinstalled snubbers will be tagged, with the pertinent information thereon.
No items of noncompliance were identified.
7.
Licensee Identified Item - Units 1 and 2 (Closed) Reactor Vessel Support and Nozzle Loads - NCR-65D -
10 CFR 50.55(e) Item The final report concerning this item, dated November 24, 1978, was reviewed prior to this inspection. The two Westinghouse topical reports WCAP 9259, "ASME Section III, Class 1, Stress Evaluation for Sequoyah Nos. I and II, RPV Inlet Nozzle", and WCAP 9260, "ASME Section III, Class 1, Stress Evaluation for Sequoyah Nos. I and II, RPV Outlet Nozzle", were reviewed at the site for applicability.
Other documentation reviewed indicated that those topical reports were reviewed by the licensee's Civil Engineering Branch and found acceptable as reported and that no corrections to equipment or hardware would be necessary.
This item is close.
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RII Rpt. Nos. 50-327/78-42 and 50-328/78-28 I-4 8.
Exit Interview A meeting was held with the licensee's representatives identified in paragraph 1 of this report and others to discuss the results of this inspection.
Items discussed included the Units' I and 2 general progress of installation of safety-related mechanical components, reactor internals, and safety-related pipe support and restraint systems. For snubbers, an item of concern was discussed about a positive method of assuring the secureness of mechanical fasteners on these restraint systems.
The licensee was informed that TVA NCR 65 D, entitled " Reactor Vessel Support and Nozzle Loads", a 10 CFR 50.55(e) item, was closed during this inspection, and that no noncompliances or new unresolved items were identified.
The licensee acknowledged these findings.