RC-11-0178, Submittal of Twenty-Second Report to 10 CFR 50.59(d)(2) Changes
| ML11311A214 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 11/04/2011 |
| From: | Gatlin T South Carolina Electric & Gas Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LTD 342, RC-11-0178, RR-8450 | |
| Download: ML11311A214 (4) | |
Text
Thomas D. Gatlin Vice President, Nuclear Operations (803) 345-4342 November 4, 2011 A SCANA COMPANY Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir / Madam:
Subject:
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 10 CFR 50.59 BIENNIAL REPORT South Carolina Electric & Gas Company (SCE&G) hereby submits the Twenty-Second VCSNS Report pursuant to 10 CFR 50.59(d)(2).
This report contains a brief description and summary of the evaluations performed to support the changes and modifications made to the facility in accordance with 10 CFR 50.59(c) (Attachment I). This report covers the time frame from October 1, 2009 to September 30, 2011.
If you have any questions or require additional information, please contact Bruce Thompson at (803) 931-5042.
Very truly yours, Thomas D. Gatlin GAR/TDG/jw -10 CFR 50.59 Summary of Evaluations and Changes c:
K. B. Marsh B. C. Gleaves S. A. Byrne NRC Resident Inspector J. B. Archie K. M. Sutton N. S. Carns NSRC J. H. Hamilton RTS (LTD-324)
R. J. White File (818.02-8, RR 8450)
W. M. Cherry PRSF (RC-11-0178)
V. M. McCree Virgil C. Summer Station
- Post Office Box 88 -Jenkinsville, SC
- 29065 -T (803) 345-5209
ý r- '
Document Control Desk Attachment I LTD 342, RR-8450 RC-11-0178 Page 1 of 3 10 CFR 50.59 Summary of Evaluations and Changes Parent Change Evaluation Document Description Summary FSAR Update for The inadvertent fuel loading (fuel The evaluation methodology Inadvertent Fuel misload) event described in FSAR (evaluation of core power Loading Event -
Section 15.3.3 is being revised to distribution measurements) of Condition Report reflect additional Westinghouse the safety analysis with respect CR-09-05705 research on the detectability of to inadvertent loading has not various severity deviations from the been changed. It is consistent analysis of record loading pattern, with the original analysis but and to summarize the effect of this more detailed. The computer on accident analyses and the codes used to process the flux associated reload evaluation. This map measurement data have change also describes some more been changed. The codes have specific detection criteria for the been reviewed and approved by purpose of identifying potential the NRC and are direct inadvertent loading, replacements for the codes listed in the FSAR. No new accidents, failure modes or malfunctions are created by this change.
MRS-SSP-2651-CGE-This procedure describes the The process of loading ten (10)
LTA, Revision 0, V. C.
process steps required to transfer specified spent fuel rods into a Summer Cask Loading ten (10) irradiated fuel rods from a NAC International LWT cask per Procedure fuel rod storage basket to a canister the direction given in this and load the canister into a cask for procedure complies with NRC shipment.
Certificate of Compliance No.
9225, the Safety Analysis Report for the NAC LWT, 10 CFR Part 71, NUREG 0612 &
ANSI 14.6 - 1979. The travel path of the cask is restricted to the safe load path controlled by a Complex Lift Plan developed in accordance with General Maintenance Procedure GMP-100.026, "Crane Operations &
Rigging." No new accidents, failure modes or malfunctions are created by use of this procedure.
Document Control Desk Attachment I LTD 342, RR-8450 RC-11-0178 Page 2 of 3 10 CFR 50.59 Summary of Evaluations and Changes Engineering Change Request ECR-50592E/H, Electro/Hydraulic Controls (EHC)
Upgrade These ECR revisions implement portions of the EHC Digital Controls Upgrade. The specific activities in these revisions install multiple core bores in the concrete wall between the turbine building and control building while the plant is in operation. The core bores provide conduit paths for the installation of control, signal and power cables from the new control system cabinets in the turbine building to the main control room and interfacing systems in the control building.
An analysis was performed to evaluate the consequences of core bores being open coincident with a turbine building steamline break (SLB). The analysis determined that there would be negligible changes in the control building area temperatures, relative humidity and pressure for a short period of time while the SLB is in process and the core bore is open. The evaluation determined that there would be no new accidents and no effect on accidents and malfunctions previously evaluated. The evaluation also determined that there would be no increase in consequences of previously evaluated accidents and no impact on methodologies described in the FSAR.
Engineering Change Request ECR-50704B, Reactor Vessel Upflow Conversion This change reversed the downflow path in the reactor pressure vessel to an upflow path and consisted of plugging 20 existing flow holes in the core barrel and machining 20 new holes in the top former plate to redirect flow in the baffle-barrel region of the lower internals.
An analysis of this change determined that the LOCA pressure and temperature analysis is bounded by the existing peak LOCA conditions.
The evaluation determined that although there is a minimal increase in containment transient temperatures and pressures, there is no increase in the frequency of any accident and no malfunction of any structures, systems or components. The probability of an accident or malfunction is not increased. There is no effect on any fission product barrier and no change to an FSAR method of evaluation.
Document Control Desk Attachment I LTD 342, RR-8450 RC-11-0178 Page 3 of 3 10 CFR 50.59 Summary of Evaluations and Changes RTN Insert Flexure Tool Operating Procedure-STD-OP-1992-6105, Rev. 6, Change 4 This activity evaluates a revision to a vendor procedure used to remove one or more flexures due to damage that might occur during fuel assembly top nozzle removal and replacement. This procedure also allows insertion of a lock tube for combinations of two or more flexures being removed.
The evaluation determined that for the 17x1 7 fuel assembly design with a total of 24 inserts, a minimum of 20 active inserts meet the shipping and handling criteria. Flexure removal in accordance with this procedure will allow a fuel assembly to be reused in the reactor core. No new accidents, failure modes or malfunctions are created by this change.