ML20115A535
| ML20115A535 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/06/1992 |
| From: | Skolds J SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | George Wunder NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20115A538 | List: |
| References | |
| NUDOCS 9210140359 | |
| Download: ML20115A535 (2) | |
Text
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e' South Ccrohna Eltetric & G:s Company John el s Jenanmue Sc P9065 touclew Operabons
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(933) 345 4040 SCE&G October 6, 1992 Document Control Desk U. S. Nuclear Regulatory Commission Washington DC 20555 Attention:
Mr. G. F. Wunder Gentlemen:
Subject:
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)
DOCKET N0. 50/395 OPERATING LICENSE NO NPF-12 TECHNICAL SPECIFICATION CHANGE REQUEST - TSP 920002 INCREASE IN STEAM GENERATOR TUBE PLUGGING FROM 15% TO 18%
in accordance with 10CFR50.90. South Carolina Electric & Gas Company (SCE&G) is submitting an amendment request to License NPF-12 to amend the VCSNS Technical Specifications.
The proposed change is a revision to Table 2.2-1
" Reactor Trip System Instrumentation Trip Setpoints " to allow an increase in the maximum permissible average level of Steam Generator Tube Piugging (SGTP) from 15% to 18%. An increase in SGTP reduces Reactor Coolant System Minimum Measured flow (MMF) and, therefore, requires changes to a constant and a setpoint reduction penalty in the Overtemperature Delta T (OTAT) setpoint equation, the OTAT trip total allowance and Z, and the loop design flow listed in Table 2.2-1.
The change request is contained in the following attachments:
Att whment 1 List of Affected Pages and Marked up Technical Specifications A Description of the Amendment Request and supporting Safety Evaluation A Description of the Amendment Request and associated No Significant Hazards Determination As discussed with the NRC in the August 12, 1992 Steam Generator Replacement Project status meeting. SCE&G requires approval of this Technical Specification change by the start of the seventh refueling outage (scheduled for March 5. 1993). SCE&G requests a sixty day implementation period for this Technical Specification e.hange to allow for necessary procedure revisions.
This proposed Technical Specification change has been reviewed and approved by the Plant Safety Review Committee and the Nuclear Safety Review Committee.
- %Ot;OO 9210140359 921006 DR ADOCK 05000395 400,i PDR II ;
Document Control Desk 1SP 920002 Page 2 of 2 1 declare that the statements and matters set forth herein are true and correct to the best of my knowledge, information, and belief.
Should you have any questions concerning this issue, please call Ms. April R.
Rice at (803) 345-4232 at your convenience.
Very truly yours, k
(
John L. Skolds ARR:lcd Attachments c:
0.W.Dixon(w/oAttachments)
R. R. Mahan (w/o Attachments)
R. J. White S. D. Ebneter G. F. Wunder General Managers NRC Resident inspector J. B. Knotts Jr.
H. G. Shealy L. R. Cartin RTS TSP 920002)
File 813.20)
NUCLEAR EXCELLENCE - A SUMMER 1RADITION!
Attachments to Document Control Desk Letter Technical Specification Change Request - TSP 920002 Increase 'n Steam Generator Tube Plugging from 15% to 18%
ABBREVIATIONS
-i ASME American Society of Mechanical Engineers BIT Boron Injection Tank COLR Core Operating Limits Report CRDM Control Rod Drive Mechanism DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio ECCS Emergency Core Cooling System E0P Emergency Operating Procedure ECT Eddy Current Testing FAH Hot Channel Enthalpy Rise Factor F0 Total Peaking Factor FSAR Final Safety Analysis Report GPM Gallons per Minute LBB Leak Before Break LBLOCA Large Dreak Loss of Coolant Accident LOCA-Loss of Coolant Accident LOL/TT Loss of Load / Turbine Trip M/E Mass and Energy MMF Minimum Measured Flow MSSV Main Steam Safety Valve NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System 0 PAT Overpower Delta T OTAT Overtemperature Delta T PCT Peak Clad Temperature PLOF Partial Loss of Flow RC Reactor Coolant RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RSR Relative Stability Ratio RTP Rated Thermal Power RWST Refueling Water Storage Tank SBLOCA-Small Break-Loss of Coolant Accident SCE&G South Carolina Electric & Gas Company i
SIS Safety Injection System S/G Steam Generator SG1P Steam Generator Tube Plugging SGTR Steam Generator Tube Rupture TA Total Allowante TAVG RCS Average Temperature-T 0T Vessel Outlet Temperature H
TCOLD Vessel Inlet Temperature T0F Thermal Design Flow VCSNS Virgil C Summer Nuclear Station l_
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