RC-16-0094, (VCSNS) Unit 1 - 2015 ECCS Evaluation Model Revisions Report
RIS 2007-24, NRC Staff Position on Use of the Westinghouse Crossflow Ultrasonic Flow Meter for Power Uprate or Power Recovery
RC-15-0100, Submittal of 2014 Emergency Core Cooling System (ECCS) Evaluation Model Revisions Annual Report
GNRO-2014/00082, Supplemental Reload Licensing Report (SRLR) Grand Gulf Cycle 20 Extended Power Uprate (Epu)/Maximum Extended Load Line Limit Analysis Plus (Mellla+) SRLR
RC-14-0076, Emergency Core Cooling System (ECCS) Evaluation Model Revisions for 2013
NL-15-183, Watts Bar Nuclear Plant, Unit 2, Amendment 114 to Final Safety Analysis Report, Section 6.0 - Engineered Safety Features, Pages 6.3-1 - 6.9-2
W3F1-2018-0037, Closure of Commitment Associated with Control Element Assembly Drop Time License Amendment
NRC-10-0003, Proposed License Amendment to Revise the Core Spray Flow Requirement of Technical Specification Surveillance Tests SR 3.5.1.8 and SR 3.5.2.6
RC-09-0050, Submittal of 2008 Emergency Core Cooling System (ECCS) Evaluation Model Revisions Report
NRC 2008-0025, Annual Report of Changes to, and Errors Discovered in Emergency Core Cooling System Evaluation Models
NRC 2007-0017, ECCS Evaluation Model Changes
RC-07-0081, ECCS Evaluation Model Revisions Annual Report
RC-06-0205, ECCS Evaluation Model Revisions Report
RC-06-0202, V. C. Summer - ECCS Evaluation Model Revisions Report, Addresses the Effect of Using a Finer Break Spectrum
RC-05-0097, Core Operating Limits Report - Reference for Best Estimate Loss of Coolant Accident
ET 12-0023, 10 CFR 50.46 Thirty Day Report of ECCS Model Changes
ET 07-0021, CFR 50.46 Annual Report of ECCS Model Changes
RC-03-0099, ECCS Evaluation Model Revisions Annual Report
RC-02-0086, ECCS Evaluation Model Revisions Annual Report
ST-HL-AE-2159, Submits Info Re Postulated LOCA in Modes 3 & 4,per 870505 Request.Basis for Assumption That 10 Minutes Available for Operator Action Following 8 Inch or Smaller Break in Modes 3 & 4 Provided to NRC in Westinghouse Ltr NS-NRC-86-3144
ST-HL-AE-2060, Forwards Revised FSAR Tables 6.3-1, ECCS Component Parameters & 6.3-9, Normal Operating Status of ECCS Components for Core Cooling, for Consistency W/Vendor Design Documents & Other FSAR Sections
ST-HL-AE-1909, Forwards Addendum to long-term Cooling Rept Submitted by ,Resolving SER Open Item 16 & Confirmatory Item 13 Re Use of Class 1E Head Vent Sys to Promote Reactor Vessel Upper Head Mixing/Cooling During Natural Circulation
ST-HL-AE-1492, Forwards Response to Draft Ser/Fsar Items Q440.38 & Q440.39 Re long-term Cooling.Changes Will Be Incorporated in Future FSAR Amend
2CAN948409, Forwards Rept, Justification of Trip Two/Leave Two Reactor Coolant Pump Trip Strategy During Transients, in Response to Generic Ltr 83-10
RC-94-0313, Application for Amend to License NPF-12 to TS Re Changes to Current SBLOCA Evaluation Model Applicable to Plant
CPSES-200500255, Submittal of Supplement to the CPSES Loss of Coolant Accident (LOCA) Analysis Methodologies - Topical Report ERX-04-004, Revision 0
05000333/LER-1989-025, :on 891130,isolation Signal Closed Outboard Isolation Valves on Steam Supply Line to HPCI Sys,Rendering HPCI Sys Inoperable & Initiating 7-day Limiting Condition for Operation.Caused by Excess Steam
05000250/LER-2075-001, Reporting Unusual Event No. 251-75-1 of Revised ECCS Analysis, for Value for Peak Clad Temperature Is in Error
PLA-2953, Forwards Application for Proposed Amend 58 to License NPF-22,changing Tech Specs to Support Cycle 3 Reload.Unit Scheduled to Shutdown on 880305 & Restart on 880503.Reload Summary & Transient Analysis Repts Encl.Fee Paid
L-83-593, Forwards Addl Info Requested by Re Operational Limits Tech Spec Amend
L-83-492, Forwards Reanalysis of Large Break LOCA W/Ungridded Bart Model & Comparison of Results W/& W/O Grid Model,Per 830823 Telcon.Analysis Provides Addl Info for Bart Tech Spec Amend
05000250/LER-1978-002-01, /01T-0:on 780324,util Notified by NSSS Vendor That Logic Inconsistency Identified in NRC Approved 10CFR50.46 App K,Eccs Evaluation Model.Cause Not Stated.Review Underway
L-83-369, Forwards Assessment of Safety Impacts of Removal of Thermal Shield from Facility,Per 830419 Commitment.Assessment Addresses Cycle 6 Reload Safety Analysis of Plant Recovery Program
L-83-429, Forwards Corrected Safety Assessment as Result of Thermal Shield Removal,Submitted w/830623 Ltr
CPSES-200701127, Response to Request for Additional Information Related to License Amendment Request (LAR) 06-009, Revision to Technical Specification (TS) 3.8.1, AC Sources - Operating, Extension of Completion Times
CPSES-200601880, Response to RAI Relief Request A-1 for the Unit 2 Inservice Inspection for Application of an Alternative to the ASME Boiler and Pressure Vessel Code Section XI Examination Requirements for Class 1 and 2 Piping Welds
BSEP 02-0169, GE Nuclear Energy Report NEDO-33063, Safety Analysis Report for Brunswick Steam Electric Plant, Units 1 and 2, Maximum Extended Load Line Limit Analysis Plus