NEI 99-02, Regulatory Assessment Performance Indicator Guideline
L-16-345, Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Cnfpa) Standard 805
RS-16-222, Response to Request for Additional Information Related to License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair
RS-16-197, License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair
PLA-3354, Semiannual Effluent & Waste Disposal Rept for Jul-Dec 1989
L-15-150, Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805
NL-15-0075, Additional Information Regarding Seismic Hazard Curves
NLS2014015, Supplement to 60-Day Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805
CNL-14-009, Response to NRC Request for Additional Information Related to Review of Potential Breaches of Hesco Modular Flood Barriers and Earthen Embankments Affecting the Updated Hydrologic Analysis Results
NLS2014001, Day Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805
LR-N17-0088, Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 11, Radioactive Waste Management
NL-14-009, Sequoyah, Units 1 & 2 and Watts Bar, Unit 1, Response to NRC Request for Additional Information Related to Review of Potential Breaches of Hesco Modular Flood Barriers and Earthen Embankments Affecting the Updated Hydrologic Analysis Resul
DCL-11-005, Diablo Canyon, Units 1 and 2, Report on the Analysis of the Shoreline Fault Zone, Central Coastal California, Chapter 6.0 - Seismic Hazard Analysis - Appendix A-3
GO2-10-138, Response to Request for Additional Information on License Renewal Application
RA-10-002, Oyster Creek, CY-OC-170-301, Revision 3, Offsite Dose Calculation Manual
DCL-10-003, Submittal of Progress Report on Shoreline Fault Zone, Central Coastal California
1CAN060902, Request for Alternative Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716
NL-09-0332, Request for Approval of Risk-Informed/Safety Based Inservice Inspection Alternative for Class 1 & 2 Piping, VEGP-ISI-ALT-02, Version 1
L-PI-08-029, Supplement to License Amendment Request (LAR) to Add a Power Factor to the Emergency Diesel Generators (EDG) 24-Hour Load Test (Surveillance Requirement 3.8.1.9)
RC-18-0091, License Amendment Request - LAR-16-01490 National Fire Protection Association Standard 805 Program Revisions
CNRO-2007-00026, Request for Alternative GG-ISI-002, Response to Request for Additional Information
CNRO-2007-00022, Request for Alternative GG-ISI-002, Response to Requests for Additional Information
NRC-07-0014, Response to Request for Additional Information Regarding License Amendment Request for Extension of Completion Time for an Inoperable Emergency Diesel Generator - Probabilistic Risk Assessment Questions
CNL-18-020, Clinch River, Response to Portion of Request for Additional Information Related to Emergency Planning Exemption Requests in Support of Early Site Permit Application
3F0703-04, License Amendment Request 280, Rev. 0 Revised Improved Technical Specification (ITS) 3.7.9, Nuclear Services Seawater System
HBL-03-003, Annual Radioactive Effluent Release Report for 2002
LR-N02-0319, Response to Request for Change to TS One Time Extension to Increase Interval of Integrated Leak Rate Test from Ten to Twenty Years
BSEP 02-0163, Revision of Request for License Amendment - Frequency of Performance-Based Leakage Rate Testing
RC-02-0161, Application of the EPRI Risk-Informed ISI Methodology to: VC Summer (Class 1 and 2), Risk Ranking Results - Table 5-2
NRC-02-0040, Proposed License Amendment for a One-Time Deferral of the Primary Containment Integrated Leak Rate Test
BVY-98-114, Provides Response to Request for Addl Info Re Implementation of ASME Code N560 at Plant
BVY-97-137, Forwards Response to 971009 RAI Re Plans for Implementing ASME Code Case N560 at Vermont Yankee Nuclear Power Station
ST-HL-AE-4248, Forwards Changes to Emergency Response Data Sys (ERDS) Data Point Library (Dpl),Per 10CFR50 & NUREG-1394,rev 1
ST-HL-AE-4028, Forwards ERDS Data Point Library,Rev a for Conformance W/ NUREG-1394 & ERDS Communications Description & Questionnaire
A03429, Summarizes 831012 Meeting W/Nrc in Bethesda,Md Re Fuel Pin Failures & Schedule for Submitting Revised Safety Analyses to Support Cycle 6 Operation W/Revised Loading Pattern
LR-N22-0065, Submittal of Relief Request Associated with the Fourth Inservice Inspection (ISI) Interval Limited Examinations
NUREG/CR-2907 Volume 26, Radioactive Effluents from Nuclear Power Plant - Annual Report 2020
W3P87-2220, Forwards Draft B&W Contract,Replacing ORNL Contract Re post-accident Sampling Analysis Program.New Contract Constitutes Emergency Plan Change That Will Be Incorporated Into Rev 12 to Plan
NUREG/CR-2907 Volume 27, Radioactive Effluents from Nuclear Power Plants: Annual Report 2021
05000245/LER-1998-005-02, :on 980330,loss of S1 & S2 Power Supply Were Noted Due to Single Failure.Caused by Inadequate Design & Indequate Mod Review.Abt & Breaker Configuration for MCC EF3 & MCC FE3 Will Be Evaluated
05000327/LER-1987-044, :on 870724,determined That Moderate Energy Line Breaks Might Impair Components Required for Safe Shutdown. Caused by Inadequate Analysis of Breaks Due to Inadequate Communication.Bldg Walls Sealed
05000483/LER-1984-040, :on 840915,feedwater Isolation Occurred Due to High Level Signals Received from Two Steam Generator Level Indicators.On 841010,reactor Trip,Feedwater Isolation & Auxiliary Feedwater Actuation Occurred
RAIO-0719-66284, LLC Response to NRC Request for Additional Information No. 516 (Erai No. 9647) on the NuScale Design Certification Application
RAIO-0218-58544, LLC Response to NRC Request for Additional Information No. 301 (Erai No. 9236) on the NuScale Design Certification Application
GO2-85-105, Semiannual Effluent Rept,Jul-Dec 1984
L-87-147, Application for Amend to License NPF-16,revising Tech Spec 3/4 7.10, Snubbers, to Delete Snubber Listings,Per Generic Ltr 84-13.Evaluation of NSHC Encl.Fee Paid
05000483/LER-2014-005-01, 1 for Callaway Unit 1 Regarding All ECCS Accumulator Isolation Valve Operator Breakers Closed in Mode 3 with RCS Pressure Greater than 1000 PSIG
05000483/LER-2014-005, Regarding All ECCS Accumulator Isolation Valve Operator Breakers Closed in Mode 3 with RCS Pressure Greater than 1000 PSIG
L-MT-13-062, Response to an Apparent Violation in NRC Inspection Report 05000263/2013008 (EA-13-096)
05000370/LER-2010-001, Regarding Loose Connection in a Panel Board Serving a Solid State Protection System Train Maintenance Could Have Prevented Fulfillment of a Safety Function
HNP-09-086, Second Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-based Standard for Fire Protection for Light Water Reactor.
NEI 94-01, License Amendment Request Supplement Pursuant to 10 CFR 50.90: Adoption of NEI 94-01, Revision 2-A, and Extension of Primary Containment Integrated Leakage Rate Test Interval to Fifteen (15) Years
L-PI-05-076, Prairie Lsland Nuclear Generating Plant Units 1 and 2, Response to Request for Additional Information Regarding the Relief Request to Implement Risk-Informed Inservice Lnspection (ISI) Scheduling for the Fourth 10-Year Lnspection Interva
RS-02-164, Request for Amendment to Technical Specification 5.5.13, Primary Containment Leakage Rate Testing Program
LR-N21-0066, Submittal of Relief Request Associated with the Fourth Inservice Inspection (ISI) Interval Limited Examinations
05000361/LER-1986-036, :on 860619,line Voltage Regulators Caused Vital Buses to Be Inoperable.Caused by Procedural Discrepancy. Revised Procedure SO23-II-11.184
05000311/LER-1983-039-03, /03L-0:on 830723,control Rod 2C4 Inserted Fully & Could Not Be Withdrawn.Unit Placed in Hot Standby.Caused by Control Rod Drive Mechanism Cable Connector Pins Pushed Into Insulator,Resulting in Intermittent Connection