United States Nuclear Regulatory Commission - Protecting People and the Environment

Supplemental Reports Related to the US-APWR Design Certification Application

This page provides access to the following documents, which Mitsubishi Heavy Industries, Ltd. (MHI), submitted to the U.S. Nuclear Regulatory Commission’s (NRC), to supplement its design certification application for the U.S. Advanced Pressurized-Water Reactor (US-APWR):

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Topical Reports

Note that the following topical reports relate specifically to the US-APWR design certification application. For a collection of similar safety-related reports with broader applicability to prospective NRC licensees, see Topical Reports.

Date Description
03/31/08 FINDS: Mitsubishi Fuel Assemblies Seismic Analysis Code
07/20/07 Large Break LOCA Code Applicability Report for US-APWR
07/20/07 Large Mass and Energy Release Analysis Code Applicability Report for US-APWR
07/20/07 Small Break LOCA Methodology for US-APWR
07/20/07 Non-LOCA Methodology
05/25/07 Thermal Design Methodology
05/25/07 Fuel Design Criteria and Methodology
04/10/07





07/03/07
Defense in Depth and Diversity
04/10/07






07/03/07
Human System Interface (HSI) System Description and Human Factors Engineering (HFE) Process
03/02/07





07/03/07
Safety System Digital Platform - MELTAC
03/02/07





07/03/07
Safety I&C System Description and Design Process
03/02/07 Revised Version of US-APWR Advanced Accumulator Topical Report
01/26/07 US-APWR Advanced Accumulator Topical Report
01/23/07







10/15/07
Submittal of Quality Assurance Program (QAP) Description for Design Certification of the Mitsubishi Heavy Industries, LTD. US-APWR Submittal of Revised Quality Assurance Program (QAP) Description for Design Certification of the Mitsubishi Heavy Industries, LTD. US-APWR

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Technical Reports

Date Description
04/25/08 Dynamic Analysis of the Coupled RCL-R/B-PCCV-CIS Lumped Mass Stick
02/29/08 US-APWR Fuel System Design
02/29/08 Subcompartment Analyses for US-APWR Design Confirmation
02/29/08 Enhanced Information for PS/B Design
02/27/08 US-APWR Sump Strainer Performance
02/27/08 Criticality Analysis for US-APWR New and Spent Fuel Storage Racks
12/31/07 Comprehensive Vibration Assessment Program for US-APWR Reactor Internals
12/31/07 Defense-in-Depth and Diversity Coping Analysis
12/31/07 FMEA of Control Rod Drive Mechanism Control System
12/31/07 US-APWR Incore Power Distribution Evaluation Methodology
12/31/07 Justification for Deviations between NUREG-1431 Rev. 3.1 and US-APWR Technical Specifications
12/31/07 Validation of the MHI Criticality Safety Methodology
12/31/07 Probability of Missile Generation from Low Pressure Turbines
12/31/07 Qualification of Nuclear Design Methodology using PARAGON/ANC
12/31/07 Structural Analysis for US-APWR Reactor Coolant Pump Motor Flywheel
12/31/07 Mitsubishi Reload Evaluation Methodology
12/31/07 APWR Reactor Internals 1/5 Scale Model Flow Test Report
12/31/07 US-APWR Technical Report Software Program Manual
12/31/07 Probabilistic Evaluation of Turbine Valve Test Frequency
12/31/07 Small Break LOCA Sensitivity Analyses for US-APWR
12/31/07 US-APWR Fuel System Design Parameters
12/05/07 Qualification and Test Plan of Class 1E Gas Turbine Generator System
11/01/07 Common Cause Failure Potential for Safety System Digital Platform

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Page Last Reviewed/Updated Thursday, April 18, 2013