Transient & Accident Safety Analysis and the Final Safety Analysis Report (FSAR)

The Final Safety Analysis Report provides a discussion and analysis of each plant system. The FSAR for older plants must comply with 10CFR50.34. For new plants, the safety analysis requirements are specified in 10CFR52. The plant FSAR often incorporates/references the plant designer's generic Design Control Document.

Sections (ABWR example) include:

TOC Table of Contents
1 Introduction and General Description of Plant
2 Site Characteristics
3 Design of Structures, Components, Equipment and Systems
4 Reactor
5 Reactor Coolant System and Connected Systems
6 Engineered Safety Features
7 Instrumentation and Control Systems
8 Electric Power
9 Auxiliary Systems
10 Steam and Power Conversion System
11 Radioactive Waste Management
12 Radiation Protection
13 Conduct of Operations
14 Initial Test Program
15 Accident and Analysis
16 Technical Specifications
17 Quality Assurance
18 Human Factors Engineering
19 Response to Severe Accident Policy Statement
20 Question and Response Guide
21 Engineering Drawings
Each time, a change is made to a nuclear power plant, the licensee must evaluate whether an unreviewed safety question exists. The law that governs this is referred to as 10 CFR 50.59. If an unreviewed safety question exists, the licensee must obtain NRC permission to perform the change. This law also applies to certain tests.

Every year the licensee also must submit to the NRC an update of all facility changes performed within the year for which NRC approval was not needed.

Each cycle, the licensee must re-analyze the impact of various types of transients and accidents. The reason these studies are done each cycle is that the fuel characteristics may change; thus the different types of accidents and transients could affect the fuel differently. Some of the typical transients and accidents analyzed are reported in Section 14 or 15 of the Nuclear Power Plants' Final and Updated Safety Analysis Reports. Examples include:

  • Loss of Feedwater
  • Anticipated Transients without Scram
  • Bypass of the feedwater heaters
  • Small Break Loss of Coolant Accident (SB-LOCA)
  • Large Break Loss of Coolant Accident (LB-LOCA)
  • Rod Ejection Accident
  • Fuel Handling Accident (Spent Fuel Area)
  • Fuel Handling Accident (Containment)
  • Rupture of Steam Pipe (Large/Small)
  • Environmental Consequences of LOCA
  • Long Term Cooling following LOCA
  • Dilution events
  • Subcriticality Events
  • Steam Generator Tube Rupture (PWR only)
  • Uncontrolled rod withdrawal (subcritical/at power)
  • Loss of Reactor Coolant Flow
  • Loss of all AC Power
  • Control rod misalignment
  • Chemical and Volume Control System Malfunction
  • Startup of an Inactive Reactor Coolant Loop
  • Loss of External Electrical Load
  • Accidental Release of Radioactive Liquids
  • Accidental Release of Waste Gas

Sample FSARs and Design Control Documents

Sample Technical Specifications

Copyright 1996-2010. The Virtual Nuclear Tourist. All rights reserved. Revised: June 24, 2010.