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.