Information Technology Solutions

PRESCRIPTIVE VS. RISK INFORMED 

REGULATION 

The NRC initially developed many of its 

regulations without considering numerical 

estimates of risk. Rather, prescriptive, 

deterministic regulatory requirements were 

based primarily on basic operational 

experience, test results, and expert judgment. 

To develop these requirements, the NRC 

considered factors such as accepted practice, 

engineering margins and the principle of 

defense-in-depth. In this approach, the NRC 

assumed that certain off-normal events and 

accidents could occur and required plant 

designers to use redundant safety systems 

capable of preventing and/or mitigating the 

consequences of these design basis accidents 

(DBA). 

Since the early 1980’s, however, NRC and its 

licensees have expanded their knowledge and 

use of risk assessment to provide insights on 

the effectiveness of safety systems. A risk 

assessment is a systematic method for 

evaluating risk to a person, as it relates to the 

performance of a particular system (which may 

include a human component) to understand 

likely outcomes, sensitivities, areas of 

importance, system interactions and areas of 

uncertainty. Today, the NRC uses the 

insights gained from risk assessment to 

complement traditional engineering methods 

when making regulatory decisions about 

nuclear facilities. This concept is called riskinformed regulation. 

NRC POLICY STATEMENTS 

The NRC has issued two policy statements 

on safety and risk assessments. The first 

policy statement sets safety goals for the 

operation of nuclear power plants. It states 

that there should be no significant additional 

risk from nuclear power plant operations, 

which translates to an increased risk to a 

person of approximately 1E-6 per year. A 

related document discussed risk informed 

decision making for fuel cycle facilities and 

makes recommendations for safety goals. 

The second policy statement increases the 

use of probabilistic safety assessment (PRA) 

to the extent supported by the state of the 

art. The statement is focused on nuclear 

power plants. There is a related White 

Paper on risk-informed, performance based 

regulation. 

REGULATORY GAPS 

DISCUSSED IN THIS 

SECTION: 

9 Gap 5 – Risk 

Considerations for a 

Production Facility 

Licensed Under 10 CFR 

Part 70 

Safety and Risk Assessment Issues 

Development of Regulations for 

Spent Nuclear Fuel Reprocessing Facilities

One of NRC’s key responsibilities is to ensure that the operation of NRC-licensed 

facilities, including any future reprocessing facility, will present no undue risk to public 

health and safety. To achieve this goal, the NRC requires safety and risk analyses as 

part of the licensing of new facilities. Safety and risk analyses require a detailed 

understanding of the processes and operations of the facility, and an evaluation of how 

these are affected by off-normal or accident conditions. One of the tools that is used to 

understand facility operations and evaluate the risk of these activities is called a risk 

assessment. Selecting the proper risk assessment approaches for reprocessing facilities 

is an important criterion for ensuring adequate assurances of safety and protection of 

public health and safety. 
NRC needs reasonable assurance that 

public health and safety is protected 

through a comprehensive evaluation of the 

hazards and risks of reprocessing facilities.

Appropriately used quantitative risk 

assessment methods, such as PRA, may 

be needed to provide NRC the level of 

information about reprocessing risks and 

hazards necessary to make a finding of 

adequate public health and safety. 

However, the use of PRA in existing 

international reprocessing facilities is 

limited. A PRA analysis is most useful 

when there is meaningful and 

representative data to input into the risk 

analysis models. The NRC will need to 

assess the benefits and costs of a 

quantitative risk assessment requirement 

for reprocessing facilities. 

 

required to complete a probabilistic risk 

assessment. 

RISK ASSESSMENT CONSIDERATIONS 

FOR A REPROCESSING FACILITY

The NRC is evaluating alternatives for 

establishing risk assessment requirements 

for reprocessing facilities. 

The current risk assessment methods 

used to demonstrate compliance with 10 

CFR Part 70 do not fully address the 

hazards associated with a reprocessing 

facility. A reprocessing facility will likely 

handle large amounts of highly radioactive 

spent nuclear fuel, which contain fission 

products and transuranic isotopes. In 

addition, these materials generate 

significant amounts of heat. High radiation 

levels will be encountered in the spent fuel 

processing and separation areas of a 

reprocessing facility. The presence of 

these high radiation materials and 

processes requires the use of automated, 

remotely-operated, and remotelymaintained equipment, in addition to 

complex monitoring and control systems. 

These operational characteristics differ 

significantly from those encountered in 

other types of fuel cycle facilities, where 

high radiation hazards are not 

encountered. Although ISA analyses 

provide adequate safety assurance for 

these fuel cycle facilities, the hazards and 

consequences considered in these 

analyses are not as extensive as those 

pertaining to certain areas or processes of 

a reprocessing facility. Accordingly, the

RISK ASSESSMENT METHODS

The types of risk assessment methods can 

be broadly classified as follows: 

• Quantitative – an approach that 

quantifies event probabilities and 

consequences, and provides estimates 

of uncertainty (e.g., PRA) 

• Semi-quantitative – an approach that 

includes some level of quantification of 

event probability and consequences 

(e.g., risk index method typically used 

for Integrated Safety Analyses [ISA]) 

• Qualitative – an approach that uses 

little or no quantification of event 

probabilities or consequences (e.g., a 

qualitative ISA) 

Currently, fuel cycle facilities licensed 

under 10 CFR Part 70 are required to 

conduct and maintain an integrated safety 

analysis (ISA). An ISA identifies potential 

accident sequences in the facility’s 

operations, and designates items relied on 

for safety (IROFS) to prevent or mitigate 

the accidents. Various ISA methods are 

used at the different types of fuel cycle 

facilities. Some ISAs employ a semiquantitative risk index method for 

categorizing accident sequences in terms 

of their likelihood of occurrence and their 

consequences. 

Reactors licensed under Part 50 initially 

used a design-basis accident (DBA) 

approach. Many reactor licensees are 

using PRA techniques to improve DBA 

analyses. Applicants for new reactor 

licenses under 10 CFR Part 52 are 

Safety and Risk Assessment Issues

(Page 2)