By Dr. Brian Johnson, Ph.D., P.E., , Reading time: 8 minutes
How safe is safe enough? This is a long standing question in the development of nuclear reactors. The defense-in-depth philosophy mandates safety systems that are independent, diverse, and redundant. When using this philosophy of design it is hard to decide if two emergency core cooling systems are enough or if there should be three or four or more. To make these decisions, nuclear reactor designers use two methods.
The first method, known as the deterministic method, uses postulated accidents and the single failure criterion. This means that designers would assume that there was an accident, such as a loss of coolant, and a single failure of the single most important component to respond to the event, such as the failure of a single safety injection pump. Nuclear reactors were, and still are, designed such that given these circumstances and conservative calculation methods, the core will not melt, and no radiation would be released. After doing this safety analysis to show that in a series of postulated accidents and single failures there is no core melt, a reactor is deemed safe enough. The problem with this method is that multiple, more frequent events, are treated the same as rare, catastrophic failures. Additionally it gives no quantitative value to the risk posed by nuclear plants.
A second way to look at the problem is to use the probability of failure as a guide. If this probabilistic method is to be used, the best way to start answering the question of “How safe is safe enough?” is to find out what risks people accept in their daily lives.
As a probabilistic safety goal, the NRC issued the Quantitative Health Objectives (QHO) such that the risk from a nuclear power plant should constitute less than 1/1000th of the total risk to those living near the power plant. This leads to the individual fatality goal of 5x10-7 per year early fatality risk for those within one mile of the site boundary. This is 1/1000th of the accidental deaths in United States. There is also a goal of less than 2x10-6 latent cancer risk for all of those within ten miles of the site boundary, this is 1/1000th of all cancer deaths.
Due to the complexity of nuclear power plants and the rarity of failure, the risk is difficult to quantify directly from experience. While cars crash every day and statistics can be tallied, this is not true with nuclear power plants. To calculate the risk posed by nuclear reactors Probabilistic Risk Assessment (PRA) has been developed. There are four basic steps in PRA.
The end states of most interest for nuclear reactors are core damage and radiation release. Core damage is an intermediate end state that is easier to calculate than radiation release. It also means that there is the economic loss of the reactor. Radiation release is contingent on core damage and failure of radiation containment.
The initiating events fall into two categories, external events such as fires, floods, and earthquakes and internal events such as turbine trip, reactivity insertion, and loss of coolant. Event and fault trees are tools that logically connect a sequence of basic events (such as trip signals, valve operation, and availability of power) intended to respond to an initiating event with end states.
The frequency of initiating events and probability of failure of basic events are input into the event and fault trees. The frequency of end states is then calculated. The basic idea is to be able to use quantifiable failure probabilities, such as those of single pumps, valves, and power sources, and multiply these probabilities so they represent the frequency with which the appropriate combination of failures leads to core damage and radiation release. The mechanistic models of the reactor response to these failures are done using best estimate calculations that include uncertainty. As a result the quantitative results are typically presented using the mean and several percentile estimates of the risk.
To communicate the results of a PRA, complementary cumulative distribution functions (CCDFs) are often presented. These also allow for direct comparison to the QHOs. Due to the difficulty of conducting a PRA from core damage, to release, to dose to public there are three levels of PRA.
Level 1 ends at core damage. Here thermal-hydraulic and reactor physics calculations are done for each scenario with a significant frequency to determine if the core is damaged.
Level 2 ends at release with the source term defined. The probability and amount of radioactive materials released from containment given the sequences from the Level 1 PRA are calculated.
Level 3 ends at dose to public with the early and latent fatality risks explicitly calculated. The dose to the public from the source term from the Level 2 PRA is calculated using meteorological and other radioactive material transport calculations. The results of a Level 3 PRA can be directly compared to the QHOs.
Using the results of several PRAs, the NRC has developed subsidiary goals to the QHOs. This allows a designer to use a Level 1 or Level 2 PRA, which are less difficult to complete than a Level 3 PRA, to show approximate compliance with the QHOs. These subsidiary goals are a Core Damage Frequency (CDF) below 10-4 (Level 1) and a Large Early Release Frequency (LERF) of 10-5 (Level 2). Typically CDF is considered a good surrogate for the latent cancer risk and LERF is considered a good surrogate for the early fatality risk.
Another important, and currently more often used, output of a PRA is the risk importance of safety systems. Using the results of the PRA and the event and fault tree logic, the safety systems most important to safety can be found. The measures most typically used are Fussell-Vesely (FV) and Risk Achievement Worth (RAW). FV basically measures how often failure of a system is expected to be a contributor to risk. RAW measures how much higher the risk would be if a system was removed from the design. These importance measures are used to rank the risk importance of systems.
All of the reactors operating today used deterministic methods to determine whether they are safe enough. The QHOs and the subsidiary goals of CDF and LERF are not binding, but are a goal. Even so, these risk measures are often used to risk-inform regulation for in service inspection. Furthermore, importance measures may also be used to determine if a system must be subject to special treatment rules or may be exempt.
The mean risk of nuclear reactors operating in the United States typically meets the QHOs by an order of magnitude. CDF is typically estimated around 10-5 per year and LERF is typically estimated around 10-6 per year. Advanced reactors promise an even higher degree of safety as PRA has been available during the design phase to assist designers in finding the best ways to reduce risk. One major challenge presented to PRA by advanced designs is the use of passive systems. These systems rely on inherent forces such as natural circulation. Failure modes for passive systems are much different than for active systems. For example, an active system will fail when a pump fails to run or a valve fails to open and thus the safety function is not performed. A passive system will certainly operate, but due to the relatively weak driving forces and uncertainty in load, the system may not have enough capacity to perform the safety function.
Go on to our fuel cycle page to read about what goes into making nuclear fuel.