ML092870586

Exhibit G - Fact Sheet - Probabilistic Risk Assessment

ML092870586
Person / Time
Site:	Indian Point
Issue date:	10/13/2009
From:	Office of Public Affairs
To:	Atomic Safety and Licensing Board Panel
SECY RAS
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References
50-247-LR, 50-286-LR
Download: ML092870586 (4)

Text

Indian Point Nuclear Generating Units 2 and 3 Docket Nos. 50-247/ 50-286-LR NRC Staff's Response in Opposition to State of New York's Motion for Partial Summary Disposition of NYS Contention 16116A Exhibit G

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Probabilistic Risk Assessment Fact Sheet United States Nuclear Regulatory Commission Office of Publ~cAffairs Wash~ngton,D.C 20555 Telephone: 301/415-8200 E-mall: opa@nrc.gov I

Background

One of the Nuclear Regulatory Commission's key responsibilities is to ensure the operation of nuclear power plants and other NRC-licensed facilities present no undue risk to public health and safety. The agency does this by applying and enforcing a set of technical requirements on plant design and operations, described in Title 10 of the Code ofFederal Regulations (10 CFR). Generally, these are written in tenns of traditional engineering practices such as "safety margins" in design, construction, and operations. Probabilistic risk assessment (PRA) systematically looks at how the pieces of a complex system work together to ensure safety. PRA allows analysts to quantifi risk and identify what could have the most impact on safety.

Risk Risk is determined by two factors:

How often might a particular hazard arise?

How much harm is likely to result'?

Risk is reduced in two ways: by making a11 undesirable event less likely or by making its outcome less serious. The NRC and the nuclear industry use PRA as one way to evaluate overall risk.

Risk Assessment Methods To perform a PRA. analysts go through Inany steps:

1. First. they specif) the hazard the outcome(s) to be prevented or reduced. (For nuclear power plants, "core damage" is often used. The core is where the nuclear fuel is arranged, and it is the first physical barrier between radioactivity and the public.)

2. Analysts then identify initiating events those that could possibly lead to the specified consequence (e.g.. breakage of a pipe canying reactor coolant).

3. Analysts then estimate the frequency of each initiat~ngevent (answering questions such as, "How often do we expect a pipe of this size to break'?").

4. Assuming that the initiating event has occurred, analysts identify each combination of failures (e.g., pump failure and valve failure) that lead to a specific outcome.

5 . The likelihood of each combination is then computed. The probabilities of all those sequences that lead to the same outcome are added. To determine how often this outcome might occur, these probabilities are nlultiplied by the frequency of the initiating event(s).

Modern PRA uses several specific techniques to accomplish this analysis:

Event trees model the plant response to each initiating event.

Flrrilt trees model plant systems in detail. Analysts use fault trees to identify the combinations of failures that cause the overall system to fail. Then, the fault tree logic is used to calculate the overall failure probability. Analysts pay particular attention to problems that can fail inore than one component at the same time.

Hr~monr-eliohili~~ nnulj-sis (HRA) is used to evaluate human errors that are important to the outcome of an event. Analysts assess the probability of a mistake in light of factors like training.

procedures, and expected conditions during an event.

Monte Cnrlo methods are used to compute the risks. This technique allows analysts to consider variations in each factor of the analysis, imperfect knowledge, as well as the many possible ways the factors can interact.

Types of Risk Assessments Even a basic PRA model takes a lot of effort. Fortunately, the speed and power of modern computers allows analysts to use, re-use, and refine each PRA nod el to address many questions.

The most comlnon nuclear power plant PRAs deal with "internal events" - those that start inside the power plant or the electric system it serves. "External event" PRAs deal with initiating events like earthquakes and hurricanes. Special PRAs are used to address unique problems such as spent nuclear fuel storage cask design or the geology of a potential reposito~ysite for high-level radioactive waste.

In the nuclear power industry, analysts label PRAs by "level."

A Level 1 PRA estimates the frequency of core damage. It starts with conditions that are well known, usually with a reactor operating at full power. All of the systems that work to protect the reactor are modeled. Since the workings of these systems are well widerstood, the uncertainty of the result is relatively small.

A Level 2 PRA estiinates the magnitude and timing of releases. (That is, "Assuming that the core is damaged, how much radioactivity might escape into the environment'?") Uncertainty associated with how much coolant escapes the reactor systems (and how violently), as well as variation in containment system response, makes a Level 2 PRA less precise than a Level I PRA.

A Level 3 PRA assesses the in.iuries and economic losses that might result if radioactibity escaped fro111containment. Highly variable factors like wind speed and direction will affect the results.

Risk Assessment Results PRA results are complex. They do not take the form of a single number. Instead, PRA provides a spectrum of possible outcomes. The frequency with which each of these outcomes is expected is a di.li.sMhutionof values. While it is convenient to summarize the results of a PRA using a single representative value (called apoint estlmntc), it is also important to understand how much larger or smaller the achlal risks might be. This uncertainty in the result is partly because reality is Inore complex than any computer model, partly because modelers do not know everything, and partly because of chance.

A certain level of confidence that adequate safety is achieved can be met in eithcr of two ways:

1. Imposing a large enough safety margin to "bound" the risk; or

2. Reducing the uncertainty so that a s~nallermargin can provide the same (or better) confidence of safety.

Analysts can reduce uncertainty by a) refining their models to more accurately reflect the real world; b) d o ~ n gresearch to better understand the physics of what they model; or c) collecting add~tionaldata to Improve the precision of their models. The NRC does all of these.

NRC Uses of PRA The NRC developed the first nuclear power plant PRA applications in the 1970s. In the decades that followed, analysts refined their methods and developed new insights on risk. Today, the NRC uses these insights to complement traditional engineering methods when making regulatory decisions about power plants, medical uses of nuclear materials, and the handling of nuclear waste. This is a "risk-informed" approach to regulation, which has included:

NRC initiatives to have plants perform integrated plant eval~~ations (using PRA) to discover and correct subtle vulnerabilities, res~~lting in significant improvements to reactor safety.

Using PRA insights, inspection programs have shifted their focus to the plant systems, operations and human performance that are most important to safety.

The reactor oversight program includes a significance determination process (SDP) based on PRA models of each plant. When a piece of equipment fails or is taken out of service for maintenance, its safety significance is assessed. The NRC increases its resources for overseeing n~lclearplants as they encounter problems of increasing risk importance.

In writing or modifying regulations, the NRC often uses PRA to confirm that the new rules are rigorous enough to cover uncertainties - and that new rules include only justifiable requirements.

PRA is one of several neth hods used for risk-informed, performance-based regulation of nuclear materials.

PRA has been used for special analyses, for example, to assess the safety of spent fuel storage casks.

The nuclear industry uses PRA, too:

Existing plant designs are improved by reducing vulnerabilities discovered through PRA.

When multiple systenls are undergoing maintenance, PRA ensures risk remains low.

Risk-informed technical specifications and risk-informed in-service inspection programs are used to focus resources on systems and components most significant to safety.

New reactor designs can be improved prior to submitting applications for certification by the NRC.

PRA use is expected to continue growing as part of a longstanding NRC policy for increased use in all regulatory matters. This should result in a more predictable and timely regulatory approach throughout the agency.

October 2007