Probabilistic safety assessment

A TVA plant's probabilistic risk assessment is not one model but a family of them: internal events at power, internal fires and floods, seismic events, high winds, shutdown states and the containment response for each. Lise shows Hana the summary table of results. "People outside the field see one number and either panic or relax," she says. "The number is the least interesting part. What matters is what drives it, how uncertain it is, and what changes would reduce it. Your job this week is to learn to read a PRA the way the people who use it for decisions read it."

What a probabilistic safety assessment is

A probabilistic safety assessment, called a probabilistic risk assessment in the United States, is a systematic estimate of the frequency and consequences of accidents at a plant. It asks three questions: what can go wrong, how likely is it, and what are the consequences? The answers are built up from the methods of the last two lessons: initiating events are identified and grouped, event trees trace the plant response, fault trees quantify each safety function, and the results are combined into the frequencies of core damage, of releases and, in a full assessment, of health effects. Unlike deterministic analysis, a probabilistic assessment considers multiple failures and realistic plant behavior, and gives a view of risk across all events at once.

The first full assessment of commercial reactors was the Reactor Safety Study, WASH-1400, published by the US Nuclear Regulatory Commission in 1975. It found that accident risk was dominated not by the large pipe breaks that had shaped the design basis but by small breaks and transients combined with multiple failures and human errors. The Three Mile Island accident four years later followed a sequence of exactly that kind, a small loss of coolant through a stuck-open valve combined with operator misdiagnosis, and the methods became standard tools worldwide.

Three levels

A Level 1 assessment covers the sequence from initiating events to core damage and produces the core damage frequency, with the dominant sequences and contributors. A Level 2 assessment takes the core damage sequences, grouped into plant damage states, and models the progression of the severe accident in the reactor and containment: the timing and mode of vessel failure, the loads on containment from pressure, hydrogen and molten core, and the performance of containment systems. Its outputs are the frequencies of release categories, defined by the size, timing and composition of the release. A Level 3 assessment models the atmospheric dispersion of each release category, the exposure of the population, the effect of protective actions and the resulting health and economic consequences.

The NRC does not require a PRA for operating plants under Part 50, but most licensees maintain one, and PRAs are required for new designs under Part 52. Risk-informed applications rely on PRAs that meet the ASME/ANS PRA standard, endorsed by Regulatory Guide 1.200. The NRC also maintains its own SPAR models of each plant.

Risk metrics

The most widely used risk metrics are core damage frequency, which measures how often the core is expected to be seriously damaged, and a release frequency, which measures how often a release of a defined size occurs. Release frequency is defined differently in different countries: as large release frequency, large early release frequency or by explicit release limits of a reference isotope such as cesium-137. Level 3 metrics include individual risk of early and latent fatality, and collective measures such as person-sieverts or land contamination.

Importance measures

A probabilistic assessment of a real plant contains thousands of basic events and tens of thousands of minimal cut sets. Importance measures summarize which of them matter. The Fussell–Vesely importance of a basic event is the fraction of the total risk that comes from cut sets containing that event: it measures how much risk would be removed if the event became impossible. The risk achievement worth is the ratio of the risk with the event assumed to have occurred to the base risk: it measures how much the risk would rise if a component were taken out of service or failed. A component with high risk achievement worth needs strong assurance; a component with high Fussell–Vesely importance is a good target for improvement.

# Aggregate core damage frequency and compute importance measures for basic events.
# Each cut set lists basic events; CDF ~ sum of cut set frequencies (rare-event).
p = {"IE_LOOP": 3e-2, "IE_SLOCA": 5e-4, "DG_A": 3e-2, "DG_B": 3e-2, "CCF_DG": 2e-3,
     "AFW": 1e-3, "HPI": 2e-3, "OPER": 5e-3}
cutsets = [("IE_LOOP", "CCF_DG", "OPER"), ("IE_LOOP", "DG_A", "DG_B", "OPER"),
           ("IE_SLOCA", "HPI"), ("IE_LOOP", "AFW", "OPER")]
def f(cs, q=p):
    r = 1.0
    for e in cs: r *= q[e]
    return r
cdf = sum(f(c) for c in cutsets)
print(f"total CDF {cdf:.2e} per reactor-year")
for ev in ["CCF_DG", "DG_A", "HPI", "AFW", "OPER"]:
    fv = sum(f(c) for c in cutsets if ev in c) / cdf
    q1 = dict(p); q1[ev] = 1.0
    raw = sum(f(c, q1) for c in cutsets) / cdf
    print(f"{ev:7s} Fussell-Vesely {fv:5.2f}  risk achievement worth {raw:8.1f}")

total CDF 1.59e-06 per reactor-year
CCF_DG  Fussell-Vesely  0.19  risk achievement worth     95.4
DG_A    Fussell-Vesely  0.09  risk achievement worth      3.8
HPI     Fussell-Vesely  0.63  risk achievement worth    315.8
AFW     Fussell-Vesely  0.09  risk achievement worth     95.5
OPER    Fussell-Vesely  0.37  risk achievement worth     74.4

In this toy model, high-pressure injection has the largest Fussell–Vesely importance because the small-break sequence dominates. The common-cause failure of the diesel generators and the auxiliary feedwater system both have high risk achievement worth: either one being unavailable would raise the core damage frequency almost a hundredfold, because the remaining cut sets would then depend only on the loss of off-site power and an operator action. The operator action appears in several cut sets and has both a substantial Fussell–Vesely importance and a high risk achievement worth, a common finding that points to procedures and training.

Uncertainty

Every input to a probabilistic assessment is uncertain, and the result is a probability distribution rather than a single number. Aleatory uncertainty is the randomness of the events themselves, which the model represents by probabilities. Epistemic uncertainty is the analyst's lack of knowledge about the parameters, models and completeness of the analysis. Parameter uncertainty is propagated by sampling the input distributions, typically lognormal for failure rates, and running the model many times to give a distribution of core damage frequency, often reported as mean, median and 5th and 95th percentiles. The mean is usually used for comparison with goals because it reflects the whole distribution.

Model uncertainty, such as the choice of success criteria, the treatment of a phenomenon or the human reliability method, is handled by sensitivity studies. Completeness uncertainty, the possibility that something important has been left out, cannot be quantified directly. It is addressed by peer review, by comparing with operating experience and with other plants' assessments, and by keeping deterministic defense in depth alongside the probabilistic results. A probabilistic assessment does not replace engineering judgment; it informs it.

Hazards and operating modes

Early assessments covered only internal events at full power, but experience showed that other contributors can be as large or larger. Internal fires can damage cables of redundant trains in one area and cause spurious actuations; fire assessments trace cable routes and fire scenarios through every room. Internal floods from pipe failures can submerge equipment. Seismic assessments combine the site hazard curve with the fragility of each component, expressed as the probability of failure as a function of ground acceleration. High winds, external floods and other external hazards are treated in the same way. Shutdown and low-power states have different configurations, with systems out for maintenance and open primary systems, and their risk is assessed separately.

US assessments that support risk-informed applications are expected to cover the hazards and modes that significantly affect the decision. Fire PRAs following NUREG/CR-6850 were developed by many plants that adopted the NFPA 805 risk-informed fire protection program.

Using a probabilistic assessment

Probabilistic assessments are used for many decisions beyond licensing. They identify design and procedure improvements; they set priorities for inspection and maintenance; they support changes to allowed outage times and surveillance intervals; they inform the categorization of components by safety significance; they evaluate the significance of events and inspection findings; and they support emergency planning. In each case the decision is risk-informed: probabilistic results are weighed together with deterministic requirements, defense in depth, safety margins and performance monitoring.

The NRC's Regulatory Guide 1.174 sets the framework for risk-informed changes to a plant's licensing basis. Acceptable changes keep the increase in core damage frequency small, typically below about 10⁻⁶ per reactor-year as very small or up to about 10⁻⁵ with a low baseline, and the increase in large early release frequency below about 10⁻⁷ or 10⁻⁶ respectively, while maintaining defense in depth and safety margins.

The quality of a probabilistic assessment determines how far it can be relied on. Standards and peer review define what a good assessment contains: a complete list of initiating events, realistic success criteria, careful treatment of dependencies and human actions, plant-specific data where possible, and documentation that lets a reviewer reproduce the results. A probabilistic assessment that is kept current with the plant, often called a living PSA, is far more useful than one produced once for licensing and then left on a shelf.

Scope and quality of a PSA

A complete probabilistic assessment covers all modes of operation, including low power and shutdown states, all sources of radioactive material on the site, including spent fuel pools, and all initiating events and hazards, internal and external. In practice, many assessments began with internal events at power and have been extended step by step. Shutdown states can contribute a surprising share of risk, because systems are out of service for maintenance, containment may be open and the reactor coolant system may be at reduced inventory. Spent fuel pools have long time scales, but their cooling is less redundant than that of the reactor, and their fuel is outside containment.

The quality of an assessment depends on peer review, on the data and on how well the model represents the plant as built and operated. Standards define capability categories, from a model good enough to identify the main contributors to one detailed enough to support decisions that rely on small differences in risk. Licensees keep their assessments up to date as the plant changes, as data accumulate and as methods improve. An assessment that is not maintained can give misleading results, because a modification or a change in procedure may have altered a dominant sequence.

Using PSA results

Probabilistic results are used in many ways. Configuration risk management tools, built from the plant model, show operators and maintenance planners how the risk changes when equipment is taken out of service, so that combinations that raise risk sharply are avoided. Importance measures guide inspection, testing and maintenance effort toward components that matter most. Assessments support decisions about changes to technical specifications, such as extending the allowed outage time of a diesel generator, and about design modifications. Regulators use them to judge the significance of inspection findings and events, often by estimating how much an observed degradation increased the conditional core damage probability.

Probabilistic results are never used alone. Risk-informed decision making combines them with deterministic requirements, defense in depth, safety margins and performance monitoring. A change that lowers the calculated core damage frequency but removes a level of defense or erodes a margin may still be unacceptable. This integrated approach recognizes both the power of probabilistic insights and the limits of models.