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Direct Containment Heating
Published in J. T. Rogers, Fission Product Transport Processes in Reactor Accidents, 2020
During this core melt progression and eventual melt quenching the contain-ment pressure and temperature increases due to physical processes that may challenge its integrity. There have been a number of integrated studies [1–5] which have attempted to systematically identify and quantify these containment pressure/temperature loads and the possible threat of early (t < 2–4 hr) or late (t > 24 hr) containment failure. Recently, the US Nuclear Regulatory Commission has issued the Reactor Risk Reference Document [6] (NUREG-1150) containing the most recent integrated study of containment loads and failure modes. This document considers the two LWR reactor types (a pressurized water reactor, PWR, and a boiling water reactor, BWR) and the five reactor containment configurations (PWR large dry, PWR ice condenser, BWR Mark I, 11 or III) as case studies to estimate the dominant severe accident frequencies, containment loads, source term release paths and the overall risk.
Nuclear power
Published in Peter N. Nemetz, Unsustainable World, 2022
Both WASH-1250 and WASH-1400 were generic studies in that they examined American reactors as a class. In 1982, the NRC published a report (known as the CRAC-II report and with simulations conducted by Sandia Labs) which drastically revised the worst-case scenarios of an atomic reactor accident. Part of the methodology involved examining the risks associated with individual plants rather than plants generically or in the abstract. Table 8.5 compares the WASH-1400 estimates of deaths and property damage with the 1982 revisions. In a further attempt to refine their analysis of a severe accident in a nuclear power plant, the NRC published another report in 1990 numbered NUREG-1150, which focused on potential accident sequences in five American reactors (see also NRDC 2011). Still not confident in the results of these analyses, the NRC subsequently initiated yet another research project in 2007 entitled State-of-the-Art Reactor Consequence Analyses (SOARCA) Report, which was published in 2012 with the designation NUREG-1935. The general results of this more recent study suggested a much more optimistic picture of potential accidents in US nuclear reactors. The key results are listed: When operators are successful in using available on-site equipment during the accidents analyzed in the SOARCA, they can prevent the reactor from melting, or can delay or reduce releases of radioactive material to the environment.Analyses in the SOARCA indicate that all modeled accident scenarios, even if operators are unsuccessful in stopping the accident, progress more slowly and release much smaller amounts of radioactive material than indicated in earlier studies.As a result, public health consequences from severe nuclear power plant accidents modeled in the SOARCA are smaller than previously calculated.The delayed releases calculated provide more time for emergency response actions such as evacuating or sheltering for affected populations. For the scenarios analyzed, the SOARCA shows that emergency response programs, if implemented as planned and practiced, reduce the risk of public health consequences.Both mitigated (operator actions are successful) and unmitigated (operator actions are unsuccessful) cases of all modeled severe accident scenarios in the SOARCA cause essentially no risk of death during or shortly after the accident.The SOARCA’s calculated longer-term cancer fatality risks for the accident scenarios analyzed are millions of times lower than the general US cancer fatality risk.
A dynamic probabilistic risk assessment platform for nuclear power plants under single and concurrent transients
Published in Journal of Nuclear Science and Technology, 2023
M. El-Sefy, A. Yosri, A. Siam, W. El-Dakhakhni, S. Nagasaki, L. Wiebe
A nuclear power plant (NPP) is a complex system-of-systems that requires full understanding of the behavior of each of its components, their system-level behaviors, and the dynamic interaction/interdependence between these different systems and components, under both normal and abnormal operating conditions. Such understanding is now more essential than ever before because of the increased magnitude and frequency of hazard events that can exceed what was originally considered in the NPP design, causing negative impacts on the plant [1]. Probabilistic risk assessment (PRA) approaches for NPPs adopt static event tree (ET) and fault tree (FT) analysis methods [2–4] to estimate the probability of occurrence of each accident scenario and its consequences. These methods rely on the so-called effect line [5], where branching points occur based on a specific action strategy for safety systems to mitigate accident propagation. ET/FT analysis methods were previously employed to estimate the frequencies of core damage [6] and containment radioactive release [7] as well as the impact of the latter on the public and environment [8]. In the nuclear engineering field, PRA has experienced a vigorous period of development, started with performing the U.S. WASH-1400 in 1975, applying several large-scale PRAs for commercial NPPs during the period 1980–1988, and the NUREG-1150 study in 1989, which lead the PRA to be the widely accepted technique by both the industry and regulators. However, PRA methods applied in such studies have been extensively criticized [4,9–14] due to their inability to: i) account for the probabilistic time-dependent interaction among component/system behaviors and the subsequent probability of cascading failures following extreme events; and, ii) consider the dynamic (time sequence) propagation of disruptive events. Lessons learned from the Fukushima nuclear accident also highlighted the importance of enhancing existing, and developing new, PRA approaches to facilitate understanding of the complex dynamic behavior of NPPs following various independent or interrelated hazards [2]. However, precise evaluation of dynamic risks in NPPs following disruptive events requires an accurate representation of the system and the complex interactions between its components.