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2 pages, 160 KiB

Open AccessEditorial

Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability

by Gyunyoung Heo

Energies 2022, 15(2), 521; https://doi.org/10.3390/en15020521 - 12 Jan 2022

Cited by 2 | Viewed by 1194

Abstract

Since the publication of the first comprehensive Probabilistic Safety Assessment (PSA) study—known as WASH-1400—in the US, PSA has developed into an effective and systematic method of identifying hazards, and evaluating and prioritizing the risks in nuclear facilities [...] Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

Research

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17 pages, 2469 KiB

Open AccessArticle

Application of Dynamic Fault Tree Analysis to Prioritize Electric Power Systems in Nuclear Power Plants

by Sejin Baek and Gyunyoung Heo

Energies 2021, 14(14), 4119; https://doi.org/10.3390/en14144119 - 08 Jul 2021

Cited by 11 | Viewed by 2659

Abstract

Because the scope of risk assessments at nuclear power plants (NPPs) is being extended both spatially and temporally, conventional, or static fault trees might not be able to express failure mechanisms, or they could be unnecessarily conservative in their expression. Therefore, realistic assessment [...] Read more.

Because the scope of risk assessments at nuclear power plants (NPPs) is being extended both spatially and temporally, conventional, or static fault trees might not be able to express failure mechanisms, or they could be unnecessarily conservative in their expression. Therefore, realistic assessment techniques are needed to adequately capture accident scenarios. In multi-unit probabilistic safety assessment (PSA), fault trees naturally become more complex as the number of units increases. In particular, when considering a shared facility between units of the electric power system (EPS), static fault trees (SFTs) that prioritize a specific unit are limited in implementing interactions between units. However, dynamic fault trees (DFTs) can be available without this limitation by using dynamic gates. Therefore, this study implements SFTs and DFTs for an EPS of two virtual NPPs and compares their results. In addition, to demonstrate the dynamic characteristics of the shared facilities, a station blackout (SBO), which causes the power system to lose its function, is assumed—especially with an inter-unit shared facility, AAC DG (Alternate AC Diesel Generator). To properly model the dynamic characteristics of the shared EPS in DFTs, a modified dynamic gate and algorithm are introduced, and a Monte Carlo simulation is adopted to quantify the DFT models. Through the analysis of the DFT, it is possible to confirm the actual connection priority of AAC DG according to the situation of units in a site. In addition, it is confirmed that some conservative results presented by the SFT can be evaluated from a more realistic perspective by reflecting this. Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

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21 pages, 6248 KiB

Open AccessArticle

An Approach to Analyze Diagnosis Errors in Advanced Main Control Room Operations Using the Cause-Based Decision Tree Method

by Awwal Mohammed Arigi, Gayoung Park and Jonghyun Kim

Energies 2021, 14(13), 3832; https://doi.org/10.3390/en14133832 - 25 Jun 2021

Cited by 4 | Viewed by 2046

Abstract

Advancements in the nuclear industry have led to the development of fully digitized main control rooms (MCRs)—often termed advanced MCRs—for newly built nuclear power plants (NPPs). Diagnosis is a major part of the cognitive activity in NPP MCRs. Advanced MCRs are expected to [...] Read more.

Advancements in the nuclear industry have led to the development of fully digitized main control rooms (MCRs)—often termed advanced MCRs—for newly built nuclear power plants (NPPs). Diagnosis is a major part of the cognitive activity in NPP MCRs. Advanced MCRs are expected to improve the working environment and reduce human error, especially during the diagnosis of unexpected scenarios. However, with the introduction of new types of tasks and errors by digital MCRs, a new method to analyze the diagnosis errors in these new types of MCRs is required. Task analysis for operator diagnosis in an advanced MCR based on emergency operation was performed to determine the error modes. The cause-based decision tree (CBDT) method—originally developed for analog control rooms—was then revised to a modified CBDT (MCBDT) based on the error mode categorizations. This work examines the possible adoption of the MCBDT method for the evaluation of diagnosis errors in advanced MCRs. We have also provided examples of the application of the proposed method to some common human failure events in emergency operations. The results show that with some modifications of the CBDT method, the human reliability in advanced MCRs can be reasonably estimated. Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

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20 pages, 6947 KiB

Open AccessArticle

Sensitivity Study on the Correlation Level of Seismic Failures in Seismic Probabilistic Safety Assessments

by Geon Gyu Choi, Woo Sik Jung and Seong Kyu Park

Energies 2021, 14(10), 2955; https://doi.org/10.3390/en14102955 - 20 May 2021

Cited by 4 | Viewed by 1492

Abstract

It is popular that correlated seismic failures spread over the fault tree of a seismic probabilistic safety assessment (PSA) for a nuclear power plant (NPP). To avoid the calculational difficulty of core damage frequency (CDF), the fault tree has been simplified by replacing [...] Read more.

It is popular that correlated seismic failures spread over the fault tree of a seismic probabilistic safety assessment (PSA) for a nuclear power plant (NPP). To avoid the calculational difficulty of core damage frequency (CDF), the fault tree has been simplified by replacing correlated seismic failures with one typical seismic failure by assuming a full correlation among the correlated seismic failures. Then, the approximate seismic CDF of a seismic single-unit PSA (SUPSA) has been calculated for decades with this simplified SUPSA fault tree. Furthermore, current seismic multi-unit PSAs (MUPSAs) have been performed with imperfect seismic MUPSA models that were generated by combining such imperfect seismic SUPSA fault trees. The authors of this study recently developed a method that can calculate an accurate seismic CDF by converting correlated seismic failures into seismic common cause failures (CCFs). In this study, accurate and imperfect MUPSA models were created and their seismic CDFs were compared. The results of this study show that the seismic CDFs in SUPSA and MUPSA are drastically distorted and safety margins are accordingly distorted when the full correlation assumption is employed. Thus, this study shows that very careful attention should be paid to calculating and interpreting seismic CDFs for the single-unit and multi-unit NPP regulations. Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

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24 pages, 6789 KiB

Open AccessFeature PaperArticle

Dynamic Probabilistic Risk Assessment Based Response Surface Approach for FLEX and Accident Tolerant Fuels for Medium Break LOCA Spectrum

by Asad Ullah Amin Shah, Robby Christian, Junyung Kim, Jaewhan Kim, Jinkyun Park and Hyun Gook Kang

Energies 2021, 14(9), 2490; https://doi.org/10.3390/en14092490 - 27 Apr 2021

Cited by 16 | Viewed by 2207

Abstract

After the Fukushima Daiichi Accident, the safety features such as accident tolerant fuel (ATF) and diverse and flexible coping strategies (FLEX) for existing nuclear fleets are being investigated by the US Department of Energy under the Light Water Reactor Sustainability Program. This research [...] Read more.

After the Fukushima Daiichi Accident, the safety features such as accident tolerant fuel (ATF) and diverse and flexible coping strategies (FLEX) for existing nuclear fleets are being investigated by the US Department of Energy under the Light Water Reactor Sustainability Program. This research is being conducted to quantify the risk-benefit of these safety features. Dynamic probabilistic risk assessment (DPRA)-based response-surface approach has been presented to quantify the FLEX and ATF benefits by estimating the risk associated with each option. ATFs with multilayered silicon carbide (SiC), iron-chromium-aluminum, and chromium-coated zirconium cladding were considered in this study. While these ATF candidates perform better than the current zirconium cladding (Zr), they may introduce additional failure modes in some operating conditions. The fuel failure analysis modules (FAMs) were developed to investigate ATF performance. The dynamic risk assessments were performed using RAVEN, a DPRA tool, coupled with RELAP5 and FAMs. A cumulative distribution function-based index provided a mean of comparing the benefits of safety enhancements. For medium break loss of coolant accidents, FLEX operational timing window for each fuel type was estimated. Among these ATF candidates, SiC-type ATF was the most beneficial candidate for an increased safety margin than Zr-based fuel and was found to complement FLEX strategies in terms of risk and coping time. Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

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13 pages, 2262 KiB

Open AccessArticle

A Method to Avoid Underestimated Risks in Seismic SUPSA and MUPSA for Nuclear Power Plants Caused by Partitioning Events

by Woo Sik Jung

Energies 2021, 14(8), 2150; https://doi.org/10.3390/en14082150 - 12 Apr 2021

Cited by 4 | Viewed by 1502

Abstract

Seismic probabilistic safety assessment (PSA) models for nuclear power plants (NPPs) have many non-rare events whose failure probabilities are proportional to the seismic ground acceleration. It has been widely accepted that minimal cut sets (MCSs) that are calculated from the seismic PSA fault [...] Read more.

Seismic probabilistic safety assessment (PSA) models for nuclear power plants (NPPs) have many non-rare events whose failure probabilities are proportional to the seismic ground acceleration. It has been widely accepted that minimal cut sets (MCSs) that are calculated from the seismic PSA fault tree should be converted into exact solutions, such as binary decision diagrams (BDDs), and that the accurate seismic core damage frequency (CDF) should be calculated from the exact solutions. If the seismic CDF is calculated directly from seismic MCSs, it is drastically overestimated. Seismic single-unit PSA (SUPSA) models have random failures of alternating operation systems that are combined with seismic failures of components and structures. Similarly, seismic multi-unit PSA (MUPSA) models have failures of NPPs that undergo alternating operations between full power and low power and shutdown (LPSD). Their failures for alternating operations are modeled using fraction or partitioning events in seismic SUPSA and MUPSA fault trees. Since partitioning events for one system are mutually exclusive, their combinations should be excluded in exact solutions. However, it is difficult to eliminate the combinations of mutually exclusive events without modifying PSA tools for generating MCSs from a fault tree and converting MCSs into exact solutions. If the combinations of mutually exclusive events are not deleted, seismic CDF is underestimated. To avoid CDF underestimation in seismic SUPSAs and MUPSAs, this paper introduces a process of converting partitioning events into conditional events, and conditional events are then inserted explicitly inside a fault tree. With this conversion, accurate CDF can be calculated without modifying PSA tools. That is, this process does not require any other special operations or tools. It is strongly recommended that the method in this paper be employed for avoiding CDF underestimation in seismic SUPSAs and MUPSAs. Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

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14 pages, 3660 KiB

Open AccessArticle

Development of a Software Tool for Seismic Probabilistic Safety Assessment Quantification with a Sufficiently Large Number of Bins for Enhanced Accuracy

by Ji Suk Kim and Man Cheol Kim

Energies 2021, 14(6), 1677; https://doi.org/10.3390/en14061677 - 17 Mar 2021

Cited by 2 | Viewed by 1364

Abstract

Quantification in seismic probabilistic safety assessment (PSA) includes the convolution of hazard and fragility curves. Because it is difficult to find the closed-form integration for the convolution of two curves, numerical methods are widely used in practice. In practical applications, the number of [...] Read more.

Quantification in seismic probabilistic safety assessment (PSA) includes the convolution of hazard and fragility curves. Because it is difficult to find the closed-form integration for the convolution of two curves, numerical methods are widely used in practice. In practical applications, the number of ground motion level bins in numerical methods is limited, and it is not clear whether the limited number of bins leads to conservative or optimistic results. In this study, the effect of the number of bins on the quantification results with simplified assumptions is investigated. It is found that the quantification results mostly decrease as the number of bins increases. To enhance accuracy in the quantification results of seismic PSA, a method and a software tool that enable a sufficiently large number of bins to be used for the quantification of seismic PSA models are developed. The application of the developed software tool to an example seismic PSA model demonstrates how the quantification results approach accurate results as the number of bins increases. The software tool developed in this study is expected to enhance the accuracy of seismic PSA quantification results. Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

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15 pages, 4597 KiB

Open AccessArticle

Mathematical Formulation and Analytic Solutions for Uncertainty Analysis in Probabilistic Safety Assessment of Nuclear Power Plants

by Gyun Seob Song and Man Cheol Kim

Energies 2021, 14(4), 929; https://doi.org/10.3390/en14040929 - 10 Feb 2021

Cited by 3 | Viewed by 1392

Abstract

Monte Carlo simulations are widely used for uncertainty analysis in the probabilistic safety assessment of nuclear power plants. Despite many advantages, such as its general applicability, a Monte Carlo simulation has inherent limitations as a simulation-based approach. This study provides a mathematical formulation [...] Read more.

Monte Carlo simulations are widely used for uncertainty analysis in the probabilistic safety assessment of nuclear power plants. Despite many advantages, such as its general applicability, a Monte Carlo simulation has inherent limitations as a simulation-based approach. This study provides a mathematical formulation and analytic solutions for the uncertainty analysis in a probabilistic safety assessment (PSA). Starting from the definitions of variables, mathematical equations are derived for synthesizing probability density functions for logical AND, logical OR, and logical OR with rare event approximation of two independent events. The equations can be applied consecutively when there exist more than two events. For fail-to-run failures, the probability density function for the unavailability has the same probability distribution as the probability density function (PDF) for the failure rate under specified conditions. The effectiveness of the analytic solutions is demonstrated by applying them to an example system. The resultant probability density functions are in good agreement with the Monte Carlo simulation results, which are in fact approximations for those from the analytic solutions, with errors less than 12.6%. Important theoretical aspects are examined with the analytic solutions such as the validity of the use of a right-unbounded distribution to describe the uncertainty in the unavailability/probability. The analytic solutions for uncertainty analysis can serve as a basis for all other methods, providing deeper insights into uncertainty analyses in probabilistic safety assessment. Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

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Review

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17 pages, 974 KiB

Open AccessReview

Reliability Assessment of Passive Safety Systems for Nuclear Energy Applications: State-of-the-Art and Open Issues

by Francesco Di Maio, Nicola Pedroni, Barnabás Tóth, Luciano Burgazzi and Enrico Zio

Energies 2021, 14(15), 4688; https://doi.org/10.3390/en14154688 - 02 Aug 2021

Cited by 13 | Viewed by 3295

Abstract

Passive systems are fundamental for the safe development of Nuclear Power Plant (NPP) technology. The accurate assessment of their reliability is crucial for their use in the nuclear industry. In this paper, we present a review of the approaches and procedures for the [...] Read more.

Passive systems are fundamental for the safe development of Nuclear Power Plant (NPP) technology. The accurate assessment of their reliability is crucial for their use in the nuclear industry. In this paper, we present a review of the approaches and procedures for the reliability assessment of passive systems. We complete the work by discussing the pending open issues, in particular with respect to the need of novel sensitivity analysis methods, the role of empirical modelling and the integration of passive safety systems assessment in the (static/dynamic) Probabilistic Safety Assessment (PSA) framework. Full article

(This article belongs to the Special Issue Advancements in Probabilistic Safety Assessment of Nuclear Energy for Sustainability)

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