Large Early Release Frequency Research Articles

On the Language of Reliability: A System Engineer Perspective

In its classical definition, risk is defined by three elements: what can go wrong, what are its consequences, and how likely is it to occur? While this definition makes sense in a regulatory-based framework where for the current fleet of operating light water reactors (LWRs), the risks associated with nuclear power plants typically are characterized in terms of core damage and large early release frequency (LERF), this approach does not provide a useful snapshot of the health of the plant from a broader perspective. This is due to the very narrow context in which the term “risk” typically is defined as nuclear safety aspects that have the potential to impact public health. In this paper, we take the viewpoint of nuclear safety that is reflective of the current fleet of operating LWRs for which core damage frequency and LERF are appropriate metrics. For other advanced reactor designs, other more applicable technology neutral metrics of reactor safety metrics would be specified. A possible alternate path would start by redefining the word risk with a broader meaning that better reflects the needs of a system health and asset management decision-making process. Rather than asking how likely an event could occur (in probabilistic terms), we can ask how far this event is from occurring. Our approach starts by defining and quantifying component and system health in terms of a “distance” between its actual and limiting conditions, i.e., determination of the margin that exists between the current state/condition and the state where the component/system is no longer capable of achieving its intended function. A margin is a measure that is more reflective of the current state or performance of a component, and therefore more closely tied to decisions that are made on an ongoing basis. We will show how, given the data available from plant equipment reliability and monitoring (e.g., pump vibration data) and prognostic (e.g., component remaining useful life estimation) data, a margin can be described and determined for all types of maintenance approaches (e.g., corrective or predictive maintenance). We show how classical reliability models (e.g., fault trees) can be used to quantify the system margin provided component margin values. In the approach described in this paper, the propagation of margin values through classical reliability models are not performed using classical probabilistic calculations applied to sets (as performed in a typical plant probabilistic risk assessment). Instead, we show how it is possible to propagate margin values through Boolean logic gates (i.e., AND and OR operators) through distance-based operations.

Unanticipated protection failure scenarios optimistically bias reactor safety metrics

Risk management of hazardous technological processes is properly understood on filtered probability spaces where the temporal flow of engineering information is represented. The consequence of ignoring a filtration carrying the time evolution of engineering information includes computing risk metrics that overlook the possibility of unanticipated protection failure scenarios. Ignoring the possibility of as yet undiscovered failure scenarios clearly leads to an optimistic bias when computing risk metrics such as Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) using Probabilistic Risk Assessment (PRA). When properly framed on a filtered probability space, it is impossible to quantify the optimistic bias of computed risk metrics.

Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

A conceptual methodology for site exclusion boundary and site large early release frequency for multi-unit nuclear power plants site

Probabilistic Safety Assessment (PSA) risk metrics Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) are currently applied for single NPP unit. The purpose of the work is to arrive at quantitative value for LERF for a site consisting of many NPPs. For this objective, the work proposes to define an exclusion boundary for the whole site (site exclusion boundary) which is different from exclusion boundary for one NPP. At a site, the exclusion boundaries of the individual NPPs may overlap or can be separate. In the present research, various concepts of total exclusion boundary length for the site are delineated upon using the exclusion radius/boundary of individual NPPs. A case study for a hypothetical site consisting of three NPP units was considered for the demonstration of the proposed methodology. The NPPs have same power level and equal radius of exclusion boundary. LERF results are available for individual NPPs and those accident sequences are relevant that (initiating events are independent events or common cause events) that cause release of more than 4% of core I-131 radioactive nuclide within 10 h of accident. The radionuclide dispersion was calculated using Gaussian plume model for dose at various locations on the site exclusion boundary for the multiple NPP site. The calculation of site LERF was carried out for various locations on the site exclusion boundary for a specific configuration of NPPs at site.

MODELING LOFW IN A PWR USING MELCOR

The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for both operational as regulatory activities. The PSA identifies those components or plant systems whose unavailability contributes significantly to the Core Damage Frequency (CDF) and to the Large Early Release Frequency (LERF) of radioactive material. Based on the PSA Level 1 results for a reference plant under design, the Analysis, Evaluating and Risk Management Laboratory (LabRisco), located in the University of São Paulo (USP), Brazil, started the analytical investigation of severe accident phenomena using the US Nuclear Regulatory Commission (NRC) MELCOR2.2 code – focusing on the qualification of a group of specialists who will subsidize a PSA Level 2 for the same plant. This PSA Level 1 shows that the accident with large CDF contribution is the Loss of Feed Water Accident (LOFW). Therefore, the initial objective of the investigation was to model the progression of severe accidents during a LOFW for the reference Pressurized Water Reactor (PWR) and to analyze the response of the plant under these accident scenarios. During the course of the hypothetical LOFW in the reference plant, hydrogen was generated – by a reaction between the high temperature steam water and the fuel-cladding inside the reactor pressure vessel (RPV) but not representing a serious threat to the RPV integrity.

Time-dependent reliability analysis of the reactor building of a nuclear power plant for accounting of its aging and degradation

The ultimate barrier to prevent contamination of the environment due to a release of radioactivity from a Nuclear Power Plant (NPP) is the reinforced concrete (RC) Reactor Building (RB) which encloses the nuclear reactor. The integrity of this barrier is the main focus of Probabilistic Risk Assessment (PRA)-Level 2, in which accident scenarios that might affect this barrier are modeled in terms of their consequences and their probabilities of occurrence. Traditionally, aging and degradation of the RB are not explicitly considered in the modeling. In this paper, a time-dependent reliability approach is adopted to explicitly model aging and degradation, and the effects on the RB resistance to the accidental stresses and eventually its failure probability. A Finite Element Model (FEM) of the RC is developed and coupled with a degradation model. By this, risk measures, like the Large Early Release Frequency (LERF) and its increase in time due to aging (ΔLERF), are actualized on the basis of the condition monitoring data related to the reactor building and the time-dependent risk of failure is quantified. A case study of an internal overpressure due to a hydrogen explosion is considered to exemplify the methodology.

Consistency issues in quantitative safety goals of nuclear power plants in Korea

Abstract As the safety level of nuclear power plants (NPPs) relates to the safety of individuals, society, and the environment, it is important to establish NPP safety goals. In Korea, two quantitative health objectives and one large release frequency (LRF) criterion were formally set as quantitative safety goals for NPPs by the Nuclear Safety and Security Commission in 2016. The risks of prompt and cancer fatalities from NPPs should be less than 0.1% of the overall risk, and the frequency of nuclear accidents releasing more than 100 TBq of Cs-137 should not exceed 1E-06 per reactor year. This paper reviews the hierarchical structure of safety goals in Korea, its relationship with those of other countries, and the relationships among safety goals and subsidiary criteria like core damage frequency and large early release frequency. By analyzing the effect of the release of 100 TBq of Cs-137 via consequence analysis codes in eight different accident scenarios, it was shown that meeting the LRF criterion results in negligible prompt fatalities in the surrounding area. Hence, the LRF criterion dominates the safety goals for Korean NPPs. Safety goals must be consistent with national policy, international standards, and the goals of other counties.

Perspectives of Optimization of Technical Services and Maintenances Using Risk-Informed Decision-Making at NPPs in Ukraine

In most cases, when it goes about maintenance and repair optimization, the repair concept based on technical state of equipment is considered. This article describes the approach to maintenance and repair optimization at nuclear power plants by their transfer from the outage period to unit operation at power. The analysis of maintenance and repair optimization at foreign NPPs is carried out. The U.S. experience in reducing the outage duration is described. The article presents information on the pilot maintenance and repair optimization project at Zaporizhzhya NPP unit 2. The US Department of Energy through Argonne National Laboratory provides support in the implementation of the project in Ukraine focused on maintenance optimization of safety important systems based on risk-informed configuration management. The scope of this Project involves technical engineering analyses, plant personnel training, a study of improved economic performance indicators, and development of new strategies for NPP configuration, maintenance and operation. In order to optimize plant oeration, it is necessary to analyze plant commitments and requirements using effective, unbiased engineering assessments. Plant specific Probabilistic Risk Assessment combined with other engineering analyses techniques will be used for these assessments. This article also describes the results of the analysis of the preparedness of the Ukrainian regulatory framework for the implementation of the Project on maintenance optimization. The analysis focused on the U.S. and Ukrainian regulations and documents of the IAEA and WENRA international organizations related to maintenance optimization using riskinformed management of plant configuration. Based on the comparative analysis results, the authors identified possible limitations stipulated by non-compliance of the Ukrainian nuclear and radiation safety regulatory framework with the U.S. regulatory and methodological documents used for maintenance and repair optimization at the U.S. NPPs. In particular, the overview of USA regulations leads to the conclusion that they contain exhaustive requirements for the application of maintenance using risk-informed plant configuration management at all stages of its implementation, starting with the requirements for the alternative approach to maintenance and ending with the requirements for maintenance performance monitoring. Ukrainian regulations in turn contain only general provisions and areas for potential application of risk-informed approaches in operation and inspection. It should also be noted that the U.S. maintenance optimization methodology uses criteria based on the core damage frequency and large early release frequency. However, besides core damage frequency Ukraine uses the emergency release frequency criterion that does not correspond to the definition for large early release frequency used in the U.S.

Integrated Level 1–Level 2 decommissioning probabilistic risk assessment for boiling water reactors

Abstract This article describes an integrated Level 1–Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1–Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1–Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1–Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1–Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1–Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

A Study on Methods for Establishing Surrogate Safety Goals for Nuclear Power

Quantitative risk values for nuclear power plants (NPPs) can be obtained by conducting a probabilistic safety assessment (PSA). However, people cannot judge the risk level without comparing the risk values from PSA with the standards of acceptable risk in society. Acceptable risk standards are affected by many factors, and those factors are preferentially considered in specified applications. There are many methods used to establish acceptable risk, and a comparative method is easily understood and accepted by the public. In the United States, both qualitative safety goals and quantitative health objectives (QHOs) for the current generation of light water reactors are established by a comparative method and are described in the Safety Goals Policy Statement published by the U.S. Nuclear Regulatory Commission. The evaluations of Level 1 PSA or Level 2 PSA are enough for most regulatory decisions and engineering practices.In order to use PSA as a useful tool for regulation, establishing surrogate safety goals based on QHOs is necessary. But, there is no clear derivation process. First, this paper introduces the process of how to derive QHOs from qualitative safety goals and a model of quantitative health risk. Then, models using core damage frequency (CDF) and large early release frequency (LERF) based on the QHOs are introduced. The situations of nuclear power for each country—the number of plants, the types of reactors, the weather conditions, the population distribution, and the off-site emergency response plan—are different for each country. This paper considers two representative situations. The first situation is that a society has only a single NPP. The maximum consequence method is used to determine the surrogate safety goals for this situation. The second situation is that a society has multiple types of NPPs and the off-site environments of the plants are different from each other. The statistical tolerance intervals method is used to determine the surrogate safety goals for this situation. Data of individual early fatality and cancer fatality risk in China from 2004 to 2013 are collected and analyzed, and then, Chinese, U.S., Korean, and Japanese QHOs are compared. Chinese QHOs and some data from the reference are used to establish surrogate safety goals for the two situations, which are compared with existing surrogate safety goals CDF = 1E-04 per reactor and LERF = 1E-05/reactor-year.

Progression of Severe Accidents, Containment Performance, and Estimates of Releases of Radionuclides-Level-2 Probabilistic Safety Assessment IPHWR

Probabilistic safety assessment (PSA) of nuclear power plants is performed to yield insights into the safety, design, and performance of the plants and their potential environmental effects. This includes the identification of dominant risk contributors, determination of the vulnerabilities of plant and containment systems, and comparison of options for risk reduction. Three levels of PSA are recognized. Level-1 addresses the identification of plant failures leading to core damage and their frequencies of occurrence. Level-2 addresses the assessment of containment response leading together with level-1 results to the determination of containment release frequencies. A level-2 PSA analyses the challenges to the containment, the possible containment responses and their estimated probabilities, and an assessment of the consequent releases to the environment. Level-3 is the assessment of off-site consequences leading, together with the results of level-2 analysis, for estimation of public risks. A comprehensive level-2 PSA study of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is performed to assess the challenges to the containment, the possible containment responses and their estimated probabilities, and consequent releases to the environment. The dominating sequences consist of small-break loss of coolant accident (SBLOCA) and station black out (SBO) followed by containment isolation failure. The results of this are used as an input for developing the severe accident management guidelines (SAMG) measures. All the SAMG measures incorporated in this study have been found as beneficial and resulted in reduced large early release frequency (LERF).

Development of a Fully-Coupled, All States, All Hazards Level 2 PSA at Leibstadt Nuclear Power Plant

Abstract This paper describes the development process, the innovative techniques used and insights gained from the latest integrated, full scope, multistate Level 2 PSA analysis conducted at the Leibstadt Nuclear Power Plant (KKL), Switzerland. KKL is a modern single-unit General Electric Boiling Water Reactor (BWR/6) with Mark III Containment, and a power output of 3600MW th /1200MW e , the highest among the five operating reactors in Switzerland. A Level 2 Probabilistic Safety Assessment (PSA) analyses accident phenomena in nuclear power plants, identifies ways in which radioactive releases from plants can occur and estimates release pathways, magnitude and frequency. This paper attempts to give an overview of the advanced modeling techniques that have been developed and implemented for the recent KKL Level 2 PSA update, with the aim of systematizing the analysis and modeling processes, as well as complying with the relatively prescriptive Swiss requirements for PSA. The analysis provides significant insights into the absolute and relative importances of risk contributors and accident prevention and mitigation measures. Thanks to several newly developed techniques and an integrated approach, the KKL Level 2 PSA report exhibits a high degree of reviewability and maintainability, and transparently highlights the most important risk contributors to Large Early Release Frequency (LERF) with respect to initiating events, components, operator actions or seismic component failure probabilities (fragilities).

Site core damage frequency for multi-unit Nuclear Power Plants site

The nuclear generating sites around the world are mostly twin unit and multi-unit sites. The PSA risk metrics Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) currently are based on per reactor reference. The models for level 1 and level 2 PSA have been developed based on single unit. The Fukushima accident has spawned the need to address the issue of site base risk metrics, Site Core Damage Frequency (SCDF) and Site Large Early Release Frequency (SLERF), on the site years rather than reactor years. It is required to develop a holistic framework for risk assessment of a site. In the context of current study, the holistic framework refers to integration of risk from all units, dependencies due to external events and operation time of individual units. There is currently no general consensus on how to arrive at site-specific risk metrics. Some documents provide suggestions for site CDF and site LERF. This paper proposes a new method of aggregation of risk metric from the consideration of operating time of individual units under certain assumptions with a purpose to provide a new conceptual aspect for multi-unit PSA. The result of a case study on hypothetical data shows that site level CDF is not sum of CDF of all units but around 18% higher than unit level CDF. When the CDF is considered to be a random variable then, the new methodology produces site CDF as 50% higher than single unit CDF. These two approaches have been detailed in the paper. For a general data set of CDF for individual units, site CDF would more than individual unit CDF however, it would not be multiples of a single unit value.

严重事故时蒸汽发生器传热管蠕变断裂风险评估

严重事故时，一回路将会出现蒸汽自然循环冷凝回流现象；高温蒸汽不断进入热管段，波动管和蒸汽发生器传热管，最终可能导致蒸汽发生器传热管发生蠕变断裂失效。本文首先使用严重事故分析计算严重事故时传热管蠕变断裂参数和传热管断裂的概率。然后使用简化二级概率安全分析模型评估传热管蠕变断裂导致的大量早期释放频率。最后评价一回路卸压和恢复二次侧给水等严重事故缓解措施对防止蠕变断裂的有效性。可以发现在中压和高压的情况下，一回路卸压和恢复二次侧给水能够有效地防止传热管蠕变断裂，降低大量早期释放频率。 Under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg, surge line and steam generator tubes, SG tubes integrity could be threatened by creep rupture, particularly if cracks are present in the tube walls. In this study, the first step we perform thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step we apply the creep rupture models to test the potential for the degraded SG to rup-ture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in simplified Level-2 PSA model, and the overall LERF (Large and Early Release Frequency) risk due to the Induced SGTR was calculated. In the final step, imple-mentation of severe accident management guidance, such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the severe accident process under the condition of high and medium pressure, and reduce LERF effectively.

The Analysis of Severe Accident Induced Steam Generator Tube Rupture and LERF Risk

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg and SG tubes. The first step performs thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step applies the creep rupture models to test the potential for the degraded SG to rupture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in LERF (Large and Early Release Frequency) model, and the overall LERF risk due to the Induced SGTR was calculated. In the final step, implementation of severe accident management guidance (SAMG), such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the result of induced SGTR and LERF effectively.

A STUDY ON METHODOLOGY FOR IDENTIFYING CORRELATIONS BETWEEN LERF AND EARLY FATALITY

The correlations between Large Early Release Frequency (LERF) and Early Fatality need to be investigated for riskinformed application and regulation. In Regulatory Guide (RG) -1.174, while there are decision-making criteria using the measures of Core Damage Frequency (CDF) and LERF, there are no specific criteria on LERF. Since there are both huge uncertainties and large costs needed in off-site consequence calculation, a LERF assessment methodology needs to be developed, and its correlation factor needs to be identified, for risk-informed decision-making. A new method for estimating off-site consequence has been presented and performed for assessing health effects caused by radioisotopes released from severe accidents of nuclear power plants in this study. The MACCS2 code is used for validating the source term quantitatively regarding health effects, depending on the release characteristics of radioisotopes during severe accidents. This study developed a method for identifying correlations between LERF and Early Fatality and validates the results of the model using the MACCS2 code. The results of this study may contribute to defining LERF and finding a measure for riskinformed regulations and risk-informed decision-making.

Quantitative evaluation of change in core damage frequency by postulated power uprate: Medium-break loss-of-coolant-accidents

In general, both deterministic and probabilistic safety analyses are used to evaluate the impact of significant plant modifications such as power uprate. While it must be demonstrated that all deterministic acceptance criteria will be satisfied after plant modifications, the fact that the plant response could get closer to the acceptance limits for several safety variables suggests a potential increase of core damage frequency and other possible risk indicators such as departure from nucleate boiling ratio, large early release frequency, and so on. This paper presents an analysis to quantify such impacts of plant modifications performed within the framework of the Safety Margin Application and Assessment. The study reported in this paper is focused on the medium-break loss-of-coolant-accident scenario of the Zion nuclear power plant with a hypothetical power uprate of 10%. The methodology employed in this analysis is based on the combination of the probabilistic and deterministic approaches. Dominant sequences expected to be impacted by plant modifications are screened out and analyzed by the best-estimate plus uncertainty methodology. The core damage frequency is calculated on the basis of the exceedance probability of safety measures estimated from uncertainty analysis results. The key aspects of the analysis are a reduction of bounding assumptions, the explicit treatment of uncertainties in thermal hydraulic calculations and the quantification of the exceedance probability. The results demonstrate the methodology is capable of quantifying small changes in core damage frequency.

A study on the effects of ESFAS STI modification on the unavailability of ESF actuated components

The nuclear industry has made so great efforts to attain safe and economically viable Nuclear Power Plants all over the world. Probabilistic Safety Assessment (PSA) applications are required now more than ever in order to operate Nuclear Power Plants. In this paper, the unavailability model of Engineered Safety Feature (ESF) actuated components, which is part of the Korean Standard Nuclear Power Plant (KSNPP) PSA model, is used to analyze the effects of Surveillance Test Interval (STI) extension of Engineered Safety Feature Actuation System (ESFAS) on the unavailability of the ESF actuated components. To develop a more accurate and realistic PSA model, the loop controller modules of Plant Control System (PCS) are also modeled on the KSNPP PSA model. The fault tree developed by each ESF actuated component includes the component failure, ESFAS signal failure, and loop controller module failure of PCS. When the ESFAS STI is extended from one month to three months, the unavailability of ESF actuated components partially increased by 10%. This shows the STI extension of ESFAS has minor effects on the unavailability of the ESF actuated components. The effects of ESFAS STI extension of KSNPP on the unavailability of the ESF actuated components as well as Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) is less than expected. That result indirectly indicates that the ESFAS in KSNPP is designed well enough to extend the ESFAS STI.

Comparison of Nuclear Power Plant Risk using Dose-based Risk Measure

In the current risk-informed framework, core damage frequency (CDF) and large early release frequency (LERF) are being widely used as two representative alternative measures to a plant risk. As both measures are focused on frequency itself rather than the plant risk, a simplified and more risk-relevant measure (SiRD) based on both the frequency and the individual dose is proposed as a surrogate for CDF and/or LERF. The whole-body dose and thyroid dose at the exclusion area boundary are introduced to define the surrogate plant risk measure. The site atmospheric dispersion factor, the dose conversion factor for the semi-infinite cloud model for fission products, the breathing rate, and the total activity of the fission product are considered for the dose calculations. The SiRD indices have been calculated for the source term release categories that have been defined at UCN 3&4 (PWR) and WS 1 (PHWR) probabilistic safety assessments (PSAs) in Korea and they were used to compare the nuclear power plant risk as a relative risk measure. The results show that the individual risk at WS 1 is less than that at UCN 3&4, due to the lower release fraction of fission products and the lower frequency of the initiating events.

Implication of PSA uncertainties on risk-informed decision making

Nuclear power plants risk-informed policy is introduced in order to improve safety decision making and regulatory efficiency. The corresponding regulatory guides define the acceptable risk measures and their changes resulting from the modifications in the licensed design of the nuclear power plant. The risk measures used in the acceptance guidelines are the core damage frequency and large early release frequency. The risk measures and their corresponding changes are assessed by the Probabilistic Safety Assessment. The uncertainties of Probabilistic Safety Assessment should be appropriately addressed in the context of the decision making, considering their implication on the obtained results. The Probabilistic Safety Assessment uncertainties include epistemic uncertainties resulting from parameter, model, and completeness uncertainties. The paper presents the obtained results from the uncertainty analysis of the Probabilistic Safety Assessment of the reference nuclear power plant and their implication on risk-informed decision making. The paper focuses particularly on parameter and model uncertainties. The analysed modification is extension of the test interval of the emergency diesel generators. The core damage frequency is the used risk measure in the analysis. The need for the appropriate consideration of the uncertainties in the Probabilistic Safety Assessment in order to adequately support the risk-informed decision making is identified. The deficiency of usage of percentile measures is identified and acknowledged. The need for the adaptation of the risk-informed decision-making principles considering new nuclear power plants is recognized.