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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

A Comparison of RESRAD and GoldSim Models for Assessing Radiological Dose for a RCRA Landfill - 20418

US Ecology Idaho (USEI) operates a Resource Conservation and Recovery Act (RCRA) Subtitle C, Hazardous Waste Treatment, Storage, and Disposal (TSD) Facility on a 640-acre property in Owyhee County, Idaho. USEI accepts a wide variety of RCRA-exempt low activity radioactive wastes, including naturally occurring radioactive material (NORM) and technologically enhanced NORM (TENORM) (USEI 2009). The performance of the landfill with regard to radiological operating permit requirements was previously assessed using a model constructed with the residual radioactivity (RESRAD) computer program, developed by Argonne National Laboratory for the U.S. Department of Energy (DOE). RESRAD supports the evaluation of several environmental transport pathways related to a radionuclide-contaminated soil source term, but it was not conceived as a model for the evaluation of landfill radiological performance. A more comprehensive and realistic model is desired in order to better support submissions to both the State of Idaho's Department of Environmental Quality as well as the U.S. Nuclear Regulatory Commission. A Performance Assessment (PA) computer model was developed using GoldSim software for the USEI RCRA Subtitle C landfill. The RESRAD computer model was specifically developed for calculating soil cleanup criteria and radiological dose and cancer risk from residual radioactive material in soil. To evaluate site-specific conditions, users may select from among a number of available environmental transport and exposure pathways and modify 'default' parameter values. In GoldSim, a user must construct the model 'from scratch.' However, this modeling is facilitated by a number of specialized elements available in GoldSim to support a radiological mass transport model. These include elements for defining radiological decay and ingrowth, container failure and radiological release, advective and diffusive transport, and other processes. Advantages afforded by the use of RESRAD for modeling radiological dose for a RCRA Subtitle C landfill include ease of use and reasonable flexibility in specifying site-specific conditions. GoldSim allows for considerably more flexibility and site-specificity than RESRAD, including evaluation of potentially relevant environmental transport processes not supported in RESRAD. The probabilistic modeling capabilities of GoldSim also far exceed those of RESRAD. The pros and cons of RESRAD and GoldSim for this modeling problem are explored with a focus on identifying approaches and critical factors in identifying the appropriate platform. The capabilities of RESRAD and GoldSim for mathematically modeling the disposal system will be contrasted and compared. For example, the existing RESRAD radiological safety assessment evaluated potentially complete exposure pathways related to infiltration to groundwater and upwards diffusion of radon. Additional transport pathways identified in the USEI Idaho facility Conceptual Site Model include deposition of radon decay products in cover material, root uptake of radionuclides in disposed wastes by native plants, and mixing of cover material by burrowing animals. Results of the two models and how the different transport and dose pathways affect results will be discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Updates to a Preliminary Probabilistic Performance Assessment Model for Radiological and Chemical Contamination at the West Valley Site, New York - 20420

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States at the WNYNSC. NFS built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (640 metric tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations halted in 1972 and never restarted, leaving behind radioactive and chemical wastes. The U.S. Department of Energy (DOE) was required to complete certain waste management activities under the West Valley Demonstration Project (WVDP) Act of 1980 including decommissioning of WVDP facilities. As collaborating agencies, NYSERDA and the DOE are tasked with making decisions about decommissioning and risk reduction for the West Valley Site. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) model to assist the agencies in their decision making process for decommissioning the WVDP and WNYNSC. One important tool that is needed in order to inform the decision-making process is a science-based model of the West Valley Site that evaluates potential future consequences for human health and the environment. This forms the core of the spatial domain of the West Valley PPA Model. The PPA Model, developed using the GoldSim system modeling software, is a tool intended to provide support for decision making that evaluates uncertainty, in a manner that is transparent, defensible, and robust. The PPA Model includes contaminant transport and health effects components, and is organized around geographically-grouped contaminated facilities. These include the waste disposal areas licensed by the U.S. Nuclear Regulatory Commission and the State of New York, a waste tank farm for storage of high level radioactive waste resulting from reprocessing operations, and several areas contaminated with radioactive and chemical constituents. The PPA Model evaluates contaminant transport from these sources to points of exposure across the site and into receiving surface waters and sediments. Hypothetical people and wildlife could be exposed to contamination at these locations, and the effects of these exposures are evaluated. Contaminant transport processes to be evaluated in the PPA Model include groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The evaluation of exposures to people in this preliminary model is limited to a resident farmer scenario, and ecological assessment is performed at the level of a screening analysis. The results of these preliminary evaluations inform future model developments. PPA Model results are subjected to sensitivity analysis in order to determine those pathways and parameters that are most significant in influencing the results. This information allows analysts and decision makers to focus on those aspects of Site behavior and processes. With this information, the decision makers can drive informed, defensible decisions regarding decommissioning of the Site. This paper includes an update of the information presented at WM2019 [1]. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Generic Post-Closure Safety Assessment for Deep Borehole Disposal of Heat-Generating Waste - 20451

The deep borehole disposal (DBD) concept for high-activity waste consists of drilling a large-diameter borehole into crystalline basement rock to a depth of about 5,000 m, placing waste packages in the lower portion of the borehole - the waste emplacement zone - and sealing and plugging the upper portion of the borehole with a combination of bentonite, cement plugs, and sand or crushed rock ballast. The waste emplacement zone is several times deeper than typical mined repositories and is well below the typical maximum depth of fresh groundwater resources. Recent work describes a generic reference case for deep borehole disposal of Cs and Sr capsules stored at the Hanford Site. Two scenarios are described: an undisturbed (nominal) scenario, and disturbed scenario, in which a waste package is stuck adjacent to a transmissive fracture zone in crystalline basement. Simulations of the processes affecting radionuclide release and transport in the two scenarios provide the basis for quantitative post-closure performance assessment (PA) of the reference case. This paper presents a probabilistic PA of the stuck waste package scenario, i.e. uncertainties affecting radionuclide mobilization and transport are propagated through 200 realizations of a predictive model to quantify uncertainties in predicted radionuclide concentrations within the borehole and the fracture zone. Simulations predict no radionuclide transport in the borehole to distances greater than 100 m above the stuck waste package. An assumed lateral pressure gradient results in predictions of limited radionuclide transport in the fracture zone at a lateral distance of 200 m from the stuck waste package. These results complement previous results for the nominal scenario and contribute to the development of a generic safety case for DBD. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Analysis of Hydrology and Interim Measure Performance in the Chromium Plume at Los Alamos National Laboratory - 20498

Hexavalent chromium, Cr(VI), is present in the regional aquifer beneath Los Alamos National Laboratory (LANL) at concentrations greater than the New Mexico groundwater standard (50 ppb). The source of the Cr is blowdown from a power plant where potassium dichromate was used as a corrosion inhibitor from 1956 to 1972. Blowdown comingled with effluent from another source and migrated approximately 3 miles down a canyon, primarily as surface flow, until reaching an alluvial infiltration zone. The infiltrating water moved vertically through a thick (approximately 900 ft), complex stratigraphic sequence of unsaturated tuffs, basalt, and unconsolidated alluvial deposits. Horizontal transport also took place within spatially limited perched groundwater horizons in the vadose zone (VZ). Breakthrough at the regional aquifer water table is thought to have occurred in multiple hydraulic windows or 'drip points' located approximately 3 miles from the initial release site at the power plant. The regional aquifer is highly heterogeneous and anisotropic, and displays unconfined behavior in an upper zone and confined behavior at depth. The Cr(VI) contamination has so far remained confined to the upper approximately 25 m of the aquifer. The ambient hydraulic gradient in the regional aquifer is toward the east/southeast, but the water table is extremely flat in the Cr plume area compared to surrounding areas. There is some local variability in flow direction, potentially related to effects of seasonal pumping from nearby Los Alamos county water-supply wells, which draw from the regional aquifer several hundred feet below the water table. An interim measure (IM) is currently operating at the site to achieve hydraulic control of plume migration using a series of extraction wells, ion-exchange treatment, and injection along the downgradient portion of the plume. The migration of Cr contamination in the aquifer has been previously evaluated using a calibrated probabilistic numerical model, which has also been used to design the spatial configuration of extraction and injection wells for the IM, and to identify locations for monitoring wells. The model is built using the Finite Element Heat and Mass (FEHM) transfer code, and the Model Analysis and Decision Support (MADS) software is used for analyses. To date, the portion of the IM system that has operated in a continuous manner has been very effective at hydraulic control. Modifications to the flow field associated with the influence of IM pumping and injection are evaluated here, with modeling constraints to be provided by recent observations of tracers injected into both monitoring wells and IM injection wells, as well as opportunistic observations of geochemical signatures of treated ion-exchange effluent injected into IM injection wells. The tracer and geochemical data provide insights into modified flow directions within the Cr plume area that cannot be otherwise obtained. Results of the modeling work are being used to inform adaptive management of IM operations and will be a valuable tool for development and refinement of long-term remediation strategies. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

A Dose Assessment Model for Radioactive Waste Exposed by Gully Erosion at West Valley - 20513

The Western New York Nuclear Service Center (WNYNSC), located approximately 48 km south of Buffalo, New York, is the site of a former nuclear fuel reprocessing and radioactive waste disposal facility. Spent nuclear fuel was processed there from 1966 to 1972, leaving behind radioactive and chemical wastes in two disposal areas and a waste tank farm. Site operations also resulted in releases of radioactivity to site soils, groundwater, and to surface waters draining the site. The New York State Energy Research and Development Authority (NYSERDA) and the U.S. Department of Energy (DOE) are collaborating in a process of decision making for decommissioning those facilities remaining at the WNYNSC following the completion of Phase 1 decommissioning. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) computer model to assist the agencies in this process. The PPA Model includes a contaminant transport component focusing on the movement of contaminants within and among environmental media including groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The model also includes evaluation of potential exposure and health effects for a Resident Farmer exposure scenario, where the Resident Farmer represents a critical group, described as that group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The West Valley Site is located in the glaciated Allegheny Plateau region of western New York State. The waste reprocessing and disposal areas were constructed on a relatively fat area of plateau dissected by drainages of Buttermilk Creek, including Erdman Brook, Franks Creek, and Quarry Creek. An important aspect of the contaminant transport component of the PPA model is consideration of erosive processes such as slumping of the stream slopes and the advance of gullies from these streams. These erosion processes remove material from the plateau, growing the size of the creek valleys and making them wider and deeper. Of particular interest for the impact of erosion is radioactive waste contained in the Nuclear Regulatory Commission (NRC)-Licensed Disposal Area (NDA), the New York State-Licensed Disposal Area (SDA), and residual radiological inventory in the underground storage tanks at the Waste Tank Farm (WTF). The PPA Model is organized around geographically-defined facilities which were constructed upon the plateau, including the NDA, SDA, and WTF. Ongoing stream erosion processes will potentially transfer radioactive waste and residual inventory from these facilities to the ground surface on adjacent hillslope areas where erosion has breached the facility. Hence, it is important to evaluate the consequences of potential exposures to a Resident Farmer on the hillslopes below a breached facility. Two interrelated aspects of the dose assessment model related to hillslope exposure are discussed: 1) representation of the physical processes related to transport of radionuclides from facilities onto the hillslopes, and from the hillslopes into adjoining creeks where contaminated material migrates downstream with surface water and sediment, and; 2) adaptation of the activities associated with the Resident Farmer scenario to assess potential exposures to contamination in the hillslope areas. This discussion will cover the conceptual basis of the hillslope exposure and transport models, and also implementation in the PPA computer model. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Seismic fragility analysis for lateral supports of HTR-PM RPV - 18521

Since Fukushima accident, there has been increasing interest in China to study the risks of nuclear power plants under seismic events that are beyond the design basis. Two methodologies are commonly being adopted for this purpose, namely the seismic probabilistic risk assessment (SPRA) and the seismic margin assessment (SMA). Fragility analysis plays an important role as it not only provides fragility curves as input for the SPRA, but also can deduce HCLPF values for the SMA. In this paper, seismic fragilities of the lateral supports of the HTR-PM reactor pressure vessel (RPV) are studied. The lateral supports are crucial to the safety of HTR-PM RPV under seismic events. As they are located at a high elevation in the reactor building, it is expected they are one of the major contributors to the seismic risk of HTR-PM. The lateral supports are first briefly introduced in terms of structural features, load conditions and stress acceptance criteria in the engineering design. Their fragilities are then analyzed by identifying basic fragility analysis variables and determining values in terms of the median and standard deviations. The associated fragility curves are finally calculated and a HCLPF value is reported. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Preliminary structural assessment of VHTR graphite components according to ASME code - 18532

In this study, a preliminary structural assessment on a graphite component of a VHTR was done in accordance with the semi-probabilistic assessment procedures of the latest 2015 edition of the ASME Boiler and Pressure Vessel Code, Section III, Div. 5, Subpart HHA-Graphite Materials. Establishment of the Weibull distribution for the tensile strength is essential to set up the stress limits based on the probability of failure of the graphite materials. All parameters required in the ASME code were identified and evaluated for a nuclear graphite material, IG-110. The assessment of design level was done on one of fuel elements in a Korean VHTR design by two methodologies in the ASME code: (1) Simplified assessment and (2) Full assessment. The structural integrity of the fuel element was assured by both assessment methods. Also, the effect of the number of integration volume groups on the probability of failure in the full assessment was investigated and a recommendation for the conservative results was suggested. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The Innograph irradiations; HTR graphite material properties from 0 to 25 DPA - 18541

In the fifth, sixth, and seventh Framework Program of the European Commission, the Nuclear Research and Consultancy Group (NRG) have conducted a large graphite irradiation program for the development of High Temperature Reactor technology. The graphite irradiation program comprises five irradiations executed in the High Flux Reactor (HFR) in Petten, resulting in design curves at irradiation temperatures between 650 deg. C and 950 deg. C. The neutron doses of these irradiations, called INNOGRAPH, range from 1 dpa to 23 dpa. The irradiation facilities contained present day available graphite grades: NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, LPEB, IG-110 and IG-430. The post-irradiation examination of the INNOGRAPH samples is focused on dimensional changes, dynamic Young s modulus, coefficient of thermal expansion (CTE) and coefficient of thermal conductivity. Irradiation behaviour of HTR candidate graphite grades are compared with respect to manufacturing process and raw coke material. Accuracy and reliability of irradiation parameters and post irradiation examination methods are essential for safety and performance relevant material property determination, in particular for probabilistic material behaviour such as graphite. Considerations on quality requirements with regards to graphite material property determination are therefore included in the paper. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Fire compartment partitioning approaches for fire probabilistic safety analysis of a high-temperature gas-cooled reactor - 18545

Fire compartment partitioning is the foundational task in Fire Probabilistic Safety Analysis (FPSA), that requires approaches able to balance analytical accuracy and computational burden. In this work, two fire compartment partitioning approaches are illustrated and applied to an electric building of a High-Temperature Gas-Cooled Reactor (HTGR, e.g., HTR-PM), and the quantified risks of the compartments obtained with the two approaches are compared. The results show that risk is insensitive to the utilized fire compartment partitioning approach, making the FPSA tasks easier to be performed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Mechanistic source terms for the XE-100 reactor XSTERM Code Development Approach - 18587

X-Energy is developing the Xe-100 pebble bed modular High Temperature Gas Reactor (HTGR) as an advanced nuclear power source. To achieve its safety objectives, the Xe-100 reactor relies on materials inherent characteristics and passive safety features. Key features include the inherent high temperature characteristics of TRISO-coated fuel particles, graphite moderator, and helium coolant, along with passive heat removal capability of a low-power-density core. This paper will describe the approach adopted by X-Energy to develop and validate mechanistic source terms for the Xe-100. The Mechanistic Source Term (MST) approach for the Xe-100 reactor will build upon that of the prior DOE Next Generation Nuclear Program (NGNP) (Refs. 1, 2, 3 and 4). MST development is closely related to Fuel Qualification (FQ), Licensing Basis Event (LBE) selection and Probabilistic Risk Assessment (PRA). This paper provides a brief overview of the MST road map process. Software plays a crucial role in establishing and quantifying MSTs required for licensing the Xe-100. An insight into X-Energy s comprehensive, integrated software development program for calculating time-dependent radionuclide releases from the Xe-100 over a wide range of events is provided. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Construction of a cyber attack model for nuclear power plants - 115

The consideration of how a compromised digital component can impact neighboring components is critical to understanding the progression of cyber attacks. The degree of influence that one component may have on another depends on a variety of factors, including the sharing of resources such as network bandwidth or processing power, the level of trust between components, and the inclusion of segmentation devices such as firewalls. The interactions among components via mechanisms that are unique to the digital world are not usually considered in traditional probabilistic risk assessment (PRA). This means potential sequences of events that may occur during an attack may be missed if one were to only look at conventional accident sequences. This paper presents a method where, starting from the initial attack vector, the progression of a cyber attack can be modeled. The propagation of the attack is modeled by considering certain attributes of the digital components in the system. These attributes determine the potential vulnerability of a component to different classes of attack and the capability gained by the attackers once they are in control of the equipment. The use of attributes allows similar components (components with the same set of attributes) to be modeled in the same way, thereby reducing the computing resources required for analyzing large systems. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Complexity and error potential of digital I and C - 135

Utilization of digital I and C for application in nuclear power plants led to a growing discussion on how to create methods for a probabilistic evaluation on such systems. Issues arise on the one hand due to a lack of commonly accepted models for quantitative evaluation of availability and reliability and on the other hand due to the integration of such models into the evaluation of reliability of an overall system. Nevertheless programmable and even more complex systems are implemented to satisfy growing functionality requirements. Therefore the question for a method to evaluate the reliability of systems based on software or on programmable logic became even more urgent. The approach described is based on the commonly accepted assumption that more complex designs are less reliable. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Probabilistic risk evaluation of cyber-attacks on a nuclear power plant safety - 205.5

By adopting digital technology in nuclear power plants (NPPs), the cyber-attack has been introduced as one of the newest dangerous threats. It is required to develop efficient defense strategies against possible cyber-attacks on NPPs. However, it is difficult to expect when, where, and how the cyber-attacks will occur because they are intended attacks. Therefore, the risk of cyber-attacks on NPP is getting more and more important. Probabilistic risk assessment (PRA), which is one of the most widely used methods in NPPs to assess the risk, is introduced. However, it is difficult to evaluate the risk of NPPs by cyber-attacks from only the existing PRA method. In this research, a new PSA method, which is based on existing PRA, is used for evaluating of cyber-attacks on NPP. In addition, the case studies were performed for the feasibility study using nuclear power plant PRA model. Through this study, the risk assessment of the cyber-attacks on the NPP is tried by using a previous PRA model. Also, several scenarios are assumed to proceed the risk assessment. The risk-assessment should be performed in future work containing more specific cyber-attack scenarios, and if possible, the risk assessment model should be developed. Finally, the risk of cyber-attacks on NPPs could be evaluated quantitatively. Also, the defense strategies could be deducted by a risk assessment based on the risk assessment results which is based on the proposed method. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Risk assessment of safety-critical data communication in digital safety feature control system - 205

As one of the safety-critical systems in Advanced Power Reactor-1400 (APR-1400) nuclear power plant (NPP), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for transmitting safety component actuation to effectively facilitate various safety-critical field controllers. However, data communication failure risk in the ESF-CCS has yet to be fully quantified, the probability that a safety-critical system in digitalized NPP becomes unsafe due to a network failure must be evaluated to quantify the risk of digital I and C system. In this study, a fault tree model was developed to assess the data link and data network failure-induced unavailability of a safety-critical system function used to generate an automated control signal for actuating design basis accident mitigation equipment. Especially, the risk of data communication failure in a digital safety feature control system was analyzed in consideration of interconnection between safety-critical controllers and the fault-tolerant algorithm implemented in the network configuration. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to safety data link and network failure and their risk effect on safety signal generation unavailability. This study is expected to provide risk information on the safety-critical data communication in a digitalized NPP instrumentation and control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Concrete structural health monitoring in nuclear power plants - 212

Assessment and management of aging concrete structures in nuclear power plants require a more systematic approach than simple reliance on existing code margins of safety. Health monitoring of concrete structures is performed in order to understand the current condition of a structure based on heterogeneous measurements and produce high-confidence actionable information regarding structural integrity. This information can be used to support operational and maintenance decisions. This paper describes a framework of research activities for health monitoring of nuclear power plant concrete structures. The proposed framework consists of four elements: (1) health monitoring, (2) data analytics, (3) uncertainty quantification, and (4) prognosis. The goal of this framework is to enable plant operators to make risk-informed decisions about structural integrity, remaining useful life, and performance of concrete structures across the nuclear fleet. Among several concrete degradations, research focuses on degradation caused by the alkali-silica reaction (ASR). ASR is a chemical reaction that, in the presence of sufficient moisture, forms a gel that increases in volume and exerts an expansive pressure inside the material, thus causing micro to macro-cracks. This paper also presents development of controlled concrete samples with reactive aggregates to cause ASR degradation and application of different non-destructive evaluation techniques to detect the damage caused by ASR. Initial results are presented in the paper along with a potential path forward. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Identification of the accident-related critical digital assets based on probabilistic safety analysis results - 225

Digital I and C systems have been developed and installed in nuclear power plants, and due to installation of the digital I and C systems, cyber security concerns are increasing in nuclear industry. However, there are too many critical digital assets (CDAs) to be inspected in digitalized NPPs. In order to reduce the inefficiency of regulation in nuclear facilities, the critical digital assets that are directly related to an accident are elicited by using the probabilistic safety assessment results. Target initial events are selected, and their heading are analyzed through event tree analysis. Among the headings, the headings that can be proceeded directly to the core damage by the cyber-attack are finally selected as the target of deriving the minimum cut-sets. Based on success criteria of each heading, we analyze the fault trees and derive the minimum set-cuts. In terms of original PSA, the value of probability for the cut-sets is important but the probability is not important in terms of cyber security of NPPs. The important factors is the number of basic events consisting of the minimal cut-sets that is proportional to vulnerability. Finally, the process for identifying accident-related CDAs is suggested based on PSA results. The results of this study are expected to be used to derive the linkage between cyber-attack and accident, and to develop the final core digital asset identification methodology and effective regulatory method. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of cyber security test scenario for non-safety display system - 226

In order to test the cyber security of NPP related to safety, the results of probabilistic safety assessment were used. The fault tree (FT) is used to reflect the impact of the cyber-attack on the operator. The cyber security test should be performed focusing on the scenario derived from the FT. The concept of a security case has been introduced. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Operational performance risk assessment in support of a supervisory control system - 251

A supervisory control system is being developed for multiunit advanced small modular reactors to minimize human interventions during normal and abnormal operations. In the supervisory control system, control action decisions are made based on a probabilistic risk assessment that employs event trees and fault trees. Although traditional probabilistic risk assessment tools are implemented, their scope is extended to normal operations, and the application is reversed to assess the success of non-safety related systems and to enable continued operation of the plant. This extended probabilistic risk assessment approach is called operational performance risk assessment (OPRA). OPRA helps to identify available paths, combine control actions for maintaining plant conditions within operational limits, and to quantify the likelihood of success of these operational trajectories to optimize the selection of alternative actions without activating reactor protection system. In this paper, a case study of OPRA in a supervisory control system is demonstrated for the Advanced Liquid Metal Reactor (ALMR) Power Reactor Inherently Safe Module (PRISM) design, specifically the power conversion system. The scenario investigated involved a condition in which the feedwater control valve that was observed to be drifting to the closed position. Alternative plant configurations that would allow the plant to continue to operate at full or reduced power were identified using OPRA. Dynamic analyses were performed with a thermal-hydraulic model of the ALMR PRISM system using Modelica to evaluate the magnitude of safety margins. Successful recovery paths for the selected scenario were identified and quantified using the supervisory control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗