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DOE OSTI2012

The Development of A Human Systems Simulation Laboratory: Strategic Direction

The Human System Simulation Laboratory (HSSL) at the Idaho National Laboratory is one of few facilities of its kind that allows human factors researchers to evaluate various aspects of human performance and human system interaction for proposed reactor designs and upgrades. A basic system architecture, physical configuration and simulation capability were established to enable human factors researchers to support multiple, simultaneous simulations and also different power plant technologies. Although still evolving in terms of its technical and functional architecture, the HSSL is already proving its worth in supporting current and future nuclear industry needs for light water reactor sustainability and small modular reactors. The evolution of the HSSL is focused on continual physical and functional refinement to make it a fully equipped, reconfigurable facility where advanced research, testing and validation studies can be conducted on a wider range of reactor technologies. This requires the implementation of additional plant models to produce empirical research data on human performance with emerging human-system interaction technologies. Additional beneficiaries of this information include system designers and HRA practitioners. To ensure that results of control room crew studies will be generalizable to the existing and evolving fleet of US reactors, future expansion of the HSSL may also include other SMR plant models, plant-specific simulators and a generic plant model aligned to the current generation of pressurized water reactors (PWRs) and future advanced reactor designs. Collaboration with industry partners is also proving to be a vital component of the facility as this helps to establish a formal basis for current and future human performance experiments to support nuclear industry objectives. A long-range Program Plan has been developed for the HSSL to ensure that the facility will support not only the Department of Energy’s Light Water Reactor Sustainability Program, but also to provide human factors guidance for all future developments of the nuclear industry.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2012

Supercell Depletion Studies for Prismatic High Temperature Reactors

The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challenges exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Nuclear hybrid energy infrastructure

The nuclear hybrid energy concept is becoming a reality for the US energy infrastructure where combinations of the various potential energy sources (nuclear, wind, solar, biomass, and so on) are integrated in a hybrid energy system. This paper focuses on challenges facing a hybrid system with a Small Modular Reactor at its core. The core of the paper will discuss efforts required to develop supervisory control center that collects data, supports decision-making, and serves as an information hub for supervisory control center. Such a center will also be a model for integrating future technologies and controls. In addition, advanced operations research, thermal cycle analysis, energy conversion analysis, control engineering, and human factors engineering will be part of the supervisory control center. Nuclear hybrid energy infrastructure would allow operators to optimize the cost of energy production by providing appropriate means of integrating different energy sources. The data needs to be stored, processed, analyzed, trended, and projected at right time to right operator to integrate different energy sources.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2014

Design and Scaling of the Natural Convection Shutdown Heat Removal Test Facility

The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (ANL) reflects a 1/2 scale model of one conceptual design for passive safety in advanced reactors. The project was initiated in 2010 primarily to conduct ex-vessel, passive decay heat removal experiments in support of the Advanced Reactor Concepts (ARC), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs while also generating data for code validation purposes. The facility successfully demonstrated scoping objectives in late 2013, and is expected to begin testing by early 2014. The following paper summarizes some of the key design and scaling considerations used in construction of the experimental facility, along with an overview of the current instrumentation and data acquisition methods. Details of the distributed fiber optic temperature system will be presented, which introduces a level of data density suitable for CFD validation and is a first-of-its-kind for largescale thermal hydraulics facilities.

Lisowski, Darius D.↗

DOE OSTI2014

The Development of a Demonstration Passive System Reliability Assessment

In this paper, the details of the development of a demonstration problem to assess the reliability of a passive safety system are presented. An advanced small modular reactor (advSMR) design, which is a pool-type sodium fast reactor (SFR) coupled with a passive reactor cavity cooling system (RCCS) is described. The RELAP5-3D models of the advSMR and RCCS that will be used to simulate a long-term station blackout (SBO) accident scenario are presented. Proposed benchmarking techniques for both the reactor and the RCCS are discussed, which includes utilization of experimental results from the Natural convection Shutdown heat removal Test Facility (NSTF) at the Argonne National Laboratory. Details of how mechanistic methods, specifically the Reliability Method for Passive Systems (RMPS) approach, will be utilized to determine passive system reliability are presented. The results of this mechanistic analysis will ultimately be compared to results from dynamic methods in future work. This work is part of an ongoing project at Argonne to demonstrate methodologies for assessing passive system reliability.

Bucknor, Matthew↗

DOE OSTI2017

Operational Performance Risk Assessment in Support of A Supervisory Control System

Supervisory control system (SCS) is developed for multi-unit advanced small modular reactors to minimize human interventions in both normal and abnormal operations. In SCS, control action decisions made based on probabilistic risk assessment approach via Event Trees/Fault Trees. Although traditional PRA tools are implemented, their scope is extended to normal operations and application is reversed; success of non-safety related system instead failure of safety systems this extended PRA approach called as operational performance risk assessment (OPRA). OPRA helps to identify success paths, combination of control actions for transients and to quantify these success paths to provide possible actions without activating plant protection system. In this paper, a case study of the OPRA in supervisory control system is demonstrated within the context of the ALMR PRISM design, specifically power conversion system. The scenario investigated involved a condition that the feed water control valve is observed to be drifting to the closed position. Alternative plant configurations were identified via OPRA that would allow the plant to continue to operate at full or reduced power. Dynamic analyses were performed with a thermal-hydraulic model of the ALMR PRISM system using Modelica to evaluate remained safety margins. Successful recovery paths for the selected scenario are identified and quantified via SCS.

Denning, Richard S.↗

DOE OSTI2018

NuScale Power Module Instrumentation

The safe and efficient operation of any nuclear power plant (NPP) depends on accurate and timely measurement of the primary system temperature, pressure, level, flow, and neutron flux. Integral pressurized water reactor (iPWR) or small modular reactors (SMRs) present unique challenges to instrumentation and control (I&C) sensors and their maintenance strategies. More specifically, the calibration and response time of these sensors may have to be verified periodically to ensure that they maintain their required degree of accuracy and speed of response. A review of technical reports and safety analysis documentation of the NuScale Power Module (NPM) by Analysis and Measurement Services Corporation (AMS), together with NuScale engineers, confirms that the I&C sensors within the NPM will need to be testable as installed to verify their static and dynamic performance at plant operating conditions. Sensor placement and installation as well as the process conditions expected in natural circulation integral SMRs like the NPM are very different from those in large scale NPPs with conventional primary system loop piping. In particular, reactor coolant system (RCS) flows are more complex and flow rates much lower, the average containment temperature is higher, nuclear radiation levels in some areas are greater, and the accessibility of sensors for hands-on maintenance is very limited. These challenges among others related to I&C sensors can be overcome by adapting existing I&C sensor test methods, developing new techniques for non-conventional I&C sensors, and incorporating online monitoring (OLM) technologies into the I&C architecture of the plant to verify the calibration and response time of sensors within the NPM prior to startup, during operation, and during subsequent refueling outages. These methods and technologies will be demonstrated and validated through a research and development (R&D) grant recently awarded to AMS by the U.S. Department of Energy (DOE) to facilitate timely deployment of the first NuScale SMR in the United States and to enable efficient I&C maintenance strategies during the life of the plant. Lastly, this paper presents the R&D plan for the DOE project with a focus on temperature measurement needs of NuScale.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

Innovations in Renewable Energy Technologies, Systems, and Energy Analysis

Costs for renewable energy technologies have declined rapidly in the past decade and their use for residential, commercial, and utility scale electricity has therefore grown exponentially. As the energy mix diversifies, energy analysis becomes increasing important, requiring renewable energy to integrate with other reduced emission energy sources, such as small modular nuclear reactors and carbon capture and utilization. Energy analysis can also explore the expansion of clean energy into new sectors, such as industrial processes and electrification. The U.S. National Renewable Energy Laboratory (NREL) and the Joint Institute for Strategic Energy Analysis (JISEA), which is a partnership of NREL and universities, study the integration and application of clean energy technologies into existing and new sectors. This presentation includes an overview of NREL and JISEA and summarizes innovations in renewable energy technologies and energy systems analysis.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2020

Sensitivity and Uncertainty Quantification of Transition Scenario Simulations

This report documents the first collective attempt at developing and applying capabilities to quantify uncertainties, assess parametric sensitivities, and optimize multiple parameters and metrics in fuel cycle simulations generated by the SA&I Campaign. To do this, external codes that were designed to perform sensitivity analysis and uncertainty quantification (SA&UQ) needed to be coupled to the SA&I Campaign’s nuclear fuel cycle simulators (NFCS). In FY20, two approaches were pursued: 1) coupling Cyclus to an ORNL-internal code called MOT (Metaheuristic Optimization Tool) and 2) coupling DYMOND to the opensource SA&UQ tool kit Dakota. The primary objective of having these NFCS/SA&UQ coupled capabilities is to better inform DOE-NE and other stakeholders on the results generated from the NFCS. For a given set of fuel cycle strategies, policies, and technology assumptions that make up a fuel cycle scenario, these NFCS have traditionally been used by the SA&I Campaign to provide quantitative answers in terms of year-by-year mass flows, infrastructure requirements, costs, etc. With these newly developed coupled capabilities, the SA&I Campaign can now efficiently simulate hundreds or thousands of these scenarios, sample large ranges of parameters and assumptions, and use the unique features of the SA&UQ tools to process the data. This enables providing answers with known and propagated uncertainties, determining the sensitivity of important metrics to different parameters and assumptions, quantifying how much fuel cycle and technology parameters impact each other, and producing optimized fuel cycle strategies for single and multiple variables. To demonstrate these new capabilities, the Cyclus/MOT was used to model several scenarios ranging from simple fleet retirements to transitions to advanced reactors. Specifically, for a transition scenario from LWRs to SFRs and advanced LWRs, uncertainty quantification, sensitivity analysis, and optimization studies were applied to cases involving single and multiple parameter (input) and single and multiple metric (output) variations. In addition, a similar transition scenario was modeled to demonstrate how to optimize the reprocessing capacity parameter to minimize two performance metrics while taking into account uncertainties from two other parameters. Lastly, a depletion module based on SCALE/ORIGEN was added in Cyclus to simulate the third scenario that was designed to quantify the impact of the modeling assumption that all LWR used nuclear fuel have the same burnup. The newly developed DYMOND/Dakota capability was also applied to a transition scenario from the existing fleet to small modular reactors and fast reactors. This particular scenario involves not only explicit isotopic depletion via ORIGEN-2, but also includes multirecycling and utilizing the criticality search feature to determine the fresh fuel composition of recycled fuel, a feature unique to the DYMOND NFCS. A large database of simulations were run with 4 main parameters that were sampled: start date of reprocessing, reprocessing capacity, energy demand growth rate, and advanced reactor share of the fleet. The 4 main metrics were uranium consumption, enrichment requirements, waste generation, and levelized cost of electricity using data from the Cost Basis Report. The demonstrated SA&UQ results include those that inform on how to choose parameters to avoid “failed” scenarios, Sobol’ indices that inform on the importance of various parameters individually and synergistically, and Analysis of Variance (ANOVA) studies that decompose parameter ranges into groups and informs on whether variations are statistically significant.

Feng, B.↗

DOE OSTI2021

Renewable Energy for Industrial Environmental Management

Costs for renewable energy technologies have declined rapidly in the past decade and their use for residential, commercial, and utility scale electricity has grown exponentially as they become cost competitive. Simultaneously, industrial and manufacturing processes have been increasingly seeking ways to reduce emissions and operational costs in highly competitive sectors. With these combined drivers of lower cost and reduced environmental impact, renewable energy may become a viable energy provider for industrial processes such as oil and gas, mining, chemical refining, food production, and manufacturing. Renewable energy technologies may also partner with other reduced emission energy sources, such as small modular nuclear reactors and carbon capture and utilization, to create cleaner and circular industrial systems for reduced resource use. The Joint Institute for Strategic Energy Analysis (JISEA), which is a partnership of the National Renewable Energy Laboratory (NREL) and five universities and others, has been studying to potential for application of clean energy technologies to the heterogenous energy demands in industry. Dr. Jill Engel-Cox will present an overview of NREL and JISEA, the status and potential future of renewable energy technologies, and collaborations with the oil and gas industry and other industrial sectors to improve their environmental performance and reduce operational costs.

ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2022

Advanced Technology and a Resilient Electric Grid

Making the electric grid cleaner and more resilient will require investing in a variety of new and emerging technologies. Advanced technologies including hydrogen, small modular nuclear reactors, and energy storage will all have a role to play in modernizing the electric grid. A diverse energy supply helps to keep the lights on and protects critical infrastructure, however it will also be important to assess the potential vulnerabilities and risks that come with unfamiliar and new technologies. This webinar will feature experts on resilience planning around advanced technologies from multiple sectors and share resources for states as they seek to modernize and decarbonize their electric grid.

advanced technology↗

DOE OSTI2022

2.1.4.420 - Waves to Water: A Desalination Prize

The Waves to Water Prize was launched in 2019 with the purpose of creating solutions to pair wave energy power with desalination, with a specific focus on remote and islanded communities. The Waves to Water Prize administered $3.3 million in prizes over five stages to accelerate the development of small, modular, wave-powered desalination systems capable of providing potable drinking water in disaster relief scenarios and remote coastal locations. The prize supported the integration of existing and novel wave energy generation technologies with water technologies that can deliver effective, consistent, durable, and low-maintenance water delivery systems.

desalination↗

DOE OSTI2023

The Deimos Experiment: Advanced Reactor Testbed

Advanced reactor initiatives are growing significantly through programs nationwide. This research area includes small modular reactors, microreactors, and space reactors. Many of the reactors being designed are untested concepts. They include unique moderators, varying fuel types, high temperatures, and compact configurations. The shift in fuel type, from highly enriched uranium (HEU) to high assay, low enriched uranium (HALEU), is particularly important as it has driven many of the other changes. For example, lower enrichment requires advanced moderators, which in turn require different reflectors to make the systems compact. The change in materials including the transition from HEU to HALEU affects the temperature feedback of the systems. Additionally, these advanced reactor concepts generally have a thermal neutron spectrum in contrast to earlier fast spectrum advanced reactor. With the extensive changes from previous reactor designs, validation experiments are needed. The National Criticality Experiments Research Center (NCERC) is uniquely equipped to perform such experiments. The Deimos experiment, designed for execution at NCERC, will serve as a testbed for advanced reactor concepts. It will use HALEU fuel in a graphite matrix, provide the ability to use advanced moderators, and allow measurements of temperature reactivity coefficients (TRCs).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Non-Nuclear Test Facility Development to Support Verification and Validation Needs

This study focuses on developing a non-nuclear test facility to help meet the verification and validation (V&V) requirements of a water-cooled small modular reactor (SMR) system. The scaling analysis necessary for reactor system experimentation and assessment model V&V considers various reactor system accident scenarios such as main steam line break (MSLB), steam generator tube rupture (SGTR), and containment condensation and heat transfer. The goal is to identify key physics based phenomena of interest (POI), specific figures of merit (FOMs), and the state of knowledge (SOK) by employing phenomena identification and ranking table (PIRT) studies. Important physics phenomena such as heat transfer, fluid dynamics, structural integrity, and material selection must be addressed when scaling a thermal-hydraulics system. Alongside these considerations, modifications to the facility may be needed to align the scaled system with the available footprint (height and space) of the facility. These modifications may include infrastructure upgrades, changes to the piping layout, implementation of safety measures, and installation of new instrumentation and control systems. The test facility aims to provide a scaled-down version of prototypic facilities for comprehensive V&V activities by overcoming these challenges and making the necessary modifications. This approach minimizes scaling distortion, simulates prototypic plant conditions, and accurately models accident event progressions. Thus, the facility will be capable of adequately modeling a scaled SMR system to support the design, development and licensing needs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Licensing, Regulations and Developing Guidance for Nuclear Technology Deployment for Embarking countries in Africa

The African region is witnessing an increased number of countries at different stages of implementing or considering the introduction of peaceful nuclear power programs to meet the demand of their rapidly growing economies as well as clean energy needs. Grid size, cost, licensing, and regulations will be some of the challenges to deploying larger power reactors. A key element of the regulatory framework will be the inclusion of nuclear security information and assurance that security is integrated into the license review process.As each country continue to implement and consider different reactor designs from several vendors, the significance of customer knowledge of licensing, Regulations, and guidance document to facilitate the deployment and operational needs for nuclear technology cannot be over-emphasized. This special session will bring together a selected panel of key stakeholders and policymakers from embarking countries in Africa to participate in a panel discussion to share their experience in the licensing, regulations, and development of guidance for power plant deployment. The session will also consider and identify potential gaps that may affect the licensing and regulatory applications for the future nuclear infrastructure. As well, the panel will promote regional networking and communication to increase capacities in readiness for the potential deployment of additional technology, such as Small Modular Reactors.

Dahunsi, Stephen↗

DOE OSTI2012

Comparison of a NuScale SMR conceptual core design using CASMO5/simulate5 and MCNP5

A key issue during the initial start-ups of new Small Modular Reactors (SMRs) is the lack of operational data for reactor model validation. To help better understand the accuracy of the reactor analysis codes CASMO5 and SIMULATE5, higher order comparisons to MCNP5 have been performed. These comparisons are for an initial core conceptual design of the NuScale reactor. The data have been evaluated at Hot Zero Power (HZP) conditions. Comparisons of core reactivity, fuel temperature coefficient (FTC), and moderator temperature coefficients (MTC) have been performed. Comparison results show good agreement between CASMO5/SIMULATE5 and MCNP5 for the conceptual initial core design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Short-time scale behavior modeling within long-time scale fuel cycle evaluations

Typically, short-time and long-time scales in nuclear energy system behavior are accounted for with entirely separate models. However, long-term changes in system characteristics do affect short-term transients through material variations. This paper presents an approach to consistently account for short-time scales within a nuclear system lifespan. The reported findings and developments are of significant importance for small modular reactors and other nuclear energy systems operating in autonomous modes. It is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by the Bateman equations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Insights from the WGRISK workshop on the PSA of advanced and new reactors

Probabilistic Safety Assessment /Probabilistic Risk Assessment for new and advanced reactors is recognized as an essential complement of the deterministic approaches to achieve improved safety and performances of new nuclear power plants, comparing to the operating plants. However, the development of PSA to these reactors is encountered to concurrent challenges, mainly due to the limited available design information, as well as due to potentially new initiating events, accident sequences and phenomena. The use of PSA in the decision making process is also challenging since the resulting PSA may not sufficiently reflect the future as-built, as-operated plant information. In order to address these aspects, the OECD/NEA/WGRISK initiated two coordinated tasks on 'PSA for Advanced Reactors' and 'PSA in the frame of Design and Commissioning of New NPPs'. In this context, a joint workshop was organized by OECD, during which related subjects were presented and discussed, including PSA for generation IV reactors, PSA for evolutionary reactors, PSA for small modular reactors, severe accidents and Level 2 PSA, Level 3 PSA and consequences analysis, digital I and C modeling, passive systems reliability, safety-security interface, as well as the results of the surveys performed in the frame of theses WGRISK tasks. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗