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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Transformational Challenge Reactor Autonomous Control System Framework and Key Enabling Technologies

Autonomous operation of nuclear plants has the potential for significantly improving the economics of existing plants while maintaining operation within the safety envelope of the reactor. Autonomous operation technology, if developed and deployed, may enable deployment of small reactors at remote sites while requiring minimal operator support. In addition to reducing costs, the technology can add additional layers of safety through rapid decision making and actuation in off-normal situations. Within the context of the Transformational Challenge Reactor (TCR), autonomous control along with the other unique characteristics of the reactor (advanced manufacturing) is expected to enable rapid deployment. On the other hand, technology for autonomy can add risk to the system if not properly developed. There is therefore a need for systematic development and demonstration of autonomy technology for TCR. This document describes a framework for autonomy for the TCR. It defines autonomous operation and examines whether autonomy is possible for nuclear systems. The report also describes enabling sensing technologies for autonomy and describes a framework to facilitate autonomy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Molten Salt Reactor Initiating Event and Licensing Basis Event Workshop Summary

Oak Ridge National Laboratory (ORNL) hosted a workshop on identifying potential initiating events for radioactive releases as precursors to licensing basis events for a generic liquid-fuel molten salt reactor (MSR). The workshop was held on May 21 and 22, 2019. Participants included representatives from seven prospective reactor vendors, industry bodies, US and Canadian regulators, US and Canadian national laboratories, and the academic community. Accident sequence evaluation is central to deterministic and risk-informed performance-based reactor safety evaluation processes, and initiating events begin the accident sequence evaluation process. The workshop focused on how MSR initiating events feed into the risk-informed, performance-based reactor safety evaluation process described in DG-1353. This report describes the workshop activities and results, provides a generic list of initiating events involving selected systems, and presents the estimation of the relative frequency and consequences of accident sequences for a few important, high-level initiating events.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Development of a Corrosion Monitoring System for Advanced Molten Salt Reactors

Molten salt reactors (MSRs) have been identified as one of several possible next-generation advanced reactor designs that could replace the aging fleet. There are many inherent advantages to MSRs, including higher power density, lower pressure, low stored energy, and prompt negative temperature reactivity coefficient Ezell et al. [2018]. However, there are challenges related to the operating temperatures, as well as, the corrosive salts attacking the structural materials. During the Molten Salt Reactor Experiment (MSRE), which was operated at Oak Ridge National Laboratory (ORNL) (Ignatiev et al. [2014], Forsberg et al. [2007], Forsberg et al. [2008], Delpech et al. [2009], Serp et al. [2014], Ignatiev and Surenkov [2017]) corrosion was monitored by sampling the salt for chromium. However, this type of monitoring does not identify the location of corrosion. In FY18, ORNL developed a novel in-situ corrosion sensor that could be placed on the outside of the structure to monitor changes in magnetic susceptibility of salt-wetted alloys as corrosion occurs Holcomb et al. [2019]. Initial testing with this novel sensor demonstrated a correlation between the change in signal output from the sensor to the various levels of pre-corroded test specimens. However, all initial testing was performed at room temperatures, which is not realistic for the harsh MSR environment. Based on this research, FY19 funding allowed for development of a high-temperature version of the sensor to be installed at various locations on the reactor infrastructure. The high-temperature sensor can operate at up to 750°C and can monitor development of corrosion inside the pipe. This sensor measures the increase in magnetic susceptibility that occurs as chromium dissolves from the inner surface of the pipe, when paramagnetic structural alloys become ferromagnetic with the loss of chromium. This report will discuss the development of corrosion monitoring sensor and the testing with results of corroded specimens at increasing temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

A New Method to Contain Molten Corium in Catastrophic Nuclear Reactor Accidents

The catastrophic nuclear reactor accident at Fukushima damaged public confidence in nuclear energy and a demand for new engineered safety features that could mitigate or prevent radiation releases to the environment in the future. We have developed a novel use of sacrificial material (SM) to prevent the molten corium from breaching containment during accidents as well as a validated, novel, high-fidelity modeling capability to design and optimize the proposed concept. Some new reactor designs employ a core catcher and a SM, such as ceramic or concrete, to slow the molten corium and avoid the breach of the containment. However, existing reactors cannot easily be modified to include these SMs but could be modified to allow injectable cooling materials (current designs are limited to water). The SM proposed in this Laboratory Development Research and Development (LDRD) project is based on granular carbonate minerals that can be used in existing light water reactor plants. This new SM will induce an endothermic reaction to quickly freeze the corium in place, with minimal hydrogen explosion and maximum radionuclide retention. Because corium spreading is a complex process strongly influenced by coupled chemical reactions (with underlying containment material and especially with the proposed SM), decay heat and phase change. No existing tool is available for modeling such a complex process. This LDRD project focused on two research areas: experiments to demonstrate the feasibility of the novel SM concept, and modeling activities to determine the potential applications of the concept to actual nuclear plants. We have demonstrated small-scale to large-scaled experiments using lead oxide (Pb0) as surrogate for molten corium, which showed that the reaction of the SM with molten Pb0 results in a fast solidification of the melt and the formation of open pore structures in the solidified Pb0 because of CO 2 released from the carbonate decomposition.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Moly 99 Reactor Case Study

Molybdenum-99 (Moly 99) is a critical raw material for Technetium (Tc) 99m, a radioactive isotope most widely used in nuclear medicine procedures. Moly 99 has a short half-life of about six hours, which means it cannot be stockpiled. When Moly 99 decays, it turns into Tc 99m, which has a half-life of 214,000 years. The photon energy emitted from the decay of Moly 99 is used in a variety of nuclear imaging technologies such as gamma cameras. Furthermore, radiopharmaceutical manufacturers use the photon energy emitted from the decay of Moly 99 to produce generators for hospitals, clinics, and radiopharmacies. Once Moly 99 decays to Tc 99m, it is used to make individual patient doses for a variety of diagnostic imaging procedures. The Moly 99 Reactor Design is a conceptual blueprint for Moly 99 production that does not use weapon-grade uranium; instead, the reactor has a target core of low-enriched uranium. The reactor design is small, reaching a foot-and-a-half in height and diameter and consumes less than two megawatts of power. The reactor's only purpose is for medical isotope production and with every fission, Moly 99 is produced.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 1: User's Manual

Razorback is a research reactor transient analysis computer code designed to simulate the operation of a research reactor (such as Sandia National Laboratories Annular Core Research Reactor (ACRR)). The code provides a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This input manual describes how an input file is composed, and facilitates an understanding of the various code input parameters. The makeup of the various code output files is also described. This manual also provides instructions for the installation and setup of the code, and how to report bugs and/or errors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Supervisory Control System for Multi-Modular Advanced Reactors

The proposed supervisory control system (SCS) may provide considerable benefits to advanced small modular reactors, including reduced plant staffing, optimized maintenance activities, greater plant availability, and higher operating efficiency. The SCS makes risk-informed decisions based on (1) a probabilistic assessment of the likelihood of success given the status of the plant/systems and component health, and (2) a deterministic assessment between plant operating parameters and reactor protection parameters to prevent unnecessary trips and challenges to plant safety system—one measure of SCS success. The probabilistic portion of the decision-making engine of the SCS is based on the control actions associated with an advanced liquid-metal reactor (ALMR) Power Reactor, Innovative, Small Module (PRISM). Within the SCS, the probabilistic assessment provides a ranking of viable control actions; however, certain instructions generated by the probabilistic model only include an abstract notion of action without specifications. For instance, one instruction may be to reduce power without specifying how much reduction is needed. The prognostic/diagnostic models incorporate the health of components into the decision-making process. Once the control options are identified and ranked based on the likelihood of success, the SCS transmits the options to the deterministic portion of the platform. The performance-based system models assess and rank each of the probabilistically identified control actions by taking into account the physical behavior (current and projected) of the system. The performance-based decision making module receives inputs from the probabilistic decision-making module and the ERM module to generate a single solution. Interfaces to these modules are defined later in the section. A utility theory algorithm factors into the decision making by estimating the distance from and approach to a trip setpoint for each control option. If the magnitude of a negative utility value increases rapidly as the system approaches the trip setpoint, that option is not likely to be the preferred option. This can lead to a re-ranking of the control options. The SCS then transmits a control signal(s) to a component or system and informs the operator of actions taken based on the action chosen. The SCS successfully coupled probabilistic and performance-based system models to arrive at optimal control decisions based on the actual status of the plant and components. The automatic, autonomous, and real-time performance requirements for a control system were met by the SCS. The value of coupling probabilistic and performance-based system models was demonstrated by the re-ranking of control options based on the use of a utility algorithm. The use of ERM monitors provides added value to the SCS as demonstrated by the re-ranking of control options based on a components degraded state.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors (Final Project Report)

Advanced (i.e., non-light-water) reactor technologies offer developers new opportunities to enhance the reliability, efficiency, and safety of nuclear power reactors through changes in fundamental design and operation. A variety of advanced design concepts are being developed that generally trend toward increased reliance on innovative, inherent, and passive safety features. Many of these features are substantially different from components now found in large light water reactors (LWRs) that dominate the commercial operating fleet. Similarly, non-light water reactor (non-LWR) suppliers are also pursuing market niches quite different from the regional baseload power generation sites typically associated with large LWR facilities.

42 ENGINEERING↗
DOE OSTI2019

The Annular Core Research Reactor (ACRR) Description and Capabilities

The Annular Core Research Reactor (ACRR) is a unique, one-of-a-kind nuclear reactor facility operated by Sandia National Laboratories for the National Nuclear Security Administration that is considered by many to be a national treasure. There is no other research reactor in the world that has the attributes and capabilities comparable to that of the ACRR. The ACRR was specifically designed to meet the irradiation testing needs of the U.S. nuclear weapons program. The ACRR has four major attributes that make it unique: 1) a large dry central cavity; 2) an epithermal neutron flux; 3) a large pulsing capability; and 4) a fueled-ring external cavity with a larger dry cavity. This report presents the unique capabilities of the ACRR and documents some of the more important metrics associated with its operation. ACKNOWLEDGEMENTS The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Design of Microwave Resonant Cavity Transducer. Development of sensor performance model of microwave cavity flow meter for advanced reactor high temperature fluids

High-temperature fluid reactors, such as molten salt cooled reactors (MSCR) and sodium fast reactors (SFR), are a promising advanced reactor option. Measurement of high-temperature fluid process variables, in particular the flow inside the pressure vessel, is a challenging task because of harsh environment, which includes high radiation, high temperature, and contact with highly corrosive coolant fluid. We are investigating a microwave cavity-based transducer for high-temperature fluid flow sensing. This sensor is a hollow metallic cylindrical cavity, which can be fabricated from stainless steel, and as such is expected to be resilient to radiation, high temperature and corrosive environment of MSCR and SFR. The principle of sensing consists of making one wall of the cylindrical cavity flexible enough so that dynamic pressure, which is proportional to fluid velocity, will cause membrane deflection. A cavity is characterized by its resonant frequencies. Membrane deflection causes cavity volume change, which leads to a shift in the resonant frequency. Feasibility of the flow sensor is evaluated with signal sensitivity using COMSOL computer simulations. A right cylinder geometry stainless steel cavity with dimeter of 0.8in was investigated. We choose membrane thickness of 10mil, so that corrosion anticipated to proceed at the rate of 1mil/year in liquid sodium would affect no more than 10% of the membrane. Using the properties of liquid sodium fluid, and stainless-steel material property values at 500oC, we calculate frequency shift for a range of values of fluid velocity from 0.5m/s to 2m/s. Results of computer simulations indicate measurable sensitivity to flow for this cavity design. Following these simulations, we have developed a preliminary design for fabrication of a transducer operating in microwave K-band for proof-of-principle tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Nevada National Security Site Demonstration Reactor Siting and Capability Study

The study’s purpose is to identify suitable testbeds within the NNSS for the demonstration of advanced and emerging reactor technologies. These testbeds are envisioned to be existing structures, facilities, or sites that may be suitable for advanced reactor demonstrations, advanced reactor fuel fabrication, or other experimental support for advanced reactor demonstrations. This study examines siting alternatives at NNSS with the objective of identifying locations for near term ARD projects. Included in the outcomes is an evaluation of the quantitative and qualitative factors that must be addressed for each site if selected to host an ARD sponsored demonstration.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

A Verification and Validation Approach for COMSOL Multiphysics to Support High Flux Isotope Reactor (HFIR)

Over the last several decades, many reactors have successfully been converted from high enriched uranium (HEU) to low enriched uranium (LEU) fuels in United States in support of its global non-proliferation objectives. Of the reactors slated for conversion, five high-performance research reactors (HPRRs) remain. The high-flux isotope reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is one of the five. Conversion of HFIR requires the qualification of a new fuel. To aid in the conversion process as well as improve safety margins, COMSOL Multiphysics was chosen to support and eventually supplement the steady-state heat-transfer code (SSHTC). In order to use COMSOL as a supplementary tool, i.e. one used for safety basis calculations, the code needs to undergo a verification and validation process. Verification and validation are important processes for the use of any software; however, physics-based solvers have a unique challenge, in that the validation is much more rigorous and its domain is only applicable to the general scope of the problem. This document details the verification and validation of COMOSL for analysis of HFIR. This report is intended for two different users: (1) current and future users of COMSOL for HFIR and (2) for individuals interested in assessing the scope of the validation of COMSOL for HFIR. This report is divided into six sections. The first of which details the process of verification and validation. Section one provides an overview of key concepts from ASME V&V 10, 20, and 40 and how these concepts are used to provide a validation of any multiphysics software. Section two covers the approach needed in order to validate multiphysics software for a given HFIR analysis. Parts three through five split up the key concepts highlighted in section two. The final section provides concluding remarks which demonstrate the scope and limitations of the validation study and how these may be expanded and improved.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Proposed Alternate Methodology for Preparation of a Documented Safety Analysis for the Transformational Challenge Reactor

10 CFR Part 830, Subpart B, Appendix A, Table 2 lists acceptable methodologies for development of a Documented Safety Analysis (DSA) for various types of nuclear facilities. The acceptable methodology listed for a DOE reactor is U.S. Nuclear Regulatory Commission Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, or successor document. This document was developed for large commercial light-water power reactors. The difference between the Transformational Challenge Reactor (TCR) and commercial reactors in power output, radiological source term, fuel type and composition, and operational life span makes application of the Regulatory Guide to the TCR overly burdensome relative to the value added.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Qualification of System-Level Advanced Reactor Safety Analysis Software for Lead Systems: Final CRADA Report

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over forty-five years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. Although SAS4A/SASSYS-1 was developed to support the analysis of any liquid-metal-cooled nuclear reactor, it has primarily been utilized to design and analyze Sodium Fast Reactors (SFRs). As a result, most of the qualification basis for SAS4A/SASSYS-1 has utilized sodium as a coolant and geometry descriptions that are prototypic of SFR configurations. In this project, which partnered with Westinghouse Electric Company, LLC, the initial foundation for a qualification basis centered on prototypic pool-type lead-cooled systems has been established, where the end goal is to extend support for utilization of SAS4A/SASSYS-1 in Lead Fast Reactor (LFR) licensing or authorization. This project included three fundamental technical tasks: LFR V&V test suite development (Task 1); qualification support for LFRs (Task 2); and LFR modeling capabilities evaluation and improvement (Task 3)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Light Water Reactor Sustainability Program Materials Research Pathway Technical Program Plan

The Materials Research (MR) Pathway within the Light Water Reactor Sustainability (LWRS) Program is charged with performing the research and development (R&D) to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear reactors. Furthermore, it is essential to use the mechanistic understanding of degradation phenomena in materials to develop mitigation, repair, and new material alternatives for existing components. The work will provide data and methods to assess performance of systems, structures, and components (SSCs) essential to safe and sustained reactor operations. The R&D products developed from the LWRS Program will be used by utilities, industry groups, and regulators to inform operational and regulatory requirements for materials in reactor SSCs subjected to long-term operation conditions, providing key input to both regulators and industry. The intent of this research is to help in reducing the operating costs, which may be in the form of offset maintenance costs due to better predictive models for component lifetimes, improved analysis of materials through nondestructive evaluation, reduced costs for repairs, or extended performance of plants through the selection of improved replacement materials. To best achieve this, industry experience and guidance are important as is their role in coordinated or collaborative research projects. The objectives of this report are to describe the motivation and organization of the MR Pathway within the LWRS Program, provide details on the individual research tasks within the MR Pathway, describe the outcomes and deliverables of the MR Pathway, including recent technical highlights and progress, and list the requirements for performing the research.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

0D Modeling of Reactor Networks Within Zuzax

We report on progress in developing macroscopic balance equations for combustion and electrochemistry systems. A steady state solution capability is described for the macroscopic reactor network, with an associated steady state continuation method and solution storage capability added in. An example is provided of continuation of a hydrogen flame versus the equivalence ratio. The reactor modeling capability is extended to charged fluid systems, with a description of the new ChargedFluidReactor, SubstrateElement, and MetalCurrentElement reactor classes and novel setup of unknowns within these reactors that preserve charge neutrality. Zuzax's setup for electrochemistry is explained including the specification of the electron chemical potential and the adherence to the SHE Reference electrode specification. The description of the different ways to enter electrochemical reaction rates are described, contrasted, and their derivations with respect to one another are derived. An example of using the ChargedFluidReactor within corrosion problems is provided. We present a description of calculations to understand the phenomena of corrosion of copper from a micron sized droplet of NaCl water droplet, where secondary spreading occurs. An analysis of the discrepancies with experiment is carried out, demonstrating that macroscopic balances can be an important tool for understanding what major factors need to be addressed for a better understanding of a physical system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

National Demonstration Reactor Siting Study—Phase I (Revision 1)

In 2020, the Department of Energy Office of Nuclear Energy (DOE-NE) initiated the Advanced Reactor Demonstration Program (ARDP), with the goal of demonstrating two advanced reactor designs within the next five to seven years. When evaluating challenges in meeting this deadline, the newly established National Reactor Innovation Center (NRIC) determined that there was a limited number of well-characterized sites that could support this time-sensitive demonstration requirement. To close this gap, NRIC initiated a two-phase joint analytic effort including DOE, National Laboratory, and academic partners to examine demonstration reactor siting alternatives. The first phase of the study, which is documented in this report, describes the development of a methodology for site assessment and a preliminary assessment of a limited number of known siting alternatives. Revision 1 of this report includes 3 additional (13 total) siting alternatives and 2 additional (4 total) example scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Demonstration of MELCOR and MACCS Capabilities for Molten Salt Reactor Decay Heat Removal During both Normal Operations and Salt Spill Scenarios

This report provides a demonstration of MELCOR and MELCOR Accident Consequence Code System (MACCS) capabilities to perform a dose assessment for a Molten Salt Reactor (MSR) off-gas system. A primary containment system salt spill is used as the off-normal scenario, along with a normal operation dose assessment for comparison. This report discusses the tools, methods, and information used in this assessment so that it may be utilized as a starting point for future advanced reactor consequence analyses. This report also highlights several gaps, to include the need for reactor inventory information specific to advanced reactors, and the need for specific atmospheric transport models that take into account the unique deposition behaviors of tritium and carbon-14, and makes recommendations for closing these gaps. This report satisfies the DOE NE Milestone M4RD-21SN0601062.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗