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DOE OSTI2022

Advanced modeling and simulation of research reactors using dynamic mode decomposition

Full text of publication follows. Due to the ever-increasing safety requirements, the current trend of nuclear reactor analysis is shifting towards high-fidelity multi-physics models, which have a very high computational cost and modelling complexity. As the cost of even a single model run makes it impossible to analyse the behaviour and performance of these models on large-scale commercial plants, it has become even more significant to provide suitable benchmarks to validate and test them extensively. In this sense, research reactors offer a promising solution for the initial validation of high-fidelity models, as they are significantly smaller than commercial reactors and their characteristics are well known. In particular, the reactors of the TRIGA family have been used to assess and validate models and methods for Generation-IV designs, as they have some similar features (such as the dominance of natural convection as cooling mechanism and the difficulties in performing sub-channel analysis using standard codes). Still, the computational requirements of high-fidelity models make them unsuitable for real-time analysis, even following their assessment on research reactors. In this sense, Model Order Reduction (MOR) techniques give an additional strategy to reduce the computational cost of high-fidelity models (whilst preserving sufficient accuracy). In particular, this work focuses on Dynamic Mode Decomposition (DMD), a non-intrusive MOR technique that aims at representing models with explicit temporal dynamics by extracting the time-varying characteristics and the governing structures based only on a set of available data, thus without needing any underlying knowledge of the governing equations. In addition, DMD also computes a low-dimensional surrogate of the dynamic matrix of the system, making it suited for stability analysis and real-time evaluations. This work focuses on the application and validation of the DMD method on the Computational Fluid-Dynamics (CFD) model TRIGA Mark II reactor, also discussing in detail the potentiality of this algorithm as an advanced modelling tool for nuclear reactor analysis. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Multi-Physics Coupled Seismic Safety Analysis of Molten-Salt Reactors

Pool-type Molten Salt Reactors (MSRs) are a promising advanced nuclear reactor concept relying on passive physical behavior to provide increased safety characteristics. The high heat capacity, high boiling point, unpressurized liquid salt contains the nuclear fuel and as it passes through the vessel achieves criticality and produces nuclear power. The salt then carries away the heat produced along with the delayed neutron precursors towards heat exchangers and primary pumps before entering the core. This forms a highly coupled multiphysics problem between neutronics and fluid flow, which makes modeling these reactors challenging. The safety/licensing case heavily relies on modeling and simulation to prove their safety, and significant effort has been expanded by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to develop the appropriate simulation tools. Their safety/licensing will require a thorough analysis of their behavior during earthquake-based transients. Earthquakes are a transient condition commonly experienced in many regions, and are safely survived by dozens of reactors every year. Nonetheless, significant conservatisms were introduced with limited predictive modeling and simulation available. The first-of-a-kind capability developed by this research will allow for engineers to analyze ranges of design-basis and beyond-design basis accident scenarios to assess the integrity of the reactor vessel and surrounding system, possibly allowing for limiting the costly conservatisms in the design. This projects uses MASTODON, Griffin, and Pronghorn in the NEAMS ecosystem to model the coupled multiphysics problem posed by an earthquake in a molten salt reactor. It marches through a number of number of coupling schemes, from uncoupled simulations to tightly two-way coupled simulations. The appropriate level of coupling for these simulations will then be determined based on the accuracy in the quantities of interest (QoI) and computational effort involved in each scheme. The QoIs chosen for this paper include the mass flow rate across the heat exchanger (for Pronghorn simulations), power output of the reactor (for Griffin simulations), and the maximum Von Mises stress (for MASTODON simulations). This summary paper reports on the current progress of this project, with 2D tightly coupled multiphysics results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

ARC-100 Reactor Security-by-Design Summary

This report applies the security-by-design methodology developed in a previous National Nuclear Security Administration–sponsored work to the Advanced Reactor Concepts 100 (ARC-100) sodium-cooled fast reactor (SFR) design. The report contains no proprietary information specific to the ARC 100 reactor. The insights developed in this report are high-level, and generally applicable to other sodium fast reactor designs. The information presented here is the result of a qualitative safety-based analysis and would not inform any potential adversary beyond what would be found in a docketed safety analysis report. The scope of this present report covers ARC-100’s reactor core, used fuel storage, and used fuel assembly wash station. These systems are also compared to a generic SFR design assumed in the previous study. The security assessment results show changes in structures, systems, and components (SSCs) safety importance relative to the generic SFR SSCs. However, the consequence assessment results are the similar to a previously assessed generic SFR. Several SSCs have higher importance rankings than others, and it is recommended that protection efforts are prioritized for these SSCs. This work will continue in the Fiscal Year 2025 for the remaining ARC-100 systems, including cesium trap, sodium cold trap, noble gas decay tanks (dewar bottles), and used fuel dry storage facility, to provide safety-and-security-by-design insights and recommendations on non-core systems. Results from this work will furnish a technical justification for the feasibility of these solutions for the ARC reactor's design and, where applicable, identify any regulatory benefits conferred by the proactive design aspect within a risk management framework. This initiative will contribute to a more secure design of the ARC reactor and support its licensing process.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Development of Predictive Models for Advanced Reactor Autonomous Control

Advanced reactor designs including microreactors and small modular reactors will contribute to the clean production of cheap energy, and autonomous control for advanced reactors is an appealing option for reducing cost. However, there is a lack of industry experience applying autonomous control for advanced nuclear reactors. To accelerate the development and industry acceptance of autonomous control software for nuclear reactors, we aim to demonstrate autonomous control of the Purdue University research reactor (PUR-1) using INL-developed model predictive control (MPC) methods. To prepare for this demonstration, data-driven predictive models based on process data collected from PUR-1 have been developed and integrated with MPC and used to control a physics-based model of PUR-1. A data-driven dynamics model and a gated recurrent unit (GRU) network were both trained on process data from PUR-1. The dynamics model was shown to effectively control the reactor model with MPC when provided reactivity as a control variable but failed to control the model through the control rod positions. The GRU network produced more accurate predictions than the dynamics model when evaluated on operational data, and future work will include the evaluation of the GRU network in the controller.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+ ) irradiation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Coupling of Thermal Energy Storage with Small Modular Reactors

Thermal energy storage has been proposed as a load management strategy for Small Modular Reactors (SMRs) deployed on constrained grids and/or in hybrid energy systems (1). Hybrid energy systems can include process steam applications, and the presence of intermittent energy sources such as wind and solar. Under these conditions, the reactor can be subjected to significant time varying electric loads. One strategy for accommodating these loads is to operate the reactor in a load follow mode where reactor power is modulated to match the electric demand. In addition to lost energy potential, load follow operation can result in additional stresses on the fuel and other mechanical components. A more attractive approach is to operate the reactor at or near steady state and bypass excess steam to a thermal energy storage system. The thermal energy can then be recovered later, either for electric generation during periods of peak electric demand, or for uses in process steam applications. In previous work, the design of a sensible heat Thermal Energy Storage (TES) system was proposed for SMR applications (2). In this paper we examine coupling of the TES system to a SMR, typical of Integral Pressurized Water Reactor (IPWR) designs currently under development. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Performance Analysis of the Passive Residual Heat Removal System of an Integral Reactor

A demand for small modular reactors (SMRs) is gradually emerging worldwide, especially in Middle East Asia and Africa. SMART (System-integrated Modular Advanced ReacTor) developed by KAERI (Korea Atomic Energy Research Institute) is the first mover among several SMRs. It is designed to produce 100 MW of electricity, or 90 MW of electricity and 40,000 tons of desalinated water concurrently, which is sufficient for 100,000 residents. SMART is currently holding the initiative as the first licensed integral reactor in the world since 4 July 2012, when the standard design approval (SDA) was officially issued by the Nuclear safety and security commission (NSSC). After that, the SMART safety enhancement project was successfully conducted with intense support of the Government, resulting in an in-depth development of the most advanced safety features including the passive residual heat removal system (PRHRS). Recently, an agreement between KAERI and King Abdullah city for atomic and renewable energy (K.A. CARE) was officially signed for the pre-project engineering (PPE) in pursuit of the construction of SMART and developing human resources capability. Both parties are currently conducting a three-year preliminary study to review the feasibility of building SMART and to prepare for its commercialization. KAERI also takes responsibility to train R and D personnel chosen from Kingdom of Saudi Arabia. Sooner or later, SMART will open the global market of SMR and be the best competitive pioneer. Since the Fukushima accident, the passive cooling system of a nuclear reactor has received a great amount of attention. SMART is designed with passive systems in order to ensure the safety of the reactor. The PRHRS is one of the passive safety systems, which is activated after an accident to remove the residual heat from the core and the sensible heat of the reactor coolant system (RCS) through the steam generators (SGs) until the condition of the RCS reaches a safe shutdown condition. In the present study, the thermal performance of the PRHRS is predicted by using in-house code. Temperature of the RCS and the cooling rate of the RCS are traced with time to see if it is satisfactorily designed to meet the design requirements of the PRHRS. The dependency of the mass flow rate of the natural circulation on the filling ratio (FR) is also discussed. The FR is arbitrarily chosen from 0.22 to 0.50 and further parametric study is needed in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Baseline Radiological Conditions for the Transient Reactor Test (TREAT) Restart

The Transient Reactor Test (TREAT) facility, located on the Idaho National Laboratory (INL), northwest of the Materials and Fuels Complex (MFC), has been in programmatic shutdown for more than twenty-three years. Construction of TREAT by the Teller Construction Co., Portland, Oregon, was started in February, 1958, and completed in early November, 1958. The reactor first achieved criticality on February 23, 1959. Operations continued until April 28, 1994, when the reactor was shut down due to the lack of a programmatic mission. Recently, the U.S. Department of Energy (DOE) identified a need to resume transient testing of new reactor fuels. The Resumption of Transient Testing Program (RTTP) has readied the TREAT facility to resume transient testing operations. The first reactor criticality since the shut down in 1994 was achieved on November 14, 2017. This report details the radiological measurements taken in the TREAT Reactor Building and the surrounding environment prior to and just after its restart in order to characterize the radiological conditions that existed prior to and during the initial restart operations. The authors would like to acknowledge the support from RTTP Management and the TREAT Operations staff for allowing the extensive radiological surveys, as well as the health physics technicians from the INL Radiological Control organization who performed many of the measurements referenced in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Monte-Carlo Simulations of the Nuclear Energy Deposition Inside the CARMEN-1P Differential Calorimeter Irradiated into OSIRIS Reactor

The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR), under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the calorimeter were carried out. A preliminary analysis shows that the numerical results overestimate the measurements by about 20 %. A new approach has been developed in order to estimate the nuclear heating by two methods (energy deposition or KERMA) by considering the whole complete geometry of the sensor. This new approach will contribute to the interpretation of the irradiation campaign and will be useful to improve the out-of-pile calibration procedure of the sensor and its thermal response during irradiations. The aim of this paper is to present simulations made by using MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) for the nuclear heating inside the different parts of the calorimeter (head, rod and base). Calculations into two steps will be realized. We will use as an input source in the model new spectra (neutrons, prompt-photons and delayed-photons) calculated with the Monte Carlo code TRIPOLI-4{sup R} inside different experimental channels (water) located into the OSIRIS periphery and used during the CARMEN-1P irradiation campaign. We will consider Neutrons- Photons-Electrons and Photons-Electrons modes. We will begin by a brief description of the differential-calorimeter device geometry. Then the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements will be introduced. The energy deposition due to the prompt-gamma, delayed-gamma and neutrons, the neutron-activation of the device will be considered. The different components of the nuclear heating inside the different parts of the calorimeter will be detailed. Moreover, a comparison between KERMA and nuclear energy deposition estimations will be given. Finally, a comparison between this total nuclear heating Calculation and Experiment in graphite sample will be determined. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Packed Bed thermal storage for compact Light water-cooled Small Modular Reactors

Nuclear power plants (NPPs) have less flexibility to follow grid load demand than their natural-gas fueled counterparts. Most of the natural-gas power plants can supply peak-loads by adding more fuel, and thus, can generate far more revenue during those peak hours. On the other hand, use of NPPs for peak load following is quite complex due to technical constraints associated with reactor behavior. These technical challenges include the adequate handling of reactivity swings caused by time-varying fuel and moderator temperatures, a higher fuel-failure probability due to thermal-structural cycling, and spatial variations in xenon concentration. Although there are presently some reactors around the world that are operating with flexible load-following capabilities, such operation is restricted to slowly-varying powers, 2-3 times a day, and only up to 80% of the fuel cycle. Thus, a more convenient and effective method to facilitate load following by NPPs would be to integrate energy storage Authors have previously developed models and performed exergy analysis for integrating liquid type TES with LWRs and SMRs. These exergy based models show that technical feasibility index of integrating thermal energy storage is very high. But most of these previous studies include integration based on existing layout of NPPs with heat exchange between storage Heat Transfer Fluid (HTF) and Reactor Coolant (RC). However, most of the Light Water-cooled Small modular reactor (LW-SMR) based NPP designs can not accommodate additional heat exchange equipment. Although power production is only reduced to 1/20 scale, the containment volume in thus NuScale LW-SMR is reduced to {approx} 1/170 scale. As steam is the secondary or working fluid which exits out of the reactor containment in LW-SMRs, the simplest and most practical way to integrate thermal storage with LW-SMRs is to directly store the energy from steam. Indirect heat exchange between steam and another heat transfer fluid during the storage process, and then heat recovery again via indirect heat exchange is inefficient process due to heat transfer resistance. This inefficiency is due to what is called as 'Pinch point' which occurs whenever there is indirect heat transfer between two media where one undergoes phase change and other undergoes change in sensible heat. Another existing popular option is to use a steam accumulator which stores dry or wet steam directly inside the pressure vessels, however, due to large volumetric requirements it is not a viable option. This work proposes a new method of integrating packed bed TES systems with LW-SMRs, where steam produced in the secondary side of LW-SMRs can be injected into the packed bed of inert particles with sufficient thermal conductivity and capacity to efficiently store energy with high energy density. The changes in the load demand will not effect the LWSMR operation and the excess steam will be diverted to TES. When the demand increases this stored energy in hot packed beds can be recovered by injecting cold pressurized water. Although there are some existing packed bed TES designs and models which consider steam as HTFs, these models have not been extended for wide range of process parameters important for this integration. Authors have previously developed a packed bed thermal model for steam injection and validated it using experimental data. This thermal model is extended to assess behavior of packed bed thermal energy storage (TES) system under storage and recovery cycles. Steam is the only heat transfer fluid (HTF) which exits the reactor vessel, so any application such as process heat or TES integration should be evaluated with steam as process fluid. This project will develop a process system integration model of LW-SMR with packed bed type TES using steam as HTF. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

CANDLE Burning Reactor with Plutonium Fuel Start-up Core

The CANDLE burning fast reactor concept makes it possible to utilize natural uranium or depleted uranium with high burnup of up to 40% without repossessing of spent fuel from the reactor. Once the reactor reaches equilibrium burning wave condition, it needs only natural or depleted uranium, although it needs some amount of fissile material for the startup. Studies have been performed for the startup core using enriched uranium and absorbers. It is expected that plutonium obtained by reprocessing of spent fuel from light water reactors (LWR) will be an excellent fuel for the startup core. Using this fuel for the startup core will eliminate the need to produce enriched uranium, as well as make it possible to utilize the plutonium from spent fuel effectively as a useful resource. For effective utilization of plutonium as a resource, a startup core that requires only a small amount of plutonium fuel is preferable. The purpose of this study was to clarify the CANDLE burning fast reactor concept using a minimum amount of plutonium fuel from LWR spent fuel for the startup core by analysis with a simplified reactor model. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Experimental Capability for Preparation and Purification of Molten Salts to Support Advanced Reactor Concepts, Nuclear Hybrid Energy Systems and Dynamic Grid Energy Storage Concepts

High-temperature multi-fluid, multi-loop test facility to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications is being developed at Idaho National Laboratory (INL). In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. The main objectives to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers at prototypical operating conditions, flow, and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for light water reactors at prototypical conditions. Fluoride salt coolants are eutectic binary or tertiary mixtures of fluoride salts with melting points in the range of 320 to 500 deg. C. Fluoride high temperature reactors have reactor outlet temperatures of 600 deg. C or higher for high-efficiency power generation or process heat applications. To address research needs related to salt coolants experimental capability has been added to study thermodynamic properties such as melting point temperature, heat capacity, and thermal conductivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

AREVA Steam Cycle High Temperature Gas-Cooled Reactor Development and Deployment Plan - 18684

The development and deployment timeline for the AREVA SC-HTGR is among the shortest of all Generation IV reactor technologies. Once demonstrated, the SC-HTGR enters an expanding marketplace beyond electricity generation. This enables industrial growth that is today solely captive to fossil fuels, primarily natural gas supply. The SC-HTGR plant will be a hedge that insulates a wide range of process industries from fossil fuel price and supply volatility in coming decades when the SC-HTGR plant design is ready for deployment. Similar to existing light water reactors, the SC-HTGR is largely immune to fuel price swings for its expected 60 year operating life cycle. Approximately 70% of the SCHTGR plant energy cost is driven by the initial capital investment with fuel being less than 20%. This is entirely the opposite of natural gas fueled power plants where about 70% of the energy cost is directly tied to the cost and availability of fuel and the potential volatility this brings with it over a plant life cycle that could span several decades. This paper details the development and deployment strategy envisioned by AREVA for the SC-HTGR plant. The development phase includes completion of the full scale reactor module design and licensing, build and demonstrate the first commercial scale module, get design certification, and launch the commercial fleet deployment. The development plan takes full advantage of a) a large body of successful pre-application licensing interactions with the NRC under the DOE NGNP project, b) ongoing experiments and test results accumulated through the TRISO particle fuel qualification and nuclear grade graphite characterization R and D activities conducted in the U.S. and internationally, and c) operation of several small scale technology demonstration facilities and large scale commercial design projects. The abundance of the historic HTGR reactor technology demonstration projects provides a high degree of confidence that sufficient technical information is available to begin FOAK plant engineering and licensing followed by commercial demonstration of the first reactor module within a next decade. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Modeling a fast spectrum molten salt reactor in a systems dynamics fuel cycles code

Molten salt reactor (MSR) concepts are being actively developed by several private companies, with some touting the fuel cycle benefits of this technology. Simulating the deployment of MSRs and their effect on fuel cycle transitions is the role of a system dynamics fuel cycles tool. However, many current fuel cycles tools have difficulty modeling MSR isotopic variations because they were designed for modeling existing reactors based on discrete fuel elements with batch fueling schemes (i.e., charge and discharge). This work presents ORION fuel cycle modeling results using a specific high power density fast spectrum MSR design. For code verification, a single-stage MSR fuel cycle was set up in ORION and results were compared with the results from the SCALE-based reactor physics model (ChemTriton). A transition scenario was set up in ORION from the current light water reactor fleet to a future fleet of MSRs to study the characteristics of an MSR that affect its performance during transition. Several transition scenarios were set with holds and delays in place that are representative of a sodium fast reactor as modeled in Evaluation Group 23 for the Fuel Cycle Options Campaign Evaluation and Screening Study. Several MSR characteristics lend to its high performance and ease in transition from the current fuel cycle: fissile material is only loaded at initial startup; all fissile material is available as it is bred due to online fuel processing; and fissile material recirculating through the core is briefly held in tanks for processing and has very minimal out-of-core time (on the order of a minutes to a few days). These characteristics are in contrast to a solid-fueled reactor, which requires discharge, cooling, and fabrication of fuel all of which could take a few months to a few years and repeated loading of fresh fuel assemblies. In conclusion, a solid-fueled reactor only discharges a fraction of a core every year.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Scoping Analysis of Pebble-Bed Reactors for the Destruction of the Transuranic Inventory of LWR Spent Nuclear Fuel

With the forecasted increase in the construction and operation of nuclear reactors, there will be a corresponding increase in the quantity of spent nuclear fuel (SNF) that requires long-term storage. In SNF, transuranic isotopes contribute the most to the long-term radiotoxicity of the fuel and pose a proliferation risk. One option that has been explored to address these issues is the removal of the transuranic isotopes from SNF and the conversion of these isotopes into transuranic fuel (TRU fuel). Here, this work sought to determine how effective a micro-modular Pebble-Bed High-Temperature Gas-Cooled Reactor (PB-HTGR); the 10-MW High Temperature Gas-cooled Test Reactor (HTR-10); and a salt-cooled small-modular pebble-bed reactor (PBR), i.e. the generic Fluoride-cooled High-temperature Reactor (gFHR), are at reducing the inventory of transuranic isotopes while still maintaining the intrinsic safety features of the PBR designs, such as negative temperature coefficients of reactivity. Optimized pebble designs utilizing TRU fuel were found for both reactors through the adjustment for the packing fraction of fuel in each pebble. The Axial Zone Equilibrium Modeling (A-ZEM) method was used in this work to help select the optimized pebble design. Once an optimized pebble design was selected and an equilibrium model was produced, the results from the deep burn (DB) HTR-10 and gFHR designs were compared to the results of two models from the literature. While both the DB gFHR and the DB HTR-10 were able to reduce the weapons-usable transuranic inventory, the performance of these reactors did not match that of the small-modular PB-HTGRs in the literature. Therefore, a need was identified for further refinement of the gFHR design using TRU fuel, as the results for this model were more promising than those of the DB HTR-10, which was strongly limited by the high leakage intrinsic to micro-modular PB-HTGRs.

Transuranic fuel↗

DOE OSTI2023

Exclusion and Verification of Remote Nuclear Reactors with a 1-kiloton Gd -Doped Water Detector

To date, antineutrino experiments built for the purpose of demonstrating a nonproliferation capability have typically employed organic scintillators situated as close to the core as possible (typically at a distance of a few meters to tens of meters) and have not exceeded a few tons in size. One problem with this approach is that proximity to the reactor core requires accommodation by the host facility. Water Cherenkov detectors located offsite, at distances of a few kilometers or greater, may facilitate nonintrusive monitoring and verification of reactor activities over a large area. As the standoff distance increases, the detector target mass must scale accordingly. This paper quantifies the degree to which a kiloton-scale gadolinium-doped water Cherenkov detector can exclude the existence of undeclared reactors within a specified distance, and remotely detect the presence of a hidden reactor in the presence of declared reactors, by verifying the operational power and standoff distance using a Feldman-Cousins-based likelihood analysis. A 1-kton scale (fiducial) water Cherenkov detector can exclude gigawatt-scale nuclear reactors up to tens of kilometers within a year. In conclusion, when attempting to identify the specific range and power of a reactor, the detector energy resolution is not sufficient to delineate between the two.

42 ENGINEERING↗

DOE OSTI2017

VERA Core Simulator methodology for pressurized water reactor cycle depletion

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclide transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗