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DOE OSTI2016

Microstructural Characterization of In-reactor Corrosion Tested Hastelloy N{sup R} and 316 Stainless Steel

The success of the Molten Salt Reactor Experiment (MSRE) at the Oak Ridge National Laboratory (ORNL) in the period 1950's-70's has rekindled the interested in molten salt-cooled reactors. The most recent form of this type of reactor, the Fluoride salt-cooled High-temperature nuclear Reactors (FHRs) is emerging as a leading reactor concept for the next generation nuclear reactors because it offers, among other benefits, a high degree of passive safety, high thermal efficiency, and low spent fuel. Since 2011, two three-year Integrated Research Projects (IRPs) were awarded by the U. S. Department of Energy (DoE) to the Massachusetts Institute of Technology (MIT) in partnership with the University of California, Berkeley and the University of Wisconsin-Madison to develop a 'path forward' for the development of the FHR. One primary challenge in these projects is to evaluate the corrosion and compatibility of structural alloys in the potential primary coolant of molten Li{sub 2}BeF{sub 4} (FLiBe) salt at 700 deg. C in a neutron irradiation environment. In support of structural materials development for the FHR, the static in-reactor corrosion tests of two promising alloys, nickel-based Hastelloy N{sup R} and 316 stainless steel, in {sup 7}Li-enriched molten FLiBe have been conducted in both metal-lined and graphite crucibles in the MIT nuclear research reactor in 2013. This work focuses on the microstructural characterization of the in-reactor corrosion tested alloys for understanding of their performance in high-temperature molten FLiBe salt. Our preliminary results indicate significant thermal effects, graphite effects, and irradiation effects on the alloys' performance in high-temperature molten FLiBe salt, particularly for the in-reactor corrosion tests. The neutron irradiation induces tritium generation in molten salt which could combine with fluorine ions to form corrosive HF to dramatically attack alloys. Additionally, the neutron irradiation introduces structural defects into alloys which probably alter alloys' corrosion resistivity to high temperature molten FLiBe salt. The effect of graphite on the alloys' corrosion in a nuclear reactor is also possibly different from the out-of-reactor corrosion. All these issues will be better understood by characterizing the microstructure of in-reactor corrosion tested alloy samples in our near future plans. The experimental data in this study provides valuable input for evaluating structural materials performance in the FHR. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2015

Of the final end state for decommissioning of reactors at SRS: collaboration between stakeholders, regulators, and the Federal Government - 15279

The Savannah River Site (SRS) is an 802 square-kilometer (310 square-miles) United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina. Nuclear weapons material production began in the early 1950's, utilizing five (5) production reactors. In the early 1990's all SRS production reactor operations were terminated. The first reactor closure end-state declaration was documented in a Comprehensive Environmental Response and Compensation and Liability Act (CERCLA) Early Action Record of Decision. The decision for the final closure of the 29,543 square-meter (318,000 square-feet) 105-P Reactor was determined to be in situ decommissioning (ISD). ISD is an acceptable and cost-effective alternative to off-site disposal for the reactor building, which will allow for consolidation of remedial action wastes generated from other cleanup activities within the P Area. ISD is considered protective by the regulators, U.S. Environmental Protection Agency (USEPA) and the South Carolina Department of Health and Environmental Control (SCDHEC), public and stakeholders as waste materials are stabilized / immobilized, and radioactivity is allowed to naturally decay, thus preventing future exposure to the environment. Stakeholder buy-in was critical in the up-front planning in order to achieve this monumental final decision. Numerous public meetings and workshops were held in two different states (covering a 321 kilometer (200-mile) radius) with stakeholder and SRS Citizens Advisory Board (CAB) participation. These meetings were conducted over an eight- month period as the end-state decision making progressed. Information provided to the public evolved from workshop to workshop as data became available and public input from the public meetings were gathered. ISD is being considered for the balance of the four (4) SRS reactors and other hardened facilities such as the chemical Separation Facilities (canyons). Closure of the SRS P Reactor Area was precedent setting and sets the stage for closure of other reactors and hardened facilities. The DOE, USEPA, and SCDHEC engaged in active, up-front, and timely involvement with the SRS. The SRS CAB, elected officials, and citizens of South Carolina and Georgia met to discuss the P reactor area closure and associated risks. Early public involvement gave SRS and the regulators assurance that the selected decision was valid and would not be overturned. Public input improved the process and ensured that the important decision on the final end-state of the reactor facilities was not made in isolation. ISD of P Reactor will provide adequate protection of human health and the environment and the lowest cost since short-term risk is minimized to remedial workers from exposure to contaminated equipment and facilities by leaving the reactor vessel in place allowing for radiological decay over time. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Coupled CFD-DEM Analysis of Molten Salt-Cooled Pebble-Bed Reactor Experiment

The development of new nuclear reactor designs is an area of ongoing investigation, and a focus of dedicated work for various laboratories around the world. Despite the steadfast and reliable service of Light Water Reactors (LWR), the ideal power reactor, with a combination of high operational temperature, passive safety, proliferation resistance, economic build and waste minimization remains elusive. One promising design that might address these diverse, and at times contradictory requirements, is the Molten Salt Reactor (MSR). In its most radical form, a liquid fueled (LF) MSR has the nuclear fuel dissolved in the molten-salt primary coolant itself, as demonstrated by Oak Ridge National Lab (ORNL) in 1965 to 1969. An alternative fuel arrangement is to retain conventional, solid fuel elements. A proposed method is to implement pebble fuel elements, as have been used in previous gas-cooled reactors, with the gas coolant being replaced with a molten salt. This is referred to as pebble-bed, molten-salt reactor (PB-MSR). The spherical fuel pebbles used within PB-MSRs contain TRISO particles dispersed throughout the pebble's graphite matrix. Many pebbles are constrained to form a porous bed within the core region of a reactor. Coolant molten-salt passes through the pebble bed to remove heat produced due to the nuclear fission taking place within the fuel pebbles' TRISO particles. Various benefits are apparent with the PB-MSR design. When considering the choice of fuel elements, they are a geometry that has been produced and utilized in a similar way in the past, within gas-cooled reactors. The fuel pebbles are also able to be added to, and removed from, the core during operation, enabling on-line refueling. When considering the coolant choice, molten-salts have much higher boiling points (1703 K at 1 atm. for FLiBe) than conventional coolants such as water (630 K at 150 atm.), and are capable of reaching these temperatures at atmospheric pressure. This in turn removes the need for high pressure containment that is present in light water reactors and gas-cooled reactors. The future implementation of new reactors, and the search for improvements in efficiency and safety in current designs, require the advancement of simulation tools, which this work ultimately aims to address.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Particle-based high-temperature thermochemical energy storage reactors

Solar and other renewable energy driven gas-solid thermochemical energy storage (TCES) technology is a promising solution for the next generation energy storage systems due to its high operating temperature, efficient energy conversion, ultra-long storage duration, and potential high energy density. Experimental and theoretical studies suggest that the respective gravimetric and volumetric TCES energy storage densities vary from 200 to 3000 kJ kg –1 and 1–3 GJ m –3 . Solar radiation or heat generated from electric furnaces powered by renewable electricity can be stored in the form of chemical energy through endothermic reactions, while the stored chemical energy can be converted to thermal energy via an exothermic reaction when needed. The design of highly effective reactors requires a deep understanding of materials, thermodynamics, chemical kinetics, and transport phenomena. At time of writing, TCES reactors are yet to be deployed at commercially relevant scales, leaving a substantial gap between development efforts and commercial feasibility. Therefore, this review aims to examine the state-of-the-art design and performance of particle-based TCES reactors with different reactive materials. Fundamentals related to TCES reactive materials, reaction conditions, thermodynamics and kinetics, and transport phenomena are reviewed in detail to provide a comprehensive understanding of the reactor design and operation. Five major types of TCES reactors have been comprehensively reviewed and compared, including fixed, moving, rotary, fluidized, and entrained bed reactors. Most reported prototype reactors in the literature operate at lab scale with thermal inputs below 40 kW, and scaled TCES reactors (e.g., at megawatt level) are yet to be demonstrated. The nominal reactor operating temperatures range from 300 to 1500 °C, depending on the selected chemistry, reactive material, and heat sources. To evaluate their designs, the reactors are assessed in aspects of performance, cost, and durability. Discrepancies in performance indicators of energy storage density, extent of reaction, and various energy efficiencies are highlighted. The scale-up of reactors and power block integration, which hold the key to the successful commercialization of TCES systems, are critically analyzed. Furthermore, advanced materials (both reactive materials and ceramic reactor housing materials), effective particle flow control, advanced modeling tools, and novel system design may bring significant improvement to the energy efficiency, storage density and cost competitiveness of particle-based TCES reactors.

25 ENERGY STORAGE↗

DOE OSTI2019

Cardinal: A Lower Length-scale Simulator for Fluoride Cooled High Temperature Reactors

The new DOE-NE “Center of Excellence for Thermal-fluids applications in Nuclear Energy” inaugurated in April 2018 considers and researches novel new solution strategies for historically challenging flow issues that still plague the current fleet of deployed Light Water Reactor (LWR) nuclear reactors as well as predicting various fluid flow and fluid related issues with advanced reactor technologies. Our advanced thermal-fluids research and development approach synergistically combines three natural, though overlapping, length and time scales in a hierarchal multi-scale approach to avoid the temptation and pitfalls of attempting to develop a single “solve all” algorithm for physical fluid flow problems that will span 109 in spatial and temporal scales. The demonstrate the multi-scale philosophy of the center we focus first on Fluoride Cooled High Temperature Reactors (FHRs), and in particular on the Berkeley’s PB-FHR design Mark-I design. The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxiliary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids - Flibe - from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. The PB-FHR reactor relies on a pebble bed approach and pebble bed reactors are in a sense the poster child for multiscale analysis. The lower length-scale simulator for pebble reactor cores comprises three physics: neutronics (OpenMC), thermal-fluids (Nek5000) and fuel performance (BISON). As part of center of excellence ongoing research efforts we have developed “Cardinal”, a new tool platform for lower length-scale simulation. Cardinal tightly couples all three physics and leverages advances in MOOSE such as the MultiApp system and the concept of MOOSE-wrapped Apps. Moreover it is designed from the ground-up to scale on massively parallel architectures and perform well on world-class super-computing architectures. In this report we describe the design of Cardinal and present a demo of the application of Cardinal to an FHR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Integration of Control Methods and Digital Twins for Advanced Nuclear Reactors

Advanced nuclear reactors offer a new set of features to energy generation, due to their ability to adapt to variable energy demand, operate autonomously, be deployed in rural locations and monitored remotely, afford compact size and lower power ratings, and rely on novel technologies to achieve safer operations. Thus, a requirement for the success of these reactors is the use of intelligent forms of control to track changing power demands, make autonomous decisions, and reduce the need for human involvement. Regulatory requirements pertaining to control of nuclear reactors could be met via historical means of control; however, these are not expected to enable the level of highly autonomous operations desired in advanced nuclear reactors. Historical control methods rely on both logical and high-performance (HP) control. These two types of control are usually used separately, with a human element being introduced whenever decisions are cascaded from one science to another. AI/ML control, on the other hand, can replace the human element in the current U.S. fleet of nuclear power plants (NPPs) by acting as a supervisory optimizer that understands the plant internal/external variables in order to make control decisions, and can easily handle non-linear and multi-input/multi out (MIMO) decisions—another requirement for advanced nuclear reactors that could be difficult to handle via logical and HP control. Because of the harsh operating environments produced in advanced reactors, resulting in the frequent failure of sensors and other types of equipment, and considering the lack of operating history for advanced nuclear reactors, control of advanced nuclear reactors would necessitate relying on a model that can track and adapt to the actual process (i.e., a digital twin). This digital twin can make approximations when knowledge and data are unavailable and would evolve as more knowledge is gained. The reactor control must also be risk-informed to account for the high-consequence nature of advanced reactors. This report introduces a high-level (i.e., not method- or process-specific) integration of the three different control and digital twinning methods able to meet the requirements for advanced nuclear reactors. These methods could be applied during both the operational and design stages of these reactors. The aim is to demonstrate how each method interfaces with and highlights enabling solutions necessitated by the unique features of advanced nuclear reactors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2026

Radiological Releases from Novel Fuel Forms in Advanced Reactors During Severe Accidents for Consequence Analyses

Various advanced reactor developers are exploring the potential for reductions in the size of physical security forces and emergency planning zones. These reductions are based on robust fuel forms and inherently safe reactor designs. However, such reductions in physical protection measures could increase the risk of sabotage. To assess the possibility of reducing these measures, sabotage-induced radiological consequence analyses were carried out. These analyses considered accident scenarios that were beyond design basis accidents and overly conservative (Shah, 2025a; Shah, 2025b; Shah and Hartanto, 2026), yielding very large release fractions. These fractions, which can be used to evaluate physical protection and emergency planning requirements, have been crudely determined and applied as demonstrations for a sodium-cooled fast reactor (SFR) (Shah and Hartanto, 2025a), a high-temperature gas-cooled reactor (HTGR) (Shah and Hartanto, 2025b), a heat pipe–cooled reactor (HPR) (Shah and Hartanto, 2025c), and a molten salt–cooled reactor (MSR) (Shah et al., 2026). A Sandia National Laboratories (SNL) team used MELCOR—a fully integrated severe accident analysis code—to demonstrate the code’s capability to analyze advanced (i.e., not light water–cooled) reactors (including a fluoride salt–cooled high-temperature reactor [FHR]) and calculate radiological releases to the environment during severe accidents (Wagner et al., 2022a, 2022b, 2022c, 2023a, and 2023b). Although the analyses were carried out to demonstrate MELCOR’s growing capability, the release source terms were estimated for advanced reactors, providing valuable insights into the accident progression and radiological releases. These findings from prior SNL studies, including estimated source terms and related sensitivity studies, were leveraged to derive source terms for postulated sabotage-induced accidents. Insights from these sensitivity studies informed the scaling of SNL’s estimated source terms for the defined accident scenarios. The derived release fractions for the severe accident scenarios for the respective reactor designs can be used to perform more nuanced dose consequence analyses to evaluate the reactors’ physical protection and emergency planning zone requirements. These analyses are in accordance with the risk-informed, performance-based approach proposed under 10 CFR Part 53. This study builds on the prior source term analyses and associated sensitivity studies by SNL to derive time-dependent and design-informed release fractions. Section 2 describes the diverse advanced reactor designs analyzed by the SNL team. Section 3 discusses the severe accident analyses, the release fractions calculated, and the limitations and assumptions of the demonstration project. Section 4 presents the release percentages derived for the hypothetical sabotage-induced severe accidents at the advanced reactors. Section 5 summarizes the study’s findings and conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Assessment of Survey Results from Advanced Reactor Industry Domain Experts on Nonlinear Soil-Structure Interaction Analysis Software Verification and Validation

The seismic load case has a significant impact on the design and cost of nuclear power plants. Given the need to substantially reduce cost, advanced reactor designers are looking to leverage numerical tools that enable seismic analysis of an integrated vessel/support/structure/soil system. Modern nonlinear analysis tools provide a solution, capturing dynamic coupling between components (including soil-structure interaction (SSI)) concurrent with nonlinear behavior in one or more parts of the system. Although these methods have a rich history of technical development and implementation, software quality assurance (SQA) following nuclear industry standards remains a significant burden for those wishing to adopt such methods for advanced reactor design and licensing. To reduce the SQA burden, this project will develop the guidance for SQA verification and validation (V&V) of coupled nonlinear SSI analysis tools, to include a test matrix of software features and test problems to support commercial grade dedication (CGD). The guidance is intended to be technology-neutral: supporting designers of all advanced reactors. To maximize the value of the project to reactor designers, the project team performed focused surveys of and interviews with advanced reactor designers. Additionally, the project surveyed members of the broader industry involved in reactor design and licensing. The aggregated industry feedback consists of written survey responses, live polling responses, focused interviews, and informal feedback provided after outreach presentations, collectively referred to as “survey results”. Herein, the survey results are reported, reviewed, and assessed to inform follow-on project activities. The survey results confirmed the familiarity of the industry with the coupled nonlinear SSI analysis. They also affirm the project need based upon the expressed intent to (1) incorporate nonlinear features and (2) pursue the coupled nonlinear SSI analysis to reduce seismic demands and construction costs. All reactor designers indicated their intent to incorporate two or more nonlinear features, and all reactor designer respondents opined that the coupled nonlinear SSI analysis would allow for the optimization of the reactor design and construction. However, the programmatic challenges, whether real or perceived, present a significant barrier to reactor designers. The barrier most commonly identified by the reactor designer population was regulatory risk, with 70% citing this as a reason not to pursue the approach. The perceived regulatory risk identified by the reactor designers underscores the importance of regulator engagement and dialogue in this project. Additionally, half of the reactor designers identified cost and lack of guidance as a deterrent. The survey results also provide insights to tailor specific aspects of the guidance document and test matrix. The project plans to prepare both a formal referenceable guidance document and a collaborative, web-based test matrix and problem set. The project will focus on more complete test problem definitions for the prioritized nonlinear features over shallower problem descriptions for a larger set of features. The nonlinearities prioritized as high based upon survey feedback include fluid-structure interaction and seismic isolation and energy dissipation devices. The nonlinearities prioritized as intermediate include nonlinear geomaterials, nonlinear concrete, nonlinear steel, and interface/contact nonlinearity. Deep embedment and the associated nonlinear phenomena are assigned the lowest priority based upon survey feedback.

42 ENGINEERING↗

DOE OSTI2017

Versatile Irradiation Test Reactor User Needs Assessment

An assessment of potential test reactor industry user needs has been completed. The needs of DOE R&D programs have also been assessed and included in this document. A process of discussions with industry representatives under non-disclosure agreements (NDAs) was undertaken during the last half of the calendar year. This activity was performed by staff members of the Idaho National Laboratory (INL), Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL). Contributors to this assessment were selected from a list of representative organizations. As the duration for completion was short, all reactor vendors were not included and some chose not to participate. However, the contributors represent well-established reactor vendors and reactor developers working to bring new designs to the reactor market. Equally important was the breadth of the reactor technologies that are included: sodium cooled fast reactors, lead cooled fast reactors, high-temperature gas-cooled reactors, light water reactors and molten salt reactors. Throughout this report the term Molten Salt Reactors is used to represent all of the variants of molten-salt cooled reactors with either dissolved or solid fuel: fast, thermal, dissolved and prismatic fueled.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2015

Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development at the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to the delayed gamma rays. In this paper we describe experiments in each of the three reactors and how we estimate delayed gamma rays with MIC measurements. The results and perspectives are discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Study of Reactor BN-350 Wastewater Impact on the Environment

The fast neutron reactor BN-350 had a dual purpose: steam production and plutonium accumulation. The steam produced in the reactor was used to generate electricity (up to 150 MW) and desalination of seawater (up to 120,000 tons/day). The BN-350 reactor became operational in July 1973; its design operation lifetime was 20 years. After expiration of the reactor design lifetime in July 1993, its subsequent operation was continued according to the annual decisions on extension of the reactor operation, by one year. The reactor has been idle since December 1998. On April 22, 1999 Resolution No. 456 'About decommissioning of the BN-350 reactor in Aktau, Mangistau region' was adopted by the RK Government. Thus, the total operation period of the reactor was more than 25 years. One of the important issues related to the long-term operation of a radiation-hazardous facility, such as the BN-350 reactor, is to assess its impact on the environment. The necessity and urgency of detailed studies of the radiation situation is explained by the location of the Caspian Sea and Aktau city in the area of the BN-350 reactor's potential impact. This paper describes the preliminary results of the evaluation of the BN-350 reactor's effect on the level of radionuclide contamination of surface water, including the Caspian Sea via the mechanism of radionuclide discharge in wastewater. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Comparison of Irradiated TRISO Fuel Radioactivity from Multiple Advanced Reactor Designs

The SCALE code system was used to model, deplete, and compare several different TRISO-fueled reactor designs: a helium-cooled prismatic reactor, a helium-cooled pebble-bed reactor (PBR), and a Fluoride-Lithium-Beryllium (FLIBE) molten-salt-cooled PBR. The purpose of this comparison was to understand how differences in the reactor designs affect the radioactivity of the fuel after discharge and whether those differences are significant. First, the various reactor designs were build and depleted in the TRITON module for each design and fuel enrichment. Then, the TRITON outputs were used to create burn-up dependent reactor libraries. These libraries were then used by ORIGEN to determine the activities of discharged fuel for each reactor, which were compared to generic Westinghouse 17x17 fuel. Overall, the results showed that short term decays are dominated by reactors with higher operating powers, and the reactor type, initial fuel enrichment, and maximum burn-up are of only secondary importance. Although this analysis only focuses on activities in becquerels, these dependencies are consistent with the expected behavior of decay heat. However, analysis of long- term time periods post-irradiation shows that the reactor type and maximum burn-up have strong impacts on the activities; initial fuel enrichment has a secondary impact while operating power is inconsequential. These results would be useful for analyses, such as dose assessment and modeling in post- release scenarios; normal fuel handling operations; and spent fuel transport, storage and disposal. Of particular interest, the results in this report show that analyses that focus on spent nuclear fuel of advanced reactors need to consider each parameter carefully. Unsurprisingly, if the correct operating power is not used in short term analyses, the results will not be correct. Perhaps unexpectedly, however, if the correct reactor type is not used, then the long term results will also be incorrect, especially for areas such as permanent disposal. Even though this report focuses on the total activity of nuclear fuel, it provides initial results on the effects of various input parameters and also provides a framework to extend the work into other analyses of spent fuel from advanced reactors, especially those employing TRISO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Nuclear Data to Reduce Uncertainties in Reactor Antineutrino Measurements

The large quantities of antineutrinos produced through the decay of fission fragments in nuclear reactors provide an opportunity to study the properties of these particles and investigate their use in reactor monitoring. The reactor antineutrino spectra are measured using specialized, large area detectors that detect antineutrinos through inverse beta decay, electron elastic scattering, or coherent elastic neutrino nucleus scattering; although, inverse beta decay is the only demonstrated method so far. Reactor monitoring takes advantage of the differences in the antineutrino yield and spectra resulting from uranium and plutonium fission providing an opportunity to estimate the fissile material composition in the reactor. Recent experiments reveal a deviation between the measured and calculated antineutrino flux and spectra (the reactor anomaly) indicating either the existence of yet undiscovered neutrino physics, uncertainties in the reactor source term calculation, incorrect nuclear data, or a combination of all three. To address the nuclear data that impact the antineutrino spectrum calculations and measurements, an international group of over 180 experts in antineutrino physics, reactor analysis, detector development, and nuclear data came together during the Workshop on Nuclear Data for Reactor Antineutrino Measurements (WoNDRAM) to discuss nuclear data needs and achieve concordance on a set of recommended priorities for nuclear data improvements. Three topical sessions focused on the reactor source term, the antineutrino spectrum, and the detector response, provided a forum to gain consensus amongst the participants on the most important data improvements to address two goals: 1) understand the reactor anomaly and 2) improve the ability to monitor reactors using antineutrinos. This report summarizes the outcomes of the workshop discussions and the recommendations for nuclear data efforts that reduce reactor antineutrino measurement uncertainties.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Assessment of Sensor Technologies for Advanced Reactors

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope for fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.

Korsah, Kofi↗

DOE OSTI2019

Enhancing Research Reactor Safeguards Through the Use of Mailbox Declarations

There are historic examples in several States of research reactors being misused to produce undeclared nuclear material. However, misuse of a State’s research reactor has typically only been discovered after further investigation into the State’s non-compliance related to other undeclared activities. Light water research reactors are typically designed with sample irradiation ports, where the introduction of fertile material can illicitly produce fissile material. Additionally, some light water reactors continue to be fueled by highly enriched uranium, which is a more significant safeguards concern by the International Atomic Energy Agency (IAEA). Heavy water and graphite reactors can be used for illicit plutonium production, which is of particular proliferation concern. Because research reactors are designed for flexible operations and small research reactors are less frequently inspected by the IAEA than other nuclear facilities (depending on power and fuel type, it could several years between inspections), the likelihood of timely detection by the IAEA of misuse of the research reactor is low. During a 2013 workshop on safeguarding research reactors, the United States, France, and the United Kingdom made recommendations to the IAEA on how safeguards could be strengthened for such facilities. One of the recommendations was to assess the use of “mailbox” declarations coupled with short notice random inspections or unannounced inspections by the IAEA at research reactors. An electronic mailbox declaration refers to the use of a secure information repository into which an operator submits operating reports. The near real time submittal of information enables the IAEA to better implement randomized short notice inspections. These mailboxes are not used for collecting and submitting State reports to the IAEA, but for collecting and transmitting operator data. The use of mailboxes, in conjunction with short notice random inspections, is in routine use for some other facility types. This paper assesses the potential for the practical application of mailbox declarations for various research reactor types and how the approach could be effectively used. Potential diversion scenarios are described and operator data records are discussed in connection to mailbox declarations, including discussion of how the data could indicate research reactor misuse.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Variable Nuclear Electricity with Base-Load Reactor, Closed Brayton Power Cycle and Firebrick Resistance-Heated Energy Storage (Fires)

Power cycle choices are determined by markets and available technologies. Two changes are occurring in the electricity sector. First, the world is beginning to move away from fossil fuels because of concerns about climate change. What replaces fossil fuels for variable electricity production? Second, there have been extraordinary technological advances in gas turbine technology. The changes in market requirements and technology suggest the need to reexamine nuclear power cycles. One option is to couple high-temperature base-load nuclear reactors to closed Brayton cycles (helium, nitrogen, etc.) with Firebrick Resistance-Heated Energy Storage (FIRES). The reactor operates at base-load but not the power cycle. When electricity prices are low, firebrick is electrically heated to high temperatures using base-load electricity and/or electricity from the grid. The gas turbine can use nuclear heat for base-load operation and produce added peak electricity by sending the compressed heated gas through hot firebrick after nuclear heating to further raise power cycle gas temperatures and power output from the turbine-creating a nuclear plant with a thermodynamic topping (NUTOP) cycle. If such cycles can be developed, they would increase plant revenue and enable a low-carbon electricity grid. Such cycles can be coupled to Fluoride-salt-cooled High-temperature Reactors (FHRs), molten salt reactors (MSRs), high-temperature gas-cooled reactors (HTGRs), sodium-cooled fast reactors (SFRs) and lead-cooled fast reactors (LFRs). We discuss (1) market changes, (2) previous work on Nuclear Air-Brayton Combined cycles (NACC) with base-load reactors and variable power to the electricity grid and (3) options for closed Brayton cycles with variable power while the reactor operates at base load. Market (customer) requirements ultimately drive selection of power cycles. A low-carbon electricity grid requires dispatchable nuclear electricity to replace the traditional role of fossil fuels in the electricity grid. While reactors can be operated at part load, this is expensive for a capital-intensive low-operating-cost technology. Brayton power cycles with FIRES have the potential capability to provide that variable electricity while the reactor operates at base load to (1) improve economics and (2) enable a zero-carbon electrical grid. Open Brayton cycles have been investigated. Work has now started on developing closed Brayton cycles with FIRES. Only limited work has been done to date has been done for closed cycles with FIRES. (authors)

42 ENGINEERING↗

DOE OSTI2017

NBSR reactor control room upgrade - 215

The maintenance and modernization of the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) reactor, known as the NBSR, is an ongoing process to endure until the eventual decommissioning of the reactor. As part of the renovation effort, the reactor control system is to be upgraded. Following some initial outsourcing efforts, the NBSR reactor control system upgrade has been restructured to be implemented by in-house means. The beginning stage of the upgrade involved the development of a simulator environment based on a PC system and digitization of the majority of nuclear reactor monitoring data. The NBSR reactor surveillance data consists of approximately thirteen hundred input/output points, including analog/digital sensor signals, signal conditioners, and relay logic. The aim of the current phase is to implement a new reactor data acquisition and display system (RDS) and lay out the infrastructure for the upgrade of the reactor control system (RCS). Key design criteria are to implement a reliable, redundant, diverse, modular and candid system where any complex digital manipulations are taken outside of the control room. Basically, the data analytics and custom codes are separated from reactor operation control and decision making logic. Furthermore, the new RDS is designed such that it could be maintained internally by NCNR personnel. The RDS is unique in design and modularity, minimizing custom codes and testing, making it ideal for research reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗