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NASA NTRS1987

Small reactor power system for space application

A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

Shirbacheh, M.↗

DOE OSTI2017

LOCA Analysis for the NIST Research Reactor

The research reactor (aka NBSR) at the National Institute of Standards and Technology (NIST) is a heavy water cooled, moderated, and reflected tank-type research reactor that operates at atmospheric pressure and a design power of 20 MW. Core power is generated in 30 fuel elements, each with 17 fuel plates in an upper section and a separate lower section. Power is removed using forced convection and heat exchangers in the cooling loop. The licensing analysis of the loss-of-coolant accident (LOCA) in the NBSR was addressed in the current Safety Analysis Report (SAR). The SAR discusses the low probability of a significant pipe break: 'the main piping is located in protected areas, system pressures are low, and flow rates are small so that wear is not an issue.' Therefore, the probability of a large break, including a double-ended guillotine break, is extremely low. For smaller breaks where the operator has time to take action, procedures are in place to 1) mitigate a loss of water by tripping pumps and closing control valves to isolate the reactor vessel after the falling water level in the vessel is detected by instrumentation, and 2) ensure that emergency cooling water continues to flow for as long as needed. The SAR also refers to an analysis of why there is sufficient water without forced cooling to cool the fuel elements (FEs) in a LOCA if the flow channels between two fuel plates (or a fuel plate and an outside plate) remain filled with water. That analysis was based on a flooding-limited critical power but assumed that reactor trip had taken place. A more recent study showed that some scenarios may lead to the draining of water from the coolant channels. In these cases, the cooling on the inside of the fuel element is from emergency cooling water that falls only along the inner surface of one of the side plates in the fuel elements; the remaining surfaces within the fuel element being exposed to gas. The analysis therein was done for both the present high-enriched uranium fuel (HEU) and the proposed low-enriched uranium fuel (LEU). The LOCA analyses done previously assumed that there is sufficient cooling during the short-term, i.e., before the water level reaches the top of the fuel plates and begins to drain from the coolant channels or before decay heat becomes low enough such that the flooding-limited critical power shows that there is no problem. The present study is to determine if that assumption is correct, i.e., to consider the situation immediately after the break, before and after reactor trip when the power is still relatively high.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Final report on assessment of molten salt corrosion testing of unirradiated and ion irradiated advanced manufactured high entropy alloys

Generation IV reactors and future fusion reactor designs have led to more demanding materials performance requirements due to their increased operating temperatures, corrosive coolants, and increased radiation doses compared to the current light-water reactor fleet. Among the innovative nuclear technologies under development, molten salt reactors stand out for their potential to offer superior fuel utilization, intrinsic safety characteristics, and economic viability. Of the proposed Generation IV designs, the gas fast reactor operates at 450 to 850°C and the molten salt reactor operates at 565 to 850°C, with the molten salt reactor design needing molten salt corrosion resistant materials [1, 2]. These increased temperatures and more extreme corrosion environments necessitate higher material performance, such as creep strength, radiation-tolerant microstructures, corrosion resistance, and high-temperature tensile properties. Hastelloy-N, a nickel-based alloy with additions of molybdenum and chromium, has been successfully employed to contain molten fluoride salt at temperatures up to 705°C. However, Hastelloy-N becomes embrittled upon neutron irradiation, primarily due to the accumulation of helium produced by (n,a) transmutation reactions. Furthermore, the corrosive nature of molten fluoride and chloride salts presents a formidable challenge, as these salts can react with and dissolve alloying elements such as Cr, Mo, and Fe, leading to selective leaching, loss of protective oxide layers, and accelerated degradation. High entropy alloys (HEAs) and refractory high entropy alloys (RHEAs) have emerged as a prominent area of interest, due to their ability to achieve tailored chemical compositions for specific applications. Unlike conventional alloys, HEAs are characterized by having multiple principal elements in equimolar or near equimolar ratios, leading to an unconventional alloying strategy [3]. This alloying strategy is believed to promote unique properties, such as single-phase stabilization of chemically compatible elements, lattice distortion effects due to atomic radius differences, and proposed sluggish diffusion effects. For extreme-environment applications, RHEAs have garnered much research interest because of the possibility of creating relatively ductile materials that can operate in extremely high-temperature environments, beyond the operating temperatures where other Ni-based superalloys begin to lose strength [4-6]. Idaho National Laboratory (INL) initiated a joint international effort with the Czech Republic to explore the feasibility of manufacturing HEAs for high-temperature nuclear applications using advanced manufacturing. This effort was funded at INL by the United States Department of Energy's Office of Nuclear Energy under the Advanced Reactor Technologies and Advanced Materials and Manufacturing Technologies (AMMT) Program. The HEAs were specifically designed for the corrosive and irradiation environments experienced in gas-cooled fast reactors, molten salt reactors, and fusion power. These alloys have been manufactured by multiple processes to determine the impact of manufacturing processes on the performance of the alloys in corrosive and irradiation environments. Preliminary molten salt corrosion testing showed that equimolar MoNbTiV and MoNbTi alloys exhibit exceptional performance, with arc-melted variants demonstrating only minimal degradation after 1000 hours of exposure to molten chloride salt at 700°C. Conversely, Nb2TiVZr2 showed significant molten salt corrosion susceptibility and microstructural instability during high-temperature molten salt exposures, and was, therefore deemed unfit for molten salt reactor applications. The MoNbTiV, MoNbTi, and Nb2TiVZr2 alloys were further evaluated through ion irradiation experiments conducted at the Michigan Ion Beam Laboratory at the University of Michigan. The microstructural stability and the evolution of irradiation-induced defects were characterized to assess the irradiation resistance of each of these alloys.

36 - MATERIALS SCIENCE↗

DOE OSTI2021

Design and operational considerations of catalytic membrane reactors for ammonia synthesis

Production of ammonia using hydrogen derived from renewable electricity instead of hydrocarbon reforming would dramatically reduce the carbon footprint of this commodity chemical. Novel technologies such as catalytic membrane reactors may potentially be more compatible with distributed ammonia production than the conventional Haber-Bosch process. Here, a reactor model is developed based on integrating a standard industrial iron catalyst into a catalytic membrane reactor (CMR) equipped with an inorganic membrane that is selective to NH 3 over N 2 /H 2 . CMR performance is studied as functions of wide ranges of membrane properties and operating conditions. Conversion and ammonia recovery are dictated principally by the ammonia permeance, and the benefits by using membranes become significant above 100 GPU = 3.4 × 10 –8 mol m –2 s –1 Pa –1 . To be effective, the CMR requires a minimum selectivity for ammonia of 10 over both nitrogen and hydrogen, and purity scales with the effective selectivity. Increasing the pressure of operation significantly improves all metrics, and at P = 30 bar with a quality membrane ammonia is almost completely recovered, enabling direct recycle of un-reacted hydrogen and nitrogen without need for recompression. Temperature drives conversion and scales monotonically without thermodynamic limitations in a CMR. Alternatively, the temperature may be reduced as low as 300°C while achieving conversion levels surpassing equilibrium limits at T = 400°C in a conventional reactor.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2017

Verification of severe accident management guideline (SAMG) entry condition for OPR1000 - 367

The current severe accident prevention and mitigation strategies for OPR 1000 have emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs), respectively. The main objective of SAMG is to prevent a release of radioactive materials from releasing into environment and mitigate severe accident phenomena during severe accident. The current SAMG entry condition is met when the core exit temperature (CET) reaches 650 deg. C in OPR 1000. Peak cladding temperature (PCT), which represents core integrity more precisely than CET does, cannot be used since it is not measurable in the main control room (MCR). Current SAMG entry condition for every kind of accident scenarios is not reasonable, since entry condition is determined without considering operator action time. Therefore, SAMG entry condition need to be verified with consider operator action time. Thus, an analysis methodology for SAMG entry condition was developed to verify the SAMG entry condition. A severe accident DB were also developed in order to analyze the SAMG entry condition. The analysis results of DB show that available operator action time from the SAMG entrance to RV fail is shorter than the required time in most cases. As a result, the proper SAMG entry condition was suggested through the results of DB analysis. When the SAMG entry condition is changed from CET to suggested condition, the available action time is increased about 350 to 4300 sec. Also, the suggested entry condition was verified through the simulation which considered operation action time and reflecting the suggested entry condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Radiochemical Methodology Applied To Determination of gamma-Emitting Radionuclides In Waste Samples From IEA-R1 Reactor - 17265

The Brazilian reactor IEA-R1 has been in operation since 1957. It is an open pool-type nuclear research reactor, operating at 2 to 5 MW that uses water as coolant, moderator and biological shield. Polypropylene filter cartridges are part of the water treatment system of the reactor and become radioactive as they remove suspended solid material from the cooling water during the operation of the reactor. Annually, the replacement of the filters from the reactor's water treatment system generates about 36 units. The development of a method for the characterization of the radionuclides present in the filters is required for routine operation in the Radioactive Waste Management Facility (RWMF) in order to determine the radioisotopic inventory of the waste packages. Wastes from nuclear reactors contain transuranic elements, activation and fission products, only few of them emitting gamma radiation measurable by simple gamma scanning methods. The method of the Scaling Factors (SF) uses empirically determined proportions between the concentrations of Difficult to Measure (DTM) radionuclides and that of easily measurable gamma emitters called Key Nuclide (KN), to calculate DTM concentrations. The objective of this study is to describe a method of the dissolution of the filter cartridges for determination of gamma emitting radionuclides. To determine the activity of a sample by means of of gamma spectrometry, it is necessary previous knowledge of the efficiency curve for a given geometry, which can be obtained by calibrating the efficiency and energy of the detector. The need to dissolve the filter cartridges has arisen given the difficulty in obtaining a standard source with the same filter geometry. Therefore, for this work, the calibration curve was constructed using a mixed gamma-ray standard with known activities, same geometry, density, physical state and chemical composition of the source to be analyzed. A sample with background count (BG) was prepared and analyzed under the same conditions as the samples and the result was subtracted from the results obtained for each filter sample. After many tests, it was possible to establish a method for the dissolution of filter cartridges for the determination of gamma-emitting radionuclides. A mixed gamma-ray standard was produced in liquid form, containing the following radionuclides: {sup 241}Am, {sup 137}Cs and {sup 60}Co, which have photo-peaks at the following energies: 59.54, 661.66 and 1332.49 keV, respectively. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Proposed NRCDA Groundwater Pathway Conceptual Model

As requested by Solid Waste Management (SWM), this report documents the Savannah River National Laboratory’s (SRNL’s) evaluation of updated Naval Reactor waste container and inventory projections and proposes a Naval Reactor Component Disposal Area (NRCDA) groundwater (GW) pathway modeling approach for the next E-Area Low-Level Waste Facility (LLWF) Performance Assessment (PA). Two areas within the E-Area LLWF are used as disposal sites for reactor components from the U.S. Navy. Currently, components arrive by rail and are moved by crane to the 643-26E at-grade gravel disposal pad. Prior to operational closure, reactor components were stored and ultimately disposed in-place on the 643-7E NRCDA (Wilhite and Flach 2004). Naval Reactor (NR) waste is comprised of highly radioactive components consisting of activated corrosion-resistant metal alloy contained within welded thick steel casks, and auxiliary equipment primarily contaminated with Activated Corrosion Products (sometimes referred to as “crud” by the U.S. Navy) at low levels and contained within thinner-walled bolted containers. The 643-7E disposal pad, which contains 41 casks and has an interim soil cover, is closed to future receipts. The latest NR waste projections for the 643-26E disposal pad are substantially different from the original estimates due to a change in reactor maintenance. The original estimate was for 50 heavily shielded, welded casks and 50 thinner-walled bolted containers (primarily shield blocks). Currently, NR Programs project 33 heavily shielded, welded casks, of which 31 are already in-place on the 643-26E pad, and 381 thinner-walled bolted containers. In addition to new waste form projections, there have been changes in facility layout and closure plans as well as PA modeling improvements. This report documents these changes as well as key NRCDA recommendations from the 2015 PA Strategic Plan (Butcher 2016).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Isotopic and Fuel Lattice Parameter Trends in Extended Enrichment and Higher Burnup LWR Fuel, Volume II: BWR

Commercial light water reactor (LWR) operators and fuel vendors in the United States are pursuing changes to nuclear fuel that include extended enrichment (EE) and accident-tolerant fuel (ATF) designs to further improve reactor safety and plant economics. Extended fuel enrichments above 5% 235 U pin enrichment and up to 10% 235 U are a subset of high assay low-enriched uranium (HALEU) that may be deployed in commercial US LWRs in the near term. ATF features, such as cladding coatings or alternative cladding materials, are designed to improve fuel system performance under accident conditions. One goal of EE is to improve fuel cycle economy by enabling fuel to be depleted to higher burnup than the typical current limits (62 gigawatt-days per metric ton of uranium [GWd/MTU]). Adoption of EE, ATF, and high burnup (HBU) fuels in the US commercial fleet requires a clear understanding of the effects on core physics parameters and used fuel isotopic content, as well as confidence in the accuracy of computer code predictions over an expanded range of materials, enrichment, and burnup. A thorough understanding of the applicability and adequacy of benchmark data (e.g., criticality, decay heat, isotopic content) for computer code validation is necessary to ensure that appropriate safety margins are maintained. As part of the US Nuclear Regulatory Commission (NRC) agreement number 31310019N0008, “SCALE Code Development, Assessment and Maintenance,” the effects of EE, ATF, and HBU are being assessed for selected representative LWR fuel designs. The project is divided into phases, and this report summarizes the findings of the Phase 1 work, which focuses on lattice physics parameter and used fuel isotopic changes for a conventional GE14 10 x 10 boiling water reactor (BWR) design with GNF-2 part length rod patterns to model a modern BWR assembly design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Irradiation Testing of Additively Manufactured Materials for LWR Applications: Alloy 718 and 316L Stainless Steel

The objective of this NSUF Project is to assess the changes in irradiated additively manufactured (AM) material properties as compared to non-irradiated material. Type 316L stainless steel and Alloy 718 samples were produced using Direct Metal Laser Melting (DMLM) fabrication. Materials produced from this fabrication method have several potential applications within the nuclear industry as reactor internal repair parts, fuel debris resistant filters, or fuel spacers within existing light water reactors (LWRs). AM materials have been shown to achieve equivalent mechanical behavior in simulated reactor environments as compared to wrought materials, but have significantly more flexibility when it comes to unique design features. The increased component design flexibility makes these AM materials an attractive choice for both current LWR applications as well as for small modular reactor (SMR) designs. Prior to use of these materials in reactor fleet operation, the industry as a whole must evaluate the effects of irradiation on their material properties. Standard 0.4 inch thick Compact Tension specimens and SSJ3 type tensile bars were neutron irradiated at the Advanced Test Reactor to ~1 dpa for the purpose of performing a variety of mechanical tests in a range of simulated environments applicable to LWRs. For the ductile austenitic Type 316L stainless steel, the irradiated data will be used to confirm that the AM process produces materials with properties that are equivalent to wrought materials under testing conditions applicable to LWR operation. Transmission electron microscopy analysis was also performed in order to understand microstructural and microchemical changes induced in each material in response to neutron irradiation. If possible, data collected from these AM 316L samples will be used to remove fluence limits from specifications of ASME code cases for this alloy, which will give vendors much more flexibility in building future components.

36 - MATERIALS SCIENCE↗

DOE OSTI2018

Design and Operation of an Electric Peaking/Industrial Steam Production Unit for SMRs

Various methods of Thermal Energy Storage (TES) can be coupled to nuclear (or renewable) power sources to help absorb grid variability caused by daily load demand changes and renewable intermittency. Our previous research has shown that coupling a sensible heat TES system to a Small Modular Reactor (SMR) allows the reactor to run at effectively nominal full power during periods of variable electric demand by bypassing steam to the TES system during periods of excess capacity. In this paper we demonstrate that this stored thermal energy can be recovered; allowing the TES system to act as a peaking unit during periods of high electric demand, or used to produce steam for ancillary applications such as desalination. For both applications the reactor can operate continuously at approximately 100% power.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

MPACT Verification and Validation Manual (rev 6)

The MPACT code provides the capability to calculate the steady-state and transient reactor core neutron distribution for operating nuclear power plant (PWR) conditions across multiple fuel cycles. As required in the CASL SQA plan, it is the responsibility of UM and ORNL as joint owners of MPACT to ensure that verification and validation activities are performed and documented in a V&V manual with supporting publications and CASL technical reports which can be distributed for reference and distribution within VERA. This document provides the fifth revision of the MPACT verification and validation (V&V) manual and describes the current state of MPACT V&V and updates the framework for future MPACT V&V activities. During FY19 MPACT V&V activities in the areas of code / solution verification and code validation have continued to mature. Several specific tasks have also been completed to include the Ganapol and IPEN Benchmarks which have provided significant improvements in MPACT verification and validation, respectively. Based on the results of V&V efforts this past year, the confidence level has increased in the ability of MPACT to model an operational Pressurized Water Reactor. Specific areas in both verification and validation were identified which require increased emphasis during FY20.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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 OSTI2017

Assessment of sensor technologies for advanced reactors - 250

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 and 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 about {approx}650 deg. 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. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Transition of Site Programs and Implementation of Decommissioning Agent Oversight - 18653

San Onofre Nuclear Generating Station (SONGS), acting as the Decommissioning Agent (DA), retained its NRC License and contracted a Decommissioning General Contractor (DGC) to manage plant programs and execute the Demolition and Dismantlement (D and D) of the site. To support this effort, over twenty-one programs transitioned to the DGC, including Occupational Health and Safety, Radiation Protection, Maintenance, and Engineering. In addition, SONGS effected a cultural shift from an operating station performing work to one of overseeing it. The framework and road map for this kind of D and D model is new to the industry. This presentation shares the SONGS' experience and lessons learned associated with this transition. SONGS and the DGC entered an Agreement that established the contractual requirements and responsibilities of the parties, the Scope of Work, and the mechanisms to ensure effective transition of DA programs. As a part of defining the Scope of Work, the contract established functional areas and divisions of responsibilities divided into the roles of Governance, Oversight, Support, and Performance (GOSP). SONGS, as the site licensee, retained the Oversight Role, with exception of those areas that, as the licensee, would be considered 'non-transferrable'. Examples of those programs were Quality Assurance, Regulatory Affairs and the Employee Concerns Program. While the Company retained responsibility for those programs the DGC developed their own 'parallel' programs that the Company provided Oversight for. In 2017, SONGS completed program transition within one-year following the notice-to-proceed. The overall strategy was driven by three high-level principles: - Allow the DGC the most unfettered execution of the Scope of Work with limited interference and maximum self-governance. With the overriding principle that transitioned programs have a direct link to a basis in regulation or law. - Identify safety and compliance risks, and then bridge the transition of programs from the SONGS to the DGC such that identified risks are minimized and/or mitigated. - Allow the SONGS organization to transition from a Performance role to a primarily Oversight role as the DA. Working collaboratively with the DGC, programs were strategically grouped and sequenced for turnover. Central to the success of transition was fostering a collaborative environment with the DGC, facilitating knowledge transfer, applying schedule rigor, and keeping the team horizontally and vertically aligned to management expectations. SONGS achieved this alignment by embedding the formal process with management expectations for review and acceptance of proposed programs. SONGS' transitioned core programs to a general contractor and transition from a 'perform' organization to an 'oversight' organization. Lessons learned from this experience may benefit other nuclear generating station owners currently evaluating options for site D and D. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

The effects of process parameters on the properties of manganese-rich carbonate precursors: A study of co-precipitation synthesis using semi-batch reactors

The synthesis of precursors for lithium- and manganese-rich xLi 2 MnO 3 ∙ (1-x)LiMO 2 (M = Ni, Mn, Co) (LMR-NMC) materials is generally carried out via co-precipitation using a continuously stirred tank reactor (CSTR) under steady-state conditions. But during the early stages of research (e.g., when screening compositional spaces of interest), using a CSTR in steady-state mode can be time consuming and wasteful. An alternative is to operate the reactor in semi-batch mode, which shortens reaction times and reduces the amount of waste. However, the effect of this mode on the product is not well documented. The present work investigates how several process variables affect the physical properties of Mn-rich NMC carbonate precursors prepared under semi-batch operation of a CSTR. The process variables examined are pH, ammonia concentration of feed solution, process temperature, stirring speed, and reaction time. Overall, the results can help guide research-scale production to support efficient development of Mn-rich carbonates for cathode oxides.

25 ENERGY STORAGE↗

DOE OSTI2017

A Fully Passive Daily Load-follow Operation in a Small PWR System

An innovative fully passive load-follow operation (PLFO) concept, based only on the negative reactivity feedbacks, is proposed by the authors. The new design concept requires no control rod adjustment or boron utilization for achieving load-follow operation in pressurized water reactors (PWRs). PLFO is motivated by the need for developing next-generation PWRs with enhanced power maneuvering capabilities, increased level of autonomism, and less dependence on active controllers. Enhanced power maneuvering capabilities are required to allow next-generation PWRs to actively contribute to electrical grid stability. Meanwhile, increased level of autonomism and less dependence on active controllers are required for safety and reliability of next-generation PWRs. However, developing an autonomous thermal reactor is challenging due to its strong thermal and xenon (Xe) reactivity feedback. Therefore, enhancing power maneuvering capabilities and simultaneously increasing the level of autonomy with less dependence on active controllers is very challenging in PWRs. However, PLFO is a promising solution. For the feasibility of PLFO, the coolant temperature coefficient (CTC) should be always clearly negative and this requires a soluble boron-free (SBF) coolant system. An SBF coolant system is also favorable in terms of reactor safety. Nevertheless, utilizing soluble boron in load-following operation is not favorable, as the change in boron concentration is a quite slow process. The power demand can be either regular (repeated pattern in a daily-basis) or irregular (varying demand). The feasibility of PLFO for regular demand was demonstrated by the authors in other works. This paper is dedicated for summarizing the feasibility of PLFO for irregular demand, and to point out the impact of Xe reactivity on PLFO for a varying power demand.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Passive Low-Energy Nuclear-Recoil Detection with Color Centers

Crystal-damage events such as tracks and point defects have been used to record and detect radiation for a long time and recently they have been proposed as a means for dark-matter detection. Color centers can be read out optically and we propose a scheme based on selective plane illumination microscopy for micrometer-scale imaging of large volumes corresponding to kilogram mass detectors. This class of detectors would be passive and would operate at room temperature and we call this the passive low-energy optical color-center nuclear-recoil (PALEOCCENE) detection method. We apply these concepts to the detection of reactor neutrinos using coherent elastic neutrino nucleus scattering (CE ν NS). Crystal-damage formation energies are intrinsically on the order of 25 eV, resulting in similarly low nuclear-recoil thresholds. Furthermore, this would enable the observation of reactor-neutrino CE ν NS with detectors as small as 10 g. Additionally, a competitive search for spin-dependent dark-matter scattering down to a dark-matter mass of 0.3 GeV could be possible. Passive crystal detectors might also be attractive for nuclear-nonproliferation safeguards if used to monitor reactor power and to put limits on plutonium production. The passive nature and small footprint of the proposed detectors implies that these might fit well within accepted reactor-safeguards operations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

NASA NTRS2025

CFD Validation of a Methanation Reactor With Cooling Jacket Used in an ISRU Propellant Production System

A digital model of an OxEon Energy methanation reactor (with a cooling jacket) planned for use in a NASA designed In-Situ Resource Utilization (ISRU) propellant production system was developed by the UTEP Aerospace Center using STAR-CCM+ software, a Siemens Computational Fluid Dynamics (CFD) package. The reactor digital model created includes two primary components: (i) a cooling jacket that maintains the required reactor chamber surface temperature and (ii) a reactor core chamber where the exothermic methanation reactions occur. The StarCCM+ digital model has two objectives: (i) maintain the surface temperature of the reactor below a set value so that the extra heat accumulation does not melt or crack the chamber, and (ii) produce the desired amount of methane (CH4) according to test cases completed by OxEon Energy for NASA JSC. The computational digital model was validated with the test data enabling the cooling jacket scale to be predicted based on the reactor's operating temperature. The StarCCM+ simulation of the digital model shows a good agreement with the cooling jacket performance and the CH4 production. One test case for the cooling jacket maintained the reactor wall temperature around 310 °C was validated and the reactor produced 185 g/hr CH 4 .

ISRU↗