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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Demonstration of Passive Phoretic Decontamination by Accident Simulation in a Scaled Reactor Containment

Reactor containment performance is a critical defense-in- depth barrier for the prevention of radioactive release in the event of a nuclear accident. One of the functions of the containment is to assist in the decontamination of aerosolized radioactive particles through both 'passive' or natural processes and 'active' or mechanical processes. Current United States Nuclear Regulatory Commission (NRC) guidance (Reg. guide 1.183) allows the use of aerosol natural deposition correlations to quantify postaccident nuclear containment decontamination factors. Large Light Water Reactor (LLWR) correlations have previously been developed to quantify the contribution of natural phenomena to post-accident aerosol decontamination. Specific correlations must similarly be developed to support the safety analyses of Small Modular Reactors (SMRs). For LLWR plants, decontamination is achieved 'actively' by mechanical systems such as containment sprays and 'passively' via natural phenomena such as gravitational settling, condensation, impaction, thermophoresis, diffusiophoresis and hygroscopic effects. Integrated pressurized water reactor (iPWR) SMR designs do not have containment sprays; therefore, they rely primarily on passive natural phenomena for decontamination in the integrated reactor vessel. Analytical studies and inferences from LLWR test data suggest that the smaller containment vessel volume to surface area ratio, coupled with the thermal hydraulic properties of iPWR plants, would significantly enhance decontamination factors for these smaller designs. The research aims to quantify the amount of decontamination from natural phenomena for the range of iPWR physical and thermal hydraulic characteristics. A preliminary assessment has shown that diffusiophoresis, thermophoresis and hygroscopic effects are the major areas with a research gap for iPWR plants; hence these areas are the foci of the research. These first-of-a-kind experiments will be described and preliminary analytical results will be presented that show the expected effects of SMR-specific correlations on passive decontamination of aerosolized radionuclides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Deployment and performance of a Low-Energy-Threshold Skipper-CCD inside a nuclear reactor

Charge Coupled Devices (CCD) are being used for reactor neutrino experiments and have already demonstrated their potential in constraining new physics models. The prospect of a Skipper-CCD experiment looking for standard and beyond standard model (BSM) physics in a nuclear reactor has been evaluated for different benchmark scenarios. Here, we report the first installation of a 2-g Skipper-CCD inside the containment building of a 2 GW th nuclear power plant and analyze its performance throughout its first 18 months of operation. The sensor was successfully deployed at Atucha II, in Argentina, 12 meters away from the center of the reactor core. We discuss the challenges involved in the commissioning of the detector and present data acquired during reactor ON and reactor OFF periods, with the sensor functioning with a sub-electron readout noise of 0.17 e - . Based on an exposure of 56.8 g day reactor ON and two reactor OFF data sets with a total exposure of 118.1 g day we characterize the system and evaluate the sensitivity to CEvNS. We achieved a background rate of 33 kdru and a low threshold of 45 eV ee . The ongoing efforts to improve sensitivities to CEvNS and BSM interaction are also discussed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Perspectives on Tailoring Neutron Energy Spectra in Material Test Reactors

Material test reactors (MTRs) are used to irradiate nuclear fuels and materials to develop data for how they endure neutron bombardment in or near reactor cores. Most historic MTRs, and all that remain operational in the US today, are water-cooled types and produce a thermalized neutron flux. New irradiation facilities are needed which can produce neutron energy spectra relevant to fast and fusion reactor environments. Construction of these facilities will take several years of steadfast funding to complete, which poses a formidable schedule challenge for current fast and fusion reactor developers. Irradiation designs which modify the neutron energy spectra delivered to test specimens in thermal spectrum MTRs, an approach referred to as “spectral tailoring”, can be used to approximate several relevant phenomena in the materials needed to enable fast and fusion reactor technologies. This approach is imperfect, but still valuable in the present situation. Here, the two highest flux MTRs operational in the United States, the Advanced Test Reactor (ATR) and High Flux Isotope Reactor (HFIR), have rich histories, ongoing developments, and new potentials for spectral tailoring that will be reviewed and discussed in this paper.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Design-informed neutronics assessment of long-lived fission product transmutation in a tokamak fusion reactor blanket

This study presents a neutronics-based assessment of the feasibility and viability of transmuting six major long-lived fission products (LLFPs) from light-water reactors, namely 99 Tc, 129 I, 79 Se, 93 Zr, 126 Sn, and 135 Cs, within the blanket region of a tokamak fusion reactor, using the MIT ARC design as a concrete fusion configuration. Monte Carlo neutronics simulations were performed to evaluate LLFP transmutation and to compare the results with a reference boiling water reactor (BWR). The results indicate that transmutation of all six LLFPs is neutronics-feasible in fusion reactors, with transmutation half-lives significantly shorter than their natural decay half-lives. For elemental targets, transmutation of 135 Cs, 126 Sn, and 93 Zr was found potentially viable, as the net mass transmuted exceeded that achievable in the reference BWR under identical target volume and irradiation time. When isotopically separated targets were considered, transmutation of 126 Sn and 93 Zr appeared potentially viable. A parametric study demonstrated that plasma geometry modifications can enhance local neutron flux, increasing the transmuted 93 Zr mass by approximately 33% and reducing the transmutation half-life from approximately 240 years to 180 years. Repositioning the target and adjusting material layer thickness reduced the transmutation half-life of 93 Zr to 67 years and increased the net mass transmuted by a factor of 50. Furthermore, these results demonstrate that fusion reactors can enable LLFP transmutation beyond the practical limits of thermal fission reactors and highlight the critical role of reactor and blanket design optimization. Engineering and fuel-cycle considerations required for deployment are beyond the scope of this neutronics-focused study.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Modeling of fission and activation products in molten salt reactors and their potential impact on the radionuclide monitoring stations of the International Monitoring System

Molten Salt Reactors (MSRs) are one of six Generation IV reactor designs currently under development around the world. Because of the unique operating conditions of MSRs, which include molten fuel and the continuous removal of gaseous fission products during operation, work was performed to analyze the potential impact of emissions on the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Simulations were performed to predict the production of IMS-relevant radionuclides in four MSR designs operating under two scenarios: (1) a sealed reactor with releases only during operational shutdown, and (2) continuous reprocessing or sparging of the fuel salt. From these production estimates the radioxenon and radioiodine signatures were extracted and compared to three current reactor designs (Pressurized Water Reactor (PWR), BWR, RBMK). In cases where continuous reprocessing of the fuel salt occurred, both the radioxenon and radioiodine signatures were nearly indistinguishable from a nuclear explosion. Estimates were also made of the potential emission rate of radioxenon for three reactor designs and it was found that MSRs have the potential to emit radioxenon isotopes at a rate of 10^15-8×10^16 Bq/d for 133Xe if no abatement is used. An assessment was made of activation products using a candidate fuel salt (FLiBe) mixed with corrosion products for the Thorium Molten Salt Reactor (TMSR-LF1).

Johnson, Christine M.↗
DOE OSTI2020

Irradiation performance of nonfertile (Pu-MA-Zr) fast reactor metal fuels

This work was part of a program begun in 2001 to develop advanced nuclear fuels, originally as carriers for plutonium and minor actinides (neptunium, curium, and americium) taken from spent commercial light-water reactors (LWR) so that the plutonium and minor actinides could be ‘burned’ or transmuted in an accelerator or a fast nuclear reactor. A central part of these experiment programs has been the development of advanced fast reactor fuels, because a fast reactor was considered the most efficient vehicle to transmute the actinide waste products, and metallic fuels is a central focus of these tests. An experiment design was developed in which a thermal test reactor, the Advanced Test Reactor (ATR), was used to test small fuel pin prototypes, by creating areas in the core shielded by cadmium filters to produce a largely epithermal and fast neutron spectrum environment in which the pins could be irradiated. The results of non-fertile metallic fuel (no uranium) tests are presented here. Pu-Am-Np-Zr fuels were irradiated to fission densities up to 33 × 10 20 fission/cm 3 and Pu-239 depletions of up to 39%. The depletions were created by roughly 2/3 by fission and 1/3 by transmutation neutron capture. Up to five fuel ‘rodlets’ were irradiated in three sealed capsules stacked axially in the core, and the peak cladding temperatures ranged from 300°C to 500°C, depending on axial location as those near the core centerline are operating hotter and to higher fission densities. Several post-irradiation examinations (precision gamma scanning and fission gas release) were similar to other historical metal fuel experiments in fast reactors. However, optical metallography indicated that two of the rodlets had breached. The exact reasons are unclear. Due to the design of this irradiation experiment a rodlet breach could have increased the temperature in others in the same capsule by contaminating the thermal gap helium with heavier and less conductive fission product gases. Finally, some of those rodlets showed high amounts of fuel/cladding chemical interaction (FCCI).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Methodology for inferring reactor core power distribution from an optical fiber based gamma thermometer array

Here, a 4-step data analytic methodology has been devised for the purpose of inferring the distribution of power in a reactor core, based on the response of an array of optical fiber based gamma thermometers (OFBGTs). This data analytic methodology is crucial for the development of a system of OFBGTs for the purpose of calibrating local power range monitors in boiling water reactors. Such a system would be an improvement to the present calibration system in boiling water reactors, in terms of safety, efficiency, and permanence. The first step of this methodology is to establish an energy balance method. In this method, one uses MCNP to determine response functions, which allow one to convert from gamma thermometer response to power. The gamma thermometers and the reactor core are segmented, such that each gamma thermometer segment provides an estimate of power for each reactor core segment. The estimates of power for the reactor core segments (hereafter referred to as fuel assembly segments) are calculated based on the measured response of the gamma thermometers. The second step of the methodology is to employ a weighting scheme to combine the estimates of the power of the various fuel assembly segments, based on the response of the various OFGBT segments and the incremental dose rates of the various fuel assembly segments to the OFBGT segments. The third step is to iteratively calculate the estimates of the power of the fuel assembly segments, until a convergence criterion is met, which indicates that the calculation has converged. The fourth and final step of the methodology is to estimate the uncertainty of the power for each fuel assembly segment. The mathematical basis for this step is not the focus of this paper. We have used a 3-D homogeneous reactor model to demonstrate the data analytic methodology; and have found that the data analytic methodology operates as intended.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Janus Reactors with Highly Efficient Enzymatic CO 2 Nanocascade at Air–Liquid Interface

Though enzymatic cascade reactors have been the subject of intense research over the past few years, their application is still limited by the complicated fabrication protocols, unsatisfactory stability and lack of effective reactor designs. In addition, the spatial positioning of the cascade reactor has so far not been investigated, which is of significant importance for biphase catalytic reaction systems. Inspired by the Janus properties of the lipid cellular membrane, in this paper we show a highly efficient Janus gas–liquid reactor for CO 2 hydration and conversion. Within the Janus reactor, nanocascades containing the nanoscale compartmentalized carbonic anhydrase and formic dehydrogenase were positioned at a well-defined gas–liquid interface, with a high substrate concentration gradient. The Janus reactor exhibited 2.5 times higher CO 2 hydration efficiency compared with the conventional gas–liquid contactor with pristine membranes, and the formic acid conversion rate can reach approximately 90%. Finally, through this work, we provide evidence that the spatial arrangement of the nanocascade is also crucial to efficient reactions, and the Janus reactor can be a promising candidate for the biphase catalytic reactions in environmental, biological and energy aspects.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Fluidized-bed atomic layer deposition reactor for the synthesis of core-shell nanoparticles

The design of a fluidized bed atomic layer deposition (ALD) reactor is described in detail. The reactor consists of three parts that have all been placed in one protective cabinet: precursor dosing, reactor, and residual gas treatment section. In the precursor dosing section, the chemicals needed for the ALD reaction are injected into the carrier gas using different methods for different precursors. The reactor section is designed in such a way that a homogeneous fluidized bed can be obtained with a constant, actively controlled, reactor pressure. Furthermore, no filters are required inside the reactor chamber, minimizing the risk of pressure increase due to fouling. The residual gas treatment section consists of a decomposition furnace to remove residual precursor and a particle filter and is installed to protect the pump. In order to demonstrate the performance of the reactor, SiO{sub 2} particles have been coated with TiO{sub 2} using tetrakis-dimethylamino titanium (TDMAT) and H{sub 2}O as precursors. Experiments with varying pulse times show that saturated growth can be obtained with TDMAT pulse times larger than 600 s. Analysis of the powder with High-Angle Annular Dark-Field Scanning Transmission Electron Microscopy (HAADF-STEM) and energy dispersive X-ray spectroscopy confirmed that after 50 cycles, all SiO{sub 2} particles were coated with a 1.6 nm homogenous shell of TiO{sub 2}.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Design improvements for a recirculating reactor: Enhanced temperature measurement and sample-isolated reactivity in steady-state kinetic studies

Building upon a previous recirculating reactor design [S.A. Tenney, K. Xie, J.R. Monnier, A. Rodriguez, R.P. Galhenage, S. Audrey, D.A. Chen, Rev. Sci. Instrum. 84, 104101 (2013)], we present significant improvements that address key limitations in steady-state kinetic measurements for heterogeneous catalysis. The enhanced reactor design features direct sample heating with a focused IR lamp and temperature measurement in direct contact with the sample, enabling more accurate temperature control and improved kinetic analysis. A critical advancement is the isolation of sample reactivity from reactor wall contributions, ensuring that only the sample contributes to measured reaction rates. This was a limitation in earlier designs where the entire reactor contributed to the observed reactivity. The system incorporates a bypass flow cell for direct comparison with powder catalysts under identical conditions using a standard plug-flow reactor configuration. We demonstrate these capabilities through CO oxidation experiments on Pt(111) single crystals and graphene-passivated Pt(111), highlighting the system's ability to differentiate catalytic activity in model systems and directly compare them with high surface area powder catalysts. This reactor is particularly suited for thin films and low surface area catalysts that are not effectively evaluated in traditional flow reactors, especially for samples with low numbers of active sites or slow reaction rates.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Review and Modeling of Integrated Energy Systems with Nuclear Reactor Coupled Desalination and District Heating

Detailed reviews of a past advanced nuclear reactor based integrated energy system, as well as other nuclear reactor and fossil fuel based integrated energy systems have been performed for this work. Review of the utilization of heat from nuclear reactors for various applications and cogeneration has been done. The heat can be utilized by extraction of the steam from the turbine while the steam is still at a desired temperature. While use of nuclear process heat for district heating in countries like Finland, France, China, Poland, and elsewhere is discussed, more focus of the review has been given on nuclear desalination processes. Integrated energy systems (IES) where distinct types of reactors like PWR, BWR, sodium cooled fast reactor, heavy water reactor and other advanced reactors are coupled with various nuclear desalination processes like multi-effect distillation (MED), multi-stage flashing (MSF) and reverse osmosis (RO) methods have been discussed. The nuclear desalination plant at Aktau has been discussed in more detail due to its decades of successful operation. The IES of the Aktau plant coupled with 5-effect MED desalination plant has been taken as a reference for modeling the Open Modelica (OM) based IES of this work. Here, the OM IES model shows good agreement with the MED plant output of Aktau and can be extended for future applications of IES.

42 ENGINEERING↗
DOE OSTI2019

Nuclear Archaeology to Distinguish Plutonium and Tritium Production Modes in Heavy-Water Reactors

Several nuclear archaeology techniques have been proposed to determine historic plutonium production in dedicated nuclear reactors. These methods rely on sampling materials from the reactor core or specially designed monitor tags, to examine suited isotopic ratios and deduce the amount of plutonium produced. However, some production reactors can be used to produce other isotopes than plutonium, such as tritium. If a reactor has been declared to produce tritium, it could be crucial to confirm that this was the case and that the reactor was not used to produce plutonium instead. In this paper, we describe how isotopic ratios discrepancies can be used to distinguish between plutonium and tritium production modes. Here, we present results obtained via the simulation of reactor lattices inspired by Savannah River Site heavy water production reactors and show that elements such as hafnium and tungsten can be conveniently used to detect undeclared production of plutonium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

High power ring methods and accelerator driven subcritical reactor application

High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g., PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the transverse beam dynamics. The results obtained allow to develop a correction scheme to minimize the tune variations of the FFAG. This is the cornerstone of a new fixed tune non-scaling FFAG that represents a potential candidate for high power applications. As part of the developments towards high power at the KURRI FFAG, beam dynamics studies have to account for space charge effects. In that framework, models have been installed in the tracking code ZGOUBI to account for the self-interaction of the particles in the accelerator. Application to the FFAG studies is shown. Finally, one focused on the ADSR concept as a candidate to solve the problem of nuclear waste. In order to establish the accelerator requirements, one compared the performance of ADSR with other conventional critical reactors by means of the levelized cost of energy. A general comparison between the different accelerator technologies that can satisfy these requirements is finally presented. In summary, the main drawback of the ADSR technology is the high Levelized Cost Of Energy compared to other advanced reactor concepts that do not employ an accelerator. Nowadays, this is a show-stopper for any industrial application aiming at producing energy (without dealing with the waste problem). Besides, the reactor is not intrinsically safer than critical reactor concepts, given the complexity of managing the target interface between the accelerator and the reactor core.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Seismic Isolation of Major Advanced Reactor Systems for Economic Improvement and Safety Assurance

Advanced reactor concepts currently under development in the United States [e.g., the High-Temperature Gas Reactor (HTGR) of X-Energy and the Molten Chloride Fast Reactor (MCFR) of TerraPower] are striving to improve the economics of nuclear power primarily by using passively and ‘walk-away’ safe technologies in the reactor. The recently published report, ‘Future of Nuclear Energy in a Carbon-Constrained World’ by the Massachusetts Institute of Technology Energy Institute (MIT EI) has shown that a significant contributor to the capital cost of NPPs is the ‘civil works’ i.e., the construction of the balance of plant (BoP: all SSCs except those involved in power generation, e.g., reactor vessel) such as the buildings, containment dome, foundation, etc., which amount to almost half of the overnight capital cost. The report suggests that using advanced construction technologies such as seismic isolation and high-performance concrete to significantly reduce the capital costs of advanced reactors. This report demonstrates the application of two cost-cutting solutions on advanced reactor designs that are currently under development. These solutions include, (1) seismic base isolation, and (2) risk- and cost-based seismic design optimization including seismic isolation of individual components.

42 ENGINEERING↗
DOE OSTI2020

Drop Analysis of the Advanced Test Reactor Fresh Fuel Shipping Container with Heavier Low-Enriched Uranium Fuel Contents

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.

42 ENGINEERING↗
DOE OSTI2021

Performance Demonstration of Self-Powered Neutron Detectors for Steady-State Reactor Operations

The irradiation testing of sensors in reactors is a crucial step towards calibrating and qualifying sensors prior to their deployment in experiments. This report details the process toward qualifying and calibrating custom-designed rhodium-based self-power neutron detectors (Rh-SPNDs) for steady-state reactor irradiations. This process serves to both demonstrate the performance capabilities of Rh-SPNDs as well as to provide experimental data for development of a sensor sensitivity model. Two designs of Rh-SPNDs were tested in various reactors to demonstrate: detection resolution in a low neutron flux environment, a delayed-response compensation technique, output linearity in a large range of neutron flux, and measurement accuracy verified with dosimetry. The detection resolution and compensation technique was demonstrated in the AGN-201m reactor at Idaho State University. The irradiation confirmed the sensors’ capability to perform steady-state operations in a low neutron flux of ~2E8 n/cm 2 -sec. Sensor output linearity coupled with the delayed-response compensation was investigated at the neutron radiography reactor at Idaho National Laboratory. A Rh-SPND was irradiated to neutron fluxes ranging from 2E8 to 2E13 n/cm2-sec range. The measured data demonstrated a wide and linear range of operation with a measured linear sensitivity of 1.0129 ×10 -13 A/W with a correlation-squared value of r 2 =0.9927. The measurement accuracy was investigated at the Advanced Test Reactor Critical reactor. The SPNDs were inserted into a test vehicle with collocated flux wires. Two irradiations with different flux levels were performed, and the SPNDs relative measurement between the two irradiations was calculated to be 1.2613 ± 0.0153 for the small SPND design and 1.1809 ± 0.0108 for the large SPND design. Both SPND measurements fell between the co-axial dosimetry result, which reported 1.218 ± 0.047. Additionally, the preliminary MCNP model for calculating SPND sensitivity was developed in parallel to this work. Modeled neutron spectrum with measured magnitude was used for inputs to determine the simulated SPND output. The results showed an overestimation of signal strength by a factor of 5, which was expected because of model simplification. This leads to future modeling work to account for signal losses from additional physical properties, including high temperature environments for FY-21.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Overview of Advanced Reactor Simulation Capabilities to Support Demonstrations

The Virtual Test Bed (VTB) was established to support the deployment of advanced reactors by facilitating the use of state-of-the-art modeling and simulation capabilities. These tools are able to solve previously prohibitive challenge problems in greater levels of fidelity and in a multi-physics framework. A VTB repository was recently set up to host models for a wide variety of reactor types and using a range of NEAMS-developed software. Table I summarizes the different capabilities already available in the repository and up-and-coming near/long term ones. The repository currently hosts models for a wide variety of reactor types, including High-Temperature Gas Cooled Reactors (HTGR), Sodium Fast Reactors (SFR), Fluoride salt cooled High-temperature Reactors (FHR), Molten Salt Reactors (MSR), and Heat-Pipe Microreactors (HP-MR). It should be noted that work is ongoing for all of the models and each example problem is at a different state of maturity.

99 GENERAL AND MISCELLANEOUS↗