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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Neutronics Calculations for EBR-II Run 137E

Neutronics calculations were performed for EBR-II Run 137E using MC2 -3/TWODANT, DIF3D, and PERSENT codes. Parameters required for system safety analysis such as the reactivity feedback coefficients and core power distribution were evaluated. The detailed results are provided in the accompanying Excel spreadsheet Run_137E_Neutronics_Results.xls. This report describes the core model of EBR-II at the beginning of Run 137E, the computation methodology, and how to interpret the results listed in the Excel spreadsheet.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

SCALE Modeling of the Sodium Cooled Fast-Spectrum Advanced Burner Test Reactor

This report documents the modeling and simulation of a sodium-cooled fast reactor (SFR) as part of a U.S. Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with the Oak Ridge National Laboratory code SCALE and the Sandia National Laboratories (SNL) code MELCOR. Based on publicly available benchmark specifications, a fully heterogeneous 3D SCALE model of the 250 MWth Advanced Burner Test Reactor (ABTR) was developed to demonstrate SCALE’s capabilities for full-core reactivity analysis, fuel inventory prediction, and decay heat analysis of an SFR. The benchmark specifications contain modeling details for the ABTR core at the beginning of equilibrium cycle (BOEC) at operating conditions; they were derived from a 2006 preconceptual design report produced by Argonne National Laboratory. The ABTR was designed to demonstrate reactor-based transmutation of transuranics, that is, to “burn” transuranics recovered from light-water reactor (LWR) spent fuel. The ABTR’s fuel is designed to operate in 4 month cycles using uranium/transuranic (U/TRU) metallic fuel, with a TRU content of approximately 20%, at a conversion ratio of approximately 0.6. Various reactivity calculations were performed with SCALE for the ABTR and, where possible, compared with results available in the open literature. Additionally, SCALE was used to perform a full-core depletion calculation over the 4 month cycle to obtain the nuclide inventory at the end of equilibrium cycle (EOEC). These nuclide inventories, decay heat, power profiles, and reactivity feedback coefficients at EOEC represent the initial conditions for analyzing severe accident scenarios with MELCOR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Field observational constraints on the controllers in glyoxal (CHOCHO) reactive uptake to aerosol

Abstract. Glyoxal (CHOCHO), the simplest dicarbonyl in the troposphere, is a potential precursor for secondary organic aerosol (SOA) and brown carbon (BrC) affecting air quality and climate. The airborne measurement of CHOCHO concentrations during the KORUS-AQ (KORea–US Air Quality study) campaign in 2016 enables detailed quantification of loss mechanisms pertaining to SOA formation in the real atmosphere. The production of this molecule was mainly from oxidation of aromatics (59 %) initiated by hydroxyl radical (OH). CHOCHO loss to aerosol was found to be the most important removal path (69 %) and contributed to roughly ∼ 20 % (3.7 µg sm−3 ppmv−1 h−1, normalized with excess CO) of SOA growth in the first 6 h in Seoul Metropolitan Area. A reactive uptake coefficient (γ) of ∼ 0.008 best represents the loss of CHOCHO by surface uptake during the campaign. To our knowledge, we show the first field observation of aerosol surface-area-dependent (Asurf) CHOCHO uptake, which diverges from the simple surface uptake assumption as Asurf increases in ambient condition. Specifically, under the low (high) aerosol loading, the CHOCHO effective uptake rate coefficient, keff,uptake, linearly increases (levels off) with Asurf; thus, the irreversible surface uptake is a reasonable (unreasonable) approximation for simulating CHOCHO loss to aerosol. Dependence on photochemical impact and changes in the chemical and physical aerosol properties “free water”, as well as aerosol viscosity, are discussed as other possible factors influencing CHOCHO uptake rate. Our inferred Henry's law coefficient of CHOCHO, 7.0×108 M atm−1, is ∼ 2 orders of magnitude higher than those estimated from salting-in effects constrained by inorganic salts only consistent with laboratory findings that show similar high partitioning into water-soluble organics, which urges more understanding on CHOCHO solubility under real atmospheric conditions.

Kim, Dongwook (ORCID:0000000221723588)↗
DOE OSTI2023

Hydrided moderators for nuclear reactors

High temperature moderators for nuclear reactors and processes for their production are disclosed. The moderators include at least one hydrided metal and/or hydride metal allow, such as yttrium hydride, thorium hydride, yttrium-cerium hydride, yttrium-gadolinium hydride, yttrium calcium hydride, cerium hydride, etc. Such metal hydrides and/or hydride alloys may have high thermal stability, a relatively low thermal neutron absorption cross section, the ability to retain hydrogen over a large temperature range, and have good mechanical properties. Such moderators may induce spectral shift in reactors which, in turn, magnifies the Doppler reactivity temperature coefficient. Such moderators to thermalize neutrons may also enhance fuel utilization and cost-effectiveness of the reactor while keeping the core portable.

Dasari, Venkateswara Rao↗
DOE OSTI2022

Initial development of a generic fluoride salt-cooled reactor model

Fluoride high-temperature reactors (FHRs) are high-temperature, low-pressure reactor concepts that use tri-structural isotropic (TRISO) fuel and molten fluoride salt coolant. These reactors have the potential to provide both electrical power and high-temperature process heat. We used generic FHR parameters for a pebble-bed FHR to develop an initial model with fresh fuel for a generic FHR (gFHR) in MELCOR and SCALE (NEWT and KENO). In this paper, we present the development of our gFHR models, which will serve as the baseline for a sensitivity and uncertainty analysis to quantify the range of possible source terms for FHRs in severe accidents. We present MELCOR results for fuel and coolant temperatures through the core, a nodalization study for the steady-state thermal hydraulic model, and development of reactor physics models in SCALE. As this work progresses, these models will be used to calculate source terms for a loss-of-forced-flow accident and to conduct a sensitivity study on this accident to establish a range of possible source terms. SCALE will provide reactor physics parameters like isotopic inventory, decay heat generation, and temperature coefficients of reactivity. Using the uncertainty quantification tools within SCALE, we will generate distributions for those parameters and will use the uncertainty quantification code RAVEN or DAKOTA to sample those distributions in MELCOR to quantify the impact of reactor physics and thermal hydraulic uncertainties on FHR source terms. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

On the characterization of biases arising from methods and approximations used for sensitivity analyses

The preliminary assessment of the performance and safety of advanced reactors, as well as the identification of research and development needs, rely on computer simulations due to the lack of operational experience. Those simulations should be based on verified and validated computational tools, calculation schemes, and nuclear data libraries, and should be accompanied of a proper characterization of the involved uncertainties, providing confidence in the computational-based predictions. Thus, sensitivity and uncertainty studies, together with integral experiments, play an essential role in that process. Sensitivities are then a critical element and guidelines about how to produce accurate enough sensitivities depending on the subsequent analyses to be performed are of interest. The present study addresses the sensitivities biases arising from the use of different nuclear data libraries, computational methods and the assumption of modelling simplifications. Not only sensitivities for multiplication factor but also for safety-relevant reactivity responses have been analyzed. In the study, propagated uncertainties are also included for the considered parameters to illustrate the impact of the sensitivity profiles obtained for each case on the final uncertainty. Then, relevant recommendations are given for sensitivity analyses of highly complex systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

The Jupiter High-240 Experiment

The Jupiter High-240 experiment was performed in May of 2019 by researchers from Los Alamos National Laboratory (LANL) at the National Criticality Experiments Research Center (NCERC) in the Device Assembly Facility (DAF) located at the Nevada National Security Site (NNSS). This experiment has previously been mentioned briefly in prior publication related to a collaborative effort with the Japan Atomic Energy Agency (JAEA) to assess lead void coefficients of reactivity in uranium- and plutonium-fueled systems with lead. This series of experiments supports JAEA’s research into the development of an accelerator driven transmutation system for spent nuclear fuel. The Jupiter High-240 experiment built upon the previous Jupiter experiment by incorporating plutonium fuel plates with higher 240 Pu content. Efforts to formally benchmark the original Jupiter experiment have continued for inclusion in the benchmark handbook of the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Whereas there is much similarity between the two Jupiter experiments, there is a desire to also evaluate and benchmark this second experiment to further contribute towards the availability of lead-sensitive benchmarks. The components utilized in these two experiments have also been used to perform other subcritical and Rossi-α measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Effects of Uranium Impurities in Downblended HEU on HTGR Performance

Many advanced reactor designs require fuel enriched between 5% and 20% 235 U. To assist in producing fuel at these enrichment levels, government-owned inventories of highly enriched uranium can be downblended. However, fuel produced from these inventories contain uranium impurities that are not often found when enriching natural uranium or accounted for when modeling reactor cores. To address this concern, this work models reactor designs like the X-energy Xe-100 and the Ultra Safe Nuclear Company’s Micro Modular Reactor, and compares their performance with fuel from enriching natural uranium to fuel from downblended highly enriched uranium. This paper evaluates the models based on the effective neutron multiplication factor, k eff , effective delayed neutron fraction, β eff , and energy- and spatially dependent neutron flux, ϕ, as well as the fuel, coolant, moderator, and total reactivity temperature feedback coefficients, α F , α C , α M , and α T . The results show that the fuel from downblended highly enriched uranium inventories leads to differences in each of the metrics, especially in the keff values. In the Xe-100–like and Micro Modular Reactor–like models, k eff changes by about 1400 pcm and up to 1200 pcm, respectively. Total reactivity feedback coefficients α T are negative with the impure fuels and the keff values remain above 1 for each core configuration and fuel composition. These results show that the impure fuel compositions do not necessarily prevent achieving key design parameters, such as cycle length, or from operating in a safe condition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sensitivity analyses of a homogeneous model and a RZ model of the MYRRHA reactor in its critical configuration

Two homogenized models of the MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) reactor in its critical configuration are presented in this study. Both models will be verified by comparing the main reactor parameters (K{sub eff}, β{sub eff} and Λ) with those of the heterogeneous model and a sensitivity analysis will be performed. A sensitivity analysis of some important parameters related to reactivity, namely the vacuum coefficient, the Doppler and the power peaking factors, is also presented. The study of the vacuum coefficient has been carried out by analyzing two possible scenarios: 50% vacuum in the coolant (lead bismuth eutectic) and 100%, obtaining a list of the 10 reactions and nuclides that most affect the k{sub eff} value. With respect to the Doppler coefficient, it will be seen that a 300 degree increase in fuel temperature (starting at 800 K) results in a reduction of the total reactivity of the system by 134 pcm.. Finally, the map of the power peaking factors will be shown, as well as a sensitivity analysis. For making the sensitivity calculations Serpent-2 reactor physics Monte Carlo code has been used and JEFF-3.3 library has been chosen in order to continue with the OECD/NEA WPEC SG46 work evaluating the nuclear data of this library. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Transient Modeling and Simulation of a Generic Stable Salt Reactor

A SAM system-level model of a generic stable salt reactor has been developed to investigate thermal-hydraulic behavior and safety performance under steady and transient conditions. The model integrates information generated from a reactor physics analysis using PROTEUS and PERSENT, and a computation fluid dynamics (CFD) analysis using STAR-CCM+. A loose, iterative coupling scheme between PROTEUS and SAM is implemented to calculate the equilibrium power and temperature distributions in the steady-state critical core condition. The converged steady-state model is then used in PERSENT to calculate the four reactivity feedback temperature coefficients (Doppler, fuel density, coolant density, and core radial expansion) and kinetic parameters that are needed in SAM to model the temperature feedback effects in transient simulations. Within the fully enclosed liquid fuel pins, natural convection is the dominant heat transfer mechanism. The STAR-CCM+ model of the fuel pin considers conjugate heat transfer from the liquid fuel salt to the pin cladding and external reactor coolant. The CFD results of the axial and radial temperature profiles are used to empirically determine an effective fuel salt thermal conductivity in the SAM fuel pin model so that the temperatures predicted by the SAM model match as closely as possible the CFD results. In the central region of the fuel pin, the effective thermal conductivity is as high as similar to 60 times the physical fuel salt thermal conductivity. The whole-plant SAM model is then used to simulate an unprotected station blackout transient. The results of this simulation showed that the large negative fuel axial expansion reactivity feedback reduces fission power to similar to 2.4% nominal power. The core is cooled by natural circulation, which removes heat in the core to the emergency heat removal system, and ultimately, to the ambient. However, peak fuel salt and cladding temperatures can potentially reach as high as 1500 K, albeit briefly, if the shutdown mechanism fails to operate.

stable salt reactor; transient simulations; system↗
DOE OSTI2020

Continuous polymer films deposited on top of porous substrates using plasma-enhanced atomic layer deposition and molecular layer deposition

Molecular layer deposition (MLD) is able to produce ultrathin polymer films with control over thickness, cross-linking, and chemical composition. With these capabilities, MLD should be useful in the fabrication of novel polymer membranes on porous supports. However, confining a continuous MLD film to the surface of porous substrates is difficult because of MLD film growth in the pores. The deposition in the pores lowers the conductance of the porous support. This paper presents a method to deposit continuous polymer films on top of porous substrates. In this method, Al2O3 plasma-enhanced atomic layer deposition (PE-ALD) using trimethylaluminum and oxygen plasma as the reactants was first used to cap the pores of the substrate. Subsequently, a polyamide MLD film was deposited on the Al2O3 PE-ALD capping layer using m-phenylenediamine and trimesoyl chloride as the reactants. The Al2O3 pore caps were then removed from the porous substrate by etching from the backside using a timed exposure to a dilute sodium hydroxide solution. This method was demonstrated using anodic aluminum oxide (AAO) and polyethersulfone (PES) porous substrates. Al2O3 PE-ALD film growth was limited to the top of the porous substrate, resulting in rapid surface recombination or high sticking coefficients for the reactive plasma species within the pores. Gas permeance measurements confirmed the pore capping of the AAO substrates. The reopening of the pores by dissolving the Al2O3 pore caps with a sodium hydroxide solution was monitored using gas permeance versus etch time. The removal of the Al2O3 pore caps from the PES substrates could also dissolve the Al2O3 layer underneath the MLD film. The loss of this Al2O3 layer led to the detachment of the MLD film from the PES substrate. However, the MLD film could be anchored to the PES support at fractures located in the Al2O3 film prior to the MLD. The Al2O3 film fracture allowed the MLD film to anchor firmly to the PES substrate by MLD in the pores of the PES porous substrate. The distance between the anchor points was a function of fracture density. This distance could be controlled by applying a tensile stress to the Al2O3 PE-ALD film to fracture the film through sample bending. This method produced firmly anchored polymer MLD films on top of the PES porous substrates.

Welch, Brian C. (ORCID:000000029747045X)↗
DOE OSTI2022

A multiphysics model of the versatile test reactor based on the MOOSE framework

The traditional modeling approach for sodium fast reactor cores relies on separate physics models, where the fuel performance, thermal–hydraulics, and neutronics calculations required to predict the core physics characteristics for nominal conditions are decoupled by relying on user-imposed boundary conditions. Here, this paper aims at evaluating the impact of multiphysics simulations for predicting the core characteristics of the Versatile Test Reactor, which is being designed as a 300-MWt sodium-cooled fast reactor. The purpose of the Versatile Test Reactor is to accelerate the testing of advanced nuclear materials in the United States. The proposed multiphysics model relies on the Griffin reactor physics code, the SAM thermal–hydraulic system code, the BISON fuel performance code, as well as generic Multiphysics Object-Oriented Simulation Environment capabilities implemented in the open-source tensor mechanics module. For k eff calculations, the introduction of a tight coupling between the neutronics, thermo-mechanical and thermal–hydraulics models induces a change of around 543 pcm in the eigenvalue, compared to the traditional standalone neutronics calculation where approximate temperature profiles are used. The multiphysics model is then employed for quantifying the impact of the thermal conductivity uncertainties on some of the key figures of merit, such as the fuel centerline temperature, assembly powers, and keff for nominal core conditions. As anticipated, uncertainties on fuel thermal conductivity mostly impact the fuel centerline temperature, and to a lesser extend the k eff .

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Serpent - Bison - THM Preliminary Multiphysics Modeling of a Nuclear Thermal Propulsion System Fuel Assembly

This work demonstrates the Monte Carlo neutronic and Thermo-Hydraulic coupling scheme using the Serpent code and MOOSE application Bison and Thermo Hydraulic Module. The coupling scheme is then applied to the reference BWX Technologies Nuclear Ther- mal Propulsion system at he fuel assembly level where it’s used to perform an analysis of the isothermal material coefficients and potential material reactivity worth. A method is developed to isolate which feedback effects should be considered for proceeding with reduced order deterministic neutronic modeling where branch off analysis must be con- ducted. The convergence behavior of the coupling scheme is demonstrated where it fol- lows the standard Picard iteration approach. Verification studies for the method of deduc- ing relevant feedback effects is also demonstrated.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Fuel molecular structure effect on autoignition of highly branched iso-alkanes at low-to-intermediate temperatures: Iso-octane versus iso-dodecane

Highly branched iso-alkanes are an important class of hydrocarbons found in conventional petroleum-derived and alternative renewable fuels used for combustion applications. Recognizing that chemical kinetics for most of these iso-alkanes, especially at low-to-intermediate temperatures, has not been well studied, an experimental and modeling investigation of two selected iso-alkanes, iso-octane (2,2,4-trimethylpentane, iC8) and iso-dodecane (2,2,4,6,6-pentamethylheptane, iC12), is conducted to understand the fuel molecular structure effect on their autoignition characteristics. Using a rapid compression machine (RCM), the ignition responses of iC8 and iC12 at varying pressures, temperatures, and equivalence ratios are characterized and compared. The newly-acquired experimental ignition delay times have been compared with the literature RCM and shock tube data, demonstrating the complementary nature of the current dataset. Further comparison of the experimental pressure traces and ignition delay times illustrates the reactivity crossover between iC8 and iC12. Namely, there exists a temperature window in the negative temperature coefficient regime within which iC12 is less reactive than iC8, but iC12 becomes more reactive outside this temperature window. Furthermore, a chemical kinetic model of iso-alkanes including both iC8 and iC12 is developed. Simulated results using this model are then compared to the experimental data obtained in this study and available in the literature, showing its ability to predict the experimental trends. Chemical kinetic analyses have also been conducted to identify the important reaction pathways controlling autoignition at varying conditions, and to elucidate the underlying mechanism leading to different reactivity trends between iC8 and iC12.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

A Review and Revalidation of Equivalency Factors in H-Canyon

In recent years, an effort has begun to validate and reaffirm long-existing limits and equivalency factors in use at the H-Canyon (HCA) facility located at the Savanah River Site. Equivalency factors (EF) are coefficients used to express the reactivity of a quantity of a given isotope in terms of an equivalent amount of another isotope’s reactivity, commonly termed Fissile Gram Equivalent (FGE). For criticality safety purposes, EFs are often used to conservatively determine if a given mixed-isotope system is within previously established limits. These EFs are calculated by taking ratios of established subcritical limits for a given system or process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Equivalency Factors and Mixed Isotope Subcritical Limits [Slides]

Equivalency factors (EF) are coefficients used to express the reactivity of a quantity of a given isotope in terms of an equivalent amount of another isotope’s reactivity. EFs are often used to conservatively determine if a given mixed-isotope system is within previously established limits. These EFs are calculated by taking ratios of established subcritical limits, also called the Rule of Fractions.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Benchmark Calculation for Turkey Point Unit 3 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

Benchmark calculations were performed for Turkey Point Unit 3 cycles 1–3 to validate the SCALE 6.3/Polaris–PARCS v3.4.2 code with the ENDF/B–VII.1 56–group library by comparing the simulated results with the measured data. The benchmark results will be used in evaluating the SCALE/Polaris–PARCS code package’s uncertainties for pressurized water reactor physics analysis. That future analysis will include key nuclear parameters such as reactivity, control bank worth, temperature coefficients, and pin and assembly power peaking factors. The present document details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS, and PARCS. Additional details are provided with respect to the input and output files produced for the benchmark calculations. The benchmark results are summarized such that they can be used in evaluating uncertainties with other benchmark results for key nuclear parameters.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Benchmark Calculation for Surry Unit 1 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

The benchmark calculations were performed for Surry Unit 1 cycles 1–3 to validate the SCALE 6.3/Polaris–Purdue Advanced Reactor Core Simulator (PARCS) v3.4.2 with the ENDF/B–VII.1 56–group library by comparing the simulated results with the measured data. The benchmark results will be used to evaluate uncertainties of the SCALE/Polaris–PARCS code package for pressurized water reactor physics analysis for key nuclear parameters such as reactivity, control bank worth, temperature coefficients, and pin and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS, and PARCS. Additional details are provided for the input and output files produced for the benchmark calculations. The benchmark results were summarized such that they can be used in evaluating uncertainties with other benchmark results for key nuclear parameters.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗