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At least 127 records · Page 7

Carbon Isotope Systematics in Mineral-Catalyzed Hydrothermal Organic Synthesis Processes at High Temperature and Pressures

Observation of methane in the Martian atmosphere has been reported by different detection techniques. Reduction of CO2 and/or CO during serpentization by mineral surface catalyzed Fischer-Tropsch Type (FTT) synthesis may be one possible process responsible for methane generation on Mars. With the evidence a recent study has discovered for serpentinization in deeply buried carbon rich sediments, and more showing extensive water-rock interaction in Martian history, it seems likely that abiotic methane generation via serpentinization reactions may have been common on Mars. Experiments involving mineral-catalyzed hydrothermal organic synthesis processes were conducted at 750 C and 5.5 Kbars. Alkanes, alcohols and carboxylic acids were identified as organic compounds. No "isotopic reversal" of delta C-13 values was observed for alkanes or carboxylic acids, suggesting a different reaction pathway than polymerization. Alcohols were proposed as intermediaries formed on mineral surfaces at experimental conditions. Carbon isotope data were used in this study to unravel the reaction pathways of abiotic formation of organic compounds in hydrothermal systems at high temperatures and pressures. They are instrumental in constraining the origin and evolution history of organic compounds on Mars and other planets.

Fu, Qi↗

Carbon Isotope Measurements of Experimentally-Derived Hydrothermal Mineral-Catalyzed Organic Products by Pyrolysis-Isotope Ratio Mass Spectrometry

We report results of experiments to measure the C isotope composition of mineral catalyzed organic compounds derived from high temperature and high pressure synthesis. These experiments make use of an innovative pyrolysis technique designed to extract and measure C isotopes. To date, our experiments have focused on the pyrolysis and C isotope ratio measurements of low-molecular weight intermediary hydrocarbons (organic acids and alcohols) and serve as a proof of concept for making C and H isotope measurements on more complicated mixtures of solid-phase hydrocarbons and intermediary products produced during high temperature and high pressure synthesis on mineral-catalyzed surfaces. The impetus for this work stems from recently reported observations of methane detected within the Martian atmosphere [1-4], coupled with evidence showing extensive water-rock interaction during Martian history [5-7]. Methane production on Mars could be the result of synthesis by mineral surface-catalyzed reduction of CO2 and/or CO by Fischer-Tropsch Type (FTT) reactions during serpentization reactions [8,9]. Others have conducted experimental studies to show that FTT reactions are plausible mechanisms for low-molecular weight hydrocarbon formation in hydrothermal systems at mid-ocean ridges [10-12]. Further, recent experiments by Fu et al. [13] focus on examining detailed C isotope measurements of hydrocarbons produced by surface-catalyzed mineral reactions. Work described in this paper details the experimental techniques used to measure intermediary organic reaction products (alcohols and organic acids).

Socki, Richard A.↗

Carbon Isotopes of Alkanes in Hydrothermal Abiotic Organic Synthesis Processes at High Temperatures and Pressures: An Experimental Study

Observation of methane in the Martian atmosphere has been reported by different detection techniques [1-4]. With more evidence showing extensive water-rock interaction in Martian history [5-7], abiotic formation by Fischer-Tropsch Type (FTT) synthesis during serpentization reactions may be one possible process responsible for methane generation on Mars [8, 9]. While the experimental studies performed to date leave little doubt that chemical reactions exist for the abiotic synthesis of organic compounds by mineral surface-catalyzed reactions [10-12], little is known about the reaction pathways by which CO2 and/or CO are reduced under hydrothermal conditions. Carbon and hydrogen isotope measurements of alkanes have been used as an effective tool to constrain the origin and reaction pathways of hydrocarbon formation. Alkanes generated by thermal breakdown of high molecular weight organic compounds have carbon and hydrogen isotopic signatures completely distinct from those formed abiotically [13-15]. Recent experimental studies, however, showed that different abiogenic hydrocarbon formation processes (e.g., polymerization vs. depolymerization) may have different carbon and hydrogen isotopic patterns [16]. Results from previous experiments studying decomposition of higher molecular weight organic compounds (lignite) also suggested that pressure could be a crucial factor affecting fractionation of carbon isotopes [17]. Under high pressure conditions, no experimental data are available describing fractionation of carbon isotope during mineral catalyzed FTT synthesis. Thus, hydrothermal experiments present an excellent opportunity to provide the requisite carbon isotope data. Such data can also be used to identify reaction pathways of abiotic organic synthesis under experimental conditions.

Fu, Qi↗

Hydrogen Isotope Measurements of Organic Acids and Alcohols by Pyrolysis-GC-MS-TC-IRMS: Application to Analysis of Experimentally Derived Hydrothermal Mineral-Catalyzed Organic Products

We report results of experiments to measure the H isotope composition of organic acids and alcohols. These experiments make use of a pyroprobe interfaced with a GC and high temperature extraction furnace to make quantitative H isotope measurements. This work compliments our previous work that focused on the extraction and analysis of C isotopes from the same compounds [1]. Together with our carbon isotope analyses our experiments serve as a "proof of concept" for making C and H isotope measurements on more complex mixtures of organic compounds on mineral surfaces in abiotic hydrocarbon formation processes at elevated temperatures and pressures. Our motivation for undertaking this work stems from observations of methane detected within the Martian atmosphere [2-5], coupled with evidence showing extensive water-rock interaction during Mars history [6-8]. Methane production on Mars could be the result of synthesis by mineral surface-catalyzed reduction of CO2 and/or CO by Fischer-Tropsch Type (FTT) reactions during serpentization [9,10]. Others have conducted experimental studies to show that FTT reactions are plausible mechanisms for low-molecular weight hydrocarbon formation in hydrothermal systems at mid-ocean ridges [11-13]. Our H isotope measurements utilize an analytical technique combining Pyrolysis-Gas Chromatograph-Mass Spectrometry-High Temperature Conversion-Isotope Ratio Mass Spectrometry (Py-GC-MS-TC-IRMS). This technique is designed to carry a split of the pyrolyzed GC-separated product to a Thermo DSQII quadrupole mass spectrometer as a means of making qualitative and semi-quantitative compositional measurements of separated organic compounds, therefore both chemical and isotopic measurements can be carried out simultaneously on the same sample.

Socki, Richard A.↗

Whole core analysis of OECD/NEA TVA-WB1 benchmark exercise 1 using continuous-energy Monte Carlo code MVP with JENDL-4.0u and ENDF/B-VII.0

The TVA WB1 benchmark problem was precisely modeled in accordance with the geometry and material specifications, and benchmark exercise 1 was solved with the MVP code and the nuclear data libraries based on JENDL-4.0u (JENDL-4.0 updated file) and ENDF/B-VII.0. The criticality, the control rod bank worth, the isothermal temperature coefficient (ITC), the integral rod worth of bank D, and differential soluble boron worth (DBW) were obtained. The criticality and the control rod bank worth were compared with the measured data and published calculation results with other codes, Serpent and KENO. The criticality and the control rod bank worth calculated by MVP and the libraries agree with the measured and the published calculation results within the criteria specified in the ANSI/ANS standard. From these results, the calculation model for the MVP and the libraries was verified. The other parameters such as ITC, the integral rod worth, and DBW were also compared between JENDL-4.0u and ENDF/B-VII.0. The comparison of the results shows JENDL-4.0u and ENDF/B-VII.0 give comparable results within the standard deviation, the difference between libraries is small. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of a calculation framework coupling Monte Carlo neutronics, subchannel thermal-hydraulics and 1.5D fuel thermo-mechanics

A coupled calculation system consisting of the Monte Carlo neutronic solver SERPENT and the fuel behaviour solver TRANSURANUS was extended with the sub-channel thermal-hydraulic solver SUBCHANFLOW to form a coupled calculation framework for multi-physics reactor analyses. The coupling between the 3 solvers was accomplished via a driver program written in Perl that runs the solvers, processes and transfers the output data between the solvers and monitors the convergence. While some features of the coupling are still under development, a steady-state coupled calculation resulting in convergence was performed for a single hexagonal fuel assembly to demonstrate that the base functionalities work as expected. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Coupled neutronic-thermal-hydraulic simulations of the European SFR core

Within the European SFR (Sodium-cooled Fast Reactor) - Safety Measures Assessment and Research Tools (ESFR-SMART) project, steady-state coupled simulation of the ESFR core has been performed using several core analysis packages, with the objective of quantifying the coupling effect. Focus is on the fuel Doppler effect and coolant expansion effect. Standalone neutronics calculations in TRACE/PARCS (PSI), DYN3D (HZDR) and WIMS (Jacobs) showed superb agreement with the reference Serpent power distribution (root mean square - rms discrepancies of 1.3%, 1.5% and 0.7% respectively). Results for COUNTHER (CIEMAT) were also in reasonable agreement, but with a somewhat higher discrepancy of 3.7%. Temperature distributions from thermal-hydraulic calculations were also compared and are found to be in good agreement. The effect of Doppler and coolant density feedback on core power distribution was predicted by TRACE/PARCS, DYN3D and WIMS to be between 0.4% and 0.8% rms difference in assembly powers. Reactivity coefficients for perturbations in the inlet temperature, flow rate and core power were shown to be negative for these three codes, with values of roughly -0.5 pcm/C. degree, -0.3 pcm/C. degree and -3.5 pcm/C. degree respectively. A preliminary investigation of differential thermal expansion effects indicates that this may have a significant effect on core power distribution of a few %, greater than anticipated a priori and may warrant inclusion in coupled core analysis to ensure the accurate calculation of power distributions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of the TREAT M2 experiment as a transient benchmark

A transient benchmark based on the Transient Reactor Test Facility (TREAT) M2 Calibration experiment (M2-CAL) is under development. TREAT, at Idaho National Laboratory, is a graphite moderated air-cooled research reactor which has been used extensively for fuel material testing under extreme and accident conditions. Accurate benchmark models are a beneficial component in the operations, experimental planning, and development of TREAT. In this work, we present a transient benchmark model for the M2-CAL experiment core loading. The benchmark model incorporates the coupling between the Monte Carlo Code SERPENT and the computational fluid dynamics code OpenFOAM to capture the temperature feedback mechanism. The M2-CAL transient 2580 was simulated in this work. A pre-transient analysis was performed to determine the optimum core composition and conditions before the beginning of the transient. The analysis was validated against historic TREAT kinetic measurements and the worth of the transient rod T-2. The axial power distribution in the flux wire for the M2-CAL experiment was determined and contrasted with the experiment. The model was then used to simulate the M2-CAL transient 2580 experiment based on the reported pre-transient and transient conditions. Several transient observables were calculated and compared to the experiment. The model shows good agreement with the experimental power traces, period, and average increase of power at the power ramp (time > 7.8 s). Inverse point kinetics analysis was performed during the period of the power ramp for the model and the experiment based on that, the temperature feedback component was isolated and contrasted with the experimental feedback. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Activation of lead coolant in Westinghouse lead fast reactor

The Westinghouse Lead-cooled Fast Reactor (LFR) is a medium-sized, passively safe, scalable reactor being developed by Westinghouse in collaboration with domestic and international organizations with the primary goals of reducing front-end capital cost and generating flexible and cost-competitive electricity while satisfying the highest standards in terms of safety and sustainability. An important aspect of new plant design is an assessment of activation level of its components so that an appropriate nuclear waste management plan can be established. As lead coolant is a distinguishing design feature of the LFR, its activation due to irradiation and the implications on establishing an appropriate waste disposal approach are important for engagement of project stakeholders, customers and nuclear safety regulators. This paper provides a brief description of the Westinghouse LFR design, a background on lead activation phenomenology, and then describes the methodology used for lead activation analysis performed with the computer code Serpent 2. It was determined that impurities naturally occurring in lead have the biggest contribution to lead activation. Corrosion and erosion products from the components immersed in lead can also be important contributors to lead activity and their impact can be limited by appropriate selection of materials and operating conditions. Polonium isotopes, resulting from bismuth generated from lead irradiation and its impurities, bring a notable contribution to the inhalation and ingestion doses during the plant operation and in the first years after the plant shutdown. However, it is not by far as important as for lead-bismuth coolant.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of k{sub eff} in the unfolding process of the LOTUS reactor design

An innovative small thermal reactor concept named LOTUS, which is a pool-type reactor aimed at heat supply. The passive concept applied in the LOTUS and it has a separable core. It means that its core can passively unfold into four parts along its central axis only with the buoyancy and gravity when some accidents happen, especially the wall rupture of the pool caused by earthquake, aircraft impact and so on. This idea can effectively enhance the ability of pool-type reactor for resisting some severe accidents. We have executed some pre-computations in order to check a non-supercritical system for the transient simulation. In this paper, the development of k{sub eff} have been calculated by both stepwise steady-state and transient modes using Serpent, as well as including some sensitivity analyses about the k{sub eff} to unfolding velocity, inserting velocity of control rod (CR) and time step. The result showed that k{sub eff} value decreased definitely as the reactor unfolded and the consistent change exposed between the stepwise steady state and transient result when the unfolding angle varied from 0 to 1 degree. However, the stepwise steady state k{sub eff} negatively declined after that angle, which demonstrated that for such a dynamic unfolding change the transient mode was more suitable, peculiarly for considering delayed neutrons and neutron moderation. For sensitivity calculation, the k{sub eff} decreased more rapidly as the unfolding velocity and inserting velocity of CR increased, and a smaller time step corresponded to a more stable change. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of the MCNP-ORIGEN activation automation tool

This paper introduces the MCNP-ORIGEN activation automation tool for streamlining the calculation of experiment source terms. This tool couples the Monte-Carlo radiation-transport solver, MCNP to the depletion tool, ORIGEN-S. To showcase its current capabilities, this paper presents an activation analysis exercise, which is conducted with Serpent, and the results are com- pared. The experiment consists in a 90%-enriched uranium sphere surrounded by light water. A non-fissile cylinder is placed in the water, representing an irradiation experiment. We conducted simulations for 2 different cylinder materials: iron and cobalt at two different temperatures: 300 K and 900 K. The exercise consists of an irradiation of 50 days at a constant power of 5 MW, followed by a decay of 50 days. These studies highlight the discrepancies between how the different tools handle nuclear data. Finally, the results underscore multiple areas of possible improvement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Investigation of the Impact of TSL Data Libraries and Geometry Variations on the MSRE Benchmark

The increasing demand for green, low-carbon energy solutions has amplified the focus on advanced nuclear reactor technologies. Among these, Molten Salt Reactors (MSRs) have been spotlighted because of their special features. Historically, Oak Ridge National Laboratory (ORNL) started the Molten Salt Reactor Experiment (MSRE) in 1956. Later, ORNL and the University of California, Berkeley (UCB) jointly developed the MSRE benchmark. This was then reviewed by the International Reactor Physics Experiment Evaluation Project (IRPhEP) committee and added to their 2019 handbook. In the current study, the benchmark model was recreated via the Serpent code. The calculated effective multiplication factor was 1.02087 ± 0.00019, which gives a deviation of about 2000 pcm compared to the benchmark/experimental result. At North Carolina State University (NCSU), recent evaluations were conducted on thermal scattering cross sections for molten salt FLiBe and 20% porous nuclear graphite, anticipated for incorporation in the ENDF/B-VIII.1 library. The influences of these thermal scattering law (TSL) data and geometry changes on the multiplication factor were examined using the ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. Introducing the TSL data for FLiBe and 20% porous graphite resulted in an effective multiplication factor of +270 pcm compared to the reference value obtained with ENDF/B-VII.1 in this study. The adjustment of the reactor vessel dimensions resulted in decrease of -448 pcm in the calculated effective multiplication factor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Progress toward the development of a high-fidelity neutronic \model of MARVEL using Griffin

The Microreactor Applications Research Validation and Evaluation (MARVEL) is a 100 kWth microreactor envisioned to start operations at Idaho National Laboratory (INL) in early 2025. This paper is focused on the development of a high-fidelity neutronic model of the MARVEL reactor using Griffin. In particular, the step-by-step verification procedure against Serpent reference calculations is here reported. In this version of the paper, the comparison is only performed in terms of effective multiplication factor, root mean square percentage error (RMSPE) of the pin power, and maximum error for pin-wise power.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Multi-Physics Investigation of a Natural Circulation Molten Salt Micro-Reactor that Utilizes an Experimental In-Pile Device to Improve Core Physics and System Thermal-Hydraulic Performance.

INL employee PhD Dissertation - The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. This thesis investigates the performance benefit of a new device invented by the doctoral candidate. The device is referred to as a Wrapped Helix around an Inclined Plane or WHIP. The WHIP is protected under a provisional patent filed with the USA Patent Office on 28 September 2021 under application number 63/261,776, BEA docket number BA-1254. The WHIP device can be located in-core or near-core to promoted enhanced thermal-hydraulics and neutronics performance. While the WHIP can be employed in a variety of solid or liquid fuel designs, this thesis investigates the device’s benefit in the application of a natural-circulation, micro-molten salt nuclear battery concept (MsNB). This thesis will specifically investigate the temperature coefficient of the MsNB fuel (FLiNaK) using novel temperature sensitivity techniques unique to Serpent particle transport code, evaluate the WHIP’s thermal-hydraulic and neutronic performance effects using both established (STAR-CCM+) and novel (Python code developed to estimate the circulation effective delayed neutron fraction, ßef f ) analytical and numerical methods, evaluate the neutron noise behavior of the MsNB and how the WHIP may alter the character of the MsNB’s transfer function, and how the WHIP affects the autonomous, load-following performance under transient power conditions using Python code developed by the candidate. The sum body of this work, in part, has been published in three journal articles as the timing of the provisional patent process has allowed. Results show that creative utilization of WHIP engineering design and function reduces reactor system volume, fuel loading, control/stability in the buoyant regime

buoyant flow↗

Summary Report Of The FY25 Reactor Physics Verification And Validation Exercises In The Advanced Reactor Technologies - Gas-cooled Reactor Program

Valdiation and verification of numerical tools is critical for ensuring reasonable predictions for design scoping, licensing, and safety analsyis. In this report, two reactor physics verification and validation exercises are presented. The first of these exercises focuses on burnup analysis with data from the Advanced Gas Reactor (AGR) program. Simulations are performed with Monte Carlo N-Particle (MCNP) and are compared with the experimental measurements for the AGR 1 and 2 experiments that utilize both UCO and UO2 fuel. The second exercises utilizes data from the HTR-Proteus experiments to perform reactor physics validation. Specifications of the experimental facility are provdied, along with a demonstration of initial modeling efforts in Serpent for one of the determistic packing experiments. Both cases are part of the Generation-IV international forum (GIF) Very High-Temperature Reactor (VHTR) Computational Methods, Validation, and Benchmarking (CMVB) program, an international collaborative organization dedicated to the verification and validation of High-Temperature Gas-Cooled Reactor (HTGR) analysis. Participation in the CMVB allows the US Department of Energy (DOE) to leverage these existing validation activities to provide extra value through benchmarking activities with other CMVB members.

and Benchmarking (CMVB) program↗

The HTR-Proteus Benchmark: Analysis and Use as a Verification and Validation Case

This presentation outlines the evaluation and application of the HTR-Proteus benchmark as a verification and validation (V&V) case for advanced reactor modeling tools. The work supports the U.S. Department of Energy’s HALEU Availability Program (HAP) and the joint DOE/NRC DNCSH project, which aims to reduce criticality safety uncertainties in commercial-scale HALEU fuel cycle and transportation systems. The HTR-Proteus experiments, conducted at the Paul Scherrer Institute, provide high-fidelity data for TRISO-fueled, graphite-moderated pebble bed reactors with high neutron leakage—conditions relevant to HALEU transport scenarios. This study focuses on Cores 4.2 and 4.3 of the HTR-Proteus benchmark, analyzing key sources of uncertainty including pebble packing, TRISO particle positioning, and core height. Using Project Chrono for realistic pebble geometries and Serpent, SHIFT, and MCNP for neutronics simulations, the study quantifies the impact of these uncertainties on the effective multiplication factor (keff). Results show that a sample size of 110 pebble configurations is sufficient to converge keff, with ±30 pcm uncertainty due to packing randomness. TRISO positioning and core height variations also significantly influence keff, highlighting the importance of detailed modeling in V&V efforts. The benchmark serves as a valuable test case for validating the Griffin reactor physics code and improving confidence in HALEU system simulations.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗

A multiscale approach to analyze the effect of radial void fraction distributions on BWR neutronic lattice parameters

The goal of this study is to analyze the effect of radial void fraction distribution, considered at three degrees of fidelity, on lattice and 2-group homogenized cross sections. The novelty of this work lies in its use of experimentally-determined radial void fraction distributions which allows for fine-scale coolant density data with no predictive uncertainty and which could be prohibitively expensive to simulate over the range of operational parameters used in the experiments. The experimentally-determined radial void fraction distributions are also characterized using a statistical tool called a semi-variogram in order to determine the scale of continuity. Void fraction distributions are averaged at the lattice and subchannel level for comparison with the fine-scale model. Models averaged at the subchannel and lattice level show a difference in about 68% of the time. Also, by comparing 2-group homogenized cross-sections, the differences between the subchannel and homogeneous cases are measurable.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Monte Carlo neutronics benchmarks on nuclear fuel depletion: A review

Monte Carlo (MC) neutronics codes are used widely for academic and industrial needs. Several schemes of coupling MC neutronics code with isotope generation and depletion code exist, which are used for performing nuclear fuel depletion simulations. These simulations can estimate the inventory of isotopes in neutron irradiated nuclear reactor fuel. However, the accuracy of these simulations shall be validated through experiments. MC codes are seldom validated by isotopic benchmarks compared to criticality benchmarks. This work compiles and analyzes the fuel depletion benchmarks and validations used to analyze the performance of MC-based fuel depletion neutronics codes. Analyses of these benchmarks and validations showed that the computed concentrations of 133 Cs, 135 Cs, 137 Cs, 148 Nd, 239 Pu, 240 Pu, and 241 Pu in the irradiated fuel by the depletion codes agreed with the measured values within 10% error. However, the computed concentrations of 125 Sb, 242 Cm, 243 Cm, 244 Cm, 245 Cm, and 246 Cm had errors more than 15% compared to the measured values. As a result, Ventina depletion code showed the most accurate predictions for the greatest number of isotope concentrations compared to ORIGEN2 and CINDER90.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗