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From Escherichia coli mutant 13 C labeling data to a core kinetic model: A kinetic model parameterization pipeline

Kinetic models of metabolic networks offer the promise of quantitative phenotype prediction. The mechanistic characterization of enzyme catalyzed reactions allows for tracing the effect of perturbations in metabolite concentrations and reaction fluxes in response to genetic and environmental perturbation that are beyond the scope of stoichiometric models. In this study, we develop a two-step computational pipeline for the rapid parameterization of kinetic models of metabolic networks using a curated metabolic model and available 13 C-labeling distributions under multiple genetic and environmental perturbations. The first step involves the elucidation of all intracellular fluxes in a core model of E. coli containing 74 reactions and 61 metabolites using 13 C-Metabolic Flux Analysis ( 13 C-MFA). Here, fluxes corresponding to the mid-exponential growth phase are elucidated for seven single gene deletion mutants from upper glycolysis, pentose phosphate pathway and the Entner-Doudoroff pathway. The computed flux ranges are then used to parameterize the same (i.e., k-ecoli74) core kinetic model for E. coli with 55 substrate-level regulations using the newly developed K-FIT parameterization algorithm. The K-FIT algorithm employs a combination of equation decomposition and iterative solution techniques to evaluate steady-state fluxes in response to genetic perturbations. k-ecoli74 predicted 86% of flux values for strains used during fitting within a single standard deviation of 13 C-MFA estimated values. By performing both tasks using the same network, errors associated with lack of congruity between the two networks are avoided, allowing for seamless integration of data with model building. Product yield predictions and comparison with previously developed kinetic models indicate shifts in flux ranges and the presence or absence of mutant strains delivering flux towards pathways of interest from training data significantly impact predictive capabilities. Using this workflow, the impact of completeness of fluxomic datasets and the importance of specific genetic perturbations on uncertainties in kinetic parameter estimation are evaluated.

59 BASIC BIOLOGICAL SCIENCES↗

Data for “Tree root nutrient uptake kinetics vary with nutrient availability, environmental conditions, and root traits: A global analysis”

This data package contains data and code used in the paper “Tree root nutrient uptake kinetics vary with nutrient availability, environmental conditions, and root traits: A global analysis”. The central product is a global dataset of root inorganic nutrient uptake rates and kinetics parameters covering temperate, boreal, and sub/tropical tree species, representing a collection of nutrient uptake data from published studies. This dataset enables tree investigation of root nutrient uptake rates across species, space, and experimental conditions. The data can also be combined with supplementary data on root and soil traits or with external datasets (e.g. R scripts contained within use data from FRED 3.0; (Iversen et al., 2021)). Contained within is the main nutrient data “uptake_data.csv” as well as 4 additional .csv files that link uptake data to supplementary measurements, source references, taxonomic information, and additional nutrient uptake measurements across nutrient gradients, and 1 .csv file that records meta-analysis results for plotting with the R scripts. There are seven R scripts that support data analysis and creation of the figures in the related publication.

54 ENVIRONMENTAL SCIENCES↗

ONE-DIMENSIONAL TIME TO EXPLOSION (THERMAL SENSITIVITY) TESTS ON PETN, PBX-9407, LX-10, AND LX-17

Incidents caused by fire and combat operations can heat energetic materials that may lead to thermal explosion and result in structural damage and casualty. Some explosives may thermally explode at fairly low temperatures (< 100 C) and the violence from thermal explosion may cause a significant damage. Thus it is important to understand the response of energetic materials to thermal insults. The One Dimensional Time to Explosion (ODTX) system at the Lawrence Livermore National Laboratory has been used for decades to measure times to thermal explosion, threshold thermal explosion temperature, and determine the kinetic parameters of thermal decomposition of energetic materials. Samples of different configurations (pressed part, powder, paste, and liquid) can be tested in the system. The ODTX testing can also provide useful data for assessing the thermal explosion violence of energetic materials. This report summarizes the results of our recent ODTX experiments on PETN powder, PBX-9407 pressed part, LX-10 pressed part, LX-17 pressed part and compares the test data that were obtained decades ago with the older version of ODTX system. Test results show the thermal sensitivity of various materials tested in the following order: PETN> PBX-9407 > LX-10 > LX-17.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and grain boundaries in the ion-irradiated alloys. More significant segregation was observed in the neutron irradiated alloys. For the more concentrated alloys, irradiation did not affect existing Cr segregation to grain boundaries and segregation to dislocation loops was not observed perhaps due to a change in the dislocation loop structure with increasing Cr concentration. Precipitation of α’ was observed in the neutron irradiated alloys containing over 9 at.% Cr. However ion irradiation appears to suppress the precipitation process. Initial low dose ion irradiation experiments strongly suggest a cascade recoil effect. The systematic analysis of grain boundary orientation on Cr segregation was significantly challenged by carbon contamination during ion irradiation or by existing carbon and therefore carbide formation at grain boundaries (sensitization). The combination of the proposed systematic experimental approach with atomistic modeling of diffusion processes directly addresses the programmatic need for developing and benchmarking predictive models for material degradation taking into account atomistic kinetics parameters

36 MATERIALS SCIENCE↗

The Enhancement of Gas Pressure Diagnostics in the P-ODTX System

The One Dimensional Time to Explosion (ODTX) system at the Lawrence Livermore National Laboratory is a useful tool for thermal safety assessment of energetic material. It has been used since 1970s to measure times to explosion, threshold thermal explosion temperature, thermal explosion violence, and determine decomposition kinetic parameters of energetic materials. ODTX data obtained for the last 40 years can be found elsewhere.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Development and Characterization of High Performance and Robust Mixed Conducting Cathodes Supportive of Lower Cost Solid Oxide Fuel Cells (Final Report)

The objective of the proposed research is to advance an existing partnership between the University of South Carolina (USC) and LG Fuel Cell Systems, Inc. (LGFCS) to research and develop mixed electronic and ionic conducting cathode materials for durable and robust low-cost solid oxide fuel cells (SOFCs). The first task is to develop mixed ionic and electronic conducting doped (Pr,Nd)NiO 4 (PNNO)-based cathodes. This task is built on a prior collaborative work with LG Fuel Cell System, which showed that doped PNNO could exhibit phase stability at high temperatures. The initial focus will be on investigating the durability and endurance of its electrochemical performance and the temporal evolution of microstructures. In-situ high temperature neutron diffraction and x-ray diffraction will be used to study the structures of these phase stable compounds. PNNO doped with Ca and transition metal elements will be synthesized. Physical properties of these materials with respect to coefficient of thermal expansion, electrical conductivity, diffusion coefficient, and nonstoichiometry will be measured. The benefits of dense versus porous barriers will be studied and the role of barrier layer will be identified and a guideline will be provided to choose the maximum porosity for durable and robust cathode performance with co-doped PNNO as the cathodes. LSCF6428 will be used to compare its power density with that of co-doped PNNO in the absence of Cr, but some studies comparing Cr tolerance will be performed towards the end of the project. The conditions at which preferred nickelate phases are stable will be defined, including sintering temperature, operating temperature, and composition range. The second task will focus on the investigation of the location of degradation by using differential relaxation time (DRT) analysis, coupled with least squares fitting and systematical post analyses. DRT analysis can provide (1) a direct access to the kinetic parameters of the underlying processes in both cathode and anode, and (2) a clearer picture of SOFC operation, which allows loss factors to the changes of microstructures or activities of either electrode to be distinguished, and a better target fuel cell development to be enabled. The third task aims at applying the accelerated testing protocol to investigate the microstructural evolution of the cathode candidates identified in this research. Technical approaches will be developed to mitigate and improve the durability of cathode materials by choosing the right chemistry and tuning cathode microstructures.

36 MATERIALS SCIENCE↗

Corrosion growth model for previously vacuum-baked LiH

In this report, a hydroxide corrosion growth model is proposed for the reaction of previously vacuum-baked LiH samples with moisture. The model is then validated experimentally by subjecting a previously vacuum-baked polycrystalline LiH sample to 35 ppm of H 2 O at room temperature while monitoring the corrosion growth as a function of time with diffuse-reflectance infrared Fourier transform (DRIFT) spectroscopy. Kinetic parameters extracted from the validating experiments can be used to make long-term prediction.

36 MATERIALS SCIENCE↗

Multiphysics Coupling of PROTEUS-NODAL and SAM for Molten Salt Reactor Simulation

The purpose of this report is to document the new features and capabilities that have been implemented in the PROTEUS-NODAL code and the coupling procedure of PROTEUS-NODAL with SAM to analyze MSRs. This report is organized as follows. Section 2 presents the variational nodal P 1 solver for cylindrical geometries and its verification test results. Section 3 discusses the transient analysis capability implemented in the P 1 solvers of PROTEUS-NODAL. The solution methods for the coupled system of TFSP and precursor drift equations, the coarse mesh finite difference (CMFD) acceleration, the kinetics parameter evaluation method, and the thermal feedback are discussed. The verification test results against the SP 3 solutions are also discussed. Section 4 describes the coupling scheme of PROTEUS-NODAL and SAM under the MOOSE framework for MSR applications. The developed MOOSE applications named “TreeFrog” and “TreeKangaroo” are discussed. Verification test results for the coupled system are also discussed by comparing the MSFR steady state solution with that obtained from a manually coupled PROTEUS-NODAL and ANSYS CFX computational fluid dynamics (CFD) calculations and the MSFR transient analysis results with the open literature solutions and the PROTEUS-NODAL solutions obtained with the standalone thermal-hydraulics solver. Section 5 provides conclusions and future works.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NRC Multiphysics Analysis Capability Deployment (FY2021 - Part 2)

This report details the progress and activities of Idaho National Laboratory (INL) on the Nuclear Regulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.” Task 4b was completed for this report. INL developed a sample problem showing how to use Serpent 2 to calculate macroscopic cross sections for use in Griffin for a typical sodium fast reactor Unprotected Loss of Flow (ULOF) transient. The complicating factor is the relatively large axial gradient of the coolant (and hence fuel) temperature and its evolution during the transient. A 3D Griffin model of the Advanced Burner Test Reactor (ABTR) is coupled to the System Analysis Module (SAM) to perform the ULOF transient. The model includes various explicit feedback mechanisms for fast reactors, including Doppler, radial expansion from the displacement of the support plate, and axial expansion from the displacement of the fuel pins. The results for the ULOF transient are consistent with published values. We observed small effects in the transient results that arise from the superhomogenization equivalence correction of uniform and nonuniform temperature data sets, but they diminish as more dominant reactivity mechanisms are added to the model. Potential improvements to the neutronics model include: adding the differential control rod and sodium feedback mechanisms, implementing cross section re-homogenization based on the partial volume of the various materials, and enhancing the kinetics parameters. Furthermore, the fluids model only includes four subchannels to represent the entire active core due to performance limitations in SAM when attempting to model one subchannel per assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Develop low-temperature diffusional moisture outgassing model for silica-filled silicones

A universal outgassing model for all silica-filled silicones would allow researchers to make moisture outgassing predictions for many different silica-filled silicones using the same sets of kinetic parameters (but with different intensities for the outgassing constituents). In FY 2020, we succeeded in developing such a model for outgassing in a vacuum/dry environment after ~ 1-2 hours of room temperature vacuum pump. That means the FY2020 model does not account for the low temperature diffusional moisture outgassing during the first few hours under vacuum pump. In FY 2021, we have used the experimental technique of dynamic vapor sorption together with the modified Page diffusion model to measure and quantify the low temperature (i. e. room temperature) diffusional moisture release to complete our universal moisture outgassing model for silica-filled silicones from low temperatures to high temperatures.

36 MATERIALS SCIENCE↗

ARDP Natrium Neutronic Methodology: Argonne Neutronic Assessment of ABR-1000

The Natrium Sodium-cooled Fast Reactor (SFR) concept developed by TerraPower, LLC was selected as one of the advanced reactor designs for demonstration under the Advanced Reactor Demonstration Program Natrium Demo project. In collaboration with TerraPower, the Argonne National Laboratory (ANL) team provide independent reviews of the neutronic methodology being used for the Natrium design. As part of this activity, both the TerraPower and ANL teams agreed to perform independent neutronics analyses of the ABR-1000 reactor problem as a demonstration on the usage of the neutronics methodologies. Comparison of the results from TerraPower and ANL would expose any methodological differences in the modeling approaches. This report describes and presents the Argonne analysis work on the ABR-1000, and will be provided to TerraPower, along with data files, to complete the comparison study. The comparison will cover the methodology to evaluate the reactivity coefficients for a safety analysis, the shutdown margin, and the reconstructed pin power distribution. The methods and models used to calculate the kinetic parameters and the reactivity coefficients (density, Doppler, axial/radial expansion, etc.) are described in detail with recommendations of specific method options and modelling techniques. The control system reactivity worth and the excess reactivity due to temperature defect are calculated for the shutdown margin analysis. The method to evaluate the excess reactivity is discussed in detail. The pin power reconstruction methodology used in SE2ANL, SE2RCT, and DASSH are discussed in this report. The pin power distribution evaluated by different methods are compared and discussed in the report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Automated Control for Nuclear Thermal Propulsion Start-Up using MOOSE-based Applications

This report presents a Griffin/Bison/RELAP-7 numerical model of a prototypical NTP system that features fuel assemblies arranged in rings, and which was designed to simulate rapid startup transients. The physics modeled include full-core neutronics, assembly-wise heat conduction, and conjugate heat transfer, with the balance of plant mainly imposed through boundary conditions. In addition, various forms of automated reactivity control were deployed by using the MOOSE to autonomously drive the model and simulate the reactor transitioning from assumed initial conditions to nominal power in a fraction of a minute. To generate the cross-sections of the neutronics model, and in an effort to simultaneously account for the tremendous axial temperature gradients in the reactor and to limit the number of state points required for cross-section generation, the average component temperatures and hydrogen densities in the cooling channels were correlated to the average fuel and moderator temperatures, and fixed axial profiles were derived for nominal conditions and then used during the transient. With this approximation, a tractable cross-section library tabulated with fuel/moderator temperatures and CD angles was generated using Serpent. The full-core SPH correction procedure and the CD decusping technology in Griffin, respectively, ensure preservation of the multiplication factor and reaction rates at state points, along with a reasonably accurate reactivity worth between tabulated CD angles, despite using a coarse mesh. Feedback from other physics was calculated by modeling one representative fuel assembly per ring, along with the corresponding fuel and moderator cooling channels. To limit power overshoots during startup, another layer of multiphysics coupling was added to the model in order to automatically control the drums. Two different technologies presented herein showed outstanding performance in this regard: (1) a novel hybrid PID controller based on both power and reactivity signals, and (2) a PGC that relies on kinetics parameters and reactivity coefficients to predict future behavior and adjust the desired signal accordingly. A challenging benchmark was devised, featuring a power demand curve that exponentially increases by a factor of 500 within 30 seconds, then levels out after that. Both control approaches create a simulated power curve that closely follows the power demand curve and limits power overshoots to 1% or less. While the former approach requires more tuning of the internal parameters, the latter requires additional knowledge of the reactivity feedback coefficients and rates of change of the corresponding variables, including fuel and moderator temperature, which could be difficult to dynamically measure for a real NTP system. Fortunately, some inaccuracy in these quantities will not drastically degrade the PGC performance. Subsequently, a more realistic startup sequence was considered, in which the mass flow rate and outlet pressures are ramped up to model bootstrap and thrust build-up phases prior to reaching steady-state conditions, demonstrating the ability of the hybrid PID and PGCs to handle such transients, with both types of controllers exhibiting very similar behavior. Nevertheless, a significant chamber temperature overshoot was observed, caused by the demanded power signal and assumed mass flow rate. This issue could be mitigated by deploying a reactor controller that follows the chamber temperature signal and actuates both the control valves and drums (rather than using a power signal based solely on the drums to control reactivity). Enhancement of the hydrogen fluid properties available in MOOSE, as well as a better understanding of prototypical initial conditions, are also needed to further enhance this startup model. Finally, a study was performed to model decay heat post-shutdown, and to prepare for extending this model to predict shutdown behavior and post-shutdown pulsed cooling requirements.

33 ADVANCED PROPULSION SYSTEMS↗

Models to Incorporate Reaction Mechanisms into DG-OSPREY: Fixed-Bed Simulations for Organic Iodides Capture Using Ag 0 Z

As one of the most potent radioisotopes released during spent nuclear fuel reprocessing, 129 I is strictly regulated and must be removed before discharge. Organic iodides (primarily alkyl iodides with different chain lengths, i.e., CH 3 I, C 4 H 9 I, and C 12 H 25 I) comprise ~2% of the total iodine in the reprocessing off-gases and are primarily present in vessel off-gas (VOG). Reduced silver mordenite (Ag 0 Z) is predominantly considered for the removal of radioiodine; however, its capture performance and underlying interaction processes with long-chain organic iodides are not fully understood. Two major tasks were accomplished in this study. First, to improve upon the previous experimental studies where Ag 0 Z was used to capture CH 3 I, C 4 H 9 I, and C 12 H 25 I at different concentrations, we comprehensively investigated the corresponding capture mechanisms by characterizing fully loaded Ag 0 Z samples. Second, computational codes were implemented to perform fixed-bed simulations that account for transport and reaction mechanisms. Scanning electron microscopy with energy dispersive X-ray analysis (SEM-EDX), powder X-ray diffraction (PXRD), UV-visible diffuse reflectance spectroscopy (UV-vis DRS), and thermogravimetric analysis (TGA) were conducted on Ag 0 Z samples that are saturated with I 2 , CH 3 I, C 4 H 9 I, and C 12 H 25 I), respectively. Results indicate that AgI is the predominant adsorption product regardless of the adsorbed iodine species, yet alkyl iodides with different carbon chain lengths may have different compositions of α- and γ-AgI. Synchrotron pair distribution function (PDF) measurements and TGA coupled with a Fourier transformed infrared detector (TGA FTIR) have been performed, and experimental data are currently analyzed. Results are expected to provide further insights into the adsorption mechanisms. The fixed-bed capture performance for CH 3 I was successfully simulated by solving mathematical equations that describe the underlying transport processes and adsorption reactions. The computational framework, Catalytic After Treatment System (CATS), that was originated in our research group, was used to solve the governing equations. Kinetic parameters, including the pore diffusivity and reaction rate constant were obtained by optimization techniques using data from thin-bed experiments performed at Oak Ridge National Laboratory. The performance of a deep bed predicted using the optimized parameters showed promising agreement with the experimental data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Outgassing of casted and 3D printed silicones

Outgassing of 3D printed and casted silica-filled silicones was studied using a standard ASTM E-595 procedure and a modified Soreq’s method that allows very sensitive measurements of the outgassing kinetics even at low temperatures, 20 – 50 °C. We studied three different silica-filled silicones, non-porous cased M9787, and 3D printed LL50 and 40P materials. The present report shows that vacuum outgassing products of polysiloxane-based materials include multiple species. In addition to chemisorbed and physisorbed water outgassing, previously studied by TPD, polymer fragments of varying sizes evaporate as well. The chemical composition of the condensable volatile species collected on a Germanium substrate, was analyzed by FTIR. It was shown that all studied materials produce mostly PDMS-based outgassing products that form stable droplet-like structure on the Ge substrate. The highest amount of the condensable volatile species was measured for LL50. Isothermal TGA in nitrogen atmosphere of the LL50 material was compared to the standard vacuum outgassing. The results of TGA showed lower mass loss and higher water regain. The kinetics of the outgassing at relatively low temperatures of 20, 30 and 50 °C was studied using Thermoelectric Quartz Crystal Microbalance (TQCM). The TQCM surface was held at -10 °C to enhance condensation of the outgassed moisture. The analysis presented here, when combined with the extensive previous work on moisture outgassing, provides a better understanding of the aging mechanism of polysiloxane-based materials. In this work we show that outgassing techniques, usually applied for qualification of polymers for space-related applications, may be used as a complementary method to study thermal degradation and aging of silica-based silicones. Our results indicate that the long-term degradation mechanics and thermal degradation products depend primarily on the chemical composition of silicone-based material and only slightly affected by the manufacturing process. Surprisingly, the surface-to-volume ratio of the 3D printed vs. casted/molded materials showed only minor effect on the outgassing kinetic parameters.

36 MATERIALS SCIENCE↗

CSEWG Meeting Minutes

Work has been done at LANL to validate various updated evaluations from ENDF/BVIII.1b0 including: Fe, Cu, Si, U, Pu. Continuous energy tables were processed with NJOY2016 and output to ACE files. For non-Pu legacy critical benchmarks, all calculations were within the experimental benchmark uncertainty, with some benchmarks having shown lower bias using ENDF/B-VIII.1b0. In HST systems, ENDF/B-VIII.0 shows reasonable results above the thermal leakage fraction. ENDF/B-VIII.0.1 is doing slightly worse than ENDF/B- VIII.0. b0.2 and it also includes a new 19 F evaluation, so more work is needed to examine its impact. 233 U thermal and intermediate bare and water-reflected benchmarks agree well with ENDF/B-VIII.0. From ENDF/B-VIII.0 to ENDF/B-VIII.0b0.1, HMF-007 C/E are increasing and the new 19 F evaluation is helping the HMF-007 cases containing Teflon which have historically had high C/E. The Pu legacy benchmark suite is within experimental uncertainty with the exception of THOR. The INDEN evaluation followed by LANL gave the lowest mean absolute bias, but it remains difficult to make a determination on which Pu evaluation is the best. INDEN performs slightly better for the JEZEBEL series, while the LANL Pu evaluation is better for THOR and flattop benchmarks. For the PMI benchmark series INDEN outperforms the LLNL evaluation. For the PST benchmark series ENDF/B-VIII.0 and LLNL Pu evaluations are better than the INDEN evaluation. C/E ratios for Jezebel and Godiva reaction rate ratios show all the Pu evaluations are very close to one another, making it again difficult to say which is best. When examining neutron reactor kinetics parameters, each Pu evaluation is once again very close. Each individual Pu evaluation seems to have advantages in their own applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Fluoride-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model

In this report we present work performed in Fiscal Year 2022 that demonstrates the modeling and simulation of a fully coupled neutronics thermal hydraulics reference plant model for a fluoride-cooled high-temperature pebble-bed reactor. The multiphysics model is developed on the Nuclear Regulatory Commission’s Comprehensive Reactor Analysis Bundle (BlueCRAB) available on the Idaho National Laboratory’s high-performance computer, which natively and seamlessly couples Griffin, Pronghorn, and the BISON Multiphysics Object-Oriented Simulation Environment based applications. Griffin provides reactor physics capabilities, including depletion to the equilibrium core, k-eigenvalue, adjoint, and transient. The unique direct equilibrium core capability in Griffin is based on a streamline methodology to spatially deplete the pebbles into burnup groups. Pronghorn solves the porous medium equations for the fluid regions and conduction in the solid regions and incorporates a fluidic diode model to simulate the transition from forced to natural convection during accident scenarios. MOOSE modules solves thermal conduction problems for the pebbles and tristructural isotropic in the pebble-bed core, thus providing the fuel and moderator spatial fields for each pebble burnup group. The neutronics feedback relies primarily on fuel, moderator, and reflector temperatures as as well as the FLiBe salt density. Here, we present results for the uncoupled equilibrium core and perform comparisons to equivalent Monte Carlo models. The power distributions and kinetic parameters obtained with Griffin are consistent with those computed with Griffin. We demonstrate a noticeable improvement with the use of discrete ordinates method (SN) transport. The coupled steady-state equilibrium core provides the initial condition for two time-dependent problems: a control rod withdrawal event and an unprotected loss of flow event. In both cases, the reactor design is self-stabilizing and the solutions are consistent with the expected physics. Although this model is prototypical regarding BlueCRAB’s capabilities, its results are consistent with published work by Kairos Power and other research entities. Significant improvements to the model are planned in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Fiscal Year 2023 Software Quality Assurance Activities for the ARC Software

The Argonne Reactor Code (ARC) software suite [1-17] has been developed by Argonne researchers for fast reactor design and analysis since the 1970s. With the ARC software suite, a user can quickly build a model of a proposed or existing fast spectrum reactor and carry out fuel cycle, nominal thermal analysis and flow requirements, and assess, as is appropriate, whether the core design and constraint system yield an acceptable mechanical behavior. For transient reactor analysis with SAS4A [18], the ARC software suite can be used to generate reactivity coefficients and kinetics parameters at any modeled fuel cycle time point which forms part of the input to SAS4A. The ARC suite was consistently being developed until the 1990s and followed a software QA program which was an appropriate standard for the time. In the 1990s, the DOE funding to fast reactor research and development was all but eliminated and the ARC software was put into maintenance mode. In the early 2000s, the software quality assurance (SQA) program for ARC was still in place to define an official version, but by 2005 it all but was abandoned as there were insufficient staff to fill the work roles. Since 2005, there has been a considerable increase in research and design work on fast spectrum reactors. The ARC software as a whole has since been exported to many universities and commercial companies and ANL support has been given to the various projects over the years [19-23]. Further, MC 2 -3, PERSENT, and DASSH were all developed after 2005 without any adherence to a software standard. In recent time, the DOE VTR project [22] paid for verification work to be done on the ARC software as part of the goal of making it NQA-1 complaint. The VTR project was not considered the appropriate pathway to fund and maintain a SQA program for the ARC software and while software developments (DASSH) were made and several manuals were updated and software verification work was carried out, the ARC software is not NQA-1 compliant. More recently the Advanced Reactor Development Program (ARDP [23]) has funded the creation of manuals for some ARC utility programs and funded additional software verification work on DIF3D [6, 7] and MC 2 -3 [2-5] for the purpose of commercial grade dedication. Because of the VTR and ARDP projects, software verification work was completed on MC 2 -3 and DIF3D, and detailed reports were created for each piece of software, which discuss the inputs and outputs from the codes that are covered by the verification work and link various analytic, code-to-code, and hand calculation based verification work presented in the report with verification test problems provided with the software. This is a key part of the commercial grade dedication work and constitutes the bulk of the cost to get the ARC software to commercial grade. The ARC software suite is a valuable asset as a fast reactor design and analysis tool set that has been reasonably well verified and validated with various fast reactor benchmark problems and experiments over decades. Some or all of the ARC software suite has been utilized for designing the IFR [20], PGSFR [21], VTR [22], and Natrium [23] reactors and we can expect it to continue to be used for advanced fast reactor design and/or confirmatory calculation purposes in the future. Due to increased interest by commercial companies and regulatory bodies, it is becoming more important to make the ARC software suite complete and ready-to-use in terms of its SQA pedigree and commercial grade dedication needs. This report discusses the achievements made towards building a new SQA program for the ARC software and dealing with outstanding identified QA gaps.

97 MATHEMATICS AND COMPUTING↗

Advancements in Multiphysics Microdepletion Analysis of an eVinci TM -like Microreactor Leveraging OpenMC-CRAB Workflow

Nuclear microreactors (MRs) are a class of nuclear reactor technology, characterized by reduced dimensions, modular design, and reduced power output in contrast to conventional Light Water Reactors (LWRs). MRs are proposed for supplying electricity and eventual process heat to remote locations, such as military installations and disaster-affected areas. Current research work sponsored by the US Department of Energy Microreactor Program (MRP) is devoted to the development of novel modeling and simulation tools to better support MR vendors and regulatory bodies. Notably, the NRC is projected to utilize the CRAB multiphysics software driver for executing both design and beyond-design-basis accident analyses. Furthermore, the NRC has been utilizing the MELCOR code to calculate mechanistic source terms during accidents. Since MELCOR relies on isotopic inventory and reactor temperature/power profiles under accident conditions, which theoretically can be derived from CRAB, the goal is to establish a comprehensive CRAB-MELCOR computational framework. Past work was focused on testing and demonstrating CRAB's capability to generate results that can be used to inform mechanistic source term calculations in MELCOR. In particular, a computational workflow leveraging OpenMC-generated microscopic cross sections and CRAB was first applied to perform multiphysics microscopic depletion calculation followed by an accident scenario for a stylized microreactor problem. In fiscal year 2024, the research work has been focused on applying the OpenMC-CRAB workflow, which was first tested in fiscal year 2023, to a realistic 3D heat-pipe cooled MR problem representative of the eVinci TM design. The latter computational problem was developed with inputs from WEC to conserve selected neutronic and thermal characteristics of the eVinci TM design without releasing proprietary data. The results of this simulation, encompassing isotopic inventory, power density distribution, and kinetic parameters, will inform both MELCOR and the WEC-developed FATE code for mechanistic source terms calculations. The results from the two codes will then be compared for code verification purposes. This report contains the design characteristics of the realist heat pipe cooled microreactor developed as a use-case for the verification exercise, and the current results for the multiphysics microscopic depletion performed with the OpenMC-CRAB workflow. The results include eigenvalue as a function of time, power distribution at EOL, in addition to nuclides inventory's time evolution and spatial distribution. Finally, we report improvements to the workflow efficiency achieved through a collaboration with the NEAMS programs. Through this collaborative effort, we were able to strongly decrease the computational time for the multiphysics microdepletion calculation (i.e., from 17.4 hours to 5.7 hours on 280 processors) in addition to simplifying the interface to generate isotopics spatial distribution utilizable by FATE and MELCOR. Future work, including the improvement of the current microscopic cross-sections' library and the simulation of an accident scenario at EOL, is also discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗