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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

A multi-group Monte Carlo core analysis method and its application in SCWR design

Complex geometry and spectrum have been the characteristics of many newly developed nuclear energy systems, so the suitability and precision of the traditional deterministic codes are doubtable while being applied to simulate these systems. On the contrary, the Monte Carlo method has the inherent advantages of dealing with complex geometry and spectrum. The main disadvantage of Monte Carlo method is that it takes long time to get reliable results, so the efficiency is too low for the ordinary core designs. A new Monte Carlo core analysis scheme is developed, aimed to increase the calculation efficiency. It is finished in two steps: Firstly, the assembly level simulation is performed by continuous energy Monte Carlo method, which is suitable for any geometry and spectrum configuration, and the assembly multi-group constants are tallied at the same time; Secondly, the core level calculation is performed by multi-group Monte Carlo method, using the assembly group constants generated in the first step. Compared with the heterogeneous Monte Carlo calculations of the whole core, this two-step scheme is more efficient, and the precision is acceptable for the preliminary analysis of novel nuclear systems. Using this core analysis scheme, a SCWR core was designed based on a new SCWR assembly design. The core output is about 1,100 MWe, and a cycle length of about 550 EFPDs can be achieved with 3-batch refueling pattern. The average and maximum discharge burn-up are about 53.5 and 60.9 MWD/kgU respectively. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Host genetic variation drives the differentiation in the ecological role of the native Miscanthus root-associated microbiome

Microbiome recruitment is influenced by plant host, but how host plant impacts the assembly, functions, and interactions of perennial plant root microbiomes is poorly understood. Here we examined prokaryotic and fungal communities between rhizosphere soils and the root endophytic compartment in two native Miscanthus species (Miscanthus sinensis and Miscanthus floridulus) of Taiwan and further explored the roles of host plant on root-associated microbiomes. Our results suggest that host plant genetic variation, edaphic factors, and site had effects on the root endophytic and rhizosphere soil microbial community compositions in both Miscanthus sinensis and Miscanthus floridulus, with a greater effect of plant genetic variation observed for the root endophytic communities. Host plant genetic variation also exerted a stronger effect on core prokaryotic communities than on non-core prokaryotic communities in each microhabitat of two Miscanthus species. From rhizosphere soils to root endophytes, prokaryotic co-occurrence network stability increased, but fungal co-occurrence network stability decreased. Furthermore, we found root endophytic microbial communities in two Miscanthus species were more strongly driven by deterministic processes rather than stochastic processes. Root-enriched prokaryotic OTUs belong to Gammaproteobacteria, Alphaproteobacteria, Betaproteobacteria, Sphingobacteriia, and [Saprospirae] both in two Miscanthus species, while prokaryotic taxa enriched in the rhizosphere soil are widely distributed among different phyla. We provide empirical evidence that host genetic variation plays important roles in root-associated microbiome in Miscanthus. The results of this study have implications for future bioenergy crop management by providing baseline data to inform translational research to harness the plant microbiome to sustainably increase agriculture productivity.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Ion-Assisted Nanoscale Material Engineering in Atomic Layers

Achieving deterministic control over the properties of low-dimensional materials with nanoscale precision is a long-sought goal. Mastering this capability has a transformative effect on the design of multifunctional electrical and optical devices. Here, we present an ion-assisted synthetic technique that enables precise control over the material composition and energy landscape of two-dimensional (2D) atomic crystals. Our method transforms binary transition-metal dichalcogenides, like MoSe2, into ternary MoS2αSe2(1-α) alloys with systematically adjustable compositions, α. By piecewise assembly of the lateral, compositionally modulated MoS2αSe2(1-α) segments within 2D atomic layers, we present a synthetic pathway toward the realization of multicompositional designer materials. Our technique enables the fabrication of advanced 2D structures with arbitrary boundaries, dimensions as small as 30 nm, and fully customizable energy landscapes. Our optical characterizations further showcase the potential for implementing tailored optoelectronics in these engineered 2D crystals.

2D materials↗
DOE OSTI2022

Variance reduction techniques for Monte Carlo neutron noise simulations

The small fluctuations of the neutron flux caused by small perturbations of the macroscopic cross-sections take the name of neutron noise. Advanced Monte Carlo methods have been recently proposed in order to solve the neutron noise equations in the frequency domain, which allows establishing reference solutions to validate faster but approximate deterministic solvers. Due to the presence of particles carrying two statistical weights (for the real and imaginary components of the noise field), both of which may be positive or negative, Monte-Carlo simulations of neutron noise pose distinct challenges in terms of variance reduction. In this work we investigate two variance-reduction techniques, namely branchless collisions and weight cancellation, and probe their effectiveness for a benchmark con- figuration concerning the noise field induced by a pin with oscillating cross sections in a fuel assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Interactive Visualization of Multi-Group Cross Sections on High-Fidelity Spatial Meshes

Core simulation tools utilizing Monte Carlo transport methods represent the 'Holy Grail' for nuclear reactor analysis. However, deterministic transport methods pose a number of advantages to Monte Carlo from a computational perspective, which seems likely to make these methods more viable in the near to intermediate term. Deterministic methods almost universally discretize the energy domain in a few to hundreds of energy groups. One of the key challenges for deterministic transport-based methods to be viable is accurate, reactor agnostic multi-group cross section generation. This paper describes tools for data processing and visualization of multigroup cross sections generated on high-fidelity spatial meshes from the OpenMC Monte Carlo code. Next generation deterministic transport reactor core modeling tools will require multi-group cross section (MGXS) generation on high-fidelity spatial meshes. High-fidelity MGXS are necessary for the accurate simulation of neutron physics in nuclear reactors with a high degree of spectral heterogeneity - e.g., large localized neutron flux gradients throughout a reactor core. Although many local (e.g., pin cell) approximations have been developed to model the scalar flux to compute spectral-averaged MGXS, these techniques cannot account for complex intra-pin and intra-assembly spatial self-shielding effects. Monte Carlo presents the most accurate method for reactor agnostic multi-group cross section generation since it does not require the use of local approximations to the flux. Monte Carlo has been proposed as a highly desirable framework for reactor agnostic cross section generation for many years. Monte Carlo methods can effectively embed all of the relevant spectral effects into the MGXS through the random particle tracking and interaction sampling processes. Previous work has applied MCNP and Serpent to compute few-group constants for diffusion calculations, and OpenMC to compute multi-group cross sections for transport-based methods. The primary difficulty with using Monte Carlo for MGXS generation is the tremendous computational resources required to adequately converge multi-group cross sections in each spatial region and energy group. Recent development of the OpenMC Monte Carlo code for high-performance computing platforms has made multigroup cross section generation from Monte Carlo more tractable in the foreseeable future. The recent release of OpenMC v0.7.1 includes a new Python-based module for MGXS generation. The MGXS generation module leverages many of the algorithms and features for 'big data' analysis of large tally datasets on complex spatial meshes. This paper describes the data management workflow and visualization tools which have been developed to explore the plethora of MGXS data generated with OpenMC. It is the authors' hope that the development of the visualization tool described in this paper will ultimately enable the design of machine learning algorithms to accelerate the convergence rate of high-fidelity multi-group cross section generated with Monte Carlo. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Evaluation and Testing of the ADVANTG Code on SNM Detection

Pacific Northwest National Laboratory (PNNL) has been tasked with evaluating the effectiveness of ORNL’s new hybrid transport code, ADVANTG, on scenarios of interest to our NA-22 sponsor, specifically of detection of diversion of special nuclear material (SNM). PNNL staff have determined that acquisition and installation of ADVANTG was relatively straightforward for a code in its phase of development, but probably not yet sufficient for mass distribution to the general user. PNNL staff also determined that with little effort, ADVANTG generated weight windows that typically worked for the problems and generated results consistent with MCNP. With slightly greater effort of choosing a finer mesh around detectors or sample reaction tally regions, the figure of merit (FOM) could be further improved in most cases. This does take some limited knowledge of deterministic transport methods. The FOM could also be increased by limiting the energy range for a tally to the energy region of greatest interest. It was then found that an MCNP run with the full energy range for the tally showed improved statistics in the region used for the ADVANTG run. The specific case of interest chosen by the sponsor is the CIPN project from Las Alamos National Laboratory (LANL), which is an active interrogation, non-destructive assay (NDA) technique to quantify the fissile content in a spent fuel assembly and is also sensitive to cases of material diversion. Unfortunately, weight windows for the CIPN problem cannot currently be properly generated with ADVANTG due to inadequate accommodations for source definition. ADVANTG requires that a fixed neutron source be defined within the problem and cannot account for neutron multiplication. As such, it is rendered useless in active interrogation scenarios. It is also interesting to note that this is a difficult problem to solve and that the automated weight windows generator in MCNP actually slowed down the problem. Therefore, PNNL had determined that there is not an effective tool available for speeding up MCNP for problems such as the CIPN scenario. With regard to the Benchmark scenarios, ADVANTG performed very well for most of the difficult, long-running, standard radiation detection scenarios. Specifically, run time speedups were observed for spatially large scenarios, or those having significant shielding or scattering geometries. ADVANTG performed on par with existing codes for moderate sized scenarios, or those with little to moderate shielding, or multiple paths to the detectors. ADVANTG ran slower than MCNP for very simply, spatially small cases with little to no shielding that run very quickly anyway. Lastly, ADVANTG could not solve problems that did not consist of fixed source to detector geometries. For example, it could not solve scenarios with multiple detectors or secondary particles, such as active interrogation, neutron induced gamma, or fission neutrons.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

FY2012 Report on Fast Reactor Toolset Work (PROTEUS-Fast)

As part of the NEAMS activity in DOE, a fast reactor simulation program was launched in April 2007, termed SHARP. The initial goal of SHARP was to develop a suite of modern simulation tools specifically for the analysis and design of sodium cooled fast reactors, and this goal has since been extended to consider all reactor types of interest. As part of that work, a high‐fidelity deterministic neutron transport code named UNIC was started, providing a common framework for fast reactor neutronics tools. Focusing on UNIC, spatial de‐homogenization increased the typical neutronics problem with a few hundred million degrees of freedom (DOF) into ones with hundreds of trillions of DOF. Such problems are beyond the capabilities of existing supercomputers. With the advent of future computer hardware, these high‐fidelity problems can be solved but only after a multiyear development program to adapt codes such as UNIC to that technology. UNIC can presently be used on smaller problems (single assembly, multiple assemblies) with today’s hardware, but it is also necessary to provide tools to solve the larger problems of interest, albeit without the aforementioned level of spatial de‐homogenization. To address the desire to demonstrate the coupling capability of NEAMS, part of the FY12 funding was focused on preparing the legacy codes, termed PROTEUS‐Fast, which are fully capable of solving full‐scale fast reactor problems. A bulk of the work this year was focused on preparing and distributing the PROTEUS‐Fast codes for use by all NEAMS users. The demonstration of a full coupling capability will require more follow on work, but significant progress was made this year preparing the interface coupling routines and preparing PROTEUS‐Fast for use. The primary targets of interest in PROTEUS‐Fast are DIF3D, DIF3D‐K (kinetics), and VARI3D (reactivity coefficients), with REBUS (fuel cycle) being an auxiliary component updated for convenience. All codes in PROTEUS‐Fast were verified and added to the BuildBot regression test. Unfortunately, all of the older codes suffered significant problems from the deprecated coding practices used within them and resulting conflicts with modern compilers. After some work, all of these issues were resolved and we feel confident in releasing them for use by others NEAMS members. With regard to specific upgrades, DIF3D‐VARIANT was upgraded with a few months of effort to allow the large scale problems of interest to be executed. After significant study, we choose not to modify VARI3D to include a transport option, but to develop a new code, PERSENT (partially funded by an external project) which uses the DIF3D‐VARIANT transport solution option. The outstanding features missing from MC 2 ‐3 were added, and MC 2 ‐3 has been released to RSICC. All of the remaining PROTEUS‐Fast codes are either finished with the copyright and licensing procedure or are actively going through it. While that process will delay the targeted August 2012 release to RSICC for a few more months, the codes are ready for immediate use and can be connected to the framework either by us or by others as the interface readers and documentation will be released with the PROTEUS‐Fast codes.

Wolters, E. R.↗
DOE OSTI2016

Groundwater–surface water mixing shifts ecological assembly processes and stimulates organic carbon turnover

Environmental transition zones are associated with geochemical gradients that overcome energy limitations to microbial metabolism, resulting in biogeochemical hot spots and moments. Riverine systems where groundwater mixes with surface water (the hyporheic zone) are spatially complex and temporally dynamic, making development of predictive models challenging. Spatial and temporal variations in hyporheic zone microbial communities are a key, but understudied, component of riverine biogeochemical function. To investigate the coupling among groundwater-surface water mixing, microbial communities, and biogeochemistry we applied ecological theory, aqueous biogeochemistry, DNA sequencing, and ultra-high resolution organic carbon profiling to field samples collected across times and locations representing a broad range of mixing conditions. Mixing of groundwater and surface water resulted in a shift from transport-driven stochastic dynamics to a deterministic microbial structure associated with elevated biogeochemical rates. While the dynamics of the hyporheic make predictive modeling a challenge, we provide new knowledge that can improve the tractability of such models.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2012

Probabilistic Modeling of Settlement Risk at Land Disposal Facilities - 12304

The long-term reliability of land disposal facility final cover systems - and therefore the overall waste containment - depends on the distortions imposed on these systems by differential settlement/subsidence. The evaluation of differential settlement is challenging because of the heterogeneity of the waste mass (caused by inconsistent compaction, void space distribution, debris-soil mix ratio, waste material stiffness, time-dependent primary compression of the fine-grained soil matrix, long-term creep settlement of the soil matrix and the debris, etc.) at most land disposal facilities. Deterministic approaches to long-term final cover settlement prediction are not able to capture the spatial variability in the waste mass and sub-grade properties which control differential settlement. An alternative, probabilistic solution is to use random fields to model the waste and sub-grade properties. The modeling effort informs the design, construction, operation, and maintenance of land disposal facilities. A probabilistic method to establish design criteria for waste placement and compaction is introduced using the model. Random fields are ideally suited to problems of differential settlement modeling of highly heterogeneous foundations, such as waste. Random fields model the seemingly random spatial distribution of a design parameter, such as compressibility. When used for design, the use of these models prompts the need for probabilistic design criteria. It also allows for a statistical approach to waste placement acceptance criteria. An example design evaluation was performed, illustrating the use of the probabilistic differential settlement simulation methodology to assemble a design guidance chart. The purpose of this design evaluation is to enable the designer to select optimal initial combinations of design slopes and quality control acceptance criteria that yield an acceptable proportion of post-settlement slopes meeting some design minimum. For this specific example, relative density, which can be determined through field measurements, was selected as the field quality control parameter for waste placement. This technique can be extended to include a rigorous performance-based methodology using other parameters (void space criteria, debris-soil mix ratio, pre-loading, etc.). As shown in this example, each parameter range, or sets of parameter ranges can be selected such that they can result in an acceptable, long-term differential settlement according to the probabilistic model. The methodology can also be used to re-evaluate the long-term differential settlement behavior at closed land disposal facilities to identify, if any, problematic facilities so that remedial action (e.g., reinforcement of upper and intermediate waste layers) can be implemented. Considering the inherent spatial variability in waste and earth materials and the need for engineers to apply sound quantitative practices to engineering analysis, it is important to apply the available probabilistic techniques to problems of differential settlement. One such method to implement probability-based differential settlement analyses for the design of landfill final covers has been presented. The design evaluation technique presented is one tool to bridge the gap from deterministic practice to probabilistic practice. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Multi-cycle reload analysis of a long cycle gas-cooled fast modular reactor

There is currently significant interest in deploying HALEU-fueled fast reactors, including the General Atomics (GA) Fast Modular Reactor (FMR). Such reactors can achieve very long fuel cycles, but with multi-batch loading will take decades to reach equilibrium. This motivates design and analysis of both the initial core and multi-cycle reload, which is typically performed using fast-running, deterministic fast reactor codes such as the Argonne Reactor Computation (ARC) codes. In this paper, multicycle reload of the GA FMR is analyzed using the ARC codes. The GA FMR utilizes 19.75 % enriched fuel in a 16 year cycle with a three-batch strategy, with twice-burned fuel placed on the core periphery. The GA FMR has a softened neutron spectrum due to reflecting elements in the core, so the neutronic solution is first benchmarked against the OpenMC Monte Carlo code. Discrepancy on k eff is 400–600 pcm, likely due to the softened neutron spectrum, heterogeneous fuel assembly design and central reflector. However, the rms discrepancy on the assembly power distribution is only 0.6 %, despite the presence of the central reflector. A reload strategy is devised for the first three cycles of such a reactor, ultimately spanning the first 45–48 years of its operation. The fresh core uses 19.75 %, 19.25 % and 16.75 % enriched fuel in place of fresh, once-burned and twice-burned and is then subsequently refueled with only 19.75 % enriched fuel. The cycle length is varied over 3 cycles of operation to balance fuel utilization and reactor availability, specifically with use of an extended 18-year Cycle 1, followed by a shortened 11-year Cycle 2. Cycle 3 is close to the target 16-year length. Finally, placing twice burned assemblies next to the GA FMR central reflector can reduce power peaking by 3 %, at the expense of slightly reducing the cycle length.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Engineering Structural Transitions in a Multilevel Molecular Switch via Intermolecular Coupling

Controlling molecular conformations with atomic precision is essential for advancing molecular functional electronics, as well as our understanding of molecular dynamics. While switching between bistable molecular conformers is common in nature, creating systems with multiple, addressable states remains synthetically challenging. Here, we demonstrate a bottom-up strategy in which intermolecular interactions give rise to multilevel functionality within a simple two-molecule assembly. Using low-temperature scanning tunneling microscopy, we show that a pyrrolidine dimer on Cu(100) exhibits six distinct adsorption conformations, exceeding the four expected from two independent bistable units. This unusual complexity arises from the interplay between intermolecular van der Waals attraction and steric repulsion, which reshapes the potential energy landscape and changes a single high-energy transition into a sequential two-step pathway. Each step is driven by low-energy inelastic electron excitations, achieving a switching efficiency an order of magnitude higher than that of the monomer. Here, by tuning the bias voltage and tip–molecule distance, we achieve deterministic control over multiple stable states, establishing a general design principle for on-demand engineering of collective molecular behavior and energy-efficient multilevel molecular devices.

Molecular interactions↗
DOE OSTI2019

Peak Ground Displacement Saturates Exactly When Expected: Implications for Earthquake Early Warning

The scaling of rupture properties with magnitude is of critical importance to earthquake early warning (EEW) systems that rely on source characterization using limited snapshots of waveform data. ShakeAlert, a prototype EEW system that is being developed for the western United States, provides real-time estimates of earthquake magnitude based on P-wave peak ground displacements calculated at stations triggered by the event. The algorithms used in ShakeAlert assume that the displacement measurements at each station are statistically independent and that there exists a linear and time-independent relation between log peak ground displacement and earthquake magnitude. Here we challenge this basic assumption using the largest dataset assembled for this purpose to date: a comprehensive database of more than 140,000 vertical-component waveforms from M4.5–M9 earthquakes occurring near Japan from 1997 through 2018 and recorded by the K-NET and KiK-net strong-motion networks. By analyzing the time-evolution of P-wave peak ground displacements for these earthquakes, we show that there is a break, or saturation, in the magnitude-displacement scaling that depends on the length of the measurement time window. We demonstrate that the magnitude at which this saturation occurs is well-explained by a simple and non-deterministic model of earthquake rupture growth. We then utilize the predictions of this saturation model to develop a Bayesian framework for estimating posterior uncertainties in real-time magnitude estimates.

58 GEOSCIENCES↗
DOE OSTI2022

On the effect of scalar flux weighing of linearly anisotropic scattering matrices in few-group transport calculations

The majority of the codes available for homogenized group constant generation for deterministic transport calculations apply the approximation of scalar flux weighting during energy group condensation of higher-order anisotropic scattering matrices. In this paper, we discuss the effect of scalar flux weighting of the linearly anisotropic scattering matrices in the frame of S P{sub 3} and S{sub 12} calculations performed for a two-dimensional VVER-440 reactor benchmark. To compare group constants generated for two neutron energy groups, an infinite pin cell was homogenized with Serpent 2 and ERANOS ECCO. Serpent 2 applies scalar flux, while ERANOS ECCO performs current weighting of the linearly anisotropic scattering matrices during energy group condensation. For analyzing the effect of the various weighting options, three simple reactor models were built assuming different core sizes using standard rectangular assemblies with 15*15 fuel pins. Diffusion, SP{sub 3} and S{sub 12} calculations were performed for the 3 models using group constants generated with Serpent 2 and ERANOS ECCO. The effect of scalar flux weighting of linearly anisotropic scattering matrices is shown by comparing the decrease in reactivity due to the decreased reactor size, as well as the assembly power distribution to reference results obtained with Serpent 2 Monte Carlo calculations. Neglecting higher than linearly anisotropic scattering and indirect application of diffusion coefficients in higher-order transport calculations is advised if angular flux-moment spectra weighted higher-order scattering matrices cannot be generated. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Deterministic Fabrication of Large-Area, High-Crystallinity Oxide Moiré Superlattices

Oxide twistronics extends moiré engineering beyond van der Waals materials, offering a promising platform for accessing emergent interfacial phenomena arising from the strong coupling of lattice, charge, and orbital degrees of freedom in complex oxides. However, deterministic fabrication of high-crystallinity oxide moiré superlattices over large lateral dimensions remains challenging due to the three-dimensional bonding network of oxides. Here, we demonstrate a scalable, generalized fabrication strategy that enables the formation of high-crystallinity oxide moiré superlattices with clean, chemically bonded interfaces and precisely controlled twist angles down to nominal values of 0.1°, achieving subdegree twist-angle accuracy across large contiguous lateral dimensions approaching the millimeter scale. Using NaNbO3 as a model system, we show that the resulting interlayer coupling drives pronounced structural reconstruction that modifies both the phase structure and ferroelectric domain configuration. Synchrotron-based X-ray 3D reciprocal space mapping reveals the emergence of a single-phase state in twisted bilayers, in contrast to the mixed-phase structure observed in single-layer membranes prior to twist assembly. The structural signatures are further consistent with gradual lattice rotation distributed along the thickness direction that may accommodate interfacial shear strain, distinct from reconstruction observed in van der Waals moiré systems which primarily occurs through in-plane stacking rearrangement. This collective lattice response is correlated with twist-dependent nanoscale electromechanical modulations observed by piezoresponse force microscopy. These results establish a scalable materials platform for oxide twistronics and support the implementation of twist-engineered functionalities in practical, macroscale device architectures.

Ghanbari, Reza [North Carolina State University (N↗
DOE OSTI2022

Rotational symmetry relation for efficient response function generation in the coarse mesh transport method COMET

The coarse mesh transport code COMET is a continuous energy hybrid stochastic-deterministic neutronics solver with high fidelity and formidable computation speed in solving reactor core problems. Its method is based on the incident flux expansion theory. In this work, we take advantage of the local geometric symmetry in many reactor cores lattices (e.g., fuel lattices and reflector blocks) to develop relations among the flux response expansion coefficients for symmetric surfaces to further improve the computational efficiency of the COMET response function generation tool (method). This is done by a rigorous derivation of the transformation matrices for the angular and spatial expansion moments resulting from a rotation of a coarse mesh by an arbitrary angle. The relations for the response coefficients for the symmetric surfaces can be then written as the Kronecker product of those transformation matrices. The method is implemented into COMET and tested on two advanced high temperature reactor (AHTR) full-length single assembly benchmark problems. The COMET results using the response function library based on the symmetry relations were compared to those using the library directly generated by continuous energy Monte Carlo for all surfaces. It was found that the eigenvalues and stripe-wise fission densities using the two libraries are in statistical agreement as expected. This indicates that the new method maintains the high fidelity of the original COMET method while improving the computational efficiency in the response function generation by 270% to 400%, depending on the local geometric symmetry. This method also reduces the size of the response function library by the same magnitude (270% to 400%). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Reassessment of uncertainties for the PU-MET-INTER-002 benchmark - 14432

Uncertainty for the PU-MET-INTER-002 benchmark of ZPR-6/10 was unusually large for a ZPR assembly and was dominated by the uncertainties in matrix tube pitch and composition uncertainties for plutonium, graphite and stainless steel plates. The original uncertainty analysis for PU-MET-INTER-002 was limited by computer resources, software and the nuclear data available in 2002, all of which have been improved since. An as-built Monte Carlo model of ZPR-6/10 was used to quantify components of the transformation from the detailed model to a simplified RZ benchmark in 2002. Compositional uncertainties were estimated based upon TWODANT calculations of the RZ benchmark model (20 group, S{sub 8} - P{sub 1}). MC{sup 2}-2 was used with END/B-V.2 data to generate homogenized cross section data for TWODANT. The MC{sup 2}-2 calculations did not account for lattice cell spatial heterogeneity which was driven by the manual process required to generate ∼80 modifications to the deterministic model and by the software limitations of MC{sup 2}-2. Improvements to DIF3D-VARIANT, development of the MC{sup 2}-3 and BuildZPRModel codes, and better computer resources make accurate 3D transport calculations of reactivity changes practical today. BuildZPRModel imports the as-built geometry and creates a 'drawer' homogenized DIF3D model where each radial DIF3D mesh is the exact size of a single ZPR matrix tube and drawer. A MC{sup 2}-3 input is constructed for each drawer to produce homogenized cross section data where the spatial cell heterogeneity is captured by solving a 1D MOC transport calculation on the ultrafine energy group level. Collapsing 3D drawers into 1D MC{sup 2}-3 drawer models follows conventional procedures and is done by BuildZPRModel. MCNP6 was used to re-compute relevant geometry components of the uncertainty using ENDF/B-VII.0 data with the as-built model. Changes in the geometry components compared with earlier results are insignificant. DIF3D calculations of the compositional uncertainties were performed using 33, 70, 116, 230, 425, and 703 group energy structures and selectively combined with P{sub 3}-P{sub 7} transport approximations and P{sub 1} and P{sub 3} scattering kernels. 3D diffusion theory calculations were also performed for comparison. Many composition uncertainty components were essentially unchanged with the updated methodology. However, the dominant experimental uncertainty for PU-MET-INTER-002 is the composition of stainless steel. The new calculations reduce the stainless steel uncertainty from ∼115 pcm to ∼70 pcm which significantly reduces the total uncertainty for PU-MET-INTER-002. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Analysis of the VERA Core Physics Benchmark Problems using cosRMC Code

In reactor physics, two kinds of numerical methods are generally employed to solve the neutron transport equations, one of which is the deterministic method and the other one is the Monte Carlo method. Monte Carlo is widely used to provide reference solution because of its ability to deal with continuous energy cross sections and complex geometry. Just like any statistical process, the Monte Carlo method requires repetition to achieve a small relative uncertainty, and therefore may inevitably have larger simulation times than deterministic method. Fortunately, the great advancements of computer technology make it possible to apply Monte Carlo method in reactor core analysis. To promote nuclear power plant system design and simulation software self-reliance development, a Monte Carlo neutron/photon transport code cosRMC has been developed jointly by State Nuclear Power Software Development Center (SNPSDC) and Tsinghua University. The cosRMC is aimed at achieving full core calculations and analysis with high fidelity and efficiency by means of advanced methodologies and algorithms as well as high performance computing techniques. The cosRMC has many functions as criticality calculation, fixed-source calculation, burnup calculation, etc. In order to ensure its reliability, several types of verification and validation (V and V) work have been and are being carried out for cosRMC. The simulation results of cosRMC based on Problem 1 to 3 of VERA Core Physics Benchmark showed its good capability of 2D lattice and 3D assembly modeling. In this paper, some progressive verification results of multi-assembly and full core problems are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗