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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Nanosecond Repetitively Pulsed (NRP) Plasmas: Relationship Between Induced Flow and Plasma Characteristics at Atmospheric Pressure (Final Technical Report)

The local flow fields induced by plasma discharges at atmospheric pressure can play a critical role in aerodynamic flow and combustion control and have significant implications for plasma-based technologies in medicine and environmental engineering. However, the coupling between the plasma characteristics, the geometry and the induced flow field is not well understood. The primary goal of this work is to determine the relationship between the plasma characteristics and the induced flow field for nanosecond pulsed discharges at atmospheric pressure. A comprehensive experimental investigation of plasmas produced by nanosecond high-voltage pulses in a pin-to-pin electrode configuration was conducted, employing several measurement techniques including microwave and laser Rayleigh scattering, optical emission spectroscopy, and high-speed imaging. The flow field induced by the nanosecond plasma was measured and characterized using measurements of the velocity (particle image velocimetry) and density (background oriented schlieren) with unprecedented spatiotemporal resolution. The experimental measurements were complemented by development of a one-dimensional model of the plasma discharge and a high-fidelity computational fluid dynamics (CFD) simulation of the subsequent plasma-induced flow field.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

MPACT Theory Manual (V.4.3)

This manual presents the theory underlying the three-dimensional (3D) whole-core, pin-resolved neutron transport calculation methodologies employed in the MPACT code. MPACT’s primary goal is to provide accurate sub-pin power distributions in a computationally efficient manner. To accomplish this, MPACT offers several different transport method options. The 2D/1D method, in which 3D problems are decomposed into an axial stack of radial “slices,” is currently the most commonly used. In this option, two-dimensional (2D) planar solutions are provided by the method of characteristics (MOC), and axial solutions are provided via one-dimensional (1D) approximate diffusion, or P3 solutions. The radial and axial solutions are coupled by (i) axial and radial transverse leakages, and (ii) a global 3D coarse mesh finite difference (CMFD) solve, which provides both acceleration and stability to the solution iteration scheme. The subsequent chapters of this manual present a range of topics, including MOC, CMFD, axial nodal transport solvers, 2D/1D, self-shielding, depletion, thermal-hydraulics, and transient methods. Development of the underlying theory of the 2D/1D method is presented, diagrams are included to highlight important algorithmic flow, and important concepts are discussed as appropriate. This manual is intended to be self-sufficient, but references to published articles and other materials are included for further reading. MPACT is a relatively new code, with new capabilities and many computational methods that did not exist until recently. This manual is an even newer document, with chapters written by several different code contributors working under time constraints. For these reasons, the manual is not yet complete and finalized. Future versions of this manual will address current deficiencies.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Assessment of the Effect of Prototypic High-Burnup Operating Conditions of Fuel Fragmentation, Relocation, and Dispersal Susceptibility

The US nuclear energy industry is investigating strategies that further reduce the cost of energy production by using its existing fleet of nuclear generating stations. Most nuclear power plant operating costs are associated with purchasing fresh fuel assemblies or the efficiency of the reactor core design. Material costs are typically beyond the operator’s control; however, the core design optimizations offer potential operational savings. The core design envelope available to operators is constrained by two primary regulatory criteria: an enrichment limit of 5% 235 U and a burnup limit of 62 GWD/tU. These constraints have resulted in renewed efforts by the nuclear industry to pursue extending the peak rod-average burnup beyond 62 GWd/tU. This effort will likely require additional safety analyses beyond what is currently accepted by the US Nuclear Regulatory Commission. The purpose of this work is to demonstrate a best estimate plus uncertainty pin-by-pin high-burnup loss of coolant accident analysis technique to assess full-core high-burnup fuel fragmentation, relocation, and dispersal (FFRD) and identify approaches for minimizing or potentially mitigating FFRD through core design optimizations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Specifications of EBR-II Neutron Radiography Method Description and Digitization Approach

All neutron radiography (NRAD) images of fuel pins in Argonne’s collection were originally generated using the NRAD imaging facility established in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The NRAD reactor facility was built in 1977 and has been operating since. The reactor is a TRIGA-type reactor operating at a power level of 250 kWth to provide a neutron source for radiography imaging. The reactor is equipped with two beam tubes (i.e., east beam tube and north beam tube) to guide the neutron beams to two radiography stations. The east radiography station is directly under the HFEF main cell and is dedicated for specimens already in the HFEF hot cell. The north radiography station is outside of the main HFEF hot cell and allows NRAD imaging of non-irradiated items. The NRAD images of EBR-II irradiated metallic fuel pins were taken in the east radiography station. Thermal neutrons have the capability to transmit through most materials and are ideal for NRAD imaging. However, because of their high thermal neutron absorption cross-section, fissile materials (e.g., highly-enriched nuclear fuels) may not be as transmissible to thermal neutrons. This is also the case for oversize specimens with extraneous thickness. Epithermal neutron imaging is therefore used as a complement to thermal neutron NRAD imaging. At HFEF’s NRAD facility, both thermal and epithermal neutrons can be used for NRAD imaging. Irradiated nuclear fuels emit high levels of γ radiation that can easily darken X-ray films, so direct exposure NRAD cannot be used to image them. Instead, an indirect NRAD imaging method was developed at HFEF’s NRAD facility. In this method, foils made of materials that can be activated by neutrons (i.e., with large neutron absorption cross section) are used to collect transmitted neutron signals. Then the activated foils are then placed against X-ray films and enclosed in a vacuum cassette so that the γ decay from the activated foils can produce images on the X-ray films. Then, general X-ray film processing procedures are used to digitize and store the images. By using different foil materials, different energy neutrons can be used for NRAD imaging. At the HFEF NRAD station, two types of films are commonly used: dysprosium (Dy) foils with thickness of 130 microns are used to capture thermal neutron signal, while indium (In) foils with thickness of 130 microns are used to capture epithermal neutron signal. A cadmium or gadolinium foil is put before the indium foil to work as a thermal neutron filter. The thermal and epithermal NRAD images can be taken simultaneously by using a Dy/Cd/In sandwiched foil combination. The typical NRAD exposure time is approximately 20 minutes. Then the exposed foils are transferred to film vacuum cassettes. The vacuum ensures that there is no gap between the foil and the film. The foil-to-film exposure time is at least three half-lives of the corresponding radioisotopes, which are 3 hours for In and 7.5 hours for Dy, respectively. Exposed films are processed using an automatic film processor to produce completed NRAD images.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Transition Core Modeling for Extended Enrichment & Accident-Tolerant Fuels Using Polaris/PARCS

Commercial light water reactor (LWR) operators and fuel vendors are currently interested in increasing the low-enriched uranium (LEU) fuel enrichments from the current limit of 5.0 $^w/_o$ $^{235}U$ up to 10 $^ w/_o$ $^{235}U$ (referred to as "LEU+") in their current fleets; they are also interested in using accident-tolerant fuel (ATF) with both LEU and LEU+ fuel. This report aims to identify modeling challenges and accuracy concerns in transition core analysis using the SCALE Polaris lattice physics code and U.S. Nuclear Regulatory Commission core simulator PARCS. At the time this study was started, no publicly available LEU+ core designs existed for boiling water reactor (BWR) or pressurized water reactor (PWR) systems. Therefore, fuel lattices were shuffled within a multi-assembly model to mimic neutronically challenging lattice combinations seen in transition cores, such as a fresh LEU+ lattice next to depleted LEU lattices. In addition to multi-assembly models, whole-core BWR transition core calculations were performed for ATF and LEU+ fuel using an existing Hatch-1 Cycle 3 core model. A whole-core BWR model was chosen due to the more heterogeneous core designs compared to those for a PWR core. Since the original core is an old checkerboard core design and no core or fuel design optimization was performed for the modeled fuel types, these core calculations were intended only to provide: (1) Comparisons of core characteristics of interest, such as the pin power distributions and peaking factors, Doppler temperature coefficients (DTCs), and control blade worths (CBWs) under challenging core designs, (2) Identification of reactor physics challenges in modeling LEU+ and ATF cores, and (3) A stress test for the Polaris/PARCS two-step modeling approach, including characterization of the relative accuracy for predicting characteristics of interest such as pin power distributions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

HTV Post Irradiation Examination (PIE) Results and Analysis

This report is responsive to 2017 Idaho National Laboratory (INL) Memorandum Purchase Order No. 00181805, statement of work 13687 “HTR Graphite R&D – FY ORNL 2017 MPO”, Deliverable 12A “issue final report on the HTV PIE to INL”, due June 15 th 2017.The HTV (High Temperature Vessel) was a HFIR (High Flux Isotope Reactor) high temperature target rod capsule in support of the DOE’s (Department of Energy) High Temperature Reactor (HTR) program. The HTV capsule contained eight sub-capsules, each with nine graphite specimens. The sub-capsules were designed to operate at ~900°C, ~1200°C or ~1600°C. The actual irradiation temperature was established by interrogating SiC and graphite TMs during PIE (post-irradiation examination. The capsule dose ranged from 1.49 dpa to 3.34 dpa, and varied with specimen position within the capsule the maximum dose occurring at the reactor mid-plane and the minimum dose occurring at the core periphery. The irradiated graphite specimens were examined during PIE to establish their high temperature dimensional and elastic (Young’s) modulus changes. The PIE results are reported and analyzed here. The dimensional and modulus changes were shown to have a strong thermal dependence. The dimensional changes were small, being of the order of a few percent or less, ranging from a maximum shrinkage of -3.1% (PCEA WG, T irr =1500°C at 3.3 dpa) to +0.131% (2114 AG, T irr = 869 at 3.2 dpa). The specimen volume shrinkages varied from –0.81 (2114, T irr = 840 at 3.1 dpa) to -9.15% (PCEA, T irr =1500°C at 3.3 dpa). The volume changes were larger at higher Irradiation temperatures. Of the six grades examined, only 2114 moved into the net positive dimensional swelling situation. Significantly, the volume change for 2114 remained negative at the doses and temperature tested here. Dimensional and volume changes were explained in terms of the temperature dependence of the crystal growth rate, the anisotropy of the crystal growth (crystallographic c-axis growth and crystallographic a-direction shrinkage) and the thermal closure of aligned micro-porosity. The modulus changes were much larger, being of the order of a few tens of percent. A trend of lesser fractional increases of Young’s Modulus with an increasing T irr was noted. The minimum increase in Young’s Modulus was 20% (H-451 WG, T irr =1500°C at 3.3 dpa) and the maximum 71% (IG-110, T irr = 840(actual) at 3.3 dpa). The rise in Young’s modulus is attributed to dislocation pinning by lattice defects produced by neutron irradiation and the temperature dependency is related to the irradiation temperature influence on the number of defect available to pin dislocations.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Accuracy Enhancement of Nuclear Power Plant Simulators Utilizing High Accuracy Simulation Predictions

More recently, reactor core simulators for core designs associated with commercial nuclear power plants that utilize what is believed to be higher fidelity models have been developed. Features such as neutronics models that utilize transport equation solvers with fine spatial meshes and many energy-groups, thermal-hydraulic models that utilize sub-channel solvers with fine spatial mesh and capable of treating a wide range of fluid conditions, and fuel-coolant chemistry interaction models capable of treating CRUD deposition are to be found in these higher fidelity core simulators. These reactor core simulators require access to higher performance computers, characterized by many processors, cores and large memory. So associated with utilization of these simulators is access to high performance computers and ability to accommodate in one’s workflow longer execution times. By contrast, currently used core simulators by the nuclear industry can execute on engineering workstations and have execution times of seconds to minutes. The desirability for having short execution times is not only desired for support of time critical tasks but supports the mental process of decision making by engineers. The goal of the work reported upon here has the objective of retaining the fidelity of higher fidelity models while retaining the ability to utilize engineering workstations. Beyond the core simulator goal, additional goals of this work include incorporating the just described core simulator capability into a Nuclear Steam Supply System (NSSS) simulator, and to incorporate the resulting capability into an environment supportive of design and operational decision making associated with nuclear power stations. The model selected for the core neutronics model is the NESTLE code, for the core thermal-hydraulic model is the CTF code utilizing coarse mesh, and for the NSSS model is the RELAP5-3D code. WSC’s proprietary 3KEYMASTERTM platform is being used to provide software coupling, user interface, visualization, and reporting. The NESTLE core neutronics simulator was first integrated with the CTF core thermal-hydraulic simulator using CTF developed communication commands which are also used for CTF to communicate with RELAP5-3D under WSC’s proprietary 3KEYMASTERTM platform. To assure NESTLE prediction consistency with higher fidelity core neutronic simulators, buffer codes have been created to automatically generate from output files written by the VERA core simulator the NESTLE nodal neutronic parameter’ library, geometry, and pin-power reconstruction input files, thereby avoiding a number of challenges associated with utilizing lattice physics codes and providing consistency with VERA predictions. To treat absorber rod effects a multi-set library is utilized, where a set refers to a specific absorber rod fully inserted pattern. A coarse spatial mesh CTF model was developed with features added that support using CTF as envisioned in the engineering quality simulator. A hybrid meshing approach was implemented to allow for automated construction of models with mixed levels of refinement. Specifically, a core model could resolve some assemblies at a nodal level (4 subchannels per assembly) and others at a pin-resolution (one subchannel per coolant subchannel in the assembly). The intention is that this will allow for better resolution of limiting conditions such as DNBR and PCT, which are based on local rod and subchannel conditions. Further development was done of features that enhance the capabilities for the envisioned engineering quality simulator that has been developed, but now for RELAP-3D. The RELAP5-3D code development includes ability to model more than 999 components and the addition of the cross-channels turbulence mixing model and the void drift model that are implemented in CTF, aiming to achieve closer prediction agreement of the two codes for transient simulations, specifically, more accurate matches of the overall mass, momentum, and energy exchanges of both the liquid and gas phases between the neighboring core assemblies. Graphics were also developed for the Instructor Station for this project under WSC’s proprietary 3KEYMASTERTM platform to facilitate design and operational decision making.

42 ENGINEERING↗
DOE OSTI2024

The Monolithic Heat Pipe Microreactor Reference Plant Model

This work introduces a reference plant model for a generic monolithic heat-pipe-cooled microreactor. The model will serve as a springboard to develop future evaluation models in the licensing process of similar microreactor designs at the U.S. Nuclear Regulatory Commission. This model has been developed with the Comprehensive Reactor Analysis Bundle and its specifications are based on open literature publications for the eVinci TM design. BlueCRAB is the U.S. Nu- clear Regulatory Commission non-light-water reactor analysis system based on the Multiphysics Object-Oriented Simulation Environment framework, which can couple the Griffin, BISON, and Sockeye applications to resolve the various physics that are essential for the safety analysis of this type of reactor system. The core specifications includes tristructural isotropic fuel, graphite monolith, graphite reflectors, and drums composed of graphite and B 4 C. No moderator or burnable poison pins are used in the design. The fuel enrichment is reduced to control excess reactivity in the core. This core design is not optimized and only serves for testing purposes, since the primary objective of this work is to exercise the multiphysics coupling for this type of reactor system. A three dimensional (3D) core heterogeneous Griffin discrete ordinates (SN) transport model allows the precise calculation of the flux distribution and pin powers. Griffin transfers the power density distribution and obtains a temperature distribution to and from BISON. The BISON model com- putes the 3D core temperature distribution and is coupled to 876 Sockeye subapplications running a heat pipe model. This 3D conduction model is coupled to the various heat pipes via heat flux boundary conditions. The model includes a small gap between the heat pipe and the monolith. Convective heat transfer boundaries with either ambient temperature or condenser temperature as heat sinks are imposed at the model boundaries. The 2D Sockeye heat pipe model uses a vapor- only methodology, which provides the needed resolution for transient calculations and allows the determination of various heat pipe limits. This approach is superior to the superconductor model traditionally used in steady-state calculations. BlueCRAB computes steady-state power and temperature distributions that serve as the initial condition for a loss-of-heat-sink transient simulation. The steady-state results show significant peaking due to the position of the control drum, but this is a characteristic of the particular design used, which is not optimized at this stage. The transient results show the reactor power slowly stabilizing towards a 3% power level after the partial loss of secondary heat removal. Several recriticalities are observed due to cooling through the secondary system but the reactor is self-stabilizing and behaves as expected.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

The Roadrunner Trap: A QSCOUT Device

The Roadrunner ion trap is a micro-fabricated surface-electrode ion trap based on silicon technology. This trap has one long linear section and a junction to allow for chain storage and reconfiguration. It uses a symmetric rf-rail design with segmented inner and outer control electrodes and independent control in the junction arms. The trap is fabricated on Sandia’s High Optical Access (HOA) platform to provide good optical access for tightly focused laser beams skimming the trap surface. It is packaged on our custom Bowtie-102 ceramic pin or land grid array packages using a 2.54 mm pitch for backside pins or pads. This trap also includes an rf sensing capacitive divider and tungsten wires for heating or temperature monitoring. The Roadrunner builds on the knowledge gained from previous surface traps fabricated at Sandia while improving ion control capabilities.

42 ENGINEERING↗
DOE OSTI2024

High-Burnup BWR LOCA Burst Analysis Framework Development and Demonstration

Nuclear power currently contributes approximately 20% of total electricity generation in the United States and more than 10% globally. Given the increasing reliance on nuclear energy to achieve our nation’s goal of reaching net-zero carbon emissions by 2050, there is significant pressure on the existing nuclear industry to extend plant operational licenses and improve efficiency. This is crucial as the existing nuclear fleet serves as a vital bridge until new light water and advanced reactors can be developed and deployed, bolstering the supply of carbon-free energy to meet domestic demands. Operational costs primarily consist of plant operation and maintenance and fuel costs, influenced by materials and reactor core designs. These factors, coupled with heavily subsidized renewable energy markets, create a challenging economic environment for the existing light water reactor fleet, as well as for new build projects. To address these economic challenges, the nuclear industry has developed a strategic blueprint aimed at enhancing nuclear power’s economic sustainability. Past initiatives, such as efforts to eliminate fuel failures by 2010 and reduce operating costs by 30% before 2020, have laid the groundwork. Optimizing core design parameters, including burnup limits and enrichment levels, can lengthen cycles, reduce outages, reduce batch reload batch fractions and spent fuel storage requirements, and lower maintenance and operating expenses, thereby enhancing economic viability. In the United States, boiling water reactors (BWRs) comprise approximately one-third of the fleet, although much of the research and development focus has traditionally been on pressurized water reactors (PWRs). Advances in modeling and simulation, particularly through the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, are crucial to the long-term viability of BWRs, just as they are for PWRs. A key research area of the high burnup/increased enriched fuel initiative is focused on addressing loss-of-coolant-accident (LOCA)-related issues. NEAMS has dedicated significant effort to enhancing tools to better support BWRs, with a current focus on showcasing the BWR framework for high-burnup LOCA analysis. This high-fidelity work will demonstrate a best estimate pin-by-pin high-burnup BWR LOCA analysis to assess full-core cladding rupture behavior. This modeling capability will help with better understanding and realistic evaluation of fuel fragmentation, relocation, and dispersal (FFRD) phenomena at BWRs, which then could be used to prevent FFRD at BWRs without penalizing operational parameters. In addition, the results of this work will help identify strategies to identify additional margins or to potentially limit cladding rupture through core design optimizations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Development of Mechanistic Fission Gas Release and Swelling Models for UN Fuels in BISON

This report describes the work in NEAMS (Nuclear Energy Advanced Modeling and Simulation) to develop a mechanistic fission gas model for uranium nitride fuels in BISON. The existing Sifgrs (Simple integrated fission gas release and swelling) model tracks the average properties of two bubble populations in the bulk and at the grain boundaries. It was recognized that dislocations play a crucial role in the fission gas swelling of UN, and an irradiation-induced dislocation density model was needed, as well as a model describing how fission gas interacts with dislocations creating a third population of bubbles along dislocations. A mechanistic model that tracks dislocation bubbles was implemented in Sifgrs. This model was used to simulate fission gas swelling and release in UN. Lower-length-scale calculations and experimental observations from carbide fuel were leveraged to help populate the model with essential parameters. As a placeholder, an empirical function was formulated for the evolution of the dislocation network. Because of the difference in evolution of this network at different temperatures, the dislocation bubbles are able to capture behavior that the bulk intragranular bubbles cannot. It was found that the bulk bubbles dominate microscopic swelling at low temperatures, and the dislocation bubbles dominate at higher temperatures. Based on this model, the transition between the two bubble types is the main factor behind the breakaway swelling phenomenon in UN. In order to test the model, a couple of assessments were run. For the lower temperature JOYO pins, the model produces reasonable fission gas release and swelling values. For the higher temperature SP1 pin, the model dramatically underestimates the fission gas release. This issue is attributed to the gas being trapped inside dislocation bubbles, unable to escape to grain boundaries to cause fission gas release, and could be remedied by a mechanistic dislocation model that allows the dislocation density to decrease at very high temperatures. A preliminary mechanistic dislocation model was developed supported by first-principle calculations. These calculations provided valuable insight into how interstitial defects cluster in UN in the {110} orientation, and may eventually form dislocation loops, leading to the conclusion that dislocation loops may nucleate from these clusters. An estimate of the dislocation line energy in UN was obtained and will be improved in future work. The free-energy cluster dynamics code Centipede was used to track interstitial and vacancy absorption at dislocation loops and calculate their growth. In addition, improvements to Centipede were made including new convergence criteria for transient simulations and sink driving force updates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Lightning Induced Interior Fields And Voltage Bounds For Coaxial Topologies

We assemble bounding formulas for the interior fields and pin voltages inside a cylindrical coaxial Faraday cage which has been struck by lightning. Approximate formulas for penetrations through a circumferential door slot with subsequent coupling to the interior center conductor structure. Fields at the opposite open end are estimated and used to drive a capped connector and estimate interior pin voltages. Finally, penetrations through small circular holes and direct diffusion through the barrier are also addressed.

42 ENGINEERING↗
DOE OSTI2025

Technologies for Enhanced Condensation in sCO 2 Power Cycles

The direct fired supercritical carbon dioxide power cycle has higher efficiency, compact machinery footprint, and is resilient under transient operation. It has been shown that improving the cycle main cooler significantly improves performance, including increased cycle efficiency, decreased cost of electricity and decreased cooler outlet temperatures. One way to improve this process is through enhanced condensation techniques and management of condensation fluid. This manuscript provides an overview of active and passive enhanced condensation technologies, with a focus on the passive approaches. compared to passive approaches, active approaches can have higher heat transfer enhancement, but the additional power and infrastructure can be problematic. The passive approaches identified include microfins (helical and herringbone), corrugated tubing, dimpled tubing and pin fins. After comparison it is found that the microfin geometry has the best performance characteristics followed by the additively manufactured pin fins. It is noted that current literature around additively manufactured enhancements is lacking, and investigation of conventional as well as novel geometries fabricated in this manner may be worth further investigation. Benchmark correlations for these geometries are identified and their limitations outlined. For further investigation, a matrix of test articles is presented for possible experimentation.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2025

Improvement and Verification of Online Cross Section Generation Capability of Griffin for TRISO-fueled Reactors

Griffin, a MOOSE-based reactor multiphysics code jointly developed by Idaho National Laboratory and Argonne National Laboratory under the DOE Office of Nuclear Energy’s NEAMS program, has pursued the development of an online multigroup cross section generation capability for a few years to enable high-fidelity, problem-dependent neutronics analyses of advanced thermal reactors. Recent advancements in Griffin’s online multigroup cross section generation capability have significantly improved the accuracy, robustness, and efficiency of self-shielding calculations for both prismatic and pebble-bed TRISO-fueled reactor applications. Key developments include a unified fuel self-shielding method applicable to both TRISO and annular compact/spherical shell fuel zone geometries; an advanced Dancoff Category-based Equivalence Theory using a bell function for non-fuel resonance treatment, achieving more than an order-of-magnitude speedup compared to the Tone method; an on-the-fly multigroup equivalence approach to mitigate group condensation errors; and a streaming correction method for pebble-bed homogenization. A proof-of-concept demonstration of on-the-fly group condensation with consistent P0 transport correction was also achieved. The method reproduced direct fine-group solutions with excellent accuracy (eigenvalue errors within 10 pcm and pin-power differences within 0.5%), but due to performance limitations of the current fixed-source solver, improvements to solver efficiency will be addressed in future work. Verification tests were performed on graphite-moderated TRISO-fueled two-dimensional core benchmark problems representing gas-cooled microreactors, heat pipe-cooled microreactors, gas-cooled pebble-bed reactors, and fluoride salt-cooled high-temperature reactors. Across all cases, Griffin showed excellent agreement with Serpent2 continuous energy Monte Carlo solutions: eigenvalue errors within 200 pcm, pin-power root-mean-square errors within 2%, and control rod and drum worth errors less than 2%. It should be noted that, for the benchmark problem, cross section generation contributed less than 3% of the total simulation times. These results demonstrate that Griffin’s online cross section generation capability delivers accurate and efficient reactor physics solutions across a wide spectrum of TRISO-fueled advanced reactor designs. With further improvements to the fine-group fixed-source solver and planned extensions to depletion, transients, and coupled neutron–gamma transport, Griffin will be well-positioned to become a powerful and comprehensive tool for advanced reactor analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Integrating BISON into VERA-CS and Fuel Temperature Calculations

Approaches to integrating BISON into VERA-CS were identified as: 1) Coupled TIAMAT calculations in which BISON is solved simultaneously with MPACT and CTF via direct coupling, 2) Inline TIAMAT in which BISON calculations are performed as part of the same VERA-CS calculation sequence as coupled MPACT-CTF calculations but there is no feedback from the BISON calculations into the MPACT-CTF calculations, and 3) Stand-alone (“Offline”) BISON calculations in which desired data is parsed from the results of an already-completed MPACT-CTF VERA-CS analysis and BISON calculations are performed afterwards. The current maximum capabilities of these approaches in the latest VERA build are a single ramp-to-power for Coupled TIAMAT, analysis of two subsequent operating cycles for a single pin in Inline TIAMAT, and BISON analysis for every fuel pin in a PWR quarter-core for multiple operating cycles in Stand-alone BISON.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Development of Transient MPACT-CTF Capability for RIA with DNB

The objective of this milestone was to continue the development, implementation, verification, and validation of the methods in VERA-CS necessary to solve the whole-core time-dependent neutron transport equations with pin-resolved detail for a practical PWR applications. This capability is necessary to provide the time dependent pin resolved neutron flux for analysis of the Reactivity Insertion Accident (RIA) as specified in the CASL RIA Challenge Problem Implementation Plan, CASL-I-2013-0060-000. The FY17 work reported here continued the FY16 work reported in CASL-U-2016-1188-000 which was primarily focused on MPACT and performed the methods, and Validation and Verification (V&V) efforts necessary to demonstrate the transient capability in VERA-CS in order to facilitate the L1 RIA milestone in FY18. This report describes the activities performed in FY17 in the three general areas of: User Support, Methods Improvements, and in V&V and Demonstration. In the area of User Support various user requested features were added to the code such as detailed reactivity edits and improvements to the VERA input. In the area of Methods Improvements, the features added included user controlled variable time stepping, a one-group CMFD Acceleration, and a Delayed Energy Deposition model; as well as improvements to the overall code robustness for the transient solution. In the area of V&V, the full suite of C5G7 Transient benchmarks was completed and significant progress was made on the solution of SPERT with VERA-CS. In the area of practical demonstration, a full core AP1000 demonstration problem was performed with VERA-CS. The work performed in FY17 enhanced the capability of performing VERACS transients for practical large-scale light water reactor problems with an acceptable computational burden and provided an important step in the completion of the CASL Reactivity Insertion Accident (RIA) Challenge Problem.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Restart File Improvements for MPACT

This report documents the work performed to address several defects and features request related to the restart capability in MPACT. The following defects were fixed as a part of this milestone: • #3616 (Defect) Error in MPACT Restart Files with BPRs, • #3637 (Defect) Restart with Semi-PC Yields Inconsistent Results, • #3652 (Defect) Fix Incorrect Exposure Value on HDF File After Restart, • #3777 (Defect) Fix State Point Number when Using Restart File, • #3986 (Defect) Pin Exposures are Incorrect, • #4106 (Defect) Pin Exposures Equal to Zero After Restart, and • #4288 (Defect) Core Average Exposure Wrong when Reading Restart Files. The new features that were added are: • Axial Re-meshing capability, • Incorporation of MAMBA-1D data necessary to restart coupled MPACT/CTF/MAMBA calculations, and • Addition of state variables to the restart file. Numerous tests were also implemented to verify and protect these features and this report was prepared to document the work performed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

MPACT Development Activities at University of Michigan

This report documents the efforts of general development tasks in MPACT performed primarily at the University of Michigan during FY17. These tasks included: 1. Unstructured CMFD - The unstructured CMFD supports the RIA and BWR challenge problems. It allows MPACT to simulate mixed lattice cores (e.g. 8x8 next to 9x9). This design feature is common to operating BWRs and also the SPERT experiments being used to validate VERA-CS for RIA. 2. Higher order spatial sources for MOC - The higher order sources are for improving the overall performance of MPACT. Higher order spatial sources in MOC provide accurate solutions with potentially significantly fewer mesh regions, thus speeding up every calculation. This method is also particularly important for BWRs. 3. Improving fine source region visualization - The improvement of the visualization of the flat source region mesh in MPACT is to support users and communication of models and results from MPACT. VERAView provides visualization of pin-wise information, however MPACT resolves the spatial solution to intra pin level. The capability provided here will let users visualize the full model geometry and components of the solution at the highest spatial resolution available from MPACT. 4. Automated spatial decomposition - The automated spatial decomposition removes the burden from the user to know how to spatially decompose the model for MPACT; a task that has been rather tricky in the past. Instead the user may give some number of processors that is a multiple of the axial mesh and MPACT will determine a nearly optimal decomposition that gives better load balance than most decompositions used currently. 5. Optimal relaxation factors for feedback - The development of optimal relaxation factor is to improve the stability and robustness of problems with feedback. Presently the default feedback values are 0.5, which may or may not be optimal. Without relaxation the fixed point iteration that is currently used is known to be unstable for large problems. The optimal relaxation factors are derived by utilizing a previous Fourier analysis. From the original milestone there was a scope change where the work on the Intel MIC was deferred for work on the automated spatial decomposition.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗