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NASA NTRS1999

Turbomachinery Design Quality Checks to Avoid Friction Induced Structural Failure

A unique configuration of the P&W SSME Alternate Fuel Turbopump turbine disk/blade assembly, combined with a severe thermal environment, resulted in several structural anomalies that were driven by frictional contact forces. Understanding the mechanics of these problems provides new quality checks for future turbo machinery designs. During development testing in 1997 of the SSME alternate fuel turbopump at Stennis Space Center, several potentially serious problems surfaced with the turbine disk/blade assembly that had not been experienced in extensive earlier testing. Changes to the operational thermal environment were noted based on analytical prediction of modifications that affected performance and on stationary thermal measurements adjacent to the rotor assembly. A detailed structural investigation was required to reveal the mechanism of distress induced by the change. The turbine disk experienced cracking in several locations due to increased thermal gradient induced stress during start and shutdown transients. This was easily predictable using standard analysis procedures and expected once the thermal environment was characterized. What was not expected was the curling of a piston ring used for blade axial retention in the disk, indentation of the axial face of the blade attachment by a spacer separating the first and second stage blades, and most significantly, galling and cracking of the blade root attachment that could have resulted in blade release. Past experience, in gas turbine environments, set a precedent of never relying on friction for help and to evaluate it only in specific instances where it was obvious that it would degrade capability. In each of the three cases above, friction proved to be a determining factor that pushed the components into an unsatisfactory mode of operation. The higher than expected temperatures and rapid thermal transients combined with friction to move beyond past experience. The turbine disk/blade assembly configuration contributed to the potential for these problems to occur by limiting the radial deflection from thermals and centrifugal loading. The cooled solid bore configuration was chosen to improve rotordynamic stability by limiting the length of rotor overhang while still protecting the roller bearing by maintaining zero slope under the inner race. During a start transient, the rim area of the disk heats rapidly and expands axially and circumferentially and requires corresponding radial and axial growth of the disk to maintain relative positioning of the disk, blades, spacers and retainer rings. The stiffness, large thermal mass, and bore cooling flow combine to severely limit the disk rim radial growth which results in the potential for relative movement between these parts. Friction then becomes a player in the determination of component stress.

Moore, Jerry H.↗

DOE OSTI2025

Impact of recent ENDF nuclear data update, high initial enrichment and high burnup fuel on critical experiments applicability determination via the integral index c k for burnup credit validation

In 2012, NUREG/CR-7109 reported on the validation of burnup credit calculations involving major and minor actinides and major fission products which was investigated for pressurized and boiling water reactor (PWR and BWR) fuel enrichments up to 5 wt% 235 U and assembly-average burnups up to 60 GWd/MTU. Recently, there has been interest in increasing the maximum enrichment used in PWR fuel as high as 8 wt% 235 U and correspondingly increasing the maximum assembly-average burnups to approximately 75 GWd/MTU. These proposed increases in enrichment and burnup necessitate reinvestigation of the validation basis for k eff calculations for this expanded application space. Additionally, the 2012 study was performed by using the Evaluated Nuclear Data File (ENDF)/B-VII.0 nuclear data with the SCALE 6 covariance library, and the effects of using the newly released ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data and covariance libraries should be evaluated. In this work, published in NUREG/CR-7309 in 2025, the validation assessment was performed consistently with NUREG/CR-7109: modeling irradiated fuel assemblies in the Generic Burnup Credit (GBC)-32 cask defined in NUREG/CR-6747. The TSUNAMI-3D sequence was used to generate sensitivity data for the application model, and the data were compared with sensitivity data from select benchmark models. The integral parameter c k is the metric of similarity used in this study and is consistent with NUREG/CR-7109, where a c k value in excess of 0.8 indicates sufficient similarity for use in validation. A new set of benchmark experiments with sensitivity data has been assembled for this effort. The number of experiments with available sensitivity data is now 2,104, compared to 474 in NUREG/CR-7109. This increase was facilitated by the efforts of the Nuclear Energy Agency to generate sensitivity data for a majority of the experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook to supplement the data available in the Oak Ridge National Laboratory (ORNL) Verified, Archived Library of Inputs and Data (VALID). The complete set of benchmarks considered here includes experiments for low-enriched uranium (LEU), intermediate enriched uranium (IEU), and a mixture of uranium and plutonium (MIX) from the ICSBEP Handbook and VALID, as well as ORNL models of the Haut Taux de Combustion (HTC) experiments and other potentially relevant models not included in VALID. The updated similarity study shows that none of the extended burnup and higher enrichment combinations considered show a significant decrease in the number of potentially applicable experiments, meaning sufficient critical experiments exist for the validation of BUC criticality safety calculations, with initial enrichments up to 8 wt% 235 U and burnups up to 80 GWd/MTU. Additionally, both the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries can be used for validation since the number of critical experiments applicable for validation increases for most cases with the most recent nuclear data compared to the previous one. As in previous BUC validation studies, the French HTC experiments are the most similar in a majority of the application cases studied, especially from representative discharge burnups ranging from 40 to 80 GWd/MTU. In conclusion, these results match the conclusions presented in NUREG/CR-7109 regarding validation of the primary actinides in BUC analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

An Experimental Investigation of Multilayer Flow in a Slide Die Coating Process

Roll-to-roll coating technology has been actively applied not only for conventional film production but also for emerging energy devices such as organic solar cells, organic light emitting diodes, batteries, smart windows, and fuel cells. The devices are multilayer structures requiring several coating and drying steps when implemented in a serial fashion, which increases capital costs and energy consumption. Simultaneous multilayer (ML) coating technologies present an opportunity for cost reductions and improved process efficiency by reducing equipment footprint and the number of process steps. In this talk, we experimentally investigate ML slide die coating flow for the manufacturing of fuel cell membrane electrode assemblies (MEAs). For ML slide coating, understanding fluid properties such as density, surface tension, and viscosity are necessary for stable flows and limiting intermixing of the miscible layers. Here, we utilize a design-of-experiment methodology - Box-Behnken design (BBD) - to correlate ink formulation parameters to fluid properties. Using the ink properties obtained from the BBD, we performed numerical flow simulations to obtain a proper process conditions to guide experimental coating trials. Finally, experimental verification is performed on a ML slide die and a degree of miscibility was observed for various ink formulation conditions.

28 EE - Advanced Manufacturing Office (EE-5A)↗

DOE OSTI2022

Transition core modeling for extended enrichment, accident tolerant fuel using PARCS/Polaris

Current plans and efforts from reactor operators and vendors to include extended enrichment (EE) and accident-tolerant fuel (ATF) in current reactor fleets motivate the study of these changes in reactor physics analysis. This work uses the US Nuclear Regulatory Commission core simulator PARCS to do the core calculation and SCALE Polaris lattice physics code to generate the required homogenized, few-group constants. The lattice model used is based on the GE-14 10x10 assemblies with UO{sub 2} fuels. Both nominal core and transition core are studied in this work, and both cores use GE-14 10x10 assemblies with UO{sub 2} fuel. The nominal core uses regular UO{sub 2} fuel with a maximum enrichment of 5 wt % and ZIRC-2 cladding. The ATF transition core uses FeCrAl cladding and regular UO{sub 2} fuel, while the EE-ATF transition core uses FeCrAl cladding with UO{sub 2} fuel with 8 max wt % enrichment. The accuracy presented in the colorset models verified the capability of the PARCS/Polaris procedures for the transition core analysis. For the whole core calculation, the ATF and EE-ATF transition core models were made in addition to the nominal core model. The core parameters to study are the core power distribution and power peaking factor, doppler temperature coefficients, and control rod worth at cold zero power and hot full power. Comparing the core parameters of the transition cores with the nominal core in PARCS, the results suggest that there is no unexpected outcome for the implementation of ATF and EE fuels. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1995

Nuclear Safety [Vol. 36, No. 2, July-December 1995]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. THE CHORNOBYL ACCIDENT: 195 The Chornobyl Accident Revisited, Part III: Chernobyl Source Term Release Dynamics and Reconstruction of Events During the Active Phase, A. R. Sich; GENERAL SAFETY CONSIDERATIONS: 218 Second ANS Workshop on the Safety of Soviet-Designed Nuclear Power Plants, R. A. Bari; 234 Elements of a Nuclear Criticality Safety Program, C. M. Hopper; 243 Rickover, Excellence, and Criticality Safety Programs, R. E. Wilson; ACCIDENT ANALYSIS: 249 Transient Analysis of the PIUS Advanced Reactor Design with the TRAC-PF1/MOD2 Code, B. E. Boyack, J. L Steiner, S. C. Harmony, H. J. Stumpf, and J. F. Lime; 278 The Hierarchy-By-Interval Approach to Identifying Important Models that Need Improvement in Severe-Accident Simulation Codes, T. J. Heames, M. Khatib-Rahbar, J. E. Kelly, R. P. Jenks-Johnson, and Y.-S. Chen; 290 RELAP5/MOD3 Code Coupling Model, R. P. Martin; 299 Missiles Caused by Severe Pressurized-Water Reactor Accidents, R. Krieg; DESIGN FEATURES: 310 Validation of COMMIX with Westinghouse AP-600 PCCS Test Data, J. G. Sun, T. H. Chien, J. Ding, and W T. Sha; ENVIRONMENTAL EFFECTS: 321 Spent Nuclear Fuel Characterization for a Bounding Reference Assembly for the Receiving Basin for Off-Site Fuel, S. D. Kahook, R. L. Garrett, L. R. Canas, and M J. Beckum; OPERATING EXPERIENCES: 332 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 335 Reactor Coolant System Blowdown at Wolf Creek on September 17, 1994, J. V. Kauffman and S. L. Israel; RECENT DEVELOPMENTS: 344 Reports, Standards, and Safety Guides, D. S. Oueener; 349 Proposed Rule Changes as of June 30, 1995; ANNOUNCEMENTS: 320 Symposium on Acceptability of Risk From Radiation—Application to Manned Space Flight; 320 24th DOE/NRC Nuclear Air Cleaning and Treatment Conference; 361 Radiation Biology and Radiation Protection— Modern Developments and Tendencies in Radiation Biology; 362 1997 IEEE Sixth Conference on Human Factors and Power Plants; 354 The Authors; 360 Reviewers of Nuclear Safety, Vol. 36.

05 NUCLEAR FUELS↗

DOE OSTI1995

Nuclear Safety [Vol. 36, No. 2, July-December 1995]

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Monolithic ATF MiniFuel Sample Capsules Ready for HFIR Insertion

Several advanced nuclear fuel concepts are being considered throughout the nuclear industry. Implementation of these concepts could increase accident tolerance or enhance performance beyond the capabilities of the UO₂-based nuclear fuel currently in use. Qualification and deployment of any new fuel requires rigorous irradiation testing to demonstrate performance under representative normal and off-normal operating conditions. The traditional approach for qualifying new fuels requires exhaustive execution of many integral fuel tests. However, due to the long timeframe for executing these integral tests and the limited number of available materials test reactors, this approach is becoming impractical. To accelerate fuel qualification, Oak Ridge National Laboratory developed the MiniFuel irradiation vehicle for use in conducting accelerated separate effects irradiation testing of a wide range of nuclear fuel materials in the High Flux Isotope Reactor (HFIR). The first MiniFuel irradiations performed in the facility tested sol gel–derived uranium nitride kernels and tristructural isotropic (TRISO)-coated particle fuels. This report describes the preparation and assembly of the first set of monolithic MiniFuel irradiations conducted to support accident-tolerant fuel (ATF) development and accelerated fuel qualification. Two irradiation targets containing a variety of UO₂ and U₃Si₂ disk fuel specimens were fabricated and assembled for irradiation to burnups of 8-10 and 28-40 MWd/kg U. The target irradiation temperature is 450-550°C. Irradiation of U₃Si₂ will provide new data regarding the irradiation performance of a candidate ATF to complement current ATF-1 integral experiments being performed in the Advanced Test Reactor. UO₂ samples were included as a reference so that the results from the MiniFuel experiments can be compared with the extensive UO₂ fuel performance database. The monolithic MiniFuel capsules were successfully assembled, welded, and tested per HFIR requirements and are ready for insertion into the reactor. Pictures of the assembly process are included in this report. The experiment is planned for insertion into the HFIR during cycle 487 in April 2020.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

IMPACT-01 Assembly Overview [Slides]

The Irradiated Material Properties Accelerated Characterization Test (IMPACT) series of experiments will irradiate three metallic fuel alloy specimens with embedded thermal conductivity probes in the Advanced Test Reactor (ATR). Metallic fuel alloys have long been under investigation for use in advanced reactors on account of their high thermal conductivity. Metallic fuel undergoes dramatic microstructural changes early in life due to fission gas swelling until ~2-3 at% burnup when pores interconnect, thus allowing fission gas to escape into the fuel pin plenum and swelling effectively ceases. The evolution of metallic fuel thermal conductivity during this early phase has never been successfully measured in situ. This experiment will be designed to use advances in measurement sciences to characterize how thermal transport properties evolve while in reactor. The IMPACT experiment consists of three metallic fuel rodlets with thermal conductivity probes axially centered within the fuel specimen. This presentation is an overview of the IMPACT-01 assembly process with a focus MFC fuel fabrication and experiment assembly steps.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2004

Reduced size fuel cell for portable applications

A flat pack type fuel cell includes a plurality of membrane electrode assemblies. Each membrane electrode assembly is formed of an anode, an electrolyte, and an cathode with appropriate catalysts thereon. The anode is directly into contact with fuel via a wicking element. The fuel reservoir may extend along the same axis as the membrane electrode assemblies, so that fuel can be applied to each of the anodes. Each of the fuel cell elements is interconnected together to provide the voltage outputs in series.

Narayanan, Sekharipuram R.↗

DOE OSTI2021

Assembly of MiniFuel Targets for Irradiation of U-Mo Fuel Specimens in the High Flux Isotope Reactor

To support the development of advanced low-enriched uranium for use in nuclear reactors, irradiation testing of U-Mo disk specimens was performed at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) to collect experimental irradiation data on this type of fuel at pressurized water reactor–relevant temperatures. U-Mo is a uranium alloy that has superior dimensional stability relative to alpha-phase uranium metal and has a substantially higher uranium density compared to UO 2 . U-Mo disks specimens were fabricated at Idaho National Laboratory and inserted into MiniFuel targets for HFIR irradiation. Three MiniFuel targets were successfully assembled, welded, tested, and delivered to HFIR, along with their quality assurance documentation. The targets were inserted into HFIR’s inner vertical experiment facility within the permanent beryllium reflector. Each target contains six disk specimens and will be irradiated in HFIR for three, four, and eight cycles, with target temperatures between 250 and 500°C. This report summarizes the experiment design, test matrix, pre-characterization of specimens, and experiment assembly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Fine Gradient Electrode and Micro Porous Layer Structures for Improved Heavy Duty Fuel Cells (Final Report)

The commercial deployment of Heavy Duty Fuel Cells(HDFC) for applications such as large trucks (for example, Class 8 capable of carrying 50,000 lb. loads) depends vitally on achieving high efficiency and durability at reasonable costs. Furthermore, the HDFC must operate under practical conditions such as with fuel and air impurities, multiple stop-start cycles, and under the extremes of climate our planet offers – from hot and cold to dry and wet. These conditions place a premium on the stability and utilization of the materials comprising the membrane electrode assemblies (MEAs) powering the fuel cell. and the current catalysts, MicroPorous Layers, and Electrode structures and additives are insufficient for these needs. In Phase I Pajarito and Advent will develop durable electrocatalysts, MicroPorous Layer and electrode additives, and electrode structures for heavy duty fuel cells designed for zero-emission long-haul trucking. The electrocatalyst products are designed to solve the challenging durability and performance needs of fuel cells designed for long life and high efficiency through a combination of uniquely structured designed catalysts as well as new MicroPorous Layers (MPLs) and electrode structure additives. These improved materials will provide Pajarito both an expanded commercial opportunity in electrocatalysts, as well as new markets for MPL and electrode additives. The resulting public benefits include improved economics of fuel cells, a leading zero-emission technology for mobility, as well as reduced reliance on the critical minerals and metals used in the heavy-duty trucks industry. Possible follow-up Phase II and III projects would add full MEA products based on the Phase I efforts catalysts and additives, with full system validation by leading Fuel Cell Truck developers.

08 HYDROGEN↗

DOE OSTI2022

System Analysis Modeling and Intermodal Transportation for Commercial Spent Nuclear Fuel

This is a technical presentation that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Nitrogen-Doped PtNi Catalysts on Polybenzimidazole-Functionalized Carbon Support for the Oxygen Reduction Reaction in Polymer Electrolyte Membrane Fuel Cells

PtM (M = 3d transition metals) alloys are known as the promising oxygen reduction reaction catalysts and have been considered as the replacement of pure Pt catalysts for the commercialization of proton exchange membrane fuel cells. Although great progress has been made in the past three decades, the performance and durability of PtM catalysts still face stringent challenges from practical applications. Functionalization of a catalyst carbon support with nitrogen-contained groups can add charges onto its surface, which can be utilized to build a more complete ionomer/catalyst interface, to reduce the catalyst particle size, and to improve particle size distribution. Nitriding of PtNi catalysts can effectively improve the catalyst activity and stability by the modification of lattice strain. Hereby, we propose a synergistic approach of combining polybenzimidazole-grafted Vulcan XC72 carbon as the catalyst carbon support and the nitriding of PtNi to develop PtNiN/XC72-polybenzimidazole catalysts. Such PtNiN/XC72-PBI catalysts exhibit the excellent performance of fuel cell membrane electrode assembly (i.e., mass activity, 440 mA mgPt –1 ; electrochemical surface area, 51 m 2 gPt –1 ; and rated power density, 836 mW cm –2 ) as well as promising catalyst stability. The developed PtNiN/XC72-PBI meets the US DOE 2020 targets of mass activity for the fuel cell catalysts. This work provides a novel approach and a promising pathway on the development of the catalyst using such a synergistic approach-modification of the catalyst structure by nitrogen doping and functionalization of carbon support by polybenzimidazole for both high performance and high durability.

25 ENERGY STORAGE↗

DOE OSTI2022

Next Generation System Analysis Model: Recently Added Features and Future Plans

To better enable informed decision-making regarding the back-end of the nuclear fuel cycle, the Integrated Waste Management System (IWMS) program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the development and application of system analysis tools capable of analyzing various system options for the management of spent nuclear fuel (SNF) and high-level radioactive waste (HLW).With these tools, IWMS architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool expressly designed to be capable of modeling features within various IWMS architectures. NGSAM imports data from Oak Ridge National Laboratory (ORNL)’s unified database (e.g., historic assembly information, thermal profiles for assembly heat, and at-reactor dry storage loadings) to ensure that each simulation initializes with a realistic representation of the state of commercial SNF in the U.S. Recent major enhancements implemented into NGSAM in the period since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort car acquisition and buffer car acquisition • The addition of heavy haul truck (HHT) and barge routes for some sites, as well as support for user-defined inter-modal routes • Updates to the logic that checks the transportation cask thermal limit maps prior to package transport • An allocation method that predicts when reactor sites would pack assemblies from their spent fuel pools for dry storage, and prioritizes shipments directly from the pools of those reactor sites in the preceding periods (before the predicted loadings to dry storage), thus reducing the number of casks loaded into dry storage at reactor sites • The addition of “reactor site family” operational limits to restrict the number of loads of SNF taken from the pool and from dry storage at a given reactor site each year • Added support for multiple canister loading map options and packages with multiple compatible transportation overpacks • Updates to the handling of non-commercial SNF, including a new database that contains data to support the updates • Updates to allow analysis of hypothetical scenarios which include repackaging at reactor sites, e.g., for possible comparative analysis with other scenarios • Implementation of additional output reports, or modification of existing ones • The ability to generate and implement user edits via the NGSAM website • The ability to model loading SNF from pool storage at an interim storage facility (ISF) into dry storage at the ISF • The ability to model consolidating SNF from different existing storage containers at a DOE site into the same DOE standard canister • The ability to model transferring SNF casks from one transportation mode to another, e.g., from HHT to rail, referred to as transloading. These new features have improved NGSAM capabilities and users’ experience with the model. Preliminary NGSAM requirements for modeling advanced reactor fuels, reprocessing, treatment, and conditioning were considered, and this paper describes them at a high level. Other nuclear fuel cycle system analysis tools developed under sponsorship of DOE-NE, like the VISION code developed at Idaho National Laboratory, might be better suited for initial high-level analysis of those technologies and advanced fuel cycles. As technologies are developed and system concepts evolve, NGSAM could provide value by providing more detailed modeling of transport, storage, and disposal of spent fuel and wastes from advanced reactors and advanced fuel cycles at the fuel element and waste container level.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Verification and Validation of START: A Spent Nuclear Fuel Routing and Decision Support Tool

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1973

Fabrication of capsule assemblies, phase 3

Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.

Keeton, A. R.↗

DOE OSTI2022

Design of the high-yield time-gated x-ray hot-spot imager for OMEGA

Time-resolved x-ray self-emission imaging of hot spots in inertial confinement fusion experiments along several lines of sight provides critical information on the pressure and the transient morphology of the hot spot on the University of Rochester’s OMEGA Laser System. At least three quasi-orthogonal lines of sight are required to infer the tomographic information of the hot spots of deuterium–tritium cryogenic layered implosions. OMEGA currently has two time-gated x-ray hot-spot imagers: the time-resolved Kirkpatrick–Baez x-ray microscope and the single-line-of-sight, time-resolved x-ray imager (SLOS-TRXI). The time-gated x-ray hot-spot imager (XRHSI) is being developed for use on OMEGA as the third line of sight for the high-yield operation of up to 4 × 10 14 neutrons. XRHSI follows the SLOS-TRXI concept; however, it will have improved spatial and temporal resolutions of 5 μm and 20 ps, respectively. The simultaneous operation of the three instruments will provide 3-D reconstructions of the assembled hot-spot fuel at various times through peak thermonuclear output. The technical approach consists of a pinhole array imager and demagnifying time-dilation drift tube that are coupled to two side-by-side hybrid complementary metal–oxide semiconductor (hCMOS) image sensors. To minimize the background and to harden the diagnostics, an angled drift-tube assembly shifting the hCMOS sensors out of the direct line of sight and neutron shielding will be applied. As a result, the technical design space for the instrument will be discussed and the conceptual design will be presented.

47 OTHER INSTRUMENTATION↗

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↗