Results for “hexagonal geometry” — Page 2

DOE OSTI2013

Study of cavity type antenna structure of large-area 915 MHz ultra-high frequency wave plasma device based on three-dimensional finite difference time-domain analysis

A large-area planar plasma source with a resonant cavity type launcher driven by a 915 MHz ultra-high frequency wave was developed. Theoretical analysis with the three-dimensional finite difference time-domain simulation was carried out to determine the optimized launcher structure by analyzing the resonant transverse magnetic mode in the resonant cavity. Numerical result expects that the resonant electric field distribution inside the cavity dominantly consists of the TM{sub 410} mode. The resonant cavity type launcher having 8 holes in an octagonal geometry was designed to fit the resonant transverse magnetic mode. Adjusting 8 hole positions of the launcher to the field pattern of the resonant TM{sub 410} mode, we found that the plasma density increased about 40%∼50% from 1.0∼1.1 × 10{sup 11} cm{sup −3} to ∼1.5 × 10{sup 11} cm{sup −3} at the same incident power of 2.5 kW, compared with the previous results with the launcher having 6 holes in the hexagonal geometry. It is also noted that the electron density changes almost linearly with the incident wave power without any mode jumps.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2014

Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Effect of lattice orientation on compressive properties of selective laser sintered nylon lattice coupons

As AM lattices become more popular in medical devices, it is important to consider how lattice design parameters affect the mechanical integrity and performance of a device. This research investigated the effect of lattice orientation on the compressive mechanical response of five common lattice geometries: Hexagonal Honeycomb (Hex), Diamond, Voronoi Tessellation Method (VTM), sheet-based Gyroid, and Face-Centered-Cubic (FCC). Samples were tested at two relative densities and eight orientations, printed in Nylon 12 (PA2200) on an EOS P396. The mechanical response was compared to simulated finite element analysis (FEA) response for Hex and FCC lattices. Hex lattices displayed significant orientation dependance. Diamond lattices demonstrated some orientation dependance when at a lower relative density. Gyroid, FCC, and VTM all displayed minimal orientation dependance. Increasing lattice relative density appeared to reduce anisotropy among tested orientations of all lattice geometries except Hex. FEA simulations were able to produce a trend response that matched the mechanical response trends based on the investigated orientations. Furthermore, these results support the need for multi axial lattice mechanical test considerations. It also bolsters the need for lattice FEA models to have a sufficient comparator when identifying “worst-case” lattice test scenarios.

36 MATERIALS SCIENCE↗

DOE OSTI2013

VARI3D & PERSENT: Perturbation and Sensitivity Analysis

The nodal diffusion method is one of the most widely used approaches in modern reactor analysis. In the nodal diffusion method, a coarse multi-group set of “homogenized” parameters is constructed such that the complex geometry of a reactor core along with the energy dependence of neutron and gamma ray cross sections in a nuclear reactor are conserved in the simpler geometry. The homogenization is typically done on a fuel assembly level as is the case in the DIF3D code developed at Argonne National Laboratory. The nodal methodology is used primarily to predict fuel cycle behavior of nuclear systems of which there is a substantial amount of validation in the literature. Another use of the nodal method is to obtain reactivity coefficients and kinetics parameters for use in a safety analysis of a given nuclear reactor. While there are many ways to obtain reactivity worth and kinetics parameters, the work presented in this manuscript is unique as it provides the user with the ability to compute reactivity worths, kinetics parameters, and cross section sensitivities with a Cartesian and hexagonal geometry based transport code. This manuscript serves as a single manual for two separate codes: VARI3D and PERSENT. The VARI3D code (VARiational 3D) is based upon the classic finite difference diffusion theory solver available in DIF3D. The PERSENT code (PERturbation and SENitivity for Transport) is based upon the variational nodal method employed in DIF3D termed VARIANT. The VARIANT solver was added to DIF3D in 1995 and has seen continued development and use for the last 18 years. Because VARI3D primarily uses deprecated coding practices, rather than incorporate the perturbation and sensitivity treatments for transport within VARI3D, a new coding development was built using modern Fortran coding. The primary purpose of this manual is to describe the theory behind PERSENT (and by convenience, that of VARI3D) and discuss the input and output of PERSENT along with giving potential users an idea of how to use it. While this manuscript does describe the input and output of VARI3D, the PERSENT code is intended to be the replacement capability of VARI3D as PERSENT can generate nearly identical (if not superior) diffusion theory results. In this manuscript, the relevant aspects of generalized perturbation theory that are relevant to both VARI3D and PERSENT are covered. The input and output of VARI3D is displayed by excerpting several of the example problems. Similarly, the input and output of PERSENT is displayed along with tips on how best to use the code. Note that the input and output of the inhomogeneous solver wrapped around DIF3D (DIF3D_IFS) is also discussed as it is needed to carry out some of the sensitivities in PERSENT such as reaction rate ratios. This manuscript describes several perturbation theory problems and sensitivity problems and the results computed using PERSENT. From these sections, potential users should find that PERSENT provides not only the typical tables of numbers desired in perturbation and sensitivity analysis work, but also can visually plot the result for a more thorough understanding of the space and energy distribution (Section 5). Overall, PERSENT is observed to produce accurate reactivity worths and sensitivities for the displayed set of test problems and clearly demonstrates the need to have a transport based sensitivity capability as evident from the thousands of percent errors observed in the 21 group hexagonal fast reactor problem (covered in Section 7).

Smith, M. A.↗

DOE OSTI2025

VARI3D & PERSENT: Perturbation and Sensitivity Analysis

The nodal diffusion method is one of the most widely used approaches in modern reactor analysis. In the nodal diffusion method, a coarse multi-group set of “homogenized” parameters is constructed such that the complex geometry of a reactor core along with the energy dependence of neutron and gamma ray cross sections in a nuclear reactor are conserved in the simpler geometry. The homogenization is typically done on a fuel assembly level as is the case in the DIF3D code developed at Argonne National Laboratory. The nodal methodology is used primarily to predict fuel cycle behavior of nuclear systems of which there is a substantial amount of validation in the literature. Another use of the nodal method is to obtain reactivity coefficients and kinetics parameters for use in a safety analysis of a given nuclear reactor. While there are many ways to obtain reactivity worth and kinetics parameters, the work presented in this manuscript is unique as it provides the user with the ability to compute reactivity worths, kinetics parameters, and cross section sensitivities with a Cartesian and hexagonal geometry based transport code. This manuscript serves as a single manual for two separate codes: VARI3D and PERSENT. The VARI3D code (VARIational 3D) is based upon the classic finite difference diffusion theory solver available in DIF3D. The PERSENT code (PERturbation and SENitivity for Transport) is based upon the variational nodal method employed in DIF3D termed VARIANT. The VARIANT solver was added to DIF3D in 1995 and has seen continued development and use for the last 18 years. Because VARI3D primarily uses deprecated coding practices, rather than incorporating the perturbation and sensitivity treatments for transport within VARI3D, a new coding development was built using modern Fortran coding. The primary purpose of this manual is to describe the theory behind PERSENT (and by convenience, that of VARI3D) and discuss the input and output of PERSENT along with giving potential users an idea of how to use it. While this manuscript does describe the input and output of VARI3D, the PERSENT code is intended to be the replacement capability of VARI3D as PERSENT can generate nearly identical (if not superior) diffusion theory results. In this manuscript, the relevant aspects of generalized perturbation theory and exact perturbation theory that apply to both VARI3D and PERSENT are covered. The input and output of VARI3D is displayed by excerpting several of the example problems. Similarly, the input and output of PERSENT is displayed along with tips on how best to use the code. Note that the input and output of the inhomogeneous solver wrapped around DIF3D (DIF3D_IFS) is also discussed as it is needed to carry out some of the sensitivities in PERSENT such as reaction rate ratios. This manuscript describes several perturbation and sensitivity problems, and the results computed using PERSENT. From these sections, potential users should find that PERSENT provides not only the typical tables of numbers desired in perturbation and sensitivity analysis work, but also can visually plot the result for a more thorough understanding of the space and energy distribution (Section 5). Overall, PERSENT is observed to produce accurate reactivity worths and sensitivities for the displayed set of test problems and clearly demonstrates the need to have a transport-based sensitivity capability as evident from the thousands of percent errors observed in the 21-group hexagonal fast reactor problem (covered in Section 7). The uncertainty calculation capability is described in Section 3 and demonstrated in Section 7.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

VARI3D & PERSENT: Perturbation and Sensitivity Analysis

The nodal diffusion method is one of the most widely used approaches in modern reactor analysis. In the nodal diffusion method, a coarse multi-group set of “homogenized” parameters is constructed such that the complex geometry of a reactor core along with the energy dependence of neutron and gamma ray cross sections in a nuclear reactor are conserved in the simpler geometry. The homogenization is typically done on a fuel assembly level as is the case in the DIF3D code developed at Argonne National Laboratory. The nodal methodology is used primarily to predict fuel cycle behavior of nuclear systems of which there is a substantial amount of validation in the literature. Another use of the nodal method is to obtain reactivity coefficients and kinetics parameters for use in a safety analysis of a given nuclear reactor. While there are many ways to obtain reactivity worth and kinetics parameters, the work presented in this manuscript is unique as it provides the user with the ability to compute reactivity worths, kinetics parameters, and cross section sensitivities with a Cartesian and hexagonal geometry based transport code.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

VARI3D & PERSENT: Perturbation and Sensitivity Analysis

The nodal diffusion method is one of the most widely used approaches in modern reactor analysis. In the nodal diffusion method, a coarse multi-group set of “homogenized” parameters is constructed such that the complex geometry of a reactor core along with the energy dependence of neutron and gamma ray cross sections in a nuclear reactor are conserved in the simpler geometry. The homogenization is typically done on a fuel assembly level as is the case in the DIF3D code developed at Argonne National Laboratory. The nodal methodology is used primarily to predict fuel cycle behavior of nuclear systems of which there is a substantial amount of validation in the literature. Another use of the nodal method is to obtain reactivity coefficients and kinetics parameters for use in a safety analysis of a given nuclear reactor. While there are many ways to obtain reactivity worth and kinetics parameters, the work presented in this manuscript is unique as it provides the user with the ability to compute reactivity worths, kinetics parameters, and cross section sensitivities with a Cartesian and hexagonal geometry based transport code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

VARI3D & PERSENT: Perturbation and Sensitivity Analysis

The nodal diffusion method is one of the most widely used approaches in modern reactor analysis. In the nodal diffusion method, a coarse multi-group set of “homogenized” parameters is constructed such that the complex geometry of a reactor core along with the energy dependence of neutron and gamma ray cross sections in a nuclear reactor are conserved in the simpler geometry. The homogenization is typically done on a fuel assembly level as is the case in the DIF3D code developed at Argonne National Laboratory. The nodal methodology is used primarily to predict fuel cycle behavior of nuclear systems of which there is a substantial amount of validation in the literature. Another use of the nodal method is to obtain reactivity coefficients and kinetics parameters for use in a safety analysis of a given nuclear reactor. While there are many ways to obtain reactivity worth and kinetics parameters, the work presented in this manuscript is unique as it provides the user with the ability to compute reactivity worths, kinetics parameters, and cross section sensitivities with a Cartesian and hexagonal geometry based transport code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Dilute Regeneration-Driven Membrane Capacitive Deionization of Synthetic Seawater Using Nanopatterned Membranes and Prussian Blue Analog Electrodes

Membrane capacitive deionization (MCDI) offers energy-efficient seawater desalination but is limited at high salinity by membrane resistance and incomplete electrode regeneration. Nanopatterned ion-exchange membranes, dilute regeneration protocols, and Prussian blue analog (PBA)-functionalized electrodes are combined in a flow-by-MCDI cell. Nanopatterned ion-exchange membranes (hexagonal, octagonal, double-ring, rectangular) enhance interfacial ion transport, with hexagonal geometry delivering ≈12.5% greater surface area and the best performance. PBA-functionalized electrodes increase salt adsorption and charge-transfer kinetic rates. The integrated system lowers the area-specific resistance by 45 Ω cm2, resulting in a 500 mV reduction in the cell voltage for a current density of 2 mA cm−2 for a 35 000 ppm NaCl feed. This improves the energy-normalized salt adsorption six fold (64–382 mmol J−1). Low salinity (2000 ppm) and mixed-salt regeneration sustains a ≈39% water recovery and stable performance for at least seven cycles. Overall, combining nanopatterned membranes, which promote confinement-enhanced ion mobility, and PBA electrodes, which enhance salt adsorption, improved the energy efficiency of MCDI.

Hasan, Mahmudul↗

DOE OSTI2014

Space-time S-matrix and flux-tube S-matrix III. The two-particle contributions

We consider light-like Wilson loops with hexagonal geometry in the planar limit of N=4 Super-Yang-Mills theory. Within the Operator-Product-Expansion framework these loops receive contributions from all states that can propagate on top of the colour flux tube sourced by any two opposite edges of the loops. Of particular interest are the two-particle contributions. They comprise virtual effects like the propagation of a pair of scalars, fermions, and gluons, on top of the flux tube. Each one of them is thoroughly discussed in this paper. Our main result is the prediction of all the twist-2 corrections to the expansion of the dual 6-gluons MHV amplitude in the near-collinear limit at finite coupling. At weak coupling, our result was recently used by Dixon, Drummond, Duhr and Pennington to predict the full amplitude at four loops. At strong coupling, it allows us to make contact with the classical string description and to recover the (previously elusive) AdS 3 mode from the continuum of two-fermion states. More generally, the two-particle contributions serve as an exemplar for all the multi-particle corrections.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2014

Constrained, aqueous growth of three-dimensional single crystalline zinc oxide structures

We study low temperature (90 °C) aqueous growth of single crystal zinc oxide structures through patterned PMMA molds of different sizes, shapes, and orientations. We demonstrate the ability to create 3D shapes with smooth vertical sidewalls. Although the unconstrained growth is influenced by the hexagonal geometry of the underlying crystal structure, the ZnO is shown to conform exactly to any shape patterned. Using electron backscatter diffraction and scanning electron microscopy we show that the mold orientation, in conjunction with control of the growth rates of the c and m planes of the ZnO, is crucial in determining the final structure shape.

36 MATERIALS SCIENCE↗

DOE OSTI2014

Optimal packing size of non-ligated CdSe nanoclusters for microstructure synthesis

Structural and electrostatic properties of nanoclusters of CdSe of diameter 1–2 nm are studied with first principle calculations to determine the optimal size for synthesizing microstructures. Based on robustness of the core structure, i.e., the retention of tetrahedral geometry, hexagonal ring structure, and overall wu{sup ¨}rtzite structure to surface relaxations, we conclude that nanoclusters of ~2 nm diameter are the best candidates to form a dense microstructure with minimal interstitial space. Se-terminated surfaces retain a zigzag structure as Se atoms are pulled out and Cd atoms are pulled in due to relaxation, therefore, are best suited for inter-nanocluster formations.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2015

COUPLING OF RELAP5-SCDAP MOD4.0 AND NEUTRONIC CODES

High-fidelity and accurate nuclear system codes play a key role in the design and analysis of complex nuclear power plants, which consist of multiple subsystems, such as the reactor core (and its fuel, burnable poisons, control elements, etc.), the reactor internal structures, the vessel, and the energy conversion subsystem and beyond to grid demand. Most commonly the interplay between these various subsystems is modeled using coupled codes, each of which represents one of the subsystems. And the most common direct coupling is that of thermal-hydraulics and neutronics codes. The subject of this paper is the coupling of codes that model not only thermal-hydraulics and neutronics, but also structural components damage. Furthermore, the neutronic component is not limited to the sole core solver. The coupled code system encompasses thermal-hydraulics, material performance of the fuel, neutronic solver, and neutronic data preparation. Thus, this paper presents a framework for coupling RELAP5/SCDAPSIM/MOD4.0 with a suite of neutron kinetics codes that includes NESTLE, DRAGON and a version of the ENDF library. The version of the RELAP5/SCDAPSIM/MOD4.0 code used in this work is one developed by Innovate System Software (ISS) as part of the international SCDAP Development and Training Program (SDTP) for best-estimate analysis to model reactor transients including severe accident phenomena. This RELAP5/SCDAPSIM/MOD4.0 code version is also capable of predicting nuclear fuel performance. It uses nodal power distributions to calculate mechanical and thermal parameters such as heat-up, oxidation and meltdown of fuel rods and control rods, the ballooning and rupture of fuel rod cladding, the release of fission products from fuel rods, and the disintegration of fuel rods into porous debris and molten material. On the neutronics side, this work uses the NESTLE and DRAGON codes. NESTLE is a multi-dimensional static and kinetic neutronic code developed at North Carolina State University. It solves up to four energy groups neutron diffusion equations utilizing the Nodal Expansion Method (NEM) in Cartesian or hexagonal geometry. The DRAGON code, developed at Ecole Polytechnique de Montreal, performs lattice physics calculations based on the neutron transport equation and is capable of using very fine energy group structures. In this work, we have developed a coupling approach to exchange data among the various modules. In the coupling process, the generated nuclear data (in fine multigroup energy structure) are collapsed down into two- or four-group energy structures for use in NESTLE. The neutron kinetics and thermal-hydraulics modules are coupled at each time step by using the cross-section data. The power distribution results of the neutronic calculations are transmitted to the thermal-hydraulics code. The spatial distribution of coolant density and the fuel-moderator temperature, which result from the thermal-hydraulic calculations, are transmitted back to the neutron kinetics codes and then the loop is closed using new neutronics results. Details of the actual data transfers will be described in the full length paper.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Verification of EvaluateFLux Utility Program

The EvaluateFlux program is a post-processing utility program for DIF3D, specifically DIF3D-VARIANT which handles Cartesian and hexagonal geometries. The EvaluateFlux program was developed to allow users to obtain flux and power traverses through the geometry domain, and its initial purpose was to facilitate foil analysis by evaluating the flux solution from DIF3D-VARIANT and combining it with foil cross section data. The EvaluateFlux program can calculate the neutron flux, as well as the reaction rates, at any user provided evaluation point. It does this by identifying the spatial mesh associated with the evaluation point and then evaluates the polynomial based neutron flux moments stored in the NHFLUX file at that point. The output of EvaluateFlux varies depending on the input setup. The maximum output includes the neutron flux and microscopic and macroscopic reaction rates at each evaluation point. The purpose of this work is to verify the outputs of EvaluateFlux. Simple models that have hand calculatable results are first defined and used to verify the EvaluateFlux outputs. More complex cases are then added where a duplicate program of EvaluateFlux that uses PrintTables outputs of the binary files is used to verify the EvaluateFlux outputs. In those complex cases, hand calculations of selected evaluation points were also displayed to confirm the software verification. For all the tests done, the hand calculations agreed well with those calculated by EvaluateFlux. For the larger complex problems, the duplicate program that can process hundreds of evaluation points was able to identify that zero points within some meshes have large errors. This aspect was attributed to the truncation error on the input provided to the duplicate program and is not a concern for the accuracy of the EvaluateFlux software.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2024

Verification of the EvaluateFlux Utility Program

The EvaluateFlux program is a post-processing utility program for DIF3D, specifically DIF3D-VARIANT which handles Cartesian and hexagonal geometries. The EvaluateFlux program was developed to allow users to obtain flux and power traverses through the geometry domain, and its initial purpose was to facilitate foil analysis by evaluating the flux solution from DIF3D-VARIANT and combining it with foil cross section data. The EvaluateFlux program can calculate the neutron flux, as well as the reaction rates, at any user provided evaluation point. It does this by identifying the spatial mesh associated with the evaluation point and then evaluates the polynomial based neutron flux moments stored in the NHFLUX file at that point. The output of EvaluateFlux varies depending on the input setup. The maximum output includes the neutron flux and microscopic and macroscopic reaction rates at each evaluation point. The purpose of this work is to verify the outputs of EvaluateFlux. Simple models that have hand calculatable results are first defined and used to verify the EvaluateFlux outputs. More complex cases are then added where a duplicate program of EvaluateFlux that uses PrintTables outputs of the binary files is used to verify the EvaluateFlux outputs. In those complex cases, hand calculations of selected evaluation points were also displayed to confirm the software verification. For all the tests done, the hand calculations agreed well with those calculated by EvaluateFlux. For the larger complex problems, the duplicate program that can process hundreds of evaluation points was able to identify that zero points within some meshes have large errors. This aspect was attributed to the truncation error on the input provided to the duplicate program and is not a concern for the accuracy of the EvaluateFlux software.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Generalization of NESTLE into a Multi-Energy N-Group Formulation

The NESTLE code is a few-group, diffusion-based core simulation code that employs the Nodal Expansion Method (NEM). NESTLE has the capability to solve both steady-state and transient problems, using either Cartesian or hexagonal geometry, and to apply thermal-hydraulic feedback including single and two-phase flow. A modern version of the NESTLE nodal-diffusion simulator is currently maintained and developed at the University of Tennessee. The original version of the code was limited to only two or four neutron energy groups, therefore, this article describes recent work carried out to expand NESTLE into a general multi-group solver. Generalization of a fixed group finite-difference diffusion theory into arbitrary N groups is a well-known problem [3], however, this paper describes the intricacies associated with the generalization of the Nodal Expansion Method, and the methodical and computational challenges encountered and managed along the way.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Use of spectral shift to greatly improve uranium utilisation in FHR designs

Fluoride-salt-cooled high-temperature reactors (FHRs) are thermal-spectrum nuclear reactors defined by their use of liquid-fluoride-salt coolants, together with tri-isotropic (TRISO)-coated particle fuels and graphite moderator materials (Forsberg, 2004). FHRs are also characterized by their improved thermodynamic efficiency relative to gas-cooled reactors. The liquid salt coolant is designed to operate with high core outlet temperature and nearly atmosphere pressure. These features allow the FHR to possess higher core power density compared to gas-cooled TRISO-based systems. However, industry fabrication limitations result in a low volumetric fraction of fuel kernel in TRISO particles. In order to obtain reasonable operation periods, the enrichment needs to be in the high LEU (low enriched uranium) range as was proposed in the original design (Greene, 2010). Recent studies (Ilas, 2014) presented newly repurposed SmAHTR-CTC reactor concept that relies on prior experience obtained from the 2010 initial SmAHTR design (Greene, 2010). This design consists of 19 hexagonal fuel assemblies arranged in a 3-ring hexagonal geometry with 18 fuel plates each. The active length of fuel plates in each fuel assembly is 4 meters. The proposed SmAHTR-CTC achieves substantial improvements from the neutronic performance standpoint over the original concept. The results indicate that a fuel cycle length of over two years with discharged fuel burnup of 40 gigawatt-day per metric ton of initial heavy metal (HM) is possible with an 8% fuel enrichment in a once-through fuel cycle. However, the lower uranium enrichment loading requires a more frequent refueling strategy and hence considerably reduces the cycle length. It is necessary to achieve reasonable cycle length values but at the same time to extend the residence time of the fuel, especially since the costs associated with TRISO-fuel are substantially higher (Shropshire et al., 2008) than traditional UO{sub 2} fuel ($10,000/kg for TRISO vs $200/kg for UO{sub 2}). This paper presents an approach to extend the operational time and improve fuel utilization of FHR by means of spectral shift control. This is achieved by incorporating movable graphite structures into the design, instead of the static graphite blocks presently envisaged. At beginning of life (BOL), the graphite structures are withdrawn. The reactor is under-moderated, with excess neutrons being primarily captured in {sup 238}U, breeding {sup 239}Pu. Towards the end of life (EOL), the graphite structures are inserted, thermalizing the neutron spectrum and increasing reactivity. The extra {sup 239}Pu bred during the cycle is then burned, allowing the cycle and discharge burnup to be extended. Such movable structures regime should be feasible from an engineering point of view, given that it is broadly equivalent to inserting and withdrawing control rods. The preliminary results indicate that such a design may considerably improve the fuel utilization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Conformal geometry of null hexagons for Wilson loops and scattering amplitudes

The cross-ratios do not uniquely fix the class of conformally equivalent configurations of null polygons. In view of applications to Wilson loops and scattering amplitudes we characterise all conformal classes of null hexagon configurations belonging to given points in cross-ratio space. At first this is done for the ordered set of vertices. Including the edges, we then investigate the equivalence classes under conformal transformations for null hexagons. This is done both for the set of null hexagons closed in finite domains of Minkowski space as well as for the set including those closed via infinity.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗